ML20153G050
| ML20153G050 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 04/26/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20153G031 | List: |
| References | |
| NUDOCS 8805110125 | |
| Download: ML20153G050 (6) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION e
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I WASHINGTON, D. C, 20665
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEN 0 MENT NO.120TO FACILITY OPERATING LICENSE NO. OPR-46 NEBRASKA FLPLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298
1.0 INTRODUCTION
In a letter fren L. G. Kunc1 to USNRC dated October 28, 1987 and supplemen-ted by a letter from G. A. Trevors to USNRC dated February 22, 1988, the Nebraska Public Power District (the licensee) proposed to amend Facility Operating i.icense No. DPR-46 for the Cooper Nuclear Generating Station (Cooper). The amendment proposes to revise the reactor coolant system pressure-temperature linits and surveillance capsule withdrawal schedule, which are contained in Section 3.6 and 4.6 of the Cooper Technical Spect-fications (TS). The RT Shift Curve (Figure 3.6.1) is to be deleted, Non-NuclearHeatup/ Coo 1EnCurve(Figure 3.6.1.a)andCoreCritical Curve (Figure 3.6.1.b) are to be applicable for 12 effective full power years (EFPY), and Pressure Test Curves are to be applicable for 8, 10 and 12 EFPY (Figure 3.6.?).
The licensee proposes to revise the capsule withdrawal schedule to recuire withdrawal of the next capsule at 15 EFPY, and the remaining capsule at 32 EFPY. The bases for these changes are the test results from the Cooper surveillance program, which are contained in a letter from G. A. Trevors to USNRC dated July 6, 1987, 2.0 DISCUSSION Pressure-Temperature Limits: Pressure-Tenperature limits must be calcu-lated in accordance with the requirements of Appendix G,10 CFR Part 50, which became effective on July 16, 1983.
Pressure-Temperature limits that are calculated in accordance with the requirements of Appendix G,10 CFR Part 50 are dependent upon the initial reference temperature (RTthe l for o
i reactor vessel and the increase in reference temperature resulting from neutron irradiation damage to the limiting beltline material. The Cooper reactor vessel was procured to earlier ASME Code requirements, which did not specify fracture toughness testing to detemine the initial RT for each vessel material. AptendixG,10CFRPart50indicatesthatVMeis l
4 fabricated to earlier ASPE Code requirenents rust provide supplementary data and analyses to demonstrate that the vessel raterial's fracture toughness data and raterial analysis requirements are equivalent to that specified in later editions of the ASME Code.
l The Cooper reactor vessel was fabricated by Combustion Engineering (CE).
I The beltline was fabricated by welding plates together and the closure flange regions were fabricated by welding plates and forgings together.
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The initial RT for plate materials vas deterinined by extrapolating the existingdata$IngacalculationrethoddevelopedbyGeneralElectric (GE). The GE method is based on test results from 24 plates reported in Welding Research Council Bulletin ?17 and from 22 plates reported in the The initial RT for the forging naterials was deterrined LaSalle FSAR.
using the method recomended b,0Tthe staff in NRC Branch Technical Position MTEB 5-2.
This branch technical position is documented in Standard Reviev Plan 5.3.2, "Pressure-Temperature Limits" of NUREG-0800, Rev. 1, July 1981. The initial RT used for weld me.tals was the two standard deviation upper bound value use$$y the staff for CE weld metals in SECY-82-465, "Pressurized Thermal Shock." These methods result in an initial RT for tFe liniting beltline base retal and weld metal of 14'F, and -22'F,NDT for the limiting closure flange region respectively, and an initial RTNDT raterial of 20'F.
The increase in RT resulting from neutron irradiation damage was estimatedbythe1Enseebyextrapolatingthesurveillancedataatthe rate documented in Regulatory Guide (R.G.) 1.99, Rev. 1. "Effe(.ts of Residual Elements on Predicted Radiation Darage to Reactor Vessel Materials."
This method of predicting neutron irradiation damage is dependent upon the predicted arount of neutron fluerce and the amounts of residual elements in the beltline raterials. The neutron fluence predictions were upper bound estiretes, which were calculated using measurements from passive neutrcn flux renitors and by analysis, which was made with the DOT two-dirensienal discrete ordinate code and the SN1D one-dirensional computer code.
Inputs into the analysis included 26 neutrcn energy groups, cross-sections from ENDF B-IV, P3 expansion of the scattering cross section, average reasured neutron spectra for BWRs at the GE Test Reactor at Vellecitos, and power distributiens representative cf time-averaged i
ccnditions derived from Cooper plant specific cycles. The neutron spectra used in the analysis are docurented in NEDO-24793, which is contained in the licensee's letter dated February 22, 1988.
The measured increase in reference temperature for the Cooper surveillance raterials are compared in the table belew, to the values predicted using R.G. 1.99, Rev. I and R.G. 1.99, Rev. 2.
which has been approvel and is traiting publication as a final guide. The increase in reference ten-perature measured from the surveillance raterial significantly exceeds the values predicted using the formula in R.G.1.99, Rev.1. The increats i
in reference terperature for the weld retal is less than the value pre-dicted using the method in R.G. 1.99, Rev. 2.
The increase in reference temperature fer the plate material is slightly greater than the value predicted using the method in R.G. 1.99, Rev. 2.
The surveillance data indicates that R.G. 1.99, Rev. I underpredicts the effect of neutron irradiation en the Cooper beltline material, while R.G. 1.99, Rev. 2 conservatively predicts the effect of neutron irradietion of the beltlire veld metal and slightly underpredicts the effect of neutron irradiatier en the Cooper beltlire plate material.
e 5
O e i Surveillance Capsula Test Results Material Increase in Increase in
- Increase in Ratio of Ref. Temp.
Ref. Temp.
Ref. Temp.
Measured to Measured from Predicted by Predicted by Predicted by Surveillance R.G. 1.99, R.G. 1.99, R.G. 1.99, Material Rev. 1 Rev. 2 Rev. 1
(*F)
('F)
('F)
('F)
Plate 74 31 69 2.39 Weld Petal 55 34 83 1.62 l
- Increase in Ref. Terp. are rean plus two standard deviation values The Pressure-Temperature linits preposed by the licensee were calculated using the increase in reference terperature fomula in R.C.1.99, Rev. I with a correction to account for the underpredication of this nethod compared to the surveillance material. The fenrula in R.G.1.99, Rev.1 consists of a chemistry factor and a fluence factor. The licerses increased the cher.istry factor in this ferrula by the ratio of the reasured increase in reference terperature from the surveillance naterial to the values predicted by the fomula in R.G.1.99, Rev.1 (this ratio is reported in thelastcolurninTable1). The licensee's method of predicting the increase in reference temperature results in adjusted reference temperature (ART) values for the liniting beltline raterial of 110*F, 102'F and 93'F at 12 EFPY, 10 EFPY and 0 EFPY, respectively. The ART is the sum of the initial RT and the increase in reference temperature resulting fren j
neutronir$1ation. The licensee indicates that this rethod results in a predicted final end-of-life ART (the ART at 32 EFPY) value of 171'F for the 1
limiting beitline material.
The ART values for the limiting beltline raterial using the femule in R.G. 1.99, Rev. 2 are 95'F, 91 F, and 86*F at 12 EFPY, 10 EFPY and 8 EFPY, l
respectively. The final AP.T value for the liriting beltline material using the fonnula in R.G.1.99, Rev. 2 is 144*F.
Since the ART values used by the licensee to calculate the proposed Pressure-Terperature limits are greater than the values predicted using thp fomula in R.G. 1.99, Rev. 2, the proposed Pressure-Ter.perature limits will reet R.G. 1.99, Pev. 2.
To confim that the Pressure-Terperature limits proposed by the licersee
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will meet the safety margins of Apperdix G, 10 CFR Part 50 for the proposed operating periods, the staff has used the method of calculating Pressure-4 Terperature limits in USFPC Standard Review Plan 5.3.2, NUREG-0800, Pev. 1, July 1981 to evaluate the proposed Pressure-Temperature limits. The staff's
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cattulatien includes the licensee's ART values. Our calculations confim l
that the proposed Precsure-Temperature limits meet the safety rargins of Appendix G, 10 CFR Part 50 for the operating periods identified on the curvis.
Surveillance Program: The amendment request includes editorial charges to the Technical Specifications requirements for the withdrawal schedule for the remaining two capsules. The existing Technical Specifications define two withdrawal schedules, both stated in terms of service life. One schedule is for use based on the ad,i,usted reference tenperature not ex-ceeding 100 deg. F over the life of the vessel. The second schedule is to be used in the event surveillance specimens indicate a shift of the Charpy V-notch fracture energy curve greater than predicted. Since the test data being "cleaned-up"greater than predicted, the Technical Specifications indicetes a shift requirements. This is a simple editorial change to the Technical Specifi-cations involving no change in the surveillance program requirements.
Also, the Technical Specifications are being revised to specify capsule remeval intervals in terms of "EFPY", instead of "service life".
Since the service life is defined by ASTM E-185-80 as 32 EFPYs, and the revised intervals in tems of EFPYs correspond to the original intervals in tems of service life, this is also a simple editorial change in actual surveil-lance program requirerents. Because these changes are considered editerial they are acceptable.
The staff believes that changes should be made to the surveillance procran.
Thepresentwithdrawalscheduleisbasedonoriginalassumptionsthat(1) the increase in reference temperature resulting from neutron expcsure would te less than 100 deg.
F., and (2) the surveillance specimens would receive greater fluence than the vessel wall. The licensee's analysis indicates that the surveillance spectren neutron exposure lags the vessel wall material. The licensee's analysis and an independent staff analysis indicate that, as noted above, the increase in reference terperature will be greater than 100 deg. F at end of life. Since the original assumptions were incorrect, the surveillance plan should be revised. Appendix H. 10 CFR Part 50 requires, to the extent practical, that the capsule withdrawal program meet the requirements of ASTM E-185-8?. When the predicted increase in reference terperature is greater than 100*F, ASTM E 185-87 recomends that 4 capsules be withdrawn and the surveillance program renitor the long tem effects of neutron irradiatien. The proposed techni-cal specifications indicate that the removal and analysis of the remaining capsules is for the second to be removed at 15 EFPY, and the third ar.d last to be removed at 32 EFPY. Our findings are that irradiation damage is in excess of predictions using Regulatory Cuide 1.90 Rev. 2 criteria, irra-diction danage will exceed 100 deg. F, and the surveillance specimens are receiving less expesure than the vessel wall. Therefore, to assure rain-tenance of safety margins beyond 12 EFPY and to support possible life exten-sion, the staff recomends that the schedule for withdrawal of the second capsule should te accelerated to 12 EFPY and the schedule for withdrawal of the third should be detemined based on the findirgs of analysis of the second capsule.
In addition, the licensee should begin planning for o
e e
5 possible insertion of a fcurth capsule into the Ceeper reactor vessel, per-haps with reconstituted specimens from the first capsule. The staff wi?1 request the licenseo to review the surveillence program in consideration cf these reccrrendations.
l 3.0 EVALUATION Based on our review, we find that:
1.
The data and analysis provided by the licensee demonstrate that the vessel naterial's fractura toughness is equivalent to that specified in later editions of the ASME Code and the initial RT values proposedfortheCooperreactorvesselnaterialareabptablefor use in calculating the Cooper Pressure-Temperature limits.
2.
The licensee's method of calculating neutron fluence is acceptable and may be used to predict the increase in reference temperature resulting frem neutron irradiation.
3.
Since ART values used to calculate the Pressure-Temperature limits were derived from the surveillance material test results and are greater than the values calculated using the rethod documented in R.G. 1.99, Rev. ?, the proposed Pressure-Terperature limits adequately account for neutron irrcdiatien.
4.
Based on the above findings and the staff's ccnfimatory calculaticns.
l the propeted Pressure-Temperature limits meet the safety rargins of Appendix G, 10 CFR Part 50.
5.
The proposed changes to the capsule withdrawal setetule are editorial i
in nature.
6.
The proposed Pressuro-Terperature lirits and capsule withdrawal schedule ray be incorporated into the Cooper Technical Specificatiens.
j However, the surveillance program shculd be prorptly reevaluated to pemit early developrent of plans to assure maintenance of safety rargins for operation teyend I? EFPY.
4.0 FfWIpCtmENTAL CONSIDERATION The amendment involves a change in the installation or use of a facility cerperent located within the restricted area as defired in 10 CFR Part 20.
The staff has detemined that the amerdront involves no sienificant increase in the amounts, and nn significant change in the types, of any eft 1uents that ray be released offsite, and that there is no significant increase in individual or curulative occupational radiation exposures.
The Ccerissien has previously issued a proposed finding that the amendment I
involves r.c sienificant harstds consideratien and there has been no public ccrrentonsuchfinding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.22(c)(9).
Pursuantto10CFR51.22(b),noenvironmentalimpactstatementorenviron-nental assessment need be prepared in connection with the issuance of the amendment.
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5.0 CONCLUSION
The staff has concluded, basod on the considerations discussed above, that:
will not be endangered by operation in the proposed manner, and (2) p (1) there is reasonable assurance that the health and safety of the
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such activitiet will be conducteu in compliance with the Comission's regulations.
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j and the issuance of the amendment will not be inimical to the comon defense i
and security or to the health and safety of the public.
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Date:
April 26,1988
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Principal Contributors:
- 8. Elliot/W Long/L. Lambros i
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