Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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

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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.

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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.

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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