ML20198J353

Notice of Violation from Insp on 980504-22 & 0608-26. Violation Noted:As of October 1997,licensee Failed to Provide Procedures to Implement Compensatory Measures for Standby Liquid Control Tank Heater Failure,Per USAR

ML20198J353
Person / Time
Site:	Cooper
Issue date:	12/17/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198J332	List: IR 05000298/1998015 ML20198J329 ML20198J353
References
50-298-98-15, NUDOCS 9812300149
Download: ML20198J353 (5)

Text

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l ENCLOSURE 1 NOTICE OF VIOLATION Nebraska Putiic Power District Docket No.: 50-298 Cooper Nuclear Station License No.: DPR 46 l

During an NRC inspection conducted on May 4-22 and June 8-26,1998, five violations of NRC requirements were identified. In accordance with the " General Statement of Policy and l Procedure for NRC Enforcement Actions," NUREG-1600, the violations are listed below:

l A. 10 CFR Part 50, Appendix B, Criterion V, states, in part, that 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. Furthermore, these instructions, procedures, or i drawings shallinclude appropriate quantitative or qualitative acceptance criteria for l determining that important activities have been satisfactorily accomplished.

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1. Contrary to the above, in July 1993, a change was made to Procedure 6.DG.602,

" Diesel Fuel Oil Availability," that established a 4-inch water level acceptance criterion in the emergency diesel generator fuel oil tanks. This procedure's l acceptance criterion was not appropriate to the circumstances in that the l maximum water level of 4-inches in the emergency diesel generator fuel oil tanks would allow water to be pumped to the day tank.

2. Contrary to the above, as of October 1997, the licensee failed to provide procedures to implement compensatory measures for a standby liquid control tank heater failure as described in the Updated Safety Analysis Report. 1 I

3. Contrary to the above, as of November 1997, Procedure 13.17," Residual Heat l Removal Heat Exchanger Performanco Evaluation," failed to provide appropriate  ;

acceptance criteria for determining that heat exchanger performance was l acceptable in that it directed that a fouling factor calculated from as-measured l

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data be compared to a fouling factor that was based on design basis accident l conditions. )

l l 4. Contrary to the above, as of December 4,1997, Surveillance  ;

Procedure 6.OG.601, " Daily Surveillance Log," Revision 9, did not include  !

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[ appropriate acceptance criteria in that the specified limit of 90 degrees F did not

' I include the instrument uncertainties of Temperature Indicator Ml-TR-3020. As a result, the use of Temperature Indicator Ml-TR-3020 could permit s.ervice water l temperatures to exceed the specified limit.

l l S. Contrary to the above, as of November 1997, Procedure 13.15.1, " Reactor Equipment Cooling Heat Exchanger Performance Analysis," was inappropriate to the circurnstances. The equation for calculating the log mean temperature difference was incorrect. Incorrect values were used to compensate for plugged tubes, instrument uncertainties were not considered, and the excel spreadsheet 9812300149 981217 PDR ADOCK 05000298 G PM i

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l used to supplement the analysis performed within the body of the procedure I contained equation errors. This resulted in erroneous test data.

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6. Contrary to the above, as of December 1997, Surveillance Test l Procedures 6.1SWBP.101, "RHR Service Water Gooster Pump Flow Test and l Valve Operability Test (DIV 1)," Revision 4, and 6.2SWBP.101, "RHR Service  !

Water Booster Pump Flow Test and Valve Operability Test (DIV 2),"  !

Revision 4,were not appropriate to the circumstances. The procedures specified I the use of gauges that did not have the appropriate range. This resulted in l Pressure Gauges SW-PI-385A through 385D being in an over-ranged condition, which could have yielded inaccurate data.

I This is a Severity Level IV violation (Supplement l} (50-298/9815-01).

B. 10 CFR Part 50, Appendix B, Criterion Ill, requires, in part, that measures be i established to assure that applicable regulatory requirements and the design basis for I those structures, systems, and components to which this appendix applies are correctly l translated into specifications, drawings, procedures, and instructions. Criterion lli further requires that design control measures provide for verifying or checking the adequacy of the design, such as, by performance of design reviews by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Criterion ill also states, in part, that design changes shall be subject to design control measures commensurate with those applied to the original design.

1. Contrary to the above, as of November 1997, design control measures commensurate with the original design were not applied to electrical loads that were plugged into the welding receptacles. Since these welding receptacles i were connected to the Class 1E ac electrical distribution system, the uses of l these receptacles were defacto modifications that were not subjected to the design controls of the design control program.

2. Contrary to the above, as of November 1997, design control measures were not commensurate with those applied to the original design, in that, Calculation 86-155, " Control Room HVAC - System Balancing," Revision 0, was not updated to reflect Design Modification 93-257," Replacement of the Control i Room Booster Fan BF-C-1 A with a Higher Capacity Fan." This modification replaced the emergency booster fan, which had a capacity of 1000 cfm.

Calculation 86-155 concluded that a flowrate of 3000 cfm was necessary to pressurize the control room envelope to 1/8-inch differential pressure with respect to the adjacent areas.

In addition, Calculation 94-134," Verification of Control Room Emergency Filter System Capability," Revision 1, incorrectly calculated the face velocity over the control room emergency filtration carbon filter beds and the resultant efficiency of this component by neglecting the reduction of face area due to the geometric influence of the supporting angle frame.

3. Contrary to the above, on October 5,1991, the licensee failed to properly translate the specifications for the service water temperature into Calculation 2

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NEDC 93-08,"RHR Heat Exchanger Fouling Factor Determination for Mode C2,"

Revision 0, which was required to be performed as part of Procedure 13.17,

" Residual Heat Removal Heat Exchanger Performance Evaluation." As a result, l

Residual Heat Removal Heat Exchanger A showed a significant mismatch in heat removed from the residual heat removal side of the heat exchanger and l heat entering the service water side of the heat exchanger.

4. Contrary to the above, as of November 1997, the licensee failed to properly verify the adequacy of the design. Calculation NEDC 93-050,"RHR Quad

Temperature," Revision 1, failed to consider heat loads from two residual heat i removal pumps operating in one room during accident conditions when the fan coil unit for the room was ir. operable. A low pressure coolant injection signal would start both residual heat removal pumps in the room if emergency power was available.

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5. Contrary to the above, on May 14,1997, the licensee failed to properly translate j specifications into calculations. Nonconservative and incorrect inputs were used '

in Calculation NEDC 97-074," Evaluation of the Service Water System,"

Revision 1. As a result, the calculation was not appropriate to demonstrate the capability of service water system to cool the reactor equipment cooling system.

This is a Sevedty Level IV violation (Supplement l} (50-298/9815-02).

! C. 10 CFR Part 50, Appendix B, Criterion XI, " Test Control," states, in part, that a test l L program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures, which incorporate the requirements and acceptance limits contained in applicable design documents.

! Contrary to the above, in November 1997, following replacement of the Emergency Diesel Generator 1 lubricating oil pump, testing was not pedormed to demonstrate that a lubricating oil pump relief valve would perform satisfactorily in service. Specifically, while the vendor recommended valve-set point was 75 psig, the actual as-found set point was found to be 90 psig.

This is a Severity Level IV violation (Supplement l} (50-298/9815-03).

D. 10 CFR 50.73(a)(2)(ii)(B) requires, in part, that the licensee submit a licensee event report within 30 days after discovery of any condition that resulted in the nuclear power plant being in a condition that was outside the design basis.

Contrary to the above, on July 24,1996, the licensee identified that the motors and operators for outboard primary Containment isolation Valves RHR-MOV-M0166A/B and l -M0167A/B at Torus Penetration X-214 hari been misclassified as nonessential and were powered from the same power source, a condition that rendered the plant configuration outside the design basis for primary containment isolation. However, the licensee failed to report this as a condition that was outside its design basis.

This is a Severity Level IV violation (Supplement 1) (50-298/9815-06).

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I l E. 10 CFR 50.59(a)(1) states, in part, that a licensee may make changes in the facility as d described in the Safety Analysis Report without prior Commission approval, unless the l proposed change involves an unreviewed safety question.

10 CFR 50.59(a)(2) states, in part, that a proposed change shall be deemed to involve an unreviewed safety question it the probability of a malfunction of equipment important l to safety may be increased.

Updated Safety Analysis Report, Volume IV, Chapter X, Section 6.3, states, "The (REC]

system shall be designed to supply an adequata supply of cooling water to the CSCS l (core standby cooling system] areas and the RHR [rosidual heat removal) pumps under i

all accident and transient conditions."

Contrary to the above, in 1991, without prior Commission approval, the licensee made a change to the facility described in the Safety Analysis Report such that the reactor equipment coo!!ng system would not have an adequate supply of cooling water to the core standby cooling system (high pressure coolant injection, core spray, reactor core isolation cooling, and residual heat removal) area room coolers under postulated accident and transient conditions. This increasea the probability of a maifunction of the core standby cooling system pumps due to high room temperatures under certain design basis conditions.

This is a severity Level IV violation (Supplement I) (50-298/9815-07).

Regarding Violations A.1, A.2, A.5, B.1, B.5, D and E above, the NRC has concluded that information regarding the reason for the violations, the corrective actions taken and planned to correct the violations and prevent recurrence, and the date when full compliance will be achieved is already adequately addressed on the docket in NRC Inspection Report 50-298/98-15. However, you are required to submit a written statement or explanation pursuant to 10 CFR 2.201 if the description therein does not accurately reflect your corrective actions or your position. In that case, or if you choose to respond, clearly mark your response

as a " Reply to a Notice of Violation," and send it to the U.S. Nuclear Regulatory Commission, ATTN

Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region IV, 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 Notice of Violation l (Notice).

Pursuant to the provisions of 10 CFR 2.201, Nebraska Public Power District is hereby required to submit a written statement or expianation 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 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 Notice 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 l reason for the violation, or, if contested, the basis for disputing the violation or severity level; j (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 correspondence,if the correspondence adequately addresses the required response. If an adequate reply is not 4

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f received within the time specified in this Notice, an order or a Demand for information may be issued as to why the 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, consideretion will be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response to the i Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.

Because your response will be placed in the NRC Public Document Room (PDR), to the exten possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be pl. iced in the PDR without redaction. If personal privacy or proprietary inforrnation is necessary to provide al acceptable response, then please provide a bracketed copy of your

, response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in

, detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by i 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial  !

information). If safeguards information is necessary to provide an acceptable response, please  ;

provide the level of protection described in 10 CFR 73.21. l l

Dated at Arlington, Texas, this 17th day of December 1998.

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