ML20199H128

Notice of Violation from Insp on 970907-1018.Violations Noted:Battery Capacity After Svc Test Was Inadequate & Check Valve V003 & V002D F & H Were Not Flow Tested in Manner That Proves That Disk Travels to Seat on Cessation

ML20199H128
Person / Time
Site:	Farley
Issue date:	11/17/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20199H100	List: ML20199H095 ML20199H128 ML20199H154
References
50-348-97-11, 50-364-97-11, NUDOCS 9711260026
Download: ML20199H128 (5)

Text

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. l NOTICE Of VIOLATION Southern Nuclear Operating Company Docket Nos.: 50-348 and 50-364 farley Nuclear Plant. Units 1 and 2 License Nos.. NPF-2 and NPF-8 During an U.S. Nuclear Regulatory Commission (NRC) inspection conducted during the period September 7 through October 18. 1997. certain violations of NRC requirements were identifled. In accordance with the " General Statement of Policy and Procedures for NRC Enforcement Actions." NUREG-1600. the violations are listed below:

A. 10 CfR 50.36(c)(3). Technical Specifications. Surveillance Requirements, states, in part, that surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, and that limiting conditions for operation will be met.

Contrary to the above. as of March 14. 1997, the Auxiliary Building.

Direct Current (DC) Distribution technical specification (TS) surveillance requirement (4.8.2.3.2.C.5). battery capacity after a service test, was inadequate to ensure that limiting conditions for operation would be met.

Specifically. the battery surveillance requirement specified in TS Section 4.8.2.3.2.c.5 was less conservative than the design voltages specified in calculation 07597-E144. Battery operation at the TS allowed voltage would not have met the design requirement for supplying adequate voltage for all safety-related components.

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

B. Technical Specification Section 4.0.5 requires inservice testing of American Society of Mechanical Engineers (ASME) Code Classes 1. 2. and 3 lumps and valves in accordance with Section XI of the ASME Boiler and 3ressure Vessel Code and applicable Addenda. The licensee is committed to inservice testing in accordance with the 1983 Edition of the Code and Summer 1983 Addenda.

Section XI. Subsection IWV-3522 of the ASME Boiler and Pressure Vessel Code requires valves whose function is to prevent reverse flow to be tested in a manner that proves that the disk travels to the seat promptly on cessation or reversal of flow.

Contrary to the above, neither the turbine-driven auxiliary feedwater pump discharge check valve V003 nor check valves V0020. F. and H. were reverse flow tested in a manner that proves that the disk travels to the seat on cessation or reversal of flow. Either check valve V003 or check valves V0020. f. and H were required to perform a safety function in the closed position.

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! This is a Severity Level IV Violation (Supplement 1).

l C. 10 CFR Part 50. Appendix B. Criterion XI. Test Controls, states, in part, that a test program shall be established to ensure that all testing required to demonstrate structures, systems, and components will 9711260026 971117 gDR ADOCK 05000348 Enclosure 1 PDR

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.e Notice of Violation 2 l

perform satisfactory is identified and performed in accordance with  !

written test procedures. j 1

Technical Specification 6.8.1,a requires that applicable written procedures recommended in Mpendix A of Regulatory Guide 1.33. Revision l

2. 1978, shall be established, implemented and maintained. Appendix A.

Section 8.b of Regulatory Guide 1.33 recommends surveillance test procedures for the Auxiliary Feedwater System.

Contrary to the above, as of March 14, 1997, no surveillance test procedures were established for the turbine-driven auxiliary feedwater pump Class 1E battery to ensure that the battery will perform satisfactorily in servict to the required battery duty cycle in accordance with design basis requirements.

This is a Severity Level IV Violation (Supplement 1).

D. 10 CFR Part 50. Appendix B. Criterion V. Instructions. Procedures, and Drawings, states, in part, that activities affecting quality be prescribed by and performed in accordance with instructions, procedures or drawings which include appropriate acceptance criteria for determining the activity is satisfactorily accomplished.

Contrary to the above, as of March 14, 1997, the following were identi fied:

1. The Unit 2 Turbine Driven Auxiliary Feedwater (TDAFW) pump battery structural and electrical component installations were not installed in accordance with drawings and instructions (7597 E10,9 29. Revision 1. U265645A, Revision 2, and U263212, Revision D).

2. The surveillance procedure for verifying the forward flow for check valve V003 did not provide appropriate acceptance criteria for determining that the activity was satisfactorily accomplished.

Specifically, the acceptance criterion of 625 gallons per minute (gpm) full desigr. flow specifled for TDAFW pump check valve V003 in ".irveillance Test Procedure (STP) FNP-1(2)-STP 22.13 was incorrect, because the acceptance criterion did not consider the flow through the minimum flow line.

This is a Severity Level IV Violation (Supplement 1).

E. 10 CFR Part 50. Appendix B, Criterion Ill. Design Control, states, in part, that the design control measures shall provide for verifying or checking the adequacy of the design.

I Licensee's Operations Quality Assurance Program UFSAR 17.2.3.1. states that design control measures will comply with the requirements of American National Standard Institute (ANSI) N45.2.ll-1974. It further 1 states-that the design control system shall include a method of I.

adequately verifying the design.

l Enclosure 1 l

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. j Notice of Violation 3 Farley Support Procedure GO H 1. " Designer Interface Document." Revision ,

3. states that calculations or analyses )erformed as a part of

• engineering support activities for FNP s1all conform to the requirements i of Section 4.2 of ANSI N45.2.11-1974. -In addition, procedure GO H-1 states that existing calculations should be reviewed to determine if the  ;

calculation must be revised.

ANSI N45.2.11 states that design analyses shall be performed in a planned. controlled and correct manner, in addition, the design activities shall be prescribed and accomplished in accordance with .

procedures which provides adequate checking or verifying the results of the activity.

Contrary to the above, on March 14, 1997.

1. The design control measures did not ensure that calculations or analyses were verified and controlled adequately to ensure original plant design basis documentation is maintained current and to prevent inappropriate use of the existing calculations for the following examples:

a. Calculation 38.06. " Determine flow rate through pipe break in CCW system." Revision 0, concerning the Component Cooling Water (CCW) surge tank low low level setpoint. supersedes Calculation 35.6. " Evaluation of CCW Surge Tank Level Setpoints." Revision 0. yet Calculation 35.5 was shown as active on the licensee's calculation index.

b. Calculation 34.5 " Component Cooling Water System NPSH (ES No. 90 1820)." Revision 0, was affected by Calculation 39.3. e "CCW Surge Tank Analytical Limit for Level Setpoint."

Revision 0, but Calculation 34.5 was not revised accordingly,

c. Calculation 25.3 "Over pressurization of AFW Piping During Overspeed Testing." Revision 0, determined that an unacceptable piping pressure could develop in the event of turbine overspeed of 125% of rated speed. This calculation was not revised to reflect the acceptable overspeed setpoint of 115% of rated speed nor was a new calculation pre)ared to supersede Calculation 25.3 when design modification )CN B- .

88-1-5003. " Change Overspeed Trip Setting for TDAFW Pump."

Revision 1 was implemented.

d. Calculation E-144. " Determination of Battery Capacity Margins for Adecuacy of Voltage at Safety-Related Components for Various Loac Profiles." Revision 2. reflected a new battery instal'.ation but calculation E 42. " Steady-State DG Loading Calculation for LOSP. SI and SBO." Revision 8, wh4ch -

used calculation E-144 as an input, was not revised accordingly,

e. Calculation 40.02. " Verification of AFW Flow Bases."

Revision 3 (Unit 1), and the equivalent calculation for Unit Enclosure 1

Notice of Violation 4 2 (38.04) provide design basis information for the AFW system: yet Calculation 35.04. " Auxiliary feedwater System Head Curves (ES 90-1831)." Revision 0, which provides similar information. was not referenced or revised to  :

indicate that it did not contain design basis information.

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f. Calculation SC-96 1211-002. "CCW Heat Exchanger Maintenance  !

Repairs." Revision 1. states that it is judged that the l increase in weight resulting from the CCW heat exchanger  ;

modifications will not affect the acceptability of the i foundation anchorages. However, the calculation for the .

anchorages was not revised nor was the increase in heat exchanger weight re'iected in the appropriate calculation. -

2. Design control measures were not 3roperly implemented to verify or '

check the adequacy of the design ) asis differential pressures specified for CCW motor operated valves (MOVs). S)ecifically, differential pressures identified in Design Basis Jocument.

U418109. Revision A for CCW system containment isolation valves MOV 3046. 3052. and 3182 were non-conservative as the effect of post Loss of Coolant Accident (LOCA) containment pressure was not considered.

This is a Severity Level IV Violation (Supplement 1).

F. 10 CFR Part 50. Appendix B. Criterion XVI states that conditions adverse to quality such as malfunctions and deviations be promptly identified and corrected.

Contrary to the above. as of March 14, 1997, the licensee failed to s promptly correct several differences between CCW system piping and instrument drawings D 205002. Sheet 1. Revision 21: Sheet 2. Revision 10: and Sheet 3. Revision 2 and standard operating procedures (SOPS) '

FNP 1(2)-SOP-23.0A " Component Cooling Water System." Revision 5. and FNP-2-SOP-2.1A " Chemical and Volume Control System," Revision 8.

concerning the existence of caps on vents and drain lines. Differences between piping and instrumentation drawings (P&lDs) and procedures were previously identified in the licensee's CCW system self-assessment in 1990, but no corrective action was identified by the licensee.

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

G. 10 CFR 73.21(d)(2) requires Safeguards Information (SGI) to be stored in a-locked security storage container while unattended. 10 CFR 73.21(e) -

requires each document that contains SGI to be marked as " Safeguards Information" in a conspicuous manner to indicate the presence of protected information.

Contrary to the above, as of September 17. 1997 SGI was stored in the main control room in an unlocked container that was routinely left unattended for short periods of time. Furthermore, some of the SGI was bound in nondescript. unmarked folders and was not marked in a conspicuous manner to indicate its presence.

Enclosure 1 i

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4 Notice of Violation 5  !

This is a Severity Level IV violation (Supplement 111).  ;

Pursuant to the provisions of 10 CFR 2.201. Southern Nuclear Operating Company is hereby required to submit a written statement or ex)lanation to the U.S. l Nuclear Regulatory Commission ATIN: Document Control )esk, Washington, DC 20555, with a copy to the Regional Administrator, Region 11, and a copy to the ,

NRC Resident inspector, Farley Nuclear Plant, 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 A through F: (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 incluue previously docketea correspondence, 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 license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause ir, shown. consideration will be given to extending the response time.

The NRC has concluded that information regarding the reason for Violation G, the corrective actions taken and planned to correct the violation and prevent recurrence is already adequately addressed on the docket in the attached inspection report. Therefore, you are not required to respond to this violation unless the description therein does not accurately reflect your corrective actions or your position. In that case or if you choose to provide additional information. you should follow the instructions specified in the enclosed Notice.

Because your res)onse will be placed in the NRC Public Document Room (PDR), to the extent possiale, it should not include any personal privacy, )roprietary, or safeguards information so that it can be placed in the PDR wit 1out 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.

Dated.at Atlanta, Georgia this 17th day of November 1997 Enclosure 1

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