ML18052B518
| ML18052B518 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/29/1988 |
| From: | Smedley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 8804040067 | |
| Download: ML18052B518 (5) | |
Text
consumers Power POW ERIN&
NllCHl&AN'S PROliRESS General Offices: 1945 West Pernell Road, Jackson, Ml 49201 * (517) 788-0550 March 29, 1988 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
NUREG-0737, ITEM II.D.l, PERFORMANCE TESTING OF RELIEF AND SAFETY VALVES (TAC NO. 44604) - ADDITIONAL INFORMATION NRC letter dated October 20, 1987 requested 10 items of additional information regarding NUREG-0737, Item II.D.l - Performance Testing of Relief and Safety Valves.
Consumers Power Company letter dated November 17, 1987 informed the staff that we would respond to Items 3-10 before April 1, 1988 and to Items 1 and 2 before July 1, 1988.
This letter responds to Items 3, 4, 5 and 10.
Items 6-9 relate to analyses performed by our contractor, Impel! Corporation (previously named EDS Nuclear).
Impel! has been contracted to supply our response to the NRC questions regarding the EDS Nuclear analysis; however, a condition beyond their control has caused a delay.
We, therefore, have re-scheduled our response for Items 6-9 to be on or before June 1, 1988.
The following are NRC Items No's 3, 4, 5 & 10 as contained in the October 20, 1987 NRC letter and the Consumers Power Company response.
- 3.
Question/Item "Dresser Ind., in March 1976, recommended to Metropolitan Edison Co. that the PORV block valve be closed at pressures below 1000 psig to prevent steam wirecutting of the PORV seat and disk.
Testing by Dresser later showed the 1000 psig pressure limit to be overly conservative and that the PORV as designed was qualified to system pressure of 100 psig.
Below 100 psig the deadweight of the lever on the pilot valve was sufficient to keep the pilot valve open.
Dresser recommends that heavier springs be used under the main and pilot disks to ensure closure, if the plant is to operate at pressures below 100 psig.
Without the heavier springs recom-mended by Dresser, the PORV should not be used at system pressure below 100 psig.
Since the minimum operating pressure at Palisades is below 100 psig, verify that Consumers Power has installed the heavier springs in its PORVs consistent with Dresser recommendations."
Response
The response to NRC Question 2 of the NRC letter dated August 6, 1985 as contained in our December 30, 1985 submittal dockets our position on this
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matter of heavier closure assist springs (Type 2 Internals) used in lieu of the lighter (Type 1 Internals) closure assist spring being used.
In summary, there is no overriding need to change to the Type 2 internals.
Palisades does not experience low pressure leakage which would make this modification urgent.
If the presently installed PORVs are not changed to 2
a different type or larger size to facilitate plant operations, and if the presently installed Dresser 31533VX-1 valves exhibit excessive low pressure leakage, Type 2 internals (heavier spring) will be installed.
- 4.
Question/Item NUREG-0737 requires the PORV control circuitry be qualified for post-accident environments.
None of the submittals from Consumers Power on Palisades have addressed this requirement.
The Nuclear Regulatory Commis-sion staff has agreed that meeting the licensing requirements of 10 CFR 50.49 for this circuitry is satisfactory and that specific testing per NUREG-0737 requirements is not required.
Therefore verify whether the PORV circuitry has been reviewed and accepted under the requirements of 10 CFR 50.49.
If the PORV circuitry has not been q~alified to the requirements of 10 CFR 50.49, provide information to demonstrate that the control circuitry is qualified per the guidanc*e provided in Reg. Guide 1. 89, Revision 1, Appendix E.
As. an alternative, the staff has determined that the requirements of NUREG-0737 regarding the qualification of the PORV control circuitry may be satisfied if one or more of the following conditions is met.
A.
The PORVs are not required to perform a safety function to mitigate the effects of any design basis event in the harsh environment, and failure in the harsh environment will not adversely impact safety functions or mislead the operator (PORVs will not experience any spurious actuations and, if emergency operating procedures do not specifically prohibit use of PORVs in accident mitigation, it must be ascertained that PORVs can be closed under harsh environment conditions).
B.
The PORVs are required to perform a safety function to mitigate the effects of a specific event, but are not subjected to a harsh environ-ment as a result of that event.
C.
The PORVs perform their function before being exposed to the harsh environment, and the adequacy of the time margin provided is justified; subsequent failure* of the PORVs as a result of the harsh environment will not' degrade other safety functions or mislead the operator (PORVs will not experience any spurious actuations and, if emergency operating procedures do not specifically prohibit use of PORVs in accident mitigation, it must ~e ascertained that PORVs can b~ closed under harsh environment conditions).
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3 D.
The safety function can be accomplished by some other designated equipment that has been adequately qualified and satisfies the single-f ailure criterion.
Response
Consumers Power Company satisfies the requirements of NUREG-0737 regarding qualification of the PORV circuitry by meeting condition "D".
The use of the PORVs for pressure control and/or heat removal is not required unless the steam generator heat sink is lo$t. It takes more than a single failure for the Palisades Plant to lose the steam generators as a heat sink.
Therefore, the safety function of the PORVs can be accomplished by some other designated equipment that has been adequately qualified and satisfies the single failure criterion.
- 5.
Question/Item None of the submittals have identified the code used to evaluate the pipe supports.
The load combinations were given in the December 30, 1982 submittal, but not the allowable stress associated with them.
Identify the code governing the support analysis and the applicable code allowables.
In the December 30, 1985 submittal, Consumers Power noted that the piping analysis was revised and the EPRI load combinations applied to the piping system.
Were the EPRI support load combinations also used? If not, justify not including a seismic load (SSE) with the safety valve discharge load in the support load combinations listed in the December 30, 1982 submittal.
Response
The evaluation of the piping supports associated with the power operated relief valve and safety valve discharge system has been conducted according to the Palisades Plant FSAR Update.
The evaluation of the supports upstream of the valves (the safety related segments of the piping) has demonstrated that those supports meet EPRI or FSAR Update load allowables with consistent EPRI or FSAR Update load combinations.
The downstream supports were explicitly analyzed* to FSAR Update loads and load combinations in order to maintain consistency with the analysis employed on safety related piping systems in the plant.
The FSAR Update methodology is consistent with that employed in IE Bulletin 79-02/79-14 implementation with regard to anchor bolts, baseplates, welds and steel support components.
Vendor catalog component data employed for load allowables was that provided by the vendor for the load condition of concern.
It was noted during a* recent review of tne support load eval-uation results that safety valve discharge loads strongly dominated seismic loads for supports downstream of the valves where the addition of discharge loads significantly increased the original support design loads.
Therefore, the SRSS combining of dynamic load responses *(per EPRI) would have had a negligible impact on total support *loads and use of the EPRI OC0388-0073-NL02
load combination would, therefore, not have influenced* the result of the support evaluation.
4 The Palisades Plant FSAR Update (Section 5.9.1.1.2 for anchor bolts and baseplates and Section 5.10.1.2 for all.other support components) provides only.two sets of load combinations. for support load evaluation.
The load allowables employed for the evaluation of safety valve discharge loads on downstream supports were the allowables used for SSE responses at Palisades.
These allowables are factored loads based upon vendor or AISC requirements and are intended to keep the components and support members elastic.
Conformance to these allowables ensures dimensional stability of the supports and provides further assurance of the applicability of elastically computed piping response.
The support evaluation showed considerable margin in the original support designs.
However, very large discharge loads required that four supports be.redesigned.
One design has been implemented and three modifications will be made during the next refueling outage.
- 10.
Question/Item The Combustion Engineering (CE) inlet conditions report listed the FSAR transients and accidents for each plant which result in a peak pressure greater than the safety valve setpoint.
For some plants this list included the feedwater line break (FWLB), but.for other plants the FWLB was not i~cluded. Palisades was a plant that did not include the FWLB in its list of transients and accidents that challenge.the safety valves.
From the CE report it was not clear whether the FWLB was missing because the accident did not challenge the safety valves or because Palisades was licensed prior to the issuance of Regulatory Guide 1.70, Rev. 2 and, therefore, the FWLB was not analyzed as part of Palisades design basis.
Discuss why the FWLB was not listed for Palisades. If the FWLB was not listed because of the second reason discussed above, it is the staff position that the Palisades submittal is incomplete.
Item II.D.l in NUREG-0737 specifically requires that PORVs and safety valves *be qualified for fluid conditions resulting from transients and accidents referenced in Regulatory Guide 1.70, Rev. 2.
The FWLB is specifically defined in.Regulatory Guide 1.70, Rev. 2. Additionally, from the staff review of other plant-specific response to Item II.D.1, it is clear that for many plants the FWLB accident is the limiting case for providing high pressure liquid to the safety valves, a fluid for which they were not specifically designed originally.
This is exactly the type of concern that NUREG-0737, II.D.1 was established to address.
In accordance with the requirements of the NUREG, we require that information be provided to demonstrate that the PORVs and safety valves will function as required to assist in safe shutdown of the plant and will not experience any degradation that would inhibit safe plant shutdown if exposed to the FWLB.
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Response
Palisades was licensed prior to issuance of Reg. Guide 1.70, and analyses of record have not included the Feedwater Line Break (FWLB) as part of the design basis.
However, latest analyses performed by Advanced Nuclear Fuels (ANF) Corporation (Reference 1) contain an evaluation of the FWLB event.
Reference 1 concludes that the FWLB event for the Palisades plant begins as a cooldown event and later becomes a heatup event.
The cooldown is bounded by the main steam line break event.
The heatup is bounded by the Loss of Load event.
Thus, based on Reference 1 conclusions, the FWLB event does not require analysis because the Loss of Load event presents the greatest challenge to safety valve operation.
Therefore, although FWLB was not evaluated when the the Loss of Load event was selected as the transient which results in peak pressure, later analyses (References 1 & 2) show that a Loss of Load event results in a higher peak pressure than the FWLB event.
Reference 2 also shows that the Loss of Load event does not cause the pressurizer liquid level to rise to the elevation of the safety valves.
This agrees with the C-E Loss of Load analysis (Reference 3) used in our earlier submittals.
Reference
- 1. Advanced Nuclear Fuels Report ANF-87-150(P) Volume 1, "Palisades Modified Reactor Protection System Report - Disposition of Standard ReviewPlan Chapter 15 Events", November, 1987.
- 2. Advanced Nuclear Fuels Report ANF-87-lSO(P) Volume 2,,;Palisades.
Modified Reactor Protection System Report:
Analysis of Chapter 15 Events", February, 1988.
- 3. C-E Report CEN-227, "Summary Report on the Operability of Pressurizer Safety Valves in C-E Designed Plants", December, 1982.
Richard W Smedley Staff Licensing Engineer CC Administrator, Region III, NRC NRC Resident Inspector - Palisades OC0388-0073-NL02 5