ML19319D918
| ML19319D918 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/21/1978 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| Shared Package | |
| ML19319D913 | List: |
| References | |
| NUDOCS 8003270592 | |
| Download: ML19319D918 (10) | |
Text
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7590-01 0
UNITED STATES OF AMERICA l
NUCLEAR REGULATORY COMMISSION l
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In the Matter of SACRAMENTO MUNICIPAL UTILITY DISTRICT
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Docket No. 50-312 1
i Rancho Seco Nuclear Generating Station ORDER FOR MODIFICATION OF LICENSE I.
The Sacramento Municipal Utility District (the licensee), is the holder of Facility Operating License No. DPR-54 which authorizes the operation of the nuclear power reactor known as Rancho Seco Nuclear Generating Station (the facility) at steady state reactor power levels not in excess i
of 2772 megawatts thennal (rated power). The facility consists of a l
Babcock and Wilcox Company (B&W) designed pressurized water reactor (PWR) located at the licensee's site in Sacramento County, California.
i II.
In accordance with the requirements of the Comission's Emergency Core Cooling System (ECCS) Acceptance Criteria,10 CFR 50.46, the licensee submitted on July 8,1975, an ECCS evaluation for the facility.
The ECCS perfonnance submitted by the licensee was based upon an ECCS Evaluation Model developed by B&W, the designer of the Nuclear Steam Supply System for this facility.
The B&W ECCS Evaluation Model had been previously 8003270 5 7 2
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found to conform to the requirements of the Commission's ECCS Acceptance I
Criteria,10 CFR Part 50.46 and Appendix K.
The evaluation indicated that with the limits set forth in the facility's Technical Specifications, I
the ECCS cooling performance for the facility would conform with the i
criteria contained in 10 CFR 50.46(b) which govern calculated peak clad temperature, maximum cladding oxidation, maximum hydrogen generation, coolable geometry and long-term cooling.
On April 12, 1978, B&W informed the Comission that it had detennined that in the event of a small break Loss of Coolant Accident (LOCA) on the the discharge side of a reactor coolant pump, high pressure inject.on (HPI) flow to the core could be reduced somewhat.
Subsequent calcu-lations 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 sucticn line of the reactor coolant pump.
Recent analyses have shown that the dis-charge line break is more limiting than the suction line break.
The Rancho Seco Nuclear Generating Station has an ECCS configuration which consists of two HPI trains.
Each train has a HPI pump and the train injects into two of the four reactor coolant system (RCS) cold legseithedischargesideofthe'T.Spump.
(There is also a third HPI pumpinstalled.) The two parallel HPI trains are connected but are kept isolated by manual valves (known as the crossover valves) that are t
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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 pcwer and the worst single failure (failure of diesel.o start) only one HPI pump would be available and two of the feu '
' ion 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 injectien flow injected into this line (about half of the cutput of one high pressure pump) could flew cut the break.
Therefore, for the worst ccmbina'tien of break location and single failure, only one-half of the ficw rate of a single high pressure ECCS pump would contribute to maintaining the ccolant inventory in the reactor vessel.
This situation had not been previously analyzed and 3&'.4 had indicated that the limits specified in 10 CFR 50.46 may be exceeded.
Folicwing discovery of this proble5, B&W stated that they had analyzed a spectrum of small breaks in the pump discharge line and had detarmined that to meet the limits of 10 CFR 50.a6, ocerater action was required to open the two manually operated crossover valves and to manually align the two motor driven isolation valves which had failed to coen.
This
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wou1d allcw the flew frcm the one HPI pump to feed all four reacter l
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f coolant legs.
B&W assumed that 30% of the flow would be lost thrcugh the break and 70% would refill the core.
By letters dated April 14 and 21,1978, supplemented by discussions with the staff, the licensee committed to provide for the necessary operator actions within the i
required time frame. That is, in the event of a small break and a limiting single failure, manual action would be taken to begin opening f
these valves within five minutes and have them fully opened and an adequate flow split obtained within an additional 10 minutes.
To f
facilitate this o,'eration, the licensee ccmmitted to maintain one of the i
series-connected, manually operated crossover valves normally open.
The analyses performed by B&W assumed that the flow split was established
'i at 650 seconds by operator action.
We tnerefore.ccncluded that the i
modeling of operator action used in the analyses was a reasonable apprcxi-mation of the operator action that actually will be taken, provided specific procedure;s were prepared and followed to assure such acticn.
Based on the B&W analyses availabls at that time, unrelated plant condi-tions which limited the attainable pcwer level, and the licensee's commitment to provide operator action consistent with that assumed in the analyses, we issued an Order for Modification of License dated April 26, 1978, which amended the facility license to: (1) limit the maximum rea'ctor power level to 2080 Mwt, (2) require operation in accordance with the procedures committed to by the licensee, and (3) require submission of further analyses as soon as possible.
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By letter dat'd July 18, 1978, B&W submitted a sumary further analyses of this event.
This sumary (the 2772 Summary) described the methods used and the results obtained for small breaks in the pump discharge piping for a reactor power of 2772 Nt, which is the rated power level of Rancho Seco.
The results provided in this 2772 Summary were obtained by incorporating two modifications in the B&W ECCS Evaluation Model. These modifications, which involve use of a two node inner vessel simulation and phase distributional multipliers for bubble rise in all control volumes within the reactor vessel, were described in a B&W letter to the staff, dated May 26, 1978, and have i
been reviewed and approved by the staff.
By letter dated July 18, 1978, the licensee stated that he had reviewed the B&W submittal of July 18, 1978, and had fou.M t:1e conclusions accept-able and applicable to Rancho Seco.
Based on this review, the licensee requested authorizatior,) to operate Rancho Seco at 100% full power (2772 ht).
In a submittal dated July 7,1978, the licensee also confinned that procedures for operator action consistent with the assumptions of the B&W analyses had been implemented, that drills had been conducted which verified that the required operations could be completed in less time than assumed in the B&W analyses and that all five operating shifts had been trained in the procedures.
Representatives of the Commission's regional office state that they have verified the licensee's implementation i
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of the precedures and have inspected the licensee's training records to verify that training in the procedures was conducted.
Based on the above, we conclude that the procedures implemented by the licensee relative to operator action in the event of a small break are accept-able.
In addition, in his letter dated April 21, 1978, the licensee
'. comitted to submit by July 21, 1978, a proposal for any 1cng-tenn modification (to eliminate the need for prcmpt operator action) considered appropriate.
Regarding the licensee's request for authorization to operate the facility at full power (2772 ?%t), we have reviewed the S&W submittal of July 18, 1978.
This submittal presents the results of analyses performed for reactor ecolant pump discharge line break sizes of 0.15, 0.10, 0.085, 0.07, 0.055 and 0.04 ft2 at a reactor pcwer level of 2772,'^at.
Based on these results, S&W states that with operator action consistent 2 discharge line break is with that medeled in the arilysis, a 0.07 ft the most limiting case.
In this case, core uncovery occurs fer about 410 seconds and the conservatively ca'leclated peak clad temperature is approximately 1092cF.
This temperature is well below the limit specified
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in 10 CFR 50.46(b).
e
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e Based on our review of these analyses, we find that the calculations support the conclusion that a 0.07 ft2 discharge line break is the most limiting case.
The analyses which have been submitted used a simplified input to the FOAM code for the distribution of steam sources which the licensee describes as conservative.
However, these analyses did not provide adequate justification that this approach was clearly conservative. Therefore, we require that the licensee provide a sensitivity analysis of the most limiting break (0.07 ft2 split on the pump discharge) using the more exact steam source distribu-tion as input to the F0AM code (normally used to predict the mixture height whenever core uncovery is predicted) to demonstrate that the assumptiens used in the submitted analysis are conservative. Accordingly, i
we cannot conclude at this tima that operation of Rancho Seco at 2772 Mwt would be fully in conformance with 10 CFR 30.46.
On the other hand, 1
for operation of this facility at power levels up to 2772 Mwt, ECCS performance calculations for the limiting small break indicate that this break has a very substantial margin on peak clad temperature below
_ the limits of 10 CFR 50.46(b) if operator action consistent with that assumed in the analyses is properly taken.
However, the additional analysis is needed to-fully confirm that the submitted calculations
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are suitably conservative and the submission of such analysis is being made a condition of this Order.
Accordingly, because of the very substantial margin on peak clad temperature below the limits of 10 CFR 50.46(b), the NRC staff believes that operation of Rancho Seco at power levels of up to 2772 Mwt in accordance with appropriate operating procedures identified herein will not endangar life or
. property or the common defense and security, and the conditions which were imposed by the Order of April 26, 1978, may be modified accordingly.
The conditions modified by this Order have been discussed with and agreed to by the licensee.
III.
Copies of.the following documents are available for inspection at the Commission's Public Document Room at 1717 H Street, Washington, D. C.
20555, and are being placed in the Commission's local public document
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room at the Sacramento City-County Library, Sacramento, California.
(1) Letters from J. J. Mattimoe to Rs W. Reid, Chief, Operating Reactors Branch #4, dated April 17 and 21,1978.
_ (2). Order for Modification of License, Docket No. 50-312, dated April 26, 1978.
(3) Letters frem J. H. Tay'.or to S. A. Varga, Chief, Light Water Reactors Branch #4, dated May 26 and July 18, 1978.
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(4) Letters from J. J. Mattimoe to R. W. Reid, Chief, Operating Reactors Branch !4, dated July 7 and 18,1978.
IV.
i Accordingly, pursuant to the Atomic Energy Act of 1954, as amended, I
and the Comission's Rules and Regulations in 10 CFR Parts 2 and 50, our ORDER FOR MODIFICATION OF LICENSE dated April 26, 1978, is super-seded effective this date and IT IS ORDERED THAT Facility Operating License No. DpR-54 is hereby amended by adding the following new
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provisions:
2.
(1) As soon as possible, the licensee shall submit a reevaluation of ECCS cooling performance calculated in accordance with the B&W Evaluation Model for operation with operating procedures described in its letters of April 14, 1978 and April 21, 1978, which is wholly in conformance with 10 CFR 50.16 except for the credit for operator action within 10 minutes after initiation of the event.
(2) The steady state reactor core power level shall not exceed 2772 Mwt,
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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, and July 7, 1978, except that the maximum time for completion of operator action shall be 10 minutes after initiation of the event, and
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7590-01 (4) As soon as possible, the licensee sha'll submit a description and safety evaluation of a proposed plant modification which will i
eliminate reifance on prompt operator action described herein.
i FOR THE NUCLEAR REGULATORY COMMISSION i
.x n Vi or St.eTTo', Jr(
rector Division of Opera' ting Reactors Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this 21st day of July 1978.
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