IR 05000213/1989005

Safety Insp Rept 50-213/89-05 on 890329-0502.Violation Noted.Major Areas Inspected:Plant Operations,Radiation Protection,Fire Protection,Security,Maint,Surveillance Testing & Licensee Events

ML20247C709
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	05/17/1989
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20247C688	List: IR 05000213/1989005 ML20247C681 ML20247C691
References
50-213-89-05, 50-213-89-5, NUDOCS 8905250025
Download: ML20247C709 (18)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /89-05 Docket N License N DPR-61

' Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06101 Facility: .Haddam Neck Plant, Haddam Neck, Connecticut Inspection'at: Haddam Neck P1 ant

. Inspection dates: March 29 through May 2, 1989 Inspectors: Andra A. Asars, Resident Inspector John T. Shediosky, Senior Resident Inspector Approved by: 9{ldE b 7//e 17, /D Ddte '

E. C. McCabe,gef, Reactbr Projects Section IB Summary: Inspection 50-213/89-05 (3/29/89 - 5/2/89)

Areas Inspected: This was a routine safety inspection by the resident inspec-tors. Areas reviewed included plant operations, radiation protection, fire protection, security, maintenance, surveillance testing, licensee events, con-trol of containment penetration boundary valves, initiation of a shutdown due to inoperable containment isolation valves, errors identified in the Post-LOCA Containment Spray and Large Break LOCA Analyses, Justification for Continued Operation with certain Westinghouse Steam Generator Tube Plugs installed, and report of a Substantial Safety Hazard concerning ASEA Brown Boveri Circuit Breaker Results: One Unresolved Item was closed. One Violation of Technical Specif1-cations resulted from licensee identification of an uncontrolled containment penetration boundary valv PDR ADOCK 05000213 o PDC

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TABLE OF CONTENTS 2 i

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PAGE t S umma ry o f Fa c i l i ty Ac t i v i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Plant Operations..................................................... 1 Plant Operations Review Committe . ..... ... .................... 2 Maintenance and Surveillance Testing................................. 2 4.1 Containment Sump Suction Valve Failure......... .. ....... ..... 2 4.2 Schedular Exemption Request for Containment Leakage Testing..... 2 Uncontrolled Containment Penetration Boundary Valve....... .......... 4 Background.................... ................................. 4 5.2 Corrective Actions........ ..... ............ .................. 4 5.3 Contributing Circumstances..... ...... ....................... 5 5.4 Previous Similar Occurrences............ ....................... 6 5.5 Conclusions... .. .. ..... .. . .... ......... ... . ... . ... 6 Events Occurring During the Inspection.. ..... .... .... ............ 7 Licensee Event Reports and Safeguards Event Reports......... ... 7 6.2 Initiation of Shutdown Due to Inoperable Containment Isolation Valves................. ....... .............................. 7 Review of Periodic and Special Reports. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 Inadequacy in Post-Accident Containment Spray Design Analysis........ 9 Defective Westinghouse Steam Generator Tube Plugs... ................ 10 10. Di scovery of Error in the Large Brea k LOCA Analysi s. . . . . . . . . . . . . . . . . . 12 10.1 Background............................ ............ ............ 13 10.2 Justification for Continued Reactor Power Operation. . . . . . . . . . . . . 14 10.3 Conclusions........ .. ....... ... ............................ 15 1 Report of a Substantial Safety Hazard (10 CFR Part 21) - ASEA Brown Bove ri C i rc u i t B rea ke rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 15 1 Exit Interview.... .... .. .. .... . . .. ....... ............. 16 i

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DETAILS l Summary of Facility Activities (71707) l At the beginning of this inspection on March 29, the plant was at 70%

power during power escalation after Containment Air Recirculation Heat Exchanger cleaning at reduced power. The power escalation was interrupted and power was reduced to 30% on March 29 for repair of a Main Feedwater pum Full power operation then resumed on April 9. On April 14, an Unusual Event was declared and a shutdown was initiated due to two in-operable containment isolation valves associated with the Heating Steam System. One valve was repaired within one hour and the plant was returned l to full powe Full power oper:'tions continue for the rest of the in- l spection perio . Plant Operations (71707, 71710)  !

l The inspectors observed plant operation during regular tours of the fol-lowing plant areas:

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Control Room --

Primary Access Point

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Primary Auxiliary Building --

Fence Line (Protected Area)

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Vital Switchgear Room --

Yard Areas l

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Diesel Generator Rooms --

Turbine Building l

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Radiological Control Point --

Intake Structure  !

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Appendix R Switchgear Building --

Auxiliary Feedwater Pump Room l

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Control room instruments were observea for correlation between channels l and for conformance with Technical Specification requirements. The in-spectors observed various alarm conditions which had been received and j acknowledged. Operator awareness and response to these conditions were i reviewed. Control room and shift manning were compared to regulatory re-quirements. Posting and control of radiation and high radiation areas were inspected. Compliance with Radiation Work Permits and use of appro- j priate personnel monitoring devices was checked. Plant housekeeping con- '

trols were observed, including control and storage of flammable material and other potential safety hazards. The inspectors also examined the con-dition of various fire protection system During plant tours, logs and records were reviewed to determine if er.tries were properly made and com-municated equipment status / deficiencies. These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, and Plant Information Reports. The inspectors observed selected aspects of plant security including access control, physical barriers, and person-nel monitorin In addition to normal utility working hours (7:00 a.m. to 3:30 p.m.),

plant operation was reviewed during portions of weekend, midnight and evening shifts. Such coverage was provided on the following days:

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April 13, 1989 through 7:30 April 22, 1989 6:00 a.m. through 5:00 No unacceptable conditions were identified. Operators were alert and dis-played no signs of inattention to duty or fatigu . Plant Operations Review Committee (40500)

The inspectors.attendcd several Plant Operations Review Committee (PORC)

meetings. Technical Specification 6.5 requirements for member attendance were verified. The meeting agenda included procedural changes, and pro-posed changes to the Technical Specifications and Plant Design Change Records. The meetings were characterized by frank discussions and ques-tioning of the proposed changes. In particular, consideration was given to assure clarity and consistency among procedures. Items for which ade-quate review time was not available were postponed to allcw committee mem-bers ample time to review and comment. Dissenting opinions were encour-aged and resolved to the satisfaction of the committee. The inspectors had no further comment . Maintenance and Surveillance Testing (61726, 62703)

The inspectors observed various maintenance and problem investigation ac-tivities for compliance with requirements and applicable codes and stand-ards, QSD (Quality Services Department) involvement, safety tags, equip-ment alignment and use of jumpers, personnel qualifications, radiological controls, fire protection, retest, and deportability. Also, the inspec-tors witnessed selected surveillance tests to determine whether properly approved procedures were in use, test instrumentation was properly cali-brated and used, technical specifications were satisfied, testing was per-formed by qualified personnel, procedure details were adequate, and test results satisfied acceptance criteria or were properly dispositione Portions of the following activities were reviewe SUR 5.1-13, Auxiliary Feedwater Pump Monthly Functional Tes SUR 5.2-96, Functional Verification of Automatic Initiation of Auxiliary Feedwater Solenoid Valve "B" Main Feedwater Pump shaft seal replacemen .1 Containment Sump suction Valve Failure On April 6, during performance of SUR 5.1-4, Core Cooling Systems Het Operational Test, the RHR (Residual Heat Removal) pump containment sump suction valve (RH-MOV-22) failed to operate electrically from the control room. The valve was declared inoperable at 7:50 a.m. An administratively controlled 72-hour action statement was entered i

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based on the inoperability of one train of RHR for Post-LOCA long term cooling. Sump suction bypass valve RH-V-808A was unlocked and >

cycled to verify an operable flow path to the "B" RHR pum j i

After the valve failed, operators manually opened it off the backseat I and cycled the breaker. The valve still failed to operat Investi-

.gation into the cause of the valve failure was conducted under AWD 1 (Automated Work Order) 89-2887. A torque switch set screw was found i loose, and the thermal overload was tripped. This locked out the  :

valve motor-operator. The torque switch set screw was tightened, the l torque switch was reset, and the valve was successfully stroke teste j i

Plant Design Change Record 955, Installation of Emergency Automatic Closure Circuitry for the Turbine Building Se"vice Water Header Iso-lation Valves, SW-MOV-1 and SW-MOV-2, had grovided automatic closure 1 for two motor-operated valves with a signal from the same relay which j supplies a close signal to RH-MOV-22. The post modification test, ST  ;

11.7-4, Preoperational Test of SW-MOV-1.and SW-MOV-2, involved manual ,

operation of the relay to simulate a safety injection signal. The '

licensee determined that operation of this relay during ST 11.7-4 gave a closed signal to the normally closed valve, causing the ther-mal overload trip to lock out the motor-operato This caused the

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valve to fail to operate electrically during the monthly surveil-lance, SUR 5.1- The inspectors noted that, had the torque switch set screw not been loose, the thermal overload trip would not have tripped and the valve )

, would have operated as required. Also, the plant design does not 1 I provide indication that a motor control center thermal overload has I tripped. In this case, that was determined through an investigation I conducted by work orde Review of the AWO and associated procedures were performed. The in-spectors also reviewed the evaluation of equipment potentially af-fected by ST 11.7-4 prior to the test. No deficiencies were iden-tdfie .2 Schedular Exemption Request for Containment Leakage Testing The licensee has requested, by letter dated April 26, 1989, a schedu-lar exemption to provide temporary relief from the requirements of 10 j CFR 50, Appendix J for Type A, B, and C tests of containment leakag ;

These tests were performed during the 1987 Refueling Outage as re- i'

quired. However, due to the extension of that outage for unantici-pated repairs to the Core Support Barrel Thermal Shield, the testing intervals expire prior to the upcoming refueling outage, which is scheduled to begin on September 9, 198 l l-l _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ -

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The Containment Integrated Leak Rate Test (Type A) interval expired on April 27, 1989 and the Local Leak Rate Tests (Type B and C) will expire shortly before the refueling outage. In most cases, perfor-mance of these tests require that the plant be shutdown and system isolated and vented. Pending resolution of the schedular relief re-quest, this item is unresolved (UNR 89-05-02). Uncontrolled Containment Penetration Boundary Valve (71707)

(Closed) Unresolved Item (89-02-01): On March 27, 1989, the licensee iden-tified a containment penetration boundary valve (PW-V-600) in the Primary Water System (PW) which was not being controlled in accordance with sta-tion procedures. The identification of this apparent error and immediate corrective actions were discussed in NRC Inspection Report 50-213/89-02, Detail .1 Background On March 27, during a walkdown inspection of Containment Penetration P-68 in support of local leak rate surveillance procedure review, PW-V-600 was identified as mislabelled and not locked closed as re-quired by SUR 5.1-126, Locked Valve Checklist. PW-V-600 was in fact open, labelled PW-V-601, and was being used as the normal source of primary demineralized water for the chemistry sample sin PW-V-600 is a manual isolation valve for the PW System flush to the chemistry sample sink. This flushing line ties into the containment penetration between the outboard isolation check valve and the con-tainment wall and is therefore described as a containment penetration boundary valve. Position and operational restrictions for this valve are not specified in the Technical Specification The valve is a one-half inch ball valve located in the Primary Auxiliary Building pipe trench, Chemistry personnel routinely use this portion of the piping to supply demineralized flushing water to the sample sin Two sample sink isolation valves downstream of PW-V-600 are normally operated to supply the flow to the sample sin The PW system is normally operated at about 150 psig and does not service containment during power operatio PW-V-601 is a drain valve for the PW supply line to the chemistry sampia sink. This drain is located downstream of PW-V-600. PW-V-601 was mislabelled, locked closed, t~f tagged as containment isolation valve PW-V-60 .2 Corrective Actions PW-V-600 was immediately placed in the locked closed position and correctly labelled as a containment isolation valv Procedures as-sociated with this penetration were also reviewed and revised as necessary to correctly specify the function and position required of I

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this valve. Additionally, the licensee conducted an administrative review of all containment boundaries and developed individual pene-tration drawings that illustrate all valves and components which com-prise the containment boundary. Utilizing these drawings, operators walked down all accessible containment isolation valves to verify correct valve position This review was completed on April 21. All valves reviewed were found to be in the correct position. Several penetration boundary valves were found without the orange Containment Isolation Valve tags and were subsequently labelle Some discrepancies were noted be-  ;

tween the P& ids (Piping and Instrument Diagrams) and as-built arrange- i ment of instrument valves. These were noted and forwarded to Engi-neering for review and resolutio Proposed long term corrective actions include evaluation of the  !

existing system for control of containment isolation valves and re-view ~of isolation and boundary valve labelling practice A root cause evaluation was also performed. The licensee's evalu-ation of the circumstances which resulted in the incorrect position-ing of this valve and corrective actions are documented in Licensee Event Report 50-213/89-03, dated April 24, 198 '

5.3 Contributing Circumstances The inspectors reviewed the associated Chemistry, Operations, and  !

Engineering procedures, P& ids, and the root cause evaluation. This matter was also discussed with plant personnel involved in manipu-  ;

lation and control of this penetratio The following circumstances '

were determined to have contributed to the lack of adequate control over this containment penetration boundary valv )

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Inconsistent P&ID revisions interchanged valve identification  !

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Valve labelling errors also interchanged valve identification numbers,

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Procedural deficiencies in which valve identification numbers and/or functional descriptions varied between disciplines, j

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Plant design in which a portion of a containment boundary is j used as the normal lineup for supply of primary demineralized  !

water to the chemistry sample sink, and

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Technical Specifications (TSs) which do not control PW-V-600 and i many other containment penetration isolation and boundary i valve j

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These factors together contributed to the mispositioning of PW-V-600 and,. subsequently, to plant operation with an uncontrolled contain-ment boundary valv .4 Previous Similar Occurrences Previously, the licensee had been cited for violations of TS 3.11, Containment, on three occasion These three violations involved manipulation of designated, locked closed, containment isolation valves during surveillance testing and normal operational evolutions when Containment Integrity was required. These violations are docu-mented in NRC Inspection Reports 50-213/86-08 issued May 22, 1986, 50-213/86-20 issued August 29, 1986, and 50-213/86-27 issued November 26, 1986. Connecticut Yankee Atomic Power Company responded to these violations by letters dated June 25, 1986, October 14, 1986, October 24, 1986, and December 24, 1986, respectivel Corrective actions for these violations included:

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TS changes were made to TS 1.8, Containment Integrity, to permit operation of several containment isolation valves with the pro-vision that an operator in communication with the control room is dedicated to close the valve within one minute in the event that a containment isolation is required,

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In 1986, a detailed review of all non-automatic containment isolation valves and associated procedures was conducted to identify and evaluate additional isolation valves which are opened periodically during operations; procedure changes were made as necessary,

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A Containment Isolation Valve tagging program was initiate These valves were labelled with a large orange tag stating that the valve is a containment isolation valve and can be operated only by order of the Shift Superviso .5 Conclusions The operatien without adequate control of PW-V-600 was identified by the licensee as part of a local leak rate surveillance procedure re-view. Prompt and thorough corrective actions were taken including the assembly of detailed penetration drawings and walkdowns of all accessible penetrations to verify containment integrit This event could have been prevented had corrective actions for the previous three violations been sufficiently comprehensive. Operation with a containment penetration boundary valve (PW-V-600) open vio-lates TS 3.11, Containment, which requires that containment integrity be maintained when the reactor coolant system is above 300 psig and

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200 F (89-05-01). Additionally, through interviews and documentation reviews, the inspectors concluded that PW-V-600 had been open for long periods of time during plant operatio Over the past two years, the inspectors have found significant im-provements in the plant staffs' awareness of the importance of main-taining positive control over the containment boundary. This has 1 been particularly evident during response to events involving con-tainment penetration boundaries, revisions t- the Inservice _Inspec-tion and Testing Programs, procedure upgrad- and, improvements in plant labellin . Events Occurring During the Inspection (71707, 92700, 93702)

6.1 Licensee Event Reports and Safeguards Event Reports The following Licensee Event Reports (LERs) and Safeguards Event Re-ports (SERs) were reviewed for clarity, accuracy of the description, root cause determination, and adequacy of corrective action. The inspectors determined whether further information was required and whether there were were generic implications. The inspectors also verified that the reporting requirements of 10 CFR 50.73, 10 CFR 73.71, and Station Administrative and Operating, and Security Proce-dures had been met, that appropriate corrective action had been taken, and that the continued operation of the facility was conducted within Technical Specification Limit S02 Safeguards Event Repor Design Deficiency in Auxiliary Feedwater Syste Manual Containment Isolation Valve Found Open.

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j 89-04 Inadequate Flood Protection of Main Diesel Fuel Oil Tan No unacceptable conditions were identifie .2 Initiation of Shutdown due to Inoperable Containment Isolation Valves On March 14 at 8:05 a.m., with the plant at 100% power, the licensee declared the two heating : team containment isolation valves (HS-TV-380 and 381) inoperable due to their failure to operate during sur-veillance testing. TS (Technical Specification) 3.11, Containment, does not specify a:tions to be taken with two containment penetration valves inoperable. The penetration was immediately isolated by closure of manual isolation valves HS-V-382 and 385. A shutdown was started at 9:03 a.m. in accordance with administrative controls. At 9:05 a.m. one of the isolation valves was declared operable. At this time, the compensatory actions specified by TS 3.11.G with one con-tainment isolation valve inoperable were taken. The power reduction l

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was discontinued at 9:11 a.m. after decreasing about 10 MWe, and the plant was returned to full power. The appropriate notifications were made to the NRC and the State of Connecticu ,

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The affected containment penetration (P-30) had been extensively'al-tered by PDCR (Plant Design Change Record) 878, Appendix A and J Penetration Modifications, during the 1987-88 Refueling Outag Changes to this penetration included rerouting of piping, removal of check valve internals, and installation of the two .'--operated iso-lation valves. HS-TV-380 and 381 are two-inch, Con .matic, air-operated ball valves with teflon coated packing and seating surface They are located in the pipe trench between the Containment and Pri-mary Auxiliary Buildings. During normal operation, these valves are open and supply a continuous Heating Steam System steam flow to the containment atmospher Until recently, these and other containment isolation valves were not stroke tested during power operation. They were scheduled for quar-terly testing during plant shutdown only. Recently, Inservice In-spection and Testing personnel re-evaluated the ability to test con-tainment isolation valves during plant operation. The impact on

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plant operation was determined to be minimal for stroke testing of HS-TV-380 and 381. The April 14 surveillance test, SUR 5.7-149, In-service Testing of Containment Heating Steam Valves, HS-TV-380 and 381, was the first stroke test of these valves with the plant at powe Both isolation valves had been tested quarterly during plant shut-downs in accordance with SPL 10.7-290, Functional and Integrity Test-ing of Penetration 7, 10, 30, 38, 41 and 60 Isolation Valves' Piping Modifications. The last stroke test had been performed on May 18, 198 Between that stroke test and the April 14, 1989 surveillance, the valves were maintained in the normally open position and were subject to steam flow of about 15 psig and 250 On April 14, during SUR 5.7-149, the valves failed to close when operated from the control room. Operators were stationed locally and observed that the valves appeared to be sticking. Both valves were exercised and successfully tested on April 1 Currently, both valves are in the closed position and their control room handswitches are caution-tagged. The air has been removed from HS-TV-380. Also, the manual isolation valves HS-V-382 and 385 are closed. The licensee has elected to maintain a closed penetration until the cause of the valve failure can be determine _ _ _ - _ - _ _ _ - _ - _ - ___ -__ ___-_-_ - _ _

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Engineering reviews of the isolation valve design verified that these valves are appropriate for normal Heating Steam System condition This failure has also been discussed with the valve vendor. No con-clusions have been reached. Evaluation is continuing. The licensee noted that the containment isolation valves for P-60, Component Cool-ing Water to the Neutron Shield Tank Cooler, are identical to the failed valves. These were tested to verify no generic concerns with this type of valve; the test results were satisfactor The inspectors observed control room activities following the initial surveillance testing including the power reduction, valve trouble-shooting, and notifications. The associated surveillance testing and the ongoing failure evaluation were discussed with Engineering per-sonnel. No deficiencies were identifie . Review of Periodic and Special Reports (90712)

Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewed. This review verified that the reported information was valid and included the NRC required data; that test re-sults and supporting information were consistent with design predictions and performance specifications; and that planned corrective actions were adequate for resolution of the problem. The inspectors also ascertained whether any reported information should be classified as an abnormal oc-currence. The following reports were reviewed:

Haddam Neck Plant Fire Suppression Systems Inoperable Diesel Fire Pump Special Report, dated April 6, 1989 Haddam Neck Plant Monthly Operating Report No, 89-03, for the period March 1, 1989 through March 31, 1989 No discrepancies were identifie . Inadequacy in Post-Accident Containment Spray Design Analysis During February 1989, the NRC conducted a Procurement and Vendor Interface Inspection of Haddam Neck (NRC Inspection Report 50-213/89-200). This inspection focused partly on licensee evaluation and action in response to vendor supplied information concerning the facility and equipment. Evalu-ations and actions (as appropriate) for several Westinghouse Technical Bulletins were reviewed. Further evaluation was deemed appropriate for '

Technical Bulletin NSID-TB-86-08, Post-LOCA Long Term Cooling Boron Re-quirements, dated December 1, 1986. This bulletin discussed the need for sufficient containment sump boron concentration to assure that the reactor core remains subcritical during the Post-LOCA long term recirculation cooling phas .

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This review of the Technical Bulletin noted that the Containment Spray System (CS) is a source of unborated water during the long term recircu-lation phase. CS is an auxiliary system used for the reduction of con-tainment pressure by spraying either water from the Refueling Water Stor-age Tank via the LPSI (Low Pressure Safety Injection) Pumps or the Fire Suppression System. This system is not required for the reduction of Post-Accident (Steam Line Break, LOCA, and Control Rod Ejection) contain-ment pressures and is not taken credit for in the Accident Analysis. The Containment Air Recirculation System is designed to maintain containment pressures below design pressure (39.6 psig). CS is intended for use only 4 after the containment has exceeded its design pressur An uncertainty in the wide range containment pressure transmitter indica-tion coulo cause operators to initiate containment spray prematurel There are two qualified, wide range containment pressure transmitters with a range of 5.0 to 120.0 psi Their maximum uncertainty was calculated to be 7.58% or 8.4 ps Emergency Operating Procedures (EOPs) require that, when containment pressure exceeds 40 psig, initiate spray and ter-minate that spray when pressure decreases below 35 psi '

Additionally, the licensee identified that, if LPSI is used for spraying containment, a large portion of the flow to the reat'or core would be diverted for spray. This could result in insufficient core cooling in this scenario. A detailed evaluation of the affects of reduced LPSI flow to the reactor core is in progres The licensee concluded that the use of CS in these two scenarios might prevent the fulfillment of the safety functions needed to maintain the reactor in a safe shutdown condition and mitigate the consequences of a accident. This was reported to the NRC on the Emergency Notification System in accordance with 10 CFR 50.72(b)(2)(iii) on April 1 The licensee elected to change the E0Ps to provide for a delayed initi-ation and termination of containment spray to account for the calculated instrument uncertainties. Containment Spray will only be initiated if the containment design basis pressure (37.3 psig) has been exceeded. The spray initiation pressure was increased to 50 psig. The spray termination pressure was increased to 40 psig if spraying with LPSI and 48 psig if spraying with the Fire Suppression System. Preshift training of operators was in the form of informal discussions and Read-and-Sign documentation describing the procedure changes. The inspectors observed portions of this training and reviewed the associated reading materia No deficien-cies were identifie . Defective Westinghouse Steam Generator Tube Plugs Some plants have experienced primary water stress corrosion cracking leaks in Westinghouse mechanical steam generator tube plugs. Recently, plants have experienced mechanical tube plug failures following plant I

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transients. This phenomenon was discussed in NRC Information Notice 89-33, Potential Failure of Westinghouse Steam Generator Tube Mechanical Plug Westinghouse Electric Co. review of the February 1989 North Anna Unit I tube rupture event identified that primary water stress corrosion cracking of the tube plug led to a circumferential fracture which allowed the top of the plug to be released during a plant transient. Reactor Coolant Sys-tem-pressure provided the motive force to project the plug piece through the U-bend portion of the steam generator tub Steam Generator tube plugs are manufactured from Inconel 600. Westing-house informed licensees that three heats of the Inconel 600 material (NX-3513,-3279,-3962) are susceptible to this type of failure. An algorithm was developed to predict the time necessary for plug cracks to cause an event similar to that at North Anna Unit 1. Elements factored into the algorithm were temperature and (naterial microstructure. Cold leg-tube plugs were identified as being much less susceptible to this phenome-no The licensee reviewed their records for steam generator tube plug instal-lation. The only susceptible plugs installed were from the 3513 heat and were installed during the 1984, 1986 and 1987 outages. Westinghouse cal-culations and confirmatory calculations by the licensee indicate that Haddam Neck has 53 hot leg tube plugs which could, during the current cycle, have similar cracking to that which resulted in the Steam Generator Tube Rupture at North Anna Unit 1. These tubes were installed during the 1984 Refueling Outage and are in the No. 2, 3 and 4 steam generator On March 21, the licensee determined that, if the cracked steam generator tube plugs suddenly break or cause tube ruptures, they can no longer meet the original design basi This was evaluated to be reportable because it could constitute a serious degradation of a principle safety barrier or a condition which alone could prevent the fulfillment of the safety function of a structure needed to control the release of radioactive material. The appropriate reports were made to NRC in accordance with 10 CFR 50.7 A. Justification for Continued Operation (JCO) was also prepared for plant operation until the upcoming refueling outage in September 1989. The JC0 concluded that plant operation is justified with the 53 susceptible plugs installed for the following reasons:

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It is unlikely that a broken tube plug top would be released during a plant transient because it would be captured in the unexpanded por-tion of the steam generator tube shee The majority of the cracked plugs are expected to exhibit leakage prior to release of the plug. This leakage would allow differential pressure equalization and therefore limit the energy with which the

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plug is released. Additionally, Haddam Neck has the ability to de-tect primary-to-secondary leakages as low as 2.5 gallons per day be-cause stainless steel fuel clad provides highly sensitive leak de-tection capabilitie If a single, high energy tube plug top was released, it could result in a tube rupture. However, the tube plug top would not have suffi-cient energy to penetrate any adjacent tube If a broken tube plug top were released into the steam generator tube and. punctured the tube, the primary-to-secondary leak rate would be limited by the flow constriction through the tube plug expander which remained in the tube sheet. Eleven tube plug tops would have to be released in this manner to exceed the analyzed primary-to-secondary leak rate for Haddam Neck. The probability of this happening con-currently is extremely lo Based on their evaluation, the licensee concluded that continued operation is justified because the postulated incidents are within the design basis of Haddam Nec The licensee is currently evaluating inspection and repair evolutions for the upcoming refueling outage. These efforts will be reviewed during future inspection . Discovery of Error in the Large Break LOCA Analysis The licensee discovered an error within the input data which was used to develop the large diameter pipe break LOCA (Loss of Coolant Accident)

analysis which first supported operation in 1986. An incorrect value was used for the reactor vessel volume below the active fuel. The error oc-curred in 1985, during data transfer between the licensee and Westing-house. The Atomic Energy Commission IAC (Interim Acceptance Criteria)

model required that the vessel bottom be reflooded before crediting ECCS (Emergency Core Cooling System) flow to core cooling. A larger vessel bottom volume therefore results in a greater increase in peak clad tem-perature. The report of the error was made to the NRC in accordance with 10 CFR 50.72(b)(1)(ii)(B) on April 2 Reanalysis concluded that a change in the reactor core axial offset limits and reductions in the linear heat generation rate limits were required to maintain fuel Peak Clad Temperature (PCT) below the 2300 F limit of the AEC Interim Core Cooling Criteria.

l This error was discovered during qualification testing of a Large Break LOCA model written to meet 10 CFR 50.46 and Part 50, Appendix This analysis was undertaken to support future conversion to Zircaloy clad fue . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ ,

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10.1 Background This calculational error appears to have been introduced in a program intended to improve confidence in the analysis. Although the West-inghouse IAC model has been used for LOCA transient analysis since 1972, the licensee supplied input data which was traceable to con-trolled engineering calculations to support Operating Cycle 14, This was done because much of the basis for the data in use at that time had been los The licensee took advantage of a quality controlled data base devel-oped for a new analysis to meet 10 CFR 50, Appendix This was undertaken to support future conversion from Stainless to Zircaloy clad fuel. Some of this data, which had been qualified for use in non-LOCA transient analysis, was supplied to Westinghouse by memo dated September 10, 1985 written in response to a request dated June 27, 1985. Reactor Vessel volume distribution and elevation inforea-tion was among that neede The breakdown of the volume fractions needed were not adequately de-scribed in the above referenced communications. Although the IAC model used the total reactor vessel volume below the active core for its calculation of reflood time, this volume was not needed by non-LOCA transient analysis program for which the data was develope The data supplied from the non-LOCA analysis had separate listings for both the reactor vessel lower plenum volume and the vessel down-comer area (the annular region outside of the core support barrel above the lower core support plate) volum The lower plenum volume alone was used in the large break LOCA cal-culations. Missing was a fraction of the downcomer volume from the annular downcomer region between the bottom of the active fuel and the core support plate. This would have added 85.9 cu.ft. (cubic feet) to the volume below the core. The effect of undersizing the volume for the most limiting ECCS flow delivery rate was to decrease the reflood time by 6.08 second The design basis analysis resulted in a peak clad temperature of 2296 F, and the additional time re-quired to refill the volume below the core was recognized as enough to allow the fuel cladding to exceed the AEC IAC limit of 2300 The inspector examined the Westinghouse request (Serial CYW-85-562)

and portions of the licensee's response dated September 10, 198 Although it was not clear that one of the parameters needed was the reactor vessel volume below the active fuel, the volumes supplied to Westinghouse for downcomer and lower plenum volume were not appropri-ate for what was needed by the IAC model. The response to the re-quest presented volumes and elevations used in the licensee's non-LOCA transient analysis model, RETRAN. It presented data and draw-ings which described the downcomer region as extending to the bottom of the core support barre Therefore the lower plenum volume, also

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described, did not include all of the needed volume. This was pre-sented both schematically in drawings and in lists of elevations and heights. The inspector also noted that a Westinghouse letter (Serial CYW-86-504) dated January 8,1986 supplied an addendum to an earlier report and documented that licensee supplied data for the lower plenum was 780 cu.ft., a -8.4's change from that of 851 cu.ft, used in 1982. That letter did not trigger questioning within the licensee or Westinghouse organization as to the change noted in reactor vessel volume. The later calculations and their application were assumed to be correct because of the engineering assurance of their origin. The inspector also noted that Section 15.3.1.1 of the Updated Final Safety Analysis, Large Break LOCA Analysis Development, addresses the recalculation of volume distributio .2 Justification for Continued Reactor Power Operation The licensee provided revised downcomer and lower plenum volumes to Westinghouse on April 17 for performance of blowdown, refill and fuel rod heatup analysis. The results show that the redistribution of the reactor vessel volume fractions yields a slight improvement in fuel rod temperature (5 F) at the end of the blowdown phase However, the adiabatic heatup period during refill was extended due to the change in volume below the fuel. The IAC analysis concluded that, in order to maintain fuel PCT below the limit of 2300 F, the Linear Heat Gene-ration Rate (LHGR) needed to be reduced from the Technical Specifi-cation (TS) 3.17.2.1 limit of 14.6 kilowatts per foot (kw/ft) to a more limiting value of 13.7 kw/ft for the period of core exposure from 250 Effective Full Power Days (EFDP) to the end of cycl This also required a revision to the core power distribution TS for axial offset; the negative value at full power was reduced from -15 to -1 Based on the immediate implementation of these new limits, the Plant Operations Review Committee approved a Justification for Continued Operation (JCO) on April 25. These new limits were implemented through administrative procedure changes which included Administra-tive and Surveillance Procedure Changes. At the end of this inspec-tion period, the licensee dic; not intend to submit a request to amend the Operating License Technical Specification This JC0 was supplied to the NRC Office of Nuclear Reactor Regula-tio The licensee had proposed changes to the TS on April 14, 1989 to sup-port reactor operation beyond the end of core life, that is, an operating condition in which all control rods are fully withdrawn, essentially zero boron concentration and average RCS (Reactor Coolant System) temperature below normal. The inspector was informed on May 2 that the power coastdown analysis which was the basis for this re-quest remained valid following reanalysis with the correct reactor vessel volume fraction _ _____________ a

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10.3 Cor.lusions Although the licensee discovered this long standing problem, it ap-peared to the inspectors that the error stemmed from poor communi-cations between the licensee and Westinghouse. The data needed was not adequately defined, and the data supplied was not adequately checked for applicability. Also, there was insufficient review of the analysis result The changed vessel volume fraction was noted to have yielded more favorable results without initiating a second check of why this design value changed by 8.4% from the original valu The inspectors found that the updated limits were promptly implement-ed following the April 25 safety committee meeting. Additionally, the PORC noted that results of the most recent core power distribu-tion surveillance indicated that the current LHGR is about 9.5 kw/f Values displayed for axial offset were not limitin . Report of a Substantial Safety Hazard (10 CFR Part 21) - ASEA Brown Boveri Circuit Breakers On April 3, verbal notifications were made to the NRC that defects found in new electrical circuit breakers were determined to be a Substantial Safety Hazard as defined by 10 CFR Part 21. A written report concerning this finding was dated April 4, 198 The circuit breakers were 480 volt load center power circuit breakers, Model Numbers K-1600 and K-3000, manufactured by ASEA Brown Boveri, In Although there was no immediate safety concern, since the equipment had not yet been placed in service, several of the defects, if left uncorrect-ed, could have resulted in failures that could disable safety-related equipment when required to operate. The licensee determined this to be a major degradation of safety-related equipmen The circuit breakers were procured for a new electrical switchgear instal-lation which will provide the separate power distribution required by 10 CFR Part 50, Appendix R. The licensee had kept the Resident Inspectors appraised of their findings which were gathered during detailed preinstal-lation inspection and testing. The description of the eleven (11) defects listed in the April 4 report were consistent with the information provided to the inspector The list is a compilation of the defects found o eleven (11) of the thirteen (13) new breakers. One of these was of a 3000 amp. rating (Model K-3000); the others were of a 1600 amp. rating (Model K-1600). The defects included the following: An unusual number of loose screws and bolt Several components were found loose in the boxe _ - _ _ _ _ _ - _ _ _ _ _ -

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16 Five (5) breakers failed to meet the manufacturer's acceptance cri-teria for tripper bar load; all breakers were found inconsistent during repeated measurement . The shunt trip devices on six (6) breakers could not be set in accor-dance with the manufacturer's instruction . The magnetic latch devices on three (3) breakers could not be set in accordance with the manufacturer's instruction . Control wire terminations on one (1) breaker were poor, one to the point where the conductor pulled out of the lu * The "B" phase moving contact separated from one (1) breaker during a tripping operation due to a loose retaining rin . The terminal strips of two (2) breakers were cracke . The charging motor disconnect switches on four (4) breakers were broke * The upper portion of the "C" phase primary disconnect contact jaw separated from one (1) breaker due to a loose retaining rin . The model K-3000 breaker instruction booklet did not address the tripper bar loa . Two (2) arc chutes on one (1) model K-3000 breaker were broke i The licensee determined that two (2) of these defects (Nos. 6 and 9, as-terisked above), if left uncorrected, could have resulted in a breaker

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fault or failure. These two circuit breakers will supply power to the "D" Service Water Pump and the "B" Residual Heat Removal Pump. An uncorrected

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i defect could have disabled those safety-related pumps. This equipment is

[ powered by one of the two safeguards electrical divisions and is required to operate if the other division is inoperabl It also is protected

, against an Appendix R fire.

l The inspectors identified no deficiencies in the licensee's investigatio . Exit Interview During this inspection, meetings were held with plant management to dis-cuss the findings. No proprietary information related to this inspection was identified.

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