ML20133H780

Safety Evaluation Supporting Amend 70 to License DPR-61

ML20133H780
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	10/16/1985
From:	Office of Nuclear Reactor Regulation
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Shared Package
ML20133H766	List: ML20133H764 ML20133H776 ML20133H780
References
NUDOCS 8510180110
Download: ML20133H780 (5)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l

SUPPORTING AMENDMENT NO. 70 TO FACILITY OPERATING LICENSE NO. OPR-61

, CONNECTICUT YANKEE ATOMIC POWER COMPANY l

HADDAM NECK PLANT l DOCKET NO. 50-213 i

1.0 INTRODUCTION

In a letter from J.F. Opeka to J.A. Zwolinski, dated June 11, 1985, the Connecticut Light and Power Company (the licensee) requested an amendment to the Technical Specifications for the Haddam Neck Plant (HNP). The proposed amendment would update the reactor vessel pressure-temperature limits to 22.0 effective full power years (EFPY). The licensee provided additional information and revised curves in a letter from J.F. Opeka to J.A. Zwolinski dated June 27, 1985.

The last survelliance capsule report submitted to the staff by the licensee was WCAP-10236, " Analysis of Capsule D from the Connecticut Yankee Radiation Surveillance Program." This report was submitted in a letter from W.G. Council to H.R. Denton dated July 6, 1983.

A Notice of Consideration of Issuance of Amendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing related to the requested action was pubitshed in the Federal Reafster on August 14, 1985 (50 FR 32790). No comments or requests for hearing were received.

( 2.0 EVALUATION Pressure-temperature limits must be calculated in accordance with the l requirements of Appendix G, 10 CFR Part 50. Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G, 10 CFR Part 50 are dependent upon the initial RT for the limiting materialsinthebeltlineandclosureflangeregYNsofthereactor vessel and the increase in RT damagetothebeltlinemateri$N.resultingfromneutronirradiation The HNP reactor vessel was procured to ASME Code requirements, which did not specify fracture toughness testing to determine the RT for each reactor vessel material. Hence, the initial RT formateriaY9 Tin the closure flange l andbeltlineregionoftheHNPreactorUNselcouldnotbedeterminedin accordance with the test requirements of the ASME Code. Therefore, the initial RT for these materials must be estimated from test data from other similar maINialsusedforfabricationorreactorvesselsinthenuclearindustry.

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The licensee indicates that the limiting closure flange region materials are the closure flange forgings, which were fabricated to ASME Code SA 336 requirements and were heat treated to the quenched and tempered condition.

Based on their chemical composition these forging are similar to that of ASME Code SA 508 Class 2 material. A conservative estimate of the initial RT of the licensee's closure flange base material may be based upon a coNNrvativeestimateofRT f r quenched and tempered SA 508 Class 2 material. AccordingtoTabYST4.4 of NUREG-0577, " Potential for Low Fracture Toughness and Lamellar Tearing of PWR Steam Generator and Reactor Coolant Pump Supports," the upper bound RT for quenched and tempered ASME SA 508 Class 2 material is 40*F. Thus, aN Nnservative estimate of the RT NDT f r the closure flange region forgings is 40'F The beltline of HNP reactor vessel contains A302B plate materials and weld material, which was fabricated by Combustion Engineering using RAC0 3 weld wire and ARCOS B5 flux. The initial RT N for the weld materials were estimated by the licensee as -56*F with gTstandard deviation of 17'F. These ~

initial RT N and standard deviation values were recommended by the staff in Commission Nport SECY 82-465, " Pressurized Thermal Shock" for welds fabricated by Combustion Engineering which have high initial Charpy upper shelf properties. Since the HNP beltline welds were fabricated by Combustion Engineering and have high Charpy initial upper shelf properties, the initial RT and standard deviation values estimated by the licensee for the HNP beYNineweldsareacceptable.

The increase in RT resulting from neutron irradiation damage was estimated bythelicenseeusYktheempiricalrelationshipdocumentedinCommission Report SECY 82-465. This method of predicting neutron irradiation damage is dependent upon the predicted amount of neutron fluence and the amounts of copper and nickel in the beltline material.

The licensee has used the calculated neutron fluences at 22 EFPY for determining the predicted neutron irradiation damage. The test data reported in the Radiation Analysis and Neutron Dosimetry Section of the Capsule D test report indicates that the calculated neutron fluences are conservative.

The Fe54 (n.p) Mn foil reaction in Capsule D resulted in a neutron flux of 4.03x1080 (E>l MeV) n/cm2 .sec. The calculated neutron flux for this foil reaction was 7.20x1010 (E)1 MeV) n/cm2 -sec. Since the neutron flux from the foil dosimetry is less than the calculated value, the calculated values for the neutron fluencer will be conservative.

In letters dated Novmeber 4, 1977, February 17, 1984 and June 27, 1985 the licensee reported the chemical composition of each beltline material that is in the high flux region of the HNP reactor vessel. Based upon the initial RT and the amounts of copper, phosphorus and nickel in each beltline sabial,andlimitingbeltlinematerialsduring22EFPYofoperationwould be the HNP beltline plates.

Capsule D in the HNP reactor vessel surveillance material program contains material identified as W9807-4, Correlation Monitor, and weld metal. These surveillance materials are representative of the materials in the HNP reactor vessel beltline because they were fabricated to the same specification as the materials in the HNP reactor vessel beltline. Hence, the irradiated material data from the surveillance capsule may be used to demonstrate that the model used to predict the increase in RT resulting from neutron irradiation damageisapplicabletotheHNPreborvesselbeltlinematerials. In Table I we have compared the increase in RT predicted by the method documented in CommissionReportSECY82-465tothNDdbservedincreasefromcapsulematerial.

Since the observed values are less than the predicted values, the prediction method will provide conservative estimates of the increase in RT resulting NDT from neutron irradiation damage.

The staff has used the method of calculating pressure-temperature limits in USNRC Standard Review Plan 5.3.2, NUREG-0800, Rev.1, July 1981, to evaluate the proposed pressure-temperature limits. The amount of neutron irradiation damage to the beltline materials was estimated using the method recommended by the staff in Commission Report SECY 82-465. Our conclusion is that the proposed pressure-temperature limits meet the safety margins of Appendix G, 10 CFR Part 50 for 22 EFPY and may be incorporated into the plant's technical specifications.

The licensee has also requested that the curves establishing the maximum pressure for RHR System operation and minimum pressure for operation of a reactor coolant pump be deleted from the technical specifications. The licensee stated that these curves are not required or applicable to Appendix G compliance but rather provide operational restrictions for the plant operators.

,g The licensee also indicated that these restrictions were incorporated in the plant operating procedures and therefore were no longer necessary in the technical specifications.

The staff has reviewed the licensee's request to remove the curves described above and concludes that the restrictions on plant operation will be unchanged and that control of operations within those restrictions will be by procedures subject to the Administrative Controls provisions of the Technical Specifications.

Therefore, we conclude that the proposed changes are acceptable.

4 Table I l

Increase in RT NOT for Capsule D Surveillance Material Surveillance Material Neu Increase in RT NDT Increase in RT x10}gonFluence n/cm2 Observed from Predicted by NOT (E>l Mov) Capsule test data Comm. Report

• (*F) SECY 82-465 Method (*F)

W9807 (Plate) 2.22 78 116 1.58 67 110

.471 80 93 Correlation Monitor 2.22 140 168 (Plate) 1.58 127 157 '

.471 80 127

.239 85 114 Weld Metal 2.22 110 168

.239 95 114 b

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3.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change to a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(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 this amendment.

4.0 CONCLUSION

The staff has 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.

5.0 ACKNOWLEDGEMENT This Safety Evaluation has been prepared by Mr. B. Elliot, and F. Akstulewicz.

Dated: October 16, 1985 l

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