ML20132G134

Notice of Violation from Insp on 961104-1204.Violation Noted:Licensee Discovered Reset Function for Valve RCI-MOV- M014 Powered by Ac Motor & Facility Not Operated as Described in FSAR & Written Evaluation of FSAR Not Performed

ML20132G134
Person / Time
Site:	Cooper
Issue date:	12/20/1996
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20132G007	List: ML20132G004 ML20132G134 ML20132G165
References
50-298-96-30, NUDOCS 9612260185
Download: ML20132G134 (3)

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ENCLOSURE 1 ,

1 NOTICE OF VIOLATION l- ,

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Nebraska Public Power District Docket No.: 50-298 j Cooper Nuclear Station License No.: DPR-46 l EA 96-487 l EA 96-488 During an NRC inspection conducted on November 4 through December 4,1996, three f violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violations are listed below: '

A. 10 CFR 50.63, Loss of all alternatina current power, requires that each nuclear power plant must be able to withstand for a specified duration and recover from a station blackout as defined in 10 CFR 50.2. It further states that the reactor core and associated coolant, control, and protection systems . . . . must provide sufficient capacity and capability to ensure that the core is cooled . . . .

I in response to NRC questions regarding the licensee's Station Blackou (SBO)

Analysis, the licensee responded in a September 30,1991, letter to the Commission that the "SBO evaluation for CNS assumes that HPCI and RCIC are available and that both start automatically at the onset of the event. However, the current evaluation assumes that HPCIis secured after one cycle of operation to further conserve battery energy. Thereafter, reactor vessel level would be primarily controlled by automatic operation of the RCIC system. One cycle of HPCl is sufficient to stabilize level, with no loss of coolant coverage of the core expected.

The RCIC system is considered sufficient to maintain water level beyond this point."

Contrary to the above, on August 20,1994, the licensee discovered that the reset i function for Valve RCIC-MOV-MO14, the RCIC turbine trip / throttle valve, was powered by an ac, not a de motor. Reactor vessellevel could not be automatically controlled by the RCIC system and core cooling could have been affected (01013).

This is a Severity Levellli violation. (Supplement 1)(298/96030-01) ,

l No response is required to this violation. j B. 10 CFR 50.59(a)(1) states, in part, that a licensee may make changes in the facility I as described in the safety analysis report without prior Commission Approval unless the change involves a change in the Technical Specifications incorporate in the license or an unreviewed safety question.10 CFR 50.59(b)(1) states, in part, that the licensee shall maintain records of changes in the f acility, to the extent that i these changes constitute changes in the facility as described in the safety analysis, and that these records must include a written safety evaluation which provides the basis for the determination that the change did not involve an unreviewed safety question.

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1 On October 20,1995, the licensee's FSAR, Section 8.5.6, stated, in part, that "the i residual heat removal (RHR) system can be intertied witil the Fuel Pool cooling I system if required. This capability increases the cpent fuel pool cooling capacity in '

the event that such additional capr:ity is necessitated by removal from the core of an unusually large number of fuel elements. The BHR system - fuel pool cooling ,

system intertie is sized to remove an emergency haat load . . . from the fuel pool which corresponds to fu'l core off-loading plus the batch of spent fuel discharged at the previous refueling outage.

In the NRC's safety evaluation supporting License Amendment 52 dated September 29,1978,it was indicated in Section 2.2 that the RHR cooling would be available when performing full core offloads.

Contrary to the above, on October 20,1995, the licensee changed the facility as described in the safety analysis report in that the facility was not operatt.d as described in the FSAR and a written safety evaluation of the change fro.n the FSAR had not been performed to deterniine whether this change involved ar. unreviewed safety question. Specifically, the licensee was in the process of rariorming a full core offload, and the RHR system was not available to assist the fuel pool cooling system in removing what the FSAR characterized as an emergency offload (02014).

This is a Severity Level IV violation. (Supplement 1)(298/96030-02)

C. Criterion ill of Appendix B to 10 CFR Part 50 requires that regulatory requirements and the design basis, as defined in 10 CFR 50.2 and as specified in the license application, for those structures systems and components to which the appendix applies are correctly translated into specifications, drawings, procedures, and instructions.

In the safety evaluation report which accompanied Amendment 52 to the facility operating license, the NRC staff acknowledged that the licensee's spent fuel pool and cooling systems were capable of handling the heat load associated with a full core discharge. However, this acknowledgement was based on certain design assumptions. In the Safety Evaluation Report, the staff stated that the maximum fuel pool heatload was associated with an offload that would occur 13 days after shutdown.

Contrary to the above, the desig,n basis ossumption that the maximum heat load was associated with full core discharge which was completed in 13 days was not translated into procedures. Procedure 2.3.2, " Fuel Pool Cooling and Demineralizer System," contained no administrative controls to ensure that fuel was not loadad at a rate that would exceed the 13-day assumption. In October 1995, the licenseo did exceed this offload rate (03014).

This is a Severity Level IV violation. (Supplement 1)(298/96030-03) l I

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No response is required for Violation A. For Violations B and C, a response is required in accordance with regulations as described below.

Pursuant to the provisions of 10 CFR 2.201, Nebraska Public Power District is hereby

required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN

Document Control Desk, Washington, D.C. 20555 with a copy to the

Regional Administrator, Region IV,611 Ryan Plaza Drive, Suite 400, Arlington, Texas I 76011, and a copy to the NRC Resident inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Not ce of Violation (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each violation

(1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correst oadence, if the correspondence adequately ,

addresses the required response, if an adequate reply is not received within the time  !

specified in this Notice, an order or a Demand for Information may be issued as to why the l license should not be modified, suspended, or revoked, or why such other action as may <

be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. However, if you find it necessary to include such information, you should clearly indicate the specific information that you desire not to be placed in the PDR and provide the legal basis to support your request for withholding the information from the public.

Datedgt Arlington, Texas i this/4 day ofM 1996 l l

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