[Federal Register Volume 65, Number 173 (Wednesday, September 6, 2000)]
[Notices]
[Pages 54083-54094]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-22779]
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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 August 14, 2000, through August 25, 2000.
The last biweekly notice was published on August 23, 2000 (65 FR
51346).
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 October 6, 2000, 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 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
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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).
Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: August 10, 2000.
Description of amendment request: The requested amendment proposes
to change the Technical Specifications for operations involving
positive reactivity addition. The proposed changes revise the Required
Actions and Limiting Condition for Operation (LCO) Notes to limit the
introduction of reactivity such that the required SHUTDOWN MARGIN (SDM)
or refueling boron concentration will remain satisfied.
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:
Carolina Power & Light (CP&L) Company has evaluated the proposed
Technical Specifications change and has concluded that it does not
involve a significant hazards consideration. The CP&L conclusion is
in accordance with the criteria set forth in 10 CFR 50.92. The bases
for the conclusion that the proposed change does not involve a
significant hazards consideration are discussed below.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures or components. The proposed change revises
ACTIONS in the H. B. Robinson Steam Electric Plant (HBRSEP) Unit No.
2 Technical Specifications (TS) that require suspending operations
involving positive reactivity additions and several Limiting
Condition For Operation (LCO) Notes that preclude reduction in boron
concentration. The change revises these ACTIONS and LCO Notes to
limit the introduction of reactivity such that the required SHUTDOWN
MARGIN (SDM) or refueling boron concentration will still be
satisfied. The proposed change ensures that the SDM of LCO 3.1.1 and
minimum boron concentration requirements of LCO 3.9.1 are met.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated in the Safety Analysis Report (SAR) because the
accident analysis assumptions and initial conditions will continue
to be maintained.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures or components. The proposed change, which
allows positive reactivity additions that do not result in SDM or
the refueling boron concentration being exceeded, does not introduce
new failure mechanisms for systems, structures or components not
already considered in the SAR [Safety Analysis Report]. Therefore,
the possibility of a new or different kind of accident from any
accident previously evaluated is not created because no new failure
mechanisms or initiating events have been introduced.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change will allow positive reactivity additions,
but the reactivity additions will not result in a[n] SDM or
refueling boron concentration outside of the associated design basis
limits. Allowing positive reactivity additions that do not result in
the SDM or the refueling boron concentration being exceeded will not
significantly reduce 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: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602
NRC Section Chief: Richard P. Correia
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: June 19, 2000
Description of amendment request: The proposed amendment would
revise the technical specifications to remove their applicability
related to the Boron Dilution Protection System (BDPS) after the next
refueling outage for each unit. During the refueling outages,
modifications are scheduled to be made which will permit the licensee
to mitigate a boron dilution event without the use of the BDPS.
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 proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The only accident potentially impacted by the proposed changes
is the inadvertent boron dilution event.
The Boron Dilution Protection System (BDPS) is not considered an
initiator of any analyzed event. The BDPS performs detection and
mitigative functions for the
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inadvertent boron dilution event. Therefore, the proposed changes
have no impact on the probability of an event previously analyzed.
Therefore, the proposed change does not involve a significant
increase in the probability of occurrence of an accident previously
evaluated.
The proposed changes impact the consequences of an inadvertent
dilution event due to the new requirement to manually reposition the
Chemical and Volume Control System (CVCS) valves that isolate the
boron dilution sources and that re-start boration of the Reactor
Coolant System (RCS) in Modes 3, 4, and 5 (i.e., Hot Standby, Hot
Shutdown, and Cold Shutdown, respectively). The revised detection
and mitigation methodology being proposed achieves the same basic
function as the existing BDPS, i.e., to prevent a return to critical
during an inadvertent boron dilution event. The proposed changes
will provide an improved response to the inadvertent boron dilution
event compared to the BDPS, and thereby will prevent a return to
critical. Therefore, the proposed changes do not involve a
significant increase in the consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes to manually isolate potential dilution
sources and to re-start boration of the RCS do not create the
potential for a new or different kind of accident because the change
results in plant configurations that have always been allowed. In
conjunction with these proposed changes, enhancements to plant
hardware, revisions to procedures, and administrative controls will
be implemented. The proposed enhancements to plant hardware include
the addition of two new redundant Volume Control Tank (VCT) high
level alarms, which are passive in nature (i.e., do not provide any
control function), and therefore do not create the possibility of a
new or different kind of accident. Administrative controls and
revisions to procedures will increase the operator's awareness of a
potential boron dilution event and will provide the steps necessary
to respond to a boron dilution event. As a result, the
administrative controls and revisions to procedures do not create
the possibility of a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The design criterion and margin of safety for the existing BDPS
is that the inadvertent boron dilution event is terminated within a
specified period prior to the complete loss of shutdown margin. This
criterion will continue to be satisfied following implementation of
the proposed changes. The proposed changes were evaluated to ensure
that the plant operators prevent criticality in Modes 3, 4 and 5
following an inadvertent boron dilution event, based on the revised
analytical methodology previously discussed with the NRC and found
to be feasible as documented in a letter from L. R. Wharton (U.S.
NRC) to Licensees (Commonwealth Edison, Texas Utilities Electric,
Union Electric, Wolf Creek Nuclear Operating Corporation, and
Westinghouse), ``Utility Subgroup Technical Approach to Modify or
Delete the Boron Dilution Mitigation System,'' dated February 8,
1993. The proposed method of detecting and mitigating this event has
been shown by the analysis supporting this Technical Specifications
change request to prevent a return to critical following an
inadvertent boron dilution event, and meets the same NRC acceptance
criteria as specified in the Standard Review Plan (SRP), NUREG-0800,
Section 15.4.6, ``Chemical and Volume Control System Malfunction
That Results in a Decrease in Boron Concentration in the Reactor
Coolant (PWR),'' dated July 1981, as applicable to the existing
BDPS. 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
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
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: June 1, 2000
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.6.16 Reactor Building; and TS 5.5.11
Ventilation Filter Testing Program. It will also revise Bases Sections
3.6.10, 3.6.16, 3.7.12, and 3.7.13. The amendments will: (1) Enhance
the ability to determine that reactor building annulus outside air
inleakage is within the maximum assumed design value used in the dose
analyses. Administrative limits are currently imposed at Catawba to
limit inleakage in order to ensure that the dose analyses remain
conservative. The amendments also request changes for the Unit 2
Annulus Ventilation System (AVS) in-place penetration and bypass
leakage criteria in TS 5.5.11. This portion of the amendments affects
TS Bases 3.6.10, TS 3.6.16 and Bases, and TS 5.5.11; (2) Describe the
alignment the Auxiliary Building Filtered Ventilation Exhaust System
(ABFVES) filtered exhaust units should be tested in and request
appropriate TS 5.5.11 limits in order to ensure that the ABFVES will
continue to meet its design basis functions. Similar to Item 1 above,
the amendments also request changes for the Unit 2 ABFVES in-place
penetration and bypass leakage criteria in TS 5.5.11. This portion of
the amendments affects TS Bases 3.7.12 and TS 5.5.11; and (3) Modify
the TS Bases for the Fuel Handling Ventilation Exhaust System (FHVES)
and similar to Items 1 and 2 above, the amendments also request changes
for the Unit 2 FHVES in-place penetration and bypass leakage criteria
in TS 5.5.11. This portion of the amendments affects TS Bases 3.7.13
and TS 5.5.11.
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 following discussion is a summary of the evaluation of the
changes contained in this proposed amendment against the 10 CFR
50.92(c) requirements to demonstrate that all three standards are
satisfied. A no significant hazards consideration is indicated if
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.
First Standard
Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Neither the AVS, nor the ABFVES, nor the FHVES
is capable of initiating any accident. The AVS, ABFVES, and FHVES,
which are responsible for maintaining an acceptable environment in
the annulus, the auxiliary building, and the fuel building during
normal and accident conditions, will continue to function as
designed, and in accordance with all applicable TS. The design and
operation of the systems are not being modified by this proposed
amendment. Therefore, there will be no impact on any accident
probabilities or consequences.
Second Standard
Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of NRC approval of this amendment request. No changes
are being made to the plant which will introduce any new accident
causal mechanisms. This amendment request does not impact any plant
systems that are accident initiators and does not impact any safety
analyses.
[[Page 54086]]
Third Standard
Implementation of this amendment would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The performance of these fission
product barriers will not be impacted by implementation of this
proposed amendment. The performance of the AVS, the ABFVES, and the
FHVES in response to normal and accident conditions will not be
impacted by this proposed amendment. The changes to the AVS
surveillances will provide for a better method to ensure that the
assumptions of the dose analyses are met. There is no risk
significance to this proposed amendment, as no reduction in system
or component availability will be incurred. No safety margins will
be impacted.
Based upon the preceding discussion, Duke has concluded that the
proposed amendment does not involve a 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
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Richard L. Emch, Jr.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: August 10, 2000.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to allow an alternate storage
configuration of fuel assemblies adjacent to the walls within Region 1
of the spent fuel pool.
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:
Criterion 1--Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The probability of fuel handling accidents (dropped assemblies,
misplaced/misloaded assemblies, etc.) is not changed by utilizing
the previously described vacant spaces that are face adjacent to the
SFP [spent fuel pool] walls in Region I [Region 1] to store design
basis assemblies that are less reactive than RI A [Region 1
Configuration A] type assemblies. Fuel assemblies of different types
are presently stored face adjacent to these walls. This proposal
will allow additional assemblies to be located face adjacent to the
Region I SFP walls and does not effect the precursors to any
postulated spent fuel pool accidents.
The consequences of an accident different than that previously
analyzed additionally remains unchanged. Evaluations have
demonstrated that the fuel handling accident reactivity values will
remain less than the 0.95 Keff acceptance criteria in the
event of a fuel handling accident, assuming an initial SFP boron
concentration of 1000 ppm. The boron concentration limit is
additionally bounded by ANO-2 [Arkansas Nuclear One, Unit 2] TS
[Technical Specification] Limiting Condition for Operation (LCO)
3.9.12.c which limits SFP boron to greater than 1600 ppm at all
times.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
As discussed previously, the proposed SFP configuration will not
result in exceeding the acceptance criteria of 0.95 Keff
during normal or accident conditions. Since fuel assemblies are
currently located along the Region I SFP walls, no new or different
kind of accident than that previously evaluated exists. Locations
required to be vacant will remain physically blocked. In the event
that a ``misloading'' type accident occurs in this region,
evaluations have shown that the fuel handling accident reactivity
values will remain well below 0.95 Keff when initial SFP
boron concentrations are at or above 1000 ppm, which is
significantly less than the TS boron limit of 1600 ppm.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does Not Involve a Significant Reduction in the
Margin of Safety.
As previously discussed, the proposed configuration will not
result in exceeding the 0.95 Keff acceptance criteria
during normal operations that assume zero concentration of boron at
the maximum water density in the SFP or during accident conditions
that assume an initial SFP boron concentration of at least 1000 ppm.
Furthermore, ANO-2 TS 3.9.12.c requires SFP boron to be maintained
greater than 1600 ppm at all times. Fuel assemblies are presently
stored along the Region I SFP walls; therefore, storing additional
assemblies along these same walls will not significantly reduce the
margin to safety since it has been shown that the current CSA
[criticality safety analysis] remains valid.
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, Winston and Strawn,
1400 L Street, NW., 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: May 12, 2000, as supplemented June 19,
2000.
Description of amendment request: The proposed amendment would
revise the Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and 2),
calculated doses and associated descriptions/information listed in the
Updated Final Safety Analysis Reports (UFSARs) for the Design Basis
Accidents (DBAs). An evaluation of all of the BVPS-1 and 2 dose
calculations was completed which reviewed the input parameter values,
the input assumptions, and the methodologies used. Some of the input
parameter values, input assumptions and methodologies used in the DBA
dose calculations were revised. The resultant DBA dose calculation
revisions necessitate associated revisions to the UFSARs. Additionally,
some changes would be made in response to Generic Letter 99-02. For
BVPS-1, the requested amendment would affect the analyses for the
following DBAs: loss of offsite AC power, fuel-handling accident,
accidental release of waste gas, steam generator tube rupture, major
secondary system pipe rupture, rod cluster control assembly ejection,
single reactor coolant pump locked rotor, and loss of reactor coolant
from small ruptured pipes/loss-of-coolant accidents. For BVPS-2, the
requested amendment would affect the analyses for the following DBAs:
steam system piping failures, loss of AC power, reactor coolant pump
shaft seizure, rod cluster control assembly ejection, failure of small
lines carrying primary coolant outside containment, steam generator
tube rupture, loss-of-coolant accidents, waste gas system failure, and
fuel-handling accidents.
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?
Following a reevaluation of the calculated dose values for BVPS
Unit 1 and Unit 2
[[Page 54087]]
design basis accidents (DBAs) as described in their respective
[Updated Final Safety Analysis Report] UFSAR, several calculated
dose values were identified to be increased. These increases were
small and remained within the applicable DBA previously approved
regulatory limit.
The increases for each DBA were as a result of revised plant
data being used in the dose calculation, revised calculation
assumptions, or new methodology. These changes were not the result
of plant hardware changes. The changes were intended to ensure that
accurate, current and conservative licensing basis information and
assumptions were used for DBA dose analyses. The UFSAR changes are
proposed to reflect the revised analyses results for the Unit 1 and
Unit 2 UFSAR.
Since the calculated DBA radiological doses remain within the
applicable DBA previously approved regulatory limit, these
calculated dose values do not involve a significant increase in the
probability or consequences of an accident as previously evaluated
in the BVPS Unit 1 and Unit 2 UFSAR.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
BVPS Unit 1 and Unit 2 calculations which are used to determine
DBA calculated dose values were revised. The changes were as a
result of revised plant data being used in the dose calculation,
revised calculation assumptions or new methodology. The changes were
intended to ensure that accurate, current and conservative licensing
basis information and assumptions were used for DBA dose analyses.
The DBA events themselves remain the same postulated events as
previously described within the BVPS Unit 1 and Unit 2 UFSARs. The
revised dose calculations do not create the possibility of a new or
different kind of accident from the DBA accidents previously
evaluated in the UFSAR. These changes were not the result of plant
hardware changes. The changes were only in the calculations. The
UFSAR changes are proposed to reflect the revised analyses[']
results for the Unit 1 and Unit 2 UFSAR.
3. Does the change involve a significant reduction in a margin
of safety?
This amendment request addresses only proposed changes to the
Unit 1 and Unit 2 UFSAR, which was determined to involve an
Unreviewed Safety Question pursuant to 10 CFR 50.59. This request
does not propose modifying any Technical Specification criteria.
This request proposes that several calculated dose values for BVPS
Unit 1 and Unit 2 DBAs be increased following a reevaluation of
their design basis calculations. These proposed increases are small
and remained within the applicable DBA previously approved
regulatory limit. Thus, the proposed changes to the UFSAR which
originated from revised BVPS DBA dose calculations [do] not involve
a significant reduction in the margin of safety for BVPS Unit 1 and
Unit 2 because the Technical Specifications will not be altered and
the increase in calculated dose values is small and remains within
regulatory approved 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, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Marsha Gamberoni
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: July 31, 2000
Description of amendment request: The amendment would change
Technical Specifications 3.8.1.1, ``Electrical Power Systems--A.C.
Sources--Operating,'' and 3.8.1.2, ``Electrical Power Systems--A.C.
Sources--Shutdown.'' The index and the Bases for these Technical
Specifications will be modified as a result of the proposed changes.
The proposed changes will allow certain emergency diesel generator
(EDG) surveillance requirements to be performed when the plant is
operating instead of shut down as currently required. Additional
changes will remove EDG accelerated testing and special reporting
requirements, and the surveillance requirement to perform EDG
inspections. EDG inspections will still be performed as recommended by
the manufacturer. The proposed changes will not adversely impact the
type and amounts of effluents that may be released off site.
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's analysis is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed Technical Specification changes are associated with
the surveillance requirements for the Emergency Diesel Generators
(EDGs) and will not affect the ability of the EDGs to perform their
intended safety function. Therefore, the proposed Technical
Specification changes will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Since there are no changes in components, component operation, or
system operation, this change does not create the possibility of an
accident of a different type.
3. Involve a significant reduction in a margin of safety.
The proposed changes will have no adverse effect on plant operation
or equipment important to safety. The plant response to the design
basis accidents will not change and the accident mitigation equipment
will continue to function as assumed in the design basis accident
analysis. Therefore, there will be no significant reduction in a margin
of safety.
Based on the staff's analysis, 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: Lillian M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut
NRC Section Chief: James W. Clifford
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: June 14, 2000
Description of amendment request: The proposed amendments would
revise Vogtle's Surveillance Requirements (SR) 3.8.1.9 and 3.8.1.14 to
reduce the emergency Diesel Generator (EDG) loading requirements from
6800 kW and 7000 kW to 6500 kW and
7000 kW. These changes will make the above SRs consistent
with SR 3.8.1.3 and 3.8.1.13 which are in the current Technical
Specifications (TS). In addition, the proposed amendments would revise
TS section 5.6.7, ``EDG Failure Report'', to correct a typographical
error.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed change to section 5.6.7 is administrative only
since it does nothing more than correct a typographical error. The
proposed changes to the DG loading requirements specified in SRs
3.8.1.9 and
[[Page 54088]]
3.8.1.14 have no impact on or relationship to the probability of any
of the initiating events assumed for the accidents previously
evaluated. Therefore, the proposed changes do not involve a
significant increase in the probability of any accident previously
evaluated. Furthermore, since the proposed loading requirements
bound the maximum expected loading for the DGs, SRs 3.8.1.9 and
3.8.1.14 will continue to demonstrate that the DGs are capable of
performing their safety function. Since the proposed changes do not
adversely affect the capability of the DGs to perform their safety
function, the outcome of the accidents previously evaluated (i.e.,
radiological consequences) will not be affected. Therefore, the
proposed changes do not involve a significant increase in the
consequences of any accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed change to section 5.6.7 is administrative only
since it does nothing more than correct a typographical error. The
proposed changes to the DG loading requirements specified in SRs
3.8.1.9 and 3.8.1.14 will not introduce any new equipment or create
new failure modes for existing equipment. Other than the reduced
loading requirements for the DGs, the proposed changes will not
affect or otherwise alter plant operation. The DGs will remain
capable of performing their safety function. No other safety related
or important to safety equipment will be affected by the proposed
changes. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
No. The proposed change to section 5.6.7 is administrative only
since it does nothing more than correct a typographical error. The
proposed changes reduce the loading requirements of SRs 3.8.1.9 and
3.8.1.14. The new loading requirements bound the maximum expected
loading of the DGs under the worst case scenario, and they are
consistent with the regulatory guidance found in Regulatory Guide
(RG) 1.9, Revision 3, ``Selection, Design, and Qualification of
Diesel-Generator Units Used as Standby (Onsite) Electric Power
Systems at Nuclear Power Plants,'' July 1993. Reduction in wear and
tear should inherently increase the reliability of the DGs.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Conclusion
Based on the above evaluation, the proposed changes do not
involve a significant hazard as defined in 10 CFR 50.92.
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: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216
NRC Section Chief: Richard L. Emch, Jr.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of amendment request: August 11, 2000 (TS-400).
Description of amendment request: The proposed amendment would
change the Units 2 and 3 Technical Specifications to revise the testing
frequency for certain isolation valves of a type known as excess flow
check valves (EFCV). The proposed testing frequency would allow a
representative sample to be tested every 24 months, such that each EFCV
is tested at least once every 120 months. The current specification
requires that each EFCV be tested at least once every 24 months.
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:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The current excess flow check valve (EFCV) frequency requires
that each reactor instrument line EFCV be tested every 24 months.
The EFCVs are designed to automatically close upon excessive
differential pressure including failure of the down stream piping or
instrument and will reopen when appropriate. This proposed change
will allow a reduction in the number of EFCVs that are verified
tested every 24 months, to approximately 20 percent of the valves
each cycle. BFN and industry operating experience demonstrates high
reliability of these valves. Neither the EFCVs or their failure is
capable of initiating a previously evaluated accident. Therefore,
there is no increase in the probability of occurrence of an accident
previously evaluated.
The instrument lines going to the Reactor Coolant Pressure
boundary with EFCVs installed have flow restricting devices upstream
of the EFCV. The consequences of a unisolable failure of an
instrument line has been previously evaluated and meets the intent
of NRC Safety Guide 11. The offsite exposure has been calculated to
be substantially below the limits of 10 CFR 100. Additionally,
coolant lost from such a break is inconsequential compared to the
makeup capabilities of normal and emergency makeup systems. Although
not expected to occur as a result of this change, the effects of a
postulated failure of an EFCV to isolate and instrument line break
as a result of reduced testing are bounded by TVA analysis.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This proposed change reduces the number of EFCVs tested each
operating cycle. No other changes to the TS are being proposed. BFN
and industry operating experience demonstrates that these valves are
highly reliable, a proposed reduction in test frequency is bounded
by previous evaluation of a line rupture. The change will not alter
the operation of process variables, structures, systems or
components described in the BFN Updated Final Safety Analysis
Report. Therefore, reduction in the number of EFCVS tested each
cycle does not result in the possibility of a new or different kind
of accident.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The consequences of an unisolable rupture of an instrument line
has been previously evaluated and meets the intent of NRC Safety
Guide 11. The proposed amendment does not involve a significant
reduction in a margin of safety.
Therefore, the proposed revised surveillance frequency does not
adversely affect the public health and safety, and does not involve
any significant safety hazards.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET I0H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 4, 2000 (TS 99-20)
Brief description of amendments: The proposed amendments would
change the Sequoyah Nuclear Plant (SQN) Technical Specifications (TS),
Section 6.2.2, to change the title of various shift members and to
change the Shift Technical Advisor requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the
[[Page 54089]]
issue of no significant hazards consideration, which is presented
below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The title change of Shift Operations Supervisor to Shift Manager
is administrative. The elimination of TS 6.2.2.b and Table 6.2-1 is
considered an administrative change. These two items contain similar
requirements as those contained in 10 CFR 50.54(m)(2)(iii), 10 CFR
50.54(m)(2)(i), and 10 CFR 50.54(k). These sections are considered a
duplicate of the requirements contained in the Code of Federal
Regulations. This request also eliminates the title of Shift
Technical Advisor (STA) but will not eliminate or reduce licensee
responsibilities in this area. This request is based on an NRC
policy statement, contained in Generic Letter 86-04, that supports
the transition of engineering expertise from the STA position to
another individual on shift who possesses the mandated education
qualifications. The proposed administrative and organizational
changes do not result in any increase in the probability or
consequences of an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
As described above, the proposed changes are administrative and
organizational in nature and cannot create the possibility of a new
or different kind of accident from any accident previously
evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
As described above, the proposed changes are administrative and
organizational in nature. The proposed changes are based on approved
NRC guidance. The margin of safety is, therefore, not reduced.
The NRC 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
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.
AmerGen Energy Company, LLC., et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: July 21, 2000.
Description of amendment request: The amendment requests approval
to remove a shutdown requirement with regard to the relief valve
position indication system in Section 3.13 of the Technical
Specifications.
Date of publication of individual notice in Federal Register:
August 2, 2000 (65 FR 47520).
Expiration date of individual notice: September 1, 2000.
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).
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: July 21, 2000.
Description of amendment request: The proposed amendment revises
the Oyster Creek Nuclear Generating Station Technical Specifications
Section 3.13 to remove a shutdown requirement with regard to the relief
valve position indication system.
Date of issuance: August 21, 2000.
Effective Date: As of date of issuance to be implemented within 30
days.
Amendment No.: 214.
Facility Operating License No. DPR-16: This amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (65 FR 47520) August 2, 2000. That notice provided
an opportunity to submit comments on the Commission's proposed no
significant hazards consideration determination. No comments have been
received. The notice also provided for an opportunity to request a
hearing by September 1, 2000, but indicated that if the Commission
makes a final no significant hazards consideration determination any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment finding of
exigent circumstances, state consultation, and final determination of
no significant hazards consideration determination are contained in a
Safety Evaluation dated August 21, 2000.
Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, N.W., Washington, D.C. 20036-5869.
NRC Section Chief: Marsha Gamberoni.
[[Page 54090]]
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 0-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: June 6, 2000.
Brief description of amendments: The amendments revise the
information in Figure 3.5.5-1, ``Minimum Required RWT Volume in TS
3.5.5, Refueling Water Tank (RWT),'' for the three units. The
amendments relocate design information to the Bases of the TSs,
truncate the lower end of the RWT limit curve at 210 deg.F, retitle
the right-hand ordinate from ``minimum useful volume required in the
RWT'' to ``RWT Volume,'' and delete the two footnotes and the
references to the footnotes.
Date of issuance: August 18, 2000.
Effective date: August 18, 2000, to be implemented within 45 days
of the date of issuance.
Amendment Nos.: Unit 1-127, Unit 2-127, Unit 3-127.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 12, 2000 (65 FR
43043).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 18, 2000.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket No. 50-237, Dresden Nuclear Power
Station, Unit 2, Grundy County, Illinois
Date of application for amendment: April 30, 1999.
Brief description of amendment: The amendment revised the
expiration date of the operating license to allow 40 years of operation
from the original date of issuance of the Provisional Operating
License.
Date of issuance: August 24, 2000.
Effective date: August 24, 2000.
Amendment No.: 178.
Facility Operating License No. DPR-19: The amendment revised the
Facility Operating License. Date of initial notice in Federal Register:
March 22, 2000 (65 FR 15376).
The Commission's related evaluation of the amendment is contained
in an Environmental Assessment dated June 1, 2000, and a Safety
Evaluation dated August 24, 2000.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington
Date of application for amendment: November 18, 1999, as
supplemented by letter dated June 7, 2000.
Brief description of amendment: The amendment changes Technical
Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP)'' to
include the requirement for laboratory testing of engineered safety
feature ventilation system charcoal samples per American Society for
Testing and Materials D3803-1989 and the application of a safety factor
of 2.0 to the charcoal filter efficiency assumed in the plant design-
basis dose analyses.
Date of issuance: August 25, 2000.
Effective date: August 25, 2000.
Amendment No.: 167.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73088).
The June 7, 2000, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 25, 2000.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: April 9, 1998, as supplemented by
letters dated January 13, 1999, and June 28, 2000.
Brief description of amendment: The amendment consists of changes
to the River Bend Station (RBS) Facility Operating License, paragraph
2.C(13). The amendment allows RBS to operate with final feedwater
temperature reduction in order to extend the fuel cycle by maintaining
the core thermal power at or close to rated power, thus delaying the
start of normal coastdown. The January 13, 1999, letter provided a
revised proprietary version of the licensee's analysis submitted in its
original April 9, 1998, application and the June 28, 2000, letter
provided additional information to support staff review of the original
application, and did not affect the initial finding of no significant
hazards consideration determination dated May 20, 1998 (63 FR 27762).
Date of issuance: August 22, 2000.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 112.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: May 20, 1998 (63 FR
27762).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 22, 2000.
No significant hazards consideration comments received: No.
Entergy Operations, Inc. Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2, Pope County, Arkansas
Date of amendment request: July 14, 1999, as supplemented by
letters dated February 24, 2000, and July 17, 2000.
Brief description of amendments: The proposed amendments delete
requirements from the Technical Specifications to maintain a Post
Accident Sampling System (PASS). Licensees were required to implement
PASS upgrades as a result of NUREG-0737, ``Clarification of TMI [Three
Mile Island Nuclear Station] Action Plan Requirements,'' and Regulatory
Guide 1.97, Revision 3, ``Instrumentation for Light Water Cooled
Nuclear Power Plants to Assess Plant and Environmental Conditions
During and Following an Accident.'' Implementation of these upgrades
were an outcome of the Nuclear Regulatory Commission's lessons learned
from the accident that occurred at TMI, Unit 2. The staff has concluded
that the information obtained using PASS is not required for the
development of protective action recommendations or for core damage
assessment.
Date of issuance: August 17, 2000.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 208 and 218
Facility Operating License Nos. DPR-51 and NPF-6: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43773). The supplements dated February 24 and July 17, 2000, did not
change the scope of the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 17, 2000.
No significant hazards consideration comments received: No
[[Page 54091]]
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: June 1, 2000, as supplemented by
letter dated June 30, 2000.
Brief description of amendment: The amendment approves a proposed
modification that changes the Perry Nuclear Power Plant as described in
the Updated Safety Analysis Report by installing inflatable seals that
surround the Emergency Service Water (ESW) alternate intake sluice
gates. This modification is necessary so that the licensee may use
inflatable seals to minimize leakage of warm water into the ESW forebay
from the Service Water discharge and thus maintain the ESW temperature
below the design limit.
Date of issuance: August 22, 2000
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 114
Facility Operating License No. NPF-58: This amendment authorizes
revision of the Updated Safety Analysis Report.
Date of initial notice in Federal Register: June 14, 2000 (65 FR
37414) The supplemental information contained clarifying information
and did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register Notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 22, 2000.
No significant hazards consideration comments received: No
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of application for amendment: May 10, 1999, as supplemented
April 6, April 26, and June 5, 2000.
Brief description of amendment: Changes Technical Specifications to
establish the actions to be taken for an inoperable ``Standby Filter
Unit'' (SFU) System due to a degraded control building boundary.
Date of issuance: August 11, 2000
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 233
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 14, 1999 (64 FR
38029). The April 6, April 26, and June 5, 2000, submittals provided
additional clarifying information that did not change the initial
proposed no significant hazards consideration determination or expand
the scope of the application beyond the initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 11, 2000.
No significant hazards consideration comments received: No
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: June 7, 2000
Brief description of amendment: The amendment revised the Technical
Specifications, Section 3.10.8, ``SHUTDOWN MARGIN (SDM) Test --
Refueling,'' correcting an administrative error introduced when
Amendment No. 92, dated March 2, 2000, was issued.
Date of issuance: August 24, 2000
Effective date: As of the date of issuance to be implemented
concurrently with Amendment No. 92.
Amendment No.: 93
Facility Operating License No. NPF-69: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 16, 2000 (65 FR
37807)
The staff's related evaluation of the amendment is contained in a
Safety Evaluation dated August 24, 2000.
No significant hazards consideration comments received: No
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: February 1, 2000, as
supplemented on April 13, 2000
Brief description of amendment: The amendment temporarily suspends
the technical (TSs) requirements for TSs 3.7.7 and 3.7.8 in order to
conduct testing of the cable spreading room that will pressurize the
area to a pressure that exceeds the adjacent control room envelope
area.
Date of issuance: August 22, 2000
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 181
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 31, 2000 (65 FR
34748)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 22, 2000.
No significant hazards consideration comments received: No
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: October 29, 1999, as
supplemented March 14 and April 25, 2000
Brief description of amendment: The amendment conforms the license
to reflect the transfer of possession under Operating License No. DPR-
22 to a newly formed utility operating company subsidiary of Northern
States Power Company merged with New Century Energies, Inc., as
approved by Order of the Commission dated May 12, 2000.
Date of issuance: August 18, 2000
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 111
Facility Operating License No. DPR-22. Amendment revised the
Operating License.
Date of initial notice in Federal Register: February 10, 2000 (65
FR 6641)
The March 14 and April 25, 2000, supplements were within the scope
of the initial application as originally noticed.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 12, 2000.
No significant hazards consideration comments received: No
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: February 29, 2000, as
supplemented July 10, 2000
Brief description of amendment: The amendment (1) approves
continued use of two exceptions previously granted by the Nuclear
Regulatory Commission (NRC) to the American Society of Mechanical
Engineers N510-1989 testing requirements for the emergency filtration
train (EFT) system, (2) revises the Technical Specifications (TSs) to
reflect modifications to the EFT system that eliminate the need for
additional test exceptions, (3) revises the TSs to be consistent with
the guidance of NRC Generic Letter 99-02, and (4) revises the TSs to
include operability requirements for the EFT system during operations
that could result in a fuel handling accident.
Date of issuance: August 18, 2000
Effective date: As of the date of issuance and shall be implemented
within 45 days.
[[Page 54092]]
Amendment No.: 112
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 5, 2000 (65 FR
17917)
The July 10, 2000, supplemental letter provided clarifying
information that was within the scope of the original application and
did not change the staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 18, 2000.
No significant hazards consideration comments received: No
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, and Docket No. 72-10,
Prairie Island Independent Spent Fuel Storage Installation, Goodhue
County, Minnesota
Date of application for amendments: October 29, 1999, as
supplemented March 14 and April 25, 2000.
Brief description of amendments: The amendments conform the
licenses to reflect the transfer of possession under Operating Licenses
Nos. DPR-42 and DPR-60 and Materials License No. SNM-2506 to a newly
formed utility operating company subsidiary of Northern States Power
Company merged with New Century Energies, Inc., as approved by Order of
the Commission dated May 12, 2000.
Date of issuance: August 18, 2000.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 154 and 145.
Facility Operating Licenses Nos. DPR-42 and DPR-60 and Materials
License No. SNM-2506: Amendments revised the Operating Licenses and
Materials License.
Date of initial notice in Federal Register: February 10, 2000 (65
FR 6642)
The March 14 and April 25, 2000, supplements were within the scope
of the initial application as originally noticed.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 12, 2000.
No significant hazards consideration comments received: No
PECO Energy Company, Public Service Electric and Gas Company Delmarva
Power and Light Company; and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: August 11, 1999, as
supplemented June 29, 2000.
Brief description of amendments: The Updated Final Safety Analysis
Report (USFAR) was updated to reflect credit for use of a limited
amount of containment overpressure in calculations of net positive
suction head available for emergency core cooling pumps.
Date of issuance: August 14, 2000.
Effective date: As of Date of issuance.
Amendments Nos.: 233 and 237.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
authorized changes to the UFSAR.
Date of initial notice in Federal Register: April 19, 2000 (65 FR
21038). The June 29, 2000, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated August 14, 2000.
No significant hazards consideration comments received: No
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: June 4, 1999, as supplemented
October 22, 1999.
Brief description of amendment: The amendment revises the license
and Technical Specifications to reflect changes related to the transfer
of the license for the Hope Creek Generating Station, to the extent
held by Public Service Electric and Gas Company, to PSEG Nuclear
Limited Liability Company.
Date of issuance: August 21, 2000
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 129
Facility Operating License No. NPF-57: This amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: June 30, 1999 (64 FR
35193). The October 22, 1999, supplement provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the scope of the original
Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 16, 2000.
No significant hazards consideration comments received: No
Public Service Electric & Gas Company, 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: April 13, 2000
Brief description of amendments: The amendments deleted Technical
Specification (TS) 3/4.1.3.2.2 which is related to shutdown and control
rod group demand position indication in Modes 3, 4, and 5.
Date of issuance: August 17, 2000
Effective date: As of the date of issuance, and shall be
implemented within 60 days of issuance.
Amendment Nos.: 232 and 213
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 28, 2000 (65 FR
39960)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 17, 2000.
No significant hazards consideration comments received: No
Public Service Electric & Gas Company, 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: June 4, 1999, as supplemented
October 22, 1999.
Brief description of amendments: The amendment revises the license
and Technical Specifications to reflect changes related to the transfer
of the license for the Salem Nuclear Generating Station, Unit Nos. 1
and 2, to the extent held by Public Service Electric and Gas Company,
to PSEG Nuclear Limited Liability Company.
Date of issuance: August 21, 2000
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 233 and 214
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the License and Technical Specifications.
Date of initial notice in Federal Register: June 30, 1999 (64 FR
35192). The October 22, 1999, supplement provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the scope of the original
Federal Register notice.
[[Page 54093]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 16, 2000.
No significant hazards consideration comments received: No
Public Service Electric & Gas Company, 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: January 24, 2000, as
supplemented April 19 and May 31, 2000.
Brief description of amendments: The amendments revise the
radiological effluent technical specifications (RETS) and
administrative controls requirements (i.e., Sections 3/4.3,
Instrumentation, 3/4.11, Radioactive Effluents, 3/4.12, Radiological
Environmental Monitoring, 6.0, Administrative Controls, and the table
of contents and definitions) in the Technical Specifications (TSs) by
implementing programmatic controls for RETS in the administrative
controls section and relocating procedural details of the RETS, with
various changes, to the offsite dose calculation manual (ODCM) or to
the process control program (PCP). The proposed changes follow the
guidance and requirements in NRC Generic Letter 89-01, ``Implementation
of Programmatic Controls in the Technical Specifications for
Radiological Effluent Technical Specifications (RETS) in the
Administrative Controls Section of the Technical Specifications and the
Relocation of Procedural Details of RETS to the Offsite Dose
Calculation Manual or to the Process Control Program,'' that was issued
in 1989. There is also the change to add the word ``oxygen'' to the
title of ``Radioactive Gaseous Effluent Monitoring Instrumentation.''
Date of issuance: August 24, 2000
Effective date: August 24, 2000
Amendment Nos.: 234 and 215
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 2000 (65 FR
11094) The supplemental letters dated April 19 and May 31, 2000,
provided clarification that did not alter the scope of the proposed
action or the initial no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 24, 2000.
No significant hazards consideration comments received: No
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: August 24, 1999, as
supplemented on December 29, 1999, and June 16, 2000
Brief description of amendments: The amendments revised Technical
Specification 3.3.2 ``Engineered Safety Features Actuation System
(ESFAS) Instrumentation'' to relax the slave relay test frequency from
quarterly to every refueling not to exceed 18 months.
Date of issuance: August 22, 2000
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 114 and 92
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 22, 2000 (65 FR
15386). The supplemental letters dated December 29, 1999, and June 16,
2000, provided clarifying information only, and did not change the
scope of the August 24, 1999, application nor the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 22, 2000.
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: March 6, 2000
Brief description of amendment: Revised the Technical Specification
(TS) and associated Bases for Limiting Condition for Operation 3.9.4,
``Refueling Operations--Containment Penetrations,'' to allow the
containment personnel airlock doors and certain containment
penetrations to be open during refueling activities under appropriate
administrative controls.
Date of issuance: August 24, 2000
Effective date: August 24, 2000
Amendment No.: 26
Facility Operating License No. NPF-90: Amendment revises the TS.
Date of initial notice in Federal Register: May 17, 2000 (65 FR
31361)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 24, 2000.
No significant hazards consideration comments received: No
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendments: June 22, 2000, as supplemented
July 25, 2000
Brief description of amendments: The amendments revise the
Technical Specifications Sections 3.4.1.4, 3.4.1.6, 4.4.1.4, and
4.4.1.6.1; add Sections 4.4.1.6.4 and 4.4.1.6.5; and revise Bases
Section 3/4.4.1 for Units 1 and 2. These changes will allow for the
implementation of a vacuum-assisted backfill technique when returning
an isolated Reactor Coolant System (RCS) loop to service, and provide
the necessary controls for temperature and boron concentration of the
isolated RCS loop to ensure the required shutdown margin is maintained.
Date of issuance: August 25, 2000
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 223 and 204
Facility Operating License Nos. NPF-4 and NPF-7: Amendments change
the Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46019). The letter dated July 25, 2000, contained clarifying
information only, and did not change the initial no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 25, 2000.
No significant hazards consideration comments received: No
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: May 19, 2000, as supplemented
August 3, 2000
Brief description of amendments: These amendments eliminate one of
the license conditions and associated implementation dates from
Appendix C to the licenses. The license condition required the licensee
to submit a license amendment application and supporting radiological
dose analyses demonstrating compliance with General Design Criterion 19
dose limits without reliance on potassium iodide.
Date of issuance: August 15, 2000
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 198 and 203
[[Page 54094]]
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Operating Licenses.
Date of initial notice in Federal Register: June 6, 2000 (65 FR
35966)
The August 3, 2000, supplemental letter provided clarifying
information that was within the scope of the original application and
did not change the staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 15, 2000.
No significant hazards consideration comments received: No
Dated at Rockville, Maryland, this 30th day of August 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 00-22779 Filed 9-5-00; 8:45 am]
BILLING CODE 7590-01-P