ML17305B170

Responds to NRC Request for Info Re Basis for Operation W/ Steam Bypass Control Sys Outside Design Basis

ML17305B170
Person / Time
Site:	Palo Verde
Issue date:	10/24/1990
From:	Conway W ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
102-01875-WFC-T, 102-1875-WFC-T, NUDOCS 9011080037
Download: ML17305B170 (7)

Text

TION. SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9011080037 DOC.DATE: 90/10/24 NOTARIZED:

NO DOCKET FACIL:STN-50-528 Palo Verde Nuclear Station, Unit 1, Arizona Publi 05000528 STN-50-529 Palo Verde Nuclear Station, Unit 2, Arizona Publi 05000529 STN-50-530 Palo Verde Nuclear Station, Unit 3, Arizona Publi 05000530 AUTH.NAME AUTHOR AFFILIATION CONWAY,W.F.

Arizona Public.Service Co.

(formerly Arizona Nuclear Power RECIP.NAME

'ECIPIENT AFFILIATION

. MARTIN,J.B.

Region 5 (Post 820201)

I

SUBJECT:

Responds to NRC request for info re basis for operation steam bypass control sys outside design basis.

DISTRIBUTION CODE:

IE01D COPIES RECEIVED:LTR g ENCL~ SIZE:

TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response

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NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT'CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU'DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED:

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PHOENIX. ARIZONA85072-3999 WILLIAMF. CONWAY EXECUTIVEVICEPRESIDENT NUCLEAR 102-01875-WFC/TRB/GWS October 24, 1990 Mr. John B. Hartin Regional Administrator, Region V U.

ST Nuclear R'egulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596-5368

Reference:

Telephone call between A'PS and NRC on October 23, 1990 Subj ect:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2 and 3

Docket No.

STN 50-528/529/530 Request for Information on Steam Bypass Control System File 90-056-026 In a phone conversation with members of NRC Region V staff on Tuesday, October 23, 1990, it was requested that APS provide its basis for interim operation of the three PVNGS Units in the condition with the Steam Bypass Control System (SBCS) outside its design basis until a formal justification for continued operation can be finalized.

This condition had been reported to NRC on October 22, 1990.

This letter is provided in response to this request.

Included in this letter are:

I.

A.short description of the Unit 3 trip on October 20, 1990.

II.

A summary of the evaluation performed and the basis for the restart of Unit 3.

III.

A description of the design basis for the SBCS used in the Safety Analysis.

IV.

The basis for APS'ecision to continue operation of the SBCS with

.greater than one (1) valve in service.

EVENT

SUMMARY

PVNGS Unit 3 was operating normally at 1008 power on October 20,

1990, when numerous alarms were received in the Control Room and Steam Bypass Control System (SBCS) valves were observed to be modulating open.

The SBCS was 'taken to "emergency off" mode when it was determined that a SBCS actuation was not required.

The reactor tripped when the Core Protection Calculators (CPCs) generated a variable overpower auxiliary trip to the Plant Protection System (PPS) as expected for an overpower event.

All systems functioned as designed 4

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Mr. J.

B. Martin Page 2 of 6 102-01875-WFC/TRB/GWS October 24, 1990 to stabilize the plant in Mode 3.

The'vent was classified as an uncomplicated, reactor trip.

Engineering analysis and troubleshooting found that the initial alarms and spurious SBCS valve modulations were the result of a power distribution module failure in a balance of plant instrument cabinet.

The module failed in such a

way as to cause a spike in the steam header pressure instruments.

The spike resulted in a SBCS modulation demand which was terminated by placing the system in '"emergency off".

APS investigated the Unit 3 trip in accordance with the requirements of the Incident Investigation Program.

An event classified as an uncomplicated reactor trip is investigated with "the Plant Manager assuming the position of the Investigation Director..

The process used in the investigation is:

Assembl'e the Investigation Team.

Collect plant data, Operators statements, and'ther relevant information.

Assign actions into functional areas including:

Nuclear Safety Assessment Control System Response Personnel Performance Plant Protection System Response Overall Plant Performance Identify restart requirements and identify and assign Post Restart Actions.

Complete the restart portion of the Incident Investigation.

Restart Complete the Incident Investigation No significant anomalies were noted in the control system or overall plant response to the event which adversely affected plant oper'ation.

Several minor equipment actuation and indication problems were identified as well as one document control issue, but these were determined to not adversely affect plant operation and were not considered restart items.

UNIT 3 RESTART AUTHORIZATION A meeting of the Incident.Investigation Team comprised of the Plant Manager; the Director of Site Technical Support; the STA Supervisor; the System Engineering I&C Supervisor; Unit Operations, Work Control and Maintenance

Managers, and additional Site Technical Support staff was held on Sunday, October 21, 1990, to review the resul'ts of the troubleshooting action plan and discuss any other potential restart issues.

A Nuclear Safety Assessment had been performed and was reviewed by the Investigation Team.

Mr. J.

B. Martin Page 3 of 6 102-01875-WFC/TRB/GWS October 24, 1990 The analysis of Core Protection Calculator (CPC) trip buffers showed that significant margin to Specified Acceptable Fuel Design Limits (SAFDLs) existed throughout the event.

This demonstrated, that the CPCs were effective in terminating the event, prior to reaching fuel design limits.

An evaluation of the initiating event against the FSAR design basis accidents was performed and concluded the event was bounded by,the Main Steam Line Break acc'ident analysis.

A previous occurrence where opening of multiple SBCVs occurred was discussed by the Incident Investigation Team.

In the previous event the SBCS functioned as designed.

All SBCVs received an open signal due to a fault external to the'SBCS.. It was determined in the review of the previous event that the occurrence was.bounded by a Main Steam Line Break.

In addition, evaluation of the, SBCS during the Unit 3 trip concluded that the control system operated as expected in response

'to the external power distribution module failure, therefore, it was concluded the SBCS met its design requirements.

No nuclear safety concerns were identified.

The team was concerned however,. that, in spite of the determination that the SBCS system operated as expected,

'a failure at the interface of the system had caused multiple valve openings.

That situation, although bounded by the accident analysis, was considered to be. a significant reliability concern

.which required further engineering review post restart.

The power distribution module was replaced..

Based on the Incident Investigation Team review, Uni't restart was authorized, by the Plant, Manager in accordance with station admini'strative -control procedures.

The Unit entered Mode 2 at 1632 on October 21,

1990, and was put back on line at 0313 on October 22, 1990.

POST RESTART EVALUATION On Monday, October 22, 1990, follow-up review raised a concern regarding the design basis for the SBCS that was used in the performance of the FSAR Chapter 15 analysis.

The currently analyzed event for an "Inadvertent Opening of a Steam, Generator Relief or Safety Valve" (FSAR Section 15.1.4) assumed that only one Atmospheric Dump Valve or Steam Bypass Control Valve (SBCV) would open due to a failure of the control system. It was not.known if the Safety, Analysis had considered the opening of all eight SBCV's due to a single failure as an Anticipated Operational Occurrence (AOO).

To address this question, internal APS Engineering organizations and Combustion Engineering (CE) were contacted to research the original design requirements for the SBCS and determine how these were incorporated into the Safety Analysis.

An effort was also started to perform an analysis that would consider the opening of seven, (7)

SBCVs.which is consistent with normal unit operating configuration.

A single case was chosen from full power at end of cycle initial conditions.

Thermal hydraulic conditions were chosen so that the event was initiated at a Power Operating Limit with minimum margin to the DNBR SAFDL.

Information from APS engineering departments indicated by late

Mr, J.

B. Martin Page 4 of 6 102-01875-WFC/TRB/GWS October 24, 1990 Monday that the PVNGS units were in an unanalyzed condition.

Preliminary informati'on from the analysis in progress indicated that margin would exist to the SAFDLs following an event of this kind.

Subsequent CE review of this analysis confirmed this conclusion on October 23, 1990.

Combustion Engineering confirmed on the morning of October 23, 1990, that the opening, of more than one SBCV was not analyzed in the current Safety'nalysis and that the design criteria 'for the SBCS was that no single failure would result in more than one valve opening.

The current, analysis basis for PVNGS 'identifies the most limiting event for an Increase in Heat Removal by the Secondary System (FSAR Section 15.1) to be.a Main Steam Line Break, which results in limited fuel damage and acceptable offsite dose.

Opening of all eight SBCVs is bounded by the consequences of this event.

The AOO analyzed in the FSAR Section 15.1. is an inadvertent opening of one steam bypass control or atmospheric dump valve.

The basis for this was. the assumption that,no single failure would result in more than one valve opening.

The Unit 3 trip demonstrated that a single failure, outside the

SBCS, can result in all in service valves opening.

Preliminary information from the analysis described above indicates that no fuel damage or offsite dose would result.,

EVALUATION OF CONTINUED OPERATION The Plant Review Board met on October 23, 1990, to review the continued operation of the Units considering the potential for multiple steam bypass control valves opening due to a single fail'ure.

Based on the preliminary information available, the PRB concluded that an unreviewed safety question does not exist and that continued'peration of the Units was justified on an interim basis.

The basis for that conclusion follows:

The event that occurred on Unit 3 on October 20,

1990, had been

'evaluated as a result of the post trip review and the results indicated that a significant margin to fuel design limits (DNBR and Local Power Density) existed during the event.

This indicated that the CPC's provide adequate protection for this event for the conditions existing at Unit 3.

2.

As evaluated in the post trip review, an event resulting, in the opening of multiple steam bypass valves is bounded by the Main Steam Line Break accident analysis.

Thus the consequences are acceptable for protection of the health and safety of the public.

3.

Preliminary information from an analysis currently in,progress which considers the worst case operating parameters for the current fuel cycle of all Units while operating at 100% power, indicates, there is adequate margin to the SAFDLs for the case of

Mr. J.

B. 'Martin Page 5 of 6 102-01875-'WFC/TRB/GWS October 24, 1990 seven (7)

SBCVs openi'ng spuriously.

This calculation is currently under independent review (Note:

The revi'ew has subsequent'ly been completed.as stated on page 4).

The Unit 3 event on October 20, 1990, resulted in a significant margin to fuel design limits, providing confidence that the calculation is accurate.

Further, confidence that the health and safety of the public would not be in jeopardy exists since the event is bounded by the Main Steam Line Break accident analysi's, should the calculation results change.

4.

Station management has taken compensatory measures to provide additional mitigation of the effects of a spurious SBCS actuation.

On October 22,

1990, a dedicated operator was stationed at the SBCS panel in.each units'ontrol 'Room with instructions to monitor SBCS operation and take action to immediately close SBCVs that open spuriously.

This provides additional confidence that the consequences of,spurious opening of multiple SBCVs will be minimized and the results will be acceptable.

Subsequent to the above evaluation, CE completed a technical review of the APS analysis of the inadvertent opening of seven SBCVs.

APS has also completed an analysis of the inadvertent opening of eight SBCVs which showed acceptable results in that adequate margin, to exceeding SAFDLs exists.

CE has conducted a technical review of this analysis which substantiates

.the APS results.

CE and APS are performing a 'QA'eview of the inadvertent opening of eight SBCVs which is expected to be complete on October 25, 1990.

The cases that have been analyzed were evaluated at 100% power and APS recognizes that certain lower power cases could be more limiting.

As an interim measure, a

5% power penalty will'be installed in the Core Operating Limit Supervisory System (COLSS) by October 25, 1990.

APS and CE have determined this penalty will ensure sufficient margin exists for these low power cases.

Evaluation of the applicable reduced power cases will be completed in an expeditious manner.

APS has determined that an evaluation of the potential impact of coincident Chapter 15 analyzed events in conjunction with inadvertent opening of multiple SBCVs should be performed.

This evaluation will be completed'ithin the next 30 days.

The probability of experiencing a Chapter 15 event in conjunction with the inadvertent opening of multiple SBCVs is judged ~to be sufficiently small during the evaluati'on period to justify continued operation.

.Should a problem be identi'fied as a result of these

analyses, APS will place the units in an analyzed condition within the current SBCS design basis.

This condition involves the isolation of seven SBCVs, leaving one valve per unit in service.

Operation in thi's m'ode has been analyzed in the Chapter 15 Sa'fety Analysis with acceptable results and APS has procedures available which can be put into effect promptly to implement this mode of operation.

Mr. J.

B. Martin Page 6 of 6 102-01875-WFC/TRB/GWS October 24, 1990 Although operation with one SBCV in service is acceptable, APS prefers not to operate in this manner unless necessary as it 'increases the potential severity of plant transients that could'ccur.

APS has concluded that the above evaluation and compensatory actions make continued operation with seven SBCVs in service acceptable during the period the identified analyses are being compIeted.

Should further analysis disprove this conclusion, APS will promptly implement the single SBCV.mode of operation to place the Units in an analyzed condition.

Very truly yours, WFC/TRB/GWS/dmn cc:

Document Control Desk S.

R. Peterson C.

M. Trammell D. H.

Coe A. H. Gutterman, A.

C.

Gehr 4n