ML20199K237
| ML20199K237 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/20/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | City of Dalton, GA, Georgia Power Co, Municipal Electric Authority of Georgia, Oglethorpe Power Corp |
| Shared Package | |
| ML20199K242 | List: |
| References | |
| DPR-57-A-126, TAC 60542 NUDOCS 8607090089 | |
| Download: ML20199K237 (13) | |
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION D
'j W ASHINGTON, D. C. 20555 "s.,
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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATICN MUNICIPAL El.ECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET N0. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 i
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
License No. DP 2g/
?
1.
The Nudlear Regulatory Comission (the Commission) has found that:
B A.
The application for amendment by Georgia Power Company, et al.,
(the licensee) dated January 7, 1986 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and, regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby j
amended to read as follows:
8607090089 860620 t
PDR ADOCK 05000321 P
. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.126, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
F R THE NUCLEAR REGULATORY COMMISSION Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 20, 1986 t
.i..
0 ATTACHMENT TO LICENSE AMENDMENT NO. 126 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET N0. 50-321 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed paaes. The revised paaes are identified by Amendment number and contain vertical lines indicating the area of cahnge.
Remove Intert X
X 3.6-1 3.6-1 3.6-2 3.6-2 3.6-15 3.6-15 3.6-16 3.6-16 3.6-17 3.6-17 Figure 3.6-1 Figure 3.6-1 Figure 3.6-2 Figure 3.6-2 Figure 3.6-3 Figure 3.6-3 Figure 3.6-4 (BlankPage) i i
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- LIST OF FIGRES Fistre Title 1.1-1 mee 'Ihermal Power Safety Limit Verals Core Flow Rate 2.1-1 Reactor Wamal Meter Invels 4.1-1 Graphical Aid for the Selection of an Adequate Interval i
Between Tests 4.2-1 System Unaval1=hil4ty 3.4-1 Sodium Pentaborate Solution Volume Veraas Concentration Reg 1irements 3.4-2 Sodium Pentaborate Solution Tenperature Vermas concentration Realirements 3.6-1 Pressare 'yerass Minimum Temperature for Pressare Tests, Such as Recpired by ASMC Section XI 3.6-2 Press 2re Derals Minialm Tungerature for Non-N2 clear Heatup/Oooldown and Iow abwer Ihysics Test 3.6-3 Presaire verals Mininum Temperature for mre Critical Operation other than Iow Power. Physics Tests (Includes 40DF Margin Realired by 10GR50 Appendix G) 3.6-5 Thermal Power Limitat' ions During Operation with Less Than Two Reactor Coolant System Recirculation Loops in Operation.
3.H-1 (Sheet 1) Limiting Value for APIEGt (R1el Type 3) 3.u-l (Sheet 2) Limiting value for APIBGR (Rzel Types 1 and 2)
- 3. u-2 deleted
- 3.H-3 Kg Factor i
i 3.15-6 tkirestricted Area Beundary 6.2.1-1 offsita organization 6.2.2-1 tkilt organization 1
1 BA1M - UNIT 1 x
Amendment No. JJ0, 126
_seeQ M
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I LDGTING CDBOITIDNS FOR OPERATDI SURVEILIANCE REGJIREMENIS L 3.6 PRIMARY SETEM B00tOARY 4.6 PRDERY SETEM BOUtDARY Applicability Applicability The Limiting (bnditions for h alrveillance Rea.lirarnants apply Operation apply to the oper-to the periodic examination and ating status of the reactor testing recuirements for the coolant syntam.
reactor coolant ' system.
Obiective Chiective l
The objective of the Limiting
'Jhe objective of the alrveillance Cbnditions for Operation is to Recuirements is to determine the asaare the integrity and safe condition of the reactor coulant operation of the reactor coolant system and the operation of the systan.
safety devices related to it.
~
Specifications' Specifications A.
Reactor Cholant Heat-Up and A.
Reactor Coolant Heat-Up and (boldown
%1%
h average rate of reactor
'Jhe reactor coolant system coolant temperature change tangerature and presaire daring normal heatup or cool-shall be determined to be down shall tot enomed 100*F/hr within the limits of when averaged over a one-hcur Specifim tions 3.6.A. and 3.6.B.
period.
at least once every 30 mirutes daring reactor coolant o
i heatup and cooldown.
4 rij B.
Reactor Messel 'Dummerature and B.
Reactor Vessel 'Denperature and 1
Pressare Pressare
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1.
'the reactor vessel shell temper-l Reactor v===*1 metal tangerature atures daring inservim hydro-at the astside airface of the static or leak tasting shall be bottom head in the vicinity of l
at or above the tauperatures g
the control rod drive hassing shown on the carve of Figure 3.6-1. I and reactor vessel shell adjacent to shell flange shall be re-j oorded at least every 15 airutas daring ir.-;_.-ion ij hydrostatic or leak testing when l;
thevesselpressareisg312 5
psig.
Amendment No. 59,126 BMG - IMIT 1 3.6-1
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LDGTING (DtOITIONS FOR OPERATIDN SURVEILIANCE REQJIRDENIS j
3.6.B.
Reactor Vessel Tsuperature and 4.6.B.
Reactor Vessel Tuncerature and Press 2re (Contimed)
Presaare (Contimed) l Test specimens representing the 2.
D2 ring heatup by non-mclear means, cooldown following melaar reactor vessel, base weld and weld stutdown or low level physics tests, heat affected zone metal were l
the reactor vessel shall and fluid installed in the reactor vessel tangerat0res of Specification 4.6.A.
adjacent to tte vessel wall at shall be at or above the tangeratures the core midphne level before shown on the carve of Fig 2re 3.6-2.
the start of operation. 1he mmber and type of specimens are 3.
Daring all operation with a critical l
in accordance with GE report cor% otner than for low level physics NEDO-10ll5. The specimens meet tasts, the reactor vessel shall and the intent of AS1M E185-70.
fluid temperatures of Specification 4.6.A. shall be at or above the teenper -
The next marveillance capsule atures shown on-the curve of shall be removed from the ves-Fig 2re 3.6-3.
sel at apperwimately 15 EFPY of operation, as rerv==nded in AS1M E185-82, tut not to exceed 16 EFPY.
. t 4
3.6.C.
Reactor Vessel Head Stud C.
Reactor Vessel Head Stud Tensioning Tensioning The reactor vessel head bolting M wn the reactor vessel head studs shall not be under tension studs are under tension and the unless the temperWre'of the reactor is in the Cbid Stutdown weasel head flange and the head Cbndition, the reactor vessel is greater than 76'F.
I shell temperature immediately below the head flange shall be permanently recorded.
D.
Idle Reciroslation Icop Startup D.
Idle Reciroslation Icop Startup 1he susp in an idle recirculation Prior to and daring startup of an Icop shall not be started unless idle reciro11ation loop, the tem-the tangeratures of the coolant,
perature of the reactor cnolant within the idle and operating re-in the operating and idle loops ciro21ation loops are within 50*F shall be r = pared and permanently t
of each other.
recorded.
8 l.
Amendment No. 59,126 HATCH - INIT 1 3.6-2 v - - -
se,-------
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BASESFORLIMITINGCONDITIONSFOROPERATIONANDSUR5cILLANCEREQUIREMENTS o
3.6 PRIMARY SYSTEM BOUNDARY j
A.
Reactor Cbolant Beatup and Cooldown j
The vessel has been analyzed for stresses ca1 sed by thermal and F.1 presaire transients. Heating and cooling transients thra.ighalt j
plant life at uniform rates of 1000F per hour were considered 1
in the temperature range of 100 to 5460F and were shown to be
]
within the reazirements for stress intensity and fativ e limits of Section III of the ASME Boiler and Presaire Vessel Code (1965 Biition including Winter 1966 addenda).
l B.
Reactor Vessel TIusperature and Presaire Operating limits for the reactor vessel presaire and tenperature, 3
daring normal heatup and cooldown, and daring inservice hydro-static and leak testing were established using 10CER50 Appendix G, May 1983 and Appendix G of the Winter 1984 Addenda to Section III
. of the ASME aoiler and Presaare Vessel Code. In addition, operat-ing limits reflecting discontimity effects were calo21ated by ad)2 sting BNR/6 discontimity analyses to reflect the appropriate Batch 1 RI)gyr values. 1bgether, these operating limits asaare that a postulated marface flaw, having a depth of 0.24 inch at the F1-.,. 3-r-- -1 $2nction and one-gaarter of the unterial thickness at all other reactor vessel locations can be safely M tad.
9br the ptrpose of setting these operating limits, the Riigyr fo the vessel material was estimated from impact test data taken in i
1 accordance with restirements of the Code to which this vessel was j
c designed and ammfactured (1965 Edition including Winter 1966 j
Addenda). A General Electric Chapany procedtre, designed to j
evaluate fracture toughness restirements for older plants where 4
'information may be incomplete, was used to estimate R1)gyr alues v
on an estivalent basis to the new regairements for plants which have construction permits after Auyst 15, 1973.
J l
1 The limiting initial RI)gyr value of the RW core beltline region is 10DF, based on Charpy V-Notch data for plate material. The clamare flange region RTWyr is limited by the upper vessel shell i
plate with a value of 16PF based on Charpy data. She non-beltline j
discontimity limits for hydrotest (03rve A in a Tbpical Report,
{
' NEDC-30997) are based on the R2)gyr for the steam aatlet nozzle of 40DF, based on the dropuoight test temperature. She non-beltline discontimity limits for hastup/enaldawn (Carve B in 21bpical 3
Report IEDC-30997) and core critical operation (01rve C in GE Tbpical j
Report IEDC-30997) are based on the 40DF R2)gyr of the steem nat-a let nozzle, determined by Charpy data.
6 Figure 3.6-1 establishes minim 1m temperature remairements for leak testing and hydrostatic testing regaired by the ASME Boiler and Pressare Vammal Code,Section XI.
3
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l-BATG - UNIT 1 3.6-15 Amendment No. 57, 126 i
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BASES KR LIMITING CDtOITIONS KR OPERATION AND SURVEILIANCE RECUIRDENIS Test press 2res for inservice hydrostatic and leak testing req 2 ired by the ASME B&PV Cbde,Section XI, are a fbnction of testing tem-i perature and component material. Pbr the Hatch 1 reactor presmire i
vessel, the ISI hydrostatic test presa2re would be approximately 1.1 times operating pressare, or aboat 1106 psig, depending on the reactor water temperature. h tauperatures for pressares above 440 psig i
are determined by the RPV core beltline with a shift in Rh of 1230F, appropriate for operation up to 16 effective full power i,
years (EPPY).
Figure 3.6-2 provides appropriate limitations for plant heatup and cooldown when the reactor is not critical. Figare 3.6-2 is also applicable to low power physics tests. 'Jhese a2rves asaame heatup and cooldown rates up to 1000F per hour. Tenperatures for presa2res above 300 psig represent the limits of the RPV core beltline with a shift in R2)gyr of 1230F, appropriate for 16 EFPY of operation.
Figare 3.6-3 ' establishes operating limits when the core is critical.
I Fig 2re 3.6-3 is not applicable to low power physics tests. 3hese limits include a margin of 400F as recuired by 10CFR50 Appendix G.
In accordance with the May 1983 revision of 10CFR50 Appendix G, core txitical operation any be initiated at tenperatures at or above (R2)gyr + 600F) of the cloalre flange region, or 760F. Tenpera-tures for pressares above 300 psig represent the limits of the RPV core beltline with a R2)gyr shift of 1230F, appropriate for 16 EFPY of operation.
h fracture toaghness of all ferritic steels graciaally and uni-formly decreases with expomare to fast neltrons above a threshold i
value, and it is prudent and conservative to account for this in j
the operation of the RPV. Two types of information are needed in this analysis:
(a) a relationship between the change in fracture tough-j ness of the RPV steel and the naltron fluence (integrated nastron i
flux); and (b) a monasre of the neutron fluence at the point of interest in the RPV wall. A method of relating shift in RT)gyr to
'4 arrumilated fast ruetron (>l MeV) fluence is contained in Regalatory Glide 1.99, Revision 1.
Experimental results of irradiated mar-weillance specimens taken from the RPV show a shift in RT)gyr greater than predicted by Regalatory G2ide 1.99, so the marveillance resalts were used with the methods of 1.99 to establish the RI)gyr shift.
'Ibe shift for 16 EFPY was added to the unirradiated RPV core beltline curves, rea11 ting in the beltline being the limiting region in the vessel for higher prenaire-tenperature conditions.
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l BATCH - IMIT 1 3.6-16 Amendment No. 57, I
= = ~~ = --
=
BASES ftR LIMITING CDEOITIONS FOR OPERATION AND SURVEILLANCE REQUIRDENIS I
3.6.B.
Esactor Vessel Tamerature and Press 2re (Cbntin2ed) 1 h expected nastron fhence at the reactor vessel wan can be i
determined at any point daring plant life based on the linear relationship between the reactor thermal power outp2t and the 4 j corresponding rumber of ne2trons prod 2ced. Accordingly, neatron i
flux wires were removed from the reactor vessel with the a2 veil-lance test specimens to establish the correlation at the ca,.mle i
location by experimental methods. h flux distrib2 tion at the vessel wall and 1/4 T depth was analyticany determined as a function of core height and azimath to establish the peak flux location in the vessel and the lead factor of the marveillance specimens. Relating the fhx wire data to the vessel peak fhx analysis location gives a conservative estimate of maximan 1/4 T depth flux of 1.86 x 109 (ry'anZ-sec).
i Se first capalle containing test specimens was withdrawn in November 1984 after 5.75 EPPY of operation. Se specimens were tested according to AS'IM EL85-82 and the res21ts are in GE report j
NEDC-30997. h carves of Figures 3.6-1 through 3.6-; include the findings of the test report related to the copper-phosphon2s content of the RPV core beltline materials, the flux wire test and fluence i
, distrib2 tion analysis res21ts, and the Charpy V-Notch specimen test results.
i C.
Reactor Vessel Head Stud Tensioning i
Se recalirements for cold bolt up of the reactor vessel cloazre are 1
based on the R'4gyr temperature plus 60DF which is derived from j
the recuirements of the ASME Code to which the vessel was built.
- ]
Se =mvimm R'4gyr of the clomare flanges, adjacent head and shell j
seterial and stud seterial is 160F. h minials temperature for i
bolt-up is therefore 16 + 60 = 760F. Se neltron radiation fluence at the clomare flanges is won below 1017 not (>l Mev) and therefore radiation effects will be minor and will not influence i
this tasperature.
,t D.
Idle Recirculation Icop Startup Roosiring the coolant tenparature in an idle recirculation loop to be within 50*F of the operating loop temperature before a recirc21ation pasp is started prevents the potential seimare of the paup inpeller I
within the wear rings bocasse of the more rapid dimensional increase l
of the iwpallar daring heatup arising from thermal capacity.
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Amendment No. 59, 126 tl.
HMG - IMIT 1 3.6-17 l
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t I
1000 1
vauO TO to EPPECTivt putL POWER YEARS OF OPERATION 1400 i
i IE e
en a
O ADJUSTED CORE SELTLINE.
l 1/4 7 PLAW. RTway = to*P.
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iRRAOiATiON ewi,T. ins *,
l a.cW E
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o eaa l
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at 843 VERTICAL LIMIT LINE POR PRESSURE l
400 ABOVE 20E MYOROTEST I:12 pmel,,
mansn ON 10CFRSO APPENOlX G l
REQUIREMENT OF IRTNOT
- 80*FI.
PLANGE REGION RTNOT
- 88'P m--i l.
g
-SOLT PRELOAD TEMPERATURE OF s
I 7
yeep 34gg0 ON RECOMMENOED i
IRTNOT + eoeri pOR oL24 lN. rLAw I
IN CLOSURE FLANGE REGION, RTNOT*18'F e
o too ano ano a
son m
essNeasuu VESSEL aAETAL TEMPERATURE l*PI Figure 3.6 1 Pressure versus Minimum Temperature for Pressure Tests, Such as Required by ASME Section XI s
l HATCH - UNIT 1 Amendment No. 59,126
- - - - - - ~ - - -
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e i
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1000 I
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VAUO TO 18 EPPECTIVE FULL I
POWER YEARS OP OPER ATION I
1400 e
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1 310 g
b e
E
.w 1000 C
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E ADNSTED CORE SELTLINE 1/4 T PLAW. RTNOT*10ep W
IRMActAT10N SMtPT = 1220P 500 r
=
l
=
t e$
S g
ano s
ae h
W u
400 PEEOWATER NOZZLE TEMPERATURE LIMif POR 1/4 T PLAW (SWR /S RESULTS AOJUSTED TO 408F RTNOTI J
l ao MINIMUM OPERATING TEMPERATURE OP 70sp gAgEO ON RECOMMENDEO g
(RTNOT + 00*PI POR CL24 IN. PLAW IN
/
CLOSURE PLANGE REGION.
RTNOT
- 18'F 2
e t
E.,
aNNiMUM vessel METAL TEMPERATURE *P i
Figure 3.6-2 Pressure versus Minimum Temperature for Non-Nuclear Heatup/Cooldown and Low Power Physics Tests I
1 HATCH - UNIT 1 Amendment No. 59, 126
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1800 VAUD TO f E EPPECTIVE PULL POWER YEARS OF OPERATION 1400 7
>j g
1200 I
5b
=
C5
=
dh.
zs.
ADJUSTED CORE SELTUNE.
1
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1/4 T PLAW. RTwoT = 1Cor.
w IARADIATION SHIFT e 123er E
N soo W
81-d
.y 000 g
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t; l
5 ec W
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/
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PEEDWATER NOZZLE TEMPERATURE UMIT POR 1/4 7 PLAW (OWR /S RESULT 5 ADJUSTED TO 4087 RTNOTI 300 ABINIMUM OPERATING TEMPERATURE
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UMif OF M*P PROM 10CPRec APPENDIX 0 REQUIREMENT TNAT(Tugg = RTNOT + 00epp, g
PLANGE nTNOT - te=P
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.00 400
.00
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$$0N800UM VESSEL META'. TEMPERATURE ('PI '
- I.
Figure 3.6.3 Pressure versus Minimum Temperature for Core Critical
'I Operation other than Low Power-Physics Tests (Includes l
40*F Margin Required by 10CFR50 Appendix G) l s
HATCH - UNIT 1 Amendment No. 59, 126 a *** ~
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Amendment No. 59, 126 EM05 INIT 1 1
a nagk UNITED STATES
[
p, NUCLEAR REGULATORY COMMISSION 5
l WASHINGTON, D. C. 20555
/
SAFETY EVALUATION BY THE OFFICE OF NbCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.126 TO FACILITY OPERATING LICENSE N0. DPR-57 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 DOCKET N0. 50-321 Introduction By letter dated January 7, 1986, Georgia Power Company (GPC) requested modifica-tion of the Hatch Unit 1 Technical Specifications (TSs) regarding the reactor vessel temperature and pressure limits for pressure tests, non-nuclear heatup/
cooldown, core critical operation and vessel stud tensioning. The proposed modifications reflect the decrease in materials fracture toughness as measured in the first surveillance capsule removed from Hatch Unit 1 after Fuel Cycle 10 in November, 1984. As a supporting document to the request, GPC enclosed General Electric Report NEDC-30997 dated October 1985.
Evaluation The capsule received a fluence of 2.4 x 10 n/cm2 (E> 1.0MeV) during 5.75 17 effective full power years (EFPY) irradiation, based on dosimetry wire measure-i i
ments. Chemical analyses of reactor vessel materials indicated the following:
Limiting plate: 0.17% Cu, 0.011% P Limiting weld:
0:28% Cu, 0.013% P Surveillance plate: 0.13% Cu, 0.010% P The shift in fracture toughness properties is a function of fluence and chemical analyses, particularly copper and phosphorus, and is calculated by a specific relationship prescribed in Regulatory Guide 1.99, Revision 1.
The shift in RT in the surveillance plate was 47 F, compared to the calculated shift of 17hIusingthemethodologydescribedintheRegulatoryGuide. The shift in g
l l
RT in the weld metal was not measured because the fracture toughness properties ofNeunirradiatedweldwerenotavailable. Since the measured shift in the N
l surveillance plate exceeded the predicted shift by a factor of 47/17, the coefficient representing the materials from the equation in Regulatory Guide 1.99, Revision 1, was increased by an amount equivalent to a factor of 2.76.
The proposed changes to the Technical Specification relating to the pressure-temperature limits for reactor vessel test and operation meet the requirements of 10 CFR 50, Appendices G and H. ASTM E-185, Regulatory Guide 1.99, Revision 1, and Appendix G,Section III of the ASME Code.
8607090093 860620
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PDR ADOCK 05000321 p
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