Text

Insp Rept 50-213/97-03 on 970408-0707 & 0805.Violations Noted.Major Areas Inspected:Operations,Engineering,Maint & Plant Support
	ML20217J406
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	10/09/1997
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20217J396	List: IR 05000213/1997003; ML20217J392; ML20217J398;
References
	50-213-97-03, 50-213-97-3, NUDOCS 9710210072
	Download: ML20217J406 (61)
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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No.: 50 213 License No.: DPR-61 Report No.: 50 213/97-03 Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 061410270 Facility: Haddam Neck Station Location: Haddam, Connecticut Dates: April 8 - July 7,1997, and August 5,1997 Inspectors: William J. Raymond, Senior Resident inspector John H. Lusher, Emergency Preparedness Speclatist Approved by: Richard J. Conte, Chief, Projects Branch 8 Division of Reactor Projects

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EXECUTIVE SUMMARY Haddam Neck StSon NRC Inspection Report No. o0-213/97 03 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a period of resident inspection;in addition, it includes the results of announced inspections by a regional specialists in the area of emergency preparednes Plant Operations Operators performed well to rnonitor the status of operating plant equipment, and those systems in a layup condition. Operators showed good regard for and use of plant procedures during the conduct of evolutions. Spent fuel pool cooling system remained in service to maintain SFF i.emperature below the limits of TS 3.9.15. There was good operator attention to pool conditions, and monitoring of system status. Operators and support personnel performed well to assure adequate SFP cooling during operation on a bypass during the period the SFP cooling system was inoperable. Plant procedures reviewed were generally satisf actory and adequate to accomplish the intended purpos While procedures were generally good and improved, NRC findings show varied results in procedure quality, and suggests the need for enhanced review of activities being undertaken for the first time, initiatives to improve the effectiveness of PORC reviews are appropriate to assure consistent implementation of programs and processes for operating the " spent fuel pool nuclear island" and safely conducting decommissioning,

, Maintenance:

Plant personnel performed routine and non-routine activities wellin resolving problems, including the special test of a new check valve in the SW system, the calibration of radiation effluent monitors, and troubleshooting problems in diesel fire water pump, the EG-2A shutdown circuit and the seismic monitors. Plant personnel completed routine tests of plant equipment well, recognized degraded conditions, and initiated actions to complete troubleshooting and repairs. Good work controls were noted, including good pre job briefs, control of tagouts, adherence to work packages and work plans. There was good coordination with and support from health physics and engineering supp personne Workers demonstrated good skills and knowledge of systems under test ter repai The persistence of some problems remains a concern, such as the problems on the diesel fire pump, and the EG 2A shutdown sequence. A significant exception to good performance in the maintenance area was an operator's f ailure to remove the jacking tool during a routine surveillance test of EG-2A. The error was significant in that it impacted the aveilability (by 7 days) of the preferred emergency power source for the spent fuel pool cooling system. The f ailure to follow PMP 9.1-31 during the May 21 test of EG 2A was a violation of TS 6.8.1, and a repeat occurrence of a similar event that occurred in November,1996, which was the subject of an NRC enforcement action. Other recent findings concerned the occurrence of personnel errors and the failure to follow procedures over a wide spectrum of plant activities. The recurrent f ailure to follow procedures during ii

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diesel testing, along with the protracted response to the broader concern of human performance issues, was an example of inadequate corrective actions and was contrary to 10 CFR 50 Appendix B, Criterion XV Enaineerina:

In general, engineering provided timely and effective support to plant operations during the period to address several ist,ues important to shutdown operations and for transitioning to the decommissioning mode. Assessments and safety evaluations (SE) completed in support of the plant activities were technically sound and adequately documented. The revised process to perform safety evaluations per 10 CFR 50.59 was enhanced to include criteria to address 50.82 and to address past weaknesses. Training on the new process was of high quality. Some procedure clarity issues were identified which if left uncorrected could result in safety evaluation process weaknesses. Mixed performance was noted in the support to ship new fuel offsit Effective engineering support was also noted to identify the corrosion in the SW system, to assist in the prompt assessment and repair of degraded piping, to evaluate the significance of the ongoing corrosion relative to continued use of the SW system, and to provide recommendations for future actions. Corrective actions, upon discovery of the B SFP heat exchanger being 50% fouled, were comprehensive to address the operational, design an licensing issues. The NRC will follow actions to revise TS 3.9.15 snd to complete the short term and long term corrective action A SW system design discrepancy made the SFP cooling system inoperable, and constituted operation outside tho design basis from 1975 to March 1997. The NRC staff is exercising l

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discretion on this violation. However, once identified by CY in August 1996, untirnely corrective actions by plant engineering to promptly address the issue was an example of inadequate corrective actions and was contrary to 10 CFR 50 Appendix B, Criterion XV The related ACR thoroughly identified underlying causes along with corrective actions for'

the delay in resolving the service water problem but the corrective actions were not comprehensive for all causes, e.g., staff turnover proble A change to the facility as described in UFSAR 15.2.9 in dealing with a dedicated operator for a feedwater regulating valve deficiency created an unreviewed safety question and was contrary to 10 CFR 50.59. The NRC staff exercised discretion for this issu Plant Suonort:

A chemist was injured when a radioactive sample vial broke on June 2, resulting in a spill-of radioactive material on the floor of the laboratory. Poor initial technicians' response resulted in the spread of contamination to other individuals in the lab and on the Ir-e atory floor areas. Once recognized, the situation was controlled and the staff followup was good to minimize the further spread of contamination, administer first aid to the injured technician, perform a dose assessment, and clean up from the event. Following the loss of security equipment on June 13, the security force responded well to set up and maintain compensatory measures. The response by support personnel was timely and effective to identify and correct the cause of the proble iii

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-Significant progress was made to address discrepancies in the calibration of the effluent radiation monitors, with four of five monitors returned to service, Progress was made after the assignment of additional resources to plan, coordinate, schedule and implement an-action plan to complete the calibration. A longstanding issue with the SCANRAD computer has been resolve The licensee continues to maintain the emergency preparedness program in accordance with the 10 CFR Part 50 requirements until it submits and receives approval for the shutdown defueled emergency plan.- Overall, the training was effective, based upon the review of lesson plans and examinations and by player performance demonstrated in the table top scenarios. Additionally, the crews also demonstrated their ability to make appropriate PARS using the new EPlP 4428G " Protective Action Recommendations" procedure, which has been developed, approved, and will be implemented in the near future. The critiques by the participants and controllers were thorough and self critica Therefore, the violation dealing with ability to correctly classify emergency events (50 213/96 07 01)is close iv

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TABLE OF CONTENTS EX EC UTIV E S U M M AftY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 T AB L E O F CO NT E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v R EPO RT D ET AI LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Summ ary o f Plant St atus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . 1 1. O p e r a t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01 Conduct of Ope rations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 Operating Activities and Status of Operating Systems . . . . . . . . . . . . . . 1 01.2 Spent Fuel Pool Cooling and Building Ventilation (NCV 97-03 01) . . . . . . 3 03 Operations Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 03.1 Procedure Quality fer Shutdown Operations (NCV 9 7-0 3-0 2) . . . . . . . . . 5 I

08 Previous Operations Open issues (92901) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 08.1 (Closed) IFl 95 27 02, RCS Leak Rate Determinations . . . . . . . . . . . . . . - 7 08.2 (Closed) IFl 95 27 01: Daily Technical Specification Channel Checks . . . . 7 -

08.3 (Closed) VIO 97 01-02: Configuration Control . . . . . . . . . . . . . . . . . . . . 8 08.4 (Closed) URI 94 27-02: Hydrazine Release . . . . . . . . . . . . . . . . . . . . . . 8 08.5 (Closed) IFl 94-05-04: Service Water System Lineups .............. 8 08.6 (Closed) IFl 95-02 02: Diesel Tagging Error Causes Flood . . . . . . . . . . . , 9 11. M ai n t e n a nc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 M1 Conduct of M aintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 M1.1 Mainten a nc e Activitie s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 M 1.2 Surveillance Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . 12 ;

M1.3 Conclusions for Maintenance Activities . . . . . . . . . . . , . . . . . . . . . . 14 M4 Maintenance Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . 14 M4.1 - Emergency Diesel Generator Testing (VIO 9 7-03-03 a) . . . . . . . . . . . . . 14 M8 Status of Previous inspection Findings (92902) . . . . . . . . . . . . . . . . . . . . . . . 16 M8.1 (Closed) VIO 96-13-01: Diesel Run with Crank Tool Installed ........ 16 M8.2 (Closed) URI 96-06-05: Actions to Address MIC Corrosion . . . . . . . . . . 16 111. Eng i n e e ri ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 El Condu ct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 E1.1 Engineering Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 E1.2 Categorization of Plant Systems for Decommissioning . . . . . . . . . . . . . 17 v

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E2 Engineering Support of Faciliths and Equipment ...................... 19 E2.1 Service Water System Waterhammer (VIO 9 7 03-03 b) . . . . . . . . . . . . . 19 E2.2 Service Water System Corrosion . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 E2.3 SFP Heat Exchanger Fouling (URI 97 03 04, URI 97-03 05) ......... 22 E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . 26 E3.1 Procedures to Conduct Safety Evaluations per 10CFR50.59 (37001) . . . 26 E3.2 Design Discrepancy Inadequate SFP Pump NPSH (NCV 97-03 06) . . . . 31 E3.3 Inadequate Safety Evtluation Operator Actions (NCV 97 03-07) ..... 33 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 E8.1 Status of Previous inspection items . . . . . . . . . . . . . . . . . . . . . . . . . . 35 E8.2 Escalated Enforcement Management, Design and Corrective Action Issues............................................... 36 E8.3 Review of Licensee Event Reports (LERs) . . , , . . . . . . . . . . . . . . . . . 39 I V. Pl a n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . . . . . 43 R1,1 Spill of Cs 137 Source Meterial and Personnel Contamination . . . . . . . . 43 R2 Status of RP&C Facilities and Equipment ........................... 44 R2.1 Inoperable Ef fluent Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 P3 EP Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . -. . . . 45 P4 Staf f Knowledge and Performance in EP . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 i

P8 Miscellaneous EP Issues . . . . . . . . . . . . . . . . . . . . . '. . . . . . . . . . . . . . . . . 47 S1 Conduct of Security and Safeguards Activities . . . . . . . . . . . . . . . . . . . . . . . 48 S1.1 Loss of Security Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48-V. Management Meetings . . . ....................................... 49 X1 Exit Meeting Summary - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49

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REPORT DETAILS Summarv of Plant Status The Haddam Neck reactor remained defueled with the spent fuel stored in the spent fuel pool. The licensee maintained controls to assure the pool conditions were stable and the fuel adequately cooled. The spent fuel cooling system was made operable on April 17 following modifications to address postulated water hammer in the service water cooling lines, and repairs to address corrosions induced defects in the service water piping. There were no significent changes in the plant systems not required to support the spent fue The licensee continued activities needed to prepare the post shutdown decommissioning activity report and to plan for plant decommissionin Changes in the licensee organization during the period .ncluded the appointment of M Richard Sexton as the Health Physics Manager effective on April 9,199 NRC inspections during the period included the reviews by the resident inspector of post operating activities, and the preparations for decommissioning. A special review was conducted of the process for performing safety evaluations per 10 CFR 50.59. Resident reviews were supplemented by inspections by region based personnel in the areas of j radiological controls and emergency preparednes NRC activities at the site included the following visits and plant tours: on April 1213, by Mr. John Rogge, Chief of Reactor Projects Branch #8; on April 29 30 and June 23-27, by Messrs. Mort Fairtile and T. Fredericks of the NRR Office of Decommissioning Projects; and, on May 12, by Mr. John White, Chief of the Radiation Safety Branch. NRC personnel also attended meetings of the Citirens Decommissioning Advisory Committee on April 28,-

May 29, and June 30,1997, l. Operations 01 Conduct of Operations'

Using Inspection Procedure 71707, the inspector conducted periodic reviews of plant status and ongoing operations. Operator actions were reviewed during 1 periodic plant tours to determine whether operating activities were consistent with the procedures in effect, including the alarm response procedure .1 Operatina Activities and Status of Operatina Systems insoection Scope (71701)

The purpose of this inspection was to review the licensee activities to maintain the plant in the defueled condition, and to prepare for decommissioning activitie '

Topical headings such as 01, M8, etc., are used jn accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic . . .

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2 Observations and Findinas Operating activities during this period included those operations needed to maintain stable plant conditions with the reactor defueled, to maintain adequate level in the spent fuel pool, and to assure adequate cooling of the spent fuel. Operators implemented procedure NOP 2.0-4, Layup of Systems and Components, which provided the guidance for the general alignment and preservation of systems and components during extensive plant shutdowns and outage h The licensee maintained one SW pump operating (pumps were rotated to equalize the run times) to support spent fuel pool cooling. One component cooling water pump, and one of two turbine building closed cooling water pumps remained in >

service. The normal and emergency electric distribution system remained in service (except for periods of testing and repair) to support spent fuel pool cooling and plant operations. A number of control board annunciccors (approximately 100) !

remained illuminated all the times with the plant shutdown and defueled. The inspector reviewed the status of each annunciator and the reason it existed, and determined that the conditions were normal for the shutdown and defueled condition of the plant. Operator actions in response to off normal conditions were reviewed and were found to t'e consistent with the respective annunciator procedur The inspector observed operator actions for several activities during the period, and-reviewed operator adherence to procedures. The operating activities observed included: actions in April to monitor flooding of the Connecticut River, AOP 3.2-24; alignment of SFP cooling system following maintenance in April, NOP 2.101;

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- transferring, processing and discharging water from the borated waste storage

- tanks and the boric acid mix tank in May, NOP 2.6-3; monitoring and responding to high plant temperatures in June, NOP 2.291; and, monitoring plant status in the defueled condition throughout the period per SUR 5.1-0A. The operators also monitored air quality conditions to follow restrictions in testing on site diesel generators due to new clean air regulations imposed in Jun _Qonfiauration Control - Taaaina The purpose of this inspection was to review the licensee process to control the physical configuration of the plant. Licensee actions to issue and/or remove tags under the following clearances were reviewed: 96-748,96 877,96-843,97 189, 96 658,96-1100,96-1102 and 97-169. This review included the implementation of the tagging process during the conduct of work activities, and the control of systems removed from service due to plans to decommission the plant, Conclusions Operators performed well during the period to monitor the status of operating plant equipment, and those system in a layup condition. Operators showed good regard for and use of plant procedures during the conduct of evolutions. The

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implementation of the tagging program to control of the plant system configuration was acceptabl '

01.2 _ Soent Fuel Pool Coolina and Buildina Ventilation (NCV 97-03-01)

a.- Insoection Scooe (71707)

The purpose of this inspection was to review the licensee activities to monitor the status of fuel stored in the spent fuel pool and assure the adequate cooling of spent

. fuel, Observations and Findinag The spent fuel ventilation and cooling systems remained operating per normal operating procedures (NOP) 2.10-1 and 2.15 3. The spent fuel pool cooling system remained in service with at least one heat exchanger and one pump aligned to the pool. The licensee maintained pool temperature below the limit of 150 F per-technical Specification (TS) 3.9.15; the temperature was in the range of 80 to 95

degrees F when cooling was supplied by the normal service water (SW) system, and in the range of 95 to 110 degrees F when operating on the alternate cooling

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system (fire hoses). ' Spent fuel pool temperature increases while operating on the normal service water system followed the trend in temperature of the Connecticut Rive The SW side of the SFP cooling system was maintained using either the normal SW piping or through a temporary bypass per NOP 2.24-3 during times when the normal

- cooling lines were not availablo. The bypass included the use of fire hoses to-provide service water to the heat exchangers. The bypass was in service at the -

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start of the inspection period and operated until the licensee completed piping repairs, modifications and test activities on April 17 to address corrosion and design issues in the SW lines to the SFP heat exchangers. The licensee exited the action statement for TS 3.9.15 at 3:13 pm on April 17 when the normal SW supply was -

restored to the SFP heat exchangers. Sections E1.1 and E1.2 describes licensee actions to address the SW deficiencie The inspector reviewed licensee activities to assure compliance with the following

. Technical Specifications (TS): TS 3.9.11, SFP Water Level: TS 3.9.12, Spent Fuel Building Air Cleanup System; and, TS 3.9.15, SFP Cooling. There were no activities during this period involving the movement of fuel or heavy loads over the spent fuel pool. The licensee conducted routine surveillance of the spent fuel pool '

and building, which included the tours by the nuclear side operators once each shift per SUR 5.1-0 A deficiency in procedure NOP 2.10-1 was identified by an operator on April 17 while restoring the SFP cooling system to the normal lineup after removal of the temporary bypass. While following the instructions in Attachment 8 to shift to a

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desired alignment with the A pump operating with the B heat exchanger, the A pump (P21-1 A) began to cavitate as the B heat exchanger was placed on line in

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parallel with the A heat exchanger. The operator promptly recognized the adveise condition, and immediately terminated the cavitation by closing the valve on the discharge of the A heat exchanger (SF V-813) and throttling the valve on the discharge of the B heat exchanger (SF V-851). After consulting with the Shift Manager, the operator restorsd the SFP alignment with the A pump operating with the B heat exc! ranger pending a review of the event and revision of NOP 2.10-1 to address the deficiency. The operators issued temporary procedure change TPC 97-102 to add a step to have the operator throttle the discharge valve on the operating SFP cooling pump when opening the B heat exchange outlet valv The operators submitted adverse condition report ACR 97194 to describe the above procedure discrepancy and to assure plant engineering reviewed the event to verify cavitation was expected under the conditions existing in the SFP (pool temperature was 107 F and both suction paths from the pool to pump P21-1 A were open via valves SF-V 810 and 811). Licensee design engineering concluded that pump cavitation was expected under the high flow conditions that existed with the B heat exchanger discharge valve full open. The inr7ector independently confirmed this conclusion by review of the pump vendor operating curve for both SFP pumps (Gould curve dated 2/10/61 and Dean Brothers Pump curve dated 12/10/62). See Section E3.1 of this report for further concerns related to the design basis of the SFP pumps and adequate net positive suction head (NPSH).

Licensee engineering noted degradation in the heat capacity of the inservice spent-fuel pool cooling system heat exchanger E-5-18. The licensee took timely actions to revise SUR 5.1-08 on July 1,1997 (TPC 97-180) which identified additional hetat exchanger parameters for the operators to monitor for heat exchanger fouling on a shiftly basis. The inspector verified the operators implemented and trended the heat exchanger performance in accordance with the new requirements. No inadequacies were identifie The inadequate procedure to align SFP cooling was an example of violation of 10 CFR 50, Appendix B, Criterion V. This non-repetitive, licensee identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the enforcement policy (NCV 97-03 01).

c. Conclusions In summary, the spent fuel pool cooling system remained in service to maintain SFP temperature below the limits of TS 3.9.15. A procedure deficiency caused insufficient NPSH in the operating SFP pump while restoring pool cooling to the normal alignment. The operator responses were good to take prompt actions to protect the SFP cooling pump, iaentify the necessary procedure changes, and to involve plant engineering in the review of an adverse condition. The adequacy of plant procedures was a previously identified concern that remains under NRC review (see Section O3.1). Licensee actions to trend fouling of the B SFP heat exchanger remains under NRC review (see Section E2.3).

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03 Operations Procedures and Documentation 03.1 Procedure Quality for Shutdown Ooerations (NCV 97-03-02) Insoection Scone IP 42700 The purpose of this inspection was to review plant procedures governing activities to support testing, maintenance and operation of plant in a shutdown and defueled condition. The review also included procedures used to begin the transition from an operating status to the decommissioning process, Observations and Findinas This inspection was performed in response to past NRC concerns in which some procedures for shutdown operations were inadequate. The inspection was performed by inspector review and walkdown of selected procedures, and by followup of events that occurred during the period. The procedures covered by this review are listed in Attachment 1. The following observations were made based on the procedures reviewe Procedures reviewed were generally good and adequate to accomplish intended purpose. Examples of good quality in plant procedures included the newly defined processes in ENG 1.7-156 to classify plant systems for decommissioning, and ACP-1,2-2.42, with provided a detailed and comorehensive description of the process for completing safety evaluations per 10 CFR 50.59. However, the inspector noted examples were procedures were weak:

(1) NOP 2.101 (SFPCS) had a valve lineup / sequencing problem that resulted in cavitating the operating SFP cooling pump on April 1 (2) ACP 1.2-2.42 (50.59s) contained several clarity issues which, if left uncorrected, could result in inadequate safety evaluations or the failure to conduct SE's per 50.59 and 50.82. The weaknesses involved the need to clarify procedure guidance regarding margin of safety, malfunction of equipment, the definition of change, the increase in probability and the handling of temporary procedure change (3) SNM 1,4-20 (new fuel) contained weaknesses identified by the PORC that required clarification to assure that the new fuel could be shipped in accordance with the Certificate of Compliance. One deficiency identified by the NRC irivolved the failure by the plant staff to fully integrate consideration the conditions within the certificate of compliance while developing the procedur While the quality of prccedures was generally good, the above findings show mixed quality in some procedures. The findings also point to weak oversight and reviews by PORC. The NRC staff's concern for the effectiveness of PORC oversight was recently highlighted by the deficiencies found in the procedures used to calibrate the

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radiation monitoring systern (reference Inspection 97 02). Further examples were noted during this inspection of weaknesses in PORC administration or oversigh The exar7ples included:

(a) PORC reviews in a meeting on May 29 relative to the calibration of RMS 22 per SUR B.%0. The PORC missed an opportunity in several places in the procedure (steps 5.2, 6.1.11,6.4, 6.4.2,6.4.9) to add details that would have strengthened the program by reducing the reliance on skills of staf (b) PORC reviews in a meeting on April 14 relative to the test of check valve 963 per SUR 5.7-217 Rev 1. Similar performance was noted in the PORC reviews of SNM 1.4-20 during the period of July 1 10. The PORC did a good job to critique the procedure methods, and to followup on comments for resolution. However, the PORC appeared to be too involved in writing the procedures and without focusing on the bigger picture to review for unreviewed safety question (c) Several recent adverse condition reports (ACRs) on PORC administration:

(1) TS administrative violations that were not reviewed by PORC (ACR 97-187). The ACR coordinator had a system for tracking and bringing these items to PORC for review as a group - this practice was stopped. The licensee will assure it is part of the new ACR process and that ACRs that were missed will be brought to POR (2) The PORC function was subverted when a proposed technical specification change request was brought to the NSAB before the package was reviewed and approved by the PORC (ACR 97183),

(3) A finding by the quality assurance organization (audit CY 97-A06-02)

that the licensee had failed to maintain the correct composition of the PORC throughout the transition of the site organization (ACR 97-393). The deficiency included the lack of continuity in the expertise for the disciplines of reactor engineering (starting in June 1997) and instrument and controls (starting in February 1997). Short term actions were taken to provide the necessary expertise, including deferring PORC action on certain matters until the required expertise was availabl The PORC composition deficiency was an example of violation of TS 6.5.1.2. This non-repetitive, licensee identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the enforcement policy (NCV 97-03-02).

These PORC issues were discussed in meetings with the Unit Director periodically during the inspection. The licensee noted the inspectors comments and described planned actions to address the effectiveness of PORC reviews. The actions )

included plans to train PORC members on review techniques, which included a

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video presentation training on topics such ss " group think." The licensee also intended to use the QAS group in additional assessments of this area, Conclusions The quality of procedures reviewed was generally good, but some findings show varied quality in some procedures. The findings also show the need to address the adequacy of PORC oversight, administration and reviews. The licensee initiatives to improve the effectiveness of PORC reviews are appropriate to assure consistent strong implementation of programs and processes for operating the nuclear island and safely conducting decommissionin Previous Operations Open issues (92901)

08.1 (Closed) IFl 95 27-02. RCS Leak Rate Determinations This inspection item concerned the adequacy of the alternate methods used to measure leakage from the reactor coolant pressure boundary, and specifically, whether leakage determinations using the containment radiation monitors met the =

criteria for quantifying leakage. Af ter shutting down and defueling the reactor in 1996, the requirements of Technical Specification 3.4.6.1 were no longer applicable (below Operational Mode 1 through 4), and the RCS leakage detection systems were not required to be operable. Due the reactor has been permanently defueled -

and the reactor coolant system will no longer be operated with nuclear fuel, the question of the accuracy of the RCS leak rate determinations using the' alternate methods was no longer relevant. This issue is close .2 (Closed) IFl 95-27 01: Daily Technical Soecification Channel Checks This item concerned the ability to perform channel checks as defined in technical specification 1.5 for the following instruments: flood alarm detectors, stack flow, and power operated relief valve (PORV) block valve position indication. TS definition 1.5 states that the channel check is a qualitative assessment of channel behavior during operation by observation. The determination shallinclude, where possible, a comparison of the channelindication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. The operability of the PORV block valves was important during operation in Modes 1 through 4, and is no longer relevant to the plant safety with the reactor defueled and the vessel head remove The licensee identified alternative means to verify the proper performance of the stack flow and flood alarm ins'rumentation circuits. The alternative methods available to plant operators were described in a memorandum ODM 97-112 dated June 9,1997. Each area for which flood detection is provided is monitored by two detectors that alarm on the main control board panel FA. Additionally, potential leakage into the primary auxiliary ouiloing is drained to either the aerated drains tank (ADT) or to sump pumps that pump to the ADT. The ADT levelindication and high

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level alarms provide an indirect backup means to check the operability of the ficod detection circuit Similarly, there is only one channel to monitor total plant stack flow rate. However, the following alarms would provide indirect methods for the operator to verify the operability of the stack flow rate: PAB supply and purge air flow low, total stack flow low, and fans not running. Any of the alarms would alert the operator to check for problems with the PAB ventilation system,incluWng the stack flow rat Finally, the following localindicators could also be used to check the operability of the total stack flow rate: flow to the stack as measured on HIC 1101; PAB exhaust flow per HIC 1102; containment purge flow per HIC 1103; and, waste gas exhaust flow per F11105 Based on the above, the inspector concluded that the licensee had alternate means to check the operability of stack flow rate and flood detection circuits, and that the alternate methods were suitable to meet the technical specification definition of performing channel checks to the extent possible. TFM item is close .3 (Closed) VIO 97 0102: Confiauration Control The licensee responded to inspection 97 01 by letter dated June 6,1997 (CY 97-058), which described actions to address each deficioney noted in the inspection, and to reduce personnel errors and improve human performance. While the licensee completed actions as specified la the June 6 submittal, the corrective actions were not effective in preventing a recurrence of operator errors while testing the diesel This matter is discussed further in Section M4.1 below. NRC concerns in this area will be tracked as part of Inspection item 97 03 03. This item is close .4 (Closed) URI 94 27 02:Hvdrarine Release This item was open pending the completion of an investigation of the source of the hydrazine leak into the auxiliary building. The root cause investigation was completed es part of the followup to PIR 95-09, and was approved by the PORC on April 27,1995. The licensee found that the hydrazine leak occurred me to mispositionod valves from in the ventilation connection to the PAB process plenu The cause of the mispositioned valves was not determined. There were no subsequent leaks of hydrazine during plant operations. The hydrazine originated from the steam Jet air ejector exhaust, which is no longer a source with the plarit permanently shutdown. This item is close .5 (Closed) IFl 94 05-04: Service Water System Lincuos This item concerned an error in filing procedure changes in the control room, which resu;ted in the completion of SUR 5.1 152 with a page missing. The corrective actions were documented in response to PIR 94 057, and included an audit of all workin;; copies of procedures in the control room and counseling operators to ensure all pages of a procedure are present when performing a task. No further similar discrepancies were subsequently noted. This item is close _- --

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08.6 (Closed) IFl 95-02-02: Diesel Tanninn Error Causes Flood This item was also reported in LER 95 02. The licensee responded to the violation by letter dated May 3,1995. Previous NRC inspections (Inspection 95 02)

reviewed licensee actions to seal the floor drains in the diesel rooms to preclude the ,

potential for common modo flooding. The cause for the violation was found to be a '

combination of proceduto deficiencies, poor supervisoiy methods, and inapproptiate work practices. Corrective actions taken included personnel discipline, procedure changes to clarify tagging instructions, conducting briefings on the diesel spray down event, and instituting a temporary work hold on tagging pending the completion of short term actions. There were no subsequent tagging errors associated with the diesels. However, licensee corrective actions wers r'.:t totally effective to address procedure inadequacios and personnel errors. NRC followup of these concerns is tracked by inspection item 97 03 03. This item is closed, 11. Mainttnannt M1 Conduct of Maintenance Using Inspection Procedure 71707,61726 and 62703,the inspector conducted periodic reviews of plant status and ongoing maintenance and surveillanc M 1.1 Maintenance Activities Insnection Scope (62703)

I The inspectors observed all or portions of the following work activities:

EG 2A Repair (AWO 97 2028)

EG 2A Troubleshooting (AWO 97 873)

EG 2B Operational Test and Troubleshooting (SUR t;.1 178)

EG 2A Operational Test (SUR 5.1 17A, AWO 97 3742,3577)

Troubleshooting Diesel Fire Pump Start Failure (AWO 97 2083)

Condensate Receiver Tank Level (AWO 97 2066)

SUR 5.1 17B The purpose of this work on April 25 was to test and troubleshoot the operation EG 2B as it was returned to service following preventive maintenance during the period from April 21-25. The test was completed successfully to start, load and shutdown the engine in the normal sequence. However, the nuclear side operator (NSO) noted an alarm condition when the engine was shutdown indicating trouble with the fuel transfer to the skid mounted fuel tank. The Shift Manager declared EG 2B available but not operable, since the requirements of SUR 5.1 179, Attachment 3, could not be met with an alarm condition in effec During followup testing on April 30, EG 2B operated satisf actorily while the licensee investigated the operation of the fuel transfer system. One level switch on the skid mounted tank operates the transfer pump, while a separate levelinstrument

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actuates the fuel transfer alarm based on tank level at either the low or high limit The test showed that t% fuel transfer system operated satisf actorily to maintain the skid mounted tank inventory in the proper range (310 to 490 gallons per procedure NOP 2.16 68), but that the alarm setpoint had drif ted low to actuate with about 460 gallons in the tank. The licensee initiated action to recalibrate the level setpoint per PMP 9.2102 to set the high level alarm to the required point of 497 gallon Plant petsonnel performed well on April 30 to complete the testing per SUR B, and per a troubleshooting plan completed per WCM 2.14. The site technical group provided good engineering support to investigate the cause of the problem and to recommend a solutio AWOs 97 3742 and 3577 The purpose of this work on May 12 21 was to complete preventive maintenance on the skid mounted heat exchangers for EG 2A, Both heat exchangers were cleaned and inspected by eddy current testing to verify the integrity of the tubes. Although minor pitting was noted, no tubes were required plugging to be removed from servic However, degradation of the tubesheet divider plates was noted, as documented on nonconformance reports NCR 97 012 and 97 013, for the north and south heat exchangers, respectively. The degraded areas was corroded f abrication material (weld buildup) installed on the original tubesheet. The NCRs were dispositioned on May 14 by plant engineering as "use as is' with a recommendation to remove loose materials and to compensate for any gapu up to 1/8 inch deep with additional gasketing material. The adequacy of the repair was verified by the performance of additional testing per procedure ENG 1.7114, hydraulic resistance testing, which was completed on May 21, and repeated weekly; and, per ENG 1.7134A, thermal performance test, which was conducted quarterly to detect gasket f ailure or an -

adverse trend in gasket performance. The inspector reviewed the engineering basis for the NCR recommendations and identified no inadequacle AWO 97 2028 The purpose of this work activity was to repair damage to the E A main flywheel after it was damaged on May 21 when the engine was run with the jacking toolinstalled (See Section M4.1). The licensee recovered the disassembled pieces of the Jacking gear mechanism from inside the engine compartment. The licensee inspected the condition of the engine flywheel and the air start pinions. No degradation was found on the steel components of the stating air pinions, but there was scoring on each tooth of the engine ring gear (flywheel)-

and some gear teeth had rough edges and burs. The jecking tool gear was made of aluminum, and was destroyed by the even The EG.2A damage was documented in nonconformance report (NCR) 9714 and evaluated by plant engineering. The gear is made of SAE 1045 steel and has 265 teeth which are 1.5 inches wide. The engine is equipped with 4 starting air motors and is designed to start on two motors, with a starter pinion engagement of 0.75 inches with the main engine ring. The starting motors and Jacking tool work on opposite ends of the engine ring gear. The damage to the ring gear from the Jacking tool was limited to about 0.6251nches from the edge on the Jacking gear sid __ _

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Thus, the damage to the ring gear teeth would not affect the air start p nion gear engagement, since any impact was outside the area used by the alt start motor The recommended dispositiun was to repair the ring gear by filing a smooth surf ace and remove rough edges on any damaged tooth. The ring gear teeth were filed smoot The licensee installed a spare Jacking toolin the EG 2A room. The inspector observed the conditions of the damaged Jacking tools, the EG 2A engine gear and the air start motor gears, and verified the licensee actions to repair EG 2A. EG 2A was retested following maintenance and was declared operable at 6:28 p.m. on May 28. The inspector completed a post event review of the standby status of E3-2A and EG 2B, and confirmed both engines were properly aligned for standby operatio AWO 97 873 The purpose of this work on May 21 was to investigate the suspected faulty relays in the shutdown sequence for EG 2A. On three occasions during recent testing, the engine continued to run at 900 rpm versus the expected 450 rpm for the 11.5 minute cooldown period after the operator depressed the engine stop pushbutton. Relays in the shutdown circuit were monitored and operated several times during the May 21 test as the diesel was tagged out for maintenance. All relays operated satis *utorily. Following a test run of EG 2A for-t heat exchanger performance testing on May 28, the shutdown circuit operated satisf actorily. The troubleshooting did not identify the cause for the past anomalie Licensee engineering issued an evaluation of EG 2A in memorandum CY TS 97 28 The licensee concluded that EG 2A remained operable, but that an intermittent problem remained in the shutdown circuit. Engineering identified two options to address the problem: take additional measurements of the shutdown circuit during+

subsequent diesel runs to monitor the relay output when the shutdown sequence-falls to operate; or, replay relay ECRA which was deemed to be the most likely cause of the problem. Licensee actions were in progress at the conclusion of the inspection period to investigate this proble AWO 97 2083 The purpose of this work was to troubleshooting the starting circuit for the diesel fire pump, which experienced intermittent start f ailures during routine testing. While troubleshooting the unit, another problem was identified that kept the unit from starting. This problems was related to a relay in the local control panel that was not fully inserted into the circuit board. The cause of this condition was related to the mounting of the relay in its holder, which was correcte Subsequent testing demonstrated satisf actory operation of the diesel pump, but the initial problem resulting in intermittent problems in the starting sequence remained unresolved, Licensee efforts to troubleshoot this problem continued at the end of the inspection period. The diesel fire pump was considered operabl AWO 97 2006 The purpose of this work was to investigate the cause for the failure of the high level alarm on condensate receiver tank TK 351 A, which contributed to an overflow of the tank during past operations (reference Inspection 97 01, Section 01.9). The licensee found on June 4 that the high level switch was

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not wired to the alarm circuit in the main control room. This condition was contrary to plant drawings (reference Drawings 16103 32001, Sheets 6RB,6RC, ONJT and 16103 32112, Sheet 57). The licensee initiated actions included documenting the problem in adverse condition report ACR 97 280. The issue was referred to plant engineering to investigate the loss of configuration control for this non safety related circuit, and to evaluate a design change that would either restore the circuit to the intended configuration, or make a corrective updete to the drawing M1.2 Surveillance Observations lasoection Scooe The inspector observed portions of the following surveillance activities:

Inservice Test of SW CV 963, SUR 5.7 217, Rev 1 & 2, (AWO 971543)

i * Radiation Monitor Calibration, SUR 5.2 81.0, Rev 4 (AWO 95 9245)

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Radiation Monitor Calibration, SUR 5.4-49, Rev 0 (AWO 95 9245)

EG 2B Operational Test, SUR 5.1 170

FDS 2, Containment Smoke Detector Testing, SUR 5.9 40

SW Radiation Monitor (R 18) Calibration, SUR 5.4 49 (AWO 95 9245)

Radiation Monitor Testing (R18), SUR 5.2 81.6 (AWO 95 9245)

Radiation Monitor Calibration (R-22), SUR 5.4 50

Strong Motion Instrument Test, SUR 5.2 66 Observations and Findinns SUR 5.7 217 The purpose of this test was to demonstrate that the back leakage from check valve SW CV 963 was acceptable to perform the safety function to preclude water hammer within the SW system. The test method created a SW system configuration that simulated conditions following a postulated loss of normal power. A system pressure of about 17 psig (adjusted for height of the test gage)

was placed on the system downstream of SW CV 963 to assure the check valve leakage would be measured with the elevation head of water in the down stream pipin The test method was to provide system flow through the check valve to assure the header was full; create a test boundary around the check valve by closing supply and return side valves; use a hydrostatic pump on the return side of the check valve to establish test pressure; and, vent the piping upstrearn of the check valve, collecting any leakage as a measure of back flow through the check valv The first test trial (SUR 5.7 217, Rev 1) was performed on April 13, but was not successful due to an inability to adequately isolate the test boundary. Leakage past supply side isolation valves SW V 838A and SW V-239 kept the test header pressurized at the service water system operating pressure of about 80 psi Testing was suspended at Step 6.3.10 pending development of a method to prevent pressurization of the test boundar ;

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An alternate test method proposed by plant engincering on April 14 was technically acceptable, but was not approved by the plant operations review committee (PORC)

due to the plan to temporarily interrupt SFP cooling during the test. Engineering revised the test plan to allow continuous operation of the operating SW pump with no interruption in SFP cooling, and to divert SW leakage from the test boundar The inspector noted this example of poor plant material conditions, leakage past SW isolation valves, was a recurrent problem at Haddam Neck (reference in pection 96-80,9611 and 97 01) and caused difficulties in the successful completion of the tas The second test trial (SUR 5.7 217, Rev 2) wr.s performed on April 16. The measured leakage in the reverse direction through SW CV 963 was O gpm, which was well below the acceptance criteria limit of 2 gpm. The inspector independently verified that the required back pressure was maintained during the leakage measurement, and that the acceptance criteria was met. The pre job brief and coordination with all parties involved with the test were good. Test personnel were knowledgeable of the procedures and test configurations. Technical support personnel provided good engineering oversight of the evolution. The PORC oversight of test activities assured that test objectives were met while minimizing l the challenges to SFP coolin HUR 5.2 81.6 and SUR 5.4 49 The purpose of these tests on April 1 and May 6 was to conduct a new calibration on the service water system effluent radiation monitor, RM 18,in accordance with new procedures as part of the corrective actions for concerns identified in NRC Inspections 97 01 and 97 02. The new chemistry calibra:lon procedure (SUR 5.4 49, Rev 0) provided an in situ primary calibration of the moriitor using two radiation sources, Cs 137 and Co 60. The calibration was completed with source bombs containing new calibration source liquids fabricated for the to:t. The licensee also completed on March 31,1997 a new electronic calibration of the detector channel using the revised methodology in SUR 5.2-8 Test personnel were knowledgeable of the test methodology and equipment, and showed good procedure adherence. Supervisory and technical oversight for the calibration was good, as was the health physics support. The test was completed successfully to restore the effluent monitor to an fully functional status, which provided online readouts on the control room. Channel RM 18 remained inoperable per the technical specifications pending resolution of problems with the Scanrad '

compute SUR 5.2 66 The purpose of the test was to verify the proper operation of the Strong Motion monitor following a spurious alarm on May 13. Test personnel identified nothing degraded with the channel, which was retested satisf actoril During a routine test of the channel at the end of the inspection period, the print unit f ailed, which caused the operators to declare the channelinoperable per TS 3.3.3.3. The channel remained functionalin the capability to record a seismic event; however, the channel readouts could not be read out on site which would impede the ability of the operator to implement the abnormal operating procedure I

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following a postulated seismic event. Licensee actions were in progress to repair or replace the printer unit. The seismic monitoring system has a history of repetitive maintenance problems and was difficult to maintain because of obsolete components. Engineering was reviewing a request to identify options upgrade the system for implementation of the nuclear islan M1.3 Conclusions for Maintenance Activities Plant personnel performed routine and non routine activities wellin resolving problems, including the special test of a new check valve in the SW system, the calibration of radiation effluent monitors, and troubleshooting problems in diesel fire water pump, the EG 2A shutdown circuit and the seismic monitors. Plant personnel completed routine tests of plant equipment well, recognized degraded conditions, and initiated actions to complete troubleshooting and repairs. Good work controls were noted, including good pre job briefs, control of tagouts, adherence to work packages and work plans. There was good coordination with and support from health physics and engineering support personnel. Workers demonstrated good skills and knowledge of systems under test or repair. The persistence of some problems remains a concern, such as the problems on the diesel fire pump, and the EG 2A shutdown sequence. As in the past, poor plant material conditions challenge plant operators and impede the successful completion of testin M4 Maintenance Staff Knowledge and Performance M4.1 Ememency Diesel Generator TestinglVIO 97 03 03a) Insoection Scone (71707)

The inspector reviewed licensee action to test emergency diesel generator EG 2A on May 21,199 Observations and Findinns Following scheduled preventive maintenance on EG 2A, the licensee tested the diesel on May 21,1997 in accordance with procedures ENG 1.7114 (heat exchanger performance), PMP 9.131 (pre start Jacking and in-leakage checks), and SUR 5.1 17A (operability run). Operators operated EG 2A from the local excitation panel, which started 6:58 p.m. and ran successfully. However, the licensee did not complete the planned one hour test run. After starting the engine, a control operator identified an unusual noise and discovered that the Jacking gear was still instbiled on the engine. The operators informed the shift manager and shutdown EU 2A The Shift Manager responded to the scene to supervise recovery actions, and later prepared adverse condition report (ACR) 97 252 to describe the even The nuclear side operator (NSO) who performed PMP 9.131 had lef t the Jacking gear mechanism installed after Jacking the engine. EG 2A was jacked per steps 6.1.10 through 6.1.12 of PMP 9.131, which is classified as a continuous use procedure which must be in-hand during the conduct of the test and signed off as

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each step is completed. Af ter installing the tool and Jacking the engine, the NSO had signed step 6.1.12 as complete, Indicating that the crank tool had been removed and placed on its storage bracket inside the engine room. In f act, only the electric motor portion of the tool had been stored and the gear mechanism was left installed on the engine. The NSO failed to notice the gear as he reinstalled the cover over the Jacking area inside the engine compartment. The NSO failed to complete step 6.1.10 of PMP 9.131 because of a personnel error in his failure to assure the step was completed in accordance with the procedure requirement The licensee's investigation found damage to EG 2A, which remained unavailable for service until May 28,1997. Licensee actions to address the damage are described in Section M1.1 above (AWO 97 2028). The immediate corrective actions included: relieving the NSO from duty pending a review of the event the l NSO was subsequently reassigned duties outside of operations; changed procedure l PMP 9.131 (along with procedures SUR 5.1 17A, SUR 5.1 178, NOP 2.1 16a and

) 2.1 168) to require double verification that the jacking tool is removed prior to i

running a diesel generator; and, conducting a brief of the event with each operating crew. The licensee also conducted a root cause investigation of the May 21 event to identify contributing causes to the operator error. The licensee planned further actions to prevent recurrence of the event, including procedure changes to improve the clarity of the instructions to Jack the engine, and a plan to modify the jacking tool',o to make it impossible to reinstall the gear ring cover with the jacking tool installe The inspector identified no inadequacies regarding the immediate corrective actions, nor in the licensee's conclusions regarding the cause of the event. The f ailure to follow PMP 9.131 during the conduct of EG 2A testing on May 21 was a violation of Technical Specification 6.8.1 that was identified by the licensee. The May 21 event was a repeat occurrence of this problem in that a similar event occurred on November 27,1996 (ACR 961322),which was the subject of an NRC enforcement action (VIO 96 13 01). This event was another example of a recurring problem of deficiencies in worker performance during the conduct of routine activities, as noted in Inspection 9613 and 97 0 The licensee corrective actions to address human oerformance errors were in the process of development and implementation when the May 21 event occurre However, the May 21 event should have been prevented by a more thorowth or timely licensee review and response to the November 1996 ovent, The failure to correct a condition adverse to quality was one of two violations of 10 CFR 50,

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Appendix B, Criterion XV,1 (VIO 97 03-03a). Also see section E Human Performance issues - Personnel Errors Several recent inspection issues (Inspection items 9613 01,97 0102) concerned the occurrence of personnel errors and the failure to follow procedures over a wide spectrum of plant activities. The licensee was requested to respond to inspection Item 97-01-02 by NRC letter dated May 8,1997. Other examples of poor personnel performance were noted during the period which involved the I

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performance by various plant personnel, including operators. The licensee First Quarter 1997 Trend Report (CYCA 97 014) issued on June 10,1997, found that personnel error and procedure noncompliance was the most frequent reason for initiating ACRs during the first three months of 1997 (i.e.,32 of 162 ACRs).

The licensee responded to inspection 97 01 by letter dated June 6,1997 (CY 97-058), which described actions to reduce personnel errors and improve human performance. NRC concerns regarding human performance in routine operations and maintenance activities were discussed in a meeting with licensee management

, at the NRC Regional Office on May 28,1997. The NRC concerns regarding l personnel errors and procedure noncompliance remain, Conclusions Poor procedural adherence was demonstrated during a test of EG 28 on May 21, 1997. Licensee corrective actions to address a similar error during a test of EG 2B in November 1996 were ineffective. The occurrence of personnel performance errors in the conduct of routine activities remains an NRC concern that warrants further licensee actio M8 Status of Previous inspection Findings (92902)

M8.1 (Closed) VIO 9613 01: Diesel Run with Crank Tool Installed This item concerned an problem on November 27,1996 when an operator f ailed to follow a test procedure, resulting in tne operation of EG 2B with a jacking tool installed. The licensee responded to this matter by letter dated February 25,1997 to describe the corrective actions taken relative to the individual involved in the event. The licensee considered the November 27 incident to be an Isolated even During this inspection on May 21, an operator f ailed to follow a test procodure, resulting in the running of EG 2A with the Jacking toolinstalled and damage to the EG 2A ring gear, causing the diesel to remain unavailable for service for 7 days (See Section M4.1 above). The error on May 21 was a continuation of past performance problems, and demonstrated that past licensee corrective actions were ineffectiv NRC concerns regarding the correct performance of routine activities were addressed in inspection item 97 0102 and were the subject of a management meeting with the licensee on May 28,-1996. Licensee action to address human performance issues will be tracked as part of inspection item 97 03 03. This item is closed, M8.2 (Closed) URI 96-06-05: Actions to Address MIC Corrosion This item was open pending further NRC review of licensee actions to implement the MIC mitigation program and to address degraded conditions. This area was reviewed in inspection 97 01 and Section M1 above. Improvements were noted in the licensee efforts to monitor MIC degraded pipe sections and to make timely operability determinations for adverse findings. The licensee began startup and operation of the Bulaab injection system to mitigate the MIC problem. The licensee

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maintained the Bulaab system on line in accordance with a planned treatment schedule that involved injections into the SW system for two hours, three times per l week. Based on the above, this item is close ,

111. Enaineerina E1 Conduct of Engineering

E Enaineerina Sucoort

Engineering provided timely and effective support to plant operations during the period to address severalissues important to shutdown operations and for transitioning to the decommissioning inod ,

i in support of operations, engineering helped assure the successful resolution of the following: developed and implemented PDCR 97 002 & associated TE/SEs to address potential water hammer in the SW lines to the SFPCS; the completion of inspections and repairs to service water piping caused by MIC corrosion; the development of systematic plans to troubleshoot maintenance problems on the

< EDGs and the diesel fire pump; the Identification and prompt operability assessment of SFP pump NPSH issues (a CMP finding); the identification of fouling in the B SFP heat exchanger (HX); a timely HX operability assessment, and the development of short term corrective action and comprehensive long term followup actions; and, resolved a long standing deficiency in the SCANRAD Computer to improve RMS operability and reliability.

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in support of decommissioning, engineering: developed and implemented the process in ENG 1.7156 to categorize plant systems; implemented a revised process to conduct 50.59 SEs; made continual progress for the completion of decommissioning planning (GRPIs); developed and submitted the defueled TSs, Certified Fuel Handler TSs, Emergency Plan, Security Plan and QA Plan; and,

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developed procedures for shipping new fuel to a offsite vendor. Mixed performance i

was noted in the support to ship new fuel. Despite the staff efforts to prepare for the shipment, the licensee failed to assure full compliance with the Certificate of Compliance 685 Assessments and safety evaluations completed in support of the above activities were reviewed by the inspector while in progress and were technically sound and

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adequately documente E1.2 Cateaorization of Plant Systems for Decommissionina a. insoection Scone (73051)

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The purpose of this inspection was to review the licensee engineering evaluations

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used to categorlu, plant systems for decommissionin _ __ .

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18 Qbservations and Findings The licensee developed a process per procedure ENG 1.7150 to identify the proposed system changes necessary to support decommissioning, which resulted in the categorization of all systems on plant critical drawings in the following manner:

Onerable Those systems which provided a safety function in the decommissioning mode (maintained spent fuelin a safe condition), or were required to be fully operable in order to comply with the technical specifications, the technical requirements manual, or other regulatory requirements. The systems in this category would be maintained in full compliance with regulatory commitment Available Those systems that were required to remain available to perform a non-safety related plant support function (such as radwaste processing, heating and ventilation, etc). The function may be modified or limited from the originallicensing design function. These systems may be operated and maintained to different requirements depending on the system operating demand Lav up Those systems that would be isolated and laid up in a safe condition. These systems may be used again in the decommissioning mode, such as to support decontamination of reactor systems. There would be some effort directed toward system maintenance, but licensee commitments would be revised as needed to

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reflect the current needs.

l Abandoned Those systems that were no longer required to support any plant l activity through decommissioning. These systems could be isolated and placed in a l safe condition. All system operating and maintenance guidelines (procedures, preventive maintenance) and outstanding work items (trouble reports, nonconformance reports, work orders, action items) would be cancelle The licensee implemented the process this period by issuing the first system categorized under the new procedure in April,1997. The first system evaluated was the turbine generator lube oil system, which was selected sinco portions of this system would be classified in several different catta rie The inspector reviewed implementation of the process in the safety evaluations for the following system categorizations:

SY EV 97 04F, Spent Fuel Pit Cooling System

Sy EV 97 048,Feedwater & Condensate System

SY.EV 97 026, Fuel Oil to Diesel Generator and Auxiliary Boiler

SY EV 97-010, Hydrazine Feed System

SY EV 97-016, Circulating Water and Vacuum Priming System

SY EV 97 032, Safety injection System The safety evaluations (SE) listed above properly classified the associated systems in accordance with regulatory requirements, plant safety functions, and the future needs for the systems. The GEs were completed in accordance with the

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requirements of ENG 1.7156. The inspector concurred with the licensee's determinations that no unreviewed safety questions were created by the changes to the f acility as described in the UFSA Conclusiorg Engineering provided timely and effective support to plant operations through the development and implementation of ENG 1.7150 to assist in the smooth transitioning to the decommissioning mod E2 Engineering Support of Facilities and Equipment E2.1 Service Water System Waterhammer (VIO 97 03 03b) Inspection Scope 13755_1)

The purpose of this inspection was to review the licensee evaluations and resolution of potential two phase flow problems in the service water (SW) system, and to complete modifications to preclude postulated waterhammer events.

' Observation and Findinas in February 1997, inspection 97-01 provided the NRC review of licensee actions to

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address potential waterhammer in the SW cooling lines to the SFPCS, Actions were i completed during this period to prevent water hammer by the installation of a check valve in the common SW supply line to the SFPCS Check Valve SW CV 963 and associated test connections were installed and demonstrated to be operable per design change request DCR 97 002. The safety and technical evaluations for DCR 97 002 showed that waterhammer would be prevented as long as water leakage back through SW CV 963 was less than 2 gpm during the 45 second period following a loss of normal power (LNP) ct<ent while the emergency diesels starte The check valve leakage was measured at much less that 2 gp Inspection item 97 01-07 was open pending actions to resolve the design discrepancy, complete a review of the causes, and make a report under 50.73. LEH 97 07 dated April 23,1997 reported the event and provided the licensee's assessment of the significance of the uncorrected design discrepancy. The licensee rooorted the discrepancy per 50.73(a)(ii)B and (aMi)B as operation outside the design basis and a condition prohibited by the technical specification A delay in resolving this technical issue occurred from the time the issue wcs identified to the engineering staff on August 14,1996 (upon receipt of Report TM-1788a) until the development of an appropriate opvability and reportability evaluation in March 1997 following NRC review of the matter. The licensee's root cause investigation was completed on April 16,1997. The causes for the untimely followup included assigning tbo work as a low priority, weaknesses in the CAP (issue tracking), and poor control of work turnover during the period of staff instability af ter the decision to decommission the plan _ _

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Following the identification of the design discrepancy in February 1997, the licensee took timely and appropriate corrective actions to review other NRC open issues and engineering department open Action Requests. Th!s review showed that no other safety significant discrepancies were had been overlooked during the transition in station staff. As shown in the key performance indicators for eng;neering, progress has been made to reduce the backlog of open engineering work (the backlog reduced from greater than 200 to less than 100 items in July 1997).

CY concluded from an engineering assessment that the SW cooling lines to the SFPCS could not be shown to remain operable under the postult'ed design basis transient (LNP). The safety significance of the event was deemed to be low based on the low (current) decay heat rate in the SFP (1.5 f/hr) and the time available to implement compensatory measures (34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />) prior to reaching 150 degrees. The LER safety assessment addressed the present heat load. This matter was discussed with the licensee, who stated that it was expected that the analysis in License Amendment #188 for totalloss of SFP cooling under worst case neat load conditions was bounding for the postulated f ailure. During the exit meeting on July 15, the Site Director stated that an LER supplement would be issued to address this assessmen This original design deficiency resulted i.1 the SFPCS being inoperable under certain conditions, and resulted in plant operation contrary to TS 3.9.15. However, the delay from August 1996, resulted in operation with the deficiency during the full core offload in November 1996. The untimely licensee response war a violation (second of two) of 10 CFR 50, Appendix B, Criterion XVI (VIO 97 03-03b). (See section M4.1)

CY initiated an ACR to address the problem o, delay on these corrective action The ACR root cause analysis produced adequate corrective actions for the following underlying problems: 1) the department engineering supervisor's f ailure to follow administrative control procedure for ACR by not writing an ACR vwhen the issue came in the summer of 1996; 2) the previous action tracking system not effective in that the specific task assignment for this problem was never acknowledged (for unknown reasons) and the process was cumbersome to use. This analysis also noted that the department supervisor's staff turnover (leaving the organization due to decommissioning status of the plant) was not effective in assuring the issue was properly resolved in a timely manner. However, the ACR does not address corrective actions related to the staff turnover problem, c. Conclusions Engineering support was effective this period to complete evaluations in support of a design change to eliminate the potentiel for waterhammer in the SW system, and to complete corrective actions once the design discrepancy was realized. Past engineering support was poor resulting in inadequate control of the plant desig The failure to complete timely operability and reportability evaluations following discovery of the technicalissue in August,1996 was an example of a violation of regulatory requirements. The related ACR thoroughly identified underlying causes

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along with corrective actions for the delay in resolving the service water problem

, but the corrective actions were not comprehensive for all causes, e.g., staff

turnover problem.

J E2.2 Service Water System Corrgilqa inspection Scope (37551)

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The purpose of this inspection was to review the licensee evalisations of corrosion in the service water (SW) system. This review was in followup of Inspection item 97 01 08.

, Observations and Findinas Licensee engineering completed an assessment that was approved by the Plant i Operations Review Committee on March 26 to evaluate the acceptability of the continued use of the normal service water supply and return piping. The licensee found it acceptable to continue to use the normal service water system to cool the SFP pending the completion of repairs. The assessment included considerations for the tyres of deficiencies (physical or analytical) postulated in the supply and return piping, the consequences of pipe f ailure, including postulated flooding if a line i failed, and the additional measure to maintain the SFP area under augmented

] operational surveillance.

! The licensee reported this matter to the NRC in LER 97 08 and the final engineering

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evaluation of the SW system was described in a technical evaluation contained in an April 15 memorandum from the Engineering Director, and reviewed by the PORC l on April 15. The licensee expects that the SW supply and return lines will remain

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susceptible to degradation from microbiologically influenced corrosio The SW pipe and weld degradation monitoring program was effective to identify the recent pipe wall degradations and will be implemented to provide ongoing evaluation of the status of the lines. The inspections were conducted quarterly and include examinations of locations with past degradation. The results of the monitoring i program to date have not shown unacceptable degradation on the supply pipin Two locations in the return piping were found with significant degradation and were

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repaired. The licensee completed repairs of all known degraded piping in the SW supply and return lines. The degraded pipe sections were evaluated using the methodologies of Generic Letter 90 05, including the use of augmented inspections to identify the scope of susceptible pipin The SW piping was evaluated for structuralintegrity. Even with the most severe

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degradation, the return line had approximately a 20% margin from the code allowables for seismic loads. The piping was acceptable for loads under normal

operation and had margins for seismic loads. Thus, no failure was expected even under seismic load . . . - - . . - _ - -- - - .-

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22 The safety function of the SW supply and return lines is to cool the SFP. This function is accomplished if the supply line is structurally intact. The most significant degradation was observed on the return line. The consequence of return line f ailure is flooding, which was evaluated and found to have no effect on spent fuel cooling and no significant affect on the operation of the pool. Thus, there would be no safety consequences as a result of return line leakage or failur The licensee has contingency plans in place to monitor, isolate and replace supply or return lines with temporary hoses if leakage or failure occurs. The alternate cooling method can be placed in service before the pool temperature teaches the TS 3.9.15 limit of 150 degrees The licensee plans to implement a " spent fuel pool cooling nuclear Island" concept that will eliminate the rellance on the SW system. The conceptual design change-for the first phase of the nuclear Is!and was developed. The licensee expects to eliminate reliance on the SW piping by implementing phase 1 of the nuclear Island by December 1,1997. The corrosion in the SW piping will continued to be monitored. Ti:e projected increase of corrosion related degradation is not expected to make the rotern line inoperable prior to implementation of the design chang Should corrosion be worst than expected, the licensee will address the system in accordance with GL 90-0 Based on the above considerations, the licensee concluded that the SW supply to the SFP cooling system could be placed back in service followlig repairs and be considered operable to perform its intended design functions. Ongoing monitoring .

programs will detect further corrosion degradation and assure the degraded pipe can be repaired and returned to servic Conclusions Engineering support was timely to identify the corrosion in the SW system, to assist in the prompt assessment and repair of degraded piping, to evaluate the significance of the ongoing corrosion relative to continued use of the SW system, and to provide recommendations for future actions. The above actions addressed in part the NRC concerns in Inspection item 97 010 . E2.3 SFP Heat Exchanaer Foulina (URI 97-03-04, URI 97 03 05) * InsoecJon Scone The purpose of this inspection was to review licensee actions to evaluate the operation of the spent fuel cooling system and to monitor the thermal performance of the heat exchenger _ . _ _ _ . _ . _ _ _ _ . . . . . _ _ _ _ _ _ _ . _ _ . , __

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b. Observations and Findinos Thermal Performance Testina The licensee measured the thermal performance of both SFP heat exchangers per procedure ENG 1.7102in August 1996, as part of the program to monitor service water system performance per Generic Letter 8913. The August tests were conducted to assure the spent fuel pool cooling system was ready to support the core offload for the refueling operation. In particular, the test demonstrated that the B heat exchanger was capable of performing its design function (as defined in License Amendment 188) to maintain spent fuel temperatures below 150 degrees F during a full core offloa Both heat exchangers were verified operating within the design thermal requirements. The heat transfer coefficient for the B heat exchanger was measured l

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at 542 BTU /hr ft8 degrees F, which was larger than the design value of 37 BTU /hr ft2 degrees F. Similarly, the heat transfer coefficient for the A heat exchanger was measured at 460 BTU /hr f t' degrees F, which was greater than the design value cf 217 BTU /hr ft' degrees F. Heat exchanger fouling was measured to be at 41% of the design limit, and was found acceptable at that tim The licensee performed a full core offloau in the Fall of 1996, and the reactor was completely defueled on November 15,1996. The 8 heat exchanger was in operation and maintained SFP temperature well below the Technical Specification 3.9.15 limit of 150 degree Eystem Walkdown Evaluation During routine walkdown of the spent fuel pool cooling system (SFPCS) on June 16,1997, the system engineer (SE) evaluated operating conditions on the in service B SFP heat exchanger (E 1018), and noted higher that normal differential pressures. The differential pressure was 27 psid, which was much higher than the design value of 7 psid for a clean heat exchanger. The SE initiated a trouble report to have operators back wash the heat exchanger in an attempt to reduce the fouling; the backwash was completed from 3:00 p.m. on June 24 until 10:00 on June 26. The heat exchanger differential pressure decreased to 20 psid following the back wash, which was not effective to significantly reduce the fouling. On June 27, the licensee initiated ACR 97 361 to initiate an address this adverse condition, and to perform operability and reportability evaluation The operability evaluation, documented in memorandum CY TS 97 0335 dated June 27,1997, concluded that the 8 heat exchanger was functional with margin to remove the present spent fuel decay heat loads, but was not operating per its design basis. Although the heat exchanger fouling exceeded the maximum allowed fouling f actor specified in the manuf acturer's component specification, E 10-1B was capable of maintaining the spent fuel pool temperatures below the TS limits for the current pool heat load. The current heat load was less than 3.2 X 10 + 6 BTU /hr and decreasing monotonically with time (reference CYAPCo proposed

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revision to the operating license provided in letter CY 97 000 to the NRC dated May 30,1997). An engineering evaluation estimated the heat exchanger under conditions existing on June 27 was capable of removing 5.0 X 10 + 0 BTU /hr, and was operating with at fouling factor three times the design value and a heat transfer coef ficient of 211 BTU /hr f t' degrees The licensee estimated that E 101B could remove about 10 X 10+ 0 BTU /hr if the service water and spent fuel poolr were operating at the design limits. Thus, at the existing fouling levels and differential temperatures at the heat exchanger inlets, the spent fuel pool temperature was estimated to roach 111 degrees F if the Connecticut River temperatures reached the design maximum of 90 degrees F during the summer months. Despite the fouling, the heat exchanger had at least 300% margin between the actual heat load and the present heat removal capabilit The operators declared the SFPCS inoperable as of June 30. The licensee reported this item to the NRC per 10 CFR 50.72(bH2Hi) on June 30 as plant operation outside the design basis. The licensee plans to report this matter to the NRC per 10 CFR 50.7 Corrective Actions The licensee established several short and long term corrective actions to address heat exchanger fouling and to address operation outside the design basi Procedure SUR 5.108 was revised by TPC 97180 on July 1,1997 to add data logging requirements to the operator rounds that would monitor additional B heat exchanger data to allow continued trends and performance. The operators began logging the service water temperature at the inlet of the heat exchanger and monitored the differential temperature between the heat exchanger and river temperatures. The licensee established differential temperature limit of 35 degrees F as an action point that would indicate heat exchanger fouling had reached unacceptable levels, and result in switching from the B to the A heat exchanger for SFP cooling. A site engineering evaluation of recent heat exchanger data showed that the fouling rate had leveled off as July 7 as evidenced by the resistance dif ferential tempeinture trends. The temperature trents remained well below the limit for the remainder of the inspection perio The last time E 101B was cleaned was in 1989 in conjunction with the changes made per PDCR 977 to increase the thermal capacity of the unit. The operators backwashed the heat exchanger during this period while operating the A and B heat exchangers in parallel; this effort did not significantly reduce the differential pressure. The licensee planned to repeat the backwash operation with only the B heat exchanger in service. Further, the licensee planned to review methods to clean the heat exchanger, including alternatives to either chemically clean or replace the 338 plate An assignment was made for engineering to review and revise the UFSAR 9.1.3 and Technical Specification 3.9.15 Bases to reflect the present pool decay heat load in recognition that the worst case full core offload heat load of 22.4 X 10 + 0 BTU /hr will no longer occur. Redefining the design basis would allow terminating the

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condition of continued operation on the B heat exchanger in a " condition outside the design basis".

Finally, plans were made to evaluate options on use of the 8 heat exchanger or replacement with a different unit during the transition to the nuclear island for the spent fuel pool. The licensee also recognized the need to redefine the licensing basis in TS 3.9.15 that would remove the restriction to only operate the SFPCS on the B heat exchanger, and to allow the operational flexibility to use the A heat exchanger. The A heat exchanger has excess capacity to control pool temperatures at the present decay heat levels. The A heat exchanger is less susceptible to fouling from river silting and is much easier to clean once foule The inspector reviewed the SFPCS operational configurnt n since the completion of core offload in November 1996, and concluded that the discovery of the 8 heat exchanger fouling from normal operating parameters was difficult due to the l

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reduced flows in cold weather. Licensee corrective actions upon discovery of this lssue were comprehensive to address the operational, design and licensing issues raised by the issue. In particular, technical program personnel provided good support to identify the issue, evaluate the status of the unit and to recommend corrective action Notwithstanding the above assessment of support by the technical programs, in a broader perspective, licensee actions were not timely to identify the heat exchanger

- fouling before it lost 50% of its capacity. The vulnerability of the B heat exchanger to silting was a well known problem, resulting in past actions to minimize run time -

on the unit. The operationallimitations provided in TS 3.9.15 were recognized in late 1996. This information, combined with the extensive run time on the B heat exchanger, should have prompted a more timely investigation of the potential affects on the heat exchanger. There was missed opportunity for a more timely action to revise TS 3.9.15 to add operational flexibility afforded by the actual decay heat loa Licensee actions were not timely to identify HX fouling before it lost 50% of its capacity. This matter is unresolved pending the completion of licensee action to redefine the design basis for E 101B, and to revise the operating limits in Technical Specification 3.9.15 accordingly and in a manner that would recognize and allow operation of the SFPCS with the A heat exchanger (URl 97 03 04). This matter is further unresolved pending the completion of licensee actions to establish and implement the process to trend E 101B parameters and to evaluate heat exchanger fouling and continued performance (URI 97 03 05), Conclusigng System engineering provided good support to plant operations to identify a degraded fouling conditions on the spent fuel poc! heat exchangers, and to provide an operability assessment relative to the present SFP decay heat loads, and to complete a reportability evaluation in recognition of the component ability to perform its design function. The SE also provided timely and appropriate j

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recommendations to allow continued use of the heat exchanger while monitoring the continued performance and fouling of the heat exchange Licensee actions are needed to submit a TS change request to revise TS 3.9.15; and, to implement long term corrective actions (operational trending, extended backwash and chemical cleaning options, redefine the design bases as stated in the TSs and UFSAR, obtain replacement plates to stage for repair, consider the options for the nuclear island).

E3 Engineering Procedures and Documentation E3.1 Procedures to Conduct Safety Evaluations per 10CFR50.59(37001) Inspection Spapa The purpose of this inspection was to review licensee procedures and practices to conduct evaluations of facility changes, tests and experiments in accordance with 10 CFR 50.5 Observations and Findinati

&ggam Guidance: 10 CFR 50.59 and 50.82 Safety and Environmental Evaluations The inspector reviewed the licensee's 10 CFR 50.59 safety evaluations and its 10 CFR 50.82(a)(6) environmental and funding analyses that support facility changes at a permanently shut down f acility. The inspection was performed in two phases; first, the licensee's underlying procedure covering the preparation of these evaluations was reviewed. Secondly, selected licensee safety evaluations, all related to the maintenance of spent fuelintegrity were reviewe The underlying Connecticut Yankee Atomic Power Company (CY) procedure, that was reviewed is " Safety Evaluations" (NGP 3.12, Rev.10) dated April 17,1997 and is the top level document used in the preparation of safety evaluations. The inspectors found that NGP 3.12 provided sufficient and proper guidance to the CY staff responsible for the preparation of licensee 10 CFR 50.59 safety evaluation NGP 3.12 did not require inclusion of the 10 CFR 50.82(a)(6) analyses; however, a retitled and revised version of NGP 3.12, namely, ACP 1,2 2,42, "10 CFR 50.59 Applicability Reviews and Safety Evaluations" does include these analyse On May 12,1997, the licensee revised the administrative procedures and practices for the conduct of safety evaluations per 10 CFR 50.59. The process changes in ACP 1.2 2,42, Revision 1, were made to enhance the program and to address the defueled condition of the plant, in addition to criteria that define an unroviewed safety question, the licensee added criteria that defined an unreviewed decommissioning questions as a change that would foreclose release of the site for unrestricted use, result in a significant environmentalimpact not previously reviewed in the safety analysis report, or remove reasonable assurance that sufficient funds would not be available to complete decommissionin . . -

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The revised ACP 1,2 2.42 was developed with consideration for the NRC staff positions in SECY 97 035, Proposed Regulatory Guidance Related to implementation of 10 CFR 50.59, dated February 12,1997 (subsequently issued as draft NUREG ,

1606). Generally, where the NRC staff position differed from the industry guidance i on completing safety evaluations as defined in the Nuclear Energy Institute (NEI)

document NSAC 125, the licensee adopted the staff position. The changes in this

revision, together with those made in the past to the predecessor document NGP i 3.12, were also intended to address past weaknesses in the 50.59 process that ,

i were contributors to the past poor performance and regulatory findings (reference

NRC Inspection 96 201, items 96 20102,96 201-03,96 201-04 and 96 20105),

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! The licensee provided training on the new procedure and program requirements.

l The inspector observed one training session held on May 1,1997. The licensee

provided detailed training for engineers, supervisors and managers who would likely be involved in 50.59 safety evaluations. FamlNrization training was provided for

, each plant department on the 50.59 screening process. The training was detailed, comprehensive and confirmed table top studies of sample safety evaluation Overall, the training was of high quality as evident in the instructor's familiarity in

the subject matter and good use of training aids.

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The new ACP 1.2 2.4.2 has two parts: the first lists procedural steps to follow; the second and larger part provides guidance on terminology and interpretation of the

, 10 CFR rules. Discussion of the document with the licensee noted several cl'anges that would improve the clarity of the instructions in ACP 1.2 2.42. For example, the instructions defined " malfunction" as f ailure to perform a desired function. This

definition is less inclusive than the staff guidance, which includes partial, j inadvertent, and unexpected operation as a malfunction. An expanded definition of

malfunction was included in the guidance section of the procedure that closely i followed the riaff guidance. The licensee agreed to revise the procedure to reflect

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the staff position in the instruction portion of the procedure. Other clarifications

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included definitions of " margin of safety" and " change".

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Connecticut Yankee has created a term, " safe plant change", which is not

recognized in the regulations. The definition considers a change to be safe if it causes "no significant increase in risk". The licensee representative discussing the 4 procedure understood that the term had no regulatory meaning, and that 10 CFR 50.59 requires NRC approval prior to implementing a change with any increase in risk, regardlon of size or significance. Careful reading of the procedure reveals this requirement as implicitly contained in the definition. however, additional wording will be added to the procedure to explicitly identify the rule requirement, thus assuring that a person following the instructions will not misunderstand the intent of the regulations.

Connecticut Yankee has a category of procedure change called "no@ tent change". This category is defined in a procedure separate from ACP 1.2 2.4.2

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The notion is that some changes merely represent different but equivalent ways of carrying out proceduralinstructions. The level of review and evaluation for such

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changes is much lower than for those that change the intent of a procedure. The

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safety evaluation procedure, ACP 1.2 2. Jowed "non intent changes" to avoid the safety evaluation requiremen A considermion of the definition of "non n int change" found that changing the order of procedural steps could be classifiei is "non-intent", thus not requiring a safety evaluation. However, valve lineups are a common plant evolution where the order of performance can have significant safety effects. The licensee agreed that the exclusion of "non intent changes" from the requirement for safety evaluations was not desirable and will revise the procedure to eliminate the potential weakness in the procedure guioance, in summary, the inspector found that the licensee's past and present procedures for the conduct of safety avaluation per 10 CFR 50.59 were acceptable, and improvements were made to address past weaknesses. NRC review identified several procedure clarity issue Process lmolementation The three inspected PDCRs were allissued prior to the implementation date of the revised 10 CFR 50.82 and therefors the licensee was not required to perform these evaluations. The safety evaluations are an integral part of:

PDCR No.1550 Rev. 0 - Radiation Monitoring System Upgrade PDCR No.1555 Rev. 0 - Pormanent Installation of Diesel Generator EG 7 PDCR No.1586 Rev. 0 - Upgrade the Fuel Storage Building Crane CR 51 A From E Tons to 6 Tons

Plant Design Change Record PDCR NO.1550 Radiation Monitoring System (RMS) Upgrade This modification replaced several process radiation monitors:

R11 - Containment Particulate R16A/B Steam Generator Blowdown R12 Containment Air R18 - Service Water R14A- Main Stack R22 - Waste Test Tank Discharge R15 - Air Ejector Most components of the monitors, such as detectors, piping and valves, pumps, controls, local and Main Control Room readouts, were upgraded. A new computerized display and control system was installed in the Main Control Roo The air ejector monitor was replaced with redundant channels, R15A/B. Inputs to the plant process computer were added for all the monitors except R14A. For monitors with two channels, only one channel was input to the plant process computer. Control functions were retained to close the waste gas and liquid discharge valves and steam generator blowdown valves on high radiation.

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The change was made to increase the reliability of the RMS, and to reduce the man-hours spent maintaining l!. The computer installed as part of the change provided the ability to expand the RMS in the futur The safety evaluation for this change was approved December 12,1995. The RMS has no safety function, is non-QA, non seismic, and not subject to electrical environmental qualification. Technical Specification Action Statements allow any portion of the RMS to remain inoperable as long as compensatory manual sampling and analysis of the process stream is performed. If inoperability exceeds 7 days, the licensee must report that status, along with a plan to restore operability, to NRC within 30 days. Design basis accident mitigation does not require use of the RMS, and RMS has no role in any accident initiation. Thus, the conclusion in the safety l evaluation that the replacement of the monitors and associated equipment does not

! Involve an unreviewed safety question is acceptabl Note %t inspections 97 01 and 97 02 found that the RMS was inoperable due to

alibratio.' problems. The system is functional and in use for information purpose **he i modi' cation has not yet been closed ou PDCi4 NO.1555 Permanent installation of Diesel Generator EG 7

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This change was performed to provide a tornado-proof back up supply of 4160 volt power to provide additional shutdown risk mitigation, that is, mitigation of the

" shutdown" risk while the plant was stillin operation in early 1996. This was a beyond design basis event that included loss of all offsite power and the concurrent unavailability of both of the emergency diesel generators EDG A and EDG EG 7 is an air cooled machine; thus, does not require any cooling water. However, at the present time (1997), EG 7 will be used to supply power to the Fuel Storage Building. That future use of EG 7 is not part of this inspectio The CY analysis demonstrated that there was no increase in the consequences or probability of any previously evaluated accident due to the permanent installation of EG 7. The inspectors agree with this assessment as EG 7 would be available to supply power only through a normally open switch. This switch is a OA isolation device, seismically qualified, that will not inadvertently change state even during a seismic event. Based on this, the normal electrical power system remains unchanged: therefore, the accident analyses remain unchange The potential for an unanalyzed accident is avoided by use of the isolation device, based on the above discussio The isolation device was connected through about 30 feet of cable between the switch and the transformer terminals, the cable ran through a seismically qualified raceway. This design feature in addition to the use of the isolation device ensure that there is no reduction in the margin of safety for the plant electrical syste . . -. .. _--. - _ _ _ - _ . - - _ - _ - . - . - - - - - - - - - - ~ - - -

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I 3 The inspector concluded that the CY safety evaluation in PDCR 1555 for the relocation of EG 7 to a permanent location was acceptable.

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PDCR No.1586. Upgrade the FSB Crane CR 51 A from 5 Tons to 6 Tons This modification allowed the use of the Fuel Storage Building (FGB) crane CR 51 A for the installation the new spent fuel storage racks, as approved by the NRC in License Amendment #18 The technical and safety evaluations addressed the structural and seismic adequacy of the spent fuel building, crane and support structures to adequately handle the intended loads. The maxirnum weight of the new racks was 5.75 tons. The technical evaluation demonstrated that the load carrying capability of the crane and associated components (holsts, trolley wheels, drums, bearings, wire ropes) were ,

capable to support the maximum loads and were tested to 7.5 tons. The crane i

bridge girders, beams and support brackets were demonstrated by analysis to handle the intended loads with suitable margins. The crane was analyzed for selsmic ll/l concerns and demonstrated to be acceptabl The SFB crane load test to 7.5 tons was completed per ST 11.7167 on January 26,1996. The inspector reviewed the testing conducted per ST 11.7167, along

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with the associated safety evaluation, and found both acceptable. The licensee's evaluation recognized the need to limit loads on the yard crane while the SFB crane was in use. The licensee addressed this by establishing restrictions on the use of ,

the yard crane in procedure WCM 2.2 8, Control of Heavy Loads (TPC 96 61). The

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inspector agreed with the licensee's conclusion that the changes made to CR 51 A did not create an unreviewed safety question. The inspector concluded that the CY safety evaluation in PDCR 1586 to increase the load capscity of CR 51 A was.

l acceptable.

The inspector reviewed the licensee's procedure r.no 'ssociated safety evaluations

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for moving fuel within the spent fuel pool. Tb - . tor reviewed Bypass jumper 96 31, and procedure SNM 1.4 5, in Plant Transfer of Fuel (Excluding Refueling),

inclusive of TPC's 96 309,96 311, and 96 320. NRC reviews previously found weaknesses in a May 2,1996 safety evaluation for the use of a sling on the hoist for CR 51 A (reference Inspection Item 96 04-03). Aside from that issue, no l inadequacies were identified in the c ?.er procedure changes and temporary modifications (referenced above) fo. .he spent fuel building cran Conclu.gigna The new process for the conduct of safety evaluations as defined in ACP 1.2 2.42 was acceptable and contained improvements to address past program weaknesses *

on the conduct of reviews per 10 CFR 50.59. Training on the new process was of

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good qualit NRC review identified several procedure clarity issues which, if left uncorrected, could result in inadequate safety evaluations or the f ailure to conduct SE's per

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50.59 and 50.82. The issues involved the need to clarify procedure guidance regarding margin of safety, malfunctions, change, anu the handling of temporary procedure change E3.2 Deslan Discrenanev Inadeouste SFP Pumo NPSH (NCV 97 03 06)

e, Insoection Scoce A design basis issue identified during the period appeared as historical examples of problems in defining or implementing the plant design basis, or in meeting licensing commitments. The inspector reviewed the licensee action to disposition this issu Observation and Findinas As part of the Configuration Management Plan (CMP) effort, licensee engineers reviewed the spent fuel pool and cooling system design as part of the corrective action to update the design and licensing basis of the plant. The CMP identified a-deficiency in which the SFP pump design basis did not support the licensing basi The finding was based on an April 10 engineering review which noted that -

conditions may have occurred for which the net positive suction head (NPSH) of the SFP cooling pumps might not have been assured during plant operations in the 1970s (ACR 97182).

Calculation PDC Y 169 was completed in 1969 showed that the SFP cooling pumps had adequate NPSH assuming a maximum spent fuel pool temperature of 120 .

degrees F. The calculation was performed for the A SFP pump, which required 20 -

f t of head versus an available head of 23 ft. However, since that design basis was established, the licensee had changed the SFP licensing basis to rerack the pool to accommodate larger core Inventories, in support of the expansion, a new SFP pump (P-21 1B) and heat exchanger (E 101B plate type) was added in the 1970s, and the licensing basis for the maximum pool temperature was increased to 150 During a review of the design basis in April 1980, a preliminary engineering estimate of the required NPSH at the elevated temperature showed that pump cavitation could occur at 141 F. A more refined analysis was performed using a thermal hydraulic analysis specifically modeled for the Haddam Neck spent fuel cooling system. Calculations on April 10 usirig the PROTOFLOW analyses showed that potential cavitation could occur at 153 F for the A SFP pump, and at 143 F for

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the B SFP pump, when either pump was operated in a configuration using a single suction path frem the pool. Additional PROTOFLOW runs assuming conservative atmospheric conditions (27.5 in HG) and a double suction configuration showed that the pumps would cavitate at 194 F and 168 F, for the A and B pumps, respectivel The licensee's immediate actions was to verify that the plant was in the double suction configuration and to assure plant procedures direct operating configurations that would assure NPSH margins would as maintained for temperatures up to the -

licensing basis limit of 150 F. The double suction path was in service with both

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SFP cooling valves SF V 810 and SF V 811 open. Both valves have been open since the cooling system was aligned in preparation for the full core of fload that was completed in November,1996. The licensee revised procedures to assure that the operators would throttle the pump discharge to reduce flow and preserve NPSH margin for contingency actions while recovering from a loss of pool cooling. The licensee revised the following procedures: NOP 2.101, AOP 3.2 59, and SUR A. The procedures were approved by the PORC on April 11,199 After further engineering evaluation of present and past operating configurations and procedure controls, the licensee concluded that the licensing basis requirements were met in the present plant conditions, based on configurations allowed by existing procedures. The existing procedures had assured that the double suct on path was aligned for any pool temperature above 140 F. A review of past operating configurations showed that the licensing basis was met for allowed configurations back to 1975. Prior to 1975, when a single SFP cooling train was installed, a the maximum pool temperature was calculated to be 170 F with a full core offload and assuming summer time river conditions (85 F). Although the full core was discharged to the poolin that period and no problems were experienced with spent i fuel cooling, the licensee concluded that conditions might have existed under the i

licensing basis conditions which could have resulted in cavitation of the SFP cooling pumps, and thus the loss of spent fuel pool cooling. This loss of cooling could have been mitigated by operator throttling action The licensee concludeJ that this item was reportable to the NRC per 10 CFR 50.73(a)(2)(v)(B),50.72(b)(2)(i) and 50.72(b)(2)(iii)(B) as historical conditions where l

the plant was in an unanalyzed condition, and a condition for which the fulfillment of a safety function by plant systems used to remove residual heat was compromised. The NRC Duty Officer was notified on April 15. This issue was reported as licensee event report (LER) 9710. This old design issue was an example of past inadequate engineering in support of plant operations, and was similar to the examples of deficiencies in engineering and in the control of the plant design identified in the NRC enforcement actions issued on May 12,1997. This design issue was identified as a resuit of the licensee correctives actions to address weaknesses in the licensing and design basis. This non repetitive, licensee identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the enforcement policy (NCV 97 03 06).

c, Conclusions The Configuration Management Group provided effective engineering support for plant operations in the actions to identify a historical discrepancy in the licensing design basis for the spent fuel cooling system. Engineering evaluations were timely this period to characterize the deficiency, to research the past operating configurations and present operating procedures to identify vulnerabilities, to take appropriate corrective actions, and to complete reportability evaluation .

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E3.3 Inadeauate Safety Evaluation Ooerator Actions (NCV 97 03 07) lnsocction Scona An example of an inadequate safety evaluation was identified for a historical operating condition. The inspector discussed this matter with the licensee and reviewed the licensee actions that addressed this issue, Observation and Findinas inspection 96-06 Identified a design basis discrepancy in which the feedwater regulating valves (FRV) were found to be inadequate to close during a postulated design basis event, defined as the worst case break of a main steam line together with an assumed single active failure in which the associated feedwater line isolation valve (FWlV) fails to close. The failure to meet regulatory requ' ements relative to the design basis issue was described in Inspection 96 06 and the NRC letter to the licensee dated May 12,199 Inspection 96 06 also described the licensee actions to compensate for the inability of the FRVs to close under design basis conditions. UFSAR section 15.2.9 states the following regarding a steam line rupture: " isolation of the main feedwater lines by two valves in series will occur on a safety injection actuation. This is done via closure of the feedwater isolation valve and the feedwater regulating valve after a _

short time delav. The licensee stationed a dedicated operator in the control room to provide added assurance that the main feedwater flow to the containment would be secured in the highly unlikely event of a break in a main steam line together with the failure of the FWlVs to close. Since this action substituted manual operator action for an automatic operation of the FRVs, the licensee completed this action under a change to the emergency operating procedures, and prepared a safety evaluation dated June 14,1996 per 10 CFR 50.59. The licensee's sdoty evaluation (SE) concluded that the change to the facility as described in the UFSAR was not an unreviewed safety question (USO). -

The NRR staff reviewed the licensee's June 14 SE, which was summarized in an NRC internal memorandum dated July 23,1997. The review included a site visit to -

the site by NRR personnel on two occasions in the June - July period in 1996. The licensee evaluation process included: (1) modifying the emergency operating procedures to direct operators to isolate the main feedwater (MFW) pumps; (2)

posting a dedicated operator at the MFW panet; (3) training the operators to recognize the plant conditions which would warrant MFW isolation; and, (4) using crew simulator exercises to determine if operators could take the actions in the time frame required. The evaluation was generally consistent with the guidance in GL ,

91-18 and ANSI 58.8, " Time Response Design Criteria for Safety Related Operator l Actions," 1984.

t The licensee's analysis did not, however, sufficiently consider the important contribution of operator errors of omission or commission which could have delayed proper response and potentially increase the consequences of the main steam line

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break accident. This was significant because the time frame available for operator action in this situation was exceedingly short (i.e.,45 seconds af ter trednt initiation). Although GL 91 18 does not explicitly require licensees to amine the consequences of operator performance errors and the likelihood of recovering from such errors, the staff expects licensees to consider the possibility of operator errors and determine if sufficient time exists to recover from such errors. This staff position was based, in part, on ANSI 58.8 which states that, Nuclear safety related operator actions or sequences of actions may be performed by an operator only where a single operator error of one m6nipulation does not result in exceeding the design requirements for design basis event The licensee evaluation did not consider the possibility of performance errors or the likelihood of recovering from such errors given the time frame allotted for accomplishing the manualisolation. Given such a short duration available for operator action, it was not likely that recovery from an error in performance could be achieved without exceeding the required 45 second isolation time. The NRC staff did not believe that the 45 seconds available for operator response was sufficient to (1) ensure accurate diagnosis of a transient, (2) perform the required manual actions, and (3) recover from potential operator errors. Based on the above, the NRC concluded that the licensee's change to the f acility in the use of manual operator action in this specific caso did constitute a USQ because the change could reasonably introduce the possibility of increasing the consequences of an accident, may cause a different kind of accident, or may decrease the margin of safety. The change to the facility should not have been completed without prior NRC review and approva Thus, the change to the f acility as described in UFSAR Section 15.2.9 was a USO and a violation of 10 CFR 50.59, inspection 96-04 issued a violation for the licensee f ailure to complete an adequate safety evaluation per 50.59 when changing the hoist used to move spent fuel over the spent fuel pool. inspection 90 201 identified multiple examples of the failure to complete adequate safety evaluations per 50.59. Past NRC inspections also found other weaknesses in the licensee's process to implement operability determinations in accordance with the guidance of GL 91 1 The licensee program failures resulting in this violation were similar to past NRC findings and the issues addressed in the May 12,1997 enforcement action. The f ailures resulting in this violation pre-dated the NRC's 50.59 findings and the licensee corrective actions to improve the 50.59 program. The NRC staff is exercising discretion in accordance with Section Vll B.2 of the enforcement policy since the violation was based on events prior to the plant extended shutdown and having taken significant enforcement action on May 12,1997, for the technical and safety review program inadequacies that led to this and other violation (NCV 97 03-07).

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The NRC staff noted that improvement was needed in the industry's implementation of the 10 CFR 50.59 process. The NRC staff forwarded its recommendations to the Commission in SECY 97 035," Proposed Regulatory Guidance Related to implementation of 10 CFR 50.59 (Changes, Tests and Experiments)." The licensee's revised process for conducting 50.59 teviews (ACP 1.2 2.42) recognizes the NRC guidance relative to the consideration of operator errors in 50.59 determinations, Conclusions The licensee's safety evaluation dated June 14,1996in support of the use of manual operator action as a substitute to the automatic operation of tha FRVs was inadequate. The change to the f acility as described in UFSAR 15.2.9 created an l unreviewed safety question and was a violation of 10 CFR 50.59. The NRC

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exercised discretion for this issu E8 Miscellaneous Engineering issues E8.1 Status of Previous Inspection items (Open) URI 9810-01: Audits of Soecial Nuclear Material (SNM)Inventorv This item was open in part pending the completion of licensee actions to lift 15 fuel bundles stored in the spent fuel pool to verify the bundle serial numbers. This action was defeired due to the licensee management restrictions placed on plant activities. The licensee updated the status of his plans on this matter during the

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present inspection. The licensee intends to complete the audit activity during July, 1997. Further, the licensee planned to complete a shuffle of fuel bundles within the spent fuel pool to better disperse the zircoly clad fuel with the stainless steel clad fuel. This activity was planned for September 1997. The licensee intends to conduct a complete audit of the SFP contents following the fuel shuffle in accordance with 10 CFR 70.51. This item remains open pending NRC review of the actions to complete the subject SNM audit 'losed) IFl 95 20-01: Classification of Diesel Turbocharoer als issue concerned the need for further review of the quality classification for the diesel generator turoocharger pre lubrication system. The licensee's engineering evaluation was provided in memorandum CY TS 97 367 dated July 3,1997. The evaluation addressed the various components of the turbocharger pre lubrication and preheating subsystems and identified the bases for Category 1 and no Category 1 classifications. Components in the pressure boundary are Category 1, as are the components (three engine driven pumps) relied on the lubricate the diesel while the engine is running. Although the electric pre lubricating pump , are not Category 1, a motor failure would be annunciated by a Category 1 alarm and indication circtit, that would allow for appropriate action to be taken to correct the problem. Finally, the lack of pre-lubrication could affect the 10 second start time of the engine. The f ast start capability of the emergency diesel generators was a

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design feature needad to mitigate postulated events with the plant operating at power, and is not essential with the plant in the permanently defueled state. This item is close (Closed) VIO 94-14-02: Corrective Actions,tpr Service Water This item concerned inadequacies in the licensee corrective act:vn program that resulted in degraded conditions in motor control center MCC 5 and the service water (SW) system. Past NRC inspections have reviewed the corrective actions take; to assure the reliability of MCC 5 and to address MIC corrosion in the SW system, as well as actions to improve the corrective action program (Inspections 94 25,94-27,95-16,96-01 and 96-06). Inspection item 96-06-05 concerned the need for the licens6e to conduct timely operability evaluations of corrosion related degradation of the service water systum, improvements in licensee performance in this area was described in Inspection 97-01, and is discussed further in Section M8.2 above. The recurring inadequacies in the licensee corrective action program were described in Inspections 96-08,96 201,96-11 and 96 80, and was the subject of an escalated enforcement action issued on May 12,1997. The

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resolution of NRC concerns regarding the effectiveness of the corrective action .

l program will be tracked as part of the response to the escalated enforcement (see l Section E8.1). This item is close (Open) URI 97-0108: Corrective Actions for SW Corrosion

As described in Section E2.2 of this report, the licensee completed actions during this period to repair known corrosion related degradation of the SW piping supplying cooling water to the SFP cooling system, and to complete engineering evaluations to assure the SW system will remain an acceptable source for SFP cooling as long as the SW is needed to provide that function. This item will remain open for further NRC review of licensee actions to monitor for and address corrosion related defects in the SW piping, and pending the completion of licensee actions to install an alternate cooling water system for the spent fuel poo E8.2 Escalated Enforcement - Mananement, Desian and Corrective Action Issues On May 12,1997, the NRC issued a Notice of Violation and Proposed imposition of Civil Penalty to the licensee that was based on numerous inspections of Haddam Neck operations to review several facets of plant performance. The inspections included a special team inspection by NRC headquarters staff which focused principally on engineering performance (96-201, Engineering & Licensing); a special Augmented inspection Team (AIT) inspection of the reactor vessel nitrogen intrusion event in September 1996 (96-80); an emergency preparedness inspection to review the conduct of the emergency exercise in August 1996 (96-07); and several resident inspections (95-27,96-06,96-08, and 9611).

One violation (the low pressure safety injection system flow rate less than the accident analysis value) was discussed at a conference on Fetruary 12,199 Most of the other violations were discussed at a conference on December 4,199 . _ - _ _ .

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The violations related to inspection 9611 were not discussed during the December conference after the licensee agreed that another conference was not needed to discuss these issue The violations were grouped into broad categones:

numerous deficiencies in engineering programs and practices, including plant design, design control, and engineering support, some of which led to significant safety equipment being inoperable or degraded for extended periods;

numerous operational deficiencies, including inadequate procedures, failure to follow procedures, and inadequate corrective actions, which led to the

"nitrovn intrusion" event;

inadequate implementation of the emergency preparedness program during the August 1996 exercise; and, l * numerous deficiencies in the corrective action program resulting in untimely or inadequate correction of deficiencies adverse to quality in plant operation The engineering violations included: the failure to assure that the plant was ,

maintained as designed and specified in the Updated Final Lafety Analysis Report (UFSAR); the introduction of additional design errors during design changes as a result of poor engineering; making design changes to the facility without performing adequate safety evaluations, including at least one instance where the change (removal of a flood protection floor block) irv;olved at: unreviewed safety question; not identifying or correcting adverse conditions that resulted from poor engineering, or the causes of those conditions; and not updating the UFSAR when require At the December 4 enforcement conference, the licensee acknowledged the NRC findings and noted that a number of the issues would be applicable with the reactor in a defueled stage. The licensee acknowledged that there were significant deficiencies that must be fully addressed before the start of significant decommissioning efforts at Haddam Neck. The licensee described a number of corrective actions that had been either taken or planned to address the programmatic weaknesses. Among those actions were the establishment and communication of specific management expectations; and, the implementation of new plant processes and programs. The status of the licensee corrective actions were discussed in part during management meetings held at the NRC Region I offices on February 5, May 28 and July 18,199 The licensee responded to the May 12 enforcement action by letter dated June 11, 1997 to further describe the corrective actions taken, the actions to prevent recurrence and the schedule for completing the actions to prevent recurrence. The licensee actions to date included:

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e Enhancements to the corrective action program through the acsignment of a dedicated Corrective Action Program Manager with a dedicated staff to administer the program; the development and issuance of new procedures to implement a simpler and more streamlined ACR process; improvements to root cause evaluations; the development and implementation of a new action item tracking system (ATS) to improve the tracking and resolution of commitments; the training of key plant staff on the new ACR and ATS processes; and the development of a process to trend ACRs to identify performance issues - as evident in the First Quarter 1997 Trend Report, e implementation of the configuration management plan (CMP) to redefine the licensing and design bases for systems important to plant safety. This effort began in mid 1996 using a graded systems approach to certify the plant systems needed to support plant operations at power. The scope of this effort was reduced following the licensee's certification per 10 CFR 50.82 to 1 permanently cease power operations. The licensee has grouped the remaining plant systems to support the nuclear island into the following categories: SFP Cooling and Purification; SFP Makeup; Spent Fuel System Fire Water; Fuel Building Ventilation; Spent Fuel Structures: Service _ Water; Radiation Monitoring and, Electrical and Instrumentatio e The development of a revised Design Change Manual which is scheduled to be completed by mid-1997; and, the implementation of a revised process to .

conduct safety evaluations under 19 CFR 50.59. Section E3.1 of this report describes the NRC review of the revised program. The revised processes would assure the proper program linkages were made among accident analyses, plant design changes, 50.59 safety evaluations, design calculations, and the process to apdate the UFSAR and change the licensing basis. The licensee planned to submit a UFSAR update to reflect a comprehensive summary of the defueled condition by December 1997, e Additional corrective actions were implemented tn improve the licensing / design basis information, address procedural deficiencies, improve the design control process, assign adequate resources, improve training effectiveness, enhance the communication of management standards and expectations, and to improve independent oversigh This inspection began the NRC review of the actions by CY to address the weaknesses as described in the June 11 response. As noted in Sections M4.1 and E2.1 above, the NRC noted continued problems in the development of effective corrective actions to preclude conditions adverse to quality in the engineering area and in recurring problems with personnel error and procedure adherence. This area was also reviewed during a management meeting at NRC Region I on May 28, 1997. This area remained under NRC review at the completion of this inspection perio .

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E8.3 Review of Licensee Event Reoorts (LERs) Insoection Scooe (92700,9071_2)

The purpose of this inspection was to review licensee event reports (LERs) to verify the requirements of 10 CFR 50.72 and 50.73 were met, . Observations and Findinas Revised LER Commitments By letter dated July 7,1997, the licensee submitted update LERs for 8 past events to review the commitments to complete corrective actions made in the past when t' e plant was operating. The corrective actions applied to the following LERs: 94-13-01,961101,9618-01,96 20-01,96-22-01,96 23-01,96 28 01 and 96-29-01. In a letter dated December 5,1996, the licensee certified to the NRC that CYAPCo decided to permanently cease power operations and that the fuel had been normanently removed from the reactor. The supplemental LERs issued with the July 7 letter retracted the originally proposed corrective actions that would no longer be implemented due to the permanently defueled state of the plant. The corrective actions in the following supplemental LERs submitted during the period also were not applicable due to the defueled status: LERs 9515-01,9612-01, and 90-08 0 The inspector determined that the supplemental LERs accurately described the -

l issues and that the retraction of the corrective actions was appropriate for the l defueled pla it statu Beoorts of Individual Events LER 94-21-01: Reactor Shutdown Due to IRPI Inaccuracies This LER supplement was issued to update tha NRC on the status of corrective actions to address deficiencies in the individual rod position indication (IRPI) syste The long term plan was to .eplace the IRPI system as a plant modification during the 1996 refueling outage. Since the licensee issued certifications on December 5, 1996 to permanently cease power operations and to remove fuel from the core, no modifications to the IRPI were made and no further corrective action will be implemented. This LER is close LER 9614-01: High inverter Temperatures The licensee issued a supplemental report dated April 24,1996 to described the revised corrective actions for the overtemperature alarm feature on the vital inverters. The originally proposed action was to modify the over temperature feature to alarm only. After further analysis, the licensee justified defeating the overtemperature. alarm feature entirely. This action was based on an evaluation that showed that the maximum temperature that the inverter could experience under worst case environmental and operational conditions was 135 degrees F. The inverter internal components were rated for 150 degrees F. The probability of

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inverter failure due to high temperature effects was deemed to be negligible. A failure of the fan circuit would cause an alarm, and procedures direct operator actions in response to that conditio There were no subsequent problems with inverter operation related to the temperature monitoring circuit. Heat loads in the A switchgear room have been significantly reduced since the plant permanently shutdown and was placed in the permanently defueled condition. This LER is close LER 96-17: Main Stack Sample Performed Late This LER concerned the f ailure to collect a stack sample as required during a plant shutdown. The concerns in this matter were related to those described in LER 97-02. The licensee corrective actions for these issues and the NRC findings were described in Inspection 97 01. This LER is close LER 9618: Feedwater Bypass Valves Might Not Isolate

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l This item concerned tu licensee discovery of a deshn deficiency that might l preclude the feedwater bypass valves from closing during design basis events. The technical issues was similar to those described in LER 9612, and represent another example of old design issues not previously discovered by the licensee. NRC findings for the main feedwater valves and the associated concerns with the design basis were tracked by inspection item 96-06-07. Since the licensee issued certifications on December 5,1996 to permanently cease power operations and to remove fuel from the core, no modifications to the feedwater bypass valves were made and no further corrective action will be implemented. Licensee actions to review engineering calculations to assure they support the safety analysis assumptions will be followed as part of the NRC review of the configuration management plan. This LER is close LER 96-23: Containment Air Recirculation Fans Failed Test This LER concerned the failure of the containment air recirculation fans to provide acceptable flow during testing conducted during the 1996 refueling outage. The failure to deliver acceptable flow under test conditions was a recurrent problem, for which the licensee had proposed corrective actions which included a change to the operating license that would have increased the range of acceptable flow rate NRC concerns with the adequacy of the licensee's corrective action program are tracked by other NRC inspection items, and will be addressed in response to the escalated enforcement action issued in May 12,199 Since the licensee issued certifications on December 5,1996 to permanently cease power operations and to remove fuel from the core, the licensee withdrew the proposed license amendment and no further modificationn to the CAR fan dampers were deemed necessary due to the defueled status of the plant. No further corrective action will be implemented. This LER is close m ,_ )

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LER 96 25: Spent Fuel Building Ventilation Failed Test -

This LER concemed the operability issues with the spent fuel building ventilation system. Licensee actions to correct these issues and NRC findings were described in Inspection 9611. Normal operating procedures (NOP 2.151,2.15 2 and 2.15 3)

were revised to assure the air flow in the SFB ventilation system was adequate prior to moving spent fuel or operating a crane with a load over the spent fuel poo'.' This LER is close LER 96 30: Two Workers Received Internal Exposure This LER concerned the inadvertent contamination of two workers while working in the refueling cavity on November 2,1996. NRC review of the initial event and followup of licensee corrective actions was described in Inspection Reports 9612 and 97 01. This LER is close LER 97-02: Reactor Coolant Sample Not Taken in Defueled Mode This LER concerned the failure to obtain reactor coolant samples after defueling the reactor in November 1996. The concerns in this matter were related to those described in LER 97-02. The licensee corrective actions for these issues and the NRC findings were described in inspection 97-01. This LER is closed, l.ER 97-05: Radiation Effluent Monitors l This LER concerned the failure to perform adequate calibrations of the gaseous and -

l liquid radiological effluent monitors RM-18, RM 22, RM 14 A and RM-148, rendering the monitors inoperable per Technical Specifications 3.3.3.7 and 3.3.3.8. Since the historical operability of the monitors could not be proved, the licensee reported this matter in accordance with 10 CFR 50.73(a)(2)(1)(B) as past operation in a condition prohibited by the technical specifications. The LER provided an acceptable description of the event, along with the licensee's safety assessment. Further licensee actions are in progress at the end of this inspection period to address this matter, as described Section R2.1 below. In the LER, the licensee committed to

- calibrate all affected radiation monitors with new procedures, the review of the cause of the event, and to complete of an assessment of the as found settings on the monitors to assess the impact on accuracy. The licersee plans to address the- '

completion of these items in a supplemental LER. Inspection 97-01 and 97-02 address further NRC reviews in this area, as well as the licensee's performance to meat regulatory requirements (reference inspection item 97-02-01). LER 97-05 is close LER 97-06: RMS-18 Inoperable Due to Low Sensitivity This LER concerned the discovery that the service water effluent monitor RM-18 did not have the sensitivity needed to meet the requirements of the Offsite Dose Calculation Manual (ODCM) with the plant operating in the defueled mode with no circulating water pumps in operation. The monitor was declared inoperable per

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Technical Specification 3.3.3.7, but was already inoperable for the conditions reported in LER 97 05. The licensee reported this matter in accordance with 10 CFR 50.73(a)(2)(1)(B) as plant operation in a condition prohibited by the technical specifications. The LER was acceptable to provide a description of the event, along with the licensee's safety assessment. Further licensee actions are in progress at the end of this inspection period to address this matter, as described Section F. below. The licensee planned to revise the ODCM methodology for calculating the monitor seipoints to assure the value used was applicable to the plant operating configuration. Inspection 97 01 and 97 02 describe further NRC reviews in this area. LER 97 06is close LER 97-07: Potential for SW System Water Hammer This LER concerned the discovery that the service water (SW) lines supplying the spent fuel cooling system were inoperable due to a postulated waterhammer event -

following a loss of normal power (LNP). The cause of the event was that the original plant design and earlier analyses did not recognize the potential for waterhammer. Short term and long term corrective actions were taken to assure adequate pool cooling, including the installation of a check valve in the SW supply e line that would hold the line full of water following a LNP, and thus prevent water ,

hammer. This matter was reviewed in inspection 97 01 and in Section E2.1 abov This LEH is close LER 97-08: Wall Thinning in SW Pipe This LER concerned the discovery that the SW lines providing cooling to the spent fuel pool cooling system were inoperable due to pipe wall thinning caused by microbiologically influenced corrosion. Corrective actions were taken to assure pool cooling. The degraded pipe sections were discovered by exams completed as part of the program to followup past indications of SW corrosion. The degraJed pipe sections were replaced, end additional pipe areas were examined. This matter was described in Inspection 97-01 and in Section E2.2 above. This LER is close LER 97-03: Design Deficiency in Stack Radiation Monitor This LER concerned the discovery of a design deficiency in radiological effluent monitor RM 14A. The sample nozzle in the ventilation stack housing was not drawing an isokenetic sample of the effluents in the pathway. The monitor was declared inoperable per Technical Specification 3.3.3.7, but was already inoperable for the conditions reported in LER 97-05. The licensee reported this matter in accordance with 10 CFR 50.73(a)(2)(ii)(B) as plant operation in condition outside the design basis. The LER was acceptable to provide a description of the event, along with the licensee's safety assessment. Further licensee actions were in progress at the end of this inspection period to address this matter, as described Section R2.1 below. - The licensee was evaluating an alternate sample system that will address the nozzle issue, by using RM 14A as a monitor for gaseous effluents only, and uring a modified RM-148 for isokinetic sampling of stack particulate and lodine effluents, inspection 97-01 and 97-02 address further NRC reviews in this

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area, as well as the licensee's performance to meet regulatory requirements (reference inspection item 97-02-01). LER 97 09 is close LER 97-10: SFP Pump NPSH Inadequate This LER concerned the potential historical conditivn in which the NPSH for the SFP pump may have been inadequate. This item is described further in Section E above. The LER provided an acceptable description of this event, as well as corrective actions to address the issue. This LER is close IV. Plant Sesood R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Spill of Cs-137 Source Material and Personnel Contamination Inspection Scoce (83729)

The purpose of this inspection was to review the licensee actions in response to a =

spill of radinactive material in the chemistry laboratory on June 2, which resulted in a contaminated wound on one techrician and the spread of contamination to six technician Observations and Findinas While preparing a radioactive source to calibrate radiation monitor RM-19 on June 2, a chemistry technician opened a glass vial (ampule) containing 5 milliliters of a 25 microCi solution of Cs-137. The top of the ampute was scored to allow it to be-opened by snapping off the top. When the technician opened the ampule at 9:50 a.m., the top snapped off, but the glass vial shattered, spilling some of the solutio The technician was working over a sink and was wearing a lab coat, two pairs of gloves, and held the ampule in paper towels. However, the shattered ampute punctured the technicians left thumb, causing a wound that became contaminated with Cs 137, subsequently measured at 1000 corrected counts per minute (cpm).

The technician, startled by the incident, walked several feet from the sink towards a contaminated trash container in the laboratory, spilling some of the Cs 137 solution on the floor. Realizing he was injured, the technician called for assistance. Six other technicians in the laboratory responded to help the injured technician and were careful to avoid the immediate area of the trash container. However, over the course of the next several minutes, all six technicians inadvertently tracked through the spill on the floor and became conti.minated. The personnel contamination and spread of radioactive material was idantified as one technician tried to exit the laboratory but was found to be contaminated when he alarmed the PCM 1 personnel monitor. The laboratory supervisor recognized the spread of contamination at that point, controlled personnel movements in the laboratory, and summoned for outside assistance by medical and health physics personne .-

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'44 Health physics and medical personnel responded to assist the injured technician and to establish augmented health physics controls within the laboratory by setting up boundaries for a contamination area, Licensee surveys after the incident identified contamination levels on the laboratory floor in the range from 5,000 to 10,000 dpm/100 cm'. The laboratory areas affected by the spill were cleaned so that all areas were less that 1000 dpm/100 cm 2. The licensee described the event and immediate corrective actions in ACR 97 271. The licensee planned to investigate the need for additional corrective actions. The licensee's dose assessment for the contamination on the technicians finger was reviewed in Inspection 97-06, Conclusions Chemistry personnel and the laboratory became contaminated on June 2 when an ampute broke while preparing a calibration sample. Poor initial response by the responding technicians resulted in the spread of contamination to these individuals and on the laboratory floor. Once recognized, the situation was controlled and staff followup was good to minimize the further spread of contamination, administer first aid to the injured technician, to perform a dose assessment, and to clean up from the event No violations were identifie R2 Status of RP&C Facilities and Equipment R2.1 inoperable Effluent Monitors Insoection Scooe (71750)

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The purpose of this inspection was to review licensee actions to address -

c'eficiencies in the radiation monitoring system first identified by inspectio W 02: Observations and Findinos The licensee declared all technical specification monitors inoperable on February 5, 1997 as a result of an NRC inspection which found inadequacies in the calibration program, (Reference Inspections 97-01 and 97-02). The licensee continued to implement compensatory measures per Technical Specifications 3.3.3,7 and 3.3.3,8 throughout this inspection period as the RMS issues ware addresse The licensee assembled a team comprised of l&C, chemistry, engineering personnel and developed an Action Plan to address all RMS deficiencies in a comprehensive manner Although not addressed in the technical specifications for monitoring plant effluents, spent fuel heat exchanger monitor RM-19 was also affected by the inadequate calibration methodology and was included it the corrective action plan, Addition 61 technical issues were identified and reported to the NRC in licensee event reports. LER 97-06 concerned the discovery that the service water effluent monitor RM-18 did not have the sensitivity needed to meet the requirements of the Offsite Dose Calculation Manual (ODCM) with the plant operating in the defueled mode with no circulating water pumps in operation. LER 97-09 concerned the discovery l

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on April 4,1997 of a design deficiency in radiological effluent monitor RM 14 The sample nozzle in the ventilation stack housing was not drawing an isokenetic sample of the effluents in the pathwa Licensee plans and schedules to recalibrate the monitors were described in letters to the NRC dated February 24 and March 27,1997. To address the calibration deficiencies, the licensee actions included: revising calibration methods and procedures to conduct an insitu primary calibration of the channels; the use of new calibration test stands; the procurement of new sources (Cs 137 and Co-60) to improve the calibration method; and, additional reviews with the RMS vendor to improve the reliability of the Scanrad computer. The inspector reviewed licensee actions periodically during this inspection period to address the RMS deficiencie This review included the development of the action plan, periodic attendance at the RMS meetings, observation of the activities to complete electronic and source calibrations of the RMS, and the licensee actions to assess the as found accuracy of the radiation monitor char nels relative to periods of past plant operation As of July 7, the licensee had procured new sources, constructed new calibration equipment and revised the procedures to calibrate RM 22, RM 14A, RM-19 and RM-18. All four monitors were restored to a fully operable status by the end of the inspection period. A longstanding issue with the SCANRAD computer was resolved through the installation of new software, which was tested and found to provide trouble free operation. Licensee actions continued to develop procedures and to calibrate the stack high range noble gas monitor, RM 14B. The licensee planned to complete the calibrations on all channels, complete the operability assessment and to summarize the progress on the issue in a letter to the NRC. These actions were in progress at the conclusion of the inspectie c. Conclwions The licensee made significant progress to address the RMS deficiencies after the assignment of additional resources to plan, coordinate, schedule and implement an action plan to complete the calibrations. NRC concerns in this area are tracked by inspection 97-02, which is stiil open pending NRC review of the completed calibration work, and a review of the historical operability assessment of the monitor P3 EP Procedures and Documentation Insoection Scope (92904)

The inspector reviewed the recent emergency plan changes to assess the impact on the effectiveness of the EP program.

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b. Observation and Findinos The inspector reviewed revision 34 to the emergency plan during the inspectio The changes to the eniergency plan were made to reflect the change for removal of the onsite duty officer position due to the shutdown defueled status of the plant, c. Conclusions The inspector review emergency plan revision 34 and found that it met the requirements for 10 CFR 50.54(q) and the change will continued to be reviewed during future inspection P4 Staff Knowledge and Performance in EP a. Insoection Scone (92904)

To establish that adequate corrective measures have been taken for violation 50-213/96-07-01(Ability to correctly classify emergency events). That the licensee's root cause has been identified, and that the program procedures and practices have been appropriately strengthened to prevent recurrence. That the licensee has addressed communication problems with the State of Connecticut (Inspection item IFl 96-07 04),

I b. Observation and findinas t

l The onsite inspection was accomplished using table top scenarios for three crews consisting of a shift manager (SM), shift manager staff assistant (SMSA), assistant director technical support (ADTS), assistant director emergency operations facility (ADEOF), and the director of site emergency operations (DSEO). Each crew was given two scenarios (one of the scenarios was a common scenario which was given to all of the crews and was considered a control scenario), which required about an hour-and a half to complete. A critique was conducted where the crew members and the controllers commented on the scenarios and the strengths and areas for improvement, Additionally, the inspector discussed the current plant conditions and the emergency action levels (EALs) which were applicable with the current plant condition of being completely defueled and all fuel stored in the fuel poo The inspector also reviewed lesson plans for EALs and protective action recommendations (PARS), as well as for components and procedures related to the spent fuel pool and spent fuel building ventilation The lesson plans were detailed and appropriate for the present plant conditions. There was an strong emphasis on EAls applicable for the defueled condition. Training effectiveness was measured by evaluating trainees via written examinations or table top exercise The inspector noted that the licensee had taken actions to address the communication interfaces with the State of Connecticut during an emergency. The licensee installed a dedicated phone line in the emergency operations facility for 'ise by the DSEO when making a PAR recommendation to the state. The dedicated

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phone line was installed and has been operational since September 1996 to provide

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a direct link to Department of Environmental Protection personnel. The licensee tested the phone line monthly to assure its continued operational status. The new communications like was also established within the EOF facility at Millstone Station. Licensee actions to sddress this icrue were satisfactory. However, this

, item will remain open pending NRC review of the new communicatio.iinterface during the conduct of the emergency preparedness exercise scheduled for August 21,1997, c. Conclusions Overall, the inspector concluded that training was effective, based upon the review of lesson plans and examinations and by player performance demonstrated in the table top scenarios mentioned above. Additionally, the crews also demonstrated their ability to make appropriate PARS using the new EPIP 4428G " Protective Action Recommendations" procedure, which has been developed, approved, and will be implemented in the near future. The critiques by the participants and controllers were thorough and self assessing. Therefore, violation 50 213/96-07-01(Ability to correctly classify emergency events) is closed.

P8 Miscellaneous EP issues I a. Inspection Scooe (92904)

To ensure that the site emergency response organization (SERO) is maintained in l accordance to the current approved emergency plan.

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The inspector reviewed the following adverse condition reports (ACRs):

ARC 961371 Timely remedial training of an individual in the SERO who had failed a requalification exam.

J ACR 97-061 Poor SERO response to quarterly radiopager test conducted on

, Februar ACR 97-067 The poor SERO response to a followup radiopager test conducted on February ACR 97-088 The inadvertent termination of site access of an employee who transferred to Millstone, which might have delayed the response to Haddam Neck to perform SERO dut ACR 97-100 The poor SERO response to a quarterly radicpager test on February 26.

ACR 97-122 The findings by Quality Assuranca Services regarding emergency plan training in the development of table top exercise . .

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The licensee evaluated each ACR as it occurred and found that none would have precluded implementation of the emergency plan,

' Conclusions The inspector reviewed the corrective measures taken for each ACR and found that the corrective measures were appropriate.

! S1 Conduct of Security and Safeguards Activities S 1.1 Loss of Security Eautoment insoection Scope (71750)

The inspector reviewed the licensee response to a loss of security equipment o i June 13,1997.

' Observations and Findinas

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The plant experienced a loss of security equipment at 11:32 a.m. on June 13 which affected the ability to monitor the status of plant security. The licensee implemented security procedure SCP 1.8-18 and established compensatory 1 measures per the procedure by 11:, . a.m., which included the posting of additional secuiity personnel to monitor the site and assure the continued safety of essential plant equipment.' The inspector toured the site and interviewed licensee and contract security personnel to review compensatory measures. There was no loss of security effectiveness. The inspector reviewed the licensee's response and found it to be in accordance with the procedure. The inspector reviewed the

- planned use of overtime to verify that any extra hours worked by security was approved in accordance with security procedures and that no overtime limits were exceeded. The inspector also reviewed the impact of the degraded security

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equipment with operations personnel and noted there was no impact on plant operational safety. The security system failures were described in adverse condition reports (ACRs)97-299,97-300, and 97-30 Plant support personnel responded to the site during the afternoon of June 13 to investigate the status of security equipment and the cause of the tailures. Security equipment remained in a substantially degraded mode until about 11:00 a.m. on i

June 14 when support personnel identified and corrected a deficiency that allowed restoration of most monitoring functions. Compensatory measures remained in effect until repairs and testing were compl6ted at 9:03 p.m. on June 14. The compensatory measures partially lifted after operational tests to assure the affected equipment was restored to'a proper working order. Some compensatory measures remained in effect until 1:42 p.m. on June 16 pending the repair of test equipment needed to test some equipment (ACR 97 298). The inspector identified no inadequacies in the licensee's response to degraded conditions. Licensee corrective

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actions continued to critique the equipment and personnel performance, and to identify and address the causes of the failures, Conclusions Equipment failure resulted in the loss of security eq*iipment on June 13. The security force response was good to set up and maintain compensatory measure The response by support personnel was timely and effective to identify and correct the cause of the problem V. Manaaement Meetinaa X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on July 15,1997 and on August 5,1997. The licensee acknowledged the findings presente The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified, in addition to the final exit summary, management briefings were conducted during the -

inspection period as NRC reviews were completed. Pre-exit briefings were also conducted on the following dates:

Inspection Reporting Area Dates insoector Insoected April 21-25 Lusher / Silk Emergency preparedness April 29-30 Fairtile/Fredericks 10 CF3 50.59 Safety Evaluations June 9-20 Nimitz Radiological Controls (Inspection 50-213/97-06)

June 24-27 Fairtile/Fredericks 10 CFR 50.59 Safety Evaluations

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PARTIAL LIST OF PERSONS CONTACTED Licensee ,

K. Burgess, Lead Emergency Preparedness Coordinator Gary Bouchard, Unit Director J. Deveau, Emergency Preparedness Coordinator Noah Fetherston, Decommissioning Project Manager John Haseltine, Engineering Director R. Johannes, Director, Nuclear Training Services, Millstone Douglas Heffernan, Maintenance Manager Russell Mellor, Director, Site Operations and Decommissioning James Pandolfo, Security Manager Richard Sexton, Radiation Protection Manager A. Vomastek, Technical Training Supervisor l_ Tom McCance, Nuclear Licensing l G. van Noordennen, Licensing Manager Gerry Waig, Operations Manager J. Warnock, Quality Assurance Manager A. Nerriccio, Public Information E. Maclean, Emergency Plan Training J. Higatti, Technical Training Manager P. Stroup, Director, Emergency Planning Services l F. Crimi, Connecticut Yankee Nuclear Safety Advisory Board P. Bauchmann, Emergency Preparedness Technician M Mort Fairtile, Haddam Neck Project Manager l

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INSPECTION PROCEDURES USED IP 40500: - Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems iP 62703: Maintenance Observation IP 64704: Fire Protection Program IP 71707: Plant Operations IP 73051: Inservice inspection Review of Program IP 73753: Inservice Inspection IP 83729: Occupational Exposure During Extended Outages IP 83750: Occupational Exposure i IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92902: Followup Engineering IP 92903: Followup Maintenance IP 92904: Followup - Plant Support IP 93702: Prompt Onsite Response to Events at Operating Power Reactors ITEMS OPEN, CLOSED, AND DISCUSSED Opened 97 03-03 VIO Inadequate Corrective Action - EDG Testing, SW Water Hammer 97 03-04 URI Degraded SFP Heat Exchanger Performance - TS Change 97-03-05 URI Degraded SFP Heat Exchanger Performance - Corrective Action Closed - New 97-03-01 NCV Inadequate SFP Procedure 97 03-02 NCV Failure to Maintain PORC Composition 97 03-06 NCV Inadequate NPSH for SFP Cooling Pumps 97 03-07 NCV inadequate Safety Evaluation - Operator Action Closed - Previous 95 27-02 IFl RCS Leak Rate Determinations 95-27-01 IFl Daily Technical Specification Channel Checks E7-01-02 VIO Configuration Control 94-27 02 URI Hydrazine Release 94-05-04 IFl Service Water System Lineups 95-02-02 IFl Diesel Tagging Error Causee Flood 96-13-01 VIO Diesel Run with Crank Tool Installed 96-06-05 URI - Actions to Address MIC Corrosion 94-21-01 LER Reactor Shutdown Due to IRPI Inaccuracies 96-14-01 LER High Inverter Temperatures 96-17 LER Main Stack Sample Performed Late

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96 18 LER Feedwater Bypass Valves Might Not isolate 96 23 LER Containment Air Recirculation Fans Failed Test-96 25 LER . Spent Fuel Building Ventilation Failed Test 96 30 LER Two Workers Received Internal Exposure 97 02 LER Reactor Coolant Sample Not Taken in Defueled Mode 97-05 LER Radiation Effluent Monitors 97-OS LER RMS 18 Inoperable Due to Low Sensitivity 97 07 LER Potential for SW System Water Hammer 97 08 LER Wall Thinning in SW Pipe 97-09 LER Design Deficiency in Stack Radiation Monitor 97 10 LER SFP Pump NPSH Inadequate 96-07-01 VIO. Ability to classify emergency events 96-07-02 VIO Ability to provide PARS to the state Discussed 96-10-01- URI Audits of Special Nuclear Material 96-07-04 IFl Communication Link with the State 97 01-08 URI Corrective Actions for SW Corrosion

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LIST OF ACRONYMS USED ACP Administrative Control Procedure ACR Adverse Condition Report ADM Administrative Procedure ANN Annunciator Response Procedure

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AOP Abnormal Operating Procedure CAR Containment Air Recirculation CFR Code of Federal Regulations CMP Corrective Maintenance Procedure CMP Corrective Management Plan CYAPCo Connecticut Yankee Atomic Power Company DCR Design Change Request EDG Emergency Diesel Generator ENG Engineering Procedure F Fahrenheit GL Generic Letter HP Health Physica IR Inspection Report KPl Key Performance Indicators l LDB Licensing Design Basis

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LER Licensee Event Report LNP Loss of Normal Power Event LPE Liquid Penetrant Examination MCC Motor Control Center MIC Microbiologically influenced Corrosion NGP Nuclear Generation Procedure NOP Normal Operating Procedure NOV Notice of Violation NPDES Nuclear Pollution Discharge Elimination System NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation NSO Nuclear Site Operator ODCM Offsite Dose Calculation Manual ODM Operations Department Memorandum PAB Primary Auxiliary Building PDR Public Document Room PMP Preventive Maintenance Procedure PORC Plant Operations Review Committee QA Quality Assurance OAS Quality Assurance Surveillance QC Quality Control RCA Radiological Controllad Area RCS Reactor Coolant System RHR Residual Heat Removh!

RM Radiation Monitors RP&C Radiological Protection & Chemistry

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RWST Refueling Water Storage Tank _

SERO - Site Emergency Response Organization SFB Spent Fuel Building SFP Spent Fuel Pool SFPCS Spent Fuel Pool Cocling System SM- Shift Manager SUR Surveillance Procedure SW Service Water TPC Temporary Procedure Change TRM Technical Requirement Manual TS Technical Specification TSS Technical Specification Surveillance UFSAR Updated Final Safety Analysis Report URI Unresolved item UT Ultrasonic Inspection VIO Violation YTD Year to Date

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E5 Attachment 1 List of Procedures Reviewed The following procedures were reviewed during this inspection as part of the review of procedure quality and adequacy for plant shutdown operatio Operating Procedures NOP 2.101, Spent Fuel Pit Cooling system Operation, Revision 14 NOP 2.151, Startup of PAB Ventilation NOP 2.15 2, Shutdown of PAB Ventilation NOP 2.0-9, Disposition of Control Room and Local Panel Annunciators Maintenance Procedures PMP 9.2-20, Calibration of IST Gages, Revision 12 SNM 1.4-20, New Fuel Assembly and RCCA Packaging and Shipping, Revision 0,1 Surveillance Procedures SUR 5.1-126, All Modes Locked Valve Checklist, Revision 24 (TPC 96 609)

SUR 5.7-217, inservice Test of SW Supply to SFP Check Valve, Revision 2-SUR 5.4 49, Service Water Radiation Monitor (R-18) Calibration, Revision 0 SUR 5.1-0B, Steady State Operations Surveillance (Defueled), Revision 0 Engineering Procedures ENG 1.7-156, Plant System Categorizations ENG 1.7-102, Spent Fuel Pit Heat Exchanger Performance Test ENG 1.7-134A, EG-2A Heat Exchanger Operability Test Administrative Procedures i ACP 1,2-2.42,10CFR50.59 Applicability Reviews and Safety Evaluations, Revision 1

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v t e Inspection Report - Haddam Neck - 1997003
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1997 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) Security Operator Exams Emergency Planning Other

IR 05000213/1997003

Text

REGION I

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