39430 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

application of the standard would result

in a diminution of safety to the miners.

In addition, the petitioner asserts that

the proposed alternative method would

provide at least the same measure of

protection as would the mandatory

standard.

9. Performance Coal Company

[Docket No. M–95–100–C]

Performance Coal Company, P.O. Box

69, Naoma, West Virginia 25140 has

filed a petition to modify the

application of 30 CFR 75.350 to its

Upper Big Branch South Mine (I.D. No.

46–08436) located in Raleigh County,

West Virginia. The petitioner proposes

to install a low-level carbon monoxide

detection system as an early warning

fire detection system in all belt entries

used as intake air courses. The

petitioner asserts that the proposed

alternative method would provide at

least the same measure of protection as

would the mandatory standard.

10. Performance Coal Company

[Docket No. M–95–101–C]

Performance Coal Company, P.O. Box

69, Naoma, West Virginia 25140 has

filed a petition to modify the

application of 30 CFR 75.1700 to its

Upper Big Branch South Mine (I.D. No.

46–08436) located in Raleigh County,

West Virginia. The petitioner proposes

to plug and mine through oil and gas

wells. The petitioner asserts that the

proposed alternative method would

provide at least the same measure of

protection as would the mandatory

standard.

11. C.L.D., Inc.

[Docket No. M–95–10–M]

C.L.D., Inc., 2765 East 500 South,

Vernal, Utah 84078 has filed a petition

to modify the application of 30 CFR

57.4760(a) to its Cowboy No. 1 and 2

Mine (I.D. No. 42–02096) located in

Uintah County, Utah. The petitioner

requests a variance from the mandatory

safety standard because the mining

methods used at its gilsonite mines do

not provide a physical means to comply

with the standard. The petitioner states

that a gilsonite mine uses an opentrench method of mining and that the

mines are connected to other mines and

are self-ventilating. The petitioner

asserts that the proposed alternative

method would provide at least the same

measure of protection as would the

mandatory standard.

Request for Comments

Persons interested in these petitions

may furnish written comments. These

comments must be filed with the Office

of Standards, Regulations and

Variances, Mine Safety and Health

Administration, Room 627, 4015 Wilson

Boulevard, Arlington, Virginia 22203.

All comments must be postmarked or

received in that office on or before

September 1, 1995. Copies of these

petitions are available for inspection at

that address.

Dated: July 25, 1995.

Patricia W. Silvey,

Director, Office of Standards, Regulations and

Variances.

[FR Doc. 95–18947 Filed 8–1–95; 8:45 am]

BILLING CODE 4510–43–P

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 July 7, 1995,

through July 21, 1995. The last biweekly

notice was published on Wednesday,

July 19, 1996 (60 FR 37084).

Notice Of Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, and should cite

the publication date and page number of

this Federal Register notice. Written

comments may also be delivered to

Room 6D22, Two White Flint North,

11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m.

Federal workdays. Copies of written

comments received may be examined at

the NRC Public Document Room, the

Gelman Building, 2120 L Street, NW.,

Washington, DC. The filing of requests

for a hearing and petitions for leave to

intervene is discussed below.

By September 1, 1995, the licensee

may file a request for a hearing with

respect to issuance of the amendment to

the subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39431

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1-(800) 248-5100 (in Missouri

1-(800) 342-6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, and to the attorney for the

licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Arizona Public Service Company, et al.,

Docket Nos. STN 50-528, STN 50-529,

and STN 50-530, Palo Verde Nuclear

Generating Station, Units Nos. 1, 2, and

3, Maricopa County, Arizona

Date of amendments request: July 3,

1995

Description of amendments request:

The proposed Technical Specification

(TS) amendment temporarily adds new

ACTION Statements 3.8.1.1.f and

3.8.1.1.g to TS 3.8.1.1, ‘‘A.C. Sources -

Operating,’’ to provide a method of

responding to sustained degraded

switchyard voltage. Bases 3/4.8.1, ‘‘A.C.

Sources,’’ 3/4.8.2, ‘‘D.C. Sources,’’ and

3/4.8.3, ‘‘Onsite Distribution Systems,’’

are also being revised to provide

guidance on how and why degraded

offsite power voltage and the number of

startup transformers in service affect

compliance with GDC 17 and to give the

basis for the additional ACTION

statements.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change does not significantly

increase the probability of an accident

previously evaluated in the Updated Final

Safety Analysis Report (UFSAR). The safety

function of the Electrical Distribution System

(EDS) is to provide sufficient capacity and

capability to assure that 1) specified

acceptable fuel design limits and design

conditions of the reactor coolant pressure
39432 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

boundary are not exceeded as a result of

anticipated operational occurrences and 2)

the core is cooled and containment integrity

and other vital functions are maintained in

the event of postulated accidents. In

addition, it shall have sufficient

independence, redundancy, and testability to

perform its safety function assuming a single

failure. The proposed ACTIONs will restore

the EDS to conformance with General Design

Criterion (GDC) 17 of Appendix A to 10 CFR

50. Once in conformance with GDC 17, the

system will be capable of performing its

safety function as analyzed in Chapters 6 and

15 of the UFSAR. The proposed temporary

change has no effect on the probability of

accident initiation, therefore, the probability

of an accident previously evaluated has not

been significantly increased.

The consequences of an accident

previously evaluated in the UFSAR will not

be significantly increased. Restoring one train

to OPERABLE, by blocking Fast Bus Transfer

(FBT), within one hour is consistent with the

response time of Technical Specification (TS)

ACTION 3.0.3. The second train will be

restored to OPERABLE by having its

Emergency Diesel Generator (EDG) started,

loaded, and separated from offsite power

within two hours or FBT will be blocked

within two hours. Action within two hours

is consistent with the plant—s TS since TS

ACTION 3.8.2.1.a, ‘‘D. C. Sources -

Operating,’’ would be the most limiting

requirement with one train of inoperable

electric power. In a degraded voltage event,

the ability of the Class 1E 125VDC battery

chargers to perform their function is

indeterminate, therefore, the Class 1E

125VDC batteries must be assumed to

provide the 125VDC control power to the

Class 1E Engineered Safety Features (ESF)

circuit breakers for both of their sequences.

The battery capacity calculations assume

only one sequence. Once one train is restored

to OPERABLE and the other train—s EDG

demonstrated to be OPERABLE by loading

and separating from the grid, ACTION

3.8.1.1.a, for one INOPERABLE offsite power

supply, allows operation to continue for up

to seventy-two hours. If both trains are

blocked, then both trains are OPERABLE.

The proposed change will ensure that the

train that blocks FBT will be in conformance

with GDC 17 should a subsequent accident

occur. As such, that train of ESF equipment

will be supplied Class 1E preferred and

standby power in the manner assumed by

Chapters 6 and 15 analyses. Starting, loading,

and separating the other train—s EDG from

offsite power ensures that the second train of

ESF equipment is prepared to respond to any

subsequent accident. This configuration

presents one OPERABLE offsite circuit and

two OPERABLE EDGs to any subsequent

accident, and would be capable of

withstanding the single failures in the

UFSAR Table 15.0-0, ‘‘Single Failures.’’

Optionally, with both trains blocked, both are

OPERABLE and would be capable of

withstanding the single failures in the

UFSAR Table 15.0-0, ‘‘Single Failures.’’

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Given the current licensing basis, the

proposed temporary TS change does not

create the possibility of an accident of a new

or different kind. The plant is currently

licensed to have both trains of FBT blocked

when low switchyard voltages exist in order

to prevent the loss of power generated by the

nuclear power unit from causing the loss of

the preferred power circuits. The proposed

temporary TS ACTIONs 3.8.1.1.f and

3.8.1.1.g are being added as ACTIONs to

prevent a double sequencing event from

occurring. The train that is blocked is

consistent with previous UFSAR Chapter 6

and Chapter 15 safety analyses since it will

conform to GDC 17 prior to the onset of the

accident. Under this condition it will be able

to contribute to the mitigation of an accident

and withstand the effects of any single failure

equal to its ability when initially analyzed

and licensed. The EDG which is loaded and

isolated from offsite power also contributes

to GDC 17 compliance since the entire

system can withstand a Loss of Offsite Power

(LOP) and a single failure of an EDG. With

both trains blocked, the EDS is in compliance

with GDC 17 and is analyzed.

It is understood that an accident of a

different kind will exist if a degraded voltage

condition occurs coincident with an accident

(e.g., LOCA [versus the analyzed LOP +

LOCA]). Should such an accident occur, the

manual action described in the proposed

ACTION statements could not be credited to

protect the plant. However, the purpose of

proposed ACTIONs 3.8.1.1.f and 3.8.1.1.g is

to provide an appropriate response to

degraded voltage prior to an accident by

eliminating the malfunction of a different

type (double sequencing) and an accident of

a different type (e.g., degraded voltage +

LOCA) for one train within one hour and for

the second train within two hours. This

duration of response is consistent with the

required responses currently in the TSs 3.0.3,

3.8.2.1.a, and 3.8.1.1.a.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety has not been reduced

in that the train which has FBT blocked prior

to the onset of an accident will be in

conformance with GDC 17 (which is the basis

to TS 3/4.8.1). Since the blocked train is in

conformance with GDC 17 prior to the onset

of an accident, it will support the single

failure analyses and the safety analyses to the

extent previously analyzed and licensed. The

train not blocked will have its EDG started,

loaded, and separated from offsite power

prior to the end of the second hour. Action

within two hours is consistent with TS

3.8.2.1.a. The proposed action recovers one

train of A.C. sources in one hour and places

the plant in a configuration of one less power

source than is required by LCO 3.8.1.1 within

two hours. Currently, TS ACTION 3.8.1.1.a

(one power source inoperable) has a duration

of seventy-two hours. The proposed ACTION

requires responses within time frames

consistent with TSs 3.0.3, 3.8.2.1.a, and

3.8.1.1.a, and therefore, does not reduce the

margin of safety. Optionally, restoration of

the second train by blocking FBT within two

hours is also consistent with response times

required by TS 3.0.3 and 3.8.2.1.a and

therefore, also does not reduce the margin of

safety. TS 3.8.1.1.a would not be required

with both trains of FBT blocked as all four

AC power sources would then be

OPERABLE.

Regulatory Guide 1.93, ‘‘Availability of

Electric Power Sources,’’ Revision 0,

December 1974 recognizes that under certain

conditions it may be safer to continue

operation at full or reduced power for a

limited time than to effect an immediate

shutdown based on the loss of some of the

required electric power sources. In an effort

to minimize the risk to the health and safety

of the public, the proposed ACTIONs

3.8.1.1.f and 3.8.1.1.g balance the risk of a

forced shutdown against the risk of

remaining at power with a degraded

switchyard voltage.

Probabilistic Risk Analysis (PRA) has

compared the probability of a core melt event

for 1) blocking fast bus transfer in one train

after one hour for the next seventy-one hours,

and in the second train after two hours for

the next seventy hours; 2) blocking fast bus

transfer in one train after the first hour for

the next seventy-one hours, and supplying

power to the other train from the EDG after

the second hour for seventy hours; and 3) a

normal shutdown assuming the plant is in a

normal configuration and no other transients

or accidents except an uncomplicated reactor

trip occurs during the shutdown process.

Seventy-two hours was chosen for

comparison purposes as the proposed

ACTIONs would allow operation for up to

seventy-two hours with one offsite circuit

INOPERABLE.

The PRA has shown that the probability of

a core melt event during power operation

with FBT blocked in one train after one hour

for the next seventy-one hours, and in the

second train after two hours for the next

seventy hours is approximately 1.91E-6. The

probability of a core melt event during power

operation with FBT blocked in one train after

one hour for the next seventy-one hours and

the EDG powering the opposite train after the

second hour for the next seventy hours (the

proposed configuration) is between

approximately 1.91E-6 and 1.93E-6. A range

is provided because the current PRA model

can only model blocking both trains or the

EDGs supplying both trains. The risk lies

somewhere between the two values. The

probability of a core melt event due to a

normal shutdown assuming the plant is in a

normal configuration and no other transients

or accidents except an uncomplicated reactor

trip occurs during the shutdown process is

2.4E-6. The risk can not be calculated for a

forced shutdown with degraded switchyard

voltage present but it is expected to be

higher. Therefore, the analysis provided is

conservative.

The NRC staff has reviewed the

licensee’s analysis and, based on that

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendments request

involve no significant hazards

consideration.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39433

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072-3999

NRC Project Director: William H.

Bateman

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request: July 14,

1995

Description of amendment request:

The proposed amendment would

change the scram insertion times,

Section 3.3.C, Minimum Critical Power

Ratio section, Section 4.11.C and the

associated bases in Section 2.1.1 and 3/

4.3.

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:

Section 2.1 Bases - Safety Limits

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated because equivalent fuel cladding

protection (99.9 percent of all fuel rods do

not experience transition boiling following a

design basis transient) is provided.

2. The operation of Pilgrim Station in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the proposed

change does not affect the function of any

structure, system or component.

3. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant reduction in a

margin of safety because the utilization of

current General Electric fuel designs provides

an equivalent margin of safety. As stated

previously, equivalent fuel cladding

protection is provided and ensures that 99.9

percent of all fuel rods will not experience

transition boiling following a design basis

transient.

Section 3.3.C - Scram Insertion Times

1. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant increase in the

probability of consequences of an accident

previously evaluated. The correlation of the

scram insertion times with the actual notch

position will simplify the surveillance

procedure while maintaining the accuracy of

the test.

2. The operation of Pilgrim Station in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated because no physical

modifications are associated with the

proposed change and it does not affect the

function of any structure, system or

component.

3. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant reduction in a

margin of safety. The notch positions were

chosen to coincide with the relative insertion

values specified in the Technical

Specifications. Use of the proposed

combination of notch positions and scram

insertion times will maintain the existing

margins of safety that 99.9 percent of all fuel

rods will not experience transition boiling

following a design basis transient.

Section 4.11.C - Minimum Critical Power

Ratio (MCPR) Calculation Method

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated because the method used to

calculate the measured scram speed

distribution is consistent with the PNPS

[Pilgrim Nuclear Power Station] licensing

basis.

2. The operation of Pilgrim Station in

accordance with the proposed amendment

will not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the proposed

change does not affect the function of any

structure, system or component.

3. The operation of Pilgrim Station in

accordance with the proposed amendment

will not involve a significant reduction in the

margin of safety because the proposed

changes provide equivalent fuel

cladding protection which ensures that

99.9 percent of all fuel rods will not

experience transition boiling following a

design basis transient.

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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199.

NRC Project Director: Ledyard B.

Marsh

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendment

requests: September 17, 1993, as

supplemented July 20, 1995

Description of amendment requests:

As a result of findings by a Diagnostic

Evaluation Team inspection performed

by the NRC staff at the Dresden Nuclear

Power Station in 1987, Commonwealth

Edison Company (ComEd, the licensee)

made a decision that both the Dresden

Nuclear Power Station and sister site

Quad Cities Nuclear Power Station

needed attention focused on the existing

custom Technical Specifications (TS)

used.

The licensee made the decision to

initiate a Technical Specification

Upgrade Program (TSUP) for both

Dresden and Quad Cities. The licensee

evaluated the current TS for both

Dresden and Quad Cities against the

Standard Technical Specifications (STS)

contained in NUREG-0123, ‘‘Standard

Technical Specifications General

Electric Plants BWR/4.’’ The licensee’s

evaluation identified numerous

potential improvements such as

clarifying requirements, changing TS to

make them more understandable and to

eliminate interpretation, and deleting

requirements that are no longer

considered current with industry

practice. As a result of the evaluation,

ComEd has elected to upgrade both the

Dresden and Quad Cities TS to the STS

contained in NUREG-0123.

The TSUP for Dresden and Quad

Cities is not a complete adaption of the

STS. The TSUP focuses on (1)

integrating additional information such

as equipment operability requirements

during shutdown conditions, (2)

clarifying requirements such as limiting

conditions for operation and action

statements utilizing STS terminology,

(3) deleting superseded requirements

and modifications to the TS based on

the licensee’s responses to Generic

Letters (GL), and (4) relocating specific

items to more appropriate TS locations.

The September 17, 1993, and July 20,

1995, applications proposed to upgrade

only Section 3/4.7 (Containment

Systems) of the Dresden and Quad

Cities TS.

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 a

significant increase in the probability or

consequences of an accident previously

evaluated because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Implementation

of these changes will provide increased

reliability of equipment assumed to operate

in the current safety analysis, or provide

continued assurance that specified

parameters remain within their acceptance

limits, and as such, will not significantly

increase the probability or consequences of a

previously evaluated accident.

Some of the proposed changes represent

minor curtailments of the current
39434 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

requirements which are based on generic

guidance or previously approved provisions

for other stations. The proposed amendment

for Dresden and Quad Cities Station’s

Technical Specification Section 3/4.7 is

based on STS guidelines or later operating

BWR plants’ NRC accepted changes. Any

deviations from STS requirements do not

significantly increase the probability or

consequences of any previously evaluated

accidents for Dresden or Quad Cities

Stations. The proposed amendment is

consistent with the current safety analyses

and has been previously determined to

represent sufficient requirements for the

assurance and reliability of equipment

assumed to operate in the safety analysis, or

provide continued assurance that specified

parameters remain within their acceptance

limits. As such, these changes will not

significantly increase the probability or

consequences of a previously evaluated

accident.

The associated systems that make up the

Containment Systems are not assumed in any

safety analysis to initiate any accident

sequence for Dresden or Quad Cities Stations;

therefore, the probability of any accident

previously evaluated is not increased by the

proposed amendment. In addition, the

proposed surveillance requirements for the

proposed amendments to these systems are

generally more prescriptive than the current

requirements specified within the Technical

Specifications. The additional surveillance

requirements improve the reliability and

availability of all affected systems and,

therefore, reduce the consequences of any

accident previously evaluated, as the

probability of the systems outlined within

Section 3/4.7 of the proposed Technical

Specifications performing their intended

function is increased by the additional

surveillances.

Create the possibility of a new or different

kind of accident from any previously

evaluated because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Other changes

represent minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. These changes do not

involve revisions to the design of the station.

Some of the changes may involve revision in

the operation of the station; however, these

provide additional restrictions which are in

accordance with the current safety analysis,

or are to provide for additional testing or

surveillances which will not introduce new

failure mechanisms beyond those already

considered in the current safety analyses.

The proposed amendment for Dresden and

Quad Cities Station’s Technical Specification

Section 3/4.7 is based on STS guidelines or

later operating BWR plants’ NRC accepted

changes. The proposed amendment has been

reviewed for acceptability at the Dresden or

Quad Cities Nuclear Power Stations

considering similarity of system or

component design versus the STS or later

operating BWRs. Any deviations from STS

requirements do not create the possibility of

a new or different kind of accident

previously evaluated for Dresden or Quad

Cities Stations. No new modes of operation

are introduced by the proposed changes.

Surveillance requirements are changed to

reflect improvements in technique, frequency

of performance or operating experience at

later plants. Proposed changes to action

statements in many places add requirements

that are not in the present technical

specifications. The proposed changes

maintain at least the present level of

operability. Therefore, the proposed changes

do not create the possibility of a new or

different kind of accident from any

previously evaluated.

The associated systems that make up the

Containment Systems are not assumed in any

safety analysis to initiate any accident

sequence for Dresden or Quad Cities Stations.

In addition, the proposed surveillance

requirements for affected systems associated

with the Containment Systems are generally

more prescriptive than the current

requirements specified within the Technical

Specifications; therefore, the proposed

changes do not create the possibility of a new

or different kind of accident from any

previously evaluated.

Involve a significant reduction in the

margin of safety because:

In general, the proposed amendment

represents the conversion of current

requirements to a more generic format, or the

addition of requirements which are based on

the current safety analysis. Other changes

represent minor curtailments of the current

requirements which are based on generic

guidance or previously approved provisions

for other stations. Some of the later

individual items may introduce minor

reductions in the margin of safety when

compared to the current requirements.

However, other individual changes are the

adoption of new requirements which will

provide significant enhancement of the

reliability of the equipment assumed to

operate in the safety analysis, or provide

enhanced assurance that specified

parameters remain with their acceptance

limits. These enhancements compensate for

the individual minor reductions, such that

taken together, the proposed changes will not

significantly reduce the margin of safety.

The proposed amendment to Technical

Specification Section 3/4.7 implements

present requirements, or the intent of present

requirements in accordance with the

guidelines set forth in the STS. Any

deviations from STS requirements do not

significantly reduce the margin of safety for

Dresden or Quad Cities Stations. The

proposed changes are intended to improve

readability, usability, and the understanding

of technical specification requirements while

maintaining acceptable levels of safe

operation. The proposed changes have been

evaluated and found to be acceptable for use

at Dresden or Quad Cities based on system

design, safety analysis requirements and

operational performance. Since the proposed

changes are based on NRC accepted

provisions at other operating plants that are

applicable at Dresden or Quad Cities and

maintain necessary levels of system or

component reliability, the proposed changes

do not involve a significant reduction in the

margin of safety.

The proposed amendment for Dresden and

Quad Cities Stations will not reduce the

availability of systems associated with the

Containment Systems when required to

mitigate accident conditions; therefore, the

proposed changes do not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment requests involve no

significant hazards consideration.

Local Public Document Room

location: for Dresden, Morris Public

Library, 604 Liberty Street, Morris,

Illinois 60450; for Quad Cities, Dixon

Public Library, 221 Hennepin Avenue,

Dixon, Illinois 61021

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690

NRC Project Director: Robert A. Capra

Connecticut Yankee Atomic Power

Company, and Northeast Nuclear

Energy Company, et al., Docket Nos.50-

213, 50-245, 50-336, and 50-423

Haddam Neck Plant, and Millstone

Nuclear Power Station, Units 1,2, and

3, Middlesex County and New London

County, Connecticut

Date of amendment request: June 6,

1995

Description of amendment request:

The proposed amendment will modify

the size of the Plant Operations Review

Committee (PORC) which will

collectively have the experience and

expertise in various areas of plant

operation, and will clarify the

composition of the Site Operations

Review Committee (SORC).

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 (SHC), which is presented

below:

These proposed changes do not involve an

SHC because the changes do not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The PORC is an oversight group and helps

to ensure that the units are operated in a safe

manner. To accomplish this the PORCs

provide their recommendations on the safety

related activities to the Vice President -

Haddam Neck Plant for Haddam Neck and to

the respective Nuclear Unit Directors for

Millstone. Each Millstone

Unit has its own PORC. It is proposed that

the members of the

PORC be selected by the respective Nuclear

Unit Director based on their knowledge and
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39435

expertise in specific key plant functions. The

Millstone Station has one SORC. The SORC

is also an oversight group whose charter is

to advise the Senior Vice President -

Millstone Station on all matters related to

nuclear safety at the Millstone site. The

Haddam Neck Plant, being a single unit site,

has one PORC, which advises the Vice

President - Haddam Neck Plant. The

members of the Haddam Neck Plant PORC

will be selected by the Vice President -

Haddam Neck Plant based on their

knowledge and expertise in specific key

plant functions. The PORC and SORC add to

the defense-in-depth concept provided by the

design, operation, maintenance, and quality

oversight by promoting excellence through

the conduct of their affairs and by

maintaining a diligent watch over their

responsibilities.

These administrative changes will revise

the composition section of the technical

specifications for the PORC members.

Millstone Unit individuals will be appointed

by the Nuclear Unit Directors if the

individual meets one or more of the

following areas of expertise: Plant

Operations, Engineering, Reactor

Engineering, Maintenance, Instrumentation

and Controls, Health Physics, Chemistry,

Work Planning and Control, and Quality

Services. The Haddam Neck Plant, due to its

broader scope of review also include[s] an

individual experienced in Security and

specific experience in Electrical Maintenance

and Mechanical Maintenance. The

individuals who will serve on PORC shall

continue to meet the criteria of ANSI N18.1-

1971. This approach is consistent with the

standard technical specifications and NUREG

0800, Section 13.4. For SORC at the

Millstone Station, the method of identifying

who shall serve as Vice Chairperson has been

modified for clarity. The Site Services

Director position is proposed to be

eliminated since this position no longer

exists. The functions previously performed

by this individual have been assumed by

those individuals who currently serve on

SORC. Finally, [the TS relating to] the

individual who shall represent Quality and

Assessment Services shall be modified to

allow a qualified member of Quality and

Assessment Services to serve on SORC.

The remaining portions of the technical

specifications related to PORC and SORC are

not being revised.

These modifications broaden the unit

committee participation and reflect current

organizational positions and will not increase

the probability of occurrence or the

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed administrative

enhancements to the composition of the

PORC and Millstone Station SORC will not

affect the way in which the units are

physically operated. These administrative

changes to PORC and SORC continue to meet

the guidelines of ANSI N18.7-1976. The

modifications to PORC and SORC continue to

allow these groups to provide a thorough

review of activities at the units.

The proposed modification does not

impact any initiating events, and, therefore,

cannot create the possibility of any new or

different kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

These proposed administrative changes

will not impact the margin of safety provided

by PORC and SORC. The PORC and SORC

will continue to be staffed by qualified

individuals experienced in the operation of

the plants. These administrative changes will

modify how the composition of the PORC

and SORC members are presented in the

technical specifications, but will not

adversely impact their ability to review and

comment on operations at the units.

These changes do not impact any

protective boundaries nor do they impact the

safety limits for the protective boundaries.

These proposed changes are administrative

in nature. Therefore, there is no 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, Copnnecticut 06457,

for the Haddam Neck Plant, and the

Learning Resources Center, Three Rivers

Community-Technical College, 574 New

London Turnpike, Norwich, CT 06360,

for Millstone 1, 2, and 3.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Consumers Power Company, Docket

No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: July 5,

1995

Description of amendment request:

The proposed amendment would

change the Administrative Controls

section of the Palisades Technical

Specifications. The changes involve

deleting training requirements in the

Administrative Controls section,

revising the Plant Review Committee

composition, and revising the function

and composition of the plant safety and

licensing staff review requirements.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

This change does not affect the probability

or consequences of an accident. The changes

are administrative, deleting an unnecessary

specification on staff training requirements,

eliminating the specific references to the

Nuclear Engineering and Construction

Organization (NECO) staff, and requiring that

the Plant Review Committee (PRC) chairman,

alternate chairman, and members be

designated in administrative procedures by

the Plant General Manager. Further

administrative changes clarify the function of

the Plant Safety and Licensing organization

and eliminate the numerical requirement for

five staff members to fulfill the organization

function.

The removal of an obsolete staff training

requirement does not diminish the regulatory

requirement to have an adequately trained

staff. The accredited training programs for

the plant staff ensure an appropriate level of

training is conducted to maintain an

appropriate skill and knowledge base for the

staff. The requirements of 10CFR55 provide

the necessary rules for operator licenses.

Since a trained staff will be maintained, there

will [be] no increase in the probability or

consequences of an accident as a result of

this change.

The composition of the PRC will not be

affected by this change as it will, at a

minimum, be comprised of personnel from

the operations, engineering, radiological

services and maintenance departments as

required by the Technical Specifications. The

composition of the Plant Safety and

Licensing organization as a whole may

change. The function of the organization as

it relates to these Technical Specifications,

however, will not be affected. These changes

have no affect on the plant accident analyses.

Qualified personnel will continue to conduct

the PRC and Plant Safety and Licensing

reviews. Therefore, the changes do not

increase the probability or consequences of

an accident.

B. Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed changes are administrative

and do not create the possibility of a new or

different kind of accident. Staff training will

continue to meet the accreditation

requirements of the National Academy for

Nuclear Training Accreditation Board and

the requirements for the Systematic

Approach to Training. Operators’ license

training will continue to meet the regulatory

requirements of 10CFR55. Activities

conducted by the Plant Review Committee

and the Plant Safety and Licensing staff will

continue to be accomplished by a staff which

meets the qualification requirements of the

Technical Specifications. These

administrative changes will not affect the

operation of the plant or the safety function

of plant equipment nor will it affect the

quality of the review activities. Therefore,

there will be no possibility that a new or

different kind of accident will be created.

C. Involve a significant reduction in a

margin of safety.

The changes do not affect installed plant

equipment nor do they affect plant
39436 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

operations. These administrative changes

have not affected the probability or

consequences of a previously analyzed

accident or created the possibility of a new

or different kind [of] accident from any

previously evaluated. Therefore, they do not

involve any reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201

NRC Project Director: John N. Hannon

Duquesne Light Company, et al., Docket

No. 50-334, Beaver Valley Power

Station, Unit No. 1, Shippingport,

Pennsylvania

Date of amendment request: October

11, 1994, as supplemented June 23,

1995.

Description of amendment request:

The proposed amendments would

revise Beaver Valley Power Station, Unit

Nos. 1 and 2 (BVPS-1 and BVPS-2)

Technical Specifications (TSs) 1.18,

‘‘Quadrant Power Tilt Ratio,’’ 3/4.2.4,

‘‘Quadrant Power Tilt Ratio,’’ the Table

Notation of TS Table 3.3.-1, ‘‘Reactor

Trip System Instrumentation,’’ and

associated Bases to incorporate the

guidance provided in the NRC’s

Improved Standard Technical

Specifications (NUREG-1431) applicable

to these TSs. The proposed amendments

would clarify the requirements of the

subject TSs with regard to the use of

excore power range neutron flux

detectors to monitor quadrant power tilt

ratio when an excore power range

neutron flux instrument is inoperable.

The proposed change would also make

several minor editorial changes in the

subject TSs.

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 existing quadrant power tilt ratio

(QPTR) definition and Surveillance

Requirement (SR) 4.2.4.c are inconsistent

concerning reactor power limitations when

performing QPTR surveillance requirements.

The proposed change modifies these and

related requirements to improve the

understanding and consistency by generally

incorporating the Improved Standard

Technical Specification (ISTS) requirements

of NUREG-1431.

Editorial changes have been incorporated

throughout the proposed specifications to

address ISTS or plant specific convention

and do not affect the accident analyses. The

QPTR definition has been modified to reflect

the ISTS wording and eliminate the

inconsistency with SR 4.2.4.c. This change

does not reduce the QPTR testing

requirements or affect the accident analyses

assumptions. The current action statements

require power reduction along with a

reduction in power range high neutron flux

trip setpoints when the QPTR exceeds the

limit. This ensures the core conditions are

consistent with the accident analyses

assumptions. With the modified action

statements and the QPTR exceeding the limit,

power reduction is also required along with

performing a flux map to verify the peaking

factors are within the accident analyses

assumptions. In addition, the safety analyses

must be re-evaluated to confirm the results

remain valid prior to increasing power with

an indicated tilt condition. The new action

statements provide methods different from

the current requirements. However, they

satisfy the same objective, to ensure the

conditions assumed in the accident analyses

are maintained. Therefore, these changes will

not involve significant increase in the

probability or consequences of an accident

previously evaluated.

The current surveillance requirements

define the methods and frequencies for

verifying the QPTR is within the limit

specified in the limiting condition for

operation. The proposed SRs include

associated notes that allow separation of a

power range channel into two portions madeup of the Nuclear Instrumentation System

(NIS) and the excore detector portion. If an

excore detector portion of a power range

channel is inoperable, then the power range

channel is inoperable since the detector

provides input to the NIS which inputs to the

solid state protection system. However, if the

excore detector is operable and the NIS is

inoperable, then the power range channel is

inoperable but the ability to monitor the

QPTR is unaffected. When the NIS portion of

a channel is inoperable, appropriate actions

are applied in accordance with Specification

3.3.1. The new SRs continue to require the

same testing and frequencies as the current

SRs along with reducing the need to interpret

the requirements when special conditions

exist. Therefore, the proposed SRs will not

affect the accident analyses or significantly

increase the probability or consequences of

an accident previously evaluated.

Table 3.3-1 Action 2 applies when a power

range channel is inoperable. This action has

been reformatted to incorporate changes

similar to those adopted in the QTPR SR

which allow separation of a power range

channel into the NIS portion and the excore

detector portion. Proposed Action 2.a applies

to an inoperable power range high neutron

flux channel and Action 2.b applies to ‘‘all

other channels’’ which includes the Low

Setpoint function along with the High

Positive and High Negative Rate functions.

The new action is modified by Note (3) to

allow bypassing the inoperable channel for

surveillance testing and setpoint adjustment

and by Note (4) that only requires performing

SR 4.2.4 when the power range high neutron

flux channel input to QPTR is inoperable.

The new action does not require reducing the

power range neutron flux setpoint like the

current action since the proposed action is to

perform the QPTR surveillance or shutdown

which is more conservative than the current

action requirement, otherwise, the new

action requires essentially the same steps to

be performed as the current action.

Therefore, the proposed action will not affect

the accident analyses or involve a significant

increase in the probability or consequences

of an accident previously evaluated.

These changes are proposed to allow

flexibility in plant operations by modifying

the QPTR action and surveillance

requirements to allow separation of a power

range channel into the NIS portion and the

excore detector portion. The modified action

and surveillance requirements continue to

provide monitoring of those parameters

required to ensure the core is operating

safely. Since these changes are not

significantly different from the current

requirements and no change is being

introduced that would affect the accident

analyses assumptions, we have concluded

that the 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?

The proposed changes incorporate

modifications generally consistent with the

ISTS QPTR requirements to ensure the core

power distribution is adequately monitored.

The revised action statements provide for

peaking factor verification as a logical

compensatory measure to ensure the core is

operating within required limits. This is

more conservative than the current

requirements and provides additional

assurance that Specification 3.2.4 will

continue to govern the QPTR limitations in

a manner consistent with the accident

analyses assumptions. The revised SR

provides clear and understandable testing

requirements to reduce confusion concerning

how the QPTR is to be monitored based on

plant conditions. The proposed change does

not introduce any new mode of plant

operation or require any physical

modification to the plant, therefore, this

change will 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 QPTR limit ensures that the gross

radial power distribution is maintained

within the assumptions used in the safety

analyses. The QPTR is one of the variables

that is monitored to ensure the core operates

within the bounds used in the safety

analyses. When the QPTR is maintained

below 1.02 it provides an indication that the

peaking factors are within the limiting values

by preventing and undetected change in the
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39437

gross radial power distribution. The

proposed changes ensure the required

parameters are verified during the applicable

conditions and on a consistent basis,

therefore, these changes will 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 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: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, Arkansas Nuclear

One, Unit Nos. 1 and 2 (ANO-1&2),

Pope County, Arkansas

Date of amendment request: May 19,

1995

Description of amendment request:

The proposed amendments revise the

specifications to permit the reactor

building personnel airlock doors to

remain open during fuel handling.

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:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The proposed change would allow the

containment personnel airlock doors to

remain open during fuel movement and core

alterations. These doors are normally closed

during this time period in order to prevent

the escape of radioactive material in the

event of a fuel handling accident. These

doors are not initiators of any accident. The

probability of a fuel handling accident is

unaffected by the position of the containment

personnel airlock doors.

The proposed change alters assumptions

made in evaluating the radiological

consequences of a fuel handling accident

inside the reactor containment building.

Allowing the containment personnel airlock

doors to remain open during fuel movement

and core alterations does increase, however

not significantly, the consequences of a fuel

handling accident inside containment.

Previously, the fuel handling accident inside

containment was bounded by the fuel

handling accident analysis in the spent fuel

pool area of the auxiliary building. Part of the

dose increase has been offset by the increase

in the minimum decay time before irradiated

fuel may be moved inside the reactor

containment building. Extending the

minimum decay time actually decreases the

consequences of a fuel handling accident by

reducing the radioactive inventory of the

irradiated fuel which could possibly be

released during a fuel handling accident. The

revised fuel handling accident analysis

results in maximum offsite doses of 43.4 Rem

and 41.8 Rem to the thyroid and 0.616 Rem

and 0.598 Rem to the whole body for ANO1 and ANO-2, respectively. The calculated

offsite doses are well within the limits of

10CFR Part 100. Also, the calculated doses

are larger than the actual doses which would

be expected during a fuel handling accident

because the calculation does not incorporate

the closing of at least one of the personnel

airlock doors following evacuation of

containment. The proposed change would

significantly reduce the dose to workers in

the containment in the event of a fuel

handling accident by expediting the

containment evacuation process.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

The proposed change does not involve the

addition or modification of any plant

equipment. Also, the proposed change would

not alter the design, configuration, or method

of operation of the plant.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

This proposed change has the potential for

an increased dose at the site boundary due

to a fuel handling accident; however, the

dose remains within acceptable limits. The

margin of safety as defined by 10CFR Part

100 has not been significantly reduced. There

is an increase in the calculated offsite dose

resulting from a fuel handling accident;

however, the increase is not significant and

is well within the limits specified in 10 CFR

Part 100. The overall significance will be

offset by the increased minimum decay time,

the decreased potential radiation dose to

workers, and the increased availability of the

personnel airlock door in the event of a fuel

handling accident. Closing at least one of the

personnel airlock doors following an

evacuation of containment, further reduces

the offsite doses in the event of a fuel

handling accident which partially

compensates for the higher offsite doses

calculated as a result of this proposed

change.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

BecknerEntergy Operations, Inc., Docket

No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: March

17, 1995

Description of amendment request:

The proposed amendment revises

requirements associated with channel

functional tests of the core protection

calculator following a high temperature

alarm.

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:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The core protection calculators (CPCs) are

not accident initiators, therefore this change

does not increase the probability of an

accident previously evaluated.

The core protection calculators (CPCs) are

dedicated minicomputers that receive key

parameters necessary to calculate the

departure from nucleate boiling ratio (DNBR)

and local power density (LPD) and issue a

reactor trip command prior to reaching plant

conditions that may damage the fuel in the

reactor. Subjecting a computer to elevated

temperatures may affect the reliability of the

computer calculations. This change in the

Arkansas Nuclear One-Unit 2 (ANO-2)

Technical Specifications (TS) will require a

verification of the CPC operability, by the

performance of a channel functional test, in

the event a cabinet high temperature switch

is actuated. This is a more accurate

indication of the operating environment of

the CPCs than the current requirement to

perform the test based upon room

temperature. The ability of the CPCs to

monitor DNBR and LPD and issue a trip

command when appropriate will not be

affected in any way by this change, therefore

the consequences of an accident previously

evaluated are not increased.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

Because the proposed changes do not alter

the design, configuration, or method of

operation of the plant, they do not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

These proposed changes do not alter the

acceptance criteria of any surveillance

requirements. The changes do not alter any

assumptions used in accident analysis,

change any actuation setpoints, nor allow
39438 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

operations in any configuration not

previously analyzed. This change will trigger

a verification of affected CPC operability

based on cabinet temperature instead of room

temperature, which is a more accurate

indication of the operating environment of

the CPC computer. Therefore, this change

does not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: April 4,

1995

Description of amendment request:

The proposed amendment revises

operating criteria and requirements

associated with containment personnel

air locks.

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:

Criterion 1 - Does Not Involve a Significant

Increase in the

Probability or Consequences of an

Accident Previously Evaluated.

The containment air locks are passive

components integral to the containment

structure and are not evaluated to be accident

initiators, therefore, the proposed

amendment does not involve an increase in

the probability of an accident previously

evaluated.

Each air lock door is rated for and tested

to full design pressure of the containment

building. If one door were inoperable in each

air lock, the remaining door, since required

to remain closed and locked, would provide

the necessary fission product barrier to

prevent an uncontrolled release, therefore the

amendment allowance for an inoperable air

lock door in each air lock does not increase

the consequences of any previously

evaluated accident.

During a situation where one containment

air lock door is inoperable and the operable

door is opened, a breech in containment

integrity would essentially exist while the

operable door remains open. The time

required for a containment air lock door to

be open for ingress or egress does not exceed

two to three minutes. The amendment

provision to allow unlocking and opening an

operable air lock door for ingress and egress

to facilitate air lock maintenance necessary to

restore operability does not increase the

consequences of any previously evaluated

accident since the time necessary for the door

to be open is bounded by the existing one

hour time allowance for an actual breech of

containment integrity (TS 3.6.1.1.)

The containment air lock interlock

functions to prevent simultaneous opening of

both air lock doors thereby creating a breech

in containment integrity. A dedicated

individual stationed at the air lock to

administratively control door operations, or

locking closed an operable door will

adequately assure containment integrity. The

addition of this technical specification action

statement, therefore, does not increase the

consequences of any previously evaluated

accident.

Performance of the overall air lock leakage

test requires opening the outer air lock door

for installation of the mechanical dogging

devices on the inner door. The current

technical specifications make no provisions

for this entry and thus would require a plant

shutdown if the inner door was inoperable in

an air lock. The proposed amendment

removes the requirement to shut down when

the barrel leak rate is due. The time required

for the containment air lock doors to be

opened for dog installation would be the

same as for ingress and egress as discussed

above, therefore this change does not

increase the consequences of any previously

evaluated accident.

10 CFR 50, Appendix J contains

containment leakage testing requirements,

including specific requirements for

containment building air locks. Changing the

TS surveillance requirements to refer to 10

CFR 50, Appendix J for these test

requirements will not degrade these tests,

therefore this change does not increase the

consequences of any previously evaluated

accident.

The air lock door seal pressure test is

performed any time the air lock is used for

containment access during modes of

operation when containment integrity is

required. The door seal test is intended to be

a gross test to verify that the door seals were

not damaged during the opening and closing

cycle(s). This test does not replace the

required overall barrel leakage test. Based on

information provided by the air lock vendor,

a test pressure of 10 psig is sufficient to

perform this gross seal verification. A change

in the allowable leakage rate is requested to

remove a specific numerical value from the

TS surveillance requirements section and

replace it with a fraction of LaG. This new

acceptable leakage rate remains relatively

insignificant and is bounded by the overall

air lock leakage rate. Based on these facts this

change in test pressure and associated

acceptance criteria does not increase the

consequences of any previously evaluated

accident.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

Because the proposed changes do not

change the design, configuration, or method

of operation of the plant, they do not create

the possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed changes to ANO-2 TS

involve allowing brief breaches in

containment integrity for the purpose of

repairing inoperable air lock components or

performing surveillances required by 10 CFR

50, Appendix J. These cases are adequately

bounded by the one hour allowable outage

time afforded by TS 3.6.1.1.

The addition of a specific action statement

addressing an inoperable air lock interlock

provides those actions necessary to assure

the maintenance of containment integrity.

This is achieved by locking an operable door

in the affected air lock when not in use and

stationing a dedicated individual at the air

lock, during periods of ingress and egress,

whose sole responsibility is to insure only

one air lock door is opened at a time thereby

duplicating the function of the mechanical

interlock.

The proposed changes also consist of

administrative changes removing an outdated

exemption to 10 CFR 50, Appendix J and

removing specific surveillance requirements

from the specifications, instead referring to

the controlling requirements of 10 CFR 50,

Appendix J. This is consistent with the

provisions of NUREG 1432 ‘‘Revised

Standard Technical Specifications for

Combustion Engineering Plants,’’ Rev. 0.

None of the proposed changes increase the

allowable overall air lock leakage rate, nor

affect the acceptance criteria of the overall

integrated containment leakage rate. All of

the changes are bounded by existing analyses

for all evaluated accidents and do not create

any situations that alter the assumptions

used in these analyses. Therefore, this change

does not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas

72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, D.C.

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: May 19,

1995
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39439

Description of amendment request:

The proposed amendment adds criteria

to address optional inspections of steam

generator tubes.

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:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

Steam generator tubes are inspected on a

periodic basis to reduce the probability of a

steam generator tube rupture or tube leakage.

Five special interest groups are being added

for optional inspections in addition to the

general tube inspections currently required

by the technical specifications. These special

interest groups define areas of tubes where

known or potential degradation mechanisms

may exist for which additional inspection,

above that currently required in the technical

specifications, may be beneficial. Inspection

of these special interest groups may utilize

probes which more readily detect indications

which may be found in the special interest

areas. The increased detection capability will

reduce the probability that a structurally

significant flaw will go undetected during an

inspection. The minimum sample size and

expansion criteria (should a flaw be found)

for inspections of special interest groups are

based on percentages of tubes potentially

affected by the specific degradation

mechanisms for which the special inspection

is being performed. The percentages used are

the same as used for the current general tube

inspections. The expansion criteria allow

expansion within the area of interest without

affecting the expansions of any general tube

inspection. By expanding within the area of

interest, a more complete inspection for the

defects caused by a specific degradation

mechanism can be performed than if the

expansion were conducted in tubes not

necessarily affected by the degradation

mechanism, which is possible with the

current technical specifications. Therefore,

this change does not involve a significant

increase in the probability of an accident

previously considered.

The proposed change does not increase the

amount of radioactive material available for

release or modify any systems used for

mitigation of such releases during accident

conditions. The steam generator tubing will

continue to be examined on the frequency

currently specified in the technical

specifications. This change will allow steam

generator examinations to focus on known

areas of interest without requiring

unnecessary expansion. The integrity of the

steam generators will continue to be assured

at an equivalent level. Therefore, the change

does not involve a significant increase in the

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

Special inspections such as the ones being

added to the technical specifications have

been conducted in the past at ANO-2. The

method of inspection, pushing or pulling a

probe through the steam generator tubes from

the primary side, is the same method

employed for the current technical

specification required inspections. Inspection

methodology is not being changed by

incorporation of these special interest groups

into the technical specifications. No design

or operational characteristics of the plant are

changed by the proposed amendment.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed amendment adds special

interest groups for optional inspection into

the technical specifications. These

inspections concentrate on areas of interest

using inspection methodology that is

equivalent or better at finding specific types

of flaws than the methodology used for the

currently required general tube inspections.

If the special interest groups are not

inspected, the existing technical specification

requirements for inspection still apply.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: May 19,

1995

Description of amendment request:

The proposed amendment increases the

allowed outage time for an emergency

diesel generator from 72 hours to seven

days. Additionally, the amendment

authorizes one, ten-day diesel generator

maintenance outage every fuel cycle.

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:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The emergency diesel generators (EDGs)

are backup alternating current power sources

designed to power essential safety systems in

the event of a loss of offsite power. EDGs are

not an accident initiator in any accident

previously evaluated. Therefore, this change

does not involve an increase in the

probability of an accident previously

evaluated.

The EDGs provide backup power to

components that mitigate the consequences

of accidents. The proposed changes to

allowed outage times (AOTs) do not affect

any of the assumptions used in deterministic

safety analysis.

In order to fully evaluate the EDG AOT

extension, probabilistic safety analysis

methods were utilized. The results of these

analyses indicate no significant increase in

the consequences of an accident previously

evaluated. These analyses are detailed in CE

NPSD-996, Combustion Engineering Owners

Group ‘‘Joint Applications Report for

Emergency Diesel Generators AOT

Extension.’’

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

This proposed change does not alter the

design, configuration, or method of operation

of the plant. Therefore, this change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed changes do not affect the

technical specification limiting conditions

for operation or their bases which support

the deterministic analyses used to establish

the margin of safety. Evaluations used to

support the requested technical specification

changes have been demonstrated to be either

risk neutral or risk beneficial. These

evaluations are detailed in CE NPSD-996.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, DC

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No.

50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: May 19,

1995

Description of amendment request:

The proposed amendment increases the

allowed outage time for an inoperable
39440 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

Safety Injection Tank (SIT) from one

hour to 24 hours. Additionally, the

amendment limits power operation to

72 hours when certain SIT related

instrument functions are inoperable.

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:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The Safety Injection Tanks (SITs) are

passive components in the Emergency Core

Cooling System. The SITs are not accident

initiators in any accident previously

evaluated. Therefore, this change does not

involve an increase in the probability of an

accident previously evaluated.

SITs were designed to mitigate the

consequences of Loss of Coolant Accidents

(LOCA). These proposed changes do not

affect any of the assumptions used in

deterministic LOCA analysis. Therefore, the

consequences of accidents previously

evaluated do not change.

In order to fully evaluate the effect of the

SIT Allowable Outage Time (AOT) extension,

probabilistic safety analysis (PSA) methods

were utilized. The results of these analyses

show no significant increase in the core

damage frequency. As a result, there would

be no significant increase in the

consequences of an accident previously

evaluated. These analyses are detailed in CE

NPSD-994, Combustion Engineering Owners

Group ‘‘Joint Applications Report for Safety

Injection Tank AOT/STI Extension.’’

The change pertaining to SIT inoperability

based solely on instrumentation malfunction

does not involve a significant increase in the

consequences of an accident as evaluated and

endorsed by the NRC in NUREG-1366,

‘‘Improvements to Technical Specifications

Surveillance Requirements.’’

Therefore, this change does not involve an

increase in the probability or a significant

increase in the consequences of any accident

previously evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

This proposed change does not change the

design, configuration, or method of operation

of the plant. Therefore, this change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed changes do not affect the

limiting conditions for operation or their

bases that are used in the deterministic

analyses to establish the margin of safety.

PSA evaluations were used to evaluate these

changes. These evaluations demonstrated

that the changes are either risk neutral or risk

beneficial. These evaluations are detailed in

CE NPSD-994.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, D.C.

20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., Docket No. 50-

368, Arkansas Nuclear One, Unit No. 2, Pope

County, Arkansas

Date of amendment request: May 19,

1995

Description of amendment request:

The proposed amendment increases the

allowed outage time for one train of low

pressure safety injection from 72 hours

to seven days.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1 - Does Not Involve a Significant

Increase in the Probability or Consequences

of an Accident Previously Evaluated.

The low pressure safety injection system

(LPSI) is part of the Emergency Core Cooling

System subsystem. Inoperable LPSI

components are not considered to be

accident initiators. Therefore, this change

does not involve an increase in the

probability of an accident previously

evaluated.

The LPSI system was designed to mitigate

the consequences of a large loss of coolant

accident (LOCA). These proposed changes do

not affect any of the assumptions used in

deterministic LOCA analysis.

In order to fully evaluate the LPSI AOT

extension, probabilistic safety analysis

methods were utilized. The results of these

analyses indicate no significant increase in

the consequences of an accident previously

evaluated. These analyses are detailed in CE

NPSD-995, Combustion Engineering Owners

Group ‘‘Joint Applications Report for Low

Pressure Safety Injection System AOT

Extension.’’

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Criterion 2 - Does Not Create the Possibility

of a New or Different Kind of Accident from

any Previously Evaluated.

This proposed change does not change the

design, configuration, or method of operation

of the plant. Therefore, this change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Criterion 3 - Does Not Involve a Significant

Reduction in the Margin of Safety.

The proposed changes do not affect the

technical specification limiting conditions

for operation or their bases which support

the deterministic analyses used to establish

the margin of safety. Probabilistic evaluations

used to support the requested technical

specification changes have been

demonstrated to be either risk neutral or risk

beneficial. These evaluations are detailed in

CE NPSD-995.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., Washington, D.C.

20005-3502

NRC Project Director: William D.

Beckner

GPU Nuclear Corporation, et al.,

Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of amendment request: June 26,

1995

Description of amendment request:

The amendment revises the snubber

visual inspection intervals to match the

schedule developed by the NRC staff for

use with a 24 month refueling interval.

This schedule was documented in

Generic Letter 90-09. The licensee has

made wording changes not contained in

Generic Letter 90-09. These changes are

as follows:

a) Section 4.5.Q.1 - GL 90-09 wording

’’...performance of the following

augmented inservice inspection

program in addition to the requirements

of Section 4.0.5.’’

Proposed Technical Specification

wording ’’...performance of the

following inspection program.’’

b) Section 4.5.Q.1.a - GL 90-09

wording ’’...based on the criteria of

Table 4.7.2 and the first inspection

interval determined using the criteria

shall be based upon the previous

inspection interval established by the

requirements in effect before

Amendment (*). ‘‘Proposed Technical

Specification wording ’’...based on the

criteria provided in Table 4.5.1.’’

c) Section 4.5.Q.1.b - GL 90-09

wording ’’...All snubbers found

connected to an inoperable common

hydraulic fluid reservoir shall be
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39441

counted as unacceptable for

determining the next inspection

interval.’’

Proposed Technical Specification

deletes this sentence.

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 amendment would

revise the basis for the snubber visual

inspection to be consistent with the

bases described in Generic Letter 90-09.

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change does not affect the

probability of occurrence nor does it affect

the consequences of an accident previously

evaluated as the requested visual inspection

interval has been determined generically to

be a safe and acceptable alternative to the

existing visual inspection requirements as

documented by the NRC in Generic Letter 90-

09. With the completion of over 25 years of

operating experience and only detecting one

visual inspection failure, GPU Nuclear agrees

that the existing intervals are overly

conservative and can be extended to those

described in the generic letter.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

As the requested change deals only with

the frequency of visual inspection and not

with the content, scope, or acceptance

criteria of the inspection, no new or different

type of accident has been created.

3. Involve a significant reduction in the

margin of safety.

The margin of safety as defined in the

bases of the Technical Specifications is not

reduced as the requested requirements

provide the same degree of confidence in

snubber operability at the existing

requirements.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Phillip F.

McKee

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: May 31,

1995

Description of amendment request:

The proposed amendment would

modify (by relocation to the Technical

Requirements Manual) Technical

Specification (TS) 3/4.1.2.1, Boration

Systems/Flow Paths - Shutdown, TS 3/

4.1.2.2, Boration Systems/Flow Paths -

Operating, TS 3/4.1.2.3, Charging

Pumps - Shutdown, TS 3/4.1.2.4,

Charging Pumps - Operating, TS 3/

4.1.2.5, Borated Water Sources -

Shutdown, TS 3/4.1.2.6, Borated Water

Sources - Operating, TS 3/4.4.2.1, Safety

Valves - Shutdown, and the associated

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. Does the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed change to the subject

Technical Specifications is of an

administrative nature in that the subject

Technical Specifications and Bases will be

relocated in their entirety to the Technical

Requirements Manual. Future changes to the

relocated requirements will be in accordance

with 10CFR50.59 and approved station

procedures.

Whether the listed Technical

Specifications and Bases are located in

Technical Specifications or the Technical

Requirements Manual has no effect on the

probability or consequences of an accident

previously evaluated.

The proposed change does not alter the

assumptions previously made in the listed

Technical Specifications. The proposed

change allows the Commission and the South

Texas Project more effective use of personnel

resources to control requirements that meet

the four Criteria in the Final Policy

Statement. The proposed change will not

change the dose to workers.

Since the probability of an accident is

unaffected by administratively relocating the

subject Technical Specification, and the

doses are not affected and do not exceed

acceptance limits, the proposed change does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed change to the subject

Technical Specifications is of an

administrative nature in that the subject

Technical Specifications and Bases will be

relocated in their entirety to the Technical

Requirements Manual. Future changes to the

relocated requirements will be in accordance

with 10CFR 50.59 and approved station

procedures. Whether the listed Technical

Specifications and Bases are located in

Technical Specifications or the Technical

Requirements Manual has no effect on any

previously evaluated accident. It does not

represent a change in the configuration or

operation of the plant and, therefore, does

not create the possibility of a new or different

type of accident from any accident

previously evaluated.

3. Does the proposed change involve a

significant reduction in the margin of safety?

The proposed change to the subject

Technical Specifications is of an

administrative nature in that the subject

Technical Specifications and Bases will be

relocated in their entirety to the Technical

Requirements Manual. Future changes to the

relocated requirements will be in accordance

with 10CFR50.59 and approved station

procedures. The margin of safety is not

reduced when the requirements are relocated

to a Licensee-controlled document because

the requirements to change a License Basis

Document via the 10CFR50.59 process ensure

the same questions concerning the margin of

safety required for license amendments are

asked. Therefore, this proposed change does

not significantly reduce the margin of 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 involves 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

Nebraska Public Power District, Docket

No. 50-298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: June 28,

1995

Description of amendment request:

The proposed amendment would revise

technical specifications related to the

standby liquid control (SLC) system.

The proposed changes include

increasing the required reactor pressure

vessel boron concentration and

modifying the SLC pump operability

testing surveillance frequency from

monthly to quarterly.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the
39442 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

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 current analysis requires the

SLC system to be capable of bringing the

reactor 3% delta - k subcritical assuming a

cold xenon free condition. The increase in

SLC storage tank boron concentration limits

will ensure this capability is maintained for

future reload cores using the same 3% delta

- k shutdown reactivity margin without

imposing restrictions in cycle exposure for

current and future anticipated core

configurations. The change in the

surveillance frequency for SLC pump

operability testing to once each three months

is in agreement with the ASME Code. The

relaxation of the testing interval for the SLC

pumps decreases pump degradation, and

eliminates an unnecessary burden on

personnel resources without compromising

plant safety. In addition, the administrative

changes only correct typographical and

editorial errors.

Since these proposed changes do not affect

precursors for any accident or transient

analyzed in Chapter 14 of the USAR, there

is no increase in the probability of any

accident previously evaluated. Furthermore,

since these changes will ensure the ability of

the SLC system to mitigate the consequences

of an accident for future anticipated core

designs, they do not involve a significant

increase in the consequences of any accident

previously evaluated.

2. The proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. The change in the SLC storage

tank boron concentration limits will ensure

that a cold xenon-free reload core can be

brought to a subcritical condition as

previously analyzed. The change in the

frequency of the SLC pump operability

testing to once each three months is in

agreement with the ASME Code. The

relaxation in the testing interval for the SLC

pumps decreases pump degradation, and

eliminates an unnecessary burden on

personnel resources without compromising

plant safety. In addition, the administrative

changes only correct typographical and

editorial errors.

These proposed changes do not affect the

design, function, or operation of the SLC or

any other system. Also, these changes do not

introduce any new modes of operation or

modify existing equipment design. Therefore,

they do not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The proposed changes will not create a

significant reduction in the margin of safety.

The proposed increase in the required boron

concentration in the reactor pressure vessel

will ensure the SLC system will be capable

of bringing a cold xenon-free reload core

subcritical while maintaining the 3% delta -

k shutdown reactivity margin as specified in

the previous operating cycle. The change in

the frequency of SLC pump operability

testing to once each three months is in

agreement with the ASME Code. The

relaxation in the testing interval for the SLC

pumps decreases pump degradation, and

eliminates an unnecessary burden on

personnel resources without compromising

plant safety. In fact, it increases SLC system

availability. In addition, the administrative

changes only correct typographical and

editorial errors. Therefore, it is concluded

that the requested changes do not create a

significant reduction in the existing margin

of safety as defined in the Technical

Specifications.

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: Auburn Public Library, 118

15th Street, Auburn, Nebraska 68305

Attorney for licensee: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus,

Nebraska 68602-0499

NRC Project Director: William D.

Beckner North Atlantic Energy Service

Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: June 16,

1995

Description of amendment request:

The proposed amendment would

change the minimum boron

concentration specified for the refueling

water storage tank (RWST) in Limiting

Condition for Operation (LCO) in

Technical Specification (TS) 3.1.2.5 and

would replace the minimum specified

concentration for boron with an

acceptable range of boron concentration

for the RWST and the accumulators in

the LCOs for TS 3.1.2.6, 3.5.1.1, and

3.5.4.

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. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below.

A. The changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated (10 CFR 50.92(c)(1)) because

the changes are proposed to assure that

the post-event shutdown margin

required by the Technical Specifications

will continue to be met and the

consequences of a boron dilution event

will remain as previously evaluated.

The changes do not affect the design or

manner of operation of any structure,

system, or component important to

safety.

B. The changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated (10 CFR 50.92(c)(2)) because

they do not affect the manner by which

the facility is operated and do not

involve a change to any structure,

system, or component important to

safety. The proposed changes merely

assure that station will be operated

within original design limits.

C. The changes do not involve a

significant reduction in a margin of

safety (10 CFR 50.92(c)(3)) because the

proposed changes merely assure that the

station will continue to be operated

within the original design limits.

Therefore, the acceptance criteria for

previously evaluated accidents will

continue to be met.

Based on this review, it appears that

the three standards of 10 CFR 50.92(c)

are satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Thomas Dignan,

Esquire, Ropes & Gray, One

International Place, Boston MA 02110-

2624.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company

(NNECO), Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: July 11,

1995

Description of amendment request:

The proposed amendment modifies

Technical Specification 3.5.F.7 to also

allow the use of pull-to-lock switches to

defeat the automatic initiation of the

emergency core cooling system (ECCS)

while in the refuel condition. The

proposed amendment also makes

administrative changes and makes

changes to the associated Bases section.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed change

in accordance with 10 CFR 50.92 and

concluded that the change does not involve

a significant hazards consideration (SHC).

The basis for this conclusion is that the three

criteria of 10 CFR 50.92(c) are not

compromised. The proposed change does not
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39443

involve an SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

This change to LCO [Limiting Condition

for Operation] 3.5.F.7(e) will allow an

alternative means of de-energizing power to

the selected ECCS pump motors during

refueling. The current

technical specification already allows these

motors to be de-energized. Use of the pullto-lock switches provides a safer method of

achieving this condition. The pull-to-lock

condition of the switches is annunciated in

the control room. Therefore, the switches

will not be inadvertently left in the pull-tolock position.

Deletion of the statement that the 4160 volt

supply breakers are racked in does not affect

the requirement of LCO 3.5.F.7 to ensure the

specified ECCS subsystems are OPERABLE.

Therefore, there is no change in the

probability or consequences of an accident

previously analyzed due to this change.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The use of an alternative means of deenergizing power from the selected ECCS

pump motors does not create a possibility of

a new or different kind of accident. Using the

control room pull-to-lock switch to disable

the pump motor circuit breaker has the same

effect on the ECCS pump as the removal of

the circuit breaker from the switchgear.

Deletion of the statement that the 4160 volt

supply breakers are racked in does not affect

the requirement of LCO 3.5.F.7 to ensure the

specified ECCS subsystems are OPERABLE.

3. Involve a significant reduction in the

margin of safety.

The proposed change to the Millstone Unit

No. 1 Technical Specifications does not

reduce the margin of safety. By using the

control room pull-to-lock switches to disable

the ECCS pump motors, instead of racking

out the pump motor circuit breakers, it is

possible to reenergize the ECCS pumps more

quickly in an emergency, should one occur.

The time savings can be translated into

added safety margin from a shutdown risk

perspective. The ability to disable and enable

the pumps from the control room, instead of

the switchgear area, also contributes to this

added safety margin.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company

(NNECO), Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: July 18,

1995

Description of amendment request:

The proposed amendment request will

add operability and surveillance

requirements for reactor pressure vessel

(RPV) overfill protection

instrumentation. The proposed

amendment will also add the associated

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:

NNECO has reviewed the proposed change

in accordance with 10 CFR 50.92 and

concluded that the change does not involve

a significant hazards consideration (SHC).

The basis for this conclusion is that the three

criteria of 10 CFR 50.92(c) are not

compromised. The proposed change does not

involve an SHC because the change would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The new LCO [Limiting Condition for

Operation] and surveillance requirements

ensure that the reactor high water level

feedwater pump trip instrumentation is

available. This technical specification change

does not involve the addition of new

equipment or logic. This change does not add

new surveillance requirements for the

instrumentation. This change simply

establishes requirements for the operation

and surveillance of

reactor high water level feedwater pump

trip instrumentation in the technical

specifications. The implementation of this

technical specification change will decrease

the likelihood of an RPV overfill. No other

postulated event is affected by the addition

of this instrumentation to the technical

specifications.

Thus, adding the proposed requirements to

the technical specifications will not increase

the probability or consequences of any

previously evaluated transients or accidents.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

No new failure modes are introduced by

the addition of the reactor high water level

feedwater pump trip instrumentation LCO

and surveillance requirements. Modifying the

technical specifications to formally add

surveillance requirements already being

performed in accordance with plant

procedures will not modify plant response to

any operational or transient event. Increasing

the surveillance interval of the LITS [level

indicating transmitter switches] from annual

to once per operating cycle will not

significantly affect reliability. Ensuring the

operability of installed instrumentation does

not add new or different kinds of accidents.

Therefore, the new LCO and surveillance

requirements do not create the possibility of

a new or different kind of accident.

3. Involve a significant reduction in the

margin of safety.

The surveillance requirements being added

in this change are consistent with current

surveillances being performed for this

instrumentation, with the exception that the

LITS are currently calibrated on an annual

rather than operating cycle basis. These

surveillance and shutdown requirements

ensure that protection from RPV overfill is

maintained as assumed in the safety

analyses.

Therefore, there is no impact on the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of amendment request: July 7,

1995

Description of amendment request:

The proposed change to technical

specification 3/4.7.6 is being made to: 1)

increase the allowable control room air

conditioning (CRAC) system in-leakage

from 100 cubic feet per minute (cfm) to

130 cfm; 2) provide a more conservative

value for the maximum differential

pressure across the high efficiency

particulate air (HEPA) filters and

charcoal adsorbers; 3) clarify that when

the CRAC system is shifted to

‘‘recirculation,’’ this will be performed

from the normal mode; and 4) modify

the corresponding basis to reflect the

above changes and to note that there are

certain infrequent situations during

which the control room emergency

ventilation system (CREVS) will not

automatically operate.

Basis for proposed no significant hazards consideration determination: As
39444 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration (SHC), which is presented

below:

...The proposed changes do not involve an

SHC because the changes will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously analyzed.

The CRAC system in the recirculation

mode is used to mitigate the effects of an

accident. Surveillance Requirement

4.7.6.1.e.2 has been modified to clarify that

the system will automatically switch from

the normal mode into a recirculation mode.

This change and the proposed modifications

to the acceptance criterion for the differential

pressure across the HEPA filters and charcoal

adsorbers and the increase in the control

room in-leakage have no [e]ffect on the

probability of an accident previously

evaluated. The consequences of the accidents

that have been previously evaluated have

been reviewed to determine the impact of

these proposed modifications. The increase

in the in-leakage will affect the results of

previously generated accident analysis. The

accidents evaluated, namely the Millstone

Unit No. 1 MSLB [main steam line break] and

LOCA [loss-of-coolant accident], Millstone

Unit No. 2 LOCA, both high and low wind

speed case, and Millstone Unit No. 3 LOCA

have been reviewed. The Millstone Unit No.

1 LOCA doses to the Millstone Unit No. 2

control room were qualitatively determined

to be bounded by the Millstone Unit No. 2

LOCA cases. Therefore the Millstone Unit

No. 1 LOCA was not performed. The

remaining accidents were performed. The

resultant doses are nearly identical to the

existing doses found in the Millstone Unit

No. 2 Final Safety Analysis Report and are

all within the regulatory limits. To perform

these revised control room dose calculations,

NNECO used certain new assumptions which

NNECO believes better model the control

room and the effects the accident will have

on the control room. The most significant

change with the assumptions is the use of

ICRP 30 in lieu of Regulatory Guide 1.109,

Revision 1 for iodine dose conversion factors.

The NRC has used ICRP 30 over the past 5

years for other applications and its use in this

instance is appropriate.

The change in the acceptance criterion for

the differential pressure across the HEPA

filter and charcoal adsorbers is a conservative

modification in that the value given is a plant

specific value and will be more indicative of

blocked or clogged filters in actual plant

conditions. These proposed changes do not

have any negative impact on the

consequences of any accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed modifications to

Surveillance Requirement 4.7.6.1 will clarify

a portion of a surveillance requirement and

will modify the differential pressure across

the HEPA filters and the charcoal adsorbers.

These changes will not create the possibility

of a new or different kind of accident from

any previously evaluated. The increase in the

allowable control room in-leakage value from

it[s] current level of 100 cfm to its new value

of 130 cfm also does not create the possibility

of a new or different kind of accident. The

CRAC system is used to mitigate the

consequences of an accident.

3.Involve a significant reduction in the

margin of safety.

The proposed modifications do not

decrease the margin of safety provided. Using

the new accident assumptions, the limiting

accidents were re-calculated to determine the

impact on the Millstone Unit No. 2 control

room. These values are similar to the values

found in the Millstone Unit No. 2 Final

Safety Analysis Report and the Millstone

Unit No. 2 Safety Evaluation Report and are

within the regulatory limits established for

the control room operators. Since the recalculated doses have been shown to be

within limits, it has been concluded that

there is no reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resource Center,

Three Rivers Community-Technical

College, Thames Valley Campus, 574

New London Turnpike, Norwich, CT

06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: June 8,

1995

Description of amendment request:

The Millstone Unit No. 3 Technical

Specification Section 3/4.8.4.3 requires

removal of electrical power to the safety

injection accumulator isolation valves

in Modes 1, 2, 3, and 4 in order to

protect the containment electrical

penetrations and penetration

conductors. Bases Section 3/4.8.4 states

that containment electrical penetrations

and penetration conductors are

protected by either deenergizing circuits

not required during normal plant

operation (Modes 1 through 4) or by

demonstrating the operability of

primary and backup overcurrent

protection circuit breakers during

performance of periodic surveillances. It

is proposed that Section 3/4.8.4.3 will

be deleted since the containment

electrical penetration and penetration

conductors for these circuits are

protected by primary and backup

penetration circuit breakers which are

demonstrated to be operable by periodic

surveillance testing.

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 (SHC), which is presented

below:

The proposed changes do not involve an

SHC because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The revised Technical Specification

Section 3.5.1 requirements will provide

guidance to ensure that power to the

accumulator isolation valves is removed

when the accumulators are required to be

operable and will clarify these requirements.

Removal of the electrical penetration

protection requirements of Section 3/4.8.4.3

is justified since Section 3/4.8.4.1

(Containment Penetration Conductor

Overcurrent Protective Devices) will provide

guidance to ensure that two breakers in series

protect the electrical penetrations and

penetration conductors against an

overcurrent condition and the single failure

of a circuit breaker. The two breakers in

series also protect the Class 1E buses against

a variety of overcurrent conditions including

electrical faults which may be introduced

due to the possible submergence of the

accumulator isolation valves during a LOCA

[loss-of-coolant accident].

Therefore, the proposed amendment will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The amended Technical Specification

Section 3.5.1 requirements will provide

guidance to ensure that the accumulator

isolation valves are deenergized when the

accumulators are required to be operable.

Deletion of the Technical Specifications

Section 3.5.1 requires that electrical power to

the safety injection accumulator isolation

valves (3SIL*MV8808A, B, C, D) be removed

for the accumulators to be operable. This

requirement prevents the inadvertent closure

of these isolation valves which would block

the safety function of the accumulators.

Section 4.5.1.c requires demonstrating

accumulator operability by ‘‘At least once per

31 days when the RCS [reactor coolant

system] pressure is above 1000 psig by

verifying that power to the isolation valve

operator is disconnected by removal of the

breaker from the circuit.’’ The surveillance

requirements for verifying removal of power

to the accumulator isolation valves for

Section 4.5.1.c will be changed to ‘‘At least

once per 31 days when the RCS pressure is

above 1000 psig by verifying that the

associated circuit breakers are locked in a

deenergized position or removed.’’

The proposed change will clarify

requirements for securing these breakers in
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39445

the off (tripped) position in the applicable

modes. In addition, index page xi has been

revised to reflect the deletion of Section 3/

4.8.4.3. Attachments 1 and 2 provide the

mark-up and retyped pages of the Millstone

Unit No. 3 Technical Specifications,

respectively and reflect the currently issued

version of the pages.

Millstone Unit No. 3 Technical

Specifications Section 3/4.8.4.3 will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated since two breakers in

series protect against an overcurrent

condition and a single failure of a circuit

breaker. The proposed amendment will not

result in physical plant changes and there are

no new credible failure modes. Therefore, the

proposed amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

The revised Technical Specification

Section 3.5.1 will require that the

accumulator isolation valves have their

power deenergized when the accumulators

are required to be operable. This requirement

will maintain accumulator operability by

assuring the accumulator isolation valves

remain open.

The removal of the Millstone Unit No. 3

Technical Specification Section 3/4.8.4.3 is

safe since redundant circuit breakers in series

for the accumulator isolation valves will

provide assurance that the electrical

penetration and penetration conductors are

protected against overcurrent conditions.

This will provide assurance that the

containment boundary is intact.

The proposed amendment will not

adversely impact the physical protective

boundaries (fuel matrix/cladding, RCS

pressure boundary and containment) and

therefore 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: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: June 9,

1995

Description of amendment request:

The proposed amendment relocates

Surveillance Requirement 4.6.6.1.d.3 for

attaining a negative pressure in the

secondary containment to Specification

3.6.6.2, Secondary Containment. The

Action Statement of Section 3.6.6.1 is

revised to decouple Sections 3.6.6.1 and

3.6.6.2. In addition, Definition 1.12,

‘‘Secondary Containment Boundary’’ is

deleted and included in the Bases

Section 3/4.6.6, Secondary

Containment. Bases Section 3/4.6.6.2,

Secondary Containment is expanded

using the guidance of the improved

standard technical specifications (STS)

for Westinghouse plants (NUREG-1431).

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 (SHC), which is presented

below:

The proposed changes do not involve an

SHC because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes to LCO [limiting

condition for operation] 3.6.1.2, LCO 3.6.6.1

and LCO 3.6.6.2 Action Statements,

relocation of Surveillance Requirement

4.6.6.1.d.3 to Specification 3.6.6.2, changes to

Bases Section 3/4.6.6.1, 3/4.6.6.2, and 3/

4.6.6.3, and deletion of Definition 1.12 will

resolve the conflict that currently exists

between Specifications 3.6.6.1 and 3.6.6.2.

Specifically, the requirement to establish and

maintain a negative pressure in the

secondary containment boundary included in

Specification 3.6.6.1 belongs to Specification

3.6.6.2. In the event Secondary Containment

operability is not maintained, the Action

Statement for LCO 3.6.6.2 requires that

Secondary Containment operability must be

restored within 24 hours. Twenty-four hours

is a reasonable completion time considering

the limited leakage design of containment

and the low probability of a DBA [design

basis accident] occurring during this time

period. Therefore, it is considered that there

exists no loss of safety function. The

proposed changes do not modify the LCO or

surveillance acceptance criterion, nor do they

change the frequency of the surveillances.

The proposed changes do not involve any

physical changes to the plant, do not alter the

way any structure, system, or component

functions. Therefore, the structures, systems,

or components will perform their intended

function when called upon. The proposed

changes do not affect the probability of any

previously evaluated accident. Additionally,

the proposed changes are consistent with the

new, improved STS for Westinghouse plants

(NUREG-1431).

Based on the above, the proposed changes

do not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes do not make any

physical or operational changes to existing

plant structures, systems, or components.

The proposed changes do not introduce any

new failure modes. The proposed changes

simply resolve a conflict which currently

exists between Specifications 3.6.6.1 and

3.6.6.2. Thus, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed changes do not have any

adverse impact on the accident analyses.

Also, the proposed changes resolve a conflict

which currently exists between

Specifications 3.6.6.1 and 3.6.6.2. The

structures, systems, or components covered

under Specifications 3.6.6.1 and 3.6.6.2 will

performed [sic] their intended safety function

when called upon.

Based on the above, there is no significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Northeast Nuclear Energy Company, et

al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of amendment request: June 20,

1995

Description of amendment request:

The proposed amendment relocates the

applicable requirements of Specification

3.6.3 for the main steam line isolation

valves (MSIVs) to Specification 3.7.1.5,

‘‘Main Steam Line Isolation Valves.’’ In

addition, the Applicability section of

Specification 3.7.1.5 is revised to

indicate that Specification 3.7.1.5 is

applicable in Mode 1 and in Modes 2,

3 and 4, except where all MSIVs are

closed and deactivated (i.e., in Modes 2,

3, and 4, Specification 3.7.1.5 is

applicable only if the MSIVs are open).

Also, the Action Statement for the

Limiting Condition for Operation (LCO)

3.7.1.5 has been revised using the

guidance of the improved standard

technical specifications (STS) for

Westinghouse plants (NUREG-1431).
39446 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

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 (SHC), which is presented

below:

The proposed changes do not involve an

SHC because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes to the Applicability

section, Action Statements, and Surveillance

Requirements of Specification 3.7.1.5 and the

proposed changes to Specification 3.6.3

preserve the assumptions in the existing

safety analysis. The proposed changes to the

Applicability Section of Specification 3.7.1.5

will require the MSIVs to be operable in

Mode 1 and in Modes 2, 3, and 4, except

when closed and deactivated. The closure of

the MSIVs in Modes 2, 3, or 4 is acceptable

because when they are closed, they are

already performing their safety function.

Since the MSIV closure time has not been

changed, there is no adverse impact on the

accidents previously evaluated.

The proposed changes do not involve any

physical changes to the plant, and do not

alter the way any structure, system, or

component functions. Therefore, the

proposed changes do not affect the

probability of any previously evaluated

accident. Additionally, the proposed changes

are consistent with the new, improved STS

for Westinghouse plants (NUREG-1431).

Based on the above, the proposed changes

do not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes do not make any

physical changes to existing plant structures,

systems, or components. When the MSIVs are

closed and deactivated, they are already in

the safe position; therefore, the proposed

changes do not introduce a new failure mode.

Additionally, the MSIV closure time (i.e.,

surveillance acceptance criterion) is not

changed. The purpose of the surveillance is

to ensure that the MSIVs can perform their

safety function, and this requirement is

preserved.

Thus, the proposed changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed changes do not revise the

closure time of the MSIVs. This provides

assurance that the MSIVs will perform their

design safety function to mitigate the

consequences of an accident. In addition,

when they are closed in Modes 2, 3, and 4,

they are already performing their safety

function. Therefore, there is no significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Ms. L. M.

Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company,

Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F.

McKee

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station,Unit

No. 1, Washington County, Nebraska

Date of amendment request: June 26,

1995

Description of amendment request:

This proposed amendment would revise

Technical Specification 2.3 to extend

the allowed outage time (AOT) from 24

hours to 7 days for an inoperable lowpressure safety injection pump. This

amendment request is a collaborative

effort of participating Combustion

Engineering Owners Group members

and is based on an integrated

assessment of plant operations and

deterministic and probabilistic analyses.

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 low pressure safety injection (LPSI)

system is part of the emergency core cooling

system. Inoperable LPSI components are not

accident initiators in any accident previously

evaluated. Therefore, these changes do not

involve an increase in the probability of an

accident previously evaluated.

The LPSI system is primarily designed to

mitigate the consequences of a large loss of

coolant accident (LOCA). These proposed

changes do not affect any of the assumptions

in the deterministic LOCA analysis. Hence

the consequences of accidents previously

evaluated do not change.

In order to fully evaluate the LPSI allowed

outage time (AOT) extension, probabilistic

safety analysis (PSA) methods were utilized.

The results of these analyses show no

significant increase in the core damage

frequency. As a result, there would be no

significant increase in the consequences of an

accident previously evaluated. These

analyses are detailed in CE NPSD-995,

‘‘Combustion Engineering Owners Group

Joint Applications Report for Low Pressure

Safety Injection System AOT Extension.’’

The CEOG report reviewed the risk factors

that are impacted by extending the AOT for

a single LPSI pump from 24 hours to seven

(7) days, and demonstrates that the increase

in risk is negligible. In order to perform a

more complete assessment of the overall

change in risk, an accounting for avoided

risks associated with reducing power and

going to hot or cold shutdown was also

considered. This ‘‘transition risk’’ is

important in understanding the trade-off

between the risk of shutting down the plant

compared with restoring a LPSI pump to

operability while at power.

In assessing overall plant risk, the risk

avoided based on LPSI system maintenance

while in cold shutdown must also be

considered. Every time the plant is placed in

cold shutdown, the LPSI system is required

for decay heat removal when in the

shutdown cooling mode of operation.

Maintenance performed on the LPSI system

during shutdown cooling operations may add

to the risk of a loss of shutdown cooling

event. Therefore, performing LPSI system

maintenance with the unit on-line, when the

LPSI system is not normally in demand,

represents a decrease in shutdown risk.

The CE study concluded that the change in

core damage frequency due to increasing the

LPSI AOT from 24 hours to seven (7) days

is insignificant. Additionally, when the

reduction in transition and shutdown risks

are considered, it can be shown that there is

an overall reduction in plant risk. Thus, it is

the conclusion of the study that the overall

plant impact will either be risk beneficial or

risk neutral.

Therefore, the proposed changes would not

increase the probability or consequences of

an accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There will be no physical alterations to the

plant configuration, changes to setpoint

values, or changes to the implementation of

setpoints or limits as a result of the proposed

changes. Therefore, the proposed changes do

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.

These proposed changes do not affect the

limiting conditions for operation or their

bases used in the deterministic analyses to

establish the margin of safety. PSA

evaluations were used to evaluate this

change. These evaluations demonstrate that

the changes are either risk neutral or risk

beneficial. These evaluations are detailed in

CE NPSD-995. Therefore, the proposed

changes do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39447

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: James R.

Curtiss, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-

3502

NRC Project Director: William H.

Bateman

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station,Unit

No. 1, Washington County, Nebraska

Date of amendment request: June 27,

1995

Description of amendment request:

This proposed amendment would revise

Technical Specification 2.2 on the

chemical and volume control system to

reformat, clarify the requirements, and

be more consistent with Combustion

Engineering Standard Technical

Specifications (STS) as presented in

NUREG-0212, Revision 2.

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 do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated. The proposed changes incorporate

required actions, restrictions, and

surveillance requirements for the Chemical

and Volume Control System (CVCS) similar

to Combustion Engineering Standard

Technical Specifications (NUREG-0212

Revision 2).

Technical Specification (TS) 2.2(1)

specifies the requirements for borated water

sources and flow paths when the reactor is

subcritical and fuel is in the reactor. In order

for a flow path to be operable, a charging or

high pressure safety injection pump is

required to be operable to inject the boric

acid solution into the Reactor Coolant

System. Currently this specification does not

state any operability requirements for boric

acid transfer pumps, charging pumps or high

pressure safety injection pumps. In addition,

this specification does not state any required

actions to be taken if the borated water

source or flow path is not operable.

Therefore, the proposed changes

incorporate requirements for the CVCS

during shutdown into separate Limiting

Conditions for Operations (LCOs) that will

address the requirements for borated water

sources, boric acid flow paths, charging

pumps, and boric acid transfer pumps.

The proposed changes delete operability

and surveillance requirements for level

instrumentation on the boric acid storage

tanks. Level instrumentation by itself does

not fulfill a safety function. The proposed

changes will still require verification of tank

level.

Additionally, level instrumentation on the

boric acid storage tanks does not meet any of

the four criteria for inclusion into Technical

Specifications as presented in the Final

Policy Statement on Technical Specifications

Improvements. This instrumentation is not

installed instrumentation used to detect a

significant degradation of the RCS boundary,

a design feature or operating restriction that

is an initial condition of a Design Basis

Accident, a component that is part of the

primary success path or actuates to mitigate

a DBA, nor is it a component that has been

shown to be significant to public health and

safety. Therefore, testing and maintenance of

the level instrumentation will be controlled

outside of the TS.

TS 2.2(3) specifies the Modifications of

Minimum Requirements that are allowed

during Power Operation. This specification is

inconsistent with TS 2.2(2) which states the

minimum requirements and is incomplete as

it does not address components during

Modes 3, 4, and 5. The proposed changes

incorporates consistent allowed outage times

for the various components, and additional

required actions for component inoperability

during Modes 4 and 5 when fuel is in the

reactor.

The proposed changes incorporate

additional operability requirements for the

CVCS and required actions to be taken for

CVCS component inoperability during Modes

4 and 5 when fuel is in the reactor. The

proposed changes delete inconsistencies and

clarify operability requirements for the CVCS

whenever the reactor coolant temperature

(Tcold) is greater than or equal to 210 degrees

F, and ensures that operation of the system

is consistent with its design bases. The

proposed changes also revise the allowed

outage time for CVCS components from 24

hours to 72 hours based on Standard

Technical Specifications. This change is

insignificant based on the FCS plant specific

probabilistic risk assessment. Therefore, the

proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There will be no physical alterations to the

plant configuration, changes to setpoint

values, or changes to the implementation of

setpoints or limits as a result of this proposed

change. No new modes of operation are

proposed. Therefore, the proposed change

does not create the possibility of a new or

different kind of accident from any

previously analyzed.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes incorporate

additional operability requirements, delete

inconsistencies, and clarify operability

requirements for the CVCS to ensure that

operation of the system is consistent with its

design bases. Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Attorney for licensee: James R.

Curtiss, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-

3502

NRC Project Director: William H.

Bateman

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station,Unit

No. 1, Washington County, Nebraska

Date of amendment request: July 11,

1995

Description of amendment request:

The proposed amendment would allow

up to 24 hours to restore Safety Injection

Tank (SIT) operability if the SIT is

inoperable due to level and/or pressure

outside prescribed limits or if the

associated isolation valve is in other

than the full open position. The

proposed change would also allow up to

72 hours to restore SIT operability if the

SIT is inoperable due to boron

concentration outside prescribed limits.

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 safety injection tanks (SITs) are

passive components in the emergency core

cooling system. The SITs are not an accident

initiator in any accident previously

evaluated. Therefore, this change does not

involve an increase in the probability of an

accident previously evaluated.

SITs were designed to mitigate the

consequences of a loss of coolant accident

(LOCA). These proposed changes do not

affect any of the assumptions used in

deterministic LOCA analysis. Hence the

consequences of accidents previously

evaluated do not change.

In order to fully evaluate the affect of the

SIT allowable outage time (AOT) extension,

probabilistic safety analysis (PSA) methods

were utilized. The results of these analyses

show no significant increase in the core

damage frequency. As a result, there would

be no significant increase in the

consequences of an accident previously

evaluated. These analyses are detailed in CE

NPSD-994, ‘‘Combustion Engineering Owners

Group Joint Applications Report for Safety

Injection Tank AOT/STI Extension.’’

The AOT extension based upon boron

concentration outside the prescribed limits
39448 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

does not involve a significant increase in the

consequences of an accident as evaluated and

approved by the NRC in NUREG-1432,

‘‘Standard Technical Specifications for

Combustion Engineering Plants.’’ This

proposed change is applicable to FCS.

Therefore, the proposed changes would not

increase the probability or consequences of

any accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There will be no physical alterations to the

plant configuration, changes to setpoint

values, or changes to the implementation of

setpoints or limits as a result of these

proposed changes. Therefore, the proposed

changes do 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 do not affect the

limiting conditions for operation or their

bases that are used in the deterministic

analyses to establish the margin of safety.

PSA evaluations were used to evaluate these

changes. These evaluations demonstrated

that the changes are either risk neutral or risk

beneficial. These evaluations are detailed in

CE NPSD-994. Therefore, the proposed

changes do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment 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: James R.

Curtiss, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-

3502

NRC Project Director: William H.

Bateman

Philadelphia Electric Company, Docket

No. 50-353, Limerick Generating

Station, Unit 2, Montgomery County,

Pennsylvania

Date of amendment request: June 23,

1995

Description of amendment request:

This Technical Specifications (TS)

Change Request involves a one-time

(i.e., temporary) change affecting the

Allowed Outage Time (AOT) for the

Emergency Service Water (ESW)

System; Residual Heat Removal Service

Water (RHRSW) System; the

Suppression Pool Cooling, the

Suppression Pool Spray, and Low

Pressure Coolant Injection (LPCI) modes

of the Residual Heat Removal (RHR)

System; and Core Spray System to be

extended from 3 and 7 days to 14 days

during the Limerick Generating Station

(LGS), Unit 1, sixth refueling outage

scheduled to begin January, 1996. This

proposed extended AOT will 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 ’A’ 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-somuch 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

[Safety Analysis Report] will not be

increased. With the ’A’ 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 ’A’ 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 affects on the

systems. It was then determined how the

adverse affects would impact each system’s

design basis and overall plant safety. The

consequences of any postulated accidents

occurring on Unit 2 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 ’A’ ESW and RHRSW loops

and the affected equipment can be out-ofservice. 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 safetyrelated equipment is available for response to

all accident conditions and that sufficient

decay heat removal capability is available for

a LOCA/LOOP [Loss-of-Coolant Accident/

Loss-of-Offsite Power] 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
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39449

during the extended AOT and based on the

following discussion of minimum ECCS

[Emergency Core Cooling System]/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 the UFSAR [Updated Final

SAR] reference documents NEDO-24708A

and NEDC-30936-A. These documents

describe the minimum requirements to

successfully terminate a transient or LOCA

initiating event (with scram), assuming

multiple failures with realistic conditions

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 [Automatic

Depressurization System], 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 [High

Pressure Coolant Injection], 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. A Special

Procedure will be written to ensure the

operability of specified components and that

other appropriate compensatory measures are

implemented.

Compensatory 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 compensatory

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. The Core Damage

Frequency (CDF) increased by 3.14x10-6,

from 5.11x10-6 /reactor-year to 8.25x10-6/

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: John F Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request:

September 29, 1994

Description of amendment request:

The proposed Technical Specification

changes represent revisions to Sections

3/4.3.7.2 ‘‘Seismic Monitoring

Instrumentation’’ and 3/4.3.7.3

‘‘Meteorological Instrumentation.’’ The

proposed revisions remove the

requirements from the Technical

Specifications and relocates the

appropriate descriptive information and

testing requirements to the Hope Creek

Updated Final Safety Analysis Report

(UFSAR).

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. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed changes involve no

hardware changes, no changes to the

operation of any systems or components, and

no changes to existing structures. Neither the

relocation of the seismic/meteorological

specifications to the UFSAR nor the

elimination of the Special Report

requirements represent changes that affect

plant safety or alter existing accident

analyses.

2. Will not create the possibility of a new

or different kind of accident from any

previously evaluated.

The proposed changes are procedural in

nature concerning the operability and

surveillance of instrumentation that are not

safety related and will not impact the

operation of any plant safety related

component or equipment. Therefore, these

changes will not create a new or unevaluated

accident or operating condition.

3. Will not involve a significant reduction

in a margin of safety.

In accordance with the guidance provided

by the NRC regarding the improvement of

Technical Specifications, SECY-93-067, the

proposed changes relocate the seismic and

meteorological instrumentation portions of

the Technical Specifications, with the

exception of the Special Report requirements,

to the UFSAR. These instruments are not

safety related and do not have any associated

safety margins which could be affected by

this change.

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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request:

November 23, 1994

Description of amendment request:

The proposed changes to the Technical

Specifications (TSs) would revise TS

4.8.2.1, ‘‘Electrical Power Systems D. C.

Sources, Surveillance Requirements,’’

and associated Bases Section B 3/4.8.2.

The proposed changes would (1)

increase the terminal voltage acceptance

criteria for the battery discharge test

from 106 to 108 VDC, (2) delete a ‘‘one

time only’’ test that is no longer

applicable, (3) delete the battery load

profile from the TS, and (4) revise TS

Table 4.8.2.1-1, ‘‘Battery Surveillance

Requirements,’’ to agree more closely
39450 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

with the BWR4 Standard Technical

Specification format.

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:

....will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed changes restore the

conservatism to the battery voltage

requirements by raising the minimum

acceptable terminal voltage for the 125 VDC

system in order to support proper operation

of the connected loads. This change will

cause no change in the probability of any

accident and will, by providing increased

support for connected loads, provide

assurance [that] the consequences of

previously evaluated accidents remain

within limits. Removal of the load profile

table does not affect the surveillance test

loading which is contained in the station

procedures. The (*) footnote deletion is

purely editorial and has no safety bearing.

Table changes agree with the format and

wording of the improved BWR4 Standard

Technical Specifications.

2....will not create the possibility of a new

or different kind of accident from any

previously evaluated.

The revision of the battery sizing

calculations did not change the design base

requirement to supply the designed load for

a duty cycle of 4-hours. The proposed change

to the minimum acceptable battery terminal

voltage for the 125 VDC system ensures

proper voltages at the battery loads. No other

changes to the physical plant or to the

manner in which it is operated are caused by

the proposed amendment; therefore, there is

no new or different kind of accident created

by this change.

3....will not involve a significant reduction

in a margin of safety.

The revision of the battery sizing

calculations did not change the design base

requirement to supply the designed load for

a duty cycle of 4-hours; however, battery

capacity sizing parameter of end cell voltage

was changed to a more conservative value to

account for minimum load voltage

requirements. Load profiles for these

batteries were slightly modified to

incorporate more precise yet conservative

load current values. These batteries were

evaluated using a 25% additional capacity

margin for aging as required by IEEE-450. In

addition, the batteries have a design margin

of 5 to 10% for load growth and/or less than

optimum operating condition of the battery;

thereby, maintaining safety margins.

Additionally, changes are comparable to the

format and ACTIONS of the improved BWR4

STS. Permitting 31 days to restore a battery

to within CATEGORY A and/or B limits per

the improved BWR4 STS does not involve a

reduction in any margin of safety since the

battery, in Category C, remains operable, as

discussed in the BASES.

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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request:

November 28, 1994

Description of amendment request:

The proposed Technical Specification

(TS) revisions provide as follows: (1)

The setpoints and allowable values for

the Average Power Range Monitor

(APRM) flow-biased upscale scram/

control rod block would be modified to

improve operating margin in the

Extended Load Line Limit Analysis

(ELLLA) region; (2) The proposed

changes to the Rod Block Monitor

(RBM) trip function would transfer

control of the setpoint and allowable

value for the RBM - upscale rod block

to the Core Operating Limits Report

(COLR); (3) For the Reactor Coolant

System (RCS) recirculation flow upscale

trip function, the proposed changes

would revise the trip setpoint and

allowable value to reflect 105% of rated

core flow.

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.

A. Changes to APRM Flow-Biased Scram/

Control Rod Block

The proposed changes to the Average

Power Range Monitor (APRM) flow-biased

scram/control rod block setpoints and

allowable values were evaluated using NRC

approved procedures and methods. The

results of this evaluation are demonstrated in

NEDC-31487. Application of this change in

APRM flow-biased scram/control rod block

setpoints and allowable values to Reload 5/

Cycle 6 is confirmed in General Electric

Document No. 23A7219.

Analysis presented in NEDC-31487

demonstrate that performance in the ELLLA

region is within design limits for

overpressure protection, stability, loss-ofcoolant, containment, reactor internals, flowinduced vibration, and reactor internal

pressure difference. Impact of ELLLA

operation on anticipated transients without

scram is evaluated in Section 7.6.1.7.2 of the

UFSAR. Application of ELLLA region

extension to Reload 5/Cycle 6 has been

confirmed in GE Document No. 23A7219.

Because operation with the APRM flowbiased scram/control rod block setpoints and

allowable values is within the bases reviewed

and approved by the NRC in the UFSAR

[Updated Final Safety Analysis Report], this

change does not significantly increase the

possibility or consequences of an accident

previously evaluated.

B. Transfer of RBM Setpoint Control to the

COLR

The proposed changes would transfer

control of the setpoint and allowable value

for the rod block monitor (RBM) - Upscale

rod block to the Core Operating Limits Report

(COLR). Technical Specification 6.9.1.9,

‘‘Core Operating Limits Report,’’ requires that

the analytical methods used to determine

core operating limits be those previously

reviewed and approved by the NRC and that

the core operating limits be determined such

that all applicable limits of the safety

analysis are met.

The setpoint and allowable value

incorporate a controlling value which will be

specified in the COLR and noted as such by

reference in the Technical Specifications.

Therefore, the setpoint and allowable value

would continue to be controlled in a manner

that would ensure that safety analysis limits

are met and implementation of the proposed

changes would not reduce the level of

assurance provided by the existing Technical

Specifications. Based upon the above

information, we conclude that

implementation of the proposed change

would not significantly increase the

probability or consequences of an accident

previously evaluated.

C. RCS Recirculation Flow Revisions

The original analysis used to support

operation up to 105% of rated core flow is

contained in NEDC-31487. NEDC-31487

addresses the full range of transient and

accident events associated with operation up

to 105% of rated core flow. The affects of

operation with the revised RCS recirculation

flow upscale trip setpoint and allowable

value are bounded by the analysis presented

in NEDC-31487.

In addition, cycle specific analysis

performed for Reload 5/Cycle 6, have

incorporated the assumption of operation up

to 105% of rated core flow and have

confirmed that operation is within allowable

design limits.

Based on the above information, we

conclude that the proposed change would

not significantly increase the probability or

consequences of an accident previously

evaluated.

2. Would not create the possibility of

a new or different kind of accident from

any accident previously evaluated.

A. Changes to APRM Flow-Biased Scram/

Control Rod Block

The proposed changes to the APRM flowbiased scram/control rod block setpoints and

allowable values would not alter the function

of the APRM system nor involve any type of
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39451

plant modification. In addition, operation

with the revised APRM flow-biased scram/

control rod block setpoints and allowable

values would not create any new operating

modes, accident scenarios, equipment failure

modes, or fission product release paths.

Based upon the above information, we

conclude that the proposed changes would

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

B. Transfer of RBM Setpoint Control to the

COLR

The proposed transfer of control of the

RBM setpoint and allowable value to the

COLR would not alter the function of the

RBM system nor involve any type of plant

modification. In addition, operation with the

revised setpoint and allowable value would

not create any new operating modes, accident

scenarios, equipment failure modes, or

fission product release paths. Based upon the

above information, we conclude that the

proposed changes would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

C. RCS Recirculation Flow Revisions

The proposed changes would not alter the

function of the RCS recirculation flow

upscale trip function nor involve any type of

plant modification. In addition, operation

with the revised RCS recirculation flow

upscale trip setpoint and allowable value

would not create any new operating modes,

accident scenarios, equipment failure modes,

or fission product release paths. Based upon

the above information, we conclude that the

proposed changes would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Would not involve a significant

reduction in a margin of safety.

A. Changes to APRM Flow-Biased Scram/

Control Rod Block

The Bases for Hope Creek Technical

Specification 2.2.1 state that the APRM

setpoints were selected to provide adequate

margin for the safety limits while allowing

operating margins that reduce the possibility

of unnecessary shutdowns.

The proposed changes would ensure that

these objectives are met. The Minimum

Critical Power Ratio (MCPR) operating limit

specified in the Hope Creek COLR was

determined using the APRM flow-biased

scram/control rod block setpoints and

allowable values proposed in this

amendment application and has been chosen

to ensure that the cladding safety limit would

not be violated during normal plant

operations and anticipated transients. Since

the operating limit MCPR is chosen such that

the cladding safety limit is maintained,

adequate margins for the safety limits are

ensured. The proposed changes would also

serve to ensure that the objective of avoiding

unnecessary shutdowns is met by furnishing

greater margin between the operating

envelope and the setpoint at lower flows.

Based on the above information, we

conclude that the proposed changes would

not significantly reduce a margin of safety.

B. Transfer of RBM Setpoint Control to the

COLR

The proposed transfer of control of the

RBM setpoint and allowable value to the

COLR would not affect the methodology for

establishing the core operating limits. The

setpoint and allowable value are modified to

incorporate a controlling value which will be

included in the COLR and indicated as such

by reference in the Technical Specifications.

Therefore, the setpoint and allowable value

would continue to be controlled in a manner

that would ensure that safety analysis limits

are met. We conclude that implementation of

the proposed changes would not significantly

reduce a margin of safety.

C. RCS Recirculation Flow Revisions

The HCGS was licensed to operate up to

105% of rated core flow as part of

Amendment 15. The analysis used to justify

operation up to 105% of rated core flow is

contained in NEDC-31487. NEDC-31487

addresses the full range of transient and

accident events associated with operation up

to 105% of rated core flow. The affects of

operation with the revised RCS recirculation

flow upscale trip setpoint and allowable

value are bounded by the analysis presented

in NEDC-31487.

In addition, cycle specific analysis

performed for Reload 5/Cycle 6, have

incorporated the assumptions of operation up

to 105% of rated core flow and have

confirmed that operation is within allowable

design limits.

Based on the above information, we

conclude that the proposed changes would

not significantly 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: Pennsville Public Library,190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: January

11, 1995

Description of amendment request:

The proposed Technical Specification

(TS) revision provides changes to TS

Section 3/4.3.8 ‘‘Turbine Overspeed

Protection System.’’ The proposed

revision removes these requirements

from the TS and relocates the Bases to

the Hope Creek Updated Final Safety

Analysis Report (UFSAR) and the

Surveillance Requirements to the

applicable surveillance procedures. The

Limiting Conditions for Operation

(LCOs) would be eliminated.

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. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed changes involve no

hardware changes, no changes to existing

structures, and no changes to the operation

of any systems or components. Specifically,

the deletion of the LCO’s by this submittal

will not alter established turbine overspeed

protection system operation. Procedural

guidance will be provided in the event of an

inoperable control, stop, or intermediate

valve to place the system in a safe condition.

The relocation of this specification to the

UFSAR and surveillance procedures will

continue to ensure that the probability of

unacceptable damage to safety-related

structures, systems, and components from

turbine missiles remains acceptably low.

Relocation of this specification’s Bases and

Surveillance Requirements to the UFSAR and

surveillance procedures, respectively, and

the deletion of the LCO’s represents changes

that do not affect plant safety and do not alter

existing accident analyses.

2. Will not create the possibility of a new

or different kind of accident from any

previously evaluated.

The proposed changes are procedural in

nature concerning the location of the

descriptive information and surveillance

requirements for the turbine overspeed

protection system. Removing these

specifications from the Technical

Specifications and placing them in the

UFSAR and surveillance procedures will not

alter the operation of the turbine overspeed

protection system or its ability to perform its

intended function. Procedural guidance will

be provided to assist in placing the system

in a safe condition while maintenance and

testing of this system will continue in

accordance with the turbine manufacturers

recommendations taking into consideration

plant operating experience and ASME

guidance. Therefore, these changes will not

create a new or unevaluated accident or

operating condition.

3. Will not involve a significant reduction

in a margin of safety.

The proposed changes relocate the Turbine

Overspeed Protection System portion of the

Technical Specifications to the UFSAR and

surveillance procedures in accordance with

guidance provided by the NRC Final Policy

Statement regarding the improvement of

Technical Specifications. The requirements

that will reside in the UFSAR for the turbine

overspeed protection system will ensure that

the system remains capable of protecting

against excessive turbine overspeed.

Therefore, the proposed changes will not

involve a significant reduction in any

margins of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this
39452 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: January

20, 1995

Description of amendment request:

The proposed Technical Specification

(TS) revision represents changes to TS

Section 3/4.11.2.6 ‘‘Explosive Gas

Mixture,’’ TS Table 3.3.7.11-1

‘‘Radioactive Gaseous Effluent

Monitoring Instrumentation,’’ and TS

Table 4.3.7.11-1 ‘‘Radioactive Gaseous

Effluent Monitoring Instrumentation

Surveillance Requirements.’’ The

proposed revision would remove these

TS from the Technical Specifications

and relocate the Bases to the Hope Creek

Updated Final Safety Analysis Report

(UFSAR) and the Surveillance

Requirements to the applicable

surveillance procedures. The Limiting

Conditions for Operation (LCOs) would

be eliminated.

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. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed changes involve no

hardware changes, no changes to the

operation of any systems or components, and

no changes to existing structures. The

relocation of this specification to the UFSAR

and surveillance procedures will continue to

ensure that the entrainment of hydrogen in

the main condenser is monitored and

controlled. Relocation of this specification’s

Bases and Surveillance Requirements to the

UFSAR and surveillance procedures,

respectively, and the deletion of the LCO’s

represent changes that do not affect plant

safety and do not alter existing accident

analyses.

2. Will not create the possibility of a new

or different kind of accident from any

previously evaluated.

The proposed changes are procedural in

nature concerning the location of the

descriptive information and surveillance

requirements for the explosive gas mixture

monitoring instrumentation. Removing these

specifications from the Technical

Specifications and placing them in the

UFSAR and surveillance procedures will not

alter the operation of the explosive gas

monitors or their ability to perform intended

functions. Maintenance and testing of these

monitors will continue based upon the

manufacturers’ recommendations taking into

consideration plant operating experience.

Therefore, these changes will not create a

new or unevaluated accident or operating

condition.

3. Will not involve a significant reduction

in a margin of safety.

The proposed changes relocate the

Explosive Gas Mixture specifications from

the Technical Specifications to the UFSAR

and surveillance procedures in accordance

with guidance provided by the NRC Final

Policy Statement regarding the improvement

of Technical Specifications. The

requirements that will reside in the UFSAR

and surveillance procedures for the explosive

gas mixture monitoring instrumentation will

ensure that the ability to determine main

condenser hydrogen concentrations is

properly maintained. Therefore, the proposed

changes will not involve a significant

reduction in any margins of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070.

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of amendment request: January

20, 1995

Description of amendment request:

The proposed change to the Technical

Specifications (TS) would revise TS

4.1.3.1.2.b, ‘‘Control Rods - Surveillance

Requirement’’ to change the required

action to be taken when a control rod

becomes immovable due to excessive

friction or mechanical interference from

‘‘at least once per’’ 24 hours to ‘‘within’’

24 hours. The other control rods would

be tested within 24 hours and every 7

days thereafter, as opposed to the

current requirement of testing every 24

hours.

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. Will not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

The proposed change involves no

hardware changes, no changes to the

operation of any systems or components, and

no changes to existing structures. The

revision of the control rod movement test

frequency represents a change that does not

affect plant safety and does not alter existing

accident analyses.

2. Will not create the possibility of a new

or different kind of accident from any

previously evaluated.

The proposed change is procedural in

nature concerning the frequency of control

rod movement tests for all withdrawn control

rods after a control rod has been determined

to be immovable due to excessive friction or

mechanical interference. The methodology

for determining additional immovable

control rods remain unchanged. The

proposed change while slightly increasing

the possibility of an undetected immovable

control rod will not create a new or

unevaluated accident or operating condition.

3. Will not involve a significant reduction

in a margin of safety.

The proposed change is in accordance with

recommendations provided by the NRC

regarding the improvement of Technical

Specifications. This change will result in the

perpetuation of current safety margins while

reducing regulatory burden and decreasing

equipment degradation.

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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey

08070

Attorney for licensee: M. J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502

NRC Project Director: John F. Stolz

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, Pennsylvania Power

Company, Toledo Edison Company,

Docket No. 50-440, Perry Nuclear

Power Plant, Unit No. 1, Lake County,

Ohio

Date of amendment request: June 1,

1995

Description of amendment request:

The proposed change would revise the

Technical Specifications to make them

more restrictive regarding control rod

drive (CRD) scram time testing. CRD

scram time testing would be required

following maintenance prior to

considering the CRD operable, and

could be performed at any reactor
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39453

pressure. Additional testing would be

required when reactor coolant pressure

is greater than or equal to 950 psig and

prior to 40 percent rated thermal power.

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 provides more

stringent requirements for operation of the

facility. These more stringent requirements

do not result in operation that will increase

the probability of initiating an analyzed event

and do not alter assumptions relative to

mitigation of an accident or transient event.

The more restrictive requirements continue

to ensure process variables, structures,

systems and components are maintained

consistent with the safety analysis and

licensing basis. Therefore, this 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?

The proposed change does not involve a

physical alteration of the plant (no new or

different type of equipment will be installed)

or changes in the methods governing normal

plant operation. The proposed change does

impose different requirements. However,

these changes are consistent with

assumptions made in the safety analysis and

licensing basis. 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 imposition of more restrictive

requirements either has no impact on or

increase in the margin of plant safety. As

provided in the discussion of the change,

each change in this category is by definition

providing additional restrictions to enhance

plant safety. The change maintains

requirements within safety analyses and

licensing bases. 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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037

NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: April 10,

1995

Description of amendment request:

The proposed amendment would revise

the following Technical Specifications

(TS) and their associated Bases: TS 3/

4.7.1.2, ‘‘Auxiliary Feedwater System,’’

to clarify Action ‘‘a’’ by inserting ‘‘or

both’’ steam generator≥s’’ and to remove

references to pressure indicators and

specific pressure readings and adding

performance based requirements; TS 3/

4.7.1.3, ‘‘Condensate Storage Tanks,’’ to

modify the Limiting Condition for

Operation (LCO) to more closely

conform to standard TS; and TS 3/

4.7.1.7, ‘‘Motor Driven Feedwater Pump

System,’’ to consolidate the

requirements of 2 current surveillance

requirements and clarify the operability

requirements when local manual valves

are realigned for testing purposes.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Toledo Edison has reviewed the proposed

changes and determined that asignificant

hazards consideration does not exist because

operation of the Davis-besse Nuclear Power

Station, Unit Number 1, in accordance with

these changes would:

a. Not involve a significant increase in the

probability of an accident previously

evaluated because no change is being made

to any accident initiator. No previous

analyzed accident scenario is changed, and

initiating conditions and assumptions remain

as previously analyzed. The proposed

changes are clarifications and the

incorporations of the guidance provided by

NUREG-1430. Therefore, it can be concluded

that the proposed changes do not involve a

significant increase in the probability of an

accident previously evaluated.

b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the proposed changes do

not affect accident conditions or assumptions

used in evaluating the radiological

consequences of an accident. The proposed

changes do not alter the source term,

containment isolation or allowable

radiological releases.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the proposed

changes do not change the way the plant is

operated and, no new or different failure

modes have been defined for any plant

system or component important to safety, nor

has any limiting single failure been identified

as a result of the proposed changes. No new

or different types of failures or accident

initiators are introduced by the proposed

changes.

3. Not involve a significant reduction in a

margin of safety because the proposed

changes are clarifications and the

incorporations of the guidance provided by

NUREG-1430, and continue to ensure the

availability and capability of the Auxiliary

Feedwater System, Service Water System and

the Motor Driven Feedwater Pump System

when called upon to perform their functions.

The proposed changes will not adversely

impact any safety analysis assumptions.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: June 1,

1995

Description of amendment request:

The proposed amendment would

change the allowed outage time from 72

hours to 7 days for one unavailable

emergency diesel generator (EDG) as

detailed in Technical Specification

3.8.1.1, ‘‘AC Power Sources, Operating,’’

and its associated Bases 3.0.5.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Toledo Edison has reviewed the proposed

change and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station (DBNPS), Unit No. 1, in accordance

with this change would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because the proposed change to

increase the allowed outage time for one

emergency diesel generator from three (3)

days to seven (7) days does not make a

change to any accident initiator, initiating

condition or assumption. The accident

previously evaluated in the DBNPS Updated

Safety Analysis Report (USAR) Section

15.2.9, Loss of All AC Power to the Station
39454 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

Auxiliaries (Station Blackout), is not affected

by this proposed change. The proposed

change does not involve a significant change

to the plant design or operation, only to the

allowed outage time, and based on a review

of the available alternate A.C. power sources,

the effect on probabilistic risk at power, the

effect on shutdown risk, and maintenance

planning and scheduling, this change has

been determined to be acceptable.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the proposed change does

not invalidate assumptions used in

evaluating the radiological consequences of

an accident, does not alter the source term or

containment isolation and does not provide

a new radiation release path or alter potential

radiological releases.

2. Not create the possibility of a new or

different kind of accident from any

previously evaluated because the proposed

change does not introduce a new or different

accident initiator or introduce a new or

different equipment failure mode or

mechanism.

3. Not involve a significant reduction in

the margin of safety because the proposed

change does not significantly reduce the

margin to safety which exists in the present

Technical Specification action statements.

The DBNPS USAR Section 15.2.9 evaluates

the acceptability of the loss of all A.C. power

to the station, including the loss of both

EDGs, and the margin of safety in this

analysis is not affected by the proposed

change. in addition, since the issuance of the

original DBNPS Operating License Technical

Specifications Toledo Edison has installed a

Station Blackout Diesel Generator (SBODG),

comparable in continuous rating to the EDGs

and capable of providing emergency A.C.

power in the event all three offsite 345 kV

transmission lines and the two EDGs are

unavailable. This has positive effect on

maintaining the margin to safety which exists

in the Technical Specifications with a three

day allowed outage time, which was

established prior to installation of the

SBODG.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: June 7,

1995

Description of amendment request:

The proposed amendment would revise

Technical Specification 3/4.9.4,

Refueling Operations - Containment

Penetrations, and associated Bases 3/

4.9.4, Containment Penetrations. The

proposed changes include revising the

Limiting Condition for Operation (LCO)

3.9.4.b to allow both doors of the

containment personnel airlock to be

open during core alterations or

movement of irradiated fuel within the

containment, provided that certain

specified conditions are meet.

Additional changes are proposed to

revise or clarify TS LCO 3.9.4.c, TS

Action 3.9.4.a, and TS Surveillance

Requirement 4.9.4, and modify the

Bases to reflect the requested changes.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Plant

(DBNPS), Unit No. 1, in accordance with

these changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no Updated Safety

Analysis Report (USAR) accident initiators

are affected by the proposed changes.

The proposed change to TS LCO 3.9.4.b

would allow both doors of the containment

personnel air lock to be open during core

alterations or movement of irradiated fuel

within the containment, provided that

certain specified conditions are met. The

containment personnel air lock is not an

initiator to any accident. Whether the

containment personnel air lock doors are

open or closed during fuel movement and

core alterations has no effect on the

probability of any accident previously

evaluated.

The proposed clarification of TS LCO

3.9.4.c, changing the term ‘‘outside

atmosphere’’ to ‘‘atmosphere outside

containment,’’ and the proposed change to

TS LCO 3.9.4.c.1, confirming that, in

addition to a manual or automatic isolation

valve, or a blind flange, equivalent means

may be used to close a containment

penetration, have no bearing on the

probability of an accident previously

evaluated.

The proposed changes to TS Action 3.9.4.a,

TS Surveillance Requirement (SR) 4.9.4, and

TS Bases 3/4.9.4 are administrative changes

and have no bearing on the probability of an

accident previously evaluated.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the proposed changes do

not invalidate accident conditions or

assumptions used in evaluating the

radiological consequences of any accident.

he analysis results for a fuel handling

accident inside containment, as presented in

Section 15.4.7.3 of the DBNPS USAR, are

well within the 10 CFR 100 guideline values.

Since the analysis does not take credit for

containment isolation, the status of the

personnel air lock has no impact on the

acceptability of the results. In the event of a

fuel handling accident, release of radioactive

material will continue to be minimized since

the air lock door will remain capable of being

closed. Further, the proposed change could

significantly reduce the dose to workers in

the containment in the event of a fuel

handling accident by speeding the

containment evacuation process.

Since an engineering evaluation described

in proposed Bases 3/4.9.4 will ensure that a

particular containment penetration closure

technique is capable of restricting the release

of radioactive material from a fuel handling

accident, the proposed change to TS LCO

3.9.4.c.1, confirming that an equivalent

means may be used to close a containment

penetration, has no adverse effect on the

consequences of an accident previously

evaluated.

The proposed clarification of TS LCO

3.9.4.c, and the proposed changes to TS

Action 3.9.4.a, TS SR 4.9.4, and TS Bases 3/

4,9.4 are administrative changes and have no

effect on the consequences of an accident

previously evaluated.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because there are no

new failure modes or mechanisms associated

with the proposed changes, nor do the

proposed changes involve any modification

of plant equipment or changes in plant

operational limits.

As described above, the analysis results for

a fuel handling accident inside containment

does not take credit for containment

isolation. Thus the proposed change to TS

LCO 3.9.4.b to allow both doors of the

containment personnel air lock to be open

during core alterations or movement of

irradiated fuel within the containment could

affect the release path for radioactive material

released during a fuel handling accident,

however no new or different kind of accident

will result.

3. Not involve a significant reduction in

the margin of safety.

The analysis results for a fuel handling

accident inside containment, as presented in

[Section 15.4.7.3 of] the DBNPS USAR, are

well within the 10 CFR 100 guideline values.

Since the analysis does not take credit for

containment isolation, the status of the

personnel air lock has no impact on the

acceptability of the results.

The proposed change to TS LCO 3.9.4.c.1

regarding the use of equivalent means of

containment penetration closure has no

adverse impact on the margin of safety since

an equivalent containment penetration
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39455

closure technique will provide the same

assurance of containment closure during core

alterations or movement of irradiated fuel

inside containment.

The various administrative changes and

clarifications proposed will 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 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: June 23,

1995

Description of amendment request:

The proposed amendment would

relocate Technical Specifications (TS) 3/

4.3.3.3 - Seismic Instrumentation, TS 3/

4.3.3.4 - Meteorological

Instrumentation, and TS 3/4.4.11 -

Reactor Coolant System Vents and

associated 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:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station, Unit Number 1, in accordance with

these changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no change is being made

to any accident initiator. No previous

analyzed accident scenario is changed, and

initiating conditions and assumptions remain

as previously analyzed.

The proposed changes are deletions and

relocations of specifications that do not meet

the NRC Final Policy Statement [58 FR

39132, dated July 22, 1993] criteria for

inclusion in TS. Furthermore, these

relocations and deletions are consistent with

the NRC guidance for TS provided by the

‘‘Improved Standard Technical

Specifications for Babcock and Wilcox

Plants,’’ NUREG-1430, Revision 0. Therefore,

it can be concluded that the proposed

changes do not involve a significant increase

in the probability of an accident previously

evaluated.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the proposed changes do

not affect accident conditions or assumptions

used in evaluating the radiological

consequences of an accident. The proposed

changes do not alter the source term,

containment isolation or allowable

radiological releases.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because the proposed

changes do not change the way the plant is

operated, and no new or different failure

modes have been defined for any plant

system or component important to safety, nor

has any limiting single failure been identified

as a result of the proposed changes. No new

or different types of failures or accident

initiators are introduced by the proposed

changes.

3. Not involve a significant reduction in a

margin of safety because Seismic

Instrumentation, Meteorological

Instrumentation, and Reactor Coolant System

Vents are not inputs in the calculation of any

safety margin with regard to TS Safety

Limits, Limiting Safety System Settings,

other TS Limiting Conditions for Operation,

or other previously defined margins for any

structure, system, or component important to

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus

Virginia Electric and Power Company,

Docket Nos. 50-280 and 50-281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request: July 14,

1995

Description of amendment request:

The proposed Technical Specifications

(TS) changes would provide a two-hour

allowed outage time (AOT) for one

residual heat removal (RHR) pump to

accommodate plant safety and

emergency power systems surveillance

testing and permit depressurizing safety

injection (SI) accumulators in lieu of

accumulator isolation.

Basis for proposed no significant

hazards consideration determination:As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Specifically, operation of the Surry Power

Station in accordance with the proposed

change will not:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Surveillance and testing requirements are

necessary to assure that RHR and interfacing

systems’ reliability is maintained. Existing

analyses demonstrate that adequate

shutdown cooling will be maintained with

one train of RHR Operable and in service.

Analyses also demonstrate that alternate

shutdown cooling modes remain available

with adequate decay heat removal capability.

Furthermore, the opposite train of RHR

remains available while in the two hour

surveillance AOT. The response time and

operator actions required to place the

available RHR train in service are consistent

with similar operator response times and

actions

required to place alternate shutdown

cooling modes in service. The administrative

controls and procedures in place assure

adequate shutdown cooling capability is

maintained as supported by existing

analyses.

The existing safety analyses demonstrate

that Reactor Coolant System [RCS] integrity

will be maintained when SI accumulator

pressure is below the pressurizer PORV

[power operated relief valve] LTOPS [low

temperature overpressure system] setpoint.

Therefore, SI accumulator isolation is not

required to ensure Reactor Coolant System

integrity. With RCS temperature below the

LTOPS enabling temperature, automatic

actuation of the pressurizer PORVs or other

TS specified relief paths ensure the assumed

design basis reactor vessel beltline flaw will

not propagate under design basis low

temperature overpressurization accident

conditions. System design and configuration

adequately mitigate an LTOPS actuation due

to an SI accumulator discharge with no

negative consequences regarding RCS

structural integrity or SBLOCA [small break

loss-of-coolant accidents] concerns.

Therefore, the proposed Allowed Outage

Time for an inoperable RHR loop and the

ability to depressurize the SI accumulator in

lieu of SI accumulator isolation do not

increase the probability or consequence of

any previously analyzed accidents.

(2) Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed two hour AOT for one train

of the RHR System will preclude the

possibility of a Technical Specification

violation for conditions where a train of RHR

is out of service for surveillance testing.

Calculations by Westinghouse with

evaluations and supporting analyses

performed by Virginia Power, confirm the

adequacy of decay heat removal with one

RHR train in service, and multiple alternate

shutdown cooling modes remain available.

There are no plant modifications required by

this proposed TS change. Further, the

proposed change does not invalidate any
39456 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

component design criteria or the assumptions

of the UFSAR [updated final safety analysis

report] accident analyses. The RHR System is

being operated in a manner consistent with

the design basis and configuration of the

system and is supported by existing analyses

and procedural controls.

There are no new failure modes or

mechanisms associated with the proposed

change to allow the depressurizing of a SI

accumulator to a pressure value below the

LTOPS setpoint. The LTOPS enabling

temperature remains unchanged. No

operating limits or setpoints are added or

deleted by the proposed change. Reactor

Coolant System pressure relief paths are not

affected.

Therefore, the possibility of a new or

different kind of accident is not being created

by the proposed Allowed Outage Time for an

inoperable RHR loop and the ability to

depressurize the SI accumulator in lieu of SI

accumulator isolation.

(3) Involve a significant reduction in

margin of safety.

The proposed Technical Specifications

change does not involve a reduction in a

margin of safety. The existing safety analyses

demonstrate that adequate shutdown cooling

will be maintained when a train of RHR is

out of service for up to two hours for plant

system surveillance testing, while the

operable train of RHR is operating.

Supporting analyses determined that the

RHR System meets the design cooldown

requirements for a reactor core rating of 2546

MWth [megawatt thermal] with either one or

both trains of RHR in service. Additionally,

an evaluation of the technical basis for

shutdown operations for the proposed Surry

core uprating to 2546 MWth determined that

the administrative controls and Abnormal

Procedures in place at Surry ensure adequate

decay heat removal capability during

shutdown conditions. The administrative

controls and procedure revisions are

supported by a detailed series of thermalhydraulic calculations for various loss of

RHR scenarios. There is no reduction in

shutdown cooling capability due to the

proposed TS change, and no reduction in the

capability to mitigate a loss of decay heat

removal event since the RHR train affected by

the testing is available and can be restored in

a comparable time period to that required to

restore RHR to service in the event of loss of

station power or loss of the operating train

of RHR. Consequently, system design, plant

configuration, and administrative controls

remain available to adequately mitigate a loss

of RHR event with a single train of RHR out

of service for up to two hours during plant

system surveillance testing. It may be

concluded that there is no reduction in the

margin of safety due to the proposed

Technical Specification change.

Existing safety analyses also demonstrate

that Reactor Coolant system integrity will be

maintained in the event of an inadvertent SI

accumulator discharge when SI accumulator

pressure is below the pressurizer PORV

LTOPS setpoint. Sufficient administrative

controls are maintained to ensure LTOPS is

‘‘Enabled’’ and SI accumulators are isolated

at the appropriate RCS conditions to

minimize the possibility of challenging RCS

integrity. Technical Specifications

administrative controls that prevent

inadvertent charging pump operation,

maintain adequate relief paths, and restrict

Steam Generator primary to secondary

temperature differential remain in place.

Consequently, the Technical Specifications

change ensures that an inadvertent SI

accumulator discharge cannot challenge RCS

structural integrity during LTOPS conditions

when SI accumulator pressure is below the

pressurizer PORV LTOPS setpoint.

Therefore, the proposed Allowed Outage

Time for an inoperable RHR loop and the

ability to depressurize the SI accumulator in

lieu of SI accumulator isolation does not

reduce any margin of safety as defined in the

Technical Specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: David B.

Matthews

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: June 14,

1995, as supplemented by letter dated

July 13, 1995.

Description of amendment request:

This amendment request proposes to

revise Technical Specification (TS)

3.2.3, ‘‘Nuclear Enthalpy Rise Hot

Channel Factor,’’ TS 6.9.1.9, ‘‘Core

Operating Limits Report,’’ and the

associated Bases sections. The revisions

are needed to incorporate changes

associated with the planned

implementation of advanced nuclear

and core thermal-hydraulic design

methodologies licensed from

Westinghouse Electric Corporation for

core reload design, starting with Cycle

9.

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 probability of occurrence and the

consequences of an accident evaluated

previously in the Updated Safety Analysis

Report (USAR) are not increased due to the

proposed technical specification changes.

The Technical Specification changes being

requested are to reflect revised calculational

methods to be used for core reload design,

starting with Cycle 9. There are no changes

being made to any licensed design

parameters from previous cycles. Thus, it is

concluded that the probability and

consequences of the accidents previously

evaluated in the USAR are not increased.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

There is no new type of accident or

malfunction being created. The proposed

changes only provide revised analysis

methodologies to support core reload design,

starting with Cycle 9. The requested changes

do not change the method and manner of

plant operation. The safety design bases in

the USAR have not been altered. Thus, the

requested changes do 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 do not change the

plant configuration in a way that introduces

a new potential hazard to the plant and do

not involve a significant reduction in the

margin of safety. The analyses and

evaluations discussed in the safety

evaluation (Attachment I) [Attached to Wolf

Creek Nuclear Operating Corporation’s letter

number ET 95-0051, dated June 14, 1995]

demonstrates that all applicable design

criteria continue to be met for the changes.

Therefore, it is concluded that the margin of

safety, as described in the bases to any

technical specification, is not reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, NW., Washington, D.C.

20037

NRC Project Director: William H.

Bateman

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39457

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Commonwealth Edison Company,

Docket Nos. 50-254 and 50-265, Quad

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, Illinois

Date of amendment request: February

23, 1995

Description of amendment request:

The amendment relates to

Commonwealth Edison Company’s

(ComEd) request to reflect the merger

between IIGEC, MidAmerican, Midwest

Power Systems Inc., and Midwest

Resources, Inc. By letter dated

November 21, 1994, Iowa-Illinois Gas

and Electric Company (IIGEC) requested

approval, pursuant to Section 50.80 of

Title 10 of the Code of Federal

Regulations, of the transfer of its

ownership share of 25 percent of Quad

Cities Nuclear Power Station, Units 1

and 2, to MidAmerican Energy

Company (MidAmerican).

Date of publication of individual

notice in Federal Register: July 5, 1995

(60 FR 35054)

Expiration date of individual notice:

August 4, 1995

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Connecticut Yankee Atomic Power

Company, Docket No. 50-213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: March

31, 1995

Description of amendment request:

The proposed amendment will delete

Technical Specification (TS) Sections

1.38 and 1.39, ‘‘Definitions, Fuel

Assembly Types,’’ revise TS Sections 3/

4.9.3, ‘‘Refueling Operations, Decay

Time’’ and TS 3/4.9.14, ‘‘Refueling

Operations, Spent Fuel Pool - Reactivity

Condition,’’ replace TS Sections 5.6.1.1,

‘‘Spent Fuel,’’ and TS 5.6.3, ‘‘Capacity,’’

and add a new TS Section 3/4.9.15,

‘‘Refueling Operations, Spent Fuel Pool

Cooling.’’ These changes would support

a rerack of the spent fuel pool to expand

the spent fuel pool’s storage capacity

from 1168 assemblies to 1480

assemblies so as to accommodate a fullcore-discharge through the current

validity date of the Haddam Neck

operating license (2007).

Date of publication of individual

notice in Federal Register: May 12, 1995

(60 FR 25746)

Expiration date of individual notice:

June 12, 1995

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the

local public document rooms for the

particular facilities involved.

Commonwealth Edison Company,

Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station Units 1 and 2,

Lake County, Illinois

Date of application for amendments:

June 14, 1995

Brief description of amendments: The

amendments revise the requirement to

perform an emergency diesel generator

(EDG) automatic start and sequence

loading test immediately following the

24 hour EDG endurance test.

Date of issuance: July 18, 1995

Effective date: July 18, 1995

Amendment Nos.: 166 and 154

Facility Operating License Nos. DPR39 and DPR-48: The amendments

revised the Technical

Specifications.Public comments

requested as to proposed no significant

hazards consideration determination:

Yes (60 FR 34308). This notice provided

an opportunity to submit comments on

the Commission’s proposed no

significant hazards consideration

determination. No comments have been

received. The notice also provided for

an opportunity to request a hearing by

July 31, 1995, but indicated that if the

Commission makes a final no significant

hazards consideration determination

any such hearing would take place after

issuance of the amendment. The

Commission’s related evaluation of the

amendments, finding of exigent

circumstances and final no significant

hazards consideration determination is

contained in a Safety Evaluation dated

July 18, 1995.

Local Public Document Room

location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois

60085.

Indiana Michigan Power Company,

Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

March 31, 1995

Brief description of amendments: The

amendments revise Technical

Specification section 3.9.4 to allow,

under certain conditions, both

containment personnel airlocks to be

open during core alterations.

Date of issuance: July 12, 1995

Effective date: July 12, 1995

Amendment Nos.: 197 and 182

Facility Operating License Nos. DPR58 and DPR-74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29879)The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated July 12, 1995.No

significant hazards consideration

comments received: No.
39458 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Nebraska Public Power District, Docket

No. 50-298, Cooper Nuclear

Station,Nemaha County, Nebraska

Date of amendment request: May 2,

1995

Brief description of amendment: The

amendment revised Surveillance

Requirement 4.7.A.2.f.1 to allow a onetime extension for the performance of

Type B local leak rate testing of the

drywell head and manport from July 17,

1995, until startup from Refueling

Outage 16, scheduled to commence on

October 13, 1995.

Date of issuance: July 11, 1995

Effective date: July 11, 1995

Amendment No.: 170

Facility Operating License No. DPR46. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29879)The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated July 11, 1995No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Auburn Public Library, 118

15th Street, Auburn, NE 68305.

Northeast Nuclear Energy Company, et

al., Docket No. 50-423,

MillstoneNuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of application for amendment:

January 10, 1995

Brief description of amendment: The

amendment revises the Technical

Specifications to delete the power range

negative flux trip from Tables 2.2-1, 3.3-

1, and 4.3-1, and delete the associated

Bases Section 2.0.

Date of issuance: July 11, 1995

Effective date: As of the date of

issuance to be implemented within30

days.

Amendment No.: 116

Facility Operating License No. NPF49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 1, 1995 (60 FR

11135)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 11, 1995. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Northeast Nuclear Energy Company, et

al., Docket No. 50-423,

MillstoneNuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of application for amendment:

March 29, 1995

Brief description of amendment: The

amendment revises Technical

Specification 3.10.5 to allow more than

one control bank to be fully withdrawn

from the core simultaneously in order to

conduct rod drop time response testing.

Date of issuance: July 11, 1995

Effective date: As of the date of

issuance to be implemented within60

days.

Amendment No.: 117

Facility Operating License No. NPF49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29880)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 11, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360.

Northern States Power Company,

Docket No. 50-263, Monticello

NuclearGenerating Plant, Wright

County, Minnesota

Date of application for amendment:

February 12, 1993, as supplemented by

letters dated March 22, 1993, and

August 25, 1994

Brief description of amendment: The

amendment increases the minimum

core spray pump flow to more

conservatively account for emergency

core cooling systems bypass leakage

paths. The amendment also makes

various typographical, editorial and

administrative corrections and changes.

Date of issuance: July 12, 1995

Effective date: July 12, 1995

Amendment No.: 93

Facility Operating License No. DPR22. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 4, 1993 (58 FR 41508).

The August 25, 1994 letter provided

clarifying information within the scope

of the original submittal and did not

change the staff’s initial proposed no

significant hazards considerations

determination.The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

July 12, 1995.No significant hazards

consideration comments received: No.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station,Unit

No. 1, Washington County, Nebraska

Date of amendment request:

November 11, 1994, as supplemented by

letters dated April 7, 1995, and June 26,

1995

Brief description of amendment: The

amendment implements administrative

changes to TS 5.2 and 5.5. These

changes reflect organizational changes

in OPPD senior management, delete

specific titles of personnel on the Plant

Review Committee (PRC), revise the

makeup of the PRC quorum, revise the

membership of the Senior Audit and

Review Committee (SARC), delete SARC

audit frequencies and add minor

clarifications to the descriptions of

SARC reviews and audits.

Date of issuance: July 21, 1995

Effective date: July 21, 1995

Amendment No.: 168

Facility Operating License No. DPR40. Amendment revised the

TechnicalSpecifications.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65819). The April 7, 1995, and June 26,

1995, letters provided clarifying

information and did not change the

initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated July 21, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102.

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station,Unit

No. 1, Washington County, Nebraska

Date of amendment request: March 1,

1995

Brief description of amendment: The

amendment revises TS 2.5, 2.8, 2.11,

3.2, and 3.10 and relocates

administrative controls for the

emergency and security plans from TS

5.5 and 5.8 to the plans. The relocation

is in accordance with Generic Letter

(GL) 93-07, ‘‘Modification of the

Technical Specification Administrative

Control Requirements for Emergency

and Security Plans.’’

Date of issuance: July 21, 1995

Effective date: July 21, 1995

Amendment No.: 169

Facility Operating License No. DPR40: The amendment revised the

TechnicalSpecifications.
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39459

Date of initial notice in Federal

Register: April 12, 1995 (60 FR 18627)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 21, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102.

Pacific Gas and Electric Company,

Docket No. 50-133, Humboldt Bay

Power Plant, Unit 3, Humboldt County,

California

Date of application for amendment:

April 10, 1995

Brief description of amendment: This

amendment revised License No. DPR-7,

to permit the provisions of 10 CFR 50.59

to be applied with respect to changes to

the facility or procedures described in

the Decommissioning Plan or changes to

the Decommissioning Plan, and the

conduct of tests or experiments not

described in the Decommissioning Plan.

Date of issuance: July 7, 1995

Effective date: This license

amendment is effective as of the date of

its issuance and must be fully

implemented no later than 30 days from

the date of issuance.

Amendment No.: 29Facility License

No. DPR-7: This amendment revised

License No. DPR-7

Date of initial notice in Federal

Register: June 6, 1995 (60 FR 29885)The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated July 7, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Humboldt County Library, 636

F Street, Eureka, California 95501.

PECO Energy Company, Public Service

Electric and Gas Company, Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

No. 50-278, Peach Bottom Atomic

Power Station, Unit No. 3, York County,

Pennsylvania

Date of application for amendment:

November 21, 1994

Brief description of amendment: This

amendment changes the technical

specifications (TS) by allowing the third

Type A Containment Integrated Leakage

Rate Test in the second 10-year service

period to be conducted during refueling

outage 11 scheduled for September

1997. This TS change is consistent with

a one-time exemption from Appendix J

to 10 CFR Part 50 that extends the 10-

year service period and allows the three

type A tests to be performed at intervals

that are not approximately equal.

Date of issuance: July 10, 1995

Effective date: July 10, 1995

Amendment No.: 210

Facility Operating License No. DPR56: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 23, 1995 (60 FR

27340)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 10, 1995.No significant hazards

consideration comments received: No

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.

PECO Energy Company, Public Service

Electric and Gas Company,Delmarva

Power and Light Company, and

Atlantic City Electric Company,Docket

No. 50-278, Peach Bottom Atomic

Power Station,Unit No. 3, York County,

Pennsylvania

Date of application for amendment:

June 23, 1993, as supplemented by

letters dated April 5, May 2, June 6, June

8, July 6 (two letters), July 7, July 20,

July 28 (two letters), September 16,

September 30, and October 14, 1994 and

June 22, 1995.

Brief description of amendment: The

amendment raises the authorized

maximum power level from 3293 MWt

to a new limit of 3458 MWt. The

amendment also approves changes to

the Technical Specifications to

implement operation at the increased

power limit.

Date of issuance: July 18, 1995

Effective date: As of date of issuance

and is to be implemented prior to

startup in Cycle 11, currently scheduled

for October 1995.

Amendment No.: 211

Facility Operating License No. DPR56: Amendment revised the License and

the Technical Specifications.

Date of initial notice in Federal

Register: August 29, 1994 (59 FR

44432)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 18, 1995.No significant hazards

consideration comments received: No

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.

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

April 10, 1995

Brief description of amendments:

Remove the response time limit Tables

3.3.1-2, 3.3.2-3, and 3.3.3-3 from the

Technical Specifications, and add the

information to the Final Safety Analysis

Report in accordance with Generic

Letter 93-08.

Date of issuance: July 11, 1995

Effective date: July 11, 1995

Amendment Nos.: 148 and 118

Facility Operating License Nos. NPF14 and NPF-22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 6, 1995 (60FR 29887)The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated July 11, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

November 21, 1994, as supplemented

April 6, and July 3, 1995

Brief description of amendments:

These amendments make changes

affecting the Administrative Controls

Section of the Technical Specifications.

The areas changed are Nuclear

Effectiveness and Efficiency Design

Study (NEEDS) Organization Title

Changes; Minimum Shift Crew

Composition; delete Independent

Technical Review Section from TS;

delete Nuclear Review Board (NRB)

Review Section from TS; and delete

NRB Audit Section from TS.

Date of issuance: July 18, 1995

Effective date: Units 1 and 2, as of the

date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 96 and 60

Facility Operating License Nos. NPF39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 10, 1995 (60 FR

24914)The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

July 18, 1995.No significant hazards

consideration comments received: No
39460 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

August 31, 1994, as supplemented July

3, 1995

Brief description of amendments:

These amendments modify TS Sections

3.4.9.1, 3.4.9.2, 3.9.11.1, 3.9.11.2, and

the associated Bases Sections 3/4.4.9

and 3/4.4.11, to permit the use of either

an ‘‘analytical approach’’ (i.e.,

calculation) or ‘‘demonstrations’’ to

ensure the operability of an alternate

decay heat removal method, rather than

the existing TS requirement which

stipulates that operability of the

alternate decay removal method be

demonstrated.

Date of issuance: July 18, 1995

Effective date: Units 1 and 2, as of the

date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 97 and 61

Facility Operating License Nos. NPF39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55884)The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

July 18, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

August 22, 1994, as supplemented July

3, 1995

Brief description of amendments:

These amendments revise Technical

Specification Surveillance Requirement

4.1.3.1.4a to delete the requirement that

the Scram Discharge Volume (SDV) be

determined operable by testing the SDV

vent and drain valves from a

configuration of less than or equal to

50% rod density.

Date of issuance: July 18, 1995

Effective date: Units 1 and 2, effective

as of date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 98 and 62

Facility Operating License Nos. NPF39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55881)The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

July 18, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Philadelphia Electric Company, Docket

Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

August 22, 1994, as supplemented by

letter dated July 3, 1995

Brief description of amendments: The

amendments revise the Technical

Specifications surveillance

requirements for scram insertion times

and revise the TS surveillance

requirements for control rod block and

source range monitoring

instrumentation.

Date of issuance: July 18, 1995

Effective date: Units 1 and 2, effective

as of the date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 99 and 63

Facility Operating License Nos. NPF39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 9, 1994 (59 FR

55881)The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

July 18, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania

19464.

Public Service Company of Colorado,

Docket No. 50-267, Fort St. Vrain

Nuclear Generating Station (FSV), Unit

No. 1, Platteville, Colorado

Date of application for amendment:

Amendment No. 88, April 14, 1995.

Brief description of amendment: This

amendment would revise the FSV

Decommissioning Technical

Specifications (DTS) by: revising the

FSV DTS to reflect recent organizational

changes resulting from corporate

restructuring to prepare for repowering

the site with natural gas-power turbines

and to incorporate editorial changes.

The staff has determined that the

proposed amendment does not require a

significant hazard consideration,

pursuant to 10 CFR 50.92.PossessionOnly License No. DPR-34: Amendment

revises the DTS.

Local Public Document Room

location: Weld Library District -

Downtown Branch, 919 7th Street,

Greeley, CO 80631.

Sacramento Municipal Utility District,

Docket No. 312, Rancho Seco Nuclear

Generating Station, Sacramento

County, California

Date of application for amendment:

February 28, 1995

Brief description of amendment: This

amendment relocates the quality

assurance audit frequencies from the

technical specifications to the Rancho

Seco Quality Manual and changes the

reporting frequency of the Radioactive

Effluent Release Report from semiannual to annual.

Date of issuance: July 19, 1995

Effective date: July 19, 1995

Amendment No.: 122

Facility Operating License No. NPF-1:

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR

16200)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 19, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Central Library, Government

Documents, 828 I Street, Sacramento,

California 95814.

Southern California Edison Company,

et al., Docket Nos. 50-361 and 50-362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

September 15, 1993, as supplemented

by letter dated September 6, 1994.

Brief description of amendments: The

amendments revised Technical

Specification (TS) Table 2.2-1, ‘‘Reactor

Protective Instrumentation Trip

Setpoint Limits,’’ Table 3.3-1, ‘‘Reactor

Protective Instrumentation,’’ Table 3.3-

3, ‘‘Engineered Safety Feature Actuation

System Instrumentation,’’ and Table 3.3-

4, ‘‘Engineered Safety Feature Actuation

System Instrumentation Trip Values,’’

and the associated Bases. The revisions

to the notes in these tables change the

pressure at which the low pressurizer

pressure trip bypass shall be

automatically removed to a consistent

value of ‘‘before pressurizer pressure

exceeds 500 psia (the corresponding

bistable allowable value is less than or

equal to 472 psia).’’ In addition, the

wording of the notes is revised to make

the notes more consistent with each

other.

Date of issuance: July 14, 1995
Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices 39461

Effective date: July 14, 1995, to be

implemented within 30 days of the date

of issuance.

Amendment Nos.: Unit 2 -

Amendment No. 120; Unit 3 -

Amendment No. 109

Facility Operating License Nos. NPF10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 29, 1993 (58 FR

50975). The September 6, 1994,

supplemental letter provided additional

clarifying information and did not

change the initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated July 14, 1995. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Main Library, University of

California, P.O. Box 19557,Irvine,

California 92713.

Southern California Edison Company,

et al., Docket Nos. 50-361 and 50-362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

September 3, 1992

Brief description of amendments:

These amendments revise TS 3/4.4.8

‘‘Pressure/Temperature Limits - Reactor

Coolant System,’’ and their associated

Bases, following NRC guidance

provided in Generic Letter 91-01,

‘‘Removal of the Schedule for

Withdrawal of Reactor Vessel Material

Specimens from Technical

Specifications.’’ This generic letter

allows licensees to remove the reactor

vessel material surveillance capsule

withdrawal schedules from the TS

because they are a duplication of the

requirements of 10 CFR Part 50

Appendix H.

Date of issuance: July 17, 1995

Effective date: July 17, 1995, to be

implemented within 30 days of issuance

Amendment Nos.: Unit 2 -

Amendment No. 121; Unit 3 -

Amendment No. 110

Facility Operating License Nos. NPF10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 17, 1993 (58 FR

8781)The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

July 17, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713.

Southern California Edison Company,

et al., Docket Nos. 50-361 and 50-362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

April 30, 1993, as supplemented by

letters dated July 6, 1994 (separate

letters for each unit), and letter dated

January 27, 1995.

Brief description of amendments:

These amendments revise TS 3/4.4.8.1,

‘‘Pressure-Temperature Limits,’’ TS

3.4.8.3.1, ‘‘Overpressure Protection

Systems-RCS Temperature less than or

equal to °F [for Unit 2, less than or equal

to 246°F for Unit 3],’’ TS 3.4.8.3.2,

‘‘Overpressure Protection Systems-RCS

Temperature ≤256°F [for Unit 2, ≤246°F

for Unit 3],’’ and the associated TS

Bases. The proposed change (1) revises

the reactor coolant system (RCS)

pressure-temperature (P-T) limits and

the low temperature overpressure

protection (LTOP) enable temperatures

to be effective until 20 effective full

power years (EFPY) of operation and (2)

makes minor editorial changes.

Date of issuance: July 18, 1995

Effective date: July 18, 1995, to be

implemented within 30 days of issuance

Amendment Nos.: Unit 2 -

Amendment No. 122; Unit 3 -

Amendment No. 111

Facility Operating License Nos. NPF10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: Unit 2 - July 7, 1993 (58 FR

36445); Unit 3 - June 23, 1993 (58 FR

34094). The two supplemental letters

dated July 6, 1994, and the January 27,

1995, supplemental letter provided

clarifying information and did not

change the initial no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated July 18, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713.

Tennessee Valley Authority, Docket

Nos. 50-259, 50-260, and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2,

and 3, Limestone County, Alabama

Date of application for amendments:

March 30, 1994

Brief description of amendment: The

amendments implement an analog

transmitter/trip system on BFN Unit 3,

revise the reactor vessel water level

safety limit and limiting safety system

setting for BFN Units 1 and 3, add

instrument identifiers and revise

calibration frequencies and functional

test requirements for BFN Unit 2, revise

the calibration frequency for

instrumentation actuating the

suppression chamber-reactor building

vacuum breakers, and provide editorial

changes.

Date of issuance: July 17, 1995

Effective date: July 17, 1995

Amendment Nos.: 222, 237, 196

Facility Operating License Nos. DPR33, DPR-52 and DPR-68: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: September 28, 1994(59 FR

49435) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 17, 1995.No significant hazards

consideration comments received: None

Local Public Document Room

location: Athens Public library, South

Street, Athens, Alabama 356114.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50-346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of application for amendment:

January 30, 1995

Brief description of amendment: The

proposed amendment revises reactor

coolant system pressure-temperature

curves, changes bases for Technical

Specification 3/4.4.9, Pressure

Temperature Limits, and revises License

Condition 2.C(3)(d) to reflect a change

from 10 effective full power years

(EFPY) to 21 EFPY.

Date of issuance: July 20, 1995

Effective date: July 20, 1995

Amendment No.: 199

Facility Operating License No. NPF-3.

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 15, 1995 (60 FR

14029)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 20, 1995.No significant hazards

consideration comments received: No

Local Public Document Room

location: University of Toledo Library,

Documents Department, 2801 Bancroft

Avenue, Toledo, Ohio 43606.

Union Electric Company, Docket No.

50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

September 8, 1994

Brief description of amendment: The

amendment revises Technical

Specifications (TS) 4.2.2.2, 4.2.2.4, and

6.9.19. The changes address
39462 Federal Register / Vol. 60, No. 148 / Wednesday, August 2, 1995 / Notices

incorporating a penalty in the Core

Operating Limits Report (COLR) to

account for heat flux (FQ) increases

greater than 2 percent between

measurements.

Date of issuance: July 20, 1995

Effective date: July 20, 1995

Amendment No.: 101

Facility Operating License No. NPF30. Amendment revises the Technical

Specification Surveillance

Requirements and Administrative

Controls.

Date of initial notice in Federal

Register: December 21, 1994 (59 FR

65823). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 20, 1995. No significant hazards

consideration comments received: No

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Virginia Electric and Power Company,

Docket Nos. 50-280 and 50-281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of application for amendments:

November 22, 1994

Brief description of amendments: The

amendments revised the Technical

Specifications to delete unnecessary

descriptive phrases regarding the

number of cells in the station and

emergency diesel generator batteries.

Date of issuance: July 11, 1995

Effective date: July 11, 1995

Amendment Nos.: 201 and 201

Facility Operating License Nos. DPR32 and DPR-37: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: April 12, 1995 (60 FR

18630)The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

July 11, 1995. No significant hazards

consideration comments received: No

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Dated at Rockville, Maryland, this 2nd day

of August, 1995.

For the Nuclear Regulatory Commission

Jack W. Roe, 4Director, Division of Reactor

Projects - III/IV, Office of Nuclear Reactor

Regulation

[Doc. 95–18810 Filed 8–1–95; 8:45 am]

BILLING CODE 7590–01–F

[Docket No. 50–255]

Consumers Power Company; Notice of

Partial Denial of Amendment to Facility

Operating License and Opportunity for

Hearing

The U.S. Nuclear Regulatory

Commission (the Commission) has

partially denied a request by Consumers

Power Company, (licensee) for an

amendment to Facility Operating

License No DRP–20 issued to the

licensee for operation of Palisades,

located in Covert Township, Van Buren

County, Michigan. Notice of

Consideration of Issuance of this

amendment was published in the

Federal Register on May 25, 1994 (59

FR 27053).

The purpose of the licensee’s

amendment request was to relocate

certain Technical Specifications (TS)

containing fuel cycle-specific parameter

limits that can change with core reloads

to a Core Operating Limits Report.

Several of the TS bases have also been

revised to refer to limits relocated to the

COLR.

The NRC staff has concluded that the

licensee’s request cannot be fully

granted. The removal of the power

distribution measurement uncertainty

factors in Table 3.23.3 and the addition

of certain references to TS 6.9.1.f are

denied. The licensee was notified of the

Commission’s denial of the proposed

change by a letter dated July 26, 1995.

By September 1, 1995, the licensee

may demand a hearing with respect to

the denial described above. Any person

whose interest may be affected by this

proceeding may file a written petition

for leave to intervene.

A request for hearing or petition for

leave to intervene must be filed with the

Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC by

the above date.

A copy of any petition should also be

sent to the Office of the General

Counsel, U.S. Nuclear Regulatory

Commission, Washington, DC 20555,

and to Gerald Charnoff, Esq., Shaw,

Pittman, Potts and Trowbridge, 2300 N

Street, NW., Washington, DC 20037.

For further details with respect to this

action, see (1) the application for

amendment dated April 7, 1994, as

supplemented April 27, 1995, and (2)

the Commission’s letter to the licensee

dated July 26, 1995.

These documents are available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, 20555 and at the Van

Wylen Library, Hope College, Holland,

Michigan 49423.

Dated at Rockville, Maryland, this 26th day

of July 1995.

For the Nuclear Regulatory Commission.

Marsha Gamberoni,

Project Manager, Project Directorate III–I,

Division of Reactor Projects—III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 95–18929 Filed 8–1–95; 8:45 am]

BILLING CODE 7590–01–M

[Docket Nos. 50–206, 50–361, 50–362]

In the Matter of Southern California Edison

Company (San Onofre Nuclear Generating

Station, Units 1, 2, and 3).

Southern California Edison Co.

Exemption

I

Southern California Edison Company

(SCE or the licensee) is the holder of

Facility Operating License No. DPR–13,

which authorizes possession and

maintenance of the San Onofre Nuclear

Generating Station, Unit 1 (SONGS 1)

and Facility Operating License Nos.

NPF–10 and NPF–15, which authorizes

operation of San Onofre Nuclear

Generating Station, Units 2 and 3

(SONGS 2 and 3), respectively. The

licenses provide, among other things,

that the SONGS units are subject to all

rules, regulations, and orders of the

Commission now or hereafter in effect.

The facilities consist of three

pressurized water reactors at the SCE

site located in San Diego County,

California. SONGS 1 is permanently

shut down, while Unites 2 and 3 remain

operational.

II

It is stated in 10 CFR 73.55,

‘‘Requirements for physical protection

of licensed activities in nuclear power

reactors against radological sabotage,’’

paragraph (a), that ‘‘The licenses shall

establish and maintain an onsite

physical protection system and security

organization which will have as its

objective to provide high assurance that

activities involving special nuclear

material are not inimical to the common

defense and security and do not

constitute an unreasonable risk to the

public health and safety.’’

It is specified in 10 CFR 73.55(d),

‘‘Access Requirements,’’ paragraph (1)

that ‘‘The licensee shall control all

points of personnel and vehicle access

into a protected area.’’ It is specified in

10 CFR 73.55(d)(5) that ‘‘A numbered