IR 05000213/1988003

Insp Rept 50-213/88-03 on 880301-0404.No Violations Noted. Major Areas Inspected:Plant Operations,Radiation Protection, Fire Protection,Security,Maint,Surveillance Testing,Licensee Events & Open Items from Previous Insps

ML20151V692
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	04/20/1988
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20151V682	List: IR 05000213/1988003 ML20151V679
References
50-213-88-03, 50-213-88-3, NUDOCS 8805030148
Download: ML20151V692 (12)

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U.S. NUCLEAR REGULATORY COMMISSION REGION-I Report N /88-03 Docket'N ,

License N DPR-61 Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06101

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Facility: Haddam Neck Plant, Haddam. Neck, Connecticut Inspection at: Haddam Neck Plant Inspection ~ dates: March 1, 1988 through April 4, 1988

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Inspectors; Andra A. Asars, Resident Inspector John T. Shediosky, Senior Resident Inspector- -

Approved by: , )k 20/E9 E. C. McCabe, Chief, Reactor Projects 18 Date

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Summary: Inspection 50-213/88-03 (3/1/88 - 4/4/88)

Scope. ~ Routine safety inspection (147 hours0.0017 days <br />0.0408 hours <br />2.430556e-4 weeks <br />5.59335e-5 months <br />) by the resident inspectors. Areas i reviewed included plant operations, radiation protection, fire protection, security (including'a short strike by guards), maintenance, surveillance testing, licensee events (including licensee-identified violations), and open items from previous inspections and from the Systematic Evaluation Progra Results. No violations were cited. There were two licensee identified Technical !

. Specification violations: the Residual Heat Removal System was taken out of service -

on two occasions for over one hour (Report Detail 5.2); and a surveillance to .

monitor reactor power during low power physics testing was missed (Report Detail :

5.5). These were acceptably responded to by the licensee. The licensee's contin-gency planning for and response to a guards' strike were found to meet Station Physical Protection Plan commitments (Report Detail 5.3). Licensee actions upon detecting a potential calculational error in large Break LOCA ECCS delivery rates (Report Detail 5.6) were found appropriat l

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TABLE OF CONTENTS PAGE Summary of Facility Activities....................................... 1 Plant 0peraticns..................................................... 1 Plant Operations Review Committee.................................... 2 Maintenance and Surve111ance......................................... 2 Followup on Events Occurring During the Inspection................... 3 5.1 Licensee Event Reports and Safeguards Event Reports............. 3 5.2 Premature Shutdown of Residual Heat Removal System.............. 3 5.3 Strike By Bargaining Unit Security Officers..................... 4 5.4 Reactor Trips On Spurious High Startup Rate..................... 5 5.5 Mi ssed Surveillance During Physics Testing. . . . . . . . . . . . . . . . . . . . . . 5 5.6 Large Brea k LOCA ECCS Flow Delivery Error. . . . . . . . . . . . . . . . . . . . . . . 6 Periodic and Special Reports......................................... 7 Inoperable Dropped Rod Circuitry..................................... 7 Systematic Evaluation Program Findings............................... 8 8.1 Topic II-3.B Flooding Potential And Protection Requirements.... 8 8.2 Topic III-3.C, Inservice Inspection of Water Control Structure . Low Temperature Overpressure Protection.............................. 10 10. Exit Interview....................................................... 10

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4 DETAILS Summary of Facility Activities The licensee' concluded their four.teenth refueling outage, which began on July 18, 1987. Initial criticality and initiation of Low Power. Physics Testing were on March 19. That same day, the reactor tripped automatically on a spurious high startup rate on Intermediate Range (IR) Channel 22. The reactor was made critical again on March 20. Physics testing continued. A planned manual reactor trip for physics testing was conducted at 10:25 a.m., March 21. The reactor was made critical at 12:06 p.m. the same day. A second automatic reactor trip on spurious high startup rate on IR Channel 22 occurred on March 22. The reactor was made critical later on March 22 and physics testing was completed. On March 24, while below 5% power, the licensee re-ported the discovery of an apparent error in the Large Break LOCA Analysi ,

This precipitated a limit of 40% on power operation. The licensee prepared

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a Justification For Continued Operation (JCO) at up to 80% Power while the Large Break LOCA-Analysis was reevaluate On March 30, the NRC concurred with the JCO. At the conclusion of this inspection period, the plant was i operating at 80% pcwer and the licensee had submitted a Proposed Technical Specification Change to allow full power operation with the revised Large i

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Break LOCA Analysi . Plant Operations (71707. 71711)

The inspector 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 Rooms --

Turbine Building

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Control Point --

Intake Structure and Pump Building Control 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 were inspecte Compliance with Radiation Work Permits and use of appropriate personnel monitoring devices

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were checked. 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 systems.

Logs and records were reviewed to determine if entries were properly made and

! communicated equipment status / deficiencies. These records included operating

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logs, turnover sheets, tagout and jumper logs, process competer printouts, and Plant Information Reports. The inspector observed selected aspects of I

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plant security including access control, physical barriers, and personnel monitorun In addition to normal working hours,_the review of plant opera-tions was conducted during the following midnight shifts, and weekends:

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March 13,~1988 11:00 AM to 4:00 PM

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March 18, 1988 2:30 AM to 6:00 AM

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March 20, 1988 2:00 PM to 5:00 PM '

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March 26,.1988 2:00 PM to 3:00 PM No unacceptable conditions were identified. Operators were alert and dis- ,

played no signs of inattention to duty or fatigu . Plant Operations Review Committee (PORC) (40700)

The 1,1spector attended several Plant Operations Review Committee (PORC) meet- .

ings. Technical specification 6.5 requirements for required member attendance  !

were verified. The meeting agenda included procedural changes, proposed changes to the Technical Specifications, and field changes to design change ,

packages. The meeting was 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 re- i view and comment. Dissenting opinions were encouraged. 'The inspector had l no furthec comment l 4. Maintenance and Surveillance (61726, 62700)

The inspector observed various maintenance and problem investigation activi- i ties for compliance with requirements and applicable codes and standards,  ;

QA/QC involvement, safety tags, equipment alignment and use Mf jumpers, per-

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sonnel qualifications, radiological controls, fire protection, retest, and reportability. Also, the inspector witnessed selected a rveillance tests to i determine whether properly approved procedures were i i "se, test instrumenta- .

tion was properly calibrated and used, technical specifications were satisfied,  ;

testing was performed by qualified personnel, procedure details were adequate,

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and test results satisfied acceptance criteria or were properly dispositione The following activities were reviewed:

SUR 5.1-18, Test of Emergency Diesel Generator EG-2A With Partial Loss of AC Coincident With Core Cooling Actuation SUR 5.1-19, Test of Emergency Diesel Generator EG-28 With Partial Loss of AC Coincident With Core Cooling Actuation l Routine Operability Surveillances of A & B Emergency Diesel Generators, April 4, 1988 No deficiencies were identifie ;

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5. Events Occurring During the Inspection 5.1 Licensee Event " ets (LERs) and Safeguards Event Reports (SERs)

(90713, 937 W 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 impl; cations. The inspector also verified that the report-ing requirements of 10 CFR 50.73, 10 CFR 73.71, and Station Administra-tive, Operating and Security Procedures had been met, that appropriate corrective action had been taken, and that the continued operation of the facility was concucted within Technical Specification Limit Electric Fire Fump Declared Inoperable Due To High Amperage 86-04 Dropped Red Circuitry Found Improperly Wired

• 88-05 Leak Path Renders Cable Vault CO2 System Inoperable 88-06 Leak Path Identified in Cable Vault Flood Barrier 88-501 Safeguards Event Report

• Event detailed in NRC Inspection Report 50-213/88-02 No unacceptable conditions were identifie .2 Premature Shutdown of RHR System (71707, 92702, 93702)

On March 10, during performance of an operating procedure checklist, the licensee identifiec' that, on two occasions during the station heatup from this refueling o/tage, the Residual Heat Removal System (RHR) was not in operation (with the station in mode 5) for more than the one hour period permitted by Technical Specifications (T5s).

Following each outage and during the heatup period, the licensee conducts a hydrostatic pressure test cf the Reactor Coolant System (RCS) at 245 degrees F. During this heatup period, all four RCS loops were unisolated, two reactor coolant pumps (RCPs) were operating, and two steam generato s were unisolated and available for service. Both RHR loops were operable with one operating in accordance with TS 3. 1.4.1. Heatup was hampered by the 'nability to complet.' ' iivert flou around the RHR heat exchangers and procedural limitations the number of RCPs whic' could be operecin In this configuration, end with limited decay heat due to the lengch of the outage, RCS heatup was three degrees per hou __

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The two instances of RHR being out of service for greater than one hour with the plant in mode 5 were on March 5 for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 2 minutes and on March 10 for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 22 minutes. In each case, the plant enterea mode 4 at the end of the time period; in mode 4 it is not required to have

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an RHR pump operatin The licensee conducted an extensive review of the circumstances which led to this violation of TS 3.3.1.4.1. That Specification was changed by Amendment 97 in November 1987. The previous TS permitted the RHR pumps to be shutdown in mode 5 while performing a plant heatup. When Amendment 97 was issued, the station operating procedures were changed to reflect the new TS, with a precaution statement added to to remind the operators of the new requirement. NOP 2,1-1, Cold Shutdown - Mode 5 to Hot Standby - Mode 3, was in use for the heatup which lasted several day The error is attributed to operations personnel failure to re-review procedural precautions during each shift turnover to ensure pro-cedural adherenc In order to prevent recurrence of events such as this, the licensee took several corrective measures including procedure changes and directing operators' attention to the violation and the necessity of reverifying precautions. The inspectors reviewed the procedure changes made and discussed this event with operations personnel. No inadequacies were identifie These events constitute a violation of TS 3.3.1.4.1. Because they were licensee-identified, appropriately identified to the NRC, appropriately corrected, of limited safety significance (low violation severity) be-cause the steam generators were available to remove decay heat, and judged to not be due to inadequate correction of a previous violation, no notice of violation was issue .3 Strike by Bargaining Unit Security Officers (92709, 92710)

At midnight on March 19, the station bargaining unit Security Officers went on strike af ter rejectir.g a contract renewal of fer from the security contract ,r, Burns International . In anticipation of a possible strike, the licen:,ee had implemented a contingency plan early on March 19 by placing security suparvisory personnel on shift in security officer position The picketil.g was orderly ar.d without incident, and had no effect on plant operations. There was local media coverage of the strik A tentative agreement was reached on March 21, and the security officers returned to their posts on March 2 The contract was ratified on March 2 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .- _ _ _ _ _ __ _ _ _ _ _ . _ __

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Prior to the strike, the inspectors reviewed the contingency plan with security management. During the strike and associated turnover periods, the inspectors observed selected security activities. Additionally, an NRC inspection was performed of the station physical security program implementation during the period March 14 through 17, 1988. Contingency planning was also reviewed during that inspection; details are contained in NRC Report 50-213/88-07. No deficiencies were identified in activi-ties concerning strike contingency planning in either inspectio .4 Reactor Trips on Spurious High Startup Rate (71707, 93702)

The reactor was made critical at 12:55 a.m., March 19 to complete the refueling / maintenance outage and commence low power core physics testin Control rod outward motion was found blocked at 1:57 a.m.; with no pro-gress in solving the oroblem a reactor shutdown was started at 5:24 Inward motion of control rods was not affected. At 5:30 a.m., a reactor trip occurred on high startup rate of one (channel No. 22) of the two intermediate range nuclear instrument channels. At the time, reactor power was in tse intermediate range at 2E-11 amp After extensive investigation the licensee determined that an annunciator relay created an electrical noise spike in the intermediate range channe The problem was corrected by the addition of a suppression diode across the relay coil. A failed relay was found to have caused the outward rod motiun block and was replace The reactor was again made critical at 7:38 p.m., March 2 A second automatic reaccor trip occurred at 6:24 a.m., March 22. Again the trip was initiated by the high start up rate protective circuit of Incermediate Range channel No. 22. Reactor power was in the intermediate range at 2E-9 arnps during reactivity temperature coefficient measurement The plant had been maintained in a very stable configuration prior to the trip to support of the isothermal temperature measurements. Power was allowed to change slowly during those tests and had decreased to the point where Intermediate Range Channel No. 22 energized the source range detector high voltage. The licensees corrective actions included re-placement of components within the intermediate range instrument along with a length of electrical cable at the detector. The reactor was again mada critical at 8:00 p.m., March 22 and testing of the new low pressure tur..ne bega .5 Missed Surveillance During Physics Testing (71707, 92702, 93702)

During Low Power Physics Testing, the Special Test Exceptions of Techni-cal Specification (TS) 3.24.2 require that reactor thermal power be verified te be below 5% hourly. On March 21, while conducting physics testing, $ne licensee identified that they had missed one of these hourly surveillances by 42 minutes. Test personnel immediately verified that reactor thermal power was indeed less than 5%. Actual power was 0.02%.

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The licensee attributed this missed surveillance to personnel erro Because the plant process computer calorimetric program had not yet been fully tested and turned over, test personnel had to perform hand calcu-lations. The importance of performing these calculations on time was re-emphasized to test personnel. In the future, the licensee expects to have the plant process computer calorimetric program fully operatiena When this is the case, reactor power will be automatically calculated as necessary. Should hand calculations be required again, test personnel will be reminded of the hourly calculation requiremen There were no further instances of missed calculations during the physics test The inspector concluded that this licensee-identified event was appropriately corrected and reported, of minor safety significance, and not due to corrective action inadequacy on a previous violatio Therefore, no Notice of Violation was issue .6 Large Breat LOCA ECCS Flow Delivery Error (40701, 71707, 93702]

The NRC Resident Inspectors were informed of the discovery of a pctential error in the Emergency Core Cooling System (ECCS) flow delivery calcula-tions which were used as an input parameter in the Large Break Loss of Coolant Accident (LB-LOCA) analysis. At the time, the reactor wa; cri-tical in Mode 2, with power less than 5%.

The licensee has be:n performing an in-house LB-LOCA analysis. As part of that effort, the licensee recalculated the ECCS flow delivery rates with single active components as the input to the analysis. A discre-pancy was found between the results of their calculations and those pro-vided by the Nuclear Steam Supply System vendor, Westinghouse, in t' e original ECCS performance analysis docketed in 1971. Both the licensee and Westinghouse calculated ECCS delivery flow rates with assumed active component f ailures (that is a failed emergency "iesel generator, injec-tion valve, high pressure or low pressure injection pump). The licensee determined that limiting flow would occur with a failed shut Low Pressure Safety Injection isolation / injection valve at 866 lb mess /secon For the same failure ;lestinghouse calculated 1140 lb mass /second. The limiting Westinghouse case had been calculated as the failure of an emergency diesel generator (1060 lb mass /second). At the end of the inspection period, *.he licensee and Westinghouse were working to resolve the diff.renc However, since the case with one failed emergency gene-rator was ast med to be limiting and is the basis for the current LB-LOCA licensing ant > sis, it was conservatively assumed that the later, licen-see calculatisas were correct. Westinghouse was also recalculating the l LB-LOCA analysi The results are expected to be available during the f

week of April A conf once call was conducted between the licensee and the NRC - NRR at 4:10 p.m., March 24 The licensee explained the issue and was asked to provida a Justification for Continued Operation (JCO) at up to 40%

power prior to entering Mode That document was provided early in the evening of March 25 and was Dased on sensitivity studies which concluded . .

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that a delivery rate at 80% of that previously assumed would result in post-LB-LOCA peak clad temperatures (PCT) of about 100 degrees above the Interim Acceptance Criteria (IAC) limit of 2300 degrees F (for stainless steel clad). To reduce PCT to within the IAC limit, the sensitivity studies also indicated that peak Linear Heat Generation Rate (LHGR)

needed to be reduced frora 14.3 to 13.3 kw/ft. Changes in LHGR also re-quired re-evaluation of axial offset limits, however a change in limits at 80% power and below were not recuire The licensee began a power increase and entered Mode 1 (5%) at 3:45 p.m.,

March 2 Turbine and generator voltage regulator testing were completed and plant power was held at 30% on March 28 to allow chemical cleanup of the steam generator secondary. At a 10:00 a.m., March 30 meeting the NRC - NRR concurred with the licensee's JC0 for up to 80% of rated powe Power was increased to 79% at 4:00 p.m., March 3 During this period the NRC Resident Inspectors monitored plant operations to insure that tne limiting conditions set by the licensee and agreed to by the NRC were implemente There were no unacceptable conditions identifie . Periodic and Special Reports (90713)

Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewed. This review verified that the reported in-formation was valid and included the NRC required data; that test results and supporti a information were consistent with design predictions and performance specifications; and that planned corrective actions were adequate for resolu-tion of the p;oble The inspector also ascertained whether any reported information should be classified as an abnormal occurrence. The following periodic report was reviewed:

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Monthly Operating Report 88-02, February 1-29, 1988 No inadequacies were identifie .0 Inopecable Dropped Rod Circuitry (37701, 93702)

On February 25, during additional post modification testing of the dropped rod circuitry, the licensee identified that the circuitry for rods 12 and 13 in Control Bank B was improperly wire In this configuration, had one of 1 these rods dropped, there would not have been an initiation of the P3 permis-sive. Energi ation of the P3 relay initiates a turbine load runback to 80%

power and a block of automatic control rod withdrawa Earlier in the refueling outage, the licensee had made a minor modification to delete the bypass feature of the dropped rod circuitry. This bypass fea-ture allowed the dropped rod circuitry to be bypassed when Bank B rods were below a preset height to avoid inadvertent actuation of the P3 permissiv This feature was determined to not be necessary because, when Bank B is in-serted into the core, power level is generally below 80% and, when performing

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a startup, control rods are withdrawn manually, not automatically. Post modification testing for this design change was conducted under SPL 10.2-48, Functional Verification of PDCR 924 (Dropped Rod Block Removal). This test verified circuit operability by individually energizing each bank relay (to simulate control rods inserted into the core) and verifying that the P3 relay was energize During test development and performance, test personnel ques-tiened how~ individual rod bottom signals feeding into the P3 function were verifie I&C personnel review of the rod position calibration procedure SUR 5.2-21, Rod Position Calibration, identified that the contact which feeds the rod bottom signal into the P3 function was not periodically verified. As a result, additional testing was performed on the dropped rod circuitr .

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The additional testing simulated an individual dropped rod by systematically bypassing all but one bistable associated with each bank to initiate relay operation to energize the P3 relay. On February 25, it was identified that, had rods 12 and 13 of Bank B dropped, there would not be the associated energization of the P3 relay. Complete testing of all 45 control rods iden-tified no further inadequacies. Rcpairs and retesting were successfully per-formed on the circuitry for rods 12 and 13 of Bank The licensee has attributed the error in the wiring to pear post modification testing after modifications made to the circuitry several years ago. Current post modification testing practices have greatly imoroved and have identified several errors from previous modifications during this refueling outag Additionally, the Dropped Rod Analysis in the Updated Final Safety Analysis Report does not take credit for the turbine load runback or dropped rod sto Therefore, P3 does not have a safety-related function. The licensee had ad-ministratively downgraded the P3 permissive to control grade but maintains it to provide optimal station response to a dropped rod event. The inspector had no further questions on this matte .0 Systematic Evaluation Program Findings [92719]

8.1 Topic 11-3.8, Flooding Potential and Protection Requirements This topic addresses potential floods and required protection; including the type of protection, equipmert necessary, and implementation of the protection program. The portion of this topic which had not been veri-fied complete before this inspection period is the licensee's emergency procedures to initiate flood protection i..easures. This also has been identified as Integrated Plant Safety Assessment Report (IPSAR) Item e 4.1.4, Emergency Procedure In previous inspections, this item had been verified complete and close However, after the flood of June 1984, this issue was reopened when the licensee elected to make a major procedural change. Specifically, Ab-normal Operating Procedure (AOP) 3.2-24, Flooding of the Connecticut River, had required that protective measures be initiated when river water level reaches 16 feet mean sea level (M5L). (Site grade elevation

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is 21 feet MSL.) The licensee wanted to change the action level to re-quire protective measures when the river reaches 16 feet MSL only if it is predicted to rise above 21 feet MSL. A review of the procedure, the area flooding history, and flood prediction capabilities was conducted by the NRC. It was determined adequate to initiate flood protection measures when the river reaches 16 feet NSL and there is a predicted crest above 19 feet MSL. NRC also recommended several additional changes to the A0P. The inspector reviewed the current revision (Revision 7)

of A0P 3.2-24 and verified that necessary changes have been incorporate This item is close . t. Topic III-3.C, Inservice Inspection of Water Control Structures This topic required an Inservice Inspection Program for Water Control Structures (WCSs). WCSs are used for flood protection or emergency core cooling systems. The program ensures that WCSs which are part of the ultimate heat sink are available at all times under normal and accident conditions. This issue also is IPSAR Item Final NRC review of this program was completed with the issuance of the Safety Evaluation, dated April 7, 1987. At that time, the licensee had not fully implemented commitments to assign a qualified engineer to supervise this program, maintain a central file for program inspection results, and perform a review of Regulatory Guide 1.127, Inspection of Water Control Structures Associated With Nuclear Power Plants, to ensure that Hacdam Nec'< complies to the extent possibl The inspector discussed the implementation of these commitments with the assigned engineer. Early in 1987, the licensee hired a Civil Engineer to oversee the WCS Inspection Program. The actual inspections are per-formed by this engineer and qualified corporate engineers. The licensee also reviewed Reg Guide 1.127 and elec.ted to develop several procedures to implement its recommen6 tion Fpecifically, ACP 1.0-51, Water Con-trol Structures / Flood Protection Barrier Inspection Program, was estab-lished as the controlling document for the WCS Inspection Program. It delineates the personnel responsibilities and identifies 1spection pro-cedures for WCS/ Flood Protection Barriers. The WCS Inspection Program is implerented by several Engineering and Preventive Maintenance Proce-dures. The central file for the completed inspection and maintenance i procedures is the Nuclear Records Department onsite.

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The inspector reviewed the completed inspection and maintenance proce-dures from this refueling outage and verified that any discrepancies identified were adequately addressed through the Nonconformance Report process. The inspector also verified that the corporate engineers per-forming these inspections have appropriate, documented qualifications.

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Based on this review, this item is closed.

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9.0 Low Temperature Overpressure orotection (LTOP) (25019)

Reactor vessel integrity can be threatened by sudden pressure increases com-bined with low reactor coolant system temperatures. This issue has been identified as Unresolved Safety Issue (USI) A-26: Reactor Vessel Pressure Transient Protection for Pressurized Water Reactors. Numerous reviews of licensee actions to resolve this issue have been performed. These are docu-mented in NRC Inspection Reports 50-213/85-21, 86-01, 87-06, 87-08, 87-21, and 87-27. Throughout these inspections, reviews have been made of the lic-ensee's LTOP design reviews, station drawings, administrative controls, sur-veillance and operations procedures, operator training, and Technical Speci-fication (TS) amendments. During this inspection period, the insptetor veri-fied that station procedures reflect the most recent TS amendments. Based on these reviews, the licensee has adequately addressed the concerns raised by USI A-2 . 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.