ML20083Q952
| ML20083Q952 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/10/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20083Q947 | List: |
| References | |
| TAC-53512, NUDOCS 8404230399 | |
| Download: ML20083Q952 (5) | |
Text
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WASHINGTON. D. C. 20555 s...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ISSUANCE OF AMENDMENT NO. 94 TO DPR-65 NORTHEAST NUCLEAR ENERGY COMPANY, ET AL.
MILLSTONE NUCLEAR POWEP STATION, UNIT ?
DOCKET NO. 50-336 INTRODUCTION In a letter from W. G. Counsil to J. R. Miller dated January 4,1984, the Northeast Nuclear Energy Company (NNECO) requested a revision to the Millstone Nuclear Power Station, Unit No. 2 (Millstone-2) Technical Specifications (TS).
The licensee's proposed TS changes were to Section 3.4.4.9, Pressure / Tempera-ture Limits ar.d to Table 4.4-3 Reactor Vessel Material Irradiation Surveillance Schedule.
EVALUATION The Millstone-2 pressure-temperature limits cust be calculated in accordance with the requirements of Appendix G,10 CFR 50, which becane effective on July 26, 1983. Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G,10 CFR 50, are dependent upon the initial reference temperture, RT
, f r the limiting materials in the beltline ard closure flange regions ohD[he reactor vessel, the water / metal terperature profile throuah the reactor vessel pressure boundary wall during heatup and cooldown, and the increase in RT resulting from neutron irradiation damage to the limiting beltline materiaNT The Millstone-2 reactor vessel was procured to ASME Code requirenents, which did not specify fracture toughness testing to determine the initial RT for each reactor vessel material. The initial RT iscalculatedbasedobest results perforned on transversely oriented Chby V-notch and drop weight specimens. The licensee has drop weight tested some cf the closure flange and beltline region materials, but has Charpy V-notch tested these materials with longitudinally oriented specimens rather than transversely oriented soecimens. This test data is reported in Table 4.6-1 of the Millstone-2 FSAR and a letter from D. C. Switzer to G. Lear dated December 9, 1977. The staff's review of this data indicates that the limiting closure flange region material is Vessel Flange Code No. C-500 and the limiting beltline materials are Plate Code No. C-505-2, and Ueld Seam No. 9-203. Weld Seam No. 9-203 contains naterial fabricated using Linde 0091 flux batches 3998 and 3999 and weld wire heat numbers 90136 and 10137. However, for seven (7) effective full-power years (EFPY), which is the effective period of the proposed pres-sure-temperature limits, the limiting beltline material is Plate Code No.
C-505-2.
Later in the plant's life, the weld will become liniting, because its rate of increase in RT resulting fron neutron irradiation damage is predicted to be greater thkthat of the plate material.
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- When drop weight tests have been performed, the staff uses the criteria in Section 1.1(3)b of Rranch Technical Position-MTE8 5-2, " Fracture Toughness Requirements" to determine the initial RT of a material. Based on this criteria, the initial RT for Vessel Fl$ke Code No. C-500 and Beltline PlateCodeNo.C-505-2wbdbe+10Fand+25F,respectively. These initial RT values were reported by the licensee in Table 4.6-1 of the Millstone-2 FSMTana in Table 3.2-1 of the licensee's report, " Thermal Shield Damage Recovery Program Final Report," December 1983. The criteria in Branch Tech-nical Position-MTEB 5-2 is currently under review by the staff.
If it is
,iudged nonconservative, we will request the licensee to reevaluate their pressure-temperature limits.
The increase in RT resulting from neutron irradiation damage was estimated by the licensee us k the methodology documented in Commission Report SECY-g 87-465, " Pressurized Thernal Shock." The staff's current method of determining the increase in RT resulting from neutron irradiation damage is documented inRegulatoryGuidgD{.99,Rev.2, December 22, 1983. The amount of neutron irradiation damage, which is predicted using Regulatory Guide 1.99, Rev. 2 nethodology, depends upon the amount of neutron fluence, and the amount of copper and nickel in the material.
In Table I, we have compared the mean plus two standard deviation increase measub predicted by the Regulatory Guide 1.99, Rev. 2 method with that in RT from the Millstone-2 reactor vessel beltline surveillance program.
The test results from the Millstone-2 reactor vessel beltline surveillance progran is reported in Licensee Report TR-N-MCM-008, " Evaluation of Irradiated Capsule W-97," dated April 1987 The prediction method in Regulatory Guide 1.q9, Rev. 2 provides conservative estimates for the effect of neutron ir-radiation of the Millstone-2 reactor vessel beltline materials, because the increase in RT T predicted by the Reaulatory Guide 1.99, Rev. 2 method exceedsthatfromt$esurveillancematerial.
C We have estinated the neutron fluence to be received by the reactor vessel beltline materials by extrapolating to 7 EFPY the neutron fluence estimates in Table 3.1-2 of Licensee Report, "Thernal Shield Damage Recovery Program Final Report." The neutron fluence estimates identified in this report have been reviewed by the staff.
The amount of time that pressure-temperature limits are effective depends upon the amount of material embrittlement at the 1/4 T and 3/4 T vessel l oca tioru. The measurement used by the staff to indicate the amount of materiai enbrittlement is the ad,iusted reference temperature (ART). The ART is the sum of the initial RT increase in RT caused by neutron irradiation danace, and the amoubo,f marain requirkto obtain' a conservative upper bound for neutron enbrittlement for the limiting material. The licensee calculated the ART for the limiting Millstone-2 beltline material usina the methodology in Comission Report SECY-82-465. The licensee indicated that at 7 EFPY, the ART for the limiting Millstone-2 beltline material at the 1/4 T and 3/4 T locations will be 139*F and 123 F, respectively. The staff has used the method documented in Regulatory Guide 1.99, Rev. 2 to estimate
. Table I Comparison of Regulatory Guide 1.99, Rev. 2 Predicted Increase in RT and the Observed Increase in RT NOT NDT Reported for the MNPS-2 Surveillance Material Material Increase in Reference Temperature Mean Plus Two Standard Deviations
.0bserved from MNPS-2 Using the Regulatory Guide 1.99, Surveillance Material Rev. 2 Method ( F)
Tests ( F)
Plate C-506-1 107 96 Weld Flux Lot 3998 155 76 Weld Flux Lot 3999 127 48
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_4-the ART for the limiting Millstone-2 beltline material. This method predicts the ART af ter 7 EFPY of operation will be 133*F and 109 F for the limiting Millstone-? beltline material at the 1/4 T and 3/4 T locations, respectively.
We believe these estimates are conservative, because the increase in RT observed by the surveillance material is less than that predicted by th$DT Regulatory Guide 1.99, Rev. 7 rethod (Table I).
To calculate the water /netal temperature profile durino a heat up or a cool-down, the licensee used a finite element computer code, which was modeled based on one dimensional heat transfer and infinite conductivity at the fluid / clad interface. We have evaluated the metal / water temperature profile which results from the model. We conclude that the model produces a conser-vative metal / temperature profile which may be used in calculating heat-up and cool-down curves.
CONCLUSIONS Based on our analysis, which indicates the licensee has used conservative values for (a) the initial RT of the limiting closure flance and beltline materials, (b) the ART of thekmiting beltline material, and'lc) the water /
metal temperatures profile, we conclude that the licensee's proposed pres-sure-temperature limits meet the safety margins of Appendix G,10 CFR 50, for a period of time correspondinq to 7 EFPY and may be incorporated into the Millstone-2 TS.
The licensee's proposed surveillance capsule withdrawal schedule must meet the requirements of Appendix H, 10 CFR 50. Appendix H,10 CFR 50 requires that the surveillance capsule withdrawal schedule meet the intent of ASTM E-185-82 and provide material test data throughout the life of the plant.
The staff has reviewed the licensee's proposed surveillance capsule with-drawal schedule and concludes that it neets the requirements of Appendix H, 10 CFR 50.
ADDITIONAL INFORMATION During the staff's review of the Licensee Report, " Evaluation of Irradiation Capsule W-97," it was noticed that the baseline unirradiated Charpy V-notch weld test data was generated from two (2) sets of materials. One weld material was fabricated using Linde 0091 flux batch 3998 and wire heat number 90136 and the other weld material was fabricated usina Linde 0091 flux batch 3999 and wire heat number 10137. Since these are materials which have been pre-pared using different heats of wire and batches of flux, their initial pro-perties will be different. The licensee has not considered these differences in evaluating the effect of neutron irradiation on the surveillance weld naterial. We request that the licensee provide separate baseline Charpy V-notch curves for each of these materials so that the staff nay evaluate the effect of neutron irradiation on each weld metal.
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.. Since at 7 EFPY the weld material is not limiting, the effect of neutron irradiation on the surveillance capsule weld material will not affect the staff's conclusion concernino the licensee's proposed pressure temperature limits. However, later in the plant's life, the weld material will become limiting, because its predicted rate of-neutron irradiation damage is greater than that of the plate material. At that time, the effect of neutron irradia-tion on the capsule weld material will become important since the results of these tests will be used to determine the margins required for safe opera-tion of the Millstone-2 reactor vessel.
ENVIRONMENTAL CONSIDERATION We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is in-significant from the standpoint of environmental impact and, pursuant to 10 CFR 451.5(dMa), that an environrental impact statement or necative declaration and environmental inpact appraisal need not be prepared in connection with the issuance of this amendment.
CONCLtlSION We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: April 10, 1984 Principal Contributcr:
- 8. J. Elliot,tiTEB I.
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