ML20196B023
| ML20196B023 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/19/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20196B021 | List: |
| References | |
| NUDOCS 9811300227 | |
| Download: ML20196B023 (4) | |
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1 UNITED STATES g
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. ensaa anaq SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 239 TO FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO.' 229 TO FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLFY AUTHORITY SEQUOYAH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
The Tennessee Valley Authority (TVA) requested amendments to Operating Licenses DPR-77 and DPR-79 for Sequoyah Nuclear Plant (SON), Units 1 and 2, respectively, in a letter to the U.S. Nuclear Regulatory Commission (NRC) dated April 30,1998. The amendments would modify the SQN Units 1 and 2 Technical Specification (TS) 3.4.3.2 by revising Surveillance Requirement (SR) 4.4.3.2.1.b to allow meeting the requirement by operating the power-operated relief valves (PORVs) through one complete cycle of full travel in Modes 3,4, or 5 with a steam bubble in the pressurizer. Approval of this change would supersede the one-time footnote to Unit 1 SR 4.4.3.2.1, implemented by Amendment No. 230, dated January 13,1998.
' Therefore, this request proposed deletion of the footnote in the Unit 1 TSs. The TS Bases are also being revised to reflect that representative conditions for PORV testing exist with a steam bubble in the pressurizer,
2.0 BACKGROUND
On June 25,1990, the NRC staff issued Generic Letter (GL) 90-06, " Resolution of Generic issue 70, ' Power-Operated Relief Valve and Block Valve Reliability' and Generic Issue 94,
' Additional Low-Temperature Overpressure Protection for Light Water Reactors,' Pursuant to 10 CFR 50.54(f)." The GL requested that licensees adopt the staff positions and ar; moriate TSs for their facilities.
Generi:: lasue 70, " Power-Operated Relief Valve and Block Valve Reliability," involves the evaluaLon of the reliability of PORVs and PORV block valves and their safety significance in pressurized water reactors. The GL discussed how PORVs are incraasingly being relied on to perform safety-related functions and the corresponding need to improve the reliability of both PORVs and their associated block valves. Briefly stated, the GL required the following actions
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to improve PORV and block valve reliability:
a.
Include PORVs and PORV block valves within the scope of an operational quality assurance l
program that is in compliance with 10 CFR Part 50, Appendix B.
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2 b.
Include PORVs and PORV block valves within the scope cf a program covered by subsection IWV, " Inservice Testing of Valves in Nuclear Power Plants," of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.
Also, stroke testing of PORVs should only be performed in Modes 3 or 4 and in all cases prior to establishing conditions where the PORVs are used for low temperature overpressure protection (LTOP).
c.
Include TS for PORVs and PORV block valves for operational Mcdes 1,2, and 3 to incorporate the new staff position. Included in the staff position is a requirement that plants that run with block valves closed (e.g., due to leaking PORVS) maintain electrical power to 1
the block valves so they can readily be opened from the control room upon demand.
- Specifically, position 2 in the GL stated that stroke testing of PORVs should be performed during Mode 3 or Mode 4 and should not be performed during power operation. In response to the GL, SON's TSs require that stroke testing be performed in Mode 4.
PORV testing for the ASME Code-required Pump and Valve Inservice Testing Program is performed while in Mode 5, while testing of PORVs to demonstrate TS operability in accordance with SR 4.4.3.2.1.b, as presently written, requires Mode 4 plant conditions. The TS testing duplicates much of the inservice testing that is performed in Mode 5. Required plant conditions for the ASME Code testing, including stroke time testing to detect PORV degradation, is that reactor coolant system (RCS) average temperature (Tavg) be less than or equal to 200*F. ASME Code testing is performed in Mode 5 because there is inherent risk from a potential loss of inventory standpoint when testing during operation, and there is a potential risk of a PORV sticking open even though the block valve is capable of being closed. In addition, indirect remote position verification, achievable in Mode 5, is necessary since direct observation of stem movement is not available due to SON's valve design. Changing the mode requirement to allow Mode 5 performance would provide flexibility in testing and would eliminate duplication. Changing the mode requirement to include Mode 3 allows additional flexibility within the bounds of the original NRC guidance of GL 90-06.
3.0 EVALUATION As stated in the TS Bases, the PORVs and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.
Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety relief va!ves. Each PORV has a remotely operated block valve to provide positive shutoff capability should a relief valve become inoperable. The PORVs also function to remove noncondensable gases or steam from the pressurizer.
Current TSs for controlling PORVs are a result of NRC GL 90-06. GL 90-06 and NUREG-1316,
" Regulatory Analysis for the Resolution of Generic Issue 94, Additional Low-Temperature Overpressure Protection for Light-Water Reactors," provide little insight as to why Mode 3 or Mode 4 may be desired, except to suggest that it would better simulate temperature and pressure environmental conditions. Requirements for Modes 3,4, and 5 are basically defined by RCS Tavg. Mode 3 requires > 350*F; Mode 4 requires < 350*F, but > 200*F; and Mode 5 requires Tavg 5 200*F. However, the key parameter for PORV test is pressurizer pressure and
3 temperature, not RCS Tavg. Frequently in Mode 5, a pressurizer bubble is developed for RCS pressure control. With a steam bubble in the pressurizer, the steam pressure is 2200'F, which is higher than the lowest temperature allowed for Mode 4 (200*F). As a result, representative conditions for PORV testing are present in Modes 3,4, and 5 with a steam bubble in the pressurizer. Therefore, expanding mode requirements from only Mode 4 to Modes 3,4, or 5, with a steam bubble is acceptable and within the representative test requirements assessed in j
j A possible factor related to testing in other than Mode 4 would be the effect on the low
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temperature overpressure protection (LTOP) TS 3.4.12. This TS is applicable in Modes 4 and 5, and Mode 6 with the reactor vessel head on. Since the current SR 4.4.3.2.1.b must be performed in Mode 4 and the ASME testing is performed in Mode 5, deletion of the mode i
restriction on SR 4.4.3.2.1.b would allow the two tests to be performed together in the same mode, lessening the time that potential LTOP impacts exist. Testing in Mode 3 likewise reduces potential LTOP impacts.
I The footnote to Unit i SR 4.4.3.2.1 is superseded by this request and its deletion will have no affect on plant operations.
l The staff has reviewed the licensee's proposed modifications to the SON TS. Since the proposed modifications are consistent with the sta", position previously stated in GL 90-06 and are justified in the above referenced regulatory a fais, the staff finds the proposed modifications to be acceptable.
4.'0 STATE CONSULTATION l
In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility 4
component located within the restricted area as defined in 10 CFR Part 20 and changes i
surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that i
may be released offsite, and that there is no significant increase in individual or cumulative I
occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (63 FR 38204, dated July 15,1998). Accordingly, the amendments meet the eligibility criteria for cliegorical 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 the amendments.
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6.0 CONCLUSION
The Commission 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: R. Heman Dated: tbveter 19, 1998 i
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