IR 05000213/1988011

Insp Rept 50-213/88-11 on 880517-0731.No Violations Noted. Major Areas Inspected:Plant Operations,Radiation & Fire Protection,Security Maint,Surveillance Testing,Switchgear Bldg Const & Open Items from Previous Insps

ML20151R289
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	07/31/1988
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20151R279	List: IR 05000213/1988011 ML20151R275
References
50-213-88-11, NUDOCS 8808120114
Download: ML20151R289 (20)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /88-11

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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 Plant -

Inspection dates: May 17, 1988 through July 31, 1988 Inspectors: Andra A.'Asars, Resident Inspector John T. Shediosky, Senior Resident Inspector Harold . Gregg, Senior Reactor Engineer Approved by: [/ // 1 i ef,RegbrProjectsSection18

, 1 3/ J /fd E. C7t)tCabe G 04ff Summary: Inspection 50-213/88-11 (5/17/88 - 7/31/88)

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Areas Inspected: This'was a routine safety inspection by the resident inspector Areas reviewed included plant operations, radiation protection, fire protection,

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security, maintenance, surveillance testing, switchgear building construction,

.iicensee events, open items from previous inspections, Control Rod Position Indi-cation System anomalies, Service Water System anomalies, and ECCS Single Failure Analysi Results: No violations were identified. Eleven previously identified Violations, Unresolved and Open Items were closed. One Unresolved Item wis opened (88-11-01)

relating to temperature stabilization before Rod Position Indication System re-calibratio Events reviewed during this inspection included licensee discovery of an unlocked high radiation barrier and Service Water System flow anomalie EK)2 PDR ADOCK 05000213 0 PNU

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TABLE OF-CONTENTS PAGE S umma ry o f Fa c i l i ty Ac t i v i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Review of Plant 0perations........................................... 1 Plant Operations Review Committee.................................... 2 Observation of Maintenance and Surveillance Activities............... 2 4.1 Auxiliaty Feeowater Pump Surveillance........................... 3 4.2' Post Accident Sampling System Survei11ance...................... 3 . Followup on Previous Inspection Findings............................. ~4 5.1 NRC Commitment Tracking......................................... 4 5.2 Surveillance Testing of Auxiliary Feedwater Automatic Initiation......................... .......................... 4 5.3 Switchgear Room Halon System.................................... 5 5.4 Changes to the Containment Boundary for Leak Rate Testing....... 5 5.5 Breech of Containment Integrity Ouring Surveillance Testing..... 5 5.6 Plant Design Change Record Standards............................ 6 5.7 Automated Work Order Documentation.......... ................... 6 5.8 Material and Equipment Parts List............ .................. 6 5.9 Torque Switch Setpvint.s for Motor' Operated Valves............... 7

5.10 Unauthorized Cpenir.g cf a Containecat Isolation Valve. . . . . . . . . . . 7 5.11 Compute r Room Fi re Wa i l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 Bulletin 85-03, Motor Opera ted Valve Failures. . . . . . . . . . . . . . . . . . . . . . . 7 7.- Followup on Events Occurring During th9 I n s pe t. t i o n . . . . . . . . . . . . . . . . . . . 8

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7.1 Licensee Event Reports and Sefeguards Event Reports........ . .. 8 7.2 Control Rod Position Indication Anomalies....................... 8 7.3 Power Reduction to 38% Power.................................... 9 7.4 U11ocked High Radiation Area Barrier............................ 10 7.5 Service Water Anomalies.................. ...................... 12 Review of Periodic and Special Reports............................... 13 ECCS Single Failure Conditions....................................... 14

. 1 Engineering Evaluation......................... ..................... 15 11. Generic Letter 83-28, Item 4.1, Reactor Trip System Reliability...... 16 1 Exit Interview............................................... ....... 18 i

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DETAILS

' 1. Summary of Facility Activities (71707)

At the beginning of the inspection period, the plant was in mode 5 recovering from a planned maintenance outage for repair of a reactor coolant pump seal package. The outage had been extended when two cases were identified in the charging and residual heat removal systems where system design did not meet the Single Failure Criteria of 10 CFR 50 Appendix A.' Necessary modifications and procedure changes were made and the plant was started up on May 31. On June 5, the elaat reached the milestone of 80 Billion.kw-hrs generated since being put in commercial operation on January 1,1968. A voluntary power re-duction to 38% was conducted on June 9 after discovery of an instrumentation uncertainty problem during calibration of the Axial Offset calculator Full power operations continued on June 11, after troubleshooting and recalibration of the Axial Offset instrumentation. The station operated at full power through the end of the inspection perio NRC Commissioner Kenneth M. Carr and Regional Administrator William T. Russell visited the facility on May 26. A thorough inspection tour of the plant was made with members of station management. Licensee corporate and station man-agement briefed NRC management on goals and achievement . Review of Plant Operations (71707)

The inspe: tor observed plant operation during regular tours of the following plant areas:

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Control Room --

Security Building

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Primary Auxiliary Building --

Fence Line (Protected Area)

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Vital Switchgear Room --

Yard Areas

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Diesel Generator Roons --

Turbine Building

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Control Point --

Intake Structure and Pump Building Cun',ral room instruments were observed for correlation between channels and for conformance with Technical Specification requirements. The inspector observed various alarm conditions which had been received and acknowledge Operator awareness and response to these conditions were reviewed. Control room and shift manning were compared to regulatory requirements. Posting and control of radiation and high radiation areas was inspected. Compliance with Radiation Work Permits and use of appropriate personnel monitoring devices were checke Plant housekeeping controls were observed, including control and storage of flammable material and other potential safety hazards. The inspector also examined the condition of various fire protection system During plant tours, logs and records were reviewed to determine if entries were properly made and communicated equipment status / deficiencies. These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, and Plant Information Reports. The inspector observed selected aspects of plant security including access control, physical l

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barriers, and personnel monitoring. In addition to normal working hours, the review of plant operations was conducted during the following backshifts and weekends:

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June 9, 1988 6:00 PM to 9:30 PM No unacceptable conditions were identifie Operators were alert and dis-played no signs of inattention to duty or fatigu . Plant Operations Review Committee (PORC) (40700)

'The inspectors frequently attended Plant Operations Review Committee (PORC)

meeting Technical Specification 6.5 requirements for required member at-tendance were verified. The meeting agendas included procedural changes, proposed changes to the Technical Specifications and field changes to design change packages. The meetings were characterized by frank discussions and questioning of the proposed changes. In particular, consideration was given to assure clarity and consistency among procedures. Items for which adequate review time was not available were postponed to allow committee members time to review and comment. Dissenting opinions were encouraged. The inspector had no further comment . Observation of Maintenance and Surveillance Testing (61726, 62700)

The inspector observed various r.aintenance and troubleshooting activities for compliance with requirements and applicable codes and standards, QA/QC in-volvement, safety tags, equipment alignment and use of .iumpers, personnel qualifications, radiological controls, fire protection, retest, and report-

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ability. Also, the inspector witnessed selected surveillance tests to deter-mine whether properly approved procedures were in use, test instrumentation was properly calibrated and used, technical specifications were satisfied, testing was performed by qualified personne', procedure details were adaquete, and test resuits satisfied acceptance criteria or were properly dispositione The fc11owing activities were reviewed:

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SUR 5.1-4, Core Ccoling Systans Hot Operational Test

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SUR 5.1-13, Auxiliary Feed Pump Monthly Functional Test

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SUR 5.7-19, Inservice Inspection Pump Surveillance

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PMP 9.5-116, AFW Pump Turbine Overspeed Trip Test

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CMP 8.5-40, Maintenance of the Steam Generator AFW Pumps and Terry Turbines

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SUR 5.4-34, Operation of Reactor Coolant Post Accident Sampling Module at Power Operation

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Switchgear Building Construction Activities

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4 4.1 Auxiliary Feedwater Pump Surveillance On July 5 at 8:40 am, the licensee removed the "A" Auxiliary Feedwater (AFW) Pump from service for maintenance. Technical Specification (TS) 3.8.A.2.b. requires that with one AFW pump inoperable, it be returned to service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant must in hot standby within the next six hour ~

Inservice Inspection (ISI) personnel had requested that maintenance check the alignment of the terry turbine to the pump af ter vibration data taken by Reliability Engineering indicated that they could be misaligned. The alignment check was performed in accordance with CMP 8.5-4 After re-coupling, the operability surveillance, SUR 5.1-13, was nerformed with satisfactory results. However, SUR 5.7-19, the ISI surveillance, had

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unsatisfactory result The terry turbine outboard bearing vibrations (in the y-direction) were in the required action level of the ISI Ac-ceptance Criteri The required action level is 300 mills; the vibration data indicated 333 mill An additional work order was issued for replacement of the terry turbine inboard and outboard bearings, thrust bearing, and wheel and shaf Performance of the terry turbine overspeed trip test was also specifie The pump and turbine were realigned and retested three times before testing satisfactoril The out of specification vibrations had been reduced to 294 mills (in the ISI alert range). The "A" AFW pump was declared operable and returned to service on July 7, at 5:20 pm, within the time permitted by T '

Since startup from the refueling outage, the "A" AFW pump has been in the alert range of the ISI Acceptance Criteri During the refueling ,

outage, the licensee installed a complete new rotating element in the pump. Since then, ISI has required a monthly vibration surveillance in conjunction with the monthly cperability surveillance to monitor vibra-tions of the turbine in'aoard bearin ,

The licensee has conducted additional reviews of the ISI vibration dat The reviews indicated that the pump and turbine may be misaligned by approximately four mills. Currently, the licensee plans to remove this pump from service at the next testing interva This has been scheduled for August 1, 1988. Additional realignments will be performed and a new i ISI baseline may be establishe The inspectors observed portions of these maintenance and surveillance activities; no deficiencies were identified. Licensee actions in this area will be reviewed in future inspections.

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4.2 Post Accident Sampling System Surveillance In effort to upgrade the Post Accident Sampling System (PASS) and the associated prccedures, the licensee has retained a consultant, Lehmannn Associates, to review the equipment and procedures. The inspector ob-served portions of_ the PASS operation conducted under SUR'5.4-34. 'In-

~itially, testing was inhibited by the inability to open a sample isola-tion valve from the reactor coolant system cold leg (SS-30V-67). Trouble shooting identified'that several other valves were affecte The cause was traced to a valve on the sample panel (V-4) which was leaking by nitrogen from the nitrogen blanket on the PASS system causing a back-pressure on these valves. A trouble report was issued for V-4, The nitrogen was bled off and the affected valves were cycled successfull The surveillance testing continued without incident. It was noted that the stripping pump was difficult to start and that there is no positive indication of pump operatio Surveillance test observations, as well as procedure and system modifi-(

cation recommendations will be provided by the consultant in several weeks. The licensee will evaluate these recommendations for implementa- -

tions. PASS upgrades will be foll.wed during future inspection . Follnwup on Previous Inspection Findings (92701, 92702) -

6.1 NRC Commitment Tracking (Closed) Unresolved Item (87-25-01): This item was identified as unre-solved because of an omission while upgrading an Emergency Operating Procedure (EOP) into a symptom oriented procedure (ES). A statement relating to a commitment to close loop charging valves CH-MOV-2928 and C during the recirculation phase was not transferred from E0P 3.1-4 to the new ES- The licensee has now included that requirement within ES-1.4, Transfer to Two Path Recirculation, Revision 5; this has corrected the specific nroble In addition procedure E0P 3.1-0, Emergency Response Procedures, Revision 5, provides a listing of commitments made to the NRC which af-fect these procedures. This item is closed. However, an associated item

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87-08-01 which addresses broader issues in commitment tracking remains open and will be reviewed during future inspection .2 Surveillance Testing of Auxiliary Feedwater Automatic Initiation ,

(Closed) Unresolved Item (84-02-02): This item was identified to trat k the clarification and incorporation of an existing Technical Specifica-tion into the Standardized Technical Specifications (STS). Currently, Technical Specification 4.8.3.C requires each Auxiliary Feedwater (AFW)

pump to be started from each actuation test signal at refueling intervals.

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The current draft STS revision requires that each AFW pump start auto-matica11y upon receipt of an actuatica signal in surveillance 4.7.1. The AFW Trip actuating devices are tested for the loss of all Main Feed-water Pumps during a refueling outage; also, analog channel operational testing of Steam Generator Water Level - Low in Table'4.3-2, Item 3 of the draft STS. This conforms with NUREG 0452, Revision 4, STS for Westinghouse Pressurized Water Reactors. This item will receive further review when the STS are submitted to NRC and is therefore close .3 Switchgear Room Halon System (Closed) Unresolved Item (86-17-02): The licensee had declared the switchgear room halon system inoperable when it was identified that the system did not meet the NFPA code requirements. This item was previously addressed in NRC Inspection Reports 50-213/86-17 and 87-05. During the 1987 refueling outage the halon system was upgraded to meet the accept-ance criteria. Testing was performed in October 1987. Results were satisfactory and demonstrated that electrical and mechanical components function as designed and that the required concentrations of halon can be attained and maintained. The inspectors had observed portion of the testing and reviewed test result No further deficiencies were identi-fie This item is close .4 Changes to the Containment Boundary for Leak Rate Testing (Closed) Violation (86-08-05): The licensee failed to conduct a safety evaluation when changes were mada to the containment penetration valve line-ups in support of leak rate testin The inspectors had concluded that the valve line-ups for P-80 and P-62 had reclassified two valves as containment isolation valves (RH-MOV-31 and SA-V-413). The licensee replied, by letter dated June 25, 1986, that these valves have-been de-signated as containment isolation since preoperational testing. Both valves are tested i: accordance with the Inservice Testing Progra Operation of SA-V-413 during power operations is covered administratively and described in Technical Specification (TS) 1.8.2.6, Amendment 9 Both valves are listed as containment isolation valves in the current draft Standard TS. Based on this, this item is close .5 Breech of Containment Integrity During Surveillance Testing (Closed) Violation (86-08-06): Operation of manual containment isolation valves during surveillance testing with the piant at power. During per-formance of SUR 5.1-6, Reactor Containment - Leak Monitoring, operators are required to open SA-V-413 when charging air into containment. How-ever, SA-V-413 is a locked closed containment isolation valve. Technical Specification (TS) Amendment 90 revised TS 1.8.2.b to permit manual J operation of this and selected other manual containment isolation valves only if a operator is stationed locally to close the valve within 60 L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ . _ _ _ _ _ _ _ _ J

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seconds in the event that containment isolation is necessary. Admini-strative changes incorporating this TS were reviewed i,1 NRC Inspection Report 50-213/87-27 and found to be adequate. This item is close .6 Plant Design Change Record Standards (Closed) Inspector Follow Item (85-25-01): Clarification of performance standards to improve the timeliness in supplying plant design change documents to the site. The inspectors routinely attend nieetings of the Plant Operations Review Committee and review the documentation packages prior to the meetings. They found very good performance in the area of proposed design changes over the last year. Plant Design Change Record (PDCR) packages were generally provided to the station staff in adequate time to allow review and acceptance of outage work prior to the 1987 refueling outage. At other times, packages arrived on site in a tfnely manner. This item is close .7 Automated Work Order Decementation I

(Closed) Inspector Follow Item (85-21-10): Inadequate documentation within completed work orders. A Maintenance Department Instruction has been issued which establishes the standard for documentatior, within the automated work order system. MA 1.5-1, Work Order Preparation, Work Control and Documentation, dated June 21, 1988 addressed pre-job planning, work package content and post-job documentation. In order to insure that these documents meet established standards, completed work orders are being reviewed by the maintenance supervisor. Completed work orders are also reviewed by the NRC resident inspectors on a sampling basis as a routine inspection activit This item is close .8 Material and Equipment Parts List (Closed) Violation (84-02-02): Material and Equipment Parts List (MEPL)

inaccuracies. The licensee has completed several measures to improve tne MEPL and to provide better control over activities which support the MEP Several issues of the updated Parts List have been issued since the original findin Tha latest issue has addressed components on a system basis. All components (i.e., electrical, mechanical and instru-mentation) are listed within a system. In addit;on, a system description and safety evaluation is provided to provide a basis for the documen This newly developed document is presently in distribution to the docu-ment holders. Because of its bulk and in an effort to better control the information, the MEPL is being placed in a computer data base which will become available to its users through the corporate Time Share Terminal The licensee had also initiated a program of Process and Instrumentation Drawing (P&ID) updating to reflect as-built coaditions. For this effort, the licensee has established a standard for P& ids, entered these drawings

into a computer aihd arawing (CAD) system and verified the revised

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documents through system inspection The revised drawings are used routinely by the licensee and NRC Resident Inspectors. This item is close ' 5.9 Torque Switch Setpoints for Motor Operated Valves (Closed) Unresolved Item (87-22-02): Assure that Motor Operated Valves tested under NRC Bulletin 85-03 have Torque Switch Settings Checked based on the updated master setpoint list. The master setpoint list presently reflects the required Torque Switch Settings. This item is close .10 Unauthorized Opening of a Containment Isolation Valve (Closed) Unresolved Item (86-27-05): With the plant at power, the manual containment isolation valve (CC-V-884) on the neutron shield tank fill line was opened periodically to support filling evolution This issue has been discussed in NRC Inspection Reports 50-213/86-08, 86-20, 86-27, and 87-2 Technical Specification 1.8.2.b. was revised by Amendment 90 to permit opening of this valve with a locally stationed operator able to close the valve within 60 seconds of a containment isolation signa The inspector verified that corresponding administrative controls and procedures have been revised to reflect this change. This item is close .11 Computer Room Fire Wall (Closed) Unresolved Item (86-24-01): This item was opened to track the long term resolution of the computer room / control room fire barrier un-certainties. Technical Specification Amendment 81 had approved the de-letion of the existing computer room fire detection system but questions arose concerning the integrity of the wall shared by the computer and control rooms. This issue was discussed in NRC Inspection Reports 50-t 213/86-24 and 87-06. During the 1987 refueling cutage, the licensee imp'emented Plant Design Change Record (PDCR)87-871, Computer / Control Room Fire Wall. This PDCR included an upgrading of the wall ss .eet seismic criteria and establishing a one hour fire resistanc The in-

, spectors observed portions of the PDCR implementation. No discrepancies were identified; this item is close .Bulletin 85-03, Motor Operated Valve Failures (92703)

By letter dated June 11, 1986, the licensee responded to NRC Bulletin 85-03 and addressed the methods for implementation of the Bulletin requirement The actions taken by the licensee to resolve these concerns were addressed in inspection 50-213/87-25 report paragraph 13. A final report concerning the program implementation during the 1987 outage was submitted on June 27, 1988. This report included verification of completion of the program, summary of findings and necessary ad,justments, and a summary of data. The inspectors had no further questions at this time.

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8 Followup on Events Occurring During the Inspection (71707,90712,93702)

7.1 Licensee Event Reports (LERs) and Safeguards Event Reports (SERs)

The following LERs and SERs were reviewed for clarity, accuracy of the description of cause, and adequacy of corrective action. The inspector determined whether further information was required and whether there were generic implications. The inspector also verified that the report-ing requirements of 10 CFR 50.73, 10 CFR 73.71, and Station Administra-tive and Operating, and Security Procedures had been met, that appro-priate corrective action had been taken, and that the continued operation of the facility was conducted within Technical Specification Limit * 88-12 Reactor Trip Due to Improper Installation of Turbine Stop Valve Cam Switches

• 88-13 Design Deficiency Identified in Charging Pump Motor Operated Suction Valves 88-14 Failure to Take Samples With Service Water Effluent Monitor Out of Service 88-15 Design Deficiency Identified in Residual Heat Removal Pump Seal Coolers

• • 88-503 Safeguards Event Report

• • 88-504 Safeguards Event Report

• Event detailed in NRC Inspection Report 50-213/88-08

• • Event detailed in NRC Inspection Report 50-213/88-12 No unacceptable conditions were identifie .2 Control Rod Position Indication Anomal Les For many years, the licensee has had a chronic problem with the accuracy of the analog Rod Position Indication (RPI) System. Operating experience has shown that fluctuations in the Reactor Coolant System temperature can cause variances in the analog RPI system output even with no rod movement. This issue was previously discussed in NRC Inspection Report 50-213/88-0 Before startup, with the plant at hot zero power (535 degrees F), the licensee programmed the RPI to agree with the step counters. Data was L

also taken for the ongoing study of RPI indication changes ve. sus reactor power and average temperatur .

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On May 31 at 12:15 am, with the plant at_94% power during a power in-crease to full power, operators noticed that the anslog RPI was indicat-ing control rods in Banks A, C, and D were above the reactor core (greater than 320 steps). The rods in these banks were in the full out position (320 steps) at the time. The bank average RPI indication for these banks was not within the 16 steps of the digital step counter specified by Technical Specification (TS) 3.10.2.2. The initial correc-tive action taken was to separately insert each bank five steps to verify that the analog RPI would accurately indicate rod movement corresponding to the digital step counter. This evolution successfully demonstrated that the analog RPI would track rod movement and display a relatively 4 constant difference in rod position with respect to the digital step counter. The licensee elected to continue with the power increase to full power. Plant Information Report (PIR)88-116 was filed for this 0CCurrenCO.

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On June 1, during review of the station log, the inspector questioned if the RPI had been recalibrated. The shift supervisor stated that I&C technicians were to begin that morning. The recalibration was completed by 12:00 noon. This represents approximately a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> delay for RPI recalibration. TS 3.10.2.2 requires that with more than one analog or ,

digital rod position indicator per bank inoperable, the olant must be in hot standby within six hours and the reactor trip breakers open or the control rod drive lif t coils deenergize This matter was discussed with station mar.:gemen The inspectors o re informed of the license 9's position that it had been determined that the RPI was operable and that the rods were verified full out during their investigation on May 31. This was based on the fact that RPI was shown to track rod movement and display a relatively constant difference in rod position with respect to the step counters. The licensee delayed recalibration due to uncertainty of the effect on the computer program-ming done while the plant was at 535 degrees F and to allow the tempera-ture effects on RPI to stabilize. Determination of appropriate stabili-zation time before RFI is recalibrated is Unresolved Item (88-11-01).

The licensee has established a RPI Task Force. Their charter is to evaluate the variances in RPI and determine the causes of the variations and corrective actions necessary to improve RPI accuracy. Until such time, the licensee will closely monitor RPI performance during power manipulations and recalibrate as necessary.

l 7.3 Power Reduction to 38% Power

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On June 9, reactor power was reduced to 38% power after an instr..menta-tion uncertainty was discovered following a monthly calibration check of the Reactor Core Power Distribution Axial Offset calculators. The results of that surveillance, in which manually calculated values are

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compared to indicated values, identified a difference ir two of four channels of about 7.5% and 3%. Both displayed and calculated values were i

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within Technical Specification (TS) 3.17.1 requirements for 100% power operation. However, the difference exceeded the licensee's administra-tive limit which was based on an assumed instrument accuracy of_2%.

JBecause of this, the licensee determined that they could net meet sur-ve111ance requirement a. of TS 3.17.1.1. which requires correlation of excore and incore detectors on a continuous basis. .The licensee elected to conduct a voluntary power reduction to less than 40% where axial off-set is not limited within the time restrictions of TS. The resident inspector observed the licensee's investigation and subsequent power i reductio Troubleshooting activities identified a third channel out of specifica-tion. Failures of these _ channels were attributed to two separate factor First, Channel 32 failed because an internal resistor had gone outside tolerance and failed. This drawer had previously been a spare drawer in which resistor upgrades in 1974 had been inadvertently omitted. This spare was periodically used and had been permanently installed in June 1987. Channel 32 was replaced with another spare drawer which had the correct resistor installed. Failure of Channels 33 and 34 were attri-buted to overranging the input voltage to the Internals Vibration Monitor (IVM) to axial offset isolators. These isolators had been installed this past refueling outage as part of the plant Design Change Request for the IVM. The circuit was modified to reduce the input voltage to isolator amplifiers to within specificatio The inspectors noted that this system is to be replaced during the upcoming refueling outage as part cf the Nuclear Instrumentation System replacemen The affected channels were repaired and retested satisfactoril Full power operations continued on June 1 .4 Unlocked High Radiation Area Barrier In support of the Appendix R Switchgear Building Construction, construc-tion personnel have been working the the charging pump cubicles. At times, when the pumps are operating, these cubicles have radiation areas in excess of 1000 mr/ hour and are therefore locked high radiation area The gates to these cubicles have locting gate On July 8, at 10:00 am, health physics personnel briefed and admitted construction personnel into the "A" Charging Pump Cubicle. When exiting the area for lunch at 12:05 pm, construction personnel failed to shut (and therefore lock) the gate to the "A" cubicle. The open gate was identified by a Radwaste Technician on tour at 12:30 pm. The technician immediately closed and locked the gate and informed health physics per-

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sonnel. At this time, Health Physics suspended all construction activi-

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ties in locked high radiation area Generation Construction Supervision was then informed of the incident. All construction activities, with the exception of work in the Switchgear Building, were suspendea, e

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Recent of 10 to 40 surveys of the mr/ hou work area However, :ndicate the area has general area radiation been classified as a locklevels c l" high radiation area because in a corner of the metering pump cubicle 4 (adjacent to the "A" Charging Pump) is a line to the regenerative heat exchanger which can have radiation levels greater than 1000 mr/ hou Areas with radiation levels greater than 1000 mr/ hour are required by Technical Specifications (TS) 6.13.1 to be locked to prevent unauthorized entr Additionally, the Radiation Work Permit (RWP) governing the work states that the locked high radiation area must be kept locked or con-tinuously guarded. The failure to lock the gate upon exit constitutes a violation of the RWP requirements. (At the time of the incident actual radiation levels were below that which requires locking by TS 6.13.1).

However, because the licensee identified the open gate after a relatively short period of time and took immediate corrective actions and is imple-menting extensive long term actions, no notice of violation was issued in this cas A review of the Security Computer recor's identified that only two indi-viduals (a security guard and a fire watch) ma'e entries into the Primary Auxiliary Building (PAB) during the 25 minute period in which the gate was unlocked. The licensee interviewed both of these individuals on July Both stated that they did not enter the "A" Charging Pump Cubicl Licensee review of this incident indicate that retraining, pre-job briefings, and additional supervision of the contractor construction personnel is necessary. On July ll, all construction personnel had com-pleted a read and sign which covered the requirements for work in locked high radiation areas. At that time the work suspension was lifted and construction activities resume The licensee has initiated pre-job briefings each morning for the con-struction personnel. The briefings cover all requirements and special l conditions for the work area for each work crew.

l Additional supervision is also being added to the construction effor Bechtel has brought several :upervisors with nuclear plant experience to the Jcb site. Generation Construction has assigned two additional supervisors to the job; bringing the total to five. With these additions, it is intended that each work crew will have continuous supervisio In addition, the licensee distributed a news letter around the site in-forming station personnel of this incident and reminding personnel of the importance of assuring locked high radiation areas remain locke Previous experiences and corrective actions for similar events at the Millstone Units will also be reviewed for applicability to Haddam Neck, i . -- -

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7.5 Service Water Anomalies During this inspection period, the Service Water System (SW) reached an average temperature of approximately 84 degrees F. This elevated tem-perature affected mainly cooling of balance of plant systems. This ad-ditional SW demand in the plant reduced the available SW to the Contain-ment Air Recirculation (CAR) heat exchangers (HXs). On June 23, the control room alarm signifying low SW flow to these heat exchangers alarmed when flow to the No._3 HX dropped below the 300 gpm setpoin The alarm cleared after about six hours. A low flow condition occurred again on July 9. The No. 3 HX was primarily affected, however all four HXs did experience low flows. This alarm did not clear by the end of the inspection period. When the alarms first came in, control room per-sonnel verified that there was still flow through the HXs and switched Adams Filters (the in-line water filter) as required by annunciator re-sponse procedure ANN 4.7-1 Plant Information Report (PIR)88-129 was filed for this off-normal conditio Under normal operating conditions, without elevated SW temperatures, at least 400 gpm are supplied to the CAR HXs. The basis for maintaining this flow under normal operations has been to demonstrate that adequate flow can be achieved under accident conditions although the SW system is under different configuration and demands during accident condition The accident analysis assumed 400 gpm SW flow to the CAR HXs in order to maintain containment pressure less than 40 psig during accident con-ditions. For this reason, operating logs require that operators verify at least 400 gpm SW flow each shift. A quarterly Inservice Testing (IST)

surveillance (SUR 5.7-118, SW Penetration Check Valves) is also performed to verify the operability of the SW check valves upstream of the HXs by verifying differential pressure over the check valves and HXs. For pur-poses of this test, a 12 psi pressure drop has been correlated to a 400 gpm flow rate. This was previously discussed in NRC Inspection Report 50-213/87-12, detail Engineering conducted a review to detcuaine if any HX blockage could have occurred and if continued operation with less than normal flow to the CAR HXs is within the station design conditions. The initial conclusion was that, based on a review of station logs and recent IST testing, there is no blockage. It has also been determined that the low flow condition with elevated SW temperatures is not of concern under normal operations because of the SW system configuratio The licensee plans to conduct a more detailed analysis of SW syste This analysis is to encompass a review of the SW design basis, including flow requirements to the CAR HXs under normal and accident conditions and peak river water temperatures permitted by the accident analysi The results of this analysis will be reviewed during future inspections.

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-13 Also during this period, a leak developed requiring isolation of SW to the No. 3 CAR HX. On July 21 at about 8:20 pm, a Security Guard on rou-tine tour notices a leak on the elbow tap on the SW return line from the No. 3. CAR HX. ~This flow tap supplies the associated SW flow instrumen--

tation. When an operator tried to isolate the leak, the line broke.

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SW was isolated to the HX, and the No. 3 CAR fan /HX was declared in-operable. Temporary repairs were made immediately. Maintenance person-nel were called in to repair the pip The affected CAR fan and HX were made operable at 1:26 am on July 2 Operations personnel have been instructed to minimize the time that SW is isolated to the CAR HXs for Containment Integrity consideration By letter dated August 16, 1983, the. licensee described that SW system

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pressure is maintained greater than peak accident pressure upstream of the HXs and not greater than 10 psi below accident pressure down stream of the HXs. This is equivalent to about 40 psig and 30 psig, respectively, under normal operations. M ntenance of these pressures is part of the licensee's justification for not performing 10 CFR 50 Appendix J Type C leak testing on these lines and for operations without containment isolation valves ~on these lines. This exemption from Appendix J re-quirements was reviewed under Systematic Evaluation Program Topic VI-4, Conteinment Isolation Syste During review of NOP 2.2-2, Operation at Power, Steady State Operation and Surveillance, the inspector noted that the "remarks and limits" block for the HX inlet pressure reading does not indicate that a minimum of 40 psig is required. The outlet pressure reading limit does state that the minimum outlet pressure is to be 30 psig. The-inspector brought this to operations management attention and was informed that this will be

considered in the current revition and upgrade to this procedure. The

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inspector reviewed the logs for the period between July 11 and July 26; for this period, the inlet and outlet pressures were maintained at ap-proximately 50 and .30 psig, respectively. The inspector had no further concerns at this tim . Review of Periodic and Special Reports (90713)

Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewe This review verified that the reported in-

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formation was valid and included the NRC required data; that test results and supporting information were consistent with design predictions and performance specifications; and that planned corrective actions were adequate for resolu-tion of the problem. The inspector also ascertained whether any reported in-formation should be classified as an abnormal occurrence. The following periodic reports were reviewed:

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Monthly Operating Report 88-04, Covering the Period April 1,1988 through April 30, 198 w - - ..-, - . - - - - - , _ - . - .. -- -

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Monthly Operating Report 88-05, Covering the Period May 1, 1988 through May 31, 198 Monthly Operating Report 88-06, Covering the Period June 1,1988 through June 30, 198 New Switchger Building Construction Bimonthly Progress Report No.10

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Haddam Neck Plant Core XV Startup Physics Test Report, May 24, 198 Review of Startup Test Report For Cycle XV The inspector reviewed the licensee's Cycle XV Startup Physics Test Report documented in accordance with the requirements of Technical Specification (TS) 6.9, Reporting Requirements. The results were summarized the Connecticut Yankee Atomt Power Company, Haddam Neck Plant Cc re XV Startup Physics Tcst Report, transmitted to NRC by letter dated May 24, 198 The purpose of the Startup Physics Test Program is to verify that the measured parameters comply with.TS limits, and to validate core design calculation The content of the test program was found to be consistent with the require-ments cf the Updated Final Safety Analysis Report and TS. The results of the Startu.n Ta t Program were reviewed and found to be in compliance with TS and the test program acceptance criteri . ECCS Single Failure Conditions (37700)

During the previous inspection period, see NRC Inspection Report 50-213/88-08, the licensee identified the susceptibility to single failure in the charging system which could compromise the availability of the charging pumps. As a result of this discovery, the licensee committed, by letter dated May 13, 1988, to conduct single failure analyses of mechanical, electrical and instrumenta-tion and controls systems. The mechanical system review was to be completed before station startup from the maintenance outag The mechanical system single failure review was based on the Probabilistic Risk Assessment (PRA) as it related to ECCS and critical support system Personnel participating in the review included representatives from the PRA Group, LOCA Analysis Group, Mechanical Engineering, SEP Group, and Haddam Neck Plant Operation Twenty five (25) potential single failure concerns were identifie Of these, all were dispositioned with the er.ception of two re-lating to Service Water (SW) flow to Residual Heat Removal (RHR) pump oil coolers. On May 17, the inspectors reviewed this analysis and item disposi-tions with the licensee at the corporate office in Berlin, Connecticu The potential single failures identified with the RHR system involve SW supply to the RHR pump seal water coolers, packing and bearings in accident condi-tions. Under accident conditions, the RHR heat exchangers are cooled by S Should one of the two heat exchanger SW inlet valves fail (SW-MOV-5 and -6),

cooling water to the RHR pump would short circuit through the idle heat ex-L _

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changer. 'In effect, SW cooling to the RHR pump would be bypassed and pump overheating could occur. To prevent this short circuiting, the licensee in-stalled a check valve in each line to the RHR pumps. This change was per-formed under Plant Design Change Record 88-940. Post modification testing '

included flow measurements to verify that the additional pressure drop caused by_the check valves does not affect the cooling flow to the pump Currently, the licensee is in the planning stages of the electrical and in-strument and controls single failur.' review Once a schedule is developed, it will be forwarded to NRC. The progress and results of these reviews will be reviewed during future inspections.

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1 Engineering Evaluation (37055)

The inspector reviewed site and corporate engineering organizations; including their to communication and work coordination efforts. Major work projects performed during the outage and day-to-day engineering activities were also reviewe A new site engineering supervisor has recently been assigned. This individual had held a corporate engineering supervisory position. In the daily activi-ties reviewed by the inspector, this supervisor displayed a competency in resolving engineering issues and in communications with onsite and corporate staffs. Closer ties with corporate engineering were apparen The inspector reviewed major project completed during the 1987 refueling out-age. These included the low pressure turbine rotor replacement and core barrel thermal shield repair In each project, engineering planning and interface between site, corporate and craft was eviden The inspector fol-lowed Engineering involvement in daily operational concerns such as low ser-

!. vice water flow to the containment air recirculation heat exchangers and the moisture problems with the control air dryers in the turbine building. This inspector determined that engineering resolution of these problems was ap-propriate and progressing satisfactoril The station procedures upgrade program is about 25*s complete. The inspector

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reviewed the following procedures that have been upgrade CMP 8.5-15, Maintenance of High Pressure Safety Injection Pumps

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NOP 2.5-1, Placing the Pressurizer Relief Tank in Service

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PMP 9.2-82, Installation of EEQ Crimp Style Terminations 4 The new format, the step-by-step clarity, and graphics are significant im-

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provements over the old procedures and should eliminate many of the past pro-cedural adequacy and adherence problem Site engineering has also developed new procedures to better control and provide guidance in procurement, dedica- .

tion, and upgrading of commercial grade equipment.

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Additionally, site engineering is in process of inputting into the plants PMMS computerized work tracking systems whereby engineering status can be readily determine The licensee's use of Probabilistic Risk Assessment (PRA) as a working tool continues to be effective. The PRA model is continually updated to reflect plant and procedural modifications. It has been used as a prime working tool ,

to provide an analytical basis in prioritization of wor Removal and re-placement of_ PCB oil filled main switchgear transformers with dry type trans-formers was a beneficial program initiated by PRA that exceeded regulatory requirements. Another PRA initiated change was the procedure and piping modifications to enable use of HPSI recirculation in the event of a break in the ECCS syste The procedure change has been implemented and the final HPSI piping electric tie-ins to single-failure proof recirculation path will be completed in the 1989 refueling outage. Completion of the replacement of the Nuclear Instrumentation System with modern state of the art instrumentation and the Switchgear Building electrical tie-in will have positive effects on the PRA model. Currently, these are also scheduled for the 1989 refueling outag Recently, the site requested and received recommendations from the PRA Group of changes to routine site activities then can lower the Core Melt Frequenc These recommendations were in the areas of increased surveillance frequency, training, and preventative maintenance. Implementation of these upgrades will be reviewed during future inspections as they account for a total maximum benefit of about a 2*4 reductinn in the current core melt frequenc The Quality Services (QS) organization and work performance was also reviewed during this inspection. QS was recently reorganized from four separate groups to a single organizatio This organization is different from other Quality

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Assurance organizations and is staffed with several high level senior engi-neers. Several audits of engineering have been pert'ormed. The audits re-viewed design input, calculations, in process field installations, and equip-ment testing. The recent NRC SECY 87-220 paper has been well received by the licensee and more Performance Based Verification (PBV) audits are planned for

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safety significant systems and components. QS has drafted a position paper

regarding PBV in response to the SECY paper issues. QS has already conducted PBV audits of station hardwar . Generic Letter 83-28, Item 4.1, Reactor Trip System Reliability - Vendor Related Modifications (MPA B-80) (TI 2515/91)

The NRC issued Generic Letter 83-28 on July 8,1983 following the failures of reactor trip system breakers at the Salem Nuclear Plant. In addition to the Generic Letter, an NRC Bulletin (83-01) was issued on February 25, 198 Item 4.1 of the Generic letter required that the licensee conduct a review of vendor-related reactor trip breaker modifications and verify that each modification has been completed or that a written evaluation of the technical i basis for not implementing the modification has been complete .___ _ _ -- - ._____ .~ _ _ _ , _ - _-- _ .__ _ ._ _ ___ _ _ _ __ ~

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The licensee's letter of November 8, 1983, written in response to the generic letter, stated that three Westinghouse Technical Bulletins were applicable to the Haddam Neck Plant reactor trip breakers and that the licensee verified that the necessary actions were completed. This included the replacement of undervoltage devices in 1972 which was recommended in Westinghouse LCD-LEEC-1 An April 5, 1983 letter, written in response to the bulletin, documented the plant's trip system logic which includes circuits to both undervoltage and shunt-trip devices in the reactor trip breaker The NRC staff reviewed the licensee's actions regarding this and other items of the Generic Letter and published a Safety Evaluation of its findings dated February 27, 1986. The staff found the licensee's actions acceptabl The licensee's Procurement, Receipt, Storage and Handling Programs which fol-lowed from the Generic Letter were addressed in NRC Inspection 50-213/84-26 which was conducted November 13-16, 198 Because of an NRC required inspection follow-up of Multiplant Action Item (MPA) B-80 which was published on April 6, 1987, the inspector reviewed the licensee's action taken in response to Generic Letter 83-28, Item 4.1, Vendor Related Modification The inspector reviewed previous documentation concerning this issue and the following Westinghouse Technical Bulletins and verified that bulletin recom-me1dations have been adopted within the station administrative control NCD-ELEC-18, Replacement of Undervoltage Attachments on Breakers in Reactor Trip Switchgear, dated December 17, 1971

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NSD Data Letter 74-2, Reactor Trip Breaker Maintenance, dated February 19, 1974

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NSD Technical Bulletin 74-12, Electrical Terminal Blocks, dated November 18, 1974

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Letter Nuclear Services Division to Haddam Neck, Field Lubrication of DB Breakers, Serial CYW-83-536, dated April 19, 1983

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NSD Technical Bulletin 83-02, Revision 1, 08-50 Reactor Tip Breaker Maintenance, dated September 13, 1983

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Letter Nuclear Services Division to Westinghouse Owners Group, Compre-hensive Maintenance Program for 08-50 Reactor Trip Breakers, dated December 15, 1983

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Letter Westinghouse New England Engineering Services, Operating Times of Auxiliary Switch, dated March 5,198 _ _ _ _ _ _ _ - _ _ _ __ ___ _ _ _ _ _ _ _ _ _ _ _ _ _

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There were nc unacceptable conditions identifie This method was used as there was no clear record as to the basis for the licensee's earlier conclu-sion that all applicable vendor recommended modifications had been complete . Exit Interview (30703)

During this inspection, meetings were held with plant management to discuss the findings. No proprietary information related to this inspection was identified.

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