ML18093B323

Forwards Response to Generic Ltr 88-11, Radiation Embrittlement of Reactor Vessel Matls. Util Will Submit License Amend Requests to Incorporate New Heatup & Cooldown Curves in Tech Specs for Salem Units 1 & 2 by Dec 1988

ML18093B323
Person / Time
Site:	Salem, Hope Creek, 05000000
Issue date:	11/28/1988
From:	MILTENBERGER S Public Service Enterprise Group
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, NLR-N88195, NUDOCS 8812050041
Download: ML18093B323 (7)

Text

Public Service Electric and Gas Company Steven E. Miltenberger Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4199 . Vice President and Chief Nuclear Officer November 28, 1988 NLR-N88195 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemen:

RESPONSE TO NRC GENERIC LETTER 88-11 RADIATION EMBRITTLEMENT OF REACTOR VESSEL MATERIALS SALEM AND HOPE CREEK GENERATING STATIONS DOCKET NOS. 50-272, 50-311, AND 50-354 Generic Letter 88-11 forwards Regulatory Guide 1.99, Revision 2 for implementation by reactor licensees.

Therein, the Commission requests that the results of analyses performed*in accordance with Rev. 2 of the guide be submitted for Commission review. A proposed schedule for implementation of any actions required as a result of the changes in methodology contained in Rev. 2 is also requested.

Additionally, for Pressurized Water Reactor plants, the Commission identifies the need to review the Low Temperature Overpressure Protection system for potential changes resulting from pressure/temperature limits established using Revision 2 of the guide. This latter issue is applicable to Salem Units 1 and 2 and is addressed in Enclosure 1 to this transmittal.

Accordingly, Public Service Electric and Gas Company hereby forwards Enclosures 1 and 2 in response to the subject Generic Letter. Enclosure 1 pertains to Salem Units 1 and 2. Enclosure 2 is specific to Hope Creek. If there are any questions regarding the enclosed information, please feel free to contact us. Sincerely, Enclosures

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Document Control Desk C Mr. G. W. Rivenbark Licensing Project Manager Mr. J. C. Stone Licensing Project Manager Ms. K. Halvey Gibson 2 Senior Resident Inspector

-Salem (Acting) Mr. G. W. Meyer Senior Resident Inspector

-Hope Creek Mr. W. T. Russell, Administrator Region I Ms. J. Moon, Interim Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 11-28-88 ENCLOSURE 1 PSE&G RESPONSE TO THE NRC GENERIC LETTER 88-11 SALEM UNITS 1 & 2 (1) Results Summary PSE&G has assessed the impact of Regulatory Guide 1.99, Rev. 2 and the attached Regulatory Guide 1.99, Rev. 2, on the pressure temperature limits contained in the Technical Specifications for Salem Units 1 and 2. This review was completed along with the analyses performed to monitor the changes in reactor vessel material fracture toughness based on removal of surveillance capsule Z during Salem Unit l's 7th refueling outage and removal of capsule U during Salem Unit 2's 3rd refueling outage. Results of the technical analysis which was conducted using the methods described in Regulatory Guide 1.99, Rev. 2, were submitted to the NRC previously in Westinghouse reports WCAP 11544 and 11955. These reports are specific to Salem Units 1 and 2. The heatup and cooldown curves generated as part of the Westinghouse analyses referenced above, reflect the pressure/temperature (P-T) limits for Salem Units 1 and 2 and are based on the requirements of Appendix G to 10CFR50 and Regulatory Guide 1.99, Rev. 2. The P-T curves for Salem Unit 1 were prepared by Westinghouse in September 1988. The P-T curves for Salem Unit 2 were prepared by Westinghouse in November 1987 using the draft version of Regulatory Guide 1.99, Rev. 2 available at that time. Westinghouse has reviewed the Salem Unit 2 P-T curves and concluded that they are in compliance with the final version of Regulatory Guide 1. 99, Rev. 2. Additionally, the low temperature overpressure protection (LTOP) system for Salem Units 1 and 2 has been reviewed to determine the impact of the revised P-T curves on LTOP setpoint, enable temperature, system hardware, and applicable operating procedures.

The results of that review are summarized in the following paragraphs.

At Salem Units 1 and 2, this system is referred to as the Pressurizer Overpressure Protection System (POPS). The POPS is a two train system which uses separate and independent pressure transmitters to open two pressurizer relief valves if RCS pressure exceeds a preset value of 375 psig. The POPS is required to be armed whenever the RCS is below 312°F. The POPS relief valves protect the RCS from pressure transients exceeding the limits of Appendix G to 10CFR50 when one or more RCS cold leg temperature is at or below 312°F. Either POPS Power Operated Relief Valve (PORV) has adequate relieving capacity to protect the RCS from 1 e Enclosure 1 (Continued) overpressurization as a result of the limiting heat input or mass input cases; i.e. (1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50°F above RCS cold leg temperature or (2) the start of a safety injection pump and its injection into the water solid RCS. Several provisions presently exist for prevention of pressure transients when the RCS temperature is below 312°F. Current Technical Specification 3.4.1.3 for startup of an RCP requires that a steam bubble must be established in the pressurizer prior to pump start or the SG/RCS delta-T be verified to be less than 50°F. Also, Technical Specification

3.5.3 allows

a maximum of one safety injection pump to remain operable and power to all inoperable safety injection pumps must be removed by racking out the power supply breakers when the RCS temperature is below 312°F. Also the shutdown procedure requires that a steam bubble be maintained in the pressurizer during plant cooldowns.

The Residual Heat Removal (RHR) System is put into service once the RCS temperature is below 350°F but above 312°F. The RHR system provides relief capacity comparable to that of a POPS valve. However, no credit has been taken in the low temperature overpressure analysis for this relief capacity.

Based on the above analysis, it is determined that changes to the pressurizer overpressure protection system are not required as a result of the revised P-T limits. (2) Schedule For Implementation of R.G. 1.99, Rev. 2 As indicated previously, R.G. 1.99, Rev. 2 has been used for the preparation of revised P-T curves and operating limits for Salem Units 1 and 2. PSE&G will submit License Amendment Requests to incorporate new heatup and cooldown curves in the Technical Specifications of Salem Units 1 & 2 by December 1988. As part of those License Amendment requests, appropriate changes to the Bases section will also be provided.

Any required modifications to the Salem Updated Final Safety Analysis Report will be submitted as part of the next routine annual update presently scheduled for July 1989. 2 ENCLOSURE 2 PSE&G RESPONSE TO NRC GENERIC LETTER 88-11 HOPE CREEK (1) Results Summary The impact of implementing R.G. 1.99 Rev. 2 can best be determined by comparing the RTNDT adjusted reference temperature (ART) values based on R.G. 1.99, Rev. 1 and R.G. 1.99, Rev. 2. Table 1 shows the ART values at 4 EFPY and at 32 EFPY for each beltline material in Hope Creek. These EFPY values are selected to be representative of conditions existing early in the operating life and near the end of the operating life of the plant. The following conclusions are drawn from the results in this table: (a) The R.G. 1.99, Rev. 2 ART values at 32 EFPY for this unit are below 200°F, which is the allowable limit in 10CFR50, Appendix G. Therefore, implementation of R.G. 1.99, Rev. 2 will not result in any additional analysis, testing or provisions for thermal annealing. (b) The ART value which applies to the pressure-temperature (P-T) curves in the Technical Specifications is 39°F at 32 EFPY as determined using R.G. 1.99, Rev. 1 methods. The maximum R.G. 1.99, Rev. 2 ART value in Table 1 is 65.8°F at 32 EFPY. Therefore, at 32 EFPY the R.G. 1.99, Rev. 1 P-T curves are less conservative than P-T curves that would be generated with R.G. 1.99, Rev. 2. However, the current P-T curves are applicable up to 6.7 EFPY even if the ART is calculated according to R.G. 1.99, Rev. 2 methods. Therefore, the P-T curves presently contained in the Technical Specifications are conservative for several more years of operation. (c) The worst case LPCI nozzle and weld are also included in this beltline region analysis due to their predicted fluence value at 32 EFPY. Since the maximum ART of 13.1°F at 32 EFPY based on Rev. 2, is less than the 40°F RTNDT value applied to the limiting vessel discontinuity curves, the LPCI nozzle is bounded by the limiting vessel discontinuity curves. Therefore, the vessel discontinuity limits on the P-T curves will not be changed through 32 EFPY of operation.

(2) Schedule For Implementation Of R.G. 1.99, Rev. 2 Generic Letter 88-11 requires that R.G. 1.99, Rev. 2 be implemented within two outages after May, 1988. Based on the results of our analyses, the following implementation schedule is proposed: (a) The current P-T curves are conservative for up to 6.7 EFPY of operation.

However, because the 1 e Enclosure 2 (Continued) implementation of R.G. 1.99, Rev. 2 results in calculated 32 EFPY ART values more conservative than those of R.G. 1.99, Rev. 1 at 32 EFPY, the P-T curves will be revised within two refueling outages (but not later than 6.7 EFPY). Further, since a neutron dosimeter capsule has been removed for testing, revision of the P-T curves will be combined with the results of this testing effort. In this way, implementation of R.G. 1.99, Rev. 2 will be based upon verified, plant specific fluence predictions. (b) Upon completion of the revisions to the P-T curves, appropriate changes will be made to Section 5.3 and Appendix 5A of the Hope Creek Updated FSAR, and also to Section 3/4.4.6 and the corresponding Bases section of the Hope Creek Technical Specifications.

The Technical Specification changes will be submitted prior to completion of the second Refueling outage. Changes to the Updated FSAR will be incorporated as part of the next routine annual update following submittal of the Technical Specification changes. 2

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\f * --' * "' Table 1 COMPARISON OF REV 1 AND REV 2 ART VALUES FOR HOPE CREEK 4 EFPY Rev 1 Rev 2 Rev 1 32 EFPY Rev 2 Beltline Component ART c*r> ART C-F> ART C°F) ART {°F) Plates: 5K3025-l 26.1 29.8 39.0 64.2 5K2608-l 22.0 24.5 27.5 42.2 5K2698-l 22.8 25.2 29.8 45.0 5K2963-l -4.7 2.5 4.9 29.7 5K2530-l 24.9 33.5 35.6 65.0 5K3238-1 15.2 23.4 30.2 59.3 51(3230-1

-4.l 2.5 6.6 29.7 6C35-l -4.0 5.4 8.9 41. 3 6C45-l 5.7 15.5 14.3 47.0 Beltline Welds: 510-01205

-33.0 -9.2 -20.1 58.0 053040/1125-02205

-24.l 0.1 -13.4 65.8 519-01205

-46.3 -47.1 -41.3 -41.0 504-01205

-28.0 -29.l -22.5 -23.0 055733/1810-02205

-35.4 -27.9 10.6 053040/1810-02205

-44.6 -36.9 -36.7 1.6 LPCI Nozzles and Weld: 19468-1 (nozzle) -8.0 7.1 10024-1 (nozzle) -6.1 13.1 001-01205 (weld) -32.4 -31. 5