ML20198E741

Proposed Tech Specs Pages Re LAR-98-15,involving Relocation of TS 3/4.7.10 & Associated TS Table 3.7-3 to Technical Requirements Manual

ML20198E741
Person / Time
Site:	Seabrook
Issue date:	12/16/1998
From:	NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:
Shared Package
ML20198E716	List: ML20198E710 ML20198E729 ML20198E741
References
NUDOCS 9812240136
Download: ML20198E741 (17)

Text

.

INDEX N'

LIMITING CONDITIONS FOR OPERATION AND SURVFill ANCF RF0lllRFMFNTS SECTION PEE 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE' LIMITATION . .

3/4 7-11 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM , . . . . . . . . 3/4 7-12 3/4.7.4 SERVICE WATER SYSTEM . . . . . .. . . . . . . . 3/4 7-13 3/4.7.5 ULTIMATE HEAT SINK . . . . . . . . . . .. .... 3/4 7-14 3/4.7.6 CONTROL ROOM SUBSYSTEriS Emergency Nakeup Air and Filtration ..... ... 3/4 7-16 f

Air Conditionin 3/4.7.7 SNUBBERS . . . . g . . . .............. 3/4 7-18a (/

.................... 3/4 7-19 3/4.7.8 SEALED SOURCE CONTAMINATION ............. 3/4 7-20 3/4.7.9 (This specification number is not used) ........

3/4.7.10 0@EAN 9E#TURE MON 1 mKIN7 M. P'd"Tr*d dytm, tt,*r us") 3/4 7-23 3/4 7-22 TABLE 3.7-3 GRWTEPERATUJE #0Np@-......... .. 3/4 7-24 pn retc wset.s.1mrME$

3/4 B FIFCTRICAI POWER SYSTEMS 3/4.8.1 A.C. SOURCES Operating ............ ......... 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE . . . . . . . . . 3/4 8-10 Shutdown

....................... 3/4 8-11 3/4.8.2 D.C. SOURCES Operating .............. 3/4 8-12 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS ......... 3/4 8-14 Shutdown . . . . . . . . . . . ............ 3/4 8-15 3/4.8.3 ONSITE POWER DISTRIBUTION Operating ...... . . . ........ ... 3/4 8-16 Shutdown . . . . . . . ... ....... 3/4 8-18 Trip Circuit for Inverter I-2A . ..........

3/4 8-19 3/4.8.4 ELECTRICAL EQUIPMENT NOTECTIVE DEVICES A.C. Circuits Inside Primary Containment . . ... . 3/4 8-20 Containment Penetration Conductor Overcurrent Protective Devices and Protective Devices for Class IE Power Sources Connected to Non-Class 1E Circuits .............

Motor-Operated Valves Thermal Overload Protection

.... .. 3/4 8-21

... 3/4 8-24 3/4 9 RFFUFlTNG OPERATIONS m

3/4.9.1 BORON CONCENTRATION 3/4.9.2 INSTRUMENTATION

,....... .. ... 3/4 9-1 3/4.9.3 DECAY TIME . . .

.. . .. . .. 3/4 9-2

... . 3/4 9-3 SEABROOK - UNIT 1 viii Amendment No.

9812240136 981216 DR ADOCK O y3

.- v. -

l. .

. LEX BASES r-

, SECTION TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS . ... . . . . B.

. .3/4

. . 4-11 3/4.4.10 STRUCTURAL INTEGRITY . . . . . . . .. .. . . . . . . B 3/4 4-16 3/4.4.11 REACT 09. COOLANT SYSTEM VENTS . . ... .....

... B 3/4 4-16 3/4.5 FMFRGFNCY CORC C00fING SYSTFMS 3/4.5.1 3/4.5.2 ACCUMULATORS and 3/4.5.3 ECCS SUBSYSTEMS

...................... B 3/4 5-1 B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK .. . ...

. . . . . . . . .B. 3/4

. 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS .

3/4.6.3 B 3/4 6-3 l

CONTAINMENT ISOLATION VALVES . . . ... .. .. ..... ......  !

3/4.6.4 COMBUSTIBLE GAS CONTROL ........ B 3/4 6-3  !

3/4.6.5 . . . . B 3/4 6-3 CONTAINMENT ENCLOSURE BUILDING . . . . .... .... B 3/4 6-4 3/4 7 PIANT SYSTEMS 3/4.7.1 TURBINE CYCLE

.- 3/4.7.2 ..................... B 3/4 7-1 3/4.7.3 STEAM GENERATOR PRESSURE / TEMPERATURE PRIMARY COMPONENT COOLING WATER SYS

... LIMITATION B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM . .. ... .. .. ..TEM B 3/4 7-3

. . . .. ... . . .B .3/4 . 7-3 3/4.7.5 ULTIMATE HEAT SINK . . . . . . . . . . . . . . .... B 3/4 7-3 3/4.7.6 CONTROL ROOK SUBSYST 3/4.7.7 SNUBBERS . . . . . . EMS ................ B 3/4 7-4 3/4.7.B SEALED SOURCE CONTAMINATION.................. B 3/4 7-4 3/4.7.9 LThis specification number is not u .............. B 3/4 7-5

. . . .... B 3/4 7-5 3/4.7.10GRErliarHATURE)OlH UJRWsf . . sed) B 3/4 7-5 3/4 8 t e r .tre n na m .a u m ges u su r m. . Q . . . . . . . . \

FlFCTRICAI-P0kFR SYSTEMS .

3/4.8.1. 3/4.B.2. and 3/4.8.3 A.C. 300RCES. D.C. SOURCES, and 3/4.8.4 ELE CAL QP IV$6EVICES'.'....'. .

. B 3/4.9 REFUFlING OPFRATIONS 3/4.9.1 BORON CONCENTRATION 3/4.9.2 INSTRUMENTATION ... .................. B 3/4 9-1 3/4.9.3 DECAY TIME . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING P ................. ....

B 3/4 9-1 3/4.9.5 COMMUNICATICNS . . . .ENETRATIONS . .... B 3/4 9-1

........... . .... B 3/4 9-1 3/4.9.6 REFUELING MACHINE 3/4.9.7 .................. B 3/4 9-1 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING B 3/4. 9-2 X

SEABROOK - UNIT 1 x1 AmendmentNo.f 99%

PLANT SYSTEMS (Thr #mricd"d 4 "B" 'I & MN9

' 3/4.7.10 PERAIREMONITORINd

/ .

- 7' ~

! TIMIT /

CONDITI R OPERAT

[ ,

10 Th mperatu each ar own in or moref t an 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, r by more an 30*F. le 3.7- hall not e exceeded ,

APPLICABILITY:

/

enever the uipment i an affec d area is OP RABLE. quired to b

' ACTION:

a With one.cr more arop exceedi in Ta a'3.7-3 for,4cre than the tempe ature limit (p shown hours, pr re and sub ft to the Come sion withirV30 days, 5 al Repor hat provi s arsuant recor to J ecificati 6f the cumu) ive

.8.2, a time a e amount f hich thef emperature n the affepted area (s) .

/ exceededlimit th (s)f nd d an anal is to demo strate the c tinued '

)

OPERABIL'.IIV cation 3 0.3 areof the ffected f.equj nt. The rovisions o Specifi-t applicab /

b. Wit.h e or my areas exceeding the aparature 1 it(s) show n Tatsle 3.7-3,by more tha /30 *F, pre 'and submi a Special - ort is renuired by ACTION . above an

^ the araa('s) to with the temp tureithin 4 ho s either re ore limit ( or declare he equip-ment i the affec d area (s) operable.

RVEILLAN REQUIREME#tS

, ,. ,/

4.7.1 he temp ure in epch of the shown in e 3.7 al de p ined to b ithin ity' limit at le t once per 1 ours.

l i

[.

i

^. ,

SEABROOK - UNIT 1 3/4 7-23 i

l pgy go, 99%

(THtt 14ss 4%o(4. (Mr use% TABLE 3.7-3

{

.r ARE[TEMPERATUKNONITORIf AR

/

-TEMPERAT RE LIMIT /(*F)

. Contr Room

2. Cabl Spreadin 75 oom

3. Sw chgear Ro s 4. - Train A 104 i S itchgear R m - Train 104 i 5. attery Roo s - Train .

104 i L

6..

Battery R ms - Trai 97 } i i 7. ECCS Equ ment Vaul 9 Train A ECCS E ipment Vau} -Trainp d4

! 9.

10.

Centr ugal Char Yng Pump R e - Train 04 Cent fugal Char ing. Pump om - Trai B 104

11. EC Equipment ault Sta 104

12. E S Equipmer Vault St ell - Tr 'n A 104

13. CW Pump A a ell - Tr in 8 104

! 14. Cooling To er Switch 104 15 r Room - rain A

1. Cooling T wer Switc ear Room - Train B 1 Cooling ower SW P p Area 4

7. SW Pu 27

j 18. ouse Elec SW P phouse ricalRo El trical Roo - Train 104 i

~

/ 19. SW mp Area - Train 104

20. 104 Di sel Gener or Room - rain A  ?

f21.

22.

esel Gene tor Room FW Pumpho se

' Train B '

12 10

23. Electric 04

2. Electrip'4al Penetrat Penetran ion AreaArea

- rain- TrainA B 100 J$. Fuel Sforage Bui 85 l

/ Pum rea ing Spent uel Poo Cooling 104

/ 26. Mai Steam an eedwater ipe Cha -

27. M n Steam a East 13

28. Feedwate Pipe Ch se - West 1

drogen An yzer Roo 1 l 29. SFW East ipe Chase lectric 4

, Room 104 r

~ .

/

I i

r.

SEABROOK - UNIT 1 3/4 7-24 J

fMeadmee r 40

' 99%

_ _ . . . ._ _ _ _ __ _ . _ _. __ ~. _ _

e .

PLANT SYSTEMS C BASES 3/4.7.7 SNUBBERS (Continued)

Surveillance to demonstrate OPERABIllTY is b ments of an approved inservice inspection program.y performance of the require-Permanent or other exemptions from the surveillance program S. individual snubbers may be granted by the Commission if a justifiable basis for e.emption is presented and, if applicable, snubber life destructive testing was p?rformed to qualify the snubbers for the applicable design conditions at either the com-pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.

The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.

bases for future consideration of snubber service life.These records will provide stat 3/4.7.8 SEAITD SOURCF CONTAMINATION r

The limitations on removable contamination for sources requiring leak testing.

plutonium. including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for This limitation will ensure that leakage from Byproduct. Source, and Special Nuclear Material sources will not exceed allowable intake values.

Il' I

i Sealed sources are classified into three groups according to their use, l

with Surveillance source in that group. Requirements comensurate with the probability of damage to a >

to be tested more often than those which are not.Those sources which are fr I Sealed sources which are continuously enclosed within a shielded mechanism (i.e.. sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4 7.9 (This specification number is not used.)

f

^

3/4.7 10+ N 7FMIERATidlE MON ORING e ar a tempe ture mitsti s ensur that fety-r ated e ipmen wil

! not  !

{ te eratur s. Exp sure t exces vei temp sub cted to emper ures excess f thei enviro - ntal ualifi atto c se a ss of i s OPE ILITY The t ratures .ay de ade eg' pment nd n r ins rument ror o 4.5* . eratur limit includ an al owan 5 Mnac5S b)

h. _

SEABROOK - UNIT 1 B 3/4 7-5 Amendnent No M y

, - .-_a . - . - . . _ _ . . - . . . .-, .. .. . . . . - . .

l SECTION III Retype of Proposed Changes The attached retype reflects the currently issued version of the Technical Specifications. Pending -

Technical Specification changes or Technical Speci3 cation 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.

i s

I 4

1 Page 6

t

j. -INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS _;

p - SECTION.- PAGE l-L 3/4.7.2-  : STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION . . . 3/4 7-11 3/4.7.3 L PRIMARY COMPONENT COOLING WATER SYSTEM . . . . ... 3/4 7-12

-3/4.7.4 SERVICE WATER SYSTEM .. ..... . . . . ... 3/4 7 4 3/4.7.51 ULTIMATE HEAT SINK ........ . . . . . . ... C/4 7-14  ;

, 3/4.7.6 CONTROL ROOM SUBSYSTEM L Emergeny Makeup Air and. Filtration . . . . . . ... 3/4 7-16 Air Conditioning ..... . . . . . . ... 3/4 7-18a

' 3/4'.7.7 SNUBBERS ............. . . . . . . . .. 3/4 7-19

.3/4.7.8 SEALED SOURCE' CONTAMINATION . . . . . . . . . . . . 3/4 7-20 3/4.7.9 -(THIS SPECIFICATION NUMBER IS.NOT USED) . . . . . .. 3/4 7-22 ,

l 3/4.7.10 (THIS SPECIFICATION NUMBER IS NOT USED) . . . . . . .. -3/4 7-23 . .

L TABLE 3.7-3 .(THIS TABLE NUMBER IS NOT USED) . . . . . . .. 3/4 7-24 '

~ 3/4.8 ELECTRICAL POWER SYSTEMS'

!! 3/4.8.1 -A.C. SOURCES f Operating . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-1

~

[ .

TABLE 4 8-1 DIESEL GENERATOR lEST SCHEDULE

. . . . . . . . . .. 3/4 8-10 L' Shutdown . . . . . . . . . . . . . . . . . . . . . . 3/4 8-11

. 3/4.8.2 .D.C.-SOURCES- i

' Operating . . . . . . . . . . . . . . . . . . . .. ' 3/4 8-12 j TABLE 4.8-2 BATTERY SURVEILLANCE RE0VIREMENiS

. . . . . . . . . .. 3/4 8-14 1

' 1 Shutdown ............ . . . . . . . . .. 3/4 8-15 3/4.8.3 ONSITE POWER DISTRIBUTION 0)erating ................... .. 3/4 8-16

Slutdown. . . . . . . . . . . . . . . . . . . . .. 3/4 8-18 l Trip Circuit for Inverter I-2A ..........., 3/4 8-19 i

!- 3/4'.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES P

!- A.C. Circuits-Inside Primary Containment . . . . . .. 3/4 8-20 I Containment Penetration Conductor Overcurrent

. Protective Devices and Protective Devices for Class 1E Power Sources Connected to Non-Class t 1E Circuits . . . . . . . . . . . . . . . . . .. 3/4 8-21 l . Motor-0perated Valves Thermal Overload Protection . . . 3/4 8-24 y

.3/4.9 REFUELING OPERATIONS.

3/4.9.1- BORON CONCENTRATION . . . . . . . . . . . . . . . . . . 3/4 9-1

'i 3/4.9.2 INSTRUMENTATION ............. . . .. 3/4.9-2 .

! 3/4.9.3 DECAY TIME . . . . . . . . ... . . . . . . . .. 3/4 9-3 1

I SEABROOK - UNIT l' viii Amendment No. M.

F 6

1. 5

, _..cy.-r.- ~_ .m... , . - , . ng -

, ,.. - ,._,-~--u,_.-m.- , -

INDEX

' BASES SECTION A EAGE TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS . . . . . . B 3/4'4-11 3/4.4.10 STRUCTURAL INTEGRITY . . . . . . . . . . . . . . . . B 3/4 4-16

.3/4.4.11 REACTOR COOLANT SYSTEM VENTS . . . .. ... B 3/4 4-16 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.-1 ACCUMULATORS ........ ........ .. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS . . . . . . .. . B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK .. .... .. B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT . . . . . . . . . ... ... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ... ... B 3/4 6-3

' 3/4.6.3 CONTAINMENT ISOLATION VALVES .. . . . .. B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL , . . . . . . . . . . . . . . B 3/4 6-3 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING . ...... .. B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE . . . . . . . . . . . . . . . . . . . . B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION .. B 3/4 7-3 3/4 7.3 PRIMARY COMPONENT COOLING WATER SYSTEM .. ... B.3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM ................ B 3/4 i-3 3/4.7.5 ULTIMATE HEAT SINK .. ......... .... B 3/4 7-3 3/4.7.6 CONTROL ROOM SUBSYSTEMS .... ..... . B 3/4 7-4 3/4.7.7 SNUBBERS ......... .... .... . B 3/4 7-4 3/4.7.8 SEALED SOURCE CONTAMINATION . . . . . . . . . ... B 3/4 7-5 3/4.7.9 (THIS SPECIFICATION NUMBER IS NOT USED) . .. ... B 3/4 7-5 3/4.7.10 -(THIS SPECIFICATION NUMBER IS NOT USED) . . ..... B 3/4 7-5 [

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2. and 3/4.8.3 A.C. SOURCES. D.C. SOURCES and DNSITE POWER DISTRIBUTION . . . . . . . . . . . . . B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES . . .. ... B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION .. .... . ...... B 3/4 9-1 3/4.9.2 INSTRUMENTATION . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4. 9. 3 -- DECAY TIME ............... ..... B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS . . . . . .... B 3/4 9-1 3/4.9.5 COMMUNICATIONS .......... ........ B 3/4 9-1 3/4.9.6 REFUELING MACHINE . . . . . . . . . . . . . .... B 3/4 9-1

- 3/4.9.7- CRANE TRAVEL SPENT FUEL STORAGE BUILDING ..... B 3/4 9-2 L

SEABROOK -- UNIT 1 xi Amendment 66- '

- . . . - . . - . . - . . = . .

l* . ,

l PLANT SYSTEMS 3/4.7.10 (THIS SPECIFICATION NUMBER IS NOT USED) 1 l '

I i

1 l

1 SEASR00K.-' UNIT 1 3/4 7-23 Amendment No.

. j TABLE 3.7-3 (THIS TABLE NUMBER IS NOT USED) 1 l

l l

l 1

l l

)

l r

a-SEABROOK - UNIT 1 3/4 7-24 Amendment No M,66, 4

PLANT SYSTEMS BASES 3/4.7.7 SNUBBERS (Continued)

Surveillance to demonstrate OPERABILITY is by performance of the require-ments of an approved inservice inspection program.

Permanent or other exemptions from the surveillance 3rogram for individuai snubbers may be granted by the Commission if a justifia)le basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either *e com-pletion of their fabrication or at a subsequent date. Snubbers so exempced shall be listed in the list of individual snubbers indicating the extent of the exemptions.

The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.

3/4.7.8 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct. Source, and Special Nuclear Material sources will not exceed allowable Mtake values.

Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed .tources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.7 9 (THIS SPECIFICATION NUMBER IS NOT USED.)

3/4.7.10 (THIS SPECIFICATION NUMBER IS NOT USED.)

l SEABROOK - UNIT 1 B 3/4 7-5 Amendment No. M , 66,

i i

l l

Section IV l

Deterrnination of Significant IIazards for Proposed Change l

Page 7

~ - .- . .- -. .. - - .- --

IV. ' DETERMINATION OF SIGNIFICANT IIAZARDS FOR PROPOSED CIIANGES License Amendment request (LAR) 9815 proposes a change to the Seabrook Station Technical Speci6 cations on the basis of the NRC's " Final Policy Statement on Technical Speci6 cations Improvements for Nuclear Power Reactors"(58 FR 39312), issued in July 1993, and; the results of the l NRC Staff review of WCAP-11618, " Methodically Engineered, Restructured and Improved Technical Specifications, Merits Program - Phase II Task 5, Criteria Application," including Addendum 1, as documented in a letter dated May 9,1988 to R. A. Newton. Chairman of the Westinghouse Owners Group (generically referred as the NRC STS Split Report). The proposed change involves the relocation of Technical Specification (TS) 3/4.7.10, Area Temperature Monitoring," and associated TS Table 3.7-3, to the Seabrook Station Technical Requirements Manual (SSTR). The SSTR is referenced in the Seabrook Station Updated Final Safety Analysis Report and is the implementing manual for the Technical Specifications Improvement Program referenced in Section 6.7 of the Technical Specifications.

Note: The SSTR does not provide provision for incorporation of bases information, therefore, the associated Bases for TS 3/4.7.10 will be deleted and will not be incorporated into the SSTR.

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 NRC STS Split Report, Appendix B, Table 2," Westinghouse Standard Technical Specifications LCOs Which May Be Relocr.ted," specifically references the Area Temperature Monitoring LCO as meriting relocation (to a licensee-controlled document).

l In accordance with 10 CFR 50.92, Nortn Atlantic has reviewed the attached proposed change and has l concluded that it does not involve a significant hazards concideration (SHC). The basis for the l conclusion that the proposed change does not involve a SHC is as follows:

1. The proposed chana,e does not involve a significant increase in the probability or consequences of an aeddent previously evaluated.

The proposed change does not adversely affect accident initiators or precursors nor alter the j design assumptions, conditions, configuration of the facility or the manner in which the plant is I

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 i initiating event within the acceptance limits assumed in the Updated Final Safety Analysis l Report (UFSAR). The proposed change is administrative in nature and does not decrease the

! effbetiveness of programmatic contiols or the procedural details of assuring operation of the facility in a safe manner.

d 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 4 failure of or presents a challenge to the integrity of a fission product barrier, nor is area l 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 Page 8 4

w

a l

l 1

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 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 noi create the possibility of a new or different kind of accident from any previourly 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 Uf'3AR. P 'ng system and component redundancy is not being changed by the proposed change. The e ., posed 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 ftmetion 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, co-figuration of the facility, or the manner in which the plant is operated and surveilled, the preposed change does not create the possibility of a new or different kind of accident from any previously analyzed.

l Page 9 1

9

s ,

l l

l Future changes to area temperature monitoring requirements will be reviewed and approved in c.ccordance 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. )

l There is no adverse impact on equipment design or operation and there are no changes being j made to the Technical Specification required safety limits or safety system settings that would l adversely affect plant safety. The proposed change is administrative in nature and does not

! change the level of programmatic controls and procedural details asseciated with area j l temperature monitoring to ensure that environmentally qualified equipment will not be exposed l to temperatures beyond that which they were originally qualified.

Future changes to the area temperature monitoring requirements will be leviewed 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.

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

i i

t i

Page 10 j

i

6 s o Sections V & VI Proposed Schedule for License Amendment Issuance and Effectiveness and EnvironmentalImpact Assessment l

l l

l l

l Page1i

s 4 l

V.

• PROPOSED SCIIEDULE FOR LICENSE AMENDMENTISSUANCE AND EFFECTIVENESS North Atlantic requests NRC Staff review of License Amendment Request 98-15 and issuance of a license amendment by June 15, 1999, having immediate effectiveness and implementation required within 90 days.

VI. ENVIRONMENTAL IMPACT ASSESSMENT North Atlantic has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed changes do not involve a significant hazards consideration, nor increase the types and amounts of effluent that may be released offsite, 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.

Page1 L

_