ML20114A924
| ML20114A924 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 08/17/1992 |
| From: | PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | |
| Shared Package | |
| ML20114A915 | List: |
| References | |
| NUDOCS 9208240110 | |
| Download: ML20114A924 (23) | |
Text
{{#Wiki_filter:3 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTI0_f.! PAGE 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-3 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM.<..................... 3/4 12-5 3.0/4.0 BASES 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION
. CONTR0L.......................................... B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS................................... B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R........................ B 3/4 2-2 3/4.2.4 QUADRANT POWER TI LT RATI0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-3 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-4 hj 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION..................... .......... B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION.................... ......... B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION............. B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-1 3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-2 3/4.4.6 REACT 0P COOLANT SYSTEM LEAKAGE................... ........ B 3/4 4-3 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-7 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE.................................. B 3/4 4-9 FIGURE B 3/4.4 EFFECT OF--R#ENGE-AND-GBPPE&GOMTENT ON 5"!FT
-Of--RT"# FOR-REAGTOR-VESSEt+-EXP05E&-TO-5509 . . . . . . . .B. .3/4 . 4-10
( T J u Q a w as,- m . n w n ) 4.) SEABROOK - UNIT 1 x 9208240110 920817 3 PDR ADOCK 0500
Controlling material: Base metal
- Copper content: Consemtively asumed-te--k+10-WiF-factual- 0.0G> We/
/0 . untent-=-Or0frNT%P S$h RT initial: 40*F D NDT 1/4T, HOSP los
- F RTNOT af ter16.1 46- EFPY: 3/4T ,89^F- g op Curve applicable for heatup rates up to 60 F/hr for the service period uo to 46-EFPY and contains margins of 10*F and 60 psig for possible instrument er ~
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800 , I . TEMPERATURE (255 F)7g II I I I[ II III FOR THE SERVICE PERIOD 11_ 600- ! / I iI UP TO 4G EFPY iI __ _/ ' l l 1l (fllLi 3 !Ii 400 ' II! l I I l lll li ! I Illi l i I i Ill 11 I iiIl I
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lll 11 Illi 0 . . 100 200 300 35e 400 500 RCS ' TEMPERATURE (*F) (10 F PER DIVISION) FIGURE 3.4-2
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REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO M EFPY (u673 SEABROOK - UNIT 1 3/4 4-31 i
MATER 8AL PROPERTY BASIS
. Controlling material: Bat.e metal Copper content: -40 servat 4v4y-awteed-tMe440-WT%-{ec4ue o.oc C4 h*5?e --eentent-a-M6-4T%)
RT NDT initial: 40 F RT af ter -it EFPY: 1/4T,1199 f os 'F NDT 11.\ 3/4T, W f W F Curve applicable for cooldown rates up to "r/hr for the service period up to 46 FPY nd contains margins of 10 F ano 90 psig for possible instrument error {s f f, j llli l' l l 2600 --- I I I 2400 , 1, !IlI l 2200 -C00;L2QWN LIMLT5- / ' ll
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l 400 a ,.*y II ' g j; ____ 200 l l ~ g 'l l lill Il 100 200 300 ste 400 RCs , TEMPERATURE (F) 00 F PER DIVISION) FIGURE 3,4-3 l).1
.c REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO M EFPY tl:~
SEABROOK - UNIT 1 , 3/4 4-32 u ...4__y_.
R i 1ACTORCOOLANTSYSTEM l 9 BASES 4), - 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in che ASME Boiler and Pressure Vessel Code, Section III, Appendix G:
- 1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
- a. Allowable combinations of pressure and temperature for specific
, temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those pre-sented may be-obtained by interpolation; and b '. Figures 3.4-2 and 3.4-3 defint limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates tnat can be achieved over certain pressure-tecperature ranges.
- 2. These limit lines shall be calculated periodically using methods pro-
.- vide'd below, b '
- 3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70 F,
- 4. The pressurizer heatup and cooldown rates shall not exceed 100"F/h and 200 F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 F, and
- 5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance witt) the requirements of ASME Boiler and Pretsure Vesse d estinn L ople, ory GWJ I.4%=Rms ,io A Q3 The fracture toughness propertiIihsf7MTeWi em T'al a actor vessel are determined in accordance with the NRC Standar44v4ct. % n, ^5TM
. 41M4h and in accordance with-additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1972 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, "Baf' for Heat"p and Ceoldown Limit Curves," April 1975.
! Heatup and cooldown limit curves are calculated using the most limiti
*alue of the nil-ductility reference temperature, RTNDT, at the end of.l e fec-tive full power years-(EFPY) of service life. TheM FPY service life period d
. c. rn mr l SEABROOK - UNIT 1 B 3/4 4-7 L .. . - . .-- - .
I REACTOR COOLANY SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) is chosen such that the limiting RT NOT at the 1/4T location in the core region is greater than the RTNOT of the limiting unirradiated material. The selection of such a limiting RT NDT assures that all components in the Renctor Coolant System will be operated conservatively in accordance with applicable Code ! requirements. l
^
l The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor opera-ty tion ano resultant fast neutron (E greater than 1 MeV) irradiation can cause g , an djuste refer. ce peraw based an' increase in the RTNOT.fiberef d rupon th luent coppercontkt,an osp us co (nt of e rial 1 . qh stion, nb using 3/4. and th larges va of AR comp t,,ed e b edictid,kegJla%igure\ ther tory Guid K 99, isio - "Ef of
' Resid Element on edicted R iat Damagi o Re or Ves.1 Mat i s ,"
\ the W ngho Copper s Trend rves(s n in . ure /4.4-2. The tp an cooldo s imit .4-2 n.d 3.4- ciud redict adjus ment or thi sshift 1rvesof{aigure T justme ( N thee4ofIb(FPYasw11as
) poss ile er b in t NDT ressu ai.d th rature sensin- nstrum ts.
\
lue RT etermin)4 in th'i manne y (edunt the res ts N
, fr the ateriAsurveb'lncephram,evauated gordin o ASTM 5, are aval able. Cap!,ulb will b remove n acc ance w the r 4 irement of '
'AcTM El -7? 10 'Part 0 Appe x H. Y lead 4ctor re I tb, tion p bqeen t fastnbtronfl densi at theh cetiono(esents the ca sule\ d t e in the rea'b oc vesst . 'i h fore, t resul obtained hoom t (surh{neNwall 11 ante spec im can used t redic future iation
- mage 1 he reactor V ssel ateria y urin he les actor the w hdrawal me f thexqaps e. T heat no coo n cur at must e recalc ted when th .AR T deyrmine. fro the su iliance sule h ceeds calcul (ed e f
ART NOTNortheehyivaletcapsuletaiationexesure.
. Allowable pressure-temperature relationships for varices heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as' required by Appenoix G '
to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A. The general method for calculating heatup and cool'down limit curves is based-upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures, a semielliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T p VM SEABROOK - UNIT 1 B 3/4 4-8 11 9! + _ 1 1 A ,, ,F -, . , _ - .
g p:J Therefore, an adjusted reference temperature, based upon the fluence, copper content, and nickel content of the mater.*1 in question, can be predicted using Figure B 3/4.4-1 and the value of ARTun computed by Regulatory ijuide 1.99, Revision 2. ' Radiation Embrittlement of Reactor Vessel Materials.' The heatup and cooldown limit curves of Figures 3.4-2 and .i.4-3 include predicted adjustments for this shif t in RTyg at the end of 11.1' EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments. Values of ARTyn determined in this manner may bt used until the results from the material surveillance program.. evaluated according to ASTH E185, are available. Capsules will be removed in accordance with the 3 requirenento of ASTM E185-73 and 10CFR50, Appendix H. The lead factor represents the relationship between the fast neutron flux density at the location of the capsule. and the inner wall l of tne reactor vessel. Therefore, the results obtained from the surveillance specimens can be used
+o predict future radiation damage to the reactor vessel material by using the-lead factor and-the withdrawal timo of-the capsule. Evaluation of surveillance .. capsule _ data will be conducted in accordance with NRC Regulatory Guide 1.99, Revision 2.
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III. RetYDe 6f Proposed ChanRtp, I See attached retype of proposed changes to Technical Specifications. The attached ratype reflects the currently issued version of Tachnical Specifications. Pending Technical Specification changes or Technical Specification changes issued subsequent to this submittal are not reflected i in the enclosed retype. The enclosed retype shauld be checked for i continuity with Technical Specifications prior to issuance. Revision bars are provided in the right hand margin to designate a change , in the text. No revision bars are utilized when the page is changed solely A to accommodate the shifting of text due to additions or deletiors, t t
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_ _ _ _ - . _ - _ . . . _ _ -. _ _. -_ _ _ _ _ _ _ _ _ _ _ _ - .~___ _ _
..- .- INDEX LIMITING' CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS S.CCTION PAGE 3/4.12.2 LAND USE CENSUS. ................... 3/4 12-3 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM . . . . . . . . . . .- 3/4 12-5 3.0/4J. BASES l 3/4.0 APPLICABILITY. . . . . . . . . . . . . . . . . . . . . . . B 3/4 u l l
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORAT10N CONTROL . . . . . . . . . . . . . . . . . . . . B 3/4 1-1 3/4.1.2 BORAT10N SYSTEMS . . . . . ._. . . . . . . . . . . . . . B 3/4 1-2 1 3/4.1.3 MOVABLE CONTROL ASSEMBLIES . . . . . . . . . . . . . . . B 3/4 1-3 3/4;2 POWER DISTRIBUTION LIMITS. . . . . . . . . . . . . . . . . B 3/4 2 1 3/4.2,1 AXIAL FLUX DIFFERENCE. . . . . . . . . . . . . . . . . . B 3/4 2 1 3/4.2.2 and-3/4.2.3 HEAT FLUX H0T CHANNEL FACTOR AND NUCLEAR _ ENTHALPY RISE HOT CHANNEL FACTOR. . . . . . . . . . . . B 3/4 2-2 3/4.2.4 QUADRANT POWER TILT RATIO. . . . . . . ........ B 3/4 2-3 3/4.2.5 DNB PARAMETERS . . . . . . . . . . . . . . . . . . , . . B 3/4 2-4 3/4.3 -INSTRUMEN1ATION
-3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY -
FEATURES ACTUATION SYSTEM INSTRUMENTATION . . . . . . B 3/4 3 1 3/4.3.~3 MONITORING INSTRUMENTATION . .............. B 3/4 3-3 3/4'.3.4 -TURBINE _0VERSPEED PROTECTION . . . . . . . . ... . . . . B 3/4 3-6 3/4.4- REACTOR COOLANT SYSTEM 3/4.4.1. REACTOR COOLANT LOOPS AND C0OLANT CIRCULATION . . . . . B 3/4 4-1 3/4,4.2 SAFETY VALVES. . . . . . . . . . . . . . . . . . . . . . B 3/4 4-1
~3/4.4.3 PRESSURIZER. . . . . . . . . . . . . . . . . . . .-. . .
B 3/4 4-2 3/4.4.4 REllEF VALVES. . . . . . . . . . . . . . . . , . . . . B 3/4 4-2
-3/4.4.5 -STEAM GENERATORS-, . . . , . . . . . . . . . . . . .
. B 3/4 4-2 3/4.41.6 -REACTOR COOLANT SYSTEM LEAKAGE . . . ._. . . . . . . . .
B 3/4 4-3 3/4.4.7 CHEMISTRY. . . . . . . . . . c . . . . . . . . . . . . . B 3/4 4-5 3/4.4.8: SPECIFIC ACTIVITY. . . . . . . . . . . . . . . . . . . . B-3/4 4-5 3/4.4.9 PRESSURE /TEMPdRATURE LIMITS. . . . , . . . . . . . . . B 3/4 4 7 , FIGUREB3/4.4-l FAST NEUTRON-FLUENCE (E>1MeV).AS A FUNCTION OF - p FULL-POWER SERVICE LIFE . . . . . . . . . . , . . . . B 3/4 4-9 FIGURE:8 3/4.4 2 TThis figure number not used) . . . . . . . . . B 3/4 4-10 SEABROOK - UNIT 1 x r
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Controlling material: Base metal
* . Copper content: 0.06 WT%
PTm initial: 40'F RTm after ll.) EFPY: 1/4T, 108'F 3/4T, 86'F Curve applicable for heatup rate.; up to 60'F/hr for the service period up to 11.1 CFPY and contains margins of 10'F and 60 psig for possible instrument errors 2800 .
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(10 F PER OlVISION) FIGt'RE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP T0 11.1 EfPY SEABROOK - UNIT 1 3/4 4-31
7 l MATERIAL PROPERTY BASIS Conn olling material: Base metal l Copper content: 0.06 WT% I RT initial: 40*F o NDT 11.1 EFPY: 1/4T,108 F ' RTNDT after 3/4T, 86 F Curve-applicable for cooldown rates up to 100*F/hr for the service ceriod up to I 11.1 EFPY and contains margins of 10*F and 60 psig for possible inst.rumsnt errors i Illl' lilllill!' lllil!!!il 'I: li; ;i 2600 I
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g i ill 11 i li' 11 i Illi i 11 100 200 300-. 35e 40u o RCS . TEMPERATURE '(*F) j (10- F FER DIVISION)- FIGURE 3.4 REACTOR COOLANT SYSTEM-C00LDOWN LIMITATIONS - APPLICABLE UP .TO 11.1 EFPY SEABROOK - UNIT 1: , 3/4 4-32 . l:- l
r REACTOR C001 ANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cocidown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G:
- 1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be ;
limited in accordance with Figures 3.4-2 and 3.4-3 for the service i period specified thereon: I
- a. _ Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right i of the-limit lines shown. Limit lines for cooldown rates l between those presented may be obtained by interpciation; and
- b. Figures 3.4-2 and 3.4-3 define limits to assure prevention -l ofnon-ductile failure only. For normal operation, other
. inherent plant characteristics, e.g., pump heat addition and ,
pressurizer heater capacity, may limit the heatup and cooldown ' rates that can be achieved over certain pressure-temperature i ranges.
- 2. These limit lines shall be calcr'sted periodically using methods provided below,
- 3. The secondary side of the steam generator must not be pressured c 'above 200 psig-if the temperature of the steam generator is below 70'F,.
- 4. The pressurizer heatup and cooldown rates shall not exceed 100'F/h and 200*F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater that. 320*F, and
- 5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accoidance.with the requirements of ASME Boiler and Pressure Vessel Code Section XI.
The fracture toughness:proporties of the ferritic materials in the reactor
' vessel are determined'in accordance with the NRC Regulatory Guide 1.99, Revison 1 2, and in accordance-with additional reactor vessel requirements. 'These '
properties are then evaluated in accordance with Appendix G of the 1972 Winte. Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April-1975. Heatup and. cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTm at the end of 11.1 effec-tive full power years (EFPY) of service life. The 11.1 EFPY service life period SEABROOK - UNIT 1 B 3/4 4-7 v - --,,ar,,-ww-, w, - n -w a , -+, m ,s,--se ,--,,-~:'-w-,r--,,~e-+,e-whn ,wwn,- em+- s- w or w s v- --,N--o ~~e -
3EACTOR COOLPT_SYSIM BASES 3/4.4.9 PRESSURF/ TEMPERATURE LIMITS (Continued) is chosen such that the limiting RTer at the 1/4T location in the core region is greater than the RTer of the limiting unirradiated material. The selection assures that all components in the Reactor Coolant of System suchwill a limiting be operate RT'3 conservatively in accordance with applicable Code requirements, lhe reactor vessel materials have been tested to determine their initial RTor; the results of these tests are shown in Table B 3/4.41. Rea tor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTor. Therefore, an adjusted reference tempcrature, based upon the fluence, copper content, and nickel content of the material in question,- can be predicted using figure B 3/4.4-1 and the value of ARTer computed by either Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials." The heatup and cooldown limit curves of figures 3.4 2 and 3.4-3 include predicted adjustments for this shift in RTer at the end . . of 11.1 EFPY as well as adjustments for possible errors in the pressure and temperature-sensing instruments. Values of ADTor determined in this manner may be used until the results from the material surveillance program, evaluated according to ASTM ElG5, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10CFR50, Appendix H. The lead factor represents the relationship between the fast neutron flux density at the location of the cap-sule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict. future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. Evaluation of surveillance capsule data will be conducted in tecordance with NRC Regulatory Guide 199 Revision'2. . , Allowable pressure-temperature relationships forivarious heatup and
' cooldown rates are calculated using methods derived from Appendix G in Section 111 of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the_ linear clastic fracture mechanics (LEFM) technology. In the calculation procedures, a semielliptical surface defect - with a' depth of one-quarter of the wall thickness, T, and a length of 3/2T i SEABROOK -- VNIT 1- B 3/4 4-8 ,
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SEABROOK - UNIT 1 B 3/4 4-10
IV. Safety p aluntion of Proposed Channen Seabrook Station Technical Specification 3/4.4.9, " Pressure / Temperature Limits' are currently calculated based on the mathodology described in Revisien 1 to Regulatory Guide 1.99. The proposed revisions to Technical Specification 3/4.4.9, mandated by t he NRC in Generic Letter 88-11, are consistent with the methodology described in Revision 2 to Regulatory Guive 1.99. Revision 2 to Regulatory Guide 1.99 describes the general procedures acceptable to the NRC staff for calculating the affects of neutron rsdiation embrittlement of low alloy steels currently used f or light wa*.er reactor vessels. Based on calculation performed for North Atlantic by the Yankee Atomic Electric Company Nuclear Services Divisions (YNSD) utilizing the methodology described in Revision 2 to Regulatory Cuide 1.99 it was concluded that current heatup and couldown curves do not require revision if the service period is revised to 11.1 Ef f ective Full Power Years (EFPY)
~
versus the current 16 EFPY. The YNSD calculations are available f or raiew at Seabrook Station. The proposed revision to Figutes 3.4-2, Reactor Coolant System lleatup Limitations sud 3.4-3, Reactor Coolant System Cooldown Limitations indicate that the curves are applicable 11.1 EFPY versus 16 EFPY, revises the RTyn after 11.1 EFPY at the 1/4T and 3/4T reactcr vessel locations slightly downward to 108'F and 86'F respectively, and indicates that the copper content of the t sntrolling material is 0.06WT2. This results from a more conservative approach being taken regarding radiation embrittlement of reactor vessels and does not adversely affect plant safety. The Bases for Technical Specification 3/4.4.9 are revised to indicate that the heatup and cooldown curves are applicable for 11.1 EFPY and that the methodology described in Revision 2 to Regulatory Guide 1.99 was used to calculate the Reference Temperature for Nil Ductility Transition. Additionally, Figure B 3/4.4-2, Effect of Fluence and Copper Content on Shift of RTun for Reactor Veasels Exposed to 550*F is deleted. The - deleted figure was included in the original Seabrook Station Technical Specifications as an alternative to Regulatory Guide 1.99 for predicting the effects of radiation embrittlement on reactor vessel material. This alternative method is oot used and its .clusion in Technical Specifications serves no purpose. General Design Criterion (GDC) 31, " Fracture Prevention of Reactor Coolant Pressure Boundary," of Appendix A, ' General Design Criteria for Nuclear Power Plents," to 10CFR50, ' Domestic Livonsing of Production and Utilization Facilities', requires in part, that the reactor coolant pressure boundary be designed with suf fi.cient margin to ansure that, when stressed under operating, maintenance, testing, and postulated accident conditions. (1) the boundary beheves in a non-brittle manner and (2) the probability of a rapiuly propagating f racture la minimized. The proposed changes to Technical Specification 3/4.4.9 and its associated Bases ensure that the above criterion continues to be met. 6
e.* .- Therefore, an adjusted reference temperature _ based upon the fluence, copper content, and pickel content of the material in question, can be predicted using Figure B 3/4.4-1 and the value of .iRTgar computed by Regulatory Guide ' 1.99 Revision 2 " Radiation Embrittlement of Reactor Vessel Materials.' The heatup and cooldown limit curves of Figures 3.4 2 and 3.4-3 include predicted adjustments for possible errors in the pressure and temperature sensing instruments. Values of t2RTgar determined in this manner may be used until the results l from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10CFR50, Appendix H. The lead factor represents the relationship between the fast neutron flux density at the l location of the capsule and the irmer wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiatior damage to the reactor vessel material by using ? the lead factor and.the withdrawal time of the capsule. Evaluation of surveillance capsule data will be conducted in accordance with NRC Regulatory Guide 1.99 Revision 2. Generic Letter 88-11 required that revisions be made to Seabrook Station Technical Specifications based on the methodology described in Revision
-2 to. Regulatory Guide 1.99. The revision described herein to the Seabrook Station. Technical Specifications is consistent with the methodology described in Revisiot. 2 to Regulatory Guide 1.99. The proposed revisio_n ,
results in a greater degree of conservatism regarding the effects of radiation' embrittlement to reactor vessels and the impact- on plant operations, i-
~
l l 7
. . .u . . . ._ _ _ _ . - . . __.___. - _ _ _._....,_,_..;__-.,___..-_. c_._ _ . . - . _ .
b a V. Determination of Significant IInzards for License Amendment Request 92-06 Proposed Channes i i a (1) The proposed revision does not involve a significant increase in the i probability or consequences of an accident previously evaluated. The proposed revision to Technical Specification 3/4.4.9 Pressure / Temperature Limits imposes a more restrictive condition on the time of applicability of the pressure / temperature operating limits curves. This revisjon is the result of a more conservative calculation of the affects of radiation embrittlement of reactor vessels and its impact on plant operations using the current calculational methodology delineated in NRC Regulatory Guide 1.99 Revision 2. Since the plant respense to an accident will not change there is no change in the potential for a release of radiation to , the public. As there is no change in the potential for an increase in the release of radiation to the public it follows that the consequences of an accident, measured in terms of dose, will not increase due to the proposed conservative revisions to the heatup , and cooldown curves. The proposed revisions to the heatup and cooldown curves do not change the function or operation of any plant equipment or effect the response of that equipment if it is called upon to operate. Since the plant will continue to function as designed there will be no significcnt increase in the probability of an accident previously evaluated. (2) The proposed chat.g e s do not create the possibility of a new or different kind of aceldent from one previously evaluated. The- proposed revisions to the heatup and cooldown curves do not change plant danign or function, effect the operation of any plant equipment or introduce any new failure mechanisms. The proposed changes to Technical- Specification 3/4.4.9' provide a more conservative estimate of radiation embrittlement of the reactor vessel and reduces the time of applicability of the pressure / temperature operating limits curve. The previous accident analyses are unchanged and bound all expected plant transients and there are no new or different accident scenarios created. Therefore, the proposed revisions do not create the possibility of a new or different kind of accident from one previously evaluated. (3) The proposed changes do not result in a significant reduction in the l' margin of' safety. The Bases of Technical Specification 3/4.4.9, Pressure / Temperature
- Limits is to assure that the Reactor Coolant System is designed and operated in a manner such that a non-ductilo condition is not reached and that the Reactor Coolant System is conservatively operated in 8
_ .... ..- . . . - . .- . - . - . . - . . . ._ .. . . - . . - ..-. . - .. . - _ . ~ . ~ . - ~ . - . . - . . - _
* -e-accordance with applicable Code requirements. The proposed revision is consistent with the methodology described in Revision 2 to Regulatory Guide 1.99 and does not significantly reduce the margin of safety defined in the Bases. The proposed revisions provide increased conservatiam regarding radiation embrittlement of reactor
. vessels and its impact on plant operations.
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4 VI. Proposed Schedule for License Amendment Issuance and Effectiveness The North Atlantic response to Generic Letter 88-11 [ Reference (c)) stated that the existing Seabrook Station heatup and cooldown curves are correct until' eleven Effective Full Power Years and' committed to submit changes ' to the curves utilizing the guidance of Generic Letter 88-11 and the methodology of Regulatory Guide 1.99, Revision 2 prior to the tart of the second refueling outage. License Amendment Request 92-06 completes North > Atlantic's commitments regarding Generic Letter 88-11. Licenae Amendment Request 92-06 does not propose a specific date for ;
^
i s er c.~ ., of a license amendment. The existing Technical Specification heatup and cooldown curves are correct until 11 E"ective Full Power years. i If operation at a cumulative capacity factor 01 80Z from August 1990 is _ assumed, 11 Effective Full i'ower Years of operation will be achieved in r the fourth quarter of 2003. L L b 9 E' _ ____. . . = . _ . _ _ _ . -.;_ _ _ . . . . . _ , , ._- _ . . _ . -
. . *. . - 4 a 1 4
VII. Environmental Impact Aerossment North Atlantic has reviewed the proposed license amendment against the criteria of 10CPR$1.22 f or environmental considerations. The proposed changes do not invol.ve a significant hazards consideration, nor increase the types and amounts of eiflusnt 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 l categorical exclusion from the requirements for an Environmental Impact Statement. I T p
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VIII. Other SupportinP Documentation NHY Letter FYN-88155, ' Response to USNRC Generic Letter 88-11" dated November 30, 1988. G. S. Thomas to USNRC. 1 M 9 1em 12 i
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George S. Thomas s dl
#98 vice n esident Nuci cr Production 9 il .
Pub 6c Service of New Hampshire New Hampshire Yankee Divistori NYN 88155 November 30, 1988 United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Document Control Desk
' References (a) Facility Operating License No. NPF-56, Docket No. 50-443 (b) USNRC Generic Letter . dated July 12, 1988, 'NRC Position on Radiation Embrittl ; of Reactor Vessel Materials and its Impact on Plant Operations" Subjects- R .ponse to USNRC Generic Letter 88 '1 Gentlemen:
New Hampshire Yankee (NHY)'has reviewed Reference (b) and has determined
-that in order to maintain compliance with Section V of 10CFR50 Appendix G, a change to the Technical Specifications will be required prior to exceeding eleven (11) Effective Full Power Years (EFPY) of operation. This change will affect Reactor Coolant System Technical Specification 3/4.4.9, and associated
. BASES discussion regarding Revision 2 to Regulatory Guide 1.99 computational methods for determining RTNDT shift which is then used to develop plant heattcl.
dnd - Cooldown ' Ct.r 70s . The current heatup and cooldown curves are basea upon Revision'l to t(egulatory Guide 1.99 and, as stated above, are correct unti eleven EFPY. Generic Letter 88-11 requires that_any required changes tc, Technical Specifications be submitted for approval within two refueling outages after the effective date of Revision 2 to Regulatory Guide 1.99. New Hnmpshire Yankee anticipates submittal of these required changes prior to its second refueling outage. L rs .t%V
, -mD__1 e
>nO ,, P.O. Box 300. Seabrook. NH 03S74 . Telephone (603) 474-9574 O 6 i p-
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United States Nuclear Regulatory Commission November 30, 1983
- - Attention: Document, Control Desk Page 2 Should you require additional information please contact Mr. Robert A. Gwinn at (603) 474-9574, extension 4056.
Very truly yours, cYf eorg . Thomas ces Mr. William T. Russell
- Regional Administrator Region I
-United-States Nuclear Regulatory Commission 475~Allendale Road King of: Prussia, PA 19406 Mr.; Victor-Nerses. Project Manager Project Directorate I-3 Division of. Reactor Projects United States Nuclear Regulatory Commission Washington, DC 20S55
-Mr. David G. Ruscitto.
'NRC Senior Resident Inspector P,0. Box 1149-Seabrook.--NH 03874 r<., y}}