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

UNITED STATES 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 March 21, 1994, through April 1, 1994. The

last biweekly notice was published on March 30, 1994 (59 FR 14884).

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 P-223, Phillips Building, 7920 Norfolk Avenue,

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

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

Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555.

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

intervene is discussed below.

By May 13, 1994, the licensee may file a request for a hearing with

respect to issuance of the amendment to the subject facility operating

license and any person whose interest may be affected by this

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

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

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

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

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

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

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

Building, 2120 L Street, NW., Washington, DC 20555 and at the local

public document room for the particular facility involved. If a request

for a hearing or petition for leave to intervene is filed by the above

date, the Commission or an Atomic Safety and Licensing Board,

designated by the Commission or by the Chairman of the Atomic Safety

and Licensing Board Panel, will rule on the request and/or petition;

and the Secretary or the designated Atomic Safety and Licensing Board

will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope

of the amendment under consideration. The contention must be one which,

if proven, would entitle the petitioner to relief. A petitioner who

fails to file such a supplement which satisfies these requirements with

respect to at least one contention will not be permitted to participate

as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the

above date. Where petitions are filed during the last 10 days of the

notice period, it is requested that the petitioner promptly so inform

the Commission by a toll-free telephone call to Western Union at 1-

(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to (Project Director): petitioner's name

and telephone number, date petition was mailed, plant name, and

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and

to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

Street, NW., Washington, DC 20555, and at the local public document

room for the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-

529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.

1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: February 18, 1994

Description of amendment requests: The proposed amendment would

modify Technical Specifications (TS) 5.3.1, Fuel Assemblies, and TS

5.6.1, Criticality. In addition, the proposed Amendment would add a new

Technical Specification 3/4.9.13, Boron Concentration-Storage Pool, and

its associated BASES. This proposed amendment is requested to allow

credit to be taken for burnup of spent fuel assemblies in establishing

storage locations within the PVNGS spent fuel pools.

Basis for proposed no significant hazards consideration

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

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

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

consequences of an accident previously evaluated.

This amendment does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Radiological consequences of the fuel handling accident are not

impacted by the formation of new storage regions since the fuel

assembly design is unchanged. However, even though the probability of

occurrence of a fuel misplacement error has increased slightly, the

consequences are markedly reduced by the crediting of 2150 ppm of

soluble boron in the spent fuel storage pool. The increase is also not

significant because of the types of administrative controls being put

into place in Regions 2 and 3. Furthermore, a fuel assembly

misplacement error is not considered an accident, as defined in the

UFSAR.

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

accident from any accident previously evaluated.

This amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated. No

changes are being made to the fuel assemblies or the storage racks, and

controls will be employed to control the placement of assemblies in

Regions 2 and 3. As such, there is no possibility of a new or different

kind of accident being created. The existing design basis covers all

possible accident scenarios in the spent fuel storage pool.

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

This amendment request will not involve a significant reduction in

a margin of safety. There is no reduction in the margin of safety since

a keff less than or equal to 0.95 is met under all analyzed

conditions using conservative assumptions which do not credit the

soluble boron in the spent fuel storage pool except under some accident

conditions, as allowed by NRC guidelines. The original mechanical

analyses are unchanged for thermal and seismic/structural

considerations, as these analyses were originally performed for a fully

loaded spent fuel storage pool.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: Theodore R. Quay

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 request: March 11, 1994

Description of amendment request: The proposed amendments would

revise Technical Specification 3/4.7.D, ``Primary Containment Isolation

Valves.'' The proposed amendments will add check valves installed in

the reference leg instrumentation line. The valves have been installed

as part of the modifications required to meet NRC Bulletin (IEB) 93-03.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The proposed license amendment adds the reference leg backfill

check valves to the Technical Specifications. As such, the proposed

amendment does not change the probability nor does it change the

consequences of any previously evaluated accident for Dresden and

Quad Cities Stations.

The proposed modifications (and proposed Technical Specification

amendments) add reference leg backfill instrument lines and check

valves to the reactor vessel level instrumentation. The proposed

modifications will eliminate the phenomenon described in IEB 93-03

(dissolved gases in the [Reactor Vessel Instrumentation System]

RVLIS piping may produce uncertainties in the level instrumentation

during RPV depressurization) by providing degassed Control Rod Drive

(CRD) water to the RVLIS reference leg piping. The proposed design

ensures that a continuous column of water, free of non-condensible

gases is maintained in the RVLIS reference leg piping. As such, the

proposed modifications do not affect any accident precursors or

initiators. Therefore the proposed modifications for the reference

leg backfill instrument lines do not increase the probability of any

previously evaluated accidents for Dresden Station and Quad Cities

Station.

The proposed plant modifications for the reference leg backfill

check valves will not increase the radiological consequences of any

previously evaluated accident. The radiological impact from a

reference leg backfill instrument line break is bounded by Dresden's

and Quad Cities' Instrument Line Break analysis (UFSAR Section

15.6.2). Therefore, the proposed plant changes will not increase the

consequences of any previously evaluated accident.

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

accident from any accident previously evaluated because:

The proposed modification connects the non-safety-related CRD

system to each safety-related division of RPV instrumentation and

Feedwater Level Control System. The backfill check valves will

eliminate the potential for reference leg leakage if CRD piping

integrity is lost. These check valves are classified as safety-

related and will be maintained and controlled such that overall

plant safety is maintained. The addition of the reference leg

backfill check valves to the Technical Specifications does not

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

Dresden Station or Quad Cities Station.

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

because:

Primary containment integrity is not compromised by the addition

of a pair of check valves that provide isolation for the reference

leg backfill lines. These valves have been demonstrated to meet the

intent of the criteria specified in General Design Criterion (GDC)

55. The maintenance and control applied toward all the reference leg

backfill check valves ensures that overall plant safety is

maintained. Therefore, the addition of the reference leg backfill

valves to Technical Specification 3.7.D.1 and 3.7.D.2 does not

reduce the margin of safety for Dresden Station or Quad Cities

Station.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: for 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: James E. Dyer

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 request: March 26, 1993

Description of amendment request: The proposed amendments would

revise Technical Specification 3/4.6 for Dresden and Quad Cities

Stations to allow Single Loop Operation (SLO) with the recirculation

loop suction and discharge valves open. The amendments would also

delete outdated and unnecessary portions of Technical Specification

3.6.H for Dresden, Units 2 and 3, and provide more consistency to the

BWR Standard Technical Specifications (NUREG-0213, Revision 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, which is presented below:

Commonwealth Edison has evaluated this proposed amendment and

determined that it involves no significant hazards considerations.

According to 10 CFR 50.92(c), a proposed amendment to an operating

license involves no significant hazards considerations if 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 proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because:

The proposed change to delete the requirement to close the

suction valve of the idle loop during SLO potentially affects two

transient or accident analysis previously evaluated. The first is

the Loss-of-Coolant Accident (LOCA) which has been analyzed for the

full range of break sizes, from a small rupture, where the makeup

flow is greater than the coolant loss rate, to the largest, a highly

improbable circumferential recirculation line break. The design

basis LOCA at Dresden and Quad Cities is the double-ended guillotine

break in a recirculation line. LPCI is one of the Emergency Core

Cooling Systems that would be initiated to flood the core following

the Design Basis LOCA accident.

The LOCA analysis for Quad Cities, Units 1 and 2, takes no

credit for closure of the recirculation suction valve to properly

direct LPCI flow into the lower plenum of the reactor. Instead, the

LPCI loop selection logic is relied upon to automatically close the

recirculation discharge valve of the selected intact loop. For

Dresden, Units 2 and 3, LPCI is not credited to inject because the

limiting failure is the LPCI injection valve. The LOCA ECCS analyses

previously performed for SLO remain applicable and the severity of a

postulated LOCA event has not increased. The proposed changes do not

physically change the plant in any manner that would increase the

probability of a LOCA.

The second transient considered is the inadvertent startup of an

idle recirculation pump in an unisolated loop. This event is

precluded, however, when the loop is unisolated because the

discharge valve must be closed for the pump to start. To further

decrease the probability of the occurrence of this transient,

Section 3.6.H.3.e is added to require the pump to be electrically

prohibited from starting. In addition, leaving the loop unisolated

results in an increase in the temperature of the water in the loop,

and a correspondingly lower reactivity insertion should the

transient occur. For these reasons, neither the probability nor the

consequences of an inadvertent idle pump start have increased.

The additional requirements introduced in Section 3.6.H.5.a-b

and 4.6.H.5 to monitor temperatures between the two loops and the

reactor coolant do not cause an increase in the probability or

consequences of an accident because they limit stresses in the

vessel and primary piping system to acceptable levels.

The procedures that are currently followed at Dresden, Units 2

and 3, regarding inadvertent entrance into a region of instability,

defined as Region A, B, and C in Reference (e), are more

conservative than those recommended by the NRC Bulletin, and do not

allow operation in the stability regions defined in the Dresden

Technical Specifications. Removing Sections 3.6.H.3.b-c and Section

4.6.H.3 only removes outdated material from the Dresden Technical

Specifications and does not increase the probability or consequences

of an accident previously evaluated.

The removal of Section 3.6.H.4 allowing operation without forced

circulation below 25% of rated power at Dresden, Units 2 and 3, will

not increase the probability or consequences of an accident

previously evaluated. This change is conservative because it will

prohibit operation in a condition susceptible to instabilities. This

section also is not included in the Standard Technical

Specifications. In the same manner, Section 2.1.A.4 is removed from

the Quad Cities Technical Specifications.

The removal of Section 3.6.H.3.a from the Dresden Technical

Specifications will not increase the probability or consequences of

an accident, because a one-pump run-up transient is bounded by the

two-pump run-up transient.

The change in initiation time for SLO requirements for Quad

Cities from 12 hours to 24 hours does not represent a significant

change, and still allows adequate time to implement the

requirements. Therefore, no increase in the probability or

consequences of an accident will be caused by this change.

For the reasons stated above, no increase in the probability or

consequences of an accident previously evaluated is introduced by

the proposed changes.

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

different kind of accident from any accident previously evaluated

because:

The proposed change to eliminate the requirement to close the

recirculation suction valve of the idle recirculation loop during

SLO only removes unnecessary conservatism which is not required to

ensure proper LPCI injection into the vessel during a LOCA. The LPCI

loop selection logic already ensures that the intact loop's

recirculation pump discharge valves will close when selected for

LPCI injection. Since all ECCS functions will continue to perform as

designed, no new accident scenarios are created. Also, by requiring

the idle loop to be electrically prohibited from starting, the

possibility of a new event is not created.

Section 3.6.H.5.a-b and 4.6.H.5 provide for temperature

monitoring prior to starting an idle pump, and monitoring to

maintain acceptable primary system stress levels. Since the changes

[do] not adversely affect the performance of any safety related

systems, no new accident scenarios are created.

The change eliminating the requirements for actions when a

region of instability is entered during SLO will not create a new or

different type of accident because procedures are already in place

that are consistent with NRC guidance in this area. These procedures

are more conservative than the current Technical Specifications.

The removal of Section 3.6.H.4 (Dresden) and Section 2.1.A.4

(Quad Cities) does not create the possibility of a new or different

kind of accident because the units will not be allowed to operate

without forced circulation with these sections removed, and Section

3.6.H.4 added. The possibility of accidents occurring from operating

in this mode has been eliminated, and no new types of accidents are

created.

There is no possibility of a new type of accident being created

from the removal of Section 3.6.H.3.a from the Dresden Technical

Specifications. The analysis behind the reduced flow MCPR curves

provides more thermal margin during SLO than two-loop operation,

because the one-pump run-up transient is less severe than the two-

pump run-up transient.

The speed requirement change (Section 3.6.H.3.d for Dresden) for

the operating recirculation pump prior to idle loop startup is in a

conservative direction, and no new types of accidents are created.

The increase in allowed time to initiate SLO requirements for

Quad Cities (Section 2.1.A.4) does not represent a significant

change, and still allows adequate time to implement necessary

requirements. No new types of accidents are created by this change.

The proposed changes do not involve a significant reduction in a

margin of safety because:

The change to eliminate the requirements to close the

recirculation suction valves of the idle loop during SLO maintains

the assumptions of the LOCA analyses. The LPCI loop selection logic

will automatically close the recirculation pump discharge valve of

the unbroken loop to ensure proper LPCI injection. Therefore, the

current MAPLHGR limits at Dresden and Quad Cities will continue to

ensure that Appendix K criteria are satisfied.

During normal dual loop operation, LPCI loop selection logic is

relied upon to close the discharge valve of the unbroken loop

following a LOCA. This function is performed during SLO, provided

the discharge valve and the logic that automatically closes this

valve upon the occurrence of a LOCA signal remain operable. Since

the assumptions of the accident analysis are preserved by the

proposed change, there is no reduction in any safety margin.

The safeguards in place preventing the inadvertent start of an

idle recirculation pump are more than adequate protection against

this transient. Three concurrent failures are required for this

transient to occur. The transient would also be less severe due to

the warmer water in the loop. Therefore, no reduction in a margin of

safety will occur with this change.

The addition of Sections 3.6.H.5.a-b and 4.6.H.5 will not

decrease margin to safety, since the temperature monitoring

requirements will maintain acceptable stresses in the primary system

during idle pump starts.

For Dresden, the current procedures for entrance into a region

of stability provide more margin to safety than the current

Technical Specifications require, because operation in a stability

region is not allowed.

The elimination of Section 3.6.H.4 in the Dresden Technical

Specifications and Section 2.1.A.4 in the Quad Cities Technical

Specifications will not decrease a margin of safety because it

prohibits operations in a potentially unstable region. This change

is in a conservative direction.

The elimination of Section 3.6.H.3.a in the Dresden Technical

Specifications does not cause a decrease in margin to safety,

because there is more thermal margin to SLO than two-loop operation.

Changing the active loop speed requirement from 65% to 43% prior

to idle loop startup for Dresden is in the conservative direction;

therefore, margin to safety is increased.

The increase in allowed time to initiate SLO requirements for

Quad Cities still provides adequate time to implement these

requirements, and is not a significant change. Margin to safety is

not decreased by this change.

Margin of safety does not, therefore, decrease due to the

proposed Technical Specification amendment.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: for 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: James E. Dyer

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities

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

Date of application for amendments: March 11, 1994

Description of amendment requests: The proposed amendments would

provide surveillance requirements for new hydraulic snubbers, which

will be installed on the Main Steam Lines (MSLs) during the current

Unit 1 refueling outage. This outage began on March 13, 1994, and it is

scheduled to end on July 3, 1994. These snubbers will also be installed

on Unit 2 during the Unit 2 refuel outage (Q2R13) currently scheduled

for the first quarter of 1995.

The amendment request would also change the Snubber Visual

Inspection Intervals and Corrective Actions in Technical Specifications

Sections 3.6.1 and 4.6.1 to the format and content of the BWR

Standardized Technical Specifications (STS), as revised by the

provisions of Generic Letter (GL) 84-13 ``Technical Specification for

Snubbers'', dated May 3, 1984 and GL 90-09 ``Alternative Requirements

for Snubber Visual Inspection Intervals and Corrective Actions'', dated

December 11, 1990.

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:

Commonwealth Edison Company has evaluated the proposed Technical

Specification Amendment and determined that it does not represent a

significant hazards consideration. Based on the criteria for

defining a significant hazards consideration established in 10 CFR

50.92, operation of Quad Cities Station Units 1 and 2 (Quad Cities)

in accordance with the proposed amendment will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The proposed changes adopt the format and content of the BWR-

STS, as modified by the provisions of GL 84-13 and GL 90-09. As

such, these proposed changes are administrative in nature and have

no effect on the accident analyses or system operation.

The proposed schedule for snubber visual inspection intervals

described in GL 90-09 will maintain the same level of confidence as

the existing schedule as documented in Generic Letter 90-09,

Alternative Requirements for Snubber Visual Inspection Intervals and

Corrective Actions, dated December 11, 1990. Also, the surveillance

requirement and schedule for snubber functional testing remains the

same providing a 95 percent confidence level that 90 to 100 percent

of the snubbers operate within the specified limits. The proposed

visual inspection schedule is separate from functional testing and

adds to the confidence level that the installed snubbers will serve

their design function and are being maintained operable. Accident

analyses assume that snubbers are initially operable. Compliance

with the Technical Specification Surveillance Requirements for

functional testing in conjunction with the revised visual inspection

schedule assures continued operability of the snubbers. Therefore,

no initial assumptions are being changed and thus neither the

probability nor consequences of any accidents previously evaluated

are significantly increased.

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

accident from any accident previously evaluated because:

The proposed changes adopt the format and content of the BWR-

STS, as modified by the provisions of GL 84-13 and GL 90-09. As

such, these proposed changes are administrative in nature and have

no effect on the accident analyses or system operation.

The proposed schedule for snubber visual inspection intervals

will maintain the same level of confidence as the existing schedule

as documented in Generic Letter 90-09, Alternative Requirements for

Snubber Visual Inspection Intervals and Corrective Actions, dated

December 11, 1990. Also, the surveillance requirement and schedule

for snubber function testing remains the same providing a 95 percent

confidence level that 90 to 100 percent of the snubbers operate

within the specified limits. The proposed visual inspection schedule

is separate from functional testing and adds to the confidence level

that the installed snubbers will serve their design function and are

being maintained operable. As a result, the supported piping,

components, etc. will be maintained operable, so that supported

safety systems will perform as designed. Therefore, the possibility

of a new or different kind of accident is not created.

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

because:

The proposed changes adopt the format and content of the BWR-

STS, as modified by the provisions of GL 84-13 and G[L] 90-09. As

such, these proposed changes are administrative in nature and have

no effect on the accident analyses or system operation. In addition,

the proposed amendment maintains the same level of confidence as the

current technical specification that snubbers are operable through

the current snubber functional testing and the revised snubber

visual inspection

schedule and the associated corrective action requirements. Therefore

the proposed changes do not impact 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: 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: James E. Dyer

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

Plant, Middlesex County, Connecticut

Date of amendment request: February 25, 1994

Description of amendment request: The proposed amendment will add a

new Technical Specification 3/4.7.12, ``Ultimate Heat Sink.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve an SHC [significant hazards

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The ultimate heat sink (Connecticut River) provides the cooling

water necessary to ensure the removal of the normal heat loads and

normal cooldown loads of the plant and to mitigate the effects of

accidents at the plant within acceptable limits. By placing a

technical specification limit on the maximum temperature of the

ultimate heat sink for plant operation, CYAPCO will assure that

sufficient heat removal capacity is available. The intake structure

draws water from the ultimate heat sink for circulation by the

service water system and circulating water system. By adding this

new requirement to the technical specifications, CYAPCO will ensure

that the design basis, as stated in the Final Safety Analysis

Report, for the ultimate heat sink is not violated.

The change does not affect any initiating event. Thus, the

change does not affect the probability of occurrence of any design

basis accidents previously evaluated.

There are no adverse impacts on the design basis accidents due

to the addition of the ultimate heat sink temperature limitation.

This administrative change has no effect on the consequences of the

previously evaluated accidents.

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

from any previously analyzed.

Currently, the Haddam Neck Plant controls the ultimate heat sink

temperature limit to less than 90 deg.F for Mode 1, 2, 3, and 4 via

a plant procedure. The proposed change institutes a technical

specification in place of this administrative control.

As such, the administrative change is consistent with the

current plant practice and has no effect on plant operation. Since

there are no changes in the way the plant is operated, there is no

possibility of an accident of a different type than previously

evaluated due to the change.

3. Involve a reduction in a margin of safety.

The proposed change does not impact the physical protective

boundaries, nor does it affect the performance of safety systems.

There is no degradation in operability and surveillance requirements

for the ultimate heat sink. Therefore, there will be no adverse

impact on the margin of safety as defined in the basis for any

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: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457.

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

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

3499.

NRC Project Director: John F. Stolz

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

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

Date of amendment request: January 28, 1994

Description of amendment request: This amendment request is an

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

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

the Technical Specifications Section 1.0, Definitions, to accommodate a

24-month fuel cycle and which proposed the extension of the test

intervals for specific surveillance tests. This amendment proposes

extending the surveillance intervals to 24 months for leak testing

containment isolation valves. The changes requested by the licensee are

in accordance with Generic Letter 91-04, ``Changes in Technical

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

addition, the request corrects an administrative error.

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. A significant increase in the probability or consequences of

an accident previously evaluated will not occur.

It is proposed that the interval between leakage tests of the

containment isolation valves listed in the Technical Specifications

be revised from 24 months to 24 months (+25%), consistent with an

exemption request to 10 CFR [Part] 50 Appendix J for type C tests

which requests an identical extension in the interval between tests.

The proposed changes do not involve a significant increase in

the probability or consequences of a previously analyzed accident.

These changes propose extending the surveillance intervals for

containment leakage testing. The changes do not involve any physical

changes to the plant or alter the way equipment functions. Other

system testing (e.g., on-line tests) provides assurance of system

operability. An evaluation of past equipment performance provides

additional assurance that the longer surveillance intervals will not

degrade system performance. The 25% increase in the surveillance

interval for type C leak rate testing is compensated for by a

proportionate increase in the margin between specified leakage limit

and the allowable leakage limit. Valves that are sealed with fluid

are exempted from the 10 CFR [Part] 50, Appendix J leakage

requirements. The Technical Specifications establish separate

acceptance criteria for such cases base[d] on system design

considerations. Additionally, in most cases, containment isolation

valve redundancy (two valves in series) provides additional

assurance that actual leakage would be lower than the test results

would indicate.

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

any accident previously evaluated has not been created.

The proposed license amendment does not create the possibility

of a new or different kind of accident. These changes propose

extending the surveillance intervals for containment leakage

testing. The changes do not involve any physical changes to the

plant or alter the way equipment functions. Other system testing

(e.g., on-line tests) provides assurance of system operability. An

evaluation of past equipment performance provides additional

assurance that the longer surveillance intervals will not degrade

system performance. The 25% increase in the surveillance interval

for type C leak rate testing is compensated for by a proportionate

increase in the margin between specified leakage limit and the

allowable leakage limit. Also, containment isolation valve

redundancy (two valves in series) provides additional assurance, in

most cases, that leakage would be lower than the test results would

indicate.

3. Does the proposed amendment involve a significant reduction

in the margin of safety?

The proposed change[s] does[do] not involve a significant

reduction in the margin of safety. These changes propose extending

the surveillance intervals for containment leakage testing. Other

system testing (e.g., on-line tests) provides assurance of system

operability. An evaluation of past equipment performance provides

additional assurance that the longer surveillance intervals will not

degrade system performance. The 25% increase in the surveillance

interval for type C leak rate testing is compensated for by a

proportionate increase in the margin between specified leakage limit

and the allowable leakage limit. Also, containment isolation valve

redundancy (two valves in series), in most cases, provides

additional assurance that leakage would be lower than the test

results would indicate.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

New York, New York 10003.

NRC Project Director: Robert A. Capra

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: November 11, 1993

Description of amendment request: The amendments would change the

Technical Specification (TS) surveillance requirements for the

emergency core cooling system (ECCS) subsystems. Specifically, the

changes would revise the minimum developed head requirement for the

centrifugal charging pumps (CCPs), the safety injection pumps (SIPs),

and the residual heat removal pumps (RHRPs); revise the sum of the

minimum injection flowrates for the CCPs, SIPs, and the RHRPs; and

revise the total maximum pump flowrate (runout limit) for the CCPs and

the SIPs.

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:

(Amendment would not) involve a significant increase in the

probability or consequences of an accident previously evaluated.

The TS [technical specification] changes proposed by this

amendment request are not considered to be initiators of any Design

Basis Accidents (DBA). During normal operation the SIPs and the

RHRPs are in standby, they are not operating. In the event of an

accident resulting in an Engineered Safeguard (ES) actuation, the

pumps would start to provide flow to the reactor vessel. The minor

changes proposed for these pumps (SIPs and RHRPs) would not cause

any accidents or events that have been previously evaluated.

During normal operation, a CCP is operating. The proposed minor

changes provided by this submittal only impact the performance of

these pumps in response to an ES actuation. The proposed changes do

not affect, in any way, how these pumps are operated during normal

operation. As such, the minor changes proposed for the CCPs would

not cause any accidents or events that have been previously

evaluated. Accordingly, the proposed TS changes would not increase

the probability of an accident that has been previously evaluated.

The purpose of the ECCS subsystem is to ensure sufficient flow

is provided to the core in the event of a LOCA [loss of coolant

accident], that is to mitigate the consequences of a LOCA. A LOCA

analysis was performed to determine the impact of the proposed TS

changes. The analysis was performed in accordance with the NRC

approved LOCA methodology for McGuire Nuclear Station. The results

of the analysis demonstrate that the acceptance criteria of 10 CFR

50.46 are still satisfied. Further, the purpose of the proposed TS

changes are to prevent runout of the ECCS subsystem pumps during the

injection and recirculation phases of a LOCA. Accordingly, the

proposed TS changes would not increase the consequences of an

accident that has been previously evaluated.

(Amendment would not) create the possibility of a new or

different kind of accident from any kind of accident previously

evaluated.

The proposed TS changes would not require any modifications to

any structures, systems or components at McGuire Nuclear Station.

Some minor changes to certain testing procedures for the ECCS

subsystem pumps would be necessary. These minor changes would only

involve specific values identified within the procedure and would

not result in any changes on how the test would be performed. No

other changes to procedures on how the station is operated or

maintained would occur. Accordingly, the proposed TS change would

not create a new or different kind of accident than what has been

previously evaluated.

(Amendment would not) involve a significant reduction in a

margin of safety.

The results of the analysis that was performed to determine the

impact of the proposed TS changes would have in mitigating a LOCA

indicate that the acceptance criteria of 10 CFR 50.46 are still

satisfied. The analysis that was performed demonstrate that the Peak

Clad Temperature (PCT) would remain below 2200 deg.F. The proposed

changes ensure that the ECCS subsystem pumps will be operated within

the limits specified by the manufacturer. Accordingly, the proposed

TS changes would not significantly reduce any margins of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: David B. Matthews, Director

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: February 25, 1994

Description of amendment request: The amendments would add four

instruments to the Technical Specification (TS) Tables 3.3-10 and 4.3-7

as part of the accident monitoring instrumentation, and delete five

instruments from the TS Tables that are not part of the accident

monitoring instrumentation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

RESPONSE: No

This proposed change does not involve any significant increase

in the probability or consequences of any accident previously

evaluated because no changes in the types, categories, hardwares and

setpoints of the instruments involved were made; only the

designation of which instruments should be listed in the T/S

[technical specification] Tables and labeled as PAM [Post-Accident

Monitoring] in the control room is changed through this proposed

change.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different kind of

accident from any accident previously evaluated?

RESPONSE: No

This proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated

because no changes in the types, categories, hardwares and setpoints

of the instruments involved were made; only the designation of which

instruments should be listed in the T/S Tables and labeled as PAM in

the control room is changed through this proposed change.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

RESPONSE: No

This proposed change does not involve a significant reduction in

a margin of safety because no changes in the types, categories,

hardwares and setpoints of the instruments involved were made; only

the designation of which instruments should be listed in the T/S

Tables and labeled as PAM in the control room is changed through

this proposed change.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: David B. Matthews, Director

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: March 3, 1994

Description of amendment request: This amendment removes

restrictions from the Arkansas Nuclear One, Unit No. 1 (ANO-1)

technical specifications (TSs) that prohibit use of the auxiliary

building crane to move spent fuel shipping casks.

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 ANO procedures, load paths, crane equipment certification,

operator training and other related heavy load handling topics were

evaluated as part of the control of heavy loads issue and

found acceptable. Spent fuel cask handling is discussed in Section

9.6.2.6 of the ANO-1 SAR [safety analysis report], which shows that the

cask will never travel over spent fuel. ANO-1 SAR Section 9.6.2.6

further evaluates the unlikely event of a cask drop accident and shows

that the consequences are acceptable. Deletion of TS 3.8.15 to allow

handling of a spent fuel shipping cask by the auxiliary building crane

will have no actual impact on the cask drop or any other previously

analyzed accident and therefore, 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 amendment will allow handling of a spent fuel

shipping cask by the auxiliary building crane where [sic] previously

prohibited pending NRC evaluation of the spent fuel cask drop

accident and crane design. The cask handling methods and cask drop

accident are discussed and evaluated in ANO-1 SAR Section 9.6.2.6.

Additionally, the NRC performed an independent evaluation of the

radiological consequences of a cask drop accident, as documented in

the ANO-1 SER [safety evaluation report] dated June 6, 1973. The

evaluation of the unlikely event of a cask drop accident included

assessment of equipment failures and has shown the consequences to

be within acceptable bounds. Since no new accident scenarios can be

identified related to the proposed amendment request, this change is

bounded by the analysis described in the SAR and 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.

Although allowing use of the Auxiliary Building crane where

[sic] previously prohibited by the TS could increase the possibility

of a cask drop accident, the margin of safety is preserved in that

the acceptable consequences of the cask drop accident evaluation in

SAR Section 9.6.2.6 are not affected by this change. The proposed

amendment request will not adversely affect the adequacy and

conservatism of the cask drip accident evaluation. Therefore, the

cask handling issue at ANO-1 continues to exhibit an acceptable

margin of safety and 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

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,

St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: February 22, 1994

Description of amendment request: The proposed amendments will

relocate the instrument response time limits for the Reactor Protective

System and Engineered Safety Features Actuation System from the

Technical Specifications (TS) to the Updated Safety Analysis Report for

both units. The proposed changes are line-item TS improvements and

conform to the guidance given in Enclosures 1 and 2 of NRC Generic

Letter 93-08.

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:

Pursuant to 10 CFR 50.92, a determination may be made that a

proposed license amendment involves no significant hazards

consideration if 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. Each standard is discussed as

follows:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments for St. Lucie Units 1 and 2 simply

relocate tables of response time limits for instrumentation of the

Reactor Protective System (RPS) and Engineered Safety Features

Actuation System (ESFAS) from the Technical Specifications (TS) to

the Updated Safety Analysis Report (UFSAR). The proposed amendments

conform to the guidance given in Enclosures 1 and 2 of USNRC Generic

Letter 93-08 (GL 93-08). Neither the response time limits nor the

surveillance requirements for performing response time testing will

be altered by this submittal. The overall RPS and ESFAS system

functional capabilities will not be changed and assurance that

actions of the protective and engineered safety features systems are

completed within the time limits assumed in the accident analyses is

unaffected by the proposed TS changes. Therefore, operation of the

facility in accordance with the proposed amendment will not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

2. Operation of the facility in accordance with the proposed

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

of accident from any accident previously evaluated.

The proposed amendment will not change the physical plant or the

modes of plant operation defined in the Facility License. The change

does not involve the addition or modification of equipment nor does

it alter the design or operation of plant systems. Therefore,

operation of the facility in accordance with the proposed amendment

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

accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The measurement of instrumentation response times at the

frequencies specified in the TS provides assurance that actions

associated with the protective and engineered safety features

systems are accomplished within the time limits assumed in the St.

Lucie Units 1 and 2 accident analyses. The response time limits, and

the measurement frequencies remain unchanged by the proposed

amendments. The proposed changes do not alter the basis for any

other Technical Specification that is related to the establishment

of or maintenance of a nuclear safety margin. Therefore, operation

of the facility in accordance with the proposed amendment would not

involve a significant reduction in a margin of safety.

Based on the discussion presented above and on the supporting

Evaluation of Proposed TS Changes, FPL has concluded that this proposed

license amendment involves no significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Attorney for licensee: Harold F. Reis, Esquire, Newman and

Holtzinger, 1615 L Street, NW., Washington, DC 20036.

NRC Project Director: Herbert N. Berkow.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: February 28, 1994.

Description of amendment request: The proposed amendments will

delete the minimum frequency criteria prescribed for quality assurance

audits from Administrative Controls sections 6.5.2.8 and 6.8.4 of the

Technical Specifications (TS). Audit periodicity will thereby be

controlled by the program described in the Florida Power and Light

Company (FPL) Topical Quality Assurance Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Pursuant to 10 CFR 50.92, a determination may be made that a

proposed license amendment involves no significant hazards

consideration if 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. Each standard is discussed as follows:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed amendment relocates administrative control criteria

for minimum audit frequencies from the facility Technical

Specifications to the FPL Quality Assurance (QA) Program. The QA

Program is described in the FPL Topical Quality Assurance Report

pursuant to 10CFR50, Appendix B. The change does not alter the bases

upon which assurance is provided that safety-related activities are

performed correctly nor does it involve the conditions and assumptions

utilized in the analyses of plant transients and accidents. Therefore,

operation of the facility in accordance with the proposed amendment

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

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

of accident from any accident previously evaluated.

The proposed amendment will not change the physical plant or the

modes of plant operation defined in the Facility License. The change

does not involve the addition or modification of equipment nor does it

alter the design or operation of plant systems. Therefore, operation of

the facility in accordance with the proposed amendment would not create

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

accident previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendment does not alter the bases for assurance that

safety-related activities are performed correctly or that compliance

with the required Limiting Conditions for Operation will be achieved.

The change does not alter the basis for any Technical Specification

that is related to the establishment of or maintenance of a nuclear

safety margin. Therefore, operation of the facility in accordance with

the proposed amendment would not involve a significant reduction in a

margin of safety.

Based on the discussion presented above and on the supporting

Evaluation of Proposed TS Changes, FPL has concluded that this proposed

license amendment involves no significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Attorney for licensee: Harold F. Reis, Esquire, Newman and

Holtzinger, 1615 L Street, NW., Washington, DC 20036.

NRC Project Director: Herbert N. Berkow.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

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

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

Georgia

Date of amendment request: March 1, 1994.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3.2.4, ``Quadrant Power Tilt

Ratio,'' to add an exception to the requirements of TS 3.0.4.

Specifically, to add ACTION statement ``d. The provisions of

Specification 3.0.4 are not applicable.''

Basis for proposed no significant hazards consideration

determination: In 1992, the Vogtle TSs were amended in accordance with

the recommendations of Generic Letter 87-09 to revise the wording of TS

3.0.4 and to delete from TS 3.2.4 the statement that the provisions of

TS 3.0.4 are not applicable. With the revised wording of TS 3.0.4, the

statement of the non-applicability of the provisions of TS 3.0.4 was

redundant for many individual specifications, and its deletion caused

no change in ACTION requirements. However, in the case of TS 3.2.4, the

deletion had the unintended effect of prohibiting power escalation

above 50% rated thermal power (RTP) whenever the quadrant power tilt

ratio (QPTR) exceeds 1.0.2. This unnecessarily delays power escalation.

The proposed amendment would correct this error and restore the

originally intended meaning of TS 3.2.4. The intent of TS 3.2.4 is to

permit the escalation of reactor power above 50% RTP for limited times

and

under specified conditions when the QPTR is greater than 1.02.

With the original requirements of TS 3.2.4 restored, plant

operation and power escalation during startup would be the same as

previously approved. Therefore, the proposed change (1) does not

involve a significant increase in the probability or consequences or an

accident previously evaluated, (2) does not create the possibility of a

new or different kind of accident than previously evaluated, and (3)

does not involve a significant reduction in the margin of safety.

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 to the Technical Specifications does not

involve a significant increase in the probability or consequences of

an accident previously evaluated because it does not allow operation

in a condition that is not already allowed by the Technical

Specifications.

2. The proposed change to the Technical Specifications does not

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

any accident previously evaluated because it will not allow

operation under conditions different from those already allowed by

the technical specifications.

3. The proposed addition to the Technical Specifications does

not involve a significant reduction in the margin of safety because

the action requirements will continue to be met in the same manner

as currently required by the Technical Specifications.

Accordingly, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

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

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

Georgia 30308.

NRC Project Director: David B. Matthews, Director.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: February 7, 1994.

Description of amendment request: The purpose of the request is to

require the TMI-1 annual radioactive effluent release report for the

previous calendar year be required to be submitted prior to May 1 of

each year. Changing the TMI-1 due date to prior to May 1 can enable the

licensee to combine the reports for TMI-1 and TMI-2 into a single

report with a common due date.

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. Submission of the annual radioactive effluent release report

on or before April 30 each year in accordance with the change, as

compared to the current requirement of March 1, does not involve a

significant increase in the probability of occurrence or the

consequences of an accident previously evaluated. The date on which

the report is due has no impact on plant operations or effluents. It

does not change the control of plant activities or the monitoring of

plant effluents.

2. Operation of TMI-1 in accordance with the proposed amendment

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

accident from any accident previously evaluated. The proposed

amendment has no impact on plant operations, plant effluents or the

control of plant operations or effluents. Therefore, there is no

potential to create a new or different kind of accident.

3. Operation of TMI-1 in accordance with the proposed amendment

does not involve a significant reduction in a margin of safety.

There is no specified margin of safety in regards to the due date

for the annual radioactive effluent release report.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

NRC Project Director: John F. Stolz.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: March 2, 1994.

Description of amendment request: The purposes of the request are

to change the plant Technical Specifications (TS) to modify Operational

Safety Instrumentation requirements to specify completion times which

allows for performance of maintenance or surveillance within a

reasonable time and to be consistent with the allowable outage time for

other safety-related equipment when only one train is affected.

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. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequence of an accident

previously evaluated.

The proposed amendment permits time to restore instrumentation

channels to operable status which is consistent with existing times

allowed for outage of other safety-related equipment affecting one

train. With regard to the 1 hour timeclock, this time is sufficient

to perform the required action necessary to restore minimum required

conditions. Allowing 6 hours to reduce reactor power in an orderly

manner without challenging plant systems is reasonable, based on

operating experience. Thus, the proposed amendment maintains an

adequate degree of equipment availability without requiring

unnecessary initiation of a plant shutdown for partial equipment

outages.

Therefore, it can be concluded that the proposed changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment affects the [Reactor Protection System]

RPS and the [Engineered Safeguards Actuation System] ESAS by

providing timeclocks to perform corrective actions. During this

timeclock period the safety function of the RPS can still be

completed by the remaining minimum required channels. If an accident

occurred while one ESAS train was inoperable due to faulty pressure

switches or a faulty manual actuation channel, the redundant train

would complete the safety function. The proposed time allowed for

the pressure switches in one train or the faulty manual actuation

channel to be out of service is bounded by the allowable time for

other safety-related equipment such that only one train is affected.

The proposed 8 hour timeclock associated with the [Reactor Building]

RB purge radiation monitor, RMA-9, provides adequate time to confirm

a problem exists and perform minor troubleshooting. The containment

isolation function for the RB purge valves would still be maintained

by a redundant 4 psig ESAS signal and a redundant reactor trip

containment isolation signal.

Therefore, the proposed amendment does not create the

possibility of a new or different accident.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

As noted in the Bases for Specification 3.5.1, every reasonable

effort will be made to maintain all safety instrumentation in

operation. If RPS or ESAS instrumentation is found to be inoperable

or require maintenance to assure reliability, the proposed amendment

will allow the performance of maintenance and surveillance in a

reasonable time period. The change does not result in a significant

reduction in a margin of safety for the RPS because the automatic

functions and various alternative manual trip methods are still

available. Also, this change does not result in a significant

reduction in a margin of safety for the ESAS because at least one

train of safety features is required for continued operation within

the specified timeclocks with automatic and manual trip functions.

Thus, operation of the facility in accordance with the proposed

amendment does not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

NRC Project Director: John F. Stolz.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island

Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: March 11, 1994.

Description of amendment request: The purposes of the request are

to (1) change the plant Technical Specifications (TS) to specify an

allowable outage time for the Emergency Feedwater (EFW) Pumps during

surveillance activities and (2) change the requirement to test

redundant components for operability to a requirement to ensure

operability based on verification of completion of appropriate

surveillance activities. 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. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident

previously evaluated.

The proposed amendment specifies an allowable outage time for

testing of the EFW pumps. Also, this proposed change reflects the

current NRC staff position regarding the need for additional testing

to assure OPERABILITY. The allowable outage time this change

provides for EFW pump testing is acceptable because the operator

action required to make the motor driven EFW pump OPERABLE is

minimal and can be performed in a very short time by the Control

Room Operator who is continuously present during the time by the

motor-driven EFW pump is in the Pull-To-Lock position.

The changes affecting OPERABILITY determinations of redundant

train/components for reactor building isolation valves and the

control room air treatment systems reflect the current NRC staff

position. Verifying that the required periodic surveillance testing

is current and there are no known reasons to suggest the redundant

train/component is inoperable, provides adequate assurance of system

OPERABILITY.

Therefore, it can be concluded that the proposed changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment specifies an allowable outage time and

deletes unnecessary redundant equipment testing. These changes do

not change system operational requirements or response to system

transients. Therefore, the proposed amendment does not create the

possibility of a new or different kind of accident.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendment specifies an allowable outage time and

replaces redundant equipment testing with verification that

surveillance is current as an adequate means to ensure OPERABILITY.

These changes do not involve any activities associated with the

margin of safety envelope. Thus, operation of the facility in

accordance with the proposed amendment does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

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

NRC Project Director: John F. Stolz.

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

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

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

County, Texas

Date of amendment request: March 16, 1994.

Description of amendment request: The licensee proposes to amend

the South Texas Project technical specifications (TS) by modifying TS

3.4.9.3, ``Reactor Coolant System--Overpressure Protection Systems,''

Figure 3.4-4, ``Nominal Maximum Allowable PORV Setpoint for the Cold

Overpressure System,'' for the cold overpressure mitigation system

(COMS) with a revised setpoint curve. The proposed amendment would

account for the pressure losses of the reactor coolant flow through the

reactor core with either two or four reactor coolant pumps (RCP)

operating. It was determined that the original COMS setpoint curve

neglected reactor coolant pressure losses due to flow through the

reactor core with the RCPs operating. The resulting pressure at the

reactor vessel downcomer at the elevation equivalent to the core mid-

plane was higher than the pressure at the sensing point located in the

residual heat removal system suction line connected to the reactor

coolant system (RCS) hot leg. The proposed amendment would lower the

power operated relief valve (PORV) setpoint limit by a quantity equal

to the pressure difference between the pressure at the reactor vessel

downcomer at the elevation of the core mid-plane and the pressure at

the location of the residual heat removal system pressure transmitters.

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 a previously evaluated

accident.

The presently existing pressurizer Power Operated Relief Valves

(PORVs) setpoints, provided by the Cold Overpressure Mitigation

System Curve (Figure 3.4-4) of Technical Specification 3.4.9.3, are

nonconservative in that they do not account for reactor coolant

pressure losses due to flow through the reactor core with Reactor

Coolant Pump (RCPs) in operation. When RCP operation is considered,

the pressure at the reactor vessel downcomer, at an elevation

equivalent to core midplane, is higher than the pressure sensing

point located in the Residual Heat Removal System suction line

connection to the Reactor Coolant System (RCS) hot leg. Houston

Lighting & Power Company became aware of this condition and re-

analyzed the Cold Overpressurization Event for the South Texas

Project. The re-analysis has resulted in modifications to Figure

3.4-4 of Technical Specification 3.4.9.3. The re-analysis reduced

the PORVs setpoint to account for the pressure losses and provides

for a setpoint for two Reactor Coolant Pump operation and for four

Reactor Coolant Pump operation.

The proposed decrease in the PORVs setpoint reduces the pressure

versus temperature limit for the RCS under start-up and shut-down

operations. The decreased PORV setpoints for post-overpressure

incidents will ensure that RCS pressure will be maintained within

acceptable limits during low temperature water solid operation for

both two and four pump operation.

The proposed change is based on a re-analysis which accounts for

reactor coolant pressure losses through the reactor core. Reflecting

actual reactor coolant pressure losses and adjusting the PORV

setpoint as necessary has no adverse effect on the probability or

consequences of an accident previously evaluated. Therefore, the

proposed changes not only do not involve a significant increase in

the probability or consequences of an accident previously evaluated,

but actually maintain the original design basis.

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

or different kind of accident from any previously evaluated.

The proposed decrease in the PORV setpoints ensures that

staggered operation of the two PORVs are maintained, thus minimizing

the potential for large pressure undershoots resulting from multiple

valve operation which may compromise the Reactor Coolant Pump No. 1

Seal integrity. It also restricts the total number of discharge

ports at any given moment to that absolutely necessary for pressure

control. In addition, operation of either PORV provides the required

design basis relief capacity and the required redundancy necessary

to meet single failure criteria.

The proposed change is the result of a re-analysis of a

previously evaluated accident. Therefore, the proposed changes do

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

from any accident previously evaluated.

(3) The proposed change does not involve a significant reduction

in the margin of safety.

The proposed change corrects an error present in the original

analysis by accounting for reactor coolant pressure losses through

the reactor core. The revised COMS curves are the result of a re-

analysis of the original COMS analysis. The new analysis preserves

the originally intended margin of safety. 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 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 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, NW, Washington, DC 20036.

NRC Project Director: Suzanne C. Black.

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: March 21, 1994.

Description of amendment request: The licensee proposes to revise

Technical Specifications 3.1.2.3 ``Reactivity Control Systems Charging

Pumps--Shutdown'' and 3.1.2.1 ``Boration Systems Flow Paths--

Shutdown.'' The amendment would allow energizing of an inoperable

centrifugal charging pump in preparation for switching of the

centrifugal charging pumps, provided the pump discharge is isolated

from the reactor coolant system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. 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:

(1) The proposed change does not involve a significant increase in

the probability or consequences of a previously evaluated accident.

The proposed change is to modify the note which permits energizing

of an inoperable centrifugal charging pump for testing purposes,

provided the pump discharge is isolated from the reactor coolant

system, to include pump energization for switching purposes.

The proposed changes can potentially impact two events during Modes

4, 5, and 6: (1) Cold overpressurization of the reactor coolant system,

and (2) boron dilution resulting in a return to criticality. The

requirements of Specification 3.1.2.3 with regard to the cold

overpressure mitigation system analysis would remain valid because the

inoperable centrifugal charging pump would be isolated from the reactor

coolant system. A return to criticality would be prevented because the

action statement of Specification 3.1.2.1 would be entered if the boron

injection flow path could not be restored following centrifugal

charging pump switching. Therefore, allowing energization of an

inoperable pump for switching would have an insignificant effect on the

probability of an overpressurization and boron dilution accident.

Energization of an inoperable pump is currently permitted for

testing purposes provided the pump discharge is isolated from the

reactor coolant system. It is operationally desirable to maintain flow

to the reactor coolant pump seals during the centrifugal charging pump

switching process. This proposed change will not only protect the

reactor coolant system from overpressurization at low temperatures, but

will also provide the capability of maintaining reactor coolant pump

seal injection flow during the switching process.

Therefore, there is no increase in the probability or consequences

of a previously evaluated accident.

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

different kind of accident from any previously evaluated.

Required boron injection flow paths would be maintained in Modes 4,

5, and 6 except during centrifugal charging pump switching. In the

event the boron injection flow path could not be restored after

centrifugal charging pump switching, the action statement of

Specification 3.1.2.1 would be entered. The proposed changes would not

affect the operability of safety-related equipment and reactor coolant

pump seal injection flow could be maintained. The plant operators are

knowledgeable of the potential situation being created by energizing

two centrifugal charging pumps and will follow direct administrative

controls to isolate the pumps from the reactor coolant system.

Therefore, the possibility of a new or different kind of accident is

not created.

(3) The proposed change does not involve a significant reduction in

the margin of safety.

Cold overpressure mitigating system requirements in Specification

3.1.2.3 would continue to be maintained as a result of the proposed

change. Thus, 10 CFR 50 Appendix G limits will not be affected.

Although the boron injection flow path required by Specification

3.1.2.1 may briefly be compromised, there is no significant reduction

in a margin of safety because core alterations would be halted and

positive reactivity changes would not be made if the boron injection

path could not be maintained after centrifugal charging pump switching.

This action, coupled with the short time period required for

centrifugal charging pump switching, would preclude a return to

criticality event. Therefore, there is no significant reduction in a

margin of safety.

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, NW, Washington, DC 20036.

NRC Project Director: Suzanne C. Black.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of amendment requests: February 16, 1994 (Reference LAR 94-

05).

Description of amendment request: The proposed amendments would

revise the combined Technical Specifications (TS) for the Diablo Canyon

Power Plant Unit Nos. 1 and 2 to revise TS 3/4.7.2.1, ``Steam Generator

Pressure/Temperature Limitation,'' 3/4.7.7.1, ``Snubbers,'' 3/4.7.8.1,

``Sealed Source Contamination,'' 3/4.7.11, ``Area Temperature

Monitoring,'' and 3/4.7.13, ``Flood Protection,'' in accordance with

the Commission's Final Policy Statement on TS Improvements for Nuclear

Power Reactors. These TS would be relocated to plant administrative

controls and the final safety analysis report by reference.

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. Do the changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

These proposed changes simplify the TS, meet regulatory

requirements for relocated TS, and implement the recommendations of

the Commission's Final Policy Statement on TS Improvements. Future

changes to these requirements will be controlled by 10 CFR 50.59.

The proposed changes are administrative in nature and do not involve

any modifications to any plant equipment or affect plant operation.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

b. Do the changes create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes are administrative in nature, do not

involve any physical alterations to any plant equipment, and cause

no change in the method by which any safety-related system performs

its function.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

c. Do the changes involve a significant reduction in a margin of

safety?

The proposed changes do not alter the basic regulatory

requirements and do not affect any safety analyses. 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 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: California Polytechnic State

University,

Robert E. Kennedy Library, Government Documents and Maps Department,

San Luis Obispo, California 93407.

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and

Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Project Director: Theodore R. Quay.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of amendment requests: February 16, 1994 (Reference LAR 94-

04).

Description of amendment request: The proposed amendments would

revise the combined Technical Specifications (TS) for the Diablo Canyon

Power Plant Unit Nos. 1 and 2 to revise TS 4.2.2, ``Heat Flux Hot

Channel Factor--FQ(z),'' and 6.9.1.8, ``Core Operating Limits

Report,'' to implement the revised methodology for calculating the

penalty to FQ(z). The specific TS changes proposed are as follows:

(1) The 2 percent FQ(z) penalty listed in TS 4.2.2.2.e.1)

would be deleted and the statement revised to indicate the use of an

appropriate factor to be specified in the Core Operating Limits Report

(COLR).

(2) TS 6.9.1.8.b.1. would be changed to reference Revision 1 of

WCAP 10216-P-A, ``Relaxation of Constant Axial Offset Control

FQ(z) Surveillance Technical Specification,'' dated February 1994.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

a. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The Heat Flux Hot Channel Factor, FQ(z), is not involved in

the initiation of any accident. Verifying FQ(z) is below its

limit ensures initial conditions for accident analyses are met. The

proposed changes have been previously approved by the NRC and

provide for application of a more conservative FQ(z) penalty

which will ensure that possible FQ(z) margin decreases are

adequately accounted for. Therefore, if the FQ(z) does exceed

its limit, the appropriate actions in TS 3.2.1 and TS 3.2.2. will be

taken and are adequate to ensure design basis accidents analyses

assumptions are met.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

FQ(z) is not involved in the initiation of any accident.

The FQ(z) surveillance provides assurance that the initial

conditions for accident assumptions are met. FQ(z) is a

measurement of a physical property and is not involved in the

initiation of any accident.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

of safety?

The FQ(z) surveillance ensures that certain core parameters

are maintained consistent with supporting assumptions regarding the

core for postulated accidents. The methodology used in Revision 1 to

WCAP-10216-P-A adequately accounts for FQ(z) increases between

monthly flux maps. Using the methodology of Revision 1 to WCAP-

10216-P-A results in a FQ(z) penalty which is more conservative

than the current TS FQ(z) penalty of 2 percent. If the

FQ(z) increases above the TS limit, appropriate actions in TS

3.2.1 and TS 3.2.2. are adequate to ensure design basis accidents

analyses assumptions are met.

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 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: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and

Electric Company, P.O. Box 7442, San Francisco, California 94120

NRC Project Director: Theodore R. Quay

Portland General Electric Company, et al., Docket No. 50-344, Trojan

Nuclear Plant, Columbia County, Oregon

Date of amendment request: January 27, 1993, revised March 8, 1994.

Description of amendment request: The proposed amendment, as

revised, by Portland General Electric Company, PGE or the licensee,

would change the Trojan Nuclear Plant (Trojan) Appendix A Technical

Specifications to reflect the permanently defueled status of the

facility. The permanent cessation of power generation at Trojan and the

May 5, 1993 amendment to the license which granted the licensee a

Possession Only License for the facility has rendered many of the

existing provisions of the current Appendix A Technical Specifications

inappropriate. PGE has developed Permanently Defueled Technical

Specifications (PDTS) for Trojan using NUREG-1431, ``Standard Technical

Specifications, Westinghouse Plants,'' as a basis for the PDTS scope

and format.

Based on a series of discussions between the PGE staff and the NRC

on February 9, 1994 and February 28, 1994 the licensee has revised

several requirements contained in the original June 27, 1993 amendment

request. These revisions were forwarded to the NRC staff by letter

dated March 8, 1994. The March 8, 1994 revision updates the June 27,

1993 submittal deleting reference to sections that had been relocated

out of the Technical Specifications by amendments granted since June

1993. It also provided supplemental information concerning the deletion

and/or relocation of certain existing Trojan Technical Specifications

requested by the NRC staff. The March 8, 1994 submittal also clarified

the long term organization at the site, established a line of

succession for the operational command and control function, modified

the review and audit functions performed by the Independent Review and

Audit Committee assuring their independence of review, required

independent review of certain programs and manuals, limited annual and

quarterly doses and operability and usage of the effluent treatment

systems to conform to Appendix I to 10 CFR Part 50, continued the

existing requirement for surveillance of former structural

modifications to the facility, and required the submission of an annual

radioactive effluent release report in accordance with 10 CFR 50.36a.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.92(a), the licensee has

provided an analysis of the issue of no significant hazards

consideration. In accordance with the requirements of 10 CFR 50.92,

Issuance of Amendment, this license amendment request, as revised, is

judged to involve no significant hazards consideration based upon the

following:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The licensee analysis shows that the worst case design basis

accident for this plant, in its permanently shutdown defueled state,

is a fire in the radioactive waste annex building. The licensee has

also identified a second design basis accident scenario, a fuel

handling accident in the vicinity of the Trojan spent fuel pool.

Other Trojan Final Safety Analysis Report (FSAR) accident scenarios

addressed in Chapter 15 are no longer applicable to Trojan in the

permanently defueled mode. The proposed amendment, as revised, does

not lessen any of the requirements associated with either the

radioactive waste annex building or the spent fuel pool therefore

the probability of either accident occurring is unchanged. The

proposed amendment, as revised, does not change the consequences of

the accident since it does not affect the magnitude, detection, or

mitigation of either accident scenario. Additionally, the ability of

the radioactive waste annex building and the spent fuel pool to

withstand other applicable FSAR events, natural phenomena, and fires

is either unchanged from the existing licensing basis or is improved

during the permanently defueled condition.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

Maintaining the permanently defueled facility in accordance with

the PDTS, as revised by the March 8, 1994 letter, does not create

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

previously considered. Most of the existing plant systems and

functions will not be operational in the permanently defueled

condition since power operations are prohibited and all of the fuel

at Trojan is stored in the spent fuel pool. However,

all structures, systems and components that are necessary for safe fuel

handling and storage activities will be maintained operable during the

permanently defueled condition. The proposed PDTS, as revised, provide

operation and surveillance requirements and administrative controls

which are sufficient to ensure that the required structures, systems

and components will be maintained operable in the permanently defueled

condition.

3. Operation of the facility in accordance with the proposed

amendment does not involve a significant reduction in a margin of

safety.

The proposed PDTS, as revised by the March 8, 1994 letter, are

sufficient to ensure no reduction in a margin of safety, in part,

because of the reduced range of design basis accidents against which

the facility must be protected now that the facility is prohibited

from power operations and is permanently defueled. Only a fire in

the radioactive waste storage facility or a fuel handling accident

are relevant during the permanently defueled condition. The margins

of safety for both of these accidents will remain the same or

improve by maintaining the facility in accordance with the proposed

PDTS, as revised. None of the other Chapter 15 FSAR accidents are

applicable since power operations are prohibited and the facility is

permanently defueled. Additionally, the margins of safety for other

applicable FSAR events, natural phenomena, and fires are either

unchanged from the existing licensing basis or is improved during

the permanently defueled condition.

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, as revised by the March 8, 1994 supplement, involves

no significant hazards consideration.

Local Public Document Room location: Branford Price Millar Library,

Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,

Portland, Oregon 97207.

Attorney for licensee: Leonard A. Girard, Esq., Portland General

Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.

NRC Project Director: Seymour H. Weiss

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama

Date of amendments request: October 14, 1993

Description of amendments request: The proposed changes would

revise surveillance test intervals and allowed outage times for reactor

trip system (RTS) and engineered safety feature actuation system

(ESFAS) instrumentation. The proposed changes would also revise certain

RTS/ESFAS functions, minimum channels operable, channel calibration,

and channel functional test requirements to ensure they are in concert

with the Westinghouse Standard Technical Specifications and WCAP-10271,

``Evaluation of Surveillance Frequencies and Out-of-Service Times for

Reactor Protection Instrumentation Systems.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes to the RTS/ESFAS STIs [surveillance

test intervals] and AOTs [allowed outage times], and Minimum

Channels Operable, Channel Calibration and Channel Functional Test

requirements will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The determination that the result of the proposed changes

associated with STI and AOT are within all acceptable criteria

[that] has been established in the SERs [Safety Evaluations]

prepared for WCAP-10271; WCAP-10271 Supplement 1; WCAP-10271

Supplement 2; and WCAP-10271 Supplement 2, Revision 1.

Implementation of the proposed changes results in a slight increase

in the Reactor Trip System yearly unavailability. This slight

increase, which is primarily due to less frequent surveillance,

results in a slight increase in Core Damage Frequency (CDF) and

public health risk. The values determined by the WOG and presented

in the WCAP for the increase in CDR were verified by Brookhaven

National Laboratory (BNL) as part of an audit and sensitivity

analyses for the NRC staff. Based on the small value of the increase

compared to the range of uncertainty in the CDF, the increase is

considered acceptable. Increasing STIs and AOTs is not expected to

affect the probability or consequences of previously evaluated

accidents.

The change associated with the Minimum Channel Operable

requirement for the RTS Turbine Trip by Turbine Throttle Valve

Closure provides additional operating flexibility based on the

Westinghouse Standard Technical Specifications, Revision 5. The new

action statement ensures that any inoperable channel is placed in

trip, and the remaining operable channels fulfill the necessary

reactor trip diversity function. The change associated with the

ESFAS Minimum Channel Operable requirement for Containment

Pressure--High-High assures that the Technical Specifications

reflect the correct as-built design actuation logic while ensuring

the function continues to meet the single failure criteria. The

proposed change to the turbine trip reactor trip function Channel

Calibration reflects the assumptions in WCAP-10271 and current

Farley calibration practices. The proposed change to the safety

injection ESF [engineered safety feature] input for the reactor trip

Channel Functional Test is consistent with the assumptions in WCAP-

10271 and current Farley surveillance testing practices. The

proposed changes to the ESF permissive interlocks Channel

Calibration and Channel Functional Test requirements are in concert

with the NRC SER for WCAP-10271 and Farley surveillance practices.

The proposed change to the ESF manual initiation functions Channel

Functional Test reflects the proper surveillance requirements for a

Westinghouse Solid State Protection System (SSPS). The proposed

changes to these RTS/ESFAS Channel Calibration and Channel

Functional Test requirements are also consistent with Westinghouse

Standard Technical Specifications.

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

new or different kind of accident from any accident previously

evaluated. The proposed changes to not involve hardware changes and

do not result in a change in the manner in which the RTS/ESFAS

provides plant protection or the manner in which surveillance

testing is perform to demonstrate operability. Therefore, a new or

different kind of accident will not occur as a result of these

changes.

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

in a margin of safety. The proposed changes do not alter the manner

in which safety limits, limiting safety system setpoints, or

limiting conditions for operation are determined. The impact of

reduced testing, other than as addressed above, is to allow a longer

time interval over which instrument uncertainties (e.g., drift) may

act. Evaluations have been performed to assure that the plant

setpoints properly account for these instrument uncertainties over

the longer time interval. RTS diversity is still provided by the

Turbine Throttle Valve closure logic circuits. Steam Line Isolation

diversity continues to be provided by the ESFAS Containment

Pressure--High-High. Changes to certain RTS/ESFAS Channel

Calibration and Channel Functional Test surveillances clarify what

tests are required and when the tests are performed. Implementation

of the proposed changes is expected to result in an overall

improvement in safety as noted below.

a. Less frequent testing will potentially result in fewer

inadvertent reactor trips and ESF component actuation.

b. Longer allowed outage times provide for better assessments of

problems and easier repairs, ultimately resulting in better

equipment performance.

c. Less frequent distraction of the plant operator and shift

supervisor to attend to and support instrumentation testing will

improve the effectiveness of the operating staff in monitoring and

controlling plant operation.

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: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Attorney for licensee: James H. Miller, III, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201

NRC Project Director: S. Singh Bajwa

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power

Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: June 29, 1992, as supplemented on

February 22, 1994

Description of amendment request: The proposed amendment would

modify the following Technical Specifications and their associated

bases to permit longer allowable outage times (AOT) and increase

surveillance testing intervals from monthly to quarterly: 3/4.3.1

``Reactor Protection System Instrumentation;'' 3/4.3.2 ``Isolation

Actuation Instrumentation;'' 3/4.3.3 ``Emergency Core Cooling System

Actuation Instrumentation;'' 3/4.3.4. ``Recirculation

Pump Trip Actuation Instrumentation ATWS Recirculation Pump Trip System

Instrumentation;'' 3/4.3.5 ``Reactor Core Isolation Cooling System

Actuation Instrumentation;'' 3/4.3.6 ``Control Rod Block

Instrumentation;'' 3/4.3.9 ``Plant Systems Actuation Instrumentation;''

and 3/4.4.2 ``Safety Valves.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The staff's review is

presented below.

The proposed changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated because

the proposed changes do not involve a change to the plant design or

operation. The changes simply involve the frequency at which testing of

the instrumentation is performed and the AOT for instruments. There may

be a small increase in average instrument failure frequency as a result

of increasing the surveillance interval. However, the proposed changes

will require that a check be made for the majority of instruments to

assure that making them inoperable for surveillance testing or repair

does not result in a loss of function. This added check assures that a

loss of function has not occurred, or if it has, that the appropriate

ACTION statement be entered promptly. Therefore, these proposed changes

do not result in a significant increase in either the probability or

consequences of any accident previously analyzed.

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

different kind of accident from any accident previously evaluated

because the proposed changes do not result in any change in the plant

design or operation. The changes are for increased AOT and decreased

frequency of instrumentation testing. The instrumentation involved are

those instruments which sense plant problems and/or accidents, and then

initiate systems or alarms to respond to the plant problem/accident.

The proposed changes do not modify any of the instruments, or the

initiation logic formed by the instruments. Therefore, no new or

different type of an accident has been created.

The proposed changes do not involve a significant reduction in a

margin of safety because the small increase in average instrument

failure frequency is offset by safety benefits such as a reduction in

the number of inadvertent test-induced scrams, a reduction in wear due

to excessive equipment test cycling, and better optimization of plant

personnel resources. In addition, the proposed changes will require

that a check be made for the majority of instruments to assure that

making them inoperable for surveillance testing or repair does not

result in a loss of function. This added check assures that a loss of

function has not occurred, or if it has, that the appropriate ACTION

statement be entered promptly. Therefore, these proposed changes do not

result in a significant reduction in a margin of safety.

Based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW., Washington, D.C. 20037.

NRC Project Director: John N. Hannon.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: November 15, 1993

Brief description of amendments: The proposed amendment would

revise the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2

technical specifications to increase the maximum permitted power at

which the post-refueling power ascension reactor coolant system (RCS)

flow verification can be performed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of a previously evaluated accident.

The proposed change increases the power at which the

verification of the Reactor Coolant System flow rate with a

precision heat balance can be performed. The only potentially

relevant concern for this change is the possibility of having

insufficient flow to support the accident analyses at the higher

(85%) power level. Power level and RCS flow are important parameters

in determining the severity of an event but have no impact on the

initiation of an event or accident. Thus, the change does not

involve a significant increase in the probability of any previously

analyzed accident.

Although accidents tend to be more severe at higher initial

power levels, the acceptance criteria of the applicable safety

analyses continue to be met [even when the test is conducted at the

higher power level]. Thus, the proposed change would not involve an

increase in the consequences of any previously analyzed accident.

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

or different kind of accident from any accident previously

evaluated.

The proposed change merely increases the power at which the

initial post-refueling startup verification of RCS flow with a

precision heat balance may be performed. Since the new power level

is within the normal operating range of the reactor, it does not

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

any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The proposed change increases the power which can be attained

prior to verification of the Reactor Coolant System flow with a

precision heat balance. The power value is increased from 75% RTP

[rated thermal power] to 85% RTP. An analysis was performed which

demonstrates that the safety analysis DNB [departure from nucleate

boiling] limit will not be exceeded if the initial power is 85% or

less, even with a significant reduction in flow. A flow reduction

significantly different from the expected flow is highly unlikely,

however, since RCS flow is verified by measurement of elbow tap

differential pressure prior to operation in Mode 1. This flow

measurement, although less accurate than the precision heat balance,

is sufficient to assure adequate flow at 85% power. Adequate

limitations on power level, F delta-H verification, and the power

range neutron flux-high setpoint are imposed during post-refueling

power ascension to ensure that if the RCS flow is not verified with

a precision heat balance until 85% RTP, the results of the accident

analyses would remain valid.

The evaluation of the DNB-limited events, initiated from a power

level of 85% RTP, considered the limitations imposed during the

power ascension. Based on this evaluation, it is concluded that,

even though 85% RTP is a more severe initial condition, the

applicable event acceptance criteria would continue to be met;

therefore, the proposed 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: University of Texas at

Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.

Box 19497, Arlington, Texas 76019

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036

NRC Project Director: Suzanne C. Black

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: March 30, 1994

Brief description of amendments: The proposed amendments would

revise the Comanche Peak Steam Electric Station (CPSES) Units 1 and 2

Technical Specifications to allow the use of an alternative method for

verifying that the emergency diesel generator fuel oil meets

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:

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

the probability or consequences of a previously evaluated accident.

Either the current testing option, which remains valid for fuel

oil with ASTM Color 5 or less, or the alternative testing method,

provide the necessary assurance that the water and sediment quantity

in the new fuel oil is acceptable. As the performance of the

quantitative water and sediment test of ASTM-D1796-1968, maintains

essentially the same attribute qualities of the new fuel oil, there

should not be any undetected degradation in the Diesel Generator

fuel oil supply.

Therefore, since the fuel oil supply will be maintained at its

present quality level, there should be no increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

As stated in (1) above, the proposed amendment only provides a

quantitative method for the acceptability determination of new

Diesel Generator fuel oil. The proposed testing will continue to

verify the high quality and acceptability of the fuel oil supply.

There should not be any possibility that a new or different kind of

accident from those previously evaluated is created.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The only margin associated with this amendment is the margin

between the acceptance limit on water and sediment in the fuel oil

supply and the quantity of water and sediment that could impact

Diesel Generator operation. The ``clear and bright'' testing per

ASTM-D4176-1982 and the proposed quantitative testing per ASTM-

D1796-1968 are both written to detect and reject fuel oil containing

water or sediment at essentially the same level. The margin of

safety to Diesel Generator impairment is therefore not reduced. This

amendment does not involve a significant reduction in a margin of

safety.

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

this review, it appears

that the three standards of 10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that the amendment request involves

no significant hazards consideration.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 701 South Cooper, P.O.

Box 19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036.

NRC Project Director: Suzanne C. Black.

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

Callaway County, Missouri

Date of amendment request: February 21, 1992

Description of amendment request: The proposed amendment would

revise Technical Specifications Appendix B, Environmental Protection

Plan (Non-radiological), by removing Sections 2.3 and 4.3, ``Cultural

Resources.'' Union Electric has developed and maintains a management

plan for the protection of cultural resources on the Callaway Plant

site. The amendment request summarizes the plan that provides the

status and disposition of each portion of the current Appendix B

sections related to cultural resources.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change[s] do[es] not involve a significant hazards

consideration because operation of Callaway Plant in accordance with

these change[s] would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated (10 CFR

50.92(c)(1)).

The proposed changes are administrative in nature and have no

impact on safety-related structures, systems or components.

Therefore, there is no impact on the probability of occurrence or

the consequences of an accident or malfunction of equipment

previously evaluated.

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

from any previously evaluated (10 CFR 50.92(c)(2)).

The proposed change[s] do[es] not affect any of the assumptions

used in previous accident evaluations. All accidents continue to be

bounded by previous analyses and deletion of satisfied Environmental

Protection Plan requirements is administrative and will therefore

not introduce the possibility of any new or different kind of

accident.

3. Involve a significant reduction in a margin of safety (10 CFR

50.92(c)(3)).

The proposed changes are administrative and do not affect the

margin of safety as defined in the basis for any Technical

Specification.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

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

NRC Project Director: John N. Hannon.

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

Callaway County, Missouri

Date of amendment request: July 16, 1993

Description of amendment request: The amendment would revise

Technical Specifications 3/4.8.1.1 and 3/4.8.1.2, A. C. Sources,

Operating and Shutdown. The proposed revision changes the minimum

required storage volume of the Emergency Fuel Oil storage and day tanks

from 85,300 gallons and 390 gallons to 80,400 gallons and 510 gallons.

These changes are the result of inconsistencies found by Union

Electric, in the calculations and T/S Bases for tank capacities, while

performing a self-initiated Electrical Distribution System Functional

Assessment (EDSFA). The Electrical Distribution System Functional

Inspection (EDSFI) performed by the NRC re-examined this issue and it

was determined that NUREG-1431 provided bases for the change to day

tank level. This submittal is a T/S enhancement. 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 to Technical Specifications 3/4.8.1.1 and

3/4.8.1.2 do not involve a significant hazards consideration because

operation of Callaway Plant with these changes would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The current minimum required volume of the day tanks as given in

T/S 3.8.1.1b.1 and 3.8.1.2b.1 is based on fuel oil with a specific

gravity value in the upper range of values allowed by T/S.

Calculations made using fuel oil with a specific gravity of 39

degrees API, which is the minimum allowed specific gravity, indicate

a larger minimum required volume is needed in the day tanks. The

increased minimum required volume provides additional conservatism

for the day tanks to perform their intended safety function based on

the possibility of using different specific gravity fuel oil. The

current minimum required volume of the storage tanks as given in T/S

3.8.1.1b.2 and 3.8.1.2b.2 was calculated by using an overly

conservative Net Positive Suction Head (NPSH) available calculation.

The original calculation assumed a static head was needed in order

for the NPSH available to exceed the NPSH required and assure proper

operation of the fuel oil transfer pumps. Union Electric

calculations have determined that the NPSH required for the transfer

pumps is maintained as long as the pump suction is submerged. Based

on this calculation the minimum required volume of the storage tanks

can be reduced and the diesel generators can still achieve their

required seven days of operation at continuous rating plus an

additional volume available to be used for testing.

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

accident from any previously evaluated.

There is no new type of accident or malfunction being created

and the method and manner of plant operation remains unchanged. The

safety design bases in the FSAR have not been altered, the

requirement for continuous operation remains unchanged.

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

There are no plant design changes involved and no changes are

being made to the safety limits or safety system settings that would

adversely impact plant safety. The minimum required storage volumes

of the day and storage tanks are being changed based on calculations

using conservative data.

Based on the above discussions, it has been determined that the

requested Technical Specification revisions do not involve a

significant increase in the probability or consequences of an

accident or other adverse condition over previous evaluations; or

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

condition over previous evaluations; or involve a significant

reduction in a margin of safety. Therefore, the requested license

amendment does not involve a significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

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

NRC Project Director: John N. Hannon.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: February 25, 1994

Description of amendment request: The proposed change would revise

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

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

modify the surveillance frequency of the nozzles in the Quench Spray

and the Recirculation Spray Systems at NA-1&2.

The proposed changes to the surveillance requirements for the

nozzles in the Quench Spray System and Recirculation Spray Systems are

consistent with the intent of Generic Letter 93-05, ``Line-Item

Technical Specifications Improvement to Reduce Surveillance

Requirements for Testing During Power Operation,'' dated September 27,

1993, which is to improve safety, decrease equipment degradation, and

reduce unnecessary burden on personnel resources by reducing testing

requirements that are marginal to safety.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Specifically, operation in accordance with the proposed

Technical Specifications changes will not:

1. Involve a significant increase in the probability of

occurrence or consequences of an accident previously evaluated.

The proposed reduced testing frequency of the Spray Systems

nozzles does not change the way the systems are operated or the

Spray Systems operability requirements. The proposed change to the

surveillance frequency of safety equipment has no impact on the

probability of an accident occurrence nor can it create a new or

different type of accident. NUREG-1366, ``Improvements to Technical

Specifications Surveillance Requirements,'' dated December 1992,

concluded that the corrosion of stainless steel piping is negligible

during the extended surveillance interval. Since the Spray Systems

are maintained dry there is no additional mechanism that could cause

blockage of the spray nozzles. Thus, the nozzles in the Spray

Systems will remain operable during the ten year surveillance

interval to mitigate the consequence of an accident previously

evaluated. To date, no clogging or blockage of the nozzles in the

Spray [System] during the five year surveillance tests [has] been

observed at . . . North Anna. Testing of the Spray Systems nozzles

at the proposed reduced frequency will not increase the probability

of occurrence of a postulated accident 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 reduced frequency testing of the Spray Systems'

nozzles does not change the way the Spray Systems are operated. The

reduced frequency of testing of the spray nozzles does not change

plant operation or system readiness. The reduced frequency testing

of the Spray Systems' nozzles does not generate any new accident

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

accident previously evaluated is not created by the proposed changes

in surveillance frequency of the Spray Systems nozzles.

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

Reduced testing of the Spray Systems' nozzles does not change

the way the Systems are operated or the Spray Systems' operability

requirement. NUREG-1366 concluded that the corrosion of stainless

steel piping is negligible during the extended surveillance

interval. Since the Spray Systems are maintained dry there is no

additional mechanism that could cause blockage of the Spray Systems'

nozzles. Thus, the proposed reduced testing frequency is adequate to

ensure spray nozzle operability. The surveillance requirements do

not affect the margin of safety in that the operability requirements

of the Spray Systems remains unaltered. The existing safety analysis

remains bounding. Therefore, no margins of safety are adversely

affected.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: Herbert N. Berkow.

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

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

Date of amendment request: February 25, 1994

Description of amendment request: The proposed changes would revise

the surveillance frequency from 5 years to 10 years for the spray

nozzles in the containment spray and recirculation spray systems.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Specifically, operation of Surry [Units 1 and 2] in accordance

with the proposed Technical Specifications changes will not:

1. Involve a significant increase in the probability of

occurrence or consequences of an accident previously evaluated.

The proposed reduced testing frequency of the Spray Systems'

nozzles does not change the way the systems are operated or the

Spray Systems' operability requirements. The proposed change to the

surveillance frequency of safety equipment has no impact on the

probability of an accident occurrence nor can it create a new or

different type of accident. NUREG-1366, ``Improvements to Technical

Specifications Surveillance Requirements,'' dated December 1992,

concluded that the corrosion of stainless steel piping is negligible

during the extended surveillance interval. Since the Spray Systems

are maintained dry there is no additional mechanism that could cause

blockage of the spray nozzles. Thus, the nozzles in the Spray

Systems will remain operable during the 10 year surveillance

interval to mitigate the consequence of an accident previously

evaluated. To date, no clogging or blockage of the nozzles in the

Spray Systems during the five year surveillance tests have been

observed at Surry. Testing of the Spray Systems' nozzles at the

proposed reduced frequency will not increase the probability of

occurrence of a postulated accident 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 reduced frequency testing of the Spray Systems'

nozzles does not change the way the Spray Systems are operated. The

reduced frequency of testing of the spray nozzles does not change

plant operation or system readiness. The reduced frequency testing

of the Spray Systems' nozzles does not generate any new accident

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

accident previously evaluated is not created by the proposed changes

in surveillance frequency of the Spray Systems nozzles.

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

Reduced testing of the Spray Systems' nozzles does not change

the way the Systems' are operated or the Spray Systems' operability

requirement. NUREG-1366 concluded that the corrosion of stainless

steel piping is negligible during the extended surveillance

interval. Since the Spray Systems are maintained dry there is no

additional mechanism that could cause blockage of the Spray Systems'

nozzles. Thus, the proposed reduced testing frequency is adequate to

ensure spray nozzle operability. The surveillance requirements do

not affect the margin of safety in that the operability requirements

of the Spray Systems remains unaltered. The existing safety analysis

remains bounding. Therefore, no margins of safety are adversely

affected.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: Herbert N. Berkow.

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of amendment request: February 17, 1994

Description of amendment request: The amendment proposes to modify

the Technical Specifications (TS) to reflect management and

organizational changes at the Washington Public Power Supply System

(the licensee) for operation of the WNP-2 facility. The proposed

changes would (1) modify the reporting responsibility of the quality

assurance organization from the Managing Director to the Assistant

Managing Director, Operations (AMDO), and (2) modify the appointment

authority for the Corporate Nuclear Safety Review Board (CNSRB) from

the Managing Director to the AMDO. These changes are proposed to

reflect the current designation of the AMDO as the licensee's

designated official with corporate responsibility for overall plant

nuclear safety, and as the direct report for the CNSRB.

In addition, the proposed change would (1) delete the specific

requirement for health physics/chemistry program procedures, (2) modify

the titles of two positions on the Plant Operations Committee (POC) to

reflect revised organizational titles, (3) modify the CNSRB quorum

requirements from nine personnel to a minimum of nine personnel, and

(4) delete the requirement that the CNSRB Executive Secretary be

designated from the CNSRB membership.

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:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

All of the proposed changes are administrative in nature, and do

not involve any change in the design or operation of the plant. None of

the changes affect any initiating events, nor do they affect plant

response to postulated events already analyzed. The proposed changes do

not, therefore, affect the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes do not affect the design, operation,

maintenance, or testing of the plant. They do not, therefore, create

the possibility for any new or different kind of accident from any

accident previously evaluated.

3. Does the amendment involve a significant reduction in a margin

of safety?

The proposed changes do not affect any accident analyses, and do

not, therefore, affect any of the margins of safety affected by the

design or operational limitations of the plant. The proposed changes do

not, therefore, affect any margin of safety.

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: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

Attorney for licensee: Nicholas S. Reynolds, Esq., Winston &

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

NRC Project Director: Theodore R. Quay

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: February 23, 1994

Description of amendment request: The proposed amendment revises

Technical Specification 3.8.1.1, AC Sources Operating, and 3.8.1.2, AC

Sources Shutdown, to increase the minimum volume of fuel oil required

for the emergency diesel generator fuel oil day tanks. Several other

changes have been proposed to correct editorial errors related to

previously issued amendments.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee 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 minimum required volume of the day tanks as given in

Technical Specifications 3.8.1.1 and 3.8.1.2 is based on fuel oil

with a specific gravity value in the upper range of values allowed

by the technical specifications. Calculations made using fuel oil

with a specific gravity of 39 degrees API, which is the minimum

allowed specific gravity, indicate a larger minimum required volume

is needed in the day tanks. The increased minimum required volume

provides additional conservatism for the day tanks to perform their

intended safety function based on the possibility of using different

specific gravity fuel oil. The proposed change will not prevent the

Emergency Diesel Generator fuel oil system from performing its

design function, nor require the system to be operated in a manner

different than that for which it was designed. Therefore, the

proposed change will 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 previously evaluated. There

is no new type of accident or malfunction being created and the

method and manner of plant operation remains unchanged. The safety

design bases in the Updated Safety Analysis Report (USAR) have not

been altered, and current operating requirements of the Emergency

Diesel Generators remain unchanged. Thus, this change does not

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

any previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The basis for the existing emergency fuel oil day tank level

requirements is to ensure that sufficient fuel oil is available to

meet the operational requirements specified in ANSI N195-1976. This

proposed change to the minimum required storage volume of the day

tanks is based on revised calculations performed in accordance with

ANSI N195-1976 using conservative data. This proposed change will

not change the operation of the plant. Thus, the proposed change

will continue to ensure the Emergency Diesel Generator operating

requirements. There are no changes being made to the safety limits

or safety system settings that would adversely impact plant safety.

Therefore, the proposed change will not cause a reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and

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

NRC Project Director: Suzanne C. Black

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: February 24, 1994

Description of amendment request: The proposed amendment revises

Technical Specification 3.9.4, Containment Building Penetrations, to

allow use of temporary alternate closure methods for the emergency

personnel escape lock and containment wall penetrations, during

alterations of the core or movement of irradiated fuel within the

containment.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The probability of occurrence of a previously evaluated accident

is not increased because failure to maintain containment closure is

not an initiating condition for fuel handling an accident. The use

of temporary alternate closure methods for the emergency personnel

escape lock and containment wall penetrations does not introduce any

new potential accident initiating condition during refueling

operation.

The consequences of an accident previously evaluated is not

increased because the use of

temporary alternate closure methods for the emergency personnel escape

lock and containment wall penetrations will provide the assurance of

containment closure during refueling activities. The ability of [the]

containment to restrict the release of any fission product

radioactivity to the environment remains unchanged.

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

or different kind of accident from any previously evaluated.

The failure of the temporary alternate closure methods for the

emergency personnel escape lock and containment wall penetrations

during refueling will not result in a malfunction of any other plant

equipment. The sole purpose of establishing containment closure for

refueling is to restrict the release of any fission product

radioactivity in the event of a fuel handling accident.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The temporary alternate closure methods for the emergency

personnel escape lock and containment wall penetrations will provide

the same assurance of containment closure during refueling for

credible accident scenarios. The ability of containment to restrict

the release of any fission product radioactivity to the environment,

should a fuel handling accident occur, remains unchanged.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and

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

NRC Project Director: Suzanne C. Black

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: February 24, 1994

Description of amendment request: The proposed amendment revises

Technical Specification 4.7.1.2.1.a to require that the turbine-driven

and motor-driven auxiliary feedwater pumps be tested at least quarterly

on a staggered test basis instead of the currently required testing

once per 31 days on a staggered test basis. In addition, the proposed

changes would revise Technical Specification Bases 3/4.7.7, Emergency

Exhaust System--Auxiliary Building, and 3/4.9.13, Emergency Exhaust

System--Fuel Building, to eliminate the reference to the use of

automatic control for the emergency exhaust system heaters.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The staff has reviewed the licensee's analysis against

the standards of 10 CFR 50.92(c). The NRC staff's review is presented

below.

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

the probability or consequences of an accident previously evaluated.

This change only revises the surveillance requirement for the

auxiliary feedwater pumps. The purpose of this surveillance requirement

is to prove that the pumps are operable. The longer test interval

should result in greater availability by reducing the rate of test

induced failures which should offset any loss in reliability. The

revised surveillance requirement does not affect the probability of

accident initiation. The operability of the pumps is maintained and

therefore the consequences of evaluated accidents is unaffected by the

proposed change. The revised surveillance frequency is consistent with

the guidance issued in Generic Letter 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements for

Testing During Power Operation.''

Changes to the emergency exhaust system technical specifications

eliminate the words ``using automatic control'' associated with the

humidity controlling heaters. These changes reflect the current method

in which the fuel building emergency exhaust system heaters are

controlled in that humidity sensors are bypassed to allow continuous

operation of the heaters whenever the emergency exhaust system fans are

operating. The proposed changes do not affect the probability of

initiating an accident previously evaluated. The ability of the

emergency exhaust systems to mitigate the consequences of an accident

are likewise unaffected since the heaters remain available and

operating to control humidity.

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

different kind of accident from any previously evaluated.

Verification of pump operability is still maintained with the

change to the frequency of the surveillance requirement. No system

configuration changes are being implemented in order to perform the

surveillance testing and any potential accidents that may be associated

with the surveillance testing were previously considered. The changing

of the surveillance frequency does not introduce any additional failure

modes for the auxiliary feedwater system.

The proposed revisions to the emergency exhaust system technical

specifications are limited to the automatic control of the humidity

controlling heaters. The proposed change does not introduce any new

potential challenges to fission product barriers or any new failure

modes which might prevent the emergency exhaust systems from fulfilling

their accident mitigating functions.

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

the margin of safety.

The inservice testing program will continue to ensure that

auxiliary feedwater pumps operational readiness criteria are consistent

with the requirements of ASME Section XI. System performance

surveillances will continue to be conducted in accordance with the

technical specifications. The proposed change does not significantly

reduce the margin of safety because the availability and reliability of

the auxiliary feedwater system are not significantly decreased and the

heat removal requirements for the system are unchanged.

The proposed changes to the emergency exhaust system technical

specifications affect the requirements to maintain heater control in

the automatic mode. The proposed changes do not decrease the exhaust

systems' actual ability to control humidity or alter the other

functional requirements for the system. Therefore, the proposed change

does not significantly reduce any margin of safety related to the

performance of the emergency exhaust systems.

Based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and

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

NRC Project Director: Suzanne C. Black

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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: March 14, 1994

Description of amendment request: The proposed change to the

Millstone Unit 2

Technical Specifications (TS) would provide a one-time extension of the

surveillance frequency from the required 18-month to the next refueling

outage but no later than September 30, 1994, of the power operated

valves in the service water system (TS 4.7.4.1.b) and in the boron

injection flow path (TS 4.1.2.2.c). This would extend the surveillance

for these valves approximately 5 months. Date of publication of

individual notice in Federal Register: March 23, 1994 (59 FR 13751).

Expiration date of individual notice: April 22, 1994

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, Connecticut 06360.

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances prov

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