ML19332D392
| ML19332D392 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 11/27/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19332D389 | List: |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, NUDOCS 8912010127 | |
| Download: ML19332D392 (3) | |
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, ENCLOSURE
-SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO GENERIC LETTER 88-11 RESPONSE LO_U_lSIANA. POWER AND LIGHT COMPANY WATERFORD UNIT 3 DOCKET NO. 50-382
'1.0 ; INTRODUCTION
'In response to Generic Letter 68-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Effect on Plant Operations,"
the Louisiana Power and Light (the licensee) requested permission to retain the pressure /ter.perature (PT) limits in the Waterford Unit 3 Technical Specifications, Section 3.4.
The request was docun'ented in a letter from the licensee dated November 1,1988. This revision also n.aintains the applicability of the current P/T limits to 10 calendar
' years. The P/T limits were developed based on Section 1 of Regulatory Guide (RG) 1.99, Revision 2, and provide up-to-date P/T limits for the operation of the reactor coolant system during heatup, cooldown, criticality, and hydrotest.
Tc evaluate the P/T limits, the staff uses the following NRC regulations and l
guidance: Appendices G and H of 10 CFR Part 50; the ASTM Standards and the ASME 1
Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2); RG 1.99, L,
Rev. 2; Standard Review Plan (SRP) Section 5.3.2; and Generic Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide Technical Specifications for the operation of the plant.
In particular,10 CFR 50.36(c)(2) requires that limiting conditions of operation be L
included in the Technical Specifications. The P/T limits are among the limiting l.
conditions of operation in the Technical Specifications for all commercial nuclear
. plants in the U.S.
Appendices G and H of 10 CFR Part'50 described specific re-quirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P/T limits. An acceptable method for constructing 7
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the P/T limits is described in SRP Section 5.3.2.
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Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements
'for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the sarveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards.
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These tests cefine the extent of vessel embrittlement at the time of capsule L
. withdrawal in terms of the increase in reference temperature. Appendix G also requires the licensee to predict tht effects of neutron irradiation on vessel 8912010127 891127 PDR ADOCK 05000382 I
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embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter 88-11 requested that licensees and permittets use the methods in RG 1.99, Rev. 2, to predict the effect of neutron l
. irradiation on reactor vessel materials. This guide defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin.to account for uncertainties in the prediction method..
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw-surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens madefromplate, weld,andheat-affected-zone (HAZ)materialsofthereactor beltline.
I 2.0'EVALUAQ0N The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Waterford 3 reactor vessel. The arount of irradiation erbrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 10 calendar years was the 4
lower shell plate (M-1004-2) with 0.03% copper (Cu), 0.58% nickel (N1), and an initial RT f 22*F.
ET The licensee has not removed any. surveillance capsules from the Waterford 3 reactor pressure vessel. All six surveillance capsules contained Charpy irpact i
specimers and tensile specimens made from base metal, weld metal, and HAZ metal.
1 For tre limitirg beltline material, plate M-1004-2, the staff calculated the ART to be 55.2'F at 1/4T (T = reactor vessel beltling thickness) for 10 calendar years.
The staff used a neutron fluence of 0.92E19 n/cm at 1/4T. The ART was determined by Section 1 of RG 1.99, Rev. 2, because no surveillance capsules have been removed from the Waterford 3 reactor pressure vessel.
The licensee used the method in RG 1.99, Rev. 2, to calculated an ART of 88*F for 40 calendar years at 1/4T for the same limiting plate material.
Substituting the l
conservative ART of 88'F into equations in SRP 5.3.2, the staff verified that the existing P/T limits for heatup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
The current P/T limits are conservative enough to be valid for perhaps 40 calendar years; however, the staff only evaluated the P/T limits for 10 calendar years because that is the applicable period of the current P/T limits.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.Section IV.2 of Appendix G states that when the pressure exceeds 20%
of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference
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temperature of the material in those regions by at least 120*F for normal opera-tion and by 90'F for hydrostatic pressure tests and leak tests.
Based on the flange reference temperature of 20'F, the staff has determined that the P/T limits satisfy Section IV.2 of Appendix G.
Section IV.B of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft.-lb.
The material with the lowest unirradiated USE is intermediate shell plate M-1003-3 with 138 ft.-16. in the longitudinal' direction. Based on its 0.02% Cu and Figure 2 of RG 1.99, Rev. 2 the end-of-life (E0L) USE is predicted to be 102 ft.-lb.
To predict the EOL USE in the transverse direction, the 102 ft.-lb. was multiplied by 0.65 per SRP 5.3.2.
This resulted in a predicted E0L USE in the transverse direction of 66.4 ft.-lb.
This is greater than 50 ft.-lb.
and, therefore, is acceptable.
3.0 CONCLUSION
The staff concludes that the existing P/T limits for the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 10 calendar -
years because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The licensee's submittal also satisfies Generic Letter 88-11 because the licensee used the method in RG 1.99, Rev. 2, to calculate the ART.
Hence, the P/T linnts in the current Waterford 3 Technical Specifications need not be changed.
Principal Contributor:
John Tsao, EMCB/DET Date: November 27, 1989 i
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