Biweekly Notice

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from August 4, 1995, through August 18, 1995. The

last biweekly notice was published on August 16, 1995 (60 FR 42597).

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 Rules Review and

Directives Branch, Division of Freedom of Information and Publications

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

Washington, DC 20555, 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 September 29, 1995, 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.,

[[Page 45173]]

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, 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

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, 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 Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendments request: August 3, 1995

Description of amendments request: The proposed amendment changes

would add the analytical method supplement entitled ``Fuel Rod Maximum

Allowable Gas Pressure,'' CEN-372-P-A, dated May 1990, and its

associated Nuclear Regulatory Commission Safety Evaluation Report,

dated April 10, 1990, to the list of analytical methods in TS 6.9.1.10

used to determine the PVNGS core operating limits.

Basis for proposed no significant hazards consideration

determination: As requied 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 does not involve any change to the

configuration or method of operation of any plant equipment that is

used to mitigate the consequences of an accident. The proposed

change adds an NRC approved methodology and its associated Safety

Evaluation Report (SER), to the list of analytical methods used to

determine the core operating limits. The use of this methodology

ensures that the consequences of an accident remain within the

limits established by existing analyses. They do not alter any of

the assumptions or bounding conditions currently in the UFSAR.

The U3C6 ECCS performance analysis included the analysis of the

impact of the maximum calculated fuel rod gas pressures on the

timing of cladding rupture and on the peak cladding temperature.

This analysis concluded that the peak cladding temperature for Cycle

6 remained below that of the analysis of record and that the peak

cladding temperature continued to occur at

[[Page 45174]]

low burnup, specifically the burnup corresponding to the maximum

initial fuel stored energy.

In addition to the LOCA analysis a DNB propagation analysis was

performed to demonstrate that DNB propagation does not occur during

postulated accidents that experience DNB when pressure in a fuel pin

is higher than the system pressure. This analysis was performed

using the fuel rod strain model described in CEN-372-P-A.

Based on these analyses, there is no 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 change does not involve any change to the

configuration or method of operation of any plant equipment that is

used to mitigate the consequences of an accident. Accordingly, no

new failure modes have been defined for any plant system or

component important to safety nor has any new limiting failure been

identified as a result of the proposed change. The intent of the

proposed change is to utilize a new analytical method to ensure that

the consequences of any equipment malfunction remain within the

limits of existing analyses resulting in no impact on radiological

consequences.

The impact of the maximum fuel rod gas pressures calculated for

U3C6 was evaluated as part of the Cycle 6 ECCS performance analysis.

Except for the highest burnup analyzed, the time of cladding rupture

decreased as the initial fuel rod gas pressure increased with

burnup. However, the peak cladding temperature occurred at the

burnup with the maximum initial fuel stored energy. The analysis

also determined that the ECCS performance analysis for U3C6 is

bounded by that of the reference cycle analysis.

An evaluation was conducted to ensure that fuel would not

experience DNB propagation when the pressure in a fuel pin is higher

than the system pressure. DNB was shown not to propagate by

demonstrating that the degree of cladding deformation is no more

than the limit defined by the fuel rod maximum pressure Topical

Report (CEN-372-P-A).

Therefore, it can be concluded that the proposed change to

Section 6.9.1.10 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.

The proposed change adds an NRC approved Topical Report

(methodology) and its associated SER, to the list of analytical

methods used to determine core operating limits. The use of the new

methodology ensures that safety margins are maintained within the

results of existing calculations. Since the core operating limits

will continue to be established by an NRC approved methodology and

will provide adequate core protection, the proposed amendment does

not involve a significant reduction in the margin of safety.

Analyses were conducted to determine the impact of higher fuel

rod pressure on ECCS performance and DNB propagation. The results of

the analyses show that the effects of higher fuel rod pressure are

bounded by previous results.

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

that review, it appears that the three standards of 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: July 13, 1995

Description of amendments request: The proposed amendments would

revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2,

Technical Specifications (TSs) Section 5.2.1, ``Fuel Assemblies.'' The

current TSs only allow fuel that is clad with either zircaloy or ZIRLO.

The proposed change would allow the use of cladding material other than

zircaloy or ZIRLO with an approved exemption. Thus, the proposed change

will eliminate the need for future amendments to allow the use of

different cladding material for which the Commission has issued an

exemption.

Basis for proposed no significant hazards consideration

determination: As requied 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.

Calvert Cliffs Technical Specification 5.2.1, Fuel Assemblies,

states that fuel rods are clad with either zircaloy or ZIRLO. This

reflects the requirements of 10 CFR 50.44, 50.46, and 10 CFR [Part]

50, Appendix K, which also restrict fuel rod cladding materials to

zircaloy or ZIRLO. Baltimore Gas and Electric Company proposes to

insert fuel assemblies into Calvert Cliffs Unit 1 which have some

fuel rods clad in zirconium alloys that do not meet the definition

of zircaloy or ZIRLO for testing purposes and has applied for an

exemption to the regulations to allow that change. The proposed

change to the Calvert Cliffs Technical Specifications will allow the

use of cladding materials that are not zircaloy or ZIRLO with an

approved exemption in accordance with 10 CFR 50.12.

The proposed change to the Unit 1 and Unit 2 Technical

Specifications will allow the use of fuel rod cladding materials

other than zircaloy or ZIRLO as long as those materials have been

approved by an exemption to the regulations. To obtain approval of

new cladding materials, 10 CFR 50.12 requires that the applicant

show that the proposed exemption is authorized by law, is consistent

with the common defense and security, will not present an undue risk

to the public health and safety; and is accompanied by special

circumstances.

Under the proposed change, any fuel rod cladding materials that

are not zircaloy or ZIRLO must still be approved by the Nuclear

Regulatory Commission (NRC) prior to use under 10 CFR 50.12. This

change to the Technical Specifications allows the NRC to approve the

use of cladding materials that are not either zircaloy or ZIRLO

under 10 CFR 50.12 and not require an additional approval under 10

CFR 50.90. As such, the proposed change eliminates a duplicative

regulatory requirement and would have no effect on the probability

or consequences of an accident.

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 change eliminates a duplicated approval requirement

and would have no effect on the possibility of a new or different

type of accident. The proposed change to the Technical

Specifications would allow the NRC to approve the use of fuel rod

cladding materials that are not either zircaloy or ZIRLO under 10

CFR 50.12 and not require an additional approval under 10 CFR 50.90.

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 proposed change eliminates a duplicated approval requirement

and will have no effect on the margin of safety. The proposed change

to the Technical Specifications would allow the NRC to approve the

use of fuel rod cladding materials that are not either zircaloy or

ZIRLO under 10 CFR 50.12, and not require an additional approval

under 10 CFR 50.90.

Therefore, the proposed change does not involve a significant

reduction in a margin 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

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

[[Page 45175]]

Trowbridge, 2300 N Street, NW., Washington, DC 20037

NRC Project Director: Ledyard B. Marsh

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Units 1 and 2, Will County, Illinois Docket Nos.

STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, Ogle

County, Illinois Docket Nos. 50-237 and 50-249, Dresden Nuclear

Power Station,Units 2 and 3, Grundy County, Illinois Docket Nos.

50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle

County, Illinois Docket Nos. 50-254 and 50-265, Quad Cities Nuclear

Power Station, Units 1 and 2, Rock Island County, Illinois Docket

Nos. 50-295 and 50-304, Zion Nuclear Power Station, Units 1 and 2,

Lake County, Illinois

Date of application for amendment requests: April 24, 1995

Description of amendment requests: The licensee proposes to amend

Section 6 of the Technical Specifications of all ComEd stations to make

the following changes: (1) delete the ``Review, Investigative and Audit

Functions'' sections, in their entirety, and relocate these

requirements to appropriate sections of the ComEd Quality Assurance

Topical Report, (2) change titles to reflect the reorganization of

ComEd's Nuclear Operations Division, and (3) miscellaneous

administrative and editorial changes.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

A. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

(1) The proposed relocation of the ``Review, Investigative and

Audit Functions'' sections of Technical Specifications to the QA

Topical Report does not affect any accident initiators or

precursors, and does not change or alter the design assumptions for

the systems and components used to mitigate the consequences of an

accident.

The relocation of these sections is consistent with the

recommended changes specified in the October 25, 1993 letter from W.

T. Russell (USNRC) to the Chairpersons of the Owner Groups'

Technical Specifications Committees, entitled, ``Content of Standard

Technical Specifications, Section 5.0, Administrative Controls''.

Relocating these requirements to the QA Topical Report will

continue to ensure that proposed future changes to these

requirements will receive proper regulatory oversight. NRC review of

the Quality Assurance Program is governed by 10CFR50.54.

10CFR50.54(a)(3) states: ``Changes to the quality assurance program

description that do not reduce the commitments must be submitted to

the NRC in accordance with the requirements of 50.71. Changes to the

quality assurance program description that do reduce the commitments

must be submitted to NRC and receive NRC approval prior to

implementation, ...'' Based on these 10CFR50.54 requirements,

appropriate licensee and regulatory control of the requirements in

the subject relocated Technical Specification sections will be

maintained.

(2) The proposed title and organizational changes to Section 6

of Technical Specifications do not affect any accident initiators or

precursors and do not change or alter the design assumptions for the

systems or components used to mitigate the consequences of an

accident.

Commonwealth Edison's organizational changes allow for increased

senior management attention and oversight of station activities.

Position titles and associated responsibilities have changed to

increase the company's efficiency in the management of its nuclear

stations. These administrative changes do not reduce any

requirements or commitments. The proposed changes enhance the

administrative controls necessary to ensure safe plant operation.

(3) Other proposed administrative/editorial changes simply make

corrections or provide needed clarification prompted by the

reorganization. These changes provide consistency with station

procedures, programs, other Technical Specifications, and Standard

Technical Specifications. They are administrative in nature and do

not impact any accident previously evaluated in the UFSAR.

In conclusion, none of the proposed changes involve a

significant increase in the probability or consequences of an

accident previously evaluated.

B. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

(1) The proposed relocation of the ``Review, Investigative and

Audit Functions'' sections of Technical Specifications to the QA

Topical Report does not affect the design or operation of any

system, structure, or component in the plant. There are no changes

to parameters governing plant operation and no new or different type

of equipment will be installed that could give rise to a new or

different kind of accident that was previously evaluated.

The proposed changes are considered to be administrative or

programmatic in nature and do not affect equipment or components

that could initiate an accident. All administrative commitments

being relocated to the QA Topical Report will continue to receive

appropriate regulatory oversight pursuant to 10CFR50.54.

(2) The proposed title and organization changes do not affect

the design or operation of any system, structure, or component in

the plant. There are no changes to parameters governing plant

operation; no new or different type of equipment will be installed.

The proposed changes are considered to be administrative changes

that will enhance the performance of organizations responsible for

the safe operation of the plant to respond to plant transients or

emergencies. All responsibilities described in Technical

Specifications for management activities will continue to be

performed by qualified individuals.

(3) All other proposed changes are administrative in nature and

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

from any accident previously evaluated.

In conclusion, the proposed changes do not create the

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

previously evaluated.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

(1) The proposed changes are administrative or programmatic in

nature and do not affect the margin of safety for any safety

parameters and setpoints addressed in Technical Specifications. The

assumptions, initial conditions and methodologies used in the

accident analyses remain unchanged, therefore, accident analyses

results are not impacted.

Placing these requirements in QA Topical Report will continue to

ensure that proposed future changes to these requirements will

receive proper regulatory oversight pursuant to 10CFR50.54.

(2) The proposed title and organizational changes are

administrative in nature and do not affect the margin of safety for

any Technical Specification. The initial conditions and

methodologies used in the accident analyses remain unchanged,

therefore, accident analyses results are not impacted.

(3) All other proposed changes are administrative in nature and

have no impact on the margin of safety for any Technical

Specification.

In conclusion, 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: for Braidwood, the Wilmington

Public Library, 201 S. Kankakee Street, Wilmington, Illinois 60481; for

Byron, the Byron Public Library District, 109 N. Franklin, P.O. Box

434, Byron, Illinois 61010; for Dresden, Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450; for LaSalle,

Jacobs Memorial Library, Illinois Valley Community College, Oglesby,

Illinois 61348; for Quad Cities, Dixon Public Library, 221 Hennepin

Avenue, Dixon, Illinois 61021; for Zion, Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085

[[Page 45176]]

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

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 Docket Nos. 50-373 and 50-374, LaSalle County

Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: June 8, 1995

Description of amendment request: The proposed amendments would

revise Technical Specifications Section 3/4.8, Electrical Power

Systems, and the associated Bases for LaSalle County, Byron, and

Braidwood Stations. The proposed changes revise surveillance and

administrative requirements associated with emergency diesel generators

(EDGs) in accordance with the guidance of NRC Generic Letter 94-01,

``Removal of Accelerated Testing and Special Reporting Requirements for

Emergency Diesel Generators,'' Generic Letter 93-05, ``Line-Item

Technical Specifications Improvements to Reduce Surveillance

Requirements for Testing During Power Operation,'' and Regulatory Guide

(RG) 1.9, ``Selection, Design, Qualification, and Testing of Emergency

Diesel Generator Units Used as Class 1E Onsite Electric Power Systems

at Nuclear Power Plants.'' The proposed changes include: (1)

eliminating increased testing requirements for EDGs, (2) eliminating

special reporting requirements for EDGs, (3) eliminating the semi-

annual fast load test and replacing it with a requirement to load EDGs

semi-annually in accordance with the vendor recommendations for all

test purposes other than the refueling outage Loss of Offsite Power

(LOOP) tests, (4) de-coupling the 24-hour endurance run and the LOOP/

loss-of-coolant (LOCA) (LOOP only for LaSalle) sequencing requirements

for the hot start test, (5) removing RG 1.108 references to testing

requirements, (6) eliminating testing requirements when an EDG becomes

inoperable due to an inoperable support system, an independently

testable component, or preplanned maintenance or testing, or if there

is not a potential common mode failure for the remaining diesel

generator, (7) deleting the requirement for inspecting the EDGs in

accordance with procedures prepared in conjunction with its

manufacturer's recommendations, and (8) making editorial changes.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated:

The proposed changes do not affect accident initiators or

precursors and do not alter the design assumptions affecting the

ability of the EDGs to mitigate the consequences of an accident.

Deleting the special reporting requirements from the Technical

Specifications is administrative. ComEd will continue to notify the

Commission of significant EDG failures in accordance with 10 CFR

50.72 and 50.73 criteria.

Excessive testing requirements have proven to be a contributor

to increased equipment degradation. Removing inappropriate and

redundant requirements increases EDG reliability and enhances the

ability of EDGs to mitigate the consequences of an accident.

Implementing ComEd's alternative to the maintenance rule for the

EDGs provides additional assurance that high EDG performance will be

maintained.

EDG equipment degradation will be reduced by eliminating the

semi-annual fast load test for EDGs in accordance with the vendor

recommendations for test purposes other than the refueling outage

Loss of Offsite Power (LOOP) tests. This improves EDG reliability

and availability and further enhances their ability to mitigate the

consequences of an accident. The LOOP test would still be performed

to provide assurance that the EDG is capable of responding to a LOOP

as assumed in the accident analyses.

De-coupling the 24 hour endurance test and the LOOP/LOCA (for

LaSalle, LOOP) sequencing test requirements for the hot start test

has no effect on accident mitigation. Demonstrating diesel generator

hot restart capability without loading the engine does not

invalidate or reduce the effectiveness of the hot restart test. The

hot restart test can be conducted in any plant condition since its

performance at power will have no adverse effect on plant

operations.

The proposed editorial changes are administrative in nature.

They improve readability and provide consistency with current

industry guidance.

Therefore, the proposed changes do not involve an increase in

the probability or consequences of an accident previously evaluated.

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

from any accident previously evaluated:

The proposed changes do not alter the ability of the EDGs to

perform their intended function to mitigate the consequences of an

initiating event within the acceptance limits assumed in plant

safety analyses. The proposed changes have no impact on component or

system interactions, or the plant design basis.

Instrumentation setpoints, starting, sequencing, and loading

functions associated with EDGs are not affected by the proposed

changes. Furthermore, combining the alternate EDG system maintenance

rule implementation program with the proposed amendment will enhance

both the availability and the performance of the EDGS.

Therefore, there is not a potential for creating the possibility

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

evaluated.

3) Involve a significant reduction in a margin of safety:

The proposed changes do not increase the probability or

consequences of an accident, and there is no impact on equipment

design or operation. The proposed changes do not affect the results

of accident and transient analyses. Plant and system response to an

initiating event will remain in compliance within the assumptions of

safety analyses. There is no associated change to the type, amount,

or control of radioactive effluents, nor is there an associated

increase in individual or cumulative occupational radiation

exposure. There is no effect upon the capabilities of the associated

systems to perform their intended functions within the allowed

response times assumed in safety analyses.

The proposed changes are compatible with plant operating

experience and are consistent with the guidance provided in NUREG-

1366, Generic Letters 93-05 and 94-01, and Regulatory Guide 1.9. In

two instances ComEd's proposed changes deviate from these guidance

documents. However, the changes are consistent with the intent of

the documents or other NRC guidance documents. Eliminating excessive

testing requirements can improve safety by reducing challenges to

plant systems and reducing equipment wear and degradation. While the

proposed changes affect surveillance intervals; there are no changes

to the methods used to perform the surveillances.

EDG reliability and availability will be improved by the

proposed changes. The surveillances will continue to demonstrate the

ability of the EDGs to perform their intended function of providing

electrical power to the emergency safety systems needed to mitigate

design basis transients. No margin of safety is reduced.

Guidance has been provided in ``Final Procedures and Standards

on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,

for the application of standards to license change requests for

determination of the existence of significant hazards

considerations. This document provides examples of amendments which

are and are not considered likely to involve significant hazards

considerations. These proposed amendments most closely fit the

example of a change which may either result in some increase to the

probability or consequences of a previously analyzed accident or may

reduce in some way a safety margin, but where the results of the

change are clearly within all acceptance criteria with respect to

the system or component specified in the standard review plan.

This proposed amendment does not involve a significant

relaxation of the criteria used to establish safety limits, a

significant

[[Page 45177]]

relaxation of the bases for the limiting safety system settings, or a

significant relaxation of the bases for the limiting conditions for

operations. The proposed change does not reduce the margin of safety

as defined in the basis for any Technical Specification.

Therefore, based on the guidance provided in the Federal

Register and the criteria established in 10 CFR 50.92(c), ComEd has

concluded that the proposed change does not constitute a significant

hazards consideration.

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

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

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

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; for LaSalle, Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,

Units 1 and 2, Rock Island County, Illinois

Date of application for amendment requests: August 30, 1994, as

supplemented August 4, 1995.

Description of amendment requests: As a result of findings by a

Diagnostic Evaluation Team inspection performed by the NRC staff at the

Dresden Nuclear Power Station in 1987, Commonwealth Edison Company

(ComEd, the licensee) made a decision that both the Dresden Nuclear

Power Station and sister site Quad Cities Nuclear Power Station needed

attention focused on the existing custom Technical Specifications (TS)

used.

The licensee made the decision to initiate a Technical

Specification Upgrade Program (TSUP) for both Dresden and Quad Cities.

The licensee evaluated the current TS for both Dresden and Quad Cities

against the Standard Technical Specifications (STS) contained in NUREG-

0123, ``Standard Technical Specifications General Electric Plants BWR/

4.'' The licensee's evaluation identified numerous potential

improvements such as clarifying requirements, changing TS to make them

more understandable and to eliminate interpretation, and deleting

requirements that are no longer considered current with industry

practice. As a result of the evaluation, ComEd has elected to upgrade

both the Dresden and Quad Cities TS to the STS contained in NUREG-0123.

The TSUP for Dresden and Quad Cities is not a complete adaption of

the STS. The TSUP focuses on (1) integrating additional information

such as equipment operability requirements during shutdown conditions,

(2) clarifying requirements such as limiting conditions for operation

and action statements utilizing STS terminology, (3) deleting

superseded requirements and modifications to the TS based on the

licensee's responses to Generic Letters (GL), and (4) relocating

specific items to more appropriate TS locations.

The August 30, 1994, and August 4, 1995, applications proposed to

upgrade only Section 3/4.2 (Instrumentation) of the Dresden and Quad

Cities TS.

Basis for proposed no significant hazards consideration

determination: As requied 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

because:

In general, the proposed amendment represents the conversion of

current requirements to a more generic format, or the addition of

requirements which are based on the current safety analysis.

Implementation of these changes will provide increased reliability

of equipment assumed to operate in the current safety analysis, or

provide continued assurance that specified parameters remain within

their acceptance limits, and as such, will not significantly

increase the probability or consequences of a previously evaluated

accident.

Some of the proposed changes to the current Technical

Specifications (CTS) represent minor curtailments of the current

requirements which are based on generic guidance or previously

approved provisions for other stations. The proposed amendment for

Dresden and Quad Cities Station's Technical Specification Section 3/

4.2 are based on BWR-STS (NUREG-0123, Revision 4 ``Standard

Technical Specifications General Electric Plants BWR/4) guidance or

NRC accepted changes at later operating BWR plants. Any deviations

from BWR-STS and CTS requirements do not significantly increase the

probability or consequences of any previously evaluated accident for

Dresden and Quad Cities Station. These proposed changes are

consistent with the current safety analyses and have been previously

determined to represent sufficient requirements for the assurance

and reliability of equipment assumed to operate in the safety

analysis, or provide continued assurance that specified parameters

remain within their acceptance limits. As such, these changes will

not significantly increase the probability or consequences of a

previously evaluated accident.

The associated systems that make up the Instrumentation Systems

are not assumed in any safety analysis to initiate any accident

sequence for both Dresden and Quad Cities Stations; therefore, the

probability of any accident previously evaluated is not increased by

the proposed amendment. In addition, the proposed surveillance

requirements for the proposed amendments to these systems are

generally more prescriptive than the current requirements specified

within the Technical Specifications. These more prescriptive

surveillance requirements increase the probability that the

Instrumentation Systems will perform their intended functions.

Therefore, the proposed TS will improve the reliability and

availability of all affected systems and reduce the consequences of

any accident previously evaluated.

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

from any previously evaluated because:

In general, the proposed amendment represents the conversion of

current requirements to a more generic format, or the addition of

requirements which are based on the current safety analysis. Others

represent minor curtailments of the current requirements which are

based on generic guidance or previously approved provisions for

other stations. These changes do not involve revisions to the design

of the station, other than technically valid trip setpoint changes.

Some of the changes may involve revision in the operation of the

station; however, these changes provide additional restrictions

which are in accordance with the current safety analyses, or are to

provide for additional testing or surveillances which will not

introduce new failure mechanisms beyond those already considered in

the current safety analyses. Therefore, these changes will not

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

any accident previously evaluated.

The proposed amendment for Dresden and Quad Cities Station's

Technical Specification Section 3/4.2 is based on BWR-STS guidelines

or NRC accepted changes at later operating BWR plants. The proposed

amendment has been reviewed for acceptability at the Dresden and

Quad Cities Nuclear Power Stations considering similarity of system

or component design versus the BWR-STS or later operating BWRs. Any

deviations from BWR-STS or CTS requirements do not create the

possibility of a new or different kind of accident than previously

evaluated for Dresden and Quad Cities Stations. No new modes of

operation are introduced by the proposed changes. Various

surveillance requirements are changed to reflect improvements in

technique, frequency of performance or operating experience at later

plants. Proposed changes to action statements in many places add

requirements that are not in the present technical specifications or

adopt

[[Page 45178]]

requirements that have been used at other operating BWRs with designs

similar to Dresden and Quad Cities. The proposed changes maintain at

least the present level of operability. Therefore, the proposed

changes do not create the possibility of a new or different kind of

accident from any previously evaluated.

The associated systems that make up the Instrumentation Systems

are not assumed in any safety analysis to initiate any accident

sequence for Dresden or Quad Cities Stations. In addition, the

proposed surveillance requirements for affected systems associated

with the Instrumentation Systems are generally more prescriptive

than the current requirements specified within the Technical

Specifications; therefore, the proposed changes do not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. Involve a significant reduction in the margin of safety

because:

In general, the proposed amendment represents the conversion of

current requirements to a more generic format, or the addition of

requirements which are based on the current safety analysis. Others

represent minor curtailments of the current requirements which are

based on generic guidance or previously approved provisions for

other stations. Some of the later individual items may introduce

minor reductions in the margin of safety when compared to the

current requirements. However, other individual changes are the

adoption of new requirements which will provide significant

enhancement of the reliability of the equipment assumed to operate

in the safety analysis, or provide enhanced assurance that specified

parameters remain within their acceptance limits. These enhancements

compensate for the individual minor reductions, such that taken

together, the proposed changes will not significantly reduce the

margin of safety.

The proposed amendment to Technical Specification Section 3/4.2

implements present requirements in accordance with the guidelines

set forth in the BWR-STS. Any deviations from BWR-STS and CTS

requirements do not significantly reduce the margin of safety for

Dresden and Quad Cities Stations. The proposed changes are intended

to improve readability, usability, and the understanding of

technical specification requirements while maintaining acceptable

levels of safe operation. The proposed changes have been evaluated

and found to be acceptable for use at Dresden and Quad Cities based

on system design, safety analysis requirements and operational

performance. Since the proposed changes are based on NRC accepted

provisions at other operating plants that are applicable at Dresden

and Quad Cities and maintain necessary levels of system or component

readability, the proposed changes do not involve a significant

reduction in the margin of safety.

The proposed amendment for Dresden and Quad Cities Stations will

not reduce the availability of systems associated with the

Instrumentation Systems when required to mitigate accident

conditions; therefore, the proposed changes do not involve a

significant reduction in the margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: for Dresden, Morris Public

Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,

Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of amendment request: March 8, 1995, as supplemented June 1,

1995

Description of amendment request: The proposed amendments would

revise the secondary undervoltage setpoint.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. The proposed amendment does not involve an increase in the

probability of occurrence or consequences of any accident previously

evaluated.

The proposed amendment does not change the fundamental function

or capability of the Secondary Undervoltage protection as described

in UFSAR section 8.3. Inadvertent or spurious operation of the

Secondary Undervoltage protection function will initiate loading of

the safe shutdown loads on the diesel generators and is not assumed

to initiate an accident. The proposed Secondary Undervoltage

setpoints are low enough to prevent spurious actuations given the

expected off site grid voltages.

This change does not affect the initiators or precursors of any

accident previously evaluated. This change will not increase the

likelihood that a transient initiating event will occur because

transients are initiated by equipment malfunction and/or

catastrophic system failure. The change in setpoints for the

Secondary Undervoltage protection system does involve some changes

to existing plant equipment (such as transformer tap changes and

Circulating Water pump excitation circuit changes). However, all

changes to existing plant equipment have been or will be evaluated

in accordance with the requirements of 10CFR50.59 prior to

installation, to determine that no unreviewed safety questions exist

with regard to the plant changes.

Since any design changes have been or will be determined to be

acceptable per 10CFR50.59 prior to installation and no new plant

equipment will be installed, the probability of occurrence of

accidents previously evaluated will not increase.

With Zion Station's new Auxiliary Power System configuration and

the proposed Secondary Undervoltage setpoints, the probability of a

Loss of Off-Site Power (LOOP) is actually reduced since the original

Auxiliary Power System configuration and Secondary Undervoltage

setpoints required a higher grid voltage to ensure that safety

related loads would be powered from Off-Site power sources during a

design basis accident.

The consequences of accidents previously evaluated are not

increased. The proposed change does not affect the required level of

availability or systems required to mitigate the accidents

considered in the Analyses. Administrative controls will be in place

to ensure that the installed setpoints are low enough to ensure that

the Emergency Diesel Generators are not unnecessarily challenged.

The proposed changes will increase the level of confidence that the

ESF equipment will be capable of starting and operating during a

design basis accident with degraded off-site grid voltage. The

increase in the level of confidence is the result of the more

rigorous methodology used to determine limited ESF bus voltages,

given the minimum expected off-site AC voltage. Based on the

previous discussion, it is determined that there will be no

significant increase in the consequences of any accident previously

evaluated.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any previously analyzed.

The proposed Secondary Undervoltage setpoint change does not

change the design of the Secondary Undervoltage protection system or

its function to protect against degraded offsite power. Actuation of

the Secondary Undervoltage protection system will initiate a

sequence of events that will start the Emergency Diesel Generator

(EDG) for the associated ESF bus, strip all loads from the bus, open

all feed breakers to the bus, close the Emergency feed breaker (thus

energizing the bus from the EDG), and initiate sequenced starting of

the Safe Shutdown equipment supplied by the bus, including a Service

Water pump, Component Cooling Water pump, Auxiliary Feedwater pump,

and Reactor Containment Fan Cooler(s), as applicable.

The proposed change does not involve the addition of any new or

different types of equipment, nor does it involve the operation of

equipment required for safe operation of the facility in a manner

different from those addressed in the Final Safety Analysis Report.

No safety related equipment or function will be altered as a result

of this proposed change. Because no new failure modes are

introduced, the proposed amendment does not create a new or

different kind of accident from any previously analyzed in the

UFSAR.

Based on the above discussion, the proposed amendment does not

create a new or different kind of accident from any previously

analyzed in the UFSAR.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

[[Page 45179]]

The proposed amendment will allow the Secondary Undervoltage

setpoint to be conservatively established based on new engineering

calculations which consider the lowest expected offsite grid voltage

and operation of all required ESF equipment under design basis

accident loading conditions.

The proposed Secondary Undervoltage setpoints will provide

increased confidence that adequate bus voltage will be available to

support starting and operation of all required ESF loads. The

proposed setpoint includes worst case instrument error to ensure

that the lowest possible voltage will not be lower than the degraded

voltage analytical limits. Additionally, the proposed setpoints are

low enough to prevent spurious actuations due to expected

fluctuations in the grid voltage. The new setpoints are based on a

minimum expected grid voltage of 343 kV, with added margin. The

proposed changes will provide an increase in the level of protection

that currently exists and will ensure the margin of safety is

adequately maintained.

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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085

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

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County, Connecticut

Date of amendment request: August 3, 1995

Description of amendment request: The proposed amendment will add

an one-time footnote to Technical Specification (TS) Section 3/4.7.12,

``Ultimate Heat Sink,'' to increase the allowed outage time from 6

hours to 18 hours for the months of August and September. In addition,

also for the months of August and September, the maximum service water

limit will be elevated from 90 deg.F to 95 deg.F.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed addition of a 12 hour time period to monitor the

ultimate heat sink temperature to the Technical Specification

Limiting Condition for Operation action statements does not involve

an increase in the probability of an accident previously evaluated.

The probability of an accident previously evaluated is not increased

by a short-term increase in the ultimate heat sink temperature. An

evaluation has been performed that safe shutdown will be achieved

and maintained for a loss of normal AC power event with the

additional consideration of a single failure with service water

inlet temperatures as high as 95 deg.F. In addition, an evaluation

of the credible FSAR Chapter 15 events with AC power available and

no isolation of non-essential service water loads has been performed

that demonstrates that safe shutdown will be achieved and

maintained. There has been no significant increase in the

consequences of these events previously evaluated.

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

from any previously analyzed.

The proposed technical specification change does not create the

possibility of a new or different kind of accident previously

analyzed. The addition of a 12 hour time period to monitor the

ultimate heat sink temperature increases the amount of time that is

allowed for the plant to be in Hot Standby from 6 to 18 hours should

the ultimate heat sink temperature increase above 90 deg.F. This

extension of the time allowed for the plant to be in Hot Standby

does not change the plant configuration. As such, the change does

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

previously evaluated.

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

The proposed technical specification change does not involve a

significant reduction in the margin of safety. The addition of a 12

hour time period to monitor the ultimate heat sink temperature

increases the time required for the plant to be in Hot Standby from

6 to 18 hours should the ultimate heat sink temperature exceed

90 deg.F. An evaluation has been performed to demonstrate that the

risk significance associated with the increased action time is very

low. In addition, safe shutdown capability has been demonstrated for

service water inlet temperatures as high as 95 deg.F.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457

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

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

CT 06141-0270

NRC Project Director: Phillip F. McKee

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of amendment request: May 5, 1995

Description of amendment request: The proposed amendment would

change the surveillance frequency of radiation area, and effluent and

process monitors from monthly to quarterly; and the required frequency

for minimum exercise of control element assemblies also from monthly to

quarterly.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration. The

NRC staff has reviewed the licensee's analysis against the standards of

10 CFR 50.92(c). The staff's review is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Extending surveillance test intervals as proposed will

reduce the probability of inadvertent reactor scrams and ensuing

challenges to safety systems. This is accomplished by reducing the

occasions and thus the total time that the subject systems are

removed from their ``normal'' configuration and placed into the

required ``test'' configuration. In addition, the probability of

test-induced failures, or failures caused by human error, is

likewise decreased. Thus, 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.

Extending surveillance test intervals as proposed will not

require installation of any new or different equipment, and will not

alter or otherwise modify existing plant equipment. Thus, 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.

Independent research has found that equipment failures and

personnel errors during several types of surveillance tests caused a

significant number of reactor scrams and attendant unnecessary

challenges to safety equipment. The results of this research have

been corroborated by the licensee's plant specific operating

experience. The licensee concludes that the reduced test intervals

proposed in this amendment remain sufficient to ensure known

phenomena, such as instrument setpoint drift and random hidden

failures, remain within the assumptions of the safety analysis.

Thus, 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)

[[Page 45180]]

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME 04578

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road, Brunswick, ME 04011

NRC Project Director: Phillip F. McKee

North Atlantic Energy Service Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: July 24, 1995.

Description of amendment request: The proposed amendment would

delete Table 3.4-1, ``Reactor Coolant System Pressure Isolation

Valves'' from the Seabrook Station, Unit No. 1 Technical Specification

section 3.4.6.2. Reference to Table 3.4-1 also would be deleted from

Limiting Condition for Operation 3.4.6.2 f and from Surveillance

Requirement 4.4.6.2.2. The information contained in Table 3.4-1 would

be relocated to the Technical Requirements Manual.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration. The

NRC staff has reviewed the licensee's analysis against the standards of

10 CFR 50.92(c). The NRC staff's review is presented below.

A. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

(10 CFR 50.92(c)(1)) because they do not in any way alter the

operability or surveillance requirements for pressure isolation

valves. The proposed changes merely delete a listing of valves which

are designated as pressure isolation valves in accordance with the

definition provided in 10 CFR Part 50. Therefore, neither the

probability nor consequences of previously evaluated accidents are

affected.

B. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously evaluated

(10 CFR 50.92(c)(2)) because they do not affect in any way the

manner by which the facility is operated or make any changes in

structures, systems, or components which could affect the

operational characteristics of the facility.

C. The proposed changes do not involve a significant reduction

in a margin of safety (10 CFR 50.92(c)(3)) because the proposed

changes do not affect the operability requirements or surveillance

testing of any pressure isolation valve and do not affect in any way

the manner by which the facility is operated or involve equipment or

features which affect the operational characteristics of the

facility.

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: Exeter Public Library,

Founders Park, Exeter, NH 03833

Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast

Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270

NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,

Millstone Nuclear Power Station, Unit 1, New London County,

Connecticut

Date of amendment request: July 28, 1995

Description of amendment request: The proposed amendment adds

Technical Specifications (TS) to Section 3.10, Refueling and Spent Fuel

Handling. Specifically, the proposed TS (with applicability, action,

and surveillance requirements) will require that: (1) the reactor be

subcritical for at least 100 hours before the start of reactor

refueling operations, (2) the spent fuel pool bulk temperature be

maintained less than or equal to 140 deg.F, and (3) two trains of

shutdown cooling be operable during reactor refueling operations. In

support of the request, NNECO proposes to: (1) use the ORIGEN2 code to

more accurately predict decay heat loads from the spent fuel, (2) use

the ONEPOOL code to credit the effect of evaporative cooling on the

spent fuel pool bulk temperature, and (3) take credit for both trains

of shutdown cooling to assist the spent fuel pool cooling system during

refueling outages. In addition, the proposed amendment modifies the

table of contents and associated Bases section to reflect the changes.

Basis for proposed no significant hazards consideration

determination: As requied 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 changes in accordance with

10CFR50.92 and concluded that the changes do not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

compromised. The proposed changes do not involve an SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed license amendment will allow NNECO to use the

shutdown cooling system (SCS) to assist the spent fuel pool cooling

(SFPC) system to cool the spent fuel pool during refueling outages.

This amendment request does not affect: the number of spent fuel

bundles allowed in the spent fuel pool, spent fuel pool criticality

analysis, structural analysis of the spent fuel pool, or

radiological release scenarios.

The proposed license amendment also allows NNECO to use ORIGEN2

and ONEPOOL codes. The ORIGEN2 code more accurately predicts decay

heat loads from the spent fuel in the spent fuel pool. The ONEPOOL

code credits the effect of evaporative cooling on the spent fuel

pool bulk temperature. The use of these codes will improve the

accuracy of predicting spent fuel pool bulk temperatures during

normal and abnormal refueling scenarios.

The use of the SCS to assist the SFPC system to cool the spent

fuel pool will allow the movement of spent fuel to begin 100 hours

after reactor shutdown. The existing accident analysis for a dropped

spent fuel bundle during refueling bounds this situation as the

analysis assumed a decay time of 24 hours.

The three new proposed technical specifications will provide

sufficient controls on the movement of spent fuel into the spent

fuel pool, bulk temperature of the spent fuel pool and operability

of the shutdown cooling system to operate within analysis

assumptions during refueling operations at Millstone Unit No. 1.

Therefore, based on the above, the use of the SCS to assist the

SFPC system to cool the spent fuel pool during refueling outages,

the use of the ORIGEN2 code, the use of the ONEPOOL code, and the

addition of three technical specifications will not involve a

significant increase in the probability or consequences of an

accident previously analyzed.

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

from any previously analyzed.

The proposed license amendment to use the SCS to assist the SFPC

system to cool the spent fuel pool will allow SCS train B to cool

the spent fuel pool in a method similar to train A.

The proposed license amendment to use ORIGEN2 and ONEPOOL codes

to predict spent fuel pool bulk temperatures will increase the

accuracy of analyzing normal and abnormal refueling scenarios.

The three new proposed technical specifications will

sufficiently control refueling operations to support analyzed

accident scenarios.

Therefore, the use of the SCS to assist the SFPC system to cool

the spent fuel pool, the use of the ORIGEN2 code, the use of ONEPOOL

code and the addition of three technical specifications do not

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

any previously analyzed.

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

The proposed license amendment to use the SCS to assist the SFPC

system to cool the spent fuel pool will allow the crediting of the

SCS and SFPC system to remove heat from

[[Page 45181]]

the spent fuel pool during normal refueling scenarios. The analysis

demonstrates that this cooling configuration will maintain the spent

fuel pool bulk temperature below the pool design limit of 140 deg.F

with a postulated single active failure.

The addition of the train B SCS cross-tie does not adversely

affect the existing design basis of the SCS to remove sensible and

decay heat from the reactor water, cool it from 280 deg.F to

125 deg.F within 24 hours, and to maintain the reactor water at

125 deg.F.

The proposed license amendment to use ORIGEN2 and ONEPOOL codes

will improve the accuracy of predicting spent fuel pool bulk

temperatures during normal and abnormal refueling scenarios.

The thermal hydraulic analysis most limiting time to boil

calculation of 5.4 hours for loss of all forced cooling to the spent

fuel pool is consistent with assumed operator response times for

similar scenarios.

The three new proposed technical specifications will ensure that

the margin of safety established by engineering analysis of

refueling operations is maintained.

Therefore, based on the above, the use of the SCS to assist the

SFPC system to cool the spent fuel pool, the use of the ORIGEN2

code, the use of the ONEPOOL code, and the addition of three

technical specifications does not involve a significant reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360

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

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

CT 06141-0270

NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-

336 and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2,

and 3, New London, Connecticut

Date of amendment request: August 4, 1995

Description of amendment request: The proposed license amendments

will modify the Administrative Controls Section (Section 6) of the

Millstone Unit Nos. 1, 2, and 3 Technical Specifications to allow the

Plant Operations Review Committee (PORC) and Site Operations Review

Committee (SORC) to direct its efforts in the review of more critical

safety matters which affect day-to-day operation. This will be

accomplished by the establishment of a Station Qualified Reviewer

Program (SQRP) and the reassignment of certain procedure approvals to

designated managers in lieu of approval by PORC/SORC.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration

(SHC), which is presented below:

...These proposed changes do not involve an SHC because the

changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

These changes are administrative in nature. They do not involve

any modifications to plant systems and do not alter the method of

operation of any plant equipment. The change involves the

establishment of a SQRP for the review of plant procedures, programs

or changes thereto that do not involve a 10CFR50.59 evaluation.

Implementing a SQRP will not result in a degradation of the

current level of procedure review. PORC/SORC will retain the

responsibility for reviewing any document for which a 10CFR50.59

evaluation is required. Personnel selected to be SQRs [Station

Qualified Reviewers] will possess the technical experience and

expertise to provide a thorough technical review as required by

plant procedures. These personnel, and the managers authorized to

approve these procedures, will be designated in writing by the Unit

Director or the Senior Vice President - Millstone Station.

Procedures or classes of procedures that can be reviewed per the

SQRP will be specified in writing by the Unit Director or the Senior

Vice President - Millstone Station. Procedures will receive an

appropriate cross-disciplinary review when necessary.

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

from any accident previously evaluated.

The proposed technical specification changes do not change the

design or function of any plant structure, system, or component, nor

do they introduce any new failure modes. As stated above, the

implementation of a SQRP will not degrade the quality of plant

procedures.

There are no modifications to plant structures, systems, or

components associated with these proposed changes, and the operation

of plant equipment and systems remain unchanged. Since the changes

proposed in this license amendment request do not revise existing

plant structures, systems, or components, do not change the manner

in which the plant is operated and, do not change the manner in

which the plant will respond to any design basis accidents, the

proposed changes do not create the possibility of a new or different

kind of accident from any previously analyzed.

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

The changes proposed in this proposed license amendment request

do not affect the ability of any system to perform its safety-

related function. As described above, these proposed changes are

administrative in nature. They do not change any plant operating

parameters or design features and do not reduce the level of

effectiveness of any existing administrative controls. The proposed

change will not result in changes to the bases for any technical

specification. The establishment of the SQRP will continue to

provide for the adequate review of procedures. In addition, another

direct benefit of this program is that the amount of material

presented to PORC/SORC will decrease. The reduction in the amount of

material presented to PORC/SORC for review will allow the PORC/SORC

to focus on safety significant issues. Since none of the assumptions

in the technical specifications bases are affected by the changes

presented in this license amendment request, the margin of safety

which exists in the current technical specifications is not reduced.

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, CT 06360

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

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

CT 06141-0270

NRC Project Director: Phillip F. McKee

Northern States Power Company, Docket No. 50-263, Monticello

Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: June 22, 1995

Description of amendment request: The proposed changes modify the

facility requirements for thermal-hydraulic instability avoidance and

protection to address concerns over reactor fuel performance during

instability events. Changes are proposed to the Technical

Specifications to utilize the flow biased Average Power Range Monitor

high neutron flux scram and a power-flow map exclusion region

consistent with one of the NRC approved BWR Owners' Group solutions. In

addition, a change to correct an error in the Average Planar Linear

Heat Generation Rate during single loop operation is also proposed.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards

[[Page 45182]]

consideration, which is presented below:

a. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The implementation of BWR Owner's Group long term stability

solution Option 1-D at Monticello does not modify the assumptions

contained in the existing accident analysis. The use of an exclusion

region and the operator actions required to avoid and minimize

operation inside the region do not increase the possibility of an

accident. Conditions of operation outside of the exclusion region

are within the analytical envelope of the existing safety analysis.

The operator action requirement to exit the exclusion region upon

entry minimizes the probability of an oscillation occurring. The

actions to drive control rods and/or to increase recirculation flow

to exit the region are maneuvers within the envelope of normal plant

evolutions. The flow based scram has been analyzed and will provide

automatic fuel protection in the event of a core wide instability.

Thus, each proposed operating requirement provides defense in depth

for protection from an instability event while maintaining the

existing assumptions of the accident analysis. The proposed change

to the method by which the MAPLHGR [maximum average planar linear

heat-generation rate] is obtained for single loop operation is

consistent with the analysis performed for the Average Power Range

Monitor/Rod Block Monitor Technical Specifications (ARTS) program.

The analysis performed in support of the ARTS program demonstrated

that the limits established assure compliance with fuel limits.

Therefore, this amendment will not cause a significant increase in

the probability or consequences of an accident previously evaluated

for the Monticello plant.

b. The proposed amendment will not create the possibility of a

new or different kind of accident from any accident previously

analyzed.

As stated above, the proposed operating requirements either

mandate operation within the envelope of existing plant operating

conditions or force specific operating maneuvers within those

carried out in normal operation. Since operation of the plant with

all of the proposed requirements is within the existing operating

basis, an unanalyzed accident will not be created through

implementation of the proposed change. Therefore, the proposed

amendment will not create the possibility of a new or different kind

of accident.

c. The proposed amendment will not involve a significant

reduction in the margin of safety.

Each of the proposed requirements for the plant thermal-

hydraulic stability provides a means for fuel protection. The

combination of avoiding possible unstable conditions and the

automatic flow biased reactor scram provides an in-depth means for

fuel protection. Therefore, the individual or combination of means

to avoid and suppress an instability supplements the margin of

safety. The operating limits established for the single loop

operation MAPLHGR provide an acceptable margin of safety as

demonstrated in NEDC-30492, ``Average Power Range Monitor, Rod Block

Monitor and Technical Specification Improvement (ARTS) Program for

Monticello Nuclear Generating Plant-April 1984.'' The proposed

amendment will not involve a reduction in the margin of safety.

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

this review, it appears that the three 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: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

and Trowbridge, 2300 N Street, NW, Washington, DC 20037

NRC Project Director: John N. Hannon

Northern States Power Company, Docket No. 50-263, Monticello

Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: July 5, 1995

Description of amendment request: The proposed amendment, part of

the Monticello Surveillance Test Interval/Allowed Outage Time (STI/AOT)

Program, extends the surveillance test intervals and allowable out-of-

service times for selected instrumentation. The proposed changes are

intended to minimize unnecessary testing and remove excessively

restrictive out-of-service times that could potentially degrade overall

plant safety and availability.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

a. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The maximum failure frequency change is for the ECCS Actuation

Instrumentation as identified by General Electric topical report

NEDC-30936P-A, and Monticello specific report RE-006. These reports

concluded core damage frequency changed by less than 4% when STIs

were increased to once per 3 months, AOTs for surveillance were

increased to 6 hours, and AOTs for repair were increased to 24

hours. Since this small increase was within the guideline of

acceptability stated in NEDC-30936P-A, and Monticello only proposes

to increase the repair AOT to 12 hours rather than 24 hours, this

amendment will not cause a significant increase in the probability

or consequences of an accident previously evaluated for the

Monticello plant (see RE-006).

The drift analysis determined the associated instrumentation

would not be adversely effected with the longer calibration

intervals. Pertinent process parameters including instrument drift

will still be within acceptance criteria with the longer

surveillance intervals.

The recirculation flow meters and flow instrumentation are not

used in any safety or accident analysis. Therefore, no analysis

would be changed by increasing the calibration interval to once per

cycle.

b. The proposed amendment will not create the possibility of a

new or different kind of accident from any accident previously

analyzed.

These changes only affect the instrument STI and AOT times. No

changes are being made to the functions of the instrumentation.

Therefore, the proposed amendment will not create the possibility of

a new or different kind of accident.

c. The proposed amendment will not involve a significant

reduction in the margin of safety.

These changes will improve the performance of equipment and are

intended to reduce the potential for equipment failures due to

unnecessary testing. The safety limits and the limiting safety

system setpoints will not be affected by these changes. No safety

margins are affected, therefore, the drift will remain within the

margins 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: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

and Trowbridge, 2300 N Street, NW, Washington, DC 20037

NRC Project Director: John N. Hannon

Omaha Public Power District, Docket No. 50-285, Fort Calhoun

Station,Unit No. 1, Washington County, Nebraska

Date of amendment request: August 4, 1995

Description of amendment request: This proposed amendment would

revise the Technical Specifications (TS) for the requirements for the

containment radiation high signal (CRHS) and the safety injection and

refueling water (SIRW) tank low signal (STLS) contained in TS 2.15,

Tables 2-3 and 2-4. Specification 3.1, Table 3-2 will also be revised

to include administrative changes to the CRHS surveillance

[[Page 45183]]

methods to be consistent with the applicable surveillance functions.

The Basis for Specification 2.15 is being revised to clarify that the

number of installed channels for CRHS is two. The term ``SOURCE CHECK''

is being deleted from the Definitions section.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The Omaha Public Power District (OPPD) proposes to revise

Technical Specification (TS) 2.15, Table 2-3 by revising the

requirement for placing the Safety Injection Refueling Water (SIRW)

tank low level channel(s) in the tripped condition to placing them

in the bypassed condition. Due to the derived signal, if a channel

was in the tripped condition and a single failure occurred, (that

being one channel of STLS on either A or B circuits), a premature

SIRW tank low signal (STLS) would be generated. During a design

basis accident (DBA) with a valid Containment Pressure High Signal

(CPHS) or Pressurizer Pressure Low Signal (PPLS), this single

failure would prevent the contents of the SIRW tank from being

injected into the reactor coolant system. The resulting logic of

placing the SIRW tank low level channels in BYPASS rather than TRIP

would not cause a premature switchover of the high pressure safety

injection pumps to the containment sump and it would not prevent the

switchover when needed.

OPPD also proposes to revise TS 2.15, Table 2-4, by reducing the

number of minimum operable Containment Radiation High Signal (CRHS)

channels from two to one. This proposed change revises the

requirements of TS 2.15 to coincide with changes to the TS and

Offsite Dose Calculation Manual (ODCM) that were implemented by TS

Amendment 152. The Engineered Safety Feature (ESF) actuation system

supervisory A and B safeguard initiation channels will not be

affected by this proposed TS change. The minimum level of engineered

safeguards performance acceptable for the DBA, (i.e., minimum

safeguards) will continue to be maintained in accordance with IEEE

279 - 1971, ``Criteria for Protection Systems for Nuclear Power

Generating Stations.''

Included in this change are administrative revisions to TS 3.1,

Table 3-2, for replacing the current surveillance methods for

checking and testing the CRHS instrumentation with the defined terms

``CHANNEL CHECK'' and ``CHANNEL FUNCTIONAL TEST,'' respectively.

These proposed revisions are administrative in nature and reflect

TS-defined terminology for the instrumentation surveillance methods

utilized to ensure that the CRHS instrumentation is operable. A

channel check requires a qualitative determination of acceptable

operability by observation of channel behavior during normal plant

operation. A channel functional test requires the injection of a

simulated signal into the channel to verify that it is operable,

including any alarm and/or trip initiating actions. Other proposed

administrative changes include deleting the term ``SOURCE CHECK''

from the TS Definitions section as source check will no longer be

used in the FCS TS and adding verbiage to the TS 2.15 Basis for

clarifying that the number of installed channels for CRHS is two.

Therefore, the proposed change, as described above, would not

increase the probability or consequences of an accident previously

evaluated.

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

from any accident previously evaluated.

There will be no physical alterations to the plant

configuration, changes to setpoint values, or changes to the

implementation of setpoints or limits as a result of the proposed

changes to TS 2.15, Tables 2-3 and 2-4. The proposed revisions to TS

3.1, Table 3-2 are administrative changes to make the TS more

accurately reflect defined terminology and the methods utilized to

ensure that the CRHS instrumentation is operable. The proposed TS

revisions do not require any changes to the present methods of

verifying CRHS instrumentation operability. Therefore, the proposed

change does not create the possibility of a new or different kind of

accident from any previously evaluated.

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

There are no changes to the equipment or plant operations as a

result of the changes being made to the number of minimum operable

CRHS channels. The proposed changes to the STLS will require that

the inoperable channel be placed in BYPASS rather than TRIP. This

action would ensure that a single failure would not cause a

premature safety injection switchover to the containment sump and

would not prevent switchover when needed. Therefore, this proposed

change does not reduce a margin of safety.

The proposed revisions to TS 3.1, Table 3-2 are administrative

changes to make the TS more accurately reflect defined terminology

and the methods utilized to ensure that the CRHS instrumentation is

operable. The proposed TS revisions do not require any changes to

the present methods of verifying CRHS instrumentation operability.

The proposed changes to the Definitions and TS 2.15 Basis sections

are administrative in nature. Therefore, these 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

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: June 22, 1995

Description of amendment request: The amendments would revise the

Technical Specifications 3.4.1.4 and 3.9.8.2 by deleting footnotes and

associated information regarding Service Water header operation and its

support function for Residual Heat Removal operation. These footnotes

and associated information had been placed in the Technical

Specifications because of the concern about Service Water system piping

integrity in the mid-1980's.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. Do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Even though one service water loop will be out for maintenance,

both loops of residual heat removal (RHR) will be kept operable,

consistent with the requirements of STS (NUREG 1431). A minimum of

two RHR, two component cooling (CC), and two service water (SW)

pumps, powered from two different vital busses, will be kept

operable.

Only one component cooling heat exchanger will be operable since

only one service water loop is operable. The CC heat exchangers for

both Units 1 and 2 have a very high reliability. The primary heat

transfer surfaces of the heat exchangers are made of titanium; no

material problems have been experienced in ten years of service.

The remaining active components that, through misoperation,

could potentially defeat RHR capability are, (1) the motor operated

valves in RHR or SW that could develop a ``hot short'' and

subsequently close and (2) the air operated temperature/ flow

control valves of the CC heat exchangers. Additional actions will be

taken to effectively eliminate the possibility of these single point

valves from failing and defeating RHR capability. The motor operator

breakers will be tagged open during MODES 5 and 6, except for

flooding the cavity, when the RHR suction valves must be closed. The

CC Heat Exchanger air operated temperature/flow control valves fail

open, or as is, on loss of air which is the safe position. Operators

will monitor critical temperatures; this equipment is accessible if

any corrective action is required. Thus, with one service water

header out of service, the intent of the

[[Page 45184]]

technical specifications as defined in the bases section (to have a

single failure proof RHR system) is met with the proposed system

configuration. Therefore, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. Do not create the possibility of a new or different kind of

accident from any previously evaluated.

The catastrophic failure of a moderate energy Class 3 piping

system is not a credible event, based on the upgraded reliability of

the system, the redundancy of active components, the elimination of

single failure points, and on the industry and regulatory positions

established for this type of system. Since SW is a Class 3 moderate

energy system, the only postulated passive failure mode is a leakage

crack. In accordance with Generic Letter (GL) 91-18 and GL 90-05, a

leak in the SW system, following acceptable evaluation, does not

constitute a failure that causes the loss of capability to perform

it's intended safety function. A moderate energy Class 3 piping leak

does not cause the system to be declared inoperable. Therefore, the

proposed changes do not create the possibility of a new or different

type of accident from any previously evaluated.

3. Do not involve a significant reduction in a margin of safety.

RHR redundancy is maintained; no credible single failure point

exists that could cause a nonrecoverable loss of SW. 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, New Jersey 08079

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: September 15, 1992, as supplemented

April 20, 1993, April 26, 1995, and July 27, 1995.

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) 3.1.1.4, 3.1.1.6, and 4.3.4, and

add a Basis to address Generic Letter (GL) 90-06. GL 90-06 represents

the technical resolution of Generic Issue (GI) 70, ``Power Operated

Relief Valve and Block Valve Reliability,'' and GI 94, ``Additional Low

Temperature Overpressure Protection for Light Water Reactors.'' The

resolution of these issues proposes new requirements and TS changes

that enhance the reliability of power-operated relief valves (PORVs)

and block valves along with TS changes that will provide additional

low-temperature overpressure protection (LTOP).

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

There is no significant increase in the probability or

consequences of an accident previously evaluated because the

accident conditions and assumptions are not significantly affected

by the proposed change.

The proposed change to action statement 3.1.1.4a(i) [proposed to

be renumbered to 3.1.1.6c] to include the removal of power from a

closed block valve will provide additional assurance to preclude any

inadvertent opening of the block valve at a time in which the PORV

may not be operable to assure RCS [reactor coolant system]

integrity.

The provision of the generic letter requires, with one or both

PORV(s) inoperable to initiate shutdown actions if PORV operability

is not restored within 72 hours or 1 hour respectively. RG&E

[Rochester Gas and Electric Corporation] does not address these

shutdown actions, but rather will concentrate on re-establishing

valve operability. If the block valve(s) and power are not removed

within 1 hour shutdown provisions must be initiated. [***].

Proposed action statement 3.1.1.4a(ii) [proposed to be

renumbered to 3.1.1.6d] includes a provision to place the block

valves associated PORV(s) switch in manual control due to an

inoperable block valve(s). This requirement precludes the automatic

opening for an overpressure event to avoid the potential for a

stuck-open PORV at a time that the block valve is open and

inoperable. [***].

The proposed change of maintaining power to closed block valves

could potentially increase the probability of an inadvertent opening

of a block valve. The safety impact is, however, not significant

since the proposed changes are only applicable if the PORV is

inoperable due to excessive seat leakage (proposed action 3.1.1.6b).

[***].

Proposed action statement 3.1.1.6b establishes reactor coolant

pressure boundary integrity for a PORV that has excessive seat

leakage and is therefore considered operable to perform its intended

safety function. [***].

Proposed Surveillance Requirement 4.3.4.3 addresses operability

of the Nitrogen System by demonstration of the PORVs at least once

per 18 months by operating the PORVs through a complete cycle of

full travel. [***].

Based on the above efforts, the proposed amendment does not

involve a significant increase in the probability or consequences of

any accident previously evaluated.

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

previously evaluated is not created. In matters related to nuclear

safety, all accidents continue to bound previous analyses. The

proposed changes do not add or modify any equipment design nor do

the proposed changes involve any significant operational changes to

any plant systems.

The proposed amendment does not involve a significant reduction

in the margin of safety as defined in the basis for any technical

specification because the results of the accident analyses which are

documented in the UFSAR [Updated Final Safety Analysis Report]

continue to bound operation under the proposed changes so that there

is no safety margin reduction. [***].

Therefore, the proposed changes do not involve a significant

reduction in 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: Rochester Public Library, 115

South Avenue, Rochester, New York 14610

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005

NRC Project Director: Ledyard B. Marsh

Sacramento Municipal Utility District (SMUD), Docket No. 50-312,

Rancho Seco Nuclear Station, Sacramento County, California

Date of amendment request: June 20, 1995 and as amended August 14,

1995

Description of amendment request: The proposed amendment (PA-191)

would permit SMUD to change the Fuel Storage Building load handling

limits to allow placing the shield plugs on the dry shielded cannisters

in order to permit transfer of spent fuel assemblies from the spent

fuel pool (SFP) to the Rancho Seco Independent Spent Fuel Storage

Installation.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

PA-191 will not create a significant increase in the probability

or consequences of an accident previously evaluated in the Safety

Analysis Report (SAR), because dropping the dry shielded canister

(DSC) top shield plug over a DSC loaded with 24 spent fuel

assemblies is not considered a credible event. Also, the gantry

crane is designed such

[[Page 45185]]

that it can only handle loads over the SFP cask pit area and can not

move a load over the SFP fuel storage racks.

PA-191 will not create the possibility of a new or different

type of accident than previously evaluated in the SAR, because the

proposed Permanently Defueled Technical Specification heavy load

handling exceptions do not create a new credible accident scenario.

Dropping the DSC top shield plug and damaging spent fuel assemblies

is not considered a credible event.

PA-191 will not involve a significant reduction in the margin of

safety, because the proposed heavy load handling exceptions do not

create a credible accident scenario.

The NRC staff has reviewed the licensee's analyses of June 20, 1995

and August 14, 1995. The August 14 submittal enhanced these analyses by

providing design details regarding the significant safety factors built

into the crane and other lifting hardware. 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: Central Library, Government

Documents 828 I Street, Sacramento, CA 95814

Attorney for licensee: Dana Appling, Esq. Sacramento Municipal

Utility District, P. O. Box 15830, Sacramento, CA 95852-1830

NRC Project Director: Seymour H. Weiss

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 amendment requests: July 17, 1995

Description of amendment requests: The licensee proposes to revise

surveillance requirements associated with Technical Specifications 3/

4.3.1, ``Reactor Protective Instrumentation,'' and 3/4.3.2,

``Engineered Safety Feature Actuation System Instrumentation.'' The

surveillance interval is to be increased to 120 days for performance of

channel functional tests for certain reactor protective system and

engineered safety feature actuation system instrumentation. The

proposed change also revises Bases 3/4.3.1, ``Reactor Protective and

Engineered Safety Features Actuation System Instrumentation,'' to

reflect the new interval.

Basis for proposed no significant hazards consideration

determination: As requied 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 would extend the current sequential Channel

Functional Test (CFT) surveillance interval for Plant Protective

System (PPS) instrumentation and Nuclear Instrumentation (NI). This

change does not involve any changes to plant equipment or operation.

The proposed change actually maintains or decreases the PPS system

unavailability. PPS uncertainty and setpoint modifications will

account for the new surveillance interval. Therefore, the proposed

change will not involve a significant increase in the probability or

consequences 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.

This amendment request does not involve any change to plant

equipment or operation. The PPS system is used for monitoring and

mitigation of evaluated accidents. Increasing the availability of

the PPS system, as proposed in this amendment request, 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 amendment does not change the manner in which safety

limits, limiting safety settings, or limiting conditions for

operation are determined. This amendment request will increase

Reactor Protective System and Engineered Safety Features Actuation

System availability. Therefore, this amendment 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 50.92(c) are

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, P. O. Box 19557, Irvine, California 92713

Attorney for licensee: T. E. Oubre, Esquire, Southern California

Edison Company, P. O. Box 800, Rosemead, California 91770

NRC Project Director: William H. Bateman

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: August 7, 1995 (TS 95-12)

Description of amendment request: The proposed change would correct

various errors of an editorial nature that have been identified in the

technical specifications and remove the provisions that have exceeded

their allowed time interval for implementation or the required

conditions no longer exist.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed revisions do not change the TS requirements, plant

setpoints or functions, or plant operating practices. These changes

provide clarifications to the existing TSs by correcting editorial

errors and removing provisions that no longer apply in the

specifications. The probability or consequences of an accident will

not be increased by providing the proposed verbiage corrections that

are editorial and nonintent.

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

from any previously analyzed.

No plant functions or compliance activities associated with the

TS requirements have been affected by the proposed editorial

changes. Therefore, the possibility of a new or different kind of

accident is not created.

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

The proposed changes will not alter TS setpoint values or

functions. The proposed corrections will enhance the application of

TS requirements and will support the margin of safety provided by

the TSs. Therefore, the margin of safety will not be reduced by the

proposed revisions.

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

thisreview, 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: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

[[Page 45186]]

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: August 7, 1995 (TS 95-17)

Description of amendment request: The proposed change would

relocate the heat flux hot channel factor penalty of two percent from

Surveillance Requirement 4.2.2.2.e.1 to the Core Operating Limits

Report and add a reference to the factor to Specification 6.9.1.14.5.

Also, Specification 6.9.1.14.a.2 would be revised to reference Revision

1A of Westinghouse Commercial Atomic Power (WCAP) 10216-P-A,

``Relaxation of Constant Axial Offset Control - FQ Surveillance

Technical Specifications,'' dated February 1994.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change involves only the manner in which the

penalty factors for FQ(Z) would be specified (i.e, a burnup-

dependent factor specified in the Core Operating Limits Report

[COLR] versus a constant factor specified in the TS). This is simply

used to account for the fact that FQ(Z) may increase between

surveillance intervals. These penalty factors are not assumed in any

of the initiating events for the accident analyses. Therefore, the

proposed change will have no effect on the probability of any

accidents previously evaluated. The penalty factors specified in the

COLR will be calculated using NRC-approved methodology and will

therefore continue to provide an equivalent level of protection as

the existing TS requirement. Therefore, the proposed change will not

affect the consequences of any accident previously evaluated.

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

from any previously analyzed.

The proposed change does not involve a physical alteration to

the plant (no new or different kind of equipment will be installed)

or alter the manner in which the plant would be operated. Thus, this

change 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 proposed change will continue to ensure that potential

increases in FQ(Z) over a surveillance interval will be

properly accounted for. The penalty factors will be calculated using

NRC-approved methodology. Therefore, the proposed change will not

involve a reduction in margin of safety.

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

thisreview, 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: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: August 7, 1995 (TS 95-18)

Description of amendment request: The proposed change would revise

the titles of various administrative positions found in Section 6.0 of

the Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c).

Operation of Sequoyah Nuclear Plant (SQN) in accordance with the

proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes only involve the administrative titles of

management positions in TVA [Tennessee Valley Authority]. Plant

equipment and operating practices are not affected by the proposed

administrative changes. Therefore, there is no increase in the

probability or consequences of an accident.

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

from any previously analyzed.

Plant features are not impacted by the proposed revision;

therefore, this revision can not create the possibility of a new or

different accident.

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

Plant setpoints and features that establish and maintain the

margin of safety for SQN are not involved in the proposed

administrative TS change. Therefore, the margin of safety is not

reduced by the proposed change.

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

thisreview, 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 Romm location: Chattanooga-Hamilton County Library,1101

Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: August 7, 1995 (TS 95-03)

Description of amendment request: The proposed change would modify

Technical Specifications (TS) 3/4.1.3, ``Movable Control Assemblies,''

and Bases 3/4.1.3. The proposed change addresses operation with a rod

urgent failure condition (the control rods are out-of-service because

of failures external to the individual rod drive mechanisms; i.e.,

programming circuitry, but the control rods remain operable), including

limited operation with one control or shutdown bank inserted up to 18

steps below its insertion point. In addition, the surveillance interval

for rod movement verifications would be increased from 31 days to 92

days.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c).

Operation of Sequoyah Nuclear Plant (SQN) in accordance with the

proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Allowing for continued operation during diagnosis and repair as

a result of electronic or electrical malfunctions of the rod control

system is acceptable, since the design safety function of the

control rods (reactor trip will remain unaffected during the

diagnosis and repair period. During the extended

[[Page 45187]]

troubleshooting and repair period, the requirements for control rod

alignment, insertion limits (except for a small allowed deviation

for one bank) and shutdown margin will be maintained. The small

deviation from the control rod insertion limits allowed for one

bank, for up to 72 hours, will not adversely impact the current TS

requirements for normal operation core power distributions. The

proposed changes do not affect the ability of the control rods to

perform their intended safety function (rods remain trippable) when

a safety system setting is reached. No new or unique accident

precursors be introduced by the proposed changes. Therefore, the

probability and consequences of accidents related to or dependent on

control rod operation will remain unaffected.

The proposed change will result in a small increase in the

probability, that at any given time, a control or shutdown bank will

be inserted slightly below (i.e., up to 18 steps) its insertion

limit. However, by design, the control and shutdown banks will

continue to meet the safety analysis criterion for steady state and

American Nuclear Society (ANS) Condition II (moderate frequency)

transients. The allowed insertion is not a malfunction of equipment

important to safety in this case; therefore, the probability of such

a malfunction is not increased. Limiting the allowed time for

operation with the rod control system out-of-service, but with the

rods trippable and with a control or shutdown bank below the

insertion limit, eliminates the need for consideration of this

condition coincident with any of the low frequency (ANS Condition

III or IV) design basis accidents.

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

from any previously analyzed.

There are no new failure mechanisms associated with plant

operation for an extended period to perform diagnosis and repair on

the rod control system. Limited periods of operation with immovable,

but trippable control rods, does not involve any modification to the

operational limits or physical design of the involved systems. There

are no new accident precursors created because of the allowed

diagnosis and repair period.

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

The results of the current accident analyses are not impacted by

the change. In addition, the margin of safety as defined in the

basis of the TS has not been reduced because current core design

limits continue to be met for the accidents of concern. Therefore,

the margin of safety is not impacted.

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

thisreview, 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: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: June 23, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Surveillance Requirements

4.1.3.1.2, 4.4.6.2.2.b, 4.4.3.2, 4.6.2.1.d, 4.6.4.2, and Table 4.3-3 in

accordance with guidance provided in NRC Generic Letter (GL) 93-05,

``Line Item Technical Specification Improvements to Reduce Surveillance

Requirements for Testing During Power Operations.'' Additionally, the

proposed amendment would revise TS 4.1.1.1.1, 4.1.1.2, 3/4.1.3.1 and

associated Bases to implement portions of the Standard Technical

Specifications - Westinghouse Plants, NUREG-1431.

Basis for proposed no significant hazards consideration

determination: As requied 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 Technical Specification changes do not involve a

significant hazards consideration per 10 CFR 50.92 because operation

of Callaway Plant with the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

All changes are in accordance with the recommendations of NRC

Generic Letter 93-05, Line-Item Technical Specifications

Improvements to Reduce Surveillance Requirements for Testing During

Power Operation or NUREG 1431, Standard Technical Specifications -

Westinghouse Plants. None of the changes affects accident initiators

and each has been evaluated against Callaway Plant operating

experience.

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

from any previously evaluated.

The proposed Technical Specification changes do not modify any

equipment nor create any potential accident initiators. The changes

per GL 93-05 involve Technical Specification surveillance

frequencies and do not alter the methodology nor associated

acceptance criteria. The changes per NUREG-1431 do not create any

accident initiators and are consistent with Callaway design and

operation.

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

The surveillance frequency changes were recommended via GL 93-05

and are compatible with Callaway Plant experience. The changes per

NUREG-1431 do not impact the margin of safety. The Shutdown margin

requirements and associated safety margins are unaffected by these

changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

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

NRC Project Director: Gail H. Marcus

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: June 26, 1995

Description of amendment request: The proposed amendment would

revise the allowed outage time for component cooling water motor

operated containment isolation valves, remove the list of containment

isolation valves, and allow containment penetration check valves to be

used as isolation devices.

Basis for proposed no significant hazards consideration

determination: As requied 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 revision to TS 3/4.6 to remove the listing of

containment isolation valves, revise the ACTION Statement for the

CCW MOVs, and credit penetration check valves as isolation devices

does not involve a significant hazards consideration because

operation of Callaway Plant with this change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes simplify the TS, meet the regulatory

requirements for control of containment isolation and are consistent

with the guidelines of GL 91-08. The information contained in Table

3.6-1 has not been changed, but only relocated to a different

controlling document. This is an administrative change which should

result in improved plant practices and have no impact on plant

operations. Addition of the footnote to allow up to 12 hours for

valve testing does not affect the severity of any accident

previously evaluated. The proposed revision to the TS will not

adversely impact plant safety since the second barrier of the two

required is still available to provide isolation between the

containment atmosphere or the reactor coolant system and the outside

atmosphere.

[[Page 45188]]

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

from any previously evaluated.

There are no design changes being made that would create a new

type of accident or malfunction and the method and manner of plant

operation remain unchanged. Addition of the footnote to allow up to

12 hours for valve testing does not affect the severity of any

accident previously evaluated. The additional time provides

assurance that the inoperable valve is in proper working order prior

to returning it to OPERABLE condition.

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

There are no changes being made to the safety limits or safety

system settings that would adversely impact plant safety.

Containment isolation will still be maintained as provided by the

second isolation valve to ensure that the release of radioactive

material to the environment will be consistent with the assumptions

used in the analyses for a LOCA. This will assure that containment

integrity is maintained.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

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

NRC Project Director: Gail H. Marcus

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: July 25, 1995

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.8.1 and its associated Bases to

improve overall emergency diesel generator reliability and

availibility.

Basis for proposed no significant hazards consideration

determination: As requied 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 changes do not involve a significant hazards

consideration because operation of Callaway Plant with these changes

would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

These proposed changes do not involve a change in the

operational limits or physical design of the emergency power system.

Emergency diesel generator operability and reliability will continue

to be assured while minimizing the number of required emergency

diesel generator starts. Also, emergency diesel generator

reliability will be enhanced by minimizing severe test conditions

which can lead to premature failures.

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

from any previously evaluated.

These proposed changes do not involve a change in the

operational limits or physical design of the emergency power system.

The performance capability of the emergency diesel generator will

not be affected. Emergency diesel generator reliability and

availability will be improved by the implementation of the proposed

changes. There is no actual impact on any accident anaiysis.

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

These proposed changes do not involve a change in the

operational limits or physical design of the emergency power system.

The performance capability of the emergency diesel generator will

not be affected. Emergency diesel generator reliability and

availability will be improved by the implementation of the proposed

changes. No margin of safety is reduced.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

& 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 29, 1994

Description of amendment request: The proposed change would revise

and update the NA-1&2 Environmental Protection Plan (EPP) to reflect

current obligations to the Commonwealth of Virginia, revise portions of

the transmission corridor rights-of-way erosion control program for

clarification and to be consistent with the state regulations,

eliminate inconsistencies, and delete obsolete material. Specifically,

references to National Pollutant Discharge Elimination System (NPDES)

permits are changed to reflect the correct permit title, Virginia

Pollutant Discharge Elimination System (VPDES). Vegetation and aquatic

biota studies referred to in the EPP were satisfactorily completed on

or before June 24, 1986. The discussion of the detailed subject matter

in these studies is removed because it is extraneous information. A

reference to 10 CFR 51.5(b)(2) (which does not exist) is corrected to

10 CFR 51.60(b)(2). The explicit reporting requirements for unusual or

important environmental events are replaced with the reporting

requirement which the NRC has required pursuant to 10 CFR 50.72

(b)(2)(vi). Therefore, the reporting inconsistency (EPP requires report

to NRC within 24 hours, whereas the 10 CFR 50.72 requires a four hour

report to the NRC) is resolved. The description of the audit program to

be utilized for auditing the EPP is replaced by referring to the Audit

Program established in accordance with 10 CFR 50, Appendix B. Another

inconsistency is eliminated by revising the two year records retention

requirement for erosion control inspection field logs to five years.

This makes the requirement consistent with EPP Section 5.2, Records

Retention. References to the State Water Control Board are updated to

that agency's successor, the Department of Environmental Quality.

Additionally, the licensee's obligation to comply with Virginia

regulations concerning erosion and sediment control within the

transmission corridor rights-of-way are recognized to eliminate

redundancy with previous EPP commitments. The Virginia Soil and Water

Conservation Board is recognized as the regulatory authority concerning

erosion within the transmission corridor rights-of-way. The Virginia

Soil and Water Conservation Board reviews and approves erosion and

sediment control specifications submitted by utilities on an annual

basis.

Basis for proposed no significant hazards consideration

determination: As requied 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 EPP changes will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated. The likelihood

that an accident will occur is neither increased or decreased by the

proposed changes to the EPP. Sufficient controls are established to

ensure that environmental controls impacting safety-related

structures, systems, and components are maintained current and

accurate. The only potentially credible accident which might be

affected is the Loss of Offsite Power (if erosion were severe

[[Page 45189]]

enough to undermine the bases of a transmission tower). Each of the

three 500 KV transmission lines connected to North Anna Power

Station can supply sufficient power to the site. This limits the

effect that one transmission tower has on safe operation of the

nuclear facility. However, the erosion noted to date has not been

severe enough to make such an accident credible. Additionally, each

of the 500 KV transmission lines are inspected for material

condition annually. Although the intent of this inspection is not

soil erosion (the annual erosion inspections are currently conducted

by another group who specializes in land management), evidence of

severe erosion would be noted and addressed as appropriate.

Therefore, this EPP 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. Nuclear station systems, equipment, or components are

not affected by the proposed changes. Thus, the consequences of a

malfunction of equipment important to safety 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

changes do not involve changes to the physical plant or operations.

... the proposed EPP changes do not contribute to accident

initiation and therefore do not produce a new accident scenario or

produce a new type of equipment malfunction. Also, this EPP change

does not alter any existing accident scenarios. The proposed changes

do not affect nuclear plant equipment or its operation, and thus do

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

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

new or different kind of accident.

(3) Involve a significant reduction in a margin of safety. The

EPP does not have a formal basis description other than the

discussion in the FES-OL [Final Environmental Statement-Operating

License]. The FES-OL discusses the non-radiological impacts of

facility construction and operation on the environment. The

discussion indicates that the environment will be managed to a

stabilized condition during the operations phase, and a program will

be implemented to maintain the environment in a stabilized

condition. This intent is not altered by the proposed changes to the

EPP. The proposed changes do not affect nuclear plant equipment or

its operation, and thus do not involve any reduction in the margin

of safety.

Therefore, use of the proposed EPP would 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: David B. Matthews

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: July 26, 1995

Description of amendment request: The proposed changes would revise

the Technical Specifications (TS) for the North Anna Power Station,

Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed changes

would increase the pressurizer safety valve lift setpoint tolerance as

well as reduce the pressurizer high pressure reactor trip setpoint and

allowable value.

The licensee has prepared a safety evaluation which justifies

increasing the current TS pressurizer safety valve (PSV) at-power

(Modes 1-3) lift setpoint tolerance from plus or minus 1% as-found and

plus or minus 1% as-left to +2%/-3% average as-found with no single

valve outside plus or minus 3% as-found and plus or minus 1% per valve

as-left. The as-found value is based on testing, the results of which

are expressed as an error (i.e., positive or negative percentage

deviation from the nominal lift setpoint). The errors of the tested

valves are summed and the result divided by the number of valves

tested. This result is compared to the acceptable range of +2% to -3%.

No single valve is allowed to be outside of the plus or minus 3%

tolerance.

The safety evaluation also supports an increase to the Hot Shutdown

(Mode-4) required PSV lift setpoint tolerance from plus or minus 1% as-

found and plus or minus 1% as-left to plus or minus 3% per valve as-

found and plus or minus 1% per valve as-left. These proposed changes

will provide greater operational flexibility in meeting periodic test

requirements established by the safety analyses.

A concurrent reduction in the pressurizer high pressure reactor

trip setpoint and allowable value of TS Table 2.2-1 are also proposed.

These changes ensure that the analysis results for the loss of external

load accident continue to meet the acceptance criteria with the higher

PSV tolerance.

The Loss of Load, Locked Rotor, and Rod Withdrawal event analyses

demonstrate that increasing the at-power PSV lift setpoint tolerance to

+2%/-3% average as-found with no single valve outside plus or minus 3%

as-found and plus or minus 1% per valve as-left does not result in a

transient pressure in excess of the overpressure safety limit. Further,

the increased setpoint tolerance does not adversely impact the DNBR

[departure from nucleate boiling ratio] results of any North Anna UFSAR

[Updated Final Safety Analysis Report] Chapter 15 transient analysis.

Mode 4 overpressure protection is adequate with one PSV with a

tolerance of plus or minus 3%.

Finally, the increased PSV setpoint tolerances and reduction of the

high pressurizer pressure reactor trip setpoint do not present any

operational considerations which would significantly impact the

performance of the plant during normal operation or during postulated

accident conditions. In summary, each pertinent safety criterion was

evaluated for the proposed TS changes, and all were found to be

acceptable.

Basis for proposed no significant hazards consideration

determination: As requied 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 North Anna Power Station in

accordance with the proposed Technical Specifications changes will

not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

Affected safety related parameters were analyzed for a change to

North Anna 1 and 2 Technical Specifications 3.4.2 and 3.4.3 and

Table 2.2-1 item 10. It was determined that the overpressure safety

limits would not be exceeded in the most limiting overpressure

transients (Loss of Load, Locked Rotor, and Rod Withdrawal events)

with the as-found pressurizer safety valve lift setpoint tolerance

increased to an average of +2%/-3%, no single valve outside of [plus

or minus] 3%, and the 25 psi reduction in the Pressurizer High

Pressure Reactor Trip setpoint. The DNBR results of transients

impacted by the proposed setpoint tolerance increase meet the

acceptance criterion after accounting for the impact of the proposed

changes. The increased setpoint tolerance will not result in an

inadvertent opening of the pressurizer safety valves. Mode 4

overpressure protection is adequate with one PSV with a tolerance of

[plus or minus] 3%.

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

from any accident previously identified.

The proposed change to North Anna 1 and 2 Technical

Specifications 3.4.2 and 3.4.3 and Table 2.2-1 item 10 does not

involve any changes which would introduce any new or unique

operational modes or accident precursors. Only the allowable

tolerance about the existing PSV lift setpoint will be changed,

along with a reduction in the

[[Page 45190]]

pressurizer high pressure reactor trip setpoint.

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

It was determined that the most limiting overpressure transients

do not result in maximum pressures in excess of the overpressure

safety limits. The DNBR results of transients impacted by the

proposed setpoint tolerance increase meet the acceptance criterion

after accounting for the impact of the proposed changes. Therefore,

the margin of safety is unchanged by the proposed increase in the

safety valve setpoint tolerances.

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: David B. Matthews

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: July 26, 1995

Description of amendment request: The proposed change would revise

the Technical Specifications (TS) for the North Anna Power Station,

Units No. 1 and No. 2 (NA-1&2). Specifically, the change would clarify

the TS to allow switching of charging and low-head safety injection

pumps during unit shutdown conditions. The proposed changes would also

allow additional methods of rendering these same pumps incapable of

injecting into the reactor coolant system (RCS) when required for low-

temperature conditions. NA-1&2 is equipped with three charging pumps.

These charging pumps provide inventory control, normal boration to the

RCS, and flow to the reactor coolant pump seals. They also act as the

high-head safety injection pumps during accident conditions. During

certain shutdown conditions, it is necessary to render two of the three

charging pumps inoperable to maintain the low-temperature overpressure

protection (LTOP) design bases assumptions. This provides assurance

that a mass addition pressure transient can be relieved by the

operation of a single pressurizer power-operated relief valve (PORV).

Low-temperature overpressure protection for each NA-1&2 unit is

provided by two pressurizer PORVs.

During shutdown conditions, periodic surveillance testing of the

charging pumps is required by the NA-1&2 TS. Also during shutdown

conditions, it may be desirable to switch from one charging pump to

another to allow for other activities such as maintenance or testing.

The current NA-1&2 TS associated with charging pumps during

shutdown conditions are very restrictive and do not allow sufficient

latitude for surveillance testing or pump switching. The current NA-1&2

TS specifically state in the surveillance requirements that the method

used to render a charging pump inoperable is to place the pump control

switch in the pull-to-lock position. This requirement would not allow

for surveillance or post-maintenance testing of the inoperable charging

pumps since this switch is used to start those pumps.

Therefore, the licensee proposes to modify NA-1&2 TS to allow more

than one charging pump to be operable and capable of injecting into the

RCS for pump switching operations. Additionally, the methods used to

render charging pumps inoperable will be expanded to allow for post-

maintenance and surveillance testing.

Basis for proposed no significant hazards consideration

determination: As requied 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 North Anna Power Station in

accordance with the proposed Technical Specifications changes will

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Allowing more than one charging pump to be operable and capable

of injecting into the RCS during RCS low temperature operation for

pump switching for post-maintenance and surveillance testing does

not increase the probability of occurrence or the consequences of

any previously analyzed accident. Pump switching operations will be

under the direct administrative control of a licensed operator and

will only be for a short duration of time. Any situation that could

result in an excessive RCS mass addition would be immediately

recognized by the operator and remedial action would be taken to

prevent challenges to RCS integrity. Using methods such as opening

the charging pump power supply breaker or closing the charging pump

discharge valve(s) to render a charging pump inoperable will ensure

that these pumps will not be capable of injecting water into the

RCS. These alternate methods are as effective as placing the control

switches in the pull-to-lock position.

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

from any accident previously evaluated.

Allowing more than one charging pump to be operable and capable

of injecting into the RCS during low-temperature operation for pump

switching for post-maintenance and surveillance testing does not

involve any physical modifications of the plant nor result in a

change in a method of operation. Licensed operator control of

charging pump switching operations will continue to ensure that the

RCS will not be challenged by excessive mass addition events. Using

methods other than placing charging pump control switches in the

pull-to-lock position to render the pump inoperable will still

ensure that only one pump will be capable of injecting into the RCS

during low temperature operations. Therefore, a new or different

type of accident is not made possible.

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

Allowing more than one charging pump to be operable and capable

of injecting into the RCS during RCS low temperature operation for

pump switching for post-maintenance and surveillance testing does

not affect any safety limits or limiting safety system settings. The

alternate methods of rendering pumps inoperable provide the same

level of assurance that the pump is incapable of flowing into the

RCS as placing the pump control switch in the pull-to-lock position.

System operating parameters remain unaffected. The availability of

equipment required to mitigate or assess the consequence of an

accident is not reduced. Safety margins are, therefore, not

decreased.

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: David B. Matthews

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: July 20, 1995

Description of amendment request: The proposed amendments would: 1)

revise three Reactor Protection System/Engineered Safety Features

Actuation Systems channel trip setpoint limits, 2)

[[Page 45191]]

add a new setpoint limit for high high steam generator water level, and

3) incorporate editorial changes to revise the measurement units of one

setpoint limit and to delete certain references to two-loop operation.

Basis for proposed no significant hazards consideration

determination: As requied 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 Surry Power Station

with the proposed change will not:

(1) Involve a significant increase in either the probability of

occurrence or consequences of any accident or equipment malfunction

scenario which is important to safety and which has been previously

evaluated in the Updated Safety Analysis Report (UFSAR). The effect

of the proposed change is to ensure that actual plant setpoints

remain conservative consistent with respect to accident analysis

assumptions. The proposed change requires safety system actuation

limits that are more conservative than those currently in Technical

Specifications. The change does not invalidate currently implemented

station setpoints or currently applicable accident analysis

assumptions regarding these setpoints. Consequently, the results and

conclusions of the current UFSAR accident analyses are not affected

by these changes. The proposed Technical Specifications change

revises setpoints used to mitigate accidents and therefore has no

bearing on the probability of an accident. Further, the change

ensures that the setpoints used to mitigate an accident bound the

setpoints used in the accident analyses. Therefore, the probability

of an accident or consequences of an accident is not adversely

affected as a result of this change.

(2) Create the possibility of a new or different type of

accident than those previously evaluated in the UFSAR. Implementing

the proposed Technical Specifications setpoint limits cannot create

the possibility of an accident of a different type than was

previously evaluated in the UFSAR. Since actual plant setpoints are

not being affected, new accident precursors will not be introduced.

Furthermore, spurious challenges to safety systems are also not

expected to increase in frequency as a result of these changes since

actual setpoints installed in the plant are not being changed.

Consequently, no new accident precursors are created as a result of

the new Technical Specifications setpoint limits.

(3) Involve a significant reduction in a margin of safety. Since

the results of the existing UFSAR accident analyses remain bounding,

safety margins are not impacted. The proposed Technical

Specifications setpoint limits ensure plant setpoints remain

conservative and consistent with design base accident analysis

assumptions including appropriate instrument channel uncertainties

due to harsh environmental conditions. Therefore, the margin of

safety as defined in the Technical Specifications bases is

unaffected.

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: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219

NRC Project Director: David B. Matthews

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey County, Kansas

Date of amendment request: July 25, 1995

Description of amendment request: This license amendment request

proposes to revise Technical Specification 4.0.5a and Bases Section 3/

4.4.10 to delete the clause ``(g), except where specific written relief

has been granted by the Commission pursuant to 10 CFR Part 50, Section

50.55a(g)(6)(i).'' This proposed change is consistent with NUREG-1482,

``Guidelines for Inservice Testing and Nuclear Power Plants.''

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This proposed change would remove the wording ''...(g), except

where specific written relief has been granted by the Commission

pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice

Inspection and Testing Programs are described in the technical

specifications pursuant to 10 CFR 50.55a. In addition, the proposed

change, in accordance with NUREG-1431 and NUREG-1482, would provide

relief to the ASME Code requirement in the interim between the time

of submittal of a relief request until the NRC has issued a safety

evaluation and granted the relief. The change being proposed is

administrative in nature and does not affect assumptions contained

in plant safety analyses, the physical design and/or operation of

the plant, nor does it affect any technical specification that

preserves safety analysis assumptions. Any relief from the approved

ASME Section XI Code requirements will require a 10 CFR 50.59

evaluation to ensure no technical specification changes or

unreviewed safety questions exist. Therefore, operation of the

facility in accordance with the proposed change would not affect the

probability or consequences of an accident previously analyzed.

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

or different kind of accident from any accident previously

evaluated.

This proposed change would remove the wording ''...(g), except

where specific written relief has been granted by the Commission

pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice

Inspection and Testing Programs are described in the technical

specifications pursuant to 10 CFR 50.55a. In addition, the proposed

change, in accordance with NUREG-1431 and NUREG-1482, would provide

relief to the ASME Code requirement in the interim between the time

of submittal of a relief request until the NRC had issued a safety

evaluation and granted the relief. The change being proposed is

administrative in nature and will not change the physical plant or

the modes of operation defined in the facility license. The change

does not involve the addition or modification of equipment nor does

it alter the design or operation of plant systems. Any relief from

the approved ASME Section XI Code requirements will require a 10 CFR

50.59 evaluation to ensure no technical specification changes or

unreviewed safety questions exist. Therefore, operation of the

facility in accordance with the proposed change would 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.

This proposed change would remove the wording ''...(g), except

where specific written relief has been granted by the Commission

pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).'' The Inservice

Inspection and Testing Programs are described in the technical

specifications pursuant to 10 CFR 50.55a. In addition, the proposed

change, in accordance with NUREG-1431 and NUREG-1482, would provide

relief to the ASME Code requirement in the interim between the time

of submittal of a relief request until the NRC has issued a safety

evaluation and granted the relief. The change being proposed is

administrative in nature and will not alter the bases for assurance

that safety-related activities are performed correctly or the basis

for any technical specification that is related to the establishment

or maintenance of a safety margin. Any relief from the approved ASME

Section XI Code requirements will require a 10 CFR 50.59 evaluation

to ensure no technical specification changes or unreviewed safety

questions exist. Therefore, operation of the facility in accordance

with the proposed change would 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 locations: Emporia State University,

[[Page 45192]]

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

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consid

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