ML021280516

License Amendment Request 02-02, Relocation of Certain Engineered Safety Features Pump Values from Technical Specifications to Technical Requirements Manual - Request 2

ML021280516
Person / Time
Site:	Seabrook
Issue date:	04/24/2002
From:	Feigenbaum T North Atlantic Energy Service Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NYN-02048
Download: ML021280516 (24)

Text

r.'* North Atlantic Energy Service Corporation North P.O. Box 300 Seabrook, NH 03874 Atlantic (603) 474-9521 The Northeast Utilities System April 24, 2002 Docket No. 50-443 NYN-02048 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 Seabrook Station License Amendment Request 02-02, "Relocation of Certain Engineered Safety Features Pump Values From Technical Specifications To The Technical Requirements Manual - Request 2" North Atlantic Energy Service Corporation (North Atlantic) has enclosed herein (Enclosure 1)

License Amendment Request (LAR) 02-02. LAR 02-02 is submitted pursuant to the requirements of 10CFR50.90 and 10CFR50.4.

LAR 02-02 proposes administrative changes to the Seabrook Station Technical Specification (TS) Surveillance Requirements (SR) 4.6.2.1, "Containment Spray System;" and 4.7.1.2.1b, "Auxiliary Feedwater System." In addition, Bases 3/4.7.1.2, "Auxiliary Feedwater System," is revised to provide clarification to current surveillance testing of the steam-driven emergency feedwater pump.

The proposed changes would relocate specific pressure, differential pressure and flow values, as well as specific test methods, associated with certain Engineered Safety Features (ESF) pumps from the Technical Specifications to the Seabrook Station Technical Requirements (SSTR)

Manual. The SSTR is referenced in the Seabrook Station Updated Final Safety Analysis Report and is the implementing manual for the Technical Specification Improvement Program.

Relocation of the specific criteria to the SSTR would afford North Atlantic operational flexibility to revise the criteria, if required (e.g., changes as a result of power uprate or testing to newer versions of the ASME Code), without need for requesting an amendment to the operating license.

The Seabrook Station Technical Requirements Manual (SSTR) is a licensee-controlled 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 6.7.1 .i, and as outlined in the SSTR. Specifically, all changes to the SSTR require an evaluation pursuant to 10 CFR 50.59 and review and approval by the Station Operation Review Committee (SORC) prior to implementation.

U.S. Nuclear Regulatory Commission NYN-02048/ Page 2 The proposed changes are based on the Improved Standard Technical Specifications (ITS),

NUREG-143 1, Revision 2, "Standard Technical Specifications - Westinghouse Plants," which itself is based on the NRC's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (58 FR 39312), issued in July 1993. contains a copy of the pending Technical Requirement.

The Station Operation Review Committee and the Nuclear Safety Audit Review Committee have reviewed LAR 02-02.

North Atlantic has determined that LAR 02-02 meets the criteria of 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement (see Section VI of Enclosure 1).

As discussed in LAR Section IV of Enclosure 1, the proposed change does not involve a significant hazard consideration pursuant to 10CFR50.92. A copy of this letter and the enclosed LAR has been forwarded to the New Hampshire State Liaison Officer pursuant to 10CFR50.91(b).

North Atlantic requests NRC Staff review of License Amendment Request 02-02 and issuance of a license amendment by April 30, 2003, becoming effective immediately and implemented within 60 days thereafter.

Should you have any questions regarding this letter, please contact Mr. James M. Peschel, Manager - Regulatory Programs, at (603) 773-7194.

Very truly yours, NORTH ATLANTIC ENERGY SERVICE CORP.

Ted C. Feigenba

• Executive Vice President and Chief Nuclear Officer cc: H. J. Miller, NRC Region I Administrator R.D. Starkey, NRC Project Manager, Project Directorate 1-2 G. F. Dentel, NRC Senior Resident Inspector Mr. Donald Bliss, Director New Hampshire Office of Emergency Management State Office Park South 107 Pleasant Street Concord, NH 03301

ENCLOSURE 1 TO NYN-02048 North Atlantic SEABROOK STATION UNIT 1 North Atlantic Energy Service Corporation pursuant to 10CFR50.90 submits License Amendment Request 02-02. The following information is enclosed in support of this License Amendment Request:

0 Section I Introduction and Safety Assessment for Proposed Change 0 Section II - Markup of Proposed Change

& Section III - Retype of Proposed Change

• Section IV - Determination of Significant Hazards for Proposed Change

• Section V - Proposed Schedule for License Amendment Issuance and Effectiveness

• Section VI - Environmental Impact Assessment I, Ted C. Feigenbaum, Executive Vice President and Chief Nuclear Officer of North Atlantic Energy Service Corporation hereby affirm that the information and statements contained within this License Amendment Request are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed before me this 6 ,

. L- day of ,2002 Ted C. FeigejjIaum

"-* / / - ý'Notary Public Executive Vice President and Chief Nuclear Officer

Section I Introduction and Safety Assessment for the Proposed Change Page 1

I. INTRODUCTION AND SAFETY ASSESSMENT OF THE PROPOSED CHANGE A. Introduction License Amendment Request (LAR) 02-02 proposes administrative changes to the Seabrook Station Technical Specification (TS) Surveillance Requirements (SR) 4.6.2.1, "Containment Spray System," and 4.7.1.2.1b, "Auxiliary Feedwater System." In addition, Bases 3/4.7.1.2, "Auxiliary Feedwater System,"

is revised to provide clarification to current surveillance testing of the steam-driven emergency feedwater pump.

The proposed changes would relocate specific pressure, differential pressure and flow values, as well as specific test methods, associated with certain Engineered Safety Features (ESF) pumps from the Technical Specifications to the Seabrook Station Technical Requirements (SSTR) Manual. The SSTR is referenced in the Seabrook Station Updated Final Safety Analysis Report and is the implementing manual for the Technical Specification Improvement Program. Relocation of the specific criteria to the SSTR would afford North Atlantic operational flexibility to revise the criteria, if required (e.g., changes as a result of power uprate or testing to newer versions of the ASME Code), without need for requesting an amendment to the operating license.

The Seabrook Station Technical Requirements Manual (SSTR) is a licensee-controlled 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 6.7.1.i, and as outlined in the SSTR. Specifically, all changes to the SSTR require an evaluation pursuant to 10 CFR 50.59 and review and approval by the Station Operation Review Committee (SORC) prior to implementation.

B. Safety Assessment Currently TS SR 4.6.2.1 and 4.7.1.2. lb provide details describing ESF pump acceptance criteria and test methods (e.g., testing on recirculation flow) associated with the performance surveillance test. It is proposed that these details be relocated to the SSTR Manual. These details are not necessary to ensure the operability of the Containment Building Spray (CBS) System and the Emergency Feedwater (EFW)

System. The requirements of the applicable Limiting Condition for Operation (LCO) and the associated Surveillance Requirements for these systems, as well as the definition of OPERABILITY, are adequate to ensure the CBS and EFW systems are maintained operable. As a result, these details are not necessary to ensure the CBS and EFW systems can perform their intended safety function and are not required to be in the TS to provide adequate protection of the public health and safety. The relocation of these details maintains the consistency with NUREG-1431. Any change to these details will be made in accordance with 10 CFR 50.59, as specified in North Atlantic's programs and procedures governing changes to the SSTR Manual.

The proposed changes are based on the improved Standard Technica Specifications (ITS), NUREG 1431, Revision 2, "Standard Technical Specifications - Westinghouse Plants," which itself is based on the NRC's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (58 FR 39312), issued in July 1993.

Page 2

The NRC's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" provided a specific set of four (4) objective criteria to determine which of the design conditions and associated surveillances should be located in the TSs as limiting conditions for operation.

The Final Policy Statement noted that implementation of these additional criteria, as amended to 10 CFR 50.36, may cause some requirements presently in TSs to no longer merit inclusion in TSs.

The specific pump performance verification criteria currently within the Technical Specifications may be removed from TS because they do not meet the four specific criteria in 10 CFR 50.36. Specifically:

"Pumpperformance verification criteria are not considered installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Thus, the specific pump performance verification criteria do not satisfy Criterion 1 (as amended in 10 CFR 50.36) for retention;

"* Pump performance verification criteria are not process variables that are initial conditions of a Design Basis Accident (DBA) or Transient Analysis that assumes either the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the specific pump performance verification criteria do not satisfy Criterion 2 (as amended in 10 CFR 50.36) for retention;

" Pump performance verification criteria are not a structure, system or component (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, the specific pump performance verification criteria do not satisfy Criterion 3 (as amended in 10 CFR 50.36) for retention;

"* Pump performance verification criteria are not considered to be significant risk contributors.

Therefore, the specific pump performance verification criteria do not satisfy Criterion 4 (as amended in 10 CFR 50.36) for retention in the Technical Specifications.

Though it is recognized that proper ESF pump performance is necessary to ensure the safety analysis assumptions remain valid, the specific values for determining proper ESF pump performance need not be contained within the Technical Specification Surveillance Requirement itself. Simply stating within the Surveillance Requirement that pump OPERABILITY must be verified in accordance with a certain Specification (e.g., 4.0.5) and/or criteria contained within a certain Technical Requirement (containing criteria based on the safety analysis) is sufficient to ensure verification of proper ESF pump performance.

In conclusion, the specific details controlled by the subject specifications do not need to be included within the scope of the Technical Specifications. The subject details will be adequately controlled in the Seabrook Station Technical Requirements Manual. The inclusion of the subject details in Technical Specifications is not specifically required by 10 CFR 50.36, or other regulations. Additionally, the activities controlled by the subject specification do not pose a threat to the public health and safety.

Therefore, the proposed changes to the subject Technical Specifications Surveillance Requirements do not affect plant safety.

Page 3

Section II Markup Of Proposed Change The attached markup reflects the currently issued revision of the Technical Specifications and Bases listed below. Pending Technical Specifications or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed markup.

The following Technical Specifications and Bases are included in the attached markups:

Technical Specification Title Page(s)

Specification 4.6.2.1 Containment Spray System 3/4 6-14 Specification 4.7.1.2. lb Auxiliary Feedwater System 3/4 7-4 Bases 3/4.7.1.2 Auxiliary Feedwater System B 3/4 7-2 Page 4

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST* and automatically trans ferring suction to the containment sump.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise 04110,t... secured in osition is in its correct position;

b. By verifyin a n i 1 1 each p ep tested pursuant to Specification 4.0.5;

c. At least once per 18 months during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Prelsure-HI-3 test signal, and

2) Verifying that each spray pump starts automatically on a Containment Pressure-Hi-3 test signal.

d. At least once per 10 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

• In MODE 4, when the Residual Heat Removal System is in operation, an OPERABLE flow path is one that is capable of taking suction from the refueling water storage tank upon being manually realigned.

SEABROOK - UNIT I 3/4 6-14 Amendment No.. )'"

PLANT SYSTEM TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM SURVEI LLANCE REOUI REMENTS 4.7.1.2.la. (Continued)

3) Verifying that valves FW-156 and FW-163 are OPERABLE for alignment of the startup feedwater pump to the emergency feedwater header.

b. At least once per 92 days on a STAdGERED TEST BASIS b

1) 1 Ae motor-driven emergency feedwaterup ev ps discrge pessur 0ps'ijat flow gr o ual a 27; i**r*-* /2) r he steam turbine-driven eu a ur to27 eq~ when the .

0 eJ a~r tan psig.*

"secondar am supply pressu-re ls greater than 500 The provisions of Specification 4.0.4 are not applicable for entry into MODE 3;

3) eify rgpg ,h ssrAd. startup o! greqter feedwateror 0 pump. e elop a q',t _

Sdsch fl ea t anor equa. to S

c. At least once per 18 months during shutdown by:

1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Emergency Feedwater System Actuation test signal;

2) Verifying that each emergency feedwater pump starts as designed automatically upon receipt of an Emergency Feedwater Actuation System test signal;

3) Verifying that with all manual actions, including power source and valve alignment, the startup feedwater pump starts within the required elapsed time; and

4) Verifying that each emergency feedwater control valve closes on receipt of a high flow test signal.

tNi SFLOQ f+ý S."tCLIC Amendment No.3V' SEABROOK - UNIT 1 3/4 7-4

PLANT SYSTEMS BASES 3/4-7.1 TURBINE CYCLE (Continued) 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Coolant System can be cooled Auxiliary Feedwater System ensures that the Reactor down to less than 3500 F from conditions in the event of normal operating a total loss-of-offsite power.

The electric motor-driven ing a total feedwater flow emergency feedwater pump is of 650 capable of deliver trance of the steam generators. gpm at a pressure of 1221 psig The steam-driven emergency to the en capable of delivering a total feedwater pump is 1221 psig to the entrance of feedwater flow of 650 gpm at a pressure of serves as the third auxiliary the steam generators. The startup feedwater pump powered from an emergency bus feedwater pump and can be manually aligned to be of taking suction on the dedicated (Bus 5). The startup feedwater condensate storage tank and emergency feedwater volume pump is capable of 650 gpm at a pressure of 1221 delivering a total feedwater flow of water in the psig to the entrance of the in excess of either the main feedwater header steam generator via feedwater flow path. This or with manual alignment to capacity is sufficient to ensure thatthe emergency water flow is available to adequate feed System temperature to less remove decay heat and reduce the Reactor Coolant than 350OF when the Residual may be placed into operation.o::-- Heat Removal System 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate ume ensures that sufficient storage tank with the minimum water is available water vol of 350*F. The OPERABILITY to cool the RCS to a temperature of water following rupture of the concrete enclosure ensures of this availability missile. The contained water the condensate storage tank by a tornado generated usable because of tank discharge volume limit includes an allowance line location or other physical for water not characteristics.

3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary the resultant offsite radiation Coolant System specific activity dose ensure that 10 CFR Part 100 dose guideline will be limited to a small values in the event of a steam fraction of This dose also includes the line rupture.

tube leak in the steam generator effects of a coincident i gpm reactor-to-secondary of the affected steam line.

consistent with the assumptions These values are used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the more than one steam generatormain steam line isolation valves ensures that no rupture. This restriction will blow down in the event is of a steam line tivity effects of the Reactor required to: (1) minimize the positive reac blowdown, and (2) limit the Coolant System cooldown associated steam line rupture occurs within pressure rise within containment in with the steam isolation valves within containment. The OPERABILITY the event the the closure times of the Surveillance of the main ments are consistent with the Require assumptions used in the safety analyses.

SEABROOK - UNIT 1 a 3/4 7-2

/313V Vi

INSERT An exception to the provisions of Specification 4.0.4 allows deferral of surveillance testing for the steam-driven emergency feedwater pump in order to allow entry into Mode 3 during startup. This is necessary since a steam supply pressure of greater than 500 psig is required to provide the requisite motive force to drive the steam-driven emergency feedwater pump. Once secondary steam supply pressure exceeds 500 psig, a 24-hour time limit is administratively imposed to complete the surveillance testing activities. The 24-hour time limit is in-keeping with Specification 4.0.3 requirements.

SECTION III Retype Of Proposed Change The attached retype reflect the currently issued version of the Technical Specifications. Pending Technical Specification changes or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.

Page 5

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST* and automatically transferring suction to the containment sump.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position;

b. By verifying OPERABILITY of each pump when tested pursuant to Specification 4.0.5;

c. At least once per 18 months during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-Hi-3 test signal, and

2) Verifying that each spray pump starts automatically on a Containment Pressure-Hi-3 test signal.

d. At least once per 10 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

• In MODE 4, when the Residual Heat Removal System is in operation, an OPERABLE flow path is one that is capable of taking suction from the refueling water storage tank upon being manually realigned.

SEABROOK - UNIT 1 3/4 6-14 Amendment No. 30,

PLANT SYSTEMS TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS 4.7.1.2.1a. (Continued)

3) Verifying that valves FW-1 56 and FW-1 63 are OPERABLE for alignment of the startup feedwater pump to the emergency feedwater header.

b. At least once per 92 days on a STAGGERED TEST BASIS by verifying the following pumps develop the required discharge pressure and flow as specified in the Technical Requirements Manual:

1) The motor-driven emergency feedwater pump;

2) The steam turbine-driven emergency feedwater pump when the secondary steam supply pressure is greater than 500 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3;

3) The startup feedwater pump.

c. At least once per 18 months during shutdown by:

1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Emergency Feedwater System Actuation test signal;

2) Verifying that each emergency feedwater pump starts as designed automatically upon receipt of an Emergency Feedwater Actuation System test signal;

3) Verifying that with all manual actions, including power source and valve alignment, the startup feedwater pump starts within the required elapsed time; and

4) Verifying that each emergency feedwater control valve closes on receipt of a high flow test signal.

SEABROOK - UNIT 1 3/4 7-4 Amendment No. ,30,

BASES 3/4.7.1 TURBINE CYCLE (Continued) 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350°F from normal operating conditions in the event of a total loss-of-offsite power.

The electric motor-driven emergency feedwater pump is capable of delivering a total feedwater flow of 650 gpm at a pressure of 1221 psig to the entrance of the steam generators.

The steam-driven emergency feedwater pump is capable of delivering a total feedwater flow of 650 gpm at a pressure of 1221 psig to the entrance of the steam generators. The startup feedwater pump serves as the third auxiliary feedwater pump and can be manually aligned to be powered from an emergency bus (Bus 5). The startup feedwater pump is capable of taking suction on the dedicated emergency feedwater volume of water in the condensate storage tank and delivering a total feedwater flow of in excess of 650 gpm at a pressure of 1221 psig to the entrance of the steam generator via either the main feedwater header or with manual alignment to the emergency feedwater flow path. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350°F when the Residual Heat Removal System may be placed into operation. An exception to the provisions of Specification 4.0.4 allows deferral of surveillance testing for the steam-driven emergency feedwater pump in order to allow entry into Mode 3 during startup. This is necessary since a steam supply pressure of greater than 500 psig is required to provide the requisite motive force to drive the steam-driven emergency feedwater pump. Once secondary steam supply pressure exceeds 500 psig, a 24-hour time limit is administratively imposed to complete the surveillance testing activities. The 24-hour time limit is in-keeping with Specification 4.0.3 requirements.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to cool the RCS to a temperature of 350'F. The OPERABILITY of the concrete enclosure ensures this availability of water following rupture of the condensate storage tank by a tornado generated missile. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 1 gpm reactor-to-secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Requirements are consistent with the assumptions used in the safety analyses.

SEABROOK - UNIT 1 B 3/4 7-2 Amendment No.

Section IV Determination Of No Significant Hazards For The Proposed Change Page 6

IV. DETERMINATION OF NO SIGNIFICANT HAZARDS FOR THE PROPOSED CHANGE License Amendment Request (LAR) 02-02 proposes administrative changes to the Seabrook Station Technical Specification (TS) Surveillance Requirements (SR) 4.6.2.1, "Containment Spray System" and 4.7.1.2.lb, "Auxiliary Feedwater System." In addition, Bases 3/4.7.1.2, "Auxiliary Feedwater System,"

is revised to provide clarification to current surveillance testing of the steam-driven emergency feedwater pump.

The proposed changes would relocate specific pressure, differential pressure and flow values, as well as specific test methods, associated with certain Engineered Safety Features (ESF) pumps from the Technical Specifications to the Seabrook Station Technical Requirements (SSTR) Manual. The SSTR is referenced in the Seabrook Station Updated Final Safety Analysis Report and is the implementing manual for the Technical Specifications Improvement Program. Relocation of the specific criteria to the SSTR would afford North Atlantic operational flexibility to revise the criteria, if required (e.g., changes as a result of power uprate or testing to newer versions of the ASME Code), without need for requesting an amendment to the operating license.

The Seabrook Station Technical Requirements Manual is a licensee-controlled 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 6.7.1 .i, and as outlined in the SSTR. Specifically, all changes to the SSTR require an evaluation pursuant to 10 CFR 50.59 and review and approval by the Station Operation Review Committee (SORC) prior to implementation.

The proposed changes are based on the improved Standard Technical Specifications (ITS), NUREG 1431, Revision 2, "Standard Technical Specifications - Westinghouse Plants," which itself is based on the NRC's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (58 FR 39312), issued in July 1993.

The NRC's "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" provided a specific set of four (4) objective criteria to determine which of the design conditions and associated surveillances should be located in the TSs as limiting conditions for operation.

The Final Policy Statement noted that implementation of these additional criteria, as amended to 10 CFR 50.36, may cause some requirements presently in TSs to no longer merit inclusion in TSs. It has been determined that the specific ESF pump performance verification criteria currently within the Seabrook Station Technical Specifications may be removed from TS because they do not meet the four specific criteria stated in 10 CFR 50.36.

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes to relocate the specific ESF pump pressure and flow criteria in the aforementioned Technical Specifications surveillance requirements to the Seabrook Station Technical Requirements (SSTR) Manual are administrative in nature and do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, configuration of the facility or the manner in which it is operated. The proposed changes do not alter or prevent the ability or structures, systems, or components to perform their intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Seabrook Station Updated Final Safety Analysis Report (UFSAR).

Page 7

The subject surveillance requirement criteria relocated to the Seabrook Station Technical Requirements Manual will continue to be administratively controlled. The SSTR is a licensee controlled 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 6.7.1.i, and as outlined in the SSTR.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not alter the design assumptions, conditions, or configuration of the facility or the manner in which the plant is operated. There are no changes to the source term or radiological release assumptions used in evaluating the radiological consequences in the Seabrook Station UFSAR. The proposed changes have no adverse impact on component or system interactions. The proposed changes 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 UFSAR.

The proposed changes are administrative in nature and do not change the level of programmatic and procedural details of assuring operation of the facility in a safe manner. 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. 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 changes are administrative in nature and do not reduce the level of programmatic or procedural controls associated with the activities presently performed via the aforementioned surveillance requirements.

Future changes to the subject technical requirements will be reviewed and approved in accordance with Seabrook Station Technical Specifications, Section 6.7, and as outlined in North Atlantic's programs. Specifically, changes to the Seabrook Station Technical Requirements Manual require an evaluation pursuant to the provisions of 10 CFR 50.59 and review and approval by the Station Operation Review Committee (SORC) prior to implementation.

Therefore, relocation of the specific pump pressure and flow criteria contained in the aforementioned Technical Specifications Surveillance Requirements to the Seabrook Station Technical Requirements Manual does not involve a significant reduction in the margin of safety provided in the existing specifications.

Based on the above evaluation, North Atlantic concludes that the proposed changes do not constitute a significant hazard.

Page 8

Sections V & VI Proposed Schedule for License Amendment Issuance and Effectiveness and Environmental Impact Assessment Page 9

V. PROPOSED SCHEDULE FOR LICENSE AMENDMENT ISSUANCE AND EFFECTIVENESS North Atlantic requests NRC Staff review of License Amendment Request 02-02 and issuance of a license amendment by April 30, 2003, becoming effective immediately and implemented within 60 days thereafter. The requested date of April 30, 2003 is reflective of typical NRC review time for an administrative license amendment request of this nature.

VI. ENVIRONMENTAL IMPACT ASSESSMENT North Atlantic has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed change does not involve a significant hazards consideration, nor increase the types and amounts of effluent that may be released off-site, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, North Atlantic concludes that the proposed change meets the criteria delineated in 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement.

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ENCLOSURE 2 TO NYN-02048 Technical Requirement 28 ESF Pump OPERABILITY Requirements (Sheet 1 of 2)

LIMITING CONDITION FOR OPERATION TR28-3.1 Each ESF Pump, as listed below, shall be demonstrated OPERABLE when tested in accordance with the Inservice Test Program and/or ASME OM Code per the criteria specified herein.

APPLICABILITY: Whenever the ESF pumps are required to be OPERABLE per the Technical Specification (TS) Surveillance Requirement as tabulated below.

ACTION:

As specified per the applicable Technical Specification.

SURVEILLANCE REQUIREMENTS TR28-4.1 Demonstrate OPERABILITY of each ESF Pump as listed below:

Technical ESF Pump Operability Requirements Specification 4.1.2.3.1

• Centrifugal By verifying, on recirculation flow, that a differential pressure across the Charging pump of greater than or equal to 2480 psid is developed.

4.1.2.4

• Centrifugal By verifying, on recirculation flow, that a differential pressure across each Charging pump of greater than or equal to 2480 psid is developed.

4.5.2f.1)

• Centrifugal By verifying, on recirculation flow, that a differential pressure across each Charging pump of greater than or equal to 2480 psid is developed.

4.5.2f.2)

• Safety Injection By verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 1445 psid is developed.

4.5.2f.3)

• RHR By verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 171 psid is developed.

4.6.2.1b. ** Containment By verifying, on recirculation flow, that a differential pressure across each Spray pump of greater than or equal to 262 psid is developed.

4.7.1.2. 1b. 1) Motor-driven By verifying that the pump develops a discharge pressure of greater than

• 4 .

EFW or equal to 1460 psig at a flow of greater than or equal to 270 gpm.

4.7.1.2.1b.2) Turbine-driven By verifying that the pump develops a discharge pressure of greater than EFW or equal to 1460 psig at a flow of greater than or equal to 270 gpm when the secondary steam supply pressure is greater than 500 psig.

4.7.1.2.1b.3) Startup By verifying that the pump develops a discharge pressure of greater than Feedwater or equal to 1375 psig at a flow of greater than or equal to 425 gpm.

2-28.1 SSTR Rev. Pending

Technical Requirement 28 ESF Pump OPERABILITY Requirements (Sheet 2 of 2)

SURVEILLANCE REQUIREMENTS (continued)

TR28-4.2.* Perform a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1) For centrifugal charging pump lines, with a single pump running:

a. The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 306 gpm, and

b. The total pump flow rate is less than or equal to 549 gpm.

2) For Safety Injection pump lines, with a single pump running:

a. The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 419 gpm, and

b. The total pump flow rate is less than or equal to 669 gpm.

3) For RHR pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 4213 gpm.

BASES TR28-B3/4. Periodic surveillance testing of ESF pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by Technical Specifications which either 1) invokes inservice testing per Specification 4.0.5 pursuant to the requirements of ASME OM Code, and/or 2) require testing to ensure safety analyses criteria continue to be met. Such testing may be accomplished by measuring the pump-developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis.

• Pending approval of LAR 02-01.

• • Pending approval of LAR 02-02.

2-28.2 SSTR Rev. Pending