ML20141C981

Notice of Violation from Insp on 970302-0412.Violation Noted:On 970324,licensee Failed to Open C-S Sfpcs Pump Discharge Valve When Placing C-S Pump in Service.Sys Operated for About 2.5 Hours with Discharge Valve Closed

ML20141C981
Person / Time
Site:	Sequoyah
Issue date:	05/12/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20141C976	List: ML20141C971 ML20141C981 ML20141C988
References
50-327-97-03, 50-327-97-3, 50-328-97-03, 50-328-97-3, EA-97-083, EA-97-233, EA-97-83, NUDOCS 9705190290
Download: ML20141C981 (6)

Text

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NOTICE OF VIOLATION Tennessee Valley Authority Docket Nos. 50-327 and 50 328 Sequoyah Units 1 & 2 License Nos. DPR-77 and DPR 79

, EA 97-083 EA 97 233 b

During an NRC inspection conducted from March 2 through April 12, 1997, 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. Technical S>ecification 6.8.1.a requires, in part, that procedures shall be establisled, implemented, and maintained covering the activities recommended in Appendix A of Regulatory Guide 1.33, Revision 2 February  :

1978, " Quality Assurance Program Requirements (Operations)." Appendix A of Regulatory Guide 1.33, Section 4, includes procedures for startup and shutdown of the Fuel Storage Pool Purification and Cooling System.

System Operating Instruction (50I) 78.1, Spent Fuel Pit Coolant System (SFPCS), Revision 53, provided instruction for placing in service the C-S SFPCS pump, including opening the pump discharge valve.

Contrary to the above, on March 24, 1997, the licensee failed to open the C S SFPCS pump discharge valve when placing the C-S aum) in service.

The system subsequently operated for about 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> wit 1 t1e discharge valve closed and cooling flow to the spent fuel pool isolated.

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

B. Technical S)ecification 6.8.1.a requires, in part, that procedures shall be establisled, implemented, and maintained covering the activities t recommended in Appendix A of Regulatory Guide 1.33 Revision 2. February 1978, " Quality Assurance Program Requirements (Operations)." Appendix A of Regulatory Guide 1.33, Section 9, requires procedures for maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with documented

, instructions.

d Work Order 97-006117, was issued to provide instructions to repair electrical conductors which had been damaged during a modification to a l main control room hand switch. The Work Order required, as a prerequisite to the repair, that electrical breaker 1716 be verified open.  !

I Enclosure 1 9705190290 970512

- {DR ADOCK 05000327 PDR

• i NOV 2 Contrary to the above, on April 4, 1997, Work Order 97-006117 was not followed in that the licensee did not ensure that electrical breaker 1716 was open arior to beginning repair of the damaged conductors.

Consequently w1en repairs began, power was lost to the 1A A Shutdown Board.

This is a Severity Level IV Violation (Supplement I), for Unit 1 only.

C. Unit 2 Technical Specification.(TS) 4.3.1.1.2 requires that the total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation. This TS is applicable to components listed in Table 3.31, Reactor Trip System Instrumentation. Table 3.3 1.

Functional Unit 22, Reactor Trip System Interlocks Section "G",

contains the Reactor Trip - P-4 function.

Contrary to the above:

1. On August 29, 1996, the Unit 2 "A" reactor trip breaker total interlock function was not demonstrated to be OPERABLE at least ,

once per 18 months for the Table 3.31 Functional Unit 22.G.  !

Reactor Trip P 4 interlock, in that the "A" reactor trip breaker >

P 4 reactor trip turbine trip interlock (breaker contactor) had not been tested since May 1994 as a " spare" and was installed as i an OPERABLE component. The breaker was not adequately tested until January 18, 1997.

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2. On August 29, 1996, for the "A" reactor tria breaker and on October 22, 1996, for the "B" reactor trip areaker, the total interlock function was not demonstrated to be OPERABLE at least once per 18 months for the Table 3.3 1 Functional Unit 22.G, Reactor Trip - P 4 interlock, in that following replacement of the

. in service reactor trip breakers with spare / refurbished breakers, the reactor trip breaker auxiliary contacts to cubicle contacts were not checked to verify circuit continuity. This invalidated the previous 18 month TS 4.3.1.1.2 surveillance, which was last completed on May 31, 1996. The breakers were not adequately tested until January 18. 1997. for the "A" reactor trip breaker and on February 7, 1997, for the "B" reactor trip breaker.

This is a Severity Level IV Violation (Supplement I), for Unit 2 only.

D. Technical S>ecification 6.8.1.a requires, in part, that procedures shall be establisled, implemented, and maintained covering the activities recommended in Appendix A of Regulatory Guide 1.33, Revision 2. February 1978, " Quality Assurance Program Requirements (Operations)." Appendix A of Regulatory Guide 1.33, Section 5, requires procedures for Abnormal, offnormal or Alarm conditions.

Site Standard Practice (SSP) 8.50, Conduct of Technical Support, Revision 10, requires that the Technical Support System Engineer respond to a Technical Support Investigation Request (TSIR).

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O NOV 3 Contrary to the above, Technical Support failed to adequately respond to TSIR 96 N55 77 580, dated October 29, 1996 and TSIR No. 96 NSS 77 630, dated December 28, 1996, which identified offnormal and alarm conditions associated with the Turbine Driven Auxiliary Feedwater (TDAFW) sump pumps until March 13, 1997. On March 17, 1997, Technical Support determined that both sump pumps were significantly degraded, which potentially could have caused the TDAFW pump to overspeed.

This is a Severity Level IV Violation (Supplement I), for Unit 1 only.

E. Technical Saecification 6.8.1.a requires, in part, that procedures shall be establis1ed, implemented, and maintained covering the activities recommended in Appendix A of Regulatory Guide 1.33. Revision 2, February 1978, " Quality Assurance Program Requirements (Operations)." Appendix A of Regulatory Guide 1.33, Section 1, requires administrative equipment control.

Site Standard Practice (SSP) 12.53, Annunciator Disablement, Revision 5, requires that prior to the disablement of any annunciator the reason for the annunciator to be in the alarm condition shall be thoroughly evaluated.

Contrary to the above, on November 30, 1996, the Unit 1 Turbine Driven Auxiliary Feedwater sump high level annunciator was disabled without the reason for the alarm condition being thoroughly evaluated.

This is a Severity Level IV Violation (Supplement I), for Unit 1 only.

F. 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.

Contrary to the above, since September 18, 1990, the licensee had failed to implement promat corrective action to resolve a condition adverse to '

quality as descriaed in Problem Evaluation Report (PER) No.

SQP900372PER. The PER described reactor core design changes which had been implemented and had not been reconciled or reflected in Nuclear Engineering (NE), design basis documents, environmental qualification binders. As of April 1,1997, the EQ binders had not been revised.

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

G. 10 CFR 50, Appendix B, Criterion III, Design Control, requires in part that measures be established to assure that applicable regulatory requirements are correctly translated into drawings and procedures. The measures shall include provisions to assure that appropriate quality standards are specified and included in design documents. The design control measures shall also provide for verifying or checking the adequacy of design.

Tennessee Valley Authority Nuclear Quality Assurance Plan TVA NQA PLNB9-A. Revision 6. Section 7.0, Design Control, requires that measures be

NOV 4 established to ensure that the performance of design analysis shall be  :

planned and controlled. Additionally. it requires that measures to control plant configuration and ensure that the actual 31 ant  :

configuration is accurately depicted on drawings and otler appropriate  ;

design output documents and reconciled with the applicable design basis '

shall be established, documented, and implemented.

TVA NQA PLN89 A, through Section 7.0 and Appendix B, endorses the requirements of ANSI N45.2.11-1974, Quality Assurance Requirements for  :

the Design of Nuclear Power Plants. Section 4.0 of this standard requires that design analyses shall be performed in a planned, ,

controlled, and correct manner. Design analyses shall also be in a form suitable for reproduction, filing and retrieving.

Contrary to the above the established design control measures were i deficient'in that the following deficiencies were identified:

1. As of July 30,-1990, radiation dose values contained in design basis calculation TI RPS 48, Integrated Accident Dose Inside Primary Containment and Annulus, Revision 3, were never incorporated in calculation TI-ECS 55, Summary of Harsh Environment Conditions for Sequoyah Nuclear plant, to ensure revision of environmental data drawing series number 47E235.

Additionally, FSAR Figures 3.11.21. and 3.11.2 2 were never revised to reflect the new 100 day integrated accident doses based on a source term of 1000 EFPD. This failure to control plant configuration and ensure that actual plant configuration was accurately depicted on drawings resulted in discrepancies in design basis information listed in FSAR Table 15.1.7 1 and FSAR Figures 3.11.2-1 and 3.11.2-2.

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2. On December 12, 1991, TVA management approved design basis calculation TI-RPS 48, Revision 5, " Integrated Accident Dose Inside of Primary Containment and Annulus," to document the ,

100 day integrated beta and gamma radiation doses based on a i' source term of 650 EFPD. Radiation dose values contained in this calculation were not incorporated into calculation TI-ECS 55,

" Summary of Harsh Environment Conditions for Sequoyah Nuclear Plant. Additionally, plant modification DCN No. 508114A, Revision 16, revised environmental drawings number 1,2-47E235 sheets 45, 47, and 48 to realace radiation values that were no longer conservative. T1ese drawing revisions did not accurately depict actual plant configuration in that on the following dates listed the core average exposure for both units exceeded 650 EFPD  :

operation. t i

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NOV 5 Unit No. Cycle No. Date EFPD Exceeded 1 4 12-29 89 1 5 06 09 91 2 3 12 30 88

. 2 4 05 24 90 2 5 09 28 91 This failure to control plant configuration and ensure that actual plant configuration is accurately depicted on drawings resulted in  :

discrepancies between the units current licensing basis of 1000 EFPD burnup criterion and approved design basis information  ;

depicted on the environmental drawings.

3. From February 11, 1994, to November 15, 1996, the licensee failed to perform a calculation to determine the integrated maximum ,

hypothetical accident gamma and beta doses inside the primary containment to support a justification for continued operation for SQ PER 900372 PER.

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

Pursuant to the provisions of 10 CFR 2,201 Tennessee Valley Authority 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 II, 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 i 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 i include previously docketed correspondence, if the correspondence adequately i 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 I why such other action as may be proper should not be taken. Where good cause l is shown, consideration will be given to extending the response time. I Because your res>onse will be placed in the NRC Public Document Room (PDR), to the extent possi ale, it should not include any personal privacy, 3roprietary, or safeguards information so that it can be placed in the PDR wit 1out redaction. If personal privacy or proprietary information is necessary to provide an 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 meterial, you must s)ecifically identify the portions of your response that you seek to have with1 eld and provide in detail the bases 1

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NOV 6 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 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.

Dated at Atlanta, Georgia this 12th day of May 1997