ML20012E827
| ML20012E827 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/28/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20012E823 | List: |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, NUDOCS 9004060376 | |
| Download: ML20012E827 (4) | |
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l ENCLOSURE 1 SAFETY EVALVATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION GENERIC LETTER 88-11 RESPONSE BY THE POWER AUTHOP.1TY OF THE STATE OF NEW YORK JAMES A. TIT 2 PATRICK NUCLEAR POWER PLANT i
DOCKET NO. 50-333 INTRODUCTION In response to Generic Letter 88-11. "NRC Position on Radiation Embrittlement of Reactor Yessel Materials and its Effect on Plant Operations " the Power Authority of the State of New York (PASNY or the licensee) submitted revised pressure / temperature (P/T) limit curves for the FitzPatrick Nuclear Power plantTechnicalSpecifications(TS)inaletterdatedJune 30, 1989. The new curves are valid up to 16 Effective Full Power Years (EFPY) and were developed based on Regulatory Guide 1.99, Revision 2.
They provide up-to-date P/T limit curves for operation of the reactor coolant system during heatup, criticality, cooldown and hydrostest evolutions for evaluation in accordance with the GL.
For administrative reasons the actual amendment to incorporate the curves into the TS was submitted separately.
To evaluate the P/T limits, the staff uses the following NRC regulations and guidance: Aspendices G and H of 10 CFR Part 50; the ASTM Standards and the ASMECode,w11charereferencedinAppendicesGandH;10CFR50.36(c)(2);
RG 1.99, Rev. 2; Standard Review Plant (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 Sp(ecifications for the operation of the L
plant. In particular, 10 CFR 50.36 c)(2) requires that limiting conditions of I
operation be included in the Technical Specifications. The P/T limits are among the limiting 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 cescribe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P/T limits. An acceptable method for constructing the P/T limits is described in SRP Section l
5.3.2.
In accordance with the reactor vessel material surveillance test program specimens of the vessel base metal, the weld heat affected zone metal, an,d the I
weld metal from a reactor steel joint which simulates a welded joint in the reactor vessel, along with neutron monitor wires, were placed in capusules I
near the core mid-height prior to initial reactor startup. They were placed l
l on the reactor vessel wall where neutron exposure is similar to that of the vessel wall. Selected groups of specimens are removed at intervals over the lifetime of the reactor and tested to compare mechanical properties with the i
properties of control specimens which are not irradiated.
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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 surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to the American Society of Testing Meterials (ASTM) Standards. These tests define the extent of vessel embrittlement at the time of capsule withdrawal in terms of the increase in reference temperature. Appendix G also require *, the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter 88-11 requested that licensees and permittees use the methods in PG 1.99, Rev. 2, to predict the effect of neutron 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 made from plate, weld, and heat-effected-zone (HAZ) materials of the reactor beltline.
EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the FitzPatrick reactor vessel. The amount of irradiation enbrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 16 EFPY was the seem weld for the lower intermediate shell with (2-233A, B, and C) 0.31% copper (Cu),
0.99% nickel (Ni), and an initial RT of -22'F.
ndt The licensee has removed one surveillance capsule from the FitzPatrick reactor vessel in 1985 during the Reload 6/ cycle 7 refueling outages which corresponded to 5.98 EFPY. The results from capsule I were published in General Electric report MDE-49-0386 and submitted to the NRC by letter dated April 30, 1986 (Reference 5). All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
For the limiting beltline material, the seam weld for the lower intermediate shell (2-233A, B, and C), the staff calculated the ART to be 115'F at 1/4T (T
= reactor vessel beltline thickness) and 83 2*F for 3/4T at 16 EFPY.2 The 2
staff used a neutron fluence of 5.7E16 n/cm at 1/4T and 2.6E16 n/cm at 3/4T.
The ART was determined by the Section 1 of RG 1.99, Rev. 2, because only one capsule has been withdrawn from the FitzPatrick reactor vessel.
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-3 The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 116'T at 16 EFPY at 1/4T for the same limiting weld metal. The staff jud differenceofl'Fbetweenthelicensee'sARTof116*Fandthestaffgesthata 5 ART of t
115'F is acceptable. Substituting the ART of 116*F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for hestup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR i
Part 50.
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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 PO% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload i
must exceed the reference temperature of the materici in those regions by at least 120*F for normal operation and by 90'F for hydrostatic pressure tests and leak tests. Paragraph IV.A.3 of Appendix G states "an eFCeption may be made for boiling water reactor vessels when water level is 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 permissible temperature is 60'F (33'C) above the reference temperature of the closure flange regions that are highly stressed by the bolt preload.' Based on the flange reference temperature of 30*F, the staff has determined that the l-proposed P/T limits satisfy Section IV.2 of Appendix G.
Section IV.B of Appendix G requires that the predicted Charpy Upper Shelf Energy (USE) at end of life be no less than 50 ft-1b. For the materials with l
unirradiated Charpy USE data available, lower intermediate shell plate C3368-1 had the lowest USE of: 103 ft-lb in the longitudinal direction. Using a ratio of 65% to convert longitudinal USE data to transverse USE and applying Figure 2 of RG 1.99, Rev. 2,'it was predicted that the EOL USE would be 57.9 ft-lb. This is greater than 50 ft-lb and, therefore, is acceptable. However, because of l
the vintage of the FitzPatrick plant the licensee has no unirradiated Charpy USE data for the seam weld in tie lower intermediate shell (2-233A, B, and C) andthegirthweld(1-240). The staff will nonitor closely the irradiated USE I
of these welds and other beltline materials.
CONCLUSION The staff concludes that the proposed P/T limits for the reactor coolant system for heatup, cooldown Lt EFPY because the limits conform.k test, and criticality are valid through 20 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, it is appropriate that the proposed P/T Iimits be into the FitzPatrick Technical Specifications.
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REFERENCES 1.
Pegulatory Guide 1.99, Radiation Embrittlerent of Reactor Vessel Materials, Revision 2, May 1988 i
2.
NUREG-0800, Standard Review Plan, Section 5.3.2: Pressure-Temperature Limits 3.
James A. FitzPatrick Nuclear Power Plant, FSAR 4
May 15, 1978 Letter from L. R. Bennett (PASHY) to G. E. Lear (USNRC),
Subject:
Reactor Vessel Material Surveillance Program Data S.
April 30, 1986, Letter from J. C. Brons (PASNY) to D. R. P.uller (USNPC),
subject:
Forwards GE NDE-49-0386, "J. A. FitzPatrick Nuclear Power Plant Reactor Pressure Vessel Surveillance Materials Testing and i
Fracture Toughness Testing" 6.
June 30, 1989, Letter from J. C. Brons (PASHY) to USNRC Docunent Control Desk
Subject:
Response to Generic letter 88-11 5
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