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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from September 26, 1998, through October 8, 1998.

The last biweekly notice was published on October 7, 1998 (63 FR

53943).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period.

[[Page 56239]]

However, should circumstances change during the notice period such that

failure to act in a timely way would result, for example, in derating

or shutdown of the facility, the Commission may issue the license

amendment before the expiration of the 30-day notice period, provided

that its final determination is that the amendment involves no

significant hazards consideration. The final determination will

consider all public and State comments received before action is taken.

Should the Commission take this action, it will publish in the Federal

Register a notice of issuance and provide for opportunity for a hearing

after issuance. The Commission expects that the need to take this

action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and

Directives Branch, Division of Administration Services, Office of

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

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

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

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

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

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By November 20, 1998, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

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

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

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

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

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

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

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

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

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

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

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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 amendment request: September 23, 1998.

[[Page 56240]]

Description of amendment request: Carolina Power & Light (CP&L)

proposes to revise the Harris Nuclear Plant Technical Specification

(TS) 3/4.6.1.3, ``Containment Air Locks,'' to clarify the requirements

for locking an air lock door shut. CP&L also proposes to revise TS 3/

4.6.1.3 to be consistent with NUREG 1431, Revision 1, ``Standard

Technical Specifications, Westinghouse Plants,'' dated April 1995.

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 amendment does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

Containment Air Locks are not an accident initiating system as

described in the Final Safety Analysis Report [FSAR]. The proposed

change implements guidance for Technical Specifications associated with

air lock doors consistent with NUREG-1431, Revision 1, ``Standard

Technical Specifications, Westinghouse Plants,'' dated April 1995.

Additionally, clarification is provided to permit locking an inoperable

air lock door as required by Technical Specifications [TS]. The

proposed change does not affect another Structure, System, or

Component. The operation and design of containment air locks will not

be affected by this proposed change. The ability of containment to

mitigate an accident will not be affected by this change.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously evaluated.

Containment Air Locks are designed to form part of the containment

pressure boundary. The proposed change provides for administrative

controls and operating restrictions for air lock doors consistent with

guidance provided by the Commission. Containment Air Locks are not an

accident initiating system as described in the Final Safety Analysis

Report. The proposed change does not affect another Structure, System,

or Component. The operation and design of containment air locks will

not be affected by this proposed change.

Therefore, the proposed change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed amendment does not involve a significant reduction

in the margin of safety.

The proposed change to containment air locks does not affect any of

the parameters that relate to the margin of safety as described in the

Bases of the TS or the FSAR. Accordingly, NRC Acceptance Limits are not

affected by this change.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Pao-Tsin Kuo (Acting).

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

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: July 17, 1998 (Reference NRC-98-0044).

Description of amendment request: The proposed amendment will

revise the License to allow the licensee to possess special nuclear

material in a quantity totaling no more than 15 grams of uranium-235,

uranium-233, or plutonium, or any combination thereof and with

plutonium totaling no more than 2 curies.

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 using the standards in 10 CFR 50.92(c). The licensee's

analysis is presented below:

(1) Does the proposed change significantly increase the probability

or consequences of an accident previously evaluated?

The proposed changes do not involve a significant increase in the

probability or consequences of an accident. Possessing trace amounts of

special nuclear material cannot affect the probability of the analyzed

sodium or liquid waste accidents. The ability to possess such material

does not itself change any methods of handling liquid waste or sodium.

Possession of special nuclear material could potentially increase the

consequences of an accident if it was in use or in the vicinity if an

accident occurs. However, the increase in consequences would not be

significant due to the limitations on radioactivity content of such

special nuclear material. The special nuclear material limit is below

that requiring an emergency plan or maximum dose evaluation per 10 CFR

70.22(i). Since the quantity is below that requiring an offsite

emergency plan or evaluation, even if all the special nuclear material

allowed to be possessed by the proposed amendment were released during

a postulated accident, the consequences would not be significantly

increased. If the provision allowing for possession of more than 15

grams of special nuclear material or 2 curies of plutonium were to be

used in the future due to identified plant contamination, the

requirements of 10 CFR 70.22(i) would need to be assessed and a dose

evaluation performed or an emergency plan submitted if required to

ensure the analyzed accident is appropriately addressed and mitigated.

Any such special nuclear material would be contained in the remaining

plant contamination, since fuel and blanket material were shipped

offsite during 1973-1975. Therefore, this amendment does not involve a

significant increase in the probability or consequences of an accident.

(2) Will the proposed amendment create the possibility of a new or

different kind of accident from any accident previously analyzed?

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

different type of accident from any previously evaluated. Allowing

possession of small amounts of special nuclear material does not change

methods of monitoring the facility or operations or surveillance of any

systems at Fermi 1. The amount requested is below that requiring

criticality monitoring per 10 CFR 70.24, and the separation of the

special nuclear material will not be permitted. Thus, there is no

identified physical mechanism for creating an accident based on the

existence of such material in the quantities specified. If the

provision allowing for possession of more than 15 grams of special

nuclear material or 2 curies of plutonium if is identified in plant

contamination in the future were to be invoked, applicable provisions

to ensure public safety per 10 CFR Part 70, Part 73, and Part 74 will

apply. For these reasons, allowing Detroit Edison to possess very

limited amounts of special nuclear material at Fermi 1 will not create

the possibility of a new or different type of accident.

[[Page 56241]]

(3) Will the proposed change significantly reduce the margin of

safety at the facility?

The proposed changes do not involve a significant reduction in the

margin of safety at Fermi 1. No changes to any systems, or the status

of any systems or structures, are created by this amendment. Being able

to have a very limited amount of special nuclear material at Fermi 1

will not significantly reduce the margin of safety because a 10 CFR

Part 20 program is already in place, and the amount of special nuclear

material is being limited below criteria requiring an emergency plan,

special nuclear material control program, or criticality monitoring. If

more than 15 grams of special nuclear material or 2 curies of plutonium

is identified in plant contamination in the future, the proposed

license amendment will require the applicable portions of 10 CFR Part

70, Part 73, and Part 74 to apply for the amount identified. For these

reasons, this amendment will not significantly reduce the margin of

safety at Fermi 1.

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, 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, Esquire, Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Branch Chief: John W.N. Hickey.

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

No. 2, Pope County, Arkansas

Date of amendment request: April 30, 1998.

Description of amendment request: Arkansas Nuclear One--Unit 2

(ANO-2) Technical Specification (TS) 4.8.1.1.2.c.3 has been revised to

relocate the specific value for the single largest post-accident load

to the Bases associated with TS 4.8. The revised TS 4.8.1.1.2.c.3 would

require the licensee to verify the generator capability to reject a

load greater than or equal to its associated single largest post-

accident load.

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.

The Diesel Generators (DGs) are not identified as the initiator of

any accident previously analyzed. The design and function of the DGs

are unaffected by this proposed change. Applying more restrictive

acceptance criterion to the single largest load rejection test can not

result in an increase in the probability of accidents previously

evaluated and will provide increased assurance that the DGs will

perform as intended to support the mitigation of accidents previously

evaluated.

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 proposed change corrects information contained in the technical

specification and does not involve any design change, plant

modification, change in analyzed DG performance, or change in plant

operation. Since the DGs are not considered to be event initiators,

their accident mitigation function is unaffected, and normal operation

is unaffected, the proposed change does not result in new or different

accidents from those previously analyzed.

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.

The design and function of the DGs are unaffected by the proposed

change. Applying more restrictive acceptance criterion to the single

largest load rejection test will provide increased assurance that the

DGs will perform as intended to support the mitigation of postulated

accidents. DG performance is proposed to meet a more stringent

standard.

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: John N. Hannon.

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

No. 2, Pope County, Arkansas

Date of amendment request: May 18, 1998.

Description of amendment request: The proposed changes delete the

ANO-2 TS 3.6.2.2 and 4.6.2.2 requirements, and their associated bases,

for the sodium hydroxide addition system and add new limiting

conditions for operation, action statements, surveillance requirements,

and bases information for trisodium phosphate baskets which will be

installed during the next ANO-2 refueling outage (2R13). The capability

to add sodium hydroxide to the containment spray system during the

initial phase of a loss-of-coolant accident will be replaced with

crystalline trisodium phosphate (TSP) dodecahydrate stored in

containers located on the floor of the containment building.

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.

The proposed change modifies the method of containment spray sump

pH control. The containment spray function is important for containment

heat removal/pressure mitigation. However, this change does not affect

the probability of occurrence of the accident initiators which result

in the need for containment heat removal and pressure mitigation. Since

the TSP baskets are seismically mounted passive devices located inside

the containment, they cannot initiate a transient or affect the

probability of occurrence of any previously analyzed accident.

The proposed change only modifies the chemical composition of the

containment spray and sump fluid. The proposed changes do not affect

the heat removal/pressure mitigation functions of the system since the

spray flow rate and droplet size are unchanged. The proposed change

also will not adversely affect the radiological doses for the design

basis accident (DBA) loss-of-coolant accident (LOCA) at the exclusion

area boundary, low

[[Page 56242]]

population zone, control room, or emergency response facility. The

change does not adversely affect the calculated peak clad temperature

for the DBA LOCA or the environmental qualification (EQ) of components

located inside containment.

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 proposed change allows the use of TSP as a buffering agent for

the containment sump instead of sodium hydroxide (NaOH) added via the

containment spray system. The TSP baskets are passive devices that have

minimal impact on any other system except through water chemistry. The

change in water chemistry does not adversely affect any safety system

or required safety functions. The replacement of NaOH additive with TSP

will not change the probability of a malfunction of safety-related

equipment.

Potential malfunctions relating to the proposed modification have

been evaluated for their effect on plant safety and have been found to

be non-significant. Additionally, the transient pH behavior of the

containment spray flow does not adversely affect the EQ of components

located inside containment.

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.

The proposed change does not adversely affect the ability of the

containment spray system to perform the functions of containment heat

removal, pressure mitigation, and fission product (iodine) retention.

The proposed change does not adversely affect any equipment credited in

the safety analysis. Also, the proposed change does not increase the

peak clad temperature or the offsite doses due to the DBA LOCA.

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: John N. Hannon.

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

No. 2, Pope County, Arkansas

Date of amendment request: June 29, 1998.

Description of amendment request: The proposed amendment would

revise the as-found lift setting tolerance for the ANO-2 main steam

safety valves (MSSVs) and pressurizer safety valves (PSVs) will be

increased. The proposed increase in the lift setting tolerance is

contingent upon a reduction in a linear power level-high setpoint and

use of the latest small break loss of coolant accident (SBLOCA)

methodology for development of the Core Operating Limits Report (COLR).

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 allows for a larger 3% tolerance versus

1%, -3% as-found lift setting tolerance. The proposed

change does not involve any change to the physical characteristics of

the main steam safety valves (MSSVs) and pressurizer safety valves

(PSVs), and will have no impact on the as-left settings. During

testing, the MSSVs and PSVs will continue to adjusted to 1%

of the Technical Specification (TS) lift setting.

The impact on the Safety Analysis Report (SAR) analyses when the

as-found lift setting tolerances are increased has been evaluated and

the effects upon the impacted events have been found to be within

acceptable limits, providing the allowable linear power level with

three inoperable MSSVs is revised from 45% to 36%, and that the latest

NRC approved C-E small break loss of coolant analysis (LOCA) evaluation

model, CENPD-137, Supplement 2-P-A, is included as a methodology for

determination of operating parameters identified within the core

operating limits report (COLR). With these concurrent changes, plant

systems required for safe operation and shutdown will continue to be

available to fulfill their safety function as described in the SAR.

Steam production in excess of relief capacity is precluded by the

physical design of the plant and operation of the reactor protection

system. Revision of the MSSV as-found lift setting tolerance from

1%, 3% to 3% does not alter safety

analyses conclusions.

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.

This change does not create any new plant configuration or

operational mode. This proposal to increase the MSSV and PSV as-found

lift setting tolerance does not modify equipment or change the manner

in which the MSSVs and PSVs will be operated. ASME design requirements

for maintaining system operating pressure limits below the maximum

design pressure of 1210 psia for plant secondary systems, and 2750 psia

for the reactor coolant system (RCS) are not impacted. The reduction in

allowable linear power level when three MSSVs are inoperable assures

plant operation within current analysis assumptions. The addition of

topical report CENPD-137, Supplement 2-P-A, as a reference to develop

the COLR is bounded by assumptions within the existing safety analysis.

The cycle specific COLR analyses will continue to be performed

utilizing NRC approved methodologies. The TS changes do not require any

new equipment be included in the design basis, and current equipment

will continue to be operated in a manner consistent with its design.

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.

The upper tolerance limit for design pressure is not affected by

this change. During the most severe anticipated operational transient,

the Secondary System pressure and RCS pressure will not exceed 110% of

design pressure. The MSSV and PSV lift settings will continue to be set

within -1% of the TS lift setting during surveillance testing.

The decrease in the peak cladding temperature of the reactor fuel,

due to a change in the methodology for analysis, does not significantly

impact previous analytical results. The current and previous analytical

methodologies are approved by the Staff.

The impact of the proposed changes on the ANO-2 SAR analyses have

been evaluated. The evaluation demonstrates that the results of the

impacted events

[[Page 56243]]

remained within the acceptable limits providing the maximum linear

power level percentage for three inoperable MSSVs is reduced. This

reduction in maximum allowable linear power level assures that adequate

steam relief capacity will be available to prevent overpressurizing the

secondary steam system during the most severe anticipated operational

transient.

Addition of topical report CENPD-137, Supplement 2-P-A, will not

reduce the existing TS operability and surveillance requirements. The

cycle specific COLR limits for future reloads will continue to be

developed based on NRC-approved methodologies. The ANO-2 TSs will

continue to require that the core be operated within these limits.

The cumulative impact of all of the proposed changes and the

results of the impacted events have been found to be within acceptable

limits. The system capabilities to mitigate and/or prevent accidents

will be the same as they were prior to these changes.

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: John N. Hannon.

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

No. 2, Pope County, Arkansas

Date of amendment request: June 29, 1998.

Description of amendment request: These proposed changes are in

Technical Specification 3.4.2, ``Reactor Coolant System--Safety

Valves--Shutdown,'' and Technical Specification 3.4.12, ``Reactor

Coolant System--Overpressure Protection'' regarding the low temperature

overpressure protection system. The specific changes include modifying

the requirements for the pressurizer code safety valve requirements

specified by Technical Specification 3.4.2 and a modification of the

safety injection tank isolation requirements specified in Technical

Specification 3.4.12.

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.

The reactor coolant system (RCS) is designed with overpressure

protection devices to be used in all modes of operation. The changes to

Technical specification (TS) 3.4.2 will ensure that, if no pressurizer

code safety valves are operable, the RCS will be cooled down to the

mode of applicability of the low temperature overpressure protection

(LTOP) system (TS 3.4.12) within 12 hours. The LTOP relief valves

provide sufficient relief capacity to protect the RCS from

overpressurization when the RCS inlet temperature (Tc) less

than or equal to 220 deg. F. Therefore, this change will ensure the

proper actions will be taken that will ensure adequate overpressure

protection of the RCS. These actions are not accident initiators, and

therefore do not involve a significant increase in the probability of

any accident previously evaluated.

The proposed change to TS 3.4.12 provides additional operational

flexibility for the use of the safety injection tanks (SITs) as an

additional inventory source during Modes 4, 5, and 6 when the RCS is in

LTOP conditions. The ability to use the SITs, with a pressure less than

300 psig is within the existing LTOP analysis. The LTOP analysis

ensures that under the analyzed worst case overpressurization event,

the RCS is protected. The 300 psig SIT pressure limit, corrected for

instrument uncertainty, will prevent a challenge to the LTOP relief

valves and therefore the RCS will be assured of overpressure

protection. The SIT pressure limit will also be low enough to prevent

an inadvertent isolation of the shutdown cooling system and thus

prevent a loss of shutdown cooling due to placing an SIT in service.

The remaining changes included in this amendment request are considered

administrative in nature and are therefore considered acceptable.

Based on the above discussions, these changes do 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 proposed changes included in this amendment request provide

additional operational flexibility for the use of the SITs and specify

the proper actions to be taken that will ensure adequate overpressure

protection of the RCS. The LTOP relief valves have already been

evaluated for operation below 220 deg. F. The changes do not introduce

any new plant configurations. No new accident possibilities are being

introduced by these changes. Therefore, the proposed changes do 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.

The proposed change to the TS 3.4.2 action statement requires the

Tc be less than or equal to 220 deg. F when no pressurizer

code safety valves are available. When Tc is less than or equal to

220 deg. F, the LTOP system operability is required by TS 3.4.12. This

action will provide assurance that the RCS will be protected from an

overpressurization event and therefore increases the margin of safety.

The requirements to maintain one pressurizer code safety valve in

Mode 4 when Tc is less than or equal to 220 deg. F and in

Mode 5 has been removed by the proposed revision to TS 3.4.2. The LTOPs

provide adequate RCS over pressure protection during these modes

without reliance on the pressurizer code safeties. Maintaining the

requirement to require one pressurizer code safety to be operable at

the same time as the LTOP system is required to be operable, provides

no additional plant safety. An operable LTOP system prevents RCS

pressure from increasing high enough to challenge the pressurizer code

safety lift setpoints.

The current TS 3.4.12 LTOP limits are based on an analysis that

uses the methodology outlined in the ASME Code Case N-514. This code

case defines the margin of safety for the current LTOP limits. This

code case was utilized in the development of TS 3.4.12. The safety

factor utilized by the code case provides a reasonable vessel

overpressure allowance for conditions expected during a low temperature

transient. The margin of safety is not reduced with SITs in service and

pressurized to less than 300 psig because this condition is bounded by

the existing LTOP analysis. 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.

[[Page 56244]]

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: John N. Hannon.

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

No. 2, Pope County, Arkansas

Date of amendment request: June 29, 1998.

Description of amendment request: The proposed change to the

Arkansas Nuclear One Unit 2 Technical Specifications would provide a

range of acceptable values for the 4160 Volt bus loss of voltage

values. The present Technical Specification Table 3.3-4, item 7.a

provides a single value for both the trip and the allowable values for

the 4160 Volt bus loss of voltage requirements. These table entries do

not include an acceptable range or an explicit indication of the

allowed tolerance that the actual setting is allowed to vary from the

indicated value. The proposed change replaces the specific trip value

with an explicit range of acceptable allowable values.

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.

The two 4160 Volt (V) vital bus loss of voltage protection relays

that are provided on each of the 4160 V safety buses are provided to

detect loss of voltage, isolate the safety buses, initiate load

shedding, and start the associated emergency diesel generator. This

safety function is unchanged by the proposed setpoint revisions. The

revised settings for the loss of voltage protection relays will

continue to provide the safety function with no appreciable additional

time delay. The proposed time delays are within those assumed in the

ANO-2 safety analyses. Additionally, the lower voltage settings will

prevent unnecessary isolations from the off-site power sources which

will contribute to reducing the probability of a loss of off-site power

due to off-site power system transients.

The ANO-2 technical specifications will continue to require the

4160 V loss of voltage functions to be surveillance tested at their

present frequency without changing the modes in which the surveillance

is required or the modes of applicability for these components. The

technical specifications will continue to require the same actions as

currently exist for the inoperability of one or more of the 4160 V loss

of voltage channels. 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 proposed change introduces no new modes of plant operation or

new plant configuration. The 4160 V vital bus loss of voltage

protection relays are required to operate following a complete loss of

off-site power to initiate the bus power source transfer to on-site

power, i.e., the emergency diesel generators, to prevent a loss of all

AC power. This safety function is unchanged by the proposed setpoint

revisions, and the proposed setpoints continue to provide the required

actions consistent with the ANO-2 safety analysis. 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.

The two undervoltage relays located on each 4160 V safety bus are

provided to detect loss of voltage, isolate the safety buses, initiate

load shedding, and start the emergency diesel generators. This safety

function is unchanged by the proposed setpoint revisions.

The lower loss of voltage values do not affect the safety function

since there is no appreciable time difference in reaching the lower

setpoints during a loss of voltage event. The maximum proposed time

delay setting with the minimum loss of voltage relay setting is within

those used in the ANO-2 safety analysis. The revised settings for the

relays will continue to provide the safety function with no appreciable

additional time delay.

Removal of the trip value from the technical specifications is

consistent with that which is presented in NUREG-1432, ``Standard

Technical Specifications for Combustion Engineering Plants.'' The

current ANO-2 technical specifications and NUREG-1432 both indicate

that if the setpoint is outside the allowable value column, the

associated channel is declared inoperable. This approach is consistent

with this proposed technical specification change.

The trip and allowable values listed in the technical

specifications for the loss of voltage protection for the 4160 V buses

are presently the same. With these values being the same, if the trip

value is exceeded, the allowable value will also be exceeded. This

change provides a range of acceptable allowable values for these

relays. By relocating the trip values in the surveillance test

procedures, the procedural limits for the voltage and time delay

settings can be adjusted to ensure margin to the allowable values.

Additionally, the lower voltage settings will help to prevent

unnecessary isolation from the off-site power sources due to off-site

perturbations in the electrical grid, and thus contribute to increasing

the 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: John N. Hannon.

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

No. 2, Pope County, Arkansas

Date of amendment request: June 29, 1998.

Description of amendment request: The proposed Technical

Specification change revises the surveillance testing requirements for

the Arkansas Nuclear One--Unit 2 (ANO-2) direct current (DC) electrical

distribution system. ANO-2 is planning on modifying the 120 volt vital

alternating current (AC) electrical distribution system by installing

new inverters during the next scheduled refueling outage (2R13). This

modification will increase the normal 125 volt vital DC system loads by

adding the inverters as a normal load. The power for each 125 volt

vital DC system is normally supplied by its associated battery charger.

ANO-2 is in the process of replacing the vital DC battery chargers by

plant modification to ensure all the battery chargers are of sufficient

capacity to provide the necessary current requirements for the normal

125 volt vital DC loads. The proposed change to specification

4.8.2.3.c.4 is required to ensure the new chargers are adequately

tested to support the associated inverter replacement.

[[Page 56245]]

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.

Technical Specification (TS) surveillance requirement (SR)

4.8.2.3.b.2 requires the battery banks for each of the vital 125 volt

direct current (DC) systems to be inspected to ensure that no visible

corrosion exists at the terminals or the connectors. This SR has been

modified to allow the present corrosion inspection, or the measurement

of the resistance of the associated battery connections. The resistance

measurement provides an indication of physical damage or abnormal

deterioration that could potentially degrade battery performance and

has been an accepted alternative to the visual inspection requirement.

The Bases change associated with TS 3.8.2.3 Action ``b'' is

considered administrative in nature and simply clarifies the intent of

the action without changing the requirements of the action or its

required completion time. The station batteries are not classified as

accident initiators in the ANO-2 accident analysis. The 125 volt class

1E batteries are credited for accident mitigation in the accident

analysis. The above described changes do not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

Each battery charger is required to have sufficient capacity to

restore the battery from the design minimum charge to its fully charged

state while supplying normal steady state loads. The minimum specified

TS surveillance required charger amperage limit will ensure this

capacity. The additional charger output is presently accounted for in

the emergency diesel generator loading tables in the Safety Analysis

Report (SAR). Loss of one train of the vital 125 volt DC system is an

accident that has been evaluated in the SAR. The capacity of the

battery chargers is not a factor in the probability of this accident

occurring. Therefore, the changes associated with this technical

specification amendment request do not increase the probability of any

accident previously evaluated.

The proposed technical specification changes do not modify the

limiting condition for operation or the associated action statements

regarding operability of the battery chargers other than clarifying

these requirements. The frequency at which the battery charger

operability is demonstrated by surveillance testing is not being

modified by this technical specification change request. The proposed

battery charger surveillance testing acceptance criterion will more

appropriately demonstrate the capability of this equipment. This change

does not affect the consequences of any of the previously evaluated

accidents.

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.

Technical specification SR 4.8.2.3.b.2 requires the battery banks

for each of the 125 volt systems to be inspected to ensure that no

visible corrosion exists at the terminals or the connectors. This SR

has been modified to allow the present corrosion inspection, or to

perform resistance readings on the associated battery connections. The

visual inspection is required to detect corrosion of the battery

connections. The resistance measurement of the associated battery

connections provides an acceptable alternative to the visual inspection

requirement and provides an indication of physical damage or abnormal

deterioration that could potentially degrade battery performance.

The availability of an extra battery charger for each train

following the plant modification provides a more reliable configuration

without introduction of any new modes of plant operation. No new

accident possibilities are being introduced by the proposed change to

the surveillance testing specification for battery charger amperage.

Increasing the surveillance testing amperage limit for the battery

chargers does not create the potential for any different accident since

the new value remains within the design capacity of the components.

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.

TS SR 4.8.2.3.b.2 has been modified to allow resistance readings on

the associated battery connections or the performance of the present

visual inspection requirements. The resistance measurement of the

associated battery connections provides an acceptable alternative to

the visual inspection requirement and provides an indication of

physical damage or abnormal deterioration that could potentially

degrade battery performance without a significant reduction in the

margin of safety.

The proposed technical specification surveillance requirements for

the battery chargers continues to require testing of battery chargers

at the present duration and frequency. These requirements will also

apply to the second charger being installed for each Class 1E battery

train. Each of the new battery chargers has sufficient capacity to

restore the battery from the design minimum charge to its fully charged

state while supplying normal steady state loads. The proposed

surveillance specification change does not involve a significant

reduction in the margin to safety since the demonstrated capacity will

be of a higher amperage requirement than is demonstrated during the

surveillance test with the existing configuration. Increasing the

required amperage value assures the surveillance test will continue to

demonstrate the chargers can provide significantly more current than is

necessary to meet the design requirements. 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: John N. Hannon.

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

No. 2, Pope County, Arkansas

Date of amendment request: August 6, 1998.

Description of amendment request: The proposed technical

specification change revises the Action requirements for the Arkansas

Nuclear One--Unit 2 (ANO-2) Control Element Assembly (CEA) position

indicator channels. The Action requirements listed in Specification

3.1.3.2 are being modified consistent with the requirements of NUREG-

1432, ``Standard Technical Specifications for Combustion Engineering

Plants.'' The proposed changes also include the relocation of Technical

Specification Table 3.8-1, ``Containment Penetration Conductor

Overcurrent Protective Devices'' per

[[Page 56246]]

NRC Generic Letter 91-08, ``Removal of Component Lists From Technical

Specifications.''

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 technical specification (TS) change request contains the

relocation of Table 3.8-1, Containment Penetration Conductor

Overcurrent Protective Devices, and changes to the control element

assembly (CEA) position indication.

Generic Letter (GL) 91-08, ``Removal of Component Lists From

Technical Specifications,'' was issued as a TS line item improvement by

the NRC. Table 3.8-1 is one of the specific lists of components

contained in the GL. TS Table 3.8-1 and all its references have been

removed from Specification 3/4.8.2.5 in accordance with the GL. This

change is considered administrative in nature because the requirements

for operability, the limiting conditions for operation, the

surveillance requirements and their frequencies for the containment

penetration conductor overcurrent protective devices remains the same.

This amendment request fundamentally modifies the physical location of

the devices listed in Table 3.8-1 from the TS to the plant procedures.

These changes have no affect on the probability or consequences of any

accident previously evaluated.

The remaining changes included in this amendment request are those

relating to the CEA position indication. The Action requirements for TS

3.1.3.2 were modified to be consistent with the requirements of NUREG-

1432, ``Standard Technical Specifications for Combustion Engineering

Plants.'' The most recent revision of NUREG-1432 was used to produce

this change because it represents the latest guidance for the TS CEA

position indication requirements that are applicable to ANO-2 and

acceptable to the NRC.

The requirement was removed from TS 3.1.3.2 that restricted each

CEA group to a maximum of one CEA with less than two of the required

position indicator channels. NUREG-1432 places no requirements on the

number of CEAs in a group with less than two of the required position

indicator channels. NUREG-1432 would allow all the CEAs in a group to

have only one of the required CEA position indications operable. In

this situation, the associated CEAs with less than two of the required

position indicator channels would have to be placed at their ``Full

In'' or ``Full Out'' limits.

TS 3.1.3.2 was modified to allow the use of the ``Full In'' or

``Full Out'' limits which ensures this specification is consistent with

its bases and NUREG-1432. The TS will still maintain the requirements

for two independent means of determining CEA position with this

amendment request. With two independent means of determining CEA

position, reliable determination of actual CEA position will be

maintained.

Additionally, NUREG-1432 does not require the placement of any

other CEAs in the associated group at the ``Full Out'' limit when one

of the CEAs in the group has only one of the required position

indication systems operable. All of the remaining CEAs in the

associated group still have at least two independent means of CEA

position indication or they would already be required to be positioned

to the ``Full Out'' limit to restore the second position indication.

The TS retains the requirements for the individual and group CEA

alignment in accordance with Specifications 3.1.3.1 and 3.1.3.6. These

requirements also eliminate the need for pulling the remaining CEAs in

the group to the ``Full Out'' limit as long as the alignment

requirements are maintained.

These changes will allow the operator more time to focus on the

individual CEA position indication problem rather than moving the

remainder of the CEAs in the group unnecessarily. Anytime that a CEA is

moved, a small probability exists for it to slip or drop into the core.

If this were to occur while attempting to align the group to the ``Full

Out'' limit, a reactor transient would be initiated. Additionally,

anytime the CEAs are operated, a small probability of an error exists.

Removing the unnecessary requirement for the group withdrawal could

decrease the probability of CEA misoperation. CEA position indication

is not considered as an accident initiator. Retaining the requirements

to maintain at least two independent means of determining CEA position

will ensure the consequences of all the accidents previously evaluated

remain unchanged.

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 portions of this change that are made in accordance with GL 91-

08 are considered administrative in nature and do not result in the

creation of a new or different kind of accident from any previously

evaluated.

The bases for TS 3.1.3.2 state that the action statements

applicable to inoperable CEA position indicators permit continued

operation when the positions of CEAs with inoperable position

indicators can be verified by the ``Full In'' or ``Full Out'' limits.

Although TS 3.1.3.2 may have originally been intended to allow

continued operation using the ``Full In'' limits, it has never been

clearly addressed in the specification. NUREG-1432 allows the use of

both the ``Full In'' or ``Full Out'' limits. This amendment request

will not change the methods for CEA operation, although it will reduce

unnecessary CEA manipulations due to CEA position indication problems.

The requirements of Specification 3.1.3.1 will ensure that an

individual CEA is maintained in proper alignment with the remaining

CEAs in the group. Specification 3.1.3.6 will ensure the CEA groups are

maintained within the proper withdrawal sequence and insertion limits.

Specification 3.1.3.5 will ensure the shutdown CEA groups are

maintained in the ``Full Out'' position. The CEA position indication

changes allowed by this amendment request, including the allowance to

use the ``Full In'' limits, can produce a CEA configuration that is

different from that allowed by the current TSs. However, the allowed

configurations will be bounded by the TS 3.1.3.2 Action ``c''

requirements for compliance with Specifications 3.1.3.1, 3.1.3.5, and

3.1.3.6. Therefore, the action requirements of TS 3.1.3.2 will ensure

the CEAs are operated consistent with the safety analysis assumptions.

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.

The portions of this change that are made in accordance with GL 91-

08 are considered administrative in nature and have no effect on the

margin of safety. The remaining changes can result in a lower

probability of CEA misoperation and reduce the potential of plant

transients due to CEAs that slip or drop into the core while performing

unnecessary group realignments. These changes can also reduce

unnecessary plant shutdowns, due to unneeded restrictions on CEA

position indication. An unnecessary plant shutdown produces an

opportunity for plant

[[Page 56247]]

upsets that can be avoided by this change. The proposed TS provide an

equivalent level of safety as those specifications that currently

exist. 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: John N. Hannon.

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

No. 2, Pope County, Arkansas

Date of amendment request: September 17, 1998.

Description of amendment request: The proposed amendment addresses

a problem associated with the existing technical specifications being

inconsistent with the design of the plant protection system (PPS). The

PPS uses a design in which a single bistable is used to automatically

enable the selected core protection calculator (CPC) trip functions

whenever a permissive exists to bypass the high logarithmic power level

trip function. The technical specifications allow the bypass of the

high logarithmic power trip when power is above 10-4 percent

power and allow bypasses of the affected CPC trips when power is below

10-4 percent power. The proposed technical specification

change establishes a range for the bistable setpoint to be within such

that it is possible to meet both of its design functions while also

meeting the technical specification requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

This technical specification (TS) change request modifies the power

level at which two of the three operating bypasses can be set to

operate. This change is necessary because the present plant bistable

design requires a range for this bistable to operate within rather than

a specific setpoint as required by the present TS. The single bistable

associated with these operating bypasses is designed with an inherent

hysteresis loop and therefore requires an operating range. The band of

10-4% to 10-2% of rated thermal power provides

the bistable an adequate operating range to account for the inherent

bistable hysteresis, allow for bistable drift, and provides margin for

the applicable uncertainties. Regardless of the actual bistable

setpoint within this band, the bistable design ensures that either the

high logarithmic power level or the core protection calculator (CPC)

generated trips are available to provide reactor trip protection. The

CPC and logarithmic power operating bypasses and their setpoints are

not considered credible accident initiators and therefore modifying

their setpoints does not involve a significant increase in the

probability of an accident previously evaluated.

The automatic removal function of these operating bypasses is

designed to mitigate the consequences of accidents. As described within

the background section of the TS change request, the safety analyses

associated with operating bypasses have been reviewed for the

acceptability of these changes. This review concluded that these

changes are considered bounded by the existing safety analyses. Since

these TS changes are bounded within the present safety analyses, they

do not involve a significant increase in the consequences of an

accident previously evaluated.

The remaining changes included in this TS change request are being

made to clarify the existing requirements for the operating bypasses

and to establish consistency with the above described changes. The

remaining changes have been found acceptable because they are

considered administrative in nature and have no effect on the

probability or consequences of an accident previously evaluated.

Therefore, this change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

2. Does not create the possibility of a new or different kind of

accident from any previously evaluated.

There are no physical plant modifications being made to the plant

as a result of this change. The only function that is required by the

TS and modified by this change is associated with the allowed setpoint

for the automatic bypass removal feature of the CPCs. This feature will

still be required by the TS, but will be allowed a slightly higher

setpoint. The system connections and the reactor trip setpoints are not

affected by this change. The CPC and logarithmic power operating

bypasses and their setpoints are not considered as credible accident

initiators. 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.

The safety analyses associated with these operating bypasses have

been reviewed for the acceptability of these changes. This review

concluded that the changes associated with this TS change request are

considered bounded within the existing safety analyses. The associated

safety analyses have been considered to be acceptable because they have

produced acceptable results and thus provide an acceptable margin to

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: John N. Hannon.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 29, 1998.

Description of amendment request: The proposed changes modify

Technical Specification (TS) 3.7.6.1 (Control Room Emergency Air

Filtration System--Modes 1-4), TS 3.7.6.2 (Control Room Emergency Air

Filtration System--Modes 5 and 6), TS 3.7.6.3 (Control Room Air

Temperature--Modes 1-4), TS 3.7.6.4 (Control Room Air Temperature--

Modes 5 & 6), and TS 3.7.6.5 (Control Room Isolation and

Pressurization), and the associated Bases.

The proposed changes to the control room ventilation TS affects the

Applicability and the Actions. These changes will make the TS

consistent with NUREG-1432 (Standard Technical Specifications

Combustion Engineering

[[Page 56248]]

Plants), as applicable, and the accident analysis. The proposed changes

to the TS Bases make the Bases consistent with the TS and also clarify

that suspending movement of irradiated fuel assemblies shall not

preclude movement to a safe conservative position.

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.

The proposed changes revise the control room ventilation Technical

Specifications (TS) Actions to delete the Action statement to suspend

all operations involving positive reactivity changes, and adds an

Applicability and Action related to the movement of irradiated fuel

assemblies. The changes also add an Applicability footnote and revise

the Bases to allow irradiated fuel assemblies to be placed in a safe

conservative position when movement is required to be suspended. Other

changes to the Bases are being made to be consistent with the TS. These

changes do not affect the probability of an accident. The control room

ventilation systems (ventilation, temperature, or envelope) do not

affect the initiators of an accident; therefore, the changes do not

alter the initiators of any analyzed events.

The administrative and more restrictive changes do not affect the

consequences of an accident. The administrative changes add an

Applicability footnote and revise the TS Bases to make them consistent

with the TS. This will ensure the applicable control room ventilation

system TS are entered during movement of irradiated fuel assemblies and

that there is no confusion associated with the Bases being

inconsistent. The more restrictive change of adding the Applicability

during movement of irradiated fuel assemblies and the Action to suspend

movement of irradiated fuel assemblies eliminates the precursor to the

fuel handling accident which prevents the fuel handling accident from

occurring when the control room ventilation systems are inoperable. The

addition of this Action ensures the event that may release

radioactivity is precluded when the control room ventilation systems

are inoperable.

The less restrictive changes (deleting the requirement to suspend

positive reactivity changes and a Bases change which allows irradiated

fuel assemblies to be placed in a safe conservative position when

movement has been suspended) do not affect the consequences of an

accident because no accident mitigator is affected. The safety analysis

credits instrumentation to detect a boron dilution accident and alert

the control room staff. After the control room staff is alerted, the

accident is terminated without a radioactive consequence. These

instruments are required to be Operable and if one is inoperable,

positive reactivity changes are required to be suspended. If both

instruments become inoperable, along with suspension of positive

reactivity additions, boron concentration is required to be determined

at frequencies specified in the Core Operating Limits Report (only when

source range neutron flux monitors are inoperable). Also, the shutdown

margin (SDM) is required to be met. If the SDM requirements are not

met, action must be taken to borate (addition of negative reactivity)

until the SDM is restored. Therefore, if the control room ventilation

systems are inoperable, suspension of positive reactivity changes are

not required. The added statement in the Bases allows irradiated fuel

assemblies to be placed in a safe conservative position to preclude a

fuel handling accident from occurring. These Actions ensure that

appropriate measures are taken to preclude events that would require

the control room to be isolated when any of the control room

ventilation systems are inoperable.

Therefore, the proposed changes will not involve a significant

increase in the probability or consequences of any 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 changes revise the control room ventilation TS Actions

to delete the Action statement to suspend all operations involving

positive reactivity changes, and adds an Applicability and Action

related to the movement of irradiated fuel assemblies. The changes also

add an Applicability footnote and revise the Bases to allow irradiated

fuel assemblies to be placed in a safe conservative position when

movement is required to be suspended. Other changes to the Bases are

being made to be consistent with the TS. These changes do not alter the

design or configuration of the plant. There has been no physical change

to plant systems, structures, or components. The proposed changes will

not reduce the ability of any of the safety-related equipment required

to mitigate Anticipated Operational Occurrences (AOOs) or accidents.

Therefore, the proposed changes will not create the possibility of a

new or different kind 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 changes revise the control room TS Actions to delete

the Action statement to suspend all operations involving positive

reactivity changes, and adds an Applicability and Action related to the

movement of irradiated fuel assemblies. The changes also add an

Applicability footnote and revise the Bases to allow irradiated fuel

assemblies to be placed in a safe conservative position when movement

is required to be suspended. Other changes to the Bases are being made

to be consistent with the TS. The margin of safety is not affected

because the proposed changes to delete one Action and add an

Applicability and Action ensures the assumptions of the accident

analysis are being met. The administrative changes ensure the

applicable TS are entered and eliminate confusion associated with the

discrepancies between the TS and Bases. The more restrictive changes of

adding an Applicability and Action eliminates the precursor to an event

(fuel handling accident) that may release radioactivity when the

control room ventilation systems are inoperable. The less restrictive

changes revises the TS to rely on the instrumentation credited in the

accident analysis and to allow irradiated fuel assemblies to be placed

in a safe position to preclude a fuel handling accident. The

instruments are required to be operable per TS. Compliance with these

TS and also the SDM TS ensures that boron dilution event is precluded

or can be mitigated. Therefore, suspension of positive reactivity

changes is not required when the control room ventilation systems are

inoperable. These Actions ensure that appropriate measures are taken to

preclude events that would require the control room to be isolated when

any of the control room ventilation systems are inoperable. Therefore,

the proposed change will not involve a significant reduction in a

margin of safety.

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

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

[[Page 56249]]

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: John N. Hannon.

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

Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana

Date of amendment request: August 12, 1998

Description of amendment request: The proposed amendment will

change Technical Specifications (TS) 3.1.2.8, 3.5.1, 3.5.4, Figure 3.1-

1, and Bases 3/4.5.2 for Waterford 3. It increases the maximum boron

concentration in the Safety Injection Tanks (SITs) and the Refueling

Water Storage Pool (RWSP) from 2300 ppm to 2900 ppm.

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 consequence

of any accident previously evaluated?

Response: No.

The proposed change increases the maximum boron concentration in

the SITs and the RWSP from 2300 ppm to 2900 ppm. This change does not

affect the probability of any accident. This increase in boron

concentration affects the pH of water in the safety injection sump

during a LOCA [Loss of Coolant Accident] and the potential for boron

precipitation. The amount of TSP in containment is adequate to maintain

the pH above 7.0. The revised long term cooling analysis shows that

boron precipitation will not occur at the higher boron concentrations.

Therefore, this change will not adversely impact post-LOCA core

cooling. Thus, the consequences of a LOCA are not affected.

Therefore, the proposed change will not involve a significant

increase in the probability or consequence of any accident previously

evaluated.

2. Will operation of the facility in accordance with this proposed

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

from any accident previously evaluated?

Response: No.

The proposed change will not create any new system connection or

interactions. Thus, no new modes of failure are introduced. There is no

significant impact on the corrosion rate in the safety injection system

due to the slightly higher acidic solution with the higher boron

concentration.

Therefore, the proposed change will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. Will operation of the facility in accordance with this proposed

change involve a significant reduction in margin of safety?

Response: No.

Sufficient TSP [Trisodium Phosphate Dodecahydrate] is provided in

the containment to ensure that the pH of the safety injection sump

water during a LOCA remains above 7.0 as stated in the Technical

Specification bases. Adequate time and HPSI [High Pressure Safety

Injection] flow exist to avoid boron precipitation during a LOCA. The

higher boron concentration limit will also allow higher refueling boron

concentrations which will increase the available shutdown margin.

Therefore, the proposed change does not involve a significant

reduction in 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: John N. Hannon

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

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

Date of amendment request: August 31, 1998.

Description of amendment request: The proposed amendment would

revise Improved Technical Specification (ITS) 5.6.2.10, ``Steam

Generator (OTSG [once-through steam generator]) Tube Surveillance

Program,'' to include a new repair process, called a ``repair roll'' or

``re-roll.'' The process would be used to repair steam generator tubes

with defects within the upper tubesheet. Changes to inservice

inspection and reporting requirements are proposed for tubes which are

repaired using this process. The proposed revision would also require

inspection of both OTSGs during each inservice inspection. In addition,

several format and editorial changes are proposed to ITS 5.6.2.10 and

to ITS 5.7.2, ``Special Reports,'' for clarification purposes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed LAR [license amendment request] addresses several

editorial and format changes which do not impact accident analyses. LAR

#235 also proposes to implement the repair roll (re-roll) process.

The qualification of the re-roll joint is based on establishing a

mechanical roll length which will carry all structural loads imposed on

the tubes with required margins. A series of tests and analyses were

performed to establish this length. Tests that were performed included

leak, tensile, fatigue, ultimate load and eddy current measurement

uncertainty. The analyses evaluated plant operating and faulted loads

in addition to tubesheet bow effects. Any tube leakage will be bounded

by the main steam line break (MSLB) evaluation presented in the Final

Safety Analysis Report (FSAR). The proposed change also requires

inspections of the joints created by the repair roll process. The

addition of this inspection does not change any accident initiators.

The proposed inspections after re-roll installation, and during future

inservice inspections, assure continuous monitoring of these tubes such

that inservice degradation of tubes repaired by the re-roll process

will be detected. Based on the Framatome Technologies qualification, as

well as the history for similar industry repair rolls, there are no new

safety issues, as defined in BAW-2303P, Revision 3, associated with the

repair roll. Therefore, this change does not involve a significant

increase in the probability or consequences of any accident previously

evaluated.

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

from any accident previously evaluated.

No new failure modes or accident scenarios are created by the re-

roll process. The new pressure boundary joint created by the repair

roll process

[[Page 56250]]

has been shown by testing and analysis to provide structural and

leakage integrity equivalent to the original design and construction

for all normal operating and accident conditions. Furthermore, the

testing and analysis demonstrate the repair roll process creates no new

adverse effects for the repaired tube and does not change the design or

operating characteristics of the OTSGs. In the unlikely event that a

tube with a repair roll should fail and sever completely at the

transition of the re-roll region, the tube would remain engaged in the

tubesheet bore, preventing interaction with other surrounding tubes. In

this case, leakage is bounded by the steam generator tube rupture

(SGTR) accident analysis. Therefore, this change does not create a

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

evaluated.

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

The repair roll process effectively removes the defective/degraded

area of the tube from service. The new roll expanded interface created

with the tubesheet satisfies all the necessary structural, leakage and

heat transfer requirements. The joint is constrained within the

tubesheet bore; thus, there is no additional risk associated with tube

rupture. The accident leakage is shown to be well within the initial

assumption of the MSLB analysis of one gallon per minute primary-to-

secondary leakage. Therefore, the FSAR analyzed accident scenarios

remain bounding, and the use of the repair roll process does not reduce

the margin of safety.

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

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

satisfied.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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

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

33733-4042.

NRC Project Director: Frederick J. Hebdon.

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

Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida

Date of amendment request: August 31, 1998.

Description of amendment request: The proposed amendment will

change the Improved Technical Specifications (ITS) to add three

additional Regulatory Guide (RG) 1.97 Type A Category 1 post-accident

monitoring (PAM) instrumentation variables and one Type B Category 1

PAM instrumentation variable to ITS Table 3.3.17-1, Post-Accident

Monitoring Instrumentation. The Type A Category 1 variables added are

low pressure injection (LPI) pump run status, LPI suction from reactor

building (RB) sump isolation valves DHV-42 and DHV-43 open position,

and high pressure injection (HPI) pump run status. The Type B Category

1 variable added is reactor coolant system (RCS) low range pressure.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

The addition of post-accident monitoring instrumentation to the CR-

3 ITS and ITS Bases is to ensure instrumentation is available for use

by the operators for performing manual actions, or to verify automatic

actions have occurred, which are required to mitigate the effects of a

design basis accident. The instrumentation is used for monitoring by

the operators only after an accident occurs, performs no automatic

functions, and there are no credible failures of this instrumentation

which could initiate any accident previously evaluated. Therefore, the

probability of occurrence of any accident previously evaluated is

unaffected.

The availability and use of this instrumentation ensures that the

prescribed manual operator actions for mitigating the consequences of

an accident will be implemented when necessary, and that the operator

has sufficient information to verify required automatic actions have

occurred when necessary. Therefore, the availability and use of the

instrumentation provides assurance that the consequences of accidents

will not be greater than that previously evaluated.

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

from previously evaluated accidents?

The addition of post-accident monitoring instrumentation to the CR-

3 ITS and ITS Bases is to ensure instrumentation is available for use

by the operators for performing manual actions, or to verify automatic

actions have occurred, which are required to mitigate the effects of a

design basis accident. The instrumentation is used for monitoring by

the operators only after an accident occurs, performs no automatic

functions, and there are no credible failures of this instrumentation

which could initiate a new or different kind of accident. Therefore,

the possibility of a new or different kind of accident occurring as a

result of this passive instrumentation is not created.

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

The addition of post-accident monitoring instrumentation to the CR-

3 ITS and ITS Bases is to ensure instrumentation is available for use

by the operators for performing manual actions, or to verify automatic

actions have occurred, which are required to mitigate the effects of a

design basis accident. The instrumentation is used for monitoring by

the operators only after an accident occurs, and performs no automatic

functions. The availability and use of this instrumentation ensures

that the prescribed manual operator actions for mitigating the

consequences of an accident will be implemented when necessary, and

that the operator has sufficient information to verify required

automatic actions have occurred when necessary. These required manual

and automatic actions are necessary to preserve the margin of safety as

defined in the CR-3 ITS and ITS Bases. The availability and use of this

instrumentation provides assurance that the existing margin of safety

will be maintained, and assumptions related to the margin of safety

during mitigation of design basis accidents will be preserved.

Therefore, the existing margin of safety will not be reduced.

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

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

satisfied.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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

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

33733-4042.

NRC Project Director: Frederick J. Hebdon.

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

Generating Station, Ocean County, New Jersey.

Date of amendment request: May 5, 1998.

[[Page 56251]]

Description of amendment request: This request is to change the

licensing basis to allow for a small amount of containment overpressure

to ensure sufficient net positive suction head for the Emergency Core

Cooling System pumps under post Loss of Cooling Accident (LOCA)

conditions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change to the licensing basis does not ``Involve a

significant increase in the probability or consequences of an accident

previously evaluated * * * ''. As the strainers have no function until

after the design basis LOCA occurs, the design of the strainer cannot

affect the probability of a Large Break LOCA.

The requested change to raise the assumed containment overpressure

for suction strainer design to 1.25 psig is less than that which is

already used in LOCA analyses for offsite releases. Therefore, this

change will not increase the offsite consequences of any previously

analyzed accident. The frequency of a design basis LOCA occurrence at

the Oyster Creek Nuclear Generating Station is conservatively estimated

at 5.67 x 10-4 per year. The frequency of a design basis

LOCA with a loss of containment overpressure is conservatively

estimated at 2.46 x 10-7 per year.

Since the frequency of the design basis LOCA coincident with a loss

of containment overpressure is insignificant (2.46 x

10-7), the requested increase does not significantly impact

the probability of exceeding the existing design bases. The core damage

frequency increase due to the request for overpressure is mitigated, in

part, by the current procedural requirement to flood containment

following the design basis LOCA, thereby obviating the need for over

pressure in the long term. The risk evaluation, performed in support of

the request for over pressure, indicated a non-risk significant change

in the core damage frequency.

The proposed change to the licensing bases does not ``Create the

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

previously evaluated * * *''. Both the new and existing strainers are

passive. They function solely to prevent debris from entering the

suction of the core and containment spray pumps. The only significant

difference is that the new strainers can remove more debris without

clogging. The slight amount of containment overpressure does not affect

the operation of the strainers, and improves the ability of the core

spray and containment spray systems to continue operation. Therefore,

no new or different kind of accident is created or possible.

The proposed change to the licensing bases does not ``Involve a

significant reduction in a margin of safety * * *.'' The modification

increases the amount of debris that can be removed while maintaining

core spray system operation. The requested change takes credit for 1.25

psig of wetwell overpressure. However, as the requested change is

bounded by existing calculations for offsite release, no significant

reduction in the margin of safety can occur. Additionally, as

demonstrated in Attachment III, the probability of a LOCA with a loss

of containment overpressure is not significant.

Guidance has been provided in ``Final Procedures and Standards on

No Significant Hazards Considerations,'' Final Rule, 51 FR 7744, for

the application of standards to license change requests for

determination of the existence of significant hazards considerations.

This document provided examples of amendments which are and are not

considered likely to involve significant hazards considerations.

Based on the above evaluation and the review of 51 FR 7744, this

proposed change to the licensing basis of the Oyster Creek Nuclear

Generating Station does not involve irreversible changes, a significant

relaxation of the criteria used to establish safety limits, a

significant relaxation of the bases for the limiting safety system

settings, or a significant relaxation of the bases for the limiting

conditions for operations. Therefore, based on the guidance provided in

the Federal Register and the criteria established in 10 CFR 50.92(c),

the proposed change does not constitute a significant hazard.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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

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

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

NRC Project Director: Cecil O. Thomas.

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

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

Date of amendment request: September 9, 1998.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TS) by: (1) Changing the TS

Definitions 1.24, ``Core Operating Limits Report,'' 1.27, ``Engineering

Safety Feature Response Time,'' and 1.31, ``Radiological Effluent

Monitoring and Offsite Dose Calculation Manual (REMODCM)''; (2)

changing TS 3.0.2, ``Limiting Condition For Operation,'' by adding a

new TS 3.0.6 to the Limiting Condition For Operation TS section; (3)

changing TS 4.0.5, ``Surveillance Requirements''; (4) changing the mode

applicability of TS 3.2.3, ``Total Unrodded Integrated Radial Peaking

Factor--FrT''; (5) changing TS 3.3.2.1,

``Engineered Safety Features Actuation System Instrumentation,'' by

modifying TS Table 4.3-2 Table Notation (1) which it references; (6)

changing TS 3.4.1.1, ``Reactor Coolant System--Coolant Loops and

Coolant Circulation Startup and Power Operation'; and (7) changing TS

3.4.11, ``Reactor Coolant System--Reactor Coolant System Vents.'' The

associated TS Bases sections would also be updated to reflect the

proposed changes. The proposed changes would resolve identified

compliance issues.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Technical Specification Definitions

The minor editorial and non-technical changes to correct reference,

spelling and terminology errors contained in the definitions will not

result in any technical changes to the Millstone Unit No. 2 Technical

Specifications. The proposed changes will have no adverse effect on

plant operation. Therefore, the proposed change will not result in a

significant increase in the probability or consequences of an accident

previously evaluated.

Technical Specification 3.0.6

The new Technical Specification, 3.0.6, will provide guidance on

returning inoperable equipment to service, under administrative

control, to demonstrate operability of that

[[Page 56252]]

equipment, or the operability or other equipment. Various Technical

Specification Actions require inoperable equipment to be removed from

service, such as maintaining a containment isolation valve closed or

tripping/bypassing a failed instrument channel. An exception to these

required actions is necessary to allow the performance of testing to

demonstrate the operability of the equipment being returned to service.

Specifically, this Technical Specification addresses the situation

where the inoperable equipment has been repaired, tested to the extent

possible, and believed to be capable of performing its function. At

this point, a presumption of the operability of the equipment is

reasonable, and is supported by experience. Therefore, it is acceptable

to place the equipment in service for testing under administrative

control. Administrative controls will be used to ensure the time the

equipment is returned to service is consistent with the Action

Statements and is limited to the time necessary to perform the

surveillance requirements.

This specification will also allow the inoperable equipment to be

placed in a condition different from that required by the action

statement to demonstrate the operability of other equipment. An example

would be during the performance of an operability test on one reactor

protection channel while another channel associated with the same

function is inoperable. In this situation only one of the channels

could be in the tripped condition, otherwise a reactor trip would be

initiated. This is already permitted for reactor protection channels by

Technical Specifications 3.3.1.1, ``Instrumentation--Reactor Protective

Instrumentation,'' Action 2, and for engineered safety features

channels by 3.3.2.1, ``Instrumentation--Engineered Safety Feature

Actuation System Instrumentation,'' Action 2.

This provision is provided only to perform surveillance

requirements to prove operability, and not to provide time to perform

any other preventive or corrective maintenance. The testing will be

performed consistent with the current Technical Specification Action

Statement and will be limited to the time necessary to perform the

surveillance requirement. The proposed changes will have no adverse

effect on plant operations. Therefore, the proposed change will not

result in a significant increase in the probability or consequences of

an accident previously evaluated.

Technical Specification 4.0.5

The proposed changes will revise Technical Specification 4.0.5.a

and Bases 3/4.4.10, ``Structural Integrity,'' by removing the phrase

``(g), except where specific written relief has been granted by the

Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).'' The

changes to Technical Specifications clarify that all applicable

requirements in 10 CFR 50.55a apply. The changes relate to inservice

inspection (ISI) and inservice testing (IST) requirements which are

specified in 10 CFR 50.55a, ``Codes and Standards.'' The ISI and IST

requirements are given in 10 CFR 50.55a, which the licensee documents

via its 10 year interval program requirements. Upon finding a Code

requirement impractical because of limitations in the design (including

prohibitive dose rates), construction, or system configurations, NNECO

[Northeast Nuclear Energy Company] would be required to prepare the

determination describing the impractical condition(s) and the

applicable code requirements that cannot be met in accordance with 10

CFR 50.55a, paragraphs (f)(5)(iii) and (iv), and (g)(5)(iii) and (iv)

if within the first 12 months of a new interval. For example, 10 CFR

50.55a(f)(5)(iv), and (g)(5)(iv) allow a licensee up to a full year

after the beginning of an updated interval to inform the NRC of the new

Code requirements which cannot be met and to request relief. If an

impracticality is identified after the first 12 months, the guidance

contained in NUREG-1482 will be followed. This will eliminate

inconsistencies between the Technical Specifications and the

regulations. There will be no adverse effect on plant operations.

Therefore, the proposed changes will not result in a significant

increase in the probability or consequences of an accident previously

evaluated.

Technical Specification 3.2.3

The proposed change will change the mode of applicability for

Technical Specification 3.2.3 from Mode 1 to Mode 1 with thermal power

>20%. Data from the incore detectors are used for determining the

measured radial peaking factors to verify compliance with Technical

Specification 3.2.3. However, the accuracy of the neutron flux

information from the incore detectors is not reliable below 20% power.

The proposed change acknowledges this limitation of the incore

detectors by changing the applicability of this specification to power

levels where the data from the incore detectors is reliable. This will

have no adverse effect on plant operations since the current Technical

Specification surveillance requirements do not require the verification

of this limit until prior to operation above 70% following each fuel

loading, prior to 31 days accumulated operation in Mode 1, or if the

azimuthal power tilt limit is exceeded (Technical Specification 3.2.4

which is applicable in Mode 1 above 50% power). Therefore, the proposed

change has no impact on the initial conditions, with respect to power

distribution, assumed in the accident analysis. Thus, the proposed

change will not result in a significant increase in the probability or

consequences of an accident previously evaluated.

Technical Specification 3.3.2.1

The proposed change will add an exception to Technical

Specification 4.0.4 that will allow the channel functional test of the

automatic actuation logic associated with ESF [engineered safety

feature] actuations for safety injection, containment spray,

containment isolation, main steam line isolation, enclosure building

filtration, and containment sump recirculation to be delayed during

plant startup until the actuation blocks are removed. This will allow

entry into Mode 3 where plant conditions (sufficient pressurizer and

steam generator pressure) can be established that will automatically

remove the blocks of these ESF actuations. The channel functional test

of the automatic actuation logic, using the ATI [Automatic Testing

Insertor] circuit, will then be performed. In addition, the channel

functional tests of the automatic actuation logic must be performed

prior to entering Mode 2.

The exception to Technical Specification 4.0.4 allows a mode change

with equipment that is inoperable only because conditions [cannot] be

established to perform the SR [surveillance requirement] until after

the mode is entered. All other equipment operability requirements must

be met. Even though operability of the automatic actuation logic for

the affected ESF actuations cannot be verified prior to entering Mode

3, this equipment is still expected to be operable. The ESFAS

[engineered safety feature actuation system] will continue to function

as before. Therefore, the proposed change will not result in a

significant increase in the probability or consequences of an accident

previously evaluated.

Technical Specification 3.4.1.1

The Flow Dependent Setpoint Selector Switch was installed to allow

power operation with less than four reactor coolant pumps (RCPs) in

operation by changing the reactor trip setpoints for the variable high

power, Reactor Coolant System (RCS) low flow,

[[Page 56253]]

and thermal margin low pressure (TM/LP) reactor trips. Millstone Unit

No. 2 is not currently licensed to operate with less than four RCPs in

operation. Therefore, this switch should be maintained in the four pump

position.

The use of the switch position to ensure compliance with Technical

Specification 3.4.1.1 provides an indirect verification of LCO

[limiting condition for operation] compliance since the loss of an RCP

will result in a reactor trip when in the four pump position. The

proposed change will replace the method used for LCO verification with

one that is more consistent with the LCO. Verification of switch

position is performed as a prerequisite prior to reactor startup

(entering Mode 2). It is not necessary to verify the switch position

every 12 hours as currently required. The position of this switch is

important to the operability of the associated Reactor Protection

System (RPS) trips variable high power, RCS low flow, and TM/LP). The

operability of these RPS trips and associated setpoints is already

covered by Technical Specifications 2.2.1, ``Reactor Trip Setpoints,''

and 3.3.1.1, ``Reactor Protective Instrumentation.''

It is not necessary to verify the position of this switch fifteen

minutes prior to reactor criticality since the switch position is

verified prior to a reactor startup, and is not expected to be changed

during power operation. If surveillance testing or maintenance

activities are to be performed which may require the switch to be in

other than the four pump position, the affected RPS channels will

already have been removed from service (declared inoperable and placed

in the tripped or bypassed condition) prior to commencing the

activities. In addition, a light (``PUMP SETPOINT ERROR'') on each of

the RPS Calibration and Indication Panels will illuminate if the switch

is not in the four pump position.

It is also not necessary to verify compliance with the requirements

of Technical Specification 3.4.1.1 within fifteen minutes prior to

reactor criticality since this condition is verified prior to a reactor

startup, and the RPS will initiate a reactor trip if less than four

RCPs are in operation.

The proposed change will replace SR 4.4.1.1, verification of the

Flow Dependent Setpoint Selector Switch position, with a verification

check of the required RCS loops. This verification is more consistent

with the Limiting Condition for Operation (LCO). This will not change

the requirement that both RCS loops be operable and operating in Modes

1 and 2. The Technical Specification will continue to assure that the

initial condition, with respect to RCS loops in service, in the

accident analysis is applicable. Therefore, the proposed change will

not result in a significant increase in the probability or consequences

of an accident previously evaluated.

Technical Specification 3.4.11

The proposed change to modify the wording of SR 4.4.11.3 will not

affect the operability requirements of the RCS Vent System. This change

will provide operational flexibility to use a series of overlapping

tests to verify flow through sections of the vent system, such that

when completed, flow will be verified through all parts of the vent

system. This will minimize potential contamination of the area

surrounding the sparger and will eliminate the need to establish solid

water conditions in the RCS.

The proposed surveillance requirement will still verify the ability

of the vent valves to operate. This will provide reasonable assurance

of system operability and availability if needed to mitigate the

consequences of design basis accidents. Therefore, the proposed change

will not result in a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes have no adverse effect on any of the design

basis accidents previously evaluated or on any equipment important to

safety. Therefore, the license amendment request does not impact the

probability of an accident previously evaluate nor does it involve a

significant increase in the consequences of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes will not alter the plant configuration (no new

or different type of equipment will be installed) or require any new or

unusual operator actions. They do not alter the way any structure,

system, or component functions and do not alter the manner in which the

plant is operated. The proposed changes do not introduce any new

failure modes. Therefore, the proposed changes will not create the

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

previously evaluated.

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

The proposed changes will correct reference, spelling, and

terminology errors in various Technical Specification Definitions; add

a new Technical Specification, 3.0.6; modify Technical Specification

4.0.5 to remove an inconsistency between the Technical Specification

and the regulations; change the applicability of Technical

Specification 3.2.3; add an exception to Technical Specification 4.0.4

to Technical Specification 3.3.2.1; modify the wording of a

surveillance requirement associated with RCS Technical Specification

3.4.1.1; and modify the wording of a surveillance requirement

associated with the RCS Vent System, Technical Specification 3.4.11 to

provide operational flexibility in the performance of the test. These

changes will have no adverse effect on equipment important to safety.

The equipment will continue to function as assumed in the design basis

accident analysis. Therefore, there will be no significant reduction of

the margin of safety as defined in the Bases for the Technical

Specifications affected by these proposed changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Project Director: William M. Dean.

Philadelphia Electric Company, Docket No. 50-353, Limerick Generating

Station, Unit 2, Montgomery County, Pennsylvania

Date of amendment request: September 14, 1998.

Description of amendment request: The proposed amendment to the

Limerick Generating Station (LGS), Unit 2, Technical Specifications

(TS) would revise TS Table 4.4.6.1.3-1, ``Reactor Vessel Material

Surveillance Program--Withdrawal Schedule.'' This table provides the

schedule for withdrawing the reactor pressure vessel material

surveillance program capsules. This proposed TS change involves

revising the schedule for withdrawing the first surveillance capsule

from 8 Effective Full Power years (EFPY) to 15 EFPY, and the second

surveillance capsule from 20 EFPY to 30 EFPY.

[[Page 56254]]

A revision to TS Surveillance Requirement (SR) 4.4.6.1.4 is also

proposed. This revision will remove the reference to flux wire removal

and analysis that was originally required following the first cycle of

operation. TS SR 4.4.6.1.4 will be changed to refer to the flux wires

that are located within the surveillance capsules, which will be

removed and analyzed in accordance with the surveillance capsule

removal schedule, located in Table 4.4.6.1.3-1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes do not increase the probability of occurrence

of an accident previously evaluated in the safety analysis report and

do not affect any accident initiators as described in the Safety

Analysis Report (SAR). The change revises the withdrawal schedule for

the reactor vessel material surveillance capsules. The capsules are not

an initiator of any previously analyzed accident nor does the

withdrawal schedule of the surveillance capsules affect the probability

or consequences of any previously analyzed accident.

The proposed changes will not affect the Pressure-Temperature (P-T)

limits as specified in LGS TS Figure 3.4.6.1-1 and Updated Final Safety

Analysis Report (UFSAR) Figure 5.3-4. P-T limits are imposed on the

reactor coolant system to ensure that adequate safety margins exist

during normal operation, anticipated operational occurrences, and

system hydrostatic tests. The P-T limits are related to the

RTNDT [reference temperatures], as described in ASME Section

III, Appendix G. Changes in the fracture toughness properties of RPV

[reactor pressure vessel] beltline materials, resulting from neutron

irradiation and the thermal environment, are monitored by a

surveillance program in compliance with the requirements of 10 CFR 50

Appendix H. The effect of neutron fluence on the shift in the

RTNDT is predicted by methods given in Regulatory Guide

1.99, Rev.2.

As detailed in Attachment 3 [of the September 14, 1998, submittal],

for LGS, Unit 2, the combination of low expected RTNDT shift

for the plate material due to low predicted fluence and excellent

material chemistry; Supplemental Surveillance Program (SSP) data on

similar material; and the inherent margin in the P-T curve

calculations, with the withdrawal schedule of the first surveillance

capsule modified from 8 EFPY to 15 EFPY and the second surveillance

capsule modified from 20 EFPY to 30 EFPY, will result in more credible

sets of surveillance data, while ensuring the continued safe operation

of LGS, Unit 2.

The current LGS P-T limits were established based on adjusted

reference temperatures developed in accordance with the procedures

prescribed in Regulatory Guide 1.99, Revision 2, Regulatory Position 1,

``Surveillance Data Not Available.'' Calculation of adjusted reference

temperature by these procedures includes a conservative base fluence

estimate; power rerate adjustment of a 110% fluence multiplier from

startup, instead of a 105% fluence multiplier since 2R03 [third

refueling outage]; and a margin term to ensure conservative, upper-

bound values are used for the calculation of the P-T limits. Revision

of the first capsule withdrawal schedule will not affect the P-T limits

because they will continue to be established in accordance with

Regulatory Position 1 guidance. Also, as indicated in Attachment 3, it

is also appropriate to extend the withdrawal of the LGS, Unit 2, second

capsule. The current schedule specifies withdrawal of the second

capsule at 20 EFPY. Based upon the information provided in Attachment 3

supporting withdrawal of the first capsule at 15 EFPY, there will be an

insignificant shift in material properties at 20 EFPY, after only an

additional exposure of 5 EFPY. It is appropriate to extend this

schedule to 30 EFPY which meets the intent of ASTM E185-82, such that

the withdrawal of the second capsule occurs before the accumulated

neutron fluence of the capsule corresponds to the approximate EOL [end

of life] fluence at the reactor pressure vessel inner wall location,

and provides consistency with the LGS, Unit 1, withdrawal schedule.

In accordance with the guidance stipulated in Regulatory Guide

1.99, ``Radiation Embrittlement of Reactor Vessel Materials,'' Revision

2, Regulatory Position 2, ``Surveillance Data Available,'' the

collection of two (2) or more sets of credible surveillance data is

necessary to empirically calculate the adjusted reference temperature

(ART). Each surveillance capsule constitutes one set of credible

surveillance data. This calculated ART can be used to revise the P-T

curves (TS Figure 3.4.6.1-1). Without two (2) or more sets of credible

data, the ART must be calculated and the P-T curves revised, based upon

the calculational methodologies as provided in the Regulatory Guide

1.99, Revision 2, Regulatory Position 1, ``Surveillance Data Not

Available.'' These methodologies use plant specific chemistry and

fluence values to determine a calculated shift in RTNDT. A

``margin'' term is then added, to obtain conservative, upper-bound

values of adjusted reference temperature.

The existing LGS, Unit 2, P-T curves are based upon the Regulatory

Position 1 methodology, and are currently valid up to 10 EFPY. With

first capsule removal at either 8 or 15 EFPY, the existing P-T curves

will require a revision, prior to reaching 10 EFPY, based upon the

calculational methodologies as contained in the Regulatory Guide 1.99,

Revision 2, Regulatory Position 1, ``Surveillance Data Not Available.''

Therefore, the Technical Specification revision to the first capsule

withdrawal schedule, as supported by this Safety Evaluation [supporting

information described in attachments 1 and 3 of the September 14, 1998,

submittal], results in no impact to the calculational methodologies

that will be used for the P-T curve revision that will be necessary to

extend the curves beyond 10 EFPY.

The fluence data as determined from the surveillance capsule flux

wires at 15 EFPY will provide an accurate indication of neutron

fluence. In accordance with Regulatory Guide 1.99, Revision 2,

Regulatory Position 1 methodology, data from these flux wires will

permit an adjustment of TS Figure 3.4.6.1-1 in accordance with TS SR

4.4.6.1.3, if required, and will meet the requirements of 10 CFR 50,

Appendix H, and ASTM E-185.

The proposed changes will not affect any plant safety limits or

limiting conditions of operation. The proposed changes will not affect

reactor pressure vessel performance as it involves no physical changes

and LGS P-T limits will remain conservative in accordance with

Regulatory Guide 1.99, Revision 2, guidance. The proposed changes will

not cause the reactor pressure vessel or interfacing systems to be

operated outside of their design or testing limits.

The proposed changes do not increase the probability of the

occurrence of a malfunction, or consequences of a malfunction, of

equipment important to safety previously evaluated in the SAR. The

proposed changes do not involve any physical changes to equipment

important to safety. The potential for reactor vessel failure will be

adequately assessed by the proposed withdrawal schedule. In addition,

the results from

[[Page 56255]]

the Supplemental Surveillance Program (SSP) will provide industry data

that bounds the materials used in the LGS vessel until the data from

the first LGS capsule is available. The proposed change provides the

same level of confidence in the integrity of the vessel. The P-T curves

are currently controlled by the TS and are determined using the

conservative methodology delineated in Regulatory Guide 1.99.

Therefore, the possibility of failure of the reactor vessel is not

increased. The current P-T limit curves are inherently conservative and

will continue to be adhered to.

Therefore, the proposed TS changes do not involve an increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The proposed changes do not create the possibility of a different

type of accident than any previously evaluated in the SAR. The proposed

changes are a revision of the withdrawal schedule for the first reactor

pressure vessel material surveillance capsule from 8 EFPY to 15 EFPY,

and for the second capsule from 20 EFPY to 30 EFPY. The proposed

changes do not involve a physical modification of the design of plant

structures, systems, or components. The proposed changes will not

impact the manner in which the plant is operated as plant operating and

testing procedures will not be affected by the change. No new accident

types or failure modes will be introduced as a result of the proposed

change.

LGS's current P-T limits were established based on adjusted

reference temperatures developed in accordance with the procedures

prescribed in Regulatory Guide 1.99, Revision 2, Regulatory Position 1,

``Surveillance Data Not Available.'' Calculation of adjusted reference

temperature by these procedures includes a conservative base fluence

estimate; power rerate adjustment of a 110% fluence multiplier from

startup, instead of a 105% fluence multiplier since 2R03; and a margin

term to ensure conservative, upper-bound values are used for the

calculation of the P-T limits. Revision of the first capsule withdrawal

schedule will not affect the P-T limits because they will continue to

be established in accordance with the guidance of Regulatory Position 1

of Regulatory Guide 1.99. Also, as specified in Attachment 3, it is

appropriate to extend the withdrawal of the LGS, Unit 2, second

capsule. The current schedule specifies withdrawal of the second

capsule at 20 EFPY. Based upon the information provided in Attachment 3

supporting withdrawal of the first capsule at 15 EFPY, there will be an

insignificant shift in material properties at 20 EFPY, after only an

additional exposure of 5 EFPY. It is appropriate to extend this

schedule to 30 EFPY which meets the intent of ASTM E185-82, such that

the withdrawal of the second capsule occurs before the accumulated

neutron fluence of the capsule corresponds to the approximate EOL

fluence at the reactor inner wall location, and provides consistency

with the LGS, Unit 1, withdrawal schedule.

The existing LGS, Unit 2, P-T curves are based upon the Regulatory

Position 1 methodology, and are currently valid up to 10 EFPY. With

first capsule removal at either 8 or 15 EFPY, the existing P-T curves

will require a revision, prior to reaching 10 EFPY, based upon the

calculational methodologies as contained in the Regulatory Guide 1.99,

Revision 2, Regulatory Position 1, ``Surveillance Data Not Available.''

Therefore, the proposed TS revision to the first capsule withdrawal

schedule results in no impact to the calculational methodologies that

will be used for the P-T curve revision that will be necessary to

extend the curves beyond 10 EFPY.

The fluence data as determined from the surveillance capsule flux

wires at 15 EFPY will provide an accurate indication of neutron

fluence. In accordance with Regulatory Guide 1.99, Revision 2,

Regulatory Position 1 methodology, data from these flux wires will

permit an adjustment of TS Figure 3.4.6.1-1 in accordance with TS SR

4.4.6.1.3, if required, and will meet the requirements of 10 CFR 50,

Appendix H, and ASTM E-185.

The potential for reactor vessel failure will be adequately

assessed by the proposed withdrawal schedule. In addition, the results

from the SSP will provide industry data that bounds the materials used

in the LGS vessel, until the data from the first LGS capsule is

available. The proposed changes provide the same level of confidence in

the integrity of the vessel . The P-T curves are currently controlled

by the TS and are determined using the conservative methodology in

Regulatory Guide 1.99. Therefore, the possibility of failure of the

reactor vessel is not increased. The current P-T limit curves are

inherently conservative and will continue to be adhered to.

Therefore, the proposed TS changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

in a margin of safety.

The proposed changes to the TS do not reduce the margin of safety

as defined in the Bases for any TS. The proposed changes will not

affect any safety limits, limiting safety system settings, or limiting

conditions of operation. The proposed changes do not represent a change

in initial conditions, system response time, or in any other parameter

affecting the course of an accident analysis supporting the Bases of

any TS. The proposed changes do not involve revision of the P-T limits,

but rather a revision of the withdrawal schedule for the surveillance

capsules. The current P-T limits were established based on the adjusted

reference temperatures for reactor pressure vessel beltline materials

calculated in accordance with the guidance stipulated in Regulatory

Position 1 of Regulatory Guide 1.99, Revision 2. P-T limits will

continue to be revised as necessary for changes in adjusted reference

temperature due to changes in fluence according to Regulatory Position

1 until two (2) or more credible surveillance data sets becomes

available. When two (2) or more credible surveillance data sets become

available, P-T limits will be revised as prescribed by Regulatory

Position 2 of Regulatory Guide 1.99, Revision 2, or other NRC approved

guidance.

The current P-T limit curves are inherently conservative and

provide sufficient margin to ensure the integrity of the reactor

vessel. The changes do not adversely affect these curves. The fluence

data as determined from the surveillance capsule flux wires at 15 EFPY

will provide an accurate indication of neutron fluence. In accordance

with Regulatory Guide 1.99, Revision 2, Regulatory Position 1

methodology, data from these flux wires will permit an adjustment of TS

Figure 3.4.6.1-1 in accordance with TS SR 4.4.6.1.3, if required, and

will meet the requirements of 10 CFR 50, Appendix H, and ASTM E-185.

Therefore, the proposed TS changes do not involve a 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

[[Page 56256]]

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101.

NRC Project Director: Robert A. Capra.

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

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

Date of amendment request: April 16, 1998.

Description of amendment request: This application for amendment to

the Indian Point 3 Technical Specifications (TSs) proposes to modify a

testing requirement for the emergency diesel generators (EDGs).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident previously

analyzed?

Response:

No. The three Emergency Diesel Generators (EDG) at Indian Point 3

are designed to provide a source of power to support a safe and orderly

plant shutdown in the event that all other normal and standby sources

of power are not available, such as during a postulated Loss of Offsite

Power (LOOP). The probability of such events occurring is not affected

by the proposed amendment. Any two of the three EDGs are capable of

supplying the minimum power requirements for emergency safeguards

equipment that mitigate the consequences of postulated design basis

accident conditions. Periodic preventive maintenance and surveillance

testing are performed to provide assurance that the operability of all

three EDGs is maintained. In the event that an inoperable EDG is

identified, both the existing specification and the proposed change

provide for actions that verify the operability of the remaining 2

EDGs. Operability of 2 EDGs ensures that sufficient emergency power is

available, if needed, to mitigate the consequences of postulated

accidents. Therefore, the proposed license amendment does not involve a

significant increase in the probability or consequences of an accident

previously analyzed.

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

new or different kind of accident from any accident previously

evaluated?

Response:

No. The proposed license amendment does not involve any physical

changes to plant systems or component setpoints. Also, there are no

changes to the way in which systems or equipment are operated. The

proposed change will continue to require that the operability of the

remaining two EDGs be verified if one of the three EDGs is found to be

inoperable. The proposed change to allow the use of a common cause

failure evaluation, as an alternative to testing, to accomplish the

operability verification can benefit overall EDG reliability by

eliminating unnecessary EDG starts. Therefore, the proposed license

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

accident from any accident previously evaluated.

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

a margin of safety?

Response:

No. Important performance requirements for the EDGs include

electrical output capacity, elapsed time to start and reach rated

output, and fuel storage supply to support a minimum period of

operation. The proposed amendment does not change EDG performance

requirements. The existing specification allows a period of 24 hours in

which to verify the operability of the remaining 2 EDGs if one of the

three EDGs is found inoperable. The proposed amendment does not change

the 24-hour time limit. Operability verification, either by testing or

evaluation, within 24 hours provides assurance that this source of

emergency power is available if needed. Therefore, the proposed

amendment does not involve a significant reduction in a margin of

safety. Also, this verification method has been approved for use with

the current Standard Technical Specifications.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

York, New York 10019.

NRC Project Director: S. Singh Bajwa, Director

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

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

Date of amendment request: April 16, 1998, as supplemented August

20, 1998.

Description of amendment request: This application for amendment to

Table 4.1-1 of the Indian Point 3 Technical Specifications (TSs)

proposes to change surveillance frequency requirements for the various

instrument channels to accommodate a 24-month operating cycle. The

proposed amendment also revises Section 6 of the TSs to reflect updated

analyses.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the proposed license amendment involve a significant increase

in the probability or consequences of an accident previously analyzed?

Response:

No. The proposed license amendment to extend the calibration

surveillance frequency of the following instrument channels is being

made to support plant operation with a 24-month fuel cycle:

(a) Pressurizer Water Level

(b) Accumulator Level and Pressure

(c) Reactor Coolant System Subcooling Margin Monitor

(d) Core Exit Thermocouples

(e) Reactor Vessel Level Indication System

Changing the calibration intervals for these instrument channels

neither directly nor indirectly affects the initiation or probability

of any previously analyzed accident. The changes do not affect the

integrity of any of the principal barriers against radiation release

(fuel cladding, reactor vessel, and containment building). The ability

of the plant to mitigate the consequences of any previously analyzed

accidents is not adversely affected. Evaluation of the proposed change

to the surveillance interval demonstrates that licensing basis safety

analyses acceptance criteria and Indian Point 3 Emergency Operating

Procedure (EOP) criteria continue to be met.

Item (a) provides an input to the Reactor Protection System (RPS)

to initiate a reactor trip if the measured parameters exceed specified

values. Item (b) is used by control room operators to ensure that the

accident mitigation capability of the accumulators is maintained within

specified limits. Items (c), (d), and (e)

[[Page 56257]]

provide post-accident information to control room operators to support

recovery efforts. Item (d) is also used to monitor core performance for

fuel management activities.

The proposed new surveillance frequency for these instrument

channels was evaluated using the guidance of Generic Letter 91-04. The

basis for the changes includes a quantitative evaluation of instrument

drift. Also, loop accuracy/setpoint calculations were updated to

accommodate the extended surveillance period. Analyses and evaluations

completed to assess the proposed increase in the surveillance interval

demonstrate that the effectiveness of these instruments in fulfilling

their respective functions is maintained. Channel checks required to be

performed each shift or each day, according to Technical Specifications

for the subject channels, will continue to be performed to provide

assurance of instrument channel operability. Therefore, the proposed

amendment does not involve a significant increase in the probability or

consequences of any previously analyzed accident.

Does the proposed license amendment create the possibility of a new

or different kind of accident from any accident previously evaluated?

Response:

No. The increased calibration surveillance intervals for the above

listed instrument channels were justified based on evaluation of past

equipment performance and do not require any plant hardware changes or

changes in normal system operation. Changing the calibration intervals

for these channels neither directly nor indirectly has any means of

creating the possibility of a new or different kind of accident.

Certain alarm and EOP setpoint changes will be made consistent with the

revised uncertainty calculations for the subject channels. These new

setpoints and related operator responses support existing accident

mitigation strategies and do not create the possibility of a new or

different kind of accident from any previously analyzed. Therefore,

there are no new failure modes introduced as a result of extending

these surveillance intervals, and the proposed amendment does not

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

accident previously evaluated.

Does the proposed amendment involve a significant reduction in a

margin of safety?

Response:

No. Pressurizer water level instrumentation provides input to the

reactor protection system and to the pressurizer water level control

system. Pressurizer water level, as indicated by the selected control

channel, is used to establish the initial condition pressurizer water

level assumption for certain UFSAR [Updated Final Safety Analysis

Report] Chapter 14 safety analyses. The proposed change to the

calibration surveillance interval was evaluated using the criteria of

95% probability/95% confidence level for process sensor drift. The loop

accuracy/setpoint calculations were updated for the level channels to

demonstrate the acceptability of the proposed increase in the

surveillance interval. There are no changes required to the limiting

safety system setting (LSSS) stated in the Technical Specifications for

these channels. The LSSS for high pressurizer water level will remain

at [less than or equal to] 92% of span. The margin of safety between

the specified LSSS value required by Technical Specifications and the

safety limit used in the UFSAR Chapter 14 safety analyses is unchanged.

The instrument channels for accumulator pressure and level do not

provide input to the reactor protection system or the engineered safety

features system. These instruments provide alarms and indication to

control room operators to maintain accumulator cover gas pressure and

water volume within specified limits. They are also used for

establishing initial condition accumulator pressure and level

assumptions for certain UFSAR Chapter 14 safety analyses. Accordingly,

the process sensor drift analysis was performed using the criteria of

95% probability/75% confidence level.

The remaining three instrument channels addressed by this proposed

license change are used to provide indication of adequate core cooling

following certain hypothetical accident conditions. These instrument

channels are not associated with any margin of safety specified by the

Technical Specifications, and they are not factors in any UFSAR Chapter

14 safety analyses. However, they are factored into the calculations of

pertinent setpoints used in alarm response procedures and EOPs. The

updated drift and uncertainty calculations and evaluations for these

instrument channels demonstrate that applicable accuracy requirements

for Indian Point 3 are satisfied with the proposed new surveillance

intervals. The instrument channels will remain effective to support

plant operator implementation of the Emergency Operating Procedures,

which are consistent with the Westinghouse Owners' Group Emergency

Response Guidelines.

Changing the calibration interval for these channels does not

affect margin of safety for previously analyzed accidents. Also, the

evaluation of related changes to UFSAR Chapter 14 safety analyses input

assumptions has demonstrated that licensing basis safety analysis

acceptance criteria and EOP criteria continue to be met, and previously

existing margins based on these pertinent acceptance criteria continue

to be maintained.

Therefore, the proposed amendment does not involve a significant

reduction in a margin of safety.

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

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

satisfied. The staff has also reviewed the licensee's proposed change

to reflect updated safety analyses in Section 6 of the TSs and it

appears that the three standards of 50.92(c) are satisfied for these

changes as well. Therefore, the NRC staff proposes to determine that

the amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

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

York, New York 10019.

NRC Project Director: S. Singh Bajwa, Director.

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 amendment request: September 17, 1998.

Description of amendment request: The amendments would revise

Technical Specification (TS) 3/4.8.2, ``Electrical Power Sources--

Shutdown,'' for the AC distribution system and the 125-volt and 28-volt

DC distribution systems. Specifically, the amendments would change the

Applicability and Action Statements, if less than the complement of

equipment and busses are operable, to eliminate the need to establish

containment integrity and to add the action to suspend core

alterations, positive reactivity additions, and movement of irradiated

fuel assemblies.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will not involve a significant increase in the probability or

[[Page 56258]]

consequences of an accident previously evaluated.

In Modes 1 through 4 [power operation through hot shutdown], a

Design Basis Accident would cause the release of radioactive material

into the containment. Release of that radioactive material to the

environment is prevented during operation in Modes 1 through 4 by

maintaining containment integrity. In Modes 5 and 6 [cold shutdown and

refueling] the probability and consequences of this event are lower

because of the reduced reactor coolant pressure and temperature

limitations of these modes.

A minimum complement of electrical power sources and distribution

systems is established in Modes 5 and 6 to assure that adequate

electrical power is available to mitigate the consequences of a fuel

handling accident. Because of the lack of containment pressurization

potential during a fuel handling accident, less stringent requirements

are needed to isolate containment from the outside atmosphere. These

requirements are applied during refueling operations by Technical

Specification 3.9.4, Refueling Operations, Containment Building

Penetrations. Technical Specification 3.9.4 is applicable in Mode 6 and

establishes containment closure vice containment integrity during

refueling operation (core alterations and movement of irradiated fuel

within containment).

In Mode 5, fuel handling is generally limited to placement of new

fuel prior to core off load or movement of irradiated fuel within the

spent fuel pool. Because the Spent Fuel Pool is not located within

containment, establishment of either containment integrity or

containment closure would not help to mitigate the consequences of a

fuel handling accident in that area. Mitigation of a fuel handling

accident is accomplished through Technical Specification 3.9.12,

Refueling Operations, Fuel Handling Area Ventilation System, which

requires that the Fuel Handling Area Ventilation system be operable

whenever irradiated fuel is present in the storage pool. This insures

that all radioactive material released from the rupture of an

irradiated fuel assembly would be filtered through filtration equipment

prior to discharge to the atmosphere.

With the number of energized A.C. or D.C. power distribution

systems less than the required, sufficient power may not be available

to recover from a fuel handling accident. Consequently, the Action

statements require immediate suspension of all operations involving

core alterations, positive reactivity changes, and movement of

irradiated fuel assemblies. This precludes the possibility of a fuel

handling accident and the need for containment integrity.

Based upon the above, the proposed change will not increase the

probability or consequences of an accident previously analyzed.

2. Will not create the possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes do not require any change in the configuration

or operation of the plant. Specifically, no new hardware is being added

to the plant as part of the proposed change, no existing equipment is

being modified, and no significant changes in operations are being

introduced. Therefore, these changes will not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. Will not involve a significant reduction in a margin of safety.

The proposed change will not alter any assumptions, initial

conditions, or results of any accident analyses. The proposed

additional Applicability will ensure proper operation of the Fuel

Handling Area Ventilation system during movement of irradiated fuel in

the spent fuel pool. The proposed ACTIONS, to be taken in the event

that the LCO [limiting condition for operation] is not met, will

preclude the conditions that would lead to the need for establishing

containment integrity. The change will, therefore, 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: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Project Director: Robert A. Capra.

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 amendment request: September 29, 1998.

Description of amendment request: The proposed amendments would

revise Technical Specification 3/4.9.4, ``Refueling Operations,

Containment Building Penetrations,'' to permit the use of equivalent

methods to obtain containment closure during refueling operations.

Specifically, the proposed changes would allow the installation of an

outage equipment door or other closure devices that are capable of

providing access for temporary services needed to support maintenance

activities within containment.

In addition to the above changes, the terminology for the

Containment Equipment Hatch inside door used in LCO 3.9.4.a is being

changed. The term ``Containment Equipment Door'' is being changed to

``Containment Equipment Hatch Inside Door'' to bring it into agreement

with the terminology used in Salem design documents.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

In Modes 1 through 4 [power operations through hot shutdown], a

Design Basis Accident would cause the release of radioactive material

into the containment. The release of radioactive material from the

containment to the environment is prevented during operation in Modes 1

through 4 by maintaining CONTAINMENT INTEGRITY. In Mode 5 and 6 [cold

shutdown and refueling] the requirements to prevent releases from the

containment to the environment from postulated accidents are less

stringent because of the reduced reactor coolant pressure and

temperature limitations of these modes. In all cases, the containment

serves as a passive barrier to mitigate the consequences of accidents

analyzed. The containment is not considered to be a contributor to the

probability of those accidents. Therefore, this change, which will

permit the use of equivalent methods for establishing containment

closure during refueling operations, will not increase the probability

of an accident previously analyzed.

During refueling operations, a release of radioactive material to

the containment could occur as the result of a fuel handling accident.

Actions are taken to mitigate the consequences of a fuel handling

accident inside containment during refueling operations through

application of technical specification requirements for Refueling

Cavity water level, minimum decay time prior to CORE ALTERATIONS, and

Containment Building Penetrations.

Because of the lack of containment pressurization potential and the

reduced

[[Page 56259]]

source term during a fuel handling accident, less stringent

requirements are needed to isolate containment from the outside

atmosphere. These requirements are applied during refueling operations

by Technical Specification 3.9.4, Refueling Operations, Containment

Building Penetrations. Technical Specification 3.9.4 is applicable in

Mode 6 and establishes containment closure vice CONTAINMENT INTEGRITY

during CORE ALTERATIONS and movement of irradiated fuel within

containment. Containment closure means that all potential release paths

are closed or capable of being closed to provide an atmospheric

pressure, ventilation barrier. Since there is no potential for

containment pressurization, establishment of a pressure tight boundary

is not required.

As a part of the containment closure requirements of Technical

Specification 3.9.4, the Containment Equipment Hatch inside door must

be installed with a minimum of four bolts. In addition, each

penetration providing direct access from the containment atmosphere to

the outside atmosphere must be closed by either an isolation valve, a

blind flange, or a manual valve, or must be capable of being closed by

an OPERABLE automatic containment isolation valve.

The proposed changes will modify Technical Specification 3/4.9.4 to

permit the use of an equivalent closure device as an alternative to

installation of the inner door with a minimum of four bolts to provide

containment closure for the Containment Equipment Hatch. The proposed

change will also modify Technical Specification 3.9.4 to permit the use

of an equivalent method for containment closure for containment

penetrations providing direct access from the containment to the

outside atmosphere as an alternate method to closure by an isolation

valve, blind flange, or manual valve. Any alternate method used will be

designed, fabricated, installed, tested, and utilized in accordance

with established procedures to ensure that it is capable of providing

containment closure during a fuel handling accident to prevent the

release of fission product radioactivity to the environment. Because

the proposed technical specifications must provide equivalent

containment closure, these changes will not increase the consequences

of an accident previously evaluated.

Based upon the above, the proposed changes do not increase the

probability or the consequences of an accident previously evaluated.

2. Will not create the possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes do not require any change in the operation of

the plant. The proposed changes will permit the use of an equivalent

method to achieve containment closure for the Containment Equipment

Hatch or for individual containment penetrations that provide direct

access to the outside atmosphere. However, any equivalent method used

will be designed, fabricated, installed, tested, and utilized in

accordance with established procedures to ensure that the closure

method meets design requirements.

Based upon the above, these changes will not create the possibility

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

evaluated.

3. Will not involve a significant reduction in a margin of safety.

The proposed change will not affect the existing analysis that

forms the basis for the Technical Specifications, and does not violate

Technical Specification and Updated Final Safety Analysis Report

(UFSAR) requirements. The proposed change will not affect any design or

functional requirements of the containment, the Containment Equipment

Hatch, or containment penetrations or any conditions or assumptions of

the applicable safety analyses.

Based upon the above, the proposed changes will not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Project Director: Robert A. Capra.

Southern California Edison Company, et al., Docket No. 50-362, San

Onofre Nuclear Generating Station, Unit No. 3, San Diego County,

California

Date of amendment request: September 22, 1998.

Description of amendment request: The proposed amendment would

modify the Technical Specifications (TS) to change the parameter used

to establish and remove the bypasses for high reactor power trips. The

parameter would be changed from the current ``THERMAL POWER'' to

logarithmic power. This amendment was processed on San Onofre Nuclear

Generating Station (SONGS) Unit 2 under emergency circumstances to

allow resumption of power operations, and is being processed under

normal notice circumstances on SONGS Unit 3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed change to Technical Specification (TS) 3.3.1 does not

adversely impact structure, system, or component design or operation in

a manner which would result in a change in the frequency of occurrence

of accident initiation. The reactor trip bypass and automatic enable

functions are not accident initiators. Consequently, the proposed TS

change will not significantly increase the probability of accidents

previously evaluated. Clarifying the input process variable of the

operating bypasses and automatic bypass removals of the affected

reactor trips does not alter the setpoint nor the manner of operation

of the operating bypasses and automatic bypass removals. Therefore, the

consequences of previously evaluated accidents remain unchanged.

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

different kind of accident from any accident previously evaluated.

No new or different accidents result from clarifying the input

process variable of the operating bypasses and automatic bypass

removals of the affected reactor trips. The results of previously

performed accident analyses remain valid. Therefore, this amendment

request does not create the possibility of a new or different kind of

accident.

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

a margin of safety.

The proposed change does not alter the setpoint nor the manner of

operation of the operating bypasses and automatic bypass removals of

the affected reactor trips. The change merely replaces the

identification of the input process variable with the appropriate

identification of power. Therefore, this amendment request does not

involve a significant reduction in any margin of safety.

[[Page 56260]]

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

Attorney for licensee: Douglas K. Porter, Esquire, Southern

California Edison Company, P. O. Box 800, Rosemead, California 91770.

NRC Project Director: William H. Bateman.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: August 31, 1998.

Description of amendment request: The proposed amendment would

revise the cold overpressure mitigation curves in Technical

Specification (TS) Figure 3.4-4. This change would account for the TS

maximum allowable power-operated relief valve setpoint changes

associated with the new Model Delta 94 steam generator operating

parameters.

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 current pressurizer maximum allowable Power Operated Relief

Valve (PORV) setpoints, provided by the Cold Overpressure Mitigation

System (COMS) curves (Figure 3.4-4) of Technical Specification 3.4.9.3,

are nonconservative for application with the new Delta 94 Replacement

Steam Generators. The South Texas Project Cold Overpressure Event has

been re-analyzed as a result of changed operating parameters due to

installation of new Delta 94 Steam Generators. The re-analysis

determined that maximum allowable PORV setpoint required decreases to

ensure that the Cold Overpressure Mitigation System (COMS) continued to

provide design basis low temperature overpressure protection with Delta

94 Steam Generators. New COMS curves have been developed and are to be

incorporated into Technical Specification 3.4.9.3 by this change

request. Since the proposed COMS curves result in maximum allowable

PORV setpoint decreases to account for the changed Delta 94 Steam

Generator operating parameters, these curves are more conservative than

the existing COMS curves utilized for Model E Steam Generators.

Therefore, application of these proposed COMS curves for a unit with

Model E or Delta 94 Steam Generators ensures compliance with the

original design basis of the Cold Overpressure Mitigation System for

the South Texas Project.

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

changed operating parameters associated with the Delta 94 Replacement

Steam Generators. Reflecting actual operating parameters and adjusting

the maximum allowable PORV setpoints, as necessary, in the conservative

direction has no adverse effect on the probability or consequences of

an accident previously evaluated. Therefore, the proposed change does

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.

The proposed PORV maximum allowable setpoint changes do no create

any new operating conditions or modes. The proposed change only revises

the maximum allowable PORV setpoint curves for the Cold Overpressure

Mitigation System to account for the revised operating parameters

associated with Delta 94 Steam Generators. The actions of this system

continue to be performed in accordance with existing

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