Biweekly Notice

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

publish notice of any

[[Page 38131]]

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 23, 1997, through July 3, 1997. The

last biweekly notice was published on July 2, 1997 (62 FR 35846).

Notice Of Consideration of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Harards Consideration

Determination, And Opportunith 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

Review and Directives Branch, Division of Freedom of Information and

Publications 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 15, 1997, 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.

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

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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: Docketing and

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

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

by the above date. A copy of 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.

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

Neck Plant, Middlesex County, Connecticut

Date of amendment request: May 30, 1997, identified as CY-97-006

Description of amendment request: Changes to the Operating License,

DPR-61, and facility Technical Specifications (TS) that reflect the

permanently shut down and defueled status of the plant.

CY-97-006 contains the proposed changes to the license conditions

in DPR-61 on Fire Protection, Power Level and Fuel Movement; and

submittal of a new set of TS referred to by the licensee as the

Defueled TS (DTS). The DTS contain a revised Definitions section,

removal of the sections on Safety Limits and Limiting Safety System

Settings, Limiting Conditions for Operation and Surveillance

Requirements were modified extensively, the Design Features section was

revised, and the Administrative Controls section was modified to

reflect all the preceding changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Connecticut Yankee Atomic Power Company (CYAPCO) has reviewed

the proposed changes to the Operating License and the Technical

Specifications in accordance with 10 CFR 50.92 and concluded that

the changes do not involve a significant hazards consideration

(SHC). The basis for this conclusion is that the three criteria of

10 CFR 50.92(c) are not compromised. The proposed changes do not

involve an SHC because the changes would not:1. Involve a

significant increase in the probability of consequences of an

accident previously evaluated.

Because of the present plant configuration, many of the

postulated accidents previously evaluated (i.e., loss or coolant

accident, main steam line break, etc.) are no longer possible. The

accidents previously evaluated that are still applicable to the

plant are fuel handling accidents and gaseous and liquid radioactive

releases.

There is no significant increase in the probability of a fuel

handling accident since refueling operations have ceased. In fact,

there is more likely a decrease in probability of a fuel handling

accident since the need to move/rearrange fuel assemblies is minimal

until they are removed from the spent fuel pool (i.e., for dry cask

storage or for transferring to U.S. Department of Energy

possession).

The radiological consequences of a gaseous or liquid radioactive

release are bounded by the fuel handling accident. With the plant

defueled and permanently shutdown, the demands on the radwaste

systems is lessened since no new radioisotopes are being generated

by irradiation or fission. Therefore, there is no increase in the

probability or consequences of a gaseous or liquid radioactive

release.

The changes to the Operating License reflect the permanently

defueled condition for power level and fuel movement restrictions

and the fire protection regulation which is applicable for a

permanently defueled plant.

With respect to the Service Water System (Specification 3/

4.7.3), Electrical Power Systems (Specification 3/4.8) and spent

fuel pool makeup, the basis for placing appropriate requirements in

the Technical Requirements Manual is due to the reduced heat load in

the spent fuel pool.

The plant was shutdown on July 22, 1996 and more than 280 days

have passed since the shutdown, thus the heat load on the spent fuel

pool cooling system is greatly reduced. Present cooling performance

data as well as calculations demonstrate that either the plate or

the shell and tube heat exchanger has more than adequate heat

removal capacity. In the event of a loss of forced cooling,

calculations indicate that the spent fuel pool time to boil is

greater than 40 hours based on an initial pool temperature of

150 deg.F. The initial pool temperature of 150 deg.F is based on

Technical Specification 3/4.9.15 which has a pool temperature limit

of 150 deg.F. Even during boiling, the fuel is adequately cooled.

Once boiling commences, the operators have in excess of 18 days to

provide forced cooling and/or makeup before there is inadequate

shielding provided by the water in the pool. This allows sufficient

time to provide for alternate forced cooling or makeup to the spent

fuel pool in the event of a service water system failure. Therefore,

operability of spent fuel pool cooling does not require service

water, electrical power, or makeup water to be immediately

available.

Should failure to restore operation of the spent fuel pool

cooling system occur before boiling takes place, cooling of the

spent fuel can be accomplished by allowing the spent fuel pool to

boil and adding makeup water at a rate equal to or greater than the

boil-off rate.

CYAPCO has in place procedures to establish onsite power in the

event of a Loss of Normal Power (LNP) and in the event of a loss of

cooling to the Spent Fuel Pool. For a LNP, power can be made

available within approximately one hour. If onsite power cannot be

reestablished, due to equipment failure, at approximately 2 hours

into the LNP, limited makeup water could be provided by gravity feed

from a tank (available in approximately 30 minutes) or an unlimited

supply of water could be provided via the diesel fire pump from the

Connecticut River (available in approximately 30 minutes).

Therefore, within approximately 2 1/2 hours of the event start,

cooling and/or makeup would be reestablished to the spent fuel pool.

Historically, the longest LNP the HNP has experienced has been less

than 30 minutes.

The changes to Technical Specification 3.3.3.8, ``Radioactive

Gaseous Effluent Monitoring Instrumentation'' and Table 3.3.-10

delete the trip function from the main stack noble gas activity

monitor. The changes to Technical Specifications 3.11.2.1, Dose

Rate, and 3.11.2.3, Dose, delete the requirement to include the

radioiodine isotopes in the dose calculations. These changes are

based on the following:

There is no significant increase in the consequences of a fuel

handling accident since the accident scenarios assume an assembly

with significant amounts of radioactive iodine or noble gas. The

plant was shutdown on July 22, 1996. Except for I-125 (half-life

=59.5 days), I-129 (half-life = 1.6E7 years), and Kr-85 (half-life

=10.8 years), the spent fuel inventory of the dose contributing

radioactive iodine and noble gas isotopes has decayed more than 20

half-lives since shutdown (i.e., less than 0.0001% of the original

amount remains). In addition, the definition for ``Dose Equivalent

I-131'' (Standard Technical Specifications, Westinghouse

Plants,'' NUREG-1431) does not include I-125 and I-129 in the dose

assessment due to their negligible inventory in the spent fuel.

Except for Kr-85, the other noble gas nuclides that contribute to a

whole

[[Page 38133]]

body dose have also decayed to a negligible amount. CYAPCO has

performed fuel handling and cask drop accident dose calculations

which conclude that doses (i.e., whole body and thyroid) at the

Exclusion Area Boundary are a small fraction of the 1O CFR 100 dose

limits and the EPA PAGS. In fact, due to this decreased radioactive

inventory, there is a significant decrease in the consequences of a

fuel handling accident.

Based on the above, the proposed changes to the Operating

License and the Technical Specifications do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

There is no change in how spent fuel is stored or moved in the

spent fuel pool. Therefore, the postulated fuel handling accidents

are still bounding and are still considered as credible postulated

accidents. The bases provided in the CYAPCO analysis of previously

evaluated accidents in Section 1, above, also applies to the

possibility of new or different accidents herein.

Based on the analysis in Section 1, above, the changes to

Technical Specification related to radioactive iodine and noble gas

isotopes do not create the possibility of a new or different kind of

accident from any accident previously evaluated.

Based on these considerations, the proposed changes to the

Operating License and the Technical Specifications do not create the

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

previously evaluated.

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

With respect to the Service Water System (Specification 3/

4.7.3), Electrical Power Systems (Specification 3/4.8) and spent

fuel pool makeup, the basis for placing appropriate requirements in

the Technical Requirements Manual is due to the reduced heat load in

the spent fuel pool.

The Technical Specification basis states that the time to spent

fuel pool boiling after a loss of forced cooling following a full

core offload is 7 hours.

In accordance with the analysis set forth above under No. 1,

there is no change in how spent fuel is stored or moved in the spent

fuel pool.

Based on the above, the proposed changes to the Operating

License and the Technical Specifications do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270

NRC Project Director: Marvin M. Mendonca, Acting Director

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

Neck Plant, Middlesex County, Connecticut

Date of amendment request: May 30, 1997, identified as CY-97-024

Description of amendment request: CY-97-024 provided the proposed

technical specifications (TS) needed to implement the Certified Fuel

Handler (CFH) program at the plant. This new position will replace the

former licensed operator positions. A copy of the CFH Training Program,

``Nuclear Training Manual NTM-7.083'' was enclosed with the license

amendment request for NRC review and approval. However, this manual

will be reviewed separately from the proposed TS changes and when the

NRC review of the manual is completed a letter of approval will be sent

to the licensee.

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:

Connecticut Yankee Atomic Power Company (CYAPCO) has reviewed

the proposed changes to the Technical Specifications in accordance

with 10 CFR 50.92 and concluded that the changes do not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10 CFR 50.92(c) are not

compromised. The proposed changes do not involve an SHC because the

changes would not.

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed qualification, staffing and training requirements

are appropriate for the present plant conditions.

The plant has permanently ceased operations, the reactor has

been permanently defueled, and the spent fuel stored in the spent

fuel pool.

Because the present plant conditions, many of the postulated

accidents previously evaluated (i.e., loss-of-coolant accident, main

steam line break, etc.) are no longer possible. The accidents

previously evaluated that are still applicable to the plant are fuel

handling accidents and gaseous and liquid radioactive releases.

There is no significant increase in the probability of a fuel

handling accident since refueling operations have ceased. In fact,

there is more likely a decrease in probability of a fuel handling

accident since the need to move/rearrange fuel assemblies is minimal

until they are removed from the spent fuel pool (i.e., for dry cask

storage or for transferring to U.S. Department of Energy

possession).

There is no significant increase in the consequences of a fuel

handling accident since the accident scenarios assume an assembly

with significant amounts of radioactive iodine or noble gas. The

plant was shutdown on July 22, 1996. Except for I-125 (half-

life=59.5 days), I-129 (half-life=1.6E7 years), and Kr-85 (half-

life-10.8 years), the spent fuel inventory of the dose-contributing

radioactive iodine and noble gas isotopes has decayed more than 20

half-lives since shutdown (i.e., less than 0.0001% of the original

amount remains). In addition, the definition for ``Dose Equivalent

I-131'' (Standard Technical Specifications, Westinghouse

Plants,'' NUREG-1431) does not include I-125 and I-129 in the dose

assessment due to their negligible spent fuel inventory. Except for

Kr-85, the other noble gas nuclides that contribute to a whole body

dose have also decayed to a negligible amount. CYAPCO has performed

fuel handling and cask drop accident dose calculations which

conclude that doses (i.e., whole body and thyroid) at the Exclusion

Area Boundary and the Low Population Zone are a small fraction of

the 10 CFR 100 dose limits. In fact, due to this decreased

radioactive inventory, there is a significant decrease in the

consequences of a fuel handling accident.

The radiological consequences of a gaseous or liquid radioactive

release are bounded by the fuel handling accident. With the plant

defueled and permanently shutdown, the demands on the radwaste

systems are lessened since no new radioisotopes are being generated

by irradiation. Therefore, there is no increase in the consequences

of a gaseous or liquid radioactive release.

Based on the above, the proposed changes to the Technical

Specifications do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

from any accident previously evaluated.

There is no change in how spent fuel is stored or moved in the

spent fuel pool. Therefore, the postulated fuel handling accidents

are still bounding and are still considered as credible postulated

accidents.

Based on the above, the proposed changes to the Technical

Specifications do not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

There is no change in how spent fuel is stored or moved in the

spent fuel pool.

The plant was shutdown on July 22, 1996. Except for I-125 (half-

life=59.5 days), I-129 (half-life=1.6E7 years), and Kr-85 (Half-

life=10.8 years), the spent fuel inventory of the dose-contributing

radioactive iodine and noble gas isotopes has decayed more than 20

half-lives since shutdown (i.e., less than 0.0001% of the original

amount remains). Except for Kr-85, the other noble gas nuclides that

contribute to a whole body dose have also decayed to a negligible

amount. CYAPCO has performed fuel handling and cask drop accident

dose calculations which conclude that doses (i.e, whole body and

[[Page 38134]]

thyroid) at the Exclusion Area Boundary and the Low Population Zone

are a small fraction of the 10 CFR 100 dose limits.

Therefore, there is no significant reduction the margin of

safety. In fact, due to this decreased radioactive iodine inventory,

there is more likely an increase in the margin of safety.

Based on the above, the proposed changes to the Technical

Specifications do not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270

NRC Project Director: Marvin M. Mendonca

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

Michigan

Date of amendment request: June 20, 1997 (NRC-97-0037), as

supplemented by letter dated July 3, 1997

Description of amendment request: The proposed amendment would

relocate technical specification surveillance requirement 4.4.1.1.2 for

the reactor recirculation system motor-generator (MG) set scoop tube

stop setpoints to the Updated Final Safety Analysis Report. In

addition, the proposed amendment includes the following changes to the

surveillance testing methodology: (1) eliminating any licensing basis

requirement for the electrical stops, and (2) revising the periodicity

from a calendar basis to a situational basis (i.e., plant conditions

that would dictate a change in stop positions).

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 changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change removes from the Fermi 2 Technical

Specifications (TS) a Surveillance Requirement (SR 4.4.1.1.2) that

is an implementation detail and relocates it to the Updated Final

Safety Analysis Report (UFSAR), where it is more adequately and more

appropriately controlled in accordance with 10 CFR 50.59. In

addition, this proposed change revises the test methodology by: (1)

eliminating the requirement for the electrical stops because they

are not credited for mitigating any transients or accidents, and (2)

revising the periodicity from a calendar basis to a situational

basis to coincide with the beginning of each operating cycle or

post-maintenance. These changes do not eliminate the necessary

testing of the MG set mechanical stops. The MG set mechanical stops

will continue to remain operable because the recirculation pump MG

set mechanical speed stop settings will continue to be maintained at

or below the required limits. The MCPRf [minimum critical

power ratio] and MAPLHGRf [maximum average planar linear

heat-generation rate] limits, along with the recirculation pump MG

set mechanical speed stop settings on which they are based, are

specified in the Core Operating Limits Report and operation within

these limits is required by Technical Specifications 3.2.1 and

3.2.3. The changes described will therefore have no impact on the

probability or consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

The proposed Technical Specification change does not result in

any changes to the design (equipment/configuration) or operation of

the plant and will thus not create a new failure mode or common mode

failure. The MG set mechanical stops will continue to operate as

intended and as designed. These changes will therefore not create

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

accident previously evaluated.

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

margin of safety.

Changes in the methodology and frequency of testing will not

involve a significant reduction in the margin of safety because the

testing necessary to ensure the stops are set correctly will

continue to be performed. Additionally, the MCPRf and

MAPLHGRf limits, along with the recirculation pump MG set

mechanical speed stop setting that they are based on, are specified

in the Core Operating Limits Report, and operation within these

limits is still required by Technical Specifications 3.2.1 and

3.2.3. Therefore, the margin of safety as defined in the bases of

any Technical Specification is not reduced by relocating the

surveillance requirement from the TS to the UFSAR. In addition to

the above, relocation of the TS is consistent with the BWR Improved

Standard Technical Specification, NUREG-1433, Rev. 1.

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: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

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

2000 Second Avenue, Detroit, Michigan 48226

NRC Project Director: John N. Hannon

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,

Unit No. 2, Pope County, Arkansas

Date of amendment request: April 24, 1997

Description of amendment request: The requested amendment revises

the inservice inspection requirements associated with steam generator

tube sleeves.

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 Not Involve a Significant Increase in the Probability or

Consequences of an Accident Previously Evaluated.

This change implements a more stringent surveillance requirement

than currently exists. It incorporates a requirement to inspect a

minimum of 20% of each type of installed sleeve in each steam

generator. The 20% inspection criterion is conservative with respect

to the existing requirement of a 3% initial inspection of all steam

generator tubes. Additionally, since the process for inspections has

not changed, the probability or consequences of accidents previously

analyzed are not increased as a result of inspection activities. The

proposed changes have no impact on any previously analyzed accident

in the safety analysis report.

The administrative changes made to update the technical

specifications or to correct inconsistencies introduced in previous

amendments do not affect reactor operations or accidental analyses

and have no radiological consequences.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

2. Does Not Create the Possibility of a New or Different Kind of

Accident from any Previously Evaluated.

The changes made to increase the initial sample of sleeved tubes

inspected during a surveillance, to update the technical

specifications and to correct inconsistencies introduced in previous

amendments are administrative and do not change the design,

configuration or method of operation of the plant nor does it

introduce any new possibility for an accident.

Therefore, this change does not create the possibility of a new

or different kind of accident from any previously evaluated.

3. Does Not Involve a Significant Reduction in the Margin of

Safety.

As previously discussed, this change implements a more stringent

surveillance requirement than currently exists. The existing

technical specifications require an initial inspection of 3% of the

tubes in each steam generator while the proposed change

[[Page 38135]]

requires inspection of a minimum of 20% of each type of installed

sleeve. The 20% inspection criterion is conservative with respect to

the existing technical specification. Existing technical

specification operability and surveillance requirements are not

reduced by the proposed change, thus no margins of safety are

reduced.

The other administrative changes do not reduce technical

specification operability and surveillance requirements, and

therefore, do not reduce any margin of safety.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502

NRC Project Director: William D. Beckner

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 26, 1997

Description of amendment request: The proposed amendment will

modify Technical Specification (TS) Tables 3.7-1 and 3.7-2. Table 3.7-1

will be revised to change the Main Steam Safety Valves (MSSVs) orifice

size from 26 square inches to 28.27 square inches and to relocate the

orifice size from the TS Table to the TS Bases. The change to correct

the orifice size is an editorial change to make the TS consistent with

plant design. Table 3.7-2 will be revised by deleting the provision

that allows continued plant operation with three MSSVs inoperable. The

proposed amendment will also revise TS Bases 3/4.7.1.1 to remove the

equation used for determining the reduced maximum allowable linear

power level-high reactor trip settings of TS Table 3.7-2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No

In response to the ABB/CE report pursuant to 10CFR21 regarding

the omission of Main Steam Safety Valve (MSSV) piping pressure loss

in safety analyses, the proposed change will eliminate the ability

to operate the plant in accordance with Technical Specification

3.7.1.1 Action a with three MSSVs inoperable. The Bases to this

Technical Specification will also be revised to state that the

acceptability for operation at lower power levels with one or two

MSSVs inoperable will be determined from results obtained from a

loss of condenser vacuum accident analysis under these conditions.

Deleting the allowance for continued operation with three MSSVs

inoperable does not increase the probability of an accident. The

consequences of an accident will not be increased by these changes.

These changes are more restrictive and ensure that the MSSVs

maintain their safety function of removing adequate heat from the

steam generator in order to maintain peak steam generator pressure

and peak pressurizer pressure well below their respective acceptance

criteria during normal operation and all anticipated operational

occurrences.

Changing the MSSVs orifice size listed in TS to their actual

size and the orifice size utilized in the safety analysis, and

relocating the MSSVs orifice size to the Technical Specification

Bases does not affect the probability or consequences of an

accident. The correct orifice size was used in the safety analysis

and it is not subject to change unless a station modification is

performed which will require a 10CFR50.59 evaluation and revision of

the safety analysis. The MSSVs orifice size can be adequately

controlled in the TS Bases which will also require a 10CFR50.59 to

be changed.

Therefore, operation of Waterford 3 in accordance with this

proposed change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No

The proposed change will eliminate the ability to operate the

plant in compliance with Technical Specification 3.7.1.1 Action a

with three MSSVs inoperable. The Bases for this Technical

Specification will also be revised to state that the

acceptability for operation at lower power levels with one or two

MSSVs inoperable will be determined from results obtained from a

loss of condenser vacuum accident under these conditions. The

proposed change also revises the MSSVs orifice size to reflect the

actual orifice size and the orifice size utilized in the safety

analysis, and relocates the orifice size from Technical

Specifications to the Technical Specification Bases. The proposed

change does not involve any new equipment, components, or

modifications and does not create any new system interactions or

connections. Therefore, operation of Waterford 3 in accordance with

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

different type of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No

The proposed change will ensure that all appropriate acceptance

criteria for the MSSVs are met during normal operation and all

anticipated operational occurrences. The Technical Specification

Bases 3/4.7.1.1 will be updated to state that the acceptance

criteria for operation in accordance with Technical Specification

3.7.1.1 Action a will be determined from the results of the limiting

loss of condenser vacuum accident. This change ensures that the

transient and dynamic effects which occur during accident scenarios

are fully evaluated. These changes also ensure that the MSSVs will

maintain peak steam generator pressure and peak pressurizer pressure

well below their respective acceptance criteria during normal

operation, design basis accidents and anticipated operational

occurrences.

The proposed change also revises the MSSVs orifice size to

reflect the actual orifice size and the orifice size utilized in the

safety analysis, and relocates the orifice size from Technical

Specifications to the Technical Specification Bases. This change

corrects an editorial error in the Technical Specifications and

relocates unsurveilled design details from the Technical

Specifications. Adequate control of the orifice size will remain

adequate because any changes to the orifice size or the orifice size

listed in the Bases will require a station modification and a TS

Bases change. Station Modifications and TS Bases changes requires

evaluation in accordance with 10CFR50.59.

Therefore, operation of Waterford 3 in accordance with this

proposed change will not involve a significant reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

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

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: James W. Clifford, Acting

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: May 5, 1997

Description of amendment request: The proposed amendment to

Technical Specifications 3.9.1.2 and 3.9.13 and

[[Page 38136]]

their Bases would allow crediting soluble boron for maintaining k-

effective at less than or equal to 0.95 within the spent fuel pool

(SFP) rack matrix following a seismic event of a magnitude greater than

or equal to an operating basis earthquake (OBE).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed change in accordance with

10CFR50.92 and has concluded that the change does not involve a

Significant Hazards Consideration (SHC). The bases for this

conclusion is that the three criteria of 10CFR50.92(c) are not

satisfied. The proposed change does not involve [an] SHC because the

change would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

There is one Spent Fuel Pool accident condition discussed in

Chapter 15 of the FSAR [Final Safety Analysis Report]. The FSAR

discusses a fuel handling accident which drops a fuel assembly onto

the fuel racks during fuel movement. Degradation of the Boraflex

panels in a post-seismic condition will have no effect on the

probability of a fuel assembly drop onto the stored fuel, or the

fuel racks. Changing the way Boraflex responds to a seismic event

will have no impact on the probability of a seismic event. A

misplaced fuel assembly can be postulated in the MP3 [Millstone Unit

3] fuel pool as a result of either equipment malfunction or operator

error. Degradation of the Boraflex panels will have no effect on the

probability of a fuel misplacement event. Therefore, the degradation

of Boraflex in a post-seismic condition does not involve an increase

in the probability of an accident previously evaluated.

A fuel handling accident could cause a radioactive release of

fission gases, resulting in dose consequences. This radioactive

release of fission gases is due to the failure of a certain number

of fuel pins which are postulated to fail during the fuel handling

accident. The number of fuel pins which are postulated to fail in

this event is not affected by the degradation of the Boraflex panels

in a post-seismic condition. There are no criticality issues with

this fuel handling accident for the reasons described next. Should a

fuel handling accident occur prior to a seismic event, the existing

fuel handling accident/misloading criticality analysis is still

valid, such that 800 ppm [parts per million] of soluble boron is

sufficient to ensure that K-effective of the SFP is maintained at

less than 0.95. Although overly conservative, should a fuel handling

accident occur during or after a seismic event, even with no

Boraflex credit, the proposed 1750 ppm of soluble boron is

sufficient to ensure that K-effective of the SFP is maintained at

less than 0.95. Therefore, this proposed change does not involve an

increase in the probability or consequences of an accident

previously evaluated.

Therefore, the proposed changes do 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 in the way Boraflex in conjunction with the addition

of 1750 ppm boron responds to a seismic event does not create a new

accident. The use of soluble boron in the Spent Fuel Pool is safe

during and immediately following a seismic event, because the

balance of the equipment in the fuel building not connected to the

fuel pool which could cause a dilution (firewater, hot water

heating, and demineralized water, CCP [component cooling-plant]) are

seismic or mounted in such a fashion as to not direct unborated

water into the fuel pool should a line rupture. Non borated water

sources that are connected to the SFP will be isolated following a

seismic event of greater than or equal to [an] OBE to prevent

dilution. Therefore there is no possibility of [an] SFP boron

dilution accident coincident with a seismic event, and credit for

soluble boron is acceptable to meet the K-effective limit of 0.95

for the SFP. The crediting of soluble boron in the Spent Fuel Pool

to control K-effective following a seismic event does not create a

new accident as boron dilution of the pool can be prevented by

closing and administratively controlling the opening of dilution

paths to the pool and initiating routine sampling requirements on

SFP boron. At present the crediting of soluble boron following a

fuel misplacement event is allowed for the Millstone 3 Spent Fuel

Pool. Analysis has shown that a seismic event of greater than an OBE

level earthquake can be more limiting than a fuel misplacement

event. As such the minimum boron requirement in the fuel pool will

be increased from 800 ppm to 1750 ppm. As such, no new accident has

been created because the crediting of boron following a malfunction/

accident has always been an allowed event.

Therefore, the proposed change 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 margin of safety, as defined by MP3 Technical

Specifications, is to ensure that the K-effective of the MP3 SFP is

maintained less than or equal to 0.95 at all times. There is no

reduction in the margin of safety as the result of the degradation

of Boraflex following a greater than an OBE seismic event, because

soluble boron can be used to compensate for the loss of Boraflex. A

value of 1750 ppm of soluble boron in the SFP at all times ensures

that K-effective of the MP3 SFP is maintained less than or equal to

0.95 at all times, including this new malfunction of degraded

Boraflex following a greater than an OBE seismic event.

Eliminating the credit for the negative reactivity effect of

Boraflex panels in conjunction with the addition of 1750 ppm boron

will have no effect on the probability of a seismic event. As the

probability of a seismic event has not changed there is no increase

in the probability of an accident or malfunction due to a seismic

event. Following a seismic event operators are presently required to

make inspections of the plant to determine post seismic event plant

conditions. As a result of this change, inspections will be required

to post seismic event evaluations to review the status of the Spent

Fuel Pool and isolate potential dilution paths. These action are

consistent with present guidance in the seismic response procedure

and do not create an undue burden on the operator. To compensate for

the potential

loss of Boraflex after a seismic event, the SFP is now required

to be borated at all times to 1750 ppm to maintain the proper post

seismic [K-effective] condition. As such there is no mitigation

equipment that has to operate in the Spent Fuel Pool following a

seismic event.

Although the Boraflex in the fuel racks is assumed to fail in a

greater than an OBE seismic event, the presence of soluble boron in

the fuel pool water will compensate for the loss of Boraflex.

Surveillance requirements on SFP boron will ensure that there will

be boron present in the SFP and ensure that the SFP is not diluted

below the minimum required boron concentration during normal

operation.

As the presence of SFP soluble boron during and after a seismic

event maintains [K-effective] less than 0.95 there is no effect on

the consequences of any malfunctions evaluated. As there are no new

accidents created and there are no changes in the probability or

consequences of previously analyzed accidents there is no effect on

the consequences of any accident. There is no reduction in the

margin of safety as the result of the degradation of Boraflex

following a greater than an OBE seismic event, because soluble boron

can be used to compensate for the loss of Boraflex to maintain K-

effective less than 0.95.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

In conclusion, bases on the information provided, it is

determined that the proposed change 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,

CT 06141-0270 NRC Deputy Director: Phillip F. McKee

[[Page 38137]]

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

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: March 26, 1997

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to incorporate additional

restrictions on the operation of the main steam safety valves (MSSVs).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The Omaha Public Power District (OPPD) proposes to revise the

Fort Calhoun Station (FCS) Unit No. 1 Technical Specifications (TS)

2.1.6, ``Pressurizer and Main Steam Safety Valves,'' to incorporate

additional restrictions on the Main Steam Safety Valves (MSSVs) as a

result of recent engineering analyses.

FCS has two Steam Generators (SG), each with one 2 1/2-inch MSSV

and four 6-inch MSSVs. The purpose of the MSSVs is to limit the

secondary system pressure to less than or equal to 110% of the

design pressure of 1000 lbs. per square inch absolute (psia) when

passing 100% of design steam flow.

The pressure drops in the main steam lines were calculated. The

total losses (line losses and valve losses) of 30.5 psid (2 1/2 inch

valves) and 33.5 psid (6 inch valves) were compared to the valve

blowdown which is adjusted/checked each refueling outage as part of

the required surveillance test. The pressure losses are less than

the 39 psid and 40 psid blowdown for the 2 1/2 inch and 6 inch valve

with the lowest setpoint (respectively). Therefore, the

recommendation from the Part 21 to review blowdown settings to

preclude valve chatter was conducted and there is no concern at FCS.

A review of existing calculations for line losses in the primary

system was conducted and was determined to be 39 psid for the inlets

to the primary safety valves.

Analyses were then conducted to determine the impact of the

total line losses on previously analyzed accidents documented in the

Updated Safety Analysis Report (USAR). The scope of the analyses was

to evaluate the pressure drops in the piping run for both the

primary and MSSVs to determine the impact on the peak primary and

secondary system pressures. The applicable transient for peak

primary system pressure is the Loss of Load, and for maximum

secondary system pressure is the Loss of Feedwater. All analyses

were performed using the NRC-approved CESEC-III transient analysis

methodology and computer code.

The assumptions of the analyses were that the plant is operating

at 1535.6 MWt, (100% power + 2% uncertainty + reactor coolant pump

heat), the MSSVs lifted at +3% of their nominal setpoints, the

primary safety valve setpoints were adjusted to account for line

losses and lifting at +1% of their setpoints, and the pressure

losses in the main steam line to the SG were added to obtain the

maximum secondary system pressure within the SG. Additional cases

were evaluated with a +6% primary safety valve drift since this

possibility is described in the Bases to TS 2.1.6.

The results from these analyses confirm that the effective

increase in MSSV set pressure caused by the piping pressure losses

leading to the primary safeties and MSSVs is below the 1100 psia

design limit for the secondary system, and below the 2750 psia

design limit for the primary system. This is predicated on the fact

that only one (1) MSSV may be inoperable per SG.

Failure of a MSSV is not an initiator of any previously analyzed

accident, and therefore the proposed changes do not increase the

probability of an accident previously analyzed. The proposed change

to revise TS 2.1.6 to allow only one MSSV per SG to be inoperable

has been shown, utilizing NRC approved methodology, to

limit the design pressure to values below the design limits. An

administrative change to revise the TS setpoint value for both the

primary safety valves and MSSVs from pounds absolute to pounds gauge

is proposed to be consistent with the nameplate values of the valves

and has no effect on any analyses. Therefore the proposed changes do

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

There will be no physical alterations to the plant

configuration, changes in operating modes, setpoints, or testing

methods. The additional restrictions being incorporated into the TS

on MSSV operation will ensure that the design basis limits of 110%

of design pressure will be met for the primary and secondary systems

for analyzed accidents when considering inlet pipe pressure drops.

The possibility of valve chatter being caused by the additional

pressure losses identified in the Main Steam lines and MSSVs was

reviewed and is not a concern. This is due to the valve blowdown

(the difference between a valve's opening pressure and closing

pressure) being greater than the pressure losses. Therefore, the

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

kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change results in a peak primary pressure of 2649

psia (with 1% primary safety valve drift as allowed by TS 2.1.6) and

peak secondary pressure of 1081 psia for the loss of load event

compared to 2632 psia and 1075 psia documented in USAR Section 14.9.

The proposed change results in a peak primary pressure of 2562 psia

and peak secondary pressure of 1090 psia for the loss of feedwater

event compared to 2487 psia and 1052 psia documented in USAR Section

14.10. The analyses confirm that the primary and secondary systems

will continue to be below their respective design limits of 2750

psia and 1100 psia. Therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502

NRC Project Director: William H. Bateman

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 amendment request: May 31, 1996

Description of amendment request: This change deletes Technical

Specification 4.7.2.d.2, ``Control Room Emergency Outside Air Supply

System Surveillance Requirement,'' related to the detection of

chlorine.

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. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Review of the various design basis accidents identified in

Chapter 15 of the Susquehanna SES [Steam Electric Station] Final

Safety Analyses Report (FSAR) concluded that none of these accidents

are affected by deletion of the chlorine detection surveillance

requirement from Technical Specifications. With the elimination of

bulk quantities of gaseous chlorine from use at Susquehanna SES the

probability of control room inhabitability due to a gaseous chlorine

release has actually decreased. Therefore, this proposed change does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. This proposal does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed change involves only the deletion of the chlorine

detection system Technical Specifications based upon a plant

[[Page 38138]]

modification to remove gaseous chlorine as a biocide from

Susquehanna SES and replace it with an oxidizing biocide with non-

gaseous/non-volatile properties. The release of chlorine from an

off-site source is bounded by Reg. [Regulatory] Guide 1.95 in that

manual isolation capability for the control room ventilation system

is acceptable. Therefore, the proposed change does not create the

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

previously evaluated.

3. This change does not involve a significant reduction in a

margin of safety.

The proposed change would not alter the margins of safety

provided in the existing FSAR analysis (Sections 2.2.3.1.3 and 6.4)

for chlorine release events since the basis for the existing margin

of safety, which are the Reg. Guide 1.95 requirements, are not

altered by the change. As stated above, since gaseous chlorine is no

longer used for open cooling water treatment at Susquehanna SES and

since the biocide currently used does not pose the same personnel

inhalation threat as gaseous chlorine, safety margin has actually

increased. Therefore, the proposed change does not involve a

significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Osterhout Free Library,

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

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: John F. Stolz

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

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama

Date of amendments request: June 13, 1997

Description of amendments request: The proposed amendments would

change Technical Specification (TS) 3/4.9.13, ``Storage Pool

Ventilation (Fuel Movement),'' by adding a note in the TSs to

specifically indicate that the normal emergency power source may be

inoperable in MODE 5 or 6 provided that the requirements of TS 3.8.1.2

are satisfied and extend the TS 3.9.13 completion time allowed for

returning one out-of-service penetration room filtration system from 48

hours to 7 days. The Bases will also be modified to provide additional

detail concerning these changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated in

the FSAR [Final Safety Analysis Report]. The proposed changes have

no impact on the probability of an accident. The storage pool

ventilation system will continue to ensure that radioactive material

released as a result of a fuel handling accident in the spent fuel

pool room will be filtered through the HEPA [high efficiency

particulate air] filters and charcoal absorbers prior to discharge

to the atmosphere. There is no change in the FNP [Farley Nuclear

Plant] design basis as a result of this change and, as a result,

does not involve a significant increase in the consequences of an

accident previously evaluated.

(2) The proposed changes to the TSs do not increase the

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

already evaluated in the FSAR. No new limiting single failure or

accident scenario has been created or identified due to the proposed

changes. Safety-related systems will continue to perform as

designed. The proposed changes do not create the possibility of a

new or different kind of accident from any previously evaluated.

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

in the margin of safety. As a result of these proposed changes, the

penetration room filtration system, when it is aligned to the spent

fuel pool room, will continue to require verification of

operability. There is no impact in the accident analyses. These

proposed changes are technically consistent with the requirements of

NUREG-1431, Revision 1 which has already received the requisite

review and approval of the NRC staff. Thus the proposed changes do

not involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Houston-Love Memorial Library,

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

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

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

Alabama 35201

NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch

Nuclear Plant, Unit 1, Appling County, Georgia

Date of amendment request: April 29, 1997, as supplemented by

letter dated May 28, 1997

Description of amendment request: The amendment would revise the

Unit 1 reactor vessel pressure and temperature limits to reflect data

collected from the material sample recovered during the March 1996 Unit

1 outage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Pressure and Temperature (P/T) limits for the reactor pressure

vessel are established to the requirements of 10 CFR [Part] 50,

Appendix G to ensure brittle fracture of the vessel does not occur.

This revision changes the P/T curves in the Unit 1 Technical

Specifications to reflect the material capsule surveillance results

from the sample removed during the [s]pring outage of 1996.

The RPV [reactor pressure vessel] surveillance capsule contained

flux wires for neutron flux monitoring and Charpy V notch impact and

tensile test specimens. The irradiated material properties were

compared to available unirradiated properties to determine the

effect of irradiation on material toughness for the base and weld

materials through Charpy testing. Irradiated tensile testing results

are compared with unirradiated data to determine the effect of

irradiation on the stress-strain relationship of the materials.

The P/T curves are modified to reflect the results of the above

examination. These curves and their operating limits were evaluated

using the approved methodologies of 10 CFR [Part] 50 Appendix G and

ASME [American Society of Mechanical Engineers] Code Appendix G. The

new curves therefore represent the latest information available on

the state of the reactor vessel materials. The P/T curves are

generated for reactor vessel protection against brittle fracture,

they do not affect the recirculation piping. Accordingly, the

probability of occurrence of a design basis Loss of Coolant Accident

(LOCA) is not increased. Likewise, no other previously evaluated

accident and transients, as defined in Chapter 14 of the Final

Safety Analysis Report (FSAR) are affected by this proposed change

to the Unit 1 P/T curves. Additionally, this proposed revision does

not affect the design, operation, or maintenance of any safety

related system designed for the mitigation or prevention of

previously analyzed events.

[[Page 38139]]

Since no previously evaluated accidents or transients are being

affected by this change, their probability of occurrence is not

increased and their consequences are not made worse.

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

different type of accident from any previously evaluated?

Implementing the proposed P/T curves into the Unit 1 Technical

Specifications does not alter the design or operation of any system

or piece of equipment designed for the prevention or mitigation of

accidents and transients. As a result, no new operating modes are

introduced from which a new type accident becomes possible. Existing

systems will continue to be operated per present design basis

assumptions.

The proposed P/T limits were generated from the evaluation of

the material capsule removed during the [s]pring Unit 1 outage of

1996. As a result, these limits include the latest available

information on the reactor vessel materials. Furthermore, they will

continue to be monitored per the requirements of the Technical

Specifications and 10 CFR [Part] 50 Appendices G and H. For the

above reasons, the changes do not create the possibility of a new

type of accident.

3. Do the proposed changes involve a significant reduction in

the margin of safety?

The purpose of the P/T limits is to avoid a brittle fracture of

the reactor vessel. As such, material capsules are removed

periodically to determine the effects of neutron irradiation on

reactor vessel materials. This change to the Unit 1 P/T curves is

proposed to incorporate the evaluation results of the latest capsule

removed during the [s]pring Unit 1 outage of 1996. Accordingly,

these curves represent the latest information available on the

reactor vessel materials. Also, the curves were generated using the

approved methodologies of 10 CFR [Part] 50 Appendix G.

The pressure test curve (Figure 3.4.9-1) is also being revised

to reflect exposure dependencies. These curves were generated for

exposures of 16, 18, 20, 24, 28, and 32 EFPY [effective full-power

year]. As previously described, each of these curves were generated

using approved methodologies and all reflect the results of this

latest material capsule report.

The proposed change does not affect the evaluation of any FSAR

Unit 1 Chapter 14 transient and accident. Furthermore, the proposed

change does not affect the operation of systems or equipment

important to safety.

The Limiting Condition for Operation of Specification 3.4.9 will

not change. Also, no Technical Specification surveillances or

surveillance frequencies are revised as a result of this Technical

Specification submittal, besides the fact that the P/T surveillances

will now refer to the revised curves. Procedures regarding the

monitoring of the P/T limits during reactor startup, cooldown, and

leakage testing will not change as a result of this proposed

Technical Specification change with respect to frequency of the

surveillance or the methods used to perform the surveillances. Thus,

the P/T limits will continue to be surveilled as before per the same

procedures and the same frequencies.

No other Technical Specifications are affected by the proposed

revision. The margin of safety to any Technical Specifications

safety limit therefore is not reduced.

For the above reasons the new curves do not represent a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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

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

NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366,

Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County,

Georgia

Date of amendment request: May 30, 1997

Description of amendment request: The proposed amendments would

revise power sources to valves associated with low pressure coolant

injection (LPCI) mode of residual heat removal (RHR) system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

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

the probability or consequences of an accident previously evaluated.

The LPCI valves operate to establish and maintain adequate core

cooling following a LOCA [loss-of-coolant accident]. The proposed

changes do not alter the function or mode of operation of the LPCI

valves. Therefore, the probability of the LOCA accident is not

increased. An analysis which considered the consequences of the

various transients and accidents with the proposed change in power

supply of the LPCI valves indicates the consequences are not

increased.

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

or different kind of accident from any accident previously analyzed.

The change in power supply to the LPCI valves maintains the original

design criteria that a power supply independent of the remaining RHR

subsystem be utilized for single-failure criteria. The function of

the LPCI valves and any other existing equipment is not altered.

Operation of the valves in the proposed configuration was analyzed,

and no new failure modes exist. An analysis of the impact on the

operation and design of other systems and components indicates no

new failure modes are introduced. Therefore, these changes do not

contribute to a new or different type of accident.

3. The proposed changes do not involve a significant reduction

in the margin of safety. The change in power supply to the LPCI

valves was evaluated relative to RHR and electrical distribution

system function during normal and accident conditions. The proposed

change does not alter the performance of any system safety

functions. The results of the SAFER-GESTR LOCA analysis reconfirm

the large margins existing in fuel peak cladding temperature under

the proposed configuration. Therefore, there is no significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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

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

NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and 50-425,

Vogtle Electric Generating Plant, Units 1 and 2, Burke County,

Georgia

Date of amendment request: June 13, 1997

Description of amendment request: The proposed amendments would

revise the Technical Specification Limiting Condition for Operation

3.4.10 Pressurizer Safety Valves. Specifically, the change would reduce

the nominal set pressure by 1 percent to 2460 pounds per square inch

gauge (psig) and increase the tolerance to plus or minus 2 percent.

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:

[[Page 38140]]

1. Does the proposed change involve a significant increase in

the probability or consequences of an accident previously evaluated?

The increase in the PSV [pressurizer safety valve] tolerance

from [plus or minus] 1% with a setpoint of 2485 psig to [plus or

minus] 2% and reduction in the nominal setpoint from 2485 psig to

2460 psig has the net effect of reducing the minimum lift setting

allowed by the TS [technical specifications] from 2460 psig to 2410

psig. The effects of this change have been evaluated for its impact

on the assumed frequency of safety valve challenges and failures to

reclose, and the proposed change was found to have a negligible

impact. In other words, reducing the minimum lift setting does not

significantly increase the probability of an inadvertent actuation

of a safety valve during normal operation. Reducing the minimum lift

setting does increase the potential that the PSVs may open during an

event, but this change has been evaluated and does not adversely

impact the consequences of any accident previously evaluated. No

change to any equipment response or accident mitigation scenario has

resulted, and there are no additional challenges to fission product

barrier integrity. Therefore, the proposed change does not

significantly increase the probability or consequences of any

accident previously evaluated.

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

different kind of accident from any accident previously evaluated?

The increase in the PSV tolerance from [plus or minus] 1% with a

setpoint of 2485 psig to [plus or minus] 2% and reduction in the

nominal setpoint from 2485 psig to 2460 psig does not create the

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

previously evaluated. No new accident scenarios, failure mechanisms,

or limiting single failures are introduced as a result of this

proposed change. The proposed revision to Technical Specification

3.4.10 does not challenge the performance or integrity of any

safety-related systems. Therefore, the possibility of a new or

different kind of accident is not created.

3. Does the proposed change involve a significant reduction in a

margin of safety.

The proposed change to Technical Specification 3.4.10 does not

involve a significant reduction in a margin of safety. The

modification will have no affect on the availability, operability or

performance of the safety-related systems and components. The

increased PSV set pressure tolerance has been reviewed with respect

to the accident analysis assumptions and requirements and evaluated

or analyzed, as required. These evaluations and analyses determined

that all applicable acceptance criteria continue to be met, thus the

proposed increase in the PSV set pressure tolerance will not result

in a significant reduction in the margin of safety associated with

the acceptance criteria for the accident analyses.

The Bases of the Technical Specifications rely in part on the

ability of the regulatory criteria being satisfied assuming the

limiting conditions for operation for various systems. Conformance

to the regulatory criteria for operation with the increased PSV set

pressure tolerance is demonstrated, and the regulatory limits are

not exceeded. Hence, the margin of safety as defined in the Bases

for the Technical Specifications is not significantly reduced.

Therefore, there is no significant reduction in any margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830

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

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

Georgia 30308

NRC Project Director: Herbert N. Berkow

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: May 16, 1997 (TXX-97119)

Brief description of amendments: The licensee has proposed revised

core safety limit curves and Overtemperature N-16 reactor trip

setpoints based on analyses of the core configuration for CPSES Unit 2,

Cycle 4. These changes apply equally to CPSES Units 1 and 2 licenses

since the Technical Specifications are combined.

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?

A. Revision to the Unit 2 Core Safety Limits

Analyses of reactor core safety limits are required as part of

reload calculations for each cycle. TU Electric has performed the

analyses of the Unit 2, Cycle 4 core configuration to determine the

reactor core safety limits. The methodologies and safety analysis

values result in new operating curves which, in general, permit

plant operation over a similar range of acceptable conditions. This

change means that if a transient were to occur with the plant

operating at the limits of the new curve, a different temperature

and power level might be attained

than if the plant were operating within the bounds of the old

curves. However, since the new curves were developed using NRC

approved methodologies which are wholly consistent with and do not

represent a change in the Technical Specification BASES for safety

limits, all applicable postulated transients will continue to be

properly mitigated. As a result, there will be no significant

increase in the consequences, as determined by accident analyses, of

any accident previously evaluated.

B. Revision to Unit 2 Overtemperature N-16 Reactor Trip

Setpoints

As a result of changes discussed, the Overtemperature reactor

trip setpoint has been recalculated. These trip setpoints help

ensure that the core safety limits are protected and that all

applicable limits of the safety analysis are met.

Based on the calculations performed, no significant changes to

the safety analysis values for Overtemperature reactor trip setpoint

were required. The f(delta I) trip reset function was revised due to

more top-skewed axial power distributions predicted for this cycle.

The analyses performed show that, using the TU Electric

methodologies, all applicable limits of the safety analysis are met.

This setpoint provides a trip function which allows the mitigation

of postulated accidents and has no impact on accident initiation.

Therefore, the changes in safety analysis values do not involve an

increase in the probability of an accident and, based on satisfying

all applicable safety analysis limits, there is no significant

increase in the consequences of any accident previously evaluated.

In addition, sufficient operating margin has been maintained in

the overtemperature setpoint such that the risk of turbine runbacks

or reactor trips due to upper plenum flow anomalies or other

operational transients will be minimized, thus reducing potential

challenges to the plant safety systems.

SUMMARY

The changes in the amendment request applies NRC approved

methodologies to changes in safety analysis values, new core safety

limits and new N-16 setpoint and parameter values to assure that all

applicable safety analysis limits have been met. The potential for

an operational transient to occur has not been affected and there

has been no significant impact on the consequences of any accident

previously evaluated.

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 calculation of new reactor core

safety limits and overtemperature reactor trip setpoint resets. As

such, the changes play an important role in the analysis of

postulated accidents but 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?

[[Page 38141]]

In reviewing and approving the methods used for safety analyses

and calculations, the NRC has approved the safety analysis limits

which establish the margin of safety to be maintained. While the

actual impact on safety is discussed in response to question 1, the

impact on margin of safety is discussed below:

A. Revision to the Unit 2 Reactor Core Safety Limits

The TU Electric reload analysis methods have been used to

determine new reactor core safety limits. All applicable safety

analysis limits have been met. The methods used are wholly

consistent with Technical Specification BASES 2.1 which is the bases

for the safety limits. In particular, the curves assure that for

Unit 2, Cycle 4, the calculated DNBR is no less than the safety

analysis limit and the average enthalpy at the vessel exit is less

than the enthalpy of saturated liquid. The acceptance criteria

remains valid and continues to be satisfied; therefore, no change in

a margin of safety occurs.

B. Revision to Unit 2 Overtemperature N-16 Reactor Trip

Setpoints

Because the reactor core safety limits for CPSES Unit 2, Cycle 4

are recalculated, the Reactor Trip System instrumentation setpoint

values for the Overtemperature N-16 reactor trip setpoint which

protect the reactor core safety limits must also be recalculated.

The Overtemperature N-16 reactor trip setpoint helps prevent the

core and Reactor Coolant System from exceeding their safety limits

during normal operation and design basis anticipated operational

occurrences. However, it was shown in these calculations that the

current Unit 2 overtemperature reactor trip setpoint (presented in

the current Technical Specifications and excluding the f(delta I)

trip reset function) remains valid. The most relevant design basis

analysis in Chapter 15 of the CPSES Final Safety Analysis Report

(FSAR) which is affected by the Overtemperature reactor trip

setpoint is the Uncontrolled Rod Cluster Control Assembly Bank

Withdrawal at Power (FSAR Section 15.4.2). This event has been

analyzed with the new safety analysis value for the Overtemperature

reactor trip setpoint to demonstrate compliance with event specific

acceptance criteria. Because all event acceptance criteria are

satisfied, there is no degradation in a margin of safety.

The nominal Reactor Trip System instrumentation setpoints values

for the Overtemperature N-16 reactor trip setpoint (Technical

Specification Table 2.2-1) are determined based on a statistical

combination of all of the uncertainties in the channels to arrive at

a total uncertainty. The total uncertainty plus additional margin is

applied in a conservative direction to the safety analysis trip

setpoint value to arrive at the nominal and allowable values

presented in Technical Specification Table 2.2-1. Meeting the

requirements of Technical Specification Table 2.2-1 assures that the

Overtemperature reactor trip setpoint assumed in the safety analyses

remains valid. The CPSES Unit 2, Cycle 4 Overtemperature reactor

trip setpoint is not significantly different from the previous

cycle, and thus provides operational flexibility to withstand mild

transients without initiating automatic protective actions. Although

the value of the f(delta I) trip reset function setpoint is

different, the Reactor Trip System instrumentation setpoint values

for the Overtemperature N-16 reactor trip setpoint are consistent

with the safety analysis assumptions which have been analytically

demonstrated to be adequate to meet the applicable event acceptance

criteria. Thus, there is no reduction in a margin of safety.

Using the NRC approved TU Electric methods, the reactor core

safety limits are determined such that all applicable limits of the

safety analyses are met. Because the applicable event acceptance

criteria continue to be met, there is no significant reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: James W. Clifford, Acting

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.

Commonwealth Edison Company, Docket No. STN 50-455, Byron Station,

Unit No. 2, Ogle County, Illinois Docket No. STN 50-457, Braidwood

Station, Unit No. 2, Will County, Illinois

Date of amendment request: May 24, 1997

Description of amendment request: The amendments revise the

technical specifications related to venting of the emergency core

cooling system pumps and associated piping. The application originally

included Byron, Unit 1. However, on May 31, 1997, ComEd supplemented

the application to request an emergency license amendment for Byron,

Unit 1. Amendment No. 90 was issued on June 1, 1997.

Date of publication of individual notice in Federal Register: June

10, 1997 (62 FR 31633)

Expiration date of individual notice: July 10, 1997

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

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile

Point Nuclear Station, Unit No. 1, Oswego County, New York

Date of application for amendment: May 16, 1997

Brief description of amendment: The proposed amendment would make

an administrative change to add a supervisory position to the list of

personnel who may be required to hold a senior reactor operator

license. Date of publication of individual notice in Federal Register:

June 4, 1997 (62 FR 30625)

Expiration date of individual notice: July 7, 1997

Local Public Document Room location: Reference and Documents

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

York 13126

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.

[[Page 38142]]

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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of application for amendment: January 20, 1997, with the

proposed no significant hazards consideration submitted by letter dated

January 30, 1997, as supplemented February 27, April 11, May 14, and

June 20 (2 letters), 1997

Brief description of amendment: The amendment authorizes Boston

Edison Company (BECo) to change the UHS administrative limit from

68 deg.F to 75 deg.F, and change the Updated Final Safety Analysis

Report (UFSAR) to reflect the use of containment pressure to compensate

for the deficiency in NPSH following a design basis accident and

increase the accident analysis design UHS temperature from 65 deg.F to

75 deg.F. As part of this amendment, BECo has proposed to submit a

Technical Specification amendment for the UHS temperature by the first

quarter of 1998. In addition, within 180 days of issuance of this

amendment, BECo has committed to complete the containment analysis

using the ANS 5.1-1979 Decay Heat Curve with a 2-sigma uncertainty

added. The staff considers BECo's commitments acceptable and has

conditioned the amendment accordingly.

Date of issuance: July 3, 1997

Effective date: July 3, 1997

Amendment No.: 173

Facility Operating License No. DPR-35: Amendment revised the

Updated Final Safety Analysis Report.

Date of initial notice in Federal Register: February 26, 1997 (62

FR 8792) The February 27, April 11, May 14, and June 20 (2 letters),

1997, letters provided clarifying information that did not change the

initial proposed no significant hazards consideration determination as

submitted by letter dated January 30, 1997. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

July 3, 1997. No significant hazards consideration comments received:

No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,

North Carolina

Date of application for amendment: March 14, 1997, as supplemented

May 16, and June 17, 1997

Brief description of amendment: The amendment approves changes to

the Final Safety Analysis Report to reflect new analysis of the

radiological consequences of dropping a fuel cask.

Date of issuance: June 26, 1997

Effective date: June 26, 1997

Amendment No. 73

Facility Operating License No. NPF-63. Amendment revises the Final

Safety Analysis Report.

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

17226). The May 16, and June 17, 1997 supplemental information did not

change the original no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated June 26, 1997. No significant hazards

consideration comments received: No

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: April 11, 1997

Brief description of amendment: The amendment changes the Waterford

steam Electric Station, Unit 3, Technical Specifications (TSs) by

revising TS 3.6.2.2 and Surveillance Requirement 4.6.2.2 for the

Containment Cooling System. Also, a Surveillance Requirement is added

to verify that valves actuate on a Safety Injection Actuation Signal.

To support this addition, Technical Specification Bases 3/4.3.6.2.2 is

also included.

Date of issuance: July 3, 1997

Effective date: July 3, 1997, to be implemented within 60 days.

Amendment No.: 131

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

Technical Specifications.

Date of initial notice in Federal Register: April 22, 1997 (62 FR

19626) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 3, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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

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

Connecticut

Date of application for amendment: April 17, 1997

Brief description of amendment: The amendment modifies Technical

Specification 3.7.14 by clarifying the actions to be taken when an area

temperature exceeds its temperature limit.

Date of issuance: June 24, 1997

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

within 60 days.

Amendment No.: 141

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: (62 FR 27798 May 21,

1997) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 24, 1997. 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 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

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

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

Connecticut

Date of application for amendment: April 15, 1997

Brief description of amendment: The amendment makes changes to

Technical Specification (TS) Sections 4.3.3.6 and 4.6.4.1, which

require that the hydrogen monitors be periodically tested.

Specifically, the changes increase the testing interval of the

monitor's hydrogen sensor, correct inconsistencies

[[Page 38143]]

between the TS surveillances, and make changes to the Bases of the

surveillances.

Date of issuance: June 24, 1997

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

within 60 days.

Amendment No.: 142

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 21, 1997 (62 FR

27797) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 24, 1997. 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 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of application for amendments: April 11, 1997

Brief description of amendments: These amendments revise Technical

Specification (TS) 3/4.6.2.3, ``Containment Cooling System,'' and its

associated Bases section to ensure that the TSs properly test the

containment fan cooling units' post-accident mode of operation.

Date of issuance: June 24, 1997

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

implemented within 60 days.

Amendment Nos. 197 and 180

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 21, 1997 (62 FR

27799) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated June 24, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Tennessee Valley Authority, Docket Nos. 50-327 and 50328, Sequoyah

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

Date of application for amendments: March 13, 1997, as supplemented

on June 26, 1997 (TS 97-01)

Brief description of amendments: The amendments change the

Technical Specifications by raising the allowable U-235 enrichment, as

specified in Section 5.6.1.2, of fuel stored in the new fuel pit

storage racks from 4.5 to 5.0 weight percent.

Date of issuance: July 1, 1997

Effective date: July 1, 1997

Amendment Nos.: 225 and 216

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

revise the technical specifications.

Date of initial notice in Federal Register: May 21, 1997 (62 FR

27802). The June 26, 1997 supplement provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in an environmental assessment dated June 16,

1997, and a Safety Evaluation dated July 1, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: August 27, 1993, as supplemented

by letters dated November 9, 1993, April 26, 1996, and September 25,

1996

Brief description of amendment: The amendment revises the Technical

Specifications to incorporate the revised 10 CFR Part 20, Standards for

Protection Against Radiation.

Date of issuance: June 19, 1997

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

within 60 days.

Amendment No.: 151

Facility Operating License No. DPR-28. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 4, 1995 (60 FR

507) The November 9, 1993, April 26, 1996, and September 25, 1996,

submittals 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 19, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

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

Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of application for amendments: June 4, 1996 (TSCR 188 and

189), as supplemented August 5, September 26, October 21, November 13,

November 20, and December 2, 1996, and January 16, March 20, and April

2, 1997

Brief description of amendments: These amendments revise Technical

Specifications (TS) 15.1, ``Definitions;'' TS 15.2.1, ``Safety Limit,

Reactor Core;'' TS 15.2.3, ``Limiting Safety System Settings,

Protective Instrumentation;'' TS 15.3.1, ``Reactor Coolant System,''

Section C, ``Maximum Coolant Activity,'' and Section G, ``Operational

Limitations;'' TS 15.3.4, ``Steam and Power Conversion System;'' TS

15.3.5, ``Instrumentation System;'' TS 15.4.1, ``Operational Safety

Review;'' TS 15.5.3, ``Design Features-Reactor;'' and TS 15.6.9,

``Plant Reporting Requirements'' to reflect parameters associated with

new steam generators in Unit 2 and changes in analyses that affect both

Units 1 and 2.

Date of issuance: July 1, 1997

Effective date: July 1, 1997. The TS shall be implemented within 45

days from the date of issuance and the Final Safety Analysis Report

changes shall be implemented by June 30, 1998. Implementation of these

amendments includes incorporation of accident analyses submitted in

support of this amendment into the Final Safety Analysis Report in

sufficient detail to support future evaluations performed in accordance

with 10 CFR 50.59 and as described in the licensee's applications dated

June 4, 1996, as supplemented on August 5, September 26, October 21,

November 13, November 20, and December 2, 1996, and January 16, March

20, and April 2, 1997, and evaluated in the staff's safety evaluation

dated July 1, 1997.

Amendment Nos.: 173, 177

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34903 and 61 FR 34904) and April 9, 1997 (62 FR 17243) The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated July 1, 1997. No significant hazards consideration

comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241

[[Page 38144]]

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey County, Kansas

Date of amendment request: March 21, 1997, as supplemented by

letter dated April 15, 1997

Brief description of amendment: The amendment revises Technical

Specification 6.8.5.b to provide an exception to the examination

requirements of Regulatory Guide 1.14, Revision 1, ``Reactor Coolant

Pump Flywheel Integrity'' and delays the inspection of the ``D''

reactor coolant pump flywheel to the Fall 1997 refueling outage. A

typographical error in TS 6.8.5.c is corrected.

Date of issuance: June 24, 1997

Effective date: June 24, 1997, to be implemented within 30 days of

issuance.

Amendment No.: 106

Facility Operating License No. NPF-42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 21, 1997 (62 FR

27803) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 24, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration and

opportunity for a hearing (Exigent Public Announcement Or Emergency

Circumstances)

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 for the

amendment 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.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective 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

application for amendment, (2) the amendment to Facility Operating

License, 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 room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By August 15, 1997, 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

[[Page 38145]]

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

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

and (3) the possible effect of any order which may be entered in the

proceeding on the petitioner's interest. The petition should also

identify the specific aspect(s) of the subject matter of the proceeding

as to which petitioner wishes to intervene. Any person who has filed a

petition for leave to intervene or who has been admitted as a party may

amend the petition without requesting leave of the Board up to 15 days

prior to the first prehearing conference scheduled in the proceeding,

but such an amended petition must satisfy the specificity requirements

described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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 20555-001, Attention: Docketing and Services

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

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

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

General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC

20555-001, 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 the

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

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

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

Date of application for amendment: December 11, 1996, as

supplemented March 27, 1997, April 17, 1997, and June 17, 1997

Brief description of amendment: The amendment revises Technical

Specifications to allow extended rod position indicator deviation

limits, on-line calibration of the rod position indication and to

clarify the operability requirements during calibration.

Date of issuance: June 27, 1997

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

30 days.

Amendment No.: 194

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: No. The NRC published a public

notice of the proposed amendment, issued a proposed finding of no

significant hazards consideration and requested that any comments on

the proposed no significant hazards consideration be provided to the

staff by the close of business on June 25, 1997. The notice was

published in the Peekskill Evening Star on June 20-25, 1997.

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

exigent circumstances, consultation with the State of New York and

final no significant hazards consideration determination are contained

in a Safety Evaluation dated June 27, 1997.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610

North Atlantic Energy Service Corporation, Dockets Nos. 50-443,

Seabrook Station, Unit 1, Seabrook, Massachusetts

Date of amendment request: June 19, 1997

Brief description of amendment: The amendment revised Technical

Specification 6.8.1.6.b. to include a reference to the NRC-approved

Westinghouse Topical Report WCAP-12610-P-A, ``VANTAGE+ Fuel Assembly

Reference Core Report,'' dated April 1995.

Date of issuance: June 24, 1997

Effective date: As of the date of issuance, and to be implemented

before transition into Operational Mode 2 during startup from Refueling

Outage 5.

Amendment No.: 52

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

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: No. The Commission's related

evaluation of the amendment, finding of emergency circumstances,

consultation with the States of New Hampshire and Massachusetts, and

final no significant hazards considerations determination are contained

in the safety evaluation dated June 24, 1997.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, New Hampshire 03833

Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast

Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270

Acting

NRC Project Director: Patrick D. Milano

North Atlantic Energy Service Corporation, Dockets Nos. 50-443,

Seabrook Station, Unit 1, Seabrook, Massachusetts

Date of amendment request: May 29, 1997

Brief description of amendment: The amendment modifies Technical

Specification 5.3.1 by replacing the current term ``zircaloy'' with

terminology that explicitly identifies the NRC-approved Westinghouse

fuel assembly design in use at the Seabrook Station consisting of

assemblies with either ZIRLO or Zircaloy-4 fuel cladding material.

Date of issuance: June 24, 1997

Effective date: As of the date of issuance, and to be implemented

before transition into Operational Mode 2 during startup from Refueling

Outage 5.

[[Page 38146]]

Amendment No.: 53

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

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes. The NRC published a public

notice of the proposed amendment, issued a proposed finding of no

significant hazards consideration, and requested that any comments on

the proposed no significant hazards consideration be provided to the

staff by the close of business on June 10, 1997. The notice was

published in Foster's Daily Democrat and in the Portsmouth Herald on

June 4, 1997. Public comments were received, and they have been

addressed in the staff's safety evaluation.

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

exigent circumstances, consultation with the States of New Hampshire

and Massachusetts, and final no significant hazards determination are

contained in a safety evaluation dated June 24, 1997.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, New Hampshire 03833

Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast

Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270

Acting

NRC Project Director: Patrick D. Milano

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of application for amendment: March 22, 1997, as supplemented

by letters dated April 2, April 3, April 9, April 15, and May 14, 1997.

Additional information was also received by telefax on May 19, 1997.

Brief description of amendment: The amendment revises Surveillance

Requirement (SR) 3.3.1.1.15, Reactor Protection System (RPS) Response

Time functions 3 and 4 and SR 3.3.6.1.7, Primary Containment Isolation

System Response Time, functions 1.a, 1.b, and 1.c, adding a note to

indicate that the sensor is excluded from response time testing when

verifying that the response time is within limits. The amendment also

revises SR 3.3.5.1.7, Emergency Core Cooling System (ECCS) Response

Time by relocating the requirements to SR 3.5.1.8, ECCS Operating, and

adding a note to SR 3.5.1.8 to indicate that no actuation

instrumentation response time measurement is required. Additionally, SR

3.5.1.8 requires that the SR be met in MODES 1, 2, and 3, whereas the

previous SR 3.3.5.1.7 was required to be met in MODES 1, 2, 3, 4, and

5.

Date of Issuance: June 11, 1997

Effective date: June 11, 1997

Amendment No.: 150

Facility Operating License No. NPF-21. The amendment revised the

Technical Specifications. Press release issued requesting comments as

to proposed no significant hazards consideration: Yes. April 11, 1997.

Tri-City Herald (Washington). Comments received: No. The Commission's

related evaluation of the amendments, finding of exigent circumstances,

consultation with the State of Washington and final determination of no

significant hazards consideration are contained in a Safety Evaluation

dated June 11, 1997.

Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005-3502

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

NRC Project Director: William H. Bateman

Dated at Rockville, Maryland, this 9th day of July 1997.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects III/IV, Office of Nuclear

Reactor Regulation

[Doc. 97-18513 Filed 7-15-97; 8:45 am]

BILLING CODE 7590-01-F