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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission

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

notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of

1954, as amended (the Act), to require the Commission to publish notice

of any amendments issued, or proposed to be issued, under a new

provision of section 189 of the Act. This provision grants the

Commission the authority to issue and make immediately effective any

amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from June 22, 1998, through July 2, 1998. The

last biweekly notice was published on July 1, 1998 (63 FR 35986).

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 Chief, Rules and

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and should cite the publication date and page number of

this Federal Register notice. Written comments may also be delivered to

Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,

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

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By August 14, 1998, 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 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.

[[Page 38199]]

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 20555-0001, Attention: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

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

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1

and 2, Will County, Illinois

Date of amendment request: May 29, 1998.

Description of amendment request: The proposed amendment would

revise the technical specifications to credit the automatic function of

the pressurizer power operated relief valves (PORVs) to provide

mitigation for inadvertent safety injection at power accident. The

limiting condition for operation and surveillance requirements for the

PORVs would also be revised.

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 change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The changes to the Technical Specification (TS) Limiting Condition

for Operation (LCO), Surveillance Requirements, and Bases do not

involve an increase in the probability or consequences of the

Inadvertent Operation of Emergency Core Cooling System (Spurious SI) at

Power transient. Crediting the PORVs in the maximum pressurizer

overfill case for this transient does not increase the probability of

the occurrence of the transient since the automatic function of the

PORVs for Reactor Coolant System (RCS) pressure control is not an

initiator for the Spurious SI at Power transient. This change allows

for the NRC Standard Review Plan (NUREG-0800) acceptance criteria to be

met for the Spurious SI at Power transient, ensuring that the

consequences of this transient remain within acceptable levels.

As documented in various Safety Evaluation Reports (SERs) from the

NRC, the overpressure protection function of the PORVs was not

originally considered to be a safety related function. In response to

Generic Issue 70, the NRC performed a regulatory analysis related to

PORV and block valve reliability in Pressurized Water Reactor (PWR)

plants. This regulatory analysis is documented in NUREG-1316,

``Technical Findings and Regulatory Analysis Related to Generic Issue

70, Evaluation of Power-Operated Relief Valve and Block Valve

Reliability in PWR Nuclear Power Plants,'' where the NRC staff

concluded that it was not cost effective to backfit non-safety related

PORVs to upgrade them to safety related status to perform safety

related functions. The safety related functions were those detailed in

Section 2.1 of NUREG-1316 and any other safety related function

identified in the future. As an example, the PORVs are credited for the

cold overpressure protection function of the reactor pressure vessel

during low temperature operations. The analysis documented in this

License Amendment request demonstrates that the PORVs provide an

acceptable level of quality and performance to allow them to be

credited to mitigate the consequences of the Spurious SI at Power

transient documented in Byron and Braidwood Updated Final Safety

Analysis Report (UFSAR) Section 15.5.1. The PORVs are equipped with

safety related actuators and safety related accumulator tanks which

maintain valve function during a loss of instrument air. The position

indication and control switches in the Main Control Room (MCR) are

safety related. All pressurizer PORV open/close functions and circuitry

are supplied with uninterruptible Class 1E power supplies. The

automatic portion of the PORV circuitry which processes the high

pressurizer and high RCS pressure at low temperature is designated non-

safety related and is isolated from the safety related portions of the

circuitry by safety related interposing relays which actuate on a

faulted condition. However, both Byron and Braidwood Stations have

implemented modifications for both Units 1 and 2, which ensure that

automatic control of both PORVs is available during loss of offsite

power conditions. In addition, the PORV function is monitored within

the scope of the Maintenance Rule Program and the postulated failure of

the PORV automatic function does not result in unacceptable risk.

The probability of a Spurious SI at Power transient is not affected

by this proposed change and the above analysis demonstrates that the

PORVs will adequately function in automatic mode to mitigate the

consequences of the transient. As such, there are no changes in the

type or amount of any effluent released offsite as a result of this

change. Therefore, based on this evaluation, this proposed amendment

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated.

[[Page 38200]]

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

different accident from any accident previously evaluated. This change

would specifically allow for the PORV automatic function to be credited

in Modes 1, 2, and 3 for the Spurious SI at Power transient only. This

change allows for added assurance that the acceptance criteria as

documented in the NRC Standard Review Plan (NUREG-0800) for ANS

Condition II transients will be met. The acceptance criteria of concern

is that a Condition II transient must not lead to an event (Condition

III or IV) of more significant consequences without additional failures

occurring. The PORV automatic function is to be credited with

mitigating the maximum pressurizer overfill case for the Spurious SI at

Power transient. This case has the acceptance criteria that the

pressurizer must not go water solid prior to RCS pressure reaching the

setpoint of the pressurizer safety relief valves (PSRVs). This

conservative acceptance criteria is based on the fact that the PSRVs

are not qualified to pass subcooled water and reseat, thereby creating

a concern for an uncontrolled release path from the RCS. This proposed

change helps ensure that the acceptance criteria for this accident are

met. There is a small probability that the PORV function, either

automatic or manual, would not successfully mitigate this transient due

to the failure of one or both PORVs. However, the low likelihood of a

total failure of the PORV function during the Spurious SI at Power

transient does not create a new accident because a similar scenario is

already addressed by UFSAR Section 15.6.1, ``Inadvertent Opening of a

Pressurizer Safety or Relief Valve.'' The UFSAR analysis for the

Section 15.6.1 ANS Condition II transient indicates that the

radiological consequences of this transient are significantly less than

that of a LOCA and are therefore, acceptable. The same arguments for

radiological consequences apply to the Spurious SI at Power transient

in the event the PORV automatic function fails and water relief occurs

through the PSRVs.

The proposed change to the LCO requirements in TS Section 3/4.4.4

would allow for the PORV block valve to be closed but remain energized

in the event a PORV was considered inoperable due to the automatic

actuation circuitry. Currently, the PORV block valve is closed but

remains energized only if a PORV is considered inoperable due to

excessive seat leakage. The proposed change would extend the allowance

to include the circumstance where the PORV was inoperable due to the

automatic actuation circuitry. This allows a PORV to remain functional

in the manual mode for other safety related functions consistent with

the discussion contained in NRC NUREG-1316. However, this revised LCO

requirement would not represent a new failure mode or accident over

what has been previously evaluated.

In summary, the proposed changes documented in this TS amendment to

credit the automatic PORV function and to revise the TS LCO

requirements for PORV inoperability do not create the potential for any

new or different accidents from what was previously evaluated.

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

of safety.

The current TS bases do not credit the function of the pressurizer

PORVs for any Mode 1, 2, or 3 transients. This change would allow for

the PORV automatic function to be credited for the Spurious SI at Power

transient only. This does not represent a significant reduction in the

margin of safety. This change would allow for the conservative

acceptance criteria for the current UFSAR design analysis to be met.

The PORVs are reliable and are maintained in a manner consistent with

their proposed safety related function to mitigate the Spurious SI at

Power transient. This proposed change would not result in a significant

increase in risk or consequences, and therefore, 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Stuart A. Richards.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: May 28,1998.

Description of amendment request: The proposed amendment proposes

changes to the Final Safety Analysis Report (FSAR) to include a

description of the use of Generic Letter (GL) 87-11, ``Relaxation in

Arbitrary Intermediate Pipe Rupture Requirements,'' and NUREG/CR-2913,

``Two-Phase Jet Loads,'' as a part of the approved licensing basis and

design basis for Crystal River Unit 3. GL 87-11 will be used to

determine where high energy line breaks (HELB) are postulated to occur

for high energy lines located inside the Reactor Building (RB) and

analyzed in accordance with the guidelines described in USAS B31.1.0-

1967, ``USA Standard Code for Pressure Piping, Power Piping.'' NUREG/

CR-2913 will be used to determine the effects of the resultant jet

impingement from postulated Reactor Coolant System (RCS) piping

ruptures on safety-related systems, structures, and components (SSCs).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

The use of new design methodologies for determining postulated

break locations of RCS piping and other high energy lines located

inside containment, and the dynamic effects of postulated ruptures of

RCS piping on SSCs required for safe shutdown or accident mitigation,

does not impact the design of these high energy lines such that

previously analyzed ruptures would now be more likely to occur. The

approval of the license amendment will not result in an actual

modification to RCS piping or other high energy lines which would

reduce their design capabilities to maintain pressure boundary

integrity during normal operating and accident conditions. By using

these new design methodologies, protection of SSCs required for

accident mitigation is assured. Protection of SSCs required for

accident mitigation will continue to be assured by use of these well-

defined design methodologies if modifications to those SSCs are

implemented in the future. Therefore, there will be no reduction in the

capability of those SSCs in limiting the consequences of previously

evaluated accidents, and the proposed amendment does not significantly

increase the probability or consequence of an accident previously

evaluated.

[[Page 38201]]

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

from previously evaluated accidents?

The use of new design methodologies for determining postulated

break locations of RCS piping and other high energy lines located

inside containment, and the dynamic effects of postulated ruptures of

RCS piping on SSCs required for safe shutdown or accident mitigation,

does not impact the design of these high energy lines such that

previously unanalyzed ruptures would now occur. The approval of the

license amendment will not result in an actual modification to RCS

piping or other high energy lines which would reduce their design

capabilities to maintain pressure boundary integrity during normal

operating and accident conditions. By using these new design

methodologies, the current design of RCS piping and other high energy

lines located inside containment can be shown to include sufficient

design margin to prevent unanalyzed ruptures from occurring. Therefore,

use of these design methodologies instead of the previous licensing

basis requirements cannot create the possibility of a new or different

kind of accident.

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

The use of new design methodologies for determining postulated

break locations of RCS piping and other high energy lines located

inside containment, and the dynamic effects of postulated ruptures of

RCS piping on SSCs required for safe shutdown or accident mitigation,

does not impact the design of these high energy lines such that

unanalyzed ruptures would now occur, and cannot create a reduction in

the margin of safety for those ruptures of high energy lines previously

analyzed. The approval of the license amendment will not result in an

actual modification to RCS piping or other high energy lines which

would reduce their design capabilities to maintain pressure boundary

integrity during normal operating and accident conditions. By using

these new design methodologies, protection of SSCs required for

accident mitigation is assured. Protection of SSCs required for

accident mitigation will continue to be assured by use of these well-

defined design methodologies if modifications to those SSCs are

implemented in the future. Therefore, the capability of those SSCs to

limit the consequences of previously evaluated accidents at levels

below the approved acceptance limits will continue to be assured. As a

result, use of these design methodologies instead of the previous

licensing basis and design basis requirements cannot significantly

reduce the existing margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida

33733-4042.

NRC Project Director: Frederick J. Hebdon.

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

Generating Station, Ocean County, New Jersey

Date of amendment request: May 28, 1998.

Description of amendment request: Revision of Technical

Specification (TS) 4.5.A.1 such that the first Type A test required by

the primary containment leakage rate testing program be performed

during refueling outage 18 rather than refueling outage 17.

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 TS change does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The proposed changes does not alter the design, function or manner

of operation of any structures, systems or components. As a result, the

proposed change does not affect any of the parameters or conditions

that could contribute to initiation of any accidents.

NUREG-1493 found that the effect of containment leakage on overall

accident risk is small since risk is dominated by accident sequences

that result in failure or bypass of the containment. The major

contributor to the total identified leakage from Primary Containment

comes from Type B and C tested components. Only a small portion of the

total leakage is detectable soley through Type A testing. The leaks

that have been found by Type A tests have been only marginally above

existing requirements. In addition, Oyster Creek has two means

(monitoring nitrogen use and performing torus to drywell vacuum breaker

leak tests) of detecting gross containment leakage. The proposed change

does not alter the requirements to perform Type B and C testing in

accordance with the Primary Containment Leakage Rate Testing Program

and does not affect the ability of the facility to mitigate the

consequences of an accident.

Therefore, the proposed TS change does not involve an significant

increase in the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any previously evaluated.

Deferring the Type A test for an operating cycle does not alter the

design, function or manner of operation of any structures, systems or

components. The proposed change does not affect any of the parameters

or conditions that could contribute to initiation of any accidents nor

does it introduce any new mechanisms which could contribute to the

creation of a new or different kind of accident than previously

evaluated.

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

in a margin of safety.

The proposed change does not alter the design, function or manner

of operation of any structures, systems or components. The proposed

change does not impact the primary containment system's ability to

provide a barrier against the uncontrolled release of fission products

in the event of a break in the reactor coolant system nor does the

proposed change impact the primary containment accident leak rate. In

addition, NUREG-1493's Summary of Technical Findings states ``Reducing

the frequency of Type A tests (ILRTs) from the current three per 10

years to one per 20 years was found to lead to an imperceptible

increase in risk. The estimated increase in risk is very small because

ILRTs identify only a few potential containment leakage paths that

cannot be identified by Type B and C testing, and the leaks that have

been found by Type A tests have been only marginally above existing

requirements.'' Therefore, the proposed TS change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

[[Page 38202]]

Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

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

NRC Project Director: Cecil O. Thomas.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of amendment request: June 10, 1998.

Description of amendment request: The proposed revision to the

Millstone Unit 3 licensing basis would address post-accident mitigation

activities, vital area access travel routes, and time. NNECO determined

that the Final Safety Analysis Report (FSAR) description of post-

accident vital area routing was out of date because the radiological

control area boundary fence created an access problem on the designated

routes to the hydrogen recombiner and fuel building. The FSAR change

would revise the routes to accommodate the fence location and allow for

the time to unlock gates.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with 10 CFR

50.92 and has concluded that the revision does not involve a

significant hazards consideration (SHC). The basis for this conclusion

is that the three criteria of 10 CFR 50.92(c) are not satisfied. The

proposed revision does not involve an SHC because the revision would

not:

1. Involve a significant increase in the probability or consequence

of an accident previously evaluated.

Final Safety Analysis Report (FSAR) Section 12.3.1.3.2, Post-

accident access to vital areas, and its associated Figures and Tables

are being updated. The current FSAR descriptions are out of date and as

such do not include all required post-accident actions. Therefore, this

FSAR change adds actions to those listed in the FSAR as well as

incorporating the recalculation of the doses associated with the

required post-accident actions. The dose calculations utilize the

appropriate post-accident source terms, area access requirements and

stay times, including the appropriate routes to the areas. The

calculations show that for all design basis required post-accident

actions the calculated dose to the Operators/Emergency workers

performing those actions remains below the 5 rem criterion of General

Design Criteria (GDC) 19. The revision to the FSAR provides the

required post-accident required operator actions. Changing the FSAR to

include the current post-accident vital access requirements and

associated information for the supporting dose calculations [cannot]

cause an accident. In addition, the calculated dose to the Operators/

Emergency workers for all design basis required actions is below the

GDC 19 limit of 5 rem.

Therefore, the proposed revision does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

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

from any accident previously evaluated.

The change is to the calculated post-accident vital access dose

analyses and the FSAR description of that analyses. No new procedural

Operator/Emergency worker actions are associated with the change.

However, since the information in the FSAR was outdated, there are

Operator/Emergency actions being added to the FSAR. Dose calculations

associated with those actions have been performed utilizing the

appropriate assumptions with respect to source terms, vital area access

travel routes and stay times, and times when the post-accident actions

would be performed. The analyses confirmed that the calculated doses

associated with all required post-accident actions are less than the 5

rem limit of GDC 19. There are no changes to the Emergency Operating

Procedures associated with this change.

Therefore, the proposed revision does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

The dose calculations confirm that the calculated dose associated

with all design basis post-accident Operator/Emergency worker actions

is below the limit of 5 rem of GDC 19. There is one action, initiation

of hydrogen purge, for which the calculated dose to the Operator/

Emergency worker exceeds 5 rem. This action is a backup means of

limiting the hydrogen concentration inside containment post-accident.

This action would only be performed for multiple failures which would

disable both trains of the safety-grade hydrogen recombiner system. As

such this action is not a required design basis action and does not

need to meet the 5 rem limit. The calculated dose for this action is

below the 25 rem limit that is specified in the Station Emergency Plan

for severe accident mitigation actions.

Therefore, the proposed revision does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is determined

that the proposed revision does not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Deputy Director: Phillip F. McKee.

PECO Energy Company, Public Service Electric and Gas Company, Demarva

Power and Light Company, and Atlantic City Electric Company, Dockets

Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2

and 3, York County, Pennsylvania

Date of application for amendments: February 4, 1998.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) Surveillance Requirement (SR)

concerning Secondary Containment doors at Peach Bottom Atomic Power

Station, Units 2 and 3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

TS SR 3.6.4.1.2 will be revised to require either all inner or

outer secondary containment access doors to be closed in each air lock.

This revision will not adversely affect the ability of the Secondary

Containment to mitigate the radiological consequences of a Loss-of-

Coolant Accident or fuel handling accident, and does not involve a

[[Page 38203]]

significant increase in the probability or consequences of an accident

previously evaluated. During those times that one or more inner (or

outer) doors are open, the closed outer (or inner) doors will serve as

the Secondary Containment boundary.

Allowing certain inner or outer Secondary Containment access doors

in an air lock to be open does not compromise the design of the

Secondary Containment. No commitment is made in the UFSAR to consider

the single failure of passive structural components such as Secondary

Containment doors. As discussed in Section 1.5 of the UFSAR, ``* * *

Essential safety actions shall be carried out by equipment of

sufficient redundance and independence that no single failure of active

components can prevent the required actions''. The same UFSAR section

goes on to state that, ``For systems or components to which IEEE-279

(1968) is applicable, single failures of passive electrical components

are considered, as well as single failures of active components, in

recognition of the higher anticipated failure rates of passive

electrical components relative to passive mechanical components.''

Therefore, based on this UFSAR discussion, it is concluded that failure

of outer (inner) secondary containment doors need not be postulated

with the inner (outer) door being open.

The performance of the Secondary Containment and the Standby Gas

Treatment System is unaffected by this activity. Surveillance testing

will prove the capability to maintain Secondary Containment with only

inner or only outer doors closed. This change will not result in

greater or more frequent loading of Secondary Containment doors, and

does not result in changes that impact the reliability of the Secondary

Containment and the Standby Gas Treatment System.

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

or different kind of accident from any accident previously evaluated.

The Secondary Containment, in conjunction with the Standby Gas

Treatment System, provides the means for mitigating the radiological

consequences of an accident. The configuration of the Secondary

Containment has no effect on accident initiators which lead to a new or

different kind of accident. This change will not involve any changes to

plant systems, structures, or components which could act as new

accident initiators. The design, function, and reliability of Secondary

Containment and the Standby Gas Treatment System are also not impacted

by this change. Therefore, this change will not create the possibility

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

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

in a margin of safety.

No margins of safety are reduced as a result of this change to the

TS. No safety limits will be changed as a result of this TS change. The

Secondary Containment will continue to perform its intended safety

function of limiting the ground level release of airborne radioactive

materials and to provide a means for controlled elevated release of the

building atmosphere so that off-site doses from the postulated design

basis accidents are below the limits of 10 CFR 100. The design and

reliability of the Secondary Containment are also not impacted as a

result of this change.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105.

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

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101.

NRC Project Director: Robert A. Capra.

Power Authority of the State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: November 13, 1997.

Description of amendment request: The proposed amendment will

reduce the maximum test interval from 1 year to 6 months for the test

frequency of the main turbine stop and control valves (TS & CVs) in

Table 4.1-3 and add a footnote.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident previously

evaluated?

Response

The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident previously

evaluated. The proposed change increases the frequency of testing of

the TS & CVs by reducing the maximum allowable test interval. The

maximum test interval is reduced from one year to six months. Thus, the

proposed change will make the maximum test interval more conservative.

Therefore, the proposed change does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

(2) Does the proposed license amendment create the possibility of a

new or different kind of accident from any accident previously

evaluated?

Response

The proposed license amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated. The proposed change does not involve the addition of any new

or different type of equipment, nor does it involve the operation of

equipment required for safe operation of the facility in a manner

different from those addressed in the Final Safety Analysis Report.

(3) Does the proposed license amendment involve a significant

reduction in a margin of safety?

Response

The proposed license amendment does not involve a significant

reduction in a margin of safety. The proposed change does not adversely

affect performance of any safety related system or component,

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

increased severity of any accidents considered in the safety analysis.

The proposed change does not reduce the frequency of testing of these

valves but updates the methodology for determination of the test

frequency and reduces the maximum test interval from one year to six

months. It establishes a more conservative acceptance criteria of 5.0

x 10-6 per year than the NRC acceptance criteria of 1.0 x

10-5 for a turbine missile event. 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

[[Page 38204]]

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa, Director.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: June 16, 1998.

Description of amendment request: The proposed amendment would

relocate the Safety Review Committee review, audit and related record

keeping requirements from the Technical Specifications (TSs) to Chapter

17 of the Final Safety Analysis Report (FSAR) (i.e., Quality Assurance

Program).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident previously

analyzed?

Response

This amendment application does not involve a significant increase

in the probability or consequences of an accident previously analyzed.

The relocation of the SRC [Safety Review Committee] review, audit, and

related record keeping requirements from the TS to the FSAR does not

alter the performance or frequency of these activities. Future changes

to the QA [Qualify Assurance] program, located in Chapter 17 of the

FSAR, which constitute a reduction in commitments, are governed by 10

CFR 50.54(a). Therefore, sufficient controls for these requirements

exist and these changes do not involve a significant increase in the

probability or consequences of an accident previously analyzed.

(2) Does the proposed license amendment create the possibility of a

new or different kind of accident from any accident previously

evaluated?

Response

This amendment application does not create the possibility of a new

or different kind of accident from any accident previously evaluated.

The proposed changes involve the relocation of SRC requirements from

the TS to the FSAR. Relocation of these requirements does not affect

plant equipment or the way the plant operates. The reviews, audits, and

record keeping will continue to be performed in the identical manner as

they are currently being performed. Therefore, the proposed revisions

cannot create a new or different kind of accident.

(3) Does the proposed amendment involve a significant reduction in

a margin of safety?

Response

This amendment application does not involve a significant reduction

in a margin of safety. The requested Technical Specification revisions

relocate SRC review, audit and related record keeping requirements from

the TS to the FSAR. These requirements are not being altered by this

relocation. The reviews, audits, and record keeping will continue to be

performed in the identical manner as they are currently being

performed. Any changes to these requirements which constitute a

reduction in commitments will be processed in accordance with 10 CFR

50.54(a). Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa.

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

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

Date of application for amendments: April 30, 1998 (TS 98-01).

Brief description of amendments: The amendments would change the

Sequoyah (SQN) Technical Specifications (TSs) to allow surveillance

testing of the reactor coolant system (RCS) pressurizer power-operated

relief valves (PORVs) in Modes 3, 4, and 5.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), Tennessee Valley

Authority (TVA), the licensee, has provided its analysis of the issue

of no significant hazards consideration, which is presented below:

TVA has concluded that operation of SQN Units 1 and 2, in

accordance with the proposed change to the TSs, does not involve a

significant hazards consideration. TVA's conclusion is based on its

evaluation, in accordance with 10 CFR 50.91(a)(1), of the three

standards set forth in 10 CFR 50.92(c).

A. The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The possibility of occurrence or the consequences for an accident

or malfunction of equipment is not increased as the test conditions for

the PORVs in Mode 5 are representative conditions based on a steam

bubble being present, and testing in this mode is more conservative, if

RCS pressure is less, since there is less fluid force to aid the

solenoid force in opening the valve. Testing in Modes 3 and 4 was the

initial request of GL [Generic Letter] 90-06. No changes are proposed

to operation of the PORV block valves. Offsite dose consequences are

unchanged by this request.

B. The proposed amendment does not create the possibility of a new

or different kind of accident from any accident previously evaluated.

A possibility for an accident or malfunction of a different type

than any evaluated previously in SQN's Final Safety Analysis Report is

not created; nor is the possibility for an accident or malfunction of a

different type. A new test method is not required. No new failure modes

are introduced.

C. The proposed amendment does not involve a significant reduction

in a margin of safety.

The margin of safety has not been reduced for testing in Mode 5

since the proposed test conditions are equal to or more conservative,

if RCS pressure is less, than those currently in use with existing SRs

[surveillance requirements]. Testing in Modes 3 and 4 was the initial

request of GL 90-06. The results of the accident analysis remain

unchanged by this request.

The NRC 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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

[[Page 38205]]

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

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

Date of application for amendments: June 26, 1998 (TS 98-02).

Brief description of amendments: The amendments would change the

Sequoyah Nuclear Plant (SQN) Technical Specifications (TS) and their

Bases to lower the specific activity of the primary coolant from 1.0

microcurie/gram dose equivalent iodine-131 to 0.35 microcurie/gram, as

provided for in NRC Generic Letter 95-05, ``Voltage-Based Repair

Criteria for Westinghouse Steam Generator Tubes Affected by Outside

Diameter Stress Corrosion Cracking.'' This change allows a proportional

increase in main steam line break induced primary-to-secondary leakage

when implementing the alternate steam generator tube repair criteria,

which the NRC has already approved for Units 1 and 2.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the Tennessee Valley

Authority (TVA), the licensee, has provided its analysis of the issue

of no significant hazards consideration, which is presented below:

TVA has concluded that operation of SQN Units 1 and 2, in

accordance with the proposed change to the TS [or operating

license(s)], does not involve a significant hazards consideration.

TVA's conclusion is based on its evaluation, in accordance with 10 CFR

50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).

A. The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The proposed TS change lowers the [maximum allowable] reactor

coolant specific activity, which allows an increase in the leakage

quantity that would be postulated to occur during a MSLB accident. This

in turn allows a larger quantity of tubes with axial ODSCC to remain in

service. The methodology for identifying and defining the ODSCC and for

developing the leakage quantity remains unchanged. Therefore, the

proposed change does not result in a significant increase in the

probability of an accident.

An increase in the consequences of an accident would not occur

because the proportional decrease in reactor coolant specific activity,

while proportionally increasing the primary-to-secondary leakage during

a postulated MSLB accident, has been evaluated to confirm the amount of

activity released to the environment remains unchanged. The evaluation

uses the same methodology used to establish the original primary-to-

secondary leak limits in [Westinghouse Topical Report] WCAP-13990.

The control room dose, the low population zone dose, and the dose

at the exclusion area boundary remains bounded by the acceptance

criteria of NUREG-0800 and continue to satisfy an appropriate fraction

of the 10 CFR 100 dose limits and GDC [General Design Criterion] 19.

Therefore, the proposed TS change does not result in a significant

increase in the consequences of an accident previously analyzed.

B. The proposed amendment does not create the possibility of a new

or different kind of accident from any accident previously evaluated.

The proposed TS change does not alter the configuration of the

plant. The changes do not directly affect plant operation. The change

will not result in the installation of any new equipment or systems or

the modification of any existing equipment or systems. No new operating

procedures, conditions or modes will be created by this proposed

change. SG [steam generator] tube structural integrity, as defined in

draft Regulatory Guide 1.121, remains unchanged.

Therefore the possibility of a new or different kind of accident

from any accident previously evaluated is not created.

C. The proposed amendment does not involve a significant reduction

in a margin of safety.

Lowering the reactor coolant specific activity, while allowing the

proportional increase in the primary-to-secondary leakage during a

postulated MSLB accident, keeps the amount--of activity released to the

environment unchanged. Design basis and offsite dose calculation

assumptions remain satisfied. Therefore, the proposed change does not

result in a significant reduction in the margin of safety.

The NRC 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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

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: February 27, 1998 (TXX-98033), June 10,

1998 (TXX 98145).

Brief description of amendments: The proposed amendment would

increase the RWST Low-Low level setpoint from ``greater than or equal

to 40%'' to ``greater than or equal to 45%'' of span for CPSES, Units 1

and 2. The change raises the RWST Low-Low level setpoint in order to

increase the volume available to complete containment spray switchover

without turning off the containment spray pumps.

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. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The changes in the License Amendment Request proposes more

restrictive setpoint Allowable Values for the RWST Low-Low setpoint.

This more restrictive value assures that all applicable safety analysis

limits are being met. Changing an RWST Low-Low setpoint from greater

than or equal to 40% to greater than or equal to 45% in the Technical

Specifications has no impact on the probability of occurrence of any

accident previously evaluated. None of the accident analyses were

affected, therefore, the consequences of all previously evaluated

accidents remain unchanged.

2. Do the proposed changes create the possibility of a new or

different kind of accident from any accident previously evaluated?

The proposed changes involve the use of a more conservative value

for the RWST Low-Low setpoint. As such, none of the changes effect

plant hardware or the operation of plant systems in a way that could

initiate an accident. Therefore, the proposed changes do not create the

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

previously evaluated.

3. Do the proposed changes involve a significant reduction in a

margin of safety?

There were no changes made to any of the accident analyses or

safety analysis limits as a result of this proposed change. Further,

the proposed

[[Page 38206]]

change does not affect the acceptance criteria for any analyzed event.

ECCS, Containment spray, and the RWST will remain capable of performing

their safety function, and the new requirement will continue to provide

adequate assurance of that capability. Raising the RWST Low-Low

setpoint from 40% to 45% has no impact on the assumptions used in the

safety analysis as discussed in Chapter 15 of the FSAR. The margin of

safety established by the Limiting Conditions for Operation also

remains unchanged. Thus there is no effect on the margin of safety.

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

this review, it appears that the three standards of 10 CFR 50.92 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, 702 College, P.O. Box

19497, Arlington, TX 76019.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

Bockius, 1800 M Street, N.W., Washington, DC 20036.

NRC Project Director: John N. Hannon.

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: June 19, 1998. This amendment request

supersedes the November 5, 1997, submittal in its entirety (63 FR

19981).

Description of amendment request: The proposed Operating License

change and changes to the technical specifications (TS) would permit

the use of a temporary alternate supply line (jumper) to provide

service water (SW) to the component cooling heat exchangers. The

temporary jumper will permit maintenance to be performed on the

existing supply line.

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:

Virginia Electric and Power Company has reviewed the proposed

changes against the criteria of 10 CFR 50.92 and has concluded that the

changes do not pose a significant safety hazards consideration as

defined therein. The proposed Operating License and Technical

Specifications and Bases changes are necessary to allow the use of a

temporary, seismic, non-missile protected jumper to provide service

water (SW) to the Component Cooling Heat Exchangers (CCHXs) while

maintenance work is performed on the existing SW supply line to the

CCHXs. Since there is only one SW supply line to the CCHXs, an

alternate SW supply must be provided whenever the line is removed from

service. The temporary jumper provides this function. The jumper will

only be used for a 35-day period during each of two Unit 1 refueling

outages.

The use of the temporary jumper has been thoroughly evaluated, and

appropriate constraints and compensatory measures (including a

Contingency Action Plan) have been developed to ensure that the

temporary jumper is reliable, safe, and suitable for its intended

purpose. A complete and immediate loss of SW supply to the operating

CCHXs is not considered credible, given the project constraints and the

unlikely probability of a generated missile or heavy load drop.

Existing station abnormal procedures already address a loss of

component cooling, and the use of alternate cooling for a loss of decay

heat removal, in the unlikely event that they are required.

Furthermore, appropriate mitigative measures have been identified to

address potential flooding concerns. The minor administrative changes

merely correct a table format inconsistency and update Basis section

references.

Consequently, the operation of Surry Power Station with the

proposed amendment and license condition will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The SW and CC Systems will function as designed under the Unit

operating constraints specified by this project (i.e., Unit 2 in

operation and Unit 1 in a refueling outage), and the potential for a

loss of component cooling is already addressed by Station Abnormal

Procedures. Therefore, there is no increase in the probability of an

accident previously evaluated. The possibility of flooding due to

failure of the temporary SW supply jumper in the Turbine Building

basement has been evaluated and dispositioned by the implementation of

appropriate precautions and compensatory measures to preclude damage to

the temporary jumper and to respond to a postulated flooding event. A

flood watch will be present around-the-clock with authority and

procedural guidance to isolate the jumper, if required. Furthermore,

the CCHXs serve no design basis accident mitigating function.

Therefore, the consequences of an accident previously evaluated are not

increased.

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

from any accident previously evaluated.

The SW and CC Systems' design functions and basic configurations

are not being altered as a result of using a temporary SW supply

jumper. The temporary jumper is designed to be safety-related and

seismic with all of the design attributes of the normal SW supply line,

except for the automatic isolation function and complete missile and

heavy load drop protection. The design functions of the SW and CC

systems are unchanged as a result of the proposed changes due to (1)

required plant conditions, (2) compensatory measures, (3) a Contingency

Action Plan for restoration of the normal SW supply if required, and

(4) strict administrative control of the temporary SW isolation valve

to preclude flooding or to isolate non-essential SW within the design

basis assumed time limits. Unit 1 will be in a plant condition which

will provide adequate time to restore the normal SW supply, if

required. Therefore, since the SW and CC systems will basically

function as designed and will be operated in their basic configuration,

the possibility of a new or different type of accident than previously

evaluated in the UFSAR [Updated Final Safety Analysis Report] is not

created.

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

The margin of safety as defined in the Technical Specifications is

not reduced since an operable SW flowpath to the required number of

CCHXs is provided, and Unit operating constraints, compensatory

measures and contingencies will be implemented as required to ensure

the integrity and the capability of the SW flowpath. The use of the

temporary jumper will be limited to the time period when missile

producing weather is not expected, and Unit 1 meets specified unit

conditions. Therefore, the temporary SW jumper, under the imposed

project constraints and compensatory measures, provides the same

reliability as the normal SW supply line. Furthermore, the

Probabilistic Safety Assessment for Surry Power Station has been

reviewed relative to the use of the temporary SW jumper. It has been

determined that due to the SW restoration project's compensatory and

contingency measures, as well as the configuration restrictions that

will be imposed by the Maintenance Rule online risk matrix, the impact

on core damage frequency is negligible.

[[Page 38207]]

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: P. T. Kuo, Acting.

Westinghouse Electric Corporation (Licensee), Westinghouse Test

Reactor, Waltz Mill Site, Westmoreland, Pennsylvania, Docket No. 50-22,

License No. TR-2

Date of amendment request: December 22, 1997, supplemented on June

15, 1998.

Description of amendment request: In 1959, the Westinghouse

Electric Corporation was granted a license for the Westinghouse Test

Reactor (WTR) at Waltz Mill. On December 22, 1997, the licensee

informed the Nuclear Regulatory Commission it had changed its name to

CBS Corporation, and requested the license to be amended to reflect the

name change.

On June 15, 1998, the CBS Corporation agreed that the name of the

WTR licensee, as reflected on the license, can be revised to ``CBS

Corporation acting through its Westinghouse Electric Company

Division.'' Therefore, the purpose of this amendment is to change the

name of the licensee as indicated on the WTR license from Westinghouse

Electric Corporation to CBS Corporation acting through its Westinghouse

Electric Company Division.

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). A proposed amendment to an

operating license for a facility 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.

The staff agrees with the licensee's no significant hazards

consideration determination submitted on June 15, 1998 for the

following reason.

This corporate name change does not involve any change in the

management, organization, location, facilities equipment, or procedures

related to or personnel responsible for the licensed activities of the

WTR license. All existing commitments, obligations and representations

remain in effect.

Based on a review of the licensee's analysis, and on the staff's

analysis detailed above, 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.

Attorney for license: Lisa A. Campagna, Assistant General Counsel,

Law Department, CBS Corporation, P.O. Box 355, Pittsburgh, Pennsylvania

15230.

NRC Project Director: Seymour H. Weiss.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: February 26, 1998 (TSCR 204).

Description of amendment request: The proposed changes would modify

Technical Specifications (TS) and bases to reflect a lower containment

leakage limit, a revised program for control of primary coolant sources

outside containment, a revised control room emergency filtration

design, and the addition of the primary auxiliary building exhaust

filtration system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of this facility under the proposed Technical

Specifications will not create a significant increase in the

probability or consequences of an accident previously evaluated.

The probabilities of accidents previously evaluated are based on

the probability of initiating events for these accidents. Initiating

events for accidents previously evaluated for Point Beach include:

Control rod withdrawal and drop, CVCS [chemical volume control system]

malfunction (Boron Dilution), startup of an inactive reactor coolant

loop, reduction in feedwater enthalpy, excessive load increase, losses

of reactor coolant flow, loss of external electrical load, loss of

normal feedwater, loss of all AC [alternating current ] power to the

auxiliaries, turbine overspeed, fuel handling accidents, accidental

releases of waste liquid or gas, steam generator tube rupture, steam

pipe rupture, control rod ejection, and primary coolant system

ruptures.

This license amendment request proposes to change the limiting

conditions for operation, action statements, allowable outage times,

and surveillance requirements for the Point Beach Nuclear Plant [PBNP]

Technical Specifications associated with the maximum permissible

containment leak rate, control room emergency filtration, primary

auxiliary building exhaust filtration, and primary coolant sources

outside containment. These proposed changes do not cause an increase in

the probabilities of any accidents previously evaluated because these

changes will not cause an increase in the probability of any initiating

events for accidents previously evaluated. In particular, these changes

affect accident mitigation systems and equipment which do not cause

accidents.

The consequences of the accidents previously evaluated in the PBNP

FSAR [Final Safety Analysis Report] are determined by the results of

analyses that are based on initial conditions of the plant, the type of

accident, transient response of the plant, and the operation and

failure of equipment and systems. The changes proposed in this license

amendment request provide appropriate limiting conditions for

operation, action statements, allowable outage times, and surveillance

requirements for maximum permissible containment leak rate, control

room emergency filtration, primary auxiliary building exhaust

filtration, and primary coolant sources outside containment.

The proposed changes affect components that are required to ensure

the proper operation of accident mitigation systems and equipment. The

proposed changes do not increase the probability of failure of this

equipment or its ability to operate as required for the accidents

previously evaluated in the PBNP FSAR.

Therefore, this proposed license amendment does not affect the

consequences of any accident previously evaluated in the Point Beach

Nuclear Plant FSAR, because the factors that are used to determine the

consequences of accidents are not being changed.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of

[[Page 38208]]

a new or different kind of accident from any accident previously

evaluated.

New or different kinds of accidents can only be created by new or

different accident initiators or sequences. New and different types of

accidents (different from those that were originally analyzed for Point

Beach) have been evaluated and incorporated into the licensing basis

for Point Beach Nuclear Plant. Examples of different accidents that

have been incorporated into the Point Beach Licensing basis include

anticipated transients without scram and station blackout. The changes

proposed by this license amendment request do not create any new or

different accident initiators or sequences because these changes to

limiting conditions for operation, action statements, allowable outage

times, and surveillance requirements for maximum permissible

containment leak rate, control room emergency filtration, primary

auxiliary building exhaust filtration, and primary coolant sources

outside containment will not cause failures of equipment or accident

sequences different than the accidents previously evaluated. Therefore,

these proposed Technical Specifications changes do not create the

possibility of an accident of a different type than any previously

evaluated in the Point Beach FSAR.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are based on the design and

operation of the reactor and containment and the safety systems that

provide their protection.

The changes proposed by this license amendment request provide the

appropriate limiting conditions for operation, action statements,

allowable outage times, and surveillance requirements for maximum

permissible containment leak rate, control room emergency filtration,

primary auxiliary building exhaust filtration, and primary coolant

sources outside containment. This ensures that the safety systems that

protect the reactor and containment will operate as required. The

design and operation of the reactor and containment are not affected by

these proposed changes. Therefore, the margins of safety for Point

Beach are not being reduced because the design and operation of the

reactor and containment are not being changed and the safety systems

and limiting conditions of operation for these safety systems that

provide their protection that are being changed will continue to meet

the requirements for accident mitigation for Point Beach Nuclear Plant.

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: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

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

NRC Project Director: Cynthia A. Carpenter.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: May 28, 1998 (TSCR 203).

Description of amendment request: The proposed amendments would

revise Technical Specifications (TS) to provide a specific numerical

setting for reactor trip, reactor coolant pump trip, and auxiliary

feedwater initiation on a loss of power to the 4 kilovolt (kV) buses.

Changes to the bases for the affected TS sections are also being made.

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 Point Beach Nuclear Plant [PBNP] in accordance

with the proposed amendments will not create a significant increase in

the probability or consequences of an accident previously evaluated.

The probabilities of accidents previously evaluated are based on

the probability of initiating events for these accidents. Initiating

events for accidents potentially affected by the proposed amendments

previously evaluated for Point Beach include losses of reactor coolant

flow, loss of external electrical load, loss of normal feedwater, and

loss of all AC [alternating current] power to the auxiliaries.

This license amendment request proposes to clarify the setting

limit for the undervoltage reactor trip, auxiliary feedwater initiation

and reactor coolant pump trip by providing an actual numerical value in

place of the word ``Normal'' thereby eliminating any confusion as to

the actual value used in the setting limit for this protection

function.

This proposed change does not cause an increase in the

probabilities of any accidents previously evaluated because the change

will not cause an increase in the probability of any initiating events

for accidents previously evaluated. In particular, the proposed change

more clearly defines the actual setting limit for the 4 KV undervoltage

protection function taking into account the effects of voltage decay

and response times. This is a protection function for mitigation of

these events. Appropriate delay times are implemented in this function

to ensure momentary voltage transients do not initiate these events

while ensuring appropriate protection for these loss of power events.

Therefore, there is no significant increase in the probability or

consequences of any event previously analyzed.

The consequences of the accidents previously evaluated in the PBNP

FSAR [Final Safety Analysis Report] are determined by the results of

analyses that are based on initial conditions of the plant, the type of

accident, transient response of the plant, and the operation and

failure of equipment and systems. The changes proposed in this license

amendment request provide appropriate limiting conditions for the

setting limits for the Point Beach Nuclear Plant Technical

Specifications for the 4 KV undervoltage protection function. Thus the

analyses of the events remain valid and demonstrate that there are no

radiological consequences from these events.

Therefore, this proposed license amendment does not affect the

consequences of any accident previously evaluated in the Point Beach

Nuclear Plant FSAR, because the factors that are used to determine the

consequences of accidents are not being changed.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

New or different kinds of accidents can only be created by new or

different accident initiators or sequences. New and different types of

accidents (different from those that were originally analyzed for Point

Beach) have been evaluated and incorporated into the licensing basis

for Point Beach Nuclear Plant. Examples of different accidents that

have been incorporated into the Point Beach Licensing basis include

anticipated transients without scram and station blackout.

[[Page 38209]]

The change proposed by the amendments to provide specific

undervoltage setting limits does not create any new or different

accident initiators or sequences because the change to the 4 KV

undervoltage protection function will not cause failures of equipment

or accident sequences different than the accidents previously

evaluated. Therefore, the proposed Technical Specification change does

not create the possibility of an accident of a different type than any

previously evaluated in the Point Beach FSAR.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments [will] not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are based on the design and

operation of the reactor and containment and the safety systems that

provide their protection.

The change proposed by this license amendment request provides the

appropriate setting limit for the 4 KV undervoltage protection

function. This ensures that the safety systems that protect the reactor

and containment will operate as required. The design and operation of

the reactor and containment are not affected by these proposed changes.

Therefore, the margins of safety for Point Beach are not being reduced

because the design and operation of the reactor and containment are not

being changed.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

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

NRC Project Director: Cynthia A. Carpenter.

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.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: June 17, 1998, as supplemented June 23,

1998.

Brief description of amendment request: The proposed amendment

would revise Section 3.1.1c of the Technical Specifications (TS),

Appendix A of the Operating License for the Palisades Nuclear Plant, to

change the minimum required primary coolant system flow. The currently

specified value is 140.7 x 10 \6\ lb/hr [pounds per hour] or greater,

when corrected to 532 deg.F. The licensee proposed to revise the TS to

specify a value of greater than or equal to 352,000 gpm [gallons per

minute], which is equivalent to approximately 135 x 10 \6\ lb/hr, when

corrected to 532 deg.F.

Date of publication of individual notice in Federal Register: July

2, 1998 (63 FR 36271)

Expiration date of individual notice: August 3, 1998.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423-3698.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,

Michigan

Date of amendment request: June 26, 1998 (NRC-98-0040).

Brief description of amendment request: The proposed amendment

would provide a one-time extension of the interval for a number of

technical specification (TS) surveillance requirements that will be

performed in the sixth refueling outage. TS 4.0.2 and Index page xxii

would be revised and TS tables 4.0.2-1 and 4.0.2-2 would be replaced to

reflect the extensions.

Date of publication of individual notice in Federal Register: July

2, 1998 (63 FR 36273).

Expiration date of individual notice: August 3, 1998.

Local Public Document Room location: Monroe County Library System,

Ellis Reference and Information Center, 3700 South Custer Road, Monroe,

Michigan 48161.

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

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

Date of amendment request: June 19, 1998.

Brief description of amendment request: This amendment revises the

Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and BVPS-2),

Technical Specifications (TS) definition of a channel calibration to

add two sentences stating that (1) the calibration of instrument

channels with resistance temperature detector or thermocouple sensors

may consist of an inplace qualitative assessment of sensor behavior and

normal calibration of the remaining adjustable devices in the channel

and (2) whenever a sensing element is replaced, the next required

channel calibration shall include an inplace cross calibration that

compares the other sensing elements with the recently installed sensing

element. This proposed change would make the BVPS-1 and BVPS-2 TS

definition of channel calibration consistent with the definition of a

channel calibration contained in the NRC's improved Standard Technical

Specifications for Westinghouse Plants (NUREG-1431, Revision 1).

Date of publication of individual notice in Federal Register: June

26, 1998.

Expiration date of individual notice: July 27, 1998 (63 FR 34939).

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

663 Franklin Avenue, Aliquippa, PA 15001.

Pennsylvania Power and Light Company, Docket No. 50-388, Susquehanna

Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

Date of amendment request: June 17, 1998.

Brief description of amendment request: This amendment revises the

applicability requirement in TS Sections 3.4.2, ``Safety/Relief

Valves'' (Action c); 4.4.2; 3.3.7.5, ``Accident Monitoring

Instrumentation'' (TS Table 3.3.7.5-1, Action 80 and 4.3.7.5,

``Surveillance Requirements,'' Table 4.3.7.5-1 ``Accident Monitoring

Instrumentation Surveillance Requirements''). The change to the

referenced TSs adds the following applicability footnote:

Compliance with these requirements for the ``J'' SRV acoustic

monitor is not required for the period beginning June 15, 1998,

until the next unit shutdown of sufficient duration to allow for

containment entry, not to exceed the 9th refueling and inspection

outage.

[[Page 38210]]

Date of publication of individual notice in Federal Register: June

23, 1998 (63 FR 34200).

Expiration date of individual notice: July 23, 1998.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

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

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

the particular facilities involved.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: May 27, 1997, as supplemented

on August 1, 1997, and March 24, 1998.

Brief description of amendments: The amendments revise Technical

Specification Section 6, ``Administrative Controls,'' to incorporate

revised organizational titles and delete Unit 1 Facility Operating

License Condition 2.C.(30)(a). In addition, the amendments change the

submittal frequency of the Radiological Effluent Release Report from

semiannually to annually and make several administrative and editorial

changes.

Date of issuance: June 26, 1998.

Effective date: Immediately, to be implemented within 90 days.

Amendment Nos.: 128 and 113.

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the Unit 1 Facility Operating License and the Technical

Specifications.

Date of initial notice in Federal Register: July 30 ,1997. The

August 1, 1997, submittal provided clarifying information that did not

change the initial proposed no significant hazards consideration

determination. The March 24, 1998, submittal changed the scope of the

initial Federal Register notice. The proposed amendments were renoticed

on May 20, 1998 (63 FR 27759).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 26, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station,

Unit 1, LaSalle County, Illinois

Date of application for amendment: November 24, 1997, as

supplemented April 16, 1998.

Brief description of amendment: The amendment revises Technical

Specification 3/4 3.2, ``Isolation Actuation Instrumentation'' to add/

revise various isolation setpoints for leak detection instrumentation.

These changes are necessary due to modifications to the Reactor Water

Cleanup (RWCU) system to restore ``hot'' suction to the RWCU pumps and

due to a re-evaluation of the high energy line break analysis. In

addition, the amendment eliminates isolation actuation trip functions

for the Residual Heat Removal system steam condensing mode and shutdown

cooling mode.

Date of issuance: July 6, 1998.

Effective date: Immediately, to be implemented prior to restart

from L1F35

Amendment No.: 129.

Facility Operating License No. NPF-11: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 14, 1998 (63 FR

2278). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 6, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348

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

Plant, Middlesex County, Connecticut

Date of application of amendments: May 30, 1997, as supplemented

May 7, 1998 and June 18, 1998.

Brief description of amendments: The amendments revise the Facility

Operating License and Technical Specifications to reflect the

permanently shut down and defueled status of the reactor.

Date of issuance: June 30, 1998.

Effective date: As of the date of issuance (June 30, 1998) and

shall be implemented within 90 days.

Amendment No.: 193.

Facility Operating License No. DPR-61: The amendments revised the

Operating License and the Technical Specifications.

Date of initial notice in Federal Register: July 16, 1997 (62 FR

38132 and 62 FR 38133). The May 7, 1998, supplement relocated the

provisions of Technical Specification 3/4.9.15. The June 18, 1998,

supplement consisted of supporting technical information. The

supplements did not change the staff's initial proposed no significant

hazards consideration determination or expand the scope of the original

notice. The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 30, 1998.

No significant hazards consideration received: No.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457.

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

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: January 27, 1998 (Reference NRC-98-

0023).

Brief description of amendment: This amendment revises the Fermi 1

License to allow Detroit Edison to receive, acquire, possess, use, and

transfer byproduct material without restriction to chemical form for

sample analysis, instrument calibration, or associated

[[Page 38211]]

with radioactive apparatus, hardware, tools, and equipment, provided

the cumulative radioactive material quantity of the byproduct material

does not exceed the criteria contained in Section 30.72, Schedule C,

Quantities of Radioactive Materials Requiring Consideration of the Need

for an Emergency Plan for Responding to a Release.

Date of issuance: June 22, 1998.

Effective date: Within 60 calendar days from the date of issuance

of this amendment.

Amendment No.: 12.

Facility Operating License No. DPR-9: Amendment revised License by

adding a subpart 3 to Part 2.B.

Date of initial notice in Federal Register: April 8, 1998 (63 FR

17223).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 22, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Duke Energy Corporation, Docket Nos. 50-269 and 50-287, Oconee Nuclear

Station, Units 1 and 3, Seneca, South Carolina

Date of application of amendments: June 4, 1998.

Brief description of amendments: The amendments revise Technical

Specification 4.17.2 to allow continued operation with certain steam

generator tubes that exceed their repair limit as a result of tube end

anomalies. This action temporarily exempts these tubes from the

requirement for sleeving, rerolling, or removal from service until they

are repaired during or before the next scheduled refueling outages for

the respective unit. This action supersedes the Notice of Enforcement

Discretion that was issued by the staff on June 4, 1998.

Date of Issuance: July 1, 1998.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment Nos.: Unit 1--230; Unit 2--227.

Facility Operating License Nos. DPR-38 and DPR-55: The amendments

revised the Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: Yes. (63 FR 33097 dated June 17, 1998). The notice

provided an opportunity to submit comments on the Commission's proposed

no significant hazards consideration determination. No comments have

been received. The notice also provided for an opportunity to request a

hearing by July 16, 1998, but indicated that if the Commission makes a

final no significant hazards consideration determination, any such

hearing would take place after issuance of the amendments.

The Commission's related evaluation of the amendments, finding of

exigent circumstances, and a final no significant hazards consideration

determination are contained in a Safety Evaluation dated July 1, 1998.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 3, 1997, as supplemented by letter

dated May 1, 1998.

Brief description of amendment: The amendment changed the Appendix

A Technical Specifications (TSs) by changing the action requirements

for TS 3/4.3.2 for the Safety Injection System Sump Recirculation

Actuation Signal (RAS). It revised the allowed outage time for a

channel of RAS to be in the tripped condition from ``prior to entry

into the applicable MODE(S) following the next COLD SHUTDOWN'' to the

more restrictive time limit of 48 hours, and added a shutdown

requirement. Additionally, the TS 3.0.4 exemption was removed from the

action for the tripped condition. A change to TS Bases Section 3/4.3.2

was also included.

Date of issuance: July 2, 1998.

Effective date: July 2, 1998, to be implemented within 60 days.

Amendment No.: 143.

Facility Operating License No. NPF-38: Amendment revised the

Technical Specifications. Date of initial notice in Federal Register:

June 18, 1997 (62 FR 33124).

The additional information contained in the supplemental letter

dated May 1, 1998, was clarifying in nature and thus, it was within the

scope of the initial notice and did not affect the staff's proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 2, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

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

Date of amendment request: March 2, 1998, as supplemented by letter

dated April 21, 1998.

Description of amendment request: The amendment revised Technical

Specification 4.5.2.b.1 for the emergency core cooling system

subsystems to delete the requirement to vent the operating chemical

volume and control system centrifugal pump casing.

Date of issuance: June 24, 1998.

Effective date: As of its date of issuance, to be implemented

within 60 days.

Amendment No.: 58.

Facility Operating License No. NPF-86. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 8, 1998 (63 FR

17225).

The supplemental letter provided clarifying information that did

not change the staff's proposed no significant hazards determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 24, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: April 13, 1998.

Brief description of amendment: The amendment changes the Technical

Specifications (TSs) by adding a new TS 3.5.5, ``Emergency Core Cooling

Systems--Trisodium Phosphate (TSP).'' The TSP surveillance requirements

in TSs 4.5.2.c.3 and 4.5.2.c.4 are relocated to new TS 3.5.5 as TS

4.5.5.1 and TS 4.5.5.2, respectively. Also, the amount of TSP is

increased, the surveillance requirements are modified, a new limiting

condition of operation is included, and the applicable TS Index pages

and Bases sections are updated to reflect the changes.

Date of issuance: June 22, 1998.

Effective date: As of the date of issuance to be implemented within

60 days.

Amendment No.: 217.

Facility Operating License No. DPR-65: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR

25114).

[[Page 38212]]

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 22, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: November 13, 1997, as

supplemented on December 29, 1997, and April 8, 1998.

Brief description of amendment: The amendment changes the Technical

Specifications (TSs) by modifying TS 3.1.2.1, ``Flow Paths--Shutdown;''

TS 3.1.2.2, ``Flow Paths--Operating;'' TS 3.1.2.3, ``Charging Pump--

Shutdown;'' TS 3.1.2.4, ``Charging Pumps--Operating;'' TS 3.1.2.5,

``Boric Acid Pumps--Shutdown;'' TS 3.1.2.6, ``Boric Acid Pumps--

Operating;'' TS 3.1.2.8, ``Borated Water Sources--Operating;'' TS

3.4.1.3, ``Coolant Loops and Coolant Circulation--Shutdown;'' TS 3.4.3,

``Relief Valves;'' TS 3.4.9.1, ``Reactor Coolant System;'' TS 3.4.9.2,

``Pressurizer;'' TS 3.4.9.3, ``Overpressure Protection Systems;'' TS

3.5.3, ``ECCS Subsystems--Tavg < 300 deg.F;'' and TS

3.10.3, ``Pressure/Temperature Limitation--Reactor Criticality,'' and

their associated Bases in the areas that are affected by the modified

Low Temperature Overpressure Protection system, the updated reactor

coolant system pressure and temperature curves and heatup and cooldown

limits. Additionally, minor changes are made to correct various items,

such as, updating of redundant or outdated TSs.

Date of issuance: July 1, 1998.

Effective date: As of the date of issuance to be implemented within

60 days.

Amendment No.: 218.

Facility Operating License No. DPR-65: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4315).

The December 29, 1997, and April 8, 1998, letters provided

clarifying information that did not change the scope of the November

13, 1997, application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 1, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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 application for amendments: February 27, 1997, as

supplemented by letter dated December 4, 1997.

Brief description of amendments: The amendments revised the

combined Technical Specifications (TS) for the Diablo Canyon Power

Plant, (DCPP) Unit Nos. 1 and 2 to change Technical Specification (TS)

3/4.8.1.1, ``A.C. Sources--Operating,'' to clarify that emergency

diesel generator (EDG) testing is initiated from standby conditions

rather than ``ambient'' conditions. The associated TS Bases were

revised to discuss the temperature range that satisfies EDG standby

conditions. TS 3/4.3.2, ``Instrumentation--Engineering Safety Features

Actuation System Instrumentation'' was also changed. This revision

clarified that when one or both of the first level load shed relays, or

one or both of the second level undervoltage relays are inoperable, the

associated EDG for that bus shall be declared inoperable.

Date of issuance: June 5, 1998.

Effective date: June 5, 1998, to be implemented within 90 days from

the date of issuance.

Amendment Nos.: Unit 1--127; Unit 2--125

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 9, 1997 (62 FR

17240).

The December 4, 1997, supplemental letter provided additional

clarifying information and did not change the staff's initial no

significant hazards consideration determination. The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated June 5, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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 application for amendments: April 10, 1998, as supplemented

by letter dated May 1, 1998.

Brief description of amendments: The amendments revised the

combined Technical Specifications (TS) for the Diablo Canyon Power

Plant (DCPP) Units 1 and 2 to revise TS 6.2.2.g and TS 6.3 to change

the name of the Operations Manager to Operations Director, to add the

position of Operations Middle Manager, and to change the requirement

for the Operations Director to hold a senior reactor operator (SRO)

license.

Date of issuance: June 11, 1998.

Effective date: June 11, 1998.

Amendment Nos.: Unit 1-128; Unit 2-126.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR

25116).

The May 1, 1998, supplemental letter provided additional clarifying

information and did not change the staff's initial no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

June 11, 1998.

No significant hazards consideration comments received: Yes.

The Commission received one letter with comments which did not

change its finding and conclusion as discussed in the safety

evaluation.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments: May 12, 1998

Brief description of amendments: These amendments relocate certain

requirements related to fire protection from the TSs to the Updated

Final Safety Analysis Report. The TS sections to be relocated are: 3/

4.3.7.9, Fire Detection Instrumentation; 3/4.7.6, Fire Suppression

Systems; 3/4.7.7, Fire

[[Page 38213]]

Rated Assemblies; and 6.2.2e, Fire Brigade Staffing. The amendments

also replace License Condition 2.C.(6) for Unit 1 and License Condition

2.C.(3) for Unit 2. These amendments are consistent with the guidance

of NRC Generic Letter (GL) 86-10, ``Implementation of Fire Protection

Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements

from Technical Specifications.''

Date of issuance: June 24, 1998.

Effective date: Both units, as of date of issuance, to be

implemented within 30 days.

Amendment Nos.: 177 and 150.

Facility Operating License Nos. NPF-14 and NPF-22: The amendments

revised the Technical Specifications and Licenses.

Date of initial notice in Federal Register: May 21, 1998 (63 FR

28010).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 24, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: November 26, 1997, as

supplemented April 17, 1998.

Brief description of amendment: The amendment relocates snubber

operability, surveillance, and records requirements from the Technical

Specifications to plant controlled documents. A condition is added to

the license to require that the relocated requirements be described in

the Final Safety Analysis Report such that 10 CFR 50.59 will apply to

future changes to those requirements.

Date of issuance: June 30, 1998.

Effective date: As of the date of issuance to be implemented within

90 days.

Amendment No.: 243.

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4352).

The April 17, 1998, supplemental letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 30, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San

Diego County, California

Date of application for amendments: November 6, 1995, as

supplemented by letters dated January 9, 1998, and February 3, 1998,

for the safety injection tanks (SITs), and November 8, 1995, as

supplemented by letters dated January 9, 1998, and February 3, 1998,

for the low pressure safety injection (LPSI).

Brief description of amendments: The amendments modify the

technical specifications (TSs) to extend the allowed outage times

(AOTs) for a single inoperable SIT from one hour to 24 hours, and for a

single inoperable SIT specifically due to malfunctioning SIT water

level or nitrogen cover pressure instrumentation inoperability from one

hour to 72 hours. In addition, the amendments extend the AOT for a

single inoperable LPSI train from 72 hours to 7 days. The amendments

also add a Configuration Risk Management Program to the TSs that puts a

proceduralized probabilistic risk assessment-informed process in place

that ensures the licensee assesses the overall impact of plant

maintenance on plant risk.

Date of issuance: June 19, 1998.

Effective date: June 19, 1998, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 2--139; Unit 3--131.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 10, 1996 (61 FR

15995) and February 11, 1998 (63 FR 6991).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 19, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

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

Southern Nuclear Power Company, Inc., et al., Docket Nos. 50-424 and

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

Georgia

Date of application for amendments: September 4, 1997, as

supplemented by letters dated November 20, 1997, May 19 and June 12,

1998.

Brief description of amendments: The changes to the common

Technical Specifications allow an increase in the Unit 1 spent fuel

storage capacity from 288 to 1476 fuel assemblies.

Date of issuance: June 29, 1998.

Effective date: As of the date of issuance to be implemented on a

schedule consistent with the receipt and storage of new fuel in the

fall of 1998 for the spring 1999 refueling outage of Unit 1.

Amendment Nos.: Unit 1--102; Unit 2--80.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments

revised the Technical Specifications, Operating Licenses, and Appendix

D.

Date of initial notice in Federal Register: December 31, 1997 (62

FR 68317); and renoticed on May 11, 1998 (63 FR 25883).

The supplements dated May 19 and June 12, 1998, provided clarifying

information that did not change the scope of the September 4, 1997,

application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia.

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

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

Date of application for amendments: June 7, 1996, as supplemented

on September 26, 1997, January 21, 1998, May 28, 1998, and June 29,

1998 (TS 95-19).

Brief description of amendments: The amendments change the

Technical Specifications (TS) by relocating portions of Section 6,

``Administrative Controls,'' to the Sequoyah Nuclear Quality Assurance

Plan. This Change is consistent with NUREG-1431, ``Standard Technical

Specifications--Westinghouse Plants.''

Date of issuance: July 1, 1998.

Effective date: July 1, 1998.

Amendment Nos.: 233 and 223.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the TS.

Date of initial notice in Federal Register: July 17, 1996 (61 FR

37302).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 1, 1998.

No significant hazards consideration comments received: None.

[[Page 38214]]

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear

Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: December 23, 1997, as

supplemented by letter dated June 11, 1998.

Brief description of amendment: This amendment revises Technical

Specification (TS) Section 1.0, ``Definitions,'' to clarify the meaning

of core alteration; relocates TS Section 3/4.9.5, ``Refueling

Operations--Communications,'' and the associated bases to the Technical

Requirements Manual; and adds TS Section 3.0.6 and associated bases to

address the return to service of inoperable equipment.

Date of issuance: June 30, 1998.

Effective date: June 30, 1998.

Amendment No.: 224.

Facility Operating License No. NPF-3: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4327).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 30, 1998. .

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606.

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

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

Virginia

Date of application for amendments: March 25, 1998.

Brief description of amendments: The amendments revise the

Technical Specifications (TS) Sections 6.1.1; 6.2.1.b; 6.5.1.1;

6.5.1.6.a,d,h, and m; 6.5.1.7.c; 6.5.1.8; 6.14.1.2; 6.15.b; 6.2.3.5;

6.5.1.2; 6.5.1.7.a for Unit 1 and 6.1.1; 6.2.1.b; 6.5.1.1;

6.5.1.6.a,d,h, and m; 6.5.1.7.c; 6.5.1.8; 6.13.b; 6.14.b; 6.2.3.5;

6.5.1.2; and 6.5.1.7.a for Unit 2, changing the title of Station

Manager to Site Vice President, and the titles of the Assistant Station

Managers to Manager-Station Operation and Maintenance and Manager-

Station Safety and Licensing.

Date of issuance: June 23, 1998.

Effective date: June 23, 1998.

Amendment Nos.: 212 and 193.

Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19980).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Dated at Rockville, Maryland, this 8th day of July 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects--III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 98-18684 Filed 7-14-98; 8:45 am]

BILLING CODE 7590-01-P