ML19329D874
| ML19329D874 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/28/1978 |
| From: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| To: | OFFICE OF THE FEDERAL REGISTER |
| References | |
| NUDOCS 8004070650 | |
| Download: ML19329D874 (1) | |
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O' UNITED STATES
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WASHINGTON, D. C. 20555
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,4 seg4 OFFICE OF THE April 28, 1978 SECR ETARY Director Office of the Federal Register National Archives and Records Service Washington, D.C.
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Dear Sir:
Enclosed for publication in the Federal Register are an original
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and two certified copies of a document entitled:
SACRMiENTO MUNICIPAL UTILITY DISTRICT Docket No. 50-312 ORDER FOR MODIFICATION OF LICESSE Publication of the above document at the earliest.possible date would be apareciated.
This material is to be charged to requisition number D-149.
Sincerely, Samuel J. Chilk Secretary of the CcnTnission
Enclosures:
Original and 2 certified ccpies J
Records Facility Branch
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UtilTED STATES OF AMERICA flUCLEAR REGULATORY C0!!!11SS10!!
In the Matter of
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SACR /RE!!TO I:Ui!ICIPAL UTILITY DISTRICT
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Docket flo. 50-312
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Rancho Seco i;uclear Station, Unit fio.1
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ORDER FOR I:0DIFICAT10t! 0F LICEllSE 1.
The Sacramento fiunicipal Utility District (the licensee), is the holder of Facility Operating License flo. DPR-54 which authorizes the operation of the nuclear power reactor known as Rancho Seco fiuclear Station, Unit hio.1, (the facility) at steady reactor power levels not in excess of 2772 mecawatts thermal (rated power).
The facility consists of a Babec:t and !!ilcox Company designed pressurized water reactor (Ph'R) located at the licensee's site in Sacramento County, California.
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In accordance with the requirements of the Commission's ECCS Acceptance Criteria,10 CFR 50.46, the licensce submitted on' July 8, 1975, an ECCS
. evaluatior, 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&M), the designer of the !!uclear Steaa Supply System -
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for this facility.
The B&W ECCS Evaluation Model had been previously found to conforn to the requirements of the Commission's ECCS Acceptance Criteria,10 CFP. Part 50.46 and Appendix K.
The evaluation indicated that with the limits set forth in the facility's Technical Specifications,
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the ECCS cooling performance for the facility would confom with the
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criteria contained in 10 CFR 50.46(b) which govern calculated peak clad i
temperature, ' maximum cladding oxidation, maximum hydrogen generation, cool-able geonetry and long-tern cooling.
On April 12, 1978, B&W informed the flRC that it had determined that in the
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event of a tmall 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.
Recen 'nalyses have shown that the discharge line break is more limiting than the suction line break.
The Rancho Seco !!uclear Station, Unit flo.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 system
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(RCS) cold legs on the discharge side of the RCS pump.
(There is also a third HPI punp installed.) The two parallel HPI trains are connected but
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are kIpt isolated by nanual valves (known as the cross-over valves) that are normally closed.
Upen receiving a safety injection signal
'the HPI pumps are started and valves in the four injection lines are opened. Assuming loss of offsite power and the worst single failv ?
(failure of diesel to start) only one HPI pump would be available and two of the four injection valves would fail to open.
s 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-sure pump) could flow out the break. Therefore, 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 contr.ibute'to maintaining the coolant inventory in the reactor vessel.
This situation had not been previously analyzed and BS',,' had indicated that,the limits sp:cified in 10 CFR 50.46 nay be exceeded.
B &'n' has stated that t. 'y have analyzed a spectrum of shall breaks in the, punp discharge line and have determined that to meet the limits of 10 CFR 50.46, operator action is required to open the two nanual operated crossover valves and to nanually align the two motor driven isolation valves which had failed to open. This would allow the flo:: from the one
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HPI. pump to' feed all four reactor coolant legs.
B&W has assumed
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refill the core.- The licensee has committed to provide for the necessary operator actions within the required time 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 committed to maintain one of the series-connected, manually operated cross-over valves normally open. The analyses performed by 85H assumed that the flow split was established at 650 seconds by operato, 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 discharge line break, with the afore-nentioned operator actions, is the most ifniting case. To a,rrive at this conclusion, B&W has perforned analyses at b'reak sizes of.3,.2,.15,
.1, and.04 f t.2 The results, which were obtained using an approved Appendix K nodel for blowdcun, indicate core uncovery for about 500 b'eak.
For this brea'k size DSN has conser-seconds.for the 0.15 f t.2 r
vatively calculated the peak clad temperature to be approxinately 1760 F; well below the linits of 10 CFP, 50.t6(b).
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B&W has indicated the manner in which the calculational methods have been revised and has indicated that their revised calculations are wholly _in confornance with the requirements of 10 CFR 50.46.
- However, B&W has not yet had the opportunity to fully present the result of its calculations to the licensee for submittal to the NRC staff, and 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 b'as agreed, that operating conditions be limited to a range in which ECCS performance for sna11 break conditions is less sensitive to specific calculation inputs.
For this facility, with operation up to 2311 !!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 substantial margin on peak clad temoerature below the limits of 10CFp50.46(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 approxinately August,1978. At this lower power level, the safety nargin on peak clad temperature 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 power level specified in this Order, and in accordance with the operating procedures specified in _ this Order, will assure that the ECCS
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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 4E by the NRC staff, the licenser committed to provide the staff with B&W's reevaluation of ECCS performance applicable to the licensee's facility as promptly 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 procedures h
were described and the commitments confirmed by the licensee's letter l[{
of April 14, 1978, supplemented by letter dated April 21, 1978. The staff believes that the licensee's action, under the circumstances, is appropriate and that this action.should be confirmed by MRC Order.
Upon satisfactory completion of our assessment of the revised evaluation, we will accordir:1y modify the authorization to op'erate the facility.
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IV.
Copies of the following documents are available for inspection at the Commission's public Document Room at 1717 H Street', Uas,hington, D.C.
- P4GCC, and are being/placed in the. Commission's local public do:ument room &
the Sacramento City-County Library, Sacramento, California.
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.~E (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,
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and the Connission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS ORDERED THAT Facility Operating License No. DPR-54 is hereby amended by adding the following new provisions:
(1) As soon as possible, the licensee shall submit a reevaluation wholly in conformance with 10CFR50.46 of ECCS cooling performance calculated in accordance with the B&h' Evaluation Model for operation with operating procedures described in its letters 'of April 14, 1978, and April 21, 1978, exceot 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 20S0 U.it, and N
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(3) Ustil further authorization by the Commission, the licensee mey shall operate in accordance with the procedures described in
"~C its. letter of April 14, 1978, supplemented by letters dated i
April 21,1978, except that the maximum time for completion of
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operctor action shall be 10 minutes after initiation of the event.
FOR THE ilVCLEAR REGULATORY C0 t'11SS10!1 i=
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ictor Stello, J., Director Division of Operating Reactors Office of fluclear Reactor Regulation Dated at Bethesda, flaryland,
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this 26th day of April 1978.
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