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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from June 24, 1994, through July 8, 1994. The

last biweekly notice was published on July 6, 1994 (59 FR 34657).

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

20555. The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By August 19, 1994, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC 20555 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 20555, 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 20555, 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 (PVNGS),

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

Date of amendment requests: June 17, 1994.

Description of amendment requests: The proposed changes would

enhance the PVNGS technical specifications (TS) by removing five tables

of component lists in accordance with NRC Generic Letter (GL) 91-08,

``Removal of Component Lists from Technical Specifications.'' The

affected tables are 3.3-9B, 3.3-9C, 3.6-1, 3.8-2, and 3.8-3. The

references to these five tables will also be removed from the text of

the TS in accordance with the sample TS change amendment provided by

the NRC in GL 91-08. These five removed tables will be incorporated

into a new document, which will be administratively controlled

according to the change controls provisions of the TS.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1--Involve a significant increase in the probability or

consequence of an accident previously evaluated:

The proposed amendment will remove five tables of component lists

from the TS and add them to an administratively controlled document

which is subject to the change controls provisions of the TS. The new

location of the affected component lists is easily retrievable. The

existing TS requirements and those components to which they apply are

not altered by this TS amendment. There are no changes to the

operations, maintenance, surveillance, and/or qualification of any

component on the removed lists. Therefore, the probability of

occurrence and the consequences of any previously evaluated accident is

[sic] not changed.

Standard 2--Create the possibility of a new or different kind of

accident from any accident previously evaluated:

The TS requirements and the components to which they apply are not

altered by this amendment. The removed component lists are added to a

controlled and easily retrievable document. This amendment has no

impact on plant operations, maintenance, testing, or component

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

accident is not created by this amendment.

Standard 3--Involve a significant reduction in a margin of safety:

The removal of these five component lists from the TS does not

alter existing TS requirements or those components to which they apply.

These lists will be added to an administratively controlled document

which is subject to the controls provisions of the TS. More

specifically, there is no impact on safe plant shutdown, maintenance or

hot standby, containment isolation capability, containment leakage

rate, and/or the operability of safety related valves. Therefore,

removal of these five component lists from the TS will not involve a

reduction in the margin of safety.

The NRC staff has reviewed the licensees' 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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004.

Attorney for licensees: 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: Theodore R. Quay.

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: May 27, 1994.

Description of amendments request: The proposed amendment would

revise the Technical Specification surveillance test interval from

monthly to quarterly for several channel functional tests for the

Reactor Protective System and the Engineered Safety Feature Actuation

System (ESFAS). In addition, an administrative change to the ESFAS

table would remove an out-of-date footnote concerning the emergency

diesel generator logic circuit modifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

consequences of an accident previously evaluated.

The Reactor Protective System (RPS) and the Engineered Safety

Features Actuation System (ESFAS) provide the actuation signals to

safety equipment necessary to mitigate design basis accidents and

transients. The proposed change would increase the surveillance test

interval from monthly to quarterly for several of the RPS and ESFAS

instrumentation channel functional tests. The RPS/ESFAS instruments are

not an initiator in any previously evaluated accidents. Therefore, the

proposed changed does not involve an increase in the probability of an

accident previously evaluated. The required plant-specific setpoint

drift analysis for Calvert Cliffs demonstrated that the observed

changes in instrument uncertainties for the extended surveillance test

interval do not exceed the current 30-day setpoint assumptions. This

provides confidence the 90-day test interval will not impact the

ability to detect and monitor system degradation. Therefore, the

proposed change will not change the ability of the RPS/ESFAS

instrumentation to respond to and mitigate the consequences of any

previously evaluated accident. In addition, an obsolete footnote is

removed from ESFAS Table 4.3-2.

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 extended surveillance test interval for the RPS and

ESFAS and the removal of the obsolete footnote does not involve any

changes in equipment or the function of these instruments. The proposed

change does not represent a change in the configuration or operation of

the plant. The RPS and ESFAS setpoints will not be changed as the

instrument uncertainties resulting from the proposed surveillance test

interval (calculated using actual plant data) are less than the

instrument uncertainties assumed for the current surveillance interval.

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 will not affect the functions of the RPS or the

ESFAS instruments. The CEN-327 and CEN-327, Supplement 1, topical

reports quantified the corresponding changes in core melt frequency for

the representative fault tree models that were developed for Calvert

Cliffs. The proposed change has two principal effects with opposing

impacts on core melt frequency. The first impact is a slight increase

in core melt frequency that results from the increased unavailability

of the instrumentation in question. This assumed unavailability results

from less frequent testing. The unavailability of the tested

instrumentation components represents the potential for the failure of

the reactor to trip, an Anticipated Transient Without Scram, or a

failure of the appropriate engineered safety features to actuate when

required. The opposing impact on core melt risk is the corresponding

reduction in core melt frequency that would result due to the reduced

exposure of the plant to test-induced transients. The two changes are

nearly equal and the net result is no distinguishable effect on plant

safety. The NRC issued a Safety Evaluation Report which found that

these evaluations were acceptable for justifying the extensions in the

surveillance test intervals for the RPS and ESFAS from 30 days to 90

days.

The RPS and ESFAS setpoints will not be changed since the

instrument drift resulting from the proposed surveillance test interval

is less than the instrument drift presently assumed for the current

surveillance interval. This provides confidence the 90-day test

interval will not impact the ability to detect and monitor system

degradation. The removal of the ESFAS Table footnote only removes

obsolete information from the Technical Specifications. The conclusions

of the accident analyses in the Calvert Cliffs Updated Safety Analysis

Report remain valid and the safety limits continue to be met.

Therefore, the proposed change does not involve a significant reduction

in a margin of safety.

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

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

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,

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

NRC Project Director: Michael L. Boyle.

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: June 8, 1994.

Description of amendments request: The proposed amendments would

revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (CC1/

2), Technical Specifications (TS) 4.6.2.2.b to extend the surveillance

interval for the containment fan coolers from 18 months to 24 months.

The requested changes have been submitted in accordance with Generic

Letter (GL) 91-04, ``Guidance on Preparation of a License Amendment

Request for Changes in Surveillance Intervals to Accommodate a 24-Month

Fuel Cycle.'' These proposed amendments are part of a series of

requests that will eliminate the need for mid-cycle surveillance

outages to accommodate the existing 18 month surveillance requirements

since CC1-2 is operating on 24-month fuel cycles.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

consequences of an accident previously evaluated.

The purpose of the Containment Air Cooling (CAC) System is to cool

the containment atmosphere, and thereby limit containment pressure and

temperature, following a Loss of Coolant Accident (LOCA) or Main Stream

Line Break in containment. Failure of the CAC System is not an

initiator for any previously analyzed accident. Therefore, the proposed

change does not involve an increase in the probability of an accident

previously evaluated.

Historical CAC System reliability, monthly surveillances and

monitoring of CAC-related plant parameters provide assurance that

undetected system degradation will not occur between 24-month

surveillances, and the system will continue to perform its safety

function. Therefore, there will be no significant increase in the

consequences of accidents previously evaluated. Therefore, the proposed

Technical Specification change does not increase 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.

This requested revision to increase the interval for a CAC

surveillance from 18 to 24 months does not involve a significant change

in the design or operation of the plant. No hardware is being added to

the plant as part of the proposed change. The proposed change will not

introduce any new accident initiators. Therefore, the proposed change

would not create the possibility of a new or different type of accident

from any accident previously evaluated.

3. Does operation of the facility in accordance with the proposed

amendment involve a significant reduction in a margin of safety.

The CAC System provides a margin of safety by providing a means by

which containment pressure can be limited following a LOCA or Main

Steam Line Break. The proposed change does not affect the operation or

design of the CAC System. Historical monthly surveillances and Control

Room indications give assurance that the reduction in surveillance

frequency will not adversely affect our ability to detect degradation

in the system. Therefore, the proposed change does not involve a

significant reduction in a margin of safety.

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

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

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,

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

NRC Project Director: Michael L. Boyle.

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: June 8, 1994.

Description of amendments request: The proposed amendments would

revise the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Technical

Specifications 4.8.1.1.1.b to extend the alternate 69 kV offsite power

circuit surveillance frequency from 18 to 24 months.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

consequences of an accident previously evaluated.

The purpose of the 69 kV Southern Maryland Electric Cooperative

(SMECO) power source is to act as an independent energy source for

achieving and maintaining safe shutdown of the plant if the 500 kV

system is not available. Failure of the 69 kV SMECO system is not an

initiator for any existing accident. Therefore, the proposed change

does not involve an increase in the probability of an accident.

The 69 kV SMECO system could be used to mitigate the consequences

of accidents involving a loss of primary offsite power. However, the

accident analyses assume that if the 500 kV circuits were not

available, the Emergency Diesel Generators would be used to provide

power to maintain the plant in a safe shutdown condition. A historical

review of surveillance test results indicates the system has

experienced only one significant failure in the last ten years. In

addition, the system is routinely used.

However, the SMECO system is not assumed to function in our

accident analysis, so this change will result in no significant

increase in the consequences of accidents previously evaluated.

Therefore, the proposed Technical Specification change does not

increase 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?

This requested increase in the interval for a 69 kV SMECO

surveillance from 18 to 24 months does not involve a significant change

in the design or operation of the plant. No hardware is being added to

the plant of the proposed change. The proposed change will not

introduce any new accident initiators. Therefore, the proposed change

would not create the possibility of a new or different type of accident

from any accident previously evaluated.

3. Does operation of the facility in accordance with the proposed

amendment involve a significant reduction in a margin of safety?

The 69 kV SMECO system provides a margin of safety by providing an

alternate offsite electrical power source. The proposed change does not

affect the operation or design of the 69 kV SMECO system. Historical

surveillance data and routine use indicates that the reduction in

surveillance frequency will not adversely affect our ability to detect

degradation in the system. Therefore, the proposed change does not

involve a significant reduction in a margin of safety.

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

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

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,

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

NRC Project Director: Michael L. Boyle

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs

Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of amendment request: May 27, 1994.

Description of amendment request: This amendment would revise the

existing heatup and cooldown curves and rates to increase their

applicability to 30 Effective Full-Power Years (EFPY). The fluence

value that was used to determine the heatup and cooldown curves was

based on the peak fluence and EFPY at the end of Cycle 9 and the peak

predicted fluence for Cycles 10 and beyond. In addition, a variable-

setpoint low-temperature overpressure protection (VLTOP) system is

being installed to increase the allowable operating pressure band in

the Low-Temperature Overpressure Protection (LTOP) region. The VLTOP

system uses a variable power-operated relief valve setpoint to take

advantage of increased Appendix G pressure limits at increased reactor

coolant system (RCS) temperatures. This system will increase the

operating window in which the plant may operate during heatup and

cooldown.

Specifically, this amendment would revise the Unit No. 2 heatup and

cooldown curves and rates for the following Technical Specification

(TS) sections. TS Section 3.4.9.1.a would be revised to decrease the

maximum heatup rate from a fixed value of 75 deg.F/hr to a more

conservative variable heatup rate of 30 deg.F to 60 deg.F/hr which

varies with RCS temperature range from 70 deg.F to greater than 246

deg.F. TS Section 3.4.9.1.b would be revised to increase the RCS

temperatures for the maximum allowable cooldown rates. TS Figures

3.4.9-1 and 3.4.9-2 would be replaced by new RCS pressure Temperature

Limits. The revised curves and rates are based on the predicted fluence

value for Cycle 10 and beyond.

The following TS sections would be revised to support modifications

to the LTOP system. TS sections 3.4.9.3.a.1 and 3.4.9.3.a.2 would be

changed to ``trip setpoint below the curve in Figure 3.4.9-3*'' to

account for the VLTOP system. The footnote, ``When on shutdown cooling,

the PORV trip setpoint shall be less than or equal to 443 psia,'' has

been added for Shutdown Cooling Operation to maintain an extra setpoint

that is independent of RCS temperature and is equal to the lowest

variable setpoint. The Minimum Pressure and Temperature (MPT) Enable

would be changed from 305 deg.F to 301 deg.F. This change would

effect TS sections 3.1.2.1, 3.1.2.3, Table 3.3-3, 3.4.1.2, 3.4.1.3,

3.4.3, 3.4.9.3, 4.5.2, and 3.5.3. Due to the lower MPT Enable

temperature, the transition region at which the high pressure safety

injection pumps are placed under manual control on cooldown and

restored to automatic status on heatup would be changed from a

temperatue range of 305 deg.F-350 deg.F, to 301 deg.F-325 deg.F.

This affects TS 3.5.3 and Table 3.3-3. TS Bases Sections B3/4.4.1, B3/

4.4.9, and 3/4.5.2 would change to be consistent with the proposed

change and to provide additional clarification of some of the existing

bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

consequences of an accident previously evaluated.

The Low Temperature Overpressure Protection (LTOP) system,

including the administrative controls, ensures that the 10 CFR Part 50,

Appendix G, Pressure-Temperature (P-T) limits for the reactor pressure

vessel will not be violated while operating at low temperatures. The

heatup and cooldown curves are conservatively developed in accordance

with the fracture toughness requirements of 10 CFR Part 50, Appendix G,

as supplemented by the American Society of Mechanical Engineers Boiler

and Pressure Vessel Code Section III, Appendix G. The reactor vessel

material Adjusted RTNDT values are based on the conservative

methodology provided in Regulatory Guide 1.99, Revision 2.

Analyses show that the proposed use of a variable LTOP system will

not result in a significant increase in the probability of an

inadvertent opening of a Power-Operated Relief Valve (PORV) causing a

small break Loss-of-Coolant-Accident. The proposed heatup and cooldown

curves and associated limits continue to provide conservative

restrictions on Reactor Coolant System (RCS) pressure to minimize

material stresses in the RCS due to normal operating transients, thus

minimizing the likelihood of a rapidly propagating fracture due to

pressure transients at low temperatures. Because the proposed heatup

and cooldown curves and rates are based on conservative Appendix G

methods, and because the LTOP controls protect the Appendix G P-T

limits, the proposed curves and limits do not involve an increase in

the probability of accidents previously evaluated.

The proposed use of a variable PORV trip setpoint and the increase

in the allowable fluence at the reactor vessel wall results in the

changes to the heatup and cooldown curves and rates, the Minimum

Pressure and Temperature (MPT) Enable temperature, and high pressure

safety injection pump manual control transition temperature. These

proposed changes continue to provide sufficient margin to accommodate

postulated pressurization from mass and energy addition transients.

Calculations have been performed that predict the response to such

transients. Because the results of the analyses remain well within the

conservative acceptance limits of Appendix G, these changes do not

increase the consequences of accidents previously evaluated.

Therefore, the proposed changes do 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 new variable LTOP control system along with the proposed

changes to the Technical Specifications will ensure that the Appendix G

P-T limits will not be violated during low temperature operations.

While setpoints and curves have changed, this proposed change does not

introduce any operator actions that are significantly different from

current operator actions used at the plant. The variable LTOP system

will continue to have redundant channels to ensure that no single

equipment failure or operator error will result in violation of the P-T

limits. The use of a variable LTOP system does not create a new failure

mechanism for the PORV. The failure mechanism for the PORV continues to

be an inadvertent opening or the failure to open during a pressure

transient which has been previously evaluated. 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.

This change will ensure that the margin of safety is maintained

with respect to energy or mass addition events in that none of the

events postulated could challenge the Appendix G limits. The proposed

use of a variable PORV trip setpoint and the increase in the allowable

fluence at the reactor vessel wall necessitate the changes to the

heatup and cooldown curves and rates, the MPT Enable temperature, and

high pressure safety injection pump manual control transition

temperature. These changes ensure that the margin of safety is

maintained by protecting the Appendix G limits for all postulated

transients. Therefore, the proposed change does not involve a

significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,

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

NRC Project Director: Michael L. Boyle.

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

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

Carolina

Date of amendments request: June 17, 1994.

Description of amendments request: The proposed amendments would

revise the Technical Specifications (TS) to (1) remove the heatup and

cooldown curves from TS 3/4.4.6 and relocate them to a newly created

Pressure and Temperature Limits Report, and (2) remove the reactor

vessel material surveillance program withdrawal schedule from TS Table

4.4.6.3-1 and relocate it to the Updated Final Safety Analysis Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment [sic] does not involve a significant

increase in the probability or consequences of an accident previously

evaluated because the changes are administrative in nature. These

changes do not alter the configuration or operation of the facility.

The Limiting Safety Systems Settings and Safety Limits specified in the

current Technical Specifications remain unchanged.

2. The proposed amendment [sic] does not create the possibility of

a new or different kind of accident from any accident previously

evaluated. The safety analysis of the facility remains complete and

accurate. There are no physical changes to the facility and the plant

conditions for which the design basis accidents have been evaluated are

still valid. The operating procedures and emergency procedures are

unaffected.

3. The proposed amendment [sic] does not involve a significant

reduction in the margin of safety because these margins are established

through the Limiting Conditions of Operation, Limiting Safety System

Settings and Safety Limits specified in the Technical Specifications,

and since there are no changes to the physical design or operation of

the facility, these margins will not be changed.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Project Director: William H. Bateman.

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: January 19, 1994.

Description of amendment request: The proposed amendment would

revise the Technical Specifications by increasing the minimum reactor

coolant system temperature required for criticality from 500 degrees

Fahrenheit to 530 degrees Fahrenheit.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not result in a significant increase in

the probability or consequences of accidents previously evaluated. The

probability for an accident is independent of the changes being

proposed. Reactor criticality at 530 degrees Fahrenheit instead of the

nominal no-load Tavg of 547 degrees Fahrenheit does not affect any

of the accident initiators in the analyses, but does change one of the

initial conditions assumed in the Safety Analysis. However, the change

in initial conditions from the nominal no-load temperature of 547

degrees Fahrenheit to 530 degrees Fahrenheit does not increase the

probability of any of the events considered in the Safety Analysis. The

proposed Minimum Temperature for Criticality specification (530 degrees

Fahrenheit) will be more restrictive than the current specification

which allows reactor criticality at a temperature as low as 500 degrees

Fahrenheit. In addition, the Action Statement will require operator

response to place the reactor in a subcritical condition (Mode 3) with

15 minutes should the temperature drop below the limit for greater than

a specified amount of time (15 minutes).

Likewise, the proposed change does not significantly increase the

consequences of an accident previously evaluated. In the reanalysis of

the Zero Power accidents (Rod Withdrawal From Subcritical, Rod

Ejection, Main Steamline Break, Boron Dilution During Startup, and

Feedwater Malfunction) from an initial condition of 530 degrees

Fahrenheit, it was concluded that the results and conclusions in the

current Safety Analysis remain valid based on the fact that the current

analysis results are conservative and bounding for reactor criticality

at 530 degrees Fahrenheit. The LOCA transient analyses are unaffected

by the proposed change since they are initiated from the limiting

condition of 102 percent. Since the full power Tavg value is

unchanged by this proposed amendment, the LOCA analyses are unaffected.

The proposed change will ensure that plant parameters are within their

analyzed ranges prior to reactor criticality and appropriate operator

actions are taken should the temperature drop below the temperature

limit after reaching criticality.

The proposed administrative changes delete requirements which are

no longer applicable and will have no affect on the probability or

consequences of any accident previously evaluated in the analyses.

The proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated. The

change does not involve the addition of any new or different type 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. The proposed change will

ensure that plant parameters are within their analyzed ranges prior to

reactor criticality. The proposed administrative changes delete

requirements which are no longer applicable and will not create the

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

previously evaluated.

The proposed change does not involve a significant reduction in a

margin of safety. The proposed change does not affect any safety

related system or component operation or operability, instrument

operation, or safety system setpoints, and does not result in increased

severity of any of the accidents considered in the analysis. Operator

response to a drop in temperature after reaching criticality for a

specified period of time (greater than 15 minutes) will place the

reactor in a subcritical condition which is inherently more stable than

when critical below the Point of Adding Heat. The proposed

administrative changes are being made to clarify Technical

Specifications with no change of intent. Therefore, the proposed

changes do not create a significant reduction in a margin of safety.

In conclusion, based on the previous considerations, Commonwealth

Edison Company believes that the activities associated with this

Technical Specification amendment request satisfy the Significant

Hazards Consideration standards of 10 CFR 50.92(c) and, accordingly, a

finding that this Technical Specification amendment does not represent

a Significant Hazards Consideration is justified.

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: 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 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: June 24, 1994.

Description of amendment request: The proposed amendment would

revise the Technical Specifications by removing the containment

recirculation sump level instrumentation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant increase

in the probability of occurrence or consequences of any accident

previously evaluated.

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. Since the proposed change does not involve the

introduction of new or redesigned plant equipment, failure mechanisms

are not impacted. As a result, the probability of occurrence of

accidents previously evaluated is not increased.

The consequences of accidents previously evaluated are not

increased. Removal of [containment recirculation sump level] CRSL

instrumentation requirements from Technical Specifications does not

affect the ability to mitigate the consequences of any accident

previously evaluated. The CRSL instrumentation is used to verify that

[residual heat removal] RHR pumps have adequate [net positive suction

head] NPSH to operate in the recirculation mode following a Loss of

Coolant Accident (LOCA).

As stated in the Zion Updated Final Safety Analysis Report (UFSAR),

the changeover from the injection mode to the recirculation mode of

emergency core cooling is initiated when the low level alarm on the

RWST annunciates. This occurs when the RWST level drops to 13'-7''

(145,600 gallons).

At this point, sufficient water has been delivered to the

containment, from the Containment Spray (CS) system and, by Emergency

Core Cooling System (ECCS) injection, through the [reactor coolant

system] RCS break, to provide at least one foot of water above the

containment floor. One foot of water above the containment floor

provides sufficient volume to sustain the required NPSH of the RHR

pumps in the recirculation mode of operation. The water level in the

Containment Building during the recirculation phase will be

approximately 5 feet above the floor elevation (568') based on the

volume of the RCS, accumulators, and the minimum required volume of the

RWST. Minimum RWST volume of 350,000 gallons is required by Technical

Specification 3.8.1.F.

In summary, RWST level instrumentation, which satisfies Regulatory

Guide 1.97 qualification requirements for a Type A, Category 1

variable, provides the operator with the primary indication of the

appropriate time to initiate switchover to the recirculation mode, as

well as indication of adequate NPSH for the RHR pumps. Containment

Water Level (wide range) instrumentation which is qualified as a Type B

variable provides confirmatory indication of water level in

containment.

The proposed change does not affect the procedures controlling

operation of equipment, or systems required to mitigate the accidents

considered in the UFSAR. As such, 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 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 UFSAR. 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. Also, the methods of recovery from

accidents described in the UFSAR are not affected.

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.

No design margins are impacted and the newly chosen primary

indicator (RWST level) is both consistent with plant emergency

procedures and appropriately qualified. The proposed change will not

adversely impact the peak clad temperature, amount of fuel damage, or

offsite dose projected to occur from the design basis accidents. Thus,

the margin of safety is not diminished.

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: 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 60690.

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: June 16, 1994.

Description of amendment request: The proposed amendment will

remove a footnote applicable for Cycle 18 only regarding the

surveillance of the automatic bus transfer (ABT) system and add

surveillance requirement 4.8.3.1.2, to test the ABT once per refueling.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to delete a note from the limiting condition

for operation for the MCC-5 ABT scheme and to add the surveillance

requirement has no impact on the probability or consequences of an

accident previously evaluated. By removing the requirement to test the

scheme on-line, the probability of failure to mitigate an accident

while the Haddam Neck Plant is operational is incrementally decreased.

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

from any previously analyzed.

The proposed changes remove the requirements to disable the subject

ABT feature for testing, leaving the scheme undisturbed throughout

normal plant operation, and therefore does not create the possibility

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

analyzed.

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

The proposed changes require that the MCC-5 ABT feature be tested

during a plant shutdown rather than during normal operation. This will

place the ABT scheme in a test environment that has no significant

reduction in a margin of safety. The plant configuration that is

required to perform this test (refueling) would clearly place the

Haddam Neck Plant in a state that would be able to accept all possible

ABT scheme test outcomes, normal and abnormal. Therefore, these

proposed changes do not result in 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 10 CFR 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &

Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-

3499.

NRC Project Director: John F. Stolz.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: October 29, 1993, as supplemented on

March 28, 1994.

Description of amendment request: This amendment is an additional

followup to the amendment request of May 29, 1992, published in the

Federal Register (57 FR 30242) on July 8, 1992, which changed the

Technical Specifications Section 1.0, Definitions, to accommodate a 24-

month fuel cycle and which proposed the extension of the test intervals

for specific surveillance tests. This amendment proposes extending the

surveillance intervals to 24 months for the following additional

surveillance tests:

(1) Volume Control Tank Level Transmitter

(2) Containment High Range Area Radiation Monitors, R-25 and R-26

(3) Safety Injection System Electrical Loading

(4) Safety Injection (SI) System

(5) Reactor Coolant System Sub-Cooling Margin Monitor

The changes requested by the licensee are in accordance with

Generic Letter 91-04, ``Changes in Technical Specification Surveillance

Intervals to Accommodate a 24-Month Fuel Cycle.''

The October 29, 1993, submittal included surveillance tests for the

Auxiliary Feedwater System which duplicated a previous request which

was subsequently approved. The March 28, 1994, submittal withdrew the

duplicated request.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazard

consideration, which is presented below:

(1) Volume Control Tank Transmitter:

The proposed change does not involve a significant hazard

consideration since:

1. A significant increase in the probability or consequences of an

accident previously evaluated will not occur.

It is proposed that the channel calibration frequency for the

volume control tank level instrumentation be changed from every 18

months (+25%) to every 24 months (+25%).

A statistical analysis of channel uncertainty for a 30 month

operating cycle has been performed. Based upon this analysis it has

been concluded that sufficient margin exists between the existing

Technical Specification limits and the licensing basis Safety Analysis

limits to accommodate the channel statistical error resulting from a 30

month operating cycle. The existing margin between the Technical

Specification limit and the Safety Analysis limit provides assurance

that plant protective actions will occur as required. It is therefore

concluded that changing the surveillance interval from 18 months (+25%)

to 24 months (+25%) will not result in a significant increase in the

probability or consequences of an accident previously evaluated.

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

accident previously evaluated has not been created.

The proposed change is operating cycle length from a maximum of

22.5 months to 30 months resulting from an increased surveillance

interval will not result in a channel statistical allowance which

exceeds the current margin between the existing Technical Specification

limits and the Safety Analysis limits. Plant equipment, which will be

set at (or more conservatively than) Technical Specification limits,

will therefore provide protective functions to assure that Safety

Analysis limits are not exceeded. This will prevent the possibility of

any new or different kind of accident from that previously evaluated

from occurring.

3. A significant reduction in a margin of safety is not involved.

The change in surveillance interval from a maximum of 22.5 months

to 30 months resulting from an increased operating cycle will not

result in a channel statistical allowance which exceeds the margin

which exists between the current Technical Specification limit and the

licensing basis Safety Analysis limit. This margin, which is equivalent

to the existing margin, is necessary to assure that protective safety

functions will occur so that Safety Analysis limits are not exceeded.

(2) Containment High Range Area Radiation Monitors, R-25 and R-26:

The proposed change does not involve a significant hazard

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

It is proposed that the calibration frequency for the high-range

Containment Radiation Monitors (R-25 and R-26) be revised from every 18

months (+25%) to every 24 months (+25%).

These two monitors are redundant to each other and are used for

post accident monitoring purposes. They serve no function during normal

plant operation. Furthermore, they serve no purpose in preventing

accident initiation or mitigation. They are used for Emergency Planning

purposes to indicate a release of radioactivity to containment.

Review of past test results indicates that the devices have proven

reliable during past surveillances and there was no indication that

they would not remain operable for an extended operating cycle. In

addition, the devices are essentially redundant to each other. Each

device would respond to a release of radioactivity to Containment.

In consideration that the monitors are redundant, and in view of

the past test history of the monitors, there would be no significant

increase in the probability or consequences of an accident due to an

extended operating cycle.

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

previously analyzed has not been created.

The role of R-25 and R-26 is in the assessment of radiological

releases to Containment. In this function it is important that one of

the instruments, being high range, respond to a radiological release.

Indications from the devices are not used in a quantitative manner.

Rather they are used for qualitative purposes. Due to redundancy and

past test history, continued operability is expected. In addition, the

instruments serve no function in preventing accident initiation or

accident mitigation. Therefore, it is concluded that an extended

operating cycle for these monitors would not result in the possibility

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

3. There has been no significant reduction in the margin of safety.

Due to the qualitative function served by these two instruments as

well as their redundancy and acceptable past test history, no

significant reduction in the margin of safety due to an extended

operating cycle is expected.

(3) Safety Injection System Electrical Loading

The proposed change does not involve a significant hazard

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

The test procedure under consideration is one of the more

complicated surveillance procedures accomplished at a refueling

interval. Considering the vast number of components that are tested it

is highly improbable that some deficiencies will not occur. When such

problems are encountered it is important to note whether the failure is

time dependent and, in addition, whether the corrective maintenance

implemented prevents recurrences in the future. In consideration of the

evaluation of past test observations it is important to note that the

problems which occurred were not time dependent and that maintenance

practices have been effective in precluding future failures of the same

type. Equally important is whether the emergency power system would

have performed its intended safety function if the situation was not a

test but represented an actual demand upon the system. Test acceptance

criteria are always more stringent than required by accident scenarios

to provide margin. As discussed above the two most significant findings

were a failure of a CCW [Component Cooling Water] pump to strip from

the bus during the 1989 test and a relay which did not function within

its timing sequence. In the first instance, the diesel generator was

not overloaded. In the second instance, the relay did function albeit

not within the allotted time. In both cases, safety functions would

have been performed.

Thus, it is concluded that an extended period between surveillances

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

previously analyzed has not been created.

The deficiencies noted in the test data taken during the last

several refueling outage surveillances were not substantial in number

and would not have impacted the capability of the Safety Injection

System and its emergency power supply to perform its intended safety

function. The effectiveness of maintenance practices, both preventive

and corrective, has been proven in that deficiencies noted in one test

are not repeated in subsequent tests. The last refueling surveillance

test was completely successful where no new test failures were noted.

Because past test deficiencies do not appear to be time dependent,

extending the surveillance interval by 7.5 months is not expected to

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

accident previously created.

3. There has been no reduction in the margin of safety.

Because previous tests indicate that the engineered safety features

power supply would have performed its safety function if called upon

over the past several years, it is concluded that extending the

operating cycle by several months will not involve a significant

reduction in a margin of safety.

(4) Safety Injection System

The proposed change does not involve a significant hazard

consideration since:

1. There is no significant increase in the probability or

consequences of an accident.

The central safety objective in reactor design and operation is the

control of reactor fission products from the fuel. Four methods are

used to ensure this objective. Two of these methods are: (1) Retention

of fission products in the reactor coolant for whatever leakage occurs;

and (2) retention of fission products by the containment for

operational and accidental releases beyond the reactor coolant

boundary. The engineered safety features are the provisions in the

plant that embody these two methods to prevent the occurrence or to

ameliorate the effects of serious accidents.

The engineered safety features systems are the containment system,

the safety injection system, the containment spray system, the

containment air recirculation cooling and filtration system, the

isolation valve seal-water system, and the containment penetration and

weld channel pressurization system. Each engineered safety feature

provides sufficient performance capability to accommodate any single

failure of an active component and still function in a manner to avoid

undue risk to the health and safety of the public.

A comprehensive program of plant testing is formulated for all

equipment, systems, and system control vital to the functioning of

engineered safety features. The program consists, in part, of

integrated tests of the systems as a whole and periodic tests of the

actuation circuitry and mechanical components.

An assessment has been performed of the test results from the last

five refueling outages, covering a period in excess of seven years. In

reviewing the test results particular attention was directed towards

those test anomalies which directly impacted test acceptance criteria

and, thus, influence the capability of the safety injection system to

perform its intended safety function. Although in each test a problem

area was identified, the number of such events were minimal.

Furthermore, after corrective action these events were not repeated in

subsequent system tests. In all instances the problems were not

identified to be time dependent. Furthermore, the consequence from a

system safety function perspective was minimal. Thus, it is concluded

that extending the surveillance interval by several months will not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

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

previously analyzed has not been created.

The number of problem areas in each test have been few and of

minimal to nonexistent safety significance. In 1986, a valve failure

occurred which would have been detected by alternate means during an

extended operating cycle. In another instance, lack of valve movement

could not be repeated in a second test, leading to the conclusion that

the valve malfunction was not induced by the system but was the result

of the test process. In the last problem area, manual SI initiation, no

credit is taken within the FSAR [Final Safety Analysis Report] accident

analysis for this function. In 1989, a series of containment isolation

valves failed to stroke as required. In three instances the valves

failed closed, which is the correct position. In the other instances,

either the redundant valve did stroke to the correct position or the

valve was located in a closed system. In all these events there was

minimal impact upon safety. More importantly, after corrective action,

these failures were not repeated in the 1991 or 1993 tests. In 1991,

one breaker failed to perform within specifications and thus was

considered defective. In 1993 there were no major equipment

malfunctions, although one containment isolation valve failed to

perform as required.

In summary, although there have been anomalies in all of the tests

evaluated, none were deemed serious enough to impact the safety

function of the safety injection system or to be considered as having a

negative affect upon an increased interval of several months between

surveillances. Therefore, it has been concluded that an increased

operating cycle will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3. There has been no reduction in the margin of safety.

The results of the previous five cycles of test data have been

evaluated. None of the anomalies observed were sufficiently serious to

impact the performance of the Safety Injection System or to weigh

against an extended operating cycle. As there are no other changes to

the safety analysis parameters which are impacted by an extended

interval between surveillances, it is concluded that this change will

not involve a significant reduction in the margin of safety.

(5) Reactor Coolant System Sub-Cooling Margin Monitor:

The proposed change does not involve a significant hazards

consideration since:

1. There is no significant increase in the possibility or

consequences of an accident.

It is proposed that the channel calibration frequency for the

volume control tank instrumentation be changed from every 18 months

(+25%) to every 24 months (+25%).

The sub-cooling margin monitoring function is not relied upon

during normal operation. There is no reference to its use in the Indian

Point Unit 2 standard operating procedures. No credit is taken for this

monitoring function within the safety analysis for either the

prevention or mitigation of an accident. The increase in ``normal''

operating uncertainty, due to the longer operating cycle, as well as

``adverse'' uncertainties, is being incorporated in the EOPs [Emergency

Operating Procedures]. Therefore, the slight increase in uncertainty

associated with a longer operating cycle between surveillances will not

cause a significant increase in the probability or consequences of an

accident.

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

previously analyzed has not been created.

The sub-cooling margin serves no purpose during normal operation

for prevention of an accident. No credit is taken within the FSAR

Safety Analysis for accident mitigation. The sub-cooling margin monitor

is relied upon within the Emergency Operating Procedures. Thus, the

normal uncertainty due to a 30 month operating cycle, as supplemented

by the instrument loop error due to a post-accident harsh environment,

is being factored into the Emergency Operation Procedures in accordance

with Emergency Response Guidelines. Thus, it is concluded that the

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

analyzed has not been created.

3. There has been no reduction in the margin of safety.

Because the sub-cooling margin monitor serves no purpose during

normal operation and appropriate measures have been implemented to

reflect the additional uncertainty due to a 30 month operating cycle

into the EOPs, it is concluded that there will be no 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 request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Project Director: Michael L. Boyle, Acting.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: June 1, 1994.

Description of amendment request: The proposed amendment would

revise the Technical Specifications to allow extended Rod Position

Indication (RPI) deviation limits and on-line calibration of the RPI

channels. Specifically, Section 3.10.6.1 would be changed to allow

extended RPI deviation limits, and Section 3.10.4.4 would be changed to

allow on-line calibration of the RPI channels. The Basis for Section 10

would be changed to reflect the above, and in addition, Section

3.10.6.2 would be changed to clarify the operability requirements

during calibration. The proposed changes to Sections 3.10.6.1 and

3.10.4.4 include power limits to be included in the Core Operating

Limit Report (COLR). The use of a COLR for cycle specific core

operating limits was proposed by the licensee by submittal of October

29, 1993, and is currently under review by the NRC staff.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident previously

evaluated?

Response: Neither the probability nor the consequences of an

accident previously analyzed is increased due to the proposed changes.

All peaking factors will remain within the limits of the Technical

Specifications. Both the shutdown margin and the axial flux difference

will be maintained within the limits of the Technical Specifications.

There will be no fuel damage due to the changes. All design and safety

criteria will be met.

2. Does the proposed license amendment create the possibility of a

new or different kind of accident from any previously evaluated?

Response: The changes will not create the possibility of a new or

different kind of accident. The calibration will be performed using

plant procedures that have been reviewed and approved by Con Edison's

Safety Committees. It has been shown that even with the new RPI

deviation bands and on-line calibration, all power distribution limits

will be met.

3. Does the proposed amendment involve a significant reduction in

the margin of safety?

Response: The proposed amendment does not involve a significant

reduction in the margin of safety. There will be no change in the power

distribution limits used in the design and safety analyses and the

required shutdown margin will be maintained. It has been shown that

there is no fuel failure as a result of this change.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,

New York, New York 10003.

NRC Project Director: Michael L. Boyle.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Power Plant,

Unit 1, Monroe County, Michigan

Date of application for amendment: December 9, 1993 (Reference NRC-

93-0143)

Brief description of amendment: This Licensee Amendment Request

(LAR) proposes to revise the Enrico Fermi Power Plant, Unit 1,

Technical Specifications (TS) to bring the TS into conformance with a

revision of 10 CFR Part 20 (56 FR 23360).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

a. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes are administrative in nature and do not impact

the SAFSTOR status or design of any plant structures, systems or

components. As a result, this proposed change cannot increase the

probability or the consequences of any accident previously evaluated.

b. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes do not affect the plant SAFSTOR status as

defined. As a result, the proposed changes cannot create the

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

previously evaluated.

c. Does the change involve a significant reduction in a margin of

safety?

The proposed changes do not affect the plant SAFSTOR status. The

changes will not increase the amounts or change the types of effluents

that may be released offsite. These changes only ensure compliance with

revised 10 CFR 20. These changes do not alter any of the requirements

or responsibilities for protection of the public against radiation

hazards. As a result, these changes do not reduce 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 requests involve no significant hazards consideration.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Branch Chief: John H. Austin.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of amendment request: September 16, 1992, as superseded

February 4, 1994.

Description of amendment request: The proposed amendment would

supersede in its entirety a previous proposed amendment which was

submitted by letter dated September 16, 1992. A notice of application

and proposed no significant hazards consideration determination for the

September 16, 1992, submittal was published in the Federal Register on

January 21, 1993 (58 FR 5429); this notice supersedes the January 21,

1993, notice in its entirety.

The proposed amendment would modify the Technical Specifications

(TSs) related to containment air locks to make them as close to the

Improved Standard TSs in NUREG-1431 as the plant-specific design will

permit. The proposed changes in TS 3.6.1.1 and 3.6.1.3 would modify

surveillance requirements and limiting conditions of operation and

effect numerous administrative and format changes. The changes relate

to air lock operability, leak testing, and door interlocks.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The probability of occurrence of a previously evaluated accident is

not increased because the containment air locks do not effect the

initiation of any design basis accident [DBA]. The consequences of an

accident are also not significantly increased because the proposed

revisions to the action statements will continue to ensure that at

least one door in each air lock is maintained closed. A single door in

each air lock is capable of withstanding a pressure in excess of the

maximum expected pressure following a DBA. The structural integrity and

leak tightness of the containment will not be changed by this proposed

revision. For the brief period of time that the operable air lock door

is open and the inoperable door is providing the single containment

barrier, the consequences of [2an] accident may be increased. However,

the probability of an event occurring requiring containment integrity

is sufficiently remote to justify limited access when required.

Therefore, based on the continued ability of the containment air

locks to provide a barrier to limit leakage from containment during a

DBA, this proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

Air lock operation does not interface with the reactor coolant

pressure boundary or any other mechanical or electrical controls which

could impact the operations of the reactor or its direct support

systems.

Containment air locks are designed for the purpose of containment

entry and exit. During this operation, the air lock maintains

containment integrity by providing at least one door which is capable

of providing a leak tight barrier during a DBA.

The proposed changes will continue to ensure that air lock

operation is performed as assumed in the original design of the plant.

During the period when the operable door is open and the other door

inoperable, at least one door is being maintained closed as designed.

This condition is ensured due to the subatmospheric conditions that

exist during plant operation. The operable air lock door cannot be

safely opened unless the inoperable door is closed due to the

approximately 5 psi pressure differential that exists. The operable air

lock door would only be opened long enough to allow personnel to enter

the air lock.

Therefore, this proposed change does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. Does the change involve a significant reduction in a margin of

safety?

The applicable margin of safety consists of maintaining the primary

containment leak rates within the assumptions of the DBA analysis.

These leak rates are maintained provided at least one operable air lock

door remains closed during the event.

The proposed revisions will continue to ensure that at least one

air lock door is maintained closed. During the brief period of time

that an operable air lock door is open and the inoperable door is

providing the single containment barrier, the margin of safety is

decreased. The inoperable door may not limit containment leak rates

within the assumptions of the DBA analysis. However, the probability of

an event requiring the inoperable air lock door to limit containment

leakage occurring during this time period is sufficiently low and the

overall margin of safety would not be decreased by a significant

amount. The proposed increase in allowable door seal leakage will not

affect the overall ability of the containment air locks to restrict the

release of fission products to the environment. The overall air lock

leakage limit of less than or equal to .05 La remain unchanged.

The amount of leakage which the air lock(s) are permitted to contribute

to the combined containment leakage limit of 0.60 La remain

unchanged. Therefore, the margin of safety due to increasing the door

seal leakage limit remains unchanged.

Therefore, this proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: B.F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,

Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Walter R. Butler.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and

50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,

Georgia

Date of amendment request: March 31, 1994.

Description of amendment request: The proposed amendments would

change Technical Specification 3/4.7.1.1 and its Bases regarding

maximum allowed reactor thermal power operation with inoperable main

steam safety valves.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The new power range neutron flux high setpoint values will ensure that

the secondary side steam pressure will remain below 110 percent of its

design value following a loss of load/turbine trip (LOL/TT) when one or

more main steam safety valves (MSSVs) are declared inoperable.

Therefore, this transient will remain classified as a Condition II

probability event (faults of moderate frequency) per ANSI--N18.2, 1973

as discussed in Section 15.0.1 of the VEGP Final Safety Analysis Report

(FSAR). Accordingly, since the new power range setpoints will maintain

the capability of the MSSVs to perform their pressure relief function

associated with a LOL/TT event, there will be no effect on the

probability or consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated. The

proposed changes do not involve any change to the configuration or

method of operation of any plant equipment, and no new failure modes

have been defined for any plant system or component. The new power

range neutron flux high setpoints will maintain the capability of the

MSSVs to perform their pressure relief function to ensure the secondary

side steam design pressure is not exceeded following a LOL/TT.

Therefore, since the function of the MSSVs is unaffected by the

proposed changes, the possibility of a new or different kind of

accident from any accident previously evaluated is not created.

3. The proposed changes do not involve a significant reduction in a

margin of safety. The algorithm methodology used to calculate the new

power range neutron flux high setpoints is conservative and bounding

since it is based on a number of inoperable MSSVs per loop; i.e., if

only one MSSV in one loop is out of service, the applicable power range

setpoint would be the same as if one MSSV in each loop were out of

service. Another conservatism with the algorithm methodology is with

the assumed minimum total steam flow rate capability of the operable

MSSVs. The assumption is that if one or more MSSVs are inoperable per

loop, the inoperable MSSVs are the largest capacity MSSVs, regardless

of which capacity MSSVs are actually inoperable. Therefore, since the

power range setpoints calculated for the proposed changes using the

algorithm methodology are more conservative and ensure the secondary

side steam design pressure is not exceeded following a LOL/TT, there

will not be 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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

NRC Project Director: David B. Matthews.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and

50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,

Georgia

Date of amendment request: May 20, 1994.

Description of amendment request: The proposed amendments would

relocate the heat flux hot channel factor penalty of 2 percent in

Specification 4.2.2.2.f to the cycle-specific Core Operating Limits

Report to allow burnup-dependent values of the penalty in excess of 2

percent. The licensee also proposes to revise the reference in

Specification 6.8.1.6 to the Westinghouse FQ(Z) surveillance

methodology in order to reflect Revision 1 of WCAP-10216-P,

``Relaxation of Constant Axial Offset Control--FQ Surveillance

Technical Specification,'' approved by the NRC on November 26, 1993.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed 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 [Technical Specification]). 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. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

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. Does this change 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 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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia 30308.

NRC Project Director: David B. Matthews.

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: June 22, 1994.

Description of amendment request: The proposed amendment changes

Technical Specification Sections 1.6, 3.2.A, 3.9.F.5, and 4.2.A which

specify the Shutdown Margin (SDM) requirements that ensure the reactor

can be made subcritical and can be maintained sufficiently subcritical

to preclude inadvertent criticality in any core condition. The

amendment also proposes a new definition, Shutdown Margin, Section

1.45. The proposed changes address the requirements for SDM

demonstration and provide clarification for actions if SDM is not met.

The amendment also proposes administrative changes to Sections 1.7

and 3.2.B.2 (b). The definition, COLD SHUTDOWN CONDITION, was

simplified by stating the reactor is in the SHUTDOWN CONDITION which

eliminates the need of repeating the requirements for this condition.

The note which permitted unlimited reactor startups without the Rod

Worth Minimizer during Cycle 11 is no longer applicable. The note and

its reference are deleted from the new page 3.2-2. Starting with page

3.2-2 in Section 3.2, the pages were renumbered and repaginated to

accommodate the changes in text.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

GPU Nuclear has determined that operation of the Oyster Creek

Nuclear Generating Station in accordance with the proposed Technical

Specifications does not involve a significant hazard. The changes do

not:

1. Involve a significant increase in the probability or the

consequence of an accident previously evaluated.

The proposed SDM Limits are more restrictive and provide adequate

shutdown margin for various modes of reactor operation. Since the new

SDM limits do not modify any initial conditions for the accidents

previously evaluated in the SAR [Safety Analysis Report], the proposed

changes do not involve a significant increase in the probability or

consequences of these accidents.

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

from any previously evaluated.

The proposed TS changes do not modify the function of any

structure, system or component. The new Shutdown Margin requirements

will still meet the basic criterion that the core in its maximum

reactivity condition be subcritical with the control rod of highest

worth fully withdrawn and all operable rods fully inserted. Based on

these facts, the proposed TS changes do not create the possibility of a

new or different kind of accident from any previously evaluated.

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

The proposed changes do not reduce the margin of safety, because

the new SDM limits where the highest worth control rod is determined

analytically (0.38% delta k) or by measurement (0.28% delta k) are more

restrictive than the current Oyster Creek limit (0.25% delta k).

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, New Jersey

08753.

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

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

NRC Project Director: John F. Stolz.

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas, Docket

Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: June 6, 1994.

Description of amendment request: The licensee proposes to modify

the South Texas Project, Units 1 and 2, Technical Specification 3/

4.8.1.1, ``A.C. Sources,'' to revise the action statements and

surveillance requirements for testing of the standby diesel generators

(SDGs). The proposed amendment would eliminate excessive and

unnecessary testing of the SDGs consistent with the guidance provided

in NUREG-1366, ``Improvements to Technical Specifications Surveillance

Requirements,'' NUREG-1431, ``Standard Technical Specifications for

Westinghouse Plants,'' Generic Letter 84-15, ``Proposed Staff Actions

to Improve and Maintain Diesel Generator Reliability,'' and Generic

Letter 93-05, ``Line-Item Technical Specifications Improvements to

Reduce Surveillance Requirements for Testing During Power Operation.''

This request replaces a request for amendment dated November 23, 1993,

which was noticed on January 5, 1994 (59 FR 621). This revised

amendment request includes elimination of additional identified

unnecessary testing discovered since the original submittal. The

changes include: (1) eliminating the requirement to demonstrate the

operability of an operable SDG whenever an offsite AC power source is

determined to be inoperable, or whenever a support system or an

independently testable component of another SDG is inoperable, (2)

eliminating the requirement to load the diesel in 10 minutes during

testing, (3) replacing the minimum required loading for testing with a

load band, (4) relocating some surveillance requirements to the Diesel

Fuel Oil Testing Program, and (5) eliminating unnecessary loss of

offsite power tests.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The Standby Diesel Generators do not initiate any accidents,

therefore these changes do not increase the probability or [of] an

accident previously evaluated. The proposed changes will permit the

elimination of the unnecessary mechanical stress and wear on the diesel

engine and generator while ensuring that the diesel generators will

perform their designed function. The elimination of this mechanical

stress and wear will improve the reliability and availability of the

Standby Diesel Generators which will have a positive effect on the

ability of the diesel generators to perform their design function.

Therefore, the consequences of an accident previously evaluated are not

increased. The proposed changes are consistent with NUREG-1366, NUREG-

1431, Generic Letter 93-05, and Generic Letter 84-15.

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

different kind of accident from any previously evaluated.

The elimination of these unnecessary tests does not affect the

design bases of the SDGs, or any of the accident evaluations involving

the SDGs. The SDGs are designed to provide electrical power to the

equipment important for safety during all modes and plant conditions

following a loss of offsite power. The test schedule established in

accordance with GL 84-15 assures that operable SDGs are capable of

performing their intended safety function. The proposed changes to the

surveillance requirements are consistent with NUREG-1431, NUREG-1366,

Generic Letter 93-05, industry operating experience, and South Texas

Project operating experience. These changes are intended to improve

plant safety, decrease equipment degradation, and remove unnecessary

burden on personnel resources by reducing the amount of testing that

the Technical Specification requires during power operation. Relocating

the diesel fuel oil testing requirement to the STP Fuel Oil Monitoring

Program outside of the Technical Specifications is an administrative

change consistent with NUREG-1431 and consequently has no effect on

accident probability, consequences, or margin. Therefore, this change

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

from any previously evaluated.

3. The proposed change does not involve a significant reduction in

a margin of safety.

The proposed changes extend testing frequency and eliminate

unnecessary mechanical stress and wear on the diesel generator in an

effort to improve plant reliability and safety. These changes are

consistent with NUREG-1431, NUREG-1366, industry operating experience,

and STP operating experience and do not adversely affect the design

bases, accident analysis, reliability or capability of the SDGs to

perform their intended safety function. Relocating the diesel fuel oil

testing requirements to the STP Fuel Oil Monitoring Program outside of

the Technical Specifications is an administrative change consistent

with NUREG-1431 and consequently has no effect on accident probability,

consequences, or margin. Therefore the proposed changes do not involve

any reduction in a margin to safety.

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

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

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

request for amendments involve no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas

77488.

Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,

P.C., 1615 L Street, N.W., Washington, D.C. 20036.

NRC Project Director: William D. Beckner.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: June 18, 1993 as supplemented on

December 17, 1993 and May 5, 1994.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) by clarifying TS wording for

the Low Pressure Coolant Injection (LPCI) and Containment Spray modes

of the Residual Heat Removal (RHR) system to assure consistency with

requirements of the DAEC Updated Final Safety Analysis Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The probability or consequences of a previously-analyzed

accident will not be increased by these proposed changes to the LPCI

and Containment Spray LCOs and BASES because they merely clarify

existing TS requirements and are consistent with the DAEC UFSAR

accident analysis. The addition of the footnote clarifying LPCI

OPERABILITY during RHR system operation in the Shutdown Cooling mode is

consistent with the requirements in the NRC Standard TS (NUREG-1433).

No changes in either system design or operating strategies will be made

as a result of these changes, thus no opportunity exists to increase

the probability or consequences of a previously-analyzed accident.

(2) The possibility of a new or different kind of accident from

those previously analyzed will not be created by these changes to the

LPCI and Containment Spray LCOs and BASES because they merely clarify

existing requirements. The addition of the footnote clarifying LPCI

OPERABILITY during RHR system operation in the Shutdown Cooling mode is

consistent with the requirements in the NRC Standard TS (NUREG-1433).

No changes in either system design or operating strategies will be made

as a result of these changes, thus no possibility exists to introduce a

new or different kind of accident.

(3) The margin of safety will not be decreased as a result of these

changes because they merely clarify existing TS requirements and are

consistent with the UFSAR accident analysis. The addition of the

footnote clarifying LPCI OPERABILITY during RHR system operation in the

Shutdown Cooling mode is consistent with the requirements in the NRC

Standard TS (NUREG-1433). No changes in either system design or

operating strategies will be made as a result of these changes, thus no

possibility exists to reduce a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401.

Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,

Esquire, Newman and Holtzinger, 1615 L Street NW., Washington, DC

20036.

NRC Project Director: John N. Hannon.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: July 1, 1994.

Description of amendment request: The proposed amendment would

revise the drawdown time testing requirement of Technical Specification

(TS) 4.6.5.1.c.1 and the secondary containment inleakage testing

requirement of TS 4.6.5.1.c.2. These revisions would support a revised

design basis radiological analysis which would support an increase in

secondary containment drawdown time from 6 to 60 minutes by taking

credit for fission product scrubbing and retention in the suppression

pool. The current design basis radiological analysis does not take

credit for the pressure suppression pool as a fission product cleanup

system as permitted in NUREG-0800, Section 6.5.5, ``Pressure

Suppression Pool as a Fission Product Cleanup System.'' The proposed

amendment would also take credit for additional mixing of primary

containment and engineered safety feature systems leakage with 50

percent of the secondary containment free air volume prior to the

release of radioactivity to the environment. In the revised analysis,

mixing is assumed to occur at the onset of a Design Basis Loss-of-

Coolant Accident as the primary containment and the engineered safety

feature systems leak into secondary containment. The current analysis

takes credit for mixing within secondary containment only after

achieving a -0.25 inch water gauge (WG) pressure in secondary

containment with respect to the outside surrounding atmosphere. The

licensee's radiological evaluation for this accident, which reflects

these proposed changes and an assumed drawdown time of 60 minutes, has

determined that the radiological doses remain below 10 CFR Part 100

guidelines values and General Design Criterion 19 criteria. The revised

radiological doses, as calculated by the licensee, are lower than the

doses currently presented in the Updated Safety Analysis Report (USAR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of the Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Secondary containment and the SGTS [Standby Gas Treatment System]

are not initiators or precursors to an accident. Secondary containment

provides a pressure boundary, with limited inleakage, for the purpose

of establishing a negative pressure to prevent a ground level

unfiltered release of radioactivity. SGTS responds to accidents

involving a release of radioactivity to the secondary containment by

establishing and maintaining a negative pressure inside secondary

containment and by providing an elevated filtered release. Therefore,

changes to SECONDARY CONTAINMENT INTEGRITY surveillances cannot affect

the probability of a previously evaluated accident.

The suppression pool and secondary containment are largely passive

in nature, and the active components are suitably redundant. Therefore,

their fission product attenuation functions can be accomplished

assuming a single failure.

Currently, using an assumed drawdown time of 6 minutes, the

radiological doses for a DBA-LOCA [Design Bases Accident--Loss-of-

Coolant Accident] are below the guidelines of 10 CFR Part 100 and GDC

[General Design Criterion] 19 criteria. The calculated doses,

considering the pressure suppression pool as a fission product cleanup

system, additional mixing within secondary containment and an assumed

secondary containment drawdown time of 60 minutes, are lower than the

previously calculated doses. The new doses are below 10 CFR Part 100

guideline values and GDC 19 criteria. The revised radiological analysis

follows the source term assumptions of RG [Regulatory Guide] 1.3, with

the exception of regulatory position C.1.f as permitted by SRP

[Standard Review Plan] Section 6.5.5, and continues to provide a

conservative representation of the timing and the composition of the

release of radioactivity from secondary containment during a DBA-LOCA.

The Technical Specification SRs [Surveillance Requirements] will

ensure a continued state of readiness for the SGTS, the secondary

containment, the suppression pool and the suppression chamber/drywell

vacuum breakers. Therefore, the assumptions used in the dose assessment

will continue to bound the actual bypass of the suppression pool and

the mixing in secondary containment during a DBA-LOCA. The proposed

changes to the surveillances provide assurance that the performance of

the SGTS and secondary containment supports the radiological analysis.

Accordingly, as shown in Table 1, page 14 of 20, [of the July 1, 1994,

amendment request] operation with the SGTS and the proposed change to

the surveillances for secondary containment will not significantly

increase the consequences of an accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed change to the surveillances ensures that the SGTS and

secondary containment will be available to respond to an accident such

that the guidelines of 10 CFR Part 100 and the limits of GDC 19 are not

exceeded. The proposed change to the surveillances reflect

consideration of the pressure suppression pool as a fission product

cleanup system and credit for additional mixing in secondary

containment. The suppression pool will continue to perform its safety

functions as a pressure suppression pool and as a source of water to

support emergency core cooling system operation during a DBA-LOCA. In

addition, secondary containment will continue to perform its safety

function of controlling and minimizing radioactive leakage to the

outside atmosphere during a DBA-LOCA. Safety related equipment will

continue to be OPERABLE in the radioactive environment of secondary

containment to mitigate the consequences of a DBA-LOCA. Accordingly,

the proposed Technical Specification change will not create the

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

previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with

proposed amendment, will not involve a significant reduction in a

margin of safety.

The SGTS exhausts the secondary containment atmosphere to the

environment through the filtration system. To verify the SGTS has not

degraded, SR 4.6.5.1.c.1 verifies that each SGTS subsystem will

establish and maintain a pressure in the secondary containment that is

equal to or more negative than -0.25 inch WG within the required time

limit. To verify secondary containment is intact, SR 4.6.5.1.c.2

demonstrates that one SGTS subsystem can maintain a pressure which is

equal to or more negative than -0.25 inch WG for 1 hour at a flow rate

less than or equal to the maximum allowed inleakage. The 1 hour test

period allows secondary containment to be in thermal equilibrium at

steady state conditions. Furthermore, as an interim measure, NMPC

[Niagara Mohawk Power Corporation] implemented certain compensatory

measures through administrative controls to ensure that the

radiological consequences of a DBA-LOCA would remain within regulatory

criteria. Together, these tests and the compensatory measures assure

SGTS performance and secondary containment boundary integrity.

The proposed change to these surveillances incorporate changes to

the design basis, i.e., credit for fission product scrubbing and

retention by the suppression pool and credit for additional mixing

within secondary containment. The new inleakage is 2670 cfm which is

loss [less] than one change of the secondary containment free air

volume per day. The new drawdown time limit reflects consideration of

the proposed change in the secondary containment inleakage limit. Due

to the effects of service water temperature, inside and outside

temperature, flow measurement inaccuracies and actual test pressures,

meeting the current SRs does not by itself assure adequate SGTS

performance. Therefore, the surveillances' results are adjusted to

account for actual test conditions. Compliance with the proposed

surveillances assures that the SGTS can achieve and maintain -0.25 inch

WG in less than 60 minutes following a postulated DBA-LOCA. Achieving

-0.25 inch WG within 60 minutes assures that radiological doses will

remain below regulatory limits (see Table 1 [of the July 1, 1994,

amendment request]). Therefore, the proposed surveillances, together

with the proposed adjustments, provide adequate assurance of SGTS

performance and secondary containment boundary integrity. Accordingly,

the proposed Technical Specification change will not involve a

significant reduction in a margin of safety.

Therefore, as determined by the analysis above, this proposed

amendment involves no significant hazards consideration.

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

this review, it appears that the three standards of 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: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

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

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

NRC Project Director: Michael L. Boyle.

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

Generating Plant, Wright County, Minnesota

Date of amendment request: June 8, 1994.

Description of amendment request: The proposed amendment would

revise sections 3.7/4.7, which pertain to the Standby Gas Treatment

System (SGTS) and Secondary Containment. The proposed amendment would

revise the surveillance requirements for both SGTS and secondary

containment and revise the performance requirements for the SGTS

filters and process stream electric heaters.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The function of the SGTS and secondary containment is to mitigate

the consequences of a loss of coolant accident and fuel handling

accidents. The proposed changes maintain or improve this capability.

Therefore, this amendment will not cause a significant increase in the

probability or consequences of an accident previously evaluated for the

Monticello plant.

2. The proposed amendment will not create the possibility of a new

or different kind of accident from any accident previously analyzed.

The proposed changes to Technical Specifications for the standby

gas treatment system and secondary containment do not alter the

function of the systems or its interrelationships with other systems.

The proposed changes provide requirements to ensure the systems are

capable of performing the required functions or that actions are taken

to minimize the potential for its function being required consistent

with regulatory guidance; therefore, this amendment will not create the

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

previously analyzed.

3. The proposed amendment will not involve a significant reduction

in the margin of safety.

Improvements in the margin of safety are provided via the permanent

elimination of a potential single failure which could adversely affect

both standby gas treatment systems by deleting the reference to the

standby gas system room heaters in the technical specification bases

and providing appropriate surveillance requirements to assure system

operability. A review of the performance history of the Standby Gas

Treatment System and licensing basis assumptions has determined that

the proposed changes do not adversely affect plant safety. Changes to

the SGTS performance requirements provide greater assurance of SGTS

operability. The proposed change for the completion time to place the

plant in a cold shutdown condition if limiting conditions for operation

can not be satisfied is consistent with the time frame specified in the

current specification and is consistent with Standard Technical

Specifications. The proposed amendment will 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: 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: Ledyard B. Marsh.

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

Unit No. 1, Washington County, Nebraska

Date of amendment request: February 12, 1993, as supplemented by

letters dated August 20, 1993 and June 6, 1994.

Description of amendment request: This amendment request was

previously noticed in the Federal Register on April 14, 1993 (58 FR

19485). The June 6, 1994, submittal supplements the February 12, 1993,

application for amendment, and includes and incorporates the NRC staff

comments. The proposed amendment would modify the Technical

Specifications (TS) to implement the reactor coolant system (RCS) leak

before break (LBB) methodology detection criteria, in accordance with

the recommendations listed in Generic Letter (GL) 84-04.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes will require additional leak detection

instruments be operable to close Unresolved Safety Issue A-2,

``Asymmetrical Blowdown Loads on Reactor Primary Coolant System,'' for

the Fort Calhoun Station. Requiring additional instruments to be

operable does not increase the probability or consequences of an

accident since the safety function of the instruments is not being

altered.

The proposed changes require at least two different types of RCS

leak detection instruments, of diverse monitoring principles, be

operable or corrective actions be taken to restore the instrumentation

to operable status. Currently the Technical Specifications require only

one RCS leak detection instrument to be operable.

The probability of leaks occurring due to thermal or normal fatigue

is not affected as indicated in the fracture mechanics analysis

referenced in Generic Letter 84-04. No changes are proposed to primary

RCS piping systems or supports as a result of the proposed revision.

The proposed changes will ensure that a potential significant failure

does not go undetected within the Regulatory Guide 1.45 criteria as

noted in Generic Letter 84-04.

The Loss of Coolant Accident (LOCA) analysis will not be impacted

by the proposed change. The results of the current Fort Calhoun LOCA

analyses cited in Section 14.15 of the Updated Safety Analysis Report

(USAR) will not be impacted as a result of these changes.

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

from any previously analyzed.

It has been determined that a new or different kind of accident

will not be created due to the proposed changes since no new or

different modes of operation are created by this change. The existing

operating procedures were established to support an enhanced RCS leak

detection program. Operation of RCS leak detection instruments will not

differ from existing conditions.

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

The margin of safety as defined in the basis for the Technical

Specifications is not changed or reduced by this proposed change.

Defining adequate RCS LBB monitoring is required to meet

recommendations provided in Generic Letter 84-04.

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: LeBoeuf, Lamb, Leiby, and MacRae, 1875

Connecticut Avenue NW., Washington, DC 20009-5728.

NRC Project Director: William D. Beckner.

Philadelphia Electric Company, Docket No. 50-352, Limerick Generating

Station, Unit 1, Montgomery County, Pennsylvania

Date of amendment request: June 6, 1994.

Description of amendment request: The amendment would remove the

controls for a remote shutdown system control valve and delete the

isolation signal for certain primary containment isolation valves from

TS Tables 3.3.7.4-1 and 3.6.3-1 respectively, as a result of

eliminating the steam condensing mode of the Redidual Heat Removal

system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The RHR system steam condensing mode is a non-safety related

function of the RHR system and has been eliminated at Limerick

Generating Station, Unit 1. These proposed changes will not affect any

components required to perform the safety-related function of the RHR

system.

The ability of the RHR system to respond to an accident will not be

degraded by the proposed changes. Valve HV-51-1F011A is locked closed

with the electrical power removed. The valve's handswitch which is part

of the remote shutdown panel (RSP) controls, does not perform any

function and will be physically removed from the RSP. The deletion of

the isolation signal for valves HV-C-51-1F103A and HV-C-51-1F104B will

not affect the ability of these valves to function as primary

containment isolation valves (PCIVs), since they are locked closed

already in their safety-related position, providing containment

isolation as manual PCIVs. Therefore, the proposed TS changes do not

involve an increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously evaluated.

No new failure modes of RHR system are created by the proposed TS

changes. All valves associated with the proposed changes are dedicated

specifically for the RHR system steam condensing mode, and will not

impact the operation of any components or piping required for other

modes of operation of the RHR system. These valves are locked-closed in

their safety-related position with the electrical power removed.

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

new or different kind of accident from any previously evaluated.

3. The proposed TS changes do not involve a significant reduction

in a margin of safety.

The steam condensing mode is a non-safety related function of the

RHR system and, therefore, is not addressed in the TS. The controls for

remote shutdown system control valve HV-51-1F011A are not being used.

Presently, the valve is locked closed with the electrical power removed

and the valve's handswitch will be removed from the RSP, since it does

not perform any function. The proposed changes will not impact the safe

operation of LGS Unit 1. The deletion of the isolation signal for

valves HV-C-51-1F103A and HV-C-51-1F104B will not affect the ability of

these valves to function as primary containment isolation valves

(PCIVs), since they are locked closed already in their safety-related

position. Therefore, the proposed TS changes do not involve a reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: Charles L. Miller.

Philadelphia Electric Company, Docket No. 50-352, Limerick Generating

Station, Unit 1, Montgomery County, Pennsylvania

Date of amendment request: June 10, 1994.

Description of amendment request: The amendment involves a one-time

change affecting the Allowed Outage Time (AOT) for the Emergency

Service Water (ESW) System, Residual Heat Removal Service Water (RHRSW)

System, Suppression Pool Cooling, Suppression Pool Spray, and Low

Pressure Coolant Injection modes of the Residual Heat Removal System,

and Core Spray System to be extended from 3 and 7 days to 14 days

during the Limerick Generating Station (LGS), Unit 2 third refueling

outage scheduled to begin in January 1995. This proposed extended AOT

would allow adequate time to install isolation valves and cross-ties on

the ESW and RHRSW Systems to facilitate future inspections or

maintenance.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications changes do not involve a

significant increase in the probability or consequences of an accident

previously evaluated.

The proposed one-time TS changes will not increase the probability

of an accident since it will only extend the time period that the `B'

ESW and RHRSW loops and the affected equipment can be out-of-service.

The extension of the time duration that certain equipment is out-of-

service has no direct physical impact on the plant. The proposed

inoperable systems are normally in a standby mode while the unit is in

OPCON 1 or 2 and are not directly supporting plant operation.

Therefore, they can have no impact on the plant that would make an

accident more likely to occur due to their inoperability.

During transients or events which require these systems to be

operating, there is sufficient capacity in the operable loops to

support plant operation or shutdown, in-so-much that failures that are

accident initiators will not occur more frequently than previously

postulated.

In addition, the consequences of an accident previously evaluated

in the SAR will not be increased. With the `B' loops of ESW and RHRSW

inoperable, a known quantity of equipment is either inoperable or the

equipment is not fully capable of fulfilling its design function under

all design conditions due to certain support systems not being

operable. Based on the support functions of the ESW and RHRSW systems,

a review of the plant was performed to determine the impacts that the

inoperable ESW and RHRSW `B' loops would have on other systems. The

impacts were identified for each system, as discussed in the preceding

Safety Assessment, and it was determined whether there were any adverse

[effects] on the systems. It was then determined how the adverse

[effects] would impact each system's design basis and overall plant

safety. The consequences of any postulated accidents occurring on Unit

1 during this AOT extension was found to be bounded by the previous

analyses as described in the SAR.

The existing AOTs limit the amount of time that the plant can

operate with certain equipment inoperable, where single failure

criteria is still met. The minimum equipment required to mitigate the

consequences of an accident and/or safely shutdown the plant will be

operable or the plant will be shutdown. Therefore, by extending certain

AOTs and extending the assumptions concerning the combinations of

events and single failures for the longer duration of each extended

AOT, we conclude, based on the evaluations above, that at least the

minimum equipment required to mitigate the consequences of an accident

and/or safely shutdown the plant will still be operable during the

extended AOT. Therefore, the consequences of an accident previously

evaluated in the SAR will not be increased.

Therefore, these proposed one-time TS changes will not result in a

significant increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The proposed one-time TS changes will not create the possibility of

a different type of accident since it will only extend the time period

that the `B' ESW and RHRSW loops and the affected equipment can be out-

of-service. The extension of the time duration that certain equipment

is out-of-service has no direct physical impact on the plant and does

not create any new accident initiators. The systems involved are either

accident mitigation systems, safe shutdown systems or systems that

support plant operation. All of the possible impacts that the

inoperable equipment may have on its supported systems were previously

analyzed in the SAR and are the basis for the present TS ACTION

statements and AOTs. The impact of inoperable support systems for a

given time duration was previously evaluated and any accident

initiators created by the inoperable systems was evaluated. The

lengthening of the time duration does not create any additional

accident initiators for the plant.

Therefore, the proposed one-time TS changes will not create the

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

previously evaluated.

3. The proposed TS changes do not involve a significant reduction

in a margin of safety.

The ESW and RHRSW systems and their supported systems are designed

with sufficient independence and redundancy such that the removal from

service of a component/subsystem will not prevent the systems from

performing their required safety functions. Since removal of an ESW and

a RHRSW loop from service with one unit in operation and the other unit

in a refueling outage is allowed by the current Technical

Specifications, then the concern is the reduced margin of safety

incurred by extending the affected AOTs.

The present ESW and RHRSW AOT limits were set to ensure that

sufficient safety-related equipment is available for response to all

accident conditions and that sufficient decay heat removal capability

is available for a LOCA/LOOP on one unit and simultaneous safe shutdown

of the other unit. A slight reduction in the margin of safety is

incurred during the proposed extended AOT due to the increased risk

that an event could occur in a fourteen day period versus a three or

seven day period. This increased risk is judged to be minimal due to

the low probability of an event occurring during the extended AOT and

based on the following discussion of minimum ECCS/decay heat removal

requirements.

The reduction in the margin of safety is not significant since the

remaining operable ECCS equipment is adequate to mitigate the

consequences of any accident. This conclusion is based on the

information contained in documents NEDO-24708A and NEDC-30936-A. These

documents described the minimum requirements to successfully terminate

a transient or LOCA initiating event (with scram), assuming multiple

failures with realistic conditions and were used to justify certain TS

AOTs per UFSAR sections 6.3.1.1.2.o and 6.3.3.1. The minimum

requirements for short term response to an accident would be either one

LPCI pump or one Core Spray loop in conjunction with ADS, which would

be adequate to re-flood the vessel and maintain core cooling sufficient

to preclude fuel damage. For long term response, the minimum

requirements would be one loop of RHR for decay heat removal, along

with another low pressure ECCS loop. These minimum requirements will be

met since implementation of the proposed TS changes will require the

operability of HPCI, ADS, two LPCI subsystems (or one LPCI subsystem

and one RHR subsystem during decay heat removal) and one Core Spray

subsystem be maintained during the 14 day period.

In addition, measures will be taken prior to or during the proposed

extended AOT for those fire regions that rely on one or more safe

shutdown methods which would all be unable to safely shutdown the plant

with inoperable loops of the ESW and RHRSW systems or the inoperable

systems that ESW or RHRSW support. These measures will offset the

increased risk of a fire event occurring in the vulnerable areas,

during the fourteen day versus three day AOT period. Therefore, the

proposed extended AOT does not adversely affect the approved level of

fire protection as described in UFSAR Appendix 9A (Fire Protection

Evaluation Report).

A special procedure will be written to administratively control the

requirement to maintain the operability of specified components and

implementation of any appropriate compensatory measures which are

deemed necessary during the proposed AOT. In addition, operations

personnel are fully qualified by normal periodic training to respond to

and mitigate a Design Basis Accident, including the actions needed to

ensure decay heat removal while LGS Unit 1 and Unit 2 are in the

operational configurations described within this submittal.

Accordingly, procedures are already in place that cover safe plant

shutdown and decay heat removal for situations applicable to those in

the proposed AOTs.

A Probabilistic Safety Assessment (PSA) Study was performed for an

ESW and RHRSW loop being out-of-service for 14 days on an operating

unit. This analysis includes EDG D12 being aligned to `A' ESW and HPCI

and RCIC not requiring room cooling. No other deviations from the

bounding assumptions used in the base PSA model were made. The Core

Damage Frequency (CDF) increased by 2.7x10-6, from 5.11x10-6/

reactor-year to 7.8x10-6 /reactor-year. In absolute terms, this is

not a significant increase in risk. In addition, the modifications to

be installed during this proposed extended AOT will allow for future

maintenance and inspections to be performed on the ESW and RHRSW loops

without removing an entire loop from service, which will reduce risk in

the future. For example, if the ESW loop unavailability, due to testing

or maintenance, is reduced by half, the CDF will decrease by more than

four percent. It will also minimize the potential need for future AOT

extensions on these systems.

Therefore, the implementation of the proposed one-time TS changes

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

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: Charles L. Miller.

Philadelphia Electric Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric Company,

Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station,

Units Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: May 10, 1994.

Description of amendment request: The licensee is proposing to

revise the Technical Specifications (TS) requirements governing the

minimum low pressure cooling availability when irradiated fuel is in

the reactor vessel and the reactor is in the cold condition.

Specifically, the proposed changes are: (1) Revise the TS section

titles in the Table of Contents to agree with the TS section titles in

the body of the TS. (2) Revise TS Section 3.5.A.1 to provide proper

reference to the revised TS Section 3.5.F. (3) Revise TS Section

3.5.A.3 to provide proper reference to the revised TS Section 3.5.F.

(4) Revise TS Section 3.5.B.1 to delete the reference to TS Section

3.5.F.3. (5) Revise TS Section 3.5.F to require the limiting condition

for operation (LCO) governing minimum low pressure cooling availability

when irradiated fuel is in the reactor vessel and the reactor is in the

cold condition to be identical with the corresponding LCO in NUREG-

1433, ``Standard Technical Specifications General Electric Plants, BWR/

4.'' (6) Revise TS Section 4.5.F to require the surveillance

requirements (SR) governing minimum low pressure cooling availability

when irradiated fuel is in the reactor vessel and the reactor is in the

cold condition to be identical with the corresponding SR in NUREG-1433.

(7) Revise TS BASES 3.5.A to delete the reference to the core spray

subsystem as also providing a source for flooding of the core in case

of accidental draining because the information is being added to TS

BASES 3.5.F. (8) Revise TS BASES 3.5.F to be consistent with the

corresponding TS BASES in NUREG-1433. (9) Revise TS BASES 4.5 to be

consistent with the corresponding TS BASES in NUREG-1433. (10) Revise

TS Section 3.7.A.1 to provide proper reference to the revised TS

Section 3.5.F.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

Proposed changes 1, 2, 3, 4, 7, 8, 9 and 10 are administrative in

nature and involve no technical changes to the TS. These proposed

changes do not impact initiators of analyzed events or the assumed

mitigation of accidents or transients events. Therefore, these changes

do not involve an increase in the probability or consequences of an

accident previously evaluated.

Proposed changes 5 and 6 will not increase the probability of

initiating an analyzed event or alter assumptions relative to

mitigation of an accident or transient event. These changes will not

alter the operation of process variables or systems, structures, or

components (SSC) as described in the safety analyses. These changes do

not involve any physical changes to plant SSC. TS requirements that

govern Operability or routine testing and verification of plant

components and variables are not assumed to be initiators of any

analyzed event. The proposed changes will not alter the operation of

equipment assumed to be available for the mitigation of accidents or

transients by the plant safety analysis or licensing basis. The

proposed changes establish or maintain adequate assurance that

components are operable when necessary for the prevention or mitigation

of accidents or transients and that plant variables are maintained

within limits necessary to satisfy the assumptions for initial

conditions in the safety analysis. These changes have been confirmed to

ensure no previously evaluated accident has been adversely affected.

These changes will not allow continuous plant operation with plant

conditions during a unit outage such that a single failure will result

in a loss of any safety function. Therefore, the changes will not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

The proposed changes do not involve a physical alteration of the

plant (no new or different type of equipment will be installed or

removed) and will not alter the method used by any system to perform

its design function. The proposed changes do not allow plant operation

in any mode that is not already evaluated. Therefore, these changes

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

from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction in a

margin of safety.

Proposed changes 1, 2, 3, 4, 7, 8, 9 and 10 are administrative in

nature and will not involve any technical changes. These proposed

changes will not reduce a margin of safety because they have no impact

on any safety analysis assumptions. Because these changes are

administrative in nature, no question of safety is involved. Therefore,

these changes do not reduce the margin of safety.

Proposed changes 5 and 6 add some new requirements and make some

existing requirements more restrictive. These changes will not impact

any safety analysis assumptions. Adding new requirements and making

existing ones more restrictive either increases or does not affect the

margin of safety. As such, no question of safety is involved.

Therefore, these changes will not involve a significant reduction in a

margin of safety.

Proposed changes 5 and 6 also make three less restrictive changes.

The first change deletes the requirements for the containment cooling

system when the reactor is in the Cold Condition. The containment

cooling system is necessary to maintain primary containment Operable to

mitigate the release of radioactive material following a DBA [design

basis accident].

However, primary containment is not required to be Operable with

the reactor in the Cold Condition. As a result, the containment cooling

system is not needed to maintain the primary containment Operable with

the reactor in the Cold Condition. This change does not affect any

safety limits, operating limits, or design assumptions. This [change]

provides the benefit of allowing maintenance to be performed on the

containment cooling systems during a unit outage to ensure their

reliability during power operation. Therefore, this change does not

involve a significant reduction in a margin of safety.

The current TS requirement that both core spray systems and the

LPCI system be Operable during a refueling outage is being relaxed by

the second change to require one core spray subsystem and one LPCI

subsystem or two core spray subsystems to be Operable. This [change]

does not adversely affect any accident or transient analyses because

the change ensures adequate vessel inventory makeup is available in the

event of an inadvertent vessel draindown. The long term cooling

analysis following a design bases LOCA [loss of coolant accident]

demonstrates only one low pressure ECCS [emergency core cooling system]

injection/spray subsystem is required, post LOCA, to maintain the peak

cladding temperature below the allowable limit. This change will not

affect any safety limits, operating limits, or design assumptions. This

change provides the benefit of allowing maintenance to be performed on

the low pressure ECCS subsystems not required to be operable to ensure

their reliability during plant operation. Therefore, this change does

not involve a significant reduction in a margin of safety.

The final less restrictive change will allow low pressure

injection/spray subsystems to be inoperable during a refueling outage

if the spent fuel storage gates are removed and the water level is at

the required height over the top of the reactor pressure vessel flange.

This is acceptable because the water level requirement provides

sufficient coolant inventory to allow operator action to terminate any

inventory loss prior to fuel uncovery in the event of an inadvertent

draindown. This change will not affect any safety limits, operating

limits, or design assumptions. This change provides the benefit of

allowing maintenance to be performed on the low pressure ECCS

subsystems to ensure their continued reliability during plant

operation. Therefore, this change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

Pennsylvania 17105.

Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: Charles L. Miller.

Philadelphia Electric Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric Company,

Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station,

Units Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: June 9, 1994.

Description of amendment request: The licensee proposed the

following changes to its Technical Specifications (TS): (1) Revise TS

4.3.C.1 to require that each control rod be scram time tested after

each refueling outage or after a reactor shutdown that is greater than

120 days with reactor steam dome pressure greater than or equal to 800

psig prior to exceeding 40% of rated power. Scram time testing is not

required for control rods inserted per TS 3.3.B.1. (2) Replace TS

4.3.C.2 with the requirement to perform scram time testing with the

reactor steam dome pressure greater than or equal to 800 psig prior to

exceeding 40% of rated power for only those control rods associated

with the core cells affected by any fuel movement within the reactor

pressure vessel. (3) Add TS 4.3.C.3 to perform scram time testing for a

representative sample of control rods at least once per 120 days of

power operation with the reactor steam dome pressure greater than or

equal to 800 psig. (4) Add TS 4.3.C.4 to perform scram time testing at

any reactor steam dome pressure for individual control rods prior to

declaring them operable after work on the control rod or control rod

drive system is performed that could affect scram insertion time. (5)

Revise TS Bases 3.3.C and 4.3.C to describe: the rational for

performing scram time testing with reactor pressure greater than or

equal to 800 psig; the rationale for requiring control rods to be scram

time tested once per 120 days; what constitutes a representative sample

of control rods; examples of work that could affect scram times; and

the rational and methods for performing scram time testing following

work that could affect the scram insertion times. (6) Add TS 4.3.C.5 to

perform scram time testing with the reactor steam dome pressure greater

than or equal to 800 psig prior to exceeding 40% of rated power after

work on the control rod or control rod drive system that could affect

scram insertion time. (7) Revise TS 4.5.K.2 from performing scram time

testing of 19 or more control rods on a rotation basis to performing

scram time testing of a representative sample of control rods.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes will not involve any physical changes to plant

systems, structures, or components (SSC). These proposed changes will

not alter operation of process variables or SSC as described in the

safety analysis. The proposed changes establish or maintain adequate

assurance that components are operable when necessary for the

prevention or mitigation of accidents or transients and that plant

variables are maintained within limits necessary to satisfy the

assumptions for initial conditions in the safety analysis. In

particular, proposed change 1 is acceptable based on industry

experience with control rod scram time testing coupled with the

additional requirement in proposed change 4 that scram time testing of

any control rod on which work was performed must be satisfactorily

completed before that control rod can be declared operable. The

proposed changes will not allow continuous plant operation with plant

conditions such that a single failure will result in a loss of any

safety function. Therefore, the changes will not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

(2) The proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes do not alter the plant configuration (no new

or different type of equipment will be installed or removed) and will

not alter the method used by any system to perform its design function.

The proposed changes do not allow plant operation in any mode that is

not already evaluated. Therefore, these changes will not create the

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

previously evaluated.

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

a margin of safety.

Following a refueling outage, control rod scram time testing for

all control rods is currently required to be performed during

operational hydrostatic testing or during startup prior to

synchronizing the main turbine generator. Any control rods not tested

during operational hydrostatic testing must be tested at greater than

30% power but less than 40% power. Proposed change 1 will require that

scram time testing for all control rods be completed prior to exceeding

40% Reactor Power. This change is acceptable based on industry

experience with control rod scram time testing coupled with the

additional requirement in proposed change 4 that scram time testing of

any control rod on which work was performed must be satisfactorily

completed before that control rod can be declared operable. Proposed

changes 2, 3, 4 and 6 add some new requirements and make some existing

requirements more restrictive. The margin of safety is not reduced by

more restrictive changes. If anything, the margin of safety may

increase. Proposed change 5 revises the BASES to provide consistency

with the previously discussed SR [surveillance requirements] changes.

Proposed change 7 is administrative in nature and does not involve any

technical changes. Proposed changes 5 and 7 will not reduce a margin of

safety because they have no impact on any safety analysis assumptions.

Therefore, these changes will not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

Pennsylvania 17105.

Attorney for Licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: Charles L. Miller.

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

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

Date of amendment request: June 29, 1994.

Description of amendment request: The proposed changes would alter

the Plant Operating Review Committee's (PORC's) membership requirements

and would delegate a portion of PORC's procedure review

responsibilities for nuclear safety related procedures and procedure

changes to the line organizations. Section 6.5, ``Review and Audit,''

of the Technical Specifications (TSs) would be revised to modify the

composition of the PORC and delete review and audit responsibilities

for the Emergency and Security Plans from the TSs. The review and audit

responsibilities would be relocated to the respective Emergency and

Security Plans consistent with Generic Letter 93-07, ``Modification of

the Technical Specification Administrative Control Requirements for

Emergency and Security Plans.'' The proposed changes would also revise

Section 6.5 and Section 6.8, ``Procedures,'' of the TSs to delegate a

portion of the PORC's procedure review responsibilities for nuclear

safety related procedures to the line organizations. The PORC would

continue to perform safety reviews associated with procedures that are

of safety significance.

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:

Consistent with the criteria of 10 CFR 50.92, the enclosed

application is judged to involve no significant hazards based on the

following information:

1. Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident previously

analyzed?

Response: The proposed changes do not involve a significant

increase in the probability or consequences of an accident previously

evaluated since: (1) PORC will continue to review environmental impact

and 10 CFR 50.59 safety evaluations associated with procedures and

procedure changes; (2) Only personnel knowledgeable in the affected

functional areas will review procedures and procedure changes; (3)

Review and approval personnel (designated technical reviewers,

qualified safety reviewers and responsible procedure owners) will be

identified in the appropriate administrative procedures; (4) Designated

technical reviewers shall meet or exceed the qualifications described

in Section 4 of ANSI N18.1-1971 [* * *] for applicable positions and

are designated by the Department Managers; 5) The designated technical

reviewers will be responsible for identifying whether additional cross

disciplinary reviews are required; (6) The qualified safety reviewers

will be responsible for reviewing the procedure changes from a safety

perspective; (7) The responsible procedure owners are designated by the

Resident Manager; and (8) The responsible procedure owners will be

responsible for verifying that procedure reviews are performed in

accordance with the administrative procedure governing the procedure

review and approval process.

The proposed changes (1) will add more detailed requirements

regarding procedure review and approval to the Technical Specifications

which will strengthen the controls over the process, and (2) will free

PORC from reviewing items that are outside the charter of a ``safety

review'' committee [* * *] [because non-safety significant items can

reduce the time that PORC members can spend on matters that are safety

significant. The proposed Technical Specification change establishes a

highly structured review and approval program for procedures.]

The proposed change to the PORC membership requirements would not

significantly increase the probability or consequences of an accident

because it does not adversely affect the level of expertise applied to

the PORC review function or its effectiveness. The PORC quorum is

currently composed of five members including up to two designated

alternates and a Chairman. [* * *] [This composition is not changed by

the proposed amendment.]

The miscellaneous administrative changes not related to the

procedure review and approval process or the PORC membership

requirements cannot affect the probability or consequences of an

accident because they do not affect plant operations[,] [* * *]

equipment[, or any safety-related activity.]

2. Does the proposed license amendment create the possibility of a

new or different kind of accident from any accident previously

evaluated?

Response: [* * *] [No physical changes to the plant or changes in

plant equipment operating procedures are being proposed.] The changes

are administrative and will not have any direct effect on equipment

important to safety. Changing the process by which procedures are

reviewed and approved cannot in itself create the possibility of a new

or different kind of accident. Furthermore, a documented safety review,

utilizing screening criteria, will be performed for all nuclear safety

related procedures and procedure changes. The proposed change

establishes detailed controls while allowing PORC to spend more time on

safety significant issues.

The proposed change to the PORC membership requirements would not

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

previously evaluated since no physical alterations of plant

configuration or changes to setpoints or operating parameters are

proposed.

The miscellaneous administrative changes not related to the

procedure review and approval or the PORC Membership requirements

cannot create the possibility of an accident because they do not affect

plant operations[,] [* * *] equipment [or any safety-related activity.]

3. Does the proposed amendment involve a significant reduction in a

margin of safety?

Response: The proposed amendment does not involve a significant

reduction in the margin of safety because a program controlled by

Administrative Procedures using designated technical reviewers approved

by the Department Managers will be in place to review new procedures

and procedure changes. A 10 CFR 50.59 screening of each new procedure

and permanent procedure change will be performed by a qualified safety

reviewer, and PORC will continue to review 10 CFR 50.59 Safety and

Environmental Impact Evaluations associated with procedures and

procedure changes. Cross disciplinary reviews will be conducted as

appropriate. Thus, the margin of safety will be maintained by

implementing the new procedure review and approval process.

The proposed change to the PORC membership requirements would not

involve a significant reduction in the margin of safety since the level

and quality of PORC review will be maintained and there will not be an

adverse change to the collective educational background and work

experience of PORC. [There will not be an adverse loss of PORC

effectiveness as a result of this change.] The PORC quorum is currently

composed of five members including up to two designated alternates and

a Chairman. [* * *] [This composition is not changed by the proposed

changes.]

The miscellaneous administrative changes not related to the

procedure review and approval process or PORC membership program cannot

reduce any margin of safety because they do not affect any safety

related activity or equipment. These changes increase the probability

that the Technical Specifications are correctly interpreted by

clarifying information.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,

New York, New York 10019.

NRC Project Director: Michael L. Boyle.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: June 17, 1994.

Description of amendment request: The proposed changes would revise

Section 6.5, ``Review and Audit,'' of the Technical Specifications

(TSs) to modify the composition of the Plant Operating Review Committee

(PORC) and delete review and audit responsibilities for the Emergency

and Security Plans from the TSs. The review and audit responsibilities

would be relocated to the respective Emergency and Security Plans

consistent with Generic Letter 93-07, ``Modification of the Technical

Specification Administrative Control Requirements for Emergency and

Security Plans.'' The proposed changes would also revise Section 6.5

and Section 6.8, ``Procedures,'' of the TSs to delegate a portion of

the PORC's procedure review responsibilities for nuclear safety related

procedures to the line organizations. The PORC would continue to

perform safety reviews associated with procedures that are of safety

significance.

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:

Operation of the FitzPatrick plant in accordance with the proposed

amendment would not involve a significant hazards consideration as

defined in 10 CFR 50.92, since the proposed changes would not:

1. involve a significant increase in the probability of an accident

or consequence previously evaluated.

The proposed changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated since:

1) PORC will continue to review environmental impact and 10 CFR 50.59

safety evaluations associated with procedures and procedure changes; 2)

Only personnel knowledgeab

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