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

Federal RegisterDec 17, 1997

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

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 November 21, 1997, through December 5, 1997.

The last biweekly notice was published on December 3, 1997 (62 FR

63970).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and should cite the publication date and

page number of this Federal Register notice. Written comments may also

be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.

The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By January 16, 1998, 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

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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: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will

County, Illinois

Date of amendment request: February 28, 1997.

Description of amendment request: The proposed amendments would

revise Byron and Braidwood Technical Specifications (TS) Sections 3/

4.4.5, ``Steam Generators,'' and 3/4.4.8, ``Reactor Coolant System

Specific Activity,'' for both the Byron Station, Units 1 and 2, and the

Braidwood Station, Units 1 and 2. The intent of these proposed

revisions is to restore for both Byron, Unit 1, and Braidwood, Unit 1,

the original TS related to steam generator (SG) inspections and the

primary coolant dose equivalent iodine-131 (DEI) concentrations. These

amendments will become effective when the original steam generators

(OSG) which are Westinghouse Model D4 SGs, are removed and the

replacement steam generators (RSG) made by Babcock and Wilcox,

International (BWI), are installed. The RSGs are presently being

installed at Byron, Unit 1, while the RSGs will be installed at

Braidwood, Unit 1, in fall 1998.

The SG inspection methodology, inspection frequency, reporting

requirements and acceptance criteria for the RSGs in both Byron, Unit

1, and Braidwood, Unit 1, will revert to the TSs for the OSGs before

several prior license amendments incorporated into the TSs: (1) The

interim plugging criteria (IPC) consistent with Generic Letter (GL) 95-

05; (2) the F* criteria for the SG tube expansions into the tubesheet;

and (3) the criteria for repairing SG tubes using either Westinghouse

laser welded sleeves or Combustion Engineering tungsten inert gas (TIG)

welded sleeves. The TSs applicable to Byron, Unit 2, and Braidwood,

Unit 2, both of which have Westinghouse Model D5 SGs, remain unchanged

except for designating them in the TSs as model D5 SGs.

With respect to the limiting value of the DEI primary coolant

concentration, both the Byron, Unit 1, TSs and the Braidwood, Unit 1,

TSs will revert from their present TS limit of 0.35 to 1.0 microcuries

per gram. A license amendment request to lower the Byron, Unit 1, TS

DEI limit from 0.35 to 0.20 microcuries per gram was submitted on

January 31, 1997, but this request was

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subsequently withdrawn on November 11, 1997, because the RSGs were

being installed in the Byron, Unit 1, refueling outage which started in

early November 1997. A license amendment request to lower the

Braidwood, Unit 1, TS DEI limit from 0.35 to 0.10 microcuries per gram

was submitted on September 2, 1997. Action on this request is still

pending but in any case, will not affect the subject license amendment

request for Braidwood, Unit 1, because the September 2, 1997, request

is only applicable to the OSGs which are presently using the IPC that

were originally incorporated into the TSs on November 9, 1995. The

applicable bases sections of the Byron, Unit 1, TSs and Braidwood, Unit

1, TS will also be revised to reflect the TS changes discussed above.

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.

Due to design differences between the replacement Steam

Generators (RSGs) and OSGs, the analyses supporting the application

of the F* and voltage-based repair criteria do not apply to the

RSGs. Also, the analyses supporting sleeving repair by the

Westinghouse laser welded or Combustion Engineering Tungsten Inert

Gas (TIG) welded sleeving methodologies do not apply to the RSGs due

to the design differences. The RSG and OSG tube bundle

configurations are similar, however, the RSG tubes are smaller in

diameter, constructed of Inconel Alloy 690 instead of Alloy 600, and

supported by stainless steel lattice grids instead of the drilled

carbon steel plates used in the OSGs. The RSG tubes are

hydraulically expanded into the tube sheet during initial assembly.

The RSG upper tube bundle shape consists of tubes with continuous,

smooth, long radius bends.

The structural analysis demonstrates that the tube integrity is

maintained for a Main Steamline Break (MSLB) occurring during normal

full power operation. The structural evaluation of the tubing for

faulted conditions was performed in accordance with the ASME Boiler

and Pressure Vessel Code Section III requirements. The tube material

selection and size exceed the strength requirements of the existing

steam generators. Comparison of the Alloy 690 tube material used in

the RSGs with the Alloy 600 tube material in the OSGs show that the

RSG material strength characteristics are as good as or better than

those of the existing design. A comparison of the stress margins of

the RSG and OSG show that the stress margin in the RSG tubes exceed

the stress margin in the OSG tubes.

RSG portions of the reactor coolant pressure boundary are

designed to permit periodic inspection and testing of important

areas and features to assess structural and leak-tight integrity.

ASME Section XI, provides the depth of an allowable outside diameter

(O.D.) flaw for tubes in service. The RSG has tubing fabricated from

SB-163 material (Inconel Alloy 690) which is examined by eddy

current methods to the requirements of ASME Section III, NB-2550.

The tubing has a radius to thickness (r/t) ratio less than 8.70. In

accordance with ASME Section XI, for tubing having an r/t ratio of

less than 8.70, the depth of an allowable O.D. flaw shall not exceed

40% of the nominal tube wall thickness.

The potential for tube rupture is not increased from the OSGs as

demonstrated in the qualification analysis and testing for the RSGs.

The program for periodic inservice inspection of the steam

generators monitors the integrity of the SG tubing to ensure that

there is sufficient time to take proper and timely corrective action

if any tube degradation is detected. Therefore, installation of the

RSGs will not increase the probability of the occurrence of primary-

to-secondary leakage or a steam generator tube rupture (SGTR) during

normal or accident conditions.

The design basis doses calculated for postulated accidents

involving degradation of SG tubes, such as SGTR and MSLB accidents,

as presented in UFSAR [Updated Final Safety Analysis Report] Chapter

15 accident analysis have been evaluated and are decreased by

installation of the RSGs and restoration of the RCS activity limit

to 1.0 microcuries/gm. The decrease in offsite dose is primarily due

to the smaller RSG tube diameter and less primary-to-secondary

transfer during the event. The dose calculations are performed

consistent with NUREG-0800, ``Standard Review Plan'' and ensure site

boundary doses are within a small fraction of the Title 10 Code of

Federal Regulations Part 100 (10 CFR 100) requirements. Therefore,

the change does not involve a significant increase in the

consequences of an accident previously evaluated.

Limiting the applicability of TS provisions to a specific cycle

or SG type are administrative changes in that they provide

clarification consistent with current analyses and do not involve a

significant increase in the probability or 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. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

Restricting application of IPC, F* and sleeving methodologies to

the OSGs and reinstating an RCS activity limit of 1.0 microcuries/gm

upon installation of the RSGs will not introduce significant or

adverse changes to the plant design basis that could lead to a new

or different kind of accident being created. The RSG tubing meets

the requirements of General Design Criteria (GDC) 14, 15, 30, 31,

and 32 of 10 CFR 50, Appendix A. The RSG tubing has been designed

and evaluated consistent with ASME Code Section III criteria and the

inspection criteria for the RSGs is consistent with ASME Code

Section XI criteria. The RSGs have thermally treated Inconel Alloy

690 tubes which are hydraulically expanded into the tube sheet

during initial assembly. Alloy 690 is more resistant to stress

corrosion cracking (SCC) than Alloy 600 which is used in the OSG

tubing. Overall tube bundle structural and leakage integrity is

maintained at a level consistent with or better than the originally

supplied tubing during all plant conditions.

ComEd will continue to apply the TS maximum primary-to-secondary

leakage limit of 150 gpd (0.1 gpm) through any one SG at Byron and

Braidwood to help preclude the potential for excessive leakage

during all plant conditions. The EPRI recommended 150 gpd limit

provides for leakage detection and plant shutdown in the event of an

unexpected tube leak and precludes the potential for excessive

leakage or tube burst in the event of a Main Steam Line Break or

under Loss of Coolant Accident conditions.

Limiting the applicability of TS provisions to a specific cycle

or SG type are administrative changes in that they provide

clarification consistent with current analyses.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Restricting application of IPC, F*, and sleeving methodologies

to the OSGs for which the supporting analyses apply, does not

involve a reduction in a margin of safety. The RSG tubing has been

shown to retain adequate structural and leakage integrity during

normal, transient, and postulated accident conditions consistent

with GDC 14, 15, 30, 31, and 32 of 10 CFR 50 Appendix A. The RSG

tubing has been designed and evaluated consistent with the margins

of safety specified in ASME Code Section III. The proposed program

for periodic inservice inspection of the replacement steam

generators monitors the integrity of the SG tubing to ensure that

there is sufficient time to take proper and timely corrective action

if any tube degradation is present. The proposed program is

consistent with the Standard Technical Specifications.

The Unit 1 RCS dose equivalent I-131 limit is being raised upon

installation of the RSGs to eliminate the compensatory lower limit

that was adopted in conjunction with IPC for the existing

Westinghouse D4 SGs. With the RCS activity limit returned to the

Standard Technical Specification value of 1.0 [mu]Ci/gm, the

assessment of postulated UFSAR Chapter 15 accidents (including SGTR

and MSLB) has concluded that the calculated design basis doses

presented in Chapter 15 are not adversely impacted by the RSGs. This

ensures that the resulting 2-hour dose rates at the Byron and

Braidwood site boundaries will not exceed an appropriately small

fraction of 10 CFR 100 dose guideline values.

Limiting the applicability of TS provisions to a specific cycle

or SG type are

[[Page 66136]]

administrative changes in that they provide clarification consistent

with current analyses.

Therefore, it is concluded that this change does not involve a

significant reduction in a margin of safety with respect to plant

safety as defined in the UFSAR or the Technical Specification.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

requested amendments involve no significant hazards consideration.

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan

Date of amendment request: March 26, 1997.

Description of amendment request: The proposed amendment would

revise the containment system technical specifications (TS) contained

in TS Sections 3.6 and 4.5. The licensee has classified the changes as

``More Restrictive,'' ``Less Restrictive,'' and ``Administrative.''

``More Restrictive'' changes include reduction of the allowable

containment pressure, addition of an action statement defining action

to be taken when the containment pressure limit is exceeded, addition

of a restriction on containment temperature, and revision of the

applicable conditions for the containment purge valves to require that

the valves be operable above 210 degrees F versus the current

requirement that they be operable above 525 degrees F. ``Less

Restrictive'' changes include addition of an allowance to enter an air

lock through a locked door to perform maintenance, addition of an

allowance to open containment isolation valves under administrative

control, revision of the applicable conditions for containment pressure

to exclude the cold shutdown operating condition, and addition of an

exception to the surveillance requirement requiring verification of the

status of ``locked-closed'' manual isolation valves after a refueling

outage to exclude requiring such verification for valves opened under

administrative control. ``Administrative'' changes include the deletion

of containment isolation valve tables and component identifiers from

the TS in accordance with Generic Letter 91-08 (``Removal of Component

Lists from Technical Specifications'') and editorial restructuring of

the affected TS sections to clarify the remaining 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:

Each proposed change has been classified as ``Administrative,''

``More Restrictive,'' or ``Less Restrictive.'' ``Administrative''

and ``More Restrictive'' changes are discussed generically; ``Less

Restrictive'' changes are discussed individually.

Five of the proposed changes are classified as being ``Less

Restrictive'':

(G.1) Allowance in LCO [Limiting Condition for Operation] 3.6.1

to enter an air lock to perform maintenance.

(G.2) Allowance in LCO 3.6.1 to open containment isolation

valves under administrative control.

(I.2) Revising the applicable conditions of LCO 3.6.2,

Containment Pressure to exclude Cold Shutdown.

(J.2) Exception in SR [Surveillance Requirement] 4.5.3d for

valves opened under administrative control as allowed by LCO 3.6.1.

(P) Allowance in SR 4.5.2 to enter an air lock to perform

maintenance.

Four of the proposed changes are classified as being ``More

Restrictive'':

(I.1) Revising LCO 3.6.2 to reduce the allowable containment

pressure.

(I.3) Addition of an action statement to LCO 3.6.2, Containment

Pressure.

(K) Addition of a new LCO which restricts Containment

Temperature.

(M.2) Revising the applicable conditions for LCO 3.6.5, Purge

Valves.

The remaining changes are all classified as being

``Administrative''.

Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

1. Changes G.1, G.2, J.2, and P: Proposed changes G.1 and P

allow limited access through the operable door of an air lock when

the other door is inoperable; current Technical Specifications [TS]

do not. Proposed changes G.2 and J.2 allow unisolating containment

penetration flow paths intermittently under administrative control;

current TS do provide a similar allowance, but only for one specific

penetration. These changes cannot significantly increase the

probability of an accident because opening an air lock door or a

containment penetration is not, itself, an initiator and does not

affect the items which are initiators of any analyzed accident.

The ability to open the operable door or to open a containment

penetration, even if it means the containment boundary is

temporarily not intact, does not significantly increase the

consequences of an accident previously evaluated because of the low

probability of an event that could pressurize the containment

occurring during the short time the operable door or containment

penetration is expected to be open. In a case where containment

integrity (or containment operability) is lost due to excessive

leakage, both the Palisades Technical Specifications and the

Standard Technical Specifications [STS] allow one hour of continued

operation for its restoration. That time period is allowed without

regard to the magnitude of the potential leakage, and would be

allowed even if both personnel air lock doors [were] leaking

excessively. The additional allowance of permitting the operable

door to be opened momentarily for entry or egress when the other

door is inoperable due to excessive leakage would not significantly

add to the probability of containment leakage and the resultant

consequences of an accident. Similarly, the allowance to open any

containment penetration intermittently under administrative control,

which currently is allowed for one penetration, would not

significantly add to the probability of containment leakage and the

resultant consequences of an accident.

Therefore, operation of the Facility in accordance with proposed

changes G.1, G.2, J.2, and P would not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Change I.2: Change I.2 alters existing LCO 3.6.2, Containment

Pressure so that it no longer applies during Cold Shutdown. LCO

3.6.2 is intended to limit containment pressure to that value used

as an initial condition in the safety analysis. Containment pressure

is an initial condition in analyses which assure that containment

internal pressure will not exceed the containment design values

during a LOCA or MSLB. Containment pressure is not an initiator of

any accident previously evaluated. Neither a LOCA [loss-of-coolant

accident] nor a MSLB [main steam line break] occurring during Cold

Shutdown would pressurize the containment. Therefore, a containment

pressure LCO is not necessary, during Cold Shutdown, to assure that

containment design pressure and temperature is not exceeded. The STS

Containment pressure LCO is not applicable in Cold Shutdown.

Therefore, operation of the Facility in accordance with proposed

change I.2 would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

3. More Restrictive Changes: ``More Restrictive'' changes only

add new requirements, or revise existing requirements to result in

additional operational restrictions. The TS, with all ``More

Restrictive'' changes incorporated, will still contain all of the

requirements which existed prior to the changes. Therefore, ``More

Restrictive'' changes cannot involve a significant increase in the

probability or consequences of an accident previously evaluated.

4. ``Administrative'' changes make wording changes which clarify

existing TS requirements, without affecting their

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technical content. Since ``Administrative'' changes do not alter the

technical content of any requirements, they cannot involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Do the proposed changes create the possibility of a new or

different kind of accident from any previously evaluated?

1. Changes G.1, G.2, J.2, and P: Proposed changes G.1 and P

allow limited access through the operable door of an air lock when

the other door is inoperable; current Technical Specifications do

not. Proposed changes G.2 and J.2 allow unisolating containment

penetration flow paths intermittently under administrative control;

current TS do provide a similar allowance, but only for one specific

penetration. Opening an air lock door or a containment penetration

does not affect the operating conditions or operation of any plant

systems (other than the containment); it does not create a threat to

the integrity of any operating system or alter any system operating

practice or settings.

Since the opening of an air lock door or a containment

penetration only affects the potential leakage from the containment,

and does not affect any of the operating plant systems, operation of

the Facility in accordance with the proposed Technical

Specifications change would not create the possibility of a new or

different kind of accident from any previously evaluated.

2. Change I.2: Change I.2 alters existing LCO 3.6.2, Containment

Pressure so that it no longer applies during Cold Shutdown. LCO

3.6.2 is intended to limit containment pressure to that value used

as an initial condition in the safety analysis. Containment pressure

is an initial condition in analyses which assure that containment

internal pressure will not exceed the containment design values

during a LOCA or MSLB. Neither a LOCA nor a MSLB occurring during

Cold Shutdown would pressurize the containment. Therefore, a

containment pressure LCO is not necessary, during Cold Shutdown, to

avoid creation of a new or different kind of accident. The STS

Containment pressure LCO is not applicable in Cold Shutdown.

Therefore, operation of the Facility in accordance with proposed

change I.2 would not create the possibility of a new or different

kind of accident from any previously evaluated.

3. More Restrictive Changes: ``More Restrictive'' changes only

add new requirements, or revise existing requirements to result in

additional operational restrictions. The TS, with all ``More

Restrictive'' changes incorporated, will still contain all of the

requirements which existed prior to the changes. Therefore, ``More

Restrictive'' changes cannot create the possibility of a new or

different kind of accident from any previously evaluated.

4. ``Administrative'' changes make wording changes which clarify

existing TS requirements, without affecting their technical content.

Since ``Administrative'' changes do not alter the technical content

of any requirements, they cannot create the possibility of a new or

different kind of accident from any previously evaluated.

Do the proposed changes involve a significant reduction in a

margin of safety?

1. Changes G.1, G.2, J.2, and P: Proposed changes G.1 and P

allow limited access through the operable door of an air lock when

the other door is inoperable; current Technical Specifications do

not. Proposed changes G.2 and J.2 allow unisolating containment

penetration flow paths intermittently under administrative control;

current TS do provide a similar allowance, but only for one specific

penetration. The ability to open the operable door or a containment

penetration, even if it means the containment boundary is

temporarily not intact, does not involve a significant reduction in

a margin of safety because of the low probability of an event that

could pressurize the containment occurring during the short time the

operable door or penetration is expected to be open.

Therefore, operation of the Facility in accordance with the

proposed Technical Specifications change would not involve a

significant reduction in a margin of safety.

2. Change I.2: Change I.2 alters existing LCO 3.6.2, Containment

Pressure so that it no longer applies during Cold Shutdown. LCO

3.6.2 is intended to limit containment pressure to that value used

as an initial condition in the safety analysis. Containment pressure

is an initial condition in analyses which assure that containment

internal pressure will not exceed the containment design values

during a LOCA or MSLB. Neither a LOCA nor a MSLB occurring during

Cold Shutdown would pressurize the containment. Therefore,

elimination of a Cold Shutdown LCO for containment pressure would

not affect the post-accident pressure or temperature. Since peak

post accident [pressure] and temperature would be unaffected by the

proposed change, operation of the Facility in accordance with

proposed change I.2 would not involve a significant reduction in a

margin of safety.

3. More Restrictive Changes: ``More Restrictive'' changes only

add new requirements, or revise existing requirements to result in

additional operational restrictions. The TS, with all ``More

Restrictive'' changes incorporated, will still contain all of the

requirements which existed prior to the changes. Therefore, ``More

Restrictive'' changes cannot involve a significant reduction in a

margin of safety.

4. ``Administrative'' changes make wording changes which clarify

existing TS requirements, without affecting their technical content.

Since ``Administrative'' changes do not alter the technical content

of any requirements, they cannot 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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Project Director: John N. Hannon.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendments request: October 29, 1997.

Description of amendments request: The proposed amendments to the

Technical Specifications (TS) for the Brunswick Steam Electric Plant

(BSEP) Units 1 and 2 would revise the description of the control rod

assemblies (CRAs) in TS 5.3.2. The proposed revision was requested to

support replacement of a portion of the BSEP Unit 1 CRAs during that

unit's next refueling outage with assemblies of a different design.

Carolina Power & Light Company, the licensee, has proposed adopting the

description of CRAs used in NUREG-1433, Revision 1, ``Standard

Technical Specifications General Electric Plants, BWR/4,'' which

includes the number and shape of CRAs and a stipulation that NRC-

approved absorber material be used in CRAs. The more detailed

description in the current TS of CRAs would be relocated to the Updated

Final Safety Analysis Report. The licensee has stated that the CRA

description proposed for TS 5.3.2 will be sufficient to ensure that any

future changes in CRA design that may affect safety will require prior

NRC review and approval.

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 amendments do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Relocation of the control rod assembly descriptive information

from the Technical Specifications to the Updated Final Safety

Analysis Report will ensure that adequate control of the information

is maintained. Any changes to this design information must conform

with the requirements of 10 CFR 50.59. Restricting use of control

rod assembly absorber materials to those listed, or to materials

that have been approved by the NRC, will ensure any changes which

may affect safety to require prior NRC review and approval. Since

the information with a potential to affect safety is sufficiently

addressed by the Technical Specifications, the criteria of 10 CFR

50.36(c)(4) for

[[Page 66138]]

including the relocated information as Design Features are not met.

Because the relocated information is not required to be in the

Technical Specifications to provide adequate protection of the

public health and safety, relocation of control rod assembly

descriptive information will not increase either the probability or

the consequences of an accident previously evaluated.

2. The proposed amendments would not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

Relocation, to the Updated Final Safety Analysis Report, of the

information pertaining to the control rod assembly designs ensures

that adequate control of the information will be maintained. Since

the information with a potential to affect safety is sufficiently

addressed by the Technical Specifications, the criteria of 10 CFR

50.36(c)(4) for including the relocated information as Design

Features are not met. Because the relocated information is not

required to be in the Technical Specifications to provide adequate

protection of the public health and safety, the proposed Technical

Specification changes to relocate the control rod assembly design

information to the Updated Final Safety Analysis Report does not

create the possibility of a new or different kind of accident from

any accident previously evaluated.

3. The proposed license amendments do not involve a significant

reduction in a margin of safety.

As discussed in Items 1 and 2 above, relocation of the control

rod assembly descriptive information from the Technical

Specifications to the Updated Final Safety Analysis Report will

ensure that adequate control of the information is maintained. Any

changes to this design information must conform with the

requirements of 10 CFR 50.59. Restricting use of control rod

assembly absorber materials to those listed, or to materials that

have been approved by the NRC, will ensure any changes which may

affect safety to require prior NRC review and approval. The

information with a potential to affect safety is sufficiently

addressed by the Technical Specifications, therefore, the proposed

Technical Specification changes to relocate control rod assembly

design information to the Updated Final Safety Analysis Report 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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: James E. Lyons.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: July 16, 1997, as supplemented October

30, 1997.

Description of amendment request: The amendment would update

License condition 2.C(4) to reflect the latest revision levels of the

Oyster Creek Security Training and Qualification Plan, License

Amendment Request No. 252.

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:

GPU Nuclear has concluded that the proposed changes to the

Security Plan do not involve a significant hazard consideration. In

support of this determination, an evaluation of each of the three

standards set forth in 10 CFR 50.92 is provided below.

(1) The proposed changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Security Plan provisions are not associated with design basis

accident initiators nor do they constitute part of any mitigation

system. Therefore, the probability and consequences of accidents are

not increased.

(2) The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The Security Plan changes do not create new or change existing

physical interfaces with plant equipment. Therefore, the changes do

not create the possibility of a new or different kind of accident.

(3) The proposed changes do not involve a significant reduction

in a margin of safety.

Margins associated with reactor and fuel storage nuclear safety

are not affected by the proposed Security Plan changes since neither

physical nor procedural changes to associated systems, structures

and components are involved. Vital area security measures, which are

reduced, are compensated by commitments to hold contingency drills

at a frequency sufficient to maintain response capability for

response personnel and to use organic-type X-ray equipment.

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: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,

Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037

NRC Project Director: Ronald B. Eaton, Acting Director.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of amendment request: November 14, 1997.

Description of amendment request: The proposed change to Technical

Specification 4.5.2.d.1 will clarify the wording and increase the

setpoint for the open pressure interlock (OPI).

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:

NNECO has reviewed the proposed revision in accordance with 10

CFR 50.92 and has concluded that the revision does not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10 CFR 50.92(c) are not

satisfied. The proposed revision does not involve [an] SHC because

the revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

Increasing the Technical Specification Open Pressure Interlock

(OPI) pressure to 412.5 psia [pounds per square inch--atmospheric]

will still maintain the required function of preventing the MOVs

[motor operated valves] from opening inadvertently. The increased

pressure is within the design limits of the RHR [residual heat

removal] piping system and components. The pressure signal is

generated from a transmitter and results in an electronic input to

the bistable. This is a clarification of the conditions under which

the OPI is tested.

Therefore, the proposed revision does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

There is no change to the function of the OPI. The protection

provided by the interlock remains intact. The Technical

Specification OPI pressure has been raised to take into account

instrument accuracies and reset deadbands. The RHR system design

pressure remains protected from being exceeded by inadvertent

opening of the isolation MOVs. The method for the OPI surveillance

is clarified by clearly stating that the bistable receives a

simulated transmitter signal representative of the process pressure.

[[Page 66139]]

Therefore, the proposed revision does not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

3. Involve a significant reduction in a margin of safety.

The design pressure of the RHR system is 600 psig [pounds per

square inch--gauge]. The most limiting case is to prevent the RHR

pump developed head pressure from exceeding the design pressure when

aligned to the RCS [reactor coolant system] as suction pressure. RHR

pump testing has determined that a maximum pump differential

pressure of 195 psi [pounds per square inch] exists for deadhead/no

flow conditions. Therefore, to maintain the 600 psig design pressure

limit, RCS/suction pressure must be limited to 405 psig (420 psia,

assuming a 15 psi conversion from psig to psia). The proposed

maximum pressure, including setpoint tolerances and reset deadbands,

is less than this value; i.e. 412.5 psia. Head corrections due to

elevation differences are considered to be insignificant.

Therefore, the proposed revision does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve an SHC.

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: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

Connecticut.

NRC Deputy Director: Phillip F. McKee.

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: November 3, 1997.

Description of amendment request: The proposed amendment would

revise Sections 1, 3.1, 3.3, 4.3, and 6 of Appendix A of the Indian

Point 3 Technical Specifications. These revisions extend the Heatup-

Cooldown limits from 11 to 13 effective full power years (EFPYs),

provide the corresponding Overpressure Protection System (OPS) limits,

relocate the new pressure temperature limit curves and low-temperature

overpressurization protection (LTOP) system limits to the pressure

temperature limit report (PTLR) and include some minor revisions which

ensure specification clarity and conservatism.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response: The proposed license amendment does not involve a

significant increase in the probability or consequences of a

previously analyzed accident. The pressure-temperature limit changes

proposed by this amendment are based on supporting data and

evaluation methodologies previously submitted to the NRC in

Reference 3 [see application dated November 3, 1997] and approved as

Amendments 109 and 121 (References 4 and 5) [see application dated

November 3, 1997]. These limits are based upon the irradiation

damage prediction methods of Regulatory Guide 1.99, Revision 2. The

LTOPs changes contained in this submittal have been conservatively

adjusted in accordance with the new pressure-temperature limits, in

accordance with the methodology contained in Reference 3 and ASME

Code Case N-514.

The relocation of the pressure-temperature and LTOPs limits from

the Technical Specifications to the PTLR does not eliminate the

requirement to operate in accordance with the limits specified in 10

CFR [Part] 50, Appendix G. The requirement to operate within the

limits in the PTLR is specified in and controlled by the Technical

Specifications.

The revised version of Section 3.1.A.8 clarifies existing

requirements related to the OPS system and adds an eight hour

completion time for compensating actions, consistent with the STS

[standard technical specifications]. The changes to Section

3.1.A.1.h, i, and j revise the requirements associated with the

start of an RCP [reactor coolant pump]. These changes improve

specification clarity and do not increase the probability or

consequences of an accident.

The Technical Specification changes associated with the

restriction on SI [safety injection] pumps provides added

conservatism to the Technical Specifications and limits the

likelihood of an RHR [residual heat removal] overpressurization

event. Current plant procedures prohibit actuation of any SI pumps

when RHR is in service, except during testing, loss of RHR cooling,

or reduced inventory operations. Therefore, the change to the

Technical Specifications will not alter current plant operation.

(2) Does the proposed license amendment create the possibility

of a new or different kind of accident from any accident previously

evaluated?

Response: The proposed license amendment does not create the

possibility of a new or different kind of accident from any accident

previously analyzed. The pressure-temperature limits are updating

the existing limits by taking into account the effects of radiation

embrittlement, utilizing criteria defined in Regulatory Guide 1.99,

Revision 2, and extending the effective period to 13 EFPYs. The

updated OPS limits have been adjusted to account for the effect of

irradiation on the limiting reactor vessel material. These changes

do not affect the way the pressure-temperature or OPS limits provide

plant protection and no physical plant alterations are necessary.

The relocation of the pressure-temperature and OPS limits from the

Technical Specifications to the PTLR does not alter the requirements

associated with these limits.

The revisions to Section 3.1.A.8 concerning the OPS system

improve on the clarity of existing specifications and add a

completion time for compensating actions that is consistent with the

STS. These changes do not involve any hardware modifications and do

not affect the function of the OPS system.

The revisions concerning the operation of SI pumps bring the

Technical Specifications into line with current operating

procedures. The changes to Specification 3.1.A.1.h, i, and j provide

specification clarity and are more conservative than existing

Technical Specifications. Therefore, the changes cannot create the

possibility of a new or different kind of accident.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response: The proposed amendment does not involve a significant

reduction in a margin of safety. The margins of safety against

fracture provided by the pressure-temperature limits are those

limits specified in 10 CFR Part 50, Appendix G and ASME Boiler and

Pressure Vessel Code Section XI, Appendix G. The guidance in these

documents has been utilized to develop the pressure-temperature

limits with the requisite margins of safety for the heatup and

cooldown conditions. The new LTOP limits are based upon Reference 3

and ASME Code Case N-514. The relocation of the pressure-temperature

and OPS limits to the PTLR does not alter the requirements

associated with these limits.

The revisions to Section 3.1.A.8 clarify the requirements

associated with the OPS system. The revisions associated with the

operation of SI pumps with RHR in service (Sections 3.3.A.8, 9 and

10) and the changes regarding RCP starts (Section 3.1.A.1.h, i, and

j) are more conservative than the current Technical Specifications,

and are consistent with plant operating procedures. Therefore, they

do not reduce a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

[[Page 66140]]

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa.

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 amendment request: October 24, 1997.

Description of amendment request: The amendments would increase the

containment hydrogen analyzer surveillance frequency in Technical

Specification 4.6.4.1 from once per refueling outage to quarterly.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The containment hydrogen analyzers provide control room

indication of hydrogen concentration in the containment atmosphere.

They do not affect the probability of any previously evaluated

accident. The proposed change would increase the calibration

frequency specified in TS 4.6.4.1 to make it consistent with

manufacturer's recommendations and the current calibration frequency

at [Salem Generating Station] SGS as imposed by administrative

controls. The change in TS-required calibration frequency is in the

conservative (more frequent) direction, to ensure that potential

degradation of the sensor electrolyte over time would not result in

unacceptable performance of the hydrogen analyzers. The change in

specified frequency would not adversely affect the consequences of

any previously evaluated accident.

2. Proposed change does not create the possibility of a new or

different kind of accident from any accident previously analyzed

[evaluated].

The proposed change affects only the specified calibration

frequency of the containment hydrogen analyzers. The proposed change

does not affect the design of any SGS structure, system or

component, nor would it result in any new plant configuration.

Therefore, it does not create the possibility of a new or different

kind of accident.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change to the containment hydrogen analyzer

calibration frequency does not affect the design or operating limits

of any SGS structure, system or component. The change would make the

specified calibration frequency more conservative, to ensure the

hydrogen analyzers perform as designed over time. The proposed

change does not involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Salem Free Public library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Project Director: John F. Stolz.

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 amendment request: October 24, 1997.

Description of amendment request: The amendments would revise

Technical Specification (TS) 3/4.7.7, ``Auxiliary Building Exhaust Air

Filtration System.'' The revisions would: (1) Require both Auxiliary

Building Ventilation (ABVS) supply fans to be operable, (2) require all

three ABVS exhaust fans to be operable, (3) align ABVS TSs to be

consistent with current TS bases and recently revised system

descriptions in the Salem Updated Final Safety Analysis Report (UFSAR),

(4) assure that negative pressure is maintained in the Auxiliary

Building under all postulated single active failures, (5) clarify

required Engineered Safety Feature filter testing, (6) provide

consistency between Unit 1 and Unit 2 TSs, and (7) for Unit 2 only,

remove the requirement to verify safety injection auto-start

capabilities.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change alters the number of fans which must be

OPERABLE to ensure that a sufficient number of supply and exhaust

fans will be operable, following a most limiting single failure, to

mitigate the consequences of design basis accidents. The changes to

the ABVS surveillance requirements still provide an appropriate

means for demonstrating the operability of the ABVS.

The ABVS cannot initiate or otherwise cause any accident or

operational transient evaluated in the UFSAR. Consequently, the

probability of such events is not increased. The ABVS cannot

increase the consequences of a design basis LOCA unless: (1)

Auxiliary Building negative pressure is lost, resulting in

uncontrolled, ground level release of radioactive material; (2) ABVS

carbon adsorbers are bypassed, resulting in uncontrolled release of

radioactive iodine from the plant vent; or (3) Auxiliary Building

temperatures are not controlled, resulting in failure of accident

mitigating equipment.

By requiring OPERABILITY of all ABVS supply and exhaust fans,

the proposed changes contained in this submittal assures Auxiliary

Building negative pressure is maintained under all postulated post-

accident, single-failure scenarios. The proposed changes to ABVS

will not affect the elemental iodine adsorption capability of the

system. Finally, engineering analyses conclude that these fan

combinations, with single-active failures of the fans or their

support systems considered, provide sufficient Auxiliary Building

ventilation. Under the most limiting temperature conditions, the

fans will maintain room temperatures within design limits.

Accordingly, the consequences of a design basis LOCA, hence

applicable design basis accidents or operational transients, are not

increased.

Therefore, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

ABVS supply fans are not considered essential to the primary

safety-function of preventing or mitigating radioactive releases,

nor are they currently required to be OPERABLE. Similarly, accident

analyses take no credit for operation of supply fans. Accordingly,

malfunctions of vital buses and ABVS exhaust fans are the only

malfunctions of active ABVS related equipment important to safety

that are previously evaluated.

The probability of failure of a vital bus is not increased by

this proposal since the proposal has no direct effect on electrical

power. Neither is the probability of exhaust fan failure increased

by the proposal, since exhaust fans are not affected by this

proposal, except that the number that must be OPERABLE is increased

from two to three.

By requiring additional supply fans and exhaust fans to be

OPERABLE, no single failure of either a vital bus or ABVS fan

prevents (1) maintenance of negative Auxiliary Building pressure or

(2) maintenance of temperatures within design limits. Since ABVS

supply and exhaust fans cannot initiate accidents, increasing the

number of fans required to be OPERABLE cannot create the possibility

of a new or different kind of accident from any accident previously

evaluated. In addition, the proposed changes to the ABVS

surveillance testing concern ABVS leakage, HEPA filter and carbon

adsorber capabilities, and laboratory test methods. Therefore, the

proposed surveillance requirement changes would have no impact on

the initiation of accidents.

Thus, the proposed changes do not create the possibility of a

new or different kind of

[[Page 66141]]

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 is dependent upon the maintenance of

specific operating parameters within designated design limits. Since

iodine removal capability is not affected by the proposed changes,

and negative Auxiliary Building pressure and temperatures will

continue to be maintained within existing design limits under post-

accident conditions, including consideration of the most limiting

single active failure, the margin of safety is not reduced. By

imposing new restrictions on the allowed outage times of ABVS

components, the margin of safety is increased with the proposed

changes to the ABVS Technical Specification Limiting Condition for

Operation (LCO).

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: Salem Free Public library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Project Director: John F. Stolz.

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 amendment request: November 4, 1997.

Description of amendment request: The amendments would change

Technical Specification (TS) 3/4.6.2, ``Containment Spray System,'' to

verify on recirculation flow that the containment spray pumps develop a

differential pressure of at least 204 psi.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change revises the CS [containment spray] pump

technical specification surveillance test acceptance from pump

discharge pressure to pump differential pressure. This will account

for the effect of RWST [refueling water storage tank] level on test

results and provide acceptance criteria that verifies each CS pump

performs as assumed in the accident analyses. This surveillance test

is also being added to the Salem Unit 1 TS. The proposed change does

not alter the physical plant arrangement or the method of CS pump

inservice testing. Therefore it does not increase the probability of

an accident. There is no change to pump performance requirements as

assumed in the accident analyses. There is no change to CS system

performance in response to an accident. Therefore, the proposed

change does not involve an increase in the consequences of an

accident previously evaluated.

The proposed change also corrects a typographical error by

removing a repeated word. This change does not involve an increase

in 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 revises the Salem Unit 2 CS pump

surveillance test acceptance criteria from pump discharge pressure

to pump differential pressure. This will account for the effect of

RWST level on test results and provide acceptance criteria that

verify the CS pumps perform as assumed in the accident analyses.

This surveillance test is also being added to the Salem Unit 1 TS.

The proposed change does not alter the plant configuration. The

change does not alter the method of performing inservice testing on

the CS pumps. The change does not alter the CS pump performance

assumed in the accident analyses. Therefore, the change will not

create the possibility of a new or different kind of accident from

any accident previously evaluated.

The proposed change also corrects a typographical error by

removing a repeated word. This change will not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change ensures the CS pump Salem Unit 2 TS

surveillance test acceptance criteria verify CS pump performance as

assumed in the accident analyses accounting for RWST level effects.

This surveillance test is also being added to the Salem Unit 1 TS.

The proposal does not change the CS pump performance requirements

assumed in the accident analyses and thus does not reduce the margin

of safety.

The proposed change also corrects a typographical error by

removing a repeated word. This does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Salem Free Public library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Project Director: John F. Stolz.

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 amendment request: November 14, 1997.

Description of amendment request: The proposed changes to the

Technical Specifications (TSs) include administrative and editorial

changes to correct errors in the TSs that have either existed since

initial issuance or were introduced during subsequent changes. In

addition, surveillance requirements are added that are considered

administrative changes since the surveillances should have been

incorporated with the TS when the applicable amendment to the TSs was

approved by the NRC.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to the TS are administrative or editorial

changes to the TS and do not involve any physical changes to the

plant. The administrative changes and editorial changes do not

delete any existing surveillance requirements or delete any

requirements from the Limiting Condition for Operations (LCOs) or

Action Statements and therefore do not reduce the actions that are

currently taken in the TS to demonstrate operability of plant

structures, systems, or components (SSCs). The additional

surveillance requirements that are being added to the TS including

the new surveillances correct past administrative errors and should

have been incorporated within the TS as part of the approved

Amendments to the TS. These changes will provide additional

assurance that SSCs perform their intended safety functions.

Surveillance testing has been and is currently being performed for

the surveillance requirements that should have been incorporated and

are now administratively being added to the TS. Since these changes

do not modify any SSCs or reduce the current requirements for

demonstrating operability of these SSCs or reduce the current

requirements for demonstrating operability of these SSCs, the

proposed changes to the TS 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.

[[Page 66142]]

The proposed changes to the TS are administrative and editorial

corrections to the TS that do not affect the ability of the plant

systems to meet their current TS requirements or design basis

functions. There is no reduction in the current surveillance

requirements required to demonstrate the operability of plant SSCs.

These changes also do not involve any physical changes to plant

SSCs. Therefore the proposed changes do not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes are administrative and editorial

corrections to the TS that do not affect the ability of plant SSCs

to perform their design basis accident functions. There is no

reduction in the current surveillance requirements required to

demonstrate the operability of plant SSCs. 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: Salem Free Public library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.