ML20136B968
| ML20136B968 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/08/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20136B960 | List: |
| References | |
| NUDOCS 8511200336 | |
| Download: ML20136B968 (3) | |
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
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wasmwoTow. o. c. 20sss y.....j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 7 TO FACILITY OPERATING LICENSE NPF-29 GRAND GULF NUCLEAR STATION, UNIT 1 MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION INTRODUCTION By letter dated August 12, 1985, Mississippi Power & Light Company (the licensee) proposed to make three changes to the facility Technical Specifications: (1) change the names of two valves listed in Table 3.3.7.4-1 " Remote Shutdown Systems Controls" and four valves listed in Table 3.6.4-1, " Containment and Drywell Isolation Valves; (2) designate a different valve in the residual heat removal (RHR) to reactor head spray line as reactor coolant system pressure isolation valve (Table 3.4.3.2-1) and as containment isolation valve (Table 3.6.4-1) and make associated changes in the listing of primary conteinment penetration conductor overcurrent protective devices (Table 3.8.4.1-1), and motor-operated valve thermal overload protection (Table 3.8.4.2-1), and (3) add specifications in Table 3.3.3-1 " Emergency Core Cooling System (ECCS) Actuation Instrumentation" to incorporate interlock instru-mentation which is designed to prevent overpressurization of low design pressure ECCS piping by the reactor coolant, system, and make associated changes in other applicable Technical Specifications. This safety evaluation addresses changes (1) and (2) which are designated Item 12 in licensee's letter. Change (3) incor-poration of interlock instrumentation for ECCS injection valves and designated Item 13 in licensee's letter, is not considered in this action.
EVALUATION ~
Change (1), changing the names of valves in Technical Specification Table.s 3.3.7.4-1 and 3.6.4-1 to make the names consistent with plant nomenclature, is a purely administrative change and is therefore acceptable.
Change (2), designating a different valve (E 12 - F 394) in the RHR to reactor head spray line to serve the containment isolation function and the pressure isolation function, wa proposed because the presently designated valve (E 51 -
F 066) is not readily accessible for local leak rate testing.
The use of proposed valve (E 12 - F 394) for the pressure isolation function does not involve a change in the ASME Code piping classification. The proposed valve and piping to the valve is ASME Code Class 1, as required for reactor coolant system piping by 10 CFR 50.55a and will perform the pressure isolation function.
Valve E 51 - F 066 would continue to be stroke tested and Valve E 12 - F 394 8511200336 851108 PDR ADOCK 05000416 p
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would be both stroke tested and local leak rate tested, thus meeting the require-ments for testing pressure isolation valves in Standard Review Plan (SRP) Section 3.9.6.
Accordingly, the staff concludes that the replacement of Valve E 51 - F 066 with Valve E 12 - F 394 in Technical Specification Table 3.4.3.2-1 " Reactor Coolant System Pressure Isolation Valves" is acceptable.
The replacement of Check Valve E 51 - F 066 with Motor Operated Valve E 12 - F 394 for the inboard containment isolation function requires that containment isolation signals be included in the design of Valve E 12 - F 394. These signals will be~
the same as those previously reviewed and accepted for the out board isolation valve. Valve E 12 - F 394 is upstream of Valve E 51 - F 066 and the valve in the line for the test connection (Valve E 12 - F 344). Since leakage through Valve E 12 - F 344 would be contained by the proposed containment isolation valve, Valve E 12 - F 344 as well as Valve E 51 - F 066, would be removed from Technical Specification Table 3.6.4-1 " Containment and Drywell Isolation Valves." The licensee proposes to include in the Technical Specifications a valve closure time of 35 seconds with a provision for revising the closing time should valve testing require it. The Technical Specifications would require the NRC staff to be notified of any change to the closing time. The staff concludes that Valve E 12 - F 394 will perform the inboard containment isolation function and that the changes proposed by the licensee to Technical Specification Table 3.6.4-1 " Containment and Drywell Isolation Valves" meet General Design Criterion 55 in 10 CFR 50, Appendix A, and SRP 96.2 and are, therefore, acceptable.
The designation of Motor-Operated Valve E 12 - F 394 as the pressure isolation valve and the containment isolation valve requires the associated electrical protective devices for motor operated Valve E 12 - F 394 to be added to Table 3.8.4.1-1 " Primary Containment Penetration Conductor Overcurrent Protective Devices" and Table 3.8.4.2-1 " Motor Operated Valves Thermal Overload Protection."
The NRC staff has reviewed the proposed changes to these tables and concludes that they are in accordance with the Standard Technical Specifications and are therefore acceptable.
Licensee's letter dated August 12, 1985, stated that the proposed design of con-trols for Valve E 12 - F 394 would include a manual control switch on the remote shutdown panel in addition to the control room control circuits. Control room
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control circuits would be interlocked to prevent overpressurization of the RHR but the remote shutdown panel control circuit would not be interlocked. Since safety analyses for Grand Gulf Unit I do not require the valve to be operated from the remote shutdown panel, the licensee has not proposed the addition of this valve to the listing in Technical Specification Table 3.3.7.4-1, " Remote Shutdown System Controls." The NRC staff agrees that the remote control for Valve E 12 - F 394 need not be added to Table 3.3.7.4-1.
However, as provided in SRP 67.4.III, if a control for the valve is included on the remote shutdown panel then an interlock, or acceptable alternative must be included in the remote shutdown panel control circuit to prevent inadvertent overpressurization of the RHR system.
By letter dated October 30, 1985, licensee proposed to remove the manual control switch for Valve E 12 - F 394 from the remote shutdown panel. The NRC staff concludes that the licensee's proposal to remove the
l 3-control switch for Valve E 12 - F 394 from the remote shutdown panel is an 1
acceptable means to prevent overpressurization of the RHR piping through inadvertent operation of this valve.
The valve is not required to be operated from the remote shutdown panel because the valve need not be operated for safe shutdown in the event of any control room evacuation.
ENVIRONMENTAL CONSIDERATION The amendment involves a change of requirements of facility components located within the restricted area as defined in 10 CFR 20. The Commission made a pro-posed determination that the amendment involves no signficant hazards considera-tion, and there have been no comments on that proposal.
Based on its evaluation, the staff concludes that there is no significant change in types or significant increase in the amounts of any effluents that may be released offsite. There is no significant increase in individual or cumulative occupational radiation exposure because the changes do not affect personnel exposure. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 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 the amendment.
l CONCLUSION The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register (50 FR 34994) on August 28, 1985, and consulted with the state of Mississippi.
No public comments were received, and the state of Mississippi did not have any comments.
We have 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 issuance of th';
amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Om P. Chopra, Power Systems Branch, DSI Lester L. Kintner, Licensing Branch No. 4, DL R. W. Stevens, Instrumentation and Control Systems Branch, DSI A. Notafrancesco, Containment Systems Branch, DSI
- 0. Rothberg, Mechanical Engineering Branch, DE A. Gilbert, Reactor Systems Branch, DSI Dated: November 8, 1985 W
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