ML20100Q207
| ML20100Q207 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/29/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20100Q206 | List: |
| References | |
| NUDOCS 9202060235 | |
| Download: ML20100Q207 (4) | |
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NUCLE AR REGULATORY COMMISSION
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By letter dated June 11, 1991, the Boston Edison Cocpany (the licensee) sut ri tted a recuest for cheones te the pilor r I;uclear Pov:er Station, i
4 TechnicalSpeciiuutions(TS). The requested changes would revise tit pressure /terperature (P/T) 14r,'ts in the Pilgrin Technical Specifications, Section 3.0.
The proposed P/T limits were requested for 10, 11, 13, 15, 70, ati 37 ef fect've f t.11 pchet y e c r s ( E f h ',.
In October 1991, the EfrY is et out 9.1.
The proposed P/T idin'.ts were developed using Regulatory Guide (RG) 1.99, Fevision E.
Genet ic letter EC-11, "t1C Fosition o' Radiation Erabrittlerient of Reactor Vessel Materials and its Effect on Plant Operations," recommends FG 1.99, Rev. ?, be used in calculating P/1 linits, unless the use of different trethods can t e justified.
A P/T limits for the bottom head of the reactor vessel ure ciso requested.
To evaluate the P/T idr4ts, the staff utes the following f;PC regulations and Apptrdi os 0 and ti of 10 CfR Part 50; the ASTM Standards and the ASitE guidance:
t Code, which ere refen nted in Aprend'tes C and ti; 10 CFR E0.3f>(c)(?); RG 1.99, Fev. ?; Standard Peview Plan (SRP) Section E.3.?; and Generic Letter 80-11.
Each idctosee authorized to opet ate a nuclear power reactor is required by 10 CFR 50.30 to prov de Technical Specifications for the operation of the plant.
d In particular, ;0 CFR 50.3f(c)(?) requires that lindting conditions of operation be incluoed in the lechnical Specifications. The P/1 hmits are among the 14ritinc conditions of operation in the Technical Specifications for all corner ciel r.uc1(at ;i<nts 4r the U.S.
Appndices G and 11 of 10 CFR Part !0 describe specific requirements for fracture toughness and reactor vessel raterial surve llance that rust be considered 40 settino P/T limits.
An i
acceptable rattod ior constructing the P/T limits is described in SRI Scction 5.3.P.
C of 10 Cff. Part 50 stecdfi s fracture toughness and testing requirerents An eid4>
e f or t eactor sessel r.aterials in accordance with the ASt:E Code and, f r. partituici, that tt e t:eltline rateriels n the surveillance capsules te tested in eccordance d
with Append h b of IC Cfi fort LO. fprendi> h, in turn, refers to ASTt'. Standaids.
Tt ese test s def doe the e> tert of vessel enci dttlerient at tre tire of capsule wittirewal dr. te r r,5 of the 'nci t ase n r tference temper aturt. Irrendh C clso i
r(qu res the li censee tc pred4ct the effects of neutron irradiation on vessel i
9202060235 920129 PDR ADOCK 05000293 P
PDR 9
Charpy upper shelf er trgy (g the $djusted reference tenperature (ART) and coibrittleinent by calculatin USE). ' neric letter 80-11 requested that licensees and perrittees use the raethods in RG 1.99, Fev. T, to predict the effect of neutron drradiation on reactor vessel natorials. This guide defines the IRT as the sum of unirradiated reference teroperature, the increase in reference terperature resulting from neutron irrad% tion, ar.d a roargin to account for uncertainties in the prediction rnethod.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance prograrn to periodicelly withdraw surve411ance capsules from the reactor vessel. Appendix H refers to the ASTil Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specdtnens n.adt from plate, weld, and heat-effected zone (HA7) nattrhls of thc reactor beltline, ty letters dated /Ugust 5 and October 4,1991, the licensee requested Cases changes to Sections 3.1, 3.2, 3.3 and 4.3, 3.5.0, D & E, 3.9 ar:d 3.10.
2.0 EVAL U AT_1_00 The staff evaluatcd the effect of neutron irradhtion utbiittluant ch each beltline inaterial in the Pilgrim reactor vessel. The amount of irradiation en.brittlernent was calculated in accordante with RC 1.99, Rev. ?.
The staff has f
determined that the material with the highest ART at 32 EfPY was the lower intern!ediate shell axial weld (1-338A, B, and C) with 0.13i copper (Cu),
1.06% nickel (Ni), and an initial RT f -35'F.
ndt The licensee has removed one surveillance capsule from Pilgrim. The results f rom capsule 1 were published in Southwest Research Institute f erort SwP1 0?-5951. All surveillance capsules contained Charpy irnpact specirnens and tensile specin. ens inade frori base retal, weld netal, and HAZ retal.
for the lititing beltline r'aterial, weld 1-330A, B, and C, the staff calculated the ART to be 90.9'r at 1/4T (T = reactor vessel beltline thickness) arid 2
70.2'T for 3/4T at 32 EfrY. The staff used a neutron fluence of 9.0E17 n/cm at 1/41 and 4.9f17 n/ctrY?H at 3/4T. The ART was determ'ned by Section 1 of RG 1.99, Rev. ?, because only one capsule was renoved f rom the Pilgrim reactor pressure vtssel.
The licensee used the rcethod in RG 1.99, Rev. 2, to calculate an ART of 91*f at 32 EFPY at 1/tT for the sent l ' r14 ting weld retal.
The staff judges that the licensee's ART of 91'i is tuor e conser vat've than the staf f's ART of 90.9'f, and
=it is acceptable.
Substituting the ART of 91*r into equations in SRP 5.3.2, i k staff verifh d itet the prcrosed P/T lic.its for 37 FTPY for bettur, cooldowr, and hych ttt st. r-eet the beltline reaterial r equirer4r,ts in Appr,dh 6 oi 10 CF R TFe staff also vt:r fied thet r/T limits for 10, 11, 1?, 15, and 20 i
Part EP.
i Ef h s r eet the Appenon G (qu r eraents.
3-4 i
In addition to t'eltline nate.als, Aprendix G of 10 CfD rart 50 else imposes i
P/T lindts based on the reference temperature for the reactor vessel closure fler ge materials.
Section IV.A.2 of Appendix G states that when the pressure exceeds 20t of the pre-service system hydrcstatic test pressure, the teroperature of the closure flange regions highly stressed by the bolt preload nest exceed the reference ton.perature of the mater tal in those regions by at least IT0'T for normal operation and by 90*f f or hydrostatic pressure tests and leah tests.
Paragraph IV.A.3 of Appendix G states "an exception may be made for boiling water reactor vessels when water level 15 within the normal range for power operation and the pressure is less than 20 percent of the pre service system hydrostatic test pressure, in this case the minimum Dermissible temperature is 60*f (33'C) above the reference temperature of the closure flange regions that are highly stressed by the bolt preload."
0.ased on the flange reference temperature of -5'f, the staff has determined that the proposed P/T limits for the 11, 13, 15, 20, er.d 32 EfPYs satisfy Section IV.A.2 of Appendix G.
t In regard to the proposed bottom head P/1 limits, the staff celieves that because the bottem head of the reactor vessel doet not receive significent amount of neutron flucoce, embrittlement due to irradiatica is not of rajor concern. Tre reference temperature calculations for the reactor beltlinc materials as prescribed in RG 1.99, Rev. 2 are not applicable to the botton head P/T limits. The licensee calculated the stresses of the bottom head materials due to internal pressure, startup and cooldown transients, deadweight, and seismic loadings. The maximum stress locations are located at the junction between the lower torus and the support skirt and at control rod penetrations, frora the maximur stresses, stress intensity factors and p/T limits were calculated based on ASIE Code, Section ill, Appendix C and 10 CFR 50, Aprendix G.
The staff finds that the licensee's calcuiotion satisfy 10 CFR 60, Appendix G.
However, to saf eguard the structural integrity of the reactor beltline materials, the licensee must ensure that the pressure and temperature readings from the P/T sensors at the reactor vessel beltidne region must be within the acceptable region of the beltline P/T limit curves when the l
bottom head P/T limits are beinc used during heatup and cooldown. The bottom l
head P/T limits rnust follow the sane heatup and cooldown rate,100 degrets T per hour, as that of the telt14ne P/T linMts.
3.0 STATTCONSUITAT103 In accordance with the Comm ssion's regulations, the Comonwealth of d
Passachusetts State efficial was notified of the proposed issuance of the art ndment. The 5 tate cificiel had no contrents.
4.0 Et/Mrot:Pf t'T/t C0t:510FFE03 The arrendirent changes a requirerent w th respect to ' rata 11ation or use of a i
f ceility conponent locateo within the restricttd area es defit co in 10 Cf L The hRC staff hts determ ned that the trendment involves no i
Part 70.
significant inctease in the amounts, and ro significant char.ge in the types, of any effluents thct r,ay be released off site, and that there is no 1
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significant increase in individual or cumulative occupational radiation exposure. The Corrission has previously issued a proposed findirt that the amendnent involves r.o significant barards consideration, and there has been no public coment on such finding (56 TR 31429). Accordingly, the an.endment rieets the eligibility criteria for categorical exclusion set forth in 10 CFR 51,22(c)(9).
Pursuant to 10 CTF 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 00lnOS10N The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasor,6ble assurance that the health and safety of the public will not be endangered by operation in the proposed rianner, (2) such activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
6.0 REFERENCES
1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1!t0 2.
NUREG-0000, Standard Review Plan Section L.3.2: Pressure-Temperature 1.imits
?.
Letter from G. W. Davis (BEco) to USNRC Document Control Desk,
Subject:
Proposed Changes to the Peactor Pressure Vessel Therfral and Pressurization Technical Specification Limits, June 11, 1991 4
E. B. Norris, " Pilgrim Nuclear Power Statico Unit 1 Reactor Yesse) Irradiation Surveillance Program, 5tlR1 OP-5951,"
July 1901 Principal Contributor:
John C. Tsao l
Date:
January 29, 1992
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