ML20207B738

Notice of Violation from Insp on 880401-0527.Violation Noted:Drywell High Pressure Channel C Instrument Inoperable in That Instrument Rack Valve Partially Closed,Generating Actuation Setpoint of Approx 2.38 Psig

ML20207B738
Person / Time
Site:	Fermi
Issue date:	07/25/1988
From:	Greenman E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20207B736	List: IR 05000341/1988012 ML20207B731 ML20207B738
References
50-341-88-12, NUDOCS 8808040123
Download: ML20207B738 (4)

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I NOTICE OF VIOLATION Detroit Edison Company Docket No. 50-341 Enrico Fermi Nuclear Power Station

- Unit 2 During an NRC inspection conducted on April 1 to May 27, 1988, violations of NRC requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1988), the violations are listed below:

1. Fermi Unit 2 Technical Specifications Section 2.2.1 requires that the reactor protection system instrumentation setpoints be set consistent with the trip setpoint values shown in Table 2.2.1-1. Table 2.2.1-1 specifies the maximum setpoint for each of the f our Drywell Pressure Channels as 1.88 psig.

Contrary to the above, during the period of October 16, 1986 to March 16, 1 1988, the Drywell High Pressure Channel C instrument was inoperable in that the instrument rack valve was partially closed generating an actuation setpoint of approximately 2.38 psig.

This is a Severity Level IV violation (Supplement I).

2. Fermi Unit 2 Technical Specifications Section 6.8.1 requires that written procedures be established, implemented, and maintained covering activities such as, but not limited to, surveillance, equipment control and test activities of safety-related equipment.

P0M 44.010.039 includes restoration to service of the rack and instrument ~

valves as well as independent verification of the valve position.

P0M 41.000.09 provides guidelines to personnel for removing equipment from service, working on it, returning it to service, and requires independent verification of valve positions.

Procedure P0M 12.000.080, Conduct of Electrical Field Activities, Paragraph 7.5.1, states "An independent second check of restoring to normal shall be performed for all interim alterations performed under work orders designated as Safety-Related on the work order package Attachment A."

Contrary to the above:

a. On October 16, 1986, POM 44.010.039 was not adhered to in that the rack isolation valve for the Channel "C" RPS Drywell High Pressure Instrument (C71N0500) was not returned to its proper position.

8808040123 888072 PDR Q

ADOCN 05000341 PDC

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Notice of Violation 2

b. On March 15, 1988, a utility-non-licensed and utility-licensed individual indicated via P0M 41.000.09 that the above rack valve

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had been verified open when in fact neither individual actually verified the rack valve position.

c. On April 1,1988, utility electricians did not independently verify restoration of connectors at 0075F117 and 0075F118 associated with safety related work request 0232880209.

1 This is a Severity Level IV violation (Supplement I).

3. Technical Specification 3.3.3 Table 3.3.3-1 requires two channels per l trip system to be operable when in cold shutdown and the associated ECCS l is operable. If the number of channels is less than the minimum stated in the Table, then Action 30 of the Technical Specifications Table is invoked. The action states "For one trip system, place that trip system I in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or declare the associated ECCS inoperable."

Contrary to the above, on April 22, 1988, plant operators did not place ,

a channel of core spray reactor pressure in a tripped condition by 1718 ,

when it was discovered inoperable at 1618 in that the failure mechanism of the instrument was relied upon as the tripped condition even though there was no positive channel reset associated with the instrument channel.

This is a Severity Level IV violation (Supplement I). .

4. Technical Specification 3.7.1.2 requires that two independent emergency equipment cooling water (EECW) system subsystems shall be operable in Operational Conditions 1, 2, 3, 4, and 5.

Technical Specification 3.0.3 requires that when a Limiting Condition for Operation is not met, except as provided in the associated action requirements, within one hour action shall be initiated to place the unit in an Operational Condition in which the Specification does not apply.

Contrary to the above, on January 7 and March 3, 1987, both divisions of EECW were simultaneously placed in manual override for periods greater than one hour while the reactor was crit cal and action was not initiated within one hour to place the unit in an Operational Condition in which the Specification does not apply.

This is a Severity Level IV violation (Supplement I).

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Notice of Violation 3

5. 10 CFR 50, Appendix B, Criterion V states in part "Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings . . . ."

Procedure POM 12.000.06, "Data Collection for Diagnostic Testing," requires that Scquence of Events (SOE) tests be used to provide instructions for (1) engineering evaluation and data collection to determine the need for repairs or modifications and (2) operation of a system or component for troubleshooting purposes. The SOE is to be coordinated by the Technical Engineer and approved by the Onsite Review Organization (OSR0).

Procedure POM 23.621, "Main Control Room Annunciator and Sequence Recorder," Section 4.7.1.3. requires the permission of the Nuclear Shift Supervisor (NSS) to defeat an alarm. Attachment 1, Page 2 of 2 provides the documentation of that approval.

Procedure POM 12.000.082, "Emergency Diesel Generator Start / Failure Log Preparation and Evaluation," Section 5.1.2 requires the time /date when a diesel is unloaded and shutdown.

Plant Order EF0-8080, "Operator Aids," identifies operator graphics (otherwise known as drawings) as an operator aid, requires each operator aid to have a serial number, and requires a monthly review of the operator aid list for correctness and need.

Contrary to the above:

a. On August 15, 1987, a handwritten SOE was used for data collection and troubleshooting purposes as documented in Deviation Event Report 87-304 and the SOE was not OSR0 approved nor coordinated by the Technical Engineer.

b. Two alarms were defeated without NSS approval in that Attachment 1, Page 2 of 2 was not signed by the NSS.

c. Emergency diesel generator unloading and shotdown times were not documented for all generator runs.

d. Serial numbers had not been assigned to operator graphics.

e. The operator aid log was not being audited on a monthly basis.

This is a Severity Level IV violation (Supplement I).

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.t n Notice of Violation 4

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With respect to Item 4, the inspection showed that action had been taken to correct the identified violation and to prevent recurrence. Consequently, no reply to this violation is required and we have no further questions regarding this matter. With respect to Items 1, 2, 3, and 5, pursuant to the provisions of 10 CFR 2.201, you are required to submit to this office within thirty days of the date of this Notice a written statement or explanation in reply, including for each violation: (1) corrective action taken and the results achieved; (2) corrective action to be taken to avoid further violations; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause shown.

JUL 2 5128 Dated Edward G. Greenman, Director Division of Reactor Projects A

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