ML20206L864

Responds to Re Operation of Plants Following Steam Generator Leak at Indian Point 3.NRC Evaluated Event & Concluded That No Immediate Safety Concerns Warrant Shutdown of Facilities

ML20206L864
Person / Time
Site:	Indian Point
Issue date:	11/23/1988
From:	Zech L NRC COMMISSION (OCM)
To:	Gilman B HOUSE OF REP.
Shared Package
ML20206L872	List: ML20206L864 ML20206L875
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CCS, NUDOCS 8811300288
Download: ML20206L864 (2)

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• UNITED STATES NRC PDR 89 n NUCLEAR REGULATORY COMMISSION Local PDR

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'% November 23, 1988 TMurley/JSniezek CHAIRMAN DCrutchfield PDI-1 Rdg CVogan SVarga PDI-1 GT B8oger PErickson RCapra WRussell, RI GPA/CA The Honorable Benjamin A. Gilman SECY 0958 United States House of VStello Representatives DMossburg Washington, D.C. 20515 Dgoss MSlosson

Dear Congressman Gilman:

DNeighbors I am responding to your letter of October 20, 1988, concerning the operation of the Indian Point Nuclear Power plants following a steam generator tube leak at Indian Point 3. The Nuclear Regulatory Commission (NRC) staff has evaluated this event and has concluded that.there are no immediate safety concerns that warrant the shutdown of Inotan Point 2 and 3.

On October 19, 1988, Indian Point 3 experienced a 2-gallon-per-minute tube leak in one of its four steam generators. The Licensee, the Power Authority of the State of New York (PASNY),

declared an unusual event (the lowest of four event categories) and, in accordarce with plant Technical Specificatiens, initiated a normal plant shutdown. After isolating the affected steam generator, PASNY terminated the unusual event.

There were no significant radiological consequences from this event. The total release consisted of about 0.05 curies of noble gases with an estimated dose at the site boundary of 0.00005 millirem (one ten millionth of the general public dose limits). A steam generator tube rupture is an analyzed event within the design basis of the facility, and plant procedures are in place to respond if a leak or a rupture occurs. NRC regional and Lie a d -

quarters personnel are monitoring the corrective actions that the licensee is taking to rep:.ir the affected steam generator. The event wiil be reported ir NRC's monthly inspection report, which i is scheduled for completion in December. A copy of this report will be forwarded to you as you requested. For your information,

. all four Indian Point 3 steam generators have been scheduled for

! replacement during the upcoming refueling outaoe scheduled for

February 1989.

Indian Point 1, which is owned by Consolidated Edison Company of New York, has been shut down since October 31, 1974, the expiration date nf a variance granted to the licensee from the requirements of the Commission's "Interim Acceptance Criteria for

! Emergency Core Cooling Systemt, for Light-Water Power Reactors."

An Order was issutd to Consolidated Cdison on February 11, 1980, Originated: NRR: Slosson g 3 ,

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4 to show cause why the operating provisions of License No. OPR-5 should not be revoked. Instead of modifying the design of the facility to neet Commission regulations, the licensee elected not to operate the facility. There is no fuel in the Indian Point I reactor, and it may not operate without Commission approval.

Consolidated Edison has advised the Commission that it has no plans to restart this reactor.

During the 1987 refueling outage at Unit 2, the licensee (Consolidated Edison) discovered cracks on the upper-to-lower head transition girth weld on all four steam generators. After grinding the cracks out of the girth weld, the licensee performed an analysis to determine it the repairs were acceptable. Based on this analysis, the NRC staff on December 24, 1987, issued an interim safety evaluation allowing operation and forwarded a copy to you by letter dated January 15, 1988. The staff completed its review of the licensee's analysis and issued a final safety evaluation by letter dated October 28, 1988. A copy of this evaluation is enclosed for your information.

The NRC provides day-to-day oversight of licensed activities at Units 2 and 3 through the four NRC resident inspectors assigned to the Indian Point site. This oversight is augmented by Headquarters and region-based inspections. In addition, the overall per-formance of the facilities is evaluated once every 12 to 18 months in accordance with the Commission's Systematic Assessment of Licensee Performance (SALP) Program. Copies of our most recent SALP reports are enclosed. On the basis of this extensive inspec-tion effort and the overall operational history of the units, the NRC has concluded that the Indian Point units are being operated within the terms and conditions of their licenses and that there is reasonable assurance that the health and safety nf the public are protected I hope these comments will resolve your concerns about the safe operation of the Indian Point plants.

Sincerely, hMLU.

Lando W. Zech Jr.

Enclosures:

As stated

( om Enclosure 1

. y y$ w\g UNITED STATEh NUCLEAR REGULATORY COMMISSION e-  :; , j WASHINGTON, D. C. 20555 g Occober 28, 1988 Docket No. 50-247 Mr. Stephen B. Bram Vice President, Nucitar Power Consolidated Edison Company of New York, Inc.

Broadway and Bleakley Avenue Buchanan, New York 10511

Dear Mr. Bram:

SURJECT: STEAM GENERATOR GIRTH WELD REPAIR SAFETY EVALVATION FOR INDIAN POINT NUCLEAR GENERATING t! NIT NO. 2 (TAC 666Adi Dilring the 1987 refueling outage, Consolidated Edison discovered cracks on the upper to lower head transition girth weld on all four Indian Point 2 steam generators. The cracks were ground out of the girth welds. By letter dated December 11, 1987, as supplemented December 23, 1987, Consolidated Edison submitted ,

e report of the stress and fatigue analysis which was perfomed to support the final weld configuration after repair.

By letter dated December 24, 1987, the staff fomarded a safety evaluation to support the re-start of Indian Point 2. However, the staff required Consolidated Edison to submit a more detailed stress and fatigue life analysis and a more detailed brittle fracture analysis using representative fracture toughness values to support long-tem operation. By letters dated January 15, February 15, and April 12, 1988 Consolidated Edison provided the required infomation.

The staff has reviewed the submittals and concluded that the stress analysis performed by Consolidated Edison indicates that the maximum range of the primary and secondary stress intensity for all pertinent loading conditions meets the criteria of the ASME Code,Section III. The fatique usage further indicates that based on actual plant operating conditions to date, and for the

.nost limiting poHtion of the Downcomer Flow Resistance (DCFR) plate, the plant has 32.5 years of design life remaining. In addition, the staff has reviewed the liceesee's confimation, through simulation, that the repaired girth weld material has suitable toughness. The staff has reviewed the simulation and concluded the results of the program provide reasonable assurance that the heat affected zone of the repaired girth weld has adequate toughness to support the fracture mechanic. analysis performed.

Therefore, the staff concludes, that continued y operation of Indian Point 2 is

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. . 2 l not expected to have any adverse safety consequences as a result of the girth weld repair. Tne staff's safety evaluation is enclosed.

Sincerely, NOdd.(M Robert A. Capra. Director Project Directorate I-l i Division of Reactor Projects. I/II cc: See next page 6

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$ Mr. Stephen B. Rram Indian Point Nuclear Generating Consolidated Eoison Company Station 1/2 of Neu, York, Inc.

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Mayor, Village of Ruchanan Director, Technical Development 236 Tate Avenue Procrams

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Buchanan, New York 10511 State of New York Energy Office Agency Building 2 Ms. Donna Ross Empire State Plaza New York State Energy Office Albany, New York 12223 2 Empire State Plaza 16th Floor vr. Feter Kokolakis, Director Albany, New York 12223 Nuclear Licensing Power Authority of the State of New York 123 Main Street Pr. Jude Del Percio White Plains, New York 10601 Manager of Regulatcry Affairs Consolidated Edison Company of New York, Inc. Mr. Walter Stein Croadway and Bleakley Avenue Secretary - NFSC Puchanan, New York 10511 Consolidated Edison Company of New York. Inc.

Senior Resident Inspector 4 Irving Place - 1822 ll.S. Nuclear Regulatory Commission New York, New York 10003 Post Office Box 38 Puchanan New York 10511 Regional Administrator, Region I ll.S. Nuclear Regulatory Commission Mr. Brent L. Brandenburg 475 Allendale Road Assistant General Counsel King of Prussia, PA 19406 Consolidated Edison Company of New York. Inc. >

d Irvino Place - 1822 Charlie Donaldson, Esquire New York, New York 10003 Assistant Attorney General New York Department of Law 120 Broadway New York, New York 10271 h e e

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SAFETY EVAltfATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION CON 5OLIDATED EDISDN COMPANY INDIAN POINT UNIT 2 DOCKET NO.: 50-247 EVALUATION OF THE REPAIR OF THE STEAM GENERATOR GIRTH WELDS I. INTRODUCTION During the 1987 refueling outage the licensee, the Consolidated Edison Company, conducted a scheduled inservice inspection (ISI) of steam generator (SG) #22. Ultrasonic reflections were detected on the inside circumference of the girth weld between the transition cone and upper shell (weld No. 6).

Visual examination of the inside circumference revealed essentially horizontal intemittent linear indications around the weld length. Ultrasonic and magnetic particle examinations were extended to 100% of the girth weld on all four steam generators and a program to effect repairs by grinding was initiated. Similar crackinn of the upper shell to transition cone grith weld has been reported and reviewed by the staff at Indian Point Unit 3 and the Surry Power Station.

In a letter dated December 11, 1087, and supplemented by letter dated December 23, 1987, the licensee provided a report of the stress and fatigue analysis which was perfomed to iustify the final veld configuration after repair. By letter dated December 24, 1987, the staff completed an interim safety evaluation to suoport the re-start of Indian Point 2. However, the staff recuired the licensee to submit a more detailed stress and fatigue life analysis and and a more detailed brittle fracture analysis using representatite fracture toughness values to support lono-term operation. B.' letter dated January 15. February 15, and April 12, 1988 Consolidated Edison provided the recuired information The purpose of this safety evaluation is to document the staff's m !ew of Consolidated Edison's submittals.

II. Weld Integrity As part of the ,iustification for continued cperation of the Indian Point steam generators after repair of cracks in the transition cone girth wold, the licensee had additional work completed by Westinghcuse to confim that the weld material had suitable toughness.

Westinghouse simulated the themal history of the heat affected zone using a gleeble weld simulation machine. Hardness and microstructure checks were perfomed to confim that the material was properly simulated. Charpy V-Notch toughness tests over a range of 6emperatures gave results indicating that the simulated material had adequate toughness to justify the Westinghouse fracture mechanics analysis. The minimum upper shelf ;emperature was shown to be + 76'F and the upper shelf toughness ranged from 52 to 71 ft-lbs. for the nine specimens tested at 76'F or above. The licensee concluded that this toughness was adeouate.

We have reviewed the simulation methodology and have concluded that the material of the heat affected zone of the girth weld was properly simulated.

We have also reviewed the results of the Charpy V-Notch testing and agree with the conclusions of the licensee that the results of this program provide reasonable assurance that the heat affected zone of the repaired girth weld has adequate toughness to support the fracture mechanics analysis perfomed.

III. STRESS AND FATIAIE LIFE ANALYSIS During the Fall of 1987 refueling outege at Indian Point Unit 2 station, ultrasonic (UT) examinations were performed over one-third of the length of the steam generator #22 upper shell-to-transition cone weld (girth weld #61 as part of the plant's ASME Secticn XI Inservice Inspection Program. This examination revealed several indi:ations exceeding the allowable values specified in the acceptance tables of the ASME Code Secticn XI. The indications were confined to the vicinity of the weld and were observed % the toe of the weld-crown in the contoured region between the weld and shell/ cone regions.

Several repair configurations were evaluated on the basis of the ASME Code Section III requirements. A stress analysis and fatigue usage evaluation were perfomed for the loading conditions defined in *.he Westinghouse Design Specification.

Two unifom grind and several local grind geometrics were evaluated.

The uniform grind geometrics (360 circumferential1y) included 0.75" and 1.00" deep grinds and were selected with the purpose of conservatively enveloping the actual grinding pattern achieved. The local grind geometrics included a local grind within the envelope fomed by the uniform grind, and c local grind which exceeded the depth of the unifom grind.

A stress analysis and fatigue usage evaluation ha 1en performed to justify the acceptability of the weld grind out configurationt a the basis of the ASME Code Section III for the remaining design life objective of the S/G's. The steam cenerator shell has been analyzed as a Class I component.

The design and normal operating, steady state and transient conditions which were evaluated in the analysis are previded in the Westinghouse Design Specification.

Other loads imposed on the stean generator and internals, such as nozzle end loads and loads resulting from flow induced vibration, do not affect the girth weld region of the shell and have not been considered in the analysis. The resulting stresses due to seismic events (based on 0.37 g and 0.25 g in the horizontal and vertical directions, respectively, for the SSE, as well as 0.19 9 and 0.1? 9 for the OBE) are all less than 500 psi and are considered insignificant in the overall evaluation.

The steady state and transient pressure and temperature solutions were obtained using the finite element analysis.

Detailed heat transfer evaluation was perfomed by the licensee for two of the transients, namely, reactor trip and feedwater cycling, in the girth weld region. For reactor trip, the stean generator water level is assumed tr drop well below (100 inches) the girth weld region. When auxiliary feedwater iniection

i occurs, the cold water droos from the feedring to the Downcomer Flow Resistanca (DCFR) plate, spreads out circumferentially and flows outward to the shell wall. The water impinges against the shell wall at a temperature of 160* F.

Three DCFR plate positions have been eva'uated to assess its effect on fatigue usage and primary plus secondary stress range.

The structural evaluation of the shell and, in particular, the grind region for the above mentioned unifom grind cases was perfomed. Calculations of the local primary stress in the vicinity of the grinds indicate that the highest stress occurs in the unground region of the shell and are all within the allowable.

The maximum range of primary plus secondary stress intensities were also calculated. In some instances the stress range exceeds 35 , the limit for elastic analysis. Inthesecases,asimplifiedelastic-plisticanalysiswas performed to confirm that the range of primary plus secondary r.embrane plus bending stress intensity, excluding themal bending is less than 35 . This is found to ba acceptable, in accordance with the requirements of the ISME Code Section III, N8-3600.

The fatigue evaluation was also oerformed for each of the DCFR plate positions.

It was found that the upper plate position results in the minimum time to reach a usage of 1. 0; 19.2 years and 70.6 years for the 1.0" and 0.75" grinds, respectively.

The effect of moving the DCFR plate from the upper most position to the lowest position in three years has also been assessed. For the 1.0" grind configuration, for example, the result indicates a fatigue usage of 1.0 in 25 years.

The assessment o# the acceptability of local grind configurations has addressed two possibilities. For the first case, a single groove 1.25" deep with a 0.5" radius at the bottom, and a 2:1 taper on the sides is considered. The local grind may be of any length circumterentially up to and including the total circumference. Multiple grinds are also considered. The results indicated a maximun stress concentration factors of 1.8 and 1.7 in the axial and circumferential directions, respectively. These are substantially lower than the corresponding maximum stress concentration factors, 2.86 and 2.83, obtained for the two uniform grind profiles. Based on these results, a single or multiple grooves of up to 1.2F" depth are ecceptable and are enveloped by the stress conditions of the unifom grind profiles.

For the second case, a local grind of 0.25" deep with 0.5" radius and 2:1 taper on the sides-is as3umed to be present in the flat of the unifom grind profiles. The maximum stress concentration factors at the groove are 2.03 and 0.05 respectively. These again compare favorably with those obtained for the unifom prind profiles. In addition, the maximum total grind depth of 1.25" (due to 0.25" depth of local grind and the 1.0" unifom grind configuration) also satisfies the minimum shell wall thickness requirement of 2.2", in accordance with the ASME Code,Section III, Subsection NB, paragraph NB-3324. It is therefore, concluded that the local grind profiles are acceptable and are enveloped by the stress conditions of the unifom grind profiles, i

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l The impact of the dry-out and refill transient on the girth wele repair area was evaluited in tems of fatigue usage. Stresses for this condition are scaled from the transient gradients, T (secondaryl minus T (AUX feedwateri. These stresses are 'icorporated into the fatigue analysis for the most limiting grind depth and the most limiting DCFR plate position. The results indicated that this transient had no effect on the overall fatigue usage for the girth weld and neither did it affect the primary plus secondary stress range.

In addition, transients due to feedwater cycling and reactor t"f o for actual plant ecnditions are considered. The boundary conditions used for the earlier feedwater cycling and reactor trip transients are based on an AllX feedwater temperature of 70*F. Actual plant operation, however, may result in an AUX feedwater temperature of 40*F during the winter months. Actual plant operation also shows the number of feedwater cycling transients to be around 3800 cycles which is a significant reduction from the equipment specification (E-Spec) value of 25000. Of the 3800 cycles,1510 cycles would occur in the winter months and the remaining 2290 cycles continue to be wam water 170*F) events.

The reactor trip transient is similarly affected by the cold AUX feedwater condition.

Stresses were obtained for the cold-water feedwater cycling and reactor trip transients by scaling the stresses from the analysis based on the 70*F AUX

'tedwater. The fatigue usage calculations for these transient conditions indicated that the corresponding number of years reouired to give a fatigue usage equal to 1.0 is 32.5, compared to 19.2 years based on E-Spec conditions.

The prfmary plus secondary stress ranges previously obtained are not significantly affected by the cold-water events. ,

Based on these results, the transients as defined in the Ecuf pment Specifications can be concluded to represent a bounding set of conditions for the girth weld repair.

Based on the enrrent available infomation, no single trechanism is found to be responsible for both the initiation and growth of the indications. Stress and j fatigue usage analysis, in accordance with the ASME Code,Section I!!, would not '

support crack initiation based on design fatigue curves and normal water chemistry

and material properties. The presence of pits in the vicinity of the weld and l the conclusion from the failure analysis that the cracks apparently initiate l from the pits suggest environment factor may play a sionificant role in the overall failure mechanism.

l The stress and fatigue usage evaluation were perfomed using detailed finite element analysis for the loading corditions defined in the Westinghouse Design Specification. The design, nomal operiting, steady state and transient cordition were evaluated for both uniform and local grind configurations which are confirmed by the final ground contour configurations. The stress analysis results indicated that the maximum range of the primary and secondary stress intena;ty meets the criteria of the ASME Code,Section III. The faH gue usage evaluation indicates that the upper DCFR plate position would result in the minimum time to reach a usage of 1.0; 19.2 years and 20.6 years for the 1,0" and 0.75" grinds, respectively. Based on actual plant operation to date, the correspending remaining service life of the plant has further been detemined to be at least 32.5 years.

.* - 5-Based on the above calculated fatigue usage life remaining for the plant, we conclude that continued operation of Indian Point Urit 2 is not expected to have any adverse safety related consequences as a result of the girth weld repair.

IV. CONCLUSION Rased on the staff's review of Consolidated Edfson's reports submitted as required in the December 24, 1987 Safety Evaluation, the staff concludes that continued operation of Indian Point 2 is not expected to have any adverse safety consequences as a result of the girth weld repair.

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23 MAY 1988 Docket Nos. 50-247 Consolidated Edison Company of New York, Inc.

ATTN: Mr. Steven Bram Vice President, Nuclear Powar Indian Point Station Broadway and Bleakley Avenue Buchanan, New York 10511 Gentlemen:

Subject:

Systematic Assessment of Licensee Performance (SALP) Report Number 50-247/86-99 This refers to the evaluation of Indian Point Unit 2, operated by the Consolidated Edison company of New York, conducted by the NRC staff on March 17, 1988. This report was forwarded to you on April 4, 1988 and discussed with you in a meetitg held April 12, 1988, in the NRC Region I office, King of Prussia, Pennsylvania.

The list of attendees at the April 12 meeting is attached as Enclosure 1. The NRC SALP Report is attached as Enclosure 2. Our letter of April 4,1988 (Enclosures 3) forwarded the SALP Report and solicited your comments within thirty days of the April 12 meeting. Your letter of May 12, 1988 in response to the SALP Report is attached as Enclosure 4.

Your response of May 12, 1988 has been reviewed; it appears to address the concerns expressed in our report and during the April 12 meeting. No changas to the SALP Report were deemed necessary based on discussions during the manage-ment meeting or review of your written response.

The Indian ., int Unit 2 facility was operated safely, with improved availability when compared with the previous assessment period. Your trip raduction efforts have been particularly effective. However, some common and interrelated problems were idantified in several functional areas. It appears that resources were not effectively managed to reduce existing backlogs in maintenance, licensing and engineering support and to assure implementing of timely corrective actions.

Management should make improvements in these areas and heighten personnel sensitivity to potential safety issues. -

We believe that our meeting and interchange of information was beneficial; future meetings between Region I management and plant management to apprise us of your progress and to discuss matters of mutual interest are encouraged.

OFFICIAL RECORD COPY IP2 SALP - 0001.0.0 05/23/88 eso60totte DR '390523 V(( '\

ADOCK 05000247 DCD LO s -

ar Conso ida ed Edison Company 2 l18 lAAY 1988 No reply to this letter is required. Your actions in response to the NRC SALP i will be reviewed 4s part of the ongoing inspection program at Indian Point Unit 2.

Your cooperation is appreciated.

Sincerely, Original Signed By VILLIAM T. LUSSELL William T. Russell Regional Administrator

Enclosures:

1. List of Attendees

2. SALP Report No. 60-247/86-99 '

3. NRC Letter to Consolidated Edisen, dated April 4, 1988

4. Consolidated Edison Letter to NRC, dated May 12, 1988 cc w/ enclosures:  !

Jude G. De1Percio, Manager, Regulatory Affairs Brent L. Brandenburg, Assistant General Counse!

P. Kokolakis, Director, Nuclear Licensing (NYPA)

Walter Stein, Secretary, NFSC 4

Chairman Zech Commissioner Bernthal Commissioner Carr Commissioner Roberts t Commissioner Rogers K. Abraham, PA0 (16)

INPO, Records Center >

4 Public Document Room (POR) ,

Local Public Document Room (LPOR)

Department of Public Service, Power Division r i Nuclear Safety Information Center (NSIC)

NRC Resident Inspector State of Nrw York

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0FFICIAL RECORD COPY IP2 SALP - 0002.0.0 05/18/88

i Consolidated Edison Company 3 of New York gg bec w/encls:

Region I Docket Room Management Assistant, DRMA (w/o enc 1)

P. Swetland, DRP Section Chief Robert J. Dores, DRSS T. Murley, NRR W. Russell, RI J. A11aii, RI J. Taylor, DEDO D. Holody, RI W. Johnston, DRS T. Martin, DRSS M. 51ossen, NRR J. Lieberman, OE R. Brady, DRP Wishlist Coordfriator B'anch 2 Files

• H. Eichenholz Managment Meeting Attendees RI:DRS RI:DRP Kelley RI:DRP P tlanhWenzinger YP RI:DRA RI:RA Dudjey/rh1 at othe S V Allan Russell fac 5//4/88 5/r /88 //f/88 5/ M/88 5/J3/88 5/ /88 54J/88 0FFICIAL RECORD COPY IP2 SALP - 0003.0.0 05/18/88 i

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ENCLOSURE 1

!.IST OF ATTENDEES - SALP MANAGEMENT MEETING April 12, 1988 N,uclear Regulatory Commission W. Kane, Director, Division of Reactor Projects (DRP)

W. Johnston, Director, Division of Reactor Safety (DRS)

A. Thadant, Deputy Director, DRS E. Wenzinger, Chief, Projects Branch No. 2. DRP R. Bellamy, Chief, Facilities Radiological Safety & Safeguards, Division of Radiation Safety & Safeguards (DRSS) -

W. Pasciak, Chief, Effluents Radiation Protection Section, DRSS P. $ wetland, Chief, Reactor Projects Section 28, DRP L. Rossbach, Senior Resident Inspector P. Kelley, Resident Inspector R. Capra, Director, Project Directorate I-1, NRR M. Slosson Project Manager, NRR Consolidated Edison Company of New York G. McGrath, Executive Vice President M. Selman, Senior Vice President, Control Operations S. Bram, Vice President, Nuclear Power C. Durkin, Vice President, Engineering J. Basile, General Manager, Nuclear Power Generation M. Miele, General Manager, Envir.anmental Health and Safety and Technical Support

. A. Budnick, General Manager, Administrative Services O. Marguglio, Manager, Nuclear Power QA

s. Del Percio, Manager, Regulatory Affairs

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J ENCLOSURE 2-U.S. NUCLEAR REGULATORY COMMISSION REGION I SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE INSPECTION REPORT 50-247/86-99 CONSOLIDATED EDISON COMPANY, INC.

INDIAN POINT STATION - UNIT 2 ASSESSMENT PERIOD - AUGUST 1, 1986 TO FEBRUARY 7, 1988

SALP BOARD DATE - MARCH 17, 1988 8

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TABLE OF CONTENTS Page a

I. INTRO 1UCTION . . . . . . . . . . . . . . . . . . . . . 1 A. Pu. pose and Overview . ........... . . 1 B. SALP Board Members . . . . . . . . . . . . . ,. 1 e II.

CRITERIA . . . . . . . . . . . . . . . . . . . . . . . 3 III.

SUMMARY

OF RESULTS . . . . . . . . . . . . . . . . . . 5 I

, A. Overall Summary . ................. 5 B. Background .................... 7 C. Facility Performance Analysis Summary . ...... 11 D. Unplanned Shutdowns, Plant Trips, and Forced Outages . . . . . . . . . . . . . . . 12 IV. PERFORMANCE ANALYS!$ . . . . . . . . . . . . . . . . . 14 A. Plant Operatiens . . . . . . . . . . . . . . . . . 14 B. Radiological Controls. . . . . . . . . . . . . . . 19 C. Maintenance ............... . . . 23 D. Surveillance . . . . . . . . . . . . . . . ... 25 E. Emergency Preparedness . . . . . . . . . . . . . . 29 F. Security and Safeguards ............. 31 G. Refueling. Outage Management . .......... 35 H. Engineering Support .............. 37 -

I. Licensing Activities . . . . . . . . . . . . . . . 41 J. Training and Qualification Effsettveness. . . . . . . 44 K. Assurance of Quality. ................ 48 V. SUPPORTING DATA AND SUMMARIES ...... ...... 51 A. Investigations, Petitions and Allegations . . . . . 51 B. Escalated Enforcement Actions . . . . . . . . . . . 51 C. Management Conferences .............. 51

0. Licensee Event Reports .............. 52 TABLE 1 - INSPECTION REPORT ACTIVITIES . . . . . . . . . . . . 53 TABLE 2 - INSPECTION HOUR

SUMMARY

,............ 57 TABLE 3 - ENFORCEMENT ACTIVITY . . . . . . . . . . . . . . SB TABLE 4 - LICENSEE EVENT REPORTS . . . . . . . . . . . . . . 62 TABLE 5 -

SUMMARY

OF LICENSING ACTIVITIES ......... 66 1

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I. INTRODUCTION A. Purpose and Overview The Systematic Assessment of Licensee Performance (SALP) is an integrated NRC staff effort to collect the available observations and data on a periodic basis and to evaluate licensee performance based upon this information. The SALP program is supplemental to normal regulatory processes used to ensure compliance to NRC rules and regulations. The SALP program is intended to be sufficiently diagnostic to provide a rational basis for allocating NRC resources and to provide meaningful guidance to the licensee's management to promote the quality and safety of plant operation.

The NRC SALP Board, composed of the staff members listed below, met on March 17, 1988, to review the collection of performance observations and data to assess licensee performance.in accordance

' with the guidance in NRC Manual Chapter 0516 "Systematic Assessment of Licensee Performance." A summary of the guidance and evaluation criteria is provided in Section II of this report.

This report is the SALP Board's assessment of the licensee's safety performance at Indian Point Station, Unit 2 for the period August 1, 1936 through February 7, 1988. It is noted that the suraary findings and totals reflect an 18 month assessment period. The SALP period

' was extended by two months in order to include the a>',essment of activities associated with the steam generator dryou, event that led to the Augmented Inspection Team inspection.

The SALP Board was comprised of the following:

Chairman W. F. Kane, Director, Division of Rea; tor Projects (DRP)

Members R. A. Carra, Director, Project Directorate I-1, NRR S. J. Collins, Deputy Director, DRP (Part time)

R. N. Gallo, Chief, Operations Branch, Division of Reactor Safety (ORS) .

E. C. Wenzinger, Chicf, Projects Branch 2, DRP P. D. Swetland, Chief, Reactor Projects Section (RPS) No. 28, DRP W. J. Pasciak, Chief. Ef fluents Radiation Protection Section, 01 vision of Radiation Safety and Safeguards (ORSS)

L. W. Rossbach, Senior Resident Inspector, Indian Point 2 M. M. Slosson, Project Manager, Project Directorate I-1, NRR

. . . 2 0_ _t he r s P. W. Kelley, Resident inspector, Indian Point 2 C. J. Anderson, Chief, Plant Systems Section, DRS (Part tima)

R. R. Keimig, Chief, Safeguards Section, DRSS (Part time)

W. J. 1.azarus, Chief, Emergency Preparedness Section, DRSS (Part time)

J. E. Richardson, Acting Deputy Director, ORS (Part time)

J. R. Strosnider, Acting Chief, Engineering Branch, DRS (Part time)

5. S. Sherbini, Senior Radiation Specialist, DRSS (Part time) s 4

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11. CRITERIA Licensee performance is assessed in selected functional areas. Functional areas normally represent areas significant to nuclear safety and the environment. ,

One area:

or more of the following evaluation criteria were used to assess each  !

1. Management involvement and control in assuring qua)ity. '

2.

i Approach to resolution of technical issues from a safety standpoint.

J 3. Responsiveness to NRC initiatives. i

4. Enforcement history. P

5.

Operational events (including response to, analysis of, and corrective actions for)

6. Sta'fing (including management),

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7. Training and qualification effecttweness.

Based upon the SALP Board assessment, each functional area evaluated is classified into one of three performance categories. The definitions of these performance categories are:

! t Category 1: Reduced NRC attention may be appropriate. Licensee management attention and involvement are aggressive and oriented toward nuclear safety; licensee resources are ample and effectively used such that a high level of performance with respect to operational safety is being achieved.

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Category 2: HRC attention should be maintained at normal levels. Licensee management attention and involvement are evident and concerned with nuclear safety; licensee resources are adequate and reasonably effective I so that satisfactory performance with respect to operational safety is being achieved. i Category 3: $3th NRC and Itcensee attention should be increased. Licensee

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management attention or involvement is acceptable and considers nuclear safety, but weaknesses are evident; licensee resources appear strained or not ef fectively used such that minimally satisfactory performance with .

t respect to operational safety is being achieved. I r

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' ' The SALP Board may determine to include an appraisal of the performance trend of a functional area. Normally, this perfo.mance trend is only used where both a definite trend of performance is discernible to the Board and theperformance of Board believes level.thatImproving continuatio" of the trend may result in a change :

(dec;'ning) trend is defined as:

Licensee performance was determined to be improving (declining) near the close of the assessment period.

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5 III.

SUMMARY

OF RESULTS A. Overall Sammary Indian '/oint Unit 2 was operated safely, with improved availability wnsn compared with the previous assessment period. Lietnsee performance in many functional areas, however, was inconsistent.

While progress was evident during the assessment period in areas such as trip reduction, auxiliary building decontamination, resolution of emergency plan issues and operator requalification programs, signifi-cant weaknesses were identified in each of the functional areas except security.

Licensee efforts to raduce plant trips have been successful, however, further improvement is this area could be achieved through better procedural control of operational activities and attention to detail by the operstors. Weaknesses in management control of opt.'ations were noted during the startup from the 1987-88 refueling outage.

Improvements in the material condition of the plant were noted although a significant maintenance backlog continues to exist. Our review of unplanned shutdowns indicates that increased attention is needed in the areas of preventive maintenance to improve equipment reliability.

The surveillance testing program has generally been performed satis-factorily, however, weaknesses were identified in the post-maintenance testing and calibration areas.

Our review of radiological controls indicated that many areas of the plant are being successfully decontaminated. The high source term and contamination levels inside containment continue to cause hign integrated radiation exposures. Increased attention is required in the areas of ALARA practices, high radiation area controls, effluent monitoring and chemistry.

Good performance was recognized in the areas of security and outage management. The security programs were notably effective. In the area of outage management, increased attention is needed to reduce the outage backlog and to assure a smooth transition from outage to operational conditions.' The licensee aggressively pursued outstanding emergency preparedness issues to resolution.

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6 Engineering initiatives related to safety system functional assessments and recovery of system design basis records are commendable. Field engineering work is also generally satisfactory. However, documen-tation of the resolution of technical and safety issu15, and control of engineering calculations are significant weaknesses.

Some common, but interrelated problems were identified in several functional areas. These included strained or ineffectively used resources, slow or ineffective corrective actions, excessive backlogs, and insufficient sensitivity to potential safety issues.

Site and corporate management do not appear to have adequately addressed these weaknesses, in particular their en.phasis on the recognition of potential safety issues.

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B. Background 1.

Licensee Activities The assessment period began on August 1, 1986 with the unit operating at 100% power.

On August 2, the unit tripped automatically due to a leaking thrust bearing oil seal which resulted in low bearing oil pressure causing the main boiler feed pump to trip. The operators could not maintain steam generator levels with the remaining main boiler feed pump. The unit was restarted the following day.

On August 6, an Unusual Event was declared due to the loss of the normal and emergency power supplies to a 480 volt safety-related bus due to a fault on the bus. The bus was i

reenergized within one hour and the Unusual Event was terminated.

On September 16, while preparing to perform routine rod exercise tests, the unit was manually tripped after three rods dropped. A safety injection followed the trip due to an erroneous high I steam flow signal coincident with low average reactor coolant temperature. The high steam flow signal was due to the slow response time of the main steam flow instrumentation. No injection occurred because the design does not provide for high-head safety injection capability. The rods dropped due to a poorly designed jumper which was intended to keep an inoperable rod stationary during the test. The unit was restarted the same day.

On October 20, the unit tripped automatically from 100% power due to a loose connection on a reactor protection system relay.

The licensee set up and conducted a p.ogram to tighten electrical connections. Following the trip, a relief valve lifted on the steam supply line for the steam d-tves autt11ary feedwater (TOAN) pump due to an improcerly operating pressure control valve. As a result, the TDA N pump did not operate. One motor-driven (MD)

AFW pump also failed to start. The MOAFV pump tripped due to the overcurrent trip setpoint set too low. The unit was restarted October 23.

On Oct6ber 23, the unit was manually tripped from 38% power following the loss of a main boiler feedwater pump (MBFP),

Prior to the trip, while troubleshooting to determine why the M2FP could not attain full speed, the speed control for the MBFp was placed in manual and lef t there af ter troubleshootir.;

was complete. This troubleshooting activity was not controlled by procedures. During subsequert feed system realignment with the MBFP in manual control, the pump tripped due to high discharge pressure, it was determined that the check valve

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in the discharge line of the other MBFP L . x1 functioned.

Following the trip, the TDAFV pump tripped again due to inadequate corrective actions folleving the Oc;ober 20 trip. The pressure control valve for the TOAFW steam supply was repaired and tested.

Due to the recurrent nature of AFV problems, other concerr,s were developed about the AFN system and were addressed by the licensec.

The unit was restarted on October 26.

On November 6, the unit tripped automatically from 97*. power while performing a surveillance test. The cause of the trip was a me.! functioning f elay in the pressurizer p.* essure circuitry.

Several relays were replaced. No failure mechanism for the relays was determined. The unit was restarted on November 9.

On November 13, while attempting to fill the accumulators using a Safety Injection (SI) pump, the SI pump started, t".!t immediately tripped. The pump would not restart ar:d could not l be turned over by hand. per Technical Specifications, a plant l

shutdown was commenced. The unit reached hot shut down on l November 15 and the pump was replaced. The failure analysis is I

continuing on the failed SI pump. The unit was returned to service en November 17.

On January 21. 1987, an Unusual Event was declared after the f

licensee received information from radiographs that tne flow control valves in the Service Water (SW) system on the discharge of the Emergency Diesel Generators (EDG) were possibly obstructed / shut. A plant shutdown was commenced. Upon further #

review, the licensee determined the valves were, in fact, open.

The shutdown and the Unusual Event were terminated.

On January 30, the unit was shut down for a planned mid-cycle outage to repair the electrical generator end turns. End turn vibration was limiting power generation to 90-95 percent of full

apacity. During the shutdown. two Main Steam Isolation Valves (MSIVs) failed to close due 'o 4:k of shait lubrication. The problems were corrected during she outage. On February 7, the mid-cycle outage ended and the unit v n "extarted. Following the main generator repairs, the ur .n- i le to reach 100t power due to the reduced eed turn *? '

-"L On February 10, tl.a unit was reanually tvipped from 100% power when a licensed operator opened the wrong circuit bretter in the control room. While reenergi ing EDG control power fo'. lowing maintenance, the operator opened the reactor trip breaker control power instead. The unit was rostarted on February 13.

. 1 On June 27, the unit tripped automatically f rem 100*.' power while j attempts were being made to reset the steam generator low level  :

reactor trip setpoints. The cause of the trip was attributed to Taulty relays which were replaced.

No failure mechanism for the relays could be found. While in hot shutdown, the licensee identified that one MSIV would not close. The plant was cooled down to inspect the MSIV. It was determined that the disc stop from the valve body had failed. The unit was restarted on June 30.

On October 4, the licensae shut down the unit to begin the cycle 8 D efueling outage which was expected to last 65 days. Work to be performed included inspecting the previously identified reactor vessel flaw, and installing a new electrical generator.

During the cutage, steam generator girth weld cracks were di overed in all four steam generators.

out The cracks were ground

/he unit was heated above cold shutdcwn on January 1.  ;

From January 2 to 3, 1988, steam generator 23 was allowed to boil dry, while the plant was in hot shutdown and less than 350 degrees F. The steam generator was refilled on January 5 and 6 by c anecting its blowdown line *.o the steam generator 21 blewdown line. A Confirmatory Action Letter (CAL) placed a hold on startup and an Augmented Inspection Team (AIT) was formed.

The CAL hold was lifted on January 18, after the AIT inspection, a management meeting, and completion of augmentea short-term co"rective actions. T% unit was brought critical on January 19.

Jn Jana ry 25, the unit tripped automatically from 16'4 power, then operators increased load too quickly. The unit was returned to power the same day.

On December 1, 1987, management promotions and rotations took place which affected all vice presidential positions related to Indian point Unit 2. These changes incluced the appointment of a new on-site Vice president of Nuclear Power and a new Vice President of Engineering. The on-site Vice President of Nuclear Power new reports to the Senior Vice President of Central Operations, who was the former on-site Vice President of Nuclear Power. Numerous on-site supervisory changes took place at the end of this assessment period. On March 1, 1988, the General Manager r

of Environmental Health and Safety took on the dutie; of the ientral Manager of Technical Support and the Chairman of the Statfor A clear Safety Comittee as a collateral duty.

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2. Jnspection_ Activities

,Two NRC restoent inspectors were assigned during the inspection period. The total NRC inspection effort for the period was $616  ;

hours (3694 annualized) with a distribution in the appraisal functional areas as shown in Table 2 (Inspection Hour Summary).

The average inspection effort for plants in Region I is approximately 3150 hours0.0365 days <br />0.875 hours <br />0.00521 weeks <br />0.0012 months <br /> per year. Tabl9 1 lists every NRC inspection conducted at Indian Poir.t 2 during this period. ,

Ouring the period, NRC team inspections were cor, ducted in the  !

following areas:

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a. Electric power system load distribution (Inspection Report i

247/87-07). '

b. Emergency preparedness partial scale exercise (Inspection  !

Report 247/87-14).

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c. Assessment of the effectiveness of the licensee's quality verification organization (Inspection Report 247/87-16),

d. Assessment of startup activities (Inspection Report  !

247/88-01),

e. Augmented Inspection Team inspection of the steam generator dryout event (Inspection Report 247/88-03). ,

f. Safety System Functional Inspection (Inspection Report L

247/88-200, inspection still in progress at the end of this l assessment persod).

l Inspection Activities are summarized in Tables 1 and 2.

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4 Enforcenient activities are summarized in Table 3. i J

I This report also discusses "Training and Qualification Effec-  !

tiveness" and "Assurance of Quality" as separate functional .

3 areas. Although these topics, in themselves, are assessed in I the other functional areas through their use as criteria, the [

two areas provide a synopsis. ,

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i C. Facility Performance Analysis Summary l ,

Category Category Last Period This Period Functionai A ea 8/17f5 - 7/31786 8/1/86 - T/7738 Trend A. Plant Operations 2 2 Declining B. Radiological Controls '

and Chemistry 2 2 C. Maintenance 2 2 l D. Surveillance 1 2 I E. Emargency l Preparedness 2 2 >

F. Security and I Safeguart)s 2 1 t G. Refueling, Outagt Management 1 1 Declining

H. Engineering Support N/A' 3 I. Licensing Activities 2 2 Declining ;

I l J. Training and i Qualification 2 2  !

i Effectiveness '

4 K. Assurance of Quality 2 2 Declining I

1 L. Fire Protection 1 N/A*  ;

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. r- 12 D. Unplanned shutdowns, plant Trios, and Forced Outages  ;

Functional Orte &_ Power Level Description Cause Area 8/2/8, - 100% Automatic reactor trip Equipment Maintenance on Low steam generator failure level af 6er #22 main (Maintenance ,

boiler feed pump (MBFP) deficiency) tripped due to a previously noted thrust bearing seal leak.

9/16/86 - 200?. Manual reactor trip Inadequate Engineering after three control jumper design rods dropped. A jumper installed to alles rod movement with an in- .

operable rod did not provide the intended function. $1 actuation after the trip due to slow response time of the main steam flow instrumertation. ,

10/20/86 - 100% Automatic reactor Eqvipment Maintenance L trip due to loese failure wire in reactor

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protection circuitry; auxiliary feedwater ,

complications followed trip. An overcurrent protection problem led i to unavailability of

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ene motor-driven (MD)

AFW p9mp. Also the '

j T0AFP tripped.  !

10/23/86 - 38% Manual reactor trip Procedural Operations after MBFP #21 failure

! tripped. MOFP trouble (Procedure ,

i shooting without a non-existent)

! procedure led to the MBFP trip; auxiliary

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' feedwater complications followed the trip due i to inadequate corrective 1 action for the 10/20/86 i failure. i l

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Date & power Level functional Description Cause Area i 11/6/86 - 97% Automatic reactor Random --

trip due to op2n relay equipment contact during failure pressurizer pressure instrument surveillance testing.

11/15/86 Controlled shutdown Cause is Engineering following LCO still under expiration to replace investigation '

a failed $1 pump.

2/10/87 - 100*4 Manual reactor trip. Operator Opera tions During equipment error:

restoration, an inattentiveness operator deenergized '

reactor trip breaker control power by mistake.

6/27/87 - 1004 Automatic reactor Equipment Mainte,ince trip due to open failure: '

relay contacts during Maintenance SG 1evel calibration, program deficiency 1/25/88 - 16% Automatir eactor Operator Operations trip on intermediate error:

range high power inadequate after closing the communication and generator output supervision breakers and increasing performing load too fast. reactor operator duties i

l Note: The root cause in this Table is the opinion of the SALP Soard based on the inspector (s) description of the event; and say, in certain cases, differ from the LER.

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. 34 IV. PERFORMANCE ANALYSIS A. plant Operations (1825 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.944125e-4 months <br />, 32M

1. Analysis The previous SALP determined the operations area to be Category

2. Plant operation was well managed. However, trips were occurring at a high rate. Emergency operating procedures, although successfully implemented, needed to be thoroughly reviewed and updated in a timely manner.

Shortly before the end of the last SALP period, the licensee formed a trip response team. This team investigates each trip and identifies short-term and long-term corrective actions. The licensee's participation in the Westinghouse Owr.er's Group trip reduction sub-committee has also resulted in long-term recommendations. One recommendation implemented - the lowering of steam generator low level setpoints - has already prevented a trip.

The recent trip history shows a significant reduction in tric rate. During the current period the unit was critical for approximately 10218 hours (*26 days) and had eight trips. This gives a trip rate for the current SAlp period of 0.78 trips per 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> critical and is about one third the trip rate of 2.13 from the previous $ ALP (eleven trips in 5175 hours0.0599 days <br />1.438 hours <br />0.00856 weeks <br />0.00197 months <br /> crit' cal).

This improving trip rate was due entirely to better performance in 1987 since trips continued at a high rate during the 1986 portion of this assessment. Although the recent trip history shows a significant reduction in trip rate, the effectiveness of the licensee's trip reduction efforts must be shown by consistent performance over time.

A reactor trip in February 1957, was caused by an operator opening the wrong breaker in D.C. control panel 21 in the control room while clearing isolation from a maintenance job.

This personnel error was due to inattentiveness of the operator and the design and placement of the stop tag. This event was an isolated case, not an indicator of a general problem.

Redesigned stop tags trt now in use, ,

One other trip occurred due to operator error. This trip, which occurreJ in January 1988, was due to increasing lead too quickly, not having good communications between operators, and the senior operator manipu14 ting controls rather than standing back and supervising the startup evolution. Sta- r. trips

• .ilar to this one occurred in 1985, and resulte . the 1teensee improving the simulator modeling of plan unrtup and increasing operator training in normal plant evolutions,

15 including startup. This appears to have had a positive effect because only one such startup trip occurred during the current sal.P period, whereas three occurred in the previous assessment period.

Following the 1988 trip the licensee reviewed this event with all shifts, reemphasizing that operators should increase load slowly.

A trip on October 23, 1986 was due to inadequate planning (no procedure) for a troubleshooting evolutten to determine why #21 main boiler feed pump (MBFP) was not producing expected flow, The pump was not producing expected flow due to a failed check valve. When the evolution was stopped without returning #2)

MBFP to automatic, the resulting feedwater transient caused the only operating MBFP to *. rip. This event was indicative of a lack of procedural controls over off-normal evolutions. The licensee has since established a station policy which requires that such evolutions be controlled by test procedures or temporary operating instructions. .

Control room operators generally conducted themselves in a '

professional manner.

Effective access controls and low not.c levels contribute to a formal atmosphere in the control room.

Shift relief turnovers were effective and included thorough briefings of the relief. However, as demonstrated t,y the steam generator dryout event, board walkdowns and supervisor aurveil'ance of plant status following relief tu.novers was not always effective.

The steam generator dryout event in January 1988 showed significant problems in the control and supervision of operations.

During this event, there was a willingness to <

deviate from approved directives relating to supervisor and operator responsibilities, and operating policies. Most significant were deviatiens from policies requiring that supervisors remain cognizant of plant status, that operators properly notify supervisurs of abnormal plant conditions and activities, and that operatirg procedures be followed. Startup evolutions were not ef fectively planned and staffed to pr1 vent them f*om distracting supervisors and operators from their I primary responsibilities. These poor practices were not limited i

to any single individual or shift, and existed over several days.

This represents ineffective site management control,of operations. i Other weaknesses were also evidens in the supervision of tne operating staff. These in:1uded: excessive overtime in  :

violation of Technical Specifications, not properly certifying i licensed operators prior to their resuming licensed duties after extended absences, inadequate watchstanding and logkeeping

o 16 practices among some non-licensed operators, and isolated instances of sleeping or inattensiveness of non-licensed operators. The licensee took prompt and effective corrective action when NRC informed them of alleged sleeping by non-licensed -

operators. The licensee's increased site management and super- '

visory attention to this matter identified the few instances found. Effective corrective actions have been taken for these items. These are all supervisory issues, however, and increase -

site management attention to assure adequate staff supervision I is still warranted.

An Emergency Operating Procedure (EOP) background document was misapplied during the steam generator dryout event, demonstrating a need 4r adititional training in the applicability of E0Ps.

The dryout event also showed the need for additional training on equipment life cycles including steam generators. Other than I

these concerns, the operators demonstrated through effective use -

' of the E0Ps that they are well trained in their.use. Previous concerns that the E0Ps were not being maintained have been resolved with an effective procedure maintenance program.

Near the end of this assessment period, in response to the ,

dryout event, the licensee created an operational pienning group.

Our initial review of this group shows that it has been effective L in identifying, for operators, contingencies and precautions ,

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' needed when equipment is taken out of service or placed in tut.

The shift technical advisor (STA) program also underwent a major  :

change near the end of this assessment period, in response to the dryout event. STAS are now full-time members of the shift and perform required plant tours and control room walkdowns. This is viewed as a good initiative although it raises staffing

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a concerns as discussed below, There are 55 licensed operators; however, many of these are assig..ed to support functions in the operations department or

' work in other departments. While this use of operators has strengthened those areas, it has lef t the operating staff with only minimum statfing. Sickness and unplanned absences result in excessive use of overtime. Supervisors should be attentive to fatigue problems and to overtime scheduling with the 12-hour shift. Staf fing the operattor.al planning effort and on-shift STA positions, near the end of this assessment peripd, has returned scme licensed operators to the operations department, but it has in turn created vacancies elsewhere. No new resources have been added to fill these vacancies.

A violation was issued due to an improper 11M up of the emergency diesel generator starting air system. Operations 1

generally maintains strict control over system lineups and this is considered an isolated cast.

17 A violation was issued due to bypassing an instrument air check valve. This valve was the system's safety class baundary.

Operations review of the procedure change which bypassed this valve did not identify the need for a 10 CFR 50.59 safety review. No other examples of inadequate screening for 10 CFR 50.59 applicability were identified. Another violation was issued due to the Station Nuclear Safety Committee not promptly reviewing a post-trip review report. Corrective actions in this area have not yet been reviewed.

The licensee continues to maintain an effective fire protection and prevention program. Fire protection tystems, equipment and fire barriers are properly maintained. L:abustible materials are controlled so that they do not present a safety hazard. The licensee's fire brigade training program is generally adequate, well defined a d implemented in accordance with the intent of 10 CFR 50, Appendix R. Fire protection audits were generally complete and the audit findings were being resolved in a timely and satisfactory manner.

In the area of housekeeping, the Itcensee has taken strong corrective actions and as a rt., ult, the plant is noticeably cleaner. For example, it is ..ow possible to enter many pre-viously contaminated areas in the Primary Auxiliary Building witnout wearing anti-contamination clothing. However, in spite of past housekeeping deficiencies, there was a recurrent problem involving inadequate control of gas cylinders and trash observed in cable trays. Corrective action for the first i

deficiency consisted of upgraded training and an improved house- i neeping procedure. Failure to use the housekeeping procedure checklist, resulted in a second violation. Subsequent corrective actions in this area were found to be effective.

In summary, the trip rate is much improved. Housekeeping upgrades were evident and fire protection measures were well maintained. Significant weaknesses were evident in personnel supervision, control of operations, and adherence to policies and procedures. These proolems were particularly evident during the startup from the outage at the end of this period. Staf."ing levels appear to be strained.

2. Conclusion .

Rating: Category 2 Trend: Declinir.g

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3. board Recom.mendations Licensee: -

Develop a writtaa policy regarding the philosophy of facility operations, emphasizing

' control of operations and procedura development i and adherence. Train the staff on this policy.  !

Review the adequac;t of staffing levels. ,

NRC: None I

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S. F)diological Controls (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br />, 8*4)

1. Analysis This area was rated Category 2 last assessment period. Weaknesses were identified last period in the areas of contractor to:hnician training, high radiation area controls, procedure compliance, implementation of Radiological Occurrerce Reports (RORs), and ,

whole body counting. One other c.oncern identified last period was that total person-rem exposure was found to be above the national average for pressurized water reactors which was attributed to higher than average source ters and less than optimum equipment shield design.

Radiation Protection The licensee presantly has a well defined, adequately staffed radiation protection organization. The organization was approp-riately augmented with trained and qualified contractor personnel to support outage activities. The contractor radiation prcte: tion training program was upgraded to address weaknesses identified by the NRC during the last assessment period. The upgraced program enhanced the understanding of work, and contributed to adherence to procedures. There was a modest number of personnel errors.

In general, the iwplementation of external exposure controls was adequate. Consistent posting of radiologically controlled areas was observed. Radiation Work Permit (RWP) use helped maintain adequate control of work evolutions. The effectiveness and use of radiological occurrence reports was improved by the addition of root cause analysis and documentation of corrective actions.

These analyses have, in some cases, resulted in rewriting pro-cedures and issuing new management directives. The two violat. ns identified in this area, high radiation area control and improper release of contaminated materials from the site, were immediately investigated by the licensee and corrective actions were pron.pt and effective.

A loss of control of radioactive material resulted in inadvertant removal of material to an unrestricted area off-site. Three scaf folding timbers, contaminated with small quantities of Cobalt-58 were removed from *.ae containment equipment hatch area, placed in a clean area dumpster and sent to a local incinerator plant for disposal. The incinerator plant idertified the radio-active material with their own monitoring equipment, and sent the material back to Indian Point 2. The root cause of this incident was a lack of appropriate administrative controls and surveillance at an unsecured access potat to the radiologically contro' led area. The licensee instituted some irsediate cor-rective actions to monitor this identified release pp.th for radioactive materials. However, the pctential for inadvertent

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release of radioactive material from other 'Athways had not been fully resolved at the end of this assessment period. '

The licensee adequately addressed previous weaknesses identified '

by .he NRC during the last assessment period n the internal  ;

ex u sure control program. Improvements had been made in the whole body counting facility, including the purchase of a new '

whole body counter. Bicassay results and cir sample analyses indicated that no significant uptakes of radioactive materials occurred during the assessment period. Improvements in the self-contained breathing apparatus maintenance and control program indicated a technically sound and thorough approach to resolving previously identified deficiences in this area.

Although rteent licensee initiatives have enhanced the quality of the ALARA proDram, with aggressive decontamination efforts in f the PAC and the acquisition of aedio visual equipment to minimtre the number of personnel entries into radiation areas, the tctal station exposure continues to be high (975 person-rem). Licensee goals for collective population exposures were realistic. However, a high source term and contamination levels continue to exist in the Vapor Containme it (VC). Further, extensive work was planned

• and under-taken during tae last outage, including steam generator eddy current testing an6 tube plugging, sludge lancing, girth weld repairs, refueling evolutions, containment fan cooler unit replacement, ard snubber and velve maintenance activities. How-ever, the licensee has no immediate plans to idduce the high source term or contamination levels in the VC. Due to the con-tinuing high level of contam'aation in high dose rate areas in the plant, respiratory prot'etion continued to be required.

Extensive use of respira w ry .otection equipment increases the time nacessary for work eyelutione and results in higher personnel exposure. Since there is no significant reduction in VC work activities forseen by the licensee in the near future, saurce term reduction and VC deontamination are necessary in or';er to  !

I achieve significant overall exposure reductions.

Radioactive Waste Manacement/Effluene Centrol Program controls related to the waste processing, management and classification areas were well stated and implemented. Respon-sibilities and authorities for radwaste management were established.

There is significant awareness of radwaste reduction at the site.

Staffing in this area is adequate. Aggressive efforts to reduce radioactive waste generation through material control, waste segregation and supercompaction were noted. The radioactive gaseous and liquid effluents control program was found adequate.

The program's effectiveness was reduced by a lack of canagerent

! involvement in several areas. Specifically, the Offsite Oose

, Calculation Method (00CM) needed corrections, dose factors had

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not been updated to reflect current operating conditions, timely reviews of effluent control procedures had not been made, and samples for liquid s* fluent gross alpha activity had not been sent to the vendor for analysis. These suggest inattention to ,

detail by the site management. Important initiatives being taken include a pregram to significantly upgrade monitoring capibili-ties by the purchase of new skids for effluent control. Audits in the effluents area were thorough and department responses were timely as evidenced by NRC review of audit findings.

An inspection of the licensee's radioanalytical laboratory analysis capabtitty determined that the Itcansee's program was  ;

generally satisfactory, although.some weaknesses were noted.

The weaknesses were problems with resolving interlaboratory

' comparison and inappropriate QC data review. Staffing and procedures in the radiochemistry laboratory were generally adequite. Evidence of good responses to internal audits was apparent.

Transportation '

i program controls related to transportation were well stated and  !

implemented. Review of ongoing transportation activities '

indicated adherence to shipping requirements and licensee procedures, maintaining an effective transportation program, l with one exception. As discussed earlier, a trash shipment t i

containing surveys.

contaminated lumber was shipped offsite without proper l Close HP management attention to implementation of  :

corrective action was noted.  !

Chemistry The 'icensee's capability to monitor non-radiological chemical  !

parameters in various p1 2nt systems was reviewed. There were v some equipment limitations for which the licensee had already initiated upgrades, personnel performance, qualifications and ,

staf fing was adequate. The results of the standard measurements comparison showed five out of thirty-five disagreements. The ,

i causes related to weaknesses in instrument control, sampling i error and no use of independent standard solutions. Other weaknesses were identified involving limited use of control i charts resulting in systemmatic biases in several afeas, and use '

of single point calibrations. These weaknesses were communicated to the licensee during previous inspections. The probiens identified indicate that inadequate site management oversight is  !

being exercised, as well u poor follow-up to NRC identified concerns.

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, , - - . - . , , . . , . ~ -

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• 22 4

Summary In summary, the licensee continues to implement an acceptable  ;

radiological controls program. Strengths were noted in the aggressive decontamination efforts in the surillary building and in tie effort to reduce radioactive ve11v. However, problems ,

continue in adberence to procedures, high radiation area centrols, high radiation exposures, and inattention to details. Vaaknesses  ;

were also noted in the quality and oversight of effluent monitoring l and chemistry program activities.

2. Conclusinn  !

Rating: Category 2  !

Trend: Ncne

3. Board Recommendations '

Licensee: Assess the reasons for continued weaknesses in chemistry. Reassess available corrective I l

actions to reduce overall radiation exposure l goals. '

NRC: Ncne i

i I

[

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I

+ -__ _ - _ . _ , - - , , - . - . . . _ _ _ _ - - - - _ , , . . , _ . - . _ _ - . . - _ - - --_-_..__ . - .

. 23 C. Maintenance (657 hours0.0076 days <br />0.183 hours <br />0.00109 weeks <br />2.499885e-4 months <br />. 12%)

1. Analysis The previous SAlp rated the maintenance area as Category 2.

Maintenance was being performed satisfactcrily by competent and skilled personnel; however, a considerable backlog of maintenance existed. Additional staff and improved efficiency was recommended to help work off this backlog. The use of probabilistic risk assessment in prioritizing work wss encouraged, increased attention to post-maintenance job site restcration was needed.

Maintenance personnel are experienced and well trained. All of the licensee's maintenance training programs are now INPO accredited. Several licensed operators were tamnorarily trans-ferred to the maintenance depar* ment, in o!cer to broaden the understanding of operational problems in the department, and to t work or, special projects. The licensee has started an area raintenance program. This program assigns a senior staff member to one area of the plant. This worker is responsible for locating and reporting faulty conditions in his area. This worker also repairs minor items or leads a tecm to repair numerous items. This area maintenance concept has been used ef fectively in the auxiliary feedwater building and the diesel generator building.

The maintenance backlog continues to be of concern. While the I&C backlog has decreased over the assessment period, the mechanical maintenance bachlog has decreased only slightly.

Unrepaired equipment is a concern because it could hamper the operators in managing a situation due to equipment being unavailable or in a degraded condition. Increased resources should be considered by the licensee in order to further reduce this backlog. Maintenance management is very capable and committed to improving the material condition of the plant.

Si;e and corporate management should continue to support mainter.ance and efforts tu reduce the maintenance backlog and toprove t he material condition of the plant.

The material condition of the nlant is improving as. evidenced by fewer reactor trips, less t1se in Limiting Condition for Operation (LCO) action statraents, and fever control room annunciators contini.susly eiergized. The deficiency tagging program is being used more effectively. This has made the effort to keep the work oroer backing down more difficult, but it also appears to have reduced the unidentified maintenance backlog. Chronic problems are being fixed with long-term l

solutions, such as the rel acement of leaking boric acid transfer pumps with new pimps. Improved packi.'g is baing installed on many valves, and 1&C is replacing aging electrical

24 components when possible. Improved spare parts control contributed to the large volume of work activities whkh Ws accomplished during this assessment period.

Probabilistic Risk Assessment (PRA) is being utilized in a qualitative manner in the planning and performance of corrective maintenance. More personnel are aware of PRA and its importance in getting certain safety-related equipment back in service quickly. PRA insights are not having Any apparent impact on the prever.tive maintenance program. Increased use of PRA is delayed '

by shortages in engineering staff as discussed in the Engineerir.g Support section of this report.

The licensee is continuing to expand its use of the computerized power Plant Mainten ace Information System (PPMIS). An eztensive Operating Equipment Menu (OEM) is being put in PPMIS that would allow component level equipment tr. be tracked. The coaponent's manufJcturer, stock number, purchase order, and ,

other identifying features allows planners to accurately know what is in the field. This should improve the planner's ef fectiveness and material control, and appears to be a good hittative by the licensee.

During thi3 period, a reactor trip was associated with untimely v

caintenance. The trip occurred due to oil leaking from a worn oil seal ring on #22 main boiler feed pump thrust bearing. The resulting low bearing oil pressure tripped the pump. This seal had been leakins for some time. An aggressive caintenance program for the MBFP could have prevented the trip.

Th)re have been complications after reactor trips att-ibutable to maintenance, such es Main Steam Isolation Valves (MSIVs) failing to close due to lack of shaft lubrication. This was due to the lubrication procedure being inadvertently omittad from the preventive maintenance procedure. Two reactor trips occurred due to f aulty relays and one reactor trip occurred duc to loo:e electrical connections. The faulted relays were sent to a failures analysis laboratory, but no failure mechanism could ,

I be identified. The loose electri:a1 connections occurred in the reactor protective circuitry. The licensee checked other i circuits fo11cwing these events and found other steilar problems i and corrected them. However, the lack of such vsrifications on

• a routine basis f*cicated a weakness in preventive maintenance in this area. As a result, preventive maintenance (FM) packages were developed to check the tightness of wire terminals and to i replace or check safety-related relays.

While performing this .elay PM with the plant in cold shutdown, a loss of all 4 0 volt power occurred for several minutes. This 7 was due to inacequate instructions in the PM, and poor planning and control of this evolution by supervision. This is not a 1

_ _ _ _ . _ _ , , . . ~ _ , , -. .

+

25 general problem. The majority of maintenance activities observed by NRC were well supervised, satisfactorily performed and adequately documented.

Some instances of ineffective corrective actions have been identified. The licensee's QA/QC organization had identified several instances of documentation problems in I&C trouble-shooting work packages. The NRC also identified the same findings during the loose wire inspection. As a result QA/QC is increasing its surveillance of !&C troubleshooting. It took two months to lubricate main steam isolation valves af ter QA  :

pointed out that they were overdue. A Bulletin commitment to prepare procedures to maintain motor-operated valve switch settings was not net. Followup on identified problems needs to be teproved.

In summary, the skill and knowledge of the maintenance

- department is high. Most maintenance activities are completed satsifactorily. The mati.fal cendition of the p1t r.t is i 1.mproving, hcwever, the size of the work backlog is of eencern be:ause it af fects the operators response caca'Jility, burdens site management with reprioritization, and inhibits the ability to do an ef fective predictive / preventive mairitenance orogram.

Increased resources should be considered by the licensee to complete the identified work. More attention to preventive and 1 predictive maintenancs is needed to reduce plant trips and irprove equipment reliability. More timely teplementation of corrective actions is nceded.

2. Conclusion Rating: Category 2 Trend: None

3. Beard Recommendations t

Licensee: Increase efforts to reduce the maintenance  !

backlog, includthg review of staffing levels and work prioritiration. (Repeat recomtendation from ,

previous 5 ALP). l NRC: None i

- - _ _ _ . ._____. -_ _ -_ - . ~ _ _ - _ _ _ _ _

26 sP D. Surveillance (743 hours0.0086 days <br />0.206 hours <br />0.00123 weeks <br />2.827115e-4 months <br /> 13*)

1. Analysis The previous SALP determined performance in this area to be Category 1. Surveillance t.ests were well written and easy ti, fcilow. Completion of surveillance tests was timely and reviews of completed tests were effective in identifying pro *vlem areas.

Staffing was adequate and the program was effectively managed.

The licensee's surveillance program is generally technically adequate and appropriately reviewed. Surveillance procederes included adequate detail. NRC review of the Testing and Performance Group (T&P) activities indicated that the section is maintaining an objective and independent approach to testing and test results. It was noted that T&P personnel involved with maintenance and operations were well qualified and frequently had direct experience in the areas in which they were involved.

The licensee performed approximately 6240 Technical Specifica-tion surveillance tests during the SALP period. No trips were caused by surveillance tests. None of these tests were missed.

This shows that the licensee has effective controls over test scheduling. Two tests required by Technical Specification action statements were performed late and were idtntified by the licensee. These were considered to be isolated cases. The licensee requested and received an extension of the surveillance period for several refueling st.rveillances. This occurred despite previous extension requests being characterized as "one-time only" exte:siens.

The ultrasonic exanmation of a Reactor Pressure Yessel flaw area showed evidenc of long range planning by the associated personnel, Planning of the technical aspects of the examination was thorough, and the activities at the site appeared to be well coordinated among the various s.*oups, Management (semitment to this 15! activitity was demonstrated through the use of several different ultrasonic testing systems including state of the art techniques. The steam generator tube and girth weld ISI programs and secondary ripe inspections were also well CondVCted.

A special NRC review was conducted to assess the adequacy of the girth repairs for the four steam generators. Independent reasurements conducted by NRC personnel confireed the Itcensee's NDE inspection results. Managewent and staff demonstrated a thorough understanding of the technical issues surrounding the

,7, .

l

, 27 repairs. The licensee personnel performing the various inspec-tions were experienced, and the overall inspection plan was technically sound.  !

Several examples of inadequate post-modification testing were i identified and indicated a weakness in this area. Esamples t included testing following modifications of a p-essure control valve in the auxiliary feedwater system, a service water flow indicator in the emergency diesel generator system, a dryer in -

the instrument air system, and a change in the setting of the '

steam driven auxiliary feedwater pump speed changer. The pro)1 ems identified were that adequate test requirements were not specified and tests were not verified to be complete before '

returning equipment to service. Also, some responses to the  !

i violations cited were inadequate, in that they did not describe  ;

the corrective steps which had been taken, the results achieved, j and the date when full compliance would be achieved. One response also containcd inaccurate information, fhese problems _

were subsequently resolved. l l

t A backlog of corrective actions developed in the testing program.  !

This included the need to provide an adequate channel i l chech program, the incorporation of tests by the I&C department i into the more formal surveillance test program the resolution i i of a technical issue holding up the instrument air post-modif t- i cation test mentioned in the preceding paragraph, and completion  ;

)

of a record of plant transients and operating cycles. This .

backlog illustrates the need for better site management control 4

of corrective actions and indicates the need for additional staff. 4 I

j A staffing weakness was also apparent due to the absence of one q j person who esintained the test schedule and an open staff t

' position during part of this assessment period. The I&C PM l superviser position was also vacant until near the end of this l

1 SMP period. The test and performance engineer and the experi- r 1

enced engineer who temporarily replaced him both lef t the test l

program, near the end of this assessment period. The effect t this loss of technical experience will have on the test program i

- is of concern. Management attention should be direi:ted toward

• 1 assuring adequate resource > to accomplish the test program, i Additional resource problems were evident in the calibration  !

program. The licensee conducts many safety related tests and  !

4 calibrations using the loss formal I&C preventive maintenance

[

(PM) program instead of the more formal surveillance program, t

! Examples were identified where calibrations in the PM program were not completed on schedule. These calibrations are not  !

i  !

formally tracked and receive low priority. The if censee is now  ;

! applying more resources and is developing a tracking system for this program.  !

I f

I

. ,' /

28 Generally T&P personnel were sensitive to abnormal conditions identif'ed during surycillance activities. NRC review of the surveillance and calibration program resulted in the identifi-cation.of two problems: the failute to include acceptance limits for battery electrolyte temperature, and the failure to ensure that battery room temperature is kept within normal bounds. Thase findings led to initiation of a plant shutdown af ter Battery #21 was declared inoperable due to low electrolyte temperature. Neither the technicians nor the engineers in the T&P group were aware of the effect that temperatures have on the operation of the battery system even though the site's battery manual discusses this subject.

In summary, the surveillance testing, calibration, and inserytce inspection programs have generally been performed satisfactorily. Weaknesses were evident in post modification testing, calibration PMs and the corrective action backlog.

Staffing levels appear to be strained. Manageme.nt involvement will be needed to resolve this backlog and prevent reoccurrence of post-maintenance testing problems.

2. Conclusion Rating: Category 2 Trend: Nont

3. Board Recomendations Licensee: Review staffing levels to assure vacancies are filled, work backlog is reduced, and resources are adequate to accomplish the test and calibration program.

NPC: None

F

., 29 E. Emergency Preparedness (159 hours0.00184 days <br />0.0442 hours <br />2.628968e-4 weeks <br />6.04995e-5 months <br /> 3'4)

1. Analysis Licensee performance in this area was rated as Category 2 during the previous assessment period due to the need for more direct involvement in resolviag offsite deficiencies.

During the current assessment period, one emergency exercise was observed and an unannounced routine safety inspection was conducted. The emergency exercise, conducted on May 13, 1987, demonstrated particular strengths in the organization, manage-ment and control of emergency facilities, working knowledge of the plant, and the control roem operator's ability to mitigate further plant degradation. However, a significant weakness was observed in that the Emergency Director in the Control Room failed to classify both the Unusual Event and Alert emergency conditions. This weakness was similar to those in two previous exercises.

This weakness is now being addressed aggressively by the licensee. An interim measure has been instituted to correct ambiguity in the Emergency Action Level (EAL) tables. The final corrective measures will include redesign of the EAL tables and the integration of EAL categories in station procedures and technical specifications. Additionally, a computer system is being developed to further assist control room operators in classification of EALs. Licensee performance in the 1983 emergency exercise will demonstrate the effectiveness of improvements in this area.

The emergency response facilities were found to be adequate; however, in response to NRC concerns, Emergency Operations Factitty (EOF) expansion is being planned with construction expected to begin in early summer 1988. Other ai'as of the licensee's program were examined and were found to be acceptable with only minor weaknesses which were promptly addressed.

The licerset exh bited extraordinary efforts to mediate and successfully resolee offsite issues regarding the protection of school children as ra hed in the New York Public Interest Research Group (NYp!RG) p3tition. The licensee took the initiativt by maintaining close liaison with the State, affected counties and FEMA while seeking an acceptable resolution to 151ues regarding reception centers, training of bus drivers.

transportation resources, agreements, and public information.

Once these issues had been spe.-ifically defined, they were adequately corrected by the State and affected countie! with the lic:n e's assistance. During the process of resolving these issue the protection of publi: health and safety was maintained.

4 30 I In summary, the Itcensee has been prompt ar.d responsive in I dealing with identified concerns in the Emergency Preparedness area. There has been continued improvement throughout this assessment period. However, previous corrective actions to '

resolve recurrent weaknesses in classifying emergency conditions (

were inadequate as demonstrated by 1987 exercise findings.  !

f

2. Conclusion t

Rating: Category 2 Trend: None

3. Board Recomendations  !

Licensee; None I NRC: None  !

t F

i t

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F. Security and Safeovards (231 hours0.00267 days <br />0.0642 hours <br />3.819444e-4 weeks <br />8.78955e-5 months <br />. 4*.)

t 1. Analysis ,

During the previous SALP period, the licensee's performance in this area was Category 2. This rating was influenced by  ;

programmatic problems in the security program that resulted in an enforcement conference about midway through the period. l l

The li'ansee committed additional resources and management  ;

attention to upgrade and strengthen the acurity program, i Improvements were noted during the remainder of that assessment  !

period; however, the NRC did not conduct another region-based t security inspection prior to the end of the assessment period, i j so the overall effectiveness of the licensee's efforts to i improve the program was not evaluated.

(

3 4

In this assessment period, the Itcensee's security and j

• safeguards program was inspected during three region-based  ;

routine physical security inspections, one material control and  !

I

' accounting inspection, and continuing inspections by the  !

resident inspectors. No violations of NRC requirements were identified, t

t The security management organization, under a newly established (

licensee Program Manager, was providing positive and effective r

' oversight of the security force contractors and direction for i the security program. The licensee's management organization l 1

now includes proprietary security shift supervisors who. provide l i

around-the-clock oversight of program implementation. Corporate {

Security Mariagement continued to be actively involved in all  ;

i I security program matters, with site visits to perform program l appraisals and te provide assistance in the planning process for

' program and equipment modifications and upgrades. "hese .

initiatives go beyond NRC regulatory requirements. Security management personnel are also actively involved in the Region I f i

4 Nuclear Security Association, and other industry groups engage <' f in nuclear plant security matters, j

c During one routine NRC inspection, a potential morale problem l was identified in the security organization resulting from <

members of the contract sesurity force possibly losing l employment benefits when the security force contract was awarded to a new company. When the problem was called to the licenste's {

l attention by NRC, the matter was aggressively pursued with the '

new contractor and resolved. The initiatives taken to j strengthen management oversight of the security program and to t eggressively pursue potential problems, demonstrate noteworthy  ;

program support and attention from site and corporate j 4

management. ,

i t

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32 The licensee has developed a self-appraisal program, based on both compliance and procram performance objectives. This is in addition to the NRC required annual security program audit. The self-appraisal program is carried out by on-site corporate security representatives on a continuing basis, and provides an effective method for monitoring program implementation, security force performance, and providing feedback for overall program (nhancements. The program findings are reviewed at the corpo-rate level and forwarded to site security management for appropriate action. This initiative is also indicative of the licensee's desire to implement a highly effective security program that goes beyond complianct with NRC regulatory requirements. It is partially responsible for the excellent enforcement history during this evaluation period.

The licensee submitted five security event reports in accordance with 10 CFR 73.71 during the assessment period. One report resulted from an equipment malfunction, another resulted from failure of a plant employee to follow a security procedre. A third report resulted from the licensee's identification of ir. adequate background investigations by contractors, and another resulted from a non-credible bomb threat. The fifth was to notify the NRC of the establishment of a picket line on the main entry road to the plant in April,1987. All of the events were promptly reported and the decementation was sufficiently com-prehensive to permit NRC analysis without the need for additional information. Correctiva actions and compensatory measures were prcmptly implemented. These actions demonstrate appropriate management attention to this program area.

Members of the security force exhibited a good appearance and a professional demeanor throughout the assessment period. Staff-ing af the security organization is adequate as evidenced by a minimum use of overtime during plant operations. The security officer training and requalification program is well-developed and administered by three full-time instructors and one part-time instructor. The licensee revised the Training and Qualification Plan and completed an upgrade of the training progree, which included devel @ ent of formalized lesson plans in an offort to update and improve the program. This is another initiative that indicates the licensee's desire to implement an effective program.

Review of the licensee's maintenance supprt for security equipment during this period revealed it to be generally much improved over the prior assessment periods. However, one instance was identified by the NRC Aere compensatory measures were employed for an extended pertoa in lieu of a needed system

~

yi

[ 33 improvement. When the potential liat'ilities of extended com-pensatory mecsures were called to the licensee's attention, immediate short-term action to correct the problem was taken and an engineering review for long-term resolution was initiated.

Security facilities and spaces were adequate and well maintained. Records were readily retrievable, complete and centrally located for ease of use. These attributes demonstrate an appropriate understanding of program needs and, again, reflect the licensee's desire to implement an effective program.

' The material control and acciurti.ag inspection conducted during this period identified no deficiencies. However, the licensee  ;

had prev %usly reported that four fission counters in storage could not be accounted for. This was investigated by the licensee and it was determined that the counters were most likely dispesed of inadvertently as part of a reutine low-level radioactive waste shipment sent to a licensed disposal site.

The licensee promptly initiated cost *5ensive actions to preclude recurrence, demonstrating an intent to implement an effective program in this safeguards area. l The prior assessment report noted that licensee security representatives, on their own initiative, visited ths Region I offict to discuss major revisions to their security program t

plans for the purpose of imoroving the quC ity and to facilitate implementation. During this period, the Itce 3ee submitted ,

major revisions to all security program plar.., i.e., Security,  !

Contingency, and Training and Qualification. An explanatter. of each change was provided, together wig e cross reference to facilitate NRC review. These revision; represented an extenst*<e effort on the part of the licensee and resulted in a high quality product. The revised plans are reflective of the .

i current physical security program and equipment at the plant, ano l are structured in an orderly fashion to facilitate  ;

i implementation. Implementing procedures for the plans were also revised and were found to be much clearer and more conciss.

In sumary, the licensee's acticas 60 upgrade and strengthen the security program, have resulted in the establishment of a very ef fective program that goes beyond compliance with' regulatory l requirements and a curity plan commitments. The initiatives taken are noteworthy and indit.ative of the licensee's n.ention to implement a high quality and effective security program.

l t

. 34

2. Conclusion Rating: Category 1 Trend. None

3. Board Recommendations Licensee: N0ne NRC: None

{

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35 G. Refueling. Outage Management - (404 hours0.00468 days <br />0.112 hours <br />6.679894e-4 weeks <br />1.53722e-4 months <br />, 70)

1. Analysis The previous SAlp included Engineering Support within this area and rated this area as Category 1. During that assessment period improvements in outage management were evident and resulted in effective outage control. itartup testing was performed well. The implementation of Environmental Qualification requirements was effective.

During this assessment period, two scheduled outages were conducted: a mid-cycle outage to inspect the sain generator from January 30, 1987 to February 7, 1987 and the unit's eighth refueling outage from October 5, 1987 to January 1, 1988.

The licensee made effective use of the mid-cycle outage to improve the plant. In addition to the main generator work, the licensee reseated 40 main steam valves. The licensee made major plant improvements curing the refueling outage including replacing the fan cooler units, the main generator, and nine feedwater heaters. M:ny smaller maintenance activities were also undertaken including the repair of approximately 250 valves and the clearing of many temporary repairs. Despite all of these improvements, a large backlog of outage work still exists.

Following the 1936 refueling outage NRC inspectior.s identified several instances where safety-related work was deferred from the outage. Examples include the RHR pump power supply fire protection modifications and prever+ 4ve maintenance of the 480 volt switchgear and the 23 AFV pump motor. Continued management attention is needed to assure that safety-related outage work receives appropriate priority.

The strong outage management noted in the previous SALP has continued. Extensive pre-planning enabled management to focus their attention on major activities and unanticipated problems.

The licensee establishes an organization structure specifically for the outage. Although the outage organization worked well, a significant weakness became apparent in the transition from the refueling outage to operations. Unfinished outage. work became a burden on operations and resulted in heating up the plant without a normal supply of feedwater to two steam generators (SGs),

Irrprovements in planning and scheduling should be considered to assure that outage work is completed before startup. An operations planning group has been formed to address activities in tht!

area.

n

. a . .

36 The licensee's outage organization made good use of experienced staff and licensed operators to manage contractors and major projects. They have also continued the effective use of outage coordinators for each department, major project and work area.

These coordinators resolve issues and provide oversight, coordination, and feedback to upper management.

i Outage meetings were held before each shift to track critical path work, ensure adequate support for all tasks, and resolve any coordinatisn issues which the staff had not been able to resolve. Communications between departments, contractors and outage management was good during the outage itself. However, poor communication between operations and the outage organization contributed to the steam generator dryout event during plant ctartup.

Ceordination of modifications work has been improved by holding a pre-implementation meeting fur each modification. All groups who have responsibilities related to the modification attend this meeting and address the tasks for which they are responsible.

i Responsib111 ties for tracking modification completion have also been bettor defined. These improvements should prevent recurrence of two NRC identified problems: partially completed modifications not being centrolled per procedure, and preventive maintenance programs not being revised when equipment was modified.

In summary, outages continue to be well planned, coordinated, and managed. Continued management attention is needed to reduce the outage work backlog and assure that safety-related work is not deferred needlessly. Management attention is also needed to assure a smooth transition from major outages to operations.

Problems in this area ' led to the steam generator dryout event at the end of this assessment period.

2. Conclusion Rating: Category 1

, Trend: Declining

3. Board Recommendations ,

Licensee: None NRC: None i

S 37 H. Engineering Surport (506 hours0.00586 days <br />0.141 hours <br />8.366402e-4 weeks <br />1.92533e-4 months <br />. 9t_)

1. Analysis The previcus SALP did not assess engineer thg .s a separate functional area.

Four inspections were conducted during the SALP period which addressed engineering activities - two Environmental Quali-fication (EQ) inspections, an inspection of electric load distribution, and an inspection of motor-operated valve switch settings. A Safety System Functional Inspection, an Augmented In?.pection Team, and an Assessment of Quality Verification Effectiveness also reviewed this area, in part. Engineering support was also under routine coverage by the resident inspectors.

' Engineering involvement in assuring the quality of modifica-tions, especially recent ones, was evident. Corporate manage-eent includ.ng engir.eering support personnel were heavily involved in the implementation of all me'.'ifications. All mod i fica t io.15, includtag those proposed by site personnel are passed through the corporate engineering group to ensure that all design, procurement, and installation requirements are specified and adequately reviewed. The safety evaluations of these modifications were found to be adequate in most cases.

One exception was noted in that service water pipe repairs used methovs not sanctiered by the ASME code.

Selection el a pressure switch with an inadequate current rating resulted in an auxiliary /eeA4ter pump being out of service 1

during startup in 1988. (his c'ntributed to drying out a steam generator. *,ven a design trror appears to be an isolated case. Inadequate desijn of a jumper allowed three control rods 3

to drop and necessitated a manual reactor trip in September 1986. This occurred despite reviews of the jumper by the licensee and the manufacturer of the rod control system. This is also considered an isolated case.

In several instances, including an RHR pipe support modification and resolving service water pump cable issues, engfneering judgment was inappropriately used instead of analysis in perfgeming the safety evaluation. In these instances, this apprcach did not provide an adequate technical basis. When analyses are done, the calculations are often not well documented and controlled, nor are they properly reviewed.

Calculation issues were raised by the Quality Verification 1

I Inspection in the middle of this assessment period. At the end of this assessrent period the Safety System Functional Inspection raised similar issues with very recent calculations.

33 The ineffectiveness or lack of corrective actions in this area indicates inadequate site and corporate management attention is being paid to this area.

Site and corporate management involvement in the corrective action for Bulletin 85-03, on motor-operated valvns (MOVs),

appea.s to be inadequate. The licensee took a narrow interpiStation of the bulletin, limiting th> corrective actions to mechanical activities only, and not including electrical engineering in their response. In addition, some activities performed by the contractor (MOVATS) were not reviewed for adequacy by the licensee's engineering staff. In particular, limit switch positions set by the contractor were not reviewed by the licensee for as-left conditions. As a result, improper setpoints for limit switches were identified by NRC personnel, Maintenance procedures were also neted to lack inforeation designating limit switch settings. These factors indicate that site and corporate management did not adequately pursue initi-atives that address generic MOV problems.

The licensee's initial evaluation of the trip of an electrical auxiliary feedwater pump following the October 20, 1986 unit trip did not identify the root cause. NRC involvemer.t was necessary to heighten licensee concerns. Similarly, NRC involvement was ne:essary to adequately evaluate the effects of a possible cavity seal failure, adequately evaluate the effects of a possible f ailure of the auxiliary feedwater steam relief valve, obtain testing of the p0RV nitrogen check valves, identify the source of leakage from steam generator 23, and provide proper splices and cable separation for service water pump cables. While some instances of NRC involvement are to be expected, the relatively large number of instances during this SALP period indicates that a significant weakness exists in the licensee's process of screening for potential safety questions.

It should be pointed out that once the licensee fully recognized the existence of potential safety problems, they ensured that the issua was thoroughly resolv6d. However, continued NRC involvement was necessary throughout the assessment period to assure an appropriate recogrition of patential safety questions.

Due to auxiliary feedwater system problems following the October 20 and 23, 1956 reactor trips, the licensee initiated a program of safety system functional assessments. The first assessment was done on the auxiliary feedwater system and identified several problems including a single failure potential later described in NRC Information Notice 87-34 This safety system functional assessment project is a commendable effort; however, it is moving slowly.

, 39 l Compensatcry measures needed to meet 10CFR 50, Appendix R requirements were not communicated to operations, revised lubrication requirements for modified H5IV packing was not comunicated to maintenance, and an instrument air system modification was not testet because engineering did not resolve a question on test values. These examples indicate that communications between engineering and other groups need .

improvement. Although the modification pre-implementation i meetings described in Section G of this assessment should  !

improve communications, additional steps may be needed. -

The engineering support request backlog is high, although it appears to be staying constant. Important safety work such as resolving residual heat removal cable reuting design deficien- -

cies required to meet Appendix R were not performed promptly.  !

Resolution of instrument air system post-modification test  ;

requirements was not timely. Engineering's review of a '

centractor's analysis of a residual heat removal system pipe support modification did not occur until af ter the system was declared operable and returned to service. These examples indicate that engineering resources are strained and manage-ment controls are ineffective.

During this assessment period, the licensee completed the transfer of off-site technical support engineering groups to the i

site. This placed a heavy strain on technical support engineering, as many staff resigned or transferred within the company rather than make the move. Probabilistic Risk -

Assessment (PRA) applications including applying PRA to maintenance prioritizations were affected by the loss of technical support staff.

$1te and corporate management involvement in the EQ area ap; ears ,

to be deficient. Although Information Notice 86-53 concerning l

! improper installations of Raychem Splices was issued to the i station more than a year ago, no programmatic inspection or  ;

replacement of unqualified Raychem Splices was performed up to the time of the 1987 refueling outage. During the outage, ,

licensee inspections and tests identified unqualified Raychem and Crouse-Hind splices. These unqualified splices were repaired or replaced with qualified splices during the refueling outage. A subsequent NRC inspection identified that several '

junction boxes with conduit penetrations were not preoared in I accordance with station procedures. A substantial number of -

items (terminal blocks and electrical cables) reoviring EQ were identified by the licensee in 1986 to be either unqualified or undocumented. Although the licensee took corrective actions promptly to resolve these deficiencies, the lack of timeliness ,

+

for discovery of these deficiencies is of concern stoce the EQ '

rule requires this effort to have been completed before November i 30, 1985. Several violations have been identified from these s

f t , ..- ,-._-__-s. . - - . - - - - - - - - - ~~- - - - - * ~

40 findings.

These factors indicate weaknesses in the control and l implementation of the EQ program. These discrepancies also indicate that QA/QC involvement in the EQ area is not adequate.

On their own initiative the licensee has begun to estabitsh  :

systems engineers in order to provide more in depth knowledge I

' of systems and create a sense of ownership of system work.  !

Corporate engineering is assigning a counterpart for each '

on-site system engineer.

A The licensee has initiated a significant program to identify,  :

document and make accessiele the plant's design basis. This .

effort, which is just beginning. should be a major contribution  !

to the engineering program. Engineering documentation was often difficult to retrieve for original plant systems.

I In summary, field engineering work is generally of good quality.

An teportant effort is underway to improve the documentation and retrievability of the plant design basis. A backlog of engineering work and delays in completing engineering assign-  ;

ments indicates that resources are strained. This also contri-  ;

butes to slow or ineffective corrective actions. Improvements '

have been made in communications between engineering and other  ;

youps, however, additional steps may be needed. The process of  !

screening potential safety issues and the documentation of the technical basis for resolving safety issues needs to be l i

impreved. Control of calculations is poor.  !

2. Conclusion l Rating: Category 3

[

Trand: None

3. Boaro Recommendations Licensett -

Formalize the engineering analysis proc 6ss.

Provide the engineering staff with better guidance, safety perspective and training related to identifying, investigating, and resolving safety issues.

Conduct a third party review of the engineering program and safety philosophy.

NRC: Recommendations appear under section K.

= ,

43

!. Licensing Activities

1. Analysis During the previous assessment period, this area was rated as a Category 2. More detailed submittals and stricter aeerence tr, schedules were identified as areas needing improveeent.

Early in this SALP evaluation period, the Licensing bganization was reorganized. Licensing, which previously retorted to the Vice President Nuclear Engineering and Quality /.ssurance, was combined with the on-site Regulatory Affairs group. The com-bined group reports to the Vice Presicient Nuc' ear Power. By the end of the evaluation period, effecti$ely all of the licensing functions were conducted on-site. This has been seen as an effective management decision and has provided better integration of licensing and other sice activities. However, the merging of the two groups has presented some difficulties in that areas of responsibility have not always been clearly defined. In addition, Regulatory Affairs has lost several key licensing engineers during the evaluation period. As a result, the historic reference and overall understanding of several issues has not been tr6nsferred to other individuals. The Itcensee recognizes these problems and has been making progress in defining areas of responsibilities and in hiring contractors to make up for some of the personnel shortages. ,

In general, the licensee's management participated in licensing activities and kept abreast of current and ant ~ ipated licensing actions. However, toward the end of the perb it was not evident that proper attention was focused towled planning and followup of licensing actions. As an example, the Itcensee made a timely submittal to amend the Technical Specification: to redefine the conditions associated with the Safety Injection Test. However, additional information was required to complete the amendment process. Because Regulatory Affairs was not aware of the schedular requirements associated with this test and its relatio,1 ship to the amendment, issuance of the ar.Jndment was delayed until after the safety Injection test was performed using the new procedure. If the amendment was not issued for any reason, prior to restart, the licensee would have been Mquired to perfers a second Safety Injection test'using the old procedure. In addition, two A5ME Code relief requests were submitted during the refueling outage for which better planning would have allowed more review time. Had the NRC staff found nroblems with the Itcensee's relief requests and either denied the reliefs or required additional inspection, the licensee could have been in the position of having to reopen the steam generators and the pressurizer af ter closing them up. Had the licensee provided more review time, they would have avoided this potential problem.

42 In most instances, the licensee has been responsive to NRC initiatives. Schedules are negotiated with the licensee based

' on priorities. The licensee usually meets deadlines and noti-fies the staff in advance of schedular problems. The licensee facilitates a timely resolution in most instances. However, it should be noted that extensive delays were experienced with the

'icenset in responding to the NRC questions concerning the 0 stalled Control Room Design Review and the Hurricane Technical 5ptcification Revision.

In gineral, communications with the licensee have been good.

Commuications deteriorated towards the end of the evaluation period beciuse licensing personnel were often not available for discussions with the NRC staff. Telephone calls were repeatedly not returned. Followirg discussions with the General Manager of Technical Support, communications improved t3eporarily. The licensee has attributed this in part to inadequate staffing.

The licensee's engineering and scientific support of licensing activit,ies has been adequate in almost all cases. The licensee's technical capability is reflected in the quality of submittals made in support of, or in respoise to Itcensee or NRC-initiated actions and has minimized the need for additional inform & tion during the technical review, Powever, requests for additional information have been required on several e< stions to support the licensee's no significant hazards (NSH) determination. Although the licensee's NSH determinations have teproved since the previous evaluation period, the licensee should make sure that all areas are addressed in order to minimize requests for additional information. In demonstration of the licensee's technical capability, the licensee completed the augmented inspection of the reactor vessel. This was a significant effort using state-of-the-art technology. The itcensee's performance in this area was commendable.

The Regulatory Affairs group, which is responsible for all LER's, amendment requests, and almost all responses to NRC initiatives, holds information training sessions on topics of current and future interest, and participates in corporate training programs. The group also participates in industry wide t, raining programs provided by various organizations. This training appears adequate.

• In summary, sicensing activities are conducted by a dedicated, knowledgeable and adequately trained staff. Management involvement and overview are evident. The regulatory affairs group has been integrated with other plant activities resulting in a reduction in key licensing staff. This has reduced the licenste's ability to support NRC initiatives as evidenced by weakness in responsiveness and planning.

. 43 l a

c. Conclusion

Rating: Category 2 Trend: Declining

3. Board Reccmmendations Licensee: None NRC: None

44 .

J. Trat'.ino and Qualification Effectiveness

1. Analysis During the previous assessmer i period, this area was rated Category 2. Training effectiveness was weak in operations and security. The plant simulator needed upgrading.

Training and qualification effectiveness continued to be an evaluation criterion for each functionb area and is discussed

h. other SALP functional areas. As such, this is a synopsis of the asse35ments of training in the other areas. The security i training and qualification program was noted as oarticularly effective. No licensed operator exams were conducted during this assessment period.

The licensee increased operator simulator training for normal startup evolutions and improved the accuracy of.these  ;

i simulations. This training was at least partially effective as evidenced by there being only one reactor trip due to operator ,

errors during startup, in contrast to three during the previous SALp period.

As a result of the steam generator dryout event the licensee  !

i identified the need for additional training on the application of the bases for Emergency Operating Procedures (EOPs) for non-l .

errergency events, and on equipment life cycle design transients I a

I in the licensed operator training programs. A significant contributor to this event was a general lack of compliance with i

procedures. This indicates a need for emphasis on following procedures in all personnel training programs.

j J

The licensee has been responsive to the need to upgrade the  !

simulator. A simulator ungrade program has been established and  !

a significant amount of resources have been committed to the  :

project. The primary system model upgrade has been completed.  !

j In general, non-operations technical staff are qualified by {

their formal education and work experience, rather than by  !

i licensee training. Also, although all training programs are new  ;

INp0 accredited, the technical staf f and manager's programs have

] not been fully implemented. A significant weakness,is apparent l in engineers and managers' approach to resolution of problems  ;

from a safety standpoint and shotid be addressed through l training. It was observed that ]&C per;onnel, including l

J supervision, did not appear to be knowledgeable or well trained  ;

on the maintenance administrative control requirements. It was also noted that !&C personnel did not believe these requirements  !

were applicable to the I&C area. A licensee deficiency  ;

indicated that there was also a lack of training for mairitenance '

personnel in this area. Other than the above examples, the  ;

I ,

i

45 performance of the non-licensed staff indicates that the training and qualification program contributes to an adequate understanding of their work and adherence to procedures.

Continued upgrading and implementation of the non-licensed training program is needed. '

Management support for assuring that the quality of training is continually improving is evidenced by:

Indian Point 2 receiving INPO accreditation for the remaining six of te areas in August of 1987. ,

The simulator facility being remodeled to train l non-licensed operators as well as licensed cp9rators.

- l The Toddville Trainin Facility being renovated to expand i both classroom and of'...a space and the training shop's capability. ,

i Maintenance dedicating more than half of its Fabrication

$ hop to on-tne-job training for maintenance and !&C personnel.

The addition of veral degreed staff to the shift.

The establishment of a college degree program for non-degreed operations personnel which will lead to a bachelor's degree in nuclear technology. l

- (

The security training program being upgraded ano program administratson being improved, l i

Contractn radiation protection training was upgraded to [

address w'aknesses identified by NRC during the last l assessmen. period. I

, Training sup.evision was reduced at the end of the assessment '

t l

period by coabining the duties of the operations training  !

' administrator and the simulator upgrace project manager. The l effect of this change has not been evaluated.

During the $At.P period, several reviews of the licensed operator  !

initial and requalification programs wre conducted. These  ;

reviews involved verification of the progras's conformance to '

the requirements of 10 CFR 55 and the training program l implementing procedures. While conduct and instruction of the L

$RO upgrade program wcs coni:dered satisfactory, attention to administrative details was lacking, for example, there was no r master schedule that accuratelv reflected the candidates i' progress through the trair.s iy program. The training schedule was incomplete and inaccurate. In addition, several examples were  !'

identified where Training Administrative Directives (TRAD) had .

i J

o .

46 not been implemented. For example, no attendance records were kept for the $RO upgrade program as required; and individual question grades were lined out and changed without the grader initialing the changes. Recent NRC inspections verified that the above weaknesses have been resolved and that the $RO upgrade program is now receiving adequate administrative attention.

The licensed operator requalification program was also reviewed.  !

Although some program weaknesses were observed, the facility's in house annual written and operating examinations were evaluated as being adequate to identify generic and individual weaknesses. The revision to the r.equalification program which  :

was established in January of 1985, appears to have brought about significant improvements over the previous program. These improvements were demonstrated during the review of the course content scheduled for the 1986 Requalification Training Program, and through interviews conducted with licensed operators.

However, various discrepancits were noted in the r$ qualification l program, for example: Individuals had missed lectures and had not made these lectures up as required by the Implementing Guideltaes (1G), and instances where operators had received less than 80*. on the annual examination and had never received a follow up oral examination as required by the IG. In  ;

addition, even though annual requalification oral examination i documentation strongly reccmends remedial actions for some licensed operators who barely passed the oral examinations, no i such actions had been taken on these recommendations. Also, no feedback was provided to operstors on individual areas of weakness identified by the annual requalification written I

examination unless an examination failure occurred. Rennt NRC >

inspections confirmed that these problems were resolved. '

It was noted that of the five staff positions for licensed instructors, four positions were filled. This number of i

licensed instructors ($ adequate to conduct either a requali-fication or an initial licensing training program. If both i

) programs are to be conducted concurrently, additional training staff may be required.

During the previous $ ALP period it was determined that the }

) licensee was not properly applying the Itcensed operator l 4

recertification requirements of 10 CFR 55.31(e). Doring the current 5 ALP period a violation was issued for this previous  ;

J finding. The requirements of 10 CFR 55.31(e) have since been -

revised. A followup inspection verified that the licensee is in compliance with the revised recertification requirements.

l In conclusion, trrining at Indian Point Unit 2 is generally i adequate and has been improving for several years. Security j training programs were particularly effective. INPO

4r accreditation and the revised operator requalification program are indicative of management att'.ntion in the area of training.

Training activities are genersily effective although significant weaknesses have been identifiaj. Technical staff and site and corporate management apparentl;* lack appropriate' training related to problem resolution from a saftty standpoint.

2. Conclusion Rating: Category 2 Trend: None

3. Board Recommendations Licensee: None NRC: None 9

48 K. Assurance of Ovality (659 hours0.00763 days <br />0.183 hours <br />0.00109 weeks <br />2.507495e-4 months <br />,12*.)

2. An aly si s The primary purpcse of this functional area is to assess the ef fectiveness of the licensee's program for assuring initial quality performance, and for identifying and correcting problems.

This functional area is not an assessment of the quality assurance department, but is an overall evaluation of the effsetiveness of the licenste's initiatives, programs, and policies which affect or assure quality. A special NRC team inspection assessed the effectiveness of the licensee's quality verification activities. Various asper.ts of this area were also routinely examined as part of the resident inspector and region-based specialist inspection programs.

The previous $ Alp rated performance in this area Category 2 and concluced that quality programs are generally effective although quality assurance (CA) involvement in operations and surveillance was not as evident as in other areas.

The licensee's staff makes a conscientious effort to produce gocd quality work although workers sometimes appear frustrated with the backlog of assignrents. ResJurce limitations make it necessary to delay work until it is necessary to do it on a reactive basis instead of proactive. Workers appear knowledge-able of statien policies and usually follow procedures, although significant lapses were evident during the steam generator dryout event.

Supervisors make frequent plant tours. They are actively involved in assessing plant conditions and work in the field.

Senior supervisor and management observation of work on backshifts was increased following an allegation, and is considered adequate.

General performance in rany functional areas and particularly the security area were indicative of the quality of performance which is expected. The frequent lapses in performance identified in the operations, maintenance, surveillance and engineering sreas of this report demonstrate the need for prcept management attension to assure c consistent quality of operations.

I During this assessment period, the licensee's OA and test staff planned and executed well-controlled and technically adequate audits and surveillances in the areas of operations and performance and testing. The individuals who performed these quality verification functions were experienced and capable of conducting in-depth technical verification activities. They were also perforrance oriented and assertive in pursuing and I

1

l tracking safety issues. Staffing changes occurring at the end of this assessment period may adversely affect the ability to continue these audits and surveillances due to reassfgnment of operations qualifted auditors to shift duties. i i

The QA inspection program made significant coatributions to improving the material condition of the plant. !*provecents in pipe erosion EQ splices, motor control center panels, control room cabinets, cable separation, steam gcnerator girth welds, and the plant simulator resulted from the QA inspection program, and show that the QA inspection program is generally effective.

4 However, the Itcensee is not always responsive to significant QA findings. Examples include poor documentation of !&C trouble-shooting, delays in lubricetton of M5!V's and deficiencies in ,

i the cesign of service water pump cab 1; splices. The licensee L should assess whether the QA organization has appropriate '

j authority to resolve the issues they ra'Se.

The Station Nuclear Safety Lomittee (SNSC) has generally been effective and their regiews comprehensive. The $NSC chairman has provided st rong leavership and focuses the committee discussions on tafety, h3 exceptions were identified, however;  ;

the unit trip o.i October 20, 1986 for which SNSC approved restart without determining the root cause of auxiliary feedwater putp #12 tripping; and, approval of an RHR pipe i support modificatton prior to engin?ering reviewing a '

consultant's analysis. One area of cencern is that safety ,

evaluations of pl.nt modifications and i* pairs somettees did not describe the bases for determining whethe.' an unreviewed safety question esisted. D11ance on inadequate engineering analyses l

of a modification or v3 pair was sometimes used by the review i.

Committee in making thet* determinations, t As discussed in the engineering functioral area of this SALP  !

(Section H), the licensee is .wot alwav. effective in re:ognizing i potential safety questions. Oti.cr 6 nan this, the quality  :

verification organizations are generally effective in searching i for and identifying safety significant technical deficiencies in l'

plant systems and operations; however, management does not appear to deal promptly and completely with these def tetencies. -

Numerous examples of this weakness were identified.by the NRC l quality vertftcation team inspection. Other examples are t described in many functional areas of this SALP, As an example, i a safety injection pump failed in November 1986. The licensee  !

replaced the pump and initiated an evaluation of the failure  :

mode. Over a year later, and despite continued inspector inquiries, the cause of failure has act been finalized and the impact on the reliability of the currently installed pumps remains unknown. One deviation and six of the twenty-one violations issued during this assessmeat period were due to inadequate corrective actions for self-identified or NRC-

50 (dentified prcblems.

Further illustration of this weakness is thst sia violation responses submitted during this assessment period needed to be revised, at least in part, and resubmitted.

The root cause of these problems is a lack of management effec-tiveness in ecturing that corrective actions for safety-signift-cent issues are taken promptly and are adequate to prevent recurrence. Although these issues were raised earlier in the SALP period, they do not appear to have been dealt with effectively.

Acditional site and corporate management attention is neeced.

The licensee is developing an improved commitrent tracking system to help address some of the above concerns. This system is only a tool for management's use, it is not a substitute for effective management. Progress in developing this system has been sicw, and improvertents in corrective actions and commitment tracking should not wait for its development.

Many areas of this $ ALP conclude that staffing' levels are strained. This is reflected in what appear to be excessive backlogs. It also relates directly to effective corrective actions since it affects their timeliness and the difficulty of prioritizing assignteents.

In sumary, the quality ma tection organizations are generally ef fective in searching for, and identifying safety stynificar,t technical deficiencies in plant systems and operationst although improvements are needed in screening for potential safety issues. Management does not appear to deal proeptly and completely with deficiencies. Organizatters should become more responsive to identified quality deficiencies, esr.ecially when they are of potential safety significance.

2, Conclusion Rating: Category 2 l Trend: Declining

3. Board Recow endattens Licensee: None l

NRC: Conduct an independent assesseent tean (IPAT) inspection of the major issues in the operations. '

health physics, engin2ering, and assurance of quality areas. Specifically, the team should focus on: the formality in operations and engineering activities; safety perspective; effectiveness of corrective actions; adequacy of staffing, particularly as elated to backlogs, and, ths licensee's ALARA prograa.

L i

• i L

. $1 l

t V. SUPPORTING DATA AND

SUMMARY

A. \

Investications and A11ecations Review

1. Investications There were no Office of Investigations (01) reviews initiated at Indian Point Unit 2 during this assessment period.  :

2. Allegations  !

Sin allegations were received during this assessment period. One  !

allegation was partially subster,ttated, one allegation was not subst,antiated, and four allegations were under review at the end j of t,his assesseent per$od. The allegation which was partially ,

substantiated involved non-licensed operators sleeping on shift {

and not completing their plant tours. This issue is one of  !

several examples cf inadequate supervision that is discussed in Section IV.A of this report, i B. Escalated Enforcement Actions  !

e Confirmatory Action Letter dated January 7,1988 following the steam [

generator dry-out event.

t C. Management Conferences November 21, 1986 - $ ALP Management Meeting

{

May 1,1987 - Region I and headquarters management meettng to d*scuss l performance indicators, trip reduction efforts, simulator upgrade and [

maintenance backlog, i

i Janua ry 15, 1988 - Region I and headquarters management meeting to discuss corrective actions from the steam generator dry-out event, li i

L t

[

l t

[

I E

I

. a r

52 >

0. Review of Licensee Event Reports (LERs)

The overall quality of L .ensee Event Reports (LERs) is good. The reports have been submitted .n a timely manner. The discussions of the events and the correcth actions planned or performed has generally improvec. One area which would benefit from added attention is the submittal of supplemental LERs on planned corrective '

actions and the results received from various testing laboratories on corponent, failures. i Six LERs (86-25, 86 27, 86-35, 86-36, 87-03, 87-06) involved problems associated with the,matit feed- water or auxiliary feedwater systems.

Corrective actions taker, after the events and during the 1987 i refueling cutage should resolve these problems.  !

Six t.ERs (86-26, 86 31, 36-39, 87-02, 87-04, 87-13). involved significant personnel / procedure errors, j i

Two LERS (36-29 and d6-33) involved Emergency Diesel Generator breaker malfunctions. Repairs performed en the 480 volt breakers l i

during the 1987 refueling cutage should resolve these problems.

l Two LERs (86-37 ar.d 87-09) involved the malfunction of relays in '

reactor protective circuitry. The relays were sent to a failure ,

analysis laboratory where extensive testing showed no conclusive mode of failure. After the second event, the licensee developed a i i

cceprehensive relay raintenance program for implementation at the '

next refviling outage. ,

1 l  !

l I l t I

l i

I 1

l I

?

l l'

L.

d

! ', . L

.' . 53 TABLE 1 l

) _ INSPECTION REDORT ACTIVITIES  !

PEPORT NUMEER TYPE I TNSPECTION DATES TNTEECTION Hours DESCRIPTION l 86-24 $PECIALIST 28 i ROUTINE. UNANNOUNCED INSPECTION  !

9/22/86 9/25/86 0F LICENSEE ACTIONS TAKEN IN  !

RESPONSE TO IE8 82-02 86-25 RESIDENT 150 ROUTINE RESIDENT INSPECTION i 8/2/86 9/15/86  :

i  !

86-26 SPECIALIST 52 REQUALIFICATION AND INITIAL l 8/26/86 8/28/86 LICEN5!NG TRAINING PROGRAMS l [

FOR LICEN5ED OPERATOR $  !

1 1

' 96-27 RE510ENT 76 ROUTINE RESIDENT INSPECTION l 1

9/16/S6 10/13/86 i i 86-28 RESIDENT 320 i ROUTINE RESIDENT INSPECTION f 10/14/S6 11/21/36 l J

86-29 1

SPECIALIST 0 OPERATOR REQUALIFICATION PROGRAM 12/10/86 12/12/6L EVALUATION B6-30 5PECIAll5T 44 ROUTINE UNANNOUNCEO SAFEGUARDS i 11/17/96 11/21/86 INSPECTION i I

86-31 SPECIALIST 44 SPECIAL ANNOUNCEf INSPECTION OF  !

11/18/86 11/20/86 A LOOSE WIRE IN.i CONTROL ROOM PANEL AND LICEN5!E ACTIONS ON PRA IN5PECTION t t

86-32 RE510ENT 64 ROUTINE RE5!0ENI INSPECTION -

11/22/86 1/5/87 '

86-33 $PECIALIST 78 ROU11NE UNANNOUNCED RADIATION 12/15/86 12/19/86 PROTECTION INSPECTION I i

87-01 RESIDENT 147  !

ROUTINE RE5!0ENT INSPECTION 1/6/87 2/2/87 .

87-02 $PECIAt.15T 25 UNANNOUNCED INSPECTION OF THE 1/27/87 1/29/87 .

SURVEILLANCE TESTING AND CALIBRATION CONTROL PROGRAM l I

i 1

, 54 87-03 SPECIALIST 26 1/30/87 EVALUATION OF PERFORMANCE TEST t REhuLTS OF dos! METERS USED IN ,

PERSONNEL 005! METRY PROGRAM REMDENT 175

] h3$7 3/2/87 ROUTINE RESIDENT INSPECTION l

87-05 5PECIALIST 34 ROUTINE ANNOUNCED INSPECTION OF 3/2/87 3/5/87 THE MA010 LOGICAL CHEMISTF.?

[

PROGRAM i 87-06 5PECIALIST 37 1

ROUTINE ANN 0VNCEO IN5PECT!0N OF i

2/23/87 2/07/87 LICEh5EE ACTIONS ON PREVIOUS i INSPiCTION FINDING $

(

87 07 $PECIALIST 135 SPECIAL ANNOUNCEO TEAM IN5PECTION t

3/2/8/ 3/6/87 0F THE ELECTRIC POWER SYSTEMS 87-08 RE31 DENT 190 j ROUTJNE RES!0ENT INSPECTION 3/3/87 4/6/87 87-00 SPECIAl.15T 40 ROUTINE UNANNOUNCED INSPECTION OF i 3/16/87 3/20/87 L10010 AND GASEOUS RA0!OAC1!VF EFFLUENT CONTROL PROGRAMS

! B7-10 SPECIAL!$T 25 ROUTINE UNANNOUNCED INSPECTION OF I 3/30/87 4/1/87 ,

NUCLEAR MATERIAL CONTROL AND  !

] A.*CDUNTING

) (

87-11 $PECIALIST 44 ROUTINE UNANNOUNCED NY5' Cit 3/30/87 4/3/87 $Etut!TY INSPECTION 87-12 SPECIALIST 31 ROUTINE UNANN0UNCED FREVENT!vE 4/6/87 4/9/67 EMERGENCY PREPAREDNESS INSPECTION t

E7-13 RE510ENT 137 EiOUTINE RE5! DENT INSPECTION 4/7/87 5/4/S7

\

87-14 $PECIALIST 126 ROUTINE SAFETY INSPECTION OF  ;

5/12/87 5/14/87 ANNUAL EMERGENCY PREPAREONE55

(

EXTRCI5L t t

87-15 SPFCIALIST 38 ROUTINE INSPECTION OF  !

5/4/87 5/8/87 RA0!0 ANALYTICAL MEASUREMENTS l PROGRAM l 87-16 i 5PECIAL15T 502 SPEt!AL ANWOUNCEO INSPECTION TO L 6/1/87 6/12/87 EVALUATE THE ROLE AND FUNCTION l

OF THE QUALITY VERIFICATION OREANIZATION

e

, , SS

. i i

87-17 RESIDENT 130 ROUTINE RES!0ENT INSPECTION I 5/5/87 6/1/87 i i

87-18 RE510ENT 220 ROUTINE RESIDENT INSPECTION 6/2/87 7/6/87 .

i 87-19 RESIDENT 196 ROUTINE RE510ENT INSPECTION 7/7/87 8/3/87 87 20 SPECIALIST 34 ROUT!NE UNANNOUNCED INSPECTION 7/20/87 7/24/87 0F SERVICE WATER SYSTEM 87-21 SPECIAL15T 40 ROUTINE UNANNOUNCEO SAFETY r 8/3/S7 8/7/87 INSPECTION OF RA0 WASTE AND  !

TRANSPORTATION ACTIVITIES '

U10ENT IN5PECTION l 4 7 S/31/87 3 87-23 SPECIALIST 41 a.1NOUNCED INSPECTION OF RAYCHEM 8/17/87 8/21/87 CABLE SPLICES ,

97-24 SPECIALIST $2 ROUTINE ANNOUNCED INSPECTION AND

! S/24/87 8/28/87 FOLLOWUP ON PREVIOUS INSPECTION j

FINDINGS AND NON LICENSE 0 -

TRAINING  !

87 25 RE5! CENT 236 ROUTINE RESIDENT INSPECTION

) 9/1/87 10/5/87 (

j 87-26 $rECIALIST 95 ROUTINE UNANNOUNCEO RADIATION 11/16/87 11/20/87 PROTECTION INSPECTION

' 87 27 RE5! CENT 39 8/21/87 9/28/87 5PECIAL IN5PECT!UN TO INVESTIGATE AN ALLEGATION l

4 87 28 $PECIALI5T 50 1

10il/87 10/9/87 ROUf!NE UNANNOUNCEO $ECUPITY  !

INSPECTION

}

I

$PECIALIST 44 ROUTINE .: ANNOUNCEO FIRE

,7 10/9/87 PROTECT!;. INSPECTION RE5! DENT 219 ROUTINE RE5! DENT IN5kECTION

.. s/87 })/2/87 87-31 SPECIALIST 40 ROUTINE UNANNOUNCEO INSPECTION 10/19/87 10/28/87 0F REACTOR PRE 550RE VE55EL !!!

r 87-32 RESIDENT 201 11/3/87 11/30/87 ROUT!NL RES! DENT IN5PECTION (

5 I

L-------

. t l

. 56 l i

u 87 's $PECIAL157 38 11/16/87 11/20/87 ROUTINE ANNOUNCED INSPECTION OF !

STEAM GENERATOR 7UBE AND GIRTH  !

WELD !$1  :

87-34 CANCELLED 87-35  !

SPECIALIST SA 12/7/87 12/9/87  ?.PECIAL ANNOUNCED INSPECTION OF l STEAM GENERATOR G!RTN WELD i INDICATIONS l

87-36 CAhCELLED 87-37 RE510ENT 450 ROUTINE RE5! DENT INSPECTION 12/1/87 1/31/SS 67-38 SPECIALIST "

12/14/87 12/18/87 SPECIAL ANNOUNCEO INSPECTION OF CABLE SPLICE ENV'dONMENTAL ,

QUALIFICATION i 88-01 SPECIALIST 172 1/4/8B 1/8/SB SPECIAL ANNOUNCED INSPECTION OF  !

PLANT STARTUP t 88-02 SPECIAl.!$7 48 1/11/E8 1/15/E8 ROUTINE UNANNOUNCED IN5PECTION  !

0F LIMITORQUE SWITCH SETTINGS 88 03 SPECIALIST 389 '

AUGMENTED INSPECTION TEAM 1/BiSS 1/12/83 INSPECTION OF STEAM GENERATOR  !

DRY-0UT EVENT i 88-200

• I

$PECIALIST SAFETY 5YSTEM FUNCTIONAL 1/25/88 2/19/88 INSPECTION i

Inspection was not completed by the end of the assessment period.

57 TABLE 2 INSPECTION HOUR $ $UMMARY q

INDIAN POINT 2 Actual Annualized Arte

- HOWrs Percent Heurl OPERATIONS 1825 32 1200 RADIOLOGICAL CONTROL 5 432 8 284 MAINTENANCE 65 1 l 12 432 SURVEILLANCE 74 3

! 13 489 EMERGENCY PREP. Ilil 3 105 SEC/5AFEGUARD5 2P1 4 152 OUTAGES 404 7 266 ENGINEERING 506 l 9 333 LICEN51NG ,

TRAININ3/ QUALIFICATION ASSURANCE OF QUALITY 65v 12 433 Total 5615 300 3694

58

)

TABLE 3 IN0! M 5TRT 2 ENFORCEMENT ACTIVITY A. Violattens versus Functional Area __by Severity Level _

No. of Violatigns in fach Severity Level AREA 1 2 3 4 001 5 TOTAL  !

l OPERATIONS 6 2 8 '

RADIOLOGICAL CONTROLS 2 2 MAINiiuNCE 1 2 1 4 I r

$URVE!LLANCE 3 1 4  !

EMER3ENCY PREP.

SEC/ SAFEGUARD 5 OUTAGE 5 1 1  !

j EN3!hEERING ,UPPORT 2 1 3 & 6' f

l LICEN5!NG t

TRAINING & QUALIFICATION '

A53'JRANCE OF QUALITY " i i

TOTAL 5 15 6 1 22 & 6' I

l l

' Severity levels have not yet been determined and notices of violations have not been issued for sin violations,  !

j

Reported under other functional areas. l t

t I

i l

r l

f I

59 l Tm.3 T3 2 l S. Sumary of Violattens  !

l IN5PECTION REPORTS REQUIREMENT SEVEPITY FUNCil0NAL INSPECTION DATES _V!0 LATE 0 LEVEL A_R jj DESCRIPflCN 86-27 T.S. 6.8.1 4 O!ERATIONS EDG SYSTEM 5 NOT  !

09/16/86 10/16/86 LINE0 UP PROPERLY .

86-28 T.5 6.8.1 4 OUTAGES AND PART! ALLY COMPLETE !

10/14/86 11/21/86 t.55URANCE OF M005 NOT QUALITY CONTROM ED PER

• PROCEDURE 66-28 AN5!.18.7 4 $URVEILLANCE PCV !!39 MOD AND 1 10/14/56 11/21/86 $ PEED CHANGER I SETTING CHANGE l NOT PROPERLY l TESTED

$6-28 10CFR 55.21 5 OPERATIONS OPERATOR 5 NEE 0E0 10/14/86 11/21/S6 RECERTIFICATIONS 87-01 T.S. 6.8.1 4 5URVEILLANCE I POST MAINTENANCE 1/6/S7 2/2/&7 TEST NOT PERFORMED AFTER CHANGING LOCATION OF EMERGENCY O!E5EL GENERATOR'S SERVICE WATER 1

FLOW INDICATION 37 C6 T.S. 6.8.1 4 MAINTENANCE INADE00 ATE 2/23/87 2/27/S7 AND A55UR. CORRECTIVE ACTION OF QUALITY GA5 CYLINDER 5

!MPROPERLY SECURED AND TRA5N IN CARLE TRAYS 87-08 10CFR 50.59 4 OPERATION 5 CHECK VALVE IA-20 l 3/3/87 4/6/87 AND ASSUR. BYPA$$ED WITHOUT OF QUALITY 50.5/s REVIEW 87-08 T.5. 6.8.1 5 MAINTERANCE SERVICE '4ATER MOV 3/3/87 4/6/87 AND ASSUR. TEMPORARY AEPAlk 0F QUALITY NOT CONTR0dE0 PER ADMIN 151RATIVE PROCEDURE

i

, 60  !

87-20 CODE REPA!R$ 4 ENGINEERING USE OF i 7/20/87 7/24/87 UNAUTHORIZE0 CODE ,

REPA!R$

87-23 10CFR 50.49 4 ENGINEERING CA88.E $PLICE '

8/17/87 8/21/87 DOCUMENTATION INADEQUATE.'.ND i

!PLICES INSTALLED OVER 3 RAID!M i

87-25 T 5. 6.2.2(g) 4 OPERATION $

OPERATOR OVERTIME 3

9/1/87 10/5/87 AND A55UR. HOUR 5 NOT 8EING

0F QUALITY CONTROLLEO PER GL 82-12 P0llCY j STATEMENT 3

87-25 T.S. 6.8.1 5 $URVE!LLANCE TECH SPEC REQUIRED J

9/1/37 10/5/67 AND AS5UR. CHANNEL CHECK j

1 OF QUALITY PROGRAM h0T ,

DEFINED 01 DOCUMENTED 87-26 T.S. 6.12 4 RAD / CON LOCKED HIGH RAD [

a 11/16/87 11/20/87 AND A5504 000R NOT SECURED L OF QUALITY (

87-32 T.S. 6.8.1 5 11/3/87 11/30/87 ENG!h 1 RING APPENDIX R COMFEN5ATORY

(

r MEASURE NOT l PROLEOURALIZED ,

t 87-32 10CFR 20 4 RAD / CON CONTAMINATED -

11/3/87 11/30/87 TRA5H 5H!PPE0 0FF5!TE WITN007 5URVEYING 87 37 T.5.6.8.1 5 OPERATION 5 P05T-TRIP REVIN 12/1/87 1/31/88 AND A55URAkCE REPORT NOT f

0F QUALITY REVIEVED PA0MPTLY  !

87-38 10CFR 50.49

• ENGINEERING CROUSE HIN05 12/14/87 12/18/87 AND A$$URANCE INSTRLetNT OF QUALITY CA8LE $PLICE5 NOT QUALIFIED 87-38 10CFR 50.49

• ENGINEEk!NG CROUSE-H!N05 12/14/87 12/18/87 AND ASSURANCE CONTROL AND OF QUALITY PCVER CABLE SPLICE 5 NOT QUAllFIED 1 1

._ _a

- ^

61 87-38 10CFR 50.49

• ENGINEERING UE&C CABLE 12/14/87 12/18/87 AND ASSURANCE SPLICES NOT OF QUALITY QUALIFIED 87-38 10CFR 50.49

• ENGINEERING MOTOR LEAD 12/14/87 12/18/87 AND ASSURANCE TAPE SPLICES OF QUALITY NOT QUALIFIED 87-38 10CFR 50.49

• ENGINEERING QUALIFICATION OF 12/14/87 12/18/87 WEIDMULLER TERMINAL BLOCKS NOT DEMONSTRATED 87-38 10CFR 50 APP. 8 5 MAINTENANCE JUNCTION BOXES 12/14/87 12/18/87 AND ASSURANCE NOT SEALED PER OF QUALITY PROCEDURE B7-38 10CFR 50.49

• ENGINEERING RAYCHEM CABLE 12/14/87 12/18/87 AND ASSURANCE SPLICES NOT OF QUALITY QUALIFIED 88-0) 10CFR APP. 8 4 SbRVEILLANCE POST MAINTENANCE 1/4/88 1/8/83 TEST NOT PERFORMED BY PROCEDURE 88-02 BULLETIN 85-03 DEV MAINTENANCE LIMITORQUE 1/11/88 1/15/88 COMM:THENT **

AND ASSURANCE MAINTENANCE OF QUALITY PROGRAM NOT IN PLACE 88-03 T.S.6.8.1 4 OPERATIONS SUPERVISORS AND 1/8/88 1/12/S3 OPERATORS NOT FULFILLING THEIR RESPONSIBILITIES OR FOLLOWING OPERATING POLICIES 88-03 T.S.6.8.1 4 OPERATIONS OPERATING 1/8/88 1/12/88 PROCEDURES VIOLATED 88-03 T.S.6.8.1 4 OPERATIONS LOGKEEPING AND 1/8/88 1/12/88 AND ASSURANCE RECORDER CHART OF QUALITY RECORD REQUIREMENTS NOT MET

• Severity levels have not yet been determined and notices of violation have not been issued for six EQ violations.

• • Inspection report has not yet been issued.

y

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• - 62 TABLE 4 INDIAN POINT 2 LICENSEE EVENT REPORTS A. LER by Functional Area Number by Cause Codes FUNCTIONAL AREA A B C D E X TOTAL OPERATIONS 3 2 5 RADIOLOGICAL CONTROLS 2 2 KAINTENANCE 1 2 4 2 9 SURVEILLANCE 1 9 1 11

, EMERGENCY PREP.

SEC/ SAFEGUARDS OUTAGES 1 1 ENGINEERING SUPPORT 1 4 2 7 LICENSING t

TRAINING AND QUALIFICATION ASSURANCE OF QUALITY TOTALS: 7 4 0 5 13 6 35 Cause coces: A. Personnel Error B. Derign, Manufact; ring, Construction, or Installation Error

t. External cause D. Defective Procedure E. Component Failure

• X. Other

,- 63 TABLE 4 B. LER Synopsis INDIAN POINT 2  !

LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION 86-25 7/18/86 A LOSS OF INSTRUMENT AIR TO AFW 86-26 7/24/86 A HYOROGEN-0XYGEN MONITOR MALFUNCTION 86-27 8/2/86 B UNIT TRIP OUE TO LOSS OF M8FP 86-28 8/4/86 E RWST LEVEL INSTRUMENTS OUT OF SPEC 86-29 8/6/S6 E DIESEL GENERATOR BREAKER FAILURE 86-30 8/12/86 A VIOLATION OF CONTAINMENT INTEGRITY AT PERSONNEL AIR LOCK 86-31 9/16/86 0 REACTOR TRIP: THREE R005 DROPPE0 86-32 9/22/86 A MISSED SURVEI'. LANCE TEST - DAILY VC ATMOSPHERE GRAB 86-33 10/15/86 E EDG #22 BREAKER MALFUNCTION 86-34 10/16/86 E TWO INOPERABLE OT DELTA T AND OP DELTA T CHANNELS, CONTROLLER REPLACE 0 86-35 10/20/86 E REACTOR TRIP: LOOSE CONNECTION IN TRIP CIRCUIT, AFV PROBLEMS 86-36 10/23/86 E MANUAL REACTOR TRIP: LOSS OF MFP, AFW #22 TRIPPE0 86-37 11/6/86 E REACTOR TRIP: RELAY MALFUNCTION DURING SURVEILLANCE TEST 86-38 11/14/86 E S1 PUMP FAILURE: HOT SHUT 00VN 86-39 12/17/86 A MISALIGNMENT CAUSES INOPERABLE R005

. 64 t

'87-01 1/21/87 X EDG THOUGHT TO HAVE INSUFFICIENT SERVICE WATER 87-02 1/30/87 0 TWO MSIVS FAILED TO CLOSE OURING PLANNEO SHUTDOWN

. 87-03 2/2/87 E TWO AFW PUMPS INOPERABLE OUE TO LEAKY RECIRCULATION VLLVES '

87-03 2/2/87 E REV. 1 CWST LCV 1158 OION'T CLOSE FAST ENDUGH 87-04 2/10/87 A REACTOR TRIP: OPERATOR ERROR 87-05 3/18/87 X TWO EDG'S INOPERABLE - ONE IN PMT ONE LOST START AIR 87-06 4/30/87 8 SINGLE FAILURE POTENTIAL F0k AFW PUMPS 87-07 6/3/87 A CCR VENTILATION INOPERA8LE DUE TO 800 STER FANS 005 87-08 6/23/87 X CCR CHARCOAL FILTERS IbOIDE EFFICIENCY TOO LOW 87-09 6/27/87 E RCACTOR TRi?: SidAM GENERATOR LEVEL RELAY MALFUNCTION 87-10 10/8/87 X WCCPP ACTUATED BY NO8LE GAS  :

f, MONITOR 87-11 10/9/87 E TWO SW PUMPS FAILED SECT. XI TEST

- HEAD TOO LOW 87-12 10/19/87 X

' ELECTRICAL SPIKE CAUSES WCPPS TO ACTUATE 87-13 11/5/87 0 LOSS OF 480V POWER 00 RING SI RELAY PM 87-14 11/6/87 0 PROCEDURAL DEFICIENCY PREVENTS OBTAINING "AS FOUNO" SG LEVEL TRANSMITTER DATA 87-15 11/18/87 E PORY BACKUP N2 SUPPLY INOPERA8LE 87-16 12/11/87 E TWO OF THREE PRESSURIZER SAFETY VALVE SETPOINTS OUT OF TOLERANCE 87-17 12/1/87 B ENVIRONMENTAL QUALIFICATION OF ELECTRICAL SPLICES

s 65 l

87-18 12/4/87 B

~

COMMON RHR R2 CIRC LINE CAN LEAD .

TO PUMP FAILURE (!E NOTICE 37-59)'

87-19 12/8/87 A. '

ACCUMULATOR TANK LEVEL INSTRUMENT- >

Call 8 RAT 10N ERROR

-l 87-20 12/31/87 0 '!',

RTO TERMINAL BOXES NOT E0'O W

1 l

v . .

l

. l 1

l

. 66 TABLE 5

$UMMARY OF LICENSING ACTIVITIES

1. NRR/ LICENSEE MEETINGS SALP MANAGEMENT M!!ETING 11/21/86 AFW RELIABILITY 12/9/86 REACTOR VESSEL AUSMENTED INSPECTION 1/13/87 HURRICANE TECH SPEC 3/20/87 '

REACTOR VESSEL AUGMENTED INSPECTION 4/7/87 POWER UPRATE 10/5/87 SIMULATOR UPGRADE 10/29/87

2. NRR SITE VISITS / MEETINGS MANAGEMENT VISIT 11/16/86 SITE VISIT 3-9-10/87 MANAGEMENT VISIT 5/1/87 SITE VISIT 5/5/87 QUALITY VERIFICATION INSPECTION 6/1-12/87 REFUELING OUTAGE WORK SCOPE 9/17/87 REACTOR VESSEL AUGMENTED INSPECTION 10/21-24/87 SITE VISIT 11/5-6/87

3. COMMISSION MEETINGS NONE 4

SCHEDULAR EXTENSION GRANTED REGULATORY GUIDE 1.97 8/12/86

5. RELIEFS GRANTED ASME SECTION XI 8/7/86 ASME SECTION XI 8/7/86 ASME SECTION XI 8/28/86

6. EXEMPTIONS GRANTED APPENDIX R - HVAC EXHAUST r'ANS 3/4/87

4 3- 67 AMENOMENT NUMBER TITLE DATE

7. LICENSE AMENDMENTS ISSUE 0 115 STATION NUCLEAR SAcETY 8/7/86 COMMITTEE 116 POWER DISTRIBUTION LIMITS 10/6/86 117 QUADRANT POWER TILT 11/13/86 118 LICENSE EXTENSION 4/21/87 119 SAFETY VALVES 5/28/87 120 BATTERIES 23 & 24 6/17/87 121 PARTIAL MOVEMENT OF 8/21/87 CONTROL R005 122 SURVEILLANCE EXTENSION 9/1/87 123 ORGANIZATION 9/2/87 124 AFW PUMP LCO'S 10/14/87 125 AFV FLOW 10/15/87 126 CTMT PURGE AND VENT 10/29/87 127 ILRT 11/16/87 128 RHR PUMP OPERABILITY 11/18/87 OURING SI TEST 129 ISI AND IST 11/18/87

- ~

A,,, u, ENCLOSURE 3

'e UNITED STATES 8C 't I* NUCLEAR REGULATORY COMMISSION s  ?

• $ REGION I j.

.f 47s ALLENDALE AoAD xiNn or emuss:A rENNsvLv4NiA se4os 04 APR 1988 Oceket No. 50-247 Consolidated Edison company of New York, Inc.

ATTN: Mr. Stephen Bram Vice President, Nuclear Power Indian Point Station Broadway and Bleakley Avenue Buchanan, New York 10511 Gentlemen:

Subject:

Systematic Assessment of Licensee Performance (SALP)

The NRC Region ! SALP Board has reviewed and evaluated the perfarnance of activities 1968.

at Indian Point Unit 2 for the period of August 1,1986 - February 7, The results of the assessment are documented in the enclosed SALP Board report dated March 17, 1988. A meeting to discuss this assessment has been schedvied for April 12, 1988 at the NRC Region I office.

At the SALP meeting, you should be prepared to discuss our assessments and your plans to improve performance. This meeting is intended to be a dialogue wherein any e m.ents you may have regarding our report may be discussed.

Additiona'.ly, you may provide written comments within 30 days af ter the meetings Your cooperation is appreciated.

Sincerely, RJ1, liam T. Russell Regional Administrator

Enclosure:

SALP Report No. 50-247/86-99

\

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gTT -

s

Consolidated Edison Company 04 APR M of New York 2 cc w/ enc 1:

Jude G. Del Percio, Manager, Regulatory Affairs P. Kokolakis, Director, Nuclear Licensing Brent L. Brandenburg, Assistant General Counsel Walter Stein, Secretary - NFSC Department of Public Service, "cate of New York Public Document Room (POR)

Local Public Document Room (L90R)

Nuclear Safety Information Center (NSIC) '

NRC Resident Inspector State of New York K. Abraham, PA0 (11)

Chairman Zech Commissioner Roberts Commissioner Bernthal Commissioner Carr -

Commissioner Rogers bec w/ encl:

Region 1 Docket Room (with concurrences)

Management Assistant ORMA (w/o enc 1)

DRP Section Chief Robert J. Beres W. Johnston, DRS F. Congel, DRSS "

J. Taylor, 0500 W. Russell J. Allan '

O. Holody J. Lieberman, OE Marylee Slosson, NRC LPH Brent Clayton, Regional Coordinator, EDO H. Eichenholz Board Members l

l l

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i I

i

. - steenen s. stem

~.. va a =4 ENCLOSURE 4

. Consc* dated Edison Company of New vorm, ine inean Pont Staten Broadway & 84anwy Avenue 8venanen. NY 10511 te+onone (914) 737 8116 May 12,1988 Re: Indian Point Unit No. 2 Docket No. 50 247 Mr. William Russell Regional Administrator - Region I U.S. Nuclear Regulatory Conunission 475 Allendale Road King of Prussia, PA 19406-1498

SUBJECT:

Response to Systematic Assessment of Licensee Performance (SEP) r This letter provides a response to the issues ident Mied both in the Systematic Assessment of Licensee Performance (SEP) report for IP-2 dated April 4, 1988 and at the April 12, 1988 meeting between Con Edison and the

.iRC, held to discuss the assessments presented in that report.

The SEP report covering the performance of activities at IP-2 for the period August 1, 1986 through February 7, 1988 hs been reviewed. We acknowledge the report's assessments and concur with the SEP Board reconunendations. Actions have been or will be taken in response te each of the specific reconenendations, and with regard to the broader issues raised in the report.

Attachment A provides a brief discussion of our strategy and goals for improvement. It represents a general overview of our program to address issues identified in the SEP report. This improvement program is consistent with our 1988 Corporate Goals for Nuclear Power and Engineering. Attac!w.ent B specifically addresses further programs relating to the Engineering Support Functional Area. In addition, we will strive to further enhance our performance in areas that have already been identified as strong.

We are determined to correct weaknesses that have been identified. In order to best accomplish this to our mutual full satisfaction, con Edison upper management, together with those middle managers who are responsible for various plant-related functional areas, plan to meet with appropriate NRC staff personnel to discuss our specific betterment programs. We would request and welcome feedback relating to our programs.

Very truly you::s,

, i -

i Atta chment,

.SSO6010195 880523 ff/

PDR ADOCK 05000247 Q DCD

~ -

t' e

cc: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 Ms. Marylee M. Slosson, Project Manager Project Directorate I-1 Division of Reactor Projects I/II U.S. Nuclear Regulatory Commission ,

Mail Stop 143-2 Nashington, DC 20555 Senior Posident Inspector U.S. Nuclear Regulatory Commission P.O. Box 38 Buchanan, NY 10511

.. o e

ATTACHMENT A strategy for Improvement The goal of site and corporate management is to reaffirm throughout our nuclear organization that the highest priority for the operation of Indian Point Unit 2 is reactor safety, and to ens ure that resources are effectively allocated to quickly and thoroughly identify, address and resolve potential safety issues. The initiatives listed below have either ,

been implemented or are planned to be implemented in the near future in response to this goal.

1. A policy statement for Excellence in Operations is being developed with the full input and support.of all levels of station management.

2. An operations planning group has been established to ensure adequate planning for activities controlled by the watch sections. This planning includes advanced development of contingency plans and review of required procedures and work permits to conduct the planned evolutions.

3. Increased emphasis is being placed on tracking ecomitments and corrective actions to ensure their timely and complete resolution.

Responsibility and accountability for corrective action is being addressed at all levels within the management chain.

4. Increased emphasis has been placed on ensuring the adequacy of operating proc <Jures and strict adherence to procedure.

5. Increased emphasis has been placed on ensuring proper communica'aon.

Particular attention has been placed on cosuunication within the watch organitation and between the watch organization and operations management.

6. A review of the Nuclear Power organization and the interface between Nuclear Power and Central Engineering is being conducted. Various organisational changes will be accomplished to streamline the management process, enhance personnel ef fectiveness and improve coseunications.

7. A System Engineer program is being implemented on site to provide a single individual as a point of contact for key plant systems.

Each '

individual system expert will be cognizant of his/her assigned system.

It is expected that this program will facilitate the conduc't of design, operation, maintenance and testing of the systems and contribute to improved communications, responsibility and accountability for system perforr.ance among organizations.

8. The Chemistry area has been strengthened by increasing management involvement and oversight. A quality program is taking shape by realigning organization riporting, increasing the number of middle management personnel and specifically assigning an oversight function to a subcoreittee of the Nuclear racilities Safety Comunittee.

1

=

e

9. Radiation exposure reduction lessons learned from the past outage have been identified, reviewed and prioritized for applicability to the next three cutanes. Research and development efforts are also being taken to identity and determine the feasibility of additional means to reduce radiation dose.

10. The maintenance workload is being reviewed to reexamine those items with potential for safety-significance. High priority will be assigned to reducing this portion of the current maintenance workload.

11. A Design Basis Documentation program has been initiated to improve the completeness and retrievability of design basis information for key plant systems. It is expect.ed that this effort will provide additional capability to ensure that potential safety issues will be properly addressed.

o 2

O ATTACHKCiT S Introduction,n In addition to initiatives listed in Actachment A, con Edison has reviewed Section E (Engineering Support) of the SALP including the bar;es for the observations drawn therein. The following are descriptions of additional actions tahich have been implemented in our prograns or are planned to be implemented in the near future to addrses these observations and board recorucentlations.

Engineering Support Irtnrovemant Actions

1. Design modif>..tions are required to be presented to the Station Nue:1 ear Safety Committce (SNSC). SNSC now requires the discipline en<lineer to personally make the SNSC presentation with key beckup do:umentation to support the package. SNSC members wila ascertain the safety basis of proposed changes and assure appropriate input and assumptions are well founded and properly documented thereby ensuring a thorough review of any potential safety issues.

2. W; are evaluating the developunt of a training program specifically designed to increase tht, awarenesw and sensitivity of plant personnel in the scope of reactor and equipment safety issues.

3. Con Edison's Corporate Engineering Organization recognizes the imp:tance of complying with procedures for verification and revies of analysis and calculations, including the required documentation of these verifications and reviews. Procedures have been revised to require the assembly, submittal and check of this documentation at the tire of inodifiestion issuance.

4. A station procedure has been developed, which establishes the controls needed for the identification, verifiestion and documentation of these calculatiens which impact operation of the plant.

5. Engineering is developing procedural revisions to provide a formal mechanism for transmitting to the site potential safety significant issues identified by Engineering for inclusion into the SOR review process for reportability determination.

6. We will develop a program for pre-screening Field Engineering workload and requests. The backlog of Field Engineering Reques':s (FER) will be assigned a priority for completio . These actions will assist in (1) reducing backlogs (2) managing ti:Pely CoSpletion of FERO, and (3) steeting operating needs. ';entra l Engir,eering will track this workload to facilitate management consideration of temporary or permanent increases in Field Engineering forces at the site. Responsibilities and guidelines for this program will be incorporated into Nuclear Power and Central Engineering procedures.

3

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I l

7. The EQ program plan document, expected to be finalized by September 30, will formalize and define the interrelationship of job functions and responsibilities to support the EQ program. ,

Board Recoernendations for Engineering Support

1. Formalize the engineering analysis process. Provide the engineering staff with better guidance, safety perspective and training related to identifying, investigating, and resolving safety issues.

ACTIONSf I Compliance with formal requirements of the engineering analysis process is recognized by corporate and site engineering management. Nuclear Power has draf ted procedural requirements for review of analyses perfomed by Nuclear Power and the contrel of analysis documentation.

A project has been initiated to identify and correct deficiencies in file records for modifications performed since plant start-up.

A consultant was etained in July 1987 to provide guidance and methods for 10CFR50.59 safety reviews / evaluation. The result of this work has been incorporated into procedures for safetv impact analysis and input to safety evaluations for pisnt design modifica' ions. Evaluation of procedural changes for safety review evsluation of other types of facility changes such as jumpors, temporary repairs, and plarc procedures is underway. The development of a femalized program for train Ang/ retraining those personnel responsible for evaluation of plant changes in the concepts and criteria of 10CFR50.59 is underway.

2. Conduct a third party review of the engineering program and safety philosophy.

ACTICH:

We will develop an overall upgrade action plan to further improve the engineering program and safety philosophy. This plan will utilize third party reconeendations presented to a senior level oversight committee.

4 e

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. , Enclosure 3

. ,2 i so

. s 3sc;e.n.s' "g

-) aa Docket No. 50-286 Power Authority of the State of New York Indian Point 3 Nuclesr Power Plant ATTN: Mr. William vasiger Resident Manager P. O. Box 215 Buchanan, New York 10511 Gentlemen:

Subject:

Systematic Assessment of License Performance (SALP)~ Report No.

50-286/85 Amended Report This refers to the aisessment we conducted of the activities at the Indian Point Unit 3 Nuclear Power Plant, for the period December 1,1985 to May 31, 1987. This assessment was discussed with you at a meeting on .

November 13, 1987 at your Trainirig Center in Buchanan, New York. The list of attendees is attached as Enclosure 1. Your written comments on our report are attached as Enclosure 2. The SALP Board Report was forwarded to you for review by our letter or October 6,1987 (Enclosure 3) and is attached (amended) as Enclosure 4. Based on our review of your comments, changes to the report are detailed and explained in the attached errata sheet (Enclosure 5).

Our overall assessment of your facility operation concludes that increased management attention is needed to reduce the number of challenges to the reactor protection system, to conclude long-term corrective actions to improve procedures and update drawings, and to enhance engineering and technical support for site-related modifications.

No reply to this letter is required. Your actions in response to the NRC Systematic Assessment of Licensee Performance will be reviewed during future inspections of your licensed facility.

We appreciate your cooperation.

Sincerely, .

Original Signed By UILLIC T. P.USST.LL William T. Russell Regional Administrator 8002140319 880209 PDR ADOCK 0500029/r f//

O PDF OFFICIAL RECORD COPY SALP IP3 85-98 ANENDED - 0001.0.0 02/03/88 t ..

lll

e Power Authority of the ' 2 State of New York 00 FEB 1988 1

1

Enclosures:

1. Attendees at Indian Point Unit 3 SALP Management Meeting on November 13, 1987

2. Response letter from the fewer Authority of the State of New York to NRC dated December 28, 1987

3. Letter from the NRC to Powar Authority of the Stata of New York dated October 6, 1987

4. Amended SALP Report

5. SALP Board Report Errata Sheet cc w/encis: .

J. Phillip Bayne, President s Gerald C. Goldstein, Ass',stant General Counsel

, A. Klausmann, Senior Vica Presidant, Appraisal and Compliance Services W. Josiger, Resident Manager G. M. Wilverding, Chairman, Safety Review Committee Robert Spring, Director, Regulatory Affairs (Con Ed)

R. E. Beedle, Vt;e President Nuclear Support R. Burns, Vice President Nuclear Operations S. S. Zulla, Vice President Nuclear Engineering P. Kokolakis, Director Nuclear Licensing - PWR .

NRC Licensing Project Manager Dept. of Public Service, State of New York Public Document Room (POR) local Public Document Room (LPOR)

Nuclear Safety Information Center (NSIC)

NRC Resident Inspector State of New York Chairman Zech' Connissioner Roberts .

Commissioner Bernthal i Commissioner Carr .

Commissioner Rogers '

K. Abraham (PA0 - 8)

P N

I I

OFFICIAL RECORD COPY SALP IP3 85-98 AMENDED - 0002.0.0 02/03/88 l

~

Power Authority of the 3 09 FEB 1988 State of New York bec w/ehc1:

Region I Docket Room (with concurrences) ,

Management Assistant, DRMA (w/o enc 1) G P L .wo C;owi A '

Robert J. Bores, DRSS Section Chief DRP W. Johnston, DRS T. Martin J. Taylor, DECD '

W. Russell J. Allan

. D. Holody Management Meeting Attendees (NRC Attendees)

DRP Wishlist Coordinator R. Capra, NRR D. Neighbors, NRR G. Meyer, DRP DRP File RI:DRP .0 I hRP h

PRI: RI

, i RI:RA Meyer/mk/rh1 Jo nson nger d p6ns/ Ka an Russell 4'/h /88 /88 /88 A/B/88 a/0/88 a/y/88 a///88

( '

i OFFICIAL RECDRD CDPY SALP IP3 63-98 AMENDED - 0003.0.0 01/28/88

. t I

ENCLOSURE 1  :

Attendees at Indian point Unit 3 SALP Management Meetino (November 13, 1987)

NRC l

)

W. Russell, Regional Administrator .

R. Capra, Director, Project Directorate I-1, NRR "

J. Johnson, Chief, Reactor Projects Section 2C .

, D. Neighbors, Project Manager, NRR l

. C. Haughney, Chief, Special Inspection Branch, NRR ,

L. Norrholm, Section Chief, Special, Special Inspection Branch, NRR P. Koltay, Senior Resident Inspector, Indian Point 3 . -

R. 'Barkley, Resident Inspect:r. Indian Point 3 -

O. Hickman, Performance Evaluation Branch, NRR ,

PASNY J. Bayne, President

J. Brons, Executive Vice President, Nuclear Generation R. Burns, Vice President, Nuclear Operations R. Beedle, Vice President, Nuclear Support S. Zulla, Vice President, Nuclear Engineering C. Lipsky, Vice President Design and Analysis ,

i C. Spieler, Vice President, Public Relatic,s '

W. Josiger, Resident Manager M..Cass, Assistant to Resident Manager P. Kokolakis, Otractor, Nuclear Licensing (PWR)

K. Mavrikis 01 rector, Project Engineering i J. Kelly, Director, Radiological Health and Chemistry (WPO) '

K. Chapple, Director, Nuclear Operations and Maintenance G. Wilverding, Manager, Nuclear Safety Evaluation '

F. Pesce, Director, Quality Assurance (WPO)

W. Harrington, Director of Security i J. Russell, Superintendent of Power i S. Munoz, Technical Services Superintendent J. Vignola, Maintenance Superintendent -

J. Perrotta, Radiation and Environmental Services Superintendent i R. Tansky, Training Superintendent l

C. Mackay, Operations Superintendent i D. Halama, Quality Assurance Superintendent .

F. Findar, Quality Assurance Manager (WPO) l M. Albright, Superintendent, Instrumentation and Control j J. Hahn, Site Security Manager ,

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1. b NewYorkPower

& Authority December 28, 1987 IP3-WAJ 071Z IP3-MPC-1238 Docket No. 50 286 License No. DPR-64 Mr. William Russell Regional Administrator, Region !

U.S. Nuclear Regulatcry Comission 631 Park Avenue Kg of Prussia, PA 19406

Subject:

Systematic Assessment of Licensee Performance Report No. 50 286/85 98

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Dear Mr. Russell:

On behalf of Indian Point 3 Nuclear Power Plant and the New York Power Authority, I am writing to you concerning he recent assessment of performance presented in the subject report. Let me begin by expressing the Authority's views on the meeting held on November 13, 1987 to discuss the report. It is our opinion that the meeting was productive and afforded the opportunity for the frank exchange of ideas between our respective staffs. We also greatly appreciate your time in touring our facility and the presentation of licenses to the recent graduates of our licensed operator training program.

Regarding the SALP, we consider the evaluation to be a very positive representation of the operation at Indian Point 3. An arLa addressed in the report which we consider to be extremely important is Operations. The need to reduce the number of plant trips is recognized. The Authority has taken several positive steps to accomplish this. These steps include both hardware and programatic improvements. .

Due to the high frequency of plant trips initiated in the Feedwater System, the Authority comissioned a tasx force to review the system and recomend improvements. Hardware upgrades consisting of a new feedwater pump speed control system and trip setpoints have greatly improved the feedwater system performance and provide the operating staff with greater flexibility in response to feedwater transients.

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We also recognize that the design of certain existing systems and their testing requirements present an elevated potential for plant trips. Examples cited in the report include the instrument bus loading arrangements and turbire overspeed trip system surveillance testing. Efforts are underway to reduce or eliminate the potential for plant trips due to such system designs or testing. We are actively involved in the Westinghouse Owners Group trip reduction effort.

Programa ic enhancements are also underway. A detailed root cause analysis capability is being developed for use in'avaluating in house operating events. Improvements in shift turnover have been instituted including a revised shift relief and turnover procedure which provides for the walkdown of each control panel by the counterparts of the oncoming and offgoing shift. The practice of assembling the shift crew at the start of the watch for a comprehensive briefing of plant status and planned evolutions during the, shift has also been instituted.

The Authority is concerned with the Staff's statement on page nine of the sal.P report which implies the thoroughness of engineering evaluations is compromised to facilitate the plant's return to service following a trip. The policies and procedures which address post trip review and' restart are clear with respect to the evaluations that mu:t be completed before a restart decision is considered. Furthermore, .

the appropriate level of management is involved in all restart decisions. Plant restart decisions are based on a thorough review of the events leading up to the trip including a full understanding of the salient causes of the trip. While there are attendant economic incentives to return a plant to service, the Authority's primary, motivation is the safe operation of Indian Point 3.

We want to take this opportunity to present details to clarify maintenance and operation of a main boiler feedpump cited in the assessment of maintenance on page 16 of the ruport. The report describes a reactor trip attributed to the failure to replace marg;nal parts during routine maintenance on the pump. The maintenance referred to occurred during the 1985 refueling outage during which the pump was overhauled. An oil seal to be installed was identified as being slightly out of round. The anomaly was reviewed and, since til manufacturers tolerances were met, the seal was installed. Other work was performed on the pump by an outsioe vendor.

During operation, water intrusion into the control oil system caused a plant trip. The preliminary assessment attributed the contamination of the oil to the oil seal. Steps were taken to mitigate the potential for recentamination during operation. During the recent refueling outage a complete overhaul of the pump was performed at which time the root cause of the water intrusion to the control oil system was identifits. The water seal was determined to have been -

installed improperly by the pump vendor representative. This resulted in the water seal's failure to operate properly placing demands on the oil seal for which it was not designed.

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e Based on our review of the matter, proper decisions had been made during main boiler feedwater pump maintenance and appropriate parts ,

replacement was conducted. I We believe one point made in the SALP report to be in error.

Specifically, on page 28, "Assurancs of Quality", it is noted that the Plant Operating Review Comittee and Safety Review Comittee (SRC) failed to uncover a problem with the implementation of a Technical Specification amendment. As was discussed at the meeting, the principal method of followup , exercised by the comittees is the audit process. The failure to completely implement Amendment No. 67 to the Technical Specifications was identified by an SRC directed audit.

Also, as we discussed at the meeting, we are perplexed with the SALP evaluation conclusions in the area of Training. In reviewing the  !

report; the contribution of training in several of the functional categories is noted to be positive. With respect to the Training area specifically, the Authority has made a significant commitment including a plant specific simulator, an 80,000 square foot training '

facility, and accreditation. To date, all programs have been accredited including those programs dealing directly with the operations and maintenance areas.

An excellent indicator of the effectiveness of our training p.aoaru is the success rate in licensed operator examinations and requal-ifications. We acknowledged a weakness in our emergency operating procedure training and promptly undertook an in depth retraining program. This program was considered excellent as documented in the sal.P report.

Training at Indian Point 3 will continue to advance with the maintenance of the accredited programs, delivery of our simulator and completion of the training facility.

I am available to discuss these coments should you desire.

Sincerely, MMM. Josi Wil',.i. .-

Mr dent ana Indian P int nit 3 Nuclear Power Plant ,

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U.S. Nuclear Regulatory Comission Washington, D. C. 20555 Resident Inspector's Office Indian Point Unit 3 U.S. Nuclear Regulatory Comission P.O. Box 337  !

Buchanan, NY 10511 i i

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• ENCLOSURE 3 ,

/ \ a UN.Tsc suTEs NUCLEAR REGULATORY COMMISSION

. i l MEGION I 8., 631 PAMK AVENUE

"'g /g KING OF PRUSSIA. PENNSYLVANIA 19408 Docket No. 50-286 Power Authority of the State of New York Indian Point 3 Nuclear Power Plant ATTN: Mr. William Josiger Resident Manager P. O. Box 215 Buchanan, New York 10511 Gentlemen:

Subject:

Systemat <ssessment of Licensee Performance (SALP)

Report No. 50-286/85-98 On July 29, 1987, the NRC Region I SALP Board reviewed and evaluated the performance of activities associated with the Indian Point Unit 3 Nuclear Power Plant from December 1, 1985 to May 31, 1987. This assessment is documented in the enclosed SALP Board Report dated July 29, 1987. A meeting to discuss this assessment will be scheduled for a mutually acceptable date.

Although we have identified areas for improvement, we find that your overall performance was satisf actory. At the meeting, you should be prepared to -

discuss our assessment and ycur plans to ensure improved or continued emphasis upon those activities which would have a positive effect upon performance. In particular, because of the continued high number of reactor trips and the inconsistent performance in the opervions :nd angineering areas, you should be prepared to discuss activities and initiatives to reduce challenges to the reactor protection system and to improve your proiedures. Any other comments you may have regarding our report may be discussec. Additionally, you may provide written comments within 30 days after the mu ting.

In addition, the NRC Safety System Outage Modification Inspection (550MI) team

recently completed detailed reviews of the engineering design, installation, and testing required to support modifications implemented during the recent refueling outage. Because the majority of the SSCHI site activities (May 11 -

July 24, 1987) and your refueling / modification outage (May 1 - September 4, 1987) took place outside this SALP assessment period, e.ortain weakeasses which were iden'.ified as a result of the SSCMI, and presented to you by.the team, are not included in this Report. Similarly, the recent findings from the NRC Environmental Qualification team inspection (September 21-25, 1987) are no.

included in this Report. Nevertheless, we feel that the findings of these teams have shot i fundamental and significant weaknesses, that they are indicative of declining performance in several areas, and therefore you should also be prepared to discuss these finding during this SALP meeting.

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Power Authority of the State of New York 2

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Following our meeting and receipt of your response, the enc 1csed report, your written response (if deemed necessary), and a sumrary of our findings and planned actions will be placed in-the NRC Dublic Document Rorm. ,

Your cooperation is appreciated. l Sincerely, [

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& . 'j~, h - LR William T. Russell ' '

Regional Administrator r

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Enclosure:

NRC Region I SALP Report No. 50-286/85-98 r

cc w/ enc 1

J. Phillip Bayne, President  !

J. C. Brons, Executive Vice President - Nuclear Generation Gerald C. Goldstein, Assistant General Counsel ,

A. Klausmann, Senior Vice President, Appraisal and Compliance Services F. X. Pindar, Quality Assurance Superintendent i

G. M. Wilverding, Chairman, Safety Review Committee <

Robert Spring, Director, Regulatory Affairs (Con Ed)

R. E. Svedle, Vice President Nuclear Suppo.t R. Burns, Vice President Nuclear Operations L i S. S. Zulla, Vice Pred dent Nuclear Engineering  :

P. Kokolakis, Director Nuclear Licensing - PWR  !

1 NRC Licensing Project Manager  :

Dept. of Public Service, S*'.a of New York Public Document Room (POR) t Local Public Document Room (LPJR) ,

1 Nuclear Safety Information Center (NSIC) f I NRC Resident Inspector i j State of New York j l Chairman Zech -

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) Cc-iissioner Roberts I Commissioner Bernthal l Commissioner Carr f Commissioner Rogers ..

X. Abraham (PA0 - 8)

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ENCLOSURE 4 -

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U.S. NUCLEAR REGULATORY COMMISSION L i .

REGION I SYSTEMATIC ASSESSMENT OF k!CENSEE PERFORMANCE INSPECTION REPORT 50-286/85 AMENDED REPORf POWER AUTHORITY OF THE STATE OF NEW YORK INDIAN POINT UNIT 3 NUCLEAR POWER PLANT 3

i ASSESSMENT PERIOD - DECEMBER 1, 1985 TO MAY 31, 1987 i

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... TABLE OF CONTENTS .

.c Page I. INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . 1 A. Purpose and Overview . . . . . . . . . . . . . . . . . . . 1 B. SALP Board Members . . . . . . . . . . . . .

1 C. Background ... . . . . . . . . . . . . . . . . .. ... . . ...

... 2 II.

CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 III.

SUMMARY

OF RESULTS t

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A. Overall Facility Evaluation ......... . . . . . 6 e F:etitty o eeformance . . . . . . . . . ,. . . .. ... 7 i IV. FUNCTIONAL AREA ASSE!$McNT , . . . . . . . . . . . . . . . . . 8 I A. P l a n t Op e ra t i o n s . . . . . . . . . . . . . . . . . . . . ." t

• B. Radiological Centrols .................. 12 C. Maintenance ........... . . . . . . . . . . . 16

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0. Surveillance . . . . . . . . . . . ........c.. 19 .

E. Emargency Preparedness . , . . . . . . . . . . . . . . . . 21 F. Security and Safeguards ................. 23 L G. Trainirg and Qualification Effectiveness . . . . . . . . . 26 H. Assurance of Quality . . . . . . . . . . . . . ... ... 28 I I. Engineering and Technical Support. . ...... 31 Outages. . . . . . . . . . . . . . . . . . . . . . . . . . ,33 J. . . ..

X. . Licensing Activities . . . . . . . . . . . . . . . . . . 35 i

V. SUPPORTING DATA AND SUMMARIES ...

...,......... 37 A. Investigations and Allegation Review. . . .. . .. . . . 37

3. Escalated Enforcement Actions . . . . . . . . . . . 37 C. Management Cor.ferences. . . . . . . . . . . . . . .. .. .. . 37

0. Licensee Event Reports . . . . . . . . . . . . . . . . . . 38 E. Automatic Reactor Trips, Engineered Safeguard Actuations and Unplanned Shutdowns . . . . . . . . . . . . . . . . 39 F. Licensing Activities . . . . . . . . . . . . . . . . . . . 39  ;

TABLES "

Table 1 - Licensee Event Reports by Functional Area . ... . . . . . T1-1 [

Table 2 - Inspection Hours Summary . . . . . . . . . . . . . . . . . . T2-1 l Table 3 - Enforcement Summary . . . . . . . . . . . . . . . . . . . 73-1 -

Table 4 - Inspection Report Activities . . . . . . . . . . . . . . .*, T4-1 Table 5 - Automatic Reactor Trips. Engineered Safeguard Actuations '

and Unplanned Shutdowns ................. T5-1

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Tabl e 6 - SALP Hi s tory . . . . . . . . . . . . , .. ..... . . . . T6-1 t

Table 7 - Nunbe r of Days Shutcown. . . . . . . . . . . . . . . . . . . T7-1 I i

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I. INTkODUCTION A. Purpose and Overview The Systematic Asse:,sment of Ifcensee Performance (SALP) is an integrated NRC staff effort to collect the available observations and  ;

data on a periodic basis and to evaluate licensee performance based upon this information. SALP is supplemental to normal regulatory processes used to ensure compliance to NRC rules and regulations.

SALP is intended to be sufficiently diagnostic to provice a rational basis for allocating NRC resources and to provide meaningful guidance to the licensee's management to promote quality and safety of plant construction and operation. ,

An NRC Indian Potht Unit 3 SALP Board, composed of the stuf members listed below, met on July 29, 1987, to review the collect 4en of performance observations and data to assess the licensea. performance in accordance with the guidance in NRC Manual Chapter 0516,

) "Systematic Assessment of Licensee Performance." A summary of the guidance and evaluation criteria is provided in Section II of this

, report.

This report is the SALP Board's assessment of the licenne's safety performance at the Indian Point Unit 3 Nuclear Power Plant for the -

l period December 1, 1985 through May 31, 1987. The summary findings '

and totals reflect the eighteen month assessment period.

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B. SALD Board Me?.bers Chairman S. J. Collins, Acting Ofrector, Division of {

I Reactor Projects '

i i Me-bers J. Joyner, Acting Director, Division of Radiation Safety and Safeguards l W. Johnston, Acting Of rector, Division of r

Reactor Safety l

L. Bettenhausen, Chief, Operations Branch, DRS  :

1 R. Gailo, Chief, Projects Branch 2  !

J. Johnson, Chief Reactor Projects Section 2C  :

P. Koltay, Senior Resident inspector, Indian  !

l 'roint Unit 3 L R. Canra, Acting Ofrector Project  !

Directorate I-1 i M. Slosson, Project Manager, Project  !

Ofrectorate I-1  ;

l Other Attendees R. Barkley, Resident inspector  :

A. Weacock, Acting Chief Facilities Radiation I Protection Section i R. Struckmeyer, Acting Chief, Etfluents and  !

Radiation Protection Section [t I

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G. Meyer, Project Engineer N. Perry, Reactor Engineer N. Blumberg, Chief, h erational Programs Section J. Ourr, Acting Oeputy Director, Division of Reactor Safety

• C. Background Licensee Activities 1.

This assessmen&, period comme ^ed on December 1, 1985 w;th the unit in power ascension following a suberitical reactor trip caused by the removal of a control fuse fror the source range power supply drawer while trouble shooting equipment.

On February 28, 1986, the reactor tripped from 100% power due to low steam generator level. The plant was returned to power operations on March 1.

On March 13, 1986 the licensee conducted a partial emergency exercise. State and local authorities participated on a limited easis.

Or) April 25, 1986, the licensee initiated a contrelled shutdown tb commence a planned mid-eycle outage. During the outage, the Ifeensee completed sludge lincing, eddy current testing, and tube plugging operations on all four steam generators.

Modifications were completed in the areas of nvironmental qualifications and TMI Action Plan items.

During the unit startup on May 18, 1936 from the mid-cycle outage, three reactor trips occur *ed:

May 18 h actor trip from 2". power due to low level in No. 32

,t:am generator May 19 Reactor trip from 25% power due to high level in No.

34 steam generator May 19 Reactor trip due to the failure of a control room '

operator to block the source range high fldx trip during startap The unit was maintained in hot shutdown while the environmer.tal avalification tests of certain sections of safety-related instrument wiring were under review and was returned to power operation on May 21, 1986. *

During the next month, the following events occurred

3 May 23 Reactor trip from 73% power due to a turbine generator overspeed trip initiated while overspeed protection surveillance testing was underway May 24 Manual shutdown from 20% power to inspect turbine for high vibrations. The unit restarted on May 25.

May 26 Reactor trip from 16% power due to low level in No.

34 steam generator while conducting manual shutdown from 70% oower to investigate combustible gas accumulation in main transformer No. 32 May 27 Inadvertent safety injection signal generated while shutdown during turbine first stage pressure surveil-lance testing. No water was injected into the reactor coolant system.

May 28 The unit rusumed power operation without main trans-former No. 32 June 7 The licensee completed a manual plant shutdown from 57% power in order to return No. 32 main transformer to service Jbne 14 Another plant shutdown was accomplished from 57%

power to once again inspect No. 32 main transformer which continued to generate combustible gases.

June 15 The unit was returned to. power operations with only No. 31 transformer in service. Unit load was limited to 550 MWe. The licensee proceeded to replace the No.

32 main transformer. 1 On July 5, the failure of an expansion joint between the low pressure turbine casing and the condenser resulted in a gradual loss of ccadenser vacuum. Plant shutdown was completed that same day. During unit restart on July 6, an 18 inch piece of blading brcne off from the No. 31 low pressure turbine rotor.

On July 6, the unit entered a 57-day forced outage to repair the damage to the low pressure turbines.

The unit was returned tc service on September 2, 1986. On September 5, a reactor trip occurred from 56% power when the main generator disconnect switch opened due to a control circuit fault. On September 9, the reactor tripped from 99% power due to low steam generator level caused by a feedwater per urbation.

The unit was returnec to power operations on September 10.

• An evaluation cf the plant's performance was conducted by tne Institute of Nv;1 ear Power Operations on September 22 - October 3

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1986. The licensee was also formally inducted into the National Academy of Nuclear Training on October 30, 1986.

On November 14, 1986, a reactor trip from 100*; pu.er occurred due to the temporary loss of 125V DC control power from the No.

31 OC distribution panel. The unit returned to service the next day.

On January 31, the reactor tripped from 100% power due to low steam generator level. Power operation resumed on February 2.

On February 11, the No. 34 instrument bus output breaker opened .,

on overload, eventually resulting in an unexpected reactor trip and safety injection system actuation. The licensee later determined that there is a high probability for reactor trip and safety injection signals to be generated upon the loss of either 33 or 34 instrument busses. Power operations resumed on February 13, 1937.

An unplanned 7-day outage was initiated on March 27 to replace the seal package on No. 33 reactor coolant pump. On April 4, the reactor tripped from 10% power due to problems with the turbine governor controls. Power operations resumed later that day. On April 17, 1987, the reactor tripped from 100% power and a safety injection system actuation occurred cue to an electrical perturbation on instrument bus No. 33. The unit was returned to power operations on April 18. .

On May 1, the licensee initiated a controlled shutdown to begin .

the cycle 5/6 refueling / maintenance outage. The outage is scheduled to last 89 days.

2. Inspection Activities Aa NRC senior resident inspector was assigned for the entire assessment period; an additional residens inspector was assigned in June 1986.

During an 18 month assessment period, the NRC conducted a total of 4405 inspection hours equating to 2935 hours0.034 days <br />0.815 hours <br />0.00485 weeks <br />0.00112 months <br /> on an annualized basis. Functional Area wistribution of inspection hours is detailed in Table 2.

• During the period, three NRC team inspections were conducted in the following areas:

a. Fire Protection / alternate safe shutdown, in accordance with 10 CFR 50, Appendix R requirements.

b. Operational Assessment Team

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c. Safety System Outage Modification Inspection (First two week phase of a 4 phase inspection program was conducted during the period.)

Fire protection and housekeeping have not been assessed as a discrete functional area within this report. Evaluations based on the day-to-day observations of inspectors have been incorporated into evaluations of appropriate functional areas.

II. CRITERIA The following evaluation criteria were used to assess each functional area:

1. Management involvement and centrol in assuring quality.

2. Approach to resolution of technical issues from a safety standpoint.

3. Responsiveness to NRC initiatives.

4. Enforcement history.

L 5. Reporting and analysis of reportable events.

6. Staffing (including management).

7. Training effectiveness and cualification.

To provide consistent evaluation of licensee performance.. attributes associated with each criterion and describing the characteristics applicable to Category 1, 2, and 3 performance were applied as discussed in NRC Manual Chapter 0516, part II and Table 1.

The SALP Board conclusions were categorized as follows:

Category 1. Reduced NRC attention may be appropriate. Licensee management attention and involvement are aggressive and oriented toward nuclear safety; licensee resources are ample and effectively used so that a high level of performance with respect to operational safety or construction is being achieved. .

Category 2. NRC attention should be maintained at normal levels.

Licensee management attention and involvement arc evident and are concerned with nuclear safety; licensee resources are adeouate and reasonably effective so that satisfactory performance wis. respect to operational safety or constructico is being achieved.

Category 3. Both NRC and licensee attention should be increased.

Licensee management attention or involvement is acceptable and considers nuclear safety, but weaknesses are evident; licensee resources appear to be strained or not effectively used so that minimally satisfactory

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er,'ormance with respect to Operational safety or construction is being achieved.

The SALP Board may determine to include an appraisal of the performance trend of a functional area. Normally, this performance trend is only used' where both a definite trend of performance is discernable to the Board and i the Board believes that continuation of the trend may result in a change i of performance level. Improving (declining) trend is defined as: ,

4 Licensee performance was determined to be imoroving (declining) near the close of the assessment period. The SALP Board found no such definite trends in performance. Accordingly, no performance trends were included and the references to trend in each functional area have been deleted.

III.

SUMMARY

OF RESULTS A. Overall Facility Evaluation During this assessment period, a dedicated and competent staff continued to implement safe and acceptable practices in all plant areas. However, the staff functions in an environment thtt encourages individual initiative without relying on detailed and comprehensive procedures. Management attention is required to recognize the need for prompt development of a program resulting in the de';siopment and cohesive support of NYPA procedures.

Long term corrective actions to improve procedures have been initiated several times; however, the effort continues to encounter delays due to inadequate management support and a lack of accountability, r

i We continue to be concerned regarding the number of events resulting in challenges to the reactor protection system. NYPA has not been i

able to effectively address the problem since the 1980 SALP period.

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While this area is receiving increased management attention, more effort needs to be placed on post-trip reviews, root cause analyses, r

and the preparation of Licensee Event Reports (LERs).

Our assessment notes continued good cerformance in the areas of surveillance and maintenance. The expansion of the preventiva maintenance program has resulted in improved equipment availability.

Additionally, strong management support of the radiological protection and cecurity and safeguards areas was responsible for continued good performance.

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9. Facility Performance Functional Catecory Category Recent Area Last Period This Peried Trend A. Plant Operations 2 2 B. Radiological Controls 1 1 C. Maintenance 1 1 D. Surveillance 1 1 E .' Emergency Preparedness 1 1

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1 G. Training and Qualification Effectiveness 2 2 H. Assurance of Quality 2 2

1. Engineering &

Technical Support N/A* 2 J. Outages 1 Nc,t Rated K. Licensing Activities 2 1

• Not assessed as a distinct functional area last period.

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IV. PJNCTIONAL AREA ASSESSMENTS A. Plant Ooerations (474. 2060 hours0.0238 days <br />0.572 hours <br />0.00341 weeks <br />7.8383e-4 months <br />) 1, Analysis During the previous SALP period, this area was rated Category 2.

Licensee strengths were identified in the areas of licensed operator training, implementation of the fire protection program and house-keeping. Deficiencies were identified in root cause analysis and post-trip reviews and contributed significantly to the repetitive i reactor trips caused by longstanding design problems and personnel errors. Increased management attention to the reduction of the reactor trips was recommended by the assessment. Additional weak-nesses were noted in the areas of control room log entries, timeliness of procedure and as-huilt drawing updates, and control room decorum.

During this assessment period, plans operations were under continual review by the resident inspectors. RSgionally based inspectors assessed licensed operators' familiarity with the Emergency Operating Procedures (EOPs), and evaluated licensee response to the February ll,1987 reactor trip / safety injection. A special inspection by an Operational Assessment Team was conducted in April-May, 1987, which observ'ed and evaluated the performance of the operating staff during-a reactor shutdown for refueling. An audit of the licensee's compliance with the requirements of 10 CFR 50, Appendix R Fire protection was also conducted in July 1986.

Site management takes an active role in the day-to-day operation of the plant. Management personnel regularly tour accessible plant areas. Corporate operations management are frequently on site generally participating in activities. Daily management meetings are well attended and greatly enhance communication and departmental interfaces. Cooperation exists at all levels as evidenced through routine QA/QC involvement in all maintenance and modification activities including the balance of plant areas.

Ourtag this 18-month assessment period, 13 reactor trips and three inadvertent safety injection system actuations occurred. While design changes, enhanced maintenance and improved controls,over L contractor activities eliminated the root causes for eleven of the fifteen reactor trips of the previous SALP period, management has not been effective in the control and reduction of the number of challenges to the reactor protection system.

3ight of the reactor trips and two inadvertent safety injection a.tuatiens can be attributed to equipment failures and design deficiencies in the instrument bus loading arrangement and main boiler feed pump controls. However, known design defittencies, lack of attention to detail (e.g. failure to follow proceci r weak procedures, control recm activity during startup, failu,w to

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adequately review equipment failure trends), a sense of urgency o return to power operations, and minor errors in judgment contr uted to reactor trips.

The reactor post-trip reviews root cause analysis and uni restart decision making processes are,not riefined in the applica e proce-dures. The quality of post-trip reviews is inconsisten and tends to be poor when personnel error is involved in the event ich initiated the reactrer trip. Restart decisions &re made at the peropriate managener.t level, and are generally based on the r acoment of defee:1ve equipcent. However, in-depth review fo e',tablishing root .

caus s for equipment failures is not supported m/.nagement and was not evident following the post-trip review pro ss.

Following a series of five reactor trips du ng M/y 1986, post-trip review and root cause analy.ses became more deta17dd under their own initiative. Increased caution was exere ed during return to power operations.

However, appropriate prece ure changes to support improvements in this area have not be made. A trip recuction review committee, formed in Septembe 1986, presented its findings and recomendations to site manage nt early this year. At the end of the assessment period, the ce ttee's recommendations still have not been implemented.

Licensee Event Reports (LERs in the operations area were all generated by personnel erro initiated events. (LERs 86-02, 86-06 86-07, 86-09, 86-12 ahd 87 6). In all instances, poor judgment, lack of attention to det 1, lack of knowledge or a combination of the above were contribu ing factors to the event.

A major violation of echnical Specif t-ation requirements (EA 86-197) resulted in escala d enforcement action and a civil penalty. The violation involve licensed operators' failure to recognize opera-bility requirem ts of safety-related pumps for existing plant conditions. T event' pointed to weaknesses in the routine plant operational tivities. These included the lack of positive control over the ce rol room activities by senior operators, intermittent lack of a ention to detail by operators and reduced effectiveness of existing recedures. The same weaknesses also contributed to two other p oblems: the failure to maintain detailed control room logs and t failure to follow Administrative Procedures regulating work acti ties on safety related systems. Short term corrective actions wer timely. Long term corrective action involving procedure use, t

ining and changes in operational philosophy as it relates to rocedure uso, log keeping, and coordination of control room activi-ties is baginning to show improvement.

The lack of a formal tracking mechanism for the field implementation of Technical Specification changes resulted in a violation of Technical Specification requirements regarding Lew Temperature 7

93 adequately review equipment failure trends), and minor errors in judgment cor.tributed to reactor trips.

, The reactor post-tria reviews, root cause analysis and unit restart decision making processes are not defined in the applicable proce-dures. The quality of post-trip reviews is inconsistent and tends to i be poor when personnel error is involved in the event which initiated the reactor trip. Restart decisions are made at the appropriate management level, and are generaliy based or, the replacement of defective equipment. However, in-depth review for establishing root causes for equiement failures is not supported by management and was not evident fol' wing the post-trip review process.

Following a series of five reactor trips during May 1986, post-trip review and root cause Analyses became more detailed under their own initiative. Increased caution was exere sed during return to power operations. However, appropriate procedure changes to support improvements in this area have not been made. A trip reduction review comittee, formed in Septemeer 1956, presented its findings and reccmmendations to site management early this year. At the end of the assessment period, the committee's recommendations sttl1 have not been implemented.

Licensee Event Reports (LERs) in the operations area were all generated by personnel error initiated events. (LERs 86-02, h % ,

86-07, 86-09, S6-12 and 87-06). In all instances, poor judgwent, lack of attention to detail, lack of knowledge or a comoination of tne above were contributing factors to the event.

A major violation of Technical Specification requirements (EA 86-197) resulted in escalated enforcement action and a civil penalty. The violation involved licensed operators' failure to recognize opera-bility requirements of safety-related pumps for existing plant conditions. The event pointed to weaknessts in the routine plant operational activities. These included the lack of positive control over the control room activities by senior operators, intermittent lack of attention to detail by operators and reduced efft.ctiveness of existing procedures. The same weakntsses also contributed to two other problems: the failure to maintain detailed control room logs and the failure to follow Administrative Procsdures regulating work activities on safety-related systems. Short :.erm corrective actions were timely. Lcng term corrective action involving procedure use, training and changes in operational pt. losophy as it relates to procedure use, log keeping, and coordination of control room activi-ties is beginning to show improvr; ment.

The lack of a tormal tracking mechanism fer the field implementation of Technical Specification changes resulted in a violation of Technical Specification requirements regaMing Low Temperature

10 Overpressure protection during a controlled plant shutdown. (see Section H). The violation was identified by a recent licensee.

initiated quality assurance audit of the Operations 0 partment's activities.

The responses to NRC inspector immediate concerns are timely and technically sound. However, correctivs actions requiring long term resolutions appear to lose momentum in the planning stages and sub:equently in the field. This is evidenced in the licensee's lact of controls over long-term followup of procedural inaccuracies, iaatequate control ecom log entries, inadequate surveillance and preventative maintenance of No. 33 station battery, the updating of ,,

as-built drawings, and the equipment tagging program. Written responses to NRC concerns are detailed and timely; however, they often include extraneous information and seem to lessen the impor-tance of the event.

Licensed control room operators continued to demonstrate significant l operating experience and specific knowledge of the Operations procedures. Experienced senior operators provide thorough on-the-job training to operators who recently have been upgraded to senior licenses, and subsequently assumed sontor operster responsibilities.

Weaknesses related to control room activities identified early in the SAlp eriod, such as inconsistencies in the thosaughr.ess of shift turnovers, lack of detail in log keepfng, periodic lack of attention to detail, and allowing repair :nd surveillance teams to distract operators have shown signs of improvement. However, due to an apparent lack of continued folloc<p on these programmatic enhance-ments the improvements have been inconsistent. In general, programs directing operating activities are weak in that a number of activi-ties are conducted based on informal guidance or past practices. As a result of the absence of strong, formalized programs, several notable larses in oper 4tions have occurred (e.g., Enforcement Action 86-197).

t Control room decorum has been enhanced by the removal of uncontrolled operator aids such as graphs and sections of procedures taped to various control room furniture and control cabinets. Cont,rol room formality has improved through more stringent access controls and increased emphasis on the leadership role of senior operators.

Operating crews demonstrated their ability to respond to plant transients. Through the proper use of emergency procedures, operators stabilized the plant in a safe mode following two events which involved unexpected equipment resconses to losses of instrument bus voltage and recovered from a loss of reactor coolant pressure' transient.

t .

11 i

The licensee continues to maintain an effective fire protection and I prevention program. Corporate management involvement is evident through regular audits of site activities.

In summary, the staff has good attitudes towards safe operation of the plant and operating activities are safe and acceptable in all ,

areas. However, operations are based on past experience and informal guidance without the tupport of clear and comprehensive precedures. '

The lack of strong support for providing detailed procedures and procedural usage has contributed to cperational problems which originally appeared to be isolated cases. I i

2. Conclusion I Rating: Category 2

3. Board Recommeneatiens I Licensee.

6 Continued management attention is needed in the areas of post-trip ,

reviews, rcot cause analysis, and the resolution and complation of '

long term corrective actions. Special attention must be given to the  :

reduction of the number of reactor trips.

l NRC:

Conduct inspection and subsequent management meeting to review NYPA i trip reduction efforts.

. I r

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12 S. Badiological Controls (194, 631 hours0.0073 days <br />0.175 hours <br />0.00104 weeks <br />2.400955e-4 months <br />) ,

1. Analysis The radiological controls area was rated Category 1 during the previous assessment peried. Licensee strengths were noted in the areas of management controls and audits, organization and staffing, and training and qt'alif.ication of personnel.

• Ouring this assessment period, six routine and two special radiologi.-

cal safety inspections were conducted. The two special inspections focused on the remaining items related to NUREG-0737, (Post Accident Sampling, monitoring, and analysis), and a dosimetry program appraisal. Resident inspectors oerformed routine reviews of this program area throughout the assessment period.

Radiation Protection An u9 planned exposure to the upper arm (2.6 rem) resulted from inadequate radiological control of work. The licensee provided prompt and extensive corrective actions, and the event is considered isolate 1 and non programmatic.

There were sufficient numbers of highly qualified and experienced personnel in the radiation protection organization tc ensure a high degree of management oversight, technical review, and program implementation. One exception to this was noted during a review of ,

the dosimetry program which found an insufficient number of dedicated personnel te provide technical development and program oversight.

Review of this area found inadequate procedures, lack of sof tware development and testing, lack of resolution of technical problems, and an absence of ? system for identifying and correcting program 4

deficiencies, t

The Health Physics technician training and qualificatinn program was found to be well-defined and implem'ented with dedicated staff and i resource

• with strengths including well organized initial training of l incoming ..ealth Physics technicians, adequate lesson plans for incoming craft personnel, and frequent refresher training of plant Health Physics technicians. The periodic retraining of site personnel was also found to be well-established and thoroughly docu-mented. However, an area in need of improvement was identified, in that, the retraining program did not include practical factors (i.e.,

use of dosimetry and HP instrumentation, proper use of protective clothing, and proper frisking technique reviews).

In addition to the deficiencies in the dosimetry program procedures,  !

weaknesses were also identified in the access control procedures. The procedures were confusing and lacking in clari.ty, with incomplete

directions in the performance of certain tasks. The licensee ini-l tiated immediate corrective actions and assigned two indivicuals 1

. 13 dedicated to rewriting and reorganizing the procedures into a coherent system. -

The A MRA program was found to be highly effective due to management support and commitment, a formally established A MRA program with a dedicated AMRA Engineer, adequate staffing support and detailed procedures. Cumulative 1936 radiation exposure, which included a mid-cycle outage, totaled 185 person-rem, indicating an aggressive A uRA program. There was also ample evidence of thorough pre-planning for the 1987 ten year inservice inspection outage, with many dose-saving techniques and methods incorporated into the work  !

packages.

The external exposure control program was of high quality, with an effective radiation work permit program that established clear reouirements for work in radiation and high radiation areas. The 1936 mid-cycle steam generator inspection outage was handled compe-tently. Health Physics personnel exercised positive control over wors, activities, with the exception of one incident in which two individuals entered a high radiation area without the appropriate Health Physics controls (i.e., continual coverage with a Health Physics technician providing moattoring with a dose rate instrument

• and stay-time determination). This incident resulted in one indivi-dual exceeding the administrative whole body dose limit. It was also determined that there was a potential for ex:eeding the regulatory limit, which resulted in a Severity Level III violation. However, this was an isolated incident and was not reflective of the licensee's sustained good performe.nce in this area.

The internal exposure controls program was found to be uffectively implemented. Further, plant management fully supports en aggressive contamination control program that has minimized plant contamination and subsequent airborne activity concerns.

Radiation protection equipment and facilities were reviewed and found to be well-maintained, with ample supplies available. ,

Instrumentation is state-of-the-art and maintained in accordance with the highest industry standards. Strong management support and commitment is evidenced in the planned construr: tion of a cew instrument calibration and repair facility.

Radioactive Waste Management / Effluent Controls i The Itcensee has an adequate radiochemical measurement program. i However, weaknesses were noted in regard to the lack of an inter- and l intra- laboratory qu ' ' controls program, inconsistent use of '

control charts for m<asurement systems, and the level of detail pro-vided in a procedure for gamma spectral analysis. In adoition, QA audits in this area were weak due to the lack of experienced QA personnel in this ciscipline. Notaole improvements in these areas >

14 were evidenced near the end of this period. Also, the licensee has

' created a new chemical engineer's position that should strengthen management controls in this area. One violation of procedures was identified. Adequate corrective actions were promptly effected to prevent such a recurrence. Except for the examples noted above, procedures generally followed industry practices.

The licensee *has in place an agressive solid radioactive waste management program involving volume reduction methods and contamina-tien control. As a result, the licersee has generated less than 11,000 cubic feet of solid radwaste for burial in the last two years.

The program is well funded and fully staffed.

Review of the licensee's program for gaseous and liquid radioactive effluent controls determined that radwaste treatment systems were properly used. However, the large number of unit trips and power changes led to effluent releases of fission and activation gas curie quantities greater than the plant's historical average but within the T:rchnical Specification limits. The licensee generally demonstrated positive management conteols for conducting routine surveillances.

Quality assurance audits were performed by off-site corporate staff to determine everall program capabilities and wea'knesses. On-site quality control surveillances verified that procedural and regulatory requirements are met. Both of these type audits were generally comprehensive and timely. The licensee's responses to NRC identifi1d deficiencies in the radwaste management area were generally timely and technically sound.

Transcertation Review of ongoing transportation activities indicated adherence to shipping requirements and licensee procedures, maintaining an effective transportation program. Some weaknesses in training of tecnnicians and QA/QC inspectors were noted during the NRC inspection early in the assessment period. The licensee has since ungraded training activities and has adequately dealt with these weaknesses.

Adequate management attention to planning and implementation of the program was evident.

Water Chemistry Controls Reviews of the laboratory water chemistry control program indicated that a generally adequate program was developed and implemented.

Since the previous assessment period, the licensee developed a laboratory measurement control program incorporating the use of control charts. However, coatrol charts still have to be established for metal analyses with plasma emission spectrophotometry and for'

15 general spectrochotometry. Since the last period, the licensee also developed an inter-laboratory cross-check program with other laboratories and established an intra-laboratory cross check program as well. Thus as a result of previous NRC findings, the licensee has taken effective corrective actions.

In summary, the licensee maintains a well-implemented and well-main-tained radiological controls program, with strong management commitment to plant cleanliness, radiological controls, and ALARA.

Minor and isolated weaknesses were identified in dosimetry staffing, procedures, and quality control of effluent measurements that are not

. indicative of major programmatic problems. The licensee's responses to NRC findings continued to be prompt and effective.

2. Conclusion ,

Category: 1

3. Board Recomendation Licensee: ,

None NRC:

None e t 9

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" Ov y

C. Maintenance (144, 604 hours0.00699 days <br />0.168 hours <br />9.986772e-4 weeks <br />2.29822e-4 months <br />) y ,

1. Analysis

.During the previous assessment period, the maintenance are was rated Category 1. Licensee strengths were noted in the area of management involvement, control of utstanding work requests, and e implemen- t tation of the preventive maintenance program. Defici cies regarding the removal of maintenance related equipment followi maintenance activities were identified during the same period. ,

During this assessment period, two inspections region-based .

insptetors evaluated the licensee's maintenanc activities. One dealt with the requirements of IE Bulletin 8 02, "Degradation of Threaded Fasteners in the Reactor Coolant P essure Boundary". The other covered the Environmental Qualifica on (EQ) of Limitorque valve actuator wiring and Rsychem cable olices. The Operational Assessment Team also reviewed maintena e activities and interface between operations and e.atntenance de artments. The resident d

inspecters observed maintenance act vities on a day-to-day basis, l

J

" A high tiegree of management and gineering involvement continued to .

be evident in all maintenance d modification activities. The l

control of maintenance activi es is generally effective and super-visors are actively involve in day-to-day maintenance activities. 1 l

Improved control over cent ctor activities eliminated contractor initiated inaevertent ch lenges to the reactor protection system.

However, during the a essment period, four reactor trips can be

, attributed to mainte nee related equipment failure. One challenge to the reactor pro etion system resulted from the lack of .

conservatism, dur g routine maintenance of a main boiler feed pu.np, by failing to ri lace marginal parts. Lack V preventive maintenance programs for c .tical balance of plant equi pent, such as the main l

boiler feed mp discharge check valve, main generator output motor '

disconnect itch, and safety-related relays (see surveillance section) so resulted in reactor trips. In reference to a previous NRC conc n, long term corrective action to ensure proper housekeep-l ing fo owing maintenance activities proved to be effective. Ade-  !

quate communications with Operations and the review of work act ittes by supervisory personnel are evident. The overall {

pr grams to administer these activities are well established and are scribed in sufficient detail in station and departmental adminis-trative procedures. .

Maintenance personnel are experienced and well trained. Experienced

! personnel in the department share their experience with .iunior -

members. Juality control oversight of maintenance activities is evident and cooperation between the two departments appears to be ,

good.

16 A .

C. Maintenance _(144, 604 hours0.00699 days <br />0.168 hours <br />9.986772e-4 weeks <br />2.29822e-4 months <br />) 1

1. Analysis During the previous assessment period, the maintenance area was rated Category 1. Licensee strengths were noted in tae area of management involvement, control of outstanding work requasts, and the implemen-tation of the preventfve maintenance orogram. Deficiencies regarding the removal of maintenance related equipment following maintenaace activities were identified during the same period.

During this assessment period, two inspections Ly ragion-based inspectors evaluated the licensee's maintenance activities. One deelt with the requirements of IE Bulletin 82-02, "Degradation of Threaded Fasteners in tae Reactor Coolant Pressure Boundary". The other covered the Environmental Qualification (EQ) of Limitorque valve actuator wiring and Raychem cable splices. The Operational Assessment Team also reviewed maintenance activities and interface between operations and maintenance departments. The resident inspectors observed maintenance activities on a day-to-day basis.

A high degree of management and engineering involvement continaed to be evident in all maintenance and modification activities. The control of maintenance activities is generally effective and super-visors'are actively involved in day-to-day maintenance activities.

Improved contro* over contractor attivities eliminated contractor initiated inadvertent challenges to the reactor protection system.

However, during the assessment period, four reactor trips can be attributed to maintenance related equipment failure. One challenge to the reactor protection system resulted from the improper instal?ation of a feedwater pump seal by a contractor during the 1985 refueling outage. Lack of preventive mainten:nce programs for critical balanca of plant equipment, such as the main boiler feed pu p discharge check valve, main generator output motor disconnect switch, and safety-related relays (see surveillance section) also resulted in reactor trips. In reference to a previous NRC concern, long term corrective action to ensure proper housekeeping following maintenance activities proved te be effective. Adequate communi-cations with Operations and the review of work activities by super-visory personnel are evident. The ove all programs to adiinister these activities are well established and are described in sufficient detail in station and departmental administrative procedures.

Maintenance personnel are experienced and well trained. Experienced personnel in the department share their experience with junior members. Quality control oversight of maintenance activities is evident and cooperation between the two departments appears to be good.

7 17 .

The department ontinued to be fally :taf fed and maintained on all shifts. Six mechanics completed tre accredited 8 to 12 week training course. An addittenal 14 mechanics are currently in the training course. Specia'. courses offered for the maintenance department have been attended by 48 people. A maintenance supervisor was successful in receiving a senior operator's license, bringing increased unt.or-

$tading of operational problems to the department. '

, Work requests for equipment affected by Technical Specification ,

limitations were handled promptly. During the assessment period, all i such maintenance was accomplished within the time limitations of the  !

Technical Specifications.. Lower priority major maintenance was

accomplished during scheduled and forced outages. A computerized work control system was placed in service. This tracking system i

4 includes ar. index of machinery, a categorized machine history, the status of work, requests on the item, and trending of equipment i failures .

The maintenance p ogram associated with the degradation of ;hreaded  !

fasteners in the reactor coolant pressure boundary of PWRs (Ref. IE Bulletin 82-02) was well documented. The maintenance procedures provide a more conservative approach than the manufacturer's recom-mendations in t5e area of bolt torquing and 1.brication. This conservatism was also demonstrated in the establishment of a proca- t dure fbr the irapection and replacement of steam p crator and pressurizer m4 way bolts. An example of the conservatism used by the licensee was also observed in the material acceptance area, where minor indications of degraded bolt conditions, acceptible under the manufacturer's criteria, were rejected. ,

The preventive maintenance progras was expanded during this period  !

with 407 items completed in 1986 compared to the 183 in 1985 and 86 i in 1934. The program is fully staffed, including a full-time engineer. In response to equipment failures, preventive mainter3nce has oten extended to include critical nonsafety-related equipment such as the main boiler feed pumps and hssociated discharge check valves. Inspection programs have also been developed to detect and correct corrosion / erosion in feedwater and main steam lines. One exception to the generally good practices in this area was the failure to include the station batteries into the preventive e.'nte-nance program. This is in spite of the fact that the deteriorated condition of the batteries, most notably station battery No. 33, was brought to th', licenseo's attention try the NRC.

In summary, tw saintenance department is well staffed and e x pe ri ent,ed. Management controls and good support of the maintenance departer.nt have allowed for a minimal backlog of corrective

• mainterance.

. 18

2. Conclusion Category 1

3. Board Recommendations Licensee:

None NRC:

None 4

i ll 9

q 9

e S

9 e

9

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. 19 D. Surveillance (St, 339 hours0.00392 days <br />0.0942 hours <br />5.605159e-4 weeks <br />1.289895e-4 months <br />)

1. Analysis During the previous assessment period, this area was rated Category

1. Surveillance activities were found to be well-organized and implemented. All surveillances were accomplished within the given time constraints. Weaknesses were noted in the clarity of procedural instructions. .

This area was under routine review by the resident inspectors.

Ouring the assessment period, one inspection was performed by regior.-based inssectors which concentrated in the areas of procedural adequacy and interceparteental support and cooperation.

Surveillance tests are the responsibility of several departments depending on the type -! surveillance. The operations, instrumenta-tion and cont *ols, technical services and fire safety departments all participate in surveillance testing. Howqvir, the Instrumentation and Controls department is responsible fer the majority of routine and troubleshooting-type, surveillance activities.

The objectives of the surveillance program are clearly delineated by administrative procedures. Surveillances are accomplished in accor-dance with approximately 750 surveillance procedures. The test program is scheduled and tracked through a co"puterized system. All surveillances were accomplished within the Technical Specification requirements. Surveillance procedures are currently being revised to reflect the recommendations mac'e by the Institute of Nuclear Power Operations (INPO) and Vestinghouse Owner's Group.

Two reactor trips and one Engineered Safeguard Feature actuation occurred during the assessment period during surveillance testing conducted by Instrumentation and Control technicians. Both trips were initiated by relay failures of the turbine Independent Electrical Overspeed Protection System (IEOPS). The failec relays were not identical and each had different functional requirements.

1 Post-trip reviews included the bench testing of all identical relays in the system. It appears that the corrective actions in this case

, were shortsighted. The commonality for these relays appears tn be age (15-20 years old) and there is an increase in relay failures in other plant systems. There is no trending analysis or a preventive maintenance program in place ta address this p oblem.

1 One minor problem was identified by the NRC regardiag the failure to calibrate instruments monitoring the temperature of the boric acid storage tanks due te a design deficiency that remained uncorrected since plant construction. The licensee chose not to clatbrate the l instruments oue to the difficulty involved; however their procedures l specifically recuired them to be periodically calibrated. The l licensee promotly corrected the situation when identified.

l l

. . ao i

The Instrue.4entation and Control department is fully staffed. A formal, accredited training program was implemented during this ,

period and completed by six technicians. Eleven other technicians '

completed various phases of the program. Twenty cou*ses are given as I part of continuing training.

I t

In summary, surveillahce testing continued to be successful in l confirming that safety-related equipment will function as required. I Weaknesses were noted in the area of equipment failure trending [

affecting all departments involved with surveillance activities, t There is also a lack of a preventive maintenance program for relays f installed to control safety-related equipment. '

2. Conclusion f Ratics: Category 1

3. Board Recemmendt.ttons Licensee: {

None NRC:

Nont

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1 E. Emergency preparedness (4L 162 hourg

1. Analysis i

During the previous assessment period, the licensee was rated i Category 1 in the area of emergency preparedness (Ep). This rating ,

was based on the results of the licensee's strong performance l observed during two NRC observed exercises.

During this assessment period, two emergency exercises were aise ,

observed, allowing the opportunity to closel> assess Itcensee 1 performance in response to emergencies. The first exercise was  !

conducted on March 12, 1986. During this exercise, the licensee i demonstrated particular strengins in command and control in Emergency Response Facilities training, and effective utilization of (

j communication equipment. No significant weaknesses were observed.

' The second exercise observed during this assessment period occurred i March 18, 1987, and similar positive findings were noted. Overall e,ffectiveness of command and control, timeliness of

• protective action recommendations, and training were noted as particular strengths.  ;

' Minor 'seaknesses were noted in the areas of preparation of pubite '

inferreation notices, personnel accountability, conduct of briefings, use ref Control Room communiertorr., and Operations Support Center and Technical Support Center activation. , I i

The licensee maintained effort s to mediate and successfully resolve I offsite issues regarcing the protection of school children as raised in the New York Public Interest Research Group (NYPIRG) petition. [

The Itcensee took the initiative by maintaining close liaison with -

the State, affected counties and the Federal Emergency Management  ;

Agency'(FEMA) while seeking the resolution of issues regarding reception centers, training of bus drivers, transportation resources, t agreements, and public information. Once these is'ues had been  !

specifically defined, they were adequately corrected by the State and [

affected counties with the licensee's assistance,

, The Emergency Response Facilities are generally well designed and l

maintained. The licensee is examining the possibility of increasing the size of the present Emergency Operations Facility (EOF) and met .

I with NRC Emergency Preparedness sta'f on March 26, 1987 to describe  !

planned EOF chantes. A computer enhancsd in-plant personnel acccunt- *

} ability system is being installed and is nearing completion. l

'l In summary, the licensee has taken prompt and responsive action to ,

address weaknesses identified in observed exercises. The licensee  !

has also been instrumental in seeking the resolution of offsite l issues to the satisfaction of the State, affected counties and FEMA. [

a

22 -

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2. Conclusion Category 1

3. Board R_ec_o mer ,ation Licensee: l None r

' I:

NRC:

l None O

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23 . s '

F. Security and Safe:uards (44, 163 hours0.00189 days <br />0.0453 hours <br />2.695106e-4 weeks <br />6.20215e-5 months <br />)

1. Analysis During the previous SALP assessment period, this functional area was ra+ed Category 1. Notable strengths included strong corporate and  !

plant management attention to and involvement in the program and l frequent corporate appraisals conducted by corporate security.

Three routine physical security Mspections and one routine material control and accounting inspection were conducted by region-based .

inspectors durir.g the assessment period. Two of the routine physical i

, security inspections we. 7de to review temporary changes to the NRC-approved security plan. In addition, 4 Regulatory Effectiveness Review (RER) was conducted on September 8 12, 1986 by the Office of Nuclear Materials Safety and Safeguards with assistance from members of the Army Special Forces. Routine resident inspections were conducted throughout the assessment period. .

The licensve continued to maintain their high level of performance.

The attitude of management toward the security program and the knowled:e of program objectives and attention to detail by the security supervisory staff are reflected in the excellent compliance records and in the high morale and professionalism exhibited by

' security force personnel. There were no violations identified by  !

either the licensee or the NRC during this period.

l Manage ent initiatives itolemented during the assessment period included formal training in hostage negotiations for security supervisors, establirhing a Training Coordinator, installing addi-tional key card rescers at evacuation assembly points and installing  !

additional search equipment in the secondary access point. The licensee had planned to enlarge the primary access point to provide enhanced control at the entrance to the protected area. This was not accomplished; however, it is still included in ;,he site security t master plan for the future. The primary access point is periodically '

congested; however, the security force is able to maintain effective contrel. l The Itcensee's security systems are state-of-the-art and have a high degree of reliability. This is as a result of periodic upgrading and l

aneffectivemaintenanceandtestingprogram,therebyrejucingthe

• need for manpower intensive compensatory measures. I r

The security management and supervisory staff consists of well qualified individuals. A management representative is assigned to l l

each shift to monitor program implementation and security force '

l performance. Staffing is adequate and well deployed to implement the program. [

l

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l The licensee's traleing program is carried out by a qualified, fuil

! time professional staff. The security training program is now i

monitored and reviewed continually by the Training Coordinator who is l a memoer of tne security suparvisory. staff. L l

Corporate managen. ant conti.ves to take an active interest in the site security program as evidenced by frequent in-deptn appraisals by members of the corporate secuiity staff. Recommendations resulting from these appraisals are promptly considered for incorporation into ,

the security program. Additionally, the annual audit of the security  ;

program was conducted by qualified quality assurance personnel. '

Recommendations contained in the ' adit report were promptly acted upon by the security organization.

?

There were no security events during the period which required  !

l reporting to the NRC under 10 CFR 73.71(c). This is notable and i demonstrates the high degree of attention given to all aspects of thJ l program by the licensee and the conscientiousness of the security

. force All ecurity records and logs are clear, concise, well maintained,  :

and readily available.

I An inspection of the licensee's program for the control and  !

accounting of special nuclear material was conducted during the i period. The program was found to be well implemented.

[

Ouring the assessment period, the licar. set submitted one Security I Plan revision and three temporary changes tv the Security plan under l the provis tons of 10 CFR 50.'. .s) and provided its response to the l

August 4, 1986 Miscellaneous Amencments to 10 CFR 73.55. The t

temporary changes and revisions to the plan were technically i adequate, censistent with prov'sions of 10 CFR 50.54(p), and clearly described and marked to facilitate the NRC's review. Prior to sub-mittal of the changes, the licensee contacted Region ! Safeguardt  !

personnel to arrange for meetings on-site or in Region I, depending i on the complexity of the change, to review and discuss the planned changes and compensatory measures to ensure a full understanding i prior to implementation. The quality of the licensee's changes to  :

the security program plans further demonstrates 4 clear understanding i of NRC Security program ocjectives and high interest in maintaining "

an effective and quality security program.

In sumary, the licensee's security program continuss to ba well  !

structured and implemented. A high degree of management and i supervisory involvement and advance planning is clearly evident. The  !

Itcensee continues to place neensary emphasis on the Security and l l $4feguarcs program to assure effective implementation. i l

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I

~

. . 26 s G. Training and Qualification E'!,:tiveness T2% 82 hours9.490741e-4 days <br />0.0228 hours <br />1.35582e-4 weeks <br />3.1201e-5 months <br />)

1. Analysis '

Ouring the previous assessment period, this area was rated Category

2. Licensee management made a major commitment tS obtain INPO training accreditation of all 10 of their training mrograms and to install a site-specific simulator.

l During this assessment period, training and qualification l effectiveness continued to .be an evaluation criterion for each i functional area and is discussed as an integral part of each arsa. l Training effectiveness has been assessed primarily by the observed performance of licensee personnel by the resident inspectors, i t

In addition, on one occasion, an expanded inspection effort in the I area of Emergency Operating Procedures was concucted. Also, the  !

region based Operational Assessment Team conducted a partial l evaluation of the training activities.

Ouring this SALP period, Senior Reactor Operator (SRO) examinations were administered to 13 candidates. Twelve of the candidates were t subsequently issued SRO licenses. An evaluation of the requalification program was also conducted in which three licensed SR0s and two licensed Reactor Operators (R0s) were examined. All  :

individuals passed the examinations. The requalification program wcs j cetermined to be acceptabit. In acdition, the licensee is in the t process of traineng several experienced SR0s to corduct the bulk et '

the requalification program, i A plant specific simulator is uncer constrt.ction and scheduled for cepletion by June 1938. As a result cf the continuing divergence l between the Indfan Coint 3 plant and the Indian Peint 2 simulator l

! which the licensee uses, the NRC placed restrictions on the simulator portion of the last sentor Reactor Operator (SRO) examinat:ca. The [

licensee has agreed ta have the simulator operational prior to the i next examination of licensed operators.

) '

The licensee continues to pursue INPO accreditation of their training  !

programs. Six of the 10 training programs - Nuclear Plant Operator, Reactor Operator, Senior Reactor Operator / Shift Supervisor, Electrical Maintenance Machanic, Mechanical Maintenance Mechanic and

, Instrumentation and Controls Technician - were accredited in October l

1986 while the remaining programs received their INPO evaluation on [

June 8-12, 1937. l An inspection assessed licensed coerator familiarity with the Emergency Operating Procedures (ECPs). Weaknesses were to ntified in the operators' use of the crocedures and kncwledge of the U  !

cactground information and vare partly attributed to inconsistencies [

19 the training program designed to address the new procedures. l l

27 Subsequently, thi licensee implemented an in-depth three week classroem training program for all the operators on the E0Ps. The quality and depth of the training was reviewed and considered excellent.

With regard to ncnlicensed training, a training facility for maintenance personnel has been establish *d which contains equipment suca as a lathe, milling nachine and a pump mockup for hands-on training. The I&C technician training laboratory was also found to be adequate for initial training.

There w'ere 18 Litelsee Event Reports (LERs) submitted during this pericd. Three reports identified additional training requirements as 4

cart of the corrective actions. In all cases, such commitments acceared to be narrow in scope. NP' review of all LERs indicates that at least seven LERs should'have incluced a cor.mitment for training enhancement ir, the applicable areas. These include LER 86 Og, LER 87-02, LER 87-04 and LER 87-09.

In summary, the licensee has achieved INPO accreditation in six areas. Ccmplete accreditation of all programs.ts expected by j September 1987. Licensed operators training and requalification programs are strong. One exception was noted in the implementttien of E0P training.

2. Conclusien Category 2

3. 8, Card Recommendations Licem et:

None NEC:

None l

4

[

,y 28 o b .

H. Assurance nf Ouality h, e*

1. Analysis Assurance of Quality continued to be considered as a separate unctional trea for this SALD perfort and is a sumary of managemr.nt con ols and their effectiveness in the assurance of safe p' ant operatic s. This area was rated Category 2 in the last $ ALP with weaknesses not in the Quality Assurance audit progra i, particularly in the oversight o uperations and surveillance activities.

The various aspects of the Quality Assurance Progra requirerents have i been considered and dist:ussed as an Integral part f each functional area and the respective inspection hours are included n those areas.

Corporate and site management changes during e latter part of the previous peried resulted in continued corper e support and involvement in site activities while concurrently assurina site sanagement autoncey in t.he day-to-cay operation of the plant.

Site and corporate programs tend to be mplamented through informal management gu Lance with the support f dedicated and motivated personnel.

Mantgement attention is required to recognize the need for the p ept development of cohesive procedura support of programs. Long term efforts I to improve operations and radiol gical control procedures failed to meet I goals due to inadeguate stafft of the programs and a lack of management l followup. Existin; prececure are often incensistent and weak, lacking

! detail and accuracy. For e mole. the lack of an effective acministrative tracking systets foi Technt al Specification amenc ents result ed in a l failure to include the n Power Operated Relief Valve (POR\/ settings for the Low Temperature Ove c ressure Pretection (LTOP) system in plant documents and equipme setpoints.

Departmental ranag s and supervisors are well informed and understand the i ranagerent goals. Scre weaknesses were noted regarding the understanding by craf t pericnr 1 of departt. ental goals, particularly in the area of safety goals.

During the sessment period, the NYPA rianagement instablished and fully i staffed a perating Experience Review Group. The *,4'f includes a.  ;

I licensed enter reactor operator. The group rev$ is

• ernally.and

'l ettern y generatea operations experience and .osb c Leeropriate rc~om rdations to management. Its effectisenes;,e.. t yet been eval ated.

T e Plant Operations Review Comittet (PORC) continued to function ffectively as an advisory body to the Resident Manager. Comittee mercers exhibited good understanding of the problems at hand and con-tributed frequently to the decision making erecess. The offsite Safety Review Cemittee ($RC) also proviced effective and cceprehensive reviews and advice eencerning nuclear safety istues. An er:ection to t%15 performance was notes in that coth the SRC and W NC failed to precerly ,

assure the implem>ntation of the LTOP Tecanical Specification amencment.

l l

28 A H. Assurance of Quality

1. Analysis Assurance of Quality centinued to be considered as a separate functional area for this SALP period and is a summary of management controls and their effectiveness in the assurance of safe plant operations. This area ,

was rated Category 2 in the last SALP with weaknesses noted in the Quality Assurance audit program, particularly in the oversight of operations and surveillance activities.

The various aspects of the Quality Assurance Program requirements have been considered and discussed as an integral part of each functional area and the respective inspection hours are included in those areas.

Corporate and site management changes during the latter part of the previous period resulted in continued corporata wpport and f nvolvement in site activities while concurrently assuring siti 'anagement autonomy in the day-to-day operation of the plant.

Site and corporate programs tend to be implemented th* rough informal management guidance with the support of dedicated and motivated personnel.

Management attention is required to recognize the need for the prompt development of cohesive procedural support of programs. Long term efforts to improve bperations and radiological control procedures failed to meet goals due to inadequate staffing of the programs and a lack of management followup. Existing procedures are often inconsistent and weak, lacking detail and accuracy. For example, the lack of an ef fective administrative tracking system for Technical Specification amendments resulted in a failure to include the new Power Operated Relief Valve (PORV) settings for the Low Temperature Overpresst re Protection (LTOP) system in plant documents and equipment setpoints.

Departmental managers and supervisors are well informed and understand the management goals. Some weaknesses were noted regarding the understanding by craft personnel of cepartmental goals, particularly in the area of safety goals.

During the assesament period, the NYPA management established and fully staffed an Operating Experience Review Group. The staff includes a licensed senior reactor operator. The group reviews internally and externally generated operations experience and develops appropriate recommendations to management. Its effectiveness has not yet been evaluated.

The Plant Operations Review Committee (PORC) continued to function effectively as an advisory body to the Resident Manager. Committee memb'ers exhibited good understanding of the problems at hand and con-tributed frequently to tne decision making process. The offsite Safety I

+

pr 29 *"

The quality assurance and quality control groups, while maintai ing their organizational independence, appear to have the full support d coopera-tion of all site managers and their respective department st ervisors.

The QA and QC groups, headed by the QA superintendent, hav been reorganized and were fully staffed during the assessment riod.

The number of audits conntnued to increase in 1985 and 086. NRC insgec-tions in this area found that tne audits are detailed nd audit findings  !

are given management attention. However, an isolate incidence of a missed area. vendor surveillance audit was noted in the r dioactive effluents The expansion of quality controls into the alance of plant areas also continued. QA/QC coverage is routinely pro ided for turbine generator and main feed water system maintenan activities. ,

A weakness was identified in that quality co trol inspectors failed to y

detect improper installation of environmen lly qualified electrical splices. The licensee is committed to a ontinuing training program as i well as other improvements in this area. An audit program of licensed l operator activities in tht control roo, was initiated towards the end of 3 the assessment period. The program i enhanced by the availability of a

• quality assuranco origineer who ' comp 1 ted the senior reactor operator training program. The same QA eng eer identified the problem with the implementation of the LTOP system setpoints during the reactor shutdown at  ;

the start of the cycle 5/6 outa .

l Licensee Event Reports (LERs) fail to probe for root causes and do not identify the safety signifi nee of the everit. Cause codes were  !

improperly assigned, and i ten of the eighteen reports +.his period, no r cause codes were assigned, On one occasion, a supplemental report was necessary (LER 86-09) t :larify the operators' role and understanding of l l the event. '

'In summary, althoug the licensee has dedicated and motivated personnel, licensee managemen needs to focus on establishing formal support for existing programs Some resources have been expended to address the  ;

weaknesses iden fied in the area of post-trip reviews; however, additional i management ce tment is required to address procedural weaknesses identified in other programs and to bring these efforts to fruition.

2. Conclusion .

l-(

Category 2 '

I i

l 1

~

29 A Review C nmittee (SRC) also provided effective and comprehensive reviews and advice concerning nuclear safety issues. An exception to this performance was noted in that the PORC failed to properly assure the impleLentation of the LTOP Technical Specification amendment.

Subsequently, an SRC initiated audit of control room activities identified the LTOP problem. Apprepriate corrective action followed.

l The quality assurance and quality control groups,'while main' '" . their organizational independence, appear to have the full supp : a :ccpera-tion of all site managers and their respective department supervirors.

The QA and QC groups, headed by the QA superintendent, have been reorganized and were fully staffed during the assessment period.

The number of audits continued to increase in 1985 and 1986. NRC inspec-tions in this area found that the audits are detailed and audit findings are given management attention. However, an isolated incidence of a missed vendor surveillance audit was noted in the radioactive effluents area. The exoansion of quality controls into the balance of plant areas also continued. QA/QC coverage is routinel; provided for turbine generator ar,d main feed water system maintenance activities.

A weakness was identified in that quality control inspectors failed to

• detect improper installation of environmentally qualified electrical splices. The licensee is committed to a continuing training program as well as other improvements in this area. An audit program of licensed operator activities in the control room was initiated towards the end of the assessment period. The program is enhanced by the availability of a quality assurance engineer who completed the senine reactor operator training program. The same OA engineer identifded tne problem with tne implementation of the LTOP system setpoints during the reactor shutdown at the start of the cycle 5/6 outage.

Licensee Event Reports (LERs) fail to probe for root causes and do not identify the safety significance of the event. Cause codes were improperly assigned, and in ten of the eighteen reports this period, no cause codes were assigned. On one occasion, a supplemental report was necessary (LER 86-09) to clarify the operators' role and understanding of the event.

In summary, although the licensee has dedicated and motivated personnel, licensee management.tieeds to focus or, establishing formal support for existing programs. Some resources have been expended to address the weaknesses identified in the area of post-trip reviews; however, additional management commitment is required to address procedural weaknesses identified in other programs and to bring these efforts to fruition.

2. Conclusion Category 2

~

p.

30

3. B_ card Recommendations Licensee:

1. Establish a long-term program to upgrade facility procedures.

2. Improve the cuality of LES submittals and root cause determination (in conjunction with the deficiencies noted in the Operations functional area).

NRC:

None ,

v

.i i

,P 4

i i

E l

k t

i 1

?

31 I. Engineering and Technical Suocort (7%, 364 hour0.00421 days <br />0.101 hours <br />6.018518e-4 weeks <br />1.38502e-4 months <br /> _s_1

1. Analysis During this assessment period, Engineering and Technical Support is being considered as a functional area for the first time.

Engipeering efforts to resolve technical issues were considered as an evaluation criteria for each functional area. An NRC outage and ,

modifications inspection team completed a two week design review inspection effort at the end of the period.

At the site, engineering support is provided by the technical

• services, maintenance, operations and computer services departments.

The technical services department through its electrical, mechanical and reactor engineering sections maintains review and approval '

responsibility over all site-related modifications. At the corporate office, the nuclear support department provides radiological, chemistry, reactor and licensing engineering support-, the nuclear engineering department provides project and plant improvement support; and the nuclear operations department provides maintenance and operational engineering support. ,

Lines of responsibility to address engineering issues were not clearly defined resulting in inconsistent performance. For example, engineerfng support to meet long term commitments in the area of drawing updates lacked timeliness. The responsibility for this commitment was passed from the site to the corporate office where staffing and division of responsibility in this area appears to have delayed the program. Also, the handling of environmenta'. qualifi-cation issues appears to be lacking programmatic support and leader-ship. In at least one instance, involving the environmental qualifi-cation of electrical splices, the licensee had to redefine the scope of the program to address a previously inspected area after the specified deadline. This is contrasted by the licensee's response ,

to fire protection requirements of 10 CFR 50, Appendix R, which was conservative as supported by the NRC inspection team comments in that area.

Major modification packages are developed by corporate engineering and are subsequently approved by engineers on the site technical services staff. Strengchs in the clarity and level of engineering deteil incorporated into outage modification work packages and in the overall level of technical competence of the engineering personnel associated with both the corporate and onsite engineering groups were not.ed.

It was apparent that the design review process needs improvements' in the areas of control of design inputs, completeness of design analyses and control of the vendor interface. Corporate and site engineering support rely on a single modification procedure which is ,

inadequate to support the company's modification and design program

32 For example, design weakness were noted in the development cf calculations for service water pump capacity and battery loads.

In order to compensate for the procedural deficiencies, temporary procedure changes must be issued during major modification efforts.

In May 1987, the corporate management presented to the NRC inspectors a plan to institute a new Modification and Design Control Program.

The scope of the program has been defined and a management organization is in place to facilitate implementation of the program.

Four reactor trips were attributed to plant design deficiencies. In two instances, the design problem in the main boiler feed pumps' '

control oil system was the initiating event (see Plant Operations).

A modification is planned for the cycle 5/6 refueling outage to correct the problem. The other two reactor trips were initiated by the loss of power supply to instrument buses. An engineering review of instrument bus load distrioutions, prompted by NRC Bulletin 79-27 and Information Notice 79-02, failed to identify that a less of power '

supply to instrument bus No. 33 or bus No. 34 will result in reactor trips with a procao111ty of a safety injection signal following the reactor trip.

In summary, both the onsite and corporate office engineers and technical support staff appear technically competent. A lack of clear procedursi guidance celineating lines of responsibility '

has contributed to an inconsistent performance in this area.

2. Conclusion Category 2

3. Board Recommendations Licensee:

None NRC:

4 None t

I 33 J. Outaces

1. Analysis During the previous assessment period, the area was rated Category 1 with no deficient areas noted. During that period, the licensee completed a mid-cycle steam generator inspection outage and the cycle 4/5 refueling outage. Strong quality control and health physics coverage was noted.

During this assessment period, two scheduled outages were conducted; the mid-cycle steam generator inspection outage in April-May 1986 and the cycle 5/6 refueling outage starting May 1, 1987. The latter outage continued beyond the end of the SALP period. Outage activities were routinely observed by the resident inspectors.

Two unplanned shutdowns resulted in outages: one, a 57-day outage to inspect and repair turbine generator low pressure rotors and replace a main transformer in July - August 1986, and the other, a 7-day outage in March-April 1987 to replace two reactor coolant pump seals.

Management preplanning was evident in both scheduled outages. Job responsibilities were well defined and lines of communications between onsite and offsite departments were clearly'establisned.

Confli' cts and disagreements were promptly resolved at daily management meetings. The position of Outage Coordinator, established during the previous ceriod, also continued to enhance outage management effectiveness.

Complex activities such as eddy current testing of 100% of the hot leg tubes in all four steam generators during both the mid cycle 1986 outage and the 1987 refueling outage were conducted with few difficulties. Related CA/QC involvement in the testing process was well evidenced. A third party review of all eddy current data was performed during both outages to resolve any discrepancies between the primary and secondary inalyses.

Management responded effectively during the unplanned outages to organize and coordinate the resources necessary to facilitate repairs to the unit. Preplanning for corrective and preventive maintenance work requiring cold shutdown plant conditions was evident, permitting the licensee to perform numerous safety-related work activities during the shutoewns.

Durirg the outages, the following good practices were noted by the inspectors:

Supervisor and management presence at the work sites QC and QA involvement in work activities including those in the balance of the plant

4

, 34 Direct oversight of contractor activities by licensee employees Effective health physics coverage was provided for all radiologically sensitive activities. One notaole exception involved insufficient health physics involvement in a maintenance activity which resulted in exceeding the administrative dose limit by an individual.

(Functional Area B)

In summary, scheduled outages show evidence of preplanning.and the appropriate assignment of priorities. Management response to unplanned outages was effective in. developing schedules and organizing the necessary resources to facilitate repairs. .

2. Conclusions Not rated because of the minimal number of inspection hours due to the lack of a major refueling / modification outage.

3. Board Recommendations Licensee: i None NRC: ,

None l

a 9996 i

(

i I

I

_,,--,-,-.n---- , , - - , -

35

~

K. Licensing Activities

1. Analysis During the previous assessment period, the licensing functional area was rated Category 2. Licenses strengths were noted in the areas of technical capability and the effectiveness of the licensing staff.

During this essessment period, the licensee has shown good management overview in the area of licensing activities. The Itceasee's manage-ment demonstrated active participation in licensing activities and kept abreast of current and anticipated licensing actions. All open licensing actions are scheduled and tracked through use of the NYPA licensing status report. The licensae management and the Project  !

Manager met on several occasions to discuss 1.icensing action status.

Submittals are usually timely and show evidencr of prior plann ng and x

assignment of priorities. The licensee has met schedules or iaformed the Project Manager at an early date of schedular delays. However, during the latter portion of the rating period, the licensee sub-mitted two items for which extremely short response times were

requested. For one item, Integrated Leak Rate Testing, the licordee was not aware until close to their submittal date that a short duration test may be acceptable. This caused the subn.ittal delay and was understandable. For the second ite,m, concerning climination of arbitrary intermediate break analyses, the licensee was aware and had been discussing the item with the NRC sta'f as much as four mon,ths prior to submittal. Better planning could have avoided the request

, for an accelerated review.

The licensee maintains the technical capability in almost all

• engineering and scientific disciplines necessary to resolve items of '

concern to the NRC and the licenses. The licensee also utilizes the services of nuclear support groups to assist in the resolution of technical problems or to utilize now and improved techniques that l

will enhance the operation and safety of the plant. In addition, the good communications between the licensee and NRC staff have been beneficial both in easing the processing of licensing actions and in minimizing the-need for additional requests for information.

It should be noted that during the rating pertad that the review of the geniric issue concerning Reactor Coolant nump Trip Criteria was completed. The licensee's submittal was outstanding and the response clearly represented an attempt to provide the information needed by the NRC staff.

The licensee's findings of no significant hazards considerations ire j good and rarely require supplemental information.

t

36 The licensee has been responsive to NRC inittatives in most instances. Schedules are negotiated with the licensee based on priorities. The licensee meets deadlines and notifies the NRC staff weli in advance of any schedular problems. The licensee facilitates a t..nely resolution of most issues. Licensee submittals are

• technically sound and thorough. Acceptable resolutions are initially proposed in most instances.

Licensing activities are conducted by a well staffed and well trained group resulting in' an overall efficient operation. Management over-view is evident in that the licensing group is well integrated into other plant activities and lictnsing activities reflect a uniform approach. Upper management be:omes involved in licensing actions when necessary to assist in re solving potential deadlocks.

The licensing group has exhibited a high degree of cooperation with the NRC staff. Areas of expe"tise are well defined within the group.

In addition, the group does in excellent job of ccordinating the effort when input is requirej from the different groups within the Power Authority. The licensing manager is knowledgeable and keeps abreast of all licensing actions. The licensing group holds informal traiaing sessions on topics of current and future interest. The group also participates in industry wide training programs provided by various organizations. '

In sut. mary, the licensee's strengths are the staff's technical capability as reflected in their submittals and discu;sions with the NRC and the continued uograding of the experience, capability and effectiveness of the licensing group and the supporting cdministra-tive and technical personnel.

2. Conclusion Category 1

3. Board Recommendations Licensee:

None NRC: -

None

37 V. SUpp0RTING DATA AND SUMMARIES A. Investigations and Allecations Review

1. Investigations

• There were no Office of Investigations (OI) reviews initiated at Indian Point Unit 3 during this assessment period.

l 2. Allegations Four allegations were received and reviewed by NRC Region I during the assessment period. Three allegations were not substantiated. One allegation regarding the operation of the Fuel Storage Building Ventilation System was partially substantiated and subsequently resolved.

B. Escalated Enforcement Actions

1. Civil penalty A Civil Penal'.y of $50,000 was associated with NRC Inspection 286/86-15 conducted August 11 - September 22, 1986. The violation involved multiple inoperable safety-rtlated pumps centrary to Technical Specification requirements. (Enforcement Action No.86-197 dated January 20, 1987)

2. Enforcement Conferences An Enforcement Conference was held on June 24, 1986 regarding a violation in the area of radiological protection ider.tified in NRC Inspection Report 286/86-11 conducted May 12-16, 1986. The violation involved an unplanned exposure of a worker in excess of administrative limits with a substantial potential for exceeding regulatory limits. No civil penalty was issued.

An Enforcement Conference was held on November 10, 1986 regarding inoperable safety-related pumps in violation of Technical Specification requirements.

C. Management Conferences The SALP management meeting was conducted on April 25, 1986. In addition, regional and NRR management toured the plant and met with licensee management to discuss plant status on April 30, 1987.

s * .

38 O. Licensee Event Recorts (LERs)

Tabular Listino of NRC assigned cause codes A- Personnel Error -

6 B- Design. Man./Constr / Install. 4 C- External Cause -

0- Defactive Procedure 1 E- Compone , Failure 7 X- Other - -

TOTAL 18 l C3 gsal Analysts Two common causal chains were identified:

1. Personnel Errors Six LERs 86-02, 86-04, 86-07, 86-09, 86-12 and 87-06 can be attributed to improper or inadeauate actions taken by plant operations personnel. LER 96-07 details an inadvertent safety injection actuation during suber.itical plant conditions.

Contributing factors include lack of knowledge on the part of the surveillance group regarding expected plant conditions and poor judgment on the part of the operators allowing the conduct of the surveillance. The remaining LERs are analyzed under Table 5. The lack of attention to detail was a major contri-butor to the events. Four instances resulted in a reactor trip l or inadvertent safety injection, and one occurrence involved the violation of Technical Specification requirements. Depth of root cause analysis is inconsistent. Narratives tend to unnecessarily mitigate personnel errors and involvement.

Responsibility for corrective actions in this area is generally passed en to the training department with no specific management followup.

2. Equipment Design Deficiency Loss of Instrument Buses - LERs 87-02 and 87-04 identify unexpected equipment responses following the loss of power to instrument buses 33 and 34. Deficiencies in the load

, 39 distribution for : nase cuses were identified. Corrective actions based on an ongoing load distribution study will be initiated.

Main' Boiler Feed pumos - LERs 86-11 and 87-01 detail the two instances of main boiler feed pump trips attributed to a control oil system design deficiency. The root cause will be eliminated through a design change scheduled for completion during the cycle 5/6 refueling outage. The reactor trip event detailed in LER 86-11 was contributed to by a decision on the part of main-tenance management to allow the pump to operate with an oil seal clearance that was marginal based on pump technical manual etiteria.

E. Automatic Reactor Tries/ Engineered Safeguard Actuations and Unolanned Shutdowns During the assessment period, thirteen automatic reactor trips and five unplanned shutdowns occurred.

Engineered safeguard systems actuated three times, twice following reactor trips and once during a surveillance activity.

Table 5 summarizes the events and includes a root cause analysis.

F. Licensino Activities

1. NRR/ Licensee Meetings Toxic Gas Monitors 4/15/86 Licensing Activitiss 4/17/86 DCRDR 6/16-18/86 Appendix R Audit 7/14-18/86 Containment Leak Rate 3/17/87

2. NRR Site Visits / Meetings RER 9/8-12/86 DCRDR 6/16-18/86 Appandix R Audit 7/14-18/86 Management Discussions 11/6/86, 4/30/87

3. Commission Meetings None

4. Schedular Extensions Granted None

.,3 --- .. .

40 .

5. Reliefs Granted  !

ASME Section XI Relief 4/1/86  !

ASME Section XI Relief 7/25/86  :

ASME Section XI Relief 1/6/87.  !

ASME Section XI Relief 2/5/87  !

l 6.. Exemptions Granted Appendix R 1/7/87 i

7. Licensing Amendments Issued -

Amendment Titles Date Numbers i

63 Physical Security Plan 2/7/86

{ Changes 64 Minimum Shift Crew 3/10/86 Composition and Overtime Limitations

65 Changes to Incorporate 4/18/86

NUREG-0737 and Generic j Letter 83-37 Requirements 66 Iodine Report Requirnents 8/7/86 Operating Restrictions -

During Periods of High Coolant Activity i

! 67 Low Temperature Overpressure 9/15/86

Protection System Specification Changes J

i 68 Add an Anticipatory Reactor 10/6/86 Trip Upon a Turbine Trip i

69 Provision to Allow a 10/7/86 Temporary Closure Plate in Place of the Equipnent Hatch Ouring RefuelinJ I

70 Permit Storage of Fuel 10/19/86 l Having an Enrichment up to i 4.3 Weight Purcent U-235

• I in the New and Spent Fuel Racks 1

! 71 Organizational Changes 1/6/87 i

i

41 72 Permit Di st...; ege of More 4/2/87 Than One Region of Fuel Frem the Reactor After 162 Hours 73 Raise Fq to 2.20 4/17/87

8. Emeroency Technical Soecifications None

9. Orders Issued ,

None 9

4 e

TABLE 1 TABULAR LISTING OF LERs BY PUNCTIONAL AREA INDIAN POINT STATION, UNIT 3 Cause Code Area A B C D E X Total A. Plant Operations 6 6

8. Radiological Controls C. Maintenance 5 5 D. Surveillance 2 2 E. Emergency Preparedness

• F. Security and Safeguards G. Training and Qualification Effectiveness H. Assurance of Quality I. Engineering Support 4 1 i J. Outages K. Licensing Activities Totals 6 4 0 1 7 0 18 Cause Codes: A. Personnel Error B. Design Manufacturing, Construction, or Installation Error C. External Cause *

0. Defective Procedure E. Component Failure X. Other

o *

, T1-2 TA8LE 1 LER SYNOPSIS (12/1/85-5/31/87)

NRC Assigned LER Number Event Date Cause Code Description 86-01 2/28/86 E Reactor trip initiated by a Surveillance

• faulty test relay in the Independent Electrical Overspeed Protection System i 86-02 5/18/86 A Reactor Trip caused by ~ ~' -

^

Operations inadequate venting of the main feedwater flow transmitters prior to their return to service i

86-03 5/19/86 E Reactor trip caused'by faulty Maintenance check valve in the main feedpump discharge line &

86-04 5/19/86 A Suberitical reactor trip caused Operations l by failu e of a control room operator to block.the source range high flux trip during startup 86-05 5/23/86 E Reactor trip initiated by 2

. Surveillance failed relay in the Independent Electrical Overspeed Protection e System 86-06 5/26/86 0 Reactor trip caused by a Engineering condensate system transient .

Support during a load reduction 6

86-07 5/27/86 A Subcritical safety injection Operations system actuation due to surveillance testing curing '

inappropriate plant conditions 36-08 8/29/86 B Oiscrepancies in Enviro'nmenta11y j Engineering Qualified (EQ) equipment Support  :

86-09 9/2/86 A Control room switches for the Operat ons containment spray and

• recirculation pumps were improperly positioned curing startup i "NRC assigned functional area L i

V

t ,

T T1-3 o .

85-10 9/5/86 E Reactor trip caused by spurious i Maintenance

• open.'ng of the generator output disconnect switch 86-11 9/9/86 'E Reactor trip caused by the loss  ;

Maintenance of a main boiler feedwater pump '

due to water in its cortrol oil system 86-12 11/14/86 A Reactor trip caused by bumping Operations open a DC power supply breaker during a power panel inspection 87-01 1/31/87 E Reactor trip caused by the loss Maintenance of a main boiler feedwater pump I

due to dirt in its control. oil system

, 57-02 2/11/B7 B Reactor trip and safety injection Engineering actuation caused by loss of Support instrument bus No. 34 87-03 4/4/87 E Reactor trip caused by Maintenance malfunction of the turbine governor during startup 87-04 4/17/87 B Reactor trip and safety injection Engineering actuation caused by a loss of Support instrutent bus No. 33 87-05 4/30/87 8 Design deficiency in auxiliary Engineering boiler feed pump starting logic .

Support circuit resultiag in single  ;

failure point for pumps 87-06 5/5/87 A low temperature overpressurization t Operations system set points not updated to reflect new curve issued in  !

Technical Specifications

i l "NRC assigned functional area

. I 1

i

t

. i TABLE 2 INSPECTION HOURS

SUMMARY

-(12/1/95 - 5/31/87)

INDIAN POINT STATION - UNIT 3  !

t Hours  % of Tbse A. Plant Operations. . . . . . . . . . . 2060 47 B. Radiological Controls . . . . . . . . 631 14 C. Mainteaance . . . . . . . . ; . . . . 604 14 D. Surveillance. . . . . . . . . . . . . 339 8 E. Emergency Preparedness. . . . . . . . 162 4 F. Security and Safeguards . . . . . . . 163 4 4

G. Training and Qualification . .... 82 2 Effectiveness 4

H. Assurance of Quality. . . . . . . . . " "

I. Engineering and Technical Support . . 364 7 l J. Cutages . . . . . . . . . . . . . . . " "

i K. Licensing Activities. . . . . . . . . *

• i i 1

! Total 4405 100 1 Hours expended in facility licensing activities and operator license  !

activities not included with direct inspection effort statistics.  !

i

Hours expended in the assurance of quality and outage areas are included L in other functional areas, therefore no direct *. inspection hours are given l for these areas.  !

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TABLE 3 ENFORCEMENT

SUMMARY

(12/1/85 - 5/31/87) e A. Number and Severity level of Violations Severity level No Severity Level I O Severity Level II O Severity Level III 2 Severity level IV 2 Severity Level V 5 Total 9 B. Violations Vs. Functional Areas Severity Levels FUNCTIONAL AREAS I II ILI L, v v TOTALS A. Plant Operations 2 1 4 B. Radiological Centrols .. 1 1 2 C. Maintenance 0

0. Surveillanen 1 1 E. E.teroe.ecy Pre;:a adr es s 0 F. Security and Safeguards 0

0. Trainino 0 H. Assurance of Ouality 2 2 I. Engineerino Support 0 J. Outages 0 -

K. Licensing Activities ,,. ;_.

~

0 Violation and Deviation Totals: 2 2 5 9

~

s T3-2 C. Summary - Enforcement Cata Inspection Viol. Functional Report /Date Recuirement Level Viola *,.on

__ Area 286/87-10 Instrument 5 Surveillance Fa.ure to 03/27/87- Calibration e.411brate Boric 05/11/87 Acid Temperature

• Controllers 103 &

107 on the required interval 286/87-08 Low Temperature 4 Operations Violation of T,S.

04/27/87- Overpressure requirements for 05/06/87 Protection System Power Operated Operability Relief Valve (PORV) lift settings.

286/86-24 Environmental 5 Quality Raychem splices 12/15/86- Qualification / Assurance installed with 12/19/86 Raychem Splices less than the required 2 inch overlap 286/86-24 Environmental 5 Quality Raychem splices 12/15/86- Qualificatica/ Assurance bent greater than 12/19/86 Raychem Splices the minimum allowed radius of 5 times the diameter of the splice 286/86-23 -

Control of 4 Operations Failure of 11/18/86 Maintenance operations 01/05/87 personnel to follow administration procedures regarding the control of work activities

T3-3 286/86-21 Pump Operability 3 Operations Violatten of TS 8/11/86- requirements for 9/22/85 exceeding cold shutdown condition without the required number of containment spray and recirculation pumps operable 286/86-11 Health Physics 3 Radiological Failure to control 5/12/86- Controls Protection a high radiation area resulting in a significant potential for an overexposure 296/86-09 Environmental 5 Radiological Failurs te distill 4/14/86 TS Analysis Protection tritium samples 4/18/86 as required by procedure 286/85-27 Shift Logging 5 Operations Failure to record 12/16/85 significant events 1/27/86

• in the shift legs

r e

, TABLE 4 INSPECTION REPORT ACTIVITIES (12/1/85-5/31/87)

Reoort/ Dates Inspector Hours Areas Insoected 85-27 Resident 60 Routine, safety inspection by 12-16 resident inspectors 1-27 86-01 Specialist -

Operator Licensing Examination 3/10-11/86 86-02 Resident 104 Routine safety inspection by 1-29 resident inspectors.

3-10-86 86-03 Specialist 33 Routtae inspection of radio-2/10-14/86 active waste transportation activities .

86-04 Specialist 33 Routine inspection of the 3/3-6/86 nonradiological chemistry program 86-05 Specialist -

Operator Licensing Examination 8/6-10/36 86-06 Specialist 54 Special inspection of the 3/11-13/86 annual emergency planning exercise 86 07 Resident 112 Routine safety inspection by 3-1 resident inspectors 4-15-86 86-08 Specialist 96 Routine inspection of the 4/7-11/86 security program 86-09 Specialist 105 Routine inspection of 4/14-18/86 radicchemical measurements 86-10 Resident 85 Routine safety inspection by 4-16 resident inspectors 5-15-86 86-11 Specialist 89 Routine inspection of 5/12-16/86 radiological protection activities

T4 86-12 Specialist -

Special inspection of the 6/3-5/86 licensed operator requalification training program.

86-13 Resident 124 Routine safety inspection 5-16 by resident inspectors 7-23-86 86-14 Specialist -

Operator Licensing Examination 6-11-86 86-15 Resident 349 Routine safety inspection by 6-24 resident inspectors 8-10-86 36-16 Specialist 89 Special inspection to review 7/7-10/85 the status of open NUREG 0737 items 86-17 Team 160 Special team inspection of 10 CFR 7/14-18/86 Appendix R requirements 86-18 Specialist 72 Goerator Licensing Examination 8-13-86 86-19 Resident 287 Routine safety inspection by 9-23 resident inspectors 11-17-86 86-20 Specialist 51 Routine insptction of nonlicensed 7-28 training and cuality assurance 8-1-86 86-21 Resident 175 Routine safet/ inspection by 8-11 resident inspectors 9-22-86 86-22 Specialist 33 Routine inspection of inservice 9-15-19-86 inspection activities 86-23 Reedent 212 Routinesafetyinspect[onby 11-18 resident inspectors 1-5-87 86-24 Specialist 41 Special inspection of 12/15-19/86 environmental Qualification (EQ) requirements 87-01 Resident 276 Routine safety inspection by 1-5 resicent inspectors 2-17-87

._w

o T4-3 87-02 Specialist 44 Special inspection of dosimetry 1/27-29/87 program 87-03 Specialist -

Operator Licensing Examination 2/24-26/87 87-04 Specialist 22 Routine inspection of physical 2/10-13/87 security mearures 87-05 Resident 219 Routine safety inspection by 2-18 resident inspectors 3-26-87 87-06 Specialist 124 Routine inspection of the annual 3/17-19/87 emergency olanning exercise 87-07 Specialist 44 Routine inspection of the 3/16-20/87 radiation protection progran 87-08 Team 410 Operational Assessment Team 4-27 inspection 5-6-87 -

87-09 Specialist 19 Routine inspection of special 3-30 nuclear material accountaoilify 4-3 87

• 87-10 Resident 278 Routine safety inspection by -

3-27 resident inspectors 5-11-87 l

87-11 Specialist 35 Routine taspection of radioactive 4/20 24/87 effluent controls 87-12 Scecialist 26 Routine inspection of the 4/28-29/87 physical security program 87-13 Team 544 Safety system outage modification  !

4-22 team inspection ($$0MI) 5-29-87 t

TABLE 5 REACTOR TRIPS, ENGINEERED SAF'EGUARDS AClUATIONS AND UNPLANNED SHUTDOWNS The automatic reactor trips / engineered safeguards actuations and unplanned shutdowns occurring during this assessment period fall into two categories, personnel error and equipment failure or malfunction. The following table assesses the root cause of each trip from the NRC's eerspective within each category. The root cause was determined by the SALP Board and may not agree with LER analysis.

Personnel Errors There were five reactor trips of a total of thirteen trips and one engineered safeguards actuation attributed to personnel error. For the purposes of this table, personnel error has been broken into four groups: (1) Personnel error -

poor judgment: the individual should have known that the outcome of the action could cause a trip; (2) Cersonnel error - lacking knowledge: the individual had not been instructed, the pi*ocedure did not address the issue, or the individual did not have the knowledge that the act performed would cause a trip; (3) Personnel error - inattention to detail: the individual took a haphazard approach to an unrelated task which subsequently led to a trip; and (4) Personnel error - equipment malfunction: an equipment failu-e or malfunction in conjunction with a personnel error, where both were necessary to cause the trip.

Equie ent Malfunction / Failure There were eight of thirteen reactor trips and two engineered safeguards actuations attributed to equipment malfunction or failure. For the purposes of this table, equipment malfunction / failure has been broken into three groups; (1) Random failure - isolated failures which are not considered generic, (2)

Design deficiencies - failures attributed to equipment design, and (3)

Construction deficiencies - failures attributed to improper installation during construction.

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T5-2 .

TABLE 5 REACTOR TRIPS, ENGINEERED SAFEGUARDS ACTUATIONS AND UNPLANNED SHUTDOWNS NO. DATE POWER DESCRIPTION ROOT CAUSE LEVEL (SALP AREA) 1 2/28/86 100% Reactor trip due to low level. Equipment in No. 32 steam generator, failure -

During monthly surveillance Lack of preventive testing of the turbine independent mainte'nsnee electrical overspeed protection program for relays system..a defective relay .

(Surveillance) failed to block turbine trip signal allowing the stop valves to go closed. (LER 86-01) 3/1/86 Startup 2 5/18/86 2% Reactor trip due to steam flow / Personnel error -

feed flow mismatch coincident inattention to

• with low level in No. 32 detail. A non-steam generator. Erroneous licensed operator's feed. flow signal was generated lack of attention due to trapped air in she to the details of associated instrument lines, the applicable (LER 86-02) procedures for venting the flow instrument (Operations) 5/19/86 Startup 3 5/19/86 250 Reactor trip due to turbine trip Ecutpment caused by high level in No. 34 failure -

steam generator. The initiating Lack of preventive event was a feedwater system maintenance for perturbation due to a leose critical blanace disk in No. 32 main boiler feed of plant equipment, pump dischar (Maintenanes)

(LER 86-03) ge check valve.

5/19/86 Startup 4 5/19/86 49ESCPS Reactor trip due to source range Personnel error -

high flux level. During shutdown inattention to bank rod withdrawal, reactor detail and poor operator failed to block P-6. Judgment. Reactor (LER 86-04) operator allowed

T5-3 .

himself to be distracted during critical plant evolution. Senior reactor operator was not fully cognizan+. of ongoing control room activities.

(Operations) 5/21/86 Startup , 4 5 5/23/86 73% Reactor trip due to turbine trip. Equipment l Ouring monthly surveillance of failure -

the turbine independent Lack of preventive electrical overspeed protection maintenance system, a tested channel failed program for to clear. Upon injecting a relays, test signal in the second (Surveill. ce) channel a two out of three Unit trip logic was completed. *

(LER 86-05) 1 5/24/86 Startup 6 5/24/86 20% Unic placed in hot shutdcwn to Equipment inspect turbine for high malfunction -

l . vibrations. (Ope rati on s.)

l 5/25/86 Startup ,

l 7 5/26/86 70% Unplanned shutdown due to Equipment I

indication of accumulated malfunction -

combustible gases in No. 32 (Maintenance) main transformer.

I 8 5/26/86 16% Reactor trip due to low level Personnel error -

in No. 34 steam generator. lacking knowledge.

While reducing plant load, a Unexpected feedwater perturbation caused equipment response by the automatic stirt of two due to the failure condensate booster pumps to address the following the manual removal operational detail of a main condensate pump from c' the recently service. (LER d6-06) installed pumps (Engineering Support)

'. 4. *

. T5-4 9 5/27/86 Hot Engineerec Safeguard System Personnel error -

Shutdown actuated when a safety injection poor judgment / lack signal was generated of knowledge.

inadvertently during a Senior reactor surveillance test. (LER 86-07) operator allowed the test to commence without proper regard to ext: ing plant conditions, inadequate precautions on the associated surveillance procedure.

(Operations) 5/28/86 Startup 6/14/86 10 57% Unplanned shutdown to reinspect Equipment No. 32 main transformer, malfunction -

1

, (Maintenance)

C/.16/eC Startup with only No. 31 main transformer t- service. -

No. 32 main transfc. er out of service for replacement 11 7/5/86 574 Unplanned shutdown due to. loss Equipe.ent of condenser vacuum resulting failure -

, from failure of expansion joint Rubber gasket between the condenser and in joint failed, turbine casing. Maintenance completed j temporary repairs to the gasket.

(Ma?ntenance) 7/6/06 Startup j 12 7/6/86 0% ' Unplanned shutdown following Personnil error -

a break of one blade from No. 31 lacking knowledge.

low pressure turbine. The damage Inad<quate guidance occurred due to operation of regarcing 1.mits i

the turdine at low load with for low condenser a partial loss of condenser vacuum during plant

vacuum, opsrations.

,  ; (Operations)

9/2/86 Startup

t T5-5 #,

l 13 9/5/86 56% Reactor trip following turbine Ecuipment trip occurred when the motor failure - des gn operated disconnect switch on deficiency, the generator transmission Control ci cuit line opened. The faulty for the utpment equipment is located in the locate-switchyard which is not under under ound the itcensee's control. allo ing moisture (LER 86-10) to nter.

aintenance)

, 9/8/86 Startup 14 9/9/86 99% Reactor trip due to low steam Personnel error -

generator water levels follo ng equipment the loss of No. 32 main be er malfunction.

feed pump. (LER 86-11) Maintenance did not replace a "

, control oil seal which iarginally inet toleranc.e requirements.

Que to an identified design deficiency, protecting the control oil against contamination is critical for proper operability.

(Maintenance) 9/10/86 artup 15 11/14/86 100% Reactor trip caused by the loss Personnel error -

of D.C. control power to Train poor judgment /

8 reactor trip breaker. lack of knowledge.

(LER 86-12) Failure to take tdequate precautions during an engineerien inspectir in  !

electri 41. l (Operat 1 /15/86 Startup 16 1/31/87 100% Reactor trip due to low level in Equipment .

No. 31 steam generator caused malfunction -

by loss of No. 32 main boiler l design deficiency l feed pump. (LER 87-01) Dirt in the control f

. _ _ _ _ _ _ _ . - _ . - - - - - - - - - - - - - --- ~

!! ~

TS-5 A 13 9/5/86 56% Reactor trip following turbine Equipment  :

trip occurred when the motor failure - design .

operated disconnect switch on deficiency. '

the generator transmission Control ciredt line opened. The faulty for the equipment '

i equipment is located in the located '

. switchyard which is not under underground the licensee's control. allowing moisture l

(LER 86-10) -

to enter.

(Maintenance) 9/8/86 -

Startup

. 14 9/9/86 99% Reactor trip due to low steam Personnel error - .

generator water levels following equipment  !

the loss of No. 32 main boiler malfunction. i

, feed pump. (LER86-11) A water seal was l !

installed improperly l

, by a licensee * '

contractor during i

, a pump overhaul. I j (Maintenance) l 9/10/86 Startup ,

' i 15 11/14/S6 100% Reactor trip caused by the loss Personnel error -  !

1 of D.C. control power to Train poor judgment / i l 8 reactor trip breaker. lack of knowledge.  !

j (LER 86-12) Failure to take adequate i precautions during

an engineering

! inspection of an j electrical panel,

- t a

(Operations)  ;-

11/15/86 Startup i 15 1/31/87 100% Reactor trip due to low level in Equipment ,

No. 33 steam generator caused malfunction -  ;

by loss of No. 32 main boiler design deficien::y t feed pump. (LER 87-01) Dirc in the control  !

I P

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TS-6 -

4 oil system caused the pump to trip.

(Engineering Support) 2/2/87 Startup 17 2/11/87 100% Reactor trip due to steam flow / Equipment feed flow mismatch coincident malfunction -

with low steam generator levels design deficiency.

followed by a safety injection In the course of signal. Loss of No. 34 an instrument (

instrument % s resulted in bus loading review runback of both main boiler in response to feed rumps and the generation.Bulletin 79-27 of 1sw steam line pressure engineers did not straals. (LER 87-02) identify redundant controls and

• instrumentation  !

supplied by the ,

same instrument busses.

(Engineering

. Support) 2/13/87 Startup '

t IS 3/27/87 100*. Unplanned shutdown to replace Equipment No. S3 reactor coolant pump malfunction.  ;

, seal package, (Maintenance) 4/4/87 Startup  !

19 4/4/87 100% Reactor trip due to turbine Equipment trip caused by unstable turbine I malfunction. L governor during turbine rollup. (Maintenance) ,

(LER 87-03) 1 (

4/4/87 Startup i s

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TS-7 20 4/17/87 100*; Reactor trip due to stea.a flow / Equipment i feed flow mismatch coincident failure - design with low steam generator level, deficiency, followed by a safety injection Similiar derign -

signal. Loss of No. 33 deficiency as 17 instrument bus caused all feed above and  ;

regulating valves to go closed. Identified  :

Low steam generator pressure following the  !

on two out of four instruments reactor trip on wcs also generated due to the 2/11. i locs of instrument bus. (Engineering (LE187-04) Support) i j 4/18/87 1 et Jp  :

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TABLE 6 f

SALP HISTORY .

t 4

7/84- 1/83- 1/82 4/81-Functional Area 11/85 6/84 12/82 4/82

1. Plant Operations 2 1 1 2 i

2. Radiological Controls 1 1 1 1

3. Maintenance 1 1 1 1

4. Steam Generator Repairs N/A N/A N/A l 1

5. Surveillance 1 1 1 1 -  !

6. Fire Protection / Housekeeping N/A 1 l' 1

7. Emergency Preparedness 1 1 1 2

8. Secu,rity and Safeguards 1 1 1 2 i

9. Outages 1 1 1 N/A

10. Training and Qualification 2 N/A N/A N/A Effectiveness

11. Licensing Activities 2 2 3 2 I

- r i 12. Assurance of Quality 2 2 N/A N/A  !

]  !

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7,Luz 7 NUMBER OF DAYS SHUT 00WN INDIALPykTSTATION-UNIT 3 a- .r '330 0 J4- <y -

l'ro6 0 Fe .m L. O Marc- 1985 0 Ap.11 1936 5 May 1986 22*.

June 1986 1 July 1986 26 "

August 1986 30**

September 1986 5 October 1986 0 November 1986 0 December 19a6 0

.?anuary 1987 0 February 1987 2 March 1987 4 April 1987 3 May 1987 30"*

"Mid-Cycle Outage - Steam Generator Inspection "Turbine / Transformer Repair Outage

""Refueling Outage

4 s

ENCLOSURE 5' SAlp REPORT ERRATA SHEET' f

, PA_Q{ LINE BOARD REPORT AMENDED REPORT  !

1 9 1-3 ... a sense of urgency to ... and minor errors i return to power operations, in judgment con- I and minor errors in judgement tributed to reactor '

contributed to rea: tor trips, trips. '

Basts: The perceived sense of urgency to return to power  !

operations on the part of site management could not be l clearly identified as a contributing factor to reactor ,

trips. The intent of the subject paragraph is to t alert the licens6e's manapment to improve the post-trip review process and conduct more detailed and l timely root cause analyses, i i

16 23-25 One challenge to the reactor One challenge to the protection system resulted from reactor protection the lack of conservatism, during system resulted from routine maintenance of a main the improper  !

boiler feed pump, by failing installation of a  !

to replace margical parts. feed water pump  ;

seal t;y a contractor  :

during the 1985  !

refueling outage.

Basts: The root cause was not fully rer:ognized at the time  !

of the seal failure. Subsequently, the licensee {

determined that seal failure would have occurred r regardless of the station maintenance activity.

{

28 last ...both the SRC and the ... the FORC failed...

sentence PORC failed... Subsequently, an $RC  ;

initiated audit of l control room activities {

identifted the LTOP i problem. Appropriate {

corrective action i followed, i Basis: {

The PORC's choice to not apply the selective followup i tracking system resulted in the incomplete implementation i of a Technical Specification amendment. The weakness i was identified by an SRC initiated quality assurance j audit and followed by corrective actions. >

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• o PAGE LINE ECARD REPORT AMENDED REPORT  ;

T5-5 Entry 14 Personne' error - equipment Personnel error -

malfunction. Maintenance did equipment malfunction. I not replace a control oil seal A water seal was .

which marginally met tolerance installed improperly  ;

requirements. Due to an by a licensee ,

identified design deficiency, contractor during  ;

protecting the control oil a pump overhaul.

. against contamination is critical (Maintenance) for proper operability.

Basts: The root cause was not fully recognized at the time of the seal failure. Subsequently, the licensee determined i that the seal failure would have occurred regardless  :

of the station maintenance activity. i I

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I 2-4 all four steam generators. The cracks were ground out of the girth weld, and an analysis was perfomed to determine if the repairs were acceptable. An interim safety evaluation allowing operation was issued on December P4,1987, and was forwarded to you by letter dated January 15, 1988. The staff completed l

its review of the licensee's analysis and a final safety evaluation was issued

by letter dated Octeoer 28, 1988 A copy of this evaluation is enclosed for

. your information.

1 The Commission is concerned about the safety of the Indian Point plants as well

as the safety of all nuclear power reactors and provides day-to-day oversight M l

licensed activities through the four NRC resident inspectors assioned to the Indian Point facilities. This oversight is augmented by Headquarters and region-based inspections. In addition, the overall perfomance of the j facilities is evaluated once every 12 to 18 months in accordance with the Comission's Systematic Assessment of Licensee Performance Program, j

In sumary, the Indian Point units are being operated within the terms and conditions of their licenses. On the basis of the overall operational history i

of the units and continued NRC oversight of the licensees' activities, the NRC l

has reasonable assurance that. Indian Point Units 2 and 3 are operating without j endangering the health and safety of the public.

I trust that this letter responds teN your concerns about the safe operation of the Indian Point plants, s

~

Sincerely, 1

4 i

l Lando'W. Zech, Jr.

4 l

Enclosure:

As stated neA b

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MSlosso(mdig DNeiphhors RCapra 10D#f88 10/p/88 10/Jt/88 j fM0/J6/88

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'h' DDO:NRR JSniezek DDO:NRR TNurley FP t' 10 /88 10fj/88 1 / )/88 10/ /88 10/ /88 GPA/CA EDO OCM VStello LWZech 1 /s 10/ /88 10/ /88 10/ /88 i

i "This carrespondence addresses policy issues previously resolved by j

i the Commission, transmits factual information, or restates Commision I

policy."

3 L- ~ - . _ _ _ _ _ _ _ , . _ _