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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from 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

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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

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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

plants 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

trains 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 trains

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

[[Page 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

[[Page 39434]]

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

[[Page 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

[[Page 39436]]

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 made-up 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

[[Page 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

ANO-1 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

[[Page 39438]]

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

[[Page 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

[[Page 39440]]

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

[[Page 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

[[Page 39442]]

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

[[Page 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 pull-to-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-to-lock 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 de-energizing 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 haz- ards consideration

determination: As

[[Page 39444]]

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 re-

calculated 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

[[Page 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).

[[Page 39446]]

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 low-pressure 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

[[Page 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

[[Page 39448]]

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-so-much that failures that

are accident initiators will not occur more frequently than

previously postulated.

In addition, the consequences of an accident previously

evaluated in the SAR [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-of-service. The extension of the time duration that

certain equipment is out-of-service has no direct physical impact on

the plant and does not create any new accident initiators. The

systems involved are either accident mitigation systems, safe

shutdown systems or systems that support plant operation. All of the

possible impacts that the inoperable equipment may have on its

supported systems were previously analyzed in the SAR and are the

basis for the present TS ACTION statements and AOTs. The impact of

inoperable support systems for a given time duration was previously

evaluated and any accident initiators created by the inoperable

systems was evaluated. The lengthening of the time duration does not

create any additional accident initiators for the plant.

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

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

previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The ESW and RHRSW systems and their supported systems are

designed with sufficient independence and redundancy such that the

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

systems from performing their required safety functions. Since

removal of an ESW and a RHRSW loop from service with one unit in

operation and the other unit in a refueling outage is allowed by the

current Technical Specifications, then the concern is the reduced

margin of safety incurred by extending the affected AOTs.

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

sufficient safety-related equipment is available for response to all

accident conditions and that sufficient decay heat removal

capability is available for a LOCA/LOOP [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

[[Page 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

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