ML18053A475
| ML18053A475 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/04/1988 |
| From: | Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML18053A476 | List: |
| References | |
| NUDOCS 8808110259 | |
| Download: ML18053A475 (5) | |
Text
I'
- consumers Power l'OWERING llllCHIGAN"S l'IUJGllE55 General Offices:
1945 West Parnell Road, Jackson, Ml 49201 * (517) 788-1636 August 4, 1988 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
TECHNICAL SPECIFICATION CHANGE REQUEST -
SECONDARY SAFETY VALVES Kenneth W Berry Director Nuclear Licensing Enclosed is a request for change to the Palisades Technical Specifications.
The change proposes revising the secondary system safety valve setpoint tolerances from between 985 psig (+/- 10 psig) and 1025 (+/- 1%) psig to between 985 psig (+/- 30 psig) and 1025 (+/- 3%) psig.
The change is justified by the Advanced Nuclear Fuels Corporation report ANF-87-150(NP), Volume 2; "Palisades Modified Reactor Protection System report: Analysis of Chapter 15 events,"
submitted by Consum~rs Power Company letter of June 17, 1988.
In the report, ANF-87-150(NP), Vol. 2, the setpoint uncertainty value assumed for the steam relief valves was 3%.
The analyses concluded that with this assumed uncertainty value, in conjunction with the modified reactor protection system, the specified acceptable fuel design limits for fuel centerline melt (21 kw/ft) and minimum departure from nucleate boiling (DNBR of 1.17) would not be exceeded.
The present safety valve setpoint tolerances are based on the 1971 yersion of the ASME Boiler and Pressure Vessel Code,Section III, requirements.
These are the same requirements as in the Standard Technical Specifications.
The proposed changes are based upon the present ASME Code req~~rements in Section XI, IWV-3500, whic.h r~ferences ANSI/ASME OM-1-1981.
The acceptance criteria in OM-1 (Section 1.3.3.1.4) allows a 3% tolerance as is propo~ed in this change r~quest. Consumers Power Company requests a prompt NRC review of this proposal.
Our intent is tp implement the expanded tolerance during the upcoming refueling.outage.
We would like to know of any disagreement with.
this proposal as soon as possible to plan for resource availability.
We further request approval of this proposal prior to system operability require-ments about November 1, 1988.
A check for $150.00 is attached as required by 10 CFR 170.12.
1~aJ~
Kenneth W Berry Director, Nuclear Licensing CC Administrator, Region III, NRC NRC Resident Inspector - Palisades Attachment OCO 7 88-0079-NL02,-----r~~~;;;;-o;::;ri;:~~-~
9808110~59 880804 i
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CONSUMERS POWER COMPANY Docket 50-255 Request for Change to the Technical Specifications License DPR-20 For the reasons hereinafter set forth, it is requested the Technical Specifications contained in the Provisional Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on October 16, 1972, for the Palisades Plant be changed as described in Section I below:
I.
Change(s)
A.
Change 3.l.7c to:
"c.
Whenever the reactor is in power operation, a minimum of 23 secondary system safety valves shall be operable with their lift settings between 985 psig (+/- 30 psig) and 1025 (+/- 3%)
psig.
B.
Delete last paragraph of 3.1.7 Basis, add the following, and add reference (2):
"The overpressurization analysis for the loss of load event (2) supports the specified secondary safety valve lift pressure tolerance.
ASME B&PV Code Section XJ-1986 IWV-3500 specifies ANSI/ASME OM-1-1981 requirements which allow the specified tolerances in the lift pressures of the safety valves.
(2)
ANF-87-150(NP) Volume 2, Section 15.2.1.
II. Discussion The above proposed Technical Specifications change is requested to widen the secondary system safety valve setpoint to +/- 3% in accordance with 1986 ASME B&PV Code, Section-XI, IWV-3500 andANSI/ASME, OM-1-1981 and as justified in Advanced Nuclear Fuels Corporation report ANF-87-150(NP) which supports Palisades operation with the modified reactor protection system and up to 29.3% steam generator tube plugging.
The surveillance interval requires testing of all (24) valves on a 5-year cycle with a minimum of 20% of the valves within any 24 months.
Typically, 1/3 of the valves are tested each refueling outage.
With the failure of one valve to meet the 1% (or +/- 10 psig) acceptance criteria, other valves must be tested according to the formula (N + 12/60) Z.
Where N is the number of months from the start of the 60-month period to the end of each refueling, and Z is the number of relief valves in the program.
If more than one valve fails to meet the acceptance criteria, all additional valves must be tested.
At Palisades we have had to test all the secondary safety valves during most of the past refueling outages.
This has resulted in unnece$sary costs and truces manpower resources during their peak demand period.
Reanalysis of Standard Review Plan, Chapter 15, OC0788-0079-NL02
v 2
events i~ support of the RPS modification is documented in ANF-87-150(NP),
Volume 2.
The Loss of External Load event is the most limiting event with respect to power mismatch between the primary and secondary system and to secondary safety valve operability.
The reactor is at full power in this event, and the secondary side is not removing any energy.
The transient results in the maximum primary system pressure.
As noted in Table 15.0.8-1, of ANF-87-150(NP), Volume 2, the tolerance for the safety valves used in the analyses (all applicable analyses including loss of load event) is +/- 3%.
The objective of the Section 15.2.1 analysis is to:
- 1) demonstrate sufficient primary system relief capacity to limit pressure rise to less than 110% (2750 psia) of design pressure (Ref T.S.2.2); 2) evaluate the maximum primary/secondary differential pressure; and, 3) demonstrate MDNBR rem~ins above the XNB correlations safety limit of 1.17 (see 15.2.1.3).
The analysis*assumes full power initial conditions as listed in Table 15.2 1-1. Analysis demonstrates the "Loss of External Load" event will challenge two acceptance criteria.
The first is the primary system pressure and the second is the MDNBR specified acceptable fuel design limit (SAFDL).
This is d~e to increasing core inlet temperature and reactor power prior to scram.
The analysis demonstrates all applicable acceptance criteria are met for the "Loss 9f External Load" event. Analysis shows the maximum pressurizer pressure to be 2584.7 psia (below the limit of 2750 psia). Analysis also shows the primary/secondary maximum differential pressure to be 1604.4 psi.
Present Technical Specification Section 3.1.lc(l) lists the maximum steam generator operating transient differential pressure as 1530 psi.
However, this limit is being deleted as part of the RPS modification.
A des~ription is contained on page 28 of the CPCo submittal dated March 25, 1988.
Analysis shows the predicted. minimum DNBR equal to 1.776.
This value is above the XNB limit of 1.17.
Finally, the peak pellet LHGR for this event is approximately 12.7 kw/ft, well below the fuel centerline melt criterion of 21 kw/ft.
Analysis of No Significant Hazards Consideration Based on the ANF-87-150(P) analysis for "Loss of External Load," the change in set pressure tolerance is to+/- 30 psi for the 985_psia and+/-
3% for the 1025 psia Main Steam Relief Valves is acceptable.
Changing the set pressure tolerance as proposed does __ not_ cause the loss of external load event to occur more often than a moderate frequency event.
This. frequency is defined in Section 15.0.1.1 of the report.
Therefore, there is no increase in the probability of a previously evaluated accident.
OC0788-0079-NL02
Because all acceptance criteria are met with margin to documented limits, the.consequences of a loss of external load or the failure of equipment important to safety are not increased. Analysis shows the 3
Main Steam Relief Valves operating in conjunction with other safety equipment, as specified in the report, provide acceptable levels of protection to limit primary system pressure rise, primary to secondary system pressure, and MDNBR within the design capability of all components.
Therefore, the consequences of a previously analyzed accident are not increased.
According to the ANF report, the change to the expanded tolerance does not create the potential for a new accident.
This is because the primary and secondary systems are maintained within their design limits.
Therefore, a new or different kind of accident is not created.
The ANF report predicts a maximum differential pressure of 1604.4 psi.
However, based on the Safety Evaluation entitled, "TSCR RPS Modification" for Technical Specifications Section 3.1.lc(l), the Technical Specifications maximum transient differential pressure limit of 1530 psi is being deleted.
According to the RPS safety evaluation, the increased differential pressure value is due to improved analytical techniques, not by changes to Palisades hardware.
The structural integrity of the S/Gs is assured by appropriately selecting the tube plugging criteria.
The NRC has approved current tube plugging criteria in the SER dated June 11, 1984.
Therefore, the maximum differential pressure of 1604.4 does not significantly reduce the margins of safety for the Palisades plant.
III.
Conclusion The Pali~ades Plant Review Committee has reviewed this Technical Specif-ication Change Request and has determined that this change does not involve an unreviewed safety question and, therefore, involves no significant hazards consideration.
This change has also been reviewed under the cognizance of the Nuclear Safety Board.
A copy of this Technical Specification Change Request has been sent to the State of Michigan official designated to receive such Amendments to the Operating License.
OC0788-0079-NL02
- ~*
CONSUMERS POWER COMPANY By~
David P Hoffman, Nuclear Operation Sworn and subscribed to before me this 4th day of August, 1988.
~l~
Elaine E BuehrlrJQtarYPUbiiC Jackson County, Michigan My commission expires October 31, 1989 OC0788-0079-NL02 4