ML18038B745

Forwards Response to NRC Re Violations Noted in Insp Repts 50-259/96-05,50-260/96-05 & 50-296/96-05. Corrective Actions:Rcic Was Returned to Fully Operable Status by Implementing Design Change

ML18038B745
Person / Time
Site:	Browns Ferry
Issue date:	08/30/1996
From:	Machon R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9609050186
Download: ML18038B745 (28)

Text

CATEGORY 1 REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR,:9609050186 DOC.DATE: 96/08/30 NOTARIZED: NO DOCKET FACIL:50-259 Browns Ferry Nuclear Power Station, Unit 1, Tennessee 05000259

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50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH.NAME AUTHOR AFFILIATION MACHON,R.D.

Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Forwards response to NRC 960801 ltr re violations noted in insp repts 50-259/96-05,50-260/96-05 s 50-296/96-05.

Corrective actions:RCZC was returned to fully operable status by implementing design change.

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Tennessee vattey Auttcnty. post office Box 2000. Decatu>>. Ataxia 35K9 2cc0 R. D. (Rick) Machon Vce Presxfent, Browns Ferry Nuctear Pant August 30, 1996 U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.C.

20555 10 CFR 2

Appendix C

Gentlemen:

In the Matter of

)

Tennessee Valley Authority

)

Docket Nos.

50-259 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN)

NRC INSPECTION REPORT 50-259'~

50-260~

50-296/96-05 REPLY TO NOTICE OF VIOLATION (NOV)

On June 19,

1996, NRC issued Inspection Report 96-05 which cited one violation and identified two apparent violations.

The cited violation, involving control of overtime, was answered in a letter from R.

D. Machon to NRC, dated July 19, 1996.

The two apparent violations, one involving the failure of the BFN Unit 2 Reactor Core Isolation Cooling system to operate in response to a May 10, 1996, reactor scram and the

other, a procedural noncompliance related to American Society of Mechanical EngineersSection XI testing were issued as a

NOV on August 1, 1996.

This letter is in response to the August 1,

1996, NOV.

On July 11,

1996, TVA met with NRC to discuss the circumstances surrounding these two violations.

TVA presented the results of the root cause analysis and a

summary of the corrective actions that had been taken or which were in progress.

As discussed at that meeting, TVA recognizes the need to further improve the quality of engineering products and to ensure required post-modification testing is properly identified and completed.

TVA believes the corrective actions presented at that meeting and reiterated in this response are comprehensive and fully t

address the problems.

.-.~, 8 3 'tf609050i86 960830 PDR ADQCK 05000259 8

PDR

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U.S. Nuclear Regulatory Commission

.Page 2

August 30, 1996 The enclosure provides TVA's written response to the individual violations cited in the August 1, 1996, letter.

Enclosure 2 contains a summary of commitments made in this response.

If you have any questions regarding this reply, please contact Tim Abney at (205) 729-2636.

Sincerely, R. 'D.

hon Enclosures cc (Enclosures):

Mr. Mark S. Lesser, Branch Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road

Athens, Alabama 35611 Mr. J.

F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Paul E. Fredrickson, Chief Special Inspection Branch Division of Reactor Safety P.. 0.

Box 2257 Atlanta, Georgia 30323

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ENCLOSURE 1

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1 I 2 I AND 3 INSPECTION REPORT NUMBER 50-259, 50-260, 50-296/96-05 REPLY TO NOTICE OF VIOLATION (NOV)

RESTATEMENT OF VIOLATION A "Technical Specification 3.5.F.1 requires, in part, that the reactor core isolation cooling (RCIC) system be operable whenever there is irradiated fuel in the reactor vessel and the reactor vessel pressure is above 150 pounds per square inch gauge (psig).

If the RCIC system is inoperable, the reactor may remain in operation for a period not to exceed seven days if the high pressure cooling injection system is operable during such time.

Contrary to the above, from April 23 to May 10,

1996, the RCIC was inoperable with irradiated fuel in the reactor vessel and reactor vessel pressure greater 150 pounds per square inch gauge.

The reactor remained in operation during this period, which exceeded seven days.

(01013)

This is a Severity Level III violation (Supplement I)."

TVA'S REPLY TO VIOLATION A 1.

Reason For The Violation This violation was caused by personnel error.

Engineers responsible for a design change that replaced the Reactor Core Isolation Cooling (RCIC) system turbine exhaust check valve with a different type of check valve did not properly analyze the effects of the new valve on RCIC system operation.

Specifically, design engineers did not verify the adequacy of calculational inputs with actual plant operating data in preparing the design change package.

Also, the individuals did not demonstrate the expected level of critical thinking in evaluating the change for overall effects on RCIC operation.

During the Unit 2 refueling outage in April 1996, the RCIC turbine exhaust check valve was replaced with a lift globe check valve to improve reliability and leak tightness.

This type of replacement valve had been used at several other nuclear plants in the same application over the past several years with good results.

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Design engineers performed system calculations using the higher flow loss coefficient for the new valve to determine the impact on RCIC system operation.

The RCIC steam flow used in the calculation was 28,000 pounds per hour and was obtained from a vendor process drawing for RCIC steady state rated injection.

This value is less than the actual RCIC steam flow in the field (approximately 38,000 pounds per hour) at rated vessel injection.

Also, during RCIC autostarts, steam use peaks at approximately 50, 000 pounds per hour for a short period of time while the RCIC turbine accelerates to overcome reactor pressure for vessel injection.

Had this transient maximum steam flow been

used, the calculations would have shown that the steam pressure at the turbine exhaust would exceed the existing RCIC turbine high exhaust pressure trip setpoint of 25 pounds per square inch gauge (psig).

This result would have mandated a

reevaluation of the proposed design change to ensure RCIC reliability.

The designers, using the process drawing steam flow, erroneously concluded the design change had a minor impact on RCIC system operation.

This same information led the design engineers to determine that it was not necessary to perform a unique post-modification test.

A detailed review and root cause analysis of the design change were performed.

It was concluded that the responsible design engineers did not properly verify the adequacy of inputs to the calculations for the RCIC check valve replacement design change.

Although there was evidence that the change in operating margin to the trip setpoint was considered, it was clear that the issue was not fully evaluated against actual flow data.

Several contributing factors were identified relating to shortcomings in designer knowledge of plant operations, documentation of the bas'is for design considerations, and design verification practices.

Also, it appeared that the knowledge that the replacement valve had been successfully used at other plants may have influenced the designers into a false sense of confidence that the new valve could be used at BFN without making any other system modifications.

The post event review concluded RCIC would have been available for manual operation, if needed.

This mode of RCIC system operation is proceduralized.

In manual operation, the short term steam flow transient is avoided since the system is brought into service more slowly so peak turbine exhaust steam pressure is lower and remains below the trip setpoint.

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Corrective Actions 'Taken And Results Achieved Following the RCIC trip on May 10,

1996, RCIC was returned to a fully operable status by implementing a design change which raised the RCIC exhaust pressure trip setpoint from 25 psig to 50 psig.

This change ensures there is a large margin from steam pressures experienced during normal and transient RCIC system operation to the turbine high exhaust pressure trip setpoint.

This setpoint change was also implemented on Unit 3 although the new type check valve is not yet installed.

In addition to other near term actions, Engineering personnel involved in the RCIC design change package have been counseled on the need for rigorous adherence to the design change process procedures.

TVA also performed an extent of condition review to determine whether the causes and contributing factors associated with the RCIC design package were isolated.

Thirty design, packages were reviewed in detail using the lessons derived from the cause analysis of the RCIC design package problem.

This review concluded that the RCIC

problem, issuance of a design package with an adverse impact on plant safety, was isolated.

This review did conclude,

however, that some of the contributing factors were present in other design packages and appropriate corrective actions and enhancements were developed and implemented.

A summary of the cause

analysis, extent of condition review, and major corrective actions and improvements is presented in Table 1.

A Problem Evaluation Report (PER)

BFPER960621 is also tracking completion of additional minor corrective actions.

Corrective Ste s That

[Have Been Or],Will Be Taken To Avoid Purt er Vxo atxons The corrective actions described in Table 1 have been completed.

No further corrective actions are considered necessary to ensure full compliance.

The Engineering Review Board, discussed in corrective action 10 of Table 1, is considered a temporary measure and will remain in place pending a determination by Engineering management that expectations on quality for new design changes are being consistently met.

Nuclear Assurance will conduct an independent assessment of the effectiveness of the corrective actions within six months of this submittal.

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Date When Pull Co liance Will Be Achieved TVA believes that it is in full compliance with respect to the issues identified in Violation A.

RESTATEMENT OF VIOLATION B "Technical Specification 1.0.MM.1 requires that In-Service Testing of American Society. of Mechanical Engineers (ASME)

Code Class 1,

2, and 3 valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(f).

IWV-3200 of Section XI of the ASME Boiler and Pressure Vessel Code requires that when a valve has been replaced or repaired or has undergone maintenance that could affect its performance, and prior to the time it is returned to service, it shall be tested to demonstrate that the performance parameters, which could be affected by the replacement,

repair, or maintenance, are within acceptable limits.

Site Standard Practice-8.6, ASME Section XI In-Service Testing of Pumps and Valves, Revision 12, implements the requirements of IWV-3200 of Section XI of the ASME Boiler and Pressure Vessel Code (In-Service Testing Program).

Appendix H of Site Standard Practice-8.6 requires Procedure 2-SI-4.5.F.l.d, Revision 25, to be performed following maintenance on the Reactor Core Isolation Cooling System Turbine Exhaust Check Valve, 2-CKV-71-0580 and Procedure 2-SI-4.5.E.l.d, Revision 34, or 2-SI-4.5.E.l.d(dp),

Revision 4, to be performed following maintenance on the High Pressure Coolant Injection System Turbine Exhaust Check Valve, 2-CKV-73-0603.

Contrary to the above:

(1)

On April 23,1996, the RCIC system turbine exhaust Check Valve-2-CKV-71-0580 was returned to service after having undergone maintenance (replacement),

without Procedure 2-SI-4.5.F.l.d being performed on the valve.

(2)

On April 23,1996, the High Pressure Coolant Injection System Turbine Exhaust Check Valve, 2-CKV-73-0603 was returned to service after having undergone maintenance (replacement),

without Procedure 2-SI-4.5.E.1.d or 2-SI-4.5.E.l.d(dp) being performed on the valve.

(02014)

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

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TVA'S REPLY TO VIOLATION B 1.

Reason For The Violation This violation was caused by personnel error.

Engineers responsible for specifying the appropriate American Society of Mechanical Engineers (ASME)Section XI tests following replacement of RCIC and High Pressure Coolant Injection (HPCI) system check valves did not properly ensure required testing was identified and performed.

In November

1995, design change packages were prepared for the replacement of the RCIC and HPCI turbine exhaust check valves.

The system engineers responsible for identifying required post-modification component and program-based tests (including tests required by the site ASME Section XI procedure, Site Standard Practice (SSP)-8.6) reviewed the design packages in accordance with the standard design change package review process.

The RCIC system engineer correctly specified the rated pressure RCIC surveillance test (2-.SI-4.5.F.l.d) required by SSP-8.6 for ASME Section XI testing.

However, the HPCI system engineer specified the 150 psig surveillance test (2-SI-4.5.E.l.e) rather than the rated pressure surveillance test (2-SI-4.5.E.l.d) required by SSP-8.6.

Following the completion of the design changes in April

1996, the 150 psig surveillance tests were successfully completed for HPCI and RCIC.

These surveillance tests also satisfy Technical Specification requirements for startup from a cold condition (in this case, from the refueling outage).

The rated pressure tests were within their quarterly periodicity although they had been previously entered on the plant schedule as part of the standard post refueling outage startup sequence.

Since the rated pressure surveillances were within their periodicity, Scheduling personnel requested that the system engineers evaluate whether the rated pressure tests could be postponed until the next periodic test interval.

The HPCI system engineer reviewed the situation and approved postponement since the 150 psig test he had previously specified was complete.

The RCIC system engineer was queried on the sam0 issue on the following shift and likewise approved postponement of the RCIC rated pressure test.

During the root cause investigation, the RCIC system engineer indicated that his decision was based on the satisfactory performance of the Technical Specification required 150 psig test and that he had overlooked the association with the ASME Section XI test requirement due to the valve modification.

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The root cause analysis determined the system engineers had not been diligent in exercising their responsibilities in identifying and ensuring required testing was performed in accordance with site procedures.

Contributing factors were also identified and are discussed in greater detail below.

Corrective Actions Taken And Results Achieved TVA had successfully completed the required ASME Section XI testing prior to,NRC identification of the procedural noncompliance.

The RCIC rated pressure test was completed during testing following the turbine high exhaust pressure setpoint change discussed in Violation A.

The HPCI rated pressure test had been performed on May 19, 1996, at its next regularly scheduled surveillance test interval.

Also, the involved system engineers have been counseled on their responsibilities on handling of tests and on the conduct of decisions involving ASME Section XI requirements.

TVA performed an extent of condition review to determine if the cause and contributing factors associated with the Section XI violation were isolated.

In all, 115 design change packages were reviewed for proper inclusion of program-based testing such as ASME Section XI, 10 CFR 50 Appendix J, and Generic Letter 89-10 testing (motor operated valve testing).

Additional cases were found where Section XI external leakage checks associated with other Unit 2 HPCI and RCIC refueling outage modifications had been deferred.

Completion of these external leakage checks was linked to the rated pressure tests in the plant schedule and deferred when the rated pressure tests were postponed by the system engineers.

These tests had already been completed when TVA identified the problem; therefore, no additional testing was required.

The root cause analysis identified and the extent of condition review confirmed that system engineers needed additional training on Section XI program requirements.

Also, the analysis recommended improvements be made in the methods being used to earmark program-based test requirements during the review of design change packages and on plant work activity schedules.

A summary of the causes, contributing factors, corrective

actions, and improvement actions are shown in Table 2.

PERs BFPER960710 and BFPER960790 are also tracking minor related corrective actions.

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Corrective Ste s That

[Have Been Or] Will Be Taken To Avoid Furt er Vxo ations A training module on In-Service Test and Inspection Program has been prepared and will be provided to the System Engineers by October 4,

1996.

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Date When Full Co liance Will Be Achieved TVA believes that it is in. full compliance with respect to the issues identified in Violation B.

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ENCLOSURE 2

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 INSPECTION REPORT NUMBER 50-259, 50-260, 50-296/96-05 REPLY TO NOTICE OF VIOLATION (NOV)

COMMITMENTS 1'.

Nuclear Assurance will conduct an independent assessment of the effectiveness of the corrective, actions within six months of this submittal.

2.

Training on the In-Service Test and Inspection Program will be provided to the System Engineers by October 4,

1996.

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tTable I - Summary of Causes an orrective Actions For Violation A (Continued)

Issue Designers Concluded Design Change Had MinorImpact on System Operation (Conf.)

Root Cause and Contributin Causes Contributin Causes:

Cont.

Designers aware other utilities had used new valve, but not aware others had increased RCIC setpoint based on 1982 General Electic SIL 371 related to a different issue.

The SIL had been previously reviewed by TVAand determined that it was not required to be implemented.

Verifier/Checker Same Individual. Extent of Condition review observed same practice for other design packages.

Corrective Actions 7.

Implemented SIL setpoints (60 psig) on Units 2 and

3. Also, reemphasized need to utilize experience review information as part of design process during RCIC design package training. (Corrective Action 3) 8.

Issued policy memorandum requiring design checker and verifier be different people per Engineering Manager expectations.'.

PER process to be utilized to document and trend CheckerNerifter significant findings.'tatus Complete Complete Complete Root Cause:

10. Established an Engineering Review Board to independently review design changes, nonconformances, and corrective actions lans.*

Complete Designers Failed fo Understand Unique Test Needed to Assure Function

- Decision not to specify unique postmodlftcation test result of previous deficient calculation.

11. Corrective actions 1-7.

Complete Contributin Causes:

Designers did not exhibit consistent ownership of postmodiftcation test responsibilities.

12. Revised SSPW.3 to reinforce designers responsibilities to define unique post~odlftcation tests that ensure special design requirements are met.'omplete These items not considered regutstory commitments.

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Table 2 - Summary of Causes ana orrective Actions For Violation B Issue Root Cause and Contributin Causes Root Cause Corrective Actions Status ASME Section XITests Not Properly Identifed and Controlled Personnel Error

- System Engineer did not properly identify test during impact review asSection XI per SSP-8.6.

Also, ASME Section XItests inappropriately deferred.

1. Counseled involved employees on their responsibilities on handling oftests.

Complete Contributin Causes:

Design package not reviewed by ASME Section XI Program Owner.

2. SSP-9.3 revised to require mandatory impact review by ASME Section XI Program Owner and approval oftesting revisions.'omplete Weak System Engineer knowledge Of ASME Section XI program.

3. Training module prepared to improve System Engineer knowledge in areas of Inervice Test and Inspection Program requirements.

Training to be provided to System Engineers.

1014/96 Design package impact review form does not require earmarking tests as ASME "Section XI "tests.

Plant work activity printed schedules do not routinely highlight ties to testing basis.

4. Revised impact form to list basis for test.'.

Scheduling group to list test basis from impact reviews on plant schedules.'

These items not considered r uht commitments.

Complete Complete

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