[Federal Register Volume 62, Number 1 (Thursday, January 2, 1997)]
[Notices]
[Pages 121-136]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10102]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 9, 1996, through December 19, 1996.
The last biweekly notice was published on December 18, 1996.
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
[[Page 122]]
proposed amendment would not (1) involve a significant increase in the
probability or consequences of an accident previously evaluated; or (2)
create the possibility of a new or different kind of accident from any
accident previously evaluated; or (3) involve a significant reduction
in a margin of safety. The basis for this proposed determination for
each amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By February 3, 1997, 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 petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and
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telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. 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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of amendments request: November 6, 1996
Description of amendments request: The proposed amendment would
modify the technical specifications (TS) to require manual blocking of
one train of fast bus transfer (FBT) within the first hour of degraded
switchyard voltage should the switchyard voltage fall below the level
necessary for the electrical distribution system (EDS) degraded voltage
protection to maintain compliance with General Design Criteria (GDC)
17. The proposed amendment would further require the starting,
paralleling with the grid, loading, and then separating from the grid
the other train's emergency diesel generator (EDG) within the first
hour, rather than the current TS which allows two hours after onset of
a degraded switchyard voltage condition to start the EDG.
Alternatively, fast bus transfer can be blocked in both trains within
the first hour. The proposed amendment includes changes to the
applicable notes to reflect that these changes are no longer temporary,
but will remain as part of the long-term solution to this issue.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration. The
NRC staff's analysis is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change reduces the amount of time the second train of
electrical equipment is allowed to remain in nonconformance with GDC 17
in the TS action statement. This change only affects equipment used to
mitigate an event, and does not affect equipment assumed to initiate
any event. Thus the probability of an accident previously evaluated is
not affected.
The proposed change brings the second EDS train into compliance
with GDC 17 at least one hour sooner than the current TS. Once in
conformance with GDC 17, the consequences of accidents previously
evaluated conform to the current analysis. Thus the proposed change
does not increase the consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change only affects equipment designed to mitigate the
effects of an accident. The proposed change ensures that safety
equipment is configured as assumed in the current accident analysis.
The proposed change does not affect the conditions of structures,
systems, or components assumed in the safety analysis beyond the
existing design basis as maintained by the current TS. The proposed
change does not, therefore, 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 margin of safety affected by the proposed change is based on
calculated offsite dose consequences for postulated transients and
accidents for which the EDS provides power for equipment required to
mitigate. The proposed change reduces the time that one train of the
EDS is allowed to remain in nonconformance with GDC 17, thus
increasing the availability of the EDS prior to the onset of a
postulated accident compared to the current TS. Thus the proposed
change does not increase thecalculated offsite dose, and therefore
the proposed change does not involve a significant reduction in a
margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendments request involve no significant hazards
consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: William H. Bateman
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: November 26, 1996
Description of amendments request: The proposed amendment will
adopt Option B of 10 CFR Part 50, Appendix J, to require Type B and
Type C containment leakage rate testing to be performed on a
performance-based testing schedule. Containment leakage rate testing is
currently performed in accordance with 10 CFR Part 50, Appendix J,
Option A, ``Primary Reactor Containment Leakage Testing for Water-
Cooled Power Reactors.'' Appendix J specifies containment leakage
testing requirements, including the types of tests required, frequency
of testing, and reporting requirements. Containment leakage test
requirements include performance of Integrated Leakage Rate Tests, also
known as Type A tests, which measure overall leakage rate of the
containment; and Local Leakage Rate Tests, also known as Types B and C
tests, which measure the leakage through containment penetrations and
valves. The Nuclear Regulatory Commission (NRC) has amended the
regulations to provide an alternate performance-based option, Option B,
to the existing Appendix J. Baltimore Gas and Electric Company (BGE)
received approval to adopt Option B for Type A testing only. At this
time, BGE plans to adopt Option B for Types B and C testing, as well.
BGE is revising the Containment Leakage Rate Testing Program for
Type A testing to implement Types B and C testing of the containment as
required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B.
The revised program will be developed in accordance with the guidelines
contained in Regulatory Guide 1.163 ``Performance-Based Containment
Leak-Rate Test Program,'' dated September 1995, including errata.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
[[Page 124]]
Containment leakage rate testing is performed in accordance with
10 CFR Part 50, Appendix J, ``Primary Reactor Containment Leakage
Testing for Water-Cooled Power Reactors.'' The Appendix J
containment leakage test requirements include performance of Type A
tests, which measure the overall leakage rate of the containment,
and Types B and C tests, which measure the leakage through
containment penetrations and valves. The Nuclear Regulatory
Commission has amended the regulations to provide a performance-
based alternative, Option B, to the existing Appendix J. Baltimore
Gas and Electric Company adopted Option B for Type A testing during
the Unit 1 refueling outage earlier this year. At this time, BGE
plans to adopt Option B for Types B and C testing.
Implementation of Option B involves no physical or operational
changes to the plant structures, systems, or components.
Furthermore, leakage rate does not contribute to the initiation of
any postulated accidents; therefore, this proposed change does not
involve an increase in the probability of any previously evaluated
accidents.
Types B and C testing is necessary to demonstrate that leakage
through the containment penetrations is within the limits assumed in
the accident analyses. The only potential effect of the proposed
change to the Types B and C test frequency is the possibility that
containment penetration leakage would go undetected between tests.
To provide assurance that containment penetration leakage remains
within the limits of the Technical Specifications, BGE plans to
implement the performance-based leakage testing program in
accordance with NRC Regulatory Guide 1.163, dated September 1995
(including errata), with no exceptions.
By adopting Option B, BGE will no longer require an exemption
from 10 CFR Part 50, Appendix J, which was granted to accommodate
24-month operating cycles. The exemption increased the surveillance
interval to a maximum of 30 months, while proportionately decreasing
the combined Types B and C leakage rate acceptance criteria. Option
B to Appendix J provides the regulation necessary to accommodate an
extended fuel cycle, while maintaining the original combined Types B
and C leakage rate testing limit. Therefore, BGE has requested
revocation of the exemption to 10 CFR Part 50, Appendix J, as
adoption of Option B for Types B and C testing will enable a return
to full compliance with Appendix J. As the facility will be in full
compliance with the regulations, this change does not increase the
consequences of any previously evaluated accidents.
Implementation of Option B does not change the total allowable
containment leakage rate acceptance criteria, nor does it change the
total leakage assumed in the accident analyses. Option B allows the
implementation of a performance-based testing program to ensure that
resources are concentrated on the components most likely to exceed
administrative limits. Similarly, the changes to relocate the
procedural details, including test frequency, performance and data
conversion methodology, for containment leakage rate testing from
the Technical Specifications to the Containment Leakage Rate Testing
Program will have no effect on the total containment leakage allowed
by the Technical Specifications, or assumed in the accident
analyses. Relocating the allowable leakage rate conversions
(Standard Cubic Centimeters per Minute) to the Technical
Specification Bases does not change the allowable leakage rates (as
a percentage of the containment air volume) specified in the
Technical Specifications. Furthermore, relocation of the
programmatic controls for Types B and C testing, including the
allowable leakage rates, to the Administrative Controls section of
the Technical Specifications ensures an adequate level of regulatory
control of these criteria is retained.
Additionally, the Calvert Cliffs Individual Plant Examination
considered the effects associated with severe accidents which could
lead to containment failure. It was concluded that adopting a
performance-based testing interval will not significantly affect the
containment failure probabilities calculated for the Individual
Plant Examination. Altogether, adoption of a performance-based
testing frequency, as specified in 10 CFR Part 50, Appendix J,
Option B, will not significantly decrease the confidence in the
leak-tightness of the containment, including containment
penetrations. Therefore, this change will not result in a
significant increase in the probability of undetected containment
penetration leakage in excess of that allowed by the Containment
Leakage Rate Testing Program, or assumed in the accident analysis,
or in the consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed Technical Specification change adopts a
performance-based approach to containment penetration leakage rate
testing. This change does not add any new equipment, modify any
interfaces with any existing equipment, or change the equipment's
function, or the method of operating the equipment. The proposed
change does not affect normal plant operations or configuration, nor
does it affect leakage rate test methods. As the proposed change
would not change the design, configuration or operation of the
plant, it could not cause containment penetration leakage rate
testing to become an accident initiator.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The purpose of the existing schedule for Types B and C tests is
to provide assurance, on a regular basis, that the release of
radioactive material will be restricted to those leak paths and
leakage rates assumed in the accident analyses. The margin of safety
associated with containment penetration leakage rates is not reduced
if containment leakage does not exceed the maximum allowable leakage
rate defined in the Technical Specifications. Implementation of
Option B does not change the total allowable containment leakage
rate acceptance criteria, nor does it change the total leakage
assumed in the accident analyses. Option B only allows the
implementation of a performance-based testing program to ensure that
resources are concentrated on the components most likely to exceed
administrative limits. Similarly, the changes to relocate the
procedural details for containment leakage rate testing from the
Technical Specifications to either the Containment Leakage Rate
Testing Program or the Technical Specification Bases will have no
effect on the total containment leakage allowed by the Technical
Specifications, or assumed in the accident analyses. Furthermore,
relocation of the programmatic controls for Types B and C testing,
including the allowable leakage rates, to the Administrative
Controls section of the Technical Specifications ensures that the
same regulatory control of these criteria is retained.
Elimination of the exemption to Appendix J which reduced the
amount of combined Types B and C testing allowable leakage
redistributes that portion of the total containment leakage which
may be attributed to local leakage rate testing, but does not affect
the maximum allowable containment leakage rate, La. The
proposed change does not affect a safety limit, a Limiting Condition
for Operation, or the way in which the plant is operated.
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
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: S. Singh Bajwa, Acting Director
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: December 2, 1996 (NRC-96-0134)
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.1.4.3, Rod Block Monitor, and
Tables 3.3.6-1 and 4.3.6-1 in TS 3.3.6, Control Rod Block
Instrumentation, to expand the range of conditions under which the rod
block monitor must be operable. These changes are required to ensure
that all fuel limits are met for the core that has been loaded for
Cycle 6.
[[Page 125]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes provide requirements that are more
restrictive than the existing requirements for operation of the
facility. These changes provide assurance that the Rod Block Monitor
system is operable when necessary to prevent or mitigate transients
that could potentially threaten the integrity of the fuel cladding.
There will be no adverse impact on the probability of any accident
previously evaluated since the change provides additional assurance
that fuel thermal and mechanical design bases will be satisfied and
has no effect on any accident initiating mechanism. The additional
restrictive conditions on plant operation also ensure that the
consequences of anticipated operational occurrences are no more
severe than the most limiting conditions using the current Technical
Specifications. Therefore these changes do not involve any increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed changes will not involve any physical changes to
plant systems, structures, or components (SSC). The changes in Rod
Block Monitor operability requirements are consistent with the
current safety analysis assumptions. These requirements provide
assurance that the Rod Block Monitor will be operable if necessary
to terminate a rod withdrawal error so that fuel thermal and
mechanical design limits are satisfied. The change does not cause a
physical change to the plant or introduce a new mode of operation.
Therefore, the proposed amendment will not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. These changes maintain current assumptions within the safety
analyses and design basis. The changes provide assurance that the
Rod Block Monitor will be operable if necessary to terminate a rod
withdrawal error so that fuel thermal and mechanical design limits
are satisfied. Therefore, these changes do not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226
NRC Project Director: John N. Hannon
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: November 6, 1996
Description of amendment request: The proposed amendment would
revise the technical specifications to permit an increase in the
allowable leak rate for the Main Steam Isolation Valves (MSIVs) and
delete the Penetration Valve Leakage Control System (PVLCS) and Main
Steam-Positive Leakage Control System (MS-PLCS) 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:
(1) The operation of River Bend Station, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of any accident previously evaluated.
The proposed amendment to delete Technical Specification 3.6.1.8
and 3.6.1.9 involves eliminating the PVLCS and MS-PLCS leakage
control requirements from the Technical Specifications. As described
in Sections 9.3 and 6.7 respectively, of the USAR [Updated Safety
Analysis Report], the PVLCS and MS-PLCS are manually initialed about
20 minutes following a design basis LOCA [Loss of Coolant Accident].
Since the PVLCS and MS-PLCS are operated only after an accident
has occurred, this proposed amendment has no effect on the
probability of an accident.
Since MSIV leakage and operation of the PVLCS and MS-PLCS are
included in the radiological analysis for the design basis LOCA as
described in Section 15.6.5 of the USAR, the proposed amendments
will not affect the precursors of other analyzed accidents. The
PVLCS and MS-PLCS are not initiators of any previously analyzed
accident. The proposed amendments result in acceptable radiological
consequences of the design basis LOCA previously evaluated in
Section 15.6.5 of the USAR.
The proposed amendment to Technical Specification 3.6.1.3 does
not involve a change to structures, components or systems that would
affect the probability of an accident previously evaluated. A plant-
specific radiological analysis has been performed to assess the
affects of the proposed increase to the allowable MSIV leak rate and
deletion of the PVLCS and MS-PLCS in terms of Control Room and off-
site doses following a postulated design basis LOCA. This change
required a revision to the existing LOCA dose analysis due to the
potential leakage from the MSIVs and those valves served by the
PVLCS. Additional changes were also included in the revised dose
analysis to account for changes in regulatory guidance and dose
methodology. Leakage from the drywell to the atmosphere through the
PVLCS (secondary containment bypass valves) are both assumed to
begin at time zero. The model conservatively assumes that one
inboard MSIV fails open at time zero and the MSIVs associated with
the remaining three main steam lines are assumed to begin leakage at
2 hours with a total leak rate of 200 scfh for all four main steam
lines. The design basis leak rate of the primary containment
(excluding main steam lines and lines sealed by the PVLCS) is 0.26%
of the containment volume by weight per 24 hours for the duration of
the accident and is assumed to be released entirely to the
environment initially or the secondary containment later into the
accident. The leakage of 170,000 cc/hr (4298 sccm) at Pa
through the containment isolation valves served by the PVLCS is
considered as bypass leakage circumventing the secondary
containment. The on-site and off-site doses were determined using
the TRANSACT computer code which included the ICRP 30 dose
conversion factors. The total off-site and on-site LOCA doses for
both the airborne and liquid release pathways resulting from the
proposed change are bounded by the applicable regulatory limits.
The analysis demonstrates that dose contributions from the
proposed combined MSIV leakage rate limit of 200 scfh and from the
proposed deletion of the PVLCS and MS-PLCS result in values bounded
by the applicable regulatory limits as compared to the LOCA doses
previously evaluated for the off-site and Control Room doses as
contained in 10CFR100 and 10CFR50, Appendix A (General Design
Criteria 19), respectively. The LOCA doses previously evaluated are
discussed in Section 15.6.5 of the USAR.
The whole body (DDE [Deep Dose Equivalent]) doses at the Low
Population Zone (LPZ) is 2.82 Rem and the Control Room is 0.43 Rem.
These values are acceptable since the revised doses are bounded by
the Regulatory Guidelines (2.82 versus 25 Rem at the LPZ and 0.43
versus 5 Rem at the Control Room). The associated whole boy (DDE)
dose at the exclusion area boundary (EAB) is 4.69 Rem which also
remains bounded by the Regulatory Guideline of 25 Rem.
The thyroid CEDE [Committed Effective Dose Equivalent] dose at
the LPZ is 62.58 Rem. This is acceptable since the revised dose of
62.58 Rem is significantly less than the Regulatory Guideline (300
Rem). The EAB thyroid CEDE dose is 37.53 Rem, whereas the Control
Room thyroid CEDE dose is 11.18 Rem. These values are also
acceptable since the revised doses are well within the Regulatory
Guidelines (37.53 versus 300 Rem at the EAB and 11.18 versus 30 Rem
at the Control Room). The Control Room beta (SDE [Shallow Dose
Equivalent]) dose is 9.15 Rem which also remains bounded by the
Regulatory Guideline of 30 Rem.
In summary, the proposed changes do not result in an increase to
the radiological consequences of a LOCA previously evaluated in the
USAR. The revised LOCA doses are bounded by the Regulatory
Guidelines. The effectiveness of the proposed request even for
leakage rates greater than the
[[Page 126]]
proposed MSIV allowable leak rate ensures that off-site and Control
Room dose limits are not exceeded.
There is no physical change to the ADS/SRVs [Automatic
Depressurization System/Safety Relief Valve]. The PVLCS accumulator
tanks remain the backup air supply to the ADS/SRV accumulators. A
qualified long-term backup air supply remains but is supplied from a
difference source. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change deletes the requirements for the LCS
[Leakage Control System] isolation valves which are non-PCIVs. These
valves are eliminated and will not be performing a safety function.
The LCS lines that are connected to the PCIVs and process piping
will be welded and/or capped closed to assure primary containment
integrity is maintained. The welding and post-weld examination
procedures will be in accordance with the American Society of
Mechanical Engineers (ASME) Code, Section XI requirements. These
welds and/or caps will be periodically tested as part of the primary
Containment Integrated Leak Rate Test (CILRT) program in accordance
with the requirements of 10CFR50, Appendix J. The proposed change
does not involve an increase in the probability of equipment
malfunction previously evaluated in the USAR. In fact, the proposed
change reduces the probability of equipment malfunction since, upon
implementation, RBS will be operated with fewer process line
isolation valves and associated support equipment subjected to
postulated failure. The affected LCS MOVs [Motor Operated Valves]
will be eliminated or retained as normal system isolation or
maintenance valves having no safety or leakage control function thus
requiring no bypassing of their thermal overloads. This proposed
change has no effect on the consequences of an accident previously
evaluated since the LCS lines will be welded and/or capped closed,
thus assuring that primary containment integrity, isolation and leak
test capability are not compromised.
Therefore, as discussed above, the proposed changes do not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
(2) The operation of River Bend 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 proposed amendment to Technical Specification 3.6.1.3 does
not create the possibility for a new or different kind of accident
from any accident previously evaluated. The BWROG (Boiling Water
Reactors Owners Group) evaluated MSIV leakage performance and
concluded that MSIV leakage rates up to 200 scfh will not inhibit
the capability and isolation performance of the valve to isolate the
primary containment. There is no new modification which could impact
the MSIV operability. The LOCA has been reanalyzed at the proposed
maximum combined leakage rate of 200 scfh. Therefore, the proposed
change does not create any new or different kind of accident from
any accident previously evaluated in the USAR.
The proposed amendment to delete Technical Specification 3.6.1.8
and 3.6.1.9 does not create the possibility of a new or different
kind of accident from any accident previously evaluated because the
removal of the PVLCS and MS-PLCS does not affect any of the
remaining systems at RBS [River Bend Station) and the LOCA has been
reanalyzed with LOCA doses resulting from the proposed change
remaining bounded by the applicable regulatory limits.
The PVLCS and MS-PLCS are of low safety significance as
discussed in NUREG-1273, Technical Findings and Regulatory Analysis
for Generic Safety Issue II.E.4.3, ``Containment Integrity Check,''
and NUREG/CR-3539, ``Impact of Containment Building Leakage on LWR
Accident Risk.''
The proposed change to eliminate the LCS does not create the
possibility of a new or different kind of accident from any accident
previously evaluated because the removal of the LCS does not
adversely affect any of the remaining RBS systems or change system
inter-relationships. The associated proposed changes to delete the
LCS isolation valves does not create the possibility of a new or
different kind of accident. The affected LCS MOVs will be eliminated
or retained as normal system isolation or maintenance valves having
no safety or leakage control function thus requiring no bypassing of
their thermal overloads. The PVLCS and MS-PLCS connections to the
process piping will be welded and/or capped closed to assure that
primary containment integrity, isolation and leak testing capability
are not compromised, therefore eliminating the possibility for any
new or different kind of accident.
Therefore, as discussed above, the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) The operation of River Bend Station, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The proposed amendment to Technical Specification 3.6.1.3 does
not involve a significant reduction in a margin of safety. The
allowable leak rate limit specified for the MSIVs is used to
quantify a maximum amount of bypass leakage assumed in the LOCA
radiological analysis. Results of the analysis demonstrate
calculated doses, assuming the two single active failures of one
MSIV to close and one diesel generator to respond are bounded by the
requirements of 10CFR100 for the off-site doses and 10CFR50,
Appendix A (General Design Criteria 19) for the Control Room doses.
The calculated whole body doses are significantly reduced at the
LPZ, the Control Room, and the EAB. The calculated thyroid dose is
significantly reduced at the LPZ, the Control Room, and the EAB.
The proposed amendment to delete Technical Specification 3.6.1.8
and 3.6.1.9 for the PVLCS and MS-PLCS, does not reduce the margin of
safety. In fact, the overall margin of safety is increased. The
method is effective to reduce dose consequences of MSIV and the
PVLCS leakage over an expanded operating range and will, thereby,
resolve the safety concern that the PVLCS and MS-PLCS will not
function at leakage rates higher than their design capacity. The
method is consistent with the philosophy of protection by multiple
leak-tight barriers used in containment design for limiting fission
product release to the environment. Therefore, the proposed method
is highly reliable and effective for MSIV leakage and deletion of
the PVLCS and MS-PLCS.
The calculation shows that MSIV leakage rates up to 100 scfh per
steam line would not exceed the regulatory limits. Therefore, the
proposed method provides a substantial safety margin for mitigating
the radiological consequences of MSIV leakage beyond the proposed
Technical Specification leak rate limit of 200 scfh for all four
main steam lines (combined maximum pathway).
Minor increases in containment leakage such as the leakage
through the MSIVs, as identified in NUREG-1273, NUREG/CR-3539, and
NUREG-1493 have been found to have no significant impact on the risk
to the public. Therefore, the proposed change does not result in a
significant reduction in a margin of safety.
The proposed change to delete the LCS isolation valves does not
reduce the margin of safety. Welded and/or capped closure of the LCS
lines assure that primary containment integrity and leak testing
capability are not compromised. The affected LCS MOVs will be
eliminated or retained as normal system isolation or maintenance
valves having no safety or leakage control function thus requiring
no bypassing of their thermal overloads. The PVLCS and MS-PLCS
connections to the process piping will be welded and/or capped
closed to assure that primary containment integrity, isolation and
leak testing capability are not compromised, therefore eliminating
the possibility for a significant reduction in the margin of safety.
Therefore, as discussed above, the proposed changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
[[Page 127]]
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: November 15, 1996
Description of amendment request: The proposed amendment would
revise the technical specifications to allow the performance of the 24-
hour emergency diesel generator (EDG) maintenance run while the unit is
in either Mode 1 or Mode 2. This test for the River Bend Station (RBS)
is currently prohibited in Mode 1 and Mode 2 and allowed in Modes 3, 4,
and 5.
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 significantly increase the
probability or consequences of an accident previously evaluated.
The RBS SAR [Safety Analysis Report] assumes that the AC
[Alternating Current] electrical power sources are designed to
provide sufficient capacity, capability, redundancy and reliability
to ensure that the fuel, reactor coolant system and containment
design limits are not exceeded during an assumed design basis event.
Specifically, the SAR assumes that the onsite EDGs provide emergency
power in the event offsite power is lost to either one or all three
EDF [Engineered Safety Features]
buses. In the event of a loss of preferred power, the ESF
electrical loads are automatically connected to the EDGs in
sufficient time to provide for safe reactor shutdown and to mitigate
the consequences of a design basis accident such as a LOCA [Loss of
Coolant Accident].
The proposed change to permit the 24-hour testing of the EDGs
during power operation does not significantly increase the
probability or consequences of any previously evaluated accident.
The capability of the EDGs to supply power in a timely manner will
not be compromised by permitting performance of EDG testing during
periods of power operation. Design features of the EDGs and
electrical systems ensure that if a LOCA or LOP [Loss of Offsite
Power] signal, either individually or concurrently, should occur
during testing, the EDG would be returned to its ready-to-load
condition (i.e., EDG running at rated speed and voltage separated
from the offsite sources) or separately connected to the ESF bus
providing ESF loads. An EDG being tested is considered to be
operable and fully capable of meeting its intended design function.
Additionally, the testing of an EDG is not a precursor to any
preciously evaluated accidents.
If, during the test period, the EDG were to receive a normal
operation protective trip resulting in the actuation of a generator
lockout signal, the lockout could be reset by the operators
monitoring the test. The resulting delay does not present an
immediate challenge to the fuel cladding integrity, reactor water
level control or to containment parameters, as demonstrated by the
bounding four-hour station blackout coping analysis contained in
RBS's station blackout conformance report.
Therefore, the proposed change allowing testing of EDGs during
power operation will not significantly increase the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
As previously discussed, the proposed change to permit the
performance of EDG testing during power operation will not affect
the operation of any system or alter any system's response to
previously evaluated design basis events. The EDGs will
automatically transfer from the test configuration to the ready-to-
load configuration following receipt of a valid signal (i.e., LOCA
or LOP). In the ready-to-load configuration the EDG will be running
at rated speed and voltage, separated from the offsite source and
capable of automatically supplying power to the ESF buses in the
event that preferred power is actually lost.
The proposed change is also the same configuration currently
used for the monthly one-hour test. Therefore, testing during power
operation will not create the possibility of a new or different kind
of event from any previously evaluated.
[Surveillance Requirement] SR 3.8.1.16 demonstrated that the EDG
will automatically override the test mode following generation of a
LOCA signal. In addition, the ability of the EDGs to survive a full
load reject is verified by the performance of SR 3.8.1.9. These
existing surveillance requirements, along with system design
features, ensure that the performance of EDG testing during power
operation will not create the possibility of a new or different kind
of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The AC electrical power sources are designed to provide
sufficient capacity, capability, redundancy, and reliability to
ensure the availability of necessary power to ESF systems so that
the fuel, reactor coolant system and containment design limits are
not exceeded. Specifically, the EDGs must be capable of
automatically providing power to ESF loads in sufficient time to
provide for safe reactor shutdown and to mitigate the consequences
of a design basis accident in the event of a loss of preferred
power.
Testing of EDGs during power operation will not affect the
availability or operation of any offsite source of power. In
addition, the EDG being tested remains capable of meeting it
intended design functions. Therefore, the proposed change to the
Technical Specification Surveillance Requirement 3.8.1.13 will not
result in a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: November 15, 1996
Description of amendment request: The proposed amendment would
increase the two recirculation loop Minimum Critical Power Ratio (MCPR)
limit from 1.07 to 1.10 and the single recirculation loop MCPR limit
from 1.08 to 1.12. This change request is the result of a non-
conservative calculation identified by the fuel vendor.
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 request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The revised Safety Limit MCPR and the cycle-specific thermal
limits that are based on the revised SLMCPR have been calculated
using the methods identified in the ``Supplemental Reload Licensing
Report For River Bend Station Reload 6 Cycle 7'' (Reference 1).
These methods are within the existing design and licensing basis and
cannot increase the probability or severity of an accident. The
basis of the MCPR Safety Limit calculation is to ensure that greater
that [than] 99.9% of all fuel rods in the core avoid transition
boiling and fuel damage in the event of a postulated accident.
The SLMCPR is used to establish the Operating Limit Minimum
Critical Power Ratio (OLMCPR). Neither the SLMCPR nor the OLMCPR can
initiate an event, therefore[,] a change to the SLMCPR does not
increase the probability of a accident previously evaluated.
Maintaining the Minimum Critical Power Ratio (MCPR) at or above the
OLMCPR during normal operations precludes fuel failure due to
overheating of the fuel clad during an anticipated operational
occurrence (AOO), thus limiting the consequences of an AOO. The
proposed change will increase the SLMCPR, which will require the
OLMCPR to be increased,
[[Page 128]]
which in turn will ensure that the requirements of 10 CFR [Part] 100
are met for an AOO. Therefore, there is no increase in the
consequences of an accident previously analyzed.
The request does not create the possibility of occurrence of a
new or different kind of accident from any accident previously
evaluated.
The MCPR Safety Limit is a Technical Specification numerical
value designed to ensure that fuel damage from transition boiling
does not occur as a result of the limiting postulated accident. It
cannot create the possibility of any new type of accident.
Neither the SLMCPR or the OLMCPR can initiate an event,
therefore, a change to the SLMCPR does not create the possibility of
occurrence of a new or different kind of accident from any accident
previously evaluated.
The request does not involve a significant reduction in the
margin of safety.
The MCPR Safety Limit is a Technical Specification numerical
value designed to ensure that fuel damage from transition boiling
does not occur as a result of the limiting postulated accident. This
new Safety Limit MCPR is calculated using the methods identified in
the reference. These methods are within the existing design and
licensing basis and based on RBS specific inputs.
The margin of Safety resides between the SLMCPR and the point at
which fuel fails. The proposed change to SLMCPR (and the OLMCPR)
will in fact restore the margin of safety associated with GE's
SLMCPR methodology.
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: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 2, 1996
Description of amendment request: The proposed Technical
Specification (TS) Change Request will permit the use of 10CFR50
Appendix J, Option B, Performance-Based Containment Leakage Testing for
Type A, B and C leak rate testing. TSs 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3,
4.6.1.6 and 4.6.1.7 are revised and Section 6.15 is added establishing
the Containment Leakage Rate Testing Program. The Bases are revised to
reflect this change. Minor editorial changes are included in this
request. Waterford Steam Electric Station is planning to have a
Containment Leakage Rate Testing Program in place prior to the next
scheduled refueling outage. This program will be in accordance with the
guidelines contained in Regulatory Guide 1.163, ``Performance-Based
Containment Leak-Test Program,'' dated September 1995.
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 will not affect the assumptions, design
parameters, or results of any accident previously evaluated. The
proposed change does not add or modify any existing equipment. The
proposed changes will result in increased intervals between
containment leakage tests determined through a performance based
approach. The intervals between such tests are not related to
conditions which cause accidents. The proposed changes do not
involve a change to the plant design or operation. Therefore, this
change does not involve a significant increase in the probability of
any accident previously evaluated.
NUREG-1493, ``Performance-Based Containment Leak-Test Program,''
contributed to the technical bases for Option B of 10 CFR 50
Appendix J. NUREG-1493 contains a detailed evaluation of the
expected leakage from containment and the associated consequences.
The increased risk due to lengthening of the intervals between
containment leakage tests was also evaluated and found acceptable.
Using a statistical approach, NUREG-1493 determined the increase in
the expected dose to the public from extending the testing frequency
is extremely small. It also concluded that a small increase is
justifiable due to the benefits which accrue from the interval
extension. The primary benefit is in the reduction in occupational
exposure. The reduction in the occupational exposure is a real
reduction, while the small increase to the public is statistically
derived using conservative assumptions. Therefore, this change does
not involve a significant increase in the consequences of any
accident previously evaluated.
The proposed change does not involve modifications to any
existing equipment. The proposed change will not affect the
operation of the plant or the manner in which the plant is operated.
The reduced testing frequency will not affect the testing
methodology. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not change the performance methodology
of the containment leakage rate testing program. However, the
proposed change does affect the frequency of containment leakage
rate testing. With an increased frequency between tests, the
proposed change does increase the probability that a increase in
leakage could go undetected for a longer period of time. Operational
experience has demonstrated the leak tightness of the containment
buildings has been significantly below the allowable leakage limit.
The margin of safety that has the potential of being impacted by
the proposed change involves the offsite dose consequences of
postulated accidents which are directly related to containment
leakage rates. The limitation on containment leakage rate is
designed to ensure the total leakage volume will not exceed the
value assumed in our accident analysis. The margin of safety for the
offsite dose consequences of postulated accidents directly related
to containment leakage is maintained by meeting the 1.0 La
acceptance criteria. The proposed change maintains the 1.0 La
acceptance criteria. Therefore, the proposed change will 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 New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of amendment request: September 19, 1996
Description of amendment request: The proposed changes to Plant
Hatch Units 1 and 2 Technical Specifications would revise the
Surveillance Requirements (SRs) addressing the reactor vessel pressure
and temperature (P/T) limits. The affected SRs are 3.4.9.1, 3.4.9.2,
3.4.9.3, 3.4.9.4, 3.4.9.5, 3.4.9.6, and 3.4.9.7, and the corresponding
Units 1 and 2 Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3, which show P/T
limit curves for inservice leak and hydrostatic testing, non-nuclear
heatup and cooldown, and criticality, respectively.
The P/T curves would be changed to allow separate monitoring of the
three major regions of the reactor pressure vessel (RPV) (i.e., the
upper vessel and flange region, the beltline region, and the bottom
head region), and to extend the validity of the Unit 1 curves to 32
[[Page 129]]
Effective Full Power Years (EFPY). Separate monitoring would alleviate
the difficulties with meeting certain temperature requirements due to
the artificial limits imposed by the current P/T curves.
In support of the proposed changes, General Electric (GE) prepared
and issued GENE-523-A137-1295, ``E. I. Hatch Nuclear Power Station, P-T
Curve Modification for Unit 1 and Unit 2,'' which is provided in the
submittal.
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?
Pressure and temperature (P/T) limits for the reactor pressure
vessel (RPV) are established to ensure brittle fracture of the
vessel does not occur.
A. The proposed changes merely clarify the Applicability of the
P/T limits for each of the low pressure conditions by replacing the
word ``performed'' with ``met'', adding Notes to Surveillance
Requirements, incorporating the requirements of Notes into the
Surveillance Requirements, and modifying the Frequency statements.
Conditions 2, 3, and 4, discussed in Enclosure 1 ``justification of
changes'', [of the licensee's application] have their own
Surveillance Requirements. Temperature requirements for Condition 1
are specified in the Bases. This proposed change only clarifies
which Surveillance Requirement applies to each operating
configuration. No reduction in Surveillance Frequencies is proposed.
B. The proposed revisions to the operating limits curves for
inservice leak and hydrostatic testing, and the heatup and cooldown
allow independent monitoring of the three RPV regions; i.e., the
bottom head, the upper vessel and flange, and the core beltline. The
three Unit 1 curves, including the criticality curve, were extended
to 32 Effective Full Power Years (EFPY), and a correction to the
Unit 1 criticality curve was made. Operating limits for each of the
curves were evaluated in accordance with the methodology given in
the applicable ASME Codes; Regulatory Guide 1.99, Rev. 2, and
Appendix G of 10 CFR [Part] 50.
The actual limits in the inservice leak and hydrostatic testing
curves, and the heatup and cooldown curves were not relaxed.
Therefore, segregating the curves into the three affected vessel
regions does not represent a reduction in the actual P/T
requirements. The current P/T curves represent a composite of the
three regions, with each point representing the limiting region.
Regions of the vessel that are not limiting at a specific point are,
therefore, artificially restrained. Upon implementation of the
proposed changes, each vessel region will have its own curve, with
its own true limit.
Since the proposed changes do not affect the recirculation
piping, the probability and the consequences of a loss of coolant
accident are not increased. Likewise, no other previously evaluated
accidents or transients, as defined in Chapters 14 and 15 of the
Units 1 and 2 Final Safety Analysis Reports, are affected by the
proposed changes.
In summary, the proposed changes do not represent a relaxation
of any actual operating limit and do not reduce the Frequency of any
Surveillance. Three of the four operating configurations of the RPV
are covered by Surveillance Requirements. Temperature limitations
for the head removed from the vessel are given in the Bases. The
operating limits were developed using the approved methodology
contained in 10 CFR [Part] 50, Appendix G. Therefore, the
probability and consequences of a brittle fracture of the RPV are
not increased.
2. Do the proposed changes create the possibility of a new or
different type of accident from any previously evaluated.
Implementing the low pressure changes, or the new operating
limit curves, does not alter the design or operation of any system
designed for the prevention or mitigation of accidents. The proposed
changes do not introduce any new type of normal or abnormal
operating mode or failure mode. All P/T limits for the Unit 1 and
the Unit 2 reactor vessels continue to be monitored per the
requirements of 10 CFR [Part] 50, Appendices G and H. Therefore, the
proposed changes do not create the possibility of a new type of
accident.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
The purpose of the P/T limits is to ensure a brittle fracture of
the RPV does not occur. The proposed Technical Specifications
changes for the low pressure conditions are made for clarification
purposes. No operating limits or Surveillance Requirements are
relaxed. The wording of current Technical
Specifications SRs 3.4.9.1, 3.4.9.2, 3.4.9.5, 3.4.9.6, and
3.4.9.7 could result in overly conservative application of the
requirements. The proposed amendment is written to remove the
ambiguity in that the Applicability and Frequency of each
Surveillance Requirement are clear. Neither the acceptance criteria
nor the Surveillance Frequency of any Surveillance is reduced.
Furthermore, the four possible RPV configurations are all adequately
monitored. As a result, the margin of safety for the low pressure
conditions is not significantly reduced due to the proposed changes.
The Unit 1 operating curves were extended to 32 EFPY using
approved methodologies. More operational margin is provided, because
the three vessel regions (upper vessel and flange, beltline, and
bottom head) are being separated for the inservice leak and
hydrostatic testing curve, and the heatup and cooldown curve.
Although this separation results in more operating margin for
certain vessel regions, it does not represent a significant
reduction in the margin of safety. As described previously, the
current Technical Specifications curves represent a composite of the
three regions. Thus, the curves represent the temperature for the
limiting region at a particular point. The regions that are not
limiting at a particular point are artificially restricted.
Separating the three regions, as proposed, eliminates false limits.
The true limit for each region is preserved and uncompromised, based
on the use of approved methodologies.
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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of amendment request: October 7, 1996
Description of amendment request: The proposed changes to Plant
Hatch Unit 1 and Unit 2 Technical Specifications (TS) would revise
Surveillance Requirements (SR) 3.1.7.7 and 3.4.3.1, and Limiting
Conditions for Operation (LCO) 3.4.3, 3.5.1, and 3.6.1.6, to increase
the nominal mechanical pressure relief setpoints for all of the 11
safety/relief valves (SRV) to 1150 psig and allow operation with one
SRV and its associated functions inoperable. The proposed changes would
reduce the potential for SRV pilot leakage and the potential for forced
outages due to an inoperable SRV during a fuel cycle.
The existing TS require that during continuous operation, all of
the 11 SRVs remain OPERABLE in the safety mode, 7 in the Automatic
Depressurization System (ADS) mode, and 4 in the Low-Low Set (LLS)
mode. If one SRV is inoperable for longer than the duration specified
in the applicable Action Statements, the plant must be placed in a Cold
Shutdown Condition. Analyses have been completed which show that, with
one SRV out of service, all transient/accident criteria can still be
met. Increasing the nominal mechanical relief setpoints will increase
the simmer margin (i.e., the difference between the SRV setpoints and
the vessel steam dome pressure), thereby potentially reducing SRV pilot
leakage which may occur during a typical operating cycle.
[[Page 130]]
As a result of increasing the mechanical relief setpoints for the SRVs,
the Standby Liquid Control (SLC) System pump test discharge pressure is
increased to 1232 psig. The High Pressure Coolant Injection (HPCI) and
Reactor Core Isolation Cooling (RCIC) systems are capable of operating
at this increased pressure.
In support of the proposed changes, General Electric (GE) prepared
NEDC-32041P, ``Safety Review for Edwin I. Hatch Nuclear Power Plant
Units 1 and 2 Updated Safety/Relief Valve Performance Requirements,''
Revision 2, dated April 1996, which was included in the submittal.
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 changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The SRVs serve to mitigate postulated transients and accidents;
the proposed changes do not alter the function or mode of operation
of the SRVs. The probability of an OPERABLE or an INOPERABLE SRV
inadvertently opening or failing to open or close is not affected by
these changes. Therefore, the probability of an accident is not
increased. Analysis(a) has been performed which considers the
consequences of the various transients and accidents with the
increased setpoints and with one SRV inoperable. The analysis also
considers the impact on ECCS [Emergency Core Cooling System]
performance, including HPCI and RCIC. The analysis has shown that
the consequences of an accident with the increased SRV setpoints and
with one SRV inoperable are not increased.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously analyzed.
Revising the nominal SRV setpoint only changes when the SRV
opens in its mechanical relief mode; the operation of the SRV and
any other existing equipment is not altered. Operation with one SRV
inoperable was evaluated(a) and does not introduce any new
failure modes. The impact on the operation and design of other
systems and components has been evaluated,(a) including ECCS
and SLC. No new operating modes or failure modes are introduced.
Thus, these changes do not contribute to a new or different type of
accident.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The change in SRV setpoint and operation with one SRV inoperable
was evaluated relative to the applicable safety system settings and
found to remain acceptable. For example, the proposed changes were
evaluated against peak clad temperature limits, ECCS operation, ASME
Code overpressurization limits, the MINIMUM CRITICAL POWER RATIO
Safety Limit, and containment design limits; no significant
reduction in the margin of safety was identified(a).
(a) GE Report NEDC-32041P, ``Safety Review for Edwin I. Hatch
Nuclear Power Plant Units 1 and 2 Updated Safety/Relief Valve
Performance Requirements, Revision 2 (Proprietary), April 1996''.
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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of amendment request: October 29, 1996
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) for Plant Hatch Units 1 and 2
associated with the installation of a digital Power Range Neutron
Monitoring (PRNM) system and the incorporation of long-term stability
solution hardware.
In response to Generic Letter 94-02, ``Thermal-Hydraulic
Instabilities in Boiling Water Reactors,'' Georgia Power Company (GPC)
selected General Electric (GE) Option III as the long-term stability
solution. Option III detects core instabilities and provides a reactor
scram signal to the Reactor Protection System (RPS). The long-term
stability solution, GE Option III, is supported by the BWR Owners'
Group Topical Report NEDO-31960-A submitted to the NRC for approval in
May 1991, and NEDO-31960-A, Supplement 1, submitted to the NRC for
approval in March 1992. The NRC issued a Safety Evaluation Report (SER)
for NEDO-31960-A and Supplement 1 in July 1993. BWR Owners' Group
Topical Report NEDO-32465, submitted to the NRC in June 1995, provides
additional analysis for the detection and suppression methodology
(Option III).
To execute the stability solution software, the Average Power Range
Monitor (APRM) and Rod Block Monitor (RBM) electronics would be
replaced with a PRNM system based on digital GE Nuclear Measurements
Analysis and Control NUMAC modules. Implementation of the PRNM would
affect the RPS and Control Rod Block TS 3.3.1.1, 3.3.2.1, 3.4.1 and
3.10.8.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The purpose of the proposed amendment is to incorporate the
Power Range Neutron Monitoring (PRNM) retrofit and Oscillation Power
Range Monitor (OPRM) installation. The types of Average Power Range
Monitor (APRM) Functions that are credited to mitigate accidents
were previously evaluated. The proposed OPRM Upscale Function is
implemented in the same hardware that implements the APRM Functions.
The change to a two-out-of-four RPS [Reactor Protection System]
logic was analyzed and determined to be equal to the original logic.
The modification involves equipment that is intended to detect
the symptoms of some accidents and initiate mitigating action. The
worst case failure of the equipment involved in the modification is
a failure to initiate mitigating action (scram), but no failure can
cause an accident. As discussed in the bases for proposed changes,
the PRNM replacement system is designed to perform the same
operations as the existing Power Range Monitoring (PRM) system and
to meet or exceed all of its operational requirements. Therefore, it
is concluded that the probability of an accident previously
evaluated is not increased as a result of replacing the existing
equipment with the PRNM equipment.
* * * *
Human-machine interface (HMI) failures in the current system
could be related to incorrectly adjusted settings, incorrect reading
of meters, and failure to return the equipment to the normal
operating configuration. There are comparable failure modes for some
of these problems in the digital system where an erroneous
potentiometer adjustment in the current system is equivalent to an
erroneous digital entry in the replacement system. Certain potential
``failure to reconfigure errors'' in the current system have no
counterpart in the replacement system, because any reconfiguration''
is automatically returned to normal by the system. Also, since
parameters are available for review at any time, even if an error,
such as a digital entry error occurs, it is more likely that the
error would be almost immediately detected by recognition that the
displayed value is not the correct one.
The failure analysis of the current system assumes certain rates
of human error. The rates for the replacement system will be lower
and, hence, are bounded by the FSAR [Final Safety Analysis Report]
analysis.
Therefore, GPC [Georgia Power Company] concludes the proposed
changes do not
[[Page 131]]
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The APRM Trip Functions credited in the accident analyses are
retained in the PRNM retrofit. The response time of the new
electronics meets or exceeds the required response criteria. No new
interfaces or interactions with other equipment will introduce any
new failure modes.
The modification involves equipment that is intended to detect
the symptoms of some accidents and initiate mitigating action. The
worst-case failure of the equipment involved in the modification is
a failure to initiate mitigating action (scram), but no failure can
cause an accident. This is unchanged from the current system.
Software common-cause failures can at most cause the system to
fail to perform its safety function. In that case, it could fail to
initiate action to mitigate the consequences of an accident, but
would not cause one.
The new system is a digital system with software (firmware)
control. As such, it has ``central'' processing points and software
controlled digital processing where the current system had analog
and discrete component processing. The result is that the specific
failures of hardware and potentially common-cause software failures
are different from the current system. Also, automatic self-test
results in some cases in a direct trip as a result of a hardware
failure where the current system may have remained ``as-is''.
However, when these are evaluated at the system level, there are no
new effects. In general, FSARs assume simplistic failure modes
(relays for example) but do not specifically evaluate such effects
as self-test detection and automatic trip or alarm.
The effects of software common-cause failure are mitigated by
hardware design and system architecture. The replacement equipment
is fully qualified to operate in its installed location and will not
affect other equipment.
Therefore, GPC concludes the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The replacement equipment provides the same function as the
original electronics. Response time and operator information are
either maintained or improved. The equipment was qualified, where
appropriate, to assure its intended safety function is performed.
The replacement system has improved channel trip accuracy compared
to the current system and meets or exceeds system requirements
assumed in setpoint analysis. The channel response time exceeds the
requirements. The channel indicated accuracy is improved over the
current system, and meets or exceeds system requirements. The
replacement system meets or exceeds all system requirements.
The BWROG [BWR Owners' Group] Stability Option III was developed
to meet the requirements of GDC [General Design Criterion] 10 and
GDC 12 by providing a hardware system that detects the presence of
thermal-hydraulic instabilities and automatically initiates the
necessary actions to suppress the oscillations prior to violating
the MCPR [maximum critical power ratio] Safety Limit. The NRC has
reviewed and accepted the Option III methodology described in
Licensing Topical Report NEDO-31960 and concluded this solution will
provide the intended protection. Therefore, it is concluded that
there will be no reduction in the margin of safety as defined in the
Technical Specifications as a result of the installation of the OPRM
system and the simultaneous removal of the operating restrictions
imposed by the ICAs [item control areas].
Therefore, GPC concludes the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: November 20, 1996
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to allow the Vice President to
designate the Safety Audit and Review Committee (SARC) Chairperson, to
change the work hours limitation in accordance with guidance in GL 82-
12, ``Nuclear Power Plant Staff Working Hours;'' to change radioactive
shipments record retention requirements to comply with recent 10 CFR
Part 20 changes; and other editorial 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. The changes requested are administrative in nature. Paragraph
3.D was placed in the License by Amendment No. 155 to authorize
Omaha Public Power District (OPPD) to increase the storage capacity
of the FCS spent fuel pool. Amendment No. 155 stated that the TS as
issued would be effective when the last new rack was installed.
Since the last new rack was installed on
August 8, 1994, Paragraph 3.D is no longer necessary and should
be deleted from the License.
Table of Contents, Section 6.0, ``Interim Special Technical
Specifications,'' Subsections 6.1 through 6.4 are proposed for
deletion because all of the Specifications referred to have been
deleted by previous Amendments.
The revision proposed for TS 2.15 (Item 2C of Table 2-3 & Item
1C of Table 2-4) will insert the correct terminology (Pressurizer
Low/Low Pressure) into the Functional Unit description.
The revision proposed for TS 5.2 will require the control of
overtime worked by personnel to be in accordance with the NRC Policy
Statement on working hours (Generic Letter 82-12) in lieu of stating
the specific times requirements from the Policy as the current TS
does. This option is in accordance with NUREG-1432, Standard TS for
Combustion Engineering Plants, Specification 5.2.2e, and will allow
work groups to be on twelve hour shifts.
The revision proposed for TS 5.5.2.2 will replace the specific
title of the Chairperson of the Safety Audit and Review Committee
and replace it with ``Member as appointed by the Vice President.''
This will allow the flexibility to change chairmanship of the
committee amongst the members.
The revision to TS 5.10 concerning retention of records of
radioactive shipments will update the TS to current 10 CFR 20
requirements. Plant procedures already comply with current 10 CFR 20
record retention requirements. The addition of the Section 5.0 title
corrects a minor format discrepancy.
These proposed revisions are administrative in nature. The
proposed revisions have no effect on any initial assumptions or
operating restrictions assumed in any accident, nor do these changes
have any effect on equipment required to mitigate the consequences
of an accident. Therefore the proposed revisions do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed revisions correct minor errors, remove outdated
information, are consistent with changes in organizational
structure, 10 CFR Part 20, or NUREG-1432, ``Combustion Engineering
Standard Technical Specifications (STS). These changes will not
result in any physical alterations to the plant configuration,
changes to setpoint values, or changes to the application of
setpoints or limits. No new operating modes are proposed as a result
of these changes. Therefore the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
[[Page 132]]
3. The proposed change does not involve a significant reduction
in a margin of safety.
The revisions listed above correct minor errors, remove outdated
information, or are consistent with changes in organizational
structure, 10 CFR Part 20, or Standard TS. These changes will not
result in any physical alterations to the plant configuration,
changes to setpoint values, or changes to the application of
setpoints or limits. Therefore the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502
NRC Project Director: William H. Bateman
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: October 28, 1996
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3/4.8.1, ``A.C. Sources,''
TS Section 3/4.8.2, ``Onsite Power Distribution Systems,'' TS Table
4.8.1, ``Battery Surveillance Requirements,'' and the associated bases.
Surveillance requirements would be modified to account for the increase
in the fuel cycle, consistent with Generic Letter 91-04, ``Changes in
Technical Specification Surveillance Intervals to Accommodate a 24-
month Fuel Cycle,'' dated April 2, 1991. Administrative changes are
also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in
accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no such accidents are affected
by the proposed revisions to increase the surveillance test
intervals from 18 to 24 months for the A.C. Offsite Sources, the
Emergency Diesel Generators and the Station Batteries or the
proposed revision to remove the ``during shutdown'' restriction for
conduct of the battery performance test.
Results of the review of historical 18 month surveillance data
and maintenance records support an increase in the surveillance test
intervals from 18 to 24 months (and up to 30 months on a non-routine
basis) because no potential for a significant increase in a failure
rate of a system or component was identified during these reviews.
These proposed revisions are consistent with the NRC guidance on
evaluating and proposing such revisions as provided in Generic
Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2,
1991.
Initiating conditions and assumptions remain as previously
analyzed for accidents in the DBNPS Updated Safety Analysis Report.
These revisions do not involve any physical changes to systems
or components, nor do they alter the typical manner in which the
systems or components are operated.
The proposed revision to reflect that the battery charger
performance test will continue to be conducted on a[n] 18 month
surveillance interval is an administrative change and does not
affect previously analyzed accidents.
The proposed revision to the Bases to reflect that a change to a
24 month surveillance test interval is an exception to current
guidance is an administrative change and does not affect previously
analyzed accidents.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the source term, containment
isolation or radiological releases are not being changed by these
proposed revisions. Existing system and component redundancy is not
being changed by these proposed changes. Existing system and
component operation is not being changed by these proposed changes
and the assumptions used in evaluating the radiological consequences
in the DBNPS Updated Safety Analysis Report are not invalidated.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because these
revisions do not involve any physical changes to systems or
components, nor do they alter the typical manner in which the
systems or components are operated.
No changes are being proposed to the type of testing currently
being performed, only to the length of the surveillance test
interval and to restrictions on conducting testing only during
shutdown conditions.
Results of the review of historical 18 month surveillance data
and maintenance records support an increase in the surveillance test
intervals from 18 to 24 months (and up to 30 months on a non-routine
basis) because no potential for a significant increase in a failure
rate of a system or component was identified during these reviews.
The proposed revision to reflect that the battery charger
performance test will continue to be conducted on a[n] 18 month
surveillance interval is an administrative change and does not alter
testing currently being performed.
The proposed revision to the Bases to reflect that a change to a
24 month surveillance test interval is an exception to current
guidance is an administrative change and does not alter testing
currently being performed.
3. Not involve a significant reduction in a margin of safety
because the results of the historical 18 month surveillance data and
maintenance records review identified no potential for a significant
increase in a failure rate of a system or component due to
increasing the surveillance test interval to 24 months. Existing
system and component redundancy is not being changed by these
proposed changes.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences,
consequently there are no significant reductions 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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: November 26, 1996
Description of amendment request: The proposed changes would
eliminate the records retention requirements from the administrative
section of the Technical Specifications (TS) in accordance with NRC
Administrative Letter95-06, ``Relocation of Technical Specifications
Administrative Controls Related to Quality Assurance.''
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:
Specifically, operation of the ... North Anna Power [Station] in
accordance with the proposed Technical Specifications changes will
not:
[[Page 133]]
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated. The proposed
administrative changes do not affect equipment or its operation.
Therefore, the likelihood that an accident will occur is neither
increased nor decreased by relocating record retention requirements
from the Technical Specifications to the Operational Quality
Assurance Program. This TS change will not impact the function or
method of operation of plant equipment. Thus, a significant increase
in the probability of a previously analyzed accident does not result
due to this change. No systems, equipment, or components are
affected by the proposed changes. Thus, the consequences of any
accident previously evaluated in the UFSAR [Updated Final Safety
Analysis Report] are not increased by this change.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated. The proposed change
does not alter the design or operations of the physical plant. Since
record retention requirements are administrative in nature, a change
to these requirements does not contribute to accident initiation, an
administrative change related to this activity does not produce a
new accident scenario or produce a new type of equipment
malfunction. [These] changes do not alter any existing accident
scenarios. The proposed administrative change does not affect
equipment or its operation, and, thus, does not create the
possibility of a new or different kind of accident. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident.
(3) Involve a significant reduction in a margin of safety.
Section 6.0 of the North Anna ... Technical Specifications does not
have a basis description. The proposed administrative change does
not affect equipment or its operation, and, thus, does not involve
any 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: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Mark Reinhart, Acting
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: December 3, 1996
Description of amendment request: This amendment request proposes
to revise the technical specifications associated with the inspection
of the reactor coolant flywheel to provide an exception to the
recommendations of Regulatory Guide 1.14, Revision 1, ``Reactor Coolant
Pump Flywheel Integrity.'' The proposed exception would allow either an
ultrasonic volumetric examination or surface examination to be
performed at approximately 10-year intervals. In addition, a correction
of the issuance date of a referenced regulatory guide is included.
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 p presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The safety function of the RCP [reactor coolant pump] flywheels
is to provide a coastdown period during which the RCPs would
continue to provide reactor coolant flow to the reactor after loss
of power to the RCPs. The maximum loading on the RCP flywheel
results from overspeed following a LOCA [loss-of-coolant accident].
The maximum obtainable speed in the event of a LOCA was predicted to
be less than 1500 rpm. Therefore, a peak LOCA speed of 1500 rpm is
used in the evaluation of RCP flywheel integrity in WCAP-14535. This
integrity evaluation shows a very high flaw tolerance for the
flywheels. The proposed change does not affect that evaluation.
Reduced coastdown times due to a single failed flywheel is bounded
by the locked rotor analysis, therefore, it would not place the
plant in an unanalyzed condition. Therefore, these changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed amendment does not create the possibility of a new
or different kind of accident from any previously evaluated since
the proposed amendments will not change the physical plant or the
modes of plant operation defined in the facility operating license.
No new failure mode is introduced due to the proposed change, since
the proposed change does not involve the addition or modification of
equipment, nor do they alter the design or operation of affected
plant systems, structures, or components.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are basically unchanged by the
proposed amendment. The results of the flywheel inspections
performed have identified no indications affecting flywheel
integrity. As identified in WCAP-14535, detailed stress analysis as
well as risk analysis have been completed with the results
indicating that there would be no change in the probability of
failure for RCP flywheels if all inspections were eliminated.
Therefore these 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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: December 3, 1996
Description of amendment request: This amendment request proposes
to correct the reference to the Action Statement for Item 7.b, RWST
Level - Low-Low Coincident with Safety Injection, Table 3.3-3,
Engineered Safety Features Actuation System Instrumentation, from
Action 16 to Action 28.
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.
Changing the reference from Action Statement 16 to Action
Statement 28 for Functional Unit 7.b. of Table 3.3-3 will reduce the
probability for an automatic switchover from the RWST [refueling
water storage tank] to an empty containment sump to occur, while an
RWST level channel is inoperable or is being tested with its
bistable tripped, should an inadvertent safety injection signal
occur concurrent with a single failure of a second RWST level
channel. The design of these channels does not allow for operation
or testing in bypass, so Action Statement 16 is not applicable.
Changing to Action Statement 28 will limit
[[Page 134]]
the duration that a channel could be inoperable or be in test with
its bistable bypassed. This change does not involve any design
changes or hardware modifications, and does not introduce any new
potential accident initiating conditions. The increase in allowed
outage time for this item was evaluated and the associated
unavailability and risk was shown to be equivalent to, or less than,
that of other functional units evaluated in WCAP-10271, Supplement
2, Revision 1. Therefore, this proposed change does not increase the
probability of any accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not result in any hardware changes and
does not result in a change in the manner in which the ESFAS
[engineered safety features actuation system] provides plant
protection. This change does not alter the functioning of the ESFAS.
Rather, the likelihood or probability of the ESFAS functioning
properly is affected as described above. This change will not change
the method by which any safety-related system performs its function.
Therefore, this proposed change will not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This proposed change will not result in a significant reduction
in the margin of safety defined for any technical specification
since it does not alter the manner in which safety limits, limiting
safety system settings, or limiting conditions for operation are
determined.
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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: June 28, 1996
Brief description of amendments: The amendment would modify the
technical specifications (TS) to increase the minimum required amount
of anhydrous trisodium phosphate (TSP) in the containment baskets. TSP
is used to ensure that following a postulated design basis loss of
coolant accident (LOCA), the containment sump pH is maintained greater
than or equal to seven.
Date of issuance: December 10, 1996
Effective date: December 10, 1996, to be implemented within 45 days
from the date of issuance.
Amendment Nos.: Unit 1 - 110; Unit 2 - 102; Unit 3 - 82
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 11, 1996 (61
FR 47962) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 10, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: June 21, 1996
Brief description of amendments: The amendments revise the term
``lifting loads'' used in Technical Specification 3.9.6b.2, Manipulator
Crane, to ``lifting force.'' This revision will clarify that the static
loads associated with the lifting tool, drive rod, and control rod
weights are not included in the lifting force limit.
Date of issuance: December 12, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 171 and 153
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 11, 1996 (61
FR 47977) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 12, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: February 22, 1996, and as
supplemented by letters dated July 4 and September 20, 1996
Brief description of amendment: The amendment revises Clinton Power
Station Technical Specification 3.3.4.1, ``End of Cycle Recirculation
Pump Trip (EOC-RPT) Instrumentation,'' by deleting Surveillance
Requirement 3.3.4.1.6 which requires the RPT breaker interruption time
to be determined at least once per 60 months.
Date of issuance: December 13, 1996
Effective date: December 13, 1996
Amendment No.: 111
[[Page 135]]
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18169) The supplemental letters of July 4 and September 20, 1996,
provided clarifying information and did not include significant changes
relative to the original Federal Register notice.The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated December 13, 1996.No significant hazards consideration comments
received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: July 12, 1996, as
supplementedOctober 30, 1996.
Brief description of amendment: The amendment revises TS 6.2.2.h
regarding the administrative controls for the normal working hours of
unit staff who perform safety-related functions, and TS 6.2.2.i
regarding an organizational change. The changes authorize (1)
establishment of unit staff work schedules that average 40 hours per
week using shifts as long as 12 hours, and (2) elimination of the
positions of General Supervisor Operations and Supervisor Operations.
Date of issuance: December 12, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 158
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 14, 1996 (61 FR
42280) The October 30, 1996, letter provided supplemental information
that did not change the initial no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 12, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: July 12, 1996
Brief description of amendment: The amendment revises Technical
Specification Section 6.2.2.i regarding the administrative controls for
the normal working hours of unit staff who perform safety-related
functions. The change allows the establishment of unit staff work
schedules that average 40 hours per week using shifts as long as 12
hours.
Date of issuance: December 12, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 78
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 14, 1996 (61 FR
42281) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 12, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: July 28, 1995, as supplemented
October 25, 1995, and August 9, 1996
Brief description of amendments: The amendments revise the 250 volt
DC profiles in the Technical Specifications for the two units to
reflect new load profile calculations.
Date of issuance: December 17, 1996
Effective date: Unit 1, as of date of issuance, to be implemented
within 30 days; Unit 2, as of date of issuance, to be implemented prior
to Startup following the Eighth Refueling and Inspection Outage for
Unit 2, which is scheduled for the Spring of 1997.
Amendment Nos.: 162 and 133
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47622) The supplemental letters provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination nor the Federal Register notice.The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated December 17, 1996.No significant hazards
consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: June 12, 1992, as supplemented
September 17, 1992, March 17, 1993, August 17, 1993, August 18, 1993,
December 29, 1993, June 29, 1995, August 15, 1996, October 3,
1996,October 23, 1996, November 14, 1996, November 20, 1996 (JPN-96-
045), November 20, 1996 (JPN-96-046), and November 27, 1996.
Brief description of amendment: The amendment modifies
Facility Operating License No. DPR-59 and the James A. FitzPatrick
Nuclear Power Plant (JAFNPP) Technical Specifications (TSs) to
authorize an increase in the maximum power level of JAFNPP from 2436
MWt to 2536 MWt. The amendment also approves changes to the TSs to
implement uprated power operation.
Date of issuance: December 6, 1996
Effective date:
As of the date of issuance to be implemented upon plant startup
following the refueling outage cycle 13.
Amendment No.: 239
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4943) The letters dated September 17, 1992, March 17, 1993, August 17,
1993, August 18, 1993, December 29, 1993, June 29, 1995, August 15,
1996,October 3, 1996, October 23, 1996, November 14, 1996, November 20,
1996, (JPN-96-045), November 20, 1996, (JPN-96-046), and November 27,
1996, provided clarifying information that did not change the initial
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated December 6, 1996.No significant hazards
consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
[[Page 136]]
Public Service Electric & Gas Company, Docket No. 50-311, Salem
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey
Date of application for amendment: September 20, 1996, as
supplemented September 30, 1996
Brief description of amendment: The amendment changes Technical
Specification Surveillance Requirement 4.7.7.b.4 for the Auxiliary
Building Exhaust Air Filtration System, and its associated Bases, to
indicate that the specified flowrate applies only to system testing.
Date of issuance: December 12, 1996
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment No. 168
Facility Operating License No. DPR-75: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 23, 1996 (61 FR
55040) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 12, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: August 27, 1996, as
supplemented October 24, 1996
Brief description of amendments: The amendment to Unit 2 deletes
License Condition 2.C.(24)(a) which required establishment by June 3,
1981, of regularly scheduled 8-hour shifts without reliance on routine
use of overtime. The amendments to both Units 1 and 2 revise Technical
Specification 6.2.2 to delete the reference to Generic Letter 82-12,
``Nuclear Plant Staff Working Hours,'' and require that administrative
controls be established which will ensure that adequate shift coverage
is maintained without heavy use of overtime for individuals.
Date of issuance: December 17, 1996
Effective date: Both units, as of date of issuance, to be
implemented within 30 days.
Amendment Nos. 186 and 169
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications for both units and License for
Unit 2 only.
Date of initial notice in Federal Register: September 12, 1996 (61
FR 48175) The October 24, 1996, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or the original notice.The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated December 17, 1996.No significant hazards consideration
comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: May 29, 1996
Brief description of amendments: These amendments revise Technical
Specification (TS) Surveillance Requirement 3.5.1.4 to increase the
minimum boron concentration in the safety injections tanks from 1850
ppm to 2200 ppm.
Date of issuance: December 6, 1996
Effective date: December 6, 1996, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 2 - 135; Unit 3 - 124
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40029) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 6, 1996. No significant
hazards consideration comments received: No.Temporary
Local Public Document Room location: Science Library, University of
California, P. O. Box 19557, Irvine, California 92713
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: September 27, 1996, as
supplemented on October 25, and November 18, 1996
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant Technical Specification requirements related to the
low temperature overpressure protection (LTOP) system. Specifically,
the LTOP curve is modified to define 10 CFR Part 50, Appendix G
pressure temperature limitations for LTOP evaluation through the end of
operating cycle (EOC) 33. In addition, the LTOP enabling temperature
and the temperature required for starting a reactor coolant pump have
been changed consistent with the design basis for the LTOP system.
Finally, the TS bases were changed consistent with the changes
described above.
Date of issuance: December 13, 1996
Effective date: December 13, 1996, to be implemented within 30
days.
Amendment No.: 130
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 7, 1996 (61 FR
52472) The October 25 and November 18, 1996, submittals provided
supplemental information that did not change the initial proposed no
significant hazards consideration determination.The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated December 13, 1996.No significant hazards consideration comments
received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Dated at Rockville, Maryland, this 24th day of December 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation
[Doc. 96-33254 Filed 12-31-96; 8:45 am]
BILLING CODE 7590-01-F