ML20092F272

Proposed TS 3/4.11 Re Radioactive Effluents

ML20092F272
Person / Time
Site:	Seabrook
Issue date:	02/07/1992
From:	PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:
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ML20092F271	List: ML20092F268 ML20092F272
References
NUDOCS 9202190226
Download: ML20092F272 (3)

Text

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, 3/4.11 RADICACTIVE EFFLUENTS 3/4.11.1 t!0VID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents 6 the point of discharge from the multiport diffuser (see Figure 5.1-3) shall be limited to the concentrations spec'fied in 10 CFR Part 20. Appendix B, Table II, Column 2 for radionuclides other than dissolved or ent*ained noble gases. For dissolved or entrained noble gases, the concentratton shall be limited to 2 x 10 4 microcurie /ml total activity.

APPt.ICABILITY:; At all times.

' ACTICN:

With the concentration of radioactive material released in liquid effluents at the point of discharge from the multiport diffuser exceeding the above limits, restore-the concentration to within the-above limits within 15 minutes, SORVEILLANCEREQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes thall be-sampled and analyzed according to the sampling and analysis program specified in Part A of the 00CM.

-4.11.1.1.2 The results of the. radioactivity analyses shall be used in accordance with the methodology and. parameters in the ODCM to assure.that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1.

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RA010 ACTIVE EFFLUENTS

,..,4 LIQUID EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3,11.1.2 The ;'ose or dose commitment to a MEMBER OF THE PUBLIC from radioac-tive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (*ee Figure 5.1-3) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 mrems to any organ, and

b. During any calendar year to les, than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ.

APPLICABILITY: At all times.

ACTION:

a. Witn the calculated dose from the release of radioactive materia's in liquid effluents exceeding any of the above limits, prepare id submit to the Commission within 30 days, pursuant to specificat ;n

.i d, 6.8.2, 4 Special Report that identifies the cause(s) for exceect,g

" C,') the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

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b. The provisions of Speci ication 3.0.3 are not applicable.

i SURVEILLANCE REQUIREMENTS - 4.11.1.2 Cumulative dose contributions fron,licuid effluents for the current calendar quarter and the current calendar year shall be determined in accor-dance with the methodology and parameters in the 00CM at least once per 31 days.

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NOV 011991 W 0 71991  !

Docket No. 50 443 l UCENSlNG i Public Service Company of New Hampshire '

ATTN: Mr. Ted C. Feigenbaum President and Chief Executive Of0cer i New Hampinire Yankee Division 1 Post Of0cc Box 300 Seabrook New Hampshire 03874  :

Subject:

NRC Region ! Inspection 50 443/9129 (9/10/91 10/14/91)

This refers to the above subject safety inspection of open items and the Anal phascs of the -

Orst refueling outage at Seabrook. The inspection results are desenbed in the enclosed repon and were discussed with Mr. B. Drawbridge of your staff.

Overall, the inspection found that activities associated with the completion of the refueling  !

outage were conducted safely.

However, an apparent violation of NRC requirements is cited in the enclosed Notice of  ;

Violation (NOV) for (tilure to restore a manual valve in the demineralized water system following maintenance restilting in the leakage of reactor coolant system water into secondary systems. Please respond to the apparent violation in accordance with the directions in the NOV.

Thank you for your cooperation.

Sincerely,

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C. Lmville, ef jects Branch N . 3 '

Division of Reactor Projects

Enclosures:

1. Nntice of Violation

2. NRC Region ! Inspection Repon 50 443/9129 <

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ENCLOSL*RE1 NOTICE OF VIOLATION Public Service Company of New Hampshire Docket No. 50-443 Seabrook Unit 1 License No. NPF 5e Dunng NRC inspection from September 10 Octobe? 14, 1991, a violation of NRC requirements was identified in accordance with the ' General Statement of Policy and Prxedure for NRC Enforcement Actions," 10 CFR, part 2. Appendix C. That violation is '

listed below:  !

Technical Specification 6.7.1.a requires that the procedures recommended in Appendix A of Regulatory Guide 1.33 Revision 2, February 1978, be established, implemented, and maintained. Regulatory Guide 1.33, Revision 2 February 1973, i Appendix A, Section 1.c specifies in part that procedures be established for equipme" i control (e.g., locking and tagging). New Hampshire Yankee procedure MA4.2, Equipment Tagging and Isolation, Section 4.9 requires that tagging order boundary components be restored to their proper positions. Precedure ON1055.01, Revision 4 (Change 20), Fill and Vent of Demineralized Water System, specified the normal position of DM V301 as closed.

Contrary to the above, on about September 30,1991, & mineralized water system valve DM V301 was aligned in the open position during system restoration following completion of maintenance on the letdown line radiation monitor resulting in contamination of the demineralized water system.

This is a Severity Level IV violation (Supolement 1).

Pursuant to the provisions of 10 CFR 2.201, Public Service Company of New Hampshire is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory.

Commission A'ITN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region I, and if applicable, a copy to the NRC Resident inspector, within 30 days of the date of the letter transmitting this Notice of Violation. 'this reply .

should N clearly marked as a " Reply to a dodce of V_lolation" and should include for each '

violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved (3) the corrective steps that will be taken to avoid further violadons, and (4) the date when full compliance w.ll be achieved, If an adequate reply is not received within the tirne specified -

this Notice, an order may be issued to show cause why the license should not be modified, suspended, or revoked, or why such other action as may be ppper should not be taken, Where good cause is shown, consideration will be given to caending the response time.

Under the authority of Section 152 of the Act,42 U.S.C. 2232, this response shall be submitted under oath or affirmation.

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O. S. NUCLEAR REGUI.ATCRY COMAtlSSION REGION I Docket / Report No.: 504 0/9129 License No.:NPF 86 Licensee: Public Service Company of New Hampshire, New Hampshire Yankee (NHY) Division .

Facility: Seabrook Station, Seabrook, New Hampshire Dates: September 10 October 14. 1991 Inspectors: N. Dudley, Senior Resident Inspector. Operadons A. Cerne, Senior Resident inspector Construction S. Wookey Resident inspector D. Moy, Reactor Engineer J. Jang, Senior Radiation Specialist J. Noggle, Radiation Specialist Approved By: ^ ^ ^ * - ' d /

W. JMazarusM;411ef, Reactor Projects Section 3B Date OVERVIEW Onerations: A violation was cited for use of an uncontrolled valve position database which contributed to contamination of the demineralized water system.

Radiological Controls: The response to the contamination of the demineralized water header was excellent.

Maintenance /surntilanc=: Control and conduct of activities improved. Troubleshooting efforts on the air start distributor for diesel generator B were extensive.

Security: Implementation of program requirementa were good.

Engineering /Tachalcal Support: Support of activities required to be completed by the end of the first refueling outage was excellent.

Oumilty A===~a/Catary Verirsemelon: Resolution of safety concerns related to welding documentation and Cryofit couplings was supported by station management, i .

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U O V E R V I EW . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . *  ;

TABLE OF CONTENTS ...................................... n 1.0 S U M M A R Y OF ACTIVITI ES . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . I 1.1 NRC A c tivitie s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .  !

1.2 Plant Ac tivities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . I 2.0 OPERATIONS 2,1 Plan t Tou rs . . . . . . . . . . . . . . . . . . . . . . . . . . . .  ;

2.3 Ope ra to r A i d s . . . . . . . . . . . . . . . . . . . . . . . . . ..........

.......... 3 2.4' Demineralized Water System Contamination . . . . . . . . . . . . . . . . . 3 3.0 RA DIOLOGIC A L CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . $

3.1 Assessment of Offsite Effects from the Contamination of Demineralized Water System .................................. <

3.2 Assessment of On Site Effects From the Contamination of the Demineralized Water System . . . . . . . , . . . . . . . . . . . . . . . . . .

3.3 Control of Contamination in Containment ................ ... 9 4.0 ~ MAINTENANCE / SURVEILLANCE . . . . . . . . . . . . . . . . . . . . . . . . , . . 10 4.1- Pl an t Tov es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 4.2 Diesel Generator Air Stan Distributor . . . . . . . . . . . . . . . . . . . . . . 10 4.3 - Diesel Generator 18 hlonth Surveillance . . . . . . . . . . . , . . . . . . . .

, t l-5.0 S EC U RIT Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-6.0 ENGINEERING / TECHNICAL SUPPORT , . . . . . . . . . . . . . . . . . . . . . 12 6.1 Feedwater Check Valve Bolting Modi 6 cation . . . . . . . . . . . . . . . . .

12 6.2 .1himble Tube Thinning Bulletin No. 8849 . . . . . . . . . . . . , , . .-. . 11-6.3 - S tartup Preparations . . _ , . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . 14 6.4- Core Exit Thermocouple High Temperature Alarm . . . . . . . . . . . . . . 14 6.5 Diesel Generator Jacket Water Temperature Control . . . . . . . , , . . . 14 6.6- Westinghouse Relays and Magnetic Contactors Information Notice 91-4 5 (Closed) . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... 15 6.7 Steam Generator Level Channels Filter Card Addition . . . . . . . . . . . . 15 7.0 QUALITY ASSURANCF/ SAFETY VERIFICATIOP.! ._............... 16 7.1- Evaluation of Contamination of Nonradianctive Systems . . . . . . . . . . . Ie 7.2 Reactor Coolant System Unidentined Lankage LER 91410 (Closed) . . . 17 7.3_ - Unlocked Circuit Breaker _ LER 91005 (Closed) . . . . . . . . . . . , . . 17 7.4 _ Missing Radiographic Record for a Safety Related ASME Field Weld -

Unresolved Item 90 2 4 02 (Open) . . . . . . . . . . . . . . . . . . . . . . . . - 13 8.0 M EETI N G S . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . . . . . . . _ . . . . . < .

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l.0 SWIMARY OF ACTIVITIES 1.1 - NRC Activities i

Three resident inspectors were assigned. Three regional inspe: tors assisted in the inspection of restart commitments and a contamination event. Backshift inspec' ions were conducted on i 9/12,9/16,9/22,10/1,10/2,10/7, and 10/9. Deep backshift inspections were conducted on 9/22. 9/25,10/5,10/6,10/12,10/13. and 10/14 1.2 Plant Activities The plant was in a refueling outage. Major work included reactor vessel head replacement. I repair of diesel generator B air start distnbutor, replacement of Cryofit couplings, equipment retests, recovery from contamination of the deminerallred water header, and repair of the hydrogers seal on the turbine generator.

The plant entered Mode 4. Hot Shutdown, on October 5 and the reactor was tAken entical on October 9. Startup physics testing was completed and power was raised above 5% on  ;

October 12.

l 2.0 OPERATIONS 4

2.1 Plant Tours The inspector conducted daily control room tours, observed shift turnover, and attended da21y plan of the day meetings. The inspector reviewed mode change checklists, containment - .

integnty. compliance with Technical Specification requirements, tagging orders, and vahe lineups. No deficiencies were noted. '

i Routine tours were made of the containment, the spent fusi building, diesel generator building, service water building, switchgear rooms, and the pipe chases. Two valves in the '  ;

containment, St.V117 and CGC V51, were noted to have less than full thread engagement on  ;

one body to bonnet bolt. De Technical Support engineers provided documentation of an Engineering Evaluation completed on March 8,1990, which documented the acceptability of the thread engagement on valve SI V117. De thread engagement of the bolt on valve CGC.

• V51 was corrected. Other de6ciencies noted by the inspector were veri 6ed to have been included on the Maintenance Department clussout discrepancy lists. Other minor-  ;

housekeeping deficiencies identified to the Operations Department were corrected.

Before .tontainment closeout, the inspector walked down port!ons of the Safety injection (SI) and Residual Heat Removal (RHR) systems la the containment and verified that the SI and -

L RHR valves were properly positioned and locked.

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Planning for the outage minimized the risks identined in the Shutdown Probability Risk Assessment deseloped by New Hampshire Yankee. At least one protected train of Emergency Safety Features equipment wu maintained throughout the outage. Afier completion of the four hour operability test of diesel generator (DO) A fouowing an 18 month oserhaul, train A became the protected train and disusembly of DG.B was commenced. The next day a Nuclear Quality Group (NQG) inspector found that one of the brush nggings on DG A was cocked and scraping the slip ring. Even though DG A was operable. DG B was reassembled and tested for operability so that the cocking of the brush ngging on DO A could be corrected without entenng a Technical Speci6 cation Acuon i Statement.

The decision to maintain one diesel operable at all times and to immediately conect the identined deficiencies redected a proper safety perspective.

On September 12, with the level of water in the refueling cavity below 23' above the rea :et vessel Dange in preparation for reactor vessel head installation, the Shift Supenntendent was informed that the Inservice Inspection (ISI) of the RHR B heat exchanger ginh weld showed a potential weld defect. Train B of RHR wu in operation. The Shift Supenntendent entered Technical Speci0 cation 3.4.10, ' Structural Integrity,* which required that the structural integrity of ASME Code Clus 2 components be restored prior to increasing reactor coolant temperature above 200 degrees F. Also, Technical Specification 3.9.8.2 ' Residual Heat  :

Removal and Coolant Circulation, Low Water Level t was entered based on the assumed inoperability of RHR train B. Preparations for the reactor vessel head replacement were j stopped, RHR train B remained in operation, an evaluation of the indications of a weld defect was initiated, and the water level in the reactor cavity was maintained. After funher review of the weld and ISI data, the Technical Suppon Depanment determined that the init:aj indications were questionable and that RHR train B should be considered operable. The SN':

Supenntendent exited both Technical Specifications.

The next day, following discussions between the inspector Operations personnel, and Licensing personnel, Technical Speci6 cation 3.4.10 wu re-entered pending final resolut:en of the indication of a defect in the heat exchanger ginh weld. The indication was determined to be a result of the ultrasonic technique used and the coarse grsin boundaries in the intenor of the weld. Further details of the ISI aspects of this event are contained in NRC Inspecnon Repon 50-443/9127.

The inspector reviewed the Technical Specifications and their bases, discuated the 151 findings with Technical Suppon personnel, and observed the discussions betweert Operatons.

Licenstng and Technical Suppon personnel. The inspector concluded that the decision by tne Shift Supervisor to declare RHR train B inoperable and enter Technical Specification 3.9 8 :

was not necessary. In addition, the inspector noted that the action statements were not fully implemented. The decision to initially exit Technical Specification 3.4.10 was premature but was properly reevaluated by the Operations and Licensing Department. The inspector concluded that the requirements of the Technical Specifications were met.

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2.3 Operator Aids f

Dunng the plant walkthrough portion of operator licensing examinations, several weaknesses were noted by th1 NRC examiners in regards to operator aids and ptccedures. The Operauons Department took actions to address these weaknesses. The inspector reviend tr.e implementation of the corrective actions.

Emergency tools required for manu*lly overriding the main steam isolation vaher (MSIV) wcre asalable and labeled at each MSIV. The inspector page checked selected controlled operating and surseillance pro:edures and found that all inserted pages identified by a page number followed by an 'a' were present. Since June 1991. Station Procedures require that a revision of a procedute must tenumber all pages. An audit of selected working procedures m the nie cabinet in the rnain control room determined that the appropriate revisions of the procedures v.ere on file.

The inspector reviewed the administrative controls over five operator aids available in the main control room. Four were controlled by the " Operator Aids Control' book which indicated the aids on the index and all but one of the aids in the tabbed section of the book.

The Operations Department corrected the minor administrative deficiency. The Afth aid was controlled by procedure OX 1408.02,

• Boron injection Flow Path Monthly Valve Alignment Check", and updated monthly.

The inspector concluded that the identified weaknesses were adequately addressed.

2.4 Demineralized Water System Contamination On September 30, the demineralize- cater system became contaminated due to leakage from the Reactor Coolant System (RCS). The leakage was initiated when the pressure in the demineralized water system dropped (as a result of filling of the steam generators) below the pressure in the letdown radiation monitoring system causing the solenoid operated Sushing water valve, RV4520-02, to come off of its see: A deminerslired water manual foolation valve, DM V301, was improperly left open comp:eting the now path to the demineralized water system.

The now path remained open for approximately sit hours and an estimated 1250 gallons of RCS water entered the demineralized water system. Activity levels due to Co-58 and Co 60 ranged from 108 pC/cc at the eye wash station closest to the flushing line to 10' pC/cc in the Condensate Storage Tank. Most of the contamination ramained in the demineralized water system. Contammated water which reched tlie secondary oil / water separator sump No. 2 and the turbine building sump, was pumped through hoses to thv circulating water discharge structure. The unplanned radioactive efnuent release and the radiological conditions created by the event are discussed in Details 3.1 and 3.2, respectively.

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6 The inspector reviewed the Technical Suppon Department's Event Evaluation, station ! cgs, and alarm pnntouts. The inspector attended SORC meetings and held discussions w h pia t staff concerning the event.

The m.un control room received a high radiation monitor alarm from the turbine buildmg sump discharge monitor some time after the RCS leakage began. The alarm terminated me discharge from the sump and the sump was sampled. No activity was found. The radiatten monitor setpoint was adjusted based on the background radiation levels and the discharge 4as reestablished.

The Volume Control Tank level chart recorder in the main control room was inoperable au level was being monitored from computer points which were trended on a visual display terminal. Level was noted to be decreasing at a higher rate than normal, which was con 6rmed by sescral automatic RCS makeups. The control room operators began renew ;

all work in progress in an attempt to identify potential sources of RCS leakage. The leakage to the deminerajized system was identified after maintenance workers became contamina:ed and a second alarm was received from the turbine building sump radiation monitor. A second sample from the sump indicated activity Checks by Health Physics (HP) technic:ans of the radiation dose rate from the demineralized water piping helped identify and termina:e the source of the RCS leakage.

The licensee's Event Evaluation Team determined that the solenoid valve functioned as designed. An earlier Engiacering Evaluation determined that the design was adequate since there was no identined plant conditions that would develop the differential pressure needed w unseat the vahe.

The manual isolation valve, DM V301, was opened on September 19, 1991, in accordance with an incorrect tagging restoration sheet following maintenance on the CVCS letdown ime radiation monitor. The error was the result of specifying the restoration valve lineup soleh by reference to valve position information in a developmental, computerized tagging database. His database wu not intended to be a replacement for system alignment procedures, schematics, or P& ids, but as an informadonal tool to supplement them.

Procedure MA4.2 Section 4.9 states that components not identified on P& ids, electrical schematics, or in the procedures will be restored using other documentation. De operating procedure for the demineralized water system, ON1055.01 A, contained a system lineup checklist which indicated that DM V301 should be closed. A complete valve lineup for the demineralized water system was not conducted at the end of the outage because no work had been performed on the system and it was not recognized that the work on another system eat interfaced with the demineralized water system had affected the valve lineup. Failure to restore the deminerajized water system in accordance with procedures MA4.2 and OP1055.01, after maintenance on the letdown line radiation monitor, is a violation of Technical Speci6 cation 6.7.1.a (NV4 9129-01).

3 The Operations Department completed a valve lineup of the demineralized water sptem af:er l the event and conducted a review of the tagging restorations produced by the taggmg ,

computer to senfy proper valve lineups. No additional errors were noted in the computer i database. '

Dunng subsequent flushing of the demineralized water system, some contamination occurred in the secondary chemistry lab when a drain system backed up. The secondary chemistry lad was decontaminated and a work request was written to unclog the drajn system. Dunng filling of the condensate storage tank,200 gallons of water spilled on the ground. The '

concentraMon of radioactivity in this water was well below the level that qualified for unlimited release. The slightly contaminated soil was placed in 55 gallon drums and guidance was provided to maintain the level in aji tanks below the high level alarm setpom:

3.0 RADIOLOGICAL CONTROLS 3.1 Aswssment of Offsite Effects from the Contamination of Demineralized Water System Following the inadvertent contamination of the Demineralized Water System, the licensee's effluent and environmental radiological assessment were reviewed by a region based inspector. The inspector toured areas of the letdown radiation monitoring system, turbine >

building, auxiliary boiler, demineralized water storage tanks, condensate storage tank, oillwater separator vault No. 2, settling basin, and storm drainage systems. ~

Approximately 45,000 gallons of slightly contaminated water was released from the tutbme building sump and oillwater separator vault No. 2 to the ocean. A hose had been connected to transfer nonradioactive water from the oil / water separator vault No. 2 to the discharge building during the outage. During the event, the water in vault No. 2 became contaminated and was discharged via the hose to the discharge structure. Subsequent to the discovery of this event, the licensee identified two small leaks from the hose near a storm drain. This storm drain was connected to the settling basin from which the licensee controls the water released to the Browns River. Also, an estimated five gallons of contaminated water escaped to the auxiliary boilet building roof through the auxiliary boiler system.

The licensee took grab samples from the following areas to assess radiological impact to the public health and safety and to the environment, as well as to determine the plant contamination status.

• Turbine Building Sump

• Oil / Water Separator Vaults No. I and No. 2

• Auxiliary Boilers A and B

• Auxiliary Boiler Building Roof

• Condensate Storage Tank

• Demineralized Water Storage Tanks

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e Primary and Secondary Component Cooling Water e Reactor Makeup Water Storage Tank

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e l Storm Drain Manholes (six kcationst  !

e Settling Basin Water e

i Browns River Water Ocean Water

• Potable W ter {

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Associated Plant Systems  ;

The inspector reviewed all results of radioactive measurements and radiological dose I asscssment and noted that Co 58 was the dominant gamma emitting radionuclide. All  ;

measurement results for environmental samples, storm drain water, settling basin, Browns ,

River, end ocean (sample collected in Salisbury, Musachusetts) were less than the lower  !

limits of detection (LLD) with exception of the soil collected in the vicinity of the leaky hose  !

near the storm drain. Radiolorical analytical results of contaminated soil samples collected from the leaky hose appeared to be localized and Co 58 at a concentration of 1.53+/ 0.06 s ,

10* uCilgram was the only identified plant.related radionuclide in the soil samples. Even '

though the activity of the contaminated soil was low (almost same activity of Cs.137 in soil due to fallout, and about 3% of natural background dose), the licensee treated the

" contaminated soil as radwasta. The soil was removed and stored in 55 gallon drums.

The licenses performed a dow asussment using the Offsite Done Calculation Manual t

. methodology for the turbine building sump and oil / water separator vault No. 2 release. The licensee did not know the total volume of release, therefore, the licensee enveloped the  :

. calculated amount of release based on time of h beginning of the event. The results of the radiological dose auessments for the whole body and organ _dones were 4.76x10' mrem and j

2x10* mrem, respectively to a member of the public involved in shore activides such as swimming and boating. Those projected dows were less than one percent of monthly  ;

Technical Specification limits. >

Based on the review of the licensee's measurement techniques, analytical results, and actions, the inspector determined the following, e ne licensee has an sacellent capability to accurately measure gamma emitters. [

• De licensee's actions to monitor the possible leakage to the environment was excellent.

• The licensee has the capability to perform the necessary radiological done aswument. ,

The inspector concluded that there was no negative impact on the environment or to the public health and safety as a result of this event, O

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7 3.2 Assessment of On. Site Effects from the Contamination of the Deminerallred Water System The licensee's onsite radiological control and handling of the event was reviewed by a region based inspector. The inspector witnessed the operations and technical support recoury team in planning and beginning the recovery activities. The inspector toured the new Radiological Controlled Areas (RCA) created by the eveat, reviewed radiological surveys, and intmiewed beensee personnel in assessing the licensee's response to the esent.

On September 30,1991 at tround 5:00 p.m., the first contaminated worker assmiated witn this event was discovered. Within the next hour, two more workers were discosered to be contaminated, all apparendy assmtated with demineralized water use. Around 6:00 p.m. :re licensee deduced the possibility of a contaminated demineralized water source. The sour:e of water was posted as a contaminated area and a water sample was taken for radiochemical analysis. The water aas found to be contaminated.

Around 7:20 p.m. the process radiation monitor on the turbine buildlag sump alarmed. At 7:22 p.m. the control room made a statJon wide announcement over the public address system for personnel to refrain from using the demineralized water system until funher notice. By 8:30 p.m. the Hetith Physics Depanment made a preliminary determination that the contaminated demineralized water system extended beyond the bounds of the nonnal RCA and extended contamination monitonng and exposure monitonng controls to include the turbine building and auxiliary boiler building. All entrances and exits from these but! dings were roped off and posted resulting in two access points at which whole body contamination monitonng instrumentation were installed.

Informational flyers were produced and distributed to plant personnel entering the station informing them of the new RCA and associated monitoring requirements. Continued surveying throughout the night and next moming resulted in the posting of four contamination areas in the turbine building and one contamination area in the auxiliary building with the addition of one radiation area closely associated with the Secondary Component Cooling (SCC) Header Tank located at the top of a vertical ladder above the top floor level of the turbine building. The new RCA was completely surveyed about noon on October 1,1991 which was appmtimately eighteen houri following identification of the event. Portable Eyewash Stations were positioned in the Primary Auxiliary Building at those stations supplied by deminerallred water, ne inspector concluded thy the licensee response to the event wu appropriate and timely.

Exacture Imnact From a review of the post event survey data, the following radiological environment in the new RCA was found:

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8 LOCATION DQSE RATES CONTAMINATED AREM Turbine duiking 21 ft. 0.05 0.1 mR/hr 4 areas (5,000 dpm/100 cm9 46 ft., 50 ft. 0.010.03 mR/hr 75 ft. 0.010.02 mR/hr SCCW Head Tank 75 ft. 2.5 mR/hr Aux Boiler Building 0.1 1 mR/hr I area (5,000 dpm/100 cm9 These general area radiation levels were the initial conditions before any system decontamination flushes were performed. An average new RCA dose rate was estimated te be 0.08 mRlhr. A worker continually exposed to this radiation neld for 5:0 hours per calendar quaner could theoretically receive 42 mrem.10 CFR 20,202 does not require personnel monitonng below 312 mrem per calendar quarter. Therefore, the external exposure hazard was minimal.

8 The contamination level of 5,000 dpm/100 cm wu a low concentration of activity which was riot expected to pose any intemal exposare hazard. The licensee instituted an air sampling program in the turbine building to serify the lack of an airborne radioactivity hazard. The plant's potable water supply was repeatedly sampled during the event with no trace of radioactivity detected.

The inspector concluded that the radiological conditions found as a result of the e ent, posec no signincant additional external or intemal exposure hazard to the station personnel. A preliminary estimate suggests that there will be an increase in the radioactive waste generated in the form of spent ion exchange resins and Dry Active Waste (DAW) resulting from the cleanup of the various contaminated areas.

Nonradioactive System Controh The inspector reviewed the licensee's program for routine sampling er monitoring of the various nonradioactive systems of the plant with respect to IE Bulletin 8010 and IE Circular 80-14 IE Bulletin 8010, " Contamination of a Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release to the Environment" recommends a roe te sampling / analysis or monitoring program be established for those nonradioactive systems aat hase interfaces with radioactive systems to allow identincation of contammating events. Tha same guidance also states that if a nonradioactive system becomes contaminated, the affected system should be restncted in its use, the cause of the contarninating event identined, and the affected system components decontaminated. The guidance continues by suggesting a formai 10 CFR 50.59 safety evaluation be performed immediately if the normally nonradioacuse system is to be operated in a contaminated condition. The inspector veri 6ed curTent sampling / analyses had been performed according to the following schedule, l

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9 Nonr2dioactise Systtm Samoline Frecuency Senice Water Weekly Demir.erajired Water Header Monthly instrument Air System Monthly Sewage Treatment Sludge Annually Settling Basin Sludge Annually After acknowledgement of the event, the licensee restncted the use of the demtneralized water system and began a 10 CFR 50.59 safety evaluation to identify any unreviewed safety concerns. The licensee formed a round the clock team to prevent any further releases to tne environment, to identify the cause of the event, and to develop an operational recovery plan to effect a methodical system flush or decontamination. The inspector concluded that licensee actions were taken in accordance with stated regulatory guidance.

Eient Recoverv The licensee added five Health Physics technicians to its staff to assist in the long term monitonng and contamination control of the contaminated secondary systems. A weekly radiation and contamination surveillance program was initiated for the new RCA plant areas.

Additional radiological work controls were implemented in new areas requinng Radiation Work Permits (RWP) for any work involving breaching of a contaminated secondary system L

and for working inside any of these contaminated areas. Additionally, all vehicles exiung :ne station and any dumpsters leaving the station dunng the first few days after the event wcre sun eyed.

The licensee provided the following future recovery guidance. After the secondary system flushing is completed in October, the radiological conditions will be reasst: sed. Tne decision to remove posted signs and controls from potentially contaminated secondary areas will be made based on dose rates less than 0.6 mR/hr, contamination levels not measurable above background, and the potential for detecting uncontained radioactive material in the areas.

3.3 Control of Contamination in Containment Prior to drain down of the reactor cavity in preparation for stactor vessel head installation, personnel worked in most areas of the containment in work clothes. After the drain down of the reactor cavity, contamination from the cavity became airborne and was spread throughou:

the containment. Access to the containment was restricted until che airborne problem was eliminated. However, efforts at decontamination of the containment were unsuccessful and all personnel entering the containment we.t required to wear full anti-contamination clothing.

The He'Jth Physics Departrrent expected the contamination, comprised primanly Co 58 which nas a half life of 70 days, to decay away prior to the next refueling outage.

10 The inspector concluded that contaminadon control in the containment was good until the cavity was drained. The Health Physics Department plans to review the cavity drain down process to improve contaminatic,n control for the next refueling outage.

4.0 N!AINTENANCE/ SURVEILLANCE 4.1 Plant Tours The inspector obsersed work in progress in the containment, the primary auxiliary building.

the pipe chases, the diesel generator rooms and the tuttine building, in general, the control of matenals and processes in containment were stronger than controls in the pipe chases.

Controls and processes improved as the outage progressed. The organization and control of diesel generator B 18 month maintenance overhaul appeared strengthened as a result of the lessons learned during the 18 month maintenance overhaul of diesel generator A.

Housekeeping in non safety related areas was adequate.

The program to paint and label the doors in the protected area was cornpleted. In addition to the informadon denoting a fire or secunty door, decals were posted for heanng protection and other precaudons associated with the area about to be entered. The use of the newsletter to communicate to workers the status of outage conditions or requirerrents was a good initiadve.

4.2 Diesel Generator Air Start Distributor On September 11. 1991 diesel generator (DG) B was shutdown due to excessive vibrations and oserheadng of the right air start distnbutor. The diesel had run for over two hours as pan of a monthly surveillance test to establish operability after an 18 month maintenance overhaul. Several air line fittings from the air start distnbutor were cracked and the rotor was worn due to contacting the housing. The vendor, Colt Pielstick sent consultants to assist.

The alignment of the camshaft to which the rotor is attached was checked, and the rotor and housing were replaced.

On September 16, DG B was again shutdown during perfo" nance of the monthly surveillance test due to an oil line failure. The oil line was repaind and de surveillance test was restaned. The diesel was later shutdown due to high noise in the area of the air start distnbutor. Two air lines were found severed on the air start manifold, and the rotor was damaged due to contact with the housing. A new housing and rotor from a Unit 2 air stan distnbutor were installed. The oil mister system which provides '+ 4ation for the air start rotor during initial startup was inspected and cleaned. Blockage wu found in the oil supply lines and the vent hole in the original air distnbutor housing was found to have been clogged. The oil mister system on DG A was also inspected and an oil pump was replaced.

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i1 The air sun control valve solenoids and air lines were instrumented and vibration momtors were placed on both air stan solenoids. A series of test runs of DO B were conducted without recurrence of the high vibrations. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> operating test at fullload was conducted. After 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> of operation, the vibrations appeared and DG B was manually shutdown on September 21, 1991.

A design modification to insult an alignment bearing on the end of the air stan distnbutor housing was deseloped by New Hampshire Yankec and the parts were manufactured on site.

The modincanon wu never installed since a realignment of the coupling between the camshaft and the aar start distnbutor rotor resolved the problem. The nnal realignment specincations were provided to New Hampshire Yankee by Colt Pielstick on September :4.

The vendor stated +at each air start distnbutor housing was unique and an alignment wu necessary each time the housing and rotor were replaced.

A review of DG B vibration data indicated an increased level of engine vibration over the last year. An inspection of the foundation and foundation bolts of the diesel idenufled a cracked washer and apparent holes in the grouting. The washer was replaced and the foundation was grouted. An epoxy adhesive wu injected into the holes. The inspector reviewed the procedure (MS91 1 19) and the licensee's controls of the vendor performing the grouting. The responsible technical support engineer was kaowledgeable of the process and monitored the vendor's performance. Technical Support will monitor future diesel starts to determine if the vibration hu reduced as a result of the foundation grouting.

On September 25, DG B failed to surt due to an electrical ground caused by a temporary coaxial instrumentation lead. The ground activated the voltage imbalance relay which caused the voltage regulator to transfer to manual before full field voltage was established. The less than full neld voltage caused the potential transformers to produce less than full control voltage signals resulting in the govemor positioning the fuel racks to minimum position. As a result, the diesel was starved for fuel and stopped. Tne coa.xial lead on the temporary instrument wu replaced with an aJligator clip lead, the diesel was rolled with air, and a successful 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run was conducted.

The inspector noted that the vendor wu consulted early in the troubleshooting process and that New Hampshire Yankee independently developed design changes when implementation of initial vendor recommendations were unsuccessful.

4.3 Diesel Generator 18 Month Surwelllance The inspector observed preshift brienngs and the conduct of portions of the 18 month surveillance tests for DO A. The preshift brienngs were extensive, and copies of the procedures were available for all grsonnel involved. No problems were encountered dunng the conduct of the test. The inspector noted an improvement in the preshift brienng and test conduct over the previously conducted test on DG B.

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5.0 SECURITY During, routine plant tours, the inspector noted security practices which included p ,

access control, compensatory actions taken for impaired security bamers, the conduct of '

routine patrols, and operation of the Central Alarm Station and Secondary Alarm Stauon.

The inspector reviewed actions taken for Fitness for Duty failures and the extent of the '

licensee's teuew of previous work act;vities of the individuals who failed the tesung denciencies were noted. .

6.0 ENGINEERING /TECIINICAL SUPPORT '

6.1 Feedwater Check Valve Bolting Modification '

On April 1.1991, feedwater header check valves were found to have broken cap screws retrn the hternal valve dashplate. The design of these check valves is unique to Scabrcok, but is similar to smaller check vahes used in other facilities. The reason for th determined to be over stressing the cap screws during valve operting transients. The check valves were modified by increasing the number of cap screws on the dashplate from eigh sixteen. The modification was reviewed in NRC Inspecuon Reports 50-443/91-04 and 50-443/91 15.

On April 23,1991 Edward Valves Inc. issued a 10 CFR Part 21 report wruch recommended use of larger cap screws, a wider lower locking ring, and dashplate modifications. On September 25, Ir(11, the NRC issued a safety evaluation of the installeo short term modifications and the design modifications recommended by Edward Valves, Inc.

The NRC staff concluded that the implementation of the design changes would reduce the likelihood of failure of the check valves. The inspector reviewed the completed modincatie ,

package and the associated 50.59 safety evaluation analysis.

No transient loads were identified that could produce a sufficient load to cause elongation o failure of the sixteen 3/8 inch bolts. However, for conservansm, the dash plates and lockin rings were redesigned to withstand the system design pressure and the maximum differential pressure across the dash plate in the valve opening direction. 'Ihe inspector noted that the original design calculations did not evaluate the stresses on the dash plate dunng the open of the check valves and that later calculations based on complea design considerations were indeterminate as to the stresses that were placed on the bolts.

During receipt inspection of the redesigned dashplates, New Hampshire Yankee determined the bolt holes were improperly drilled and returned the dashplates to the vendor for correction. After the installation of the larger bolts, the Technical Support engineer determined that the clearance between the thi:ker bolt heads and the undersid4 of th piston head was insufficient. The piston head was machined to provide adequate clearance.

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13 Dunng the August 1991 outage, the check valves were inspected, No elongation of the bolts were noted. Design Change, DCR 91035, which replaced the sixteen 3/8 inch cap screws with 5/8 inch cap screws and installed redesigned dashplates and Iceking nngs provided by Edwards Vahes, Inc. was implemented.

The inspector reviewed the repairs to main steam isolation vahe MS V 92, which was supplied by Edwards Vajves, Inc. and had experienced sescral failures. The vahe was onginally supplied by Rockwell Intemational, Flow Controls Division which was later acquired by Edwards Valves, Inc. Two solenoid valves and a check valve on the hydrauhc actuator were determined to be leaking by October 28,1990. These valves were replaced with a package which included a new air driven hydraulic pump. The onginal air ensen hydraulic pump was functioning properly. Later on June 27, 1991, the new air dnsen hydraulic pump seized and was replaced.

The solenoid vahes were manufactured by Keane Controls Corporation of Fullenon, Califomia and the air dnsen hydraulic pumps were manufactured by Haskel Engineenng and Supply Company of Burbank, Califomia. The inspector determined that the failures of MS-V 92 were caused by different components which were supplied by different subcontractors.

New Hampshire Yankee conducted an inquiry into an industry reliability data system and found no problems with Edward Valves, Inc. main stevn isolation valves reponed by other utilities.

The inspector concluded that the failures of MS V 92 did not indicate a generic problem with faulty or inadequate equipment being supplied by Edwards Valves, Inc. The inspector concluded that the initial design of the main feed water isolation valves were inadequate for the application, as indicated by the Part 21 Notification. Corrective actions were taken which resolved the design deficiency. The problem was determined to be unique to Seabrook based on the site of the check va'ves; however, the reason for the design deficiency has not been resolved between Edwards Valves, Inc. and NHY. This issue is closed.

6.2 nimble Tube ninning Bulletin No. 88-09 Bulletin No. 88 09 was issued on July 26,1988 and requested the implementation of an inspection program to periodically confirm incore neutron monitoring system thimble tube integnty. The inspector reviewed New Hampshire Yankee's incore instrument thimble tube inspection and monitoring program, RS0727. Rev.0, which was developed in response to the NRC Bulletin 88 09. No deficiencies were roted.

The first incore instrumentation thimble tube inspection was successfully completed by Cramer and Lindell Engineers on September 15, 1991. A previous inspection attempt had been unsuccessful due to developmental problems associated with the test probe. De thimble tubes are of a dual wall construction with fhed incore detector wire between the walls. The diameter of the inner tube is 118 mils compared to the standard incore thimble l

_ _ _ _____.__m___ _ _ _ _ _ _ _ _ _ _ _ _ = . _ _ _ _ _ _ _ _ _ _ _ _ _ -

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'ube diameter of 188 mils. The newly developed probe was of a smaller diameter and provided eddy current data on both the inner and outer walls. A final report on the results of the eddy current test was scheduled for submittal to the NRC in 0:tober. <

6.3 Startup Preparatiorts The inspector reviewed the following plant stanup operating procedures in preparation for observing the startup.

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RS1737.1. Post Refueling Initial Criticality RS1737.2 Post Refueling 1.ow Power Physics Testing RS1738, Power Ascension Testing RN1736. Reactivity Measurement, Rev.2 RN1732, Incore Analysis, Rev.1 RN1733, Flux Mapping System Operation, Rev.!

For each of the above procedures, the inspector serined that the acceptance critena, precautions and initial conditions were adequately stated. The inspector found the procedures to be consistent with the requirements of ANSI /ANS19.6.11985, " Reload Startup Physics Test for Pressurized Water Reactor." No deviations were noted.

The inspectors observed the conduct of portions of Stanup Physics Testing. Reactor Engineering and Quality Control personnel were stationed in the Control Room working with the licensed operators during the physics testing. Pre task briefings were held before enucal steps and communications between the groups were excellent.

6.4 Core Exit Thermocouple High Temperaturt Alarm The implementation of design changes recommended by NRC Generic Letter 8817 were initially reviewed in NRC Inspection Report 50-443/91-06. The core exit thermocouple temperature high alarm was not installed at the time of the inspection. The inspector veri 6ed that the alarm wu installed in the contml room and was set at 200'F as specified in procedure OS1000.12, 'Mid Loop Operation." The inspector also reviewed pmcedure 0S1213.02, ' Loss of RHR While Operating at Reduced Inventory on Mid Loop Conditions."

No deficiencies were noted.

6.5 Diesel Generator Jacket Water Temperature Control The inspector reviewed Design Change Request DCR 90-0049 that modified the existing pneumatic control loop which modulates the diesel generator jacket water temperature contrel valve. The inspector discussed the setting for the diesel generator high temperature trip with Engineering personnel. The inspector verified installation of the derivative unit addition to the existing control loop and the addition of protective covers to the setpoint adjustment knobs for the pneumatic temperature and differential pressure controllers.

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15 Operating expenence during diesel generator surveillance testing had shown that jacket water temperature, which is controlled by a three way temperature control valve, may osershoot the tnp serpoint following a diesel start. The addition of the denvative unit to the control circuit provided a smoother transition from initial startup to steady state operauon with minimal overshoot and avoided approach to the high temperature trip serpoint.

The inspector noted that the design change was extensively reviewed prior to implementatwn A 10CFR50.59 evaluauon and an Interdiscipline Review and a RJsk/ Reliability Review were completed. Seismic qualificauon of the denvative units was dccumented. The addluon of me protecuse cosers on the controller setpoint adjustment knobs satisfied the conecuse action statement included in a September 9.1988 Diesel Generator Special Report from the licensee. DCR 90-0049 included minor system enhancements such as adding sent, drajn.

and instrument test connections, changing a setpoint, and modifying an alarm circuit. The safety evaluation included a review of the system enhancements.

The inspector concluded that the design change was beneficial to safety since the occurrence of spunous high jacket coolant temperature tnp may be avoided.

6.6 Westinghouse Relays and Magnetle Contactors . Information Notice 9145 (Closed)

As desenbed in inspection Report 50-443/91 22, there are 66 similar relays in use at Seabrook Station which may malfunction due to an epoxy compound becoming seminuid when the coil is energized for extended periods. Four of the relays are in the reactor protecuon system. The remainder are used in alarm or cornputer circuits. Dunng the refueling outage, the licensee removed and inspected the four reactor protection sptem relays. One relay showed visual evidence of flowing compound before testing. A spare relay and the four reactor protection system relays were energized at 138 volts de for two hours. At the end of the test period, each relay was dcenergized, disassembled, and inspected. Four of the relays tested satisfactorily, the one with "u found* flow exhibited obvious flowing of the compound. Westinghouse was informed of the test results and noted the results during a meeting at NRC Headquarters on September 18, 1991.

ne licensee reinstalled the four relays that satisfactonly passed the test into the reactor protection system. The relays used in alann and computer applications will remain in service since they are used in a low temperature environment and an not normally energized. The licensee plans to return spare relays to Westinghouse for testing prior to use.

6.7 Steam Generator Level Channels Riter Card Addition NRC Inspection Reports 50-443/9015 and 9017 provided background on a feedwater isolation based on erroneous steam generator level signals following a reactor tnp. The repons addressed the shon term actions taken prior to the implementation of a design

16 I change. The inspector reviewed Design Change Request 90-0041 and the completed work packages which installed a Wesunghouse recommended lag circuit. Selected winng diagrams that were : hanged as a result of the design change were reviewed.

The design change involsed instaPation of circuit cards configured for a lag funcuon in the output of the nanow range steam generator lesel transmitters. The design change meluced a modification to permit routine tesung of the steam generator lesel bistables without jumpers and remosed an unused load dependent level setpoint circuit idenufied by Westinghouse as a potennal cause for a multiple Icop feedwater malfuncuon event.

The inspector noted use of industry expenence in designing this change. The licensee developed station procedures for testing the filter cards in accordance with ISA Standard S67.06 in addition to Westinghouse recommendations. The inspector r . knowledged the posittse initiative to eliminate jumpers for routine testing and to eliminate the potential common mode failure, The work packages were comprehensive and were comp'ci:d m accordance with procedural guidelines.

7.0 QUALITY ASSL*RANCDSAFETY VERIFICATION 7.1 Evaluation of Contamination of Nonradioactive Systems The inspector reviewed the 10CFR50.59 evaluation for operation of cernin nonradioactise '

systems which became contaminated as a result of a reactor coolant system leak into the demineralized water system. The evaluation assumed that all systems which directly, or indirectly, interfaced with the demineralized water system, would operate with some lesel e:

internal contamination until activity levels decayed to less than detectible values. Some systems are normally radioactive and were not evaluated. Other normally non radioactive systems had predetermined actions under which operations could continue if they became radioactive and were not evaluated. The remaining systems were evaluated as a potennal release path to the environment.

The evaluation considered (1) expected off site liquid doses resulting from the Deminertized Water System decontamination and cleanup operations, including maximum potential impact band on the total source term transferred from the primary system, (2) the maximum hypothetical and potential airborne release via secondary system components such as the Auxiliary Boiler aerated vent, and system safety relief valves, and (3) a postulated accident leading to the failure of the outdoor Demineralized Water Storage Tanks. All potenual, as well as expected offsite doses to members of the public, were determined to be less than tre allowable dose limits established in Section 3.11 of the Station's Technical Specifications.

These dose limits were taken from both 10CFR50, Appendix I for routine operations to keep doses to the public "As Low As Reasonably Achievable", and the 40CFR190 total dose standard for combined dose contnbutions from all uranium fuel cycle sources.

17 The inspector verified that the calculated mailmum hypothetical radioactivit exceed Technical Specification effluent release limits or the offsite dose limits. He inspector concluded that the evaluation supponed continteed operations of the con systems.

and analysis.The evaluation was ennservative, thorough, and well supponed by eng ,

7.2 .

Reactor Coolant S) stem Unidentlfled teskage LER 91010 (Closed) l On September 11, 1991 t the licensee staft presented NRC Regional personnel an overview of the in.cstigation and corrective action for the Cryont tube coupling failure which result reactor coolant system leakage. Details of the meeting are provided in Attachment 1. The '

licensee submitted a Licensee Event Report (LER) Supplement on October 2.1991. The event review and the original LER response are reviewed in NRC Inspection Repons $0 443/91 19 and 9122.

The inspector discussed the scope of the investigation, testing program, and hardware changes with the Instrument and Control Engineenns Supervisor. The inspector venfied th removal of Cryofit tube couplings in the containment in instrument lines exposed to ,

temperature and high hydrogen environment. The inspector reviewed the New Hampsh Yankee Cryofit Coupling Verification Program Engineenng Evaluation and the associated procurement documents,  :

i A chemical analysis was performed to verify the material in use was the alloy specified.

Instrument tubing that was installed as a result of the cryofit removal was formed usin original tubing as a template which resulted in a reduced number of conventional tubi connections. The scope of the pip'ng replacement was extensive and included all areas of th reactor coolant boundary within containment at risk for the failure mechanisia.

The licensee expended extensive resources and significant man tem exposure in the effon to address the potential safety implications of the cryotit failures. The inspector concluded tha the Coupling Verificatico program was well defined and in.plemented in a timely manner.

The LER supplement update accurttely describes the scope of the safety related tubing replacement. This itern is closed.

7.3 Unlocked Cirruk Snaker . LER 91005 (Closed)

During performance of procedure OX 1446.07, ' Verification of Locked valves,' the breaker for the rod control cluster change fixture (1 FH RE 12) in motor control center MCC I t !

was found open and unlocked. Technical Specification 3.8.4.1. ' Containment Electrical Penetrations,' requires that the breaker be opened and locked, nis event was evaluated in NRC Inspection Report 50-443/91 15.

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18 The inspector reviewed LER 91405 and noted that the identified correctise actions appropnately addressed the esent. The inspector reviewed the implementauon of coneeuw acuens and held discussions with Engineenng Depanment personnel. Signs were placed on all TechnicaJ Specification 3.8.4.1 related circuit breakers. Procedure OS1090.05,

" Component Condguration Control,' was revised to re0cet the reduced number of !xked sabes resulting from an engineenng analysis. All locked components idenuded in OstfB0.05 were listed in the tagging computer. The inspector noted that the faalute to control the listing of components in the tagging computer contnbuted to the contamination o!

the deminerahred water systern. This LER is closed.

7.4 Silssing Radiographic Record for a Safety Related AS$1E Meld Weld -

Unresolved item 90 24 02 (Open)

By !ctter (NYN.9115) dated September 17. 1991. New Hampshire Yankee (NHY) submmed its nnal reply to the NRC Nouce of Violation issued in conjunction with NRC Inspecuon Repon (IR) 50-443/91 12. Along with the documentation of the final corTecuse accon taken in response to the violation, NHY provided a completion repon on the Weld Radiograph Reinterpretation Program (WRRIP) instituted as a result of NRC findings idenufied and documented in NRC IR 50L443/9121. NRC inspectors returned to Seabrook Stauon on September 23 24, 1991, as a continuation of IR 9121 inspection activines, to conduct additional radiograph reviews incident to the licensee WRRIP completion.

By letter dated September 24,1991,*the NRC requested additional informauon from NHY regarding the WRRIP conclusions and supporting data. NHY responded by letter (NYN-91157) dated September 27,1991, attaching supplemental contractor (i.e., Hellier) repon information and specific data and explanations to address all remaining NRC questions regarding radiographic record and film adequacy. Subsequently, the inspector was nonned tnat one weld number in Enclosure 3 to NYN 91157 had been incorrect]y documented. The incorrect weld number,1 CS 318 04, F0204, should have been listed, relative to the Table 3 reference, as weld number 1 CS 31842 F0204. The inspector detennined that this typographical error was inconsequential to the finding and conclusions of the licensee WRRIP Completion Repon.

The inspector reviewed the licensee's letters and the WRRIP data and nndings, and evaluated this information relative to the overall results of the programmatic efforts taken to revenfy the Pullman Higgins field weld records. This Weld Record Reverification Program (WRRPt planned and initiated in March 1991 at NRC request, was completed in August 1991 with the Anal NHY status repon (NYN 91134) submitted to the NRC on August 30,1991. The most '

significant WRRP findings were the identification that four weld radiographic record packages could not be located in the NHY records vault. The specific details for each missing record were inspected and are documented in previous inspection reporu addressing this open item, Subsequent corrective action, to include the re radiography of all four weldt

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19 was spot-checked by the inspector including the witness of re radiography in progress, anc the nnal radiographs were independently reviewed by qualined NRC nlm reviewers (reference: 1R 9121).

Other WRRP findings and WRRIP inspecuon items, relaung to reord deficiencies, were documented on NHY Correctise Acuon Requests (CARS) and reviewed by the NRC as ea;n CAR was closed out. The idenoned record denciencies represented issues of minor safety sigm0cance and were deemed appropnate to be processed and corTected by the licensee correctne acuon program. Dunng this inspeedon, the inspector resiewed the last fise CARS dealing with the deficiencies identined relauve to the radiographic records resiew procest These the Correctne Acuon Reports (CARS 91036 thru 91-040) all relate to record discrepancies and not weld quality problems. The inspector spot-checked the dispmtion of each CAR, noting confirmauon of the completion of all corrective measures by the NHY QA Suncillance Supervisor with final acceptance of correctne aedon venficadon by the Nuc! ear Quality Manager. Where calculations were included in validating the acceptability of radiograph quality geometric unsharpness determinadons, the inspector independently check 4 the numencal results to confirm compliance with ASME Code,Section V cntena, The inspector identined no problems with the licensee conduct or documentauon of corrective action for the identified records deficiencies.

NRC oserview of the licensee WRRP and WRRIP activities and resolution of all cortecuse acuons relative thereto is essentially complete. The results of the NRC team inspection record and film quality reviews are daumented in NRC IR 50-443/9121. No addinonal safety concerns or unacceptable findings that had not already been documented by the licensee were idenufied No evidence that inadequate final welds were accepted at Seabrcok Station has been identifbd. NHY has responded to all NRC questions and met all commitments to the NRC regarding the adequacy of Pullman Higgins field weld and related radiograph quality. Therefore, the NRC plans no funher technical review or inspecuon followup regarding this matter at Seabrook. However, NRC enforcement acdon for the identified violations is under consideration. This unresolved item remains open until the NRC enforcement deliberations and final documented actions are wmpleted.

8.0 MEETINGS The scope and findings of the inspection were discussed periodically throughout the inspection pesiod. An oral summary of the uispection findings were provided to the Plant Manager and his staff at the conclusion of the inspection penod.

A meeting was held at the Region 1 office at King of Prussia, Pennsylvania, on September 11, IMl. At the meeting, New Hampshire Yankee presented the results of their review of the failed Cryofit coupling in the pressurizer steam space sample line. The list of Attendees and the slides used during the presentation are provided as Attachment 1.

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29 Region-based. inspectors conducted the following exit meetings during this report penod. ,

l DATE SlMCT REPORT NO. NSPECTRR

. 9 13 -ISl; Eddy Current Testing 91 27 R. McBreany

- 9 24' Welding Records 91 21 . M. Modes

- 9 27 Stanup Testing 91 29' D. Moy -

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10 2- Radioactia Effluent 91 29-- J. Jang

- 10 3 H.P. Controls 91 29 J. Noggle 10-4 Operator Licensing 9130(OL) J. D' Antonio i

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At TACIBIENT 1 NEW HAMPSHIRE YANKEE PRESENTATION OF CRYC M COUPLINGS NRC REGION 10; flCE, KING OF PRUSSIA, PA Sep 'mber 11, 1991 LIST O ' ATTENDEES U.S.NRC R. Barkjey, Project Engineer, RI J. Durr, Chief Engineenng Branch, RI W. Haast, Vendor Inspection Branch, NRR H. Kaplan, Mat's Engineer, RJ W. Lanning, Deputy Director, DRS, R1 J. Linville, Chief, Projects Branch 3, RI A. Lohmeier, Reactor Engineer, RI S. Wookey, Resident inspector, SB NHY ATI'ENDEES B. Beuchel, I&C Engineering Supervisor P. Brooks, Division Manager Metals Division R. Deloach, Executive Director - Engineering & Licensing B. Drawbridge, Executive Director - Nuclear Production T. Harpster, Director Licensing Services G. Kline, Technical Suppon Manager A. Pelton, Technical Manager, Metals Division J. Vargas, Manager of Engineenns .

K. Willens,' YAEC Prmcipal Matenals Engineer i

1

. . . . _ . . _ . . = . _ _ _ . . _ .

ATTACHMENT'1 TO IR 50-4 6/91-29 tiew Harrvstm Yan kee..

CRYOFIT COUPLINGS 1

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i New Hampshire Yankee Presentation to NRC

September 11,1991 i

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Yankee PRESENTATION PARTICIPANTS Bruce Drawbridge - NHY Executive Director, Nuclear Production Jeb Deloach - NHY Executive Director, L~ngineering and Licensing Terry Harpster- NHY Director of Licensing Joe Vargas - NHY Manager of Engineering Gary Kline - NHY Technical Support Manager Bruce Beuchel- NHY I&C Engineering Supervisor Ken Willens - YAEC Mechanical Engineering Group Alan Pelton - Raychem Corporation Peter Brooks - Raychem Corporation .

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r4ew661w Yankee PURPOSE AND OVERVIEW

• Jeb DeLoach, Executive Director of Engineering .uul I.icensing

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Yankee PURPOSE Show understanding of the cause of failure

- Selective replacement appropriate and conservative Remaining couplings suitable for the application d

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Raychem Corporation tradename Joins tubing in sample and instrument lines Alloy called TINEL Shape memory effect

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. OVERVIEW Failure mechanism Corrective action Conclusion O

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Bruce Beuchel- NHY I&C Engineering Supervisor i

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m w- n..w Yankee APPLICATION Sample and instrument sensing lines 1/4", 3/8",1/2"

Safety-related and balance-of-plant Over 3000 originally installed O

ne~ ,6o, Yankee I NITIAL QUALIFICATION Raychem Qualification Package Raychem Report EDR-5116

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Raychem Qualification Package l -

ANSI B31.1 up to 2" ASME 111 up to 1" Mechanical Performance Burst Pressure Tensile Strength Fatigue, vibration, and shock Elevated temperature U.S. Navy approval Vendor-identified test programs ,

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..a Yankee Raychem Report EDR-5116 Temperature rating upgraded for pressurizer Elevated temperature tests i

Long term temperature soak Thermal cycling i

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Yankee INITIAL PROCUREMENT UE&C specification for instrument piping UE&C specification for cryogenic couplings

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- Not fraudulent material or bad lots l.

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Yankee NHY ACTIONS

- Immediate Actions

- Literature Search

- Review Third Party Testing NHY Program

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fJew HortFJw, Yankee Immediate Actions Root Cause Analysis i 1

1 Test samples from pressurizer gas space sample line Form hypothesis ;I Replace selected couplings Perform system walkdowns Develop selection criteria for additional tests Informed rest of nuclear industry

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New Ikurd Yankee Review Third Party Testing 4

1 B&W PWR primary and secondary (AVT) chemistry U-bend specimens Applicable to Seabrook Framatome PWR primary chemistry Different alloy of TINEL Confirmed hypothesis

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. NHY Program 4

i Cryofit Locations Sample Selection Process Test Specifications LOCA Test Replacement Criteria Cryofit Replacement Summary Long Term Corrosion Study

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. Cryofit Locations Main Steam and Feedwater ~1400 Sampling (primary, secondary, other) ~ 1000 Reactor Coolant ~300 Chemical and Volume Contro: ~150 Blowdown ~ 125 Hydrogen Gas - 50 l Others -250 1 Stores ~ 215

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. Sample Selection Process Both RCS and AVT chemistry Various process temperatures Various hydrogen concentrations Flow conditions (flowing / stagnant)

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- Macroscopic visual inspection

- Functional tests Leak test j Burst resi Metallography Hydrogen pickup

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Yankee Replacement Criteria Degraded couplings Couplings in lines with RCS chemistry which are:

RCS pressure boundary and not remotely isolalile Required for RPS / ESFAS Required for ECCS operability

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LOCA Test 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> w/ 4% Hydrogen

- Vir' gin Cryofit couplings 4

- Cryofit couplings removed from plant

- Results No significant hydrogen absorption

- No change in mechanical properties

- No cracking 4

Yankee Cryofit Replacement Summary RC Sample Lines in Containment Pzr gas space ~ 50 Pzr liquid space ~ 30 RCS loop 1 hot leg ~ 12 RCS loop 3 hot leg - 25 Instrument Lines Pzr level / pressure ~ 70 RCS flow ~150 RTD bypass flow - 20 Reactor vessel level 5 RCP #1 seal Ap ~ 25

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PJew ilor g>9me Yankee Long Term Corrosion Study Confirm 40 year life of Cryofit couplings Periodic sampling during outages 4

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Yankee CONCLUSIONS Rool cause evaluation indicates design application issue Implemented a thorough and effective program to resolve Remaining couplings can perform their function Periodic sampling program will ensure validity of test results-J I

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