ML19319D947
| ML19319D947 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/26/1978 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| Shared Package | |
| ML19319D943 | List: |
| References | |
| NUDOCS 8003270621 | |
| Download: ML19319D947 (8) | |
Text
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7590-01 s
UNITED STATES OF AMERICA NUCLEAR REGULATORY C0".111SSION In the Matter of
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SACRAMEtiTO fiUflICIPAL UTILITY. DISTRICT
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Docket No. 50-312
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Rancho Seco fluclear Station, Unit flo.1
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ORDER FOR P.0DIFICATION OF LICENSE I.
The Sacramento l'.unicipal Utility District (the licensee), is the holder of Facility Operating License Ho. DPR-54 which authorizes the operation of-the nuclear power reactor known as Rancho Seco Nuclear Station, Unit No.1, (the facility) at steady reactor power levels not in excess of 2772 mecawatts thermal (rated power). The facility consists of a Babcock and Wilcox Ccmpany designed pressurized water reactor (PWR) located at the licensee's site in Sacramento County, California.
II.
In accordance with the requirements of the Commission's ECCS Acceptance Criteria,10 CFR 50.46, the licensee submitted on ' July 8,1975, an ECCS evaluatfon for the facility.
The ECCS performance submitted by the li-censee was based upon an ECCS Evaluation Model developed by the Babcock
& Wilcox ~ Company (B&W), the, designe,r of the Nuclear Steam Supply System. -
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.2-for this facility. The B&W ECCS Evaluation l'odel had been previously found to conforn to the requirements of the Commission's ECCS Acceptance Criteria,10 CFR Part 50.46 and Appendix K.
The' eval ua tion. indica ted that with the limits set forth in the facility's Technical Specifications, the ECCS cooling performance for the facility would conform with the
. criter,ia contained in 10 CFR 50.46(b) which govern calculated peak clad temperature, maximum cladding oxidation, maximum hydrogen generation, cool-able geonetry and long-term cooling.
On April 12, 1978, B&W informed the NRC that it had determined that in the event of a small break LOCA on the discharge side of a reactor coolant pump, high pressure injection (HPI) flow to the core could be reduced somewhat.
Subsequent calculations indicated that in such a case the calculated peak clad temperature might exceed 2200F.
Previous small break analyses for B&W 177 fuel assembly (FA) lowered loop plants had identified the limiting small break to be in the suction line of the reactor. coolant pump.
Recent analyses have shown that the discharge line break is more' limiting than the suction line. break.
The Rancho Seco !!uclear Station, Unit !!o.1, has an ECCS configuration which consists of 'two high pressure injection (HPI) trains. Each train has a HPI' pump and the train injects into two of the four reactor coolant syst'em t
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(RCS) cold legs on the discharge side of the RCS punp.
(There is also a third HPI pump installed.) The two parallel HPI trains are connected but are kept isolated by manual valves (known as the cross-over valves) that are normally closed.
Upon receiving a safety injection signal the HPI pumps are started and valves in the four injection lines are opened. Assuming loss of off site power and the worst single failure (failure of diesel to start) only one HPI pump would be available and two of the four injection valves would fail to open.
If a small break is postulated to occur in the RCS piping between the RCS pump discharge and the reactor vessel, the high pressure injection flow injected into this line (about half of the output of one high pres-
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sure pump) could flow out the break. Therafore, for the worst combination of break location and single failure, only one-half of the flow rate of a single high pressure ECCS pump would contribute to maintaining the coolant inventory in the reactor vessel.
This situation had not been previously analyzed and B&W had indicated that,the limf ts s'pecified in 10 CFR 50.46 may be exceeded.
B&W has stateu that they have analyzed a spectrum of s' mall breaks in the pump discharge line and have deterdined that t'o meet the limits of 10 CFR 50.46, operator action is required to open the two manual operated crossover valves and to manually align the two motor driven isolation valves which had failed to open.
This would alicw the flow from the one e
i 7590-01 HPI pump to feed all four reactor coolant legs.
B&W has assumed that 30% of the flow would be lost through the break and 70% would refill the core.
The licensee has committed to provide for the necessary operator actions within the required tir.e frame. That is, in the event of a small break and a limiting single failure, manual action will be taken to begin opening these valves within five minutes and have them fully opened and an adeouate flow split obtained within 10 minutes.
To facilitate this operation the licensee has comitted to maintain one of the series-connected, manually operated cross-over valves normally open.
The analyses performed by B&'.l assumed that the flow split was established at 650 seconds by operator action. We conclude that the analyses are a reasonable approximation of the operator action that actually will be taken, provided specific procedures are prepared and followed to assure such action.
B&W has stated that a.15 ft.2 discharoe line break, iith the afore-mentioned operator actions, is the most limiting case. To a,rrive at this conclusion, B&W has perdomed analyses at b'reak sizes of.3,
.2,.15,
.1, and.04-ft.2 The results, which were obtained using an approved Appendix X model for blowdown, indicate core uncovery for about 500 seconds for the 0.15 ft.2 break.
For this brea'k size B&',l has conser-vatively calculated the peak clad temperature to be approximately 11760 F; well below the limits of 10 CFR 50.46(b).
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5-B&W has indicated the manner in which the calculational methods have
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been revised and has indicated that their revised calculations are wholly in confornance with the ' requirements of 10 CFR 50.46.
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B&W has not yet had the opportunity to fully present the result of I
its calculations to the licensee for submittal to the NRC staff, and
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the staff has accordingly not had the opportunity to fully assess the new calculations. Until the licensee and the staff have an opportunity to review the B&W revised calculations, the staff has recommended and the licensee h'as agreed, that operating conditions be limited to a range in which ECCS performance for small break conditions is less sensitive to specific calculation inputs.
For this facility, with operation up to 2311 l'wt, ECCS performance calculations for the limiting small break does not even result in core uncovering, if appropriate operator action is properly taken (as described above), thus providing a very substantiai margin on peak clad temoeraturc below'the limits of 10CFR50.46(b).
For other reas,ons which are not safety-related, however, the plant is limited to a maximum power of about 2080 megawatts thermal until approximately August,1978. At this lower power level, the safety margin on peak clad teinperature will be even greater.
Therefore, until the staff has had the opportunity to ' fully assess the B&W revised calculations, operation of the facility at the powerilevel specified in this Order, and in accordance with the operating procedures specified in this Order, will assure that the ECCS r,
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7590-01 n
6-will-conform to the performance requirements of 10 CFR 50.46(b).
' Accordingly, such limits provide reasonable assurance that the publ.ic health and safety will not be endangered.
Upon notification by the NRC staff. the ifcensee committed to provide the staff with
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B&W's reevaluation of.ECCS performance applicable to the licensee's facility as prcmptly as possible, to submit a technical specification requiring appropriate operating procedures to assure required operator action as discussed herein, and affirmed that plant operation was limited to the maximum power level specified herein.
Such p.ocedures were described and the commitments ccnfirmed by the licensee's letter of April 14, 1978, supplemented by letter dated April 21, 1978.
The staff
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believes that the licensee's action, under the circumstances, is appropriate and that this action should be confirmed by NRC Order.
Upon satisfactory completion of our assessment of the revised evaluation, we will accordingly modify the aut'horization to op'erate the facility.
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' Copies of' the following documents are available for inspection at the Commiss~ ion's public Document Room at 1717 H Street, Was,hington, D.C.
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.and are being'placed in the Commission's local, public document room at-the Sacramento City-County Library, Sacramento, California.
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7590-01 (1) Letters from J. J. Mattimoe to Mr. R. W. Reid, Chief Operating Reactors Branch #4, dated April 17 and 21,1978.
Accordingly, pursuant to the Atomic Energy Act of 1954, as amended, and the Commission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS ORDERED THAT Facility Operating License No. OPR-54 is hereby amended by. adding the following new provisions:
(1) As soon as possible, the licensee shall submit a reevaluation wnclly in conformance with 10CFR50.45 of ECC5 cooling performance calculated in accordance with the B&W Evaluation ?bdel for operation with
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operating procedures described in its letters of April'14,1978, and April 21, 1978, except that the time for completion of operator action shall be'10 minutes after initiation of the event.
(2) Until'further authorization by the Commission, the power level shall not exceed 2080 Mwt, and i
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(3) Until further authorization by the Commission, the licensee shall operate in accordance with the procedures described in its letter of April 14, 1978, supplemented by letters dated April 21,1978, exceptthatthetaaximumtimef$rcompletionof I
operator action shall be 10 minutes after initiation of the event.
FOR THE NUCLEAR REGULATORY C0:UtISSION
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.a 1ctor Stello, J., Director Division of Operating Reactors Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland, this 26th. day of April 1978.
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