ML19326C903
| ML19326C903 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/21/1978 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | NRC COMMISSION (OCM) |
| Shared Package | |
| ML19326C900 | List: |
| References | |
| NUDOCS 8004280825 | |
| Download: ML19326C903 (8) | |
Text
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UtlITED STATES OF A'iER'ICh NUCLEAR REGULATORY COMf!ISSION In the Matter of
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ARKANSAS POWER AND LIGHT COMPANY
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Docket No. 50-313
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Arkansas ~ Nuclear One - Unit No.1
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ORDER FOR MODIFICATION OF LICENSE I.
The Arkansas Power & Light Company (the licensee), is the, holder of Facility Operating License No. DPR-51 which authorizes tne operation
-of the nuclear power reactor known as Arkansas Nuclear One - Unit No 1,
(the facility) at steady reactor power levels not in excess of 2568 megawatts theraal (rated power).
The facility. consists of a Babcock and Wilcox Company designed pressurized reactor (P1lR) located at i
the-licensee's site in Pope County, Arkansas.
p II.
In accordance with the requirements of the ' Commission's FCCS Acceptance Criteria 10 CFR 50.46, the licensee submitted on July 9, 1975, an ECCS evaluation 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 designer of the Nuclear Steam Supply System
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A 8004280 MbM
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, for this facility.
The B&W ECCS Evaluation tiedel 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 evaluation indicated that with the limits set forth in the facility's Technical Specifications, the ECCS cooling performance for the facility would conform with the criteria contained in 10 CFR 50.46(b) which govern calculated peak clad tenperature, naximum cladding oxidation, maxinun 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 limiti'ng srall break to be in the suction line of the reactor coolant pump. 'For this assumed break location all high pressure ECCS flow injected 'into the system would add to the reactor coolaat inventory once the coolant level decreased to the height of the reactor ves:el nozzles.
Recently, B&W concluded that previous analyses did not consider that scne of the high pressure injection (HPI) flow would be lost if the break were located in the discherge line.
As a result, recent analyses have shown that the discharoe line break is more limiting than the suction line break.
The Arkansas Nuclear One - Unit No. I plant has an ECCS configuration wnicn consists of two hich pressure injection (HPI) trains.
Each train has a HPI ponp 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 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 offsite power and the worst single failure (failure of diesel to start) only one HPI punp 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-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 contribute to maintaining the coolant inventory in the reactor vessel.
This situation had not been previously analyzed and D&W had indicated that the limits specified in 10 CFR 50.45 nay be exceeded.
B&W has-stated that they have analyzed a spectrum of small breaks in the
., punp discharge line and have determined that to meet the linits of 10 CFR 50.46, operat 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 flcu from the one J
I l
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 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 adequate flow split obtained within 10 minutes. The analyses performed by B&W 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&U'has stated that a.1 ft.2 discharge line break, with the afore-nentioned operator actions, is the most limiting case. To arrive at this conclusion, B&W has performed analyses at break sizes cf.5,.2,
.1, and.04 f t.2, with and without the assuned operator action. The results for a power of 2568 !!wt, which were obtained using an approved Appendix K nodel for blowdown, indicate core uncovery for about 250 seconds for the 0.1 ft.2 limiting break. For this break size B&W has conservativel.y estimated the peak clad tenperature to be approximately.
1200 F; well below the'linits of 10 CFR 50.46(b).
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5-B&U has indicated the manner in which the calculational methods have been revised and has indicated that their revised calculations are wholly in conformance with the requ,irements 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 a. ' opportunity to review the B&W revised calculations, the staff has recommended and the licensee has 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 IWt, 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 Deak clad temocrature below the limits of 10CFR50.46(b).
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 spec.ified in this Order, will assure that t-ECCS f-
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will conform to tne perfornance requirenents of 10 CFR 50.46(b).
Accordingly, such limits provide reasonable assurance that the public health and safety will not be endangered.
Upon notification by the ilRC staff, the licensee conaitted to provide the staff with B&W's reevaluation of ECCS performance applicable to the licensee's facility as pronptly as possible, to limit operation as specified herein, and to subnit a technical specification requiring appropriate operating procedures to assure required operator action as discussed herein.
Such procedures were described and the commitnents confirmed by the licensee's letter of April 17, 1978, supplemented by letters 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 NRC Order.
Upon satisfactory completion of our assessment of the revised evaluation, we will accordingly modify the authorization to operate the facility.
IV.
Copies of the following document are available for inspection at the Comnission's Public Document Room at 1717 H Street, Washington, D.C.
- 20555,
, and are being placed in the Commission's local public document room at the Arkansas Polytechnic College, Russellville, Arkansas.
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7-(1 ) Letters fron Mr. Donald A. Rueter to Mr. R. W. Reid, Chief Operating Reactors Branch #4, dated April 17.and 21,1978, and from Mr. Daniel H. Williams to Mr. R. W. Reid dated April 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. DPR 51 is hereby amended by adding the folicwing new provisions:
(1) As soon as poss.ible, the li.censee shall submit a reevaluation wholly in conformance with 10CFR50.46 of ECCS cooling performance calculated in accordance with the B&W Evaluation ibdel for operation with operating procedures described in its letters of April 17, 1978, and April 21, 1978, (2) Until further authorization by the Commission, the power level shall not exceed 2311 Mwt, and O
e
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8-(3) Until further authorization by the Commission, the licensee shall-operate in accordance with the procedures described in its letter of April 17, 1978, supplemented by letters dated April-21, 1978.
FOR THE flVCLEAR REGULATORY C0llMISSIO!J A
Victor Ste 1 r.,
ctor Division of Operating Reactors Office of fluclear Reactor Regulation Dated at Bethesda, flaryland, this 21st day of April 1978.
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