[Federal Register Volume 64, Number 134 (Wednesday, July 14, 1999)]
[Notices]
[Pages 38022-38043]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-17750]
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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 June 19, 1999, through July 2, 1999. The
last biweekly notice was published on June 30, 1999 (64 FR 35199).
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 and
Directives Branch, Division of Administration 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 August 13, 1999, 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 at the local public
document room for the particular facility involved. 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
[[Page 38023]]
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: Rulemakings and
Adjudications Staff, 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 at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: June 15, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to incorporate the performance-
based 10 CFR 50 Appendix J, Option B for Type A tests (containment
integrated leakage rate tests). Option B will be implemented for Type A
testing in accordance with NRC Regulatory Guide 1.163, ``Performance-
Based Containment Leak-Test Program,'' dated September 1995, and
Nuclear Energy Institute (NEI) Guideline 94-01, Revision 0, ``Industry
Guideline for Implementing Performance-Based Option of 10 CFR Part 50,
Appendix J,'' dated July 26, 1995. Type B and C testing (containment
penetration leakage tests) will continue to be performed in accordance
with 10 CFR 50 Appendix J, Option A.
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 license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The Harris Nuclear Plant (HNP) Type A testing history provides
justification for the proposed test schedule change to one test in a
10 year period. With the successful Type A tests of September 1992
and May 1997, and a greater than 24 month elapsed time between the
two tests, CP&L considers the requirement of two consecutive Type A
tests to have been met. This testing has affirmed the acceptable
reliability of the containment structure to minimize leakage as
designed, and provides assurance that its performance to
continuously function as designed is not challenged due to this test
schedule extension to once in 10 years.
This proposed change to revise the test schedule frequency does
not impact or alter the design of any system, structure or
component. The limit on allowable leakage is not increased. Type A
testing provides periodic verification of the leak tight integrity
of the containment and the components that penetrate the containment
structure. NUREG-1493, Section 10.1.2, ``Leakage-Testing
Intervals,'' states that reducing the frequency of Type A tests from
the current three per 10 years to one per 20 years was found to lead
to an imperceptible increase in risk.
Therefore, based on these considerations, and the previous
plant-specific Type A test results, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes only incorporate the performance-based
testing approach authorized in 10 CFR 50 Appendix J, Option B, and
are justified based on previous plant-specific Type A test results.
Plant structures, systems, and components will not be operated in a
different manner as a result of these proposed changes and no
physical modifications to equipment are involved. The interval
extensions allowed by Option B of 10 CFR 50 Appendix J do not have
the potential for creating the possibility of a new or different
type of accident from any previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes do not change the allowable leak rate from
the containment; they only allow an extension of the interval
between the performance of Type A leak rate testing. NUREG-1493
provides the technical basis for the NRC's rulemaking to revise
containment leakage testing requirements for nuclear power reactors
in 10 CFR 50 Appendix J. NUREG-1493, Section 10.1.2, ``Leakage-
Testing Intervals,'' states that increasing the interval between
integrated leakage-rate tests is possible with minimal impact on
public risk.
Based on these considerations and the previous plant-specific
Type A test results, the proposed changes do not involve a 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 38024]]
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.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Section Chief: Sheri R. Peterson.
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: June 15, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification 4.7.D.6 by replacing the leakage limit
of 11.5 standard cubic feet per hour (scfh) for each main steam
isolation valve (MSIV) with a limit of 46 scfh on the total combined
leakage from all four main steam lines.
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 to the Technical Specifications, Appendix
A, modifies the allowed MSIV leakage limit to an aggregate value
with no change to the total allowed leakage rate. This change does
not affect either the automatic or manual features that would close
the MSIVs. Performance of the leakage tests do not adversely affect
any accident previously evaluated. Consequently, this proposed
amendment does 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 safety function of the MSIVs is to provide a timely steam
line isolation to mitigate the release of radioactive steam and
limit reactor inventory loss under certain accident and transient
conditions. The MSIVs are designed to automatically close whenever
plant conditions warrant main steam line isolation. Changing the
leakage limits to include an aggregate value does not affect the
isolation function. No new equipment will be installed or utilized,
and no new operating conditions will be initiated as a result of
this change. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The total allowed leakage rate for all MSIVs remains unchanged
at 46 scfh. Therefore, there will be no change in the types or
significant increase in the amounts of any effluents released
offsite, and, thus, the radiological analyses remain unchanged and
within the guidelines of 10 CFR 100 and General Design Criteria 19.
Therefore, these changes do not involve a significant reduction in
the margin of safety.
Therefore, based upon 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.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
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: May 19, 1999.
Description of amendment request: The proposed amendments would
relocate Technical Specification Section \3/4\.4.4, ``Chemistry'' from
the TS to the Updated Final Safety Analysis Report and Administrative
Technical Requirements.
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 simplify the TS, meet regulatory
requirements for relocated TS, and implement the recommendations of
the NRC's Final Policy Statement on TS improvements. The Chemistry
requirements will be relocated to the Updated Final Safety Analysis
Report (UFSAR) and Administrative Technical Requirement that has
been incorporated into the UFSAR by reference. Future changes to
these requirements will be controlled by 10 CFR 50.59. The proposed
changes are administrative in nature and do not involve any
modification to any plant equipment or affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any previously
evaluated accident.
Consequently, this proposed amendment does not involve a
significant increase in the probability or consequences of any
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 are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety related system performs
its function. Therefore, this proposed TS amendment would 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?
The proposed amendment represents the relocation of current
requirements that are based on generic guidance or previously
approved provisions for other stations. The proposed changes are
administrative in nature and do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analysis. The proposed changes have been
evaluated and found to be acceptable for use at Duane Arnold Energy
Center and Quad Cities Nuclear Power Station. Since the proposed
changes are administrative in nature, and are based on NRC accepted
provisions which have been adopted at other nuclear facilities, and
maintain the necessary levels of system reliability, 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
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.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of amendment request: June 3, 1999.
Description of amendment request: The proposed amendment would
delete sections of the Technical Specifications that no longer apply to
the Haddam
[[Page 38025]]
Neck Plant's permanently shutdown and defueled condition; increase the
weight of loads allowed over the spent fuel pool; relocate certain
definitions and requirements from the Technical Specifications to the
Technical Requirements Manual (TRM), Connecticut Yankee Quality
Assurance Program (CYQAP), or the Radiological Effluent Monitoring and
Offsite Dose Calculation Manual (REMODCM); and correct typographical
errors, renumber sections, and repaginate the Technical Specifications.
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 licensee's summary of its analysis is presented
below:
The proposed changes do not involve an SHC [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
In the present plant configuration, the reactor-related
accidents previously evaluated (i.e., LOCA [loss-of-coolant
accident], MSLB [main steamline break], etc.) are no longer
possible. The accidents previously evaluated that are still
applicable to the plant are fuel handling accidents and gaseous and
liquid radioactive releases. The following events are presently
considered as bounding of all other events:
--Fuel handling and cask drop accidents in the spent fuel building,
--Criticality in the spent fuel pool,
--Loss of spent fuel cooling,
--Resin fire (gaseous release), and
--Rupture of a tank containing radioactive liquid.
There is no significant increase in the probability of a fuel
handling accident since refueling operations have ceased, with a
corresponding decrease in the frequency of fuel movement. The
radiological consequences of a fuel handling accident, should one
occur, decrease the longer the spent fuel is allowed to decay. As
discussed previously, the spent fuel inventory of radioactive iodine
and noble gases [, with the exception of Kr-85,] have decayed more
than 20 half-lives since shutdown and are no longer a release
concern. With this reduced source the results of the fuel handling
accident show that the filters of Specification 3.9.12 are no longer
necessary. The allowed weight over the spent fuel pool is still less
than that previously [evaluated]. Therefore, there has been no
increase in the probability or consequences of a fuel handling or
cask drop accident.
Criticality controls are imposed by specifications \3/4\.9.13
and \3/4\.9.14 * * * [The requirements of these specifications have
not been changed.] Therefore, there has been no increase in the
probability or consequences of a criticality event.
Spent fuel cooling is maintained by keeping the pool temperature
below 150 deg.F. Should normal cooling be lost, the availability of
an abundant supply of water ensures that sufficient time is
available prior to boiling to restore cooling. This is controlled by
specifications \3/4\.9.11 and \3/4\.9.16 * * * [The requirements of
these specifications have not been changed. Technical specification
\3/4\.9.15 does not apply to the permanently defueled condition of
the plant. Therefore, there has been no increase in the probability
or consequences of a loss of cooling event.]
The probability of a gaseous or liquid radioactive release is
not changed by the proposed revisions. As the plant undergoes
decommissioning, the previous limiting events [such as a loss-of-
coolant-accident] are no longer applicable, and previous non-
limiting events [such as a resin fire] now become limiting. These *
* * events have not changed from how they might have occurred in the
past. The radiological consequences of a gaseous or liquid
radioactive release are bounded by the fuel handling accident during
defueled operation and a spent resin fire during processing of resin
from the reactor coolant system decontamination. The rupture of a
tank containing radioactive liquid was assessed and found to be
bounded by these events. With the plant defueled and permanently
shutdown, the demands on the radwaste systems are lessened since no
new radioisotopes are being generated by irradiation or fission.
Therefore, there is no increase in the probability or consequences
of a gaseous or liquid radioactive release.
* * * * *
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes are generally of an administrative nature
and do not have an effect on the physical plant. The events
considered bound other potential events and are considered the
limiting cases for potential gaseous or liquid releases to the
environment.
With the plant undergoing decommissioning, the types of
accidents one might be concerned with involve criticality of the
spent fuel, or draining of the spent fuel pool. None of the proposed
changes affect the possibility of such an event. Also, none of the
proposed changes could lead to a radiological release of a greater
magnitude than for the events considered, such as might occur with
the accumulation of a greater quantity of radioactive material in
one location, or with damage to a greater number of fuel assemblies
than considered in the fuel handling accident.
The proposed changes restrict the operations that can be
conducted at the plant, and do not permit any new type of activity
from what had previously been authorized. The effect on systems,
structures and components affected by the proposed changes have no
adverse impact on the storage of fuel nor on the processing of
radioactive wastes presently at the site. The present set of
limiting events are a subset of events previously considered.
Therefore these changes do not create the possibility of a new or
different kind of accident from any accident previously considered.
* * * * *
3. Involve a significant reduction in a margin of safety.
The proposed changes have no impact on the analyses of
postulated design basis events remaining applicable to the Haddam
Neck Plant. Analysis of the limiting events show that their
consequences to the public are within the limits of 10 CFR [Part] 20
and the EPA PAGs [Environmental Protection Agency Protective Action
Guides]. The consequences to members of the operating staff are
within the limits of 10 CFR [Part] 50, Appendix A, General Design
Criterion 19, ``Control Room Habitability''. Therefore there is no
reduction in a margin of safety.
* * * * *
Based on the above, the proposed changes to the operating license
and technical specifications do not involve a reduction in the margin
of safety due to the reduced decay heat load, the decay of
radionuclides since shutdown, and by maintaining the heavy load
restriction.
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.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut, 06457.
Attorney for licensee: Mr. J. A. Ritsher, Ropes & Gray, One
International Place, Boston, Massachusetts 02110-2624.
NRC Section Chief: Michael T. Masnik.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: June 10, 1999.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to adopt a Ventilation Filter
Testing Program in TS Section 6.0, ``Administrative Controls,'' and
remove the specific ventilation filter surveillance requirements from
TS 3/4.6.4.4, ``Hydrogen Purge System,'' TS 3/4.6.5.1, ``Shield
Building Emergency Ventilation System,'' and TS 3/4.7.6.1, ``Control
Room Emergency Ventilation System.''
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
[[Page 38026]]
consideration, which is presented below:
The Davis-Besse Nuclear Power Station has reviewed the proposed
changes and determined that a significant hazards consideration does
not exist because operation of the Davis-Besse Nuclear Power Station
(DBNPS), Unit Number 1, in accordance with this change 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. The replacement of the specific Technical
Specification (TS) ventilation filter testing Surveillance
Requirements for the Containment Hydrogen Purge System 3/4.6.4.4),
Shield Building Emergency Ventilation System (3/4.6.5.1), and the
Control Room Emergency Ventilation System (3/4.7.6.1), with a
reference to the newly created Ventilation Filter Testing Program
contained in TS Administrative Controls Section 6.8.4.f, Ventilation
Filter Testing Program, is a removal and relocation of certain TS
details. The proposed TS 6.8.4.f will, however, add controls to
maintain similar operation, maintenance, testing and system
operability for these three ventilation systems. The TS Bases
changes reflect the use of the Ventilation Filter Testing Program.
Therefore, it can be concluded that the proposed changes do not
involve a significant increase in the probability of an accident
previously evaluated.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
affect accident conditions or assumptions used in evaluating the
radiological consequences of an accident. No physical alterations of
the DBNPS are involved, nor are plant operating methods being
changed. 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 the proposed
changes do not change the way the plant is operated. No new or
different types of failures or accident initiators are being
introduced by the proposed changes.
3. Not involve a significant reduction in a margin of safety
because no inputs into the calculation of any Technical
Specification Safety Limit, Limiting Safety System Settings,
Technical Specification Limiting Condition for Operation, or other
previously defined margins for any structure, system, or component
important to safety are being affected by the proposed changes.
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.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Anthony J. Mendiola.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: May 10, 1999.
Description of amendment request: The proposed amendment would
correct the regulation referenced in Section 5.8, ``High Radiation
Area,'' of the Crystal River Unit 3 (CR-3) Improved Technical
Specifications (ITS). The ITS currently references 10 CFR 20, Paragraph
20.1601(2) and (3), whereas the correct reference is 10 CFR 20.1601(c).
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
The proposed change to the Crystal River Unit 3 (CR-3) Improved
Technical Specifications (ITS) is editorial in nature. The change
involves revising the incorrect reference in ITS Section 5.8.1 to
the correct Code of Federal Regulations reference that pertains to
controlling access to high radiation areas. The proposed ITS change
does not involve any change to plant design, operation, maintenance,
or procedures. As a result, no changes to the plant are being made
which would impact either the contributors to an accident or to the
consequences of an accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated?
The proposed editorial change to the ITS does not involve any
changes to any plant structure, system, or component (SSC) or to its
operation or maintenance. There is no impact on any equipment that
would be considered as contributors to either new or different
accidents. Thus, the change to the ITS does not create the
possibility of a new or different kind of accident.
3. Involve a significant reduction in a margin of safety?
The proposed change to the ITS involves a reference change and
does not involve the design or operation of any plant SSC. No
changes to the methods for controlling access to high radiation
areas are proposed. No changes to the methods for controlling
personnel and/or activities in high radiation areas are proposed.
Therefore, the proposed 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC-A 5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Section Chief: Sheri R. Peterson.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of amendment request: May 17, 1999.
Description of amendment request: The proposed amendment would
revise a note in Surveillance Requirement (SR) 3.3.8.1 in the Crystal
River Unit 3 Improved Technical Specifications (ITS). The note
currently states that, when Emergency Diesel Generator (EDG) Loss Of
Power Start instrumentation is placed in an inoperable status solely
for performance of this surveillance, entry into associated Conditions
and Required Actions may be delayed for up to four hours provided the
two channels monitoring the Function for the bus are OPERABLE or
tripped. The proposed revision to the note states that entry into the
Conditions and Required Actions of ITS Section 3.3.8 is not required
provided the applicable Conditions and Required Actions of ITS Section
3.8.1, ``Electrical Power Systems, AC Sources--Operating,'' are entered
for the EDG being made inoperable. The proposed amendment would also
delete a superceded 60-day surveillance frequency and the associated
note which indicated that the frequency was not effective after
November 23, 1997.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
Proposed Improved Technical Specifications (ITS) Change A--Revision of
Surveillance Note
The note in Surveillance Requirement (SR) 3.3.8.1 involves the
timing for placing Crystal River Unit 3 (CR-3) into the applicable
Conditions and Required Actions when
[[Page 38027]]
performing the surveillance. The proposed revision will make the
note consistent with the actual method of performing the
surveillance at CR-3. The design and testing configuration does not
allow CR-3 to use the relief provided by the note. As a result, the
Conditions and Required Actions of ITS Section 3.3.8 are entered at
the start of the surveillance performance. The design and testing
configuration requires entry into ITS Section 3.8.1 Conditions prior
to performing SR 3.3.8.1. The revised note would require entering
the applicable Conditions and Required Actions of ITS Section 3.8.1
for one Emergency Diesel Generator (EDG) being inoperable. This
approach and the proposed note are conservative relative to the
current note. The proposed note results in no changes to the method
or to the timing of performing SR 3.3.8.1. Direct entry into ITS
Section 3.8.1 Conditions will achieve the same final ITS condition
as if Section 3.3.8 Conditions and Required Actions were entered.
Therefore, the probability of occurrence and the consequences of any
accident previously evaluated are unaffected by this change.
Proposed ITS Change B--Deletion of Frequency Note
The note under Frequency in SR 3.3.8.1 involves the period of
time that the 60 day surveillance frequency would be in effect. The
60 day frequency was a temporary extension that was needed to
implement modifications to the EDG during the 1997 CR-3 design
outage. This was a one-time extension of the frequency. The note
indicates this temporary nature of the 60 day frequency. Deleting
the note is an editorial change since the surveillance has reverted
back to its 31 day frequency and the note is no longer effective.
Because the proposed deletion of the note is an editorial change,
and no change is proposed to the current 31 day frequency, the
probability of occurrence and the consequences of any accident
previously evaluated are unaffected by this change.
2. Create the possibility of a new or different kind of accident
from previously evaluated accidents?
Proposed ITS Change A--Revision of Surveillance Note
The proposed revision of the note in SR 3.3.8.1 involves only
the timing of entry into associated ITS Conditions and Required
Actions. No changes are proposed to the existing ITS Conditions and
Required Actions. The proposed change is conservative since it will
require entering the appropriate Conditions and Required Actions
immediately upon starting SR 3.3.8.1. Changing the timing for entry
into ITS Conditions and Required Actions does not create the
possibility of a new or different kind of accident from those
evaluated previously.
Proposed ITS Change B--Deletion of Frequency Note
Deletion of the note under SR 3.3.8.1 Frequency is an editorial
change since the note is no longer effective. The current frequency
for performing SR 3.3.8.1 is 31 days. This is the same frequency
that was in effect prior to the one-time, temporary change of the
frequency to 60 days. The editorial change of deleting the note that
is no longer effective does not create the possibility of a new or
different kind of accident from those evaluated previously.
3. Involve a significant reduction in a margin of safety?
Proposed ITS Change A--Revision of Surveillance Note
One manner in which a margin of safety related to a Surveillance
Requirement might be affected would be if entry into a Limiting
Condition for Operation (LCO) were delayed. The result of a delay in
entering an LCO would be an increase in the time before a Required
Action was taken, such as commencing a plant shutdown. Generally,
such allowed times reflected in ITS Required Actions are based on
some margin. Increasing the time allowed before starting a certain
Required Action might result in a reduction of a margin of safety.
However, the proposed ITS change allows entry into Section 3.8.1
Conditions immediately rather than after a delay. The proposed ITS
change does not change the final plant condition required by the
ITS. Therefore, the proposed ITS change does not result in a
reduction in a margin of safety.
Proposed ITS Change B--Deletion of Frequency Note
Another manner in which a margin of safety related to a
surveillance requirement might be affected would be if the frequency
of performance were changed. Generally, margin might be reduced if
the frequency were reduced (i.e., the interval between performing
surveillances were increased). This proposed editorial change to
delete the note in SR 3.3.8.1 Frequency does not result in a change
to the surveillance frequency. Thus, the proposed deletion of the
note does not affect the existing 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.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Section Chief: Sheri R. Peterson.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: June 11, 1999.
Description of amendment request: The proposed amendment makes
various plant organization title changes.
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
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
proposed changes are administrative in nature and do not affect
assumptions contained in plant safety analyses, the physical design
and/or operation of the plant, nor do they affect Technical
Specifications that preserve safety analysis assumptions. None of
the proposed changes involve a physical modification to the plant, a
new mode of operation or a change to the UFSAR [Updated Final Safety
Analysis Report] transient analyses. No Technical Specification
Limiting Condition for Operation, Action statement or Surveillance
Requirement is affected by any of the proposed changes. These
proposed changes do not reduce the level of qualification, authority
or accountability associated with the affected Technical
Specification responsibilities. Further, the proposed changes do not
alter the design, function, or operation of any plant component.
Therefore, the proposed amendment does not affect the probability of
occurrence or consequences of an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. The proposed changes
are administrative in nature and do not affect assumptions contained
in plant safety analyses, the physical design and/or modes of plant
operation defined in the plant operating license, or Technical
Specifications that preserve safety analysis assumptions. The
proposed changes do not introduce a new mode of plant operation or
surveillance requirement, nor involve a physical modification to the
plant. The proposed changes do not alter the design, function, or
operation of any plant components. Therefore, the proposed amendment
does not affect 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
amendment would not involve a reduction in a margin of safety. The
proposed changes are administrative in nature. There is no reduction
in the organization position qualifications, authority and
accountability associated with the affected Technical Specification
responsibilities. None of the proposed changes involve a physical
modification to the plant, a new mode of operation or a change to
the UFSAR transient analyses. No Technical Specification Limiting
Condition for Operation, Action statement, or Surveillance
Requirement is affected. Therefore, the proposed amendment does not
reduce the margin of safety.
[[Page 38028]]
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.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: S. Singh Bajwa.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: April 12, 1999.
Description of amendment request: The proposed amendment would
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)
Surveillance Requirement (SR) 3.6.1.3.7 to allow a representative
sample of reactor instrumentation line excess flow control valves
(EFCV) to be tested every 24 months, instead of testing each EFCV 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:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The current SR frequency requires each reactor instrumentation
line EFCV to be tested every 24 months. The EFCVs at DAEC are
designed so that they will not close accidentally during normal
operation, will close if a rupture of the instrument line is
indicated downstream of the valve, can be reopened when appropriate,
and have their status indicated in the control room (reference DAEC
UFSAR [updated final safety analysis report] 1.8.11). This proposed
change allows a reduced number of EFCVs to be tested every 24
months. There are no physical plant modifications associated with
this change. Industry operating experience demonstrates a high
reliability of these valves. Neither EFCVs nor their failures are
capable of initiating previously evaluated accidents; therefore
there can be no increase in the probability of occurrence of an
accident regarding this proposed change.
Instrument lines connecting to the Reactor Coolant Pressure
Boundary (RCPB) with EFCVs installed also have a flow-restricting
orifice upstream of the EFCV. The consequences of an unisolable
rupture of such an instrument line [have] been previously evaluated
in response to Regulatory Guide (RG) 1.11 (DAEC UFSAR 1.8.1.1). That
evaluation assumed a continuous discharge of reactor water for the
duration of the detection and cooldown sequence (3.5 hours).
Therefore, although not expected to occur as a result of this
change, the postulated failure of an EFCV to isolate as a result of
reduced testing is bounded by this previous evaluation. Therefore,
there is no increase in the previously evaluated consequences of the
rupture of an instrument line and there is no potential increase in
the consequences of an accident previously evaluated as a result of
this change.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
This proposed change allows a reduced number of EFCVs to be
tested each operating cycle. No other changes in requirements are
being proposed. Industry operating experience demonstrates the high
reliability of these valves. The potential failure of an EFCV to
isolate by the proposed reduction in test frequency is bounded by
the previous evaluation of an instrument line rupture. This change
will not physically alter the plant (no new or different type of
equipment will be installed). This change will not alter the
operation of process variables, structures, systems, or components
as described in the safety analysis. Thus, a new or different kind
of accident will not be created.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
The consequences of an unisolable rupture of an instrument line
[have] been previously evaluated in response to RG 1.11 (reference
DAEC UFSAR 1.8.1.1). That evaluation assumed a continuous discharge
of reactor water for the duration of the detection and cooldown
sequence (3.5 hours). The only margin of safety applicable to this
proposed change is considered to be that implied by this evaluation.
Since a continuous discharge was assumed in this evaluation, any
potential failure of an EFCV to isolate postulated by this reduced
testing frequency is bounded and 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.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869
NRC Section Chief: Claudia M. Craig
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: April 30, 1999.
Description of amendment request: The proposed amendment would
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)
Surveillance Requirement (SR) 3.4.3.1 to revise the safety function
lift setpoint tolerance limits for the main safety valves (SVs) and the
safety/relief valves (SRVs). The current tolerance bands for the SVs
and SRVs would be revised from -3% to +1% to a new band of plus or
minus 3%.
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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change allows an increase in the as-found SV and
SRVs safety mode setpoint tolerance, determined by test after the
valves have been removed from service, from +1%/-3% to plus or minus
3%.
The proposed change does not alter the TS requirements on the
nominal SV or SRV safety mode lift setpoints, the SRV relief mode
setpoints, the required frequency for the SV or SRV lift setpoint
tests, or the number of SVs and SRVs required to be operable.
Consistent with current requirements, this change continues to
require that these valves be adjusted to within plus or minus 1% of
their nominal lift setpoints following testing. This change does not
change the behavior and operation of any SV or SRV and therefore has
no significant impact on reactor operation. It also has no
significant impact on response to any perturbation of reactor
operation including transients and accidents previously analyzed in
the Updated Final Safety Analysis Report (UFSAR).
This change does not involve physical changes to the valves, nor
does it change the operating characteristics or safety function of
the valves. The proposed TS revision involves no significant changes
to the operation of any systems or components in normal or accident
operating conditions and no changes to existing structures, systems,
or components. Therefore these changes will not increase the
probability of an accident previously evaluated.
Generic considerations related to the change in setpoint
tolerance were addressed in NEDC-31753P, ``BWROG In-Service Pressure
Relief Technical Specification Revision Licensing Topical Report,''
and were reviewed and approved by the NRC in a Safety Evaluation
(SE) dated March 8, 1993. The plant specific evaluations, required
by the NRC's SE and performed to support this proposed change, show
that there is adequate
[[Page 38029]]
margin to the design core thermal limits and to the reactor vessel
pressure limits using a plus or minus 3% setpoint tolerance. They
also show that operation of the high pressure coolant injection
(HPCI) and reactor core isolation cooling (RCIC) systems will not be
adversely affected and the containment response from a loss of
coolant accident will be acceptable. The plant systems associated
with these proposed changes will still be capable of meeting all
applicable design basis requirements and retain the capability to
mitigate the consequences of accidents described in the UFSAR.
Therefore, these changes will not involve a significant increase in
the consequences of any accident previously evaluated.
Therefore, the proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
These proposed changes were developed in accordance with the
provisions contained in the NRC SE, dated March 8, 1993, for the
``BWR Owners Group Inservice Pressure Relief Technical Specification
Revision Licensing Topical Report,'' NEDC-31753P. The revised SV and
SRV setpoint tolerance limit will not adversely impact the operation
of any safety related component or equipment. Since the proposed
changes involve no significant hardware changes, no significant
changes to the operation of any systems or components, and no
changes to existing structures, systems, or components, there can be
no impact on the occurrence of any accident.
The proposed change would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change to allow an increase in the SV and SRV safety
mode setpoint tolerance from +1%/-3% to plus or minus 3% does not
alter the nominal SV or SRV lift setpoints or the number of SVs or
SRVs required to be operable. This change does not involve physical
changes to the valves, nor does it change the operating
characteristics or the safety function of the valves. The proposed
change does not involve a physical alteration of the plant. No new
or different equipment is being installed. There is no alteration to
the parameters within which the plant is normally operated. As a
result no new failure modes are being introduced. There are no
changes in the methods governing normal plant operation, nor are the
methods utilized to respond to plant transients altered.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
The proposed change does not involve a significant reduction in
a margin of safety. Establishment of the plus or minus 3% SV and SRV
setpoint tolerance limit will not adversely impact the operation of
any safety related component or equipment. Engineering evaluations
concluded that there are no significant impacts on fuel thermal
limits, safety related systems, structures or components, and no
significant impact on the accident analyses associated with the
proposed changes.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated, and the establishment of the setpoints
for the actuation of equipment relied upon to respond to an event.
The proposed change does not significantly impact the condition or
performance of structures, systems, and components relied upon for
accident mitigation. The proposed change does not significantly
impact any safety analysis assumptions or results.
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.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401.
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Claudia M. Craig.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: May 10, 1999.
Description of amendment request: The proposed amendment would
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)
Section 2.1.1.2, to revise the Safety Limit Minimum Critical Power
Ratio (SLMCPR) to support operation with GE-12 fuel with a 10x10 pin
array.
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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
There is no change to any plant equipment other than the fuel.
The SLMCPR protects the fuel in accordance with the design bases.
The SLMCPR calculations limit the bundle power to ensure the
critical power ratio remains unchanged. Therefore, there is not an
increase in the probability of transition boiling. The basis of the
SLMCPR calculation remains the same, ensuring that greater than
99.9% of all fuel rods in the core avoid transition boiling if the
limit is not violated. Therefore, there is no increase in the
probability of occurrence of a previously evaluated accident.
The fundamental sequences of accidents have not been altered.
The Minimum Critical Power Ratio (MCPR) Operating Limits are
selected such that potentially limiting accidents do not cause the
MCPR to decrease below the SLMCPR anytime during the accident.
Therefore, there is no impact on any of the limiting accidents.
Therefore there is no increase in the consequences of any accident
previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The SLMCPR values are designed to ensure that fuel damage from
transition boiling does not occur in at least 99.9% of the fuel rods
as a result of the limiting postulated accident. The values are
calculated in accordance with NRC-approved General Electric methods.
The approved General Electric methods are comprehensive for ensuring
that fuel designs will perform within acceptable bounds. The SLMCPR
ensures that the fuel is protected in accordance with the design
basis. The function, location, operation, and handling of the fuel
remain unchanged. Therefore, the possibility of a new or different
kind of accident is not created.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
The SLMCPR values do not alter the design or function of any
plant system. The new values were calculated using NRC-approved
methods to maintain the same margin of safety as presently exists
for the prevention of transition boiling. At least 99.9% of the fuel
rods will avoid transition boiling if the SLMCPR is not violated.
Therefore, a significant reduction in a margin of safety is not
involved.
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.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401.
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Claudia M. Craig.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa
Date of amendment request: May 10, 1999.
Description of amendment request: The proposed amendment would
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)
to: (1)
[[Page 38030]]
insert NOTE for Limiting Condition for Operation (LCO) 3.7.4 that would
allow intermittent opening of the control building boundary under
administrative control; (2) add a CONDITION, REQUIRED ACTION and
COMPLETION TIME to LCO 3.7.4 for when both standby filter unit (SFU)
subsystems are inoperable due to inoperable control building boundary
in MODES 1, 2, and 3; (3) re-letter items in LCO 3.7.4 for consistency;
and (4) revise LCO 3.7.4 CONDITION D (new CONDITION E) to add ``for
reasons other than CONDITION B.''
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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Requiring the plant to enter LCO 3.0.3 when the control building
pressure envelope is not intact is excessively restrictive. This
change provides less restrictive requirements for operation of the
facility. These less restrictive requirements do not result in
operation that will increase the probability of initiating an
analyzed event. The proposed change is acceptable because of the low
probability (less than 3.04 x 10-8) of a DBA [design
basis accident] occurring during the 24 hour Completion Time, and
the availability the SFU system to provide a filtered environment
(albeit with potential control room in-leakage).
Intermittent opening of the control building boundary requires
controls which consist of stationing a dedicated individual at the
opening who is in continuous communication with the control room.
This individual will have a method to rapidly close the opening when
a need for control room isolation is indicated. For entry and exit
through doors the administrative control is performed by the person
entering or exiting the area. As a result, the consequences of any
accident previously evaluated are not significantly increased.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. This change does not involve new or different equipment
being installed at the facility. The proposed change is acceptable
because of the low probability (less than 3.04 x 10-8) of
a DBA occurring during the 24 hour Completion Time, and the
availability of the SFU system to provide a filtered environment
(albeit with potential control room in-leakage).
Intermittent opening of the control building boundary requires
controls which consist of stationing a dedicated individual at the
opening who is in continuous communication with the control room.
This individual will have a method to rapidly close the opening when
a need for control building isolation is indicated. For entry and
exit through doors the administrative control is performed by the
person entering or exiting the area.
3. The proposed amendment will not involve a significant
reduction in a margin of safety. Requiring the plant to enter LCO
3.0.3 when the control room ventilation envelope is not intact is
excessively restrictive. The proposed change is acceptable because
of the low probability (less than 3.04 x I0-8) of a DBA
occurring during the 24 hour Completion Time.
Intermittent opening of the control room boundary requires
controls which consist of stationing a dedicated individual at the
opening who is in continuous communication with the control room.
This individual will have a method to rapidly close the opening when
a need for control building isolation is indicated. For entry and
exit through doors the administrative control is performed by the
person entering or exiting the area.
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.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401.
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Claudia M. Craig.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 8, 1999.
Description of amendment request: The proposed change corrects the
described method by which the Standby Gas Treatment system heaters are
to be tested. This change is necessary because the reference provided
in Technical Specification Section 5.5.7e is in 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:
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The correction of an error and clarification of a testing
method does not alter any of the precursors assumed in the CNS
[Cooper Nuclear Station] accident analysis. The proposed wording for
testing SGT [Standby Gas Treatment] heaters is in accordance with
ASME N510-1989, Section 14.5.1, ``Testing of Nuclear Air Treatment
Systems.'' Since the proposed change does not affect this portion of
plant design and operation, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change will not create the possibility of a new or
different kind of accident than evaluated in the Updated Safety
Analysis Report (USAR). The proposed change does not result in any
physical change to CNS structures, systems, or components, nor does
it change the fit, form, or function of any equipment or components
taken credit for in the accident analyses described in the USAR.
Therefore, correction of a test reference and specific description
of the testing method for the SGT heaters does not create the
possibility of a new or different kind of accident.
The proposed change will not create a significant reduction in
the margin of safety. The proposed change does not alter the design
or administrative controls necessary to ensure the required
performance of the physical barriers during acticipated operational
occurrences and postulated accidents. This conclusion is based on
the fact that the proposed change corrects an erroneous reference,
conforms to industry standards, and is consistent with past and
current operating practice at CNS; therefore, the proposed change
does not create 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.
Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 15, 1999.
Description of amendment request: The proposed change would allow
the use of the service water (SW) system to directly supply cooling
water to the reactor equipment cooling (REC) system during a loss-of-
coolant accident (LOCA) event. The present maximum allowable REC water
leakage rate is based on the requirement that there will be sufficient
water in the REC surge tank to allow the REC system to fulfill its
safety function for 30 days post-LOCA condition. A proposed Updated
Safety Analysis Report (USAR) revision would allow Cooper Nuclear
Station (CNS) to
[[Page 38031]]
revise the maximum allowable REC system leakage during normal power
operation such that the REC system surge tank would assure that the REC
would fulfill its safety function for at least the first 7 days
following a large break LOCA. The SW system would fulfill the safety
functions of the REC system, if required, for the remaining duration of
the accident.
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 licensee states that the
proposed request:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does not involve a significant increase in
the probability of an accident previously evaluated in the USAR
since there are no hardware changes associated with this USAR
change. Procedure changes associated with this USAR change are
limited to direction on which division of SW/REC backup to initiate
first, and incorporation of new system leakage limits into
surveillance procedures.
The proposed change also does not involve a significant increase
in the consequences of an accident previously evaluated in the USAR.
This conclusion is based on the safety evaluation (Attachment 2 [of
the June 15, 1999, application]) which demonstrates that the SW
system will fulfill the safety functions of the REC system in a post
LOCA condition and thus the proposed change will not affect the
performance and reliability of the REC system. The emergency systems
cooled by the REC system, the ECCS [emergency core cooling] systems
and their room coolers, will therefore also fulfill their safety
function when directly supplied by the SW system.
2. Does not create the possibility for a new or different kind
of accident from any accident previously evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated in
the USAR. The proposed license amendment does not introduce any new
equipment or hardware changes. It does, however, allow the SW system
to perform a different type of function than it is presently
licensed to perform in a post LOCA condition. This SW system post
LOCA function has been previously demonstrated to fulfill the
functions of the REC in a non LOCA emergency shutdown which are the
same as the functions required following a LOCA.
3. Does not create a significant reduction in the margin of
safety.
The proposed activity does not involve a significant reduction
in the margin to safety. The safety evaluation (Attachment 2)
demonstrates that the SW system will perform the required REC post
LOCA functions. There is an added required operator action which is
to align the SW system to directly supply cooling water to the REC
critical loops. As discussed in the safety evaluation [of the June
15, 1999, application], this action can be performed from the main
control room utilizing one control switch and there is sufficient
control room indication for the operator to be alerted to the need
for the use of service water backup. There is also sufficient time
for the operator to perform the task. Trending (prior to a
postulated LOCA) routinely provides the control room operator with
REC system leakage information. In a post LOCA situation, this
leakage information would assist the operator in taking timely
action to initiate the service water back-up before the need is
alarmed in the control room.
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.
Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 15, 1999.
Description of amendment request: The purpose of the requested
license amendment is to revise the Updated Safety Analysis Report
(USAR) to incorporate the latest analysis to demonstrate adequate net
positive suction head (NPSH) for the low pressure emergency core
cooling system (ECCS) pumps following a large break loss-of-coolant
accident (LOCA). Specifically, the change would allow (1) reliance on a
slightly larger amount of containment overpressure for residual heat
removal (RHR) and core spray (CS) pump operation during worst-case
long-term LOCA conditions (greater than 1000 seconds) while still
maintaining original license margins of 3 and 6 pounds per square inch
(psi), respectively, for the difference between minimum available
containment pressure and the pressure required for minimum pump NPSH,
(2) reliance on a small amount of containment overpressure for CS pump
runout during worst-case short-term LOCA conditions (less than 10
minutes) while still maintaining an adequate pressure margin of at
least 5 psi, and (3) the use of ANS 5.1 decay heat model in the USAR
Section 5.2.6 as currently presented based on analysis justifying the
use of this model as described in the amendment request.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does not involve an increase in the
probability of an accident previously evaluated in the USAR. There
are no changes being proposed to the maintenance, operation, or
design of plant systems or equipment postulated to initiate
accidents or transients.
The proposed change does not involve an increase in the
consequences of an accident previously evaluated in the USAR. This
conclusion is based on the conclusions of the safety evaluation
(Attachment 2 [of the June 15, 1999, application]). This safety
evaluation demonstrates that the containment overpressure is
sufficiently conservative, and that the calculated margins between
the available containment overpressure and the overpressure required
to assure adequate low pressure ECCS pump NPSH are such that ECCS
pump operation, as credited in the CNS [Cooper Nuclear Station]
accident analysis, remains unchanged.
2. Does not create the possibility for a new or different kind
of accident from any accident previously evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated in
the USAR. The proposed license amendment does not introduce any new
equipment or hardware changes. The attached safety evaluation
demonstrates that the only equipment affected by this License
Amendment are the low pressure ECCS pumps and that these will retain
their ability to function following a LOCA.
3. Does not create a significant reduction in the margin of
safety.
The proposed activity does not involve a significant reduction
in the margin of safety. The safety evaluation (Attachment 2)
demonstrates that, although there is an increased reliance on
containment overpressure to assure adequate low pressure ECCS pump
NPSH, there remains sufficient margin to provide confidence that the
ECCS pumps will operate as required. Sufficient margin is
demonstrated with the added conservatism of a 2-sigma (2 standard
deviation) uncertainty in the decay heat model, increased suction
strainer debris loading, increased RHR heat exchanger tube plugging
margin, and increases in SW [Service Water] and Suppression Pool
temperatures. The minimum margin available between available
overpressure and required overpressure is at least 5 psi for CS
(just prior to 10 minutes) and at least 3 psi for RHR (well after 10
minutes).
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 38032]]
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.
Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 23, 1999.
Description of amendment request: The proposed change to the
Technical Specifications would increase the allowed outage time for the
Control Room Air Conditioning Subsystem from 30 days to 60 days, on a
one-time basis only, for each train.
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 operational requirements for the Control Room Air
Conditioning Subsystems (CRACS) are contained in Technical
Specification 3.7.6.2 ``Control Room Subsystems Air Conditioning.''
This Limiting Condition for Operation (LCO) requires that two
independent Control Room Air Conditioning Subsystems (trains) be
operable during all modes of operation. The LCO action statement for
operational modes 1, 2, 3 and 4, with one Control Room Air
Conditioning Subsystem inoperable, states: ``restore the inoperable
system to operable status within 30 days or be in at least Hot
Standby [Mode 3] within the next 6 hours and in Cold Shutdown within
the following 30 hours.'' The LCO action statement for operational
modes 5 and 6 with one Control Room Air Conditioning Subsystem
inoperable, states: ``restore the inoperable system to operable
within 30 days or initiate and maintain operation of the remaining
OPERABLE Control Room Air Conditioning Subsystem or immediately
suspend all operations involving CORE ALTERATIONS or positive
reactivity changes.''
The proposed change adds the following note: ``* For cycle 7,
the allowable outage time may be extended to 60 days, on a one-time
basis, for each train, to implement modifications to the control
room air conditioning subsystems. The provisions of specifications
3.0.4 and 4.0.4 are not applicable during the implementation of
modifications to the air conditioning subsystems.''
This change is a one-time only change to Technical Specification
3.7.6.2 in order to facilitate the installation of a design change
to the CRACS during the present operating cycle. This change will
not affect the existing 30 [day] AOT period presently in place in
Technical Specification 3.7.6.2 which requires specific actions in
the event that the CRACS is determined to be inoperable for any
other reason. The design basis accidents are not affected as a
result of the proposed one-time change to the Technical
Specifications. The CRACS are support subsystems which can only
contribute to the initiation of an accident if the whole function is
lost. The plant would be required to shutdown before this occurred.
The proposed change does not adversely affect accident initiators or
precursors nor alter the design assumptions, conditions,
configuration of the facility (other than the CRACS) or the manner
in which the plant is operated nor does it adversely affect the
response of the plant to a transient or accident. This one-time
change is to be utilized only during the present operating cycle
(cycle 7) in order to facilitate the implementation of a design
change to modify the existing safety-related refrigerant subsystems
(one train at a time) and replace them with safety-related chilled
water subsystems. This design change is being implemented to improve
the overall reliability of the safety-related subsystems.
The consequences of an extended loss of the operating CRACS and
the non-safety related chilled water subsystem, during all modes of
operation, would result in a slow gradual rise in control room
temperature. The temperature of the control room is normally
maintained between 70 to 72 deg.F at the discretion of the Unit
Shift Supervisor utilizing a non-safety-related train of CRACS. In
the event that the control room temperature increased to a
temperature greater than 75 deg.F, plant procedures require starting
other equipment in the non-safety-related chilled water subsystem or
a safety-related train of CRACS to restore control room temperature
to its normal operating band. In the unlikely event that the non-
safety-related chilled water subsystems and the operable safety-
related train of CRACS fail during the proposed 60 day AOT period,
Technical Specification 3.7.6.2 would require that actions be
commenced to place the plant in a shutdown condition. Additionally,
alternative actions to reduce control room temperature could also be
initiated as identified in a plant procedure. It has been
conservatively determined that safety-related equipment in the
control room can be operated continuously up to 90\F in an
environment without affecting the capability of the equipment.
The exception to specifications 3.0.4 and 4.0.4 as stated in the
proposed one-time change to Technical Specification 3.7.6.2 will not
involve an increase in the probability or consequences of an
accident. TS 3.0.4 prohibits entry into a mode when the conditions
for the LCO are not met and the associated action(s) requires a
shutdown if they are not met within a specified time interval.
Surveillance Requirement 4.0.4 prohibits entry into a mode unless
the associated surveillance requirement(s) has been performed within
the stated interval. During the implementation of the modification,
when one safety-related train of CRACS is inoperable, it is possible
that a plant shutdown could occur due to reasons unrelated to the
planned modifications of the CRACS. As stated above, the CRACS are
support subsystems which do not contribute to the initiation of any
accident previously evaluated. Entry of the plant into an
operational mode from a shutdown mode as a result of the proposed
modification does not adversely affect accident initiators or
precursors nor alter the design assumptions, conditions,
configuration of the facility (other than the CRACS) or the manner
in which the plant is operated nor does it adversely affect the
response of the plant to a transient or accident. The functions of
the CRACS to provide a controlled environment inside of the control
room complex to ensure the comfort of the plant operators and to
ensure adequate climate conditions for the operability of equipment
will not be impaired in any way as a result of a plant mode change.
The remaining actions identified in TS 3.7.6.2 are unchanged as a
result of the proposed change. The risk significance involved with
removing a safety-related train of the CRACS during power operation
or during refueling conditions is low based on the short period (60
days per train) and consequences of losing this function. The CRACS
is excluded from modeling in the Seabrook Station Probabilistic Risk
Assessment (PRA) due to its extremely low risk significance.
Therefore, the proposed change does not involve 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.
The proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated
since it is a support system and the loss of function will require a
plant shutdown. The proposed change adds the following note which
pertains to both affected action statements: ``* For cycle 7, the
Allowable Outage Time may be extended to 60 days, on a one-time
basis, for each train during the implementation of modifications to
the control room air conditioning subsystems. The provisions of
specifications 3.0.4 and 4.0.4 are not applicable during the
implementation of modifications to the air conditioning
subsystems.'' As previously identified, this change is a one-time
only change to Technical Specification 3.7.6.2 in order to
facilitate the installation of a design change to CRACS during the
present operating cycle.
The CRACS are support subsystems which do not contribute to the
creation of a new or different kind of accident from any previously
evaluated nor is it used to mitigate the consequences of a transient
or accident. The functions of the CRACS are to provide a controlled
environment inside of
[[Page 38033]]
the control room complex to ensure the comfort of the plant
operators and to ensure adequate climate conditions for the
operability of equipment. The CRACS consists of two independent
safety-related air conditioning trains that provide cooling of
recirculated control room air. Due to previous reliability problems
with the CRACS, an additional non-safety chilled water subsystem has
been installed to provide control room cooling on a continuous
basis. Baseload operation of the non-safety related chilled water
subsystem to provide control room cooling reduces the operational
load on the safety-related refrigerant trains.
Implementation of the modification to the CRACS subsystems
during the 60 day AOT duration in no way affects the availability of
the non-safety-related chilled water subsystem or the operable
safety-related train of the CRACS to meet the control room cooling
requirements. The proposed modification removes freon from the
control room complex and the quantity of chilled water in the closed
loop system is too small to become a flood hazard. The consequences
of an extended loss of the operating CRACS train and the non-safety
related chilled water subsystem would result in a slow gradual rise
in control room temperature. In the event that control room
temperature increased to a temperature greater than 75 deg.F, plant
procedures require starting either the non-safety-related chilled
water subsystem or a safety-related train of CRACS to restore
control room temperature. Additionally, in the unlikely event of a
loss of the non-safety related chilled water subsystem and the
operable safety-related train of the CRACS, Technical Specification
3.7.6.2 would require that actions be taken to place the plant in a
shutdown condition.
It has been conservatively determined that safety-related
equipment in the control room can be operated continuously in an
environment up to 90 deg.F without affecting the capability of the
equipment. This proposed change will not affect the existing 30 day
AOT period presently in place in Technical Specification 3.7.6.2
which requires specific actions in the event that the CRACS is
determined to be inoperable for any other reason.
The exception to specifications 3.0.4 and 4.0.4 as stated in the
proposed one-time change to Technical Specification 3.7.6.2 will not
involve the creation of an accident of any type. During the
implementation of the proposed modification, when one safety-related
train of CRACS is inoperable, it is possible that a plant shutdown
could occur due to reasons unrelated to the planned modifications of
the CRACS. Entry of the plant into an operational mode from a
shutdown mode as a result of the proposed modification does not
adversely affect accident initiators or precursors nor alter the
design assumptions, conditions, configuration of the facility or the
manner in which the plant is operated nor the manner that it
responds to a transient or accident. The functions of the CRACS to
provide a controlled environment inside of the control room complex
to ensure the comfort of the plant operators and to ensure adequate
climate conditions for the operability of equipment will not be
impaired in any way as a result of a plant mode change. The
remaining actions identified in TS 3.7.6.2 are unchanged as a result
of the proposed change.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed one-time change to Technical Specification 3.7.6.2
will not involve a significant reduction in the margin of safety.
The functions of the CRACS are to provide a controlled environment
inside of the control room complex to ensure the comfort of the
plant operators and to ensure adequate climate conditions for the
operability of equipment. The CRACS consists of two independent
safety-related air conditioning trains that provide cooling of
recirculated control room air. Additionally, the Seabrook Station
design incorporates the use of a non-safety chilled water subsystem
(which is not within the scope of the Technical Specifications) to
provide baseload cooling of the control room on a continuous basis.
Implementation of the modification to the CRACS subsystems
during the 60 day AOT duration does not result in a significant
reduction in the plant margin of safety. As previously identified,
the CRACS is a support subsystem and the existing Technical
Specifications will require a plant shutdown on a loss of function.
The risk significance involved with removing a safety-related train
of the CRACS is extremely low based on the short period (60 days per
train) and the consequences of losing this function. The potential
that the non-safety-related chilled water subsystem and the operable
safety-related train of CRACS simultaneously fail during the
proposed 60 day AOT period of each safety-related train (120 days
total) is considered unlikely. In the event that control room
temperature increased to a temperature greater than 75 deg.F, plant
procedures require starting either the non-safety-related chilled
water subsystem or a safety-related train of CRACS to restore
control room temperature. Additionally, in the unlikely event of a
loss of the non-safety related subsystem and the operable safety-
related train of the CRACS, Technical Specification 3.7.6.2 would
require that actions be taken to place the plant in a shutdown
condition. Alternative actions to reduce control room temperature
could also be initiated as identified in a plant procedure. It has
been conservatively determined that safety-related equipment in the
control room can be operated continuously in an environment up to
90 deg.F without affecting the capability of the equipment.
The exception to specifications 3.0.4 and 4.0.4 as stated in the
proposed one-time change to Technical Specification 3.7.6.2 will not
reduce the margin of safety. During the implementation of the
proposed modification, when one safety-related train of CRACS is
inoperable, it is possible that a plant shutdown could occur due to
reasons unrelated to the planned modifications of the CRACS. Entry
of the plant into an operational mode from a shutdown mode as a
result of the proposed modification does not adversely affect
accident initiators or precursors nor alter the design assumptions,
conditions, configuration of the facility or the manner in which the
plant is operated. The functions of the CRACS to provide a
controlled environment inside of the control room complex to ensure
the comfort of the plant operators and to ensure adequate climate
conditions for the operability of equipment will not be impaired in
any way as a result of a plant mode change. The remaining actions
identified in TS 3.7.6.2 are unchanged as a result of the proposed
change.
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.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Section Chief: James W. Clifford.
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: May 17, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) section 4.4.6.2.2.e to replace the
reference to American Society of Mechanical Engineers (ASME) Code
paragraph IWV-3472(b) which pertains to the frequency of leakage rate
testing for 6-inch, nominal pipe size valves and larger with the
requirement that the surveillance interval and frequency of
surveillance leakage rate testing for these valves be performed
pursuant to the requirements of TS 4.0.5, ``Operations and Surveillance
Requirements.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided the NRC 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 review is
presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Eliminating the reference to ASME Code paragraph IWV-3427(b) and
[[Page 38034]]
performing pressure isolation valve (PIV) testing pursuant to TS 4.0.5
does not change the test conditions for PIV leakage testing and is
consistent with the currently analyzed configurations. This change
eliminates an unnecessary test requirement and incorporates
Westinghouse Owner's Group (WOG) Standard Technical Specifications
(STS) frequency requirements that are deemed to substantially reduce
the probability of an intersystem loss-of-coolant-accident. This change
in testing frequency requirements does not affect the accident
mitigation capabilities of the reactor coolant system (RCS) PIVs. This
change is bounded by existing accident analyses. Therefore, it is
concluded that, with the reduced probability of previously analyzed
accidents, and no effect on accident mitigation, the proposed revision
does not involve 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.
Eliminating the IWV-3427(b) trending for 6-inch and larger valves
(and the accompanying increased frequency testing requirement) does not
significantly change actual testing frequencies since the frequencies
continue to be addressed by the remaining TS requirements. This change
does not affect the ability of a PIV to perform its required RCS
pressure isolation safety function of limiting RCS leakage to prevent
overpressure failure of attached low pressure systems. The frequency of
testing or the testing itself are not initiating events to postulated
accidents. Therefore, the proposed revision does not create the
possibility of a new or different kind of accident from any previously
evaluated.
3. Involve a significant reduction in the margin of safety.
There is no impact on the Margin of Safety as defined in the bases
of any TS, the Updated Final Safety Analysis Report, or other licensing
basis commitments resulting from the elimination of the reference to
ASME Code paragraph IWV-3427(b). Periodic surveillances provide
continued assurance in the capability of safety related equipment to
perform its design safety (accident mitigating) function and are not
used to establish the margin of safety for accident mitigation.
Therefore, the frequency of surveillance testing of the PIVs has no
impact on the margins of safety assumed in analyzed accidents.
In its evaluation, NNECO concluded, based on its evaluation as
required by 10 CFR 50.92, that the proposed revision does not involve a
SHC.
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.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment requests: May 13, 1999.
Description of amendment requests: The proposed amendments would
modify Technical Specification (TS) 6.2.A.2, ``Onsite and Offsite
Organizations,'' to reflect a change in the organizational structure
implemented on March 1, 1999.
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 amendment[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change is administrative in nature and does not
significantly affect any system that is a contributor to initiating
events for previously evaluated accidents. Neither does the change
significantly affect any system that is used to mitigate any
previously evaluated accidents. Therefore, the proposed change does
not involve any significant increase in the probability or
consequence of an accident previously evaluated.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
The proposed change is administrative in nature and does not
alter the design, function, or operation of any plant component nor
does the proposed change install any new or different equipment,
therefore the possibility of a new or different kind of accident
from those previously analyzed has not been created.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The proposed change is administrative in nature and does not
involve a significant reduction in the margin of safety associated
with the safety limits inherent in either the fuel cladding, RCS
[reactor coolant system] boundary, reactor containment, or other
structures, systems, or components (SSCs).
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.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Claudia M. Craig.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: June 7, 1999.
Description of amendment request: The proposed amendments, if
approved, would revise Technical Specifications (TS) Section 3/4.4.3
and its associated TS Bases to reflect changes to refine and clarify
the action statement concerning inoperable reactor coolant leakage
detection systems.
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 TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS changes directly establish the minimum
acceptable level of Reactor Coolant System (RCS) leakage detection
instrumentation required to support plant power operations. The
level of RCS leakage detection capability inherent with the proposed
TS change will continue to provide acceptable early warning
detection of potential RCS pressure boundary degradation as required
under 10 CFR 50.36 (c)(2)(ii) (A) Criterion 1.
Therefore, the proposed TS changes do not involve an 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 TS changes only affect systems associated with the
detection of accidents involving degradation of the RCS
[[Page 38035]]
pressure boundary. The proposed TS changes do not involve any
physical changes to plant structures, systems, or components. The
RCS Leakage Detection Systems will continue to function as designed
in all modes of operation. No new accident type is created as a
result of the proposed changes. No new failure mode for any
equipment is created. The changes are consistent with the guidance
provided in [Standard Technical Specifications General Electric
Plants BWR/4 dated April 1995] NUREG-1433, Revision 1, pertaining to
RCS Leakage Detection.
Therefore, the proposed TS changes do 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 TS Limiting Conditions for Operation (LCO) specify for
systems and equipment important to safety, the minimum level of
operability required to permit continued power operation. The
proposed TS changes revise this minimum level of operability by
permitting long term plant operation with the removal of the Drywell
Unit Coolers Condensate Flow Rate Monitoring System from service.
Currently, this condition would permit the plant to continue to
operate for up to 30 days. This change is not a reduction in the
margin of safety since:
The proposed Technical Specification LCO change for RCS Leakage
Detection Systems maintains four (4) diverse methods of detecting
RCS leakage and permits continuous operation with the Drywell Unit
Coolers Condensate Flow Rate Monitors out of service provided that
more frequent surveillance checks are provided for the containment
atmosphere monitoring system. The proposed TS change institutes the
additional surveillance requirements.
The LGS reactor coolant pressure boundary was designed to ASME
Class 1, Seismic Category I design criteria with no special
dispensation which would warrant such additional RCS leakage
detection capability or more stringent LCO criteria than those
generically approved under the Improved Standard Technical
Specifications.
Review of the TS Bases Section and UFSAR identified no
discussions regarding margin of safety for the RCS Leakage Detection
Systems, which would be reduced by the proposed Technical
Specification LCO change. It is further demonstrated that an
acceptable margin of safety exists based on the generic regulatory
approval of the Improved Standard Technical Specifications which
will remain bounded by the proposed LGS TS changes.
Therefore, the proposed TS 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.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
Portland General Electric Company, Docket No. 50-344, Trojan Nuclear
Plant, Columbia County, Oregon
Date of amendment request: August 27, 1998.
Description of amendment request: The proposed amendment would
revise the Facility Operating (Possession-Only) License and the
Permanently Defueled Technical Specifications. Multiple license
conditions and technical specification requirements are proposed to be
deleted to reflect the transfer of the nuclear spent fuel from the
existing 10 CFR Part 50 licensed area to the 10 CFR Part 72 Independent
Spent Fuel Storage Installation (ISFSI) area.
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.
This proposed amendment reflects removal of the spent nuclear
fuel from the 10 CFR 50 licensed area and transfer of the spent
nuclear fuel to the 10 CFR 72 ISFSI licensed area. The probability
and consequences of accidents associated with storage of spent
nuclear fuel within the TNP [Trojan Nuclear Plant] ISFSI were
evaluated as part of PGE's 10 CFR 72 license application. Following
completion of the transfer of the spent nuclear fuel to the 10 CFR
72 licensed ISFSI and in light of the revised Appendix A Technical
Specification, Section 4.2, that precludes storage of spent nuclear
fuel within the 10 CFR 50 licensed area, the potential for accidents
associated with the storage and handling of fuel in the 10 CFR 50
licensed area will be eliminated. Therefore, deleting those
technical specifications associated with spent nuclear fuel will not
result in a significant increase in the probability or consequences
of accidents previously analyzed.
The proposed license amendment also relocates administrative
requirements from Section 5.0 of the Technical Specifications to
topical report PGE-8010, ``TNP Nuclear Quality Assurance Program.''
Relocation of administrative requirements follows the guidance
provided in NRC Administrative Letter 95-06. Relocating these
administrative requirements will not result in changes in method of
operation of any plant equipment, therefore these changes will not
result in a significant increase in the probability or consequences
of accidents previously evaluated.
The proposed license amendment will delete the on duty shift
manning requirements (Technical Specification 5.2.2a). With removal
of the spent nuclear fuel from the 10 CFR 50 licensed area, there
are no remaining important to safety systems required to be
monitored. With removal of the spent nuclear fuel from the 10 CFR 50
licensed area, there are no remaining credible accidents which
require the actions of a Shift Manager or non-certified operator to
prevent occurrence or mitigate consequences. Therefore, deleting the
shift manning requirements will not result in an increase in the
probability or consequences of an accident previously analyzed.
Deleting the Independent Review and Audit Committee (IRAC) is
also proposed in this license amendment request. The responsibility
of IRAC is to review and advise the Plant General Manager on matters
relating to the safe storage of irradiated fuel. Since approval of
this license amendment request is contingent upon removal of the
spent nuclear fuel from the 10 CFR 50 licensed area and a revised
Technical Specification Section 4.2 prevents future storage,
deleting IRAC will not result in an increase in the probability or
consequences of an accident previously evaluated.
This license amendment request proposes to revise and relocate
License Condition 2.C.(8), Fire Protection, to the TNP Quality
Assurance Program (PGE-8010). The revised text removes requirements
associated with making changes that could adversely impact the safe
storage of irradiated fuel. Following removal of the spent nuclear
fuel from the 10 CFR 50 licensed area and implementation of the
proposed revision to Technical Specification Section 4.2, irradiated
fuel will not be stored within the 10 CFR 50 licensed area so this
change will not result in an increase in the probability of
occurrence or consequences of accidents previously analyzed.
Relocation of the remaining requirements contained in this license
condition to the TNP Quality Assurance Program (PGE-8010) will
provide the necessary administrative control to ensure that changes
to the fire protection program will not increase the likelihood of
an offsite release of radioactive material due to a fire. Therefore,
this change will not result in an increase in the probability of
occurrence or consequence of accidents previously analyzed.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed license amendment reflects the reduced operational
risks within the 10 CFR 50 licensed area after the spent nuclear
fuel has been transferred to the ISFSI. In addition, administrative
controls contained in Section 5.0 of the Technical Specifications
will [be] relocated to PGE-8010, TNP Nuclear Quality Assurance
Program. These changes have no impact on plant equipment and only an
administrative impact on some of the procedures used for operating
plant equipment, which may still be needed within the 10 CFR 50
licensed area following the
[[Page 38036]]
transfer of the spent nuclear fuel to the 10 CFR 72 ISFSI license
area. This proposed amendment does not result in the addition of new
equipment or result in the alteration of the operation of existing
structures, systems, or components. Therefore, the proposed changes
do 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 10 CFR 50 license amendment eliminates those
technical specifications and license conditions associated with the
storage of spent nuclear fuel. Following transfer of the spent
nuclear fuel to the 10 CFR 72 ISFSI, the potential for fuel related
accidents will be eliminated from the 10 CFR 50 licensed area.
Therefore, removal of those technical specifications and license
conditions associated with the safe storage of spent nuclear fuel
will not involve a significant reduction in a margin of safety.
Relocating administrative programs in Technical Specification,
Section 5.0, ``Administrative Controls,'' follows the guidance of
NRC Administrative Letter 95-06. With the exception of deleting
those administrative controls associated with storage of spent
nuclear fuel, the administrative programs will be relocated to the
TNP Quality Assurance Program (PGE-8010). This administrative
relocation of requirements does not involve a significant reduction
in a margin of safety.
This proposed amendment also requests deleting several license
conditions and relocating License Condition 2.C.(8), ``Fire
Protection.'' The deleted license conditions were related to either
power operations or activities which have been completed and are no
longer required. Relocating License Condition 2.C.(8), ``Fire
Protection,'' to the TNP Quality Assurance Program (PGE-8010) will
continue to maintain the required level of administrative control
for the fire protection program since changes to PGE-8010 are
controlled in accordance with the requirements of 10 CFR
50.54(a)(3). Deleting these license conditions will, therefore, not
result in a 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.
Local Public Document Room location: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Attorney for licensee: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
NRC Section Chief: Michael T. Masnik.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: June 7, 1999, as supplemented by letter
dated June 24, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 2.0, Safety Limits and Limiting
Safety System Settings, TS 3.2.5, DNB [Departure from Nucleate Boiling]
Parameters, and the associated Bases, and Administrative Controls
Section 6.9.1.6, Core Operating Limits Report (COLR), by relocating
cycle-specific reactor coolant system-related parameter limits from the
TSs to the COLR. This would allow for flexibility to enhance plant
operating margin and/or core design margins without the need for cycle-
specific license amendment requests.
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 change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed amendment is a programmatic and administrative
change that does not physically alter safety-related systems, nor
does it affect the way in which safety-related systems perform their
functions. Because the design of the facility and system operating
parameters are not being changed, the proposed amendment does not
involve an increase in the probability or consequences of any
accident previously evaluated.
The cycle-specific limits in the Core Operating Limits Report
will continue to be controlled by the STP [South Texas Project]
programs and procedures. Each accident analysis addressed in the
UFSAR [Updated Final Safety Analysis Report] will be examined with
respect to changes in the cycle-dependent parameters, which are
obtained from the use of NRC-approved reload design methodologies,
to ensure that the transient evaluation of new reloads are bounded
by previously accepted analyses. This examination, which will be
conducted per the requirements of 10CFR50.59, will ensure that
future reloads will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The safety limits imposed in Technical Specification 2.1.1.1 and
2.1.1.2 are consistent with the values stated in the STP Updated
Final Safety Analysis Report. The Reactor Coolant System Flow value
in the Technical Specifications will be changed from the Minimum
Measured Flow to the Thermal Design System Flow (approved by the
Nuclear Regulatory Commission in Amendments 97 and 84 on September
29, 1998) consistent with WCAP-14483-P-A [`Generic Methodology for
Expanding Core Operating Limits Reports']. This change does not
involve an increase in the probability or consequences of any
accident previously evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Removal of cycle specific variables has no influence or impact
on, nor does it contribute in any way to the probability or
consequences of an accident. No safety-related equipment, safety
function, or plant operation will be altered as a result of this
proposed change. The cycle specific variables are calculated using
the NRC-approved methods, and submitted to the NRC to allow the
staff to continue to trend the values of these limits. The Technical
Specifications will continue to require operation within the core
operating limits, and appropriate actions will be required if these
limits are exceeded. The safety limits imposed in Technical
Specification 2.1.1.1 and 2.1.1.2 are consistent with the values
stated in the STP Updated Final Safety Analysis Report. The Reactor
Coolant System Flow value in the Technical Specifications will be
changed from the Minimum Measured Flow to the Thermal Design Flow
(approved by the Nuclear Regulatory Commission in Amendments 97 and
84 on September 29, 1998) consistent with WCAP-14483-P-A. This
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change does not involve a significant reduction in
a margin of safety.
The margin of safety is not affected by the removal of cycle
specific core operating limits from the Technical Specifications.
The margin of safety presently provided by current Technical
Specifications remains unchanged. Appropriate measures exist to
control the values of these cycle specific limits. The proposed
amendment continues to require operation within the core limits as
obtained from NRC-approved reload design methodologies, and the
actions to be taken if a limit is exceeded remain unchanged.
The development of the limits for future reloads will continue
to conform to those methods described in NRC-approved documentation.
In addition, each future reload will involve a 10CFR50.59 safety
review to assure that operation of the unit within cycle-specific
limits will not involve a significant reduction in the margin of
safety.
The safety limits imposed in Technical Specification 2.1.1.1 and
2.1.1.2 are consistent with the values stated in the STP Updated
Final Safety Analysis Report. The Reactor Coolant System Flow value
in the Technical Specifications will be changed from the Minimum
Measured Flow to the Thermal Design System Flow (approved by the
Nuclear Regulatory Commission in Amendments 97 and 84 on September
29, 1998) consistent with WCAP-14483-P-A. This proposed change does
not involve a significant reduction in the margin of safety.
[[Page 38037]]
The proposed amendment is a programmatic and administrative
change that provides assurance that plant operations continue to be
conducted in a safe manner. As stated previously, the proposed
amendment does not physically alter safety-related systems, nor does
it affect the way in which safety-related systems perform their
functions. Because the design of the facility and system operating
parameters are not being changed, the proposed amendment 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas
77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
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: June 3, 1999.
Description of amendment request: The proposed amendment would
modify the Technical Specifications to reduce the Allowable Value (Av)
used for Reactor Vessel Water Level--Low, Level 3 for several
instrument functions.
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 Reactor Vessel Water Level--Low, Level 3 functions are in
response to water level transients and are not involved in the
initiation of accidents or transients. Therefore, reducing the Level
3 Av does not increase the probability of an accident previously
evaluated. Additionally, the results of the safety evaluation
associated with the lowering of the Level 3 Av concludes that the
previously evaluated transient and accident consequences are not
significantly affected by the change. 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.
The proposed amendment to lower the BFN Units 2 and 3 Reactor
Vessel Water Level--Low, Level 3 Av does not involve a hardware
change and the purpose of the Level 3 function is not affected. The
Level 3 functions will continue to fulfill their design objective.
Therefore, reduction of the Av 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 results of the safety evaluation associated with the
reducing the BFN Units 2 and 3 Reactor Vessel Water Level--Low,
Level 3 Av concluded that transient and accident consequences remain
within the required acceptance criteria. Therefore, the margin of
safety is not reduced for any event evaluated.
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.
Local Public Document Room location: Athens Public Library, 405 E.
South Street, Athens, Alabama 35611.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Sheri R. Peterson.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station (CPSES), Units 1 and 2, Somervell County, Texas
Date of amendment request: June 23, 1999.
Brief description of amendments: The proposed license amendments
would change the way in which the Emergency Diesel Generator (EDG)
automatic trips are tested in Surveillance Requirement (SR) 3.8.1.13. A
note would also be added to specify the CPSES, Unit 2, test schedule in
SR 3.8.1.13.
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?
The emergency diesel generators are used to support mitigation
of the consequences of an accident and are not considered to be
initiator of any previously analyzed accident. Revising the
surveillance to verify the bypass of non-critical EDG trips on both
LOOP [loss of offsite power] and SI [safety injection] separately
enhances the ability of the EDG to perform its safety function by
ensuring continued operation during DBAs [design-basis accidents].
Therefore, this change will not result in an increase in the
probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change to the surveillance requirement involves an
EDG start circuit modification. The circuit modification has been
previously installed on Unit 2 during 2RF04 [CPSES Unit 2, fourth
refueling outage] for reasons other than the issue associated with
the FWLB [feedwater line break]. As a part of the Unit 2
installation a 50.59 evaluation was performed and it was determined
that the modification did not represent an unreviewed safety
question. The modification similar to Unit 2 will be implemented on
Unit 1 and therefore, as concluded in the safety evaluation for the
original modification, no new failure mechanisms will be introduced
by the proposed change. The EDGs are designed to provide electrical
power to equipment important to safety in the event of a loss of
offsite power. The proposed change to the SR enhances the confidence
that the EDGs will start and fulfill their safety related function.
Therefore, this change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed change will not alter any accident analysis
assumptions, initial conditions, or results. Revising the
surveillance requirement to verify the EDG trip bypass for the LOOP
and SI separately will enhance the confidence that the EDG starts as
assumed in the safety analyses and does not create any new failure
scenarios and no margin is reduced.
Therefore, this 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gram.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: April 16, 1999, as superseded on June 9,
1999.
[[Page 38038]]
Description of amendment request: The licensee proposed clarifying
the inservice inspection requirements for Vermont Yankee Nuclear Power
Station regarding the granting of relief from ASME Code requirements by
the NRC. The licensee also proposed changes to reflect the previous NRC
approval of the use of ASME Code Case N-560 at Vermont Yankee Nuclear
Power Station.
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 operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated?
This change is only an administrative change that: (1) clarifies
the NRC's authority to grant relief to a specific requirement, and
(2) conforms the TS language regarding GL 88-01 to agree with the
NRC's acceptance of ASME Code Case N-560 for use at VY. This
conclusion is justified in that:
(a) The pursuit of relief from the ASME code and the imposition
of alternative requirements are governed by 10CFR50.55a and require
NRC approval. There are several sections in the regulations under
which such relief can be granted. The removal of reference to a
specific section of CFR that may be used to grant relief has no
effect on plant equipment or its operation.
(b) Adding words to clarify the relationship between GL 88-01
and Code Case N-560 eliminates a contradiction in sample selection
criteria and does not affect any equipment or its operation.
These changes can be considered administrative in nature and do
not change any of the accident analyses for the facility. Thus,
there are no changes to the probability or consequences of accidents
previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The revision of the wording in the TS to generalize the granting
of relief to the ASME code does not result in any changes to the
plant equipment or its operation. Similarly, adding words to allow
use of the NRC-approved alternative to the sample selection guidance
provided in GL 88-01 does not impact plant equipment or its
operation. These changes are administrative in nature and do not
result in the creation of any new or different kinds of accidents.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
This change primarily revises the wording in the TS to clarify
the NRC's authority to grant relief to ASME Section XI requirements.
The change maintains the requirement for NRC approval to be obtained
for such relief. Secondly, this change conforms the TS language
regarding GL 98-01 to agree with a previous relevant NRC disposition
[Reference (e)]. [The staff notes that reference (e) is an NRC
letter dated November 9, 1999, which approved the use of Code Case
N-560 at Vermont Yankee Nuclear Power Station.] These administrative
changes do not result in a 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.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: June 24, 1999
Description of amendment request: The amendment clarifies the basis
for the reactor protection system bypass of the turbine stop valve
(TSV) closure and turbine control valve (TCV) fast closure scram
signals at low power. The amendment clarifies that the analytical basis
for this bypass corresponds to a fraction of reactor rated thermal
power and not other measures of power, for instance, turbine 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. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change clarifies the basis for the reactor
protection system bypass of the turbine stop valve closure and
turbine control valve fast closure scram signals. Consideration of
the bypass function itself only applies to certain pressurization
transients and not accident analyses.
The change properly states the basis for the scram bypass and
relates it to reactor thermal power and precludes potential
misinterpretation of the basis for the bypass setpoint. Turbine
power lags reactor power over the range of concern. Therefore,
changing terminology related to ``power'' to mean ``reactor power''
instead of ``turbine power'' is conservative. Accordingly, this
change can not be less restrictive.
The low power (TSV closure and TCV fast closure) scram signal
bypass does not initiate or mitigate any accident considered in the
Updated Final Safety Analysis Report. This function is enabled at
higher power to mitigate the effects of the pressurization transient
which results from TSV closure or TCV fast closure. This change will
not alter assumptions relative to the initiation or mitigation of
any accident event.
This change will not involve a significant increase in the
probability or consequences of an accident previously evaluated
since there is no physical alteration of the plant configuration or
relaxation of setpoints or operating parameters.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The reactor protection system bypass of the turbine stop valve
closure and turbine control valve fast closure scram signals is not
considered an initiator of any accident. This change to clarify the
basis for applicability of the bypass does not create any new or
different kind of accident since it does not involve any change in
the physical configuration of the plant, nor relaxation of setpoints
or operating parameters.
VY has determined that the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated because the change merely adds a more
restrictive interpretation to current terminology.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
The change involves reducing the potential for misinterpreting
the basis for the reactor protection system bypass of the turbine
stop valve closure and turbine control valve fast closure scram
signals and consequent potential for nonconservative operation of
the plant. As a result, the potential for operation of the plant in
an unsafe condition is reduced, thereby maintaining the margin of
safety.
VY has determined that the proposed change does not involve a
significant reduction in a margin of safety since operation of the
plant consistent with analytical bases of operation is further
assured.
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.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
[[Page 38039]]
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
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 at the local public document rooms for
the particular facilities involved.
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 application for amendments: March 30, 1999.
Brief description of amendments: The amendments revised license
conditions in each of the operating licenses to delete those license
conditions that no longer apply, make an editorial change in the Unit 1
license, and provide clarifying information regarding the license
condition in each license concerning equalizer valve restrictions.
Date of issuance: June 25, 1999.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 188 & 185.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the licenses.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24195).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 25, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
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: March 30, 1999.
Brief description of amendment: The amendment adds Section 4.0.2 to
allow a 24-hour grace period for performing inadvertently missed
surveillance.
Date of issuance: June 25, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 202.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27317).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 25, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
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: March 30, 1999.
Brief description of amendment: The amendment adds Section 4.0.2 to
allow a 24-hour grace period for performing inadvertently missed
surveillance.
Date of issuance: June 25, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 202.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27317).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 25, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: August 6, 1998, as supplemented
by letter dated May 18, 1999.
Brief description of amendment: The amendment approves a change to
Technical Specification (TS) 3.1.3.2, ``Position Indicator Channels--
Operating,'' which adopts requirements that are consistent with NUREG-
1432, ``Standard Technical Specifications for Combustion Engineering
Plants.'' In addition, the amendment approves the relocation of TS
Table 3.8-1, ``Containment Penetration Conductor Overcurrent Protective
Devices,'' to licensee control procedures in accordance with the
guidance provided in Generic Letter 91-08, ``Removal of Component Lists
From Technical Specifications.''
Date of issuance: June 29, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 208.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56245).
The May 18, 1999, letter provided clarifying information that did
not change the scope of the original application and the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 29, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 17, 1996, as supplemented by
letters dated October 22, 1998, and January 12 and February 5, 1999.
[[Page 38040]]
Brief description of amendment: The amendment extends the
surveillance test interval for the reactor trip circuit breakers from
monthly to quarterly and revises the appropriate Bases page.
Date of issuance: June 29, 1999.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 153.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48261).
The October 22, 1998, and January 12 and February 5, 1999, letters
provided additional information that did not extend the scope of the
original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 29, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: October 27, 1998, supplemented
March 19, 1999.
Brief description of amendment: This amendment relocates a TS
surveillance requirement from TS Section /4.6.5.1, ``Shield Building--
Emergency Ventilation System'' to TS Section 3/4.6.5.2, ``Shield
Building Integrity.'' Administrative and bases changes have also been
made.
Date of issuance: June 22, 1999.
Effective date: June 22, 1999.
Amendment No.: 233.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64125). The March 19, 1999, supplement to the application did not
expand the scope of the original application as noticed, and did not
change the staff's proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 22, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: August 31, 1998, as revised on
March 18, 1999.
Brief description of amendment: The amendment approves changes to
the Improved Technical Specifications to allow a repair roll process
which would be used to repair steam generator tubes with defects within
the upper tubesheet. Changes to inservice inspection and reporting
requirements and several format and editorial changes were also
included.
Date of issuance: June 28, 1999.
Effective date: As of date of issuance, to be implemented prior to
commencing Cycle 12 operation.
Amendment No.: 179.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56249). The revised submittal dated March 18, 1999, expanded the scope
of the amendment request as originally noticed, and the application was
renoticed on April 21, 1999 (64 FR 19557).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: February 7, 1997, as
supplemented October 24, 1998
Brief description of amendment: The amendment incorporates changes
to more accurately reflect current plant design, adopts changes in
surveillance requirements consistent with the Standard Technical
Specifications, identifies changes to plant systems and revisions to
Technical Specifications system descriptions not involving Limiting
Conditions for Operations, and makes editorial or typographical
corrections.
Date of issuance: June 21, 1999.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 212.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 25, 1998 (63 FR
14486) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 21, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: June 11, 1998.
Brief description of amendment: The amendment revises Technical
Specification 6.12.1 to allow use of an alternative high radiation area
control consistent with Regulatory Guide 8.38.
Date of issuance: July 1, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 213.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43204) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-336,
and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New
London County, Connecticut
Date of application for amendment: December 22, 1998, as
supplemented March 19, 1999.
Brief description of amendment: The amendment replaces specific
titles in Section 6.0 of the Technical Specifications of all three
Millstone units with generic titles.
Date of issuance: June 3, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 105, 235, and 171.
[[Page 38041]]
Facility Operating License Nos. DPR-21, DPR-65, and NPF-49:
Amendment revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4158). The March 19, 1999, letter provided clarifying information that
did not change the scope of the December 22, 1998, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: January 4, 1999, as supplemented
April 7, 1999.
Brief description of amendment: The amendment changes Technical
Specifications 3.5.2, ``Emergency Core Cooling Systems--ECCS
Subsystems--Tavg greater than or less than 300 deg.F;'' 3.6.2.1,
``Containment Systems--Depressurization and Cooling Systems--
Containment Spray and Cooling Systems;'' 3.7.1.2, ``Plant Systems--
Auxiliary Feedwater Pumps;'' 3.7.3.1, ``Plant Systems--Reactor Building
Closed Cooling Water System;'' and 3.7.4.1, ``Plant Systems--Service
Water System.'' The changes were made to the system pump flow
requirements to incorporate the results of revised hydraulic and
accident analyses.
Date of issuance: June 29, 1999.
Effective date: As of the date of issuance and shall implemented
within 60 days from the date of issuance.
Amendment No.: 236.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 14, 1999 (64 FR
2523).
The April 7, 1999, supplemental letter 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 June 29, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 9, 1999, as supplemented May 3,
1999
Brief description of amendments: The amendments delete the
requirement to have an independent safety engineering group (ISEG) from
the Technical Specifications and applies the substantive requirements
now applicable to the ISEG to other organizations and relocates those
requirements from the Technical Specifications to Chapter 16 of the
Operational Quality Assurance Plan (OQAP). In the letter of May 3,
1999, the licensee submitted the changes to Chapter 16 of the OQAP to
incorporate the substantive Technical Specification requirements
currently applicable to the ISEG into the OQAP in the form of an
independent technical review program, and stated that these changes to
the OQAP will become effective upon approval of the amendments.
Date of issuance: June 23, 1999. Effective date: June 23, 1999, to
be implemented within 30 days. Implementation includes incorporating
the OQAP pages into the OQAP.
Amendment Nos.: Unit 1-112 ; Unit 2-99.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 7, 1999 (64 FR
17030) The May 3, 1999, supplement provided additional clarifying
information within the scope of the original notice 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 June 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
TU Electric Company, 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 27, 1999, as supplemented May 28,
1999.
Brief description of amendments: The amendments add a footnote to
Technical Specfiication 4.8.2.1e, ``D.C. Sources--Operating,'' which
would, on a one-time basis for Unit 1 Battery BT1ED2, allow TU Electric
to substitute a performance discharge test ``...in lieu of the battery
service test required by Specification 4.8.2.1d, twice within a 60
month interval.''
Date of issuance: June 28, 1999.
Effective date: As of the date of issuance.
Amendment Nos.: 65 and 65
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (64 FR 31881 dated June 14, 1999). The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided an opportunity to request a hearing by July 14, 1999, but
indicated that if the Commission makes a final determination, any such
hearing would take place after issuance of the amendments.
The May 28, 1999, letter provided clarifying information that did
not change the scope of the original application and the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and final NSHC determination are contained in
Safety Evaluation dated June 28, 1999.
Attorney for Licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC, 20036.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: February 16, 1999.
Brief Description of amendments: These amendments revise TS Section
4.2 for Units 1 and 2. The changes relax the surveillance requirements
for reactor coolant pump (RCP) flywheels. The flywheels provide
extended reactor coolant flow coastdown capability if electric power
for the RCPs is lost. Previously, the flywheel inspections included an
ultrasonic examination (UT) of areas of high stress
[[Page 38042]]
concentration at the base and keyway every 3 years, and complete UT
every 10 years. The changes require only a 10-year UT based upon an
analysis presented in a Westinghouse topical report which has been
reviewed and accepted by the NRC staff.
Date of issuance: July 1, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 221 and 221.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24204). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, 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 at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By August 13, 1999, 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 at the local public document room for the
particular facility involved. 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
[[Page 38043]]
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. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
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: Rulemakings and
Adjudications Staff 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Unit 2, Matagorda County, Texas
Date of amendment request: July 1, 1999.
Brief description of amendment: The amendment provides for a one-
time change to Technical Specifications 3.3.2 and 3.7.8 for Unit 2 to
allow all fuel handling building exhaust air system components to be
inoperable for a period not to exceed 8 hours to facilitate repair of
the Train B exhaust booster fan.
Date of issuance: July 2, 1999.
Effective date: From the date of amendment issuance until July 14,
1999.
Amendment No.: Unit 2-100.
Facility Operating License No. NPF-80: The amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No. The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated July 2, 1999.
Local Public Document Room location: Wharton County Junior College,
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
Dated at Rockville, Maryland, this 7th day of July 1999.
For The Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-17750 Filed 7-13-99; 8:45 am]
BILLING CODE 7590-01-P