ML20043A890
| ML20043A890 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/11/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20043A889 | List: |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, NUDOCS 9005230244 | |
| Download: ML20043A890 (5) | |
Text
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t SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
RELATED TO AMENDMENT NO. 54 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 53 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT. UNIT NOS.1 AND 2 DOCKET NO.50-27S AND 50 323 1
1.0 INTRODUCTION
By letter dated July 7,1989 (Reference LAR 89-07), Pacific Gas and Electric Com)any-(PG&E or the licensee) requested amendments to the combinedTec1nicalSpecifications(TS)appendedtoFacilityOperating LicenseNos.OPR-80andDPR-82fortheDiabloCanyonPowerPlant(DCPP)
Unit Nos. I and 2, respectively.
The amendments change the TS to revise the heatup and cooldown curves and delete the surveillance capsule i
withdrawal schedule.
l Therevisedheatupandcooldowncurveswerecalculatedusingmethods described in Revision 2 to NRC Regulatory Guide (RG) l'.99, Radiation Embrittlement of Reactor Yessel Materials," as recommended by Generic Letter (GL)88-11,topredicttheeffectofneutronradiationonreactor vessel materials while continuin9 to meet the requirements of 10 CFR Part 50, Appendix >G.
Thepreviouspressure-temperature (P/T) limit curves were applicable through 6 effective full power years (EFPY) of plant operation. -The revised _TS provide up-to-date P/T limits for the t
operation of the reactor coolant system during heatup, cooldown, I
criticality, and hydrotest. The revised TS use one set of P/T limits for both units, and are applicable through 8 EFPY, To evaluate the Diablo Canyon P/T limits, the staff used the following l
NRC regulations and guidance:
A.
Appendix G of 10 CFR Part 50, Fracture Toughness Requirements."
8.
Appendix H of 10 CFR Part 50, " Reactor Vessel Material Surveillance Program Requirements."
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C.
The-ASTM Standards and the ASME Code, which are referenced in Appendices G and H to 10 CFR Part 50.
D.
Standard Review Plan (SRP) Section 5.3.2, " Pressure-Temperature Limits" (NUREG-0800, July 1981).
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E.
Regulatory Guide 1.99 " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988.
F.
Generic Letter 88-11, *NRC position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations,"
July 12, 1988.
Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide technir.41 specifications for the operation of the plant.
In particular,10 CFR 50.36(c)(2) requires'that limiting conditions of operation be included in the technical specifications.
The P/T limits are among the'11miting conditions of operation in the technical specifications for all commercial nuclear plants in the United States. Appendices G and H of 10 CFR Part 50 describe 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 5.3.2.:
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and requires 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 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 requires 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 RG 1.99,-
4 Rev. 2, to materials. predict the effect of neutron irradiation on reactor vessel 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 r egires 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 s.
startup and that they contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
The staff evaluation of the changes proposed in the licensee's letter of July 7 1989 is given below. The staff's proposed determination of no signifIcanthazardsconsiderationforthisTSchangewaspublishedin the Federal Register ori August 9,1989 at 54 FR 32715.
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2.0 EVALUATION The NRC staff has evaluated the proposed changes and finds them-acceptable, based on its review of the analyses and evaluations presented by the licensee. A discussion of each of the specific
' technical specification changes made by these amendments is given below.
l These amendments make changes in three areas.
Each area is covered
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below as a separate item.
Item A Figure 3.4-1, " Reactor Coolant System Heatup Limitations," and Figure 3.4-3, " Reactor Coolant System Cooldown Limitations,"
are revised to update the controlling chemical composition and
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P/T curves.
In reviewing these changes, the staff evaluated the effect of neutron irradiation embrittlemynt on each beltline material in the Diablo Canyon 1 and 2 reactor vessels.
The amount of irradiation embrittlement was l
calculated in accordance with RG I.99, Rev. 2.
The staff has determined that the material with the highest ART at 8 EFPY for both units is intermediate shell plate B5454-2 in Unit 2.
This plate is 0.14% copper (Cu)and0.59% nickel (Ni),andhas'aninitialRT o 6 F.
ndt The licensee has removed one surveillance capsule'each from Diablo' i
Canyon Units 1 and 2.
The results from capsule 5 from Unit 1 were published in the Westinghouse Topical Report WCAP-11567. The results from capsule U from Unit 2 were published in Westinghouse Topical Report WCAP-11851. All surveillance capsules contained Charpy impact specimens and tensile. specimens mede from base metal, weld metal, and HAZ metal.
For the limiting beltline material, plate B5454-2, the staff. calculated the ART to be 163*F at 1/4T (T = reactor vessel beltline thickness) and 142'F for 3/4T at 8 EFPY. The ART was determined using Section 1 of RG 1.99, Rev. 2.
The licensee used the method in RG 1.99, Rev. 2 to calculate an ART of 164'F at 8 EFPY at 1/4T for the same limiting plate material. The staff judges that a difference _of l'F between the licensee's ART of 164F*F and the staff's ART of 163'F is acceptable. The licensee used a technique based on Section XI of the ASME Boiler and Pressure Yessel Code to calculate the Diablo Canyon 1 and 2 P/T limits. The staff has reviewed this technique and finds it acceptable, although it produces a P/T curve that-is lower than the one the staff calculated using the methods of SRP 5.3.2.
Thus, the staff verified that the proposed P/T limits for i
heatup, cooldown, and hydrotest meet the requirements of Appendix G of-10 CFR Part 50.
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4 In addition to beltiine materials,ference temperature.for the reactor Appendix G of 10 CFR Part 50 also imposes P/T limits based on the re 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 that are highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120'F for normal operation and by 90*F for hydrostatic pressure tests.and leak tests.
Based on the flange reference temperature of $3*F, the staff has determined that the proposed 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-1b. Of'the materials that have unirradiated Charpy USE data available, plate B4106-3 in Unit I had the lowest value, 77.5 ft-lb.
Using the method in RG 1.99 Rev. 2, the predicted Charpy USEoftheplatematerialattheendoflifewillbe59.7ft-lb. This is greater than 50 ft-lb and, therefore, is acceptable. However there are no unirradiated Charpy USE data for. the intermediate to lower, shell girth weld in Unit 1 and the intermediate to lower shell girth weld and 1
Iower shell longitudinal weld seams in Unit 2.
The staff will obtain the USE of these welds in the future.
In summary, the NRC staff has concluded that the revised P/T limits for the Diablo Canyon Units 1 and 2 reactor coolant system for heatup, cooldewn, leak test, and criticality'are valid through 8 EFPY 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. Consequently, the proposed P/T limits are hereby incorporated into the Diablo Canyon 1 and 2 Technical Specifications.
Also, the' staff finds the updated controlling chemical composition acceptable because it reflects measurements made on the limiting reactor vessel plate. Therefore, the TS changes identified under Item A, above, are acceptable.
Item B TS 4.4.9.1.2 and TS Table 4.4-5, ' Reactor Yessel Material Surveillance Program Withdrawal Schedule," are deleted.
The staff has reviewed the proposed deletion of TS 4.4.9.1.2 and the associated reactor vessel surveillance capsule withdrawal. schedule (TS Table 4.4-5), and finds these changes acceptable on the basis that they are administrative rather than substantive. That is. 10 CFR 50, Appendix H, requires that a withdrawal schedule must be established and submitted to the NRC for approval prior to its implementation. The licensee has committed to include the currently approved withdrawal schedule in the next FSAR Update Revision. Repeating this requirement in the Technical Specifications is unnecessary and redundant. Therefore, the TS changes identified under Item B, above, are acceptable.
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I Item C TS Bases 3/4.4.9 are revised.to update the information I
contained therein, and to delete certain figures and tables i
that will be included in the FSAR Update.
Deletion of these tables and figures from the TS Bases is an administrative change that reflects the other TS changes made by these amendments. Therefore the TS changes under item C, above, is acceptable, i
3.0 ENVIRONMENTAL CONSIDERATION
These amendments involve changes to a requirement with respect to the installation or use of facility components located within the restricted e
area as defined in 10 CFR Part 20 and a change in surveillance requirements. At Diablo Canyon, the restricted area coincides with the site boundary.
We have determined that the amendments-involve no significant increase in the amounts, and no significant change in the i
types, of any effluents that may be released offsite, and that there is no significant increcse in individual or cumulative occupational radiation exposure. The Commission has previously-issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding.
Accordingly theseamendmentsmeettheeligibilitycriteriaforcategoricalexclusion setforthin10CFR51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no-environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.
4.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the p(2) such activities will be conducted in compliance with theublic will not be Commission'sregulationsand(3)theissuanceoftheseamendmentswill i
not be inimical to the common defense and security or the health and safety of the public.
Principal Contributors: John Tsao L
Harry Rood j
Dated: May 11, 1990 1
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