ML20148H549
| ML20148H549 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/21/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20148H540 | List: |
| References | |
| NUDOCS 8803300022 | |
| Download: ML20148H549 (4) | |
Text
'
/ga arow*o, UNITED STATES l'
'j NUCLEAR REGULATORY COMMISSION g.
E WASHINGTON, D. C. 20555 o
%, +..../
i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.120 TO PROVI,SJONAL OPER,A,TJ N,G,,LJ,CENS,E, N,0,,p,P,R,-J6 GPU NUCLEAR CORFORATION AND JERSET tlMhALT%ER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219
_I_N_TRODUCTION By letter dated January 19, 19P8, the GPU Nuclear Corporation (the licensee) proposed to revise the pressure-temperature limits in the Oyster Creek Nuclear l
Generating Station Technical Specifications through 15 effer.tive-full-power years (EFPY). The proposed pressure-temperature limits were develcped from the
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licensee's submittal, "Testing and Evaluation of Irradiated Reactor Vessel Faterials Surveillance Program Specimens," TDR-725. The limits consist of three curves that set minimuin pressure and temperature for three operating conditions -
hydrostatic and leakage test, heatup or cooldown (core not critical), and heatup or cooldown (core critical). Presently, the plant is about to reach 10 EFPY which is the current technical specification limit for the pressure-temper-ature curves. The proposed new curves will allow the operator to operate the
, reactor continucusly through 15 EFPY withcut violating the Technical Specifications.
DISCUSSION Part of the NRC's effort to ensure integrity of the reactor vessel is to periodically evaluate the reduction in fracture toughness of the vessel material due to neutron irradiation embrittlement. The effort consists of three steps.
First, the licensee is required to establish a surveillance program in accordance with Appendix H of 10 CFR 50, which requires periodic withdrawal of surveillance capsules from the reactor vessel. The capsules are installed in the vessel prior to startup and they should contain test specimens that were made from the plate, weld, and heat affected zone materials of the reactor beltline.
Secondly, the licensee is required to perfonn Charpy impact tests, tensile tests, and neutron fluence measurements of the specimens. These tests define the condition of vessel embrittlement at the tirre of capsule withdrawal in terms of the shift of the reference temperature, RT and upper shelf energy.
The licensee should also predict the future vessel Er,ittlement by calculating the adjusted RT and upper shelf energy at a specific EFPY. The licensee NDT geo3300022g$$$bi9 PDR ADOCK PDR P
may use either Revision 1 or draft Devision ? of Regulatory Guide 1.99 to calculate the adjusted RT The upper shelf energy is the averace anergy, valueforallspecimenswbe. test temperature is above the upper end of the transition temperature region.
The licensee is required by 10 CFR 50 Appendix G to assure that the adfusted RT will not exceed 200*F and that the upper shelfenergywillnotbebeinwSbpft-lb at the end of plant life.
Thirdly, the licensee is required to develop a set of pressure-temperature curves based on the adjusted RT of the limiting vessel material.
The curves should satisfy the reco M ded methods ard requirements described in 10 CFR 50, Appendix G and Standard Review Plan 5.3.2.
EVALUATION The Oyster Creek Nuclear Station is a t* oiling water reactor which has an inside diameter of 213 inches and mean wall thickness of 7.125 inches.
The reactor vessel was fabricated frcm ASTM A302, Grade B plate material.
The submerged arc weld materials were RACO #3 bare wire and ARCO 8-5 flux. Manual metal arc welding used 8018 covered electrodes.
General Electric installed three specimen capsules as a part of the reactor vessel surveillance program.
The withdrawal of the first capsule in 1971 was, unsuccessful.
Capsule No. 2 was withdrawn in March 1984 at 8.38 EFPY. After examining specimens in capsule No. 2, the licensee found several material discrepancies and that the program does not meet requirements of 10 CFR 50 Appendix H.
For example, the limiting 'naterial and the beltline welds were not included in the capsule.
The exact copper and nickel contents of several plates and welds were unavailable.
These discrepancies were partly due to the vintage of the plant and partly because the surveillance proaram was initiated before the issuance of Appendix H.
Nevertheless, the staff had reviewed the surveillance program under the Systematic Evaluation Program guidelines in the early 1980's and found it acceptable.
In this evaluation, the staff concen-trated W on review of the program itself but on the pressure-temperature curves.
The specimen capsule data showed that the G-30n-1 plate had a RT shift of 72'Fmeasuredat30gn/cm{ansitiontemperatureandhadreceive$DT lb t a neutron fluence of 7.46 x 10 Since the G-308-1 plate showed a higher RT shift than that of the weld and heat-affected-zone materials in the capsuk tha data of the G-308-1 plate were used in the RT calculation of the limiting yp7 material.
The licensee used Regulatory Guide 1.99, draft Rev. 2 to calculate the adjusted RT because the Rev. 1. calculation showed a lower RT shift.
NDT NDT
3-To calculate the highest ad,4usted RT the licensee compared the copper and nickel contents of the G-308-1 plate So,those unirradiated specimens of five g7 other beltline plates not placed in the capsule. The licensee conservatively applied the chenistry factor, reutron #1uence and measured RT of the G-304-1 platetooneofthe#iveplatespecimensthathadtheworstcNinationof cooper content, nickel content, and initial PTY.
The calculation shewed that theG-P.-6gatehgdthehighestadjustedRTn/cm,15"FPY, and 1/4T (vesseNhickness) locatio The G-8-6 125'F at the neutron fluence of 1.11x10 plate was selected as the limiting material.
The licensee also predicted the end-of-life adjusted RT upper shelf energy of 61.5 ft-lb at a neutron fluence oN of 14?'[gandthe 38 x 10 n/cn for the G-8-6 plate.
These values satisfy the 10 CFR 50 Appendix G requirements.
To construct the pressure-temperature curves, the licensee followed closely the method described in NRC's Standard Review Plan 5.3.? and ASME Section III Appendix G except in the membrane stress calculation.
To calculate the membrane stress, the licensee used the "vessel radius-thickness" relationship whereas SRP 5.3.2 prescribed the "allowable stress-design pressure" relationship.
The former gives a lower temperature profile than that of the latter; but, the former nethod is not necessarily incorrect.
The staff determined that the licensee's method was acceptable based on the stress analysis of a cylindrical container having a large radius-to-thickness ratio, (Ref. Roark, R.J.,
"Formulas for Stress and Strain," 4th edition, page 308). The lower part of the pressure-temperature curves also has to satisfy the specific requirements of 10 CFR 50 Appendix G for boiling water reactors because the boilinp water reactor vessel has an inherent pressure-temperature limitation when the ranctor water level is within the normal range for power operation and the reactor pressure is less than 10 percent of the preservice system hydrostatic test pressure.
The pressure-temperature curve is limited by the closure flange regions that are highly stressed by the bolt preload.
The minimum permissible temperature should be 60'F above the initial RT of test pressure is above 20% of the hydrotest pre $kure, the flange and when the T
the permissible tempera-ture should be 90*F above the initial PT Based on an initial PT of 40'F
- or the Oyster Creek reector flange, the $ n,imum temperature should E 100 F and the permissible test temperature should be 130*F.
Examining the lower part of the pressure-temperature curves, the staff determines that the curves satisfy the 10 CFR 50 Appendix G requirements.
The staff has reviewed the proposed pressure-temperature curves and correspond-ina paragraphs in the Technical Specifications..The licensee has applied appropriately Regulatory Guide 1.99, draft Rev. 2,10 CFR 50 Appendix G, and Standard Review Plan 5.3.2 to calculate the adjusted RT and to develop the pressure-temperature curves. The staff concludes that b proposed pressure-temperature curves are valid through 15 EFPY and may be incorporated into the Oyster Creek Nuclear Station Technical Specifications.
f
-4 ENVIRONMENTAL CONSIDERATION This amendment changes a requirement with respect to the installation or use i
of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendrent involves no significant increase in the amounts, and no significant change in the types, of any effluents '. hat may be released offsite, and that there is no signif-icant increase in individual or cumulative occupational radiation exposure.
The Commissier has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public connent on such finding.
Accordingly, the amendmort meets the eli criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9)gibility Pursuart to 10 CFR 51.22(bi, no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
CONCUJSION The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety o# the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the common defense and security nor to the health and safety of the public.
Dated: March 21, 1988 Principal Contributor:
J. Tsao f