ML17305A321

Procedure 43OP-3ZZ16,RCS Drain Operations,Not Appropriate for Circumstances

ML17305A321
Person / Time
Site:	Palo Verde
Issue date:	10/05/1989
From:	Haplin M ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:
Shared Package
ML17305A320	List: ML17305A319 ML17305A321
References
3-3-89-019, 3-3-89-19, NUDOCS 8910270018
Download: ML17305A321 (103)

Text

e PVNGS INCIDENTINVESTIGATIONPROGRAM COVER SHEET INCIDENTINVESTIGATIONRI PORT 1K98BER:

3 89 - 019

~~. Procedure 43OP-SZZ16, RCS Drain Operations, Not Appropriate for Circumstances EVENTDATE: March 11 1989 REPORT APPROVALDATE:

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ANPP INCIDENTINVESTIGATIONPROGRAM 79DP-OOPol, INCIDENTINVESTIGATIONREPORT PREPARATION - Appendix A-1 89102700i8 891005 PDR ADOCK 05000528 0

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PVNGS INCIDENTINVESTIGATIONPROGRAM REVIEW4 APPROVAL SHEET INCIDENTKGHTIGATIDMRIVORTRJMBER:

3 89 - 019 Procedure 43OP-3ZZ16, RCS Drain Operations, Not Appropriate for Circumstances

~~ed ~ M.R, Halpin Reviewed Bp Lead In tor Revimmd Bp Reamend Bp Ihwiewed Bp Reviewed Bp D te ro,y J'+(

Nrected Plant M er WH'KTDATE: March 11 1989 REPORT APIROVALDATE:

ANPP INCIDENTINVESTIGATIONPROGRAM 79DP-OOP01, INCIDENTINVESTIGATIONREPORT PREPARATION - Appendix A-3

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PVNGS INCIDENTINVESTIGATIONPROGRAM CONCURRANCE SHEET The Rllawing signatures indicate concuI~nce with the amgned

~tion. ~r do not indicate a review ofthe completeness or ofthe immmtigation process or the report.

Itexn: I 3 - Director Standax'ds &Technical Su ort R

nsible Manager Item: 2 - Director En ineerin

& Construction g/Zs Concurs ce:

Item:

Con cuI~nce:

Item:

ConcuImence:

Itexn:

Responsible Manager Item:

ce:

Responsible Manager Conaxrrence:

Responsible Manager ANPP INCIDENT1AVESTIGATIONPROGRAM 79DP-OOP01, INCIDENTINVESTIGATIONREPORT PREPARATION - Appendix A4

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PVNGS INCIDENTINVESTIGATIONPROGRAM AMMEMBERS SHEET RICIDE5FI 93VESTIGATIONREPORT NUMBER: 3 89 - 019

~~ Procedure 43OP-3ZZ16, RCS Drain Operations Not Appropriate for Circumstances INCIDENTIN'iKSTIGATIONTEAMMEMBERS Team Leader: M.R. HALPIN Print IIame-0 erationsStrandards S-~~

Date Sign tur

- Department Team Member-,

Print IIame-Team Member:

Print Ilame-N/A Signature

- Department N/A Signature

- Department Date Date Team Member:

Print Ilame-Signature

- Department Date Team Member:

Print Iiame-Team Member:

Print Name-Team Member:

Print IIame-N/A Signature

- Department N/A Signature

- Department N/A Signature

- Department Date Date Date EVEREST DMX: March 11 19S9 REPORT APPROVALDATE:

ANPP INCIDENTINVESI'IGATIONPROGRAM 79DP-OIP01, INCIDENTINVESTIGATIONREPORT PREPARATION - Appendix A-5

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~ I PVNGS INCIDENTINVESTIGATIONPROGRAM CHECKLISTSHEET INCIDENTINVESTIGATIONREPORT CHECK IST PART I H

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H-Kxxxxtive'.ha>mary Event Description Facts Sheet ld ban

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M tDKM le Orgmzuation & Individual M taMS Due Dates for all CozTectim Actions Menti6ed PART II Q

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gg PART III H-Categozy 1 &2 Events only.

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Plant Protection System Response Control System Evaluation Gx~ Page Review and Approval Page(s)

C Rg AIIDepaztments with Corrective Actions speci6ed are included on the Concurrence page.

Appropriate charts (EST or EBKP) are inchu&d.

Concern Sza~zy (Ifmultiple conceI~).

Index ofAttachments included AllAttacIhments numbered and martini Appendix Acompleted ANPP INCIDENTINVESTIGATIONPROGRAM 79DP-OOPOI, INCIDENTINVESTIGATIONREPORT PREPARATION - Appendix C

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EXECUTIVE

SUMMARY

On March 11, 1989 procedure 43OP-3ZZ16, RCS Drain Operations, was implemented to direct reduced inventory operations, including "mid-loop" evolutions, during the Unit 3 first refueling outage.

Then, during an NRC inspection, conducted from March 20 through April26, 1989, it was determined that the operating procedure 43OP-3ZZ16 was not appropriate in that the RCS temporary level versus shutdown cooling flowcorrection curve was incorrect which resulted in a Severity Level IVViolation. This resulted in a discrepancy between the two temporary level indicators and the pressurizer level indicator during the RCS drain operation.

The procedure writer incorporated the wrong data into the procedure through an oversight. The procedure writer did realize that the data received in an Engineering Action Request (EAR) was not for the same point in the system that a temporary level indicator was to be installed. This lead to a larger difference in a level error due to shutdown cooling flow than anticipated.

The corrective action is to counsel the procedure writer on attention to detail.

DETAILS On March 11, 1989 procedure 43OP-3ZZ16, RCS Drain Operations, was implemented to direct the operation of the unit during reduced inventory conditions including "mid-loop" evolutions.

This procedure was generated as a result ofNRC Generic Letters 87-12 and 88-17, Loss ofDecay Heat Removal While in a Partially Drained Condition. and the ANPP Responses to NRC Generic Letters, dated September 21, 1987 (87-12) and January 6, 1989 (88-17). The new procedure was utilized for the first time during the Unit 3 firstrefueling outage which started in March, 1989.

During an NRC inspection, conducted from March 20 through April26, 1989, it was determined that operating procedure 43OP-3ZZ16, RCS Drain Operations, was not effective in providing guidance to control RCS inventory during reduced inventory conditions. It was determined that the procedure was not appropriate in that the RCS temporary level versus shutdown cooling flow correction curve was incorrect. A Severity Level IVViolation (Supplement I) was received.

The procedure 43OP-3ZZ16, RCS Drain Operations, utilized curves to correct RCS level indication for shutdown cooling flow to determine the actual level in the reactor vessel during reduced inventory evolutions. To obtain this information a verbal request for an Engineering Action Request (EAR 88-1671) was made on March 13, 1989 though the Nuclear Engineering Department (NED), in accordance with procedure 81DP-4EE03, Task Control Within Nuclear Engineering, for a curve for the Train B temporary tygon level indicator. The request was for the Train B temporary tygon level indicator only because a Train A level compensation curve already existed in 43OP-3ZZ06, Mode 5 Operations, and 43OP-3ZZ12, Mode 6 Operations, which had provided sufficient information for previous partial drain evolutions NED answered EAR 88-1671 December 27, 1988 with a letter to J.T. Pollard dated December 23, 1988 stating that the data collected in Unit 2 for RCS level differences due to shutdown cooling fiow was applicable to all three units due to similar configurations. The Unit 2 data forTrain A and Train B level compensation curves were included in the EAR answer.

The level compensation curve for the Train B level indicator was based on data collected from drain valve SIB-V057, which was the point the temporary refueling level indicator and the permanent Refueling Water Level Indicating System modification was to be installed. The level compensation curve for the Train A level indicator was based on data collected from drain valve SIA-V056, which was the point the permanent Refueling Water Level Indicating System modification was to be installed. These curves. were incorporated into the 43OP-3ZZ16 procedure as provided in EAR 88-1671.

The existing plant design identified RCE-V214 on the Train A shutdown cooling loop as the refueling level indication connection. This connection is used for the temporary tygon level indicator which has been used to provide RCS partial drained and mid-loop level indication in the past.

When the procedure writer incorporated the level compensation curves into the procedure 43OP-3ZZ16, RCS Drain Operations, he did not realize that the Train A level compensation curve was designed for the level indicating system to be connected at SIA-V056 instead ofRCE-V214. This point was stated in the Unit 2 Shutdown Cooling (SDC) Flow Data letter but not identified on the curve attached to the letter. The procedure writer was not aware that there would be an impact on the level compensation from the difference in location between SIA-V056 and RCE-V214.

DETAILS (con't)

The procedure 43OP-3ZZ16, RCS Drain Operations, went for cross-discipline review which included the Engineering Evaluations Department (EED) System Engineer but did not include the NED Design Engineer who wrote the EAR. Per 01AC-OAP02, Review and Approval of Nuclear Administrative and Technical Procedures, a cross-discipline review is required when "more than one section has a major role in the performance of the task described by the procedure" or "an intent change is made to a system operating procedure".

The cross-discipline review "should occur when more than one section has established expertise in the area covered by the procedure and the Technical Reviewer determines the need for a confirming opinion". No comments were received concerning the flowcompensation curves on the EED cross-discipline reviews. The difference between the location of the temporary level indication, at RCE-V214, and the point the level compensation curve data was collected at SIA-V056 was also missed by the Technical Reviewer.

Per 01AC-OAP02, Review and Approval of Nuclear Administrative and Technical Procedures the Technical Reviewer "conducts a detailed technical review to ensure that the procedure: (1) accomplishes its purpose; (2) has valid acceptance criteria; (3) has clearly defined responsibilities; (4) is consistant with applicable licensing and regulatory documents, other higher tier documents, and applicable technical requirements".

The procedure was approved and implemented with the discrepancy in place.

During the evolution oflowering RCS level, the procedurally required level cross checks between level indications did not meet the "within+/-six inches" criteria between the Train B and the Train A temporary tygon level indicators. However, both indicators were within +/-six inches of the pressurizer level indication. The Train B level indication was lower than the pressurizer level indication which was expected due to the lag ofdraining the pressurizer through the surge line to the RCS and venting the pressurizer through a one inch vent line. The Train A level indication was higher than the pressurizer level indication due to the compensation error using the incorrect curve.

The drain down was stopped and the EED System Engineer was contacted at home. He recommended continuing the drain down using the temporary level indicator that did not use compensation (i.e.; the tygon level indicator on the non-operating shutdown cooling loop) since only one temporary level indicator was required until the RCS level was at the 111ft elevation. The drain down continued to'the 113ft 6in elevation.

When the EED System Engineer arrived at the site, troubleshooting was conducted to identify the problem with the temporary level indication. The troubleshooting involved switching operating trains of shutdown cooling, stopping all shutdown cooling flow and letting level stabilize (at that time both levels stabilized within 1/2 inch ofeach other). At that time the EED System Engineer determined that the Train A Level Compensation Curve for Shutdown Cooling Flow was incorrect and the original curve used in the 43OP-3ZZ12, Mode 6 Operations procedure was correct. The original curve was incorporated into the RCS Drain Operations procedure and the incorrect curve was removed through the use of a temporary procedure change (TPCN). The rest of the RCS drain down evolution was then continued.

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FACTS LIST 2.

3.

4.

5 ~

6.

10.

March 11, 1989 RCS Drain Operations procedure, 43OP-3ZZ16 was implemented.

43OP-3ZZ16 was written to implement NRC Generic Letters 87-12 &, 88-17 and the PVNGS Response Letters dated September 21, 1989 (87-12) and January',

1989 (88-17).

43OP-3ZZ16 was used for the first time during the Unit 3 Refueling Outage which started in March, 1989.

A verbal request for an Engineering Action Request (EAR) was made on March 13, 1989 though the Nuclear Engineering Department (NED) for a curve for the Train B temporary tygon level indicator, Verbal requests for EARs are in accordance with procedure 81DP-4EE03 Task Control Within Nuclear Engineering.

A compensation curve for the Train A temporary level indicator connected to RCE-V214 already existed in 43OP-3ZZ06, Mode 5 Operations, and 43OP-3ZZ12 NED answered the EAR (88-1671) on December 27, 1988.

The EAR answer stated that the data collected in Unit 2 for RCS level decreases due to shutdown cooling flow was applicable to all three units due to similar configurations.

The EAR answer included Unit 2 data forTrain A and Train B level compensation.

The level compensation curve for the Train B level indicator was based on data collected from drain valve SIB-V057.

12.

Drain valve SIB-V057 is the point the temporary refueling level indicator is connected.

Drain valve SIB-V057 is the point the permanent Refueling Water Level Indicating System modification is to be installed.

13.

The level compensation curve for the Train A level indicator was based on data collected from drain valve SIB-V056.

14.

Drain valve SIB-V056 is the point the permanent Refueling Water Level Indicating System modification is to be installed.

15.

16.

The existing plant design identified RCE-V214 on the Train A shutdown cooling loop as the refueling level indication connection used for the temporary level indicator.

The procedure writer incorporated the data from the EAR as curves directly into 43OP-3ZZ16.

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FACTS LIST (con't) 17.

EAR 88-1671 identified SIA-V056 Ec SIB-V057 as the data collection points in the letter attached to the EAR.

18.

The data sheets attached to EAR 88-1671 were identified as Level Decrease Data forTrain A and Level Decrease Data forTrain B.

19.

43OP-3ZZ16 cross-discipline review included the EED System Engineer.

20.

43OP-3ZZ16 cross-discipline review did not include the NED Design Engineer who wrote the EAR.

21.

Per 01AC-OAP02, Review and Approval ofNuclear Administrative and Technical Procedures, the Technical Reviewer shall determine the need for a cross-discipline review 22.

Per 01AC-OAP02 guidance, a cross-discipline review is required when more than one section has a major role in the performance of the task described by the procedure or an intent change is made to a system operating procedure.

23.

24.

Per 01AC-OAP02, the cross-discipline review should occur when more than one section has established expertise in the area covered by the procedure and the Technical Reviewer determines the need for a confirming opinion.

t No comments were received concerning the flowcompensation curves on the EED cross-discipline reviews.

25.

Per 01AC-OAP02, the Technical Reviewer conducts a detailed technical review to ensure that the procedure:

(1)

(2)

(3)

(4) accomplishes its purpose; has valid acceptance criteria; has clearly defined responsibilities; is consistant with applicable licensing and regulatory documents, other higher tier documents, and applicable technical requirements.

26.

The procedure was approved and implemented with the discrepancy in place.

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CONCLUSIONS The procedure writer missed the detail in the NED EAR that the Level Decrease Data for Train A flow was obtained from a point different than the point the temporary level indication was to be installed.

(Fact Number: 2,4,6,7,8,9,10,11,12,13,14,15,16,17,18,26)

The point that the procedure writer missed the detail that the Level Decrease Data forTrain A flow was obtained from a point different than the point the temporary level indication was to be installed might have been made more visable to the procedure writerifthe the individual that answered the EAR had identified on the data form that the data was for instrumentation installed at SIA-V056 instead ofjust "Train A".

(Fact Number: 4,5,6,7,8,9,10,11,12,13,14,15,16,17,18,26)

The EED System Engineer should have indentified the difference between the point that the level curve data was obtained and the point at which the temporary tygon level was installed and that this difference would effect the level indication during the cross-discipline review.

However, the procedure 01AC-OAP02, Review and Approval ofNuclear Administrative and Technical Procedures does not provide sufficient information to the cross-discipline reviewer as to what a cross-discipline review is to accomplish, the "depth" and detail the review is to take.

(Fact Number: 19,20,21,22,23,24,25,26)

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RECOMMENDED CORRECTIVE ACTIONS The procedure writer is to be counseled on the importance ofidentifying all details when preparing any portion ofa procedure.

The smallest oversight or conclusion can lead to items ofmajor impact.

(Conclusion Number. 1)

Responsibility:

Due Date:

Operations Standards Supervisor 30 days after report approval.

2.

NED shall develop specific guidelines ofensuring that engineering information transmitted to the site has clearly stated assumptions and limitations.

(Conclusion Number: 2)

Responsibility; Due Date:

Nuclear Engineering Manager 60 days after report approval.

3.

Enhance the procedure 01AC-OAP02, Review and Approval ofNuclear Administrative and Technical Procedures to provide the technical reviewer and the cross-discipline reviewer guidance and details as to what a cross-discipline review is to accomplish, the "depth" and detail the review is to take, and who should conduct the cross-discipline reviewer (i.e.; the cross-discipline review is a technical review in the cross-discipline reviewer's area of expertise: engineering to review the procedure from an engineering viewpoint verifying the accuracy, adequacy, applicability, etc., of the types of evolutions, calculations, curves, formulas, etc.; operations to review the procedure from and operations viewpoint.

verifying that the evolution is accomplished adequately, it does not create operability and/or operational concerns, chemistry to review the procedure for chemistry concerns fimpact; etc.) ICR 08713 submitted to Plant Standards and Control.

(Conclusion Number: 3,5)

Responsibility:

Due Date:

Plant Standards and Control Manager 120 days after report approval.

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ENERGY BARRIER BARRIER BARRIER BARRIER

¹4 BARRIER BARRIER BARRIER TARGET NOTE:

THIS GRAPHICALREPRESENTATION OF THE ENERGY-BARRIER-TARGETANALYSISIS ONLYINTENDEDTO INDICATEWHICH CATAGORIES OF BARRIERS WERE EFFECTIVE FOR THIS EVENTANDWHICHWERE NOT. WHILEIT IS RECOGNIZED THATTHERE ARE MANYPOSSIBLE PARALLELANDSERIES COMBINATIONSOF THESE CATAGORIES OF BARRIERS, ITIS NOTTHE INTENTOF THIS REPRESENTATION OF THE E-B-TANALYSISTO SHOW THOSE COMBINATIONS.

ENERGY-BARRIER-TARGETANALYSISFOR II'-3-89-019 ENERGY BARRIER

¹I BARRIER BARRIER BARRIER

¹4 BARRIER BARRIER BARRIER TARGET NOT USED ADEQUATE INFORMATION LESS THAN ADEQUATE UNDER EVALUATION LESS THAN ADEQUATE NOT USED ADEQUATE NOT USED BARRIER MBER AND DES RIPTI N

1. EQUIPMENTPERFORMANCE

2. PERSONNEL PERFORMANCE

3. PROCEDURES

4. TRAINING

~ 5. DESIGN

6. MANAGEMENT

7. OTHER BARRIER EFFE VENES.

ADEQUATE LESS THAN UNDPR USED ADEQUATE EVALUATION NOT USED NOT APPLICABLE QQNN ERN

~MBER N/A N/A N/A N/A N/A N/A

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. NRC Generic Letters 87-12 &

88-17 issued EVENTS AND CAUSAL FAC 01AC-OAP02 Review and Approval utilized for tech review Cross-Discipline Review completed by System Engineer ANPP responses to NRC Generic Letters 87-12 &

88-17 dated 9-21-87 1-6-89 81DPPEE03 Task Control

. WithinNuclear Engineering used for EAR Tech Review completed by former SRO/SS on PVNGS Procedure sent to EED but not NED for cross-disciplinc review Development of 43DP-3ZZ16, RCS Drain Operations starts.

11-88 Request to NED for level compensation curve forTrain B refueling level (at SIB-V057).

12-13-88 EAR 88-1671 answered by NED System Engineer.

12-27-88 Both EAR curves incorporated into 43OP-3ZZ16.

01-89 Complctcd procc-durc sent for "tech" review.

01-89 Completed proce-dure sent out for cross discipline review.

01-89 42OP-3ZZ16 approved and issued with inappropriate RCS Level vs SDC Flow Curve.

3-11-89 Contractor Procedure Writer (No previous license)

Well versed in NRC letters &

res pons cs 01AC-OOP01 Format and Contcmt 40DP-OAP01 Writer's Guide Utilized Train A Flow comp curve already existed in 43OP-3ZZ06 &

43OP-3ZZ12 at RCE-V214 EAR provided data for both Train A&B at valves SIA-V056 &

SIB-V057 Pcr 01AC-OAP02 Tech Review ensures procedure accom-plishes purpose, has valid acceptance criteria, has clearly defined responsibili-ties, is consistant with licensing and regula-tory documents and technical documents Cross discipline review when more than onc section has a major role in Ihe performance of task Cross discplinc rcviewwhcn morc than onc sccuon has expertise &

Tech Reviewer dctcrmincs nccd for matching opinion

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EVENT CATEGORIZATION WORK SHEET NOTE: ALLKEYWORDS UTILIZEDINTHIS CATEGORIZATIONMUST BE OBTAINEDFROM THE KEYWORD LIST IN APP. H OF 79DP-OIP01, INCIDENTINVESTIGATIONREPORT PREPARATION.

I.

System/Components Affected By The Event A. Component VH B. SIMS ID Number C. NPRDS Code D. Subject Primary OI'f=i4tt'isa 5;

E. System Affected Secondary

('~c. ~'Z tL Failure Mode (for component failure only) pg N/A III.

Generic Root Cause(s)

A. Major Category B. Causal Factors Categories Q C('Ivy <~

Qa C.

IV.

Plant Status Prior to Event p-9 k<~<

I V.

Reactor Trip Signal

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ESFAS Signai Generated ~ Nfa Vll.

Event Classification Vill.

Affected Unit

(,~'C 6 fX.

Reapcnalttte WOrk Graup 0 ~Q't/O <<;~rangyj f;

Type of Activity Initiating the. Event IIr.b III-~

(.><>.4M~

PV419-04DJ Aev. S>69

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ATTACHMENTS Al -

NRC Notice ofViolation May 26, 1989 A2-PVNGS Response to NRC Notice of Violation June 26, 1989 A3 -

NRC Request for Incident Investigation Report July 3, 1989 A4-Personnel Statement - Dave Faulkner August 1, 1989 A5 -

Engineering Action Request - EAR 88-1671 December 13, 1989 A6 -

ICR 08713 for Corrective Action 03

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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION V 1450 MARIALANE,SUITE 210 WALNUTCREEK, CALIFORNIA94596

~yfjjg p tt tgbl8 Docket Numbers 50-528 50-529 50-530 Arizona Nuclear Power Project P. 0.

Box 52034 Phoenix, Arizona 85072-2034 Attention:

Mr. William F.

Conway, Executive Vice President Nuclear

~I!s"t I /gal 4'<~.'entlemen:

Subject:

NRC Inspection of Palo Verde Units 1, 2 and 3

This refers to the inspection conducted by Messrs.

T. Polich, D.

Coe and G. Fiorelli of this office on March 20 through April 26, 1989, of activities authorized by NRC License Nos.

NPF-41, NPF-51 and NPF-74, and to the discussion of our findings held by the inspectors with members of the Arizona Nuclear Power Project staff at, the conclusion of the inspection.

Areas examined during this inspection.are described in the enclosed inspection report.

Within these areas, the inspection consisted of selective examinations of procedures and representative

records, interviews with personnel, and observations by the inspectors.

Based on the results of this inspection, it appears that several of your activities were not conducted in full compliance with NRC requirements, as set forth in the Notice of Violation, enclosed herewith as Appendix A.

We are particularly concerned with the adequacy of your preparation for, and execution of Reactor Coolant SysteIII (RCS) mid-loop operations.

The procedures for this activity appeared incomplete.

Engineering data was not properly incorporated, and adequate contingency actions were not specified.

Furthermore, your oversight organizations did not provide timely, critical assessments commensurate with the importance of this evolution.

We request that you address these concerns in your response to Item A of the Notice of Vio'lation.

Your response to this Notice is to be submitted in accordance with the provisions of 10 CFR 2.201 as stated in Appendix A, Notice of Violation.

In accordance with 10 CFR 2.790(a),

a copy of this letter and the enclosures will be placed in the NRC Public Document Room.

The response directed by this letter and the accompanying Notice are not subject to the clearance procedure of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511..

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APPENDIX A NOTICE OF VIOLATION Arizona Nuclear Power Project Palo Verde Units 1, 2, and 3

Docket Numbers 50-528, 50-529, and 50"530 License Numbers NPF-41, NPF-51, and NPF-?4 During an NRC inspection conducted from March 20 through April 26, 1989.

two violations of NRC requirements wer e identified.

Violation A pertains to Unit 3, while Violation B pertains to Units 1, 2, and 3.

In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1988), the violations are listed below:

A.

10 CFR Part 50, Appendix B, Criterion V states in part: "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."

Contrary to the above, on March 11, 1989, the licensee issued procedure 430P-3ZZ16, "RCS Drain Operations",

which was not appropriate to activities affecting'he quality of Reactor Coolant System (RCS) operation during reduced RCS inventory conditions.'his procedure was not appropriate to the circumstances in that (1)

Appendix D, Page 1 of 2, was an incorrect RCS level versus shutdown cooling flow correction curve for the RCS temporary level indication system configuration used, and (2) procedural provisions intended to prevent vortexing and air entrainment were ineffective, resulting in

. actual air entrainment even though procedural requirements were met..

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

B.

Technical Specification 6.8. 1 states, in par t: "Written procedures shall be established, implemented, and maintained covering...

the recommendations in Appendix A of Regulatory Guide 1.33, Revision 2,

February, 1978..."

(RG 1.33).

1.

RG 1.33 is implemented in part by ANPP procedure 01AC-OAPOl, Revision 0, "Format and Content of Nuclear Administrative and Technical Procedures,"

Section

3. 4. 2, which states:

"Each

document, or changes

thereto, shall be reviewed and approved prior to use in accordance with 01AC-OAP02, "Review and Approval of Nuclear Administrative and Technical Procedures."

,Contrary to the above, between September 1 and December 23, 1988, surveillance test procedures 72ST-9CL04, 73ST-9CL06, and 73ST-OCL07 were conducted using criteria which had not been reviewed and approved prior to use in accordance with 01AC-OAP02.

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15 corroded and to combine the efforts of EERs 88-DG-58, 59, and 60.

This EER incorporated drain plugs with zinc anodes and allowed for carbon steel plugs to be used prior to manufacture of the zinc anode plugs.

On September 9, 1988, an intercooler drain plug failed on the Unit 2 "A" DG.

This failure was attributed to the same corrosion mechanism exhibited in the Unit 1 and 3

DG intercooler plugs.

A work order to inspect/replace Unit drain plugs had not been completed prior to this second event.

As discussed in the most recent SALP report, this was an example of weak problem identification since the same event had occurred on Unit 3 less than three months before.

EER-88-DG-064 was closed on November 7, 1988.

That EER stated that as of October 4, 1988, no work order had. been initiated to install the new drain plugs in Units 1 and 3 and that the installation of the plugs should be raised to the highest priority.

Work orders initiated at Unit 2 were scheduled to be completed before October 5, 1988.

The EER also recommended establishing a Preventive Maintenance (PM) task to monitor the corrosion of the zinc anodes.

The initial frequency of the PM was suggested to be semi-annual.

The inspectors review of the April 12, 1989 intercooler elbow leak indicated that Work Order (WO) 00237201 was performed on September 15, 1987, to replace a similar elbow on Unit 2 "A" DG intercooler.

The WO indicated the elbow was removed in'ieces but did not explicitly indicate corrosion was the cause of the damage to the elbow.

However, the WO indicated water was spraying from the elbow and the drawing and part number were the same as the April 12, 1989 failure.

The inspector made the following conclusions:

o Failure of the Unit 3 drain plug was not acted on aggressively to preclude a similar occurrence at Unit 2.

The corrective action for the Unit 3 drain plug was got thorough in that it only addressed the specific problem of drain plugs and did not address other carbon steel components in the system susceptible to the same corrosion mechanism.

The matter was first discussed with the licensee at the time of the September 9, 1988, failure of the Unit 2 drain plug and in the most recent SALP report.

The subject was again discussed with the licensee's management who acknowledged the licensee's comments and indicated agreement.

No violations or deviations of NRC requirements were identified.

Mid-Loo 0 erations - Unit 3 71707)

The inspector observed mid-loop operation preparations, entry and exit in Unit 3.

The licensee's mid-loop activities, including

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16 problem resolution and responsiveness to NRC Generic Letter 88-17 "Loss of Decay Heat Removal" were reviewed.

Finally, the effectiveness of the Quality Audits and Monitoring (gA and M), and Independent Safety Engineering (ISE) oversight groups was assessed.

The inspector made the following observations:

Procedural adequacy.

Procedure 43A0-3ZZ22, "Loss of Shutdown Cooling (SDC)",

stated that if SDC flow were totally 'lost while in Mode 5, operators should feed and bleed the steam generator secondary sides to provide for reactor coolant system (RCS) heat removal.

No recommended actions existed for the Mode 5 conditions when steam generators were unavailable due to mid-loop operations.

The NRC inspector identified this discrepancy and it was corrected by the licensee prior to mid-loop operations.

2)

Procedure 430P-3ZZ16, "RCS Drain Operations",

did not provide guidance for when or how to vent the SDC system.

Precursor indications such as abnormal flow noise or the appearance of air bubbles in the tygon tube level indicator were not addressed.

Specific valve numbers, and sequencing for venting operations were not addressed.

3)

Procedure 4$0P-3ZZ16, "RCS Drain Operations",

'as originally issued, contained an incorrect correction

. factor curve for the "A" RCS loop tygon tube level indication.

The incorrect curve assumed a different tygon tube connection point to the RCS than the one actually used.

Operators discovered the error during RCS drain operations when "A" and "B" loop levels became significantly different.

They stopped draining and corrected the error before proceeding.

However, the inspector noted that correction curve data supplied by engineering had been incorrectly incorporated into the procedure.

This is considered a violation of regulatory requirements (530/89-16-01).

4)

The surveillance test calibration procedure for the SDC flow meter, used to. ensure Technical Specification minimum flow requirements, was found by the licensee to indicate approximately 160 gpm greater than actual flow due to the in-use fluid temperature of 90 degrees F being lower than the calibration temperature of 300 degrees F.

This instrument is an orifice flow restriction device with a differential pressure detector.

'The inspector noted that this was a case of engineering data incorporated into a calibration procedure which resulted in an initially unrecognized actual difference between indicated and actual flow.

The licensee subsequently determined that due to conservatism of the minimum flow requirement, the indicated flow may be used without correction.

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17 Based on the above observations, the inspector concluded that;

1) procedures related to mid-loop operations were in some cases incomplete and inaccurate, and 2) there appeared to be a lack of control over the inclusion of engineering supplied data into operations and instrument calibration procedures.

Licensee management committed to a reassessment of the mid-loop operations procedures, including loss of SDC, with the objective of reverifying Generic Letter 88"17 requirements, ensuring the adequacy of engineering input, and incorporating all lessons learned from Unit 3, and completing the necessary revisions and training prior to any further mid-loop operations with fuel in the vessel (530/89-16-02).

Second, licensee management committed to reviewing the policies and controls associated with the exchange and review of information between the engineering and standards organizations.

This item will be followed up in a future inspection (530/89-16-04).

b.

Operations during mid-loop condition.

1)

Following entry into mid-loop operation, operators attributed the appearance of "growling" and "rumbling" flow noises, emanating from specific locations in the SDC flow path, to be caused by normal flow dynamics.

Consideration of possible air entrainment was apparently not made, even though the noises appeared only after the plant was placed in a mid-loop condition.

Operators were aware of the noises for approximately two days prior to notifying a system engineer.

2)

On March 26,

1989, operators attempted to minimize or eliminate the flow noise by slightly adjusting various throttle valves.

In doing so, they increased SDC flow from 4100 gpm, the maximum flow recommended by procedure, to 4250 gpm.

-The procedure indicated that the 4100 gpm.

recommendation was based on preventing vortexing or air entrainment in the SDC flow path.

3)

On March 27, 1989, subsequent to increasing SDC flow to 4250 gpm, air bubbles appeared in the tygon tube level indicator associated with the operating SDC train.

Operators reduced SDC flow and eliminated the air bubbles.

A system engineer walked down the flow path, but made no

'immediate recommendations.

4}

On March 28, 1989, one day later, air bubbles reappeared in the same tygon tube indicator, and the system engineer concurred with operations that the SDC system should be vented.

An estimated 100 gallon volume of air was then vented from the system.

Based on the above observations, the inspector concluded that the plant experienced vortexing and air entrainment during mid-loop operations.

This is considered' violation of regulatory requirements (530/89-16-03).

In addition, operators appeared to

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18 inappropriately attempt to reduce flow noise by exceeding procedural recommendations to limit total SDC flow.

Finally, the onsite engineering staff was slow to recommend corrective action.

Licensee management restated their commitment to ensuring that all

.appropriate operations and engineering staff, including management, are briefed on the significance of these events prior to the next mid-loop operation with fuel in the vessel.

Furthermore, licensee management commited to establishing, by adequate technical

means, the actual margin to vortexing prior to the next mid-loop operation.

This is part of open item (530/89"16-02),

addressed earlier in the section.

In addition, the inspector noted that the licensee was pursuing a

change to the minimum SDC flow required by Technical Specifications.

c.

Evaluation of Oversight Group Effectiveness.

The inspector reviewed gA Monitor Report No.

MOR89-0025 and Independent Safety Engineering (ISE) surveillance report No.89-012, both covering Unit 3 mid-loop operations.

The inspector assessed the degree to which these reports formed a self critical review of the Unit 3 mid-loop operation, and their emphasis on corrective actions needed prior to another unit entering a mid-loop condition.

The inspector determined that neither report recommended any corrective action to be completed prior to the next mid-loop operation.

The gA report was critical only of some differences between the training lecture given to the Technical Staff and the final approved RCS Drain Operation procedure.

The ISE report, under "Recommendations and Future Actions", only committed the ISE group

'o evaluate the inaccuracy of the SDC flow instrument and to review changes to the licensee's commitment to monitor the tygon tube level

--indications.

Neither the gA or the ISE reports were critical of the adequacy of procedures in use.

The inspector concluded that the gA and ISE critiques were ineffective in recognizing the scope and depth of needed changes to procedures, organizational interfaces, and operating policy.

Licensee management acknowledged these concerns and stated that renewed emphasis would be given for these groups to provide more critical reviews.

In conclusion, the licensee's preparations and conduct of mid-loop operations, following their commitments to NRC Generic Letter 88-17, did not prevent several problems from occurring, including entry of the plant into a vortexing.condition which is a precursor to air binding a

SDC pump and loss of SDC fl.ow.

The licensee's corrective action in response to these concerns will be carefully reviewed.

No violations or deviations of NRC requirements were identified.

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WILLIAMF. CONWAY EXECUTIVEVICEPAESIDENT HUCI.EAA Arizona Public Service Company P.O. BOX 53999

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PHOENIX, ARIZONA85072-3999 102-01315-WFC/TDS/JJN June 26, 1989 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Reference:

Letter from M. H. Mendonca, Acting Chief, Reactor Projects

Branch, U. S. Nuclear Regulatory Commission to Arizona Nuclear Power Project, Attn.

W.

F.

Conway, Executive Vice President, dated Hay 26, 1989

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2 and 3

Docket No.

STH 50-528 (License No. NPF-41)

STN 50-529 (License No. NPF-51)

STN 50-530 (License No. HPF-74)

Reply to Notice of Violations - 528/89-16-01, 528/89-16-03, 528/89-16-04, 530/89-16-01, 530/89-16-03 File:

89-070-026 This letter is provided in response to the inspection conducted by Hessrs.

T. Polich, D.

Coe and G. Fiorelli on March 20 through April 26, 1989.

Based upon the results of this inspection, violations of NRC requirements were identified.

These violations are discussed in Appendix A of the referenced letter.

A restatement of the violations and PVNGS's responses are provided in Appendix A and Attachments 1 and 2, respectively, to this letter.

Very truly yours, Zklgu&

. Wr~~

Executive Vice Preside Nuclear WFC/TDS/JJN/kj Attachment CC:

J.

B.

H. J.

T. L.

T. J.

A. C.

Hartin Davis Chan Polich Gehr

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Doc>>me>>t Co>>carol Desk Page 1 of 3 102-01315-HFC/TDS/J JN Ju>>e 26, 1989 APPENDIX A NOTICE OF VIOLATION Arizona Nuclear Power Project Palo Verde U>>its 1, 2, and 3

Docket Numbers 50-528, 50-529, a>>d 530 License Numbers NPF-41, NPF 51, and NPF-74 During an NRC inspection conducted from March 20 through April 26, 1989 two violations of NRC requireme>>ts were ide>>ti fied.

Violatio>> A pertains to Unit 3, whiie Violatio>> 8 pertains to U>>its 1,

2, and 3.

In accordance with the "Ge>>eral Stateme>>t of Policy and Procedure for NRC E>>i'orceme>>t Actions,"

10 CFR Part 2, Appendix C,

1980, the violations are listed below:

A.

10 CFR Part 50; Appendix B, Criterion 8 states in part:

"Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumsta>>ces and shall. be accomplished in accordance with these instructions, procedures, or drawings."

Contrary to the above, on March 11,

1989, the lice>>see.issued procedure 430P-3ZZ16, "RCS Drain Operations",

which was not appropriate to act1vjties affecting the quality of Reactor Coola>>t System (RCS) operation during reduced RCS i>>ventory conditions.

This procedure was r

not appropriate to the circumstances i>> that (1) Appe>>dix D, Page 1 of 2,

was an incorrect RCS level versus shutdown cooling flow correction curve for the RCS temporary level i>>dic<~tio>> system configuration used, and (2) procedural provisions inte>>ded to prevent vor texing a>>d air

t 1

1

Dnr>>me>>l; Control Desk Page Z ol 3

102-01315-HFC/TDS/J JN Ju>>e 26, 1989 entrai>>ment were l>>effective, resulting in actual air e>>trainment even though procedural requirements were met.

This is a Severity Level IV Vio1ation (Suppiement I).

B.

Tech>>ical Specification 6.8.1 states, in part:

"llritten procedures sha11 be established, implemented, a>>d mai>>tai>>ed covering...

the recommendations in Appendix A of Regulatory Guide 1:33, Revision 2,

February, 1978..."

(RG 1.33)

RG 1.33 is implemented in part by AHPP procedure 01AC-OAP01, Revision 0, "Format a>>d Content of Nuc1ear Admi>>istrative and Technical Procedures,"

Section 3.4.2, which states:

"Each

document, or changes

thereto, shall be reviewed and approved prior to use in accordance with 01AC-OAPOZ, "Review and Approval of Nuclear Admi>>istrative and Technica1 Procedures."

Contrary to the above, between September 1 and December 23,

1988, survei11ance test procedures 72ST-9CL04, 73ST-9CL06, and 73ST-OCL07 were conducted using criteria which had not been reviewed a>>d approved prior to use i>> accorda>>ce with 01AC-OAP02.

RG 1.33 paragraph 2,

"Genera1 P1ant Operating Procedures,"

recomme>>ds procedures for "Operation at Hot Standby."

RG 1.33 is impleme>>Led in part by ANPP procedure 410P-1SG01,

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Oncrrmerrt Control Desk Page 3 of 3 IOP.-OI315-HFC/TDS/JJN J uric 26, 1989 Revisiorr 8, "Hain Steam,"

wlrich requires in part, in paragraph 4.0, "Placir>g the Hain Steam Lines in Service with tire Hain Steam Isolatiorr Valves Open,"

compietior> of Appendix C,

"Atmospheric Dump Valve Li>>e Up."

Appendix C indicates that accumulator isolat.ion valve SG-V354 is to be open.

Contrary to the above, Unit 1 Atmospheric Dump Valve (ADV) Ho.

178 nitrogen isolation valve SG-V354 was closed on April 10,

1989, rendering the ADV inoperable from the Control Room.

3.

RG 1.33, Paragraplr 9,

"Procedures For Per Forming Haintenance,"

recommerrds procedures for the control of maintenance,

repair, and replacement.

RG 1.33 is irrrpierrrented by AHPP procedure 30DP-9HP01, Revision 1,

"Conduct of Hairrtenance,"

which states irr paragraph 3.3.3 that "Haintenance and Contractor Support Personnel Shall Perform >lork in Accordance l%tlr Approved Procedures arrd tlork Documentation".

Contrary to the above, on Alrril 4,

1989, tire installat.iorr.of a fuel line on the Unit 1 "A" emergency diesel was not performed in accordance wit.lr the inst.ructions in approved work package halo.

351776, resul tin'g irr tire fuel 1 ine's disconnection from the cylinder while the engine was running.

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

Document Cor>troi Desk Page 1 of 9 102-01315-llFC/TDS/J JH Jurre 2G, 1989 ATTACHHEHT 1 Reply to tlotice of Violation 530/89-16-01, 530/89-16-03 A. I REASOH FOR VIOLATIOH (530/89-16-01) 0>> March 11,

1989, APS issued an Admi>>istrative Control procedure "Reduced Inventory Operation",

(40AC-90P20) a>>d Operatirrg Procedure "RCS Drain Operations" (430P-3ZZIG) to control plant operations and evolutions during mid loop operations.

I>> March, 1989 Palo Verde Unit 3 entered its first reduced i>>verrtory operatiorrs dur irrg a refueling outage.

This refueling outage was the first time tlrat the procedures goverrring reduced irrver>tory oper'ations were used.

Or> llarch 11,

1989, "RCS Drai>> Operations" procedure, 430P-32ZIG, was issued wlrich incorporated a correction curve for the "A" RCS loop leve) indication.

The correction curve was generated in response to an NRC generic letter, wlriclr requires two trains of temporary level indication, and is used to correct for the effects of shutdown cooling flow on tire indicated level.

However, when the correction curve was incorporated irrto tire procedure, it was not recognized that tire curve was for a permanerrt level i>>dicator to be installed in the future rather than tire location currerrtly used for'ygon tub 1 rig ~

During drain-down of the

RCS, tire level indicators for the pressurizer level and the tygon tubir>g did not meet the cross check

Document Control Desk Page 2 of 9 102-01315-HFC/TDS/ J J tl June 26, 1909 criteria of +6 inches specified in tlie procedure (430P-32Z16).

The tygon hoses were walked down to check for any kiriks or loop seals.

ljo discrepancies were noted wliich would accouiit for the approximately 1 foot difference in levels.

During the draining

process, both of the temporary level indicators arid the cold calibrated pressurizer level instrumeiit were tracking consistently.

The. System Engineer was contacted at home and i ecommended continuing the RCS drain-dowri, usirig the level indicator which did iiot require flow coliipeiisation, while he was in transit to tlie site.

This decision was acceptable based on the fact this is the accepted method of monitoring level and only one temporary level indicator is required to be used until ACS level is below the ill foot elevation.

Draiii-down was recommenced and continued'until an

'indicated pressurizer level of 1 percent was reached.

llhen the System Engineer arrived on site, troubleshooting was conducted which irivolved switcliing the operating trains of SDC and

'letting the indicated levels stabilize while both traiiis of SDC were secured for a short period of time.

When both trains were secured I

tjie levels stabilized to within I/2 incIi of eacli other.

The "A" train of SDC was tfieri started, and iiidicated level data was.

col lected while slowly increas ing flow to the normal operating flow rate.

Analysis of.. this data showed that Uie "A" train level dynamic correction curve was not correct and that.the data taken matched with the correction curve that was originally in the Node 6 General Operating Procedui e (GOP).

l J

Doc illlieriI. Cori t, ro 1

Des k Page 3 of 9 107-01315-tll C/TDS/J J N Jurie 26, 1989 I

lhe original Ilode 6

GOP curve was incorporated i>>to 430P-32216 via a

Terrrporary Procedure Cliarige Notice (TPCII) arid dr a in-dowri opera t I oris resumed.

No other problems witli level cross cliecks were noted.

Research into tlie origiris of the level correctiori curves utilized in 430P-32Z16 revealed that the curves liad been provided in response to an NRC generic letter (via an Erigineering Action Request) which required Lliat two trains of temporary level indicatioii be provided.

Tlie Engi>>eerirrg Action Request (EAR) was dispositioned and provided level correction curves For botli trains in all three Uiiits.

Bott>

curves were derived I'rom errrpirical data obtained from Uiiit 2 utilizing level indicators which were connected to the same locations that would be used for tlie permaiient level indicating system (vice the locations used for tygon tubing).

During irrcorporatioir of tliese curves iiito 430P-32Z16, tire fact that the location used for coniiecting the reference leg of the "A" train tygon level iridicator via the IIT procedure was different t,han the one which would be used for the sarrre train in tlie permanent I

installation was missed.

The dynarrric liead loss difference between these two connection poirits caused the orie foot difference between tire two level correction factors.

Tliis was overlooked during tlie preparation,

review, and approval of 430P-3ZZ16 fur Uriit 3.

An investigation of tliis event is continuing,

liowever, based on the inIor'liiation currently available, i t has been determined that the N

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DocnmenL Control Desk Page 4 of 9

]02-0] 315-tlFC/TDS/J J W June 26, 1989 request for engineering to provide correction curves for the effect of shutdow>> cooling flow did not specify Lhe location of Lhe instrument taps for the tygon 1eve1 indicator.

The letter transmitting the correction curves clearly indicated that the curves were generated for the instrument taps associated with a planned permanent level indication system.

The procedure writer did not recognize this fact or the potential effect on the correction curve.

Additiona11y, during the technical review of the procedure, the engineering organization which generated the curves was not specified as a cross disciplinary reviewer.

Administrative control procedures require a review by individuals with the requisite 4

technica1 expertise but does not provide sufficient guidance for determining which group is responsible for performing the cross disciplinary review and the reguirements for the review.

Although the procedure was reviewed by the system engineer, he did not recognize that different instrument taps were utilized for the generation of Lhe curves than are used for the tygon tubing 1evel indicator.

A.l.ll CORRECTIVE STEPS TAKEW AWD'ESULTS ACHIEVED As immediate corrective action, Unit 3 issued a

TPCN to 43pp-3ZZ16 to incorporate the origina1 tfode 6 level correction curve.

An Engineering EvaIuation Request was a1so generated to document the

W j

Document Control Desk Page 5 of 9 102-01315-WF C/T OS/J J tl June 26, 1989 cause of the level difference.

lhe results of the evaluation have been incorporated into 410P-1ZZ16 (Uni t 1),

and wi 1

1 be incorpo~ ated into 42OP-ZZZ16 (Unit 2) pr ior to their use.

A.1.111 CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER VIOLATIOlIS An incident investigation for this event is in progress.

As part of this investigation, the findings noted in Section A. 1. I will be reviewed.

Upon completion of this investigation, appropriate corrective action will be developed,

assigned, and due dales will be scheduled for implementatior>,

A.l. IV DATE llllEtl FULL COtlPLIAtlCE WILL BE ACHIEVED Full compliance was achieved or) March 27, 1989 when 430P-3ZZ16 was revised to incorporate the correct level correction curves.

1 t

Document Control Desk Page 6 of 9 102-0.1 315-Hl C/TQS/J JH Jurie 26, 1989 REASOH FOR VIOLATlOH (530/89-16-03)

After a period of ti>>ie at mid loop operations, Unit 3 personnel rioted a flow noise developing in the viciriity of the sliutdown cooliiig (SDC) i>>jection valves for tlie "A" trai>>.

Operations atte>>ipted to determiiie tire source 'of the noise by varying tlie fIow tlirougli tlie various flow control valves to try to 'determine whetlier the noise could be minimized by a particular flow path lineup.

SDC flow rate was i>>creased to approximately 4250 gpm (the band allowed by procedure is 4000 to 4400 gpm) from approximately 4100 gpm.

At tliis flow rate, operators noted small air bubbles and air slugs in tlie tygon hose leve1 iiidicator wliicli was co>>nected to tlie RCS loop with the operati>>g t.rairi of SDC.

Flow was tlien thrott1ed back to 4150 gp>>i a>>d the system allowed to stabilize.

Small bubbles were still observed iri tire zygo>> tubirig, so flow was further thrott.led to 4070 gpm.

After fur tlier stabi1ization no bubbles were observed.

The RCS water level was unchanged tliroughout tliis evolution arid the flow noise continued as before.

Since the SDC cross connect piping to tire coritai>>ment spray, (CS) pumli provides a natural liigli poiiit in a stagnant flow area for collection of non-condensable gasses, it was suspected that gas had collected in tliis area and may be contributiiig to tlie problem.

Tfie cross corlilect pi))irig was vented for approxiliiately 6-8 llliliutes before a steady stream of wal.er issued fro>>i the vent., iiidicating a gas pocket liad existed in this liigli lioi>>t of tlie SDC piping.

Tlie

i

O()clllrlr.lit Coritr o 1 Oesk Page 7 of 9 102-0 I T 15-tlFC/TOS/J Jtl Jurie'6, 1989 displacenient of the gas pocket resulted iii a rlrop of tire ACS level of approximately 3/4 of'an incli.

It is believed that Llie iiicrease iri flowrate Lo approxinraLely 4250 gpm caused additiorral air entrainment.

Tlie flow noise reniained after tlie veiiLing oper'at.iori aiid coiit.iriued unt.il RCS level was raised out of tire niid loop condition.

Womirral flow iioise is an expecLed occurrence for accept. able levels of air eritrai>>merit flowing through valves and system piping.

Addi tiorral periodic venting of the system yielded no significant amounts of gas, and veiiting was discontinued after several days.

Prior to this event, an engineering evaluat.ion of t,he allowable SDC flowrate during niid loop conditioiis was perfornied.

The flowrate specified in t.he procedure was selected to meet Techirical Specificatiori requirements (i.e.,

> 4000 gpm) and plarit operational requirement,s (i.e.,

< 4400 gpni to preveiit gas biridirig or punip failure).

Tlie uliper flowrate limit (i.e.,

4400 gpm) was determined to be accept. able based on flowrate tesLiiig and data col lection performed during startup.

Gas biiiding or punip failure did not occur up to flowrates of 4400 gpm; however, expected air entrainment occurred and was determined not to adversely effect the system at flowrates of 4000 to 4400 glim diiring mid loop olierations.

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Oorrrrrrent Control Desk Page 8 of 9 102-01315-HFC/TDS/JJtt Jurre 26, 1989 A.2.II CORRECTIVE STEPS TAKEtt AttO RESULTS ACHIEVED To rrrinimirze the effects of excessive air entrainment arrd potential adverse affects on the RCS level indication, errgirreering guidance has been provided tlrat SOC f1ow should be maintai>>ed betweerr 4000 and 4100 gpm when tire RCS leve1 is below 104 feet.

Tlris requirement Iras been irrcorporated into 430P-3ZZ16 arrd 410P-IZZ16.

This requirement and/or additional guidance will be provided in the init.ial issue of 420P-ZZZ16.

A.Z.III CORRECTIVE STEPS THAT tIILL BE TAKEW TO AVOID FURTHER VIOLATIOHS A test has beer) performed to deterrrrine the actual flow conditiorrs in the RCS and SDC pipirrg while at or near rrrid loop operaLions.

Tire t.est irrvolved varyirrg ttre SDC flowrate at various RCS levels.

Tlris Lest was perfor'rrred irr Urrit 3 after fuel off-1oad was completed and ttre RCS 1evel was 1owered back to a ririd loop corrditiorr.

The results ol'he test,irrg wi11 be used to deterrrrirre tire required SDC operating parameters for applicable RCS levels.

Tlrese results are expected to be incorporated into Lfre appropr'iate operating procedures by Septerrrber 30, 1989.

Further gur'dance.wi11 be given irr tire RCS drain operatiorrs procedure training that Auxiliary Operators (AOs) slrould be aware of increased pump noise, bubbles in ttre tygon lroses, flow noises/rurr)ble wlren tourirrg in tire areas of tire operating SDC loop.

If'frese condit,ions

I 14

Dociinii'nt Coiitrol Desk Page 9 of 9 102-01315-1IFC/ fDS/J JH Juiie 26, 1989 are iioted, tlieri venting of tlie system liigli poiiits is recommended and closer observation of tlie system wi11 be requiied to monitor for further symptoms of impeiiding vortex iiig.

Tliis tra ii)irig is expected to be completed by September 30, 1989.

A.Z. IV DATE 14HEfl FULL COHPLIAtlCE 1lILL BE ACHIEVED Although APS believes tliat 430P-3ZZ16 provided appropr iate guidance to prevent gas bindiiig and SDC pump failure, on April 17, 1989 430P-3ZZ16 was revised to requir'e sliutdow>> cool iiig flow be 1ess ttian 4100 gpm.

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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION V g450 MARIALANf SUITf 2'IO WALNUTCRffK, CALIFORNIA94595 Ju~ ia;<<,

f / -0/0-8 D/ j Q

)'vi'g Q, Docket Nos. 50-528, 50-529, 50-530 Arizona Nuclear Power Project P. 0. 8ox 52034

Phoenix, Arizona 85072-2034 Attention:

ter.

W. F.

Conway Executive Vice President Gentlemen:

Thank you for your letter of June 26, 1989,"in response to our Notice of Violation and Inspection Report No. 50-528/89-'16'0-529/89-16 d

/

ated I)ay 26, 1989, informing us of the steps you have taken to correct the items which we brought to your attention.

Paragraph A.l.III of your response states that an incident investigation of the problems that ou rovide encountered with mid-loop operation. at Unit 3 is in ro W

is in progress.

We request y

provi e us the results, of your investigation and your planned corrective actions, following the completio'n of. your investigation.

As staff we unde discussed between Nr. S. Richards. of my.staff.and M

T Sh f

e understand that your investigation will be completed by August 30,

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Our Inspection Report 50-528/89-16 and the Office of Nuclear Reactor res onse to G

Regulation letter (Chan to Karner) dated t1ay 5

1989 h'

p o

eneric Letter 88-17, "Loss of Decay Heat Removal," both have deca heat questioned whether you have thoroughly'reviewed and add d th resse e issue of y

a removal during mid-1'oop reactor coolant system o

t' pection Report 50-.528/89-..16, paragraph 12, we understand that you e

opera ion.

s corn lete are reassessing the actions taken in response"to Generic Lett 88-17 d 'll er an wi and o eratio p

this reassessment and.appropriately. brief manageme t p

ons personnel, prior to.any further mid-loop operations with fuel n

, engineering, in the reactor vessel.

We want to again reemphasize the. importance the NRC places in being properly prepared for the conduct of mid-loop operations.

inspection.

Your actions regarding the above issues will be reviewed during f t Your cooperation with us is appreciated.

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Sincerely, e.

', Ii"~l R.

P.

Ziiiimermen, Acting Oirectnr Division of Reactor Safety

, and Projects

REPORT"NUMBER:

PERSONNEL STATEMENT NAME'~e EXT. 4W++

STA.

~P~

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PositionlTitle: ~ ~>~ ~> ~u ~~~~5 r<<

Your statement should include Unit conditions prior to the event, what indications you noted that

. a problem existed, your actions as a result of those indications, noted equipment malfunctions or inadequacies and noted procedural deficiencies.

Include any information, no matter how seemingly unimportant which might be important to review of this event as well as actions you recommend to avoid recurrence, ifany.

/ 2-~~/4 -~< $LN 8 67/

/2 VO~

Signature DatelTime

L

C, REPORT NUMBER:

PERSONNEL STATEMENT I

NAME:

~ 4>8 t ken EXT.

STA.

F

~'osition/Title:

> ~

I Your statement should include Unit conditions prior to the event, what indications you noted that

. a problem existed, your actions as a result of those indications, noted equipment malfunctions or inadequacies and noted procedural deficiencies.

Include any information, no matter how seemingly unimportant which might be important to review of this event as well as actions you recommend to avoid recurrence, ifany.

0P-

+/Q Signatur DatelTime

i l

i

REPORT NUM8ER:

PERSONNEL STATEMENT NAME:

ExT. +70 sTA. 6 +7(3 PositionITitle:

Your statement should include Unit conditions prior to the event, what indications you noted that a problem existed, your actions as a result of those indications, noted equipment malfunctions or inadequacies and noted procedural deficiencies.

Include any information, no matter how seemingly unimportant which might be important to review of this event as well as actions you recommend to avoid recurrence, ifany.

~ SZ'C Signature 7

Qate/Time

REPORT NUMBER PERSONNEL STATEMENT NAME:

Ex>. 2 87 s>A. & 7Q Position/Title:

, Your statement should include Unit conditions prior to the event, what indications you noted that

. a problem existed, your actions as a result of those indications, noted equipment malfunctions or inadequacies and noted procedural deficiencies.

Include any information, no matter how seemingly unimportant which might be important to review of this event as well as actions you recommend to avoid recurrence, ifany.

Signature Date/Time

h

REPORT NUMBER:

PERSONNEL STATEMENT NAME:~~~ ~~l~~~

ExT. + 7 ~~ sTA. 4<70 Position/Title:

Your statement should include Unit conditions prior to the event, what indications you noted that

~ a problem existed, your actions as a result of those indications, noted equipment malfunctions or inadequacies and noted procedural deficiencies.

include any information, no matter how seemingly unimportant which might be important to review of this event as well as actions you recommend to avoid recurrence, ifany.

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A N P P SNGlNFERlNG ACTION REQUEST (EAR) r IGKNAVE (PLCASKPRIHTJ P.2 KARNQ'.OQ GATE:

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) Cp+prCI+L. ~( mO pR,~ W~W <C'ah'N t >4<+4@ QCCPPT DFJ4'IP 8+LJL +VCR, ) 1 I l ~, 'l fhSK TITLE: (PLKASEPRINTJ OIJR VREFEREHCKQOCUMENT(5) NO. ES PONS ISLE ENQINEERl (PLEAI5 PRINTJ u 4'E TA5K COOL PRIORITY GOOK: EXT'EO.ANAI SCIL CCMPL, I IMATE STAII QAT 44 HO>> EQUKST CONFIRMATION:IRSGIJS T(i/I'(AJJPJ HSISLESVP O ' TOTALCDT(M>>(TOO OOCTICI~~ 2Q (TAT& i~ s>e 44T(MATCO COu((l, (O(( OAT%( z z 44YIMATCOL ((NOOOO P HEM( Q'RE APPROVAL>>PVER2 0 (MAJ, OIKPOSITIQNIREKVLTSSVMMARY(INCLUOCNECESSARY FOLLOIVKJP ACT JONSR OATEl NA HP/C OTHEA +~V~>(.-~ ] pr~O jAaL '>~~ ~ ~e'E ~p/wc+8 ~&~4 id 7- ~gag W- 'Jul'pL//~(-4 g+/eu (z+ CC (O 0V z E Q IST NKW EH OOCIJMEHTS REPAREO 'TO OISPQSITIOH LT HE EAR OOCUMENTISEUEOATK f EC EOOEKIONANO ENCE]OOCUMCNF NO,OR REVIEWGOMPI ETIQNREJER LIC'EHKINOOQCUMEHT5 RESP. RCPT.J TOTAL XCZum St t DAtt ~ dF tOZQ. Aatt~ mvas 2~ CLOSEOVT APPT(OVALS (SIONhl'VRS/OATE> RE( pcu( , (AC /I Oll(AJOJ . PY< I(( 64(l R4T it4 EA LOSEQUT QA ""::;'"CW'X', Z', ~,... ~,, JMIAIJI.I((A'" I I I r ~ JUL 83 '89 14:29 ANPP23AI/E>l< v/4ae.agA$ 'in $41 ~ 4IOO vrNlax (PN440I rv "pS Arizona Nucfear Power Project r v IDg PATE: TO: StoA' 167-03167-JWR/SLG December Z3, 1988 J. T. Pollard 2706 6070 Prepared byt Signature Name/Ext,/Sta, Ravlawed Byt Signature Nome/Ext./stL. L'. Garrett/4264/70l0 J. H. Hassar/423 '010 File: 88-159-419 SDC Train B Level Decrease Data Approved by: Signature Nome/Ext. ta J. W. Rowland/4059/7010 'I I ZHEZQ'>2 The purpose of this memo ir4 to provide SDC Train B Laval Decraa¹a Data for the Refueling Water Level Indicator (Tygon Tube) as requested by Dave Faulkner (EAR 88-1671). ~ ~ P During the design, of the Refueling Water Lev'il Indication System (permanent"'" ~ system)-,.le'val decrease data dua to shutdown cooling flow was taken on Unit 2. Tha.results of this data coQ.ection era shown in Attachment 1, As can be 'oan Zrvro tha rasulta, bath Tzafn& and TraLn 5 daat.aa was takan. The Tr~in Tygon tuba was connected at valve P-SIA-V056 and tha Train B tube was 'onnoctad at valve P-SIB-V057. Per drawings 01-P-SIF-105, Rev. 10 and 23-P.SIF-105, Rev. '2, the applicabla configurations are similar for all three (3) units. Therefore;"'the Unit 2 data obtained should apply to Units 1 and 3 as well. Far your annvanfamc.a, @1nta nP hath trafnrt cata ara attached. This same sat of data will be used for the permanent'Refueling Water Level Indication System, Should any question arise, contact S. L, Garrett at extension. 4264. ~ ~ JWR/SLG/j le 1542A/2306A v ~ 1 ~ \\ v I ~ ~ v ~ ~'v I h ~ v ',.'"..': <'.,'.,i,".', "'.,'g',~g",~!'~.qj<.P.;I ~ JUL 83 '89 14:29 ANPP23AVE ~ ~ I W ~ II Page 2 J. T, Pollard I Attachments.'1) Letter !.67-02375-ECS/SLC, dated May 23, 1988 (2) Refund {ne U~t ov T 1 M 4 c o o( I A, P versus Shutdown Cooling System Flow (3) Refueling Water Level Monitoring System LQQp 8 A P versus Shutdown Cooling System Flow (4) Portion of 01-P-SIF-105,'Revision 10 (5) 'ortion of 23-P-SIF-105, Revision 12 cc: D. Faulkner E. C. Starling R. V. Burge JUL 83 '89 14:38 ANPPZ3FtVES ~ N p.5 Arlzoaz Nuctear Power ProJect o o SOX 52QS4 ~ PHOENIX. ARIZONAbSCF2~2014 16 I-02375-ECS/SLG Nay 23, 1988 Impell Co+oration 350 Lennon Lane walnut Creek, CA 94398 Attention: Emerson McFarland Pro5ect Nanager Contlemen; Subi ect; Fefueling Pater Level Indi.cation System (DCP 1/2/3PJ RC-151, Rav.

1) Level Decrease

. Data due to Shutdown Cooling Flow ANPP is horoin ittaching tho lovel deareooa deca due te shucdown cvaling flow for the Refueling Water Level Indioacion System (DCP 1/2/3FJ-RC-151, Rav. 1). Tha daea was taken by A, Hartwig on May 17, 1988 and ha confirmed that the sh'utdown cooling flow passing by the applicable taps (i,e., v056 and V057)'as the sama as that passing by the applicabla flow transmitters (i.e,~-306 end 307), Should any quoation

arise, contact S. L. Garrett at (602) 371-4264.

Very truly your~, CCS/SLC/] le Attachment E. C. Starling Manager Engineering CC; A. W. Hartwig J. M. Rowland J. H. Hasser 8, Hebison C, W. Sowars L, L. Hanson G. E. Hanson (lmpell) ~ ~ ~ 4* (I 4 I I ~ JUL 83 '89 14:38 ANPP23AVE>l< e ~ ~ ~~ELM~ ~v~ res e Tnitial level 103' 1 1/2" as read on>>B>> Train indication. SDC <~ w v saoo ca~ OA ~ -1 h>> TlaL>> s.'v<<ad*gye wsa>> 44 c-Appaw.le>>ac>> va1~r ficult to obtain because the tygon tube 100'euc 4500 GPM 99 "8 1/2 4000 GPM 100'-4-g/4>> 3500 CPM 101'-0" 3000 CPM 101 '6" 2500 CPM 101 <<.11>> 2000 GPM 102'~4-3/4" 1500 GFM 102 8 ~ 1/4>> 1000 GPM 102'-8-3/4" These readings are average

values, Due to dif iculty in maintaining a stable flow rate, the level oscillated as much as

+ 1/2>>, ~ ~ ) ~ ~j<<<<<< i; I,',' I <<>g f i ff 'UL 83 '89 14:38 ANPPZ3AVE4 ~TT t ~MEV o 8 sit O Initial level 102'-10-3/4s as read on "A" Train indication. I , 5000 CPM Leve1 ~ 99' 2" 4500 CPM Variationa of g 1 inch at this flow rate 4000 GPM 100'5-1/4" 3500 CPM 100' 10 ~ 1/2" 3000 GPM 2500 GPM 101'-5-1/2" II101'-9 3/4" 2000 GPM j 102 t 2 1/2 V ~ 1.500 cxM 102 e 7 1/2n Variations nf < 1 inc'h at thea f1aM rat~ 1000 GPM 102'9-1/2" These readings are average values Qua to difficulty in maintaining a stable flow rate,'he level oscillated between 1/8" and 1/2" except where noted. I 1 I i i ~ 8:F $8 14I31 ANPP23AVE ~:e '5 Po ) g 8 ee lUd m g 5 Cl EC g ~ 0 zgg g 5 id Pf g '6 v IA CNGi hl LU U tA LQx CDR 4p 10 0 4.75 d 'iooo REFCRENE:K zooo 3000 NO ~ QATAR RKVtSta~S SCS FLQW (GPH) 33iZS ZSiS 40'oo ':~IS ) ~ il CR CHrg SOOO 51 i5 QAK REFUEL[NO'A'fKAt.EVB.'o'~f TO%)>6 gypped~ LfifiQ A QP VCRCIIR Rkl)TQQzg Cga IjJQ ST'St~ei1 II 'wH f' A ~ .Arizona Nuclear Poser pro)erat.' r ~ i , ~ i,i I CCAL g J..! NONE Cle<QII<O tlat ATThQ.'NKNT':j: '89 14:31 AHPP23AVE ~ A lA h Q . 0 5! ~ $ 8 5 g 0 I 4 0 llqOA ~ Qt ~ ~ I DJÃ UU. (h UJ 10 0 toca 17.2S k 13 8r28 2OOO 3000 SCS Fsou {GPz> OC.OC 4000 44m 78 < 35+75 So'OO 8 5~ EJ El 5g C'46m NOi REFERENCE NO+ CATE ~ l Ihdpfl e ' REV1S1Q,'lg CR CH'.M ERG Q$ g4E ls Ar 1zona NucIeat ~ ~ <<ljh Nis REFUKl '.l/5 WATER LEVKl NCNfTORTNg gysTjg LC"R 9 b,P VERSUS SHUTOCAN CQ Q)QQ ' SYSTKH FLOW SChLC l ';. '. CRhgthQ Ngi '..' RE pouer pro) BC~ 'bloi;!E .:.,:-'-"',.AT~A.. I,~FNT P 1 ~ t ~ ~ JUt 83 '89 14:32 ANPP23AVES . 0<<iaa Of 01-F-BH'-105 Her. i% pS ttt (,t gtt (t~it( r vc, ~ ~ ~t ~ a a't g/ +v4'NCOAMf t I tffIWMiT~rtd gy~(I 'devanav ty rdt ~aSV Orru It P4((Col f fv All.i l l t s t( y J lk~ 4~ ~.tl Idr pl d~ r5e@ Vd f.It (tf I(IIIII ry tt Vo y(tt) ~Co .<a ~ v WrZ ttf ip p g" ~~V pl> ill'e DEl'AIL'0'O(lt dl ps< (]gf ~ r ~ (' Jt VI ~'l% tgC +gp Pl g(t lB ~ kt8 p 't,t (tt't( t(e(~((it'(t ViW r+9 P ~t( ~ l ~ =, ~ ~ 4 ~ t I , 'e' JUL 83 >89 i4:32 ANPP23AYEN ~ ~ ~ C ~ P. 11 Portion of 2'd-p-SIF-105 Rev> 12 ~ ~ ~ 1(l rg '( ii e ~PSV .CCOV1>(P(S SS AW( fr (1$/>V aA c,(e J >tea (s( H ~ $~,sr fiPgg P I ,eC 4' ee 1++~3 a'1 ~ .44 >e(pT 4( 1 c r~ 4% 'S,R! /'j j ace (4 gd ( ~C>>,1)'j I" 1 ('9 P 'rV ~~".(',. ~ l'l(>> >>( (P ~ pi< 4 r( er'.;>CJ( %@j>C P (" ,/ l ( ~,1 ~ (s 11 >>~ tQ 1 gi y~e~ r.>I ~ > ee a OE74>IL'0'40SO Dr <<>r +44'er V ~'s V.(44i, ~ e ie ~ >e i,r'~n~ 'i(c M~ ~ 1(( ~"f C V'( <<is'~V a ea ': u'-. ~'>r>>' r(e '~ ss ~ H( ~ $ $ 0 a ~

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~ ((c ,1!a" P4 'jr(4 ~el e e ,, ~ W(>>( 1+ (((( ~D/ a>r'. CC>C>(g *P g,r el ~ (r ~ e ~ ~ V ~ <<~er Nr>>> ar ~ e r'0 .gd J' 1 ei f ~ ~ e 4!4 $ g ~ >(e( X' ' ~ ~ I t w v e INSTRUCTION CHANGE REQUEST ORIGINATOR. ' 087i3 ORIGINATOR'S NAM STA. Noc EXT.: Q~1u SUBJECT', DATE: SHIFT: -/ 4~ o s M T.S:S RELATEDT(Check One) (Check One) Yes No QR QAG NQR PROBLEM DESCRIPTION'T-iX 3 ~- Ql i II~L ~ 3 dr '~e.~ SYSTEM: (Il4ppncehleI i <4d>M C c cue S lo -S'- &' t I V ~c~c-i C I "r 'L'C 'C Ve.u,~~., 5-i L<O-4k JH~c 0 N SUGGESTED RESOLUTI N: AHOY ')>DVidS A ~ LLc L ~J~ A IC.a ~ ~ c -'i. i C.iV~ ~ WO~C /. M.. PUwQ REF ER ENDES (P d I.O.'S. TECH. MANUALS.ETC.) xL iL. FORWARD TO ORGANIZATION: SH.. 8EVIEW FROM ORGANIZATION: a ps SH. ADDITIONALINFORMATIONATTACHED: @Yes No OAT (Check One) Priority Enhancement EXPECTED FEEDBACK DATE: EVOLUTION DATE: PRIORITY: RESPONSIBLE INDIVIDUAL: RESPONSIBLE SUPERVISOR OR DESIGNEE: ORIGINATOR CONTACTED: Yes No DATE: CHANGE COMPLETE Yes No N/A RESPONSIBLE SUPERVISOR'S COMMENTS PCN NO REV. Noc TASK REVIEW DATEI PV TIOOSBA I546I WHITE Copy - FILE ~ CANARYCopy - ORIGINATOR FEEDBACK ~ PINK Copy - ORIGINATOR RECOMMENDED CORIMCTIVEACTIONS The procedure writer is to be counseled on the importance ofidentifying all details when preparing any portion of a procedure. The smallest oversight or conclusion can lead to items ofmajor impact. (Conclusion Number. 1) Responsibility: Due Date: Operations Standards Supervisor 30 days after report approval. NED shall develop specific guidelines ofensuring that engineering information transmitted to the site has clearly stated assumptions and limitations.. (Conclusion Number: 2) Responsibility: Due Date: Nuclear Engineering Manager 60 days after report approval. Enhance the procedure 01AC-OAP02, Review and Approval of Nuclear Administrative and Technical Procedures to provide the technical reviewer and the cross-discipline reviewer guidance and details as to what a cross-discipline review is to accomplish, the "depth" and detail the review is to take, and who should conduct the cross-discipline reviewer (i.e.; the cross-discipline review is a technical review in the cross-discipline reviewer's area of expertise: engineering to review the procedure from an engineering viewpoint verifying the accuracy, adequacy, applicability, etc., of the types ofevolutions, calculations, curves, formulas, etc.; operations to review the procedure from and operations viewpoint. verifying that the evolution is accomplished adequately, it does not create operability and/or operational concerns, chemistry to review the procedure for chemistry concerns fimpact; etc.) ICR 08713 submitted to Plant Standards and Control. (Conclusion Number: 3,5) Responsibility: Due Date: Plant Standards and Control Manager 120 days after report approval. ij l II f l, I l.