[Federal Register Volume 65, Number 249 (Wednesday, December 27, 2000)]
[Notices]
[Pages 81907-81936]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-33012]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 4, 2000, through December 15, 2000.
The last biweekly notice was published on December 13, 2000.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By January 26, 2001, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and electronically from
the ADAMS Public Library component on the NRC Web site,
http://www.nrc.gov (the Electronic Reading Room). If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for
[[Page 81908]]
leave to intervene or who has been admitted as a party may amend the
petition without requesting leave of the Board up to 15 days prior to
the first prehearing conference scheduled in the proceeding, but such
an amended petition must satisfy the specificity requirements described
above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site,
http://www.nrc.gov (the Electronic Reading Room).
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: September 29, 2000.
Description of amendment request: The proposed amendments would
make various changes to the Technical Specifications (TS) to support a
change in fuel vendors from Siemens Power Corporation to General
Electric and a transition to the use of GE 14 fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Evaluation of effect on the probability of an accident previously
evaluated:
1. Administrative Changes. The revisions to Current Technical
Specifications (CTS) Sections 2.1.B, ``Thermal Power, High Pressure and
High Flow,'' and 3.6.A, ``Recirculation Loops,'' regarding the Minimum
Critical Power Ratio (MCPR) Safety Limit, the changes to CTS Section
6.9.A.6.b, ``Core Operating Limits Report,'' and the changes to the
definitions are administrative changes and will not affect the
probability of an accident previously evaluated. These changes do not
affect plant systems, structures, or components. No plant mitigating
systems or functions are affected by these changes.
2. Control Rod Operability and Scram Insertion Times Methodology.
The changes to CTS Sections 3/4.3.C, ``Control Rod Operability,'' 3/
4.3.D, ``Maximum Scram Insertion Times,'' 3/4.3.E, ``Average Scram
Insertion Times,'' 3/4.3.F, ``Group Scram Insertion Times,'' 3/4.3.G,
``Control Rod Scram Accumulators,'' 3/4.3.H, ``Control Rod Coupling,''
and 3/4.3.I, ``Control Rod Position Indication System,'' revise the
methodology for determining rod operability and control rod scram time
requirements for operation. These changes do not physically alter plant
systems, structures or components and therefore do not affect the
probability of an accident previously evaluated.
3. Control Rod Scram Times. The addition of required scram times
for General Electric (GE) analyzed cores does not physically alter
plant systems, structures or components and therefore does not affect
the probability of an accident previously evaluated.
4. Rod Worth Minimizer (RWM). The revision to CTS Section 3/4.3.L,
``Rod Worth Minimizer,'' lowers the power level at which the analyzed
rod position sequence must be followed. This change does not affect
plant systems, structures, or components. Because there is no possible
control rod configuration that results in a control rod worth that
could exceed the 280 cal/gram fuel design limit, the probability of an
accident is not increased.
5. Transient Linear Heat Generation Rate (TLHGR). The revisions to
CTS Section 3.11.B, ``Transient Linear Heat Generation Rate,'' add fuel
thermal limits that are monitored to ensure that TLHGR is not violated.
These changes do not physically alter plant systems, structures or
components and therefore do not affect the probability of an accident
previously evaluated.
Evaluation of the effect on the consequences of an accident
previously evaluated.
1. Administrative Changes. The revisions to CTS Sections 2.1.B and
3.6.A, regarding the MCPR Safety Limit, the changes to CTS Section
6.9.A.6.b regarding the COLR, and the changes to
[[Page 81909]]
the definitions are administrative changes and will not affect the
consequences of an accident previously evaluated. These changes do not
affect plant systems, structures, or components. No plant mitigating
systems or functions are affected by these changes.
2. Control Rod Operability and Scram Insertion Times Methodology.
The revisions to CTS Sections 3/4.3.C, 3/4.3.D, 3/4.3.E, 3/4.3.F, 3/
4.3.G, 3/4.3.H, and 3/4.3.I are made to ensure that appropriate scram
times are reflected in the TS for GE methodology. The scram timing
requirements ensure that the negative reactivity insertion rate assumed
in the safety analyses is preserved. CTS methods ensure this by
limiting scram times for individual rods, the average scram time, and
local scram times (i.e., a four control rod group). The proposed
revisions, based on the Improved Technical Specification (ITS) methods,
ensure this by limiting the scram times for individual rods, the number
of slow rods, and the number of adjacent slow rods. Each of these
methods ensure equivalent protection of the assumed reactivity
insertion rate. Therefore, there is no change to the consequences of a
previously evaluated accident or transient.
In addition, numerous changes to the control rod operability and
scram timing requirements were made to reflect the ITS approach to
these requirements. These revisions consist of administrative changes,
more restrictive changes, and less restrictive changes. The discussion
of each of these categories is provided below.
Administrative changes. These consist of restructuring,
interpretation, rearranging of requirements, and other changes not
substantially revising an existing requirement. Therefore, these
changes do not affect the consequences of an accident previously
evaluated.
More restrictive changes. These consist of changes resulting in
added restrictions or eliminating flexibility. The more restrictive
requirements continue to ensure that process variables, structures,
systems and components are maintained consistent with the safety
analyses and licensing basis. Therefore, these changes do not involve
an increase in the consequences of an accident previously evaluated.
Less restrictive changes. The less restrictive changes involve
increasing the time to complete actions, increasing the time intervals
between required surveillances, and deleting or revising the
applicability of certain actions. The time to complete actions and the
surveillance frequencies are not assumed in the analysis of the
consequences of any accidents previously evaluated, and therefore,
cannot increase the consequences of such accidents. The deleted or
revised actions are not assumed in the safety analyses for any
evaluated accidents. The revised scram timing methods will result in
operating thermal limits that will maintain the identical safety
limits. Thus, the consequences of the evaluated accidents will not
increase.
3. Control Rod Scram Times. Cycle-specific analyses that use the GE
methodology scram times will meet all of the same safety limit
acceptance criteria. Additionally, for the non-cycle specific events in
the Updated Final Safety Analysis Report (UFSAR), GE has determined
that there is negligible impact on results of events which are not
analyzed on a cycle-specific basis. Therefore, there is no change to
the consequences of a previously evaluated accident or transient.
4. RWM. The RWM enforces the analyzed rod position sequence to
ensure that the initial conditions of the Control Rod Drop Accident
(CRDA) analysis are not violated. Compliance with the analyzed rod
position sequence, and operability of the RWM is required in Mode 1,
``Power Operation,'' and Mode 2, ``Startup,'' when thermal power is
less than or equal to 10% Rated Thermal Power (RTP). When thermal power
is greater than 10% RTP, there is no possible control rod configuration
that results in a control rod worth that could exceed the 280 cal/gm
fuel design limit during a CRDA. Because the fuel design limit of 280
cal/gm is not exceeded, this change to lower the Low Power Setpoint
(LPSP) does not increase the consequences of an accident previously
evaluated.
5. TLHGR. The changes to this section are analytical in nature and
do not affect plant systems, structures, or components. The changes in
this section revise the description of fuel thermal limits that are
monitored to ensure that the TLHGR limit is not violated. The TLHGR
protects the fuel from 1% plastic strain and fuel centerline melt.
Because these criteria have not changed, the consequences of an
accident have not changed.
Therefore, the proposed changes to the CTS do not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
1. Administrative Changes. The revisions to CTS Sections 2.1.B and
3.6.A, regarding the MCPR Safety Limit, the changes to CTS Section
6.9.A.6.b regarding the COLR, and the changes to the definitions are
administrative changes and will not create the possibility of a new or
different kind of accident. These changes do not affect plant systems,
structures, or components. No plant mitigating systems or functions are
affected by these changes.
2. Control Rod Operability and Scram Insertion Times Methodology.
The changes to CTS Sections 3/4.3.C, 3/4.3.D, 3/4.3.E, 3/4.3.F, 3/
4.3.G, 3/4.3.H, and 3/4.3.I revise the control rod operability and
scram time requirements for operation. These changes do not physically
alter plant systems, structures or components and therefore do not
create the possibility of a new or different kind of accident.
3. Control Rod Scram Times. These changes do not physically alter
plant systems, structures or components and therefore do not create the
possibility of a new or different kind of accident.
4. RWM. The revisions to CTS Section 3/4.3.L lower the power level
at which the analyzed rod position sequence must be followed. This
change does not affect plant systems, structures, or components.
Because there is no possible control rod configuration that results in
a control rod worth that could exceed the 280 cal/gm fuel design limit,
no new or different type of accident is created.
5. TLHGR. The revisions to CTS Section 3.11.B revise the
description of fuel thermal limits that are monitored to ensure that
TLHGR is not violated. These changes are analytical in nature and do
not affect plant systems, structures, or components. Therefore, the
changes do not create the possibility of a new or different kind of
accident.
Therefore, the proposed changes to the CTS do not create the
possibility of a new or different kind of accident from any previously
evaluated.
Does the proposed change involve a significant reduction in a
margin of safety?
1. Administrative Changes. The revisions to CTS Sections 2.1.B and
3.6.A, regarding the MCPR Safety Limit, the changes to CTS Section
6.9.A.6.b, regarding the COLR, and the changes to the definitions are
administrative changes and will not reduce the margin of safety. These
changes do not affect plant systems, structures, or components. No
plant mitigating systems or functions are affected by these changes.
2. Control Rod Operability and Scram Insertion Times Methodology.
The revisions to the CTS control rod operability and scram insertion
times
[[Page 81910]]
ensure that the negative reactivity insertion rate assumed in the
safety analyses is preserved. CTS methods ensure this by limiting scram
times for individual rods, the average scram time, and the local scram
times (i.e., a four control rod group). ITS methods ensure this by
limiting the scram times for individual rods, the number of slow rods,
and the number of adjacent slow rods. Each of these methods ensure
equivalent protection of the assumed reactivity insertion rate.
Therefore, the changes do not involve a reduction in the margin of
safety.
In addition, numerous changes to the control rod operability and
scram timing requirements were made to reflect the ITS approach to
these requirements. These revisions consist of administrative changes,
more restrictive changes, and less restrictive changes. The discussion
of each of these categories is provided below.
Administrative Changes. These consist of restructuring,
interpretation, and complex rearranging of requirements, and other
changes not substantially revising an existing requirement. Therefore,
these changes do not affect the margin of safety.
More restrictive changes. These consist of changes resulting in
added restrictions or eliminating flexibility. The more restrictive
requirements continue to ensure that process variables, structures,
systems and components are maintained consistent with the safety
analyses and licensing basis. Therefore, these changes do not reduce
the margin of safety.
Less restrictive changes. The less restrictive changes involve
increasing the time to complete actions, increasing the time intervals
between required surveillances, and deleting or revising the
applicability of certain actions. The time to complete actions and the
surveillance frequencies have been extended for several reasons,
including experience showing low probability of failures, the benefit
of allowing time to perform actions without undue haste, or due to
compensating changes in other actions. The deleted or revised actions
are not assumed in the safety analyses for any evaluated accidents.
Thus, there is no significant reduction in the margin of safety.
3. Control Rod Scram Times. The addition of required scram times
for GE analyzed cores based on GE analysis methodology does not involve
a reduction in the margin of safety. For GE analyzed cores, cycle-
specific analyses using the actual averaged scram times provide MCPR
operating limits that will ensure the MCPR safety limit is not
violated. Therefore, the fuel remains appropriately protected and no
margins of safety are reduced.
4. RWM. The RWM enforces the analyzed rod position sequence to
ensure that the initial conditions of the CRDA analysis are not
violated. Compliance with the analyzed rod position sequence, and
operability of the RWM is required in Modes 1 and 2 when thermal power
is less than or equal to 10% rated thermal power (RTP). When thermal
power is greater than 10% RTP, there is no possible control rod
configuration that results in a control rod worth that could exceed the
280 cal/gm fuel design limit during a CRDA. Because the fuel design
limit of 280 cal/gm is not exceeded above 10% RTP, this change to
reduce the LPSP does not reduce a margin of safety.
5. TLHGR. The addition of the ratio of Maximum Fraction of Limiting
Power Density (MFLPD) to the Fraction of Rated Thermal Power (FRTP)
provides thermal limit protection for GE fuel. This provides equivalent
protection to ensure that the TLHGR limit is maintained. Therefore, the
revisions to CTS Section 3.11.B will not reduce a margin of safety.
Therefore, these proposed changes to the CTS do not involve a
significant reduction in the margin safety.
Proposed Changes to ITS
Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Evaluation of the effect on the probability of an accident
previously evaluated.
1. Administrative Changes. The revision to Improved Technical
Specification (ITS) Section 5.6.5, ``Core Operating Limits Report,''
and the added definitions are purely administrative changes and do not
affect the probability or consequences of an accident previously
evaluated.
2. Control Rod Scram Times. The revision to ITS Table 3.1.4-1,
``Control Rod Scram Times,'' adds scram time requirements for GE
analyzed cores. This change does not physically alter plant systems,
structures or components and therefore does not affect the probability
of an accident previously evaluated.
3. Average Power Range Monitor (APRM) Gain and Setpoint. The
revisions to ITS Section 3.2.4, ``Average Power Range Monitor (APRM)
Gain and Setpoint,'' revise the description of fuel thermal limits that
are monitored to ensure the TLHGR is not violated. The changes to this
section are analytical in nature and do not affect plant systems,
structures, or components and therefore will not affect the probability
of an accident previously evaluated.
Evaluation of the effect on the consequences of an accident
previously evaluated.
1. Administrative Changes. The revision to ITS Section 5.6.5 and
the added definitions are purely administrative changes and do not
affect the probability or consequences of an accident previously
evaluated.
2. Control rod scram times. The revisions to ITS Section 3.1.4,
``Control Rod Scram Insertion Times,'' are made to ensure the
appropriate scram times are reflected in the Technical Specifications
(TS) for GE methodology. The scram timing requirements ensure that the
negative reactivity insertion rate assumed in the safety analyses is
preserved. Cycle specific analyses that use the GE methodology scram
times will meet all of the same safety limit acceptance criteria.
Additionally, for the non-cycle specific UFSAR events, GE has
determined that there is negligible impact on the results of events
which are not analyzed on a cycle specific basis. Therefore, there is
no change to the consequences of a previously evaluated accident or
transient due to the TS changes.
3. APRM Gain and Setpoint. The revisions to ITS Section 3.2.4 will
not increase the consequences of an accident previously evaluated. The
changes to this section are analytical in nature and do not affect
plant systems, structures, or components. The changes in this section
revise the description of fuel thermal limits that are monitored to
ensure the TLHGR limit is not violated. The TLHGR protects the fuel
from 1% plastic strain and fuel centerline melt. Because these criteria
have not changed, the consequences of an accident have not changed.
Therefore, the proposed changes to the ITS do not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
1. Administrative Changes. The revision to ITS Section 5.6.5 and
the added definitions are purely administrative changes and therefore
do not create the possibility of a new or different kind of accident.
2. Control Rod Scram Insertion Times. The revisions to ITS Section
3.1.4 do not create the possibility of a new or different kind of
accident from any accident previously evaluated. The changes to these
sections revise the control rod scram time requirements for
[[Page 81911]]
operation. This change does not physically alter plant systems,
structures, or components.
3. APRM Gain and Setpoint. The revisions to ITS Section 3.2.4 will
not create the possibility of a new or different kind of accident. The
changes to this section are analytical in nature and do not affect
plant systems, structures, or components. The changes in this section
revise the description of fuel thermal limits that are monitored to
ensure the TLHGR limit is not violated.
Therefore, the proposed changes to the ITS do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Does the proposed change involve a significant reduction in a
margin of safety?
1. Administrative Changes. The revision to ITS Section 5.6.5 and
the added definitions are purely administrative changes and do not
affect the margin of safety.
2. Control Rod Scram Insertion Times. For GE analyzed cores, cycle-
specific analyses using the actual averaged scram times provide MCPR
operating limits that will ensure that the MCPR safety limit is not
violated. Therefore, the fuel remains appropriately protected and no
margins of safety are reduced.
3. APRM Gain and Setpoint. The addition of MFLPD/FRTP provides
thermal limit protection for GE fuel. This provides equivalent
protection to ensure that the TLHGR limit is maintained. Therefore, the
revisions to ITS Section 3.2.4 will not reduce a margin of safety.
Therefore, the proposed changes to the ITS do not involve a
significant reduction in the margin of safety.
Based on the above evaluation, ComEd has concluded that these
changes involve no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: November 10, 2000.
Description of amendment request: The proposed amendments would
revise several sections of the Technical Specifications (TS) and add a
new TS section to incorporate Oscillation Power Range Monitor (OPRM)
Instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the probability
or consequences of an accident previously evaluated?
The proposed changes for LaSalle County Station will delete the
thermal hydraulic instability administrative requirements and Power
versus Flow figure and references to it from the TS, and insert a new
TS for the OPRM instrumentation. The proposed TS will allow the
enabling of the OPRM instrumentation trips. The deletion of the thermal
hydraulic instability administrative requirements and Power versus Flow
figure and the requirements to have an operable OPRM instrumentation
trip does not have an effect on any accident previously evaluated or
the associated accident assumptions. Thus, the proposed changes do not
significantly increase the probability of an accident previously
evaluated.
The proposed changes do not adversely affect the integrity of the
fuel cladding, reactor coolant system or secondary containment. As
such, the radiological consequences of previously evaluated accident
are not changed. Therefore, the proposed changes do not increase the
consequences of an accident previously evaluated.
Does the change create the possibility of a new or different kind
of accident from any accident previously evaluated?
The proposed changes do not effect the assumed accident performance
of any structure, system, or component previously evaluated. The
proposed changes do not introduce any new modes of system operation or
failure mechanisms.
The OPRM instrumentation will initiate an automatic reactor trip
upon detection of an instability that could threaten the Minimum
Critical Power Ratio (MCPR) safety limit. The OPRM Instrumentation
System consists of four (4) OPRM instrumentation trip channels. When
one OPRM instrumentation module is inoperable, the remaining redundant
OPRM Instrumentation module in the associated OPRM trip channel
maintains the operability of the trip channel and thus there is no loss
of trip function redundancy.
Thus, these proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
Boiling Water Reactors are susceptible to thermal hydraulic
instabilities if operated at high power and low flow conditions. 10 CFR
50, Appendix A, General Design Criterion (GDC) 10, ``Reactor design,''
requires the reactor core and associated coolant, control, and
protection systems to be designed with appropriate margin to assure
that acceptable fuel design limits are not exceeded during any
condition of normal operation, including the effects of anticipated
operational occurrences. Additionally, GDC 12, ``Suppression of reactor
power oscillation,'' requires the reactor core and associated coolant,
control, and protection systems to be designed to assure that power
oscillations which can result in conditions exceeding acceptable fuel
design limits are either not possible or can be reliably and readily
detected and suppressed.
The detection and suppression of instability is required to insure
that the MCPR safety limit is not exceeded during a transient. The OPRM
instrumentation will initiate an automatic reactor trip upon detection
of an instability that could threaten the MCPR safety limit.
The OPRM Instrumentation System consists of four (4) OPRM
instrumentation trip channels, each trip channel consisting of two OPRM
instrumentation modules. Each OPRM instrumentation module receives
input from LPRMs. Each OPRM instrumentation module also receives input
from the RPS Average Power Range Monitor (APRM) power and flow signals
to automatically enable the trip function of the OPRM instrumentation
module.
Each OPRM instrumentation module is continuously tested by a self-
test function. On detection of any OPRM instrumentation module failure,
either a ``Trouble'' or ``INOP'' alarm is activated. The OPRM
instrumentation module provides an ``INOP'' alarm when the self-test
feature indicates that the OPRM instrumentation module may not be
capable of meeting its functional
[[Page 81912]]
requirements. When one OPRM instrumentation module is inoperable, the
remaining redundant OPRM Instrumentation module in the associated OPRM
trip channel maintains the operability of the trip channel and thus
there is no loss of trip function redundancy. The OPRM Instrumentation
System provides compliance with GDC 10 and GDC 12.
The incorporation of the OPRM instrumentation into the TS will
allow the deletion of the current thermal hydraulic instability
administrative requirements and Power versus Flow TS Figure and
associated actions. The OPRM instrumentation will provide the same
level of assurance that the MCPR safety limit will not be violated for
anticipated oscillations as that provided by the Power versus Flow TS
Figure.
The OPRM Instrumentation System enabled region of the Power versus
Flow figure was adjusted to maintain the same level of protection
against the occurrence of a thermal-hydraulic instability by
maintaining the pre-power uprate absolute power and flow coordinates. A
5% Power Uprate was approved for LaSalle County Station, Units 1 and 2,
by Facility Operating License Amendments 140 and 125, respectively, in
an NRC letter dated May 9, 2000.
The proposed changes do not affect the margin of safety as the OPRM
Instrumentation will initiate an automatic reactor trip upon detection
of an instability that could threaten the MCPR safety limit.
Thus, this proposed change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: September 29, 2000.
Description of amendment request: The proposed amendments would
make various changes to the Technical Specifications (TSs) to support a
change in fuel vendors from Siemens Power Corporation to General
Electric and a transition to the use of General Electric (GE) 14 fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Proposed Changes to Current Technical Specifications
Evaluation of effect on the probability of an accident previously
evaluated:
1. Administrative Changes. The revisions to Current Technical
Specifications (CTS) Sections 2.1.B, ``Thermal Power, High Pressure and
High Flow,'' and 3.6.A, ``Recirculation Loops,'' regarding the Minimum
Critical Power Ratio (MCPR) Safety Limit, the changes to Section 3.11B,
``Transient Linear Heat Generation Rate,'' regarding the surveillance
to monitor Transient linear heat Generation Rate (TLHGR) using either
the ratio of the Maximum Fraction of Limiting Power Density (MFLPD) to
the Fraction of Rated Thermal Power (FRTP) or the Fuel Design Limiting
Ratio for Centerline (FDLRC) Melt, and the addition of the NRC approved
RODEX2A methodology, are administrative changes and will not affect the
probability of an accident previously evaluated. These changes do not
affect plant systems, structures, or components. No plant mitigating
systems or functions are affected by these changes.
2. Control Rod Operability and Scram Insertion Times Methodology.
The changes to CTS Sections 3/4.3.C, ``Control Rod Operability,'' 3/
4.3.D, ``Maximum Scram Insertion Times,'' 3/4.3.E, ``Average Scram
Insertion Times,'' 3/4.3.F, ``Group Scram Insertion Times,'' 3/4.3.G,
``Control Rod Scram Accumulators,'' 3/4.3.H, ``Control Rod Coupling,''
and 3/4.3.I, ``Control Rod Position Indication System,'' revise the
methodology for determining rod operability and control rod scram time
requirements for operation. These changes do not physically alter plant
systems, structures or components and therefore do not affect the
probability of an accident previously evaluated.
3. Control Rod Scram Times. The addition of required scram times
for General Electric (GE) analyzed cores does not physically alter
plant systems, structures or components and therefore does not affect
the probability of an accident previously evaluated.
Evaluation of the effect on the consequences of an accident
previously evaluated.
1. Administrative Changes. The revisions to CTS Sections 2.1.B and
3.6.A, regarding the MCPR Safety Limit are administrative changes and
will not affect the consequences of an accident previously evaluated.
These changes do not affect plant systems, structures, or components.
No plant mitigating systems or functions are affected by these changes.
The changes to this section are analytical in nature and do not affect
plant systems, structures, or components. The administrative changes to
Section 3.11.B revise the description of fuel thermal limits that are
monitored to ensure the TLHGR limit is not violated. TLHGR protects the
fuel from 1% plastic strain and fuel centerline melt. Because these
criteria have not changed, the consequences of an accident have not
changed. The NRC approved burnup extension for RODEX2A has been
demonstrated to meet all applicable design criteria. Therefore, the
addition of the NRC approved RODEX2A methodology does not increase the
consequences of an accident previously evaluated.
2. Control Rod Operability and Scram Insertion Times Methodology.
The revisions to CTS Sections 3/4.3.C, 3/4.3.D, 3/4.3.E, 3/4.3.F, 3/
4.3.G, 3/4.3.H, and 3/4.3.I are made to ensure that appropriate scram
times are reflected in the TS for GE methodology. The scram timing
requirements ensure that the negative reactivity insertion rate assumed
in the safety analyses is preserved. CTS methods ensure this by
limiting scram times for individual rods, the average scram time, and
local scram times (i.e., a four control rod group). The proposed
revisions, based on the Improved Technical Specification (ITS) methods,
ensure this by limiting the scram times for individual rods, the number
of slow rods, and the number of adjacent slow rods. Each of these
methods ensure equivalent protection of the assumed reactivity
insertion rate. Therefore, there is no change to the consequences of a
previously evaluated accident or transient.
In addition, numerous changes to the control rod operability and
scram timing TS Sections were made to reflect the ITS approach to these
requirements. These revisions consist of administrative changes, more
restrictive changes, and less restrictive changes. The discussion of
each of these categories is provided below.
Administrative changes. These consist of restructuring,
interpretation,
[[Page 81913]]
rearranging of requirements, and other changes not substantially
revising an existing requirement. Therefore, these changes do not
affect the consequences of an accident previously evaluated.
More restrictive changes. These consist of changes resulting in
added restrictions or eliminating flexibility. The more restrictive
requirements continue to ensure that process variables, structures,
systems and components are maintained consistent with the safety
analyses and licensing basis. Therefore, these changes do not involve
an increase in the consequences of an accident previously evaluated.
Less restrictive changes. The less restrictive changes involve
increasing the time to complete actions, increasing the time intervals
between required surveillances, and deleting or revising the
applicability of certain actions. The time to complete actions and the
surveillance frequencies are not assumed in the analysis of the
consequences of any accidents previously evaluated, and therefore,
cannot increase the consequences of such accidents. The deleted or
revised actions are not assumed in the safety analyses for any
evaluated accidents. The revised scram timing methods will result in
operating thermal limits that will maintain the identical safety
limits. Thus, the consequences of the evaluated accidents will not
increase.
3. Control Rod Scram Times. Cycle-specific analyses that use the GE
methodology scram times will meet all of the same safety limit
acceptance criteria. Additionally, for the non-cycle specific events in
the Updated Final Safety Analysis Report (UFSAR), GE has determined
that there is negligible impact on results of events which are not
analyzed on a cycle-specific basis. Therefore, there is no change to
the consequences of a previously evaluated accident or transient.
Therefore, the proposed changes to the CTS do not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
1. Administrative Changes. The revisions to CTS Sections 2.1.B and
3.6.A, regarding the MCPR Safety Limit, the revisions to CTS Section
3.11.B to revise the description of TLHGR, and the addition of the NRC
approved RODEX2A methodology are administrative changes and will not
create the possibility of a new or different kind of accident. These
changes do not affect plant systems, structures, or components. No
plant mitigating systems or functions are affected by these changes.
2. Control Rod Operability and Scram Insertion Times Methodology.
The changes to CTS Sections 3/4.3.C, 3/4.3.D, 3/4.3.E, 3/4.3.F, 3/
4.3.G, 3/4.3.H, and 3/4.3.I revise the control rod operability and
scram time requirements for operation. These changes do not physically
alter plant systems, structures or components and therefore do not
create the possibility of a new or different kind of accident.
3. Control Rod Scram Times. These changes do not physically alter
plant systems, structures or components and therefore do not create the
possibility of a new or different kind of accident.
Therefore, the proposed changes to the CTS do not create the
possibility of a new or different kind of accident from any previously
evaluated.
Does the proposed change involve a significant reduction in a
margin of safety?
1. Administrative Changes. The revisions to CTS Sections 2.1.B and
3.6.A, regarding the MCPR Safety Limit, and the changes to CTS Section
3.11.B regarding the surveillance to monitor TLHGR, and the addition of
the NRC approved RODEX2A methodology are administrative changes and
will not reduce the margin of safety. These changes do not affect plant
systems, structures, or components. No plant mitigating systems or
functions are affected by these changes.
2. Control Rod Operability and Scram Insertion Times Methodology.
The revisions to the CTS control rod operability and scram insertion
times ensure that the negative reactivity insertion rate assumed in the
safety analyses is preserved. CTS methods ensure this by limiting scram
times for individual rods, the average scram time, and local scram
times (i.e., a four control rod group). ITS methods ensure this by
limiting the scram times for individual rods, the number of slow rods,
and the number of adjacent slow rods. Each of these methods ensure
equivalent protection of the assumed reactivity insertion rate.
Therefore, the changes do not involve a reduction in the margin of
safety.
In addition, numerous changes to the control rod operability and
scram timing TS Sections were made to reflect the ITS approach to these
requirements. These revisions consist of administrative changes, more
restrictive changes, and less restrictive changes. The discussion of
each of these categories is provided below.
Administrative Changes. These consist of restructuring,
interpretation, and complex rearranging of requirements, and other
changes not substantially revising an existing requirement. Therefore,
these changes do not affect the margin of safety.
More restrictive changes. These consist of changes resulting in
added restrictions or eliminating flexibility. The more restrictive
requirements continue to ensure that process variables, structures,
systems and components are maintained consistent with the safety
analyses and licensing basis. Therefore, these changes do not reduce
the margin of safety.
Less restrictive changes. The less restrictive changes involve
increasing the time to compete actions, increasing the time intervals
between required surveillances, and deleting or revising the
applicability of certain actions. The time to complete actions and the
surveillance frequencies have been extended for several reasons,
including experience showing low probability of failures, the benefit
of allowing time to perform actions without undue haste, or due to
compensating changes in other actions. The deleted or revised actions
are not assumed in the safety analyses for any evaluated accidents.
Thus, there is no significant reduction in the margin of safety.
3. Control Rod Scram Times. The addition of required scram times
for GE analyzed cores based on GE analysis methodology does not involve
a reduction in the margin of safety. For GE analyzed cores, cycle-
specific analyses using the actual averaged scram times provide MCPR
operating limits that will ensure the MCPR safety limit is not
violated. Therefore, the fuel remains appropriately protected and no
margins of safety are reduced.
Therefore, these proposed changes to the CTS do not involve a
significant reduction in the margin safety.
Proposed Changes to ITS
Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Evaluation of the effect on the probability of an accident
previously evaluated.
1. Administrative change. The addition of the NRC approved RODEX2A
methodology is an administrative change and will not affect the
probability of an accident previously evaluated. This change does not
affect plant systems, structures, or components. No plant mitigating
systems or functions are affected by these changes.
2. Control Rod Scram Times. The revision to ITS Table 3.1.4-1,
``Control
[[Page 81914]]
Rod Scram Times,'' adds scram time requirements for GE analyzed cores.
This change does not physically alter plant systems, structures or
components and therefore does not affect the probability of an accident
previously evaluated.
Evaluation of the effect on the consequences of an accident
previously evaluated.
1. Administrative Change. The NRC approved burnup extension for
RODEX2A has been demonstrated to meet all applicable design criteria.
Therefore, the addition of the NRC approved RODEX2A methodology does
not increase the consequences of an accident previously evaluated.
2. Control Rod Scram Times. The revisions to ITS Section 3.1.4,
``Control Rod Scram Insertion Times,'' are made to ensure the
appropriate scram times are reflected in the Technical Specifications
(TS) for GE methodology. The scram timing requirements ensure that the
negative reactivity insertion rate assumed in the safety analyses is
preserved. Cycle specific analyses that use the GE methodology scram
times will meet all of the same safety limit acceptance criteria.
Additionally, for the non-cycle specific events in the UFSAR, GE has
determined that there is negligible impact on the results of events
which are not analyzed on a cycle specific basis. Therefore, there is
no change to the consequences of a previously evaluated accident or
transient due to the TS changes.
Therefore, the proposed changes to the ITS do not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
1. Administrative Change. The addition of the NRC approved RODEX2A
methodology is an administrative change and will not create the
possibility of a new or different kind of accident. This change does
not affect plant systems, structures, or components. No plant
mitigating systems or functions are affected by this change.
2. Control Rod Scram Insertion Times. The revisions to ITS Section
3.1.4 do not create the possibility of a new or different kind of
accident from any accident previously evaluated. The changes to these
sections revise the control rod scram time requirements for operation.
This changes does not physically alter plant systems, structures, or
components.
Therefore, the proposed changes to the ITS do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Does the proposed change involve a significant reduction in a
margin of safety?
1. Administrative Change. The addition of the NRC approved RODEX2A
methodology is an administrative change and will not reduce the margin
of safety. This change does not affect plant systems, structures, or
components. No plant mitigating systems or functions are affected by
this change.
2. Control Rod Scram Insertion Times. For GE analyzed cores, cycle-
specific analyses using the actual averaged scram times provide MCPR
operating limits that will ensure that MCPR safety limit is not
violated. Therefore, the fuel remains appropriately protected and no
margins of safety are reduced.
Therefore, the proposed changes to the ITS do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Detroit Edison Energy Company, Docket No. 50-341, Fermi 2, Monroe
County, Michigan
Date of amendment request: November 21, 2000.
Description of amendment request: The proposed amendment would
approve a proposed change to the licensing basis regarding the timing
of the release of fission products following an accident. The proposed
change is based upon one of the insights established in NUREG-1465,
``Accident Source Terms for Light Water Nuclear Power Plants,'' which
recognizes that there is a delay in the release of fission products
from the reactor fuel following a postulated design-basis loss-of-
coolant accident (LOCA). The timing of fission product release from
perforated fuel rods (i.e., the gap activity release) is based on the
boiling-water reactor (BWR)-specific value of the timing of the gap
activity release phase of a LOCA as calculated in the BWR Owners Group
(BWROG) Report, ``Prediction of the Onset of Fission Gas Release From
Fuel in Generic BWR,'' NEDC-32963A, dated March 2000, as previously
approved by the NRC staff. This BWROG report would be added (as
Reference 4) to the list of references in Updated Final Safety Analysis
Report (UFSAR) Section 15.6.7. The licensing basis change to UFSAR
Section 15.6.5.5.1, ``Fission Product Release From Fuel,'' would add
the following: ``For primary containment isolation purposes, the
activity from the damaged core is assumed to be released into the
containment at 121 seconds following the accident. This timing
assumption recognizes conclusions derived from the source term studies
described in NUREG-1465, Regulatory Guide 1.183 and Reference 4. * * *
The results of this Table [15.6.5-2, which presents the airborne
activity in the containment] conservatively assume activity released
from the core enters the drywell at accident time zero.'' UFSAR Section
15.6.5.5.2, ``Fission Product Transport to the Environment,'' would be
similarly supplemented to state, ``The results in this Table [15.6.5-3,
which gives the fission product release to the environment due to
containment leakage and leakage from engineered safety feature
components outside containment] conservatively assume activity released
from the core enters the drywell at accident time zero.'' UFSAR Section
15.6.5.5.3, ``Results,'' would be supplemented to state, ``Dose
associated with coolant activity release in the first 121 seconds of
the accident is not included in this Table [15.6.5-4, which presents
the calculated exposures for the design basis analysis]. Its
contribution to the accident dose is insignificant (on the order of 2
rem [to the] thyroid at the Exclusion Area Boundary).''
The effect of the NRC staff's approval of the proposed amendment is
to allow the licensee, in accordance with 10 CFR 50.59, to increase the
automatic closure times for selected primary containment isolation
valves (PCIVs) (i.e., those PCIVs credited for limiting post-accident
doses to both control room personnel and to offsite individuals).
Valves with closure times based on other requirements (i.e., system
performance requirements, equipment qualification, high-energy line
break mitigation, or other regulatory requirements) would not be
affected by the proposed change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 81915]]
consideration, which is presented below:
1. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change takes credit for one of the alternative source
term (AST) insights contained in NUREG-1465 which recognizes that
fission product release from a fuel assembly is not instantaneous in a
design basis accident. Implementation of this change into the licensing
basis will be used to justify an increase in the maximum allowable
closure times for primary containment isolation valves. A change in the
timing of the gap release does not affect the precursors for any
accident or transient previously evaluated as part of the Fermi 2
licensing basis. Therefore, there is no increase in the probability of
any accident.
A plant specific radiological analysis has been performed to
evaluate the effects of extending the maximum allowable valve closure
times on accident dose consequences. This evaluation utilized the
insights contained in NUREG-1465 * * * and NEDC-32963A * * * to justify
no gap activity release during the initial 121 seconds of the accident.
Therefore, during this period, the only releases are from reactor
coolant activity. Assuming the maximum coolant iodine activity
permitted in the Technical Specifications, the 2-hour Exclusion Area
Boundary (EAB) dose associated with this release has been
conservatively estimated to be less than 2 rem thyroid. This dose
represents a small fraction of the LOCA dose evaluated in the UFSAR and
is significantly lower than the 300 rem thyroid dose acceptance limit
in 10 CFR Part 100.
UFSAR Figures 6.2-9 and 6.2-11 show the DBA [design-basis accident]
LOCA primary containment pressure response. These figures indicate that
drywell pressure peaks at around 5 seconds into the accident before
gradually dropping off; therefore, PCIVs would not be required to close
against increased containment pressure as a result of this change.
Utilizing all of the insights contained in NUREG-1465, would result
in a reduction in the calculated dose. However, because this request is
for a selective implementation of the AST scope, crediting only the
timing of the gap activity release, the long term dose calculations
based on TID-14844 in the UFSAR are not changed. Therefore, it is
concluded that the proposed change does not significantly increase the
consequences of a previously evaluated accident.
2. The change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The primary containment isolation system is designed to prevent the
unfiltered release of radioactive material to the environs following an
accident. Therefore, the system is relied upon to mitigate the dose
consequences of an accident. The proposed change recognizes the time
delay before fission products are released into the containment as a
result of fuel damage and allows for the adjustment of the maximum PCIV
closure times accordingly. This change does not affect the function of
the primary containment isolation system. The relaxation in valve
closure times will be applied only to valves that do not have other
system performance requirements on isolation time. Therefore, the
proposed change does not create the potential for a new or different
kind of accident from any accident previously evaluated.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed change revises the Fermi 2 licensing basis for the
offsite dose calculations during the initial 121 seconds of a LOCA
scenario. For this period of time, only coolant activity release is
postulated. No fission product release from perforated fuel rods is
assumed. All other assumptions, bases and methodologies used in the
long-term offsite dose calculations remain unchanged. The total dose
shown in UFSAR Table 15.6.5-4 does not significantly increase due to
the delay in the fission product release. The total amount of
radioactivity remains the same and is bounded by the limits established
in 10 CFR 100. The dose associated with coolant activity release in the
initial 121 seconds of the accident has been determined to be
insignificant. Therefore, the proposed change will not result in a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: Claudia M. Craig.
Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: November 22, 2000.
Description of amendment request: The proposed amendment would
change the pressure-temperature limit curves of Figures 3.6.1, 3.6.2,
and 3.6.3 of Pilgrim's Technical Specifications (TSs) to cover
operation between 20, 32, and 48 Effective Full Power Years. Also
changes to the Bases section consistent with the TS changes are
proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below:
The proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The licensee has proposed to adopt a change in the calculation
methodology for the pressure-temperature limits based upon Code Cases
N-640 and N-588. The code cases were developed using knowledge gained
through years of industry experience. Pressure-temperature curves
developed using the allowances of Code Cases N-640 and N-588 yield more
operating margin. However, the experience gained in the areas of
fracture toughness of materials and pre-existing undetected defects
show that some of the previous assumptions used for the calculation of
pressure-temperature limits are overly conservative. There are no
physical changes to the plant being introduced by the proposed changes
to the pressure-temperature curves. The proposed changes do not modify
the reactor coolant pressure boundary, (i.e., there are no changes in
operating pressure, materials or seismic loading). The proposed changes
do not adversely affect the integrity of the reactor coolant pressure
boundary such that its function in the control of radiological
consequences is affected. Therefore, providing the allowances of the
subject code cases in developing the pressure-temperature limit curves
do not involve a significant increase in the probability or
consequences of an accident previously evaluate. The proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes represent a change in the methodology in how
the
[[Page 81916]]
pressure-temperature curves were generated. The proposed changes
provide more operating margin in the pressure-temperature limit curves
for in-service leakage and hydrostatic pressure testing, non-nuclear
heatup and cooldown, and criticality. However, compliance with the
proposed pressure-temperature curves will ensure conditions in which
brittle fracture of primary coolant pressure boundary materials is
possible will be avoided because such compliance with the proposed
pressure-temperature curves provides sufficient protection against a
non-ductile-type fracture of the reactor pressure vessel. Therefore, no
new modes of operation are introduced nor will the changes create any
failure mode not bounded by the previously evaluated accidents.
Further, the proposed changes to the pressure-temperature curves do not
affect any activities or equipment and are not assumed in any safety
analysis to initiate any accident sequence. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed changes do not involve a significant reduction in a
margin of safety.
The proposed changes reflect an update of the pressure-temperature
curves. The revised curves are based on the latest U.S. Nuclear
Regulatory Commission and American Society of Mechanical Engineers
(ASME) guidance. The revised pressure-temperature limits have been
developed using the Kic fracture toughness curve shown in
the ASME Boiler and Pressure Vessel (B&PV) Code Section XI, Appendix A,
Figure A-2000-1, in lieu of the KIa fracture toughness curve
shown in ASME B&PV Code Section XI, Appendix G, Figure G-2010-1, as the
lower bound fracture toughness. The other margins involved with the
ASME B&PV Code, Section XI, Appendix G process of determining pressure-
temperature limit curves remain unchanged.
These revised pressure-temperature limits, although less
restrictive than the current limits, are established in accordance with
current regulations and the latest ASME Code information. The revised
pressure-temperature curves provide more operating margin and, thus,
more operational flexibility than the current pressure-temperature
curves. However, industry experience since the inception of the
pressure-temperature limits in 1974 confirms that some of the original
methodologies used to develop pressure-temperature curves are overly
conservative. Accordingly, ASME Code Cases N-640 and N-588 take
advantage of the acquired knowledge by establishing more realistic
methodologies for the development of pressure-temperature curves.
Therefore, operational flexibility is gained and an acceptable margin
of safety to reactor pressure vessel non-ductile type fracture is
maintained. No plant safety limits, setpoints, or design parameters are
adversely affected by the proposed changes. Therefore, the proposed
changes do not involve a significant reduction in a margin of safety.
Based on the staff's analysis, it appears that the three standards
of 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599.
NRC Section Chief: James W. Clifford.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: November 30, 2000.
Description of amendment request: The proposed amendment would
relocate the boration systems requirements from the Technical
Specifications (TSs) to the Technical Requirements Manual (TRM).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this proposed
change involve a significant increase in the probability or
consequences of an accident previously evaluated?
The boration systems, BAMT [boric acid makeup tank], Boric Acid
Makeup Pumps, and Charging Pumps, are part of the CVCS [chemical and
volume and control system], which functions to maintain Reactor Coolant
System inventory and chemistry. The boration system functions will
continue to be maintained in accordance with their associated design
requirements. During accident conditions when a boration source is
required for accident mitigation, the RWT [refueling water tank]
provides suction for the High Pressure Safety Injection (HPSI) and Low
Pressure Safety Injection (LPSI) pumps. The CVCS boration systems are
not credited in the mitigation of any accidents. Therefore, the dose
consequences associated with accident analysis will be unchanged. The
HPSI, LPSI pumps and RWT are required by Technical Specifications.
Based on an evaluation of the criterion listed in 10 CFR
50.36(c)(2)(ii), the relocation of the CVCS boration systems to the TRM
is acceptable. No changes will be made to these systems that will
affect their current operation.
Therefore, this change does not involve a significant increase in
the probability of [or] consequences of any accident previously
evaluated.
2. Will operation of the facility in accordance with this proposed
change create the possibility of a new or different kind of accident
from any accident previously evaluated?
The design and functions of the Boric Acid Makeup Tanks, Boric Acid
Makeup Pumps, Charging Pumps and associated flow paths will continue to
be maintained. These systems are not accident initiators. Because the
proposed amendment will not change the design, configuration or method
of operation of the plant, it will not create the possibility of a new
or different kind of accident.
Safety Analysis Report (SAR) Chapter 15 provides the analysis of
accidents that are considered credible. The Uncontrolled Control
Element Assemblies (CEA) withdrawal from a subcritical or a critical
condition, Boration Dilution Event, and Loss of Coolant Accident (LOCA)
were evaluated in relationship to relocating these specifications to
the TRM. Boric acid injection via the CVCS system was not credited in
mitigating any of these accidents.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Will operation of the facility in accordance with this proposed
change involve a significant reduction in a margin of safety?
The movement of these TSs to the TRM does not reduce the existing
TSs or surveillance requirements. The proposed change does not change
the design function for any of these components. Additionally, none of
the boration systems contained in these specifications are credited in
any accident analysis. The systems are used to maintain RCS [reactor
coolant system] chemistry and inventory and this function will be
maintained.
Therefore, this change does not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 81917]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: November 10, 2000.
Description of amendment request: Entergy Operations, Inc. is
proposing that the Grand Gulf Nuclear Station (GGNS) Operating License
be amended to modify those Technical Specifications (TS) required to
support GGNS Cycle 12 operation. The modifications would include a
change to the Safety Limit Minimum Critical Power Ratio (SLMCPR)
reported in TS 2.1.1.2, and the references for analytical methods used
to determine reactor core operating limits listed in TS 5.6.5.
Specifically, the proposed amendment reflects a decrease of the two
recirculation loop SLMCPR limit from 1.09 to 1.08, with the single
recirculation loop SLMCPR limit remaining unchanged at 1.10. The
proposed changes are necessary in order to reflect the Nuclear
Regulatory Commission (NRC) approved methods used in determining the
GGNS Cycle 12 core operating limits, and reflect the safety limit
changes for the Cycle 12 mixed core consisting of Siemens Power
Corporation (SPC) ATRIUM-10 reload fuel and General Electric (GE) GE-11
reactor fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this proposed
change involve a significant increase in the probability or
consequences of an accident previously evaluated?
The Minimum Critical Power Ratio (MCPR) safety limit is defined in
the Bases to Technical Specification 2.1.1 as that limit which
``ensures that during normal operation and during Anticipated
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in the
core do not experience transition boiling.'' The MCPR safety limit
satisfies the requirements of General Design Criterion 10 of Appendix A
to 10 CFR (Part) 50 regarding acceptable fuel design limits. The MCPR
safety limit is re-evaluated for each reload using NRC-approved
methodologies. The analyses for GGNS Cycle 12 have concluded that a
two-loop MCPR safety limit of 1.08, based on the application of Siemens
Power Corporation's NRC-approved MCPR safety limit methodology, will
ensure that this acceptance criterion is met. For single-loop
operation, a MCPR safety limit of 1.10 (unchanged), also ensures that
this acceptance criterion is met.
In addition to the MCPR safety limit, core operating limits are
established to support the Technical Specification 3.2 requirements
which ensure that the fuel design limits are not exceeded during any
conditions of normal operation or in the event of any anticipated
operational occurrences (AOO). The methods used to determine the core
operating limits for each operating cycle are based on methods
previously found acceptable by the NRC and listed in TS section 5.6.5.
A change to TS section 5.6.5 is requested to include the SPC methods in
the list of NRC approved methods applicable to GGNS. These NRC approved
methods will continue to ensure that acceptable operating limits are
established to protect the fuel cladding integrity during normal
operation and in the event of an AOO.
The requested Technical Specification changes do not involve any
plant modifications or operational changes that could affect system
reliability or performance or that could affect the probability of
operator error. The requested changes do not affect any postulated
accident precursors, do not affect any accident mitigating systems, and
do not introduce any new accident initiation mechanisms.
Therefore, these changes to the Minimum Critical Power Ratio (MCPR)
safety limit and to the list of methods used to determine the core
operating limits do not involve a significant increase in the
probability or consequences of any accident previously evaluated.
2. Will operation of the facility in accordance with this proposed
change create the possibility of a new or different kind of accident
from any accident previously evaluated?
The ATRIUM-10 fuel to be used in Cycle 12 is of a design compatible
with the co-resident GE-11. Therefore, the introduction of ATRIUM-10
fuel into the Cycle 12 core will not create the possibility of a new or
different kind of accident. The proposed changes do not involve any new
modes of operation, any changes to setpoints, or any plant
modifications. The proposed revised MCPR safety limits have accounted
for the mixed fuel core and have been shown to be acceptable for Cycle
12 operation. Compliance with the criterion for incipient boiling
transition continues to be ensured. The core operating limits will
continue to be developed using NRC approved methods which also account
for the mixed fuel core design. The proposed MCPR safety limits or
methods for establishing the core operating limits do not result in the
creation of any new precursors to an accident.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Will operation of the facility in accordance with this proposed
change involve a significant reduction in a margin of safety?
The MCPR safety limits have been evaluated in accordance with
Siemens Power Corporation's NRC-approved cycle-specific safety limit
methodology to ensure that during normal operation and during
Anticipated Operational Occurrences (AOO's) at least 99.9% of the fuel
rods in the core are not expected to experience transition boiling. On
this basis, the implementation of this Siemens Power Corporation
methodology does not involve a significant reduction in a margin of
safety.
Therefore, this change does not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: November 8, 2000.
Description of amendment request: The proposed amendment will
delete Technical Specification (TS) 3/4.4.1.6, ``Reactor Coolant Pump-
Startup,'' from the Beaver Valley Power Station (BVPS) TSs. This is
accompanied by moving the secondary side water temperature to
[[Page 81918]]
cold leg temperature difference Reactor Coolant Pump (RCP) start
requirement to existing Reactor Coolant System (RCS) TSs and deleting
the pressurizer level requirement from Unit 1 TS 3/4.4.1.6. Unit 2 TS
3/4.4.1.6 does not contain the pressurizer level requirement. The RCS
TSs affected are TS 3/4.4.1.2, ``Reactor Coolant System--Hot Standby,''
(for Unit 2 only) and 3/4.4.1.3, ``Reactor Coolant System--Shutdown,''
(both units).
Changes to the affected Bases of the Technical Specifications will
also be made.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes will not significantly increase the
probability of an accident previously evaluated in the BVPS Updated
Final Safety Analysis Report (UFSAR) because accident initiation
probabilities are independent of these changes. The proposed changes do
not adversely affect any accident initiating events. The assumptions of
the safety analysis are not changed by this license amendment request.
The applicable concern associated is the possibility of
overpressurizing the Reactor Coolant System (RCS) when a Reactor
Coolant Pump (RCP) is started in a non-isolated loop. Adhering to a
maximum secondary to primary side temperature difference (Technical
Specifications 3/4.4.1.2, Reactor Coolant System--Hot Standby, Unit 2
only, and 3/4.4.1.3, Reactor Coolant System--Shutdown, both units),
before an RCP is started and the operability of the OPPS (Technical
Specification 3/4.4.9.3, Overpressure Protection Systems, for both
units), which uses the PORVs as a pressure relief device, prevents
this. The existing Technical Specifications specify when the OPPS is to
be operable, the maximum secondary to primary side temperature
difference permitted, and the operability requirements imposed on the
PORVs.
The consequences associated with the starting of an RCP and
potential overpressurization of the RCS also are not changed by the
proposed license amendment. None of the accident prevention or
mitigation controls or capabilities have been changed. Reactor Coolant
Pump start restrictions are retained with the Technical Specifications,
except for the pressurizer level requirement for BVPS Unit 1. This
requirement has been shown to be unnecessary in preventing RCS
overpressurization because the analysis assumes a water solid
pressurizer when at least one PORV is operable. The safety analysis has
shown that the temperature difference requirement is sufficient to
preclude RCS overpressurization provided one PORV is available for
pressure relief. As a result, the proposed changes will not affect any
accident analysis consequences.
The Technical Specifications continue to specify the maximum
secondary to primary side temperature difference, when the OPPS is to
be enabled, and the operability requirements for the PORVs. These
requirements are not altered by this license amendment request and will
continue to assure that the OPPS analysis assumptions are met. It is
sufficient to specify the temperature difference restriction for only
Unit 2 Technical Specification 3/4.4.1.2 because the Unit 1 OPPS
enabling temperature is not within the applicability of Technical
Specification 3/4.4.1.2; i.e., Mode 3, whereas the OPPS enabling
temperature is for Unit 2. Therefore, assurance is provided that the 10
CFR 50 Appendix G limits are not exceeded and that this proposed change
is acceptable.
The Bases and editorial changes, needed to meet format requirements
and reflect the deletion of Technical Specification 3/4.4.1.6, have no
effect on accident probabilities or consequences.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not modify the manner in which any plant
equipment is maintained. The equipment used to prevent RCS
overpressurization is not altered by the proposed changes.
Specification of the number of PORVs required to be operable when the
OPPS is enabled, and at what temperature the OPPS is required, will
continue to be retained in Technical Specification 3/4.4.9.3,
Overpressure Protection Systems. The necessary RCP start restrictions
assumed in the safety analysis are not affected by the proposed
changes. It has been shown that deleting the pressurizer level
requirement for Unit 1 is consistent with the OPPS analysis. To assure
the 10 CFR 50 Appendix G limits are not violated, the necessary
requirements for starting an RCP in a non-isolated loop are retained
within the Technical Specifications. Therefore, the analysis of an
overpressurization of the RCS due to a heat input transient caused by
starting an idle RCP is not changed by this license amendment request.
The Bases and editorial changes, needed to meet format requirements
and reflect the deletion of Technical Specification 3/4.4.1.6, will not
affect the creation of accidents. The OPPS analysis has demonstrated
that an RCP can be started with a water solid RCS, provided the
secondary to primary side temperature difference requirement is met,
and a single PORV is available for pressure relief, without violating
10 CFR 50 Appendix G limits.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated for BVPS.
3. Does the change involve a significant reduction in a margin of
safety?
The margin of safety associated with starting an RCP in a non-
isolated loop is the ability of a single OPPS PORV to relieve the
potential RCS pressure increase without violating 10 CFR 50 Appendix G
limits. This is maintained by meeting the secondary side water
temperature to cold leg temperature difference and PORV operability
requirements imposed by the Technical Specifications. These Technical
Specification requirements are not altered by the proposed changes. The
only deletion being proposed is the elimination of the pressurizer
level requirement for BVPS Unit 1. This requirement has been shown to
be unnecessary in meeting 10 CFR 50 Appendix G limits because the OPPS
analysis assumes a water solid pressurizer and at least one OPPS PORV
is operable. Starting an RCP with both OPPS PORVs not operable is not
consistent with the RCS venting actions required by Technical
Specification 3/4.4.9.3. In order to comply with the venting required
actions with neither PORV operable, the RCS must be depressurized or in
the process of being depressurized. Depressurization of the RCS would
preclude starting an RCP. In order to start an RCP, the RCS must be
pressurized to ensure a minimum pressure differential exists across the
No. 1 seal of the RCP. Therefore, the PORV related requirements of Unit
1 Technical Specification 3/4.4.1.6 are sufficiently addressed by
Technical Specification 3/4.4.9.3. By eliminating PORV operability
requirements from Unit 1 Technical Specification 3/4.4.1.6,
[[Page 81919]]
the Technical Specifications become more consistent between the two
units and with the Standard Technical Specifications. All other RCP
start and OPPS requirements are retained within the Technical
Specifications and associated Bases sections.
The Bases and editorial changes, needed to meet format requirements
and reflect the deletion of Technical Specification 3/4.4.1.6, will not
affect the margin of safety. Therefore, the proposed changes do not
involve a significant reduction in a margin of safety regarding meeting
10 CFR 50 Appendix G limits.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Marsha Gamberoni.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: November 9, 2000.
Description of amendment request: The proposed changes would revise
the action statements of the Davis-Besse Nuclear Power Station
Technical Specifications (DBNPS) (TS) Limiting Condition for Operation
(LCO) 3.5.2 and 3.6.2.1. This proposal would extend the allowed outage
time for one Low Pressure Injection (LPI) System/Decay Heat Cooler
train of an Emergency Core Cooling System (ECCS) subsystem from 72
hours to 7 days (168 hours) for LCO 3.5.2. One Containment Spray System
train may be impacted by the inoperability of the associated LPI train.
Therefore, an extension of the allowed outage time for one train of the
Containment Spray System from 72 hours to 7 days for LCO 3.6.2.1 is
also being proposed, as well as new information to be added to TS Bases
Section 3/4.5.2 and 3/4.5.3 to clarify the TS LCO 3.5.2 requirements.
These proposed changes are based on the Babcock & Wilcox Owners Group
(BWOG) Topical report BAW-2295A, Revisions 1 & 2, ``Justification for
the Extension of Allowed Outage Time for Low Pressure Injection and
Reactor Building Spray System,'' dated October 9, 1998.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the
proposed changes and determined that a significant hazards
consideration does not exist because operation of the Davis-Besse
Nuclear Power Station, Unit No. 1, in accordance with these changes
would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because, as demonstrated in the Babcock &
Wilcox Owners Group's Topical Report BAW-2295A, Revisions 1 and 2,
Justification for Extension of Allowed Outage Time for Low Pressure
Injection and Reactor Building Spray Systems, no accident initiators,
conditions, or assumptions are affected by the proposed changes to
extend the allowed outage time (AOT) from 72 hours to 7 days for one
inoperable train of Low Pressure Injection (LPI) in Technical
Specification (TS) 3/4.5.2 Emergency Core Cooling Systems--ECCS
subsystems--Tavg 280 deg.F or Containment Spray
in TS 3/4.6.2.1, Containment Systems--Depressurization and Cooling
Systems--Containment Spray System. The proposed change to TS Bases
Section 3/4.5.2 and 3/4.5.3 are discussions of the present TS Limiting
Condition for Operation (LCO) which do not affect the probability of an
accident.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because an extension in the allowable
outage time from 72 hours to 7 days for one inoperable train will not
affect any previously evaluated accidents. The proposed changes to the
TS Bases discuss the present TS LCO and do not affect the consequences
of an accident. The proposed changes do not alter the source term,
containment isolation, or allowable radiological releases.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new failure
mode or transient is introduced since the proposed changes do not
involve a plant modification or allow operation of any plant systems,
structures, or components in a manner not addressed in the DBNPS Design
Basis Accident analyses.
3. Not involve a significant reduction in a margin of safety
because extending the allowed outage time to 7 days for one inoperable
train does not impact any assumptions or inputs in the DBNPS Updated
Safety Analysis Report. The proposed changes have been evaluated and
determined that the extended allowed outage time is consistent with
safe operation considering the redundant systems of required features
and the administrative controls in place for removing this equipment
from service. The proposed TS Bases changes reflect the existing TS LCO
and, therefore, do not reduce a margin of safety.
On the basis of the above, the DBNPS has determined that the
License Amendment Request does not involve a significant hazards
consideration. As this License Amendment Request concerns a proposed
change to the Technical Specifications that must be reviewed by the
Nuclear Regulatory Commission, this License Amendment Request does not
constitute an unreviewed safety question.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: November 9, 2000.
Description of amendment request: The proposed change would
relocate Technical Specification (TS) 3/4.4.9.2 to the Davis Besse
Nuclear Power Station (DBNPS) Updated Safety Analysis Report (USAR)
Technical Requirements Manual (TRM). A corresponding change to the TS
index is also proposed. Relocation of TS 3/4.4.9.2 to the USAR TRM will
allow future proposed changes to the requirements to be evaluated in
accordance with 10 CFR 50.59 and implemented if prior Nuclear
Regulatory Commission (NRC) approval is not required. The proposed
change is in accordance with the requirements of 10 CFR 50.36 and the
relocation guidance provided in the NRC's ``Final Policy Statement on
TS Improvements for Nuclear Reactors,'' dated July 22, 1993. The
proposed change is also in accordance with the guidance provided by the
improved ``Standard Technical
[[Page 81920]]
Specifications--Babcock & Wilcox Plants,'' NUREG-1430, Revision 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the
proposed changes and determined that a significant hazards
consideration does not exist because operation of the Davis-Besse
Nuclear Power Station, Unit No. 1, in accordance with these changes
would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no change is being made to any
accident initiator. No previously analyzed accident scenario is
changed, and initiating conditions and assumptions remain as previously
analyzed.
The proposed change would relocate TS 3/4.4.9.2 ``Reactor Coolant
System--Pressurizer,'' to the DBNPS Updated Safety Analysis Report
(USAR) Technical Requirements Manual (TRM). TS 3/4.4.9.2 provides
temperature limits for the Pressurizer based on its fatigue analysis
design criteria. The proposed change to remove this TS is in accordance
with 10 CFR 50.36 and the NRC's ``Final Policy Statement on TS
Improvements for Nuclear Power Reactors,'' dated July 22, 1993. The
proposed change is also consistent with the improved ``Standard
Technical Specifications--Babcock and Wilcox Plants,'' NUREG-1430,
Revision 1. A corresponding change to the TS Index page V that removes
reference to the Pressurizer Pressure/Temperature Limits is an
administrative change.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed change does not
affect accident conditions or assumptions used in evaluating the
radiological consequences of an accident. The proposed change does not
alter the source term, containment isolation or allowable radiological
releases.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new failure
mode is introduced since the proposed relocation does not involve a
modification or change in operation of any plant systems, structures,
or components. No new, or different types of failures or accident
initiators are introduced by the proposed change.
3. Not involve a significant reduction in a margin of safety
because the proposed change is administrative in nature, consisting of
the relocation of certain TS requirements into a licensee-controlled
document, and has no bearing on the margin of safety which exists in
the present TS or Updated Safety Analysis Report (USAR).
On the basis of the above, the DBNPS has determined that the
License Amendment Request does not involve a significant hazards
consideration. As this License Amendment Request concerns a proposed
change to the Technical Specifications that must be reviewed by the
Nuclear Regulatory Commission, this License Amendment Request does not
constitute an unreviewed safety question.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of amendment request: October 30, 2000.
Description of amendment request: The proposed amendment would
revise the St. Lucie Unit 1 Technical Specification (TS) 3.9.4,
Containment Penetrations. TS 3.9.4 requires a personnel airlock (PAL)
door to be closed during core alterations or movement of irradiated
fuel within containment. The proposed change would allow both
containment PAL doors to be open during core alterations and movement
of irradiated fuel in containment provided: (a) that at least one
personnel airlock door is capable of being closed; (b) the plant is in
MODE 6 with at least 23 feet of water above the fuel; and (c) a
designated individual is available outside the PAL to close the door.
Operability of the containment PAL door includes the requirements that
the door is capable of being closed and that any cables or hoses across
the PAL door have quick-disconnects to ensure the door is capable of
being closed in a timely manner.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed change to TS 3.9.4 would allow the containment
personnel airlock (PAL) doors to be open during fuel movement or core
alterations. Currently, a single PAL door is closed during fuel
movement or core alterations to prevent the escape of radioactive
material in the event of an in-containment fuel handling accident. The
PAL is not an initiator of an accident. Whether the PAL doors are open
or closed during fuel movement and core alterations has no affect on
the probability of any accident previously evaluated.
Allowing the PAL doors to be open during fuel movement or core
alterations does not significantly increase the consequences from a
fuel handling accident. The calculated offsite doses are well within
the limits of 10 CFR Part 100. In addition, the calculated doses are
larger than the expected doses because the calculation does not
incorporate the closing of the PAL doors after the containment is
evacuated. The proposed change should significantly reduce the dose to
workers in containment in the event of a fuel handling accident by
reducing the time required to evacuate the containment.
The changes being proposed do not affect assumptions contained in
plant safety analyses or the physical design of the plant, nor do they
affect other Technical Specifications that preserve safety analysis
assumptions. Therefore, operation of the facility in accordance with
the proposed amendments would not involve a significant increase in the
probability or consequences of an accident previously analyzed.
2. Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed change to Technical Specification 3.9.4, Containment
Penetrations, affects a previously evaluated fuel handling accident.
Both the current and the reanalyzed fuel handling accident analysis
assume that all of the iodine and noble gases that become airborne
within the containment escape and reach the site boundary and low
population zone with no credit taken for filtration, the containment
building barrier, or for decay or deposition taken. Since the
[[Page 81921]]
proposed change does not involve the addition or modification of
equipment, nor does it alter the design of plant systems and the
revised analysis is consistent with the fuel handling accident
analysis, the proposed change does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The margin of safety as defined by 10 CFR Part 100 has not been
reduced. The calculated dose is a well within of the limits given in 10
CFR Part 100 or NUREG-0800. The proposed changes do not alter the bases
for assurance that safety-related activities are performed correctly or
the basis for any Technical Specification that is related to the
establishment of or maintenance of a safety margin. Therefore,
operation of the facility in accordance with the proposed amendments
would not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: November 27, 2000.
Description of amendment request: The proposed amendments delete
requirements from the Technical Specifications (and, as applicable,
other elements of the licensing bases) to maintain a Post Accident
Sampling System (PASS). Licensees were generally required to implement
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the technical specifications (TS) for nuclear power
reactors currently licensed to operate. Lessons learned and
improvements implemented over the last 20 years have shown that the
information obtained from PASS can be readily obtained through other
means or is of little use in the assessment and mitigation of accident
conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated November 27, 2000.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result of
the TMI-2 accident. The specific intent of the PASS was to provide a
system that has the capability to obtain and analyze samples of plant
fluids containing potentially high levels of radioactivity, without
exceeding plant personnel radiation exposure limits. Analytical results
of these samples would be used largely for verification purposes in
aiding the plant staff in assessing the extent of core damage and
subsequent offsite radiological dose projections. The system was not
intended to and does not serve a function for preventing accidents and
its elimination would not affect the probability of accidents
previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual benefit
to post accident mitigation. Past experience has indicated that there
exists in-plant instrumentation and methodologies available in lieu of
a PASS for collecting and assimilating information needed to assess
core damage following an accident. Furthermore, the implementation of
Severe Accident Management Guidance (SAMG) emphasizes accident
management strategies based on in-plant instruments. These strategies
provide guidance to the plant staff for mitigation and recovery from a
severe accident. Based on current severe accident management strategies
and guidelines, it is determined that the PASS provides little benefit
to the plant staff in coping with an accident.
The regulatory requirements for the PASS can be eliminated without
degrading the plant emergency response. The emergency response, in this
sense, refers to the methodologies used in ascertaining the condition
of the reactor core, mitigating the consequences of an accident,
assessing and projecting offsite releases of radioactivity, and
establishing protective action recommendations to be communicated to
offsite authorities. The elimination of the PASS will not prevent an
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency
plan (EP), the emergency operating procedures (EOP), and site survey
monitoring that support modification of emergency plan protective
action recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
The elimination of PASS related requirements will not result in any
failure mode not previously analyzed. The PASS was intended to allow
for verification of the extent of reactor core damage and also to
provide an input to offsite dose projection calculations. The PASS is
not considered an accident precursor, nor does its existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within the
containment building.
Therefore, this change does not create the possibility of a new or
different kind
[[Page 81922]]
of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a neutral
impact to the margin of safety. Methodologies that are not reliant on
PASS are designed to provide rapid assessment of current reactor core
conditions and the direction of degradation while effectively
responding to the event in order to mitigate the consequences of the
accident. The use of a PASS is redundant and does not provide quick
recognition of core events or rapid response to events in progress. The
intent of the requirements established as a result of the TMI-2
accident can be adequately met without reliance on a PASS.
Therefore, this change does not involve a significant reduction in
the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: November 28, 2000.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Table 6.2.1, Minimum Shift Crew
Composition with Two Separate Control Rooms and TS Section 6.3.1 (2),
Unit Staff Qualifications for the Shift Technical Advisor (STA). The
proposed amendments would permit, as an alternative to the current
dedicated STA, an on-shift senior reactor operator (SRO) position to be
combined with the required STA position. The proposed amendments would
require an individual filling either the dedicated STA position or the
combined SRO/STA position to meet the Technical Specifications
educational requirements as described in Federal Register Notice 50 FR
43621, ``Commission Policy Statement on Engineering Expertise on
Shift.'' These proposed changes are in accordance with the
recommendations in the NRC Policy Statement on Engineering Expertise on
Shift, published on October 28, 1985 and transmitted to all power
reactor licensees and applicants by NRC Generic Letter 86-04, of the
same title as the October 28, 1985 policy statement, dated February 13,
1986. As permitted by the policy statement, FPL proposes to exercise
either of the STA options on a shift-by-shift basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Implementation of the proposed changes will not involve any
physical changes to plant systems, structures, or components (SSC), or
the manner in which these SSCs are operated, maintained, modified,
tested, or inspected. Therefore, the proposed use of either the dual
role SRO/STA position or the current dedicated STA position does not
increase the probability of an accident previously evaluated.
Implementation of the proposed changes will result in personnel with
enhanced operational knowledge being assigned to perform the STA
function of providing accident assessment expertise, and analyzing and
responding to off normal occurrences when needed.
The NRC stated preference in the October 28, 1985, Policy Statement
on Engineering Expertise on Shift, indicates that the NRC has concluded
that the individual filling the dual role SRO/STA position may perform
these functions better than a non-licensed individual filling the STA
position, even when the SRO/STA is concurrently functioning as one of
the required shift SROs. Therefore, the proposed TS changes do not
increase the consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed amendments will not change the physical plant or the
modes of plant operation defined in the facility license for either St.
Lucie unit. Changes proposed for the administrative controls do not
involve the addition or modification of equipment, nor do they alter
the design or operation of plant systems. Therefore, operation of
either facility in accordance with its proposed amendments would not
create the possibility of a new or different kind of accident from any
accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed amendments revise certain administrative controls
involving the on-site programmatic process for review and approval of
plant procedures. Neither the scope, nor the requirement to establish,
maintain, and implement procedures for activities that could affect
nuclear safety are being changed.
The NRC stated preference in the October 28, 1985, Policy Statement
on Engineering Expertise on Shift, indicates that the NRC has concluded
that the individual filling the dual role SRO/STA position may perform
these functions better than a non-licensed individual filling the STA
position, even when the SRO/STA is concurrently functioning as one of
the required shift SROs. Therefore, the proposed amendments should
involve an enhancement in a margin on safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: October 30, 2000.
Description of amendment request: The proposed amendments would
revise Technical Specification 5.3.2 for Turkey Point Units 3 and 4 to
extend the residual heat removal (RHR) pump allowed outage time (AOT)
from 72 hours to 7 days to restore an inoperable
[[Page 81923]]
RHR pump to operable status. The proposed extension is based on the
projected time required to replace a leaking or failed pump shaft seal,
perform post-maintenance testing, and complete any additional
corrective actions that may be needed to restore the pump to operable
status. The extended RHR pump AOT will provide adequate time so that
future seal repair activities are completed successfully in a safe
manner.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The RHR system is part of the Emergency Core Cooling System.
Inoperable RHR pumps are not accident initiators in any accident
previously evaluated, and an extended AOT to restore operability of an
inoperable RHR pump would not increase the probability of occurrence of
accidents previously analyzed. Therefore, this change does not involve
an increase in the probability of an accident previously evaluated.
The RHR system is primarily designed to mitigate the consequences
of the large Loss Of Coolant Accident (LOCA). In addition, the RHR
system provides for primary system heat removal during unit shutdown
conditions. The proposed changes do not affect any of the assumptions
relative to accident initiators or accident response provided in the
plant safety analyses. Accordingly, the consequences of accidents
previously evaluated do not change.
A Probabilistic Safety Assessment (PSA) was performed to evaluate
the impact of extending the allowed outage time on the RHR pump from 72
hours to 7 days. FPL concluded from the results of that assessment that
the risk contribution of the AOT extension is very small, and that the
net impact of the proposed amendment may be risk neutral.
Therefore, the change does not involve a significant increase in
the probability or consequences of any accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different kind
of accident from any previously evaluated.
The proposed change does not alter the design, physical
configuration, or modes of operation of the plant. Plant configurations
that are prohibited by Technical Specifications will not be created by
the AOT extension. Therefore, the proposed activity does not create the
possibility of a new or different kind of accident from any previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The margin of safety associated with the Emergency Core Cooling
System is established by acceptance criteria for system performance
defined in 10 CFR 50.46. The proposed amendments will not change these
acceptance criteria or the operability requirements for equipment that
is used to achieve such performance as demonstrated in the plant safety
analyses. Moreover, a Probabilistic Safety Assessment of the risk
impact of extending the AOT for a single inoperable RHR pump has
concluded that the risk contribution is very small, RHR system
reliability can potentially be improved, and the net impact of the
proposed change may be risk neutral. Therefore, the change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: December 6, 2000.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (and, as applicable,
other elements of the licensing bases) to maintain a Post Accident
Sampling System (PASS). Licensees were generally required to implement
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the technical specifications for nuclear power reactors
currently licensed to operate. Lessons learned and improvements
implemented over the last 20 years have shown that the information
obtained from PASS can be readily obtained through other means or is of
little use in the assessment and mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated December 6, 2000.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result of
the TMI-2 accident. The specific intent of the PASS was to provide a
system that has the capability to obtain and analyze samples of plant
fluids containing potentially high levels of radioactivity, without
exceeding plant personnel radiation exposure limits. Analytical results
of these samples would be used largely for verification purposes in
aiding the plant staff in assessing the extent of core damage and
subsequent offsite radiological dose projections. The system was not
intended to and does not serve a function for preventing accidents and
its elimination would not
[[Page 81924]]
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual benefit
to post accident mitigation. Past experience has indicated that there
exists in-plant instrumentation and methodologies available in lieu of
a PASS for collecting and assimilating information needed to assess
core damage following an accident. Furthermore, the implementation of
Severe Accident Management Guidance (SAMG) emphasizes accident
management strategies based on in-plant instruments. These strategies
provide guidance to the plant staff for mitigation and recovery from a
severe accident. Based on current severe accident management strategies
and guidelines, it is determined that the PASS provides little benefit
to the plant staff in coping with an accident.
The regulatory requirements for the PASS can be eliminated without
degrading the plant emergency response. The emergency response, in this
sense, refers to the methodologies used in ascertaining the condition
of the reactor core, mitigating the consequences of an accident,
assessing and projecting offsite releases of radioactivity, and
establishing protective action recommendations to be communicated to
offsite authorities. The elimination of the PASS will not prevent an
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency
plan (EP), the emergency operating procedures (EOP), and site survey
monitoring that support modification of emergency plan protective
action recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated.
The elimination of PASS related requirements will not result in any
failure mode not previously analyzed. The PASS was intended to allow
for verification of the extent of reactor core damage and also to
provide an input to offsite dose projection calculations. The PASS is
not considered an accident precursor, nor does its existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within the
containment building.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a neutral
impact to the margin of safety. Methodologies that are not reliant on
PASS are designed to provide rapid assessment of current reactor core
conditions and the direction of degradation while effectively
responding to the event in order to mitigate the consequences of the
accident. The use of a PASS is redundant and does not provide quick
recognition of core events or rapid response to events in progress. The
intent of the requirements established as a result of the TMI-2
accident can be adequately met without reliance on a PASS.
Therefore, this change does not involve a significant reduction in
the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: November 15, 2000.
Description of amendment requests: The proposed amendments would
revise the Technical Specification (TS) 3.2.6, ``Allowable Power
Level--APL,'' and TS 1.38, ``Allowable Power Level (APL),'' definitions
of APL to remove a condition that limits APL to 100 percent of rated
thermal power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
No new accident initiators or precursors are created by the
proposed T/S changes. Reactor thermal power and power distribution
within the reactor core are not initiators or precursors to any
previously evaluated accident. There are no physical changes to the
plant associated with the proposed T/S changes that would create any
new accident initiators or precursors. Therefore, the proposed T/S
changes do not increase the probability of occurrence of any accident
previously evaluated.
Reactor thermal power up to the calculated value of APL ensures
that the accident analysis results are not impacted by maintaining
reactor core power distribution within prescribed limits. Since T/S 1.3
still contains a limitation on the maximum reactor thermal power
allowed during normal operations, the normal overall operating limits
for the reactor core are not changed. Accident analyses generally
include a calorimetric error allowance of 2% or assume an initial power
level of at least 102%. Using the additional limit on reactor thermal
power based on APL ensures operation within the power distribution
limits assumed in the accident analyses. Therefore, the proposed T/S
changes do not affect operation of the reactor core and do not modify
either the maximum acceptable reactor thermal power or the maximum
allowed power distribution limits.
The proposed T/S changes do not change or alter the design criteria
for the systems or components used to mitigate the consequences of any
design basis accident. The reactor protection system (RPS), including
reactor trips based upon overall reactor thermal power and power
distribution within the reactor core, are not affected by the proposed
T/S changes. The initial conditions of the accident analyses, including
maximum reactor thermal power and worst-case power distribution within
the reactor core, are not changed. As a result, the expected operation
of the emergency core cooling systems (ECCS) are not affected by the
proposed T/S changes. Radiological consequences of previously evaluated
accidents are not increased, since overall reactor thermal power and
power distribution limits are still maintained within the assumptions
of the accident analyses, and operation of
[[Page 81925]]
the RPS and ECCS is not affected. Therefore, the proposed changes do
not increase the consequences of any accident and do not impact offsite
dose considerations.
Therefore, the probability of occurrence or the consequences of
accidents previously evaluated are not increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Reactor thermal power and power distribution within the reactor
core cannot be an initiator or precursor to an accident. There are no
physical changes to the plant associated with the proposed T/S changes
that would create any new accident initiators or precursors. The
proposed T/S changes do not degrade the reliability of any existing
system, structure, or component. No new failure modes, malfunctions, or
system interactions are created. The maximum steady state reactor core
power level as defined by T/S 1.3 is not changed. The actual power
distribution limits are not changed since the calculated value of APL
is not changed. Therefore, the accident analyses assumptions and
results are unchanged.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
The proposed T/S changes do not change either the overall maximum
reactor thermal power allowed, or the reactor core power distribution
limits allowed. Maximum reactor thermal power remains limited by T/S
1.3. The calculated value of APL in T/S 3.2.6 is not changed, and
remains as a control to ensure reactor core power distribution limits
consistent with the accident analyses are satisfied. Therefore, safety
margins related to power distribution limits are not affected. The
proposed T/S changes do not affect any of the T/S safety limits or T/S
limiting safety system settings, and RPS setpoints as defined by the T/
S are not changed or affected.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: Claudia M. Craig.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: November 28, 2000.
Description of amendment request: The proposed amendment would
establish technical specifications (TSs) for the emergency service
water system. It would also revise TS 3.0 to include general
requirements for system operability.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed amendment will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The EFT-ESW [emergency filtration train-emergency service water]
System is not an accident initiator. The proposed amendment provides
operability requirements and surveillance requirements to ensure the
ESW System is available and operable when required for accident
mitigation. The proposed operability requirements and allowed outage
times are consistent with similar requirements for the systems
supported by the EFT-ESW System. Dose to the public and the Control
Room operators are not affected by the proposed change. The proposed
general LCO [limiting condition for operation] provides direction with
respect to actions to be taken when support systems are inoperable.
The proposed Technical Specification change does not introduce new
equipment operating modes, nor does the proposed change alter existing
system relationships. The proposed amendment does not introduce new
failure modes.
Therefore, the proposed amendment will not significantly increase
the probability or the consequences of an accident previously
evaluated.
The proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously analyzed.
The proposed Technical Specification change does not introduce new
equipment operating modes, nor does the proposed change alter existing
system relationships. The proposed amendment does not introduce new
failure modes. The proposed amendment does not alter the equipment
required for accident mitigation and considers the effects on supported
systems when a support system is inoperable. When support systems are
inoperable, actions are specified to be taken consistent with safe
plant operation.
Therefore, the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed amendment will not involve a significant reduction in
the margin of safety.
The proposed amendment provides specifications for the EFT-ESW
System which are consistent with current Technical Specification
requirements for other equipment. The proposed changes ensure that the
EFT-ESW and other support systems will be available when required and
provides adequate alternative actions when the support systems are not
available. The allowed outage times for the EFT-ESW Pumps are
consistent with that allowed for other equipment that would have
similar importance to accident mitigation. The proposed general LCO
does not result in a significant reduction in the margin of safety
since it imposes requirements already in technical specifications for
support systems included in technical specifications. In cases where
support systems [are] not included in technical specifications, the
proposed general LCO does not apply and actions determined to be
required by the technical specifications will be taken for the
supported systems.
Therefore, the proposed amendment will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: September 5, 2000.
Description of amendment request: The proposed amendment would
revise the Fort Calhoun Station, Unit No. 1 (FCS) Technical
Specifications (TS) to change the definition section, TS
[[Page 81926]]
Sections 2.10, 3.10, and 5.9, and the Bases of TS 1.1 and 1.3, to allow
the use of nuclear fuel fabricated by Siemens Power Corporation at FCS.
The definition of unrodded planar radial peaking factor
(Fxy) and TS 2.10.4(3) are being deleted and TS 3.10 is
being revised to reflect the deletion of this peaking factor. TS 5.9.5
is being revised to incorporate NRC-approved methodologies necessary to
determine core operating limits with nuclear fuel from Siemens Power
Corporation. The Bases to TS 1.1 and 1.3 are being revised to delete
the discussion of the CE-1 correlation that is currently used to
calculate minimum departure from nucleate boiling ratio and the value
calculated by this method.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed amendment is to incorporate Siemens Power Corporation
topical reports for conducting reload analyses that have been
previously reviewed and approved by the NRC. The applicable FCS
Technical Specifications (TS) supported by these topical reports are
being revised. These changes are necessary to support using nuclear
fuel supplied by Siemens Power Corporation.
It is proposed to revise the Bases of TS 1.1 and 1.3 to reflect
changes in methodologies for calculating the minimum Departure from
Nucleate Boiling Ration (DNBR). The proposed methodology for
determining the minimum DNBR for fuel supplied by Siemens Power
Corporation is the NRC-approved EMF-92-153(P)(A) and Supplement 1, HTP:
Departure from Nucleate Boiling Correlation for High Thermal
Performance Fuel. As stated in the Basis of TS 1.1, Fort Calhoun
Station currently uses the NRC-approved CE-1 correlation with a minimum
DNBR value of 1.18, which provides a 95% probability at a 95%
confidence level that DNB will not occur for any operating condition.
For Siemens fuel, using the HTP correlation with a minimum DNBR of
1.14, as proposed, will continue to provide a 95% probability at a 95%
confidence level that DNB will not occur during any operating
condition. The CE-1 correlation is more restrictive than the HTP
correlation that will be used to predict the minimum DNBR limits for
the Siemens fuel. For a given set of reactor coolant conditions, the
CE-1 correlation provides a lower critical heat flux than the HTP
correlation. Therefore, this change will not significantly increase the
probability or consequences of an accident previously evaluated.
It is proposed that the total planar radial peaking factor,
FxyT, be eliminated from the Technical
Specifications. The current need for this parameter is to protect
assumptions about the maximum amount of planar peaking in the core. The
limitation on the total planar radial peaking factor,
FxyT, is provided to ensure that the assumptions
used in the analysis for establishing the Linear Heat Rate and Local
Power Density--High, Limiting Conditions for Operation, and Limiting
Safety Systems Settings set-points remain valid during operation. In a
two-dimensional set-point analysis, as currently conducted,
FxyT is combined with the maximum axial power
profile (Fz) to produce the maximum allowable peaking factor
(Fq) or equivalent Linear Heat Rate. This ensures
conservative operation relative to assumptions on linear heat rate used
as input to the loss of coolant accident and other transient analyses.
In a three-dimensional analysis, as proposed with the use of Siemens
methodology, these peaks are calculated directly during a series of
pre-determined maneuvers (axial shape oscillation, power maneuver, or
other transient).
Direct calculation of these peaks negates the need to make
inferences about the amount of planar radial peaking that occurs in any
particular plane within the core. Therefore, this change will not
significantly increase the probability or consequences of an accident
previously evaluated.
It is proposed to add NRC-approved methodologies from Siemens Power
Corporation to TS that are necessary to evaluate core parameters. The
proposed additions of NRC-approved topical reports to the TS do not
modify the manner in which the topical reports may be implemented. The
core operating limits will continue to be determined using NRC-approved
analytical methods. The plant will continue to operate within the
limits specified by the Core Operating Limits Report and will take
corrective actions as required by the current Technical Specifications
should these limits be exceeded. Therefore, these changes will not
significantly increase the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
No new or different modes of operation are proposed as a result of
these changes. The proposed revisions do not change any equipment
required to mitigate the consequences of an accident. The proposed
additions of NRC-approved topical reports to the TS do not modify the
manner in which the topical reports may be implemented. The plant will
continue to operate within the limits specified by the Core Operating
Limits Report and will take corrective actions as required should these
limits be exceeded. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
As required by TS 5.9.5, the analytical methods used to determine
the core operating limits shall be those previously reviewed and
approved by the NRC. The proposed changes incorporate methodologies
applicable for use with fuel supplied by Siemens Power Corporation that
have been approved by the NRC as documented by Safety Evaluation
Reports. Technical Specification 5.9.5 also requires that the core
operating limits shall be determined so that all applicable limits of
the safety analysis are met. These requirements will continue to be
met. Therefore, OPPD concludes that the proposed changes do not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: October 18, 2000.
Description of amendment request: The proposed amendment would
revise the Fort Calhoun Station Unit 1 (FCS) Technical Specifications
(TSs) to (1) extend the validity of the existing TS Figure 2-1A (RCS
[reactor coolant system] Pressure-Temperature Limits for Heatup) and
Figure 2-1B (RCS Pressure-Temperature Limits for Cooldown) from
[[Page 81927]]
20.0 effective full power years (EFPY) to 24.25 EFPY, (2) delete Figure
2-3 (Predicted Radiation Induced NDTT [nil ductility transition
temperature] Shift), and (3) provide replacement guidance in TSs
2.1.2(6)(a) and (b) for use of the most current fluence analysis and
Regulatory Guide 1.99, Revision 2, ``Radiation Embrittlement of Reactor
Vessel Materials,'' for projecting reference temperature nil ductility
(RTNDT) at 24.25 EFPY. The proposed amendment would also
revise the associated Bases section of TS 2.1.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The NRC previously approved Technical Specification Amendment No.
161 in March 1994 for the use of RCS Pressure-Temperature (P-T) Limits
good to 20.0 EFPY. The proposed changes in this submittal reflect the
validity of these same curves from 20.0 EFPY to 24.25 EFPY based on the
implementation of extreme low radial leakage fuel management in 1992
(Cycle 14). Significant reductions in the fast neutron flux to the
limiting 3-410 axial weld in the Fort Calhoun Station reactor pressure
vessel were obtained, thus significantly increasing the time to when
the fast neutron fluence input to the derivation of the previously
approved P-T curves will be reached. Since no inputs (including assumed
material properties of the limiting weld) to the existing analysis are
being changed, extension of the validity of the curves from 20.0 EFPY
to 24.25 EFPY is justified. In addition, deletion of Figure 2-3 and
references to it are proposed. This proposed change removes an outdated
figure which is non-operational in nature. The application of the
current Regulatory Guide 1.99, Revision 2 is more appropriate for these
purposes. Administrative changes to the Basis section of TS 2.1.2 are
proposed to reflect the extension to 24.25 EFPY.
No accidents previously analyzed are affected by these changes, and
it can be concluded that there is no significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not physically alter the configuration of
the plant and no new or different mode of operation is proposed.
Extending the validity of the P-T curves more accurately projects
reactor vessel embrittlement by accounting for improvements in FCS fuel
management which have significantly reduced the fast neutron fluence to
the limiting 3-410 axial weld, incorporates improved operating cycle
efficiency, and applies the WCAP-15443, Revision 0 fluence analysis.
The revised fluence analysis uses the ENDF/B-VI Nuclear Cross Section
Library. Deletion of Figure 2-3 represents a change which does not
affect plant operations. Figure 2-3 is administrative in nature, and
proposed revisions to Specifications 2.1.2(6)(a) and (b) provide
guidance consistent with the current Regulatory Guide for P-T curves
updates. Update of the Technical Specification 2.1.2 Basis section
represents an administrative change that does not affect plant
operation.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously analyzed.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes to extend the validity of Technical
Specification Figures 2-1A and 2-1B to 24.25 EFPY are consistent with
the extreme low radial leakage fuel management implemented in 1992
(Cycle 14) and performance/application of the updated fluence analysis
described above. With no changes to the inputs of the existing P-T
limits analysis, there is no reduction in the margin of safety. Figure
2-3 is not used to provide limits on plant operation, and deletion of
this figure, which uses a pre-Regulatory Guide 1.99, Revision 2
embrittlement correlation, is considered an improvement in the
consistency of the requirements outlined in the Technical
Specifications. This Figure is not used in plant operation and provides
only a general indication of the RTNDT shift. The TS 2.1.2
Basis section changes are administrative in nature and do not affect
the margin of safety. The changes serve to maintain consistency with
the NRC approval of Amendment No. 161.
In summary, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 27, 2000.
Description of amendment request: The proposed amendment would
revise Section 3.7 of the Fort Calhoun Station Unit 1 Technical
Specifications to eliminate item 3.7(4) ``13.8 Kv Transmission Line''
which states: ``The 13.8 Kv transmission line will be energized and
loaded to minimum shutdown requirements at each refueling outage
following installation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Eliminating the 13.8 kV testing requirement would have no impact
upon the probability of an accident previously evaluated. The circuit
breaker connecting the 13.8 kV power supply to the station electrical
busses is normally open, so this power supply could not play a role in
the initiation of any accident.
Eliminating the 13.8 kV testing requirement would have no impact
upon the consequences of an accident previously evaluated. Existing
accident analyses take no credit for the 13.8 kV power supply.
The 13.8 kV power supply is not credited for mitigation of
licensing basis transients or postulated events added to the USAR
[Updated Safety Analysis Report] by NRC requirements, such as Station
Blackout (SB0).
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The 13.8 kV power supply is only capable of supplying a limited
number of components in the unlikely event that 161 kV, 345 kV, and the
diesel-generators are unavailable. Eliminating the 13.8 kV testing
requirement would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
[[Page 81928]]
3. The proposed change does not involve a significant reduction in
a margin of safety.
Testing of the 13.8 kV power supply, as described in Technical
Specification 3.7(4), is unrelated to any margin of safety. Therefore,
deletion of the testing requirement will not reduce any margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: November 30, 2000.
Description of amendment requests: The proposed license amendments
would change Technical Specification Section 3.5.1, ``Accumulators,''
by revising the limits for accumulator borated water volume
(Surveillance Requirement (SR) 3.5.1.2) and nitrogen cover pressure (SR
3.5.1.3) to reflect analysis limits. These TS currently reflect nominal
limits. These amendments are revising TS values consistent with other
similar TS parameters which will aid in future clarity.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The accumulators only function following an accident. They cannot
initiate an accident. The proposed changes have no impact to plant
operation and are administrative in nature. Changing the technical
specification (TS) limits for accumulator volume and pressure from
nominal to analysis values will provide greater consistency within the
TS. Changing the volume limits to cubic feet verse[u]s percent level
will eliminate any potential for future revision of these limits
because of instrument tap relocation.
Plant parameters will continue to be administratively controlled
within the allowed analysis parameters. The proposed limits for tank
volume and nitrogen cover pressure are consistent with analysis values
documented in the Final Safety Analysis Report and assume that the
accident consequences remain unchanged.
There are no hardware changes or changes in the method by which any
safety-related plant system performs its safety function.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The accumulators only function following an accident. They cannot
initiate an accident. The proposed changes are administrative in
nature.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. The changes are administrative in nature so there are no new
accident scenarios, transient precursors, failure mechanisms, or
limiting single failures are [sic] introduced.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes are administrative in nature.
The proposed changes do not affect the acceptance criteria for any
analyzed event. There will be no effect on the manner in which safety
limits or limiting safety system settings are determined nor will there
be any effect on those plant systems necessary to assure the
accomplishment of protection functions.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: November 30, 2000.
Description of amendment requests: The proposed license amendments
would change the administrative controls sections of Technical
Specification (TS) 5.5.14b and 5.5.14b.2 to incorporate the changes
made to 10 CFR Part 50, Section 50.59. The proposed amendments would
replace the word ``involve'' with ``require'' in TS 5.5.14b and revise
TS 5.5.14b.2 to delete the reference to ``unreviewed safety question''
and restate the requirement as ``a change to the updated Final Safety
Analysis Report or Bases that requires NRC approval pursuant to 10 CFR
50.59.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change replaces the word ``involve'' with ``require''
and deletes reference to the term ``unreviewed safety question''
consistent with 10 CFR [Part 50, Section] 50.59. Deletion of the term
``unreviewed safety question'' was approved by the NRC with the
revision to 10 CFR 50.59. Consequently, the probability of an accident
previously evaluated is not significantly increased. Changes to the
Technical Specification (TS) Bases are still evaluated in accordance
with 10 CFR 50.59. As a result, the consequences of any accident
previously evaluated are not significantly affected.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or a
change in the methods governing plant operation. These changes are
considered administrative changes and do not modify, add, delete,
[[Page 81929]]
or relocate any technical requirements in the TS.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes will not reduce the margin of safety because
they have no effect on any safety analyses assumptions. Changes to the
TS Bases that result in meeting the criteria in paragraph (c)(2) of 10
CFR 50.59 will still require NRC approval. The proposed changes to TS
5.5.14 are considered administrative in nature based on the revision to
10 CFR 50.59.
Therefore, the proposed changes do not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: October 25, 2000.
Description of amendment request: The amendment revises the Action
Statements associated with Technical Specifications (TSs) Table
3.3.7.5-1 (``Accident Monitoring Instrumentation'') concerning the
Drywell Hydrogen/Oxygen (H2/O2) Concentration
Analyzers, and the associated TS Bases. PECO Energy proposes to add new
Action Statements 82a and 82b concerning channel operability, which
will replace the current requirements of Action Statements 80a and 80b,
respectively, for the Drywell Hydrogen/Oxygen Concentration Analyzers.
Under the existing TS Action Statements for Table 3.3.7.5-1
(``Accident Monitoring Instrumentation''), with the number of operable
accident monitoring instrumentation channels less than the ``required''
number of channels (quantity 2), restore the inoperable channels within
7 days or be in at least hot shutdown within the following 12 hours
(Action Statement 80a). Additionally, with the number of operable
accident monitoring instrumentation channels less than the ``minimum''
number of channels (quantity 1), restore the inoperable channel(s)
within 48 hours or be in at least hot shutdown within the following 12
hours (Action Statement 80b).
Proposed Action Statement 82a for Table 3.3.7.5-1 will extend the
duration from 7 to 30 days for less than the ``required'' number
operable of channels. Additionally, the proposed Action Statement 82a
will require that if the operable channel(s) cannot be restored within
the 30 days, then a Special Report shall be provided to the NRC within
the following 14 days.
Proposed Action 82b for Table 3.3.7.5-1 will extend the duration
from 48 hours to 72 hours for less than the ``minimum'' number of
operable channels. If the inoperable channel(s) cannot be restored with
the 72 hours, then be in hot shutdown with the next 12 hours.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c).
1. The proposed [technical specification] TS changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed TS changes modify the Action Statements associated
with the duration that the Drywell Hydrogen/Oxygen Concentration
Analyzers can be inoperable. The Drywell Hydrogen/Oxygen Concentration
Analyzers are not accident initiating equipment and are monitoring
devices required to be available for monitoring hydrogen and oxygen
following a LOCA. These analyzers do not perform any automatic or
control functions. Therefore, the proposed changes will not increase
the probability of an accident previously evaluated.
In the event of a failure of the Drywell Hydrogen/Oxygen
Concentration Analyzers following a LOCA, concentrations of hydrogen
and oxygen can be measured by utilizing grab samples with the post-
accident sampling system. A single failure of either analyzer package
would render that affected package inoperable with the redundant
package fully capable of performing the required function at full
capacity. Following a postulated LOCA, the hydrogen recombiners will be
utilized to ensure that the oxygen concentration in the primary
containment is maintained below the lower flammability limit as
required by plan emergency procedures.
The extended completion times are based on the passive nature of
the instrument (no critical automatic action is assumed to occur from
these instruments), the low probability of an event requiring post-
accident instrumentation during this interval, and the availability of
alternate means to obtain the required information. Therefore, the
proposed TS changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed technical specification changes modify the Action
Statements associated with the duration that the Drywell Hydrogen/
Oxygen Concentration Analyzers can be inoperable. They do not change
the design or configuration of the plant. The Drywell Hydrogen/Oxygen
Concentration Analyzers are not accident initiating equipment, and are
monitoring devices required to be available for monitoring hydrogen and
oxygen following a LOCA. The proposed changes do not create a system-
level failure mode different than those that already exist. In
addition, there are no operation or failure modes of the Drywell
Hydrogen/Oxygen Concentration Analyzers that are accident initiators.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed TS changes do not involve a significant reduction
in the margin of safety.
The proposed changes in Action Statements do not affect any safety
limits or analytical limits. There are also no changes to accident of
transient core thermal hydraulic conditions, minimum combustible
concentration limits, or fuel or reactor coolant boundary design
limits, as a result of these proposed changes. The proposed Technical
Specification changes modify the Action Statements associated with the
duration that the Drywell Hydrogen/Oxygen Concentration Analyzers can
be inoperable. Therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
[[Page 81930]]
Attorney for licensee: J.W. Durham, Sr., Esquire, Senior V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: November 29, 2000.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to reflect the enabling of the
Oscillation Power Range Monitor (OPRM) instrumentation reactor
protection system (RPS) trip function. The OPRM is designed to detect
the onset of reactor core power oscillations resulting from thermal-
hydraulic instability and suppresses them by initiating a reactor scram
via the RPS trip logic.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change specifies limiting conditions for operations,
required actions and surveillance requirements of the OPRM system and
allows operation in regions of the power to flow map currently
restricted by the requirements of Interim Corrective Actions (ICAs) and
certain limiting conditions of operation of Technical Specifications
(TS) 3.4.1. The OPRM system can automatically detect and suppress
conditions necessary for thermal-hydraulic (T-H) instability. A T-H
instability event has the potential to challenge the Minimum Critical
Power Ratio (MCPR) safety limit. The restrictions of the ICAs and TS
3.4.1 were imposed to ensure adequate capability to detect and suppress
conditions consistent with the onset of T-H oscillations that may
develop into a T-H instability event. With the installation of the OPRM
System, these restrictions are no longer required.
The probability of a T-H instability event is most significantly
impacted by power to flow conditions such that only during operation
inside specific regions of the power to flow map, in combination with
power shape and inlet enthalpy conditions, can the occurrence of an
instability event be postulated to occur. Operation in these regions
may increase the probability that operation with conditions necessary
for a T-H instability can occur.
However, when the OPRM is operable with operating limits as
specified in the COLR [Core Operating Limits Report], the OPRM can
automatically detect the imminent onset of local power oscillations and
generate a trip signal. Actuation of an RPS trip will suppress
conditions necessary for T-H instability and decrease the probability
of a T-H instability event. In the event the trip capability of the
OPRM is not maintained, the proposed change includes actions which
limit the period of time before the effected OPRM channel (or RPS
system) must be placed in the trip condition. If these actions would
result in a trip function, an alternate method to detect and suppress
thermal hydraulic oscillations is required. In either case the duration
of this period of time is limited such that the increase in the
probability of a T-H instability event is not significant. Therefore
the proposed change does not result in a significant increase in the
probability of an accident previously evaluated.
An unmitigated T-H instability event is postulated to cause a
violation of the MCPR safety limit. The proposed change ensures
mitigation of T-H instability events prior to challenging the MCPR
safety limit if initiated from anticipated conditions by detection of
the onset of oscillations and actuation of an RPS trip signal. The OPRM
also provides the capability of an RPS trip being generated for T-H
instability events initiated from unanticipated but postulated
conditions. These mitigating capabilities of the OPRM system would
become available as a result of the proposed change and have the
potential to reduce the consequences of anticipated and postulated T-H
instability events. Therefore, the proposed change does not
significantly increase the consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change specifies limiting conditions for operations,
required actions and surveillance requirements of the OPRM system and
allows operation in regions of the power to flow map currently
restricted by the requirements of ICAs and TS 3.4.1. The OPRM system
uses input signals shared with APRM and rod block functions to monitor
core conditions and generate an RPS trip when required. Quality
requirements for software design, testing, implementation and module
self-testing of the OPRM system provide assurance that no new equipment
malfunctions due to software errors are created. The design of the OPRM
system also ensures that neither operation nor malfunction of the OPRM
system will adversely impact the operation of other systems and no
accident or equipment malfunction of these other systems could cause
the OPRM system to malfunction or cause a different kind of accident.
Therefore, operation with the OPRM system does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Operation in regions currently restricted by the requirements of
ICAs and TS 3.4.1 is within the nominal operating domain and ranges of
plant systems and components for which postulated equipment and
accidents have been evaluated. Therefore operation within these regions
does not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change which specifies limiting conditions for
operations, required actions and surveillance requirements of the OPRM
system and allows operation in certain regions of the power to flow map
does not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed change specifies limiting conditions for operations,
required actions and surveillance requirements of the OPRM system and
allows operation in regions of the power to flow map currently
restricted by the requirements of ICAs and TS 3.4.1.
The OPRM system monitors small groups of LPRM signals for
indication of local variations of core power consistent with T-H
oscillations and generates an RPS trip when conditions consistent with
the onset of oscillations are detected. An unmitigated T-H instability
event has the potential to result in a challenge to the MCPR safety
limit. The OPRM system provides the capability to automatically detect
and suppress conditions which might result in a T-H instability event
and thereby maintains the margin of safety by providing automatic
protection for the MCPR safety limit while significantly reducing the
burden on the control room operators. In the event the trip capability
of the OPRM is not maintained, the proposed change includes actions
which limit the period of time before the effected OPRM channel (or RPS
system) must be placed in the trip condition. If these actions
[[Page 81931]]
would result in a trip function, an alternate method to detect and
suppress thermal hydraulic oscillations is required. Since, in either
case, the duration of this period of time is limited so that the
increase in the probability of a T-H instability event is not
significant. Operation with the OPRM system does not involve a
significant reduction in a margin of safety.
Operation in regions currently restricted by the requirements of
ICAs and TS 3.4.1 is within the nominal operating domain assumed for
identifying the range of initial conditions considered in the analysis
of anticipated operational occurrences and postulated accidents.
Therefore, operation in these regions does not involve a significant
reduction in the margin of safety.
The proposed change, which specifies limiting conditions for
operations, required actions and surveillance requirements of the OPRM
system and allows operation in certain regions of the power to flow
map, does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: November 21, 2000 (ULNRC-04346)
Description of amendment request: The proposed amendment request
would change Table 3.3.2-1, ``Engineered Safety Feature Actuation
System Instrumentation,'' of the Technical Specifications. The change
would add Surveillance Requirement (SR) 3.3.2.10 to the SRs for the
following two engineered safety feature actuation system (ESFAS)
instrumentation in the table: item f, loss of offsite power, and item
h, auxiliary feedwater pump suction transfer on suction pressure--low.
The licensee also identified that there would be changes to the Final
Safety Analysis Report (FSAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Overall protection system performance will remain within the bounds
of the previously performed accident analyses since there are no
hardware changes. The Reactor Trip System (RTS) and Engineered Safety
Feature Actuation System (ESFAS) instrumentation will be unaffected.
These protection systems will continue to function in a manner
consistent with the plant design basis. All design, material, and
construction standards that were applicable prior to the request are
maintained.
The proposed change imposes more stringent surveillance testing
requirements to ensure safety-related structures, systems, and
components are tested in a manner consistent with the safety analysis
and licensing basis.
The proposed change will not affect the probability of any event
initiators. There will be no degradation in the performance of, or an
increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident situation. There will
be no change to normal plant operating parameters or accident
mitigation performance.
The proposed change will not alter any assumptions or change any
mitigation actions in the radiological consequence evaluations in the
FSAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. This change will not affect the normal method of plant
operation or change any operating parameters. No performance
requirements will be affected; however, the proposed change does impose
additional surveillance testing requirements. These additional
requirements are consistent with assumptions made in the safety
analysis and licensing basis.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result of
this change. There will be no adverse effect or challenges imposed on
any safety-related system as a result of this change.
This change does not alter the design or performance of the 7300
Process Protection System, Nuclear Instrumentation System, or Solid
State Protection System used in the plant protection systems.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment of
protection functions. There will be no impact on the overpower limit,
departure from nucleate boiling ratio (DNBR) limits, heat flux hot
channel factor (FQ), nuclear enthalpy rise hot channel
factor (FdeltaH), loss of coolant accident peak cladding temperature
(LOCA PCT), peak local power density, or any other margin of safety.
The radiological dose consequence acceptance criteria listed in the
[NRC] Standard Review Plan [(NUREG-0800)] will continue to be met.
The imposition of more stringent surveillance requirements [in the
change] increase the margin of safety by ensuring that the affected
safety analysis assumptions on equipment response time are verified on
a periodic frequency.
Therefore, the proposed change does not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: November 22, 2000.
Description of amendment request: The proposed change to Callaway
Technical Specification (TS) 5.5.14, which ensures that a program
exists for processing changes to the TS Bases, would replace the word
``involve'' with ``require'' and deletes the phrase ``unreviewed safety
question'' as
[[Page 81932]]
defined in 10 CFR Part 50, Section 50.59.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes replace the word ``involve'' with ``require''
and deletes the phrase ``unreviewed safety question'' as defined in 10
CFR 50.59. The above changes are consistent with the revision to 10 CFR
50.59. Consequently, the probability of an accident previously
evaluated is not significantly increased. Changes to the Technical
Specification Bases are still evaluated in accordance with 10 CFR
50.59. As a result, the consequences of any accident previously
evaluated are not affected.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or a
change in the methods governing plant operation. These changes are
considered administrative changes and do not modify, add, delete, or
relocate any technical requirements in the TS.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes will not reduce the margin of safety because
they have no effect on any safety analyses assumptions. Changes to the
TS Bases that result in meeting the criteria in paragraph (c)(2) of 10
CFR 50.59 will still require NRC approval. The proposed changes to TS
5.5.14 are considered administrative in nature based on the revisions
to 10 CFR 50.59.
Therefore, the proposed changes do not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 7, 2000 (ET 00-0041).
Description of amendment request: The proposed amendment request
would change Table 3.3.2-1, ``Engineered Safety Feature Actuation
System Instrumentation,'' of the Technical Specifications (TSs). The
change would add Surveillance Requirement (SR) 3.3.2.10 to the SRs for
the following two engineered safety feature actuation system (ESFAS)
instrumentation in the table: item 6.f, loss of offsite power, and item
6.h, auxiliary feedwater pump suction transfer on suction pressure--
low. The licensee also identified that there would be changes to the
Updated Safety Analysis Report (USAR) and changes to the Bases for the
TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Overall protection system performance will remain within the bounds
of the previously performed accident analyses since there are no
hardware changes. The Reactor Trip System (RTS) and Engineered Safety
Feature Actuation System (ESFAS) instrumentation will be unaffected.
These protection systems will continue to function in a manner
consistent with the plant design basis. All design, material, and
construction standards that were applicable prior to the request are
maintained.
The proposed change imposes more stringent surveillance testing
requirements to ensure safety related structures, systems, and
components are tested in a manner consistent with the safety analysis
and licensing basis.
The proposed change will not affect the probability of any event
initiators. There will be no degradation in the performance of, or an
increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident situation. There will
be no change to normal plant operating parameters or accident
mitigation performance.
The proposed change will not alter any assumptions or change any
mitigation actions in the radiological consequence evaluations in the
USAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs its safety
function. This change will not affect the normal method of plant
operation or change any operating parameters. No performance
requirements will be affected; however, the proposed change does impose
additional surveillance testing requirements. These additional
requirements are consistent with assumptions made in the safety
analysis and licensing basis.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result of
this change. There will be no adverse effect or challenges imposed on
any safety related system as a result of this change.
This change does not alter the design or performance of the 7300
Process Protection System, Nuclear Instrumentation System, or Solid
State Protection System used in the plant protection systems.
Therefore, the proposed changes do not create a new or different
kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment of
protection functions. There will be no impact on the overpower limit,
departure from nucleate boiling ratio (DNBR) limits, heat flux hot
channel factor (FQ), nuclear enthalpy rise hot channel
factor (FdeltaH), loss of coolant accident peak cladding temperature
(LOCA PCT), peak local power density, or any other margin of safety.
The radiological dose consequence acceptance criteria listed in the
[NRC] Standard Review Plan
[[Page 81933]]
[(NUREG-0800)] will continue to be met.
The imposition of more stringent surveillance testing requirements
[in the change] increases the margin of safety by ensuring that the
affected safety analysis assumptions on equipment response time are
verified on a periodic frequency.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Northeast Nuclear Energy Company,
et al., Docket No. 50-423, Millstone Nuclear Power Station, Unit No. 3,
New London County, Connecticut
Date of application for amendment: March 19, 1999, and supplemented
by letters dated April 17, May 5, June 16, July 26, and November 21,
2000.
Brief description of amendment: The amendment changes Technical
Specification (TS) 1.40, ``Spent Fuel Pool Storage Pattern''; 1.41,
``3-OUT-OF-4 AND 4-OUT-OF-4''; 3/4.9.1.2, ``Boron Concentration''; 3/
4.9.7, ``Crane Travel-Spent Fuel Storage Areas''; 3/4.9.13, ``Spent
Fuel Pool--Reactivity''; 3.9.14, ``Spent Fuel Pool--Storage Pattern'';
5.6.1.1, ``Design Features--Criticality''; and 5.6.3, ``Design
Features--Capacity.'' In addition, the amendment revises INDEX pages
xii and xv for new figures and page numbers and replaces Figures 3.9-1
and 3.9-2 with four new figures and make changes to the TS Bases
consistent with changes to their respective TS sections.
Date of issuance: November 28, 2000.
Amendment No.: 189.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of individual notice in Federal Register: December 4, 2000 (65
FR 75736).
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: November 22, 1999, as
supplemented November 24, 1999 and September 12, 2000.
Brief description of amendments: The amendments revise Technical
Specification 5.5.11, ``Ventilation Filter Testing Program'' for
laboratory testing of charcoal in engineered safety feature ventilation
systems to reference American Society for Testing and Materials D3803-
1989 ``Standard Test Method for Nuclear-Grade Activated Carbon.''
Date of issuance: December 7, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 238 and 212.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73085)
The November 24, 1999, and September 12, 2000, submittals provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated December 7, 2000.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: September 15, 2000.
Brief description of amendments: The amendments implement Technical
Specification Task Force (TSTF)-134, Revision 1. TSTF-134 revises
Technical Specification Surveillance Requirements (SR) 3.1.7.2 which
verifies control element assembly (CEA) trip function from 50 percent
withdrawn position, by adding a note allowing SR 3.1.7.2 not be
performed if TS SR 3.1.4.6 (CEA drop time test) has been met. TSTF-134,
Revision 1, was approved by the Nuclear Regulatory Commission on April
21, 1998.
Date of issuance: December 11, 2000.
[[Page 81934]]
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment Nos.: 239 and 213.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 18, 2000 (65 FR
62384).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated December 11, 2000.
No significant hazards consideration comments received: No.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: May 5, 1999, as supplemented on
December 22, 1999, and September 18, 2000.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.5, ``Instrumentation Systems,'' for the reactor
protection system and engineered safety features actuation system
instrumentation. Specifically, the amendment: (1) Revises the allowed
outage times for the instrumentation, (2) allows on-line testing and
maintenance of instrumentation, and (3) revises the associated Bases
section. The amendment also includes several editorial changes to TS
Tables 3.5-2 and 3.5-3.
Date of issuance: November 30, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 212.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 4, 2000 (65 FR
59221).
The December 22, 1999, and September 18, 2000, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated November 30, 2000.
No significant hazards consideration comments received: No.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: August 22, 2000, as supplemented
on October 3 and 15, 2000.
Brief description of amendment: The amendment revises: (1)
Technical Specification (TS) 3.10.4, ``Rod Insertion Limits,'' to allow
on-line calibration of the rod position indicator (RPI) channels during
operating cycle 15, and (2) TS 3.10.6, ``Inoperable Rod Position
Indicator Channels,'' to allow extended RPI deviation limits during
cycle 15.
Date of issuance: December 12, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 213.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 20, 2000 (65
FR 56948).
The October 3 and 15, 2000, letters provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 12, 2000.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: February 21, 2000.
Brief description of amendment: This amendment deleted references
to stainless steel as the material for reactor coolant system and
reactor coolant pressure boundary component fasteners from Table 1.8-1
and 1.8-2 of the Beaver Valley Power Station, Unit No. 1, Updated Final
Safety Analysis Report (UFSAR).
Date of issuance: December 4, 2000.
Effective date: As of date of issuance.
Amendment No.: 235.
Facility Operating License No. DPR-66: Amendment authorized changes
to the UFSAR.
Date of initial notice in Federal Register: June 14, 2000 (65 FR
37426).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 4, 2000.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of application for amendment: July 19, 2000.
Brief description of amendment: The amendment revises the Technical
Specifications (TS) surveillance requirements of the safety-related
ventilation system charcoal consistent with the actions requested in
Generic Letter 99-02, ``Laboratory Testing of Nuclear-Grade Activated
Charcoal,'' dated June 3, 1999. Systems impacted include the control
room emergency ventilation system, the shield building ventilation
system, the emergency core cooling system area ventilation system, and
the fuel pool ventilation system--fuel storage.
Date of Issuance: December 7, 2000.
Effective Date: December 7, 2000.
Amendment No.: 167.
Facility Operating License No. NPF-16: Amendment revised the TS.
Date of initial notice in Federal Register: August 9, 2000 (65 FR
48749).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 7, 2000.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: June 21, 2000.
Brief description of amendments: These amendments relocate
Technical Specification (TS) Surveillance Requirement 4.8.1.1.2.e.1,
regarding the emergency diesel generator (EDG) inspection program, to a
licensee controlled maintenance program that will be incorporated by
reference into the next revision of the Updated Final Safety Analysis
Report for each St. Lucie unit. Upon relocation to the licensee
controlled maintenance program, the effectiveness of the maintenance on
the EDGs and support systems will be monitored pursuant to the
Maintenance Rule 10 CFR 50.65.
Date of Issuance: December 7, 2000.
Effective Date: December 7, 2000.
Amendment Nos.: 168 and 111.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the TS.
Date of initial notice in Federal Register: August 9, 2000 (65 FR
48750).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 7, 2000.
No significant hazards consideration comments received: No.
[[Page 81935]]
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: November 30, 1999, as
supplemented June 28, 2000, and November 3, 2000.
Brief description of amendment: The amendment revises Technical
Specifications Sections 3.7.2, ``Control Room Envelope Filtration
(CREF) System,'' and 5.5.7, ``Ventilation Filter Testing Program
(VFTP)'' for laboratory testing of charcoal filters to reference
American Society for Testing and Materials standard D3803-1989,
``Standard Test Method for Nuclear-Grade Activated Carbon.''
Date of issuance: December 1, 2000.
Effective date: As of the date of issuance to be implemented within
30 days of issuance.
Amendment No.: 95.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51358).
The November 3, 2000, submittal did not change the initial proposed
no significant hazards consideration determination.
The staff's related evaluation of the amendment is contained in a
Safety Evaluation dated December 4, 2000.
No significant hazards consideration comments received: No.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: May 11 and May 12, 2000, as
supplemented by letters dated June 13, June 16, July 14, September 21,
October 26, and November 3, 2000.
Brief description of amendment: The amendment grants a conforming
amendment to the License and the Technical Specifications for the
approval of the transfer of the license for the Indian Point Nuclear
Generating Unit No. 3 (IP3) held by the Power Authority of the State of
New York to Entergy Nuclear IP3, LLC. to possess and use IP3 and to
Entergy Nuclear Operations, Inc. (ENO) to possess, use and operate IP3.
Date of issuance: November 21, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 203.
Facility Operating License No. DPR-64: Amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: June 28, 2000 (65 FR
39954).
The supplemental information did not expand the scope of the
application as originally noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 9, 2000.
No significant hazards consideration comments received: No.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: May 11 and May 12, 2000, as
supplemented by letters dated June 13, June 16, July 14, September 21,
October 26, and November 3, 2000.
Brief description of amendment: The amendment grants a conforming
amendment to the License and the Technical Specifications for the
approval of the transfer of the license for the James A. FitzPatrick
Nuclear Power Plant (FitzPatrick) held by the Power Authority of the
State of New York to Entergy Nuclear FitzPatrick, LLC. to possess and
use FitzPatrick and to Entergy Nuclear Operations, Inc. (ENO) to
possess, use and operate FitzPatrick.
Date of issuance: November 21, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 268.
Facility Operating License No. DPR-59: Amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: June 28, 2000 (65 FR
39953).
The supplemental information did not expand the scope of the
application as originally noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 9, 2000.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: February 7, 2000, as
supplemented on August 9 and October 12, 2000.
Brief description of amendments: The amendments modify the Salem
Unit Nos. 1 and 2 Technical Specifications (TS), and revise
surveillance requirements associated with Auxiliary Feedwater (AFW)
Pump testing described in TS 4.7.1.2.b by replacing the current wording
with that of improved Standard TSs, NUREG-1431, ``Standard Technical
Specifications, Westinghouse Plants.''
Date of issuance: December 5, 2000.
Effective date: As of the date of issuance, and shall be
implemented within 60 days of issuance.
Amendment Nos.: 238 and 219.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 14, 2000 (65 FR
37428).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 5, 2000.
No significant hazards consideration comments received: No.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: March 8, 2000, as supplemented
April 5, 2000, and October 25, 2000.
Brief description of amendment: The amendment revises the Technical
Specifications through revision to the storage configuration
requirements within the existing storage racks and taking credit for a
limited amount of soluble boron.
Date of issuance: December 7, 2000.
Effective date: December 7, 2000.
Amendment No.: 79.
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 5, 2000 (65 FR
17918).
The April 5, 2000, and October 25, 2000, submittals provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 7, 2000.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar
Nuclear Plant Units 1 and 2, Rhea County, Tennessee
Date of application for amendment: March 10, 2000, as supplemented
November 6 and 9, 2000 and November 21, 2000 (two letters).
Brief description of amendment: Changed the Operating License to
incorporate Physical Security/Contingency Plan--Tamper Indicating/Line
Supervision Alarms Testing Frequency at Watts Bar Nuclear Plant (WBN)
Units 1 and 2.
[[Page 81936]]
Date of issuance: December 5, 2000.
Effective date: December 5, 2000.
Amendment No.: 29 and 29.
Facility Operating License No. NPF-90: Amendment revises the
Operating License.
Date of initial notice in Federal Register: September 20, 2000 (65
FR 56957). The November 6, 9, and 21, 2000, supplements provided
clarifying information that did not change the scope of the initial
proposed no significant hazards consideration determination.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: October 30, 2000, as
supplemented November 15 and 22, 2000.
Brief description of amendment: Allow a one-time-only increase in
the diesel generator Action Completion Time from 72 hours to 10 days to
facilitate repairs to an emergency diesel generator to improve
reliability.
Date of issuance: December 8, 2000.
Effective date: December 8, 2000.
Amendment No.: 30.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 3, 2000 (65 FR
66266). The November 15 and 22, 2000 supplements provided clarifying
information that did not change the scope of the initial proposed no
significant hazards consideration determimination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 8, 2000.
No significant hazards consideration comments received: No.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: August 10, 2000.
Brief description of amendments: The amendments change Technical
Specification (TS) 5.6.5, ``Core Operating Limits Report,'' to
incorporate the latest, Nuclear Regulatory Commission (NRC)-approved
methodology for analysis of large break loss-of-coolant accidents
(LBLOCAs) for Comanche Peak Steam Electric Station, Units 1 and 2. The
acceptability of this change to TS 5.6.5 is based upon the NRC staff's
conclusion that the LBLOCA analysis methodology described in TXU
Electric's Topical Report ERX-2000-002-P, ``Revised Large Break Loss of
Coolant Accident Methodology,'' March 2000, is acceptable, as addressed
in the associated Safety Evaluation.
Date of issuance: October 6, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 80 and 80.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51363).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 6, 2000.
No significant hazards consideration comments received: No.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: May 2, 2000, as supplemented August 30,
2000.
Brief description of amendments: The amendments change the CPSES
Security Plan to: (1) Allow response team members to perform
compensatory measures for protective area intrusion detection or closed
circuit television failure, and (2) to modify the patrol frequency for
the protected area.
Date of issuance: December 5, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 82 and 82.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Security Plan.
Date of initial notice in Federal Register: October 4, 2000 (65 FR
59226).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 5, 2000.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 21st day of December 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 00-33012 Filed 12-26-00; 8:45 am]
BILLING CODE 7590-01-P