ML19308A734
| ML19308A734 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/26/1978 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | NRC COMMISSION (OCM) |
| Shared Package | |
| ML19308A733 | List: |
| References | |
| NUDOCS 7912100616 | |
| Download: ML19308A734 (6) | |
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7590-01 UNITED STATES OF AMERICA gg NUCLEAR REGULATORY COMMISSION In the Matter of
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DUKE POWER COMPANY
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nackets Nos. 50-269
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50-270 Oconee Nuclear Station, Units Nos.1, 2,
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and 50-287 and 3 ORDED F02 MODIFICATION 07 LICENSE I.
The Duke Power Company (the licensee), is the hold 5r of Facility Operating Licenses Nos. OPR-35, 47, and 95 which authorize the operation of the nuclear power reactors known as Oconee Nuclear Station, Units Nos.1, 2, and 3, (the facility) at steady reactor power levels not in excess of 2568 megawatts thermal (rated cower) for each unit.
The facility consists of Babcock & Wilcox Company designed pressurized water reactors (PWR) located at the licensee's site in Oconee County, South Carolina.
II.
In accordance with the requirements of the Commission's ECCS 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 Bab~ock c
& Wilccx Company (B&W), the designer of the Nuclear Steam Supply System 4
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7590-01 for this facility.
The B&W ECCS Evaluation Model had been previously found to confora to the requirements of the Commission's ECCS Acceptance Criterf a,10 CFR Part 50.46 and Appendix K.
The eva'uation 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 temperature, maximum cladding exidation, maximun hydrogen generation, cool-able gecmetry and long-term cooling.
On April 12,1973, 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
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calculated peak clad temperature might exceed 2200F.
Previous small break analyses fo'r B&W 177 fuel assembly (FA) Icwered lecp plants had identified the l'imiting small break to be in the suction line of the reactor ccolant pump.
Recent analyses have shcwn that the discharge line break is more limiting than the suction line break.
Each Oconee Nuclear Station unit has an ECCS configuration which consists of.two high pressure injection (HPI) trains which are supplied by three HPI pumps.
Each train injects into two of the four reactor coolant system (RCS) cold legs on the discharce side of the RCS pump.
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The two parallel HPI trains are connected but are kept isolated by manual valves (known as the cross-over valves) that are normally closed. -
Duke Power has proposed to maintain all three pumps in an operable status.
The Q'conee emergency power system is designed with sufficient capacity for this mode of operation. Upon receiving a safety injection signal the HPI pumps are started and valves in the injection lines are opened.
Assuming loss of offsite pcwer and the worst single failure (the epi pump C or the HPI valve HP25), two HPI pumps would still be availaote and only one of the two 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 50% of the output of.two high pres-sure pum:s) could ficw out the break.
Therefore, for the worst c0mbination of break location and single failure, 50% of the flow rate of two high pressure ECCS oumpswould contribute to maintaining tqe coolant inventory in the reactor vessel.
This situation had not been previously analyzed and B&W had indicated that t,he limits specified in 10 CFR 50.46 may be exceeded.
B&W has stated that they have analyzed a spectrum of small breaks in the purp discharge line and have' determined that to meet the limits of-10 CFR 50.45, operator action is required to cpen the.two manual operatec e
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crossover valves and to manually align the motor driven isolation valve.
which had failed to open. This would allow the flow from the t o HPI pumps 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 The licensee has ccanitted to provide for the necessary operator core.
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 ninutes and have them fully opened and an adequate flow split obtained within the following 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 folicwed to assure such action.
B&W has stated that a.15 f t.2 discharge line break, with the afore-mentioned operator actions, is the most limiting case. 'To arrive at this conclusion, BAW has performed analyses at break sizes of
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.1, and.0a ft.2, using an approved Appendix,K medel for'blowdewn.
Additional analyses for the Oconee plants at 2568 Mwt indicate no core uncovery for the 0.15 ft.2 limiting break.
For this break si:e B&M has conservatively estimated the peak clad temperature to be well below the limitsof10CFR550.a6(b).
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g 7590-01 5-B&W has indicated the manner in which the calculational methods have been revised and has indicated that their revised calculations are 50.46.
- However, wholly in conformance with the requirements of 10 CFR B&W has not yet hao the opportunity to fully present the result of its calculations to the licensee for submittal to the NEC staff, and the staff has accordingly net hat the opportunity to fully assess the new calculations.
Therefore, until the staff has had the opportunity to fully assess the S&W revised calculations operation in accordance with the operating procedures specified in this Order, will assure that the ECCS will conform to the performance requirements of 10 CFD 50.46(b).
Accordingly, such procedures provid~e reasonable assurance that the public health and safety will not be endangered.
Upon notification by the NEC staf f, the licensee committed to provice the staf f with B&W's reevaluation of ECCS performance applicable to the licensee's facility as promptly as possible, and to s'ubmit a technical specification requiring appropriate operating procedures to assure required operator action as discussed herein.
Such procedures were described and the commitments confirmed 'by the licensee's letter of 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 acccedingly modify the authorization to operate the facility.
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IV.
Copies of the following document are available for inspection at the Comission's Public Docu ent R:cm at 1717 H Street, Washington, D.C.
- 20555, and are being placed in the Comission's local public document room at the Oc'onee County Library, 201 South Spring, Walhalla, South Carolina 29691.
fron Mr. William O. Parker, Jr. to Mr. Edsen G. Case, Acting (1 ) Letter Director, Office of Nuclear Reactor Regulation, dated April 21, 1978.
Accordingly, pursuant to the Atomic Energy Act of 1954, as amended, and tne Commission's Rules anc Regulations in 10 CFR Parts 2 and 50, IT IS ORDERED THAT Facility Operating Licenses Nos. CPR-38, 47, and 55 are hereby amended by adding the following new provisions:
(1) As soon as possible, the li.censee shall submit a reevaluation wholly in conformance with 100?R50.45 of ECCS cooling performance calculated in accordance with the S&W Evaluation Model for operation with operating procedures described in its letter of April 21, 1973, and (2) Until further authorization by the Commission.,-the licensee shall operate in accordance with the procedures described in its letter. of April 21, 1978.
FOR THE NUCLEAR REGULATORY COMMISSION lo A.
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' 'c to r Steilo,,Jr.
Director Division of 0:erating Reactors Office of Nuclear Reactor Regulation Dated at Bethesda, f'aryiand, this 26th day of April 1978.
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