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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to 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 tendency 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 January 15, 1999, through January 29, 1999.

The last biweekly notice was published on January 27, 1999 (64 FR

4152).

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

[[Page 6693]]

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Administration Services, Office of

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

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

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

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

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

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By March 12, 1999, 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 aspects(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 preset 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

[[Page 6694]]

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.

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

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

Date of amendment request: December 7, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) to permit a one-time only

extension of the steam generator tube inspection interval for fuel

cycle 14 and delete the requirement to have NRC staff concurrence of

the steam generator examination program. Specifically, TS 4.13A.2.a

would be revised with a footnote that states ``Examinations scheduled

for 1999 only, shall be conducted during the 2000 Refueling Outage

which will commence no later than June 3, 2000. The scheduled

examinations will be completed prior to return to service from the 2000

Refueling Outage.'' In addition, TS 4.13C.1 would be revised to state

``The proposed steam generator examination program shall be submitted

for NRC staff review at least 60 days prior to each scheduled

examination.''

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 59.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change does not involve any physical modifications

to the plant or modification in the methods of plant operation which

could increase the probability or consequences of previously

evaluated accidents. The proposed change permits an extension of the

current steam generator tube inservice inspection cycle. This

extension would allow the steam generator tube examinations to be

conducted during the 2000 refueling outage which will commence no

later than June 3, 2000. The basis for acceptance of this increase

in the technical specification limit is the ``non-operating'' steam

generator time between the last examination and the upcoming

examination. Extending the steam generator ``operating'' duration by

48 days would not significantly increase wear which might lead to

tube failure. No appreciable steam generator tube wear or

degradation is expected as a result of this extension. This change

will not affect the scope, methodology, acceptance limits and

corrective measures of the existing steam generator tube examination

program. The probability and consequences of failure of the steam

generators due to leaking or degraded tubes is not increased by the

proposed change. Additionally the proposed administrative change to

delete the requirement to receive NRC concurrence of the proposed

steam generator examinations will have no bearing on the actual

results of the steam generator examinations. Therefore, the

probability and the consequence of a design basis accident are not

being increased by the proposed change.

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

or different kind of accident from any accident previously

evaluated.

Plant systems and components will not be operated in a different

manner as a result of the proposed Technical Specification change.

The proposed change permits the upcoming steam generator tube

examination to be conducted during the 2000 refueling outage that

will commence no later than June 3, 2000. There are no plant

modifications or changes in methods of operation. This extension is

based upon the ``non-operating'' steam generator time between the

last examination and the upcoming examination. Extending the steam

generator ``operating'' duration by an additional 48 days would not

significantly increase wear which might lead to tube failure. The

proposed extension will not increase the probability of occurrence

of a tube rupture, increase the probability or consequences of an

accident, or create any new accident precursor. Additionally the

proposed administrative change to delete the requirement to receive

NRC concurrence of the proposed steam generator examinations will

have no bearing on the actual results of the steam generator

examinations. Therefore, the possibility of an accident of a

different type than was previously evaluated in the safety analysis

report is not created by the proposed change to the Technical

Specification.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change to Technical specification section 4.13A.2.a

will not reduce the margin of safety. This amendment involves an

extension of the current steam generator tube inservice inspection

cycle. The basis for acceptance of this increase in the technical

specification limit is the ``non-operating'' steam generator time

between the last examination and the upcoming examination. Extending

the steam generator ``operating'' duration by an additional 48 days

would not significantly increase wear which might lead to tube

failure. No appreciable steam generator tube wear or degradation is

expected as a result of this extension. Additionally the proposed

administrative change to delete the requirement to receive NRC

concurrence of the proposed steam generator examinations will have

no bearing on the actual results of the steam generator

examinations. Therefore, the accident analysis assumptions for

design basis accidents are unaffected and the margin of safety is

not decreased by the proposed Technical Specification change.

[* * *]

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

New York, New York 10003.

NRC Project Director: S. Singh Bajwa, Director.

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

Unit No. 1, Pope County, Arkansas

Date of amendment request: April 30, 1998.

Description of amendment request: The proposed amendment revises

the definition of quadrant power tilt to clearly allow the use of

either the incore detectors or the excore detectors for determining

quadrant power tilt.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1--Does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change to the quadrant power tilt (QPT) definition

will not alter any Safety Analysis Report (SAR) assumptions

established and implemented by the technical specifications. The

proposed change will allow the use of either the incore detectors or

the excore power range detectors for determining QPT. This change is

consistent with the improved Standard Technical Specifications (STS)

which has been previously approved by the NRC. QPT measured by

incore detectors provides a more accurate indication of reactor core

power distribution than the value determined from the excore

detectors. The accident prevention and mitigation features of the

plant are not affected by this proposed amendment.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

[[Page 6695]]

Criterion 2--Does not create the possibility of a new or

different kind of accident from any previously evaluated.

The proposed change to the definition of QPT does not alter the

ANO-1 SAR analysis or core operating limits report (COLR). The

change will clearly permit the use of either the incore detectors or

the excore detectors for monitoring QPT. The design and physical

configuration of the plant are not affected by this change.

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

or different kind of accident from any previously evaluated.

Criterion 3--Does not involve a significant reduction in the

margin of safety.

The proposed change to the QPT definition incorporates the

improved TS definition contained in NUREG-1430. The revised

definition allows the use of either the incore detectors or the

excore power range detectors for determination of QPT. The change

does not vary or affect any of the plant's operating parameters. The

COLR currently specifies acceptable QPT limits based upon the

measurement techniques. These limits are based upon the unique

measurement characteristics of the incore and excore power range

detectors and assure the measurement independent limit is not

violated.

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, NW., Washington, DC 20005-3502.

NRC Project Director: John N. Hannon.

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

Unit No. 1, Pope County, Arkansas

Date of amendment request: August 6, 1998.

Description of amendment request: The proposed amendment revises

the minimum and the maximum concentration limits for the sodium

hydroxide tank. The proposed change also revises the minimum specified

tank volume to refer to the parameter used in the analysis with no

allowance for instrument uncertainty and deletes the maximum specified

tank volume.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1--Does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Sodium hydroxide is not an accident initiator. It is, however, a

contributor to the mitigation of the effects of a Loss-of-Coolant-

Accident (LOCA). The proposed change in NaOH tank concentration

results in changing the expected post-LOCA reactor building sump pH.

The reduction in the lower value of sump pH, from 8.5 to 7.0, is

acceptable based on guidance contained in NUREG-0800, Standard

Review Plan, Section 6.5.2, ``Containment Spray as a Fission Product

Cleanup System Review Responsibilities,'' Revision 2, December 1988.

This guidance allows the assumption of long-term iodine retention

when the equilibrium sump pH, after mixing and dilution with the

primary coolant and ECCS injection, is above 7.0. Although the

change allows the volume of the NaOH tank to be maintained at a

lower volume, the proposed minimum volume bounds the analyses of

concern.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

Criterion 2--Does not create the possibility of a new or

different kind of accident from any previously evaluated.

Sodum hydroxide is added for iodine removal and for pH

adjustment of the borated water in the reactor building sump

following a LOCA. The proposed changes in NaOH tank concentration

and volume introduce no new mode of plant operation.

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

or different kind of accident from any previously evaluated.

Criterion 3--Does not involve a significant reduction in the

margin of safety.

The proposed change in NaOH tank concentration results in

changing the expected post-LOCA reactor building sump pH. This

proposed change does involve an incremental reduction in the margin

to safety since iodine retention is dependent on the pH of the sump/

spray solution. However, this reduction is not considered

significant in that the effect of the change in sump pH, from 8.5 to

7.0 has a relatively minor effect on iodine retention, as supported

by Standard Review Plan (NUREG-0800), Section 6.5.2, Revision 2,

dated December 1988. Although the change allows the volume of the

NaOH tank to be maintained at a lower volume, the proposed minimum

volume bounds the analyses of concern.

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, NW., Washington, DC 20005-3502.

NRC Project Director: John N. Hannon.

Entergy Operations, Inc., Et Al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi and Entergy

Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458,

River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: January 12, 1999, superceding the

amendment request in the letter of September 30, 1996, for both

stations.

Description of amendment request: The proposed amendment would add

an additional required action to the Limiting Condition for Operation

(LCO) 3.9.1, ``Refueling Equipment Interlocks,'' of the Technical

Specifications for both stations. The additional action would allow an

alternative to the current action for one or more inoperable refueling

equipment interlocks. The current action is to ``suspend in-vessel fuel

movement with equipment associated with the inoperable interlock(s).''

The alternative action proposed is to (1) insert a control rod

withdrawal block, and (2) verify all control rods are fully inserted in

core cells containing one or more fuel assemblies. The proposed

amendment would also revise the Bases for the LCO 3.9.1 actions to

describe the proposed alternative actions. The previous Federal

Register notice of the amendment request in the superceded letter of

September 30, 1996, was issued on June 16, 1996, (61 FR 31178), for

Grand Gulf Nuclear Station (GGNS).

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:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

The refueling interlocks are explicitly assumed in the GGNS

Updated Final Safety Analyses Report (UFSAR) and RBS Updated Safety

Analyses Report (USAR) analysis of the control rod removal error or

fuel loading error during refueling. This analysis evaluates the

probability and consequences of control rod withdrawal during

refueling. Criticality and, therefore, subsequent prompt reactivity

excursions are prevented during

[[Page 6696]]

the insertion of fuel, provided all required control rods are fully

inserted during the fuel insertion. The refueling interlocks

accomplish this by preventing loading fuel into the core with any

control rod withdrawn, or by preventing withdrawal of a rod from the

core during fuel loading.

When the refueling interlocks are inoperable the current method

of preventing the insertion of fuel when a control rod is withdrawn

is to prevent fuel movement. This method is currently required by

the Technical Specifications. An alternate method to ensure that

fuel is not loaded into a cell with the control rod withdrawn is to

prevent control rods from being withdrawn and verify that all

control rods required to be inserted are fully inserted. The

proposed actions will require that a control rod block be placed in

effect thereby ensuring that control rods are not subsequently

inappropriately withdrawn. Additionally, following placing the

control rod withdrawal block in effect, the proposed actions will

require that all required control rods be verified to be fully

inserted. This verification is in addition to the requirements to

periodically verify control rod position by other Technical

Specification requirements. These proposed actions will ensure that

control rods are not withdrawn and cannot be inappropriately

withdrawn because an electrical or hydraulic block to control rod

withdrawal is in place. Like the current requirements the proposed

actions will ensure that unacceptable operations are blocked (e.g.,

loading fuel into a cell with a control rod withdrawn except

following the requirements of LCO 3.10.6, ``Multiple Control Rod

Removal--Refueling,'' which is unaffected by this change).

The proposed additional acceptable Required Actions provide an

equivalent level of assurance that fuel will not be loaded into a

core cell with a control rod withdrawn as the current Required

Action or the Technical Specification Surveillance Requirement.

Therefore, the proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The change in the Technical Specification requirements does not

involve a change in plant design. The proposed requirements will

continue to ensure that fuel is not loaded into the core when a

control rod is withdrawn except following the requirements of LCO

3.10.6, ``Multiple Control Rod Removal-Refueling,'' which is

unaffected by this change.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

As discussed in the Bases for the affected Technical

Specification requirements, inadvertent criticality is prevented

during the insertion of fuel provided all required control rods are

fully inserted during the fuel insertion. The refueling interlocks

function to support the refueling procedures by preventing control

rod withdrawal during fuel movement and the inadvertent loading of

fuel when a control rod is withdrawn.

The proposed change will allow the refueling interlocks to be

inoperable and fuel movement to continue only if a control rod

withdrawal block is in effect and all required control rods are

verified to be fully inserted. These proposed Required Actions

provide an equivalent level of protection as the refueling

interlocks by preventing a configuration which could lead to an

inadvertent criticality event. The refueling procedures will

continue to be supported by the proposed required actions because

control rods cannot be withdrawn and as a result fuel cannot be

inadvertently loaded when a control rod is withdrawn except

following the requirements of LCO 3.10.6, ``Multiple Control Rod

Removal--Refueling,'' which is unaffected by this change.

Therefore, the proposed changes do not cause 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: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120, for Grand Gulf

Nuclear Station, and Government Documents Department, Louisiana State

University, Baton Rouge, LA 70803, for River Bend Station.

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

Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502, for

Grand Gulf Nuclear Station, and Mark Wetterhahn, Esq., Winston &

Strawn, 1400 L Street, NW., Washington, DC 20005, for River Bend

Station.

NRC Project Director: John N. Hannon.

Florida Power and Light Company, Et Al., Docket No. 50-335, St.

Lucie Plant, Unit No. 1, St. Lucie County, Florida

Date of amendment request: November 22, 1998.

Description of amendment request: The proposed amendment would

revise the reactor thermal margin safety limit lines and flow rates

stated in the technical specifications (TS). The amendment would also

update the reference for dose conversion factors used in Dose

Equivalent Iodine-131 calculations, and administrative changes to the

criticality analysis uncertainty described in TS 5.6.1.a.1, update the

analytical methods used in determining core operating limits listed in

TS 6.9.1.11, and revise the TS bases for the steam generator pressure-

low trip setpoint.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Replacement of the St. Lucie Unit 1 steam generators in 1997

resulted in an increase in RCS [reactor coolant system] flow. The

proposed amendment would increase the values of design minimum

reactor coolant flow and the low flow trip setpoint presently stated

in the Technical Specifications (TS). These revisions are

accompanied by a corresponding change to the Thermal Margin Safety

Limit Lines of TS Figure 2.1-1. The RCS flow related revisions do

not change the probability of any previously evaluated accident, as

they do not impact any plant component, structure or system

affecting the accident initiators. The proposed changes would

continue to maintain adequate operational margin to TS limits for

RCS flow and the low-flow trip setpoint.

The proposed changes to the thyroid dose conversion factors from

TID-14844 to ICRP-30, fuel storage TS 5.6.1.a.1, the list of

analytical methods in TS 6.9.1.11, and the Bases for Steam Generator

Pressure-Low trip setting have no relevance to the accident

initiators, and thus do not affect the frequency of occurrence of

previously analyzed transients. Additionally, there are no changes

to any active plant component due to these proposed changes.

The supporting evaluation of proposed TS changes demonstrates

acceptable results for all the accidents previously analyzed, and it

is concluded that the radiological consequences would remain within

their established acceptance criteria when including the effects of

increased RCS flow, increased low flow trip setpoint, and change to

the thyroid dose conversion factors used in the determination of

dose consequences. Proposed changes to the Bases for the Steam

Generator Pressure-Low trip setpoint, fuel storage design features,

and the list of analytical methods in TS 6.9.1.11 are administrative

in nature and do not impact current safety analyses.

Therefore, operation of the facility in accordance with the

proposed amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

(2) Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

This proposed amendment revises limiting flow parameters to

derive analysis benefits from increased RCS flow due to the

replacement stream generators, while assuring safe plant operation

commensurate with the proposed RCS flow and low flow

[[Page 6697]]

trip setpoint changes. These changes along with the proposed changes

to the Bases for the Steam Generator Pressure-Low trip setpoint,

dose conversion factors, the list of analytical methods in TS

6.9.1.11, and the fuel storage design features do not require

modifications to the plant configuration, systems or components

which would create new failure modes. There would be no change in

the modes of operation of the plant. The design functions of all the

safety systems remain unchanged. Therefore, operation of the

facility in accordance with the proposed amendment would not create

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

accident previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendment revises limiting flow parameters to

derive analysis benefits from increased RCS flow due to the

replacement steam generators, while assuring safe plant operation

commensurate with the proposed design minimum RCS flow and low-flow

trip setpoint changes. FPL has evaluated the impact of the proposed

changes on available margin to the acceptance criteria for Specified

Acceptable Fuel Design Limits (SAFDL), 10 CFR 50.46(b) requirements,

primary and secondary over-pressurization, peak containment

pressure, potential radioactive releases, and existing limiting

conditions for operation. With the proposed changes to the design

minimum RCS flow, low-flow trip setpoint, and dose conversion

factors, FPL has concluded that there would be no adverse impact to

the existing safety analyses. The proposed changes to the Bases for

the Steam Generator Pressure-Low trip setpoint, the list of

analytical methods in TS 6.9.1.11, and the fuel storage design

features are administrative in nature. Therefore, operation of the

facility in accordance with the proposed amendment would not involve

a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Project Director: Cecil O. Thomas.

Florida Power and Light Company, Et Al., Docket No. 50-389, St.

Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: December 18, 1998.

Description of amendment request: The proposed amendment would

revise the St. Lucie Unit 2 Plant Technical Specifications (TS) Index

Page III; TS 1.10, Dose Equivalent I-131; TS 2.1.1.2, Linear Heat Rate;

Bases 2.1.1, Reactor Core; Bases Figure B2.1-1, Axial Power

Distributions for Thermal Margin Safety Limits; Bases 2.2.1, Reactor

Trip Setpoints (Variable Power Level-High); TS 3.1.1.1/4.1.1.1.1,

Shutdown Margin--Tavg Greater Than 200 deg.F; TS 3/4.1.1.2, Shutdown

Margin--Tavg Less Than or Equal to 200 deg.F; TS 3.1.2.2, Boration

Systems Flow Paths--Operating; TS 3.1.2.4, Charging Pumps--Operating;

TS 3.1.2.6, Boric Acid Makeup Pumps--Operating; TS 3.1.2.8, Borated

Water Sources--Operating; Bases 3/4.1.1.1 and 3/4.1.1.2, Shutdown

Margin; Bases 3/4.1.2, Boration Systems; and TS 6.9.1.11, Core

Operating Limits Report (COLR). The core operating limits for shutdown

margin will be relocated to the St. Lucie Unit 2 COLR.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendment involves changes to the dose conversion

factors used in the thyroid dose calculations and the relocation of

the SHUTDOWN MARGIN requirements for Modes 1 through 5 from TS to

the Core Operating Limits Report (COLR). Additionally, the peak

linear heat rate value corresponding to centerline melt is deleted

from the TS. The deletion of this TS remains consistent with the

requirements of 10 CFR 50.36. Bases Figure B2.1-1 is replaced with a

new figure, consistent with the input assumptions of the safety

analysis report.

The proposed amendment addresses analytical methods changes such

as the use of HERMIT code in one dimensional mode for spatial

details, the rod bow penalty calculations using L2/I

dependence discussed in CEN-289 (A)-P, CEAW methodology change for

crediting the delta-T power trip, and the methodology for core

designs containing Gadolinia-Urania burnable absorbers (CENPD-275-P,

Revision 1-P, Supplement 1-P). None of these changes is a

contributor to the initiation of previously evaluated accidents. The

changes to TS bases and the COLR methodology changes have no impact

on the accident initiators. Accordingly, the probability of an

accident previously evaluated is not significantly increased.

The proposed changes have been evaluated by Florida Power &

Light (FPL) and Asea Brown Boveri--Combustion Engineering (ABB-CE).

The safety analyses assumed bounding physics parameters, and satisfy

all the applicable acceptance criteria. Although specification

2.1.1.2 is deleted from TS, the safety analyses continue to meet the

same centerline melt acceptance criteria as before and from which

the peak linear heat rate value is derived. Additionally, the peak

linear heat rate value (corresponding to the centerline melt) does

not meet the criteria specified in 10 CFR 50.36 for safety limits.

The changes to TS bases do not affect safety analysis results.

The relocation of SHUTDOWN MARGIN requirements to COLR does not

affect analysis results or consequences as the limits remain

unchanged. Future changes to these limits will be controlled per

Generic Letter 88-16 under the provisions of 10 CFR 50.59.

The use of HERMITE code in one dimension, for space-time loss-

of-flow simulation, has been successfully applied for other ABB-CE

plants. The use of HERMITE code in this mode, for St. Lucie Unit 2,

is acceptable since there are no fundamental core and nuclear steam

supply system (NSSS) differences between St. Lucie Unit 2 and these

plants. The analyses presented in this submittal include the use of

a supplement to the gadolinia-urania core design methodology topical

report. The change in the rod bow penalty effects similar to that

approved for another ABB-CE plant is justified for St. Lucie Unit 2

based on a comparative analysis of factors influencing the rod bow.

The change in the CEAW analysis method removes unnecessary

conservatisms as compared to the previous analysis method. The

validity of results and conclusions of this evaluation are

contingent upon NRC approval of these revised methods.

The radiological does consequences for applicable safety

analyses, using the dose conversion factors from ICRP-30, Supplement

to Part 1, satisfy the acceptance criteria established to ensure

compliance with the 10 CFR 100 dose limits.

The COLR methodology changes proposed to be listed in TS are

those previously approved for CE plants with changes as described

above. The use of these methodologies remains consistent with their

applicability for safety analyses.

Therefore, the proposed changes do not significantly increase

the probability or consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment involves changes to the Technical

Specifications for the dose conversion factors used in the thyroid

dose calculations, the deletion of TS 2.1.1.2, the replacement of

Bases Figure B2.1-1, and the relocation of SHUTDOWN MARGIN

requirements to the COLR. Additionally, there are methodology

changes related to the safety analyses reported in this submittal.

The methodology changes include the use of HERMITE code in one

dimensional mode for space-time loss-of-flow simulations, revised

rod bow DNB penalty calculations, CEAW analysis methodology change

including the use of delta-T power trip, and

[[Page 6698]]

a supplement to the methodology for core designs containing

Gadolinia-Urania burnable absorbers (CENPD-275-P Revision I-P,

Supplement I-P). None of these changes, including those of the TS

bases, will affect the plant configuration and there will be no

impact on any system performance.

Therefore, this amendment will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The proposed changes to the Technical Specifications have been

evaluated with respect to the safety analyses using either

previously approved methodology or methodology currently under NRC

review (CENPD-275-P, Revision I-P, Supplement

I-P). The use of HERMITE code in one-dimensional mode for spatial

details, for space-time loss-of-flow simulation, provides more accurate

data for thermal margin calculations and has been used for similar

applications at other plants. The calculations of rod bow DNB penalty

using L\2\/I dependence has been previously approved for another ABB-CE

plant and is justified for St. Lucie Unit 2 based on an analysis of

important factors influencing the rod bow. The CEAW methodology change

showed acceptable analysis results after conservatively accounting for

appropriate uncertainties.

The safety analyses performed with this methodology used

bounding physics parameters to allow flexibility for future cycles

core designs. The revised Bases Figure B2.1-1 is consistent with the

attached safety analysis report. Deleting TS 2.1.1.2 is justified

since the specified limit does not meet any of the criteria of 10

CFR 50.36, and the fuel centerline melt criteria applied to the

Specified Acceptable Fuel Design Limit (SAFDL) is not changed. The

setpoint analyses and safety analyses of all design basis accidents

meet the applicable acceptance criteria with respect to the

radiological consequences, SAFDLs, primary and secondary

overpressurization, and 10 CFR 50.46 requirements. The proposed

amendment, therefore, will not involve a significant reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Project Director: Cecil O. Thomas.

Florida Power and Light Company, Et Al., Docket Nos. 50-335 and 50-

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

Date of amendment request: December 16, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specification 6.3, ``Unit Staff Qualifications,'' and

add specific staff qualifications for a Multi-Discipline Supervisor

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) Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed changes are administrative in nature addressing

personnel qualification issues. The Multi-Discipline Supervisor

(MDS) position will be filled with personnel who are experienced in

one or more technical disciplines (maintenance, operations,

engineering, or other related technical discipline). Fundamental

working knowledge of tasks being performed will be acquired through

the MDS initial training program. The training concentrates on

developing the skills and knowledge of an MDS to safely oversee

tasks for multi-discipline work teams. Therefore, four years

experience in any related technical discipline or disciplines

combined with the MDS training program provide adequate technical

knowledge for proper job oversight. These proposed changes will not

involve a significant increase in the probability or consequences of

an accident previously evaluated because they do not affect

assumptions contained in plant safety analyses, the physical design

and/or operation of the plant, nor do they affect Technical

Specifications that preserve safety analysis assumptions. Therefore,

operation of either facility in accordance with its proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

(2) Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The changes being proposed are administrative in nature and do

not affect assumptions contained in plant safety analyses the

physical design and/or modes of plant operation defined in the

facility operating license, or Technical Specifications that

preserve safety analysis assumptions. These changes address

qualification requirements for the MDS position. Since the proposed

changes do not change the qualifications for those individuals

responsible for the actual licensed operation of the facility,

operation of the facility in accordance with the proposed amendments

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

accident from any accident previously evaluated. No new failure mode

is introduced due to the administrative changes since the proposed

changes do not involve the addition or modification of equipment nor

do they alter the design or operation of affected plant systems,

structures, or components. Therefore, operation of either facility

in accordance with its proposed amendment would not create the

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

previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The operating limits and functional capabilities of the affected

systems, structures, and components are unchanged by the proposed

amendments. The proposed changes to add the MDS position have

management and administrative controls associated with the required

qualification requirements. The St. Lucie Unit 1 and Unit 2

Technical Specifications will ensure that any individual filling the

MDS position has the requisite education, experience, and training.

The proposed changes do not alter the basis for any technical

specification that is related to the establishment of, or the

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

either facility in accordance with its proposed amendment would not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

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

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Project Direct: Cecil O. Thomas.

GPU Nuclear Inc. Et Al., Docket No. 50-219, Oyster Creek Nuclear

Generating Station, Ocean County, New Jersey

Date of amendment request: September 3, 1998.

Description of amendment request: The amendment would revise

Technical Specifications 3.4.A.10.e and 3.5.a.2.e to incorporate a

Condensate Storage Tank level of greater than 35 feet.

Basis for proposed no significant hazards consideration

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

[[Page 6699]]

licensee has provided its analysis of the issue of no significant

hazards consideration, which is presented below:

1. The proposed TS change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change does not alter the design or function of any

structures, systems or components and does not affect any of the

parameters or conditions that could contribute to initiation of any

accidents.

The proposed change eliminated an inconsistency between the

noted tank level and required water volume and, thereby, ensures

360,000 gallons of water are available for use. The proposed change

does not affect the volume of water required to be available, the

conditions under which it must be available nor the manner in which

it will be used. Therefore, the proposed TS change does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Eliminating an inconsistency between the noted tank level and

the required water volume does not alter the designs or function of

any structures, systems or components. The proposed tank level

requirement is within the design parameters of the tank and, as

such, does not [ ] introduce any new mechanisms which could

contribute to the creation of a new or different kind of accident

than previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed change eliminates an inconsistency between the

noted tank level and required water volume. The proposed change

ensures that an adequate makeup source is available and, in

addition, that sufficient water volume is available to support

operation of the core spray system in the event of a reactor vessel

leak. Therefore, the proposed TS change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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: William M. Dean.

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

Point Nuclear Station Unit No. 1. (NMP1) Oswego County, New York

Date of amendment request: December 30, 1998.

Description of amendment request: The footnote of current Technical

Specification (TS) Table 3.6.14-2, Radioactive Gaseous Effluent

Monitoring Instrumentation, specifies that the requirement for the

emergency condenser system to have one operable noble gas activity

monitor per vent, is applicable during reactor power operating

conditions. Note (h) of current TS Table 4.6.14-2 specifies that the

requirement to perform a sensor check once per day of the emergency

condenser system noble gas activity monitor is applicable during

reactor power operating conditions. The proposed amendment would change

the footnote of TS Table 3.6.14-2 and note (h) of TS Table 4.6.14-2 to

extend the applicability of the channel operability and daily sensor

check surveillance requirement from during reactor power operating

conditions, to during power operation conditions and whenever the

reactor coolant temperature is greater than 212 deg.F except for

hydrostatic testing with the reactor not critical. The proposed changes

would also correct a clerical error in TS 4.6.15.d. The clerical error

cited an incorrect TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The operation of Nine Mile Point Unit 1, in accordance with

the proposed amendment, will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes extend the application of operability and

daily sensor check for the Emergency Condenser Vent Noble Gas

Activity Monitors to include, in addition to power operations, the

condition when reactor coolant temperature is greater than 212

deg.F, except for hydrostatic testing. These changes will make the

conditions for Emergency Condenser Vent Noble Gas Activity Monitor

operability and daily sensor check surveillance performance

consistent with the conditions for ECS [emergency cooling system]

operability as indicated in LCO [Limiting Condition for Operation]

3.1.3.a.

The proposed changes to the Emergency Condenser Vent Noble Gas

Activity Monitor operability and daily sensor check surveillance

requirements will continue to provide assurance that the intent of

the effluent monitoring requirements of 10 CFR 50 Appendix A, GDC

[General Design Criterion] 64, is satisfied and the radiological

effluents are maintained within the dose and dose rate limits

specified in 10 CFR 50 Appendix I, 10 CFR 20, and the RETS

[Radiological Effluent Technical Specifications]. The proposed

changes will not effect the capability of the ECS to mitigate the

consequences of an accident that results in a loss of feedwater or

reactor isolation from the primary heat sink and aid the Core Spray

System and Automatic Depressurization System in providing effective

core cooling following non-limiting small breaks.

The proposed changes also correct a clerical error in the

Uranium Fuel Cycle effluent monitoring SR [surveillance

requirement]. The proposed correction simply restores the SR to the

form that existed before the error was introduced. The clerical

error did not affect the ODCM [Offsite Dose Calculation Manual]

implementing procedures or plant operation. Thus, the cumulative

dose contribution from Uranium Fuel Cycle sources will continue to

be maintained within the limits of 40 CFR 190 and the RETS.

Based on the above analysis, the proposed changes do not result

in any hardware changes or physical alteration of the plant, and the

changes will have no impact on the design or function of any

structure, system or component (SSC). As such, the SSC process

variables, characteristics, and functional performance will be

maintained consistent with the event initiator and the initial

condition assumptions for the accident analyses. Moreover, the

proposed changes will not eliminate any actions or adversely affect

any SSCs required to prevent accidents or mitigate accident

conditions, nor will the changes result in the degradation of any

fission product barriers so as to increase the radiological

consequences of an accident. It is, therefore, concluded that

operation in accordance with the proposed amendment will not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

2. The operation of Nine Mile Point Unit 1, in accordance with

the proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes do not result in any hardware changes or

physical alteration of the plant, and the changes do not impact the

design or function of any SSC. The proposed changes maintain the

capability of the ECS to respond to accidents, including non-

limiting small breaks, consistent with the current analyses. In

addition, the proposed changes provide continued assurance that the

radiological dose and dose rates will be maintained within limits.

The proposed changes do not alter the process variables,

characteristics, or functional performance of any SSC, do not

eliminate any requirements, and do not impose any new requirements

which could introduce new equipment failure modes or create new

credible accidents. It is, therefore, concluded that operation in

accordance with proposed amendment will not create the possibility

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

evaluated.

3. The operation of Nine Mile Point Unit 1, in accordance with

the proposed

[[Page 6700]]

amendment, will not involve a significant reduction in a margin of

safety.

The proposed changes do not affect the capability of the ECS to

mitigate consequences of an accident that results in a loss of

feedwater or reactor isolation from the primary heat sink, or affect

the capability of the ECS to aid the Core Spray System and the

Automatic Depressurization System in providing effective core

cooling following non-limiting small breaks. Thus, there will be no

impact on the post-accident radioactive material release analyses or

a reduction in the margin to the associated 10 CFR 100 dose limits.

In addition, the proposed changes provide continued assurance that

the intent of the effluent monitoring requirements of 10 CFR 50

Appendix A, GDC 64, is satisfied and the dose and dose rates due to

the radiological effluents are maintained within the limits

specified in 10 CFR 50 Appendix I, 10 CFR 20, 40 CFR 190, and the

RETS. Moreover, the proposed changes do not eliminate any

requirements or responsibilities, nor impose new requirements or

responsibilities, or alter any physical parameters which could

reduce the margin to an acceptance limit. It is, therefore,

concluded that operation in accordance with the proposed amendment

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 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: Reference and Documents

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

York 13126.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &

Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Project Director: S. Singh Bajwa, Director.

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

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

Date of amendment request: December 16, 1998.

Description of amendment request: The proposed editorial and

administrative changes to the Technical Specifications would either

revise references and statements that are inaccurate or provide relief

from administrative controls which provide insignificant safety

benefit.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The design basis accidents are not affected by the proposed

editorial and administrative changes. The proposed changes do not

change the level of programmatic controls or the procedural details

currently in place. The proposed changes do not revise the station

design, the response of the station to transients nor the manner in

which the station is operated, therefore, these changes have no

adverse affect to the safe operation of the station. The proposed

changes do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any previously analyzed.

The proposed changes do not alter the design assumptions,

conditions, configuration of the facility or the manner in which the

plant is operated. There are no changes to the source term,

containment isolation or radiological release assumptions used in

evaluating the radiological consequences in the Seabrook Station

UFSAR. Existing system and component redundancy is not being changed

by the proposed changes. The proposed changes have no adverse affect

on component or system interactions. The proposed changes are

editorial and administrative in nature and do not change the level

of programmatic controls and procedural details associated with the

aforementioned technical specifications. Therefore, since there are

no changes to the design assumptions, conditions, configuration of

the facility, or the manner in which the plant is operated and

surveilled, the proposed changes do not create the possibility of a

new or different kind of accident from any previously analyzed.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

There are no changes being made to the Technical Specification

safety limits or safety system settings that would adversely affect

plant safety. The changes do not affect the operation of structures,

systems or components nor do they introduce administrative changes

to plant procedures that could affect operator response during

normal, abnormal or emergency situations. Therefore, the proposed

changes do not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

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

CT 06141-0270.

NRC Project Director: William M. Dean.

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

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

Date of amendment request: December 16, 1998.

Description of amendment request: The proposed change would

relocate Technical Specifications (TS) 3/4.7.10, ``Area Temperature

Monitoring,'' and associated TS Table 3.7-3, to the Seabrook Station

Technical Requirements Manual.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed change does not adversely affect accident

initiators or precursors nor alter the design assumptions,

conditions, configuration of the facility or the manner in which the

plant is operated. The proposed change does not alter or prevent the

ability of structures, systems, or components (SSCs) to perform

their intended function to mitigate the consequences of an

initiating event within the acceptance limits assumed in the Updated

Final Safety Analysis Report (UFSAR). The proposed change is

administrative in nature and does not decrease the effectiveness of

programmatic controls or the procedural details of assuring

operation of the facility in a safe manner.

The provisions of TS 3/4.7.10 for area temperature monitoring of

the referenced selected areas is neither part of an initial

condition of a design basis accident or transient that either

assumes the failure of or presents a challenge to the integrity of a

fission product barrier, nor is area temperature monitoring relied

upon as a primary success path to mitigate such events. The

provisions for area temperature monitoring is not related to events

that are considered frequent or dominant contributors to plant risk.

Area temperature monitoring is not considered a design feature or an

operating restriction that is an initial condition of a design basis

accident or transient analysis, nor does it provide a function or

actuate any accident mitigation feature in order to mitigate the

consequences of a design basis accident or transient.

Relocating TS 3/4.7.10 to the Technical Requirements Manual will

still provide adequate controls for area temperature in those areas

designated in TS Table 3.7-3. The relocated requirements of TS 3/

4.7.10 to the Technical Requirements Manual will continue to be

administratively controlled in accordance with TS Section 6.0,

``Administrative Controls.''

The Seabrook Station Technical Requirements Manual is a

licensee-controlled

[[Page 6701]]

document which contains certain technical requirements and is the

implementing manual for the Technical Specification Improvement

Program. Changes to these requirements are reviewed and approved in

accordance with Seabrook Station Technical Specifications, Section

6.7, and as outlined in the Technical Requirements Manual.

Specifically, changes to the Technical Requirements require a 10 CFR

50.59 safety evaluation and are reviewed and approved by the Station

Operations Review Committee (SORC) and the Nuclear Safety Audit

Review Committee (NSARC) prior to implementation.

The proposed change will not degrade the ability of systems,

structures and components important to safety to perform their

safety function. The proposed change will not change the response of

any system, structure or component important to safety as described

in the Seabrook Station Updated Final Safety Analysis Report

(UFSAR). Since the plant response to an accident will not change,

there is no change in the potential for an increase in the

consequences of an accident previously analyzed. As such, 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 previously analyzed.

The proposed change does not alter the design assumptions,

conditions, configuration of the facility or the manner in which the

plant is operated. There are no changes to the source term,

containment isolation or radiological release assumptions used in

evaluating the radiological consequences in the Seabrook Station

UFSAR. Existing system and component redundancy is not being changed

by the proposed change. The proposed change has no adverse impact on

component or system interactions. The proposed change will not

adversely degrade the ability of systems, structures and components

important to safety to perform their safety function nor change the

response of any system, structure or component important to safety

as described in the Seabrook Station Updated Final Safety Analysis

Report (UFSAR). The proposed change is administrative in nature and

does not change the level of programmatic controls and procedural

details controls of assuring operation of the facility in a safe

manner. Therefore, since there are no changes to the design

assumptions, conditions, configuration of the facility, or the

manner in which the plant is operated and surveilled, the proposed

change does not create the possibility of a new or different kind of

accident from any previously analyzed.

Future changes to area temperature monitoring requirements will

be reviewed and approved in accordance with Seabrook Station

Technical Specifications, Section 6.7, and as outlined in the

Technical Requirements Manual. Specifically, changes to the

Technical Requirements require a 10 CFR 50.59 safety evaluation and

are reviewed and approved by the Station Operations Review Committee

(SORC) and the Nuclear Safety Audit Review Committee (NSARC) prior

to implementation.

Since the plant response to an accident will not change, there

is no change in the potential for an increase in the consequences of

an accident previously analyzed, nor can it create the possibility

of a new or different kind of accident from any previously

evaluated.

Relocation of the area temperature monitoring requirements to

the Technical Requirements Manual will not create the possibility of

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

3. The proposed change does not involve a significant reduction

in a margin of safety.

There is no adverse impact on equipment design or operation and

there are no changes being made to the Technical Specification

required safety limits or safety system settings that would

adversely affect plant safety. The proposed change is administrative

in nature and does not change the level of programmatic controls and

procedural details associated with area temperature monitoring to

ensure that environmentally qualified equipment will not be exposed

to temperatures beyond that which they were originally qualified.

Future changes to the area temperature monitoring requirements

will be reviewed and approved in accordance with Seabrook Station

Technical Specifications, Section 6.7, and as outlined in the

Technical Requirements Manual. Specifically, changes to the

Technical Requirements require a 10 CFR 50.59 safety evaluation and

are reviewed and approved by the Station Operations Review Committee

(SORC) and the Nuclear Safety Audit Review Committee (NSARC) prior

to implementation.

Relocation of the requirements contained in TS 3/4.7.10 to the

Technical Requirements Manual 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. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

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

CT 06141-0270.

NRC Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO), Et Al., Docket No. 50-

336, Millstone Nuclear Power Station, Unit No. 2, New London

County, Connecticut

Date of amendment request: December 28, 1998.

Description of amendment request: NNECO is proposing to change

Technical Specification 2.2.1, ``Limiting Safety System Settings--

Reactor Trip Setpoints,'' and the associated Bases to reflect revised

loss of normal feedwater (LONF) analyses. An additional Technical

Specification Bases change to the floor value for the thermal margin

low pressure reactor trip is also included. This proposed change is not

related to the revised LONF analyses.

NNECO is also seeking NRC approval to incorporate changes to the

Millstone Unit No. 2 Final Safety Analysis Report (FSAR). The proposed

changes to the FSAR, except the floor value for thermal margin low

pressure reactor trip, are associated with the revised LONF 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:

In accordance with 10 CFR 50.92, NNECO has reviewed the proposed

changes and has concluded that they do not involve a Significant

Hazards Consideration (SHC). The basis for this conclusion is that

the three criteria of 10 CFR 50.92(c) are not compromised. The

proposed changes do not involve an SHC because the changes would

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The analysis of a loss of normal feedwater (LONF) event, as

described in the Millstone Unit No. 2 FSAR Chapters 10 and 14, has

been revised. Certain key assumptions have been changed to ensure

acceptable analysis results. An evaluation of the LONF analyses

changes, and associated Technical Specification changes will be

presented. In addition, an evaluation of an additional non LONF

analyses related Technical Specification Bases and FSAR change is

included.

LONF analyses changes. The LONF analyses, contained in FSAR

Chapters 10 and 14, have been revised using a steam generator liquid

inventory assumption, at the time of reactor trip on low steam

generator water level, that is consistent with the design of the

replacement steam generators. The revised Chapter 10 and 14 LONF

analyses also incorporate a reduction in auxiliary feedwater

delivery rates resulting from a recalculation of the Auxiliary

Feedwater (AFW) System flows. The results of revised analyses

indicate that the analytical limit for the low steam generator water

level reactor trip must be raised to 43% narrow range level from the

current 34% narrow range level. This will result in a change to the

low steam generator water level reactor trip setpoint listed in

Technical Specification 2.2.1.

The revised Chapter 14 LONF analysis will now take credit for

automatic initiation of the motor driven auxiliary feedwater (MDAFW)

pumps. The current Chapter 14 LONF analysis assumes auxiliary

feedwater flow will be initiated 10 minutes after the event. The

Chapter 10 LONF analysis assumption of

[[Page 6702]]

automatic initiation of one MDAFW pump within 4 minutes, after the

low steam generator level AFW actuation setpoint is reached, has not

changed.

To demonstrate that one MDAFW pump delivers sufficient flow to

preclude steam generator dryout, the Chapter 10 LONF analysis will

not take credit for the operation of the steam generator atmospheric

dump valves, instead of the main steam safety valves as in the

current analysis. This new assumption yields lower predicted steam

generator pressures which result in an increase in the delivered AFW

flows.

LONF analyses related technical specification changes. The trip

setpoint and allowable value for the low steam generator water level

reactor trip will be changed to be consistent with the revised LONF

analyses. The revised analyses assume an analytical limit of 43%

narrow range level, instead of the current analytical limit of 34%

narrow range level. The calculation of the trip setpoint, which

includes instrument uncertainty, has determined that the trip

setpoint should be changed from [greater than or equal to] 36.0% to

[greater than or equal to] 48.5%.

The increase in the low steam generator level Reactor Protection

System (RPS) actuation setpoint from [greater than or equal to] 36%

to [greater than or equal to] 48.5% will result in an increase in

the probability of an RPS actuation on low steam generator water

level since the difference between the proposed setpoint and the

normal operating value of steam generator level will decrease. The

proposed actuation setpoint is below the normal operating level of

60 to 75%. Steam generator level is not expected to approach the

actuation setpoint during normal operation. An unexpected plant

event (e.g., loss of main feedwater or difficulty controlling steam

generator level at low power levels) would be necessary for steam

generator level to approach the actuation setpoint. To provide the

operators with advance notice of the steam generator low level

condition, the existing RPS low steam generator water level pretrip

alarm setpoint will be changed to provide approximately the same

margin between pretrip and trip as currently exists (5%). This will

ensure that the pretrip alarm is received prior to reaching the

actual record trip setpoint. Therefore, even though the proposed

change will decrease the margin between the normal operating steam

generator level and the RPS actuation setpoint, this change will not

significantly impact the probability of an RPS actuation on low

steam generator level during normal plant operations. In addition,

the proposed setpoint and allowable value change will ensure a

reactor trip signal is generated at, or before the analytical limit

used in the revised LONF analysis is reached. Therefore, the RPS

will continue to function as designed to mitigate the consequences

of the design basis accidents.

The basis for the steam generator level low reactor trip will be

modified to be consistent with the revised LONF analyses. The

discussion concerning available water inventory and time until

auxiliary feedwater is required will be removed. The proposed change

to the FSAR will include a discussion of the relationship between

the LONF analysis and the need to automatically initiate auxiliary

feedwater flow.

Non LONF analyses related technical specification bases and FSAR

change. This Technical Specification Bases and FSAR change is not

related to the revised LONF analyses.

The basis for the thermal margin low pressure (TMLP) reactor

trip (Technical Specification 2.2.1 Bases) will be modified. The

current basis states that the floor, or minimum value, for this trip

function is set at 1850 psia pounds per square inch absolute]. This

value will be changed to be consistent with instrument uncertainty

calculations that have determined that the floor should be increased

to 1865 psia. The increase in floor value is the result of greater

instrument uncertainties when harsh containment environment

conditions are included.

The increase in the TMLP floor (from 1850 psia to 1865 psia)

could result in an increase in the probability of an RPS actuation

on thermal margin low pressure since the difference between the

proposed floor setpoint and the normal operating value of

pressurizer pressure will decrease. However, the proposed actuation

setpoint is significantly below the normal operating pressure of

approximately 2250 psia. Pressurizer pressure is not expected to

approach the actuation setpoint during normal operation. A

significant plant event (e.g., loss of primary coolant) would be

necessary for a rapid pressure excursion to approach the actuation

setpoint. Since the setpoint change is small, it will not adversely

impact the probability of an RPS actuation on low pressurizer

pressure during normal plant operations. In addition the proposed

change to the floor value will ensure a reactor trip signal is

generated at, or before the analytical limit used in the respective

accident analyses is reached. Therefore, the RPS will continue to

function as designed to mitigate the consequences of the design

basis accidents.

Conclusion. The results of the revised LONF analyses contained

in FSAR Chapters 10 and 14 have concluded that the LONF event does

not result in the violation of the Specified Acceptable Fuel Design

Limits, that the peak pressurizer pressure does not exceed 110% of

the design pressure, that liquid primary coolant is not expelled

through the pressurizer safety valves, and that adequate cooling

water is supplied by the AFW System to prevent steam generator

dryout and allow a safe and orderly plant shutdown. By preventing

steam generator dryout, sufficient removal of decay heat from the

Reactor Coolant System (RCS) will occur, preventing excessive RCS

heatup and pressurization. This will ensure the steam generator

fatigue analysis remains valid, and excessive discharge of primary

coolant through the pressurizer safety valves does not occur.

Therefore, there will be no adverse effect on the consequences of a

LONF event. This is consistent with the acceptance criteria

contained in Standard Review Plan (SRP) 15.2.7, [``Loss of Normal

Feedwater Flow,'' Rev. 1--July 1981]. (Millstone Unit No. 2 is not

an SRP plant.)

The proposed changes do not alter the way any structure, system,

or component functions. The changes in actuation setpoints and

equipment used in the LONF analyses affect equipment important to

the mitigation of design basis accidents. These changes do not

affect any equipment that can cause a design basis accident to

occur. Therefore, the proposed changes do not affect the probability

of occurrence of a previously evaluated accident.

These proposed changes do not alter the way any structure,

system, or component functions. There will be no adverse effect on

any design basis accident previously evaluated, on any equipment

important to safety, or on the radiological consequences of any

design basis accident. Therefore, these proposed changes will not

adversely affect the consequences of a previously evaluated

accident.

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

from any accident previously evaluated.

Results of the proposed LONF analyses have demonstrated that the

Specified Acceptable Fuel Design Limits are not violated, that the

peak pressurizer and steam generator pressures do not exceed 110% of

the design pressure, that liquid primary coolant is not expelled

through the pressurizer safety valves, and that adequate cooling

water is supplied by the AFW System to prevent steam generator

dryout and allow a safe and orderly plant shutdown. Therefore, there

are no new or different types of failures of systems or equipment

important to safety which could cause a new or different type 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 revised FSAR Chapter 14 analysis has concluded that the

steam generator low water level reactor trip setpoint does not

provide sufficient water inventory in the steam generators at the

time of the reactor trip such that auxiliary feedwater flow will not

be required for 10 minutes. This contradicts the current Technical

Specification Basis (Technical Specification 2.2.1) for the steam

generator low water level reactor trip setpoint. Therefore, the

revised analysis reduces the margin of safety as defined in the

Bases of the Millstone Unit No. 2 Technical Specifications. However,

with the proposed changes to increase the low steam generator water

level reactor trip setpoint and taking credit for automatic AFW

System actuation, it has been shown that operation of these systems

can mitigate the LONF event, and ensure plant response is within the

acceptance criteria. Results of the proposed LONF analyses have

demonstrated that the Specified Acceptable Fuel Design Limits are

not violated, that the peak pressurizer and

[[Page 6703]]

steam generator pressures do not exceed 110% of the design pressure,

that liquid primary coolant is not expelled through the pressurizer

safety valves, and that adequate cooling water is supplied by the

AFW System to prevent steam generator dryout and allow a safe and

orderly plant shutdown. Therefore, these proposed changes do not

involve a significant reduction in a margin of safety.

The proposed change to the floor value for the TMLP reactor trip

function is the result of a revision to the instrument loop

uncertainty and setpoint calculations. The proposed change to the

Technical Specification Basis will incorporate the RPS TMLP floor

setpoint change. This change to the TMLP floor will not adversely

affect this function. The TMLP reactor trip function will still

operate as designed. The RPS will continue to function as designed

to mitigate the consequences of design basis accidents. Therefore,

this proposed change does not involve a significant reduction in a

margin of safety.

The NRC has provided guidance concerning the application of

standards in 10 CFR 50.92 by providing certain examples (March 6,

1986, 51 FR 7751) of amendments that are considered not likely to

involve an SHC. The changes proposed herein are not enveloped by any

specific example.

As described above, this License Amendment Request does not

impact the probability of an accident previously evaluated, does not

involve a significant increase in the consequences of an accident

previously evaluated, does not create the possibility of a new or

different kind of accident from any accident previously evaluated,

and does not result in a significant reduction in a margin of

safety. Therefore, NNECO has concluded that the proposed changes do

not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO), Et Al., Docket No. 50-

336, Millstone Nuclear Power Station, Unit No. 2, New London

County, Connecticut

Date of amendment request: January 18, 1999.

Description of amendment request: NNECO is proposing to change

Technical Specification 3.6.1.2, ``Containment Systems--Containment

Leakage.'' The Bases for this Technical Specification and the Final

Safety Analysis Report (FSAR) will also be modified to address the

proposed changes.

The limit for secondary containment bypass leakage specified in

Technical Specification 3.6.1.2.c will be reduced from less than 0.017

La to less than 0.0072 La. This new limit is

consistent with the value of secondary containment bypass leakage used

in the revised off-site and control room dose calculations following a

design basis loss-of-coolant accident (LOCA).

Technical Specification 3.6.1.2.c will be modified by replacing

``identified in Table 3.6-1 as'' with ``that are.'' This will allow

Table 3.6-1 to be removed. The removal of this table from Technical

Specifications and the proposed wording change are consistent with the

guidance contained in Generic Letter (GL) 91-08. It is not necessary to

maintain a list of the secondary containment bypass leakage paths in

Technical Specifications. The Millstone Unit No. 2 FSAR (Section 5.3.4)

provides the necessary information to determine the secondary

containment bypass leakage paths that must be considered to ensure that

the combined leakage rate limit contained in Technical Specification

3.6.1.2.c is met.

Technical Specification 3.6.1.2 Table 3.6-1, ``Secondary

Containment Bypass Leakage Paths,'' will be removed and the phrase

``This Page Intentionally Deleted'' will be added to Page 3/4 6-5.

The Bases for Technical Specification 3.6.1.2 will be modified to

indicate that the Millstone Unit No. 2 FSAR contains a list of the

containment penetrations that have been identified as secondary

containment bypass leakage paths.

FSAR Section 5.3.4, ``Through-Line Leakage Evaluation,'' will be

changed to include the additional secondary containment bypass leakage

paths that have been identified. The criteria used to determine the

secondary containment bypass leakage paths will be modified to be

consistent with the criteria used in the evaluation that identified the

additional leakage paths.

The discussion of the use of a leakage rate of 11 cc/hr for the

control room dose calculations will be modified. The revised control

room dose calculations will assume a total secondary containment bypass

leakage rate consistent with the proposed change to Technical

Specification 3.6.1.2.

As a result of these proposed changes, the calculated off-site and

control room doses following a design basis LOCA will change. The

calculated doses are specified in FSAR Section 14.8.4, ``Radiological

Consequences of the Design Basis Accident.'' A revision to this section

of the FSAR has been submitted to the NRC by the letter dated September

28, 1998. This submittal will be revised to incorporate the proposed

total secondary containment bypass leakage rate and the associated

change to the calculated off-site and control room doses following a

design basis LOCA.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

In accordance with 10 CFR 50.92, NNECO has reviewed the proposed

changes and has concluded that they do not involve a significant

hazards consideration (SHC). The basis for this conclusion is that

the three criteria of 10 CFR 50.92(c) are not compromised. The

proposed changes do not involve an SHC because the changes would

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to lower the limit for secondary containment

bypass leakage, as specified in Technical Specification 3.6.1.2.c,

from [less than] 0.017 La to [less than] 0.072

La will reduce the off-site doses associated with the

design basis LOCA. The proposed change to raise the limit for

secondary containment bypass leakage from 11 cc/hr to [less than]

0.0072 La will increase the dose to the Control Room

Operators following a design basis LOCA. However, the revised off-

site and control room dose calculations, using the proposed combined

secondary containment bypass leakage limit, demonstrate that the

limits of 10 CFR 100 and 10 CFR 50, Appendix A, General Design

Criteria (GDC) 19 are met. In addition, these proposed changes will

result in the use of the same limit for secondary containment bypass

leakage when determining the radiological consequences of a design

basis LOCA.

The proposed wording change to Technical Specification

3.6.1.2.c, and the associated removal of Table 3.6-1, will not

change the requirement to verify total secondary containment bypass

leakage is within the limit assumed in the determination of the

radiological consequences of the design basis LOCA. Control of the

penetrations that have been identified as secondary containment

bypass leakage paths will be maintained by the process used to

change the Millstone Unit No. 2 FSAR. This process ensures that

appropriate changes to the FSAR are evaluated in accordance with 10

CFR 50.59 to determine if NRC approval is required prior to

implementing the change. This process also ensures that the NRC is

informed of FSAR changes via regular

[[Page 6704]]

updates to the FSAR. The removal of Table 3.6-1 from Technical

Specifications and the proposed wording change are consistent with

the guidance contained in GL 91-08.

The identification and addition of more secondary containment

bypass leakage paths to the FSAR will have no impact on the

calculated off-site and control room doses following a design basis

LOCA since the combined leakage through all secondary containment

bypass leakage paths is limited to the proposed value contained in

Technical Specification 3.6.1.2. The addition of bypass leakage

paths does not change the combined leakage limit, which is now used

in the off-site and control room dose calculations.

The Bases for Technical Specification 3.6.1.2 will be modified

to indicate that the Millstone Unit No. 2 FSAR contains a list of

the containment penetrations that have been identified as secondary

containment bypass leakage paths.

The proposed changes do not alter the way any structure, system,

or component functions. These changes do not affect any equipment

that can cause a design basis accident to occur. There will be no

adverse effect on any design basis accident previously evaluated or

on any equipment important to safety. The reduction in the allowable

secondary containment bypass leakage limit will result in a decrease

in the calculated off-site doses associated with the design basis

LOCA. The use of the proposed secondary containment bypass leakage

limit will increase the calculated doses to the Control Room

Operators following a design basis LOCA. However, the calculated

doses meet the criteria of 10 CFR 100 and GDC 19. Therefore, there

will be no significant increase in the probability or consequences

of an accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes 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. Also, the response of the plant and

the operators following these accidents is essentially unaffected by

the change. The criteria used by the plant operators to terminate

containment spray following a design basis LOCA will change from

containment pressure to either time or pressure, whichever requires

longer operation. This will ensure that containment spray remains in

operation long enough to achieve the assumed iodine decontamination.

However, the operator action to terminate containment spray will

remain the same. 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 change to lower the Technical Specification limit

for secondary containment bypass leakage, to remove Table 3.6-1, and

to add more secondary containment bypass leakage paths to the FSAR

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

equipment will continue to function as assumed in the design basis

accident analysis. These changes will ensure that the secondary

containment bypass leakage paths are identified and tested to verify

that the total secondary containment bypass leakage does not exceed

the Technical Specification limit. This will ensure that the

expected off-site and control room doses following a design basis

LOCA are within the limits specified in 10 CFR 100 and GDC 19.

Therefore, there will be no significant reduction in the margin of

safety as defined in the Bases for the Technical Specification

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.

Northeast Nuclear Energy Company (NNECO) Et Al., Docket No. 50-336,

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

Connecticut

Date of amendment request: January 18, 1999.

Description of amendment request: The proposed changes will remove

the Technical Specification related to Hydrogen Purge System from the

Millstone Unit No. 2 Technical Specifications. The proposed changes

affect Technical Specifications 3/4.6.4.3, ``Containment Systems,

Hydrogen Purge System.'' The Bases of the associated Technical

Specification will be modified to address the proposed changes. The

proposed changes will allow the licensee to downgrade the hydrogen

purge system to a non-safety-related system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with 10 CFR 50.92, NNECO has reviewed the proposed

changes and has concluded that they do not involve a Significant

Hazards Consideration (SHC). The basis for this conclusion is that

the three criteria of 10 CFR 50.92(c) are not compromised. The

proposed changes do not involve an SHC because the changes would

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The Hydrogen Purge System provides a backup means to manually

control the hydrogen concentration in containment given the multiple

failure of the redundant, Seismic Category I Hydrogen Recombiner

System. The primary success path for hydr9gen control is the

Hydrogen Recombiner System. The Hydrogen Recombiner System has

redundant trains and is fully qualified to maintain hydrogen control

following a design basis accident. FSAR [Final Safety Analysis

Report] Section 14.8.3.5, ``Radiological Consequences of Purging''

is being removed from the FSAR since it is no longer required. Since

the hydrogen recombiners are fully redundant, it is not necessary to

postulate offsite doses for purge during a design basis accident.

Thus, the deletion of consequences does not represent a change in

the consequences of a design basis event. Therefore, this change

will not significantly increase the probability or consequences of

an accident previously evaluated.

Revision of Index Page VII is an administrative change. The

proposed change to Bases section 3/4.6.4 by deleting reference to

``the purge system'' is required since Technical Specification 3/

4.6.4.3 is being removed. Therefore, these changes will not

significantly increase the probability or consequences of an

accident previously evaluated.

The proposed changes do not alter how any structure, system, or

component functions. There will be no effect on equipment important

to safety. The proposed changes have no effect on any of the design

basis accidents previously evaluated. Therefore, this License

Amendment Request does not impact the probability of an accident

previously evaluated, 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 purge system is a standby purge system which is not in

service during normal operations as a hydrogen purge system (i.e.,

Charcoal Filter Heaters de-energized). Therefore, no new accident is

created either by system unavailability or actuation. The FSAR will

still address the use of the purge system as a backup to the

recombiner system, Revision of Index Page VII is an administrative

change. The proposed change to Bases section 3/4.6.4 by deleting

reference to ``the purge system'' is required since Technical

Specification 3/4.6.4.3 is being removed. Therefore, the proposed

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

of accident from any accident previously evaluated.

[[Page 6705]]

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

The margin of safety is defined in the Bases 3/4.6.4 which

states that the ``hydrogen control systems are consistent with the

recommendations of Regulatory Guide 1.7 * * * ''. Regulatory Guide

1.7 describes methods that would be acceptable in meeting the

standards for a combustible gas control system, 10 CFR 50.44,

``Standards for combustible gas control systems in light-water-

cooled power reactors.'' Regulatory Guide 1.7 acknowledges that

purging is a means of reducing the hydrogen concentration but it

should not be the primary means because of the release of

radioactivity to the environment. The regulatory guide does advise

that there be an ``installed capability for a controlled purge of

the containment atmosphere to aid in cleanup.'' Removal of the

Hydrogen Purge System Technical Specification is consistent with

Regulatory Guide 1.7. Additionally, the capability to purge is still

documented in the FSAR. Revision of Index Page VII is an

administrative change. The proposed change to Bases section 3/4.6.4

by deleting reference to ``the purge system'' is required since

Technical Specification 3/4.6.4.3 is being removed. Therefore, the

proposed changes will not result in a significant reduction in the

margin of safety as defined in the Bases for Technical

Specifications covered in this License Amendment Request.

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 proposed 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.

Northeast Nuclear Energy Company (NNECO), Et Al., Docket No. 50-

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

County, Connecticut

Date of amendment request: January 18, 1999.

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) 3/4.2.2 to be in accordance with

NRC-approved Westinghouse methodologies for the heat flux hot channel

factor--FQ(Z). In addition, the proposed amendment would

make changes to the core operating limits and the analytical methods

used to determine core operating limits contained in Section 6.9.1.6.a

and b, respectively, by adding, modifying, or deleting references.

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 signification hazards

consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with 10

CFR 50.92 and has concluded that the revision does not involve any

Significant Hazards Considerations (SHC). The basis for this

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

satisfied. The proposed Technical Specification revision does not

involve an SHC because the revision would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

To determine any potential impact, the proposed changes to the

TS are grouped into the following two categories.

(a) Changes to Technical Specification 3/4.2.2 ``Heat Flux Hot

Channel Factor--FQ(Z)''

(b) Changes that are not related to the Heat Flux Hot Channel

Factor TS, and are administrative in nature. These include defining

a new core operating limit and deleting, re-numbering, updating and

adding references to analytical methods used to determine core

operating limits in TS 6.9.1.6 ``Core Operating Limit Report

(COLR).['']

With respect to item 1.a changes related to the Heat Flux Hot

Channel Factor, FQ(Z), impact the initial conditions

assumed in the accidents analyzed for MP3 [Millstone Unit 3]. These

initial conditions are power distributions which are consistent with

reactor operation as defined in the TS. The proposed changes to the

Heat Flux Hot Channel Factor TS ensure that proper actions are taken

to maintain peaking factors within the limits assumed in the MP3

accident analysis. The proposed changes are consistent with the NRC

approved Westinghouse methodology for FQ(Z) surveillance.

Changes to the SURVEILLANCE and ACTION statements will not change

the probability of occurrence of any analyzed accidents.

Furthermore, the consequences of analyzed accidents will not change

since the power distribution assumptions will not be challenged by

reactor operation allowed by the Technical Specifications.

With respect to item 1.b the administrative changes to the

Technical Specifications do not affect existing or proposed Limiting

Conditions for Operation (LCO) or SURVEILLANCE REQUIREMENTS.

Therefore, there is no impact on the design basis accidents.

Thus it is concluded that the proposed revision does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

(a) Proposed changes to the Heat Flux Hot Channel Factor, TS 3/

4.2.2 ensure that proper actions are taken to maintain peaking

factors within the limits assumed in the MP3 accident analysis. The

proposed changes are consistent with the NRC approved Westinghouse

methodology for FQ(Z) surveillance. Maintaining safety

analysis assumptions on power distributions cannot be an initiating

event for any design basis accidents and will not create the

possibility of a different type of accident. Therefore the changes

associated with the Heat Flux Hot Channel Factor limiting condition

for operation do not represent a new unanalyzed accident.

(b) Since the administrative changes do not affect plant

operation, the potential for an unanalyzed accident is not created.

No new failure modes are introduced.

Thus, this proposed revision does not create the possibility of

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

3. Involve a significant reduction on the margin of safety.

(a) The proposed changes ensure that FQ(Z), will

remain within the safety analysis assumptions. The LCO limits and

SURVEILLANCE REQUIREMENTS are not altered. Therefore, the impact on

the consequences on the protective boundaries is unchanged. Meeting

the intent of the NRC approved Westinghouse methodology for

FQ(Z), SURVEILLANCE ensures that power distributions

assumed in the accident analysis will not be challenged by reactor

operations allowed by the Technical Specifications. Therefore,

verification of no change in the margin of safety is encompassed by

meeting the power distribution limits assumed in analyzed accidents.

(b) Since the proposed changes do not affect the consequences of

any accident previously analyzed, there is no reduction in the

margin of safety.

Thus it is concluded that the proposed revision does not involve

a significant reduction in the margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve a Significant

Hazard Consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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.

[[Page 6706]]

NRC Project Director: William M. Dean.

Northern States Power Company, Docket No. 50-263, Monticello

Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: December 31, 1998.

Description of amendment request: The proposed amendment would

revise the technical specification (TS) reactor pressure vessel (RPV)

pressure-temperature (P-T) limit curves, delete completed RPV sample

surveillance requirements, delete requirement to withdraw a specimen at

next refueling outage, and remove the standby liquid control system

(SBLC) relief valve setpoint. Associated administrative changes are

also proposed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

RPV P-T curve changes. It is proposed that P-T curves be revised

to accommodate the shift in RTNDT determined using actual

surveillance program data rather than generic data provided in

Regulatory Guide [RG] 1.99 Revision 2 (Radiation Embrittlement of

Reactor Vessel Materials). The new P-T curves will increase the

margins provided in the P-T limit curves against non-ductile failure

of the RPV. Regulatory Guide 1.99 Revision 2 encourages use of plant

specific surveillance data as data becomes available.

Eliminating prescriptive requirements to remove a RPV test

specimen sample at three fourths service life will result in an

overall improvement in the RPV surveillance program since the

limited number of remaining surveillance samples will be removed at

optimum intervals. Therefore, proposed changes will neither

significantly increase the probability or the consequences of an

accident previously evaluated.

RPV surveillance requirements. Deleting completed, one time

surveillance requirements [SRs] of SR section 4.6.B and

incorporating a discussion of the results in the Bases is an

administrative change and has no effect on probability or

consequences of accidents.

SBLC relief valve setpoint testing. The testing requirements of

TS section 4.4.A.2.c are enveloped by the current testing performed

by Monticello's IST [inservice test] Program, which implements ASME

[American Society of Mechanical Engineers] Code Section XI, approved

by 10 CFR 50.55a. The IST program requires all relief valves to be

tested to their nameplate data setpoints. Any modification to a

relief valve's nameplate data is controlled by the plant's

configuration control process which would ensure the requirements of

ASME Section XI are invoked as required by TS section 3.15. The IST

program required by TS 4.15 ensures the SBLC relief valves will be

properly tested for operability. Therefore, revising section

4.4.A.2.c to remove specific setpoints does not increase the

probability or consequences of an accident.

The proposed amendment will not create the possibility of a new

or different kind of accident from any accident previously analyzed.

RPV P-T curve change. Updated RPV P-T limit curves will not

create the possibility of a new or different kind of accident nor

alter operational standards. New limits continue a system of

operating bounds which are in place to prevent damage to reactor

vessels during normal operating conditions including hydrostatic

pressure and leakage testing, and anticipated transients. The

updated P-T curves incorporate the results of RPV surveillance

specimen testing utilizing criteria defined in RG 1.99, Revision 2.

No change is being made to the way the P-T limits provide plant

protection. No new modes of operation are involved. The changes do

not necessitate physical alteration of the plant.

RPV surveillance requirements. Deleting completed, one time

surveillance requirements of section 4.6B and incorporating a

discussion of the results in the Bases is an administrative change

and therefore has no effect on previously analyzed accidents.

SBLC Relief Valve Setpoint Testing. The testing requirements of

TS section 4.4.A.2.c are enveloped by the current testing performed

by Monticello's IST Program, which implements ASME Code Section XI,

approved by 10 CFR 50.55a. The IST program requires all relief

valves to be tested to their nameplate data setpoints. Any

modification to a relief valve's nameplate data is controlled by the

plant's configuration control process which would ensure the

requirements of ASME Section XI are invoked as required by TS

section 3.15. The IST program required by TS 4.15 ensures the SBLC

relief valves will be properly tested for operability. Therefore,

revising section 4.4.A.2.c to remove specific setpoints does not

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

any accident previously analyzed.

The proposed amendment will not involve a significant reduction

in the margin of safety.

RPV P-T curve change. The proposed RPV P-T curve changes are

designed to maintain the recommended safety factors specified in the

ASME Boiler and Pressure Vessel Code, Section III, Appendix G, and

10 CFR Part 50, Appendix G. The revised curves are based on current

NRC guidelines utilizing actual RPV surveillance program tests

results. The proposed changes shift the curves in a slightly more

conservative direction thus maintaining or increasing the previous

margins of safety.

RPV surveillance requirements. Deleting completed, one time

surveillance requirements from Section 4.6.B and incorporating a

discussion of the results in the Bases is an administrative change

and has no effect on any margin of safety.

SBLC relief valve setpoint testing. The testing requirements of

TS section 4.4.A.2.c are enveloped by the current testing performed

by Monticello's IST Program, which implements ASME Code Section XI,

approved by 10 CFR 50.55a. The IST program requires all relief

valves to be tested to their nameplate data setpoints. Any

modification to a relief valve's nameplate data is controlled by the

plant's configuration control process which would ensure the

requirements of ASME Section XI are invoked as required by TS

section 3.15. The IST program required by TS 4.15 ensures the SBLC

relief valves will be properly tested for operability. Therefore,

revising section 4.4.A.2.c to remove specific setpoints will not

reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

NRC Project Director: Cynthia A. Carpenter.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: January 4, 1999.

Description of amendment request: PECO Energy Company (PECO Energy)

is requesting Technical Specifications (TS) changes which will revise

the Administrative Section of TS pertaining to controlled access to

High Radiation Areas, and the reporting dates for the Annual

Occupational Radiation Exposure Report and the Annual Radioactive

Effluent Release Report.

The specific TS changes are as follows:

TS Section 6.12, 6.12.1, and 6.12.2 will be changed to: clarify

requirements; incorporate additional monitoring options (to allow

dosimetry and video monitoring) for entry into high radiation areas;

add the requirement that all individuals entering a high radiation area

have knowledge of the dose rates in the area; and add the requirement

that locked high radiation controls apply to each individual entering

the area.

TS Sections 6.9.1.4, 6.9.1.5(a), and 6.9.1.8 will be changed to:

support changes to the NRC reporting dates;

[[Page 6707]]

reference 10 CFR 20.2206; delete current reporting dates, and correct a

typographical error.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

in the probability or consequences of an accident previously

evaluated.

The changes are administrative in nature and do not impact the

operation, physical configuration, or function of plant equipment or

systems. The changes do not impact the initiators or assumptions, of

analyzed events, nor do they impact mitigation of accidents on

transient events. Therefore, these changes do not increase the

probability of occurrence of consequences of an accident previously

evaluated in the SAR [Safety Analysis Report].

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 are administrative in nature and do not

alter plant configuration, require that new equipment be installed,

alter assumptions made about accidents previously evaluated, or

impact the operation or function of plant equipment. Therefore,

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

kind of accident than previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The proposed changes are administrative in nature and do not

impact any safety assumptions, or potentially reduce any margin of

safety as described in the LGS TS basis. The proposed changes have

no impact on any safety analysis assumptions. Therefore, these

changes do not involve any 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.

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

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

PA 19101.

NRC Project Director: William M. Dean.

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: December 28, 1998.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to permit an increase in the

allowable leak rate for the main steam isolation valves (MSIVs) and to

delete the MSIV Sealing System. The main steam drain lines and the main

condenser would be utilized as an alternate MSIV leakage treatment

method.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to TS Section 3.6.1.2 do not involve a

change to structures, components, or systems that would affect the

probability of an accident previously evaluated in the Hope Creek

Updated Final Safety Analysis Report (UFSAR).

The proposed changes involve eliminating the Main Steam

Isolation Valve (MSIV) Steam Sealing System requirements from the

TS. As described in Section 6.7 of the UFSAR, the MSIV Steam Sealing

System is manually initiated in about 20 minutes following a design

basis Loss of Coolant Accident (LOCA). Since the MSIV Steam Sealing

System is operated only after an accident has occurred, these

proposed changes have no effect on the probability of an accident.

Since MSIV leakage and operation of the MSIV Steam Sealing System

are included in the radiological analysis for the design basis LOCA

as described in Section 15.6.5 of the UFSAR, the proposed changes

will not affect the precursors of other analyzed accidents. Analysis

of the affects of the proposed changes do, however, result in

acceptable radiological consequences for the design basis LOCA

previously evaluated in Section 15.6.5 of the UFSAR.

Hope Creek has an inherent MSIV leakage treatment capability as

discussed below. [Public Service Electric and Gas Company] PSE&G

proposes to use the drain lines associated with the main steam lines

and main turbine condenser as an alternative to the guidance in

Regulatory Guide 1.96, ``Design of Main Steam Isolation Valve

Leakage Control System For Boiling Water Nuclear Power Plants,''

Revision 0, May 1975, for MSIV leakage treatment. If approved, PSE&G

will incorporate this alternative method in the appropriate

operational procedures and Emergency Operating Procedures.

The Boiling Water Reactor Owner's Group (BWROG) has evaluated

the availability of main steam system piping and main condenser

alternate pathways for processing MSIV leakage, and has determined

that the probability of a near coincident LOCA and a seismic event

is much smaller than for other plant safety risks. Accordingly, this

proposed MSIV leakage treatment pathway will be available during and

after a LOCA. Nevertheless, the BWROG has also determined that the

design requirements applied to the Hope Creek main steam system

piping and main condenser contain substantial margin, based on the

original design requirements.

In order to further justify the capability of the main steam

piping and main condenser alternate treatment pathway, the BWROG has

reviewed limited earthquake experience data on the performance of

non-seismically designed piping and condensers during past

earthquakes. As summarized in General Electric (GE) Report, ``BWROG

Report for Increasing MSIV Leakage Rate Limits and Elimination of

Leakage Control Systems,'' NEDC-31858P, Revision 2, submitted to the

[U.S. Nuclear Regulatory Commission] NRC by BWROG letter dated

October 4, 1993, this study concluded that the possibility of a

failure that could cause a loss of steam or condensate in Boiling

Water Reactor (BWR) main steam piping or condensers in the event of

a design basis (i.e., safe shutdown) earthquake is highly unlikely,

and that such a failure would also be contrary to a large body of

historical earthquake experience data, and thus unprecedented.

PSE&G has performed a verification of seismic adequacy of the

Hope Creek main stream piping and main condenser consistent with the

guidelines discussed in NEDC-31858P, Revision 2, to provide

reasonable assurance of the structural integrity of these

components. This evaluation, ``Hope Creek Nuclear Plant Main Steam

Isolation System Alternate Leakage Treatment Pathway Seismic

Evaluation,'' clearly demonstrates that the MSIV leakage treatment

drain pathway meets the intent of 10 CFR 100 Appendix A, with

regards to seismic qualification. Except for the requirement to

establish a proper flow path from the MSIVs to the condenser, the

proposed method is passive and does not require any additional logic

control and interlocks. The method proposed for MSIV leakage

treatment is consistent with the philosophy of protection by

multiple barriers used in containment design for limiting fission

product release to the environment.

A plant-specific radiological analysis has also been performed

in accordance with NEDC-31858P, Revision 2, to assess the effects of

the proposed increase to the allowable MSIV leakage rate in terms of

Main Control Room (MCR) and off-site doses following a postulated

design basis LOCA. This analysis utilizes the hold-up volumes of the

main steam piping and condenser as an alternate method for treating

the MSIV leakage. As discussed earlier, there is reasonable

assurance that the main steam piping and condenser will remain

intact following a design basis earthquake. The radiological

analysis uses standard conservative assumptions for the radiological

source term consistent with Regulatory Guide (RG) 1.3, ``Assumptions

Used for Evaluating the Potential Radiological Consequences of a

Loss-Of-Coolant Accident for Boiling Water Reactor,'' Revision 2,

dated April 1974.

The analysis results demonstrate that dose contributions from

the proposed MSIV leakage rate limit of 200 scfh per steam line, not

to exceed a total of 400 scfh for all four

[[Page 6708]]

main steam lines, and from the proposed deletion of the MSIV Steam

Sealing System, result in an acceptable increase to the LOCA doses

previously evaluated against the regulatory limits for the off-site

doses and MCR doses contained in 10 CFR 100 and 10 CFR 50, Appendix

A, General Design Criterion (GDC) 19, respectively. However, the

calculation methodology for the revised dose exposures were

performed in a manner that included more conservative design basis

assumptions (e.g., inclusion of system response times, and increased

allowable leakage rates) than in the existing Hope Creek licensing

basis.

The whole body doses at the low population zone (LPZ) outer

boundary and MCR increase from about 0.2 rem to 0.6 rem and from

0.04 rem to 0.09 rem, respectively. These increases are not

significant since the revised doses are small fractions of the

regulatory limits of 25 rem and 5 rem, respectively. The associated

whole body dose at the exclusion area outer boundary (EAB) increases

from about 1.3 rem to 2.6 rem, which is well within the regulatory

limit of 25 rem. The revised thyroid dose at the LPZ outer boundary

increases from about 18 rem to 36 rem, which is well within the

regulatory limit of 300 rem. The revised thyroid dose at the EAB

decreases from about 175 rem to 121 rem (due to plate out on the

steam piping and condenser), which is within the regulatory limit of

300 rem. However, the MCR thyroid dose increases from about 0.3 rem

to 5.0 rem, which is well within the regulatory limit of 30 rem.

Additionally, the MCR beta skin dose increases from about 0.9 rem to

1.6, which is well within the regulatory limit of 30 rem.

The resulting revised thyroid doses discussed above are

dominated by the inorganic radioactive iodine fractions of the

accident source term used in this analysis. More than 95% of the

initial radioactive iodine inventory is assumed to be in the form of

inorganic species in accordance with the guidance in Regulatory

Guide 1.3. However, NUREG-1465, ``Accident Source Terms for Light-

Water Nuclear Power Plants,'' identifies that at least 95% of the

iodine entering containment would be in the form of particulate

iodine. Accordingly, the calculated doses discussed above are

considered to be highly conservative relative to realistic

radiological source terms resulting from a postulated LOCA.

In summary, the proposed changes discussed above do not result

in a significant increase in the radiological consequences of a LOCA

when the same assumptions and methods specified in the UFSAR are

used, recognizing that radiological consequences calculated in the

UFSAR and for these proposed changes are significantly higher than

those using more realistic assumptions and methods. Nevertheless,

the calculated off-site and MCR doses resulting from a LOCA remain

well below the regulatory limits. Although the revised LOCA doses

are higher for low MSIV leakage rates, the effectiveness of the

proposed alternate treatment method, even for leakage rates greater

than the proposed increase in the MSIV allowable leak rate, ensures

that off-site and MCR dose limits are not exceeded.

The proposed change to TS Table 3.6.3-1 involves the deletion of

MSIV Steam Sealing valves and associated main steam line drain

valves from the list of primary containment isolation valves. This

proposed change is consistent with the proposed deletion of the MSIV

Steam Sealing System. The MSIV Steam Sealing System lines and main

steam line drain valves that are connected to the main steam piping

will be welded and/or capped closed to assure primary containment

integrity is maintained. The welding and post weld examination

procedures will be in accordance with American Society of Mechanical

Engineers (ASME) Code, Section III requirements. These welds and/or

caps will be periodically tested as part of the Containment

Integrated Leak Rate Test (CILRT). This proposed change does not

involve an increase in the probability of equipment malfunction

previously evaluated in the UFSAR. This proposed change has no

effect on the consequences of an accident since the MSIV Steam

Sealing lines and associated main steam line drain valves will be

welded an/or cap closed, thus assuring that the containment

integrity, isolation, and leak test capability are not compromised.

Therefore, as discussed above, the proposed changes do not

involve a significant increase in the probability or consequences

from any accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

Although the proposed changes will introduce and take credit for

a new level of operational performance for existing plant systems

and components that have not been previously evaluated in the

accident analysis, the affect on this equipment has been evaluated

and found to provide an acceptable level of reliability that will

provide the required level of protection. This conclusion is based

on the evaluation performed in NEDC-31858P, Revision 2, and the

seismic evaluation of the proposed MSIV leakage treatment pathway.

Therefore, reliance on different equipment than previously assumed

to mitigate the consequences of an accident does not create the

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

previously evaluated.

The BWROG evaluated MSIV performance and concluded that MSIV

leakage rates up to 200 scfh per line will not inhibit the

capability and isolation performance of the MSIVs to effectively

isolate the primary containment. Implementation of the proposed

changes will not result in modifications that could adversely impact

the operability of the MSIVs. The LOCA has been analyzed using the

main steam piping and main condenser as a treatment method to

process MSIV leakage at the proposed maximum rate of 200 scfh per

main steam line, not to exceed 400 scfh total for all four main

steam lines. Therefore, the proposed change to increase the allowed

MSIV leakage rate does not create any new or different kind of

accident from any accident previously evaluated.

The proposed change to eliminate the MSIV Steam Sealing System

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

accident from any accident previously evaluated because the removal

of the MSIV Steam Sealing System does not affect any of the

remaining Hope Creek systems, and the LOCA has been re-analyzed

using the proposed alternate method to process MSIV leakage. The

associated proposed change to delete the MSIV Steam Sealing

isolation valves and associated main steam line drain valves from TS

Table 3.6.3-1 does not create the possibility of a new or different

kind of accident, since the affected main steam piping will be

welded and/or capped closed to assure that the primary containment

integrity, isolation, and leak testing capability are not

compromised.

Therefore, as discussed above, the proposed changes do not

create the possibility for any new or different kind of accident

from any accident previously evaluated.

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

in a margin of safety.

The proposed change to TS Section 3.6.1.2 to increase the MSIV

allowable leakage does not involve a significant reduction in the

margin of safety. As discussed in the current Bases for TS Section

3/4.6.1.2, the allowable leak rate limit specified for the MSIVs is

used to quantify a maximum amount of leakage assumed to bypass

primary containment in the LOCA radiological analysis. Accordingly,

results of the re-analysis supporting these proposed changes are

evaluated against the dose limits contained in 10 CFR 100 for the

off-site doses, and 10 CFR 50, Appendix A, GDC 19, for the MCR

doses. As discussed above, sufficient margin relative to the

regulatory limits is maintained even when assumptions and methods

(e.g., RG 1.3) that are considered highly conservative relative to

more realistic assumptions and methods, are used in the analysis.

Results of the radiological analysis demonstrate that the

proposed changes do not involve a significant reduction in the

margin of safety. The whole body doses, in terms of margin of

safety, are insignificantly reduced by 1.6% at the LPZ, 1.0% in the

MCR, and by 5.2% at the EAB. The margin of safety for thyroid doses

is reduced by 6.13% at the LPZ and 15.7% in the MCR, but is actually

increased by 17.3% at the EAB. The margin of safety for beta dose is

insignificantly reduced by 2.4% in the MCR. These reductions in the

margin of safety are not significant since the revised calculated

doses are highly conservative yet remain well below the regulatory

limits, and therefore a substantial margin to the regulatory limits

is maintained.

Furthermore, while the proposed changes will result in a

calculated reduction in the margin of safety, this reduction is not

significant when considering the increased reliability and

capability of the proposed MSIV leakage treatment system. The

resulting revised thyroid doses discussed above are dominated by the

inorganic radioactive iodine fractions of the accident source term

used in this analysis. More than 95% of the initial radioactive

iodine inventory is assumed to be in the form of inorganic species

in accordance with the guidance in Regulatory Guide 1.3. However,

[[Page 6709]]

NUREG-1465, ``Accident Source Terms for Light-Water Nuclear Power

Plants,'' identifies that at least 95% of the iodine entering

containment would be in the form of particulate iodine. Accordingly,

the calculated doses discussed above are considered to be highly

conservative relative to realistic radiological source terms

resulting from a postulated LOCA.

The proposed change to eliminate the MSIV Steam Sealing System

from TS does not reduce the margin of safety. In fact, the overall

margin of safety is increased. The function of this system for MSIV

leakage treatment will be replaced by alternate main steam drain

lines and condenser equipment. This treatment method is effective in

reducing the dose consequences of MSIV leakage over an expanded

operating range compared to the capability of the MSIV Steam Sealing

System and will, thereby, resolve the safety concern that the MSIV

Steam Sealing System will not function at MSIV leakage rates higher

than the Steam Sealing System's design capacity. Except for the

requirement to establish a proper flow path from the MSIVs to the

condenser, the proposed method is passive and does not require any

new logic control and interlocks. This proposed method is consistent

with the philosophy of protection by multiple barriers used in

containment design for limiting fission product release to the

environment. Furthermore, as previously identified, based on the

evaluations discussed in NEDC-31858P, Revision 2, and the seismic

evaluation performed for Hope Creek, the design of the MSIV leakage

treatment pathway meets the intent of the 10 CFR 100, Appendix A,

requirement for seismic qualification. Therefore, the proposed

method is highly reliable and effective for MSIV leakage treatment.

The revised calculated LOCA doses remain within the regulatory

limits for the off-site and the MCR doses. Furthermore, the revised

calculation shows that MSIV leakage rates greater than 200 scfh for

all four main steam lines would not exceed the regulatory limits.

Therefore, the proposed method maintains a margin of safety for

mitigating the radiological consequences of MSIV leakage beyond the

proposed TS leakage rate limit of 200 scfh per main steam line, not

to exceed a total of 400 scfh for all four main steam lines.

The proposed change to delete MSIV Steam Sealing valves from TS

Table 3.6-3-1 [3.6-3-1] does not reduce the margin of safety. Welded

and/or capped closure of the MSIV Steam Sealing lines assures that

the primary containment integrity and leak testing capability are

not compromised. These welds and/or caps will be periodically leak

tested as part of the CILRT. Therefore, the proposed deletion of the

MSIV Steam Sealing System isolation valves does not involve a

reduction in a margin of safety.

Accordingly, based on the above reasons, the proposed changes do

not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Project Director: William M. Dean.

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

Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

Date of amendment request: January 15, 1999.

Description of amendment request: The proposed amendment would

allow a one-time extension of the Technical Specification (TS)

surveillance interval to the end of fuel cycle 13 for certain TS

surveillance requirements (SRs). Specifically, (1) SR 4.3.2.1.3

requires the instrumentation response time and sequence testing of each

engineered safety features actuation system (ESFAS) function at least

once per 18 months, (2) SRs 4.8.2.3.2.f and 4.8.2.5.2.d require that

the 125 volt DC and the 28 volt DC distribution system batteries,

respectively, be capacity service tested at least once per 18 months,

during shutdown, (3) SR 4.8.3.1.a.1.a and 4.8.3.1.a.1.b require a

channel calibration and integrated system functional test for one 4.16

kilovolt reactor coolant pump circuit at least once per 18 months such

that all circuits are tested at least once per 72 months, (4) SR

4.1.2.2.c requires testing to verify that each automatic valve in the

reactivity control system flow path actuate on a safety injection (SI)

test signal at least once per 18 months during shutdown, (50 SRs

4.3.1.1, Table 4.3-1, 4.3.2.1.1, Table 4.3-2, 4.3.3.5, Table 4.3-6, and

4.3.3.7, Table 4.3-11 require, in part, the channel calibration of

pressurizer water level, pressurizer water level-high, and containment

water level-wide range, the manual solid-state protection system (SSPS)

functional input check, and the ESFAS manual initiation channel

functional test every 18 months, (6) SR 4.5.1.d requires testing to

verify each accumulator isolation valve opens automatically on an SI

test signal at least once per 18 months, (7) SR 4.5.2.e.1 requires

testing to verify that each automatic valve in the emergency core

cooling system (ECCS) flow path actuates on an SI test signal at least

once per 18 months, (8) SR 4.7.6.1.d.2 requires the control room

emergency air conditioning system to automatically actuate in the

pressurization mode on an SI test signal or control room intake high

radiation test signal at least once per 18 months, (9) SR 4.7.10.b

requires each automatic valve in the chilled water loop to actuate on

an SI signal at least once per 18 months. Further, SR 4.8.1.1.2.d.7

requires a test to verify that each emergency diesel generator operates

for at least 24 hours every 18 months, and SR 4.8.2.5.2.c.2 requires

that the 125 volt DC battery connections be verified clean, tight, and

coated with anti-corrosion material at least once per 18 months.

Because of the length of the last outage and delays in restart, the SRs

will be overdue prior to reaching the next refueling outage (1R13). The

SRs are to be completed during the 1R13 outage, prior to returning the

unit to Mode 4 (hot shutdown) upon outage completion. The proposed

amendment also make some administrative and editorial changes on some

of the pages that will be affected by above SR interval extensions.

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:

4.3.2.1.3 (Instrumentation, Engineered Safety Feature Actuation System

Instrumentation)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Deferral of the surveillance requirement does not involve any

physical changes to the plant nor does it change the way the plant

is operated. Thus the proposal does not increase the probability of

an accident previously evaluated.

The SEC [safeguard equipment control] automatic self-test

feature, the monthly functional surveillance testing and the

positive surveillance testing history provide sufficient assurance

of the operability of the system. These features also provide

assurance that a degraded condition, if it did occur, would be

detected.

Thus, it is reasonable to conclude that this proposal represents

no significant increase in the consequences of an accident

previously analyzed.

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

or different kind of accident form any accident previously

evaluated.

Deferral of the surveillance requirement does not involve any

physical changes to the plant nor does it change the way the plant

is operated.

Thus, it can be concluded that deferring the surveillance

requirement to the refueling outage cannot create the possibility of

a different kind of accident from any accident previously evaluated.

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3. The proposed change does not involve a significant reduction

in a margin of safety.

Deferral of the surveillance requirement does not involve any

physical changes to the plant nor does it change the way the plant

is operated. The self-test feature and the monthly functional

testing will provide reasonable assurance that the SECs will remain

operable during the few weeks of deferral to the refueling outage.

Also the ability to detect a degraded condition in the SEC will not

be affected during the deferral period.

Therefore, the plant's response to accident conditions during

the period of deferral will not be affected.

Thus, it can be reasonably concluded that this proposal to amend

the Salem Unit 1 Technical Specifications, on a one-time basis, to

defer surveillance requirement 4.3.2.1.3 does not involve a

significant reduction in any margin of safety.

4.8.2.3.2.f, (Electrical Power Systems, 125 Volt D.C. Distribution),

and 4.8.2.5.2.c.2 and 4.8.2.5.2.d (Electrical Power Systems, 28 Volt

D.C. Distribution)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The deferral of the battery service tests to the refueling

outage does not involve any physical changes to the power plant or

to the manner in which the power plant is operated. Therefore, the

probability of an accident previously evaluated is not increased.

Weekly and quarterly testing and performance monitoring by the

system manager along with the current condition of the batteries

(past test results demonstrating above 100% capacity) provide

assurance that battery condition and performance will not

deteriorate during the deferral period. Other positive industry

experience for similar batteries on 24 month cycles also support

this assurance. Therefore, the consequences of a loss of power

accident will not be increased due to the deferral of the

surveillance requirements.

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

or different kind of accident from any accident previously

evaluated.

The deferral of the battery service tests to the refueling

outage does not involve any physical changes to the power plant or

to the manner in which the power plant is operated. No new failure

mechanisms will be introduced by the surveillance deferral.

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 change does not involve a significant reduction

in a margin of safety.

The deferral of the battery service tests to the refueling

outage does not involve any physical changes to the power plant or

to the manner in which the power plant is operated. Continuing

weekly and quarterly testing and performance monitoring along with

the current condition of the batteries provides assurance that

battery condition and performance will be acceptable during the

deferral period and that any degradation that may occur will be

detected. Therefore, the plant's response to accident conditions

during the period of deferral will not be affected.

Thus, it can be reasonably concluded that this proposal to amend

the Salem Unit 1 Technical Specifications, on a one-time basis, to

defer surveillance requirements 4.8.2.3.2.f, 4.8.2.5.2.c.2 and

4.8.2.5.2.d does not involve a significant reduction in any margin

of safety.

4.8.3.1.a.l.a, 4.8.3.1.a.l.b (Electric Power Systems, Electrical

Equipment Protective Devices)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The deferral of inspection, calibration and meggering of 1A, 1B,

1C 460VAC transformer relays and current transformers (CT's); and

inspection, calibration and meggering of 1F 4KV Bus Overload Relays

to the refueling outage does not involve any physical changes to the

power plant or to the manner in which the power plant is operated.

Therefore, the probability of an accident previously evaluated is

not increased.

The condition of the equipment as found for the three most

recent completed surveillances (i.e. no failures or equipment

problems found, no repair actions required, and test results

satisfactory in all cases) provides assurance that equipment

condition and performance will be acceptable during the deferral

period. The subject equipment has performed well over the past

several years and has demonstrated satisfactory stability and

reliability. The plant's response to accident conditions during the

period of deferral will not be affected. Therefore, the consequences

of an accident previously evaluated will not be increased due to the

deferral of the surveillance requirements.

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

or different kind of accident from any accident previously

evaluated.

The deferral of inspection, calibration and meggering of 1A, 1B,

1C 460VAC transformer relays and current transformers (CT's); and

inspection, calibration and meggering of 1F 4KV Bus Overload Relays

to the refueling outage does not involve any physical changes to the

power plant or to the manner in which the power plant is operated.

No new failure mechanisms will be introduced by the surveillance

deferral. 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 change does not involve a significant reduction

in a margin of safety.

The deferral of inspection, calibration and meggering of 1A, 1B,

1C 460VAC transformer relays and current transformers (CT's); and

inspection, calibration and meggering of 1F 4KV Bus Overload Relays

to the refueling outage does not involve any physical changes to the

power plant or to the manner in which the power plant is operated.

The results of previous tests which demonstrate the reliable and

stable operation of the equipment over recent years provides

assurance that the equipment will operate as designed during the

deferral period. The plant's response to accident conditions during

the period of deferral will not be affected.

Thus, it can be reasonably concluded that this proposal to amend

the Salem Unit 1 Technical Specifications, on a one-time basis, to

defer surveillance requirements 4.8.3.1.a.l.a and 4.8.3.1.a.l.b does

not involve a significant reduction in any margin of safety.

4.1.2.2.c (Reactivity Control Systems, Flow Paths--Operating),

4.3.1.1.1, Table 4.3-1 (Reactor Trip System Instrumentation--

Surveillance Requirements); 4.3.2.1.1, Table 4.3-2 (Engineered Safety

Feature Actuation System Instrumentation--Surveillance Requirements);

4.5.1.d (Emergency Core Cooling Systems, Accumulators); 4.5.2.e.1

(Emergency Core Cooling Systems, ECCS Subsystems--Tave [greater than or

equal to] 350 deg.F); 4.7.6.1.d.2 (Plan Systems, Control Room

Emergency Air Conditioning System); and 4.7.10.b (Plant Systems,

Chilled Water System--Auxiliary Building Subsystem)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The deferral of the Manual Safety Injection (SI) surveillance

test to the refueling outage does not involve any physical changes

to the power plant or to the manner in which the power plant is

operated. Therefore, the probability of an accident previously

evaluated is not increased.

Other surveillance testing provides assurance that the equipment

will be reliable during the short deferral period. This testing, in

conjunction with successful previous SI test results assure that the

equipment will function properly during the short deferral period.

Therefore, the consequences of an accident previously evaluated will

not be increased due to the deferral of the surveillance

requirements.

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

or different kind of accident from any accident previously

evaluated.

The deferral of the Manual Safety Injection (SI) surveillance

test to the refueling outage does not involve any physical changes

to the power plant or to the manner in which the power plant is

operated. No new failure mechanisms will be introduced by the

surveillance deferral. 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 change does not involve a significant reduction

in a margin of safety.

The deferral of the Manual Safety Injection (SI) surveillance

test to the refueling outage does not involve any physical changes

to the power plant or to the manner in which the power plant is

operated. Other surveillance testing in conjunction with successful

previous SI test results provides assurance that the equipment will

be reliable during the short deferral period. The plant's response

to accident conditions during the period of deferral will not be

affected.

Thus, it can be concluded that this proposal to amend the Salem

Unit 1 Technical Specifications, on a one-time basis, to defer

surveillance requirements 4.1.2.2.c; 4.3.1.1.1, Table 4.3-1;

4.3.2.1.1, Table 4.3-2;

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4.5.1.d; 4.5.2.e.1; 4.7.6.1.d.2; and 4.7.10.b does not involve a

significant reduction in any margin of safety.

4.8.1.1.2.d.7 (Electrical Power Systems, A.C. Power Sources) Diesel

Generator 24 Hour Endurance Run)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Deferral of performance of the diesel generator 24 hour

endurance runs to 1R13 does not involve any physical changes to the

power plant or to the manner in which the power plant is operated.

Therefore, the probability of an accident previously evaluated is

not increased.

Based of the favorable history for previous endurance runs for

the six Sale Unit 1 & 2 emergency diesel generators, continued

normal monthly surveillance testing and the trending of engine and

generator parameters, diesel generator operability can be assured

during the deferral period. Therefore, the consequences of an

accident previously evaluated will not be increased due to the

deferral of the surveillance requirements.

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

or different kind of accident from any accident previously

evaluated.

Deferral of performance of the diesel generator 24 hour

endurance runs to 1R13. does not involve any physical changes to the

power plant or to the manner in which the power plant is operated.

No new failure mechanisms will be introduced by the surveillance

deferral. 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 change does not involve a significant reduction

in a margin of safety.

Deferral of performance of the diesel generator 24 hour

endurance runs to 1R13 does not involve any physical changes to the

power plant or other manner in which the power plant is operated.

Satisfactory endurance run history, other surveillance testing and

performance monitoring assures diesel generator operability during

the deferral period.

The plant's response to accident conditions during the period of

deferral will not be affected.

Thus, it can be conducted that this proposal to amend the Salem

Unit 1 Technical Specifications, on a one-time basis, to defer

surveillance requirement 4.8.1.1.2.d.7 does not involve a

significant reduction in any margin of safety.

4.3.1.1.1, Table 4.3-1 (Reactor Trip System Instrumentation-

Surveillance Requirements); 4.3.3.5, Table 4.3-6 (Remote Shutdown

Monitoring Instrumentation Surveillance Requirements); 4.3.3.7, Table

4.3-11 (Surveillance Requirements for Accident Monitoring

Instrumentation)

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Deferral of calibration of Pressurizer Level Channel 1, and the

Containment Sump Level devices to 1R13 does not involve any physical

changes to the power plant or to the manner in which the power plant

is operated. Therefore, the probability of an accident previously

evaluated is not increased.

Review of trends of the level channels during the current

operating cycle and continued monitoring of the channels provides

reasonable assurance that the channels will perform their design

function during the deferral period. Therefore, the consequences of

an accident previously evaluated will not be increased due to the

deferral of the surveillance requirements.

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

or different kind of accident from any accident previously

evaluated.

Deferral of calibration of Pressurizer Level Channel 1, and the

Containment Sump Level devices to 1R13 does not involve any physical

changes to the power plant or to the manner in which the power plant

is operated. No new failure mechanisms will be introduced by the

surveillance deferral. 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 change does not involve a significant reduction

in a margin of safety.

Deferral of calibration of Pressurizer Level Channel 1, and the

Containment Sump Level devices to 1R13 does not involve any physical

changes to the power plant or to the manner in which the power plant

is operated. Review of trends of the level channels during the

current operating cycle and continued monitoring provides reasonable

assurance that the channels will perform their design function

during the deferral period. There will be no effect on the response

to accident conditions during the period of deferral.

Thus, it can be concluded that this proposal to amend the Salem

Unit 1 Technical Specifications, one a one-time basis, to defer

surveillance requirements 4.3.1.1.1, Table 4.3-1, item 11; 4.3.3.5,

Table 4.3-6, item 2; and 4.3.3.7, Table 4.3-11, items 4 and 17 does

not involve a significant reduction in any margin of safety.

Administrative and Editorial Change

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes are administrative or editorial and do not

involve any physical changes to the plant. The administrative

changes and editorial changes do not delete any existing

surveillance requirements or delete any requirements from the

Limiting Condition for Operations (LCOs) or Action Statements and

therefore do not reduce the actions that are currently taken to

demonstrate operability of plant structures, systems, or components

(SSCs). The additional surveillance requirement that is being added

including the new surveillance corrects a past administrative error

and should have been incorporated within the Tech Specs as part of

an approved Amendment. This change will provide additional assurance

that SSCs perform their intended safety functions. Surveillance

testing has been and is currently being performed for the

surveillance requirement that should have been incorporated and is

now administratively being added to the Tech Specs. Since these

changes do not modify any SSCs or reduce the current requirements

for demonstrating operability of these SSCs, the proposed changes to

the Tech Specs do not

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