ML17300A689

Informs That Design,Const & Preoperational Testing Complete Except as Listed.Facility & Operating Organization Expected to Be Ready to Commence Process Leading to Fuel Loading by 870324

ML17300A689
Person / Time
Site:	Palo Verde
Issue date:	03/13/1987
From:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Harold Denton Office of Nuclear Reactor Regulation
References
161-00072-EEVB, 161-72-EEVB, NUDOCS 8703170341
Download: ML17300A689 (78)

Text

/'EQUL*TOR NFORMATION DISTRIBUTION 'EM (R IDS)

ACCESSION "NBR: 8703170341 DOC. DATE: 87/03/13 NOTARIZED: YES li DOCKET I FACIL: STN-:50-530 Palo Verde Nuc1'ear Station> Unit 3> Arizona Pub 05000530 AUTH. NAME AUTHOR AFFILIATION VAN BRUNT>.E. E. *rizona Nuclear Power Prospect (formerly Ari zona Public Serv RECIP. NAl'1E REC IP I ENT AFFILIATION DENTON> H. R. Office of Nuclear Reactor Regulation. Director (post 851125

SUBJECT:

Informs that design. const h,preoperational testing complete except as listed. Faci 1 itg 5 operating organi zation expected to be ready to commence process leading to fuel loading bg 870324.

DISTRIBUTION CODE: BOOID COPIEB RECEIVED: LTR TITLE: Licensing Submittal: PSAR/FSAR *mdts Zc I ENCL g SIZE:

Related Correspondence NOTES: Standardized plant. M. Davis> NRR: }Cg. 05000530 RECIPIENT COP IEB RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PWR-B EB 1 1 PWR-B PEICSB 2 2 PWR-B FOB 1 1 PWR-B PD7 LA 1 1 PWR-B PD7 PD 1 ~ LIC ITR*> E 04 2 2 PWR-B PEICSB PWR-B RSB 1 INTERNAL: ACRS 41 6 6 ADM/LFMB 1 0 IE FILE 1 1 IE/DEPER/EPB 36 1 1 IE/DGAVT/GAB 21 1 1 NRR BWR ADTS 1 0 NRR PWR-B *DTS 1 0 NRR ROE> M. 1 1 1 1 L'QC/HDS}

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Arizona Nuclear Power Project P.O. BOX 52034 ~ PHOENIX. ARIZONA85072-2034 March 13, 1987 161-00072-EEVB/WFQ Dr. Harold'. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Unit 3 Docket No. STN 50-530 Certification Regarding the Design, Construction and Pre-Operational Testing of Palo Verde Nuclear Generating Station Unit 3 File: 87-D-056-026

Reference:

(1) Letter from D. G. Eisenhut (NRC) to E. E. Van Brunt, Jr.

(APS) dated February 24, 1983.

Subject:

Certification Regarding the Design and Construction of the Palo Verde Nuclear Generating Stations, Units 1, 2 and 3.

(2) Letter from J. G. Haynes (ANPP) to G. W., Knighton (NRC) dated February 13, 1987 (ANPP-40123).

Subject:

FSAR Correspondence.

(3) Letter from G. W. Knighton (NRC) to E. E. Van Brunt, Jr.

(ANPP) dated February 10, 1987.

Subject:

Palo Verde Unit 3 Completion Certification.

Dear Dr. Denton:

Arizona Public Service Company (APS), the applicant for an operating license for 'Palo Verde Nuclear Generating Station (PVNGS) Unit 3, wishes to advise you that design, construction, and pre-operational testing are complete except as noted herein and that we expect Unit 3 and the 'operating organization to be operationally ready on March 24, 1987, to commence the process leading to the fuel loading of the unit in accordance with the Technical Specifications for Unit 3:

A. FSAR and Licensin Commitments Status 10 CFR Part 50'.71(e) and 50.54(a)(3) recognize that the FSAR is a living document subject to periodic updating to conform the plant description set forth therein to. its actual state at any point in time. To the best of our knowledge, Reference 2 identified the known changes necessary to update the FSAR to conform to a current status. Future discrepancies.

will be handled in a manner compatible with 10 CFR Part 50.59.

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Dr. Harold R. Denton Certification Regarding the Design, Construction and Pre-Operational Testing of Palo Verde Nuclear Generating Station Unit 3 Page 2 B. Remainin Work Items The open item tracking systems in use at Palo Verde identify and: track completion of outstanding hardware deficiencies, test deficiencies disclosed during pre-operational testing, inspection, issues. and other items requiring resolution. Entries, are reviewed and classified so as to identify restraints to plant operational modes. Those items classified as restraints to fuel load will be closed out or dispositioned, prior to commencement of fuel load. ,Similarily, items classified as restraints to any of Operational Modes 5 through 1 will be closed out or dispositioned before operation in such mode is commenced.

C. Items Not Com lete for Which Schedular Exem tions Have Been Submitted Items required by the FSAR, other licensing commitments or other regulatory requirements which will not be completed by fuel load of Unit 3 and require license conditions are listed in Attachment 1. Requests for schedular exemption and/or justifications for interim operation, respecting such items, have been submitted as shown.

D. Prep erational Tests Attachment 2 lists those Phase I (preoperational) tests which will be performed after ini'tial fuel load per PSAR Section 14.2.5.

E. Areas of 'Si nificant Interest 'to NRC Reference 3 identified a number of areas that would be of significant interest to .the NRC in the issuance of the low power operating license.

As requested by the Reference 3 letter, a discussion of these items is

'provi'ded in Attachment 3.

F'arch 6 1987 Re ion V Confirmator Acti'on Letter In connection with this confirmatory action letter, Mr. J. B. Martin (Region V) requested that we take certain actions concerning, voluntary entry into Section 3.'0.3 of the Technical Specifications.. These are to be completed to Mr. Martin's satisfaction, per his direction, prior to

.Region V concurrence with issuance of the license.

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Dr.:Har'old R. Denton Certification Regarding the Design, Construction and Pre-Operational Testing of Palo Verde Nuclear Generating Station Unit 3 Page 3 To the best of my knowledge, the foregoing addresses the remaining activities and items which bear upon your issuance of an operating license authorizing fuel loading and. low power operation of Palo Verde Unit 3: In summary, with the completion of such items as noted, except where schedular exemptions are appropriate, I presently expect that we will be ready to commence the process leading to the fuel loading of Palo Verde Unit 3 on March 24, 1987.

Accordingly, we request that a forty year operating license for Unit 3 be issued on March 24, 1987, authorizing fuel loading and low power testing of said unit.

Very truly yours, r uA E. E. Van Brunt, Jr.

Executive Vice President Project Director JGH/WFQ/dim Attachments cc: Mr. John B. Martin, Administrator (all w/a)

U.S. Nuclear Regulatory Commission Region V 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596-5368

0. M. De Michele R. P. Zimmerman G. W. Knighton F. J. Miraglia D. Canady A. C. Gehr

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STATE OF ARIZONA )

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COUNTY OF MARICOPA)

I, Edwin E. Van Brunt, Jr., represent that I am Executive Vice President, Arizona Nuclear Power Project, that the foregoing document has been signed by me on behalf of Arizona Public Service Company with full authority to do so, that I have read such document and know its contents, and that to the best of my knowledge and belief, the statements made therein are true.

Edwin E. Van Brunt, Jr.

Sworn to before ee thde~g dey of 1987.

Notary Pu ic,'-

My Commission Expires:

Mv Commission Expires April 6, 1991

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ATTACHMENT 1 Schedular Exemptions SUBJECT ANPP LETTER 0 DATE Supplement 1 to NUREG-0737 40173 February 20, 1987 Radiation Monitoring System 161-00062 March ll, 1987 Diesel Generator (GDC-. 17) 39925 February 3, 1987

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ATTACHMENT 2 PRE-OPERATIONAL TESTS TO BE PERFORMED AFTER INITIAL FUEL LOAD (To Be Performed Per FSAR Section 14.2.5 Phase I Tests)

PROCEDURE DESCRIPTION 73PE-3SA01 Load Group Assigned Verification*

73PE-3DG01 Diesel Testing" 73PE-3SS02 Gas Analyzer Test

• B Train Diesel Testing

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ATTACHMENT 3 AREAS OP SIGNIPICANT INTEREST TO THE NRC

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.CONSTRUCTION AND PREOPERATIONAL TESTING STATUS CONSTRUCTION 'STATUS Estimated Fuel Loading Date: Pirst Quarter of 1987 Estimated Commercial Operation Date: Third Quarter of 1987 Estimated'ercent Construction. Complete: '99X+

As of January 15, 1987, all subsystems have been turned over to Startup or Operations.

PREOPERATIONAL. TEST STATUS Of a total of 245 tests, 232 .are complete as of. March 6, 1987. The number of tests not yet complete reflects tests in progress that will be completed prior to fuel load, and the tests listed in Attachment 2 that will be completed after fuel load. The preoperational testing, required'or fuel load, with the exception of the B diesel testing, will be completed on the safety. related or important to safety systems prior to fuel load.

The construction and preoperational testing status of Unit 3 was discussed in great detail in ANPP-39872, dated 1/30/87, from J.. G. Haynes (ANPP) to J. B. Martin (NRC), titled'Construction Status and Operational Readiness Report".

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POWER ASCENSION'ROGRAM SCHEDULE The following schedule provides the current power ascension program schedule for PVNGS Unit 3

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UNIT 3 PAGE I ACTUAL BEV 0 FORCST FLOAT FEB HAR 87 APR HAT 87 JUN JUL 87 87 RUG SEP OCT NOV 67 DEC JAN BB REV 0 COHPAR I SON 19FEB87 FINISH FINISH FINISH HRS 2 l6 2 l6 30 13 27 ll 25 6 22 6 20 - 3 17 3l 10 26 12 26 9 '23 7 21 16 KEY PROJECT HILESTONES SIX t

FOUAtHREC 7 HO OMK ECEIVE F EL 3 I I'E RECEIVE FUEL ON SITE 03HRR87 03HAB87 182 TECH SPE RR HONITOA TECH SPEC RAD HONITOR 28FEB87 28FE887 203 ON-TECH PE ARO HONlTO NON-TECH SPEC BAD HONITOR 04RPB87 04RPB87 -266 UE LOAD FUEL LOAD 28HAR87 28HAR87 I

ACTOR H RO SSEHSLT REACTOR HEAD ASSEHBLT 20APB87 20APR87 HOOE 5 ENl HODE 5 04APR87 04APR87 ENTRl'RAIN

'O'ORK

'8'ORK WINDOW 0'4HAY87 04HAY87 N INOON TRA N 'R'ORK TBAIN 'R 'ORK WINDOW lNOON OSJUN87 OSJUN87 R PRARTIO FOA HOOE EN TAT PREPARATION FOR NODE 4 ENTRY 09JUN87 09JUN87 OST COR HFT POST CORE HFT 03JUL87 03JUL87 OOE Q TAT HODE 4 ENTRY IIJUN87 11JUN87 HOE3 REPRRATION NO ENTAT HODE 3 PREPARATION ANO ENTRY 16JUN87 16JUN87 OOE AEPARRll 0 RNO ENTAT HODE 2 PREPARATION AND ENTRY 03JUL87 03JUL87 N TIAL CAI ICALITT INITIAL CRITICALITY 08JUL87 OBJUL87 OE I PA PRRAllON AN ENTAT NODE 1 PREPARATION AND ENTRY 12JUL87 12JUL87 0 POXER PHTSICS TES INO LOW POWER PHYSICS TESTING 12JUL87 12JUL87

-20X NEA ASCENSI N TESTINO 5-20/ POWER ASCENSION TESTING 26JUL87 26JUL87 CLOSE OUTPUT BREAKER 21JUL87 21JUL87 )LOS OUTPUT BRE KER 20/ TO 50/ PLATEAU 20 I SOX PL lEAU 09AUG87 09AUG87 SOX TO 8 X PLATEAU 50% TD 80/ PLATEAU 045EP87 04SEP87 OX TO 100 PLRTERU 80/ TD 100% PLRTEAU 16SEP87 16SEP87 00X AIEAU 100% PLATEAU 26SEP87 28SEP87 RNTT RUN WARRANTY BUN 28SEP87 28SEP87 COHHERCIAL OPERATIONS HREACIAL OP AATIOHS 28SEP87 28SEP87

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CORPORATE MANAGEMENT AND PLANT STAFFING Figure 1 shows the current Palo Verde Management Organization. These positions with the exception of the Plant Services Manager, are filled. In the interim, the Fire Protection and Security Departments, which report to the Plant Services Manager, report directly to the Plant Manager. Figure 2 shows the onsite organization, and Figure 3 details the organization specific to each unit reporting to each unit superintendent.

PVNGS has an adequate number of licensed operators to staff the three units on a six shift basis. This is discussed in detail in ANPP letter 161-00056-JGH/PGN, dated March 10, 1987. The balance of the plant staff is adequate to support operation of three units. This has .been demonstrated by our recent ability to simultaneously deal with a Unit 2 outage, a Unit 1 shutdown for repair, and Unit 3 preoperational testing and completion of construction.

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FIGUIIE 1 PRESIOEIIT O. M. Oa MICIIELE EXECUTIVE V.P.

E. E. VAtl BRUtlT, JR.

V.P.

HUCLEAR PROOUCTION J. G. HAYNES MGR.

ASST. V.P. PVNGS OIR. OIR.

NUCLEAR SAFETY HUCLEAR PROOUCTIOH PLANT MANAGER TECHNICAL SERVICES CORPORATE OA ISEG/NSG J.O. ORISCOLL J. R. BYWUM J. VOREES

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R. M. BUTLER H. E. IOE MGR.

+ MGR. PROJECT MGR. MGR. ASST. OIR.

EMERGENCY PLAHNIHG OPERATIOHS UNIT 3 HUGLEAR EHGINEERIHG CORPORATE OA Yotts J. E. KIRBY

0. G. J. M. ALLEtt A. C. ROGERS A. SOUZA MGR. MGR.

MGR. +

MGR.

AOMIIL SERVICES MAlttTEIIAHCE NUCLEAR CONSTRUCTION OUALITY SYSTEMS/

EIIGltIEERIHG J. H ~ TEIICII R. K, HFLSOH 0. B. FASHACIIT O. J. WTTAS MGR. MGR. MGR. MGR.

TRAIWIWG OUTAGE MAHAGEMEIIT NUCLEAR FUELS PROCUREMENT OUALITY tt. F. FERtIOtt P. J. BRAWOJES P. F. CRAWLEY H. E. CRAIG MGR; . +

SUPERlttTEttOEHT MGR. MGR.

OVALITY AUOITS/

WRF PLANT SERVICES RECOROS MOHITORIHG

0. E. BLACKSOW B. F. GOOWIH C. tt. RUSSO MGR. MGR. MGR.

TECIINICAL SUPPORT LICENSING OUALiTY CONTROL O. J. ZERltIGUE H. F. OUIHH 0. E. FOIILER

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COIIPLIAHCE

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FIGURE 2 PVHGS PLANT MANAGER J. R. BYNUM MANAGER MANAGER MAHAGER MANAGER MAtlAGER OUTAGE MAHAGEMEHT TECHNICAL SUPPORT OPERATIONS MAIHTENAtlCE PLANT SERVICES P. J. BRANOJES 0. J. ZERINGUE J. M. ALLEN R. K. NELSOH MANAGER MANAGER SUPERIHTENOEHT SUPERINTEIIOENT MAttAGER COMPLIANCE OPS ENGIHEERIHG UNIT I IKC MAINTEHANCE OPERATIOHS SECURITY T. 0. SIIRIVER L. G. PAPHORTH R. E. YOUNGER J. F. MINHICKS 0. M. HELSOtt SUPERVISOR SUPERIHTENOENT SUPERINTEHOEHT SUPERVISOR-STA UHIT 2 MCC FIRE PROTECTION M. L. CLYOE R. J. AONEY S. M. MOYERS R, tt. MEYERS MAHAGER SUPERINTENOENT SUPERINTEHOENT RAO. PROT. K CHEM. UNIT 3 ELEC. MAINTENANCE L, E. BROW R. E. GOUGE 6, R. OLSEH SUPERINTENOENT SUPERVISOR SUPERINTENDENT OPS COMPUTER SYSTEM OPERATIONS SUPPORT STATION SERVICES

0. C. PHILLIPS F. C. BUCKINGHAM R, 0. ZERING SUPERINTEHOEHT MECH. MAINTENANCE J. H. STOUT ONSITE ORGANIZATION

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CONDENSER LEAKS PVNGS has experienced several condenser tube failures and has implemented appropriate corrective actions to help minimize future tube failures and to improve plant availability. The condenser tube failures are attributed to the following conditions:

1) High cycle fatigue failures of condenser tubes have been caused by high velocity steam flow through the condenser .and,the sequencing of the steam bypass valves into the condenser. The ori'ginal steam bypass valve sequencing was such that at low bypass demand conditions where only a couple of valves are required, the first three selected valves would all dump to the "A" condenser shell. To correct these causes of high cycle fatigue tube failures, the condenser vacuum pressure was raised to lower the steam velocity and the steam bypass valves were re-sequenced and baffled to redirect the steam flow. The re-sequencing of the steam bypass valves equalizes the loading on the condenser shells by selecting the valves such that the first three selected valves are each directed to a ,different condenser shell. Additionally, antivibration clips have been installed in the lowest pressure shell ("A" shell) to reduce tube vibrations, and anti-vibration clips have been installed in the areas of high velocity in the "B" and "C" shells.

2) Tube failures have been .experienced due to 'impingement from steam or water dumps. The individual impingement sources have been corrected as they were identified. Additionally, the operators have been made aware of the proper manner of cond'enser dump, operation in order to avoid operation outside of the design conditions of the condenser.

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3) A;.number of the condenser tube problems 'have 'been attributed to power ascension testing. Firstly, the power ascension testing phase is the first time that the condenser is subjected to normal operating loads.

Thus, tube failures are most likely to occur at this time. Secondly, the condenser is operated at partial loads for extended periods of time during the power ascension testing phase. Since the condenser is primarily designed for full power operation, some tube failures are attributable to this extended operation at partial loading conditions.

The operating experience that was gained during the startup of PVNGS Units 1 and 2 has been applied to the Unit 3 condenser by implementing modifications to help minimize the design related condenser problems. The major changes that have been made to the Unit 3 condenser involved: i) installation of baffles for the steam bypass valve condenser dumps and other miscellaneous condenser dumps, ii) install'ation of anti-vibration clips in the condenser, iii) removal of stainless steel sheet insulation on the low pressure feedwater heaters, and iv) re-sequencing of the steam bypass valves. Additionally, a lOOX eddy current examination has been conducted on the Unit 3 condenser tubes. Tubes that showed unacceptable indications, dents or gouges that may have occurred during construction or hot functional testing were plugged.

Additional improvements will be made as appropriate based on future experience with the success of these improvements.

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FEEDWATER CONTROL The Feedwater Control System (FWCS) at PVNGS is currently being evaluated by on-site and off-site ANPP departments due to technical problems having been experienced. This evaluation resulted from the observation that the predicted FWCS setpoint for the feedwater pump speed for full power operation is apparently too high and producing excessively high feedwater control valve pressure drops. In the interim, it has been at the discretion of the shift supervisor whether to institute automatic control of feedwater pump speed at low power or to remain in manual until a higher power level is reached when.

the FWCS will be more stable.

In addition to technical problems, operator errors have occurred during manual operation of the system. After the Unit 1 trip on November 19, 1986, (LER 1-86-061), an interdisciplinary meeting was held involving representatives of Operations, Operations Engineering, STA, Licensed Training, Nuclear Engineering and CE engineering to discuss the need for additional operator guidance, training enhancements on FWCS and evaluate operator action.

During an ANPP in-depth discussion of Feedwater Control system performance, design and human factors, ANPP determined that several areas needed attention:

A) Present configuration of the simulator, in some cases (i.e., for feedwater), cannot reproduce the same conditions that occur in the plant. This is due to the lack of fidelity of the simulator models with the response of the Unit to feedwater perturbations. Simulator enhancements are being implemented to provide increased simulator fidelity.

B) It was further decided that additional guidance be given to Operations in the use of the Feedwater Control System during malfunctions. In response to this decision, a night order was developed which provides control room operators with general guidelines to use in taking corrective action-C) Discussions were also held concerning the need for information to be added to the requalification training program, to allow the licensed personnel to learn from these events. This information has been passed on to the PVNGS operating crews .during their regular requalification cycles. Operators have been briefed on previous transients and the directions of the night order. In addition,,the training emphasizes the need for personal awareness while operating the FWCS to reduce the potential for personnel errors.

The technical problems associated with the FWCS have been evaluated by PVNGS Operations Engineering. It was determined that the primary cause of system problems was the excessive minimum speed of the main feedwater turbine pumps which resulted in excessive pressure differential across the feedwater control valves and made automatic operation of the system at low power difficult. A test procedure was developed to allow for dynamic evaluation and fine tuning

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of the PWCS feedwater pump speed program. Pollowing implementation of the test procedure and during power acension after the Unit 1 outage, the PWCS provided adequate automatic control of steam generator levels from approximately 3X reactor power through 60X reactor power. In addition, the PWCS provided automatic control of steam generator levels during and after the turbine trip which occurred on March 6, 1987. Data collection will continue through ascension to 100% power in Unit 1 and will be evaluated to determine if further system tuning will be required. If successful, the modification will be made in Units 2 and 3 during performance of 73TI-9SP01, "Peedwater Control System Tuning Procedure".

4i 0 STEAM BYPASS CONTROL SYSTEM Satisfactory operation of the Steam Bypass Control System (SBCS) during load rejection and loss of feedwater pump events has been demonstrated at PVNGS Unit 1 by successful completion of planned transient testing at 50X, 70X, 80X and 100X power, as well as an unplanned turbine trip at 50X power and an unplanned feedwater pump trip at 100X power. However, there have been reactor trips between the 25X to 40X reactor power levels.

As a result of'hese trips, .the NSSS vendor, Combustion Engineering (CE), and ANPP conducted an engineering evaluation of the SBCS due to ANPP concerns .that the system may not have the capability to prevent,a reactor trip following a load rejection in the range of 25X to 40X power. CE stated that the SBCS will accommodate load rejections of any magnitude when .all .control systems are in the automatic mode with full capacity and capability .available. An additional change is being evaluated to modify the modulate speed of the SBCS valves to open faster when in modulate mode. This change will be evaluated during the power ascension test program in Unit 3.

In the past, equipment unavailability (all eight bypass valves not available) and procedure inadequacies (procedures have been corrected) have prevented the SBCS from sufficiently reducing secondary pressure.

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PERSONNEL ERRORS-During 1986 Palo Verde Units 1 and 2 generated a total of 108 Licensee Event Reports (LER's). Of these LER's, 56 were categorized as Technical Specification violations. The greatest number (20) of Tech Spec violations occurred during the first quarter; then decreased slightly through the second (15) and third (14) quarters to a low of seven in the fourth quarter. Unit 1 had considerably more Tech Spec violations than Unit 2 during the first quarters (Unit 1 =,14 and violations ll; Unit 2 ~ 6 and 4), but in the last half out between .the units (Unit 1 and'econd of the year the number of evened 6 and 4; Unit 2 8 and 3).

PVNGS experienced simi'lar declines in the category of "occurrences involving personnel", including the sub-classification of procedure violations. From a high of 25 occurrences (15 procedure violations) in ,the first quarter, personnel errors decreased to 19 (ll procedure violations) for the second quarter; 14 (10 procedure violations) for the third quarter; with a dramatic decrease to only five occurrences involving personnel (two procedure violations) during the fourth quarter. Unit 1 occurrences declined over the entire year (from 15 to 2), while Unit 2 peaked at ten during the third quarter, then dropped to three occurrences in the fourth. This information is current as of February 18, 1987.

These significant improvements in performance are attributed to aggressive corrective action instituted by PVNGS management. The ANPP Maintenance IGC Department implemented a Quality Improvement Report (QIR) program in September 1985. As a result of the success of the QIR program, its concept has been incorporated into the Interdepartmental Event Investigation (IEI) program, which was implemented throughout PVNGS in October, 1986. This program prescribes the methods to be used to determine the root cause of an event, identifies remedial actions necessary and actions needed to prevent recurrence. At the discretion of the Plant Manager, a committee of PVNGS Department Heads may be convened to review an investigative report and recommended corrective actions, including disciplinary measures. The improvement in plant performance over the year reflects these vigorous actions.

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OPERATIONS PROCEDURE DEVELOPMENT When procedure problems are identified or operational difficulties are encountered by one Unit, a procedure feedback or Temporary Procedure Change Notice (TPCN) is issued to correct the problem. It is the responsibility of the PVNGS Operations Support Group to review the feedback or copy of the TPCN and ensure that the other affected procedures within that Unit and the other Units receive the same procedure changes in a timely manner. This ensures the procedures for Units 1, 2, and 3 are kept consistent. The Support Group is structured so that one individual has cognizance over the three Unit's procedures of the same type to further aid in keeping the procedures consistent throughout PVNGS. Operations Support Technical reviewers pay particular attention to ensuring that procedure changes are made to the three Unit's procedures in a timely manner. Additionally, the procedure changes originating from within the Operations Support Group are accompanied by a Notification of Procedure Change and routed to the affected Unit for review and discussion prior to actual implementation.

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EMERGENCY OPERATING PROCEDURES As a result of actions taken by operators during a reactor trip on July 12, 1986 (LER 1-86-047), the NRC staff developed concerns about the adequacy and implementation of the Emergency Operating and Recovery Procedures. The staff requested copies of PVNGS procedures for evaluation, and as a result, developed recommendations which were forwarded to ANPP on September 10, 1986.

ANPP evaluated the staff's recommendations and determined that several of the recommended changes were already being incorporated. Additional changes were made in a timely manner as a result of the NRC recommendations. A meeting was held in Bethesda, MD on February 26, 1987, to discuss the details of the procedure revisions and to resolve issues resulting from the trip and'he usage of the procedures.

During the meeting, the staff was made aware of ANPP's process for developing these procedures, the basic philosophy in their implementation and as a result, issues were resolved. The staff also learned'hat several changes had been made to the procedures prior to the receipt of the September 10 letter.

The staff was satisfied with the adequacy and timeliness of ANPP's response to the recommendations.

ANPP was verbally requested to formally submit their response to the NRC by separate letter. ANPP agreed to this request and expects ,to submit this information prior to Unit 3 fuel load.

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COMMUNICATION BETWEEN UNITS ANPP has existing programs in place which provide for information dissemination at PVNGS. The following programs are in place at PVNGS:

(1) Operations Department Experience Report (ODER)

The ODER is designed specifically to provide a communications vehicle for the Operati'ons Department to disseminate information such as lessons learned from plant trips, oprating experience, events at other nuclear power plants, etc. The ODER .program is included in the PVNGS Operations Department Guidelines. The ODER program provides a means of disseminating operating information from one unit to the other PVNGS operators.

(2) Interdepartmental Event Investigation The Interdepartmental Event Investigation is a .newly implemented program designed to provide a consistent methodology for investigating events and identifying necessary corrective actions. This program is specifically oriented to address personnel errors and procedural violations. The results of the investigations are provided to the personnel that might be subject to similar incidents. The procedure which governs the Interdepartmental Event Investigation is 71AC-OZZ03 in the PVNGS Station Manual. For further details on how this procedure is applicable to addressing personnel errors see the response to Concern 823.

(3) Quality Talks/Safety Speaks Procedure 6N417.20.00 of the ANPP Policies and Procedures Manual is the Quality Talks/Safety Speaks program. This program is designed to keep employees informed, of current project safety problems, quality problems, and other areas of conern such as work practices, violations, trends, etc.

(4) System Engineer Program This program is in place to improve overall plant performance and reliability by using system engineers. The system engineers will maintain a 'knowledge on the status of the assigned system and will be considered the engineering expert on the system. Since a system engineer is responsible for the same system on the three Units, any occurrences and resulting lessons learned from one unit can be applied to the other two units in a timely manner.

In addition to the programs described above, some PVNGS Departments are provided with their own support group which aids the department in investigating events, root causes of malfunctions, special problems and dissemination of information to each unit's dedicated staff to ensure consistency between the PVNGS units. Examples of the interdepartmental support groups are: Operations Support, Maintenance Support, Radiation Support, etc.

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In the event that difficult and/or complex problems arise in .which several departments are involved, a special task force, can be formed to. address any broad scope issues requiring focused attenti'on from the vari'ous departments.

This type of task force involves a defined task force organization to resolve the problem. A recent example of where a task force has been formed to resolve d'ifficult problems is in the area of addressing our earlier security deficiencies.

Also, Licensee Event Reports (LERs) are reviewed by the Training Department with the licensed operators during the training week of the shift cycle to enhance operator understanding.

These programs provide a means to disseminate information between the PVNGS Units'perating crews and also between other various PVNGS groups.

~ i i COMMUNICATION MITH OTHER UTILITIES ANPP management is aware of the importance of maintaining close and regular communication with other nuclear plants, particularly, the,other CPC plants.

Management ,encourages employees to develop .and ~maintain contact with their counterparts at other utilities.

ANPP has'~'lso developed formal, communications',with utilities at various levels, by. participating in formally ,established management and working groups. ANPP actively participates in the Combustion Engineering Owners Group (GEOG) and its subcommittees, NUMARC, and the Nuclear Operations Committee.

Additionally, ANPP maintains contact with INPO and has assigned an engineer to work in the Evaluation and Assistance branch. This individual, in addition to helping INPO carry out its mission, serves a known point of contact for'NPP when the need arises.

In summary, employees. at all levels within ANPP have contacts with counterparts in other utiliti'es .and communicate regularly with them.

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DIESEL GENERATOR EVALUATION AND REPAIR On December 23, 1986 the Unit 3 Diesel Generator "B" engine failed during performance of Startup Procedure 93PE3SA01. Inspections revealed that the number 9 connecting rod experienced a fatigue failure. The failed connecting rod caused damage to other engine components including the number 9 crankshaft journal, lower bearing cap, articulated rod, pistons, counterweights, main bearing, bushings, engine centerframe, right hand side engine block, and cylinder sleeves. The root cause of the failure has been determined to be a crack which initiated at an oil hole in the articulating pin bore of the master connecting rod. The crack initiation was a result of iron electroplating used to build up the bore during the manufacturing process.

The crack propagated and caused the rod failure. A review by the vendor of its manufacturing records shows that they supplied two other iron electroplated power train components to PVNGS. They were the number 2 connecting rod on the 3B engine and the number 9 connecting rod on the 2A engine. Both of these connecting rods have been replaced.

The diesel repair consists of the replacement of damaged parts except the crankshaft, centerframe and right hand side engine block. The crankshaft's number 9 connecting rod journal was milled in place and a new master rod bearing is being manufactured. The damaged areas of the centerframe and engine block are being removed and replaced with steelplate structures that will be mechanically secured with bolts and metal stitching to the parent metal. After completion of repairs and final assembly, a retesting program will be performed. Under a separate communication (ANPP-39925-JGH/JRP/98.05)

ANPP has stated that Mode 4 entry will not be made until the diesel generator repair and retesting is completed. A report on this failure was submitted by ANPP-40058-JGH-DJW/DRL-92.11, dated February 9, 1987. This diesel repair and testing program is currently under NRC review.

On February 8, 1987, a fire occurred on the Unit 2 "A" diesel generator during surveillance testing. The fire was caused by fuel oil spraying on the hot exposed portion of the exhaust manifold after the fuel injection tube at the 5R cylinder became disengaged. Several components were damaged, but no structural damage to the engine was sustained. The damaged components have been replaced. As a result of this failure the fuel lines have been changed on both diesel generators in Units 1 and 2. The fuel lines on the Unit 3 "A" diesel generator will be changed prior to fuel load. The fuel lines on the Unit 3 "B" diesel generator will be changed prior to declaring it operable.

ANPP letter 023-02100-JGH/DRL, dated March .2, 1987, transmitted the final report for this incident.

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EVALUATION OF UNIT 1 STEAM GENERATOR TUBE LEAK On January 17, 1987, Unit 1 experienced a tube leak in Steam Generator 1. The initial leakage rate was estimated to be 0.2 gpm, increasing to 0.4 prior to initiation of plant shutdown.

An eddy-current inspection determined that the tube leak occurred at Row 3, Column 2, adjacent to the second tube support in the economizer section (cold leg side) of the steam generator. Additional eddy-current inspections identified tube wall loss in several other tubes (Table 1) in Steam Generators 1 and 2 in both Units 1 and 2.

An evaluation of the eddy-current data by ANPP and Combustion Engineering determined the indications to 'be typical of that caused by tube-to-support wear resulting from flow-induced tube vibrati.on. The evaluation of this condition was discussed in more detail in ANPP-40179-JGH/BJA, dated February 23, 1987.

This condition is isolated to a small number of tubes (Figure 1), at the intersection of the tube lane and the downcomer recirculating entrance window, which are subjected to high radial cross flow velocities.

ANPP has plugged and staked the tubes with any indication of wear in the local high velocity region in Units 1 and 2. Those tubes with no wear to-date are sufficiently supported to preclude vibration and will remain in service.

The PVNGS Units 1 and 2 unplugged tubes in the region of interest .will be reinspected during the first refueling outages to confirm that wear has not initiated in these tubes.

ANPP will preventively plug and stake thirty (30) tubes per corner (60 tubes per steam generator) in Unit 3 in the pattern shown in Figure 1. This pattern was produced by overlaying the tubes with wear indications from all of the Unit 1 and 2 inspection results, thereby bounding: the area susceptible to high tube vibration. The number and pattern of plugged tubes is considered sufficient to allow'peration -to the first refueling outage. The tube plugging will"be completed, prior to .the Reactor'.Coolant is currently System fill and vent operation following fuel load. This area being reviewed by the NRC.

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TABLE 1

SUMMARY

OF INSPECTION RESULTS Unit, 1 S/G A.

r> 3I Number of Worn Tubes, 17 I

S/G 82 Number of Worn Tubes ~ 9 Unit 2

'S/G 81 Number of Worn Tubes ~ 30 S/G //2 Number of Worn Tubes '~ 21 i'

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POST TRIP REVIEW PROCESS The Post Trip Review Reporting procedure at ANPP has undergone several enhancements in the area of root cause analysis. The following is a description of the current procedure.

Following a reactor trip event, the duty Shift Technical Advisor (STA) and the Unit Superintendent (or his designee) work together to formulate a preliminary event sequence and description. The Post Trip Evaluation Team (PTET) then meets to perform a detailed event analysis. The PTET is chaired by the Unit Superintendent and its members include supervisory personnel in maintenance, operations, engineering, and technical services. Additional personnel are called in if deemed necessary by the .Unit Superintendent. Using a defined list of information, the PTET identifies concerns and allocates resources to perform root cause analyses and suggest corrective .actions. These analyses are performed to determine the adequacy of system/component response, operator actions, procedure use, and procedure effectiveness.

Pollowing completion of the investigations and analyses performed by the various working groups of the PTET, this group meets again to evaluate the results and to formulate an action plan which is submitted to the Post Trip Management Review Team (PTMRT). The PTMRT is chaired by the Plant Manager and includes upper plant management personnel. The PTMRT reviews the plan and must concur with it prior to implementation.

'When work is authorized to proceed, the PTET is kept apprised of progress.

When the root cause is identified and corrected, or the possible troubleshooting avenues have been exhausted for each safety'ignificant identified concern, the PTET recommends approval of the Post Trip Review Report (PTRR) by the PTMRT.

The PTMRT reviews concerns, actions, and resolutions and approves initial mode 2 entry according to the following criteria:

Root cause identified and corrected, or The possible troubleshooting avenues have been exhausted, and Open items are: justified with bases;as,to why, restart is satisfactory in spite of the .o'en items.

If the open items constitute a change from normal plant configuration or design, a lOCPR50.59 review,and evaluation is performed.

When the concerns are addressed and the restart criteria has been met, restart approval is given by the PTMRT, following a technical review by an STA or licensed SRO individual and approval recommendation of the PTET.

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ENHANCEMENTS TO ROOT CAUSE PROGRAM A root cause program was developed to investigate unusual component failures and respond to post trip review items. A recent effort is underway to increase the number of pro-active investigations within the root cause program and improve .the System Engineers'bility to identify and correct subtle component failure trends. ANPP is developing a PVNGS component failure database which will be used for timely identification of component failure trends, Reliability Centered Maintenance programs, and PRA activities. The system will be an online, interactive system in which the system engineer can group component failures by unit(s), system, component types, manufacturer and model number.

In addition, quarterly reports will be issued to Operations Engineering in which root cause of failure studies would be automatically initiated for components that exhibit a high number of similar failures or significantly exceed Industry Average Pailure Rates. This program is. expected to be fully functional by the fourth quarter of 1987.

ANPP is also participating with INPO in a pilot study to evaluate safety system unavailability for the APW, Diesel Generator and HPSI systems. Safety System Unavailability is also available to the system engineers as well as industry component failure information compiled by INPO.

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SECURITY PROGRAM IMPROVEMENT SINCE THE ISSUANCE OF THE UNIT 2 OPERATING LICENSE Since the issuance of the Unit 2 Operating License, ANPP has experienced increased Inspection and Enforcement activities 'in the area of Securi'ty and Safeguards. These increased activities have culminated in the issuance of an aggregate severity level III violation and civil penalty, and a Systematic Assessment of- Licensee Performance '(SALP) rating of 3'. Concurrently, ANPP has redoubled its efforts to ensure the Security Program will meet or exceed the requirements. necessary to .protect the ,public health and safety. ANPP Management conti'nues to be acutely "concerned with Security and Safeguards, and this concern has resulted in the following Security Program improvements:

1. Security radio communications have been significantly improved.

2. Extensive Security computer system hardware and software changes have been implemented.

3. Security protective lighting has been enhanced plant-wide.

4. The Security/Maintenance interface has been improved to ensure ongoing support.

5. The Security Program Audit has been made more comprehensive as a management initiative.

6. There has been an overall reduction in the number of Security violations.

7. There has been a coordinated effort between Security and other departments which has prevented continuing Security problems.

8. The Security Plan and its implementing procedures have been updated and improved.

9. Fewer incidents of a recurring nature have demonstrated improvement in management, attention to Security.

10. Changes to 10CFR 73.71(c) reporting requirements have been implemented, improving the timeliness and accuracy of Security event reports.

11. Vacancies in Security Department key positions have been filled, including some positions which provide expertise in law enforcement investigations, as well as Compliance and Quality Assurance.

12. Effective corrective actions have resulted in continuing improvements in the case of vital area barriers and access controls.

13. The security plan was implemented in Unit 3 several months ahead of the time when it was implemented in the Units 1 and 2 schedules.

it This was done to check out the security system in Unit 3 prior to when would be needed.

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STATUS OF PRA ON SINGLE FAILURE VULNERABILITY An effort is underway to identify single failure vulnerabilities at PVNGS. A single failure identification program has been developed to identify, BOP systems and components in which a single failure could result in a reactor trip. This program consists of the following subprograms:

(a) Identification of mechanical and control system single failures which have the potential for resulting in a reactor trip (e.g.

SBCS, FWCS$ PPCS).

(b) Identification of evolutions (maintenance actions, surveillance tests) which have the potential for resulting in a reactor trip.

The systems operating procedures and surveillance tests will then be reviewed for adequate operator cautions to prevent an inadvertent reactor trip.

(c) Identification of reactor protection system failures which have significant potential for resulting in spurious reactor trips.

The single failure identification program is currently in the engineering evaluation phase. The emphasis of the program to-date has been the identification of single failures within control systems. Failure Modes and Effects Analyses, (FMEA's) have been completed on the FWCS, Reactor Regulating System, Pressurizer Pressure Control System and Pressurizer Level Control System. An FMEA of the SBCS is scheduled to be completed by April, 1987.

Additionally, a screening program is being developed to identify specific areas within plant electrical and mechanical systems which have the potential to result in reactor trip. As specific areas are identified which warrent detailed analysis, the required resources will be allocated to identify single failure vulnerabilities, and design changes considered as appropriate.

Additional actions which have been completed to date include:

(1) The categorization of root cause of system failures which have previously resulted in reactor trip.

(2) The evaluation of operating experience for identification of system failures which may be precursors to reactor trips.

(3) ANPP has contacted SCE and has exchanged information on the single failure programs which each utility has been pursuing.

(4) ANPP is currently considering involvement in a GEOG program to identify the ma)or root causes of unplanned'eactor scrams.

The program includ'es participants from 'several ANPP .departments such as Operations Engineering, Maintenance, Nuclear Engineering, and Nuclear Safety.

The Nuclear Analysis group of Nuclear Engineering has the overall 1'ead and coordination duties of the project for single failure identification. The program is expected to be completed by December 31, 1987.

i' The ANPP Nuclear Engineering Nuclear Analysis Group is developing a Level 1 PRA for PVNGS. This program will provide a deterministic method for identifying potential safety-related single failure vulnerabilities and undesirable system interactions. The PRA is expected to be completed by December 31, 1988. Additionally, this program is intended to satisfy the requirements for an Integrated Plant Evaluation (IPE).

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NOBLE GAS CONCENTRATIONS IN THE AUXILIARY BUILDING There have been occurrences of high noble gas concentrations in the Auxiliary Building Low Radioactive Areas of PVNGS Units 1 & 2. These occurrences have been random, and in no specific areas.

A number of possible means of radioactive gas transport have been preliminarily identified as follows:

10 Air Balance between floors.

2~ Drain design in conjunction with Auxiliary Building HVAC design.

31 Air Balance at Containment Personnel Hatch.

'4, Venting and Draining of equipment.

In Section 12.3.3.2 of the PVNGS PSAR, item A states that the average and maximum airborne radioactivity levels to which plant personnel are exposed are as low as reasonably achievable (ALARA) and .within the limits specified in 10CPR20. Summarized below are pertinent sections of 10CPR20.

1. Areas with airborne radioactivity greater than 0.25 MPC must be clearly posted, 10CPR20 Section 20.203 (d)(ii).

2. No person should be exposed to more than 1 MPC for a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week in a 13 week period, 10CPR20 Section 20.103 (a)(3).

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3. A system should be implemented to ensure individual exposure be tracked and monitored, 10CPR20 Section 20.103 (a)(3).

4. Exposures exceeding the stated, limits must be reported to the NRC, 10CFR20': Section, 20.405 (a)'(1)'.

QA/QC Quality Investigation Report (QA Hotline Pile 886-192) indicates that radiation protection and health physics were in compliance with 10CPR20 during the reported period.

Detection of random high noble gas concentrations in -a radiation controlled area does not constitute a violation of 10CFR20 provided the requirements as stated above are met, which they have been.

FSAR Section 12.3.3.3.2 is titled "Guidelines to Control Airborne Radioactivity". Item A of this PSAR section states the guideline that airflow is directed from areas with a lesser potential for contamination to areas with greater potential for contamination. This guideline for the design of the HVAC system has been used. However, since this anomaly has been discovered, measures are being taken to determine, an acceptable solution consistent with ANPP ALARA goals.

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These measures include the following future action plan:

1. Verify original design basis of Auxiliary Building HVAC (HA)'nd Radwaste Drain (RD) systems.

2. Conduct Tracer and Air Balance testing of HA and floor drain systems.

3. Computer modelling of "HA system and Radioactive .Haste Drains to substantiate. the testing 'results. ~I

4. Evaluate results and provide any necessary design, procedural, control changes. and'dministrative, JL

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if any. w Two Plant Change Requests (PCR's) are being written .to develop 'a preliminary design change for this .anomaly. These two PCR's, will be a major part of the overall .action plan.

An estimated 20. 22 weeks is needed to complete items 1-3, evaluate results and recommend a corrective action. This estimate does not. include procurement department activities.

ANPP has a firm policy advancing AL'ARA .practices. and .we are, pursuing solutions to these random occurrences. Until the solutions can be implemented, compensatory measures are being evaluated to reduce exposure to personnel, such as covering corridor drains when equipment with potential for high radioactivity is being drained in adjacent areas.

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ANTIMONY ACTIVATION PRODUCTS IN THE RCS In October, 1985, PVNGS Unit 1 experienced a substantial increase in the concentration of antimony activation products (principally Sb-122 and Sb-124) in the Reactor Coolant System (RCS). The concentrations were closely monitored and by early 1986 the concentrations of antimony activation products in the RCS were determined to be ,8.3E-02 Ci/ml for Sb-122 and 4.6E-02 Ci/ml for Sb-124. Additionally, high radiation readings were observed throughout the RCS and on the RCS letdown purification filters.

Investigation of the antimony activity concerns by ANPP and CE personnel identified three potential sources of the activated antimony. The three potential sources were identified as being the Reactor Coolant Pump (RCP) seals, the RCP Journal bearings, and the sustainer neutron sources. The RCPs were supplied by CE and manufactured by the German firm KSB. The CE-KSB RCP is shown in Figure 1. The RCP seals and journal bearings are made of an antimony carbide material. The antimony is introduced into the primary coolant by surface erosion and leaching, and tends to be absorbed on the zirconium dioxide layer of the fuel cladding. Once the antimony is introduced into the reactor core, it becomes activated by exposure .to the high neutron flux field of the reactor core. The second potential source of antimony is the sustainer neutron sources in the reactor core. The sustainer neutron sources consist of 50K elemental antimony (Sb-121 and Sb-123) and'0X beryllium (Be-9). The purposes of the neutron sources are to i) provide a sufficient base neutron level to check that the detector indications are due to core neutrons and not due to noise, ii) ensure that the indicated neutron counts accurately describe the shutdown condition of the core, and iii) detection of changes in the core multiplication factor during core loading, permit refueling, and approach to criticality.

At this time, the evidence strongly supports the belief that the source of the activated antimony is the RCPs and not the sustainer neutron sources. This is due to the fact that the antimony activity has been observed in both PVNGS Units 1 and 2 and the probability of the sustainer neutron sources failing in both units is considered to be low. Additionally, communications have been initiated with the German utilities that utilize similar RCPs. The experience of the German utilities indicates that the Germans have experienced the same problems with antimony activity in their primary coolant systems. The Germans concluded that the antimony activity was due to the RCP seals and bearings in most cases. However, at least two additional cases were attributed to leaking neutron sources at the Biblis reactors. The Germans have developed antimony-free carbon bearing and seal material and are successfully using RCP bearings that do not contain antimony. The Germans have also developed a new RCP seal design utilizing antimony-free materials. However, this seal design has not yet received much operating experience at German. plants and has not been developed for 'he'E-KSB pump design. It should be noted that the contributi'on of antimony'rom the RCP seals has been estimated to be only a fraction of the contribution from the RCP journal bearings. Thus, the primary source of the antimony activity, is believed .to be the RCP journal bearings.

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ANPP's plans for resolution of the antimony activity problem involve the following actions:

1) Battelle Pacific Northwest Laboratories is currently analyzing a sample of PVNGS RCS fluid. The analysis is expected to confirm the belief that leaking sustainer neutron sources are not the cause of the antimony activity.

2) ANPP and CE have evaluated the antimony-free German RCP bearings and have concluded that these bearings will be acceptable for use at PVNGS. ANPP is currently trying to obtain these new bearings for installation in the PVNGS units. However, due to uncertainties involved with procurement schedules at this time, it is uncertain as to when the new bearings. can be installed.

3) Techniques to enhance antimony removal are being investigated.

4) Engineering efforts to minimize the impact of antimony during refueling outages continue.

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REACTOR'OOLANT P'UMP MOTOR FLEXIBLE COUPLING THRUST'BEAR ING ASS Fill B LY MOTOR 'SUPPORT ASSciVI6 L'Y R I G I D. CO UP L I N G SEAL ASSEMBLY JOURNAL BEARING DI FFUSFR IMPELLER CASING SUCTION

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