IR 05000213/1998003
| ML20237D414 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 08/21/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237D396 | List: |
| References | |
| 50-213-98-03, 50-213-98-3, NUDOCS 9808260148 | |
| Download: ML20237D414 (38) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.: 50-21 License No.: DPR-61 Report No.: 50-213/98-03 Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06141-0270 Facility: Haddam Neck Station i Location: Haddam, Connecticut Dates: April 14-August 3, and August 13,1998
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Inspectors: Dr. Jason C. Jang, Senior Radiation Specialist Ronald Burrows, Project Manager, NRR Thomas Fredrichs, Project Manager, NRR John Wray, Decommissioning Health Physicist Joseph Nick, Decommissioning Health Physicist William J. Raymond, Senior Resident inspector Approved by: Mark C. Roberts, Chief, Decommissioning and Laboratory Branch Division of Nuclear Materials Safety ;
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EXECUTIVE SUMNIARY Haddam Neck Station NRC Inspection Report No. 50-213/98-03 This integrated inspection included aspects of licensee operations and maintenance, spent fuel safety, decommissioning activities, and radiological controls. The report covers a four
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month period of resident inspection; in addition, it includes the results of announced inspections by regional and NRR personnel in the areas of safety evaluations per 10 CFR 50.59, the preparations for reactor coolant system (RCS) decontamination, and the followup of radiological and operational event Decommissioning Cassetieris and Maintenance: I An inadvertent release of the "A" waste test tank (WTT) occurred while discharging the
"B" WTT due to the inadvertent operation of a valve. A contributing cause was the decision to continue the release after the discharge was automatically secured without fully
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understanding the cause for the termination. One violation was identified in the failure to properly classify the release as an Unusual Event. No Technical Specification limits on radioactive liquid releases were exceeded. Licensee actions to investigate the e.ent, identify causes, and implement corrective actions were acceptable to prevent a future -
inadvertent discharge. The combination of unclear guidance in administrative controls and the failure to properly implement procedures and management expectations for self-checking of actions resulted in mixed performance in assuring proper control of the plant configuration. Several examples were noted of a failure to properly implement plant procedures, which contributed to plant events and constituted a violation of the technical specifications. Further, deficient material conditions in valves that support demineralized operation challenged plant personne Licensee performance was good with respect to testing and maintenance in support of
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decommissioning. Degraded conditions in a spent fuel pool support system (SW-CV-963)
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were properly evaluated, corrected and reported to the NR Soont Fuel Safety:
The training of personnel who perform safety evaluations was performed per the procedure requirements. Areas for improvement include the need to assure open audit issues are brought to closure, and to assure compliance with the technical specifications is included in the audit of plant modifications. An open item will track licensee actions to provide a comparative evaluation of the two calculational results, and a justification for the i conclusion that the spent fuel pool (SFP) building and associated cranes will withstand a seismic event. An open item will follow the adequacy of licensee actions to meet l administrative requirements for commitment tracking. One violation was noted in the
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licensee's process to provide accurate information in support of the licensing basis and/or licensing commitment ii l
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Decommissioning Plannino:
Engineering supported plant decommissioning through modifications for the RCS decontamination and the nuclear island. Administrative controls were adequate to assure design requirements were reflected in the as-built condition of the nuclear island modifications. -Licensee preparations to perform the RCS decontamination were evident in the development of operating procedures, ALARA planning, training and management oversight and involvement. Licensee actions were acceptable to assure the decontamination boundary was properly established after identifying a weakness in the implementation of the tagging program. However, corrective actions for configuration control were not completely effective. Further NRC evaluation continued following a 1000 gallon leak of decontamination fluid, which will be documer.ted in NRC inspection report 98-0 Plant Suncort and Radioloalcal Controls:
Health physics coverage and controls, in particular, were good as the licensee completed activities that had increasing radiological challenges. Licensee activities this period were good to complete radiological assessments at offsite locations. Surveys were thorough to assess the present radiological conditions. Areas for improved performance were identified in the need to thoroughly evaluate offsite properties for the presence of all materials to be-surveyed, assuring areas are completely remediated prior to NRC and State final status surveys, and in assuring timely communications regarding significant finding Licensee actions to address the stack flow and RMS-14Bissues were acceptable. The stack flow calibration deficiencies and the discrepancies in stack effluent monitoring were caused by past inadequate engineering, and a lack of integration in the review and oversight of system design and operating practices. One violation was identified regarding the testing of equipment per technical specification requirements. An open item will track long term corrective actions for stack monitoring end to complete a historical assessment of stack releases. Licensee controls were approp6iate to require that process buildings be maintained at negative pressures during decommissioning.
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REPORT DETAILS l
Summary of Facility Activities The Haddam Neck plant conditions remained stable with the spent fuel safely stored in the i spent fuel pool. There were no significant changes in the plant systems required to l support spent fuel cooling. The licenses continued to work on design modifications to install a new cooling system for the spent fuel pool heat exchangers as part of phase 1 of the nuclear island.
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Special NRC inspections included the review of the plans for characterization and remediation of onsite and offsite areas affected by plant radioactivity, the process to decontaminate the reactor coolant system, the modification of the spent fuel island, and the use of 10 CFR 50.59. On July 23,1998, John Greeves, Director, Division of Waste Management, Office of Nuclear Materials Safety, toured the facility. He also participated in the weekly meeting with the licensee, CT Department of Environmental Protection, and NRC to coordinate offsite characterization activities. On July 27,1998, Commissioner Nils Diaz toured the site and met with licensee management on the status of decommissioning activitie NRC personnel attended meetings of the Community Decommissioning Advisory ,
Committee (CDAC) on April 21, May 27, and July 27,1998. The NRC presented an overview of the Haddam Neck Historical Review Team Report at the April 21 meetin . Decommissioning Operations and Maintenance 01 Conduct of Operations .
01.1 Ooeratina Activities and Soent Fuel Coolina i Insoection Scone (71707,71801)
Using inspection Procedure 71801 and 71707, the inspector reviewed plant status and licensee activities to maintain the plant in the defueled condition and conduct decommissioning activitie Observations and Findinas The licensee maintained stable plant conditions, and adequate level and cooling for the spent fuel pool (SFP). The licensee operated plant equipment necessary to support the SFP, and assured the operability of support systems. The inspector vesified compliance with Technical Specifications (TS) TS 3.9.11, SFP Water Level; TS 3.9.15, SFP Cooling; and, TS 3.9.15, pool temperature below 150 F. T* sr'nt fuel pool cooling system (SFPCS) remained operating per procedure (NOP) , J 1, with the "B" SFP heat exchanger in service. Heat exchanger fouling remaine acceptable through periodic cleaning. The secondary side of the SFPCS was cooled using the service water (SW) system. The licensee conducted routine surveillance i
Topical headings such as ol, M8, etc., are used in accordance with the NRC standardized reactor inspection report outlin Individual reports are not expected to address all outline topic , _ _ _ __
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of the SFP and building per SUR 5.1-OR. Operator actions observed during periodic plant tours were consistent with the procedure The licensee determined on April 22,1998, that TS 3/4.9.15 did not apply to the 4 permanently defueled condition of the plant. On May 5,1998, the staff concluded that the licensee was correct in its determination. On May 7,1998, the licensee submitted a supplement to a pending TS amendment to revise and relocate the provisions of TS 3/4.9.15 to other parts of the TS, along with a commitment to follow the proposed provisions until the license amendment was issued, thereby reinstating the appropriate controls. Due to reduced decay heat load, it was no longer necessary to have a heat exchanger in operation at all times; however, the maximum SFP temperature limit was retained. On May 11,1998, the licensee i revised Technical Specification Clarification C-TSC-093 to reflect the proposed provisions of the TS. The NRC issued License Amendment 193 on June 30,1998, ;
which changed the requirements in the TSs to reflect the permanently shutdown and decommission status of the plant. TS 3.9.16 and 6.8.1.g were issued in Amendmant 193, which included the new TS for SFP temperature and cooling equipmv i functionalit )
Operators activities during April to July were good to monitor and operate plant systems as necessary to maintain spent fuel storage, and to assist maintenance, testing and decommissioning activities. Operators performed well to follow I procedures NOP 2.14-22 and 2.7-3 in June 1998 to transfer water in preparation )
for processing and release. Operators responded well to degraded or off normal I conditions, such as the July 7 valve alignment error that resulted in the spray of primary grade water inside the spent fuel building (SFB). However, operator errors contributed to some events which challenged stable plant conditions. Exceptions to good performance are discussed in sections 01.3 and 01.4 belo .2 Maintenance and Surveillance Activities Insoection Scone (71707(61726,62707)
Using inspecticn Procedure 71707,61726 and 62707, the inspector conducted s periodic reviews of plant status and ongoing maintenance and surveillance.
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' Observations and Findinas
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The inspector reviewed licensee activities to test and repair equipmen Maintenance personnel provided good support in response to emerging conditions, and to address degraded material conditions in preparation for the RCS decontamination. Surveillance tests were conducted well to demonstrate continued operability of SFP support system !
The inspector reviewed and observed the following activities completed in support l of the nuclear island and in preparation for the RCS decontamination:
- removal of the core barrel (PMP 9.5-306)
- diving operations to support core barrel removal (RPM 2.1-12)
- diving operations to support deluge piping installation
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- leakage monitoring of the SFP standpipe
- < inservice testing of valve SW-CV-963 (SUR 5.7-217) ,
- intermediate and spray cooling systems design change installation
- repair and/or replacement of valves and filters in the domineralizer system o installation of reactor vessel nozzle dams
Licensee actions in support of the above activities were generally good to plan and prepare for the task, conduct pre job briefings, coordinate amongst several plant working groups, and to complete the activity in accordance with procedures, work packages and design change records. The heavy load lift over the reactor vessel was very well controlled, with good regard for safety. The control of worker doses ,
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for work in or around the reactor cavity was thorough and showed extensive planning. During preparations for the RCS decontamination, the licensee identified several (diaphragm) valves that support domineralizer operation required repair or repl6 cement prior to starting the decontamination process. The licensee actions related to the stack wide range noble gas monitor, RMS-148, are discussed further in Section R2.1 belo During a routine test of the SW system per SUR 5.7-217 on July 14,1998, SW check valve SW-V-963 passed the full system flow, but could not meet the acceptance criteria for leak tightness in the check position (ACR 98-555).The licensee determined that the valve had stuck in the open position when tested on July 14, but was fully operable following additional testing and troubleshootin The valve tested satisfactorily during periodic tests over the last year. Licensee actions were good to evaluate the system operability, report the condition as operation outside the design basis (EN 34512), and take corrective actions to assure the valve remained operable (the test frequency was decreased from quarterly to monthly). ) Conclusions l
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Licensee performance was good with respect to testing and maintenance in support of decommissioning. Deficient material conditions in valves that support domineralizer operation challenged plant personnel. Degraded conditions in a SFP support system were properly evaluated, corrected and reported to the NRC. Health physics coverage and controls, in particular, were good as the licensee completed activities that had increasing radiological challenges.
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O1.3 Waste Test Tank Release Event (VIO 98-03-01) ;nspection Scooe (71707,61726 and 62707)
! .On Saturday, June 20,1998, approximately 800 gallons of radioactive liquid was inadvertently released to the Connecticut River from the "A" Waste Test Tank (WTT) while radioactive liquid was being released from the "B" WTT. A total of approximately 2200 microcuries of activity (excluding tritium and noble gases) was discharged from the "A" WTT. The simultaneous discharge from both the "A" and
"B" WTTs was not planned. The release was automatically terminated when the alarm setpoint on the discharge monitor was exceede _ - _ - _ - __ _ ______ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _
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On Monday, June 22,1998, the licensee reviewed this event and determined that it met the criterion for declaration of an Unusual Event. The licensee notified the NRC at that time pursuant to 10CFR50.72(b)(2)(vi),but did not retroactively declare an Unusual Event based on guidance given in NUREG-102 Observations and Findinas On June 19,1998, WTTs "A" and "B" were sampled and release permits prepared in accordance with applicable procedures for discharge to the Connecticut River through the normal, monitored release pathway. Release of Tank "A" was to occur first, early Saturday, June 20,1998. The liquid waste discharge line radiation monitor (RMS-22) alarm setpoint was established at 62,000 counts per minute (cpm) above background. Shortly after the release was started, RMS-22 alarmed on high activity and the release was automatically terminated. The licensee flushed the l monitor, believing the background had increased, and attempted to release WTT
"A" a second time. RMS-22 alarmed again at 62,000 cpm and the release of "A" WTT was terminated pending further evaluation. The "A" WTT contained the !
following radionuclides: )
Isotooe Concentration Mn-54 3.01E-06pCi/cc Co-60 5.84E-04pCi/cc Sb-125 1.20E-06pCi/cc Cs-134 1.86E-05 Cl/cc Cs-137 1.37E-04pCi/cc The licensee proceeded to align WTT "B" for release through the same monitored release path to the Connecticut River. At approximately 4:30 a.m. the release began and proceeded smoothly until approximately 11:20 a.m. when RMS-22 alarmed on high activity. The inspector reviewed the chart recording of the tanks volumes as a function of time. The "B" WTT recording indicated a steady release of approximately 30 gpm from 4:30 a.m. to approximately 9:40 a.m. The "A" WTT level remained constant during this time but, starting at 9:40 a.m., showed a definite release until approximately 11:20 a.m. when the monitor automatically ;
terminated the release. The release rate from the "B" WTT decreased slightly I between 9:40 a.m. and 11:20 a.m. such that the total release rate to the Connecticut River remained a constant 30 gallons per minute (gpm).
After the release was terminated, the licensee flushed the monitor and attempted to re-establish the discharge. The "B" WTT release was automatically terminated a second time when RMS-22 reached it's setpoint of 62,000 cpm. Further release of the WTTs was suspended pending evaluation. The operators completed a walkdown of the valve line-up positions and identified valve WD-V-133A (cross-l connect between "A" and "B" WTTs) was slightly open (approximately 10 degrees toward 90 degrees fully open) when it should have be closed. Adverse Condition Report (ACR) 98-0477 was written in accordance with procedure AOP 3.2-2, "High Activity Level", Rev.15, Section The inspector verified the licensee's calculation that approximately 800 gallons of water from the "A" WTT was inadvertently released to the Connecticut River
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concurrent with the "B" WTT release. Licensee operations personnel reviewed the deportability of this event (in accordance with procedure AOP 3.2-2 Section 4.6) l and determined that since there was a release permit for "A" WTT that this event was not unplanned; therefore, the event was not reportabl Emergency Plan Implementing Procedure (EPIP) 1.5-1, " Emergency Assessment Using EAL Table" sections 6.2.1,6.2.2 and Attachment 12.5 states that any unplanned release for which the total activity released (excluding tritium and noble gases) exceeds 1000 microcuries is reportable and should be classified as an Unusual Event. On Monday, June 22,1998, the licensee reviewed this release event and determined that the concurrent release of "A" and "B" WTTs was an unplanned event. At this time the licensee calculated that approximately 2200 i microcuries had been released from the "A" WTT and therefore this event should have been reported as an unplanned release of greater than 1000 microcuries (an i Unusual Event). The licensee made a notification to the NRC on June 22,1998, at 11:48 a.m. per 10CFR50.72(b)(2) (vi). At 5:37 p.m. the licensee provided an update and correction to the initial notification. The licensee did not retroactively declare an Unusual Event based on guidance given in NUREG-1022. Past NRC
- findings concerning weaknesses in event classification were described in inspection item VIC 96-07-01. The failure to recognize and classify an Unusual Event per EPIP 1.5-1 on June 20,1998 was a violation of 10 CFR 50/54(q)(VIO 98-03-01).
The inspector reviewed the event with the licensee and their plans for preventing !
recurrence. The licensee stated that there were no past inadvertent releases caused by an inadvertent bumping of waste system valves. The inspector noted the following immediate and planned long corrective actions. A Root Cause team was immediately formed. All liquid releases from the WTTs were suspended pending further evaluation. The preliminary root cause determination indicated that the
- inadvertent opening of the cross-connect valve WD-V-133A caused the unplanned release. The cause of the discharge was the accidental bumping open of the cross l
connect valve. The licensee removed the valve operator off the valve and !
reinstalled it 180 degrees from the original position making it less likely to be
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inadvertently bumped open. A list of valves was generated that if individually mispositioned could cause a similar event. Locking devices were' installed on WD- !
V-133A and 10 other valves to lessen the possibility that the valves would be i inadvertently operated. The locking devices were considered an interim solution pending the development of additional actions to either modify the valve handles or improve the design of the locking device A review of the mechanical condition of the valves was performed and no problems requiring repair were noted. The operations procedure controlling the releases from the WTTs was revised to include guidance that only one tank can be discharged at a time. This event was discussed with all operations personnel and an article describing the event was included in the site-wide daily information sheet. Plant personnel were briefed regarding the event, the need to exercise care while working around valves, and the need to contact the shift manager immediately if equipment were inadvertently operated. These corrective actions were completed prior to discharging water from the WTTs. Discharges from the recycle test tanks (RTT)
continued since the design of the valves in that system were different and could not be operated by an inadvertent " bump." The cause for the misclassification of the
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event was attributed to unclear and potentially conflicting procedure guidance. The procedures will be clarified and additional training will be provided to certified fuel handlers for emergency action level (EAL) classification of release Conclusion Licensee actions were non-conservative to: (a) continue the "B" WTT release after the discharge was automatically secured without fully understanding the cause for the termination; and, (b) properly classify this event as an Unusual Event in accordance with EPlP 1.5-1. Based on inspection of the work area near WD-V-133A, discussions with plant personnel involved with this event, and review of applicable procedures and event documents, the inspector concluded that the most likely cause of this event was the unintentional bumping open of the cross-connect valve between "A" and "B" WTTs while the contents of "B" WTT were being l properly discharged. No TS limits on radioactive liquid releases were exceeded.
l Licensee actions to investigate the event, identify causes and implement corrective actions were thorough to prevent an inadvertent discharg .4 ,lnadeauste Configuration Control NIO 98-03-02) Insoection Scooe (93702. 71707. 61726 and 62707)
The inspector reviewed licensee actions in response to events involving valve lineup errors.
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' Observations and Findinas Taaaina Error Caused Soillinside of Soent Fuel Buildina At 10:50 a.m. on July 7,1998, approximately 300 to 400 gallons of non-radioactive demineralized water leaked from a primary water system pipe header in the SFB. The spill occurred during work to install new valves as part of the modifications for the nuclear island. The leak occurred when operators added water to the volume control tank, which was thought to be isolated from the piping header to the SFB (ACR 98-532). However, the water was released because valve PW-V-108B was open, even though the valve had been red tagged closed per Clearance 980229 earlier on July 7. The spilled water sprayed several plant workers; none of whom became radioactively contaminated or injured. However, the licenses concluded the event was a significant precursor and a "near miss" for personnel injury (ACR 98-533). A small amount of water leaked from the SFB, but no I radioactive contamination was identified outside the building. The licensee stopped the work and cleaned up the spilled water. An event review team was assembled to investigate the cause of the event. The licensee stopped all work at the site except tasks important to safety or regulatory compliance (CY-GHB-98-094).
i in addition to the valving error on July 7, the inspector noted several recent valve misalignment or tagging problems, including: the inadvertent valve " bumping" error resulting in the June 20 inadvertent release from the waste test tanks (ACR 98-
, 477- See Section 01.3 above); a valve not fully seated while releasing the "B" RTT on June 29 causing level to increase in the "A" RTT (ACR 98-506); the failure to l 6
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close~a drain valve resulting in the inadvertent release of clean resins into the primary auxiliary building (PAB) floor drains on May 27 (ACR 98-397); and, the failure by operators to properly clear tags during maintenance on the Kinney filters j on March 11 (ACR 98-192). Other problems in meeting the administrative controls 1
.were described in ACRs98-557 and 558. The inspector was concerned that there was an adverse trend in the configuration control at Haddam Neck. The inspector I met with the Unit Director on July 8 to discuss these issues, and to obtain from !
CYAPCo management the actions needed to provide assurance that the RCS l decontamination activities could be conducted safely. The licensee acknowledged the above issues and agreed to a telephone conference briefing with NRC .
management, followed by a written submittal, to discuss the assessment of recent events and to describe the corrective action j l
The licensee responded to the NRC in a telephone conference between licensee and i NRC management on July 15 and in letter CY-98-121 dated July 16,1998. The corrective actions to preclude recurrence included: a work stand-down occurred 3 starting July 7, except for work deemed necessary to support safe operation; meetings were held with station personnel to discuss the event and management expectations on completing valve lineups correctly; a generic valve training program '
was initiated and all personnel involved with verifying the position of tagged valves -
were trained; procedures for tagging and verification were clarified to provide expectations for hand-on positioning and verification, and to assure the tagger and -
independent verifier do not work together; a Work Control Team was established to better control and schedule emergent work; the valve positioning and tagging for the entire RCS boundary was completed again and verified; and, actions were taken to address personnel performance and the control of work hours. Additionally, a biased audit of tagged valves was completed to assure no other similar conditions existed. This audit identified two other valves, manipulated by the same operator involved in the July 7 event, that were out of position (ACR 98-557). All valve tagging work completed by the operator was reverified. The inspector observed the completion of the RCS boundary verification on July 17-19, and independently verified the boundary was correctly establishe Despite the efforts described above to assure proper configuration control, the licensee had continued problems in achieving proper system valve lineups. A notable example occurred while aligning t'ne "B" domineralizer during the RCS decontamination. Due to the combination of " sticky" valve operation, a lack of positive visual feedback of actual valve position, and the lack of clear guidance on valve travel, the operator failed to fully open the RCS letdown post filter inlet isolation valve, LD V-238 on July 27,1998. The valve mispositioning caused flow blockage which contributed to the RCS letdown pressure transient and piping vibrations and resulted in the spill of 1000 gallons of RCS decontamination fluid (ACR 98-620). NRC review of this effort continued in NRC Inspection ! Report 98-0 The valving errors discussed above are examples of a failure to follow operating and i maintenance procedures. The failure to properly implement administrative controls l and procedures that assure system configuration control was contrary to TS 6. (VIO 98-03 02).
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l Conclusions The combination of unclear guidance in administrative controls and the failure to properly implement procedures and manag'ement expectations for self-checking of actions resulted in mixed performance in assuring proper control of the plant
, configuration. Several instances were noted in which poor personnel performance l contributed to plant events.
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02 Operational Status of Facilities and Equipment O2.1 RWST Renair and Abandonment
, Inspection Scooe (71707,62801)
l l The purpose of this review was to review licensee actions regarding the refueling water storage tank (RWST).
l Observations and Findinas
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The licensee completed work in February,1998 to seal the bottom of the RWS This activity was reviewed in Inspections 98-01 and 98-02. The temporary leak repair was accomplished by applying a sealant (InstaCote ML) to cover the tank bottom and about two feet of the side wall. The licensee planned to return the RWST to service to allow draining of the reactor cavity when decommissioning
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activities resume. During this inspection period, the licensee began tank restoration l by adding 20,000 gallons of borated water to the RWST and inspecting the tank for leak tightness. Leakage from the RWST bottom was noted, indicating that the l repair was not successful (ACR 98-419). The licensae drained the tank and left the l RWST in dry layup. There were no plans to use the tan Operations Organization and Administration 06.1 Administrative Requirements for Staffina l
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, Insoection Scoos (36801)
The purpose of this inspection was to review licensee actions to implement new I l staffing requirement I Observations and Findinas j
The NRC issued License Amendment 192 on March 27,1998, which changed the administrative requirements in the TSs, including those for personnel qualifications and shift staffing. Specifically, the requirements for operators licensed under 10 CFR Part 55 were replaced with the requirements of TS 6.2.2, which requires that Certified Fuel Handlers be assigned to the facility in the permanently shutdown condition. The licensee developed a training program for certified fuel handlers, which paralleled the requirements of the licensed operator requalification progra Those individuals included the change to certified fuel handlers remained enrolled in the ongoing requalification program. The equipment operator training program
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remained implemented and unaffected by the change to certified fuel handlers. By letter CY98 079 dated May 5,1998, the licensee notified the NRC that the operating licenses for 21 Individuals (12 operator licenses and 9 senior operator licenses) were no longer required. The licensee implemented the new staffing
. requirements during the inspection period. The inspector verified that shift staffing met the requirements of TS 6.2.2 during routine inspection tours, including the
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composition of the on duty shift per TS 6.2.2.a, and the presence of qualified personnel in the control room per TS 6.2.2.b. No inadequacies were identifie Conclusions The licensee met the staffing requirements for License Amendment 19 . Spent Fuel Safety l l
E3 Engineering Procedures and Documentation i
l E Safety Evaluations oer 10 CFR 50.59 (URI 98-03-03. URI 98-03-04) l l Insoection Scoos (37001)
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 50.59. The licensee's 10 CFR 50.59 safety evaluation program was reviewed on June 16 and 17,199 Observations and Findinas Trainina and Qualifications The inspector examined training qualification of individuals who performed 50.59 '
evaluations listed in the Annual Report. The training for 50.59 reviewers at Haddam Neck changed in 1997. The previous training qualification consisted of reading the applicable procedures and doing a safety evaluation under supervisio In-1997, the licensee revised the safety evaluation procedure. The training qualification was then revised to add a training class on using the procedure. The individual who wrote the procedure conducted the training. Training records since
- April,1997 are available at the Haddam Neck site. Records of training prior to that time are generally not maintained onsite. The prior records are kept at the Millstone site. No discrepancies were note Licensee Audits The licensee conducted an audit of Technical Specifications / Operations on June 16-27,1997 (Audit Report CY-97-A06-02). The audit report was transmitted to senior utility management on August 15,1997. Among the deficiencies noted were Technical Specification Section 5.1 and Figure 5.1-1. The " Exclusion Area" was not consistently defined and described in the section and figure. Section 5.5.5
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defines the Exclusion Area and states that the minimum distance to the boundary shall be 1740 feet. Figure 5.1-1 describes the plant site, but shows the nearest
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- distance from the reactor to the Exclusion Area boundary (along the Connecticut River) is approximately 600 feet. The audit report stated that the TS would be submitted to NRC for correction in December,1997. As of the date of the inspection, the submittal had not been mad l The licensee completed Quality Surveillance Report (OSR) No.98-014 on March 12, 1998. The OSR examined Design Change Request (DCR) CY-97-010," Spent Fuel Island Phase 1 Tie-ins and Testing". The QSR recommended that the original DCR be rewritten and proposed a number of actions to ensure consistency in the approach taken in several other DCRs that depended on DCR CY-97-01 The inspector examined the 10 CFR 50.59 Safety Evaluation performed for DCR I CY-97-010, which was dated April 15,1998. The conclusion that no TS change was required to perform the design change was based, in part, on an anticipated change to the applicable TS,3/4.9.15,"SFP Cooling System." The evaluation also stated that the work could be done within the constraints of TS 3/4.9.1 j
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However, TS 3/4.9.15 required the plate heat exchanger to be in operation, a 1 condition that could not be met within the limits of the action statement if the plate i heat exchanger was taken out of service to perform the tie-in, so the work could I not have proceeded within the constraints of TS. In licensee document C-TS , " Technical Specification Clarification", the licensee committed to consider TS 3/4.9.15 applicable until the TS was revised. OSR 98-014 did not identify that the constraints of TS 3/4.9.15 could not have been met, nor that the Technical Specification Clarification required compliance with the T !
The applicability of TS 3/4.9.15 had been discussed with the licensee during a previous site visit by the Project Manager on April 22,1998. At that time, the ;
licensee determined that TS 3/4.9.15 did not apply to the permanently defueled condition of the plant. On May 5,1998, the staff concluded that the licensee was correct in its determination. On May 7,1998, the licensee submitted a supplement to a pending TS amendment to revise and relocate the provisione of TS 3/4.9.15 to 1 other parts of the TS, along with a commitment to follow the proposed provisions until the license amendment was issued, thereby reinstating the appropriate controls. Due to reduced decay heat load, it was no longer necessary to have a 1 heat exchanger in' operation at all times, however, the maximum SFP temperature :
limit was retained. On May 11,1998, the licensee revised Technical Specification Clarification C-TSC-093 to reflect the proposed provisions of the TS. On June 30, 1998, a revision of TS to reflect the permanently defueled condition was issued, which included the new TS for SFP temperature and cooling equipment functionalit The safety evaluation for DCR CY-97-010 also stated that the structural integrity of the SFB had been evaluated, and that the work would not degrade its ability to survive a seismic event. In December,1997,the licensee received a calculation from Yankee Nuclear Services Division that indicated that a number of structural steel members of the SFB and certain components of cranes located in and near the SFB would not withstand a seismic event. QSR 98-014 did not comment on the seismic qualification of the SFB However, the licensee contracted Stevens &
Associates to perform the calculations using modeling parameters considered more realistic than those used by Yankee Nuclear Services, in April,1998, the vendor
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l provided preliminary results stating that of the 150 potential overstressed members ,
i identified by Yankee Nuclear Services,143 had been found to be within acceptance i criteria (ACR 98-519). The licensee stated that the remaining members were under l analysis, and were expected to be within acceptable stress limit )
! The licensee was requested to provide a comparative evaluation of the two i calculational results, and a justification for their conclusion that the SFB and i
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associated cranes will withstand a seismic event (URI 98-03-03). l l Commitment Trackino The licensee maintains a computerized commitment tracking system. Procedure LDI 2.02, "De-fueled Condition Commitment Tracking," requires two reviewers to
- screen each potential commitment. A cursory review of the database revealed i
! seven commitments that were processed with only one reviewer. The licensee l committed to determining how many reviews had been processed with less than l two reviewers, and reporting the result to NRC (URI 98-03-04). j i
The database contains over 13,000 commitments contained in over 8,000 letter Of the total,2080 were identified as potentially applicable to the permanently defueled condition. The licensee has been working through the list to determine !
.which commitments continue to apply in the permanently shut down condition.. The
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l i licensee distributes a newsletter to employees to keep them aware of commitments and how to obtain information about them.
l l I Conclusions The training of personnel who perform safety evaluations was performed per the procedure requirements. Areas for improvement include the need to assure open issues are brought to closure, and to assure compliance with the TSs is included in i the audit of plant modifications (QSR 98-014). An open item will track licensee actions to provide a comparative evaluation of the two calculational results, and a justification for the conclusion that the SFB and associated cranes will withstand a seismic event. , An open item will follow the adequacy of licensee actions to meet administrative requirements for commitment trackin j E3.2 Licensina Basis Discrepancies MO 98-03-05) Insoection Scope
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l The inspector reviewed licensee actions to address a deficiency in maintaining the L licensing basis.
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I Observations and Findinas On June 30,1998, the NRC issued the defueled TSs as Amendment 193 to the
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~ operating license. - The amendment was issued in response to the application provided by licensee letter CY-97-OO6 dated May 30,1997. During a licensee review of the specification and the associated safety evaluation, the licensee identified on July 20,1998 that information provided in the May 30,1997
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application regarding makeup capability to the SFP was not accurate (ACR 98-591).
In the May 1997 application, the licensee stated that..."if offsite power cannot be reestablished within approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a loss of normal power, limited makeup water to the fuel pool could be provided by gravity feed from a tank." The licensee noted that the information was no longer accurate since the tank had been abandoned. The licensee reported this matter to the NRC per 10 CFR 50.72(b)(2)(l)
on July 20,1998 as a condition that placed the plant outside of the design basi Further licensee review of the plant design basis was documented in engineering memorandum CYDE 98-0634 dated July 21,1998. The refueling water storage tank (RWST) was initially credited in the licensing basis as a seismically qualified tank that could provide makeup water to the SFP. By design, the RWST could l provide fuel pool makeup via a gravity feed. On October 9,1997, the licensee approved safety evaluation SY-EV-97-103, which deleted the RWST as the seismically qualified source of fuel pool makeup, and substituted the seismically qualified demineralized water storage tank (DWST). The DWST is not high enough to provide a gravity feed to the fuel pool. The change was justified by the licensee 1 based on other methods to provide pool makeup from the DWST in the event of a i -loss of power. Thus, the licensee concluded that the plant remained within the
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design basis, and retracted the 50.72 notification on July 21,1998. The licensee l submitted a letter to the NRC (CY-98-127 dated July 30,1998) to clarify the design
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l l basis and to correct the licensing basi The licensee engineering review (CY-97-0634) noted that the " limited" RWST ,
f makeup capability was only available from the top few feet of the tank, that is, if !
' the tank level was maintained above the 47 foot elevation. The operator logs for 1997 showed that the maximum RWST level was about 41 feet from May through October, and at 29 feet or less (empty) for the remainder of 1997. Thus, the RWST was not capable of providing the gravity feed to the fuel pool (in the event of a loss
- of normal power) at the time of the May 30,1997 application, or at any time i thereafter. The NRC Safety Evaluation issued with Amendment 193 referenced the l
gravity feed to the fuel pool capability as part of the evaluation for the changes to l
Subsections 3.8 and 4.8 of the defueled TSs.
j-l 10 CFR 50.9(a) requires that information provided to the Commission by the ;
l licensee shall be complete and accurate in all material respects. Although the inaccurate information contained in the May 30,1997 submittal was not solely l relied upon by the NRC in the evaluation of Section 3.8/4.8, the error appears as a l weakness in the licensee's process to provide accurate information in support of the licensing basis and/or licensing commitments. While licensee performance was good to identify and correct the information in the license application, the inspection record contains other recent examples of weaknesses in this area (VIO 96-01-05:
failure to provide a combustible gas detection system; eel 96-201-07:the failure to maintain the licensing basis for service water instruments; and, 97-01-03: the failure to maintain accurate training records). The failure to provide complete and accurate information in the May 30,1997 license application was a violation of 10 CFR 50.9(a) (VIO 98-03-05).
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l Conclusions Apparent weaknesses were noted in the licensee's process to provide accurate information in support of the licensing basis and/or licensing commitment E8 Miscellaneous Engineering issues E Review of LERs and Teleohonic Notifications Insoection Scoon (92700,90712)
The purpose of this inspection was to review prompt reports and licensee event reports (LERs) to verify the requirements of 10 CFR 50.72 and 50.73 were me Observations and Findinos LER 98-01, Seismic Monitor System Not Tested per the Technical Specification This event concerned the licensee's identification on April 9,1998 that the seismic monitor system was not calibrated in accordance with the requirements of TS 4.3.3.3.1. Specifically, the TS requirements for analog chennel operational tests required that the triaxial servo accelerometer and the response spectrum analyzer be
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.. tested to respond within the required range and accuracy of known input value The testing requirements were effective since Amendment 125 was approved on April 26,1990. The seismic monitors were not tested per the requirements since the installed instruments (SSA-302 accelerometer and RSA 50 analyzer) could not be adjusted as defined by the specification. The licensee installed new seismic instruments (reference Inspection 98-01). By letter CY 98-084 dated June 2,1998 the licensee submitted a proposed change to the TSs that would address the testing requirements. The cause of this event was the failure to assure the commitments in License Amendment 125 were consistent with the surveillance procedures. NRC concerns regarding the adequacy of surveillance tests is addressed further in Section R2.1. This LER is close LER 98-02, Failure to Test Fire System per Technical Specifications. This event concerned the licensee's discovery on April 27,1998 that visual inspections of a -
switchgear shaft sprinkler system had not been performed as required. Although presently a requirement of the Technical Requirements Manual, the surveillance had not been performed as required by the plant Technical Specifications during the period from April 26,1990 to February 1,1998, a historical violation of former Technical Specification 3.7.6.2. Licensee actions were appropriate during this period to address operability, deportability and inspection of the fire system in the defueled mode. The cause of this event was the failure to adequately implement
, License Amendment 125 by assuring surveillance procedures reflected the license l commitments. Licensee actions continued at the end of the inspection to review License Amendment 125 to ensure other license commitments applicable to the present plant conditions were incorporated in surveillance procedures. NRC concerns regarding the adequacy of surveillance tests is addressed further in Section R2.1. This LER is close ,
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l' LER 98-03, Failure to Sample Effluent per Technical Specifications. This event concerned the licensee's discovery on May 5,1998 that the compensatory sampling was not completed as required by Technical Specification 3.3.3.7 when radiation monitor RMS-18 was out of service. The specification requires that a grab sample be taken and analyzed once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; one sample was taken about three hours late on May 5,1998. No other samples were missed during the two days the monitor was out of service, and a licensee evaluation showed that releases were within the TS limits. The cause of this event was personnel error. Licensee immediate corrective actions were appropriate. This LER is close LERs 98-04,98-05,98-06, Design Caficiencies identified in RMS-14B and Stack Flow Monitors. The design issues associated with the stack flow and wide range radiation monitoring are discussed further in Section R2.1 below. These LERS are close In addition to the above, the inspector also reviewed the licensee actions regarding event notification made to the NRC per 10 CFR 50.72 on 5/7/98 (EN 34188),
5/12/98 (EN 34213),5/18/98 (EN 34254),6/22/98 (EN 34422),7/15/98 (EN 34512), and 7/20/98 (EN 34545). No inadequacies were identified in meeting reporting requirement In additics to the formal reporting requirements of 10 CFR 50.72 and 50.73, the -
licensee provides courtesy notification to the Resid6nt inspector of events at the site. Following a delay in notifying the NRC of an event on June 22,1998 while the Resident inspector was away from the site, the licensee developed guidelines to assure timely notification NRC Regional personnel (NL/CY-98-070), Conclusions
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Licensee reporting of events was generally acceptable and accurately described the events and followup actions. The LERs listed above are close E8.2 Followuo of Previous insoection items
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(Closed) Unresolved item 96-11-07: Spent Fuel Cooling System Single Fai!ure This item concerned the single failure vulnerability in the existing SFPCS attributed to the power supply design in which both SFPCS pumps are powered frcm the same motor control center (MCC). The licensee implemented modifications for the nuclear island that includes electrically and physically separate electrical power supplies, MCC 2A and MCC 28, that will be made operational during the final tie-in of the modifications with the plant systems. The new system addressed the design vulnerability in the original plant system. This item is close (Closed) Unresolved item 97-03-05: Performance of E10-18. The licent.ee i implemented a trending program to monitor the fouling of SFP heat exchanger E10- l 18. Tha heat exchsnger was back washed as needed. These actions were !
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' effective to assure good thermal performance, This item is close l (Closed) Inspection Issues Related to Plant Operations. Some of the previous inspection issues concerned aspects of plant operation that were important only for
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. continued opera; ion of plant components or equipment. Following the certification to the NRC on December 5,1996 of the decision to permanently shutdown the plant and cease operationr and the subsequent categorizations of plant systems, the corrective actions (plant modifications) for certain items were no longer
. appropriate. The items are listed in Attachment Ill. Decommissioning Suonort Activities E1 Conduct of Engineering E Plant Modifications for Decommissioning insoection Scone (37801f The inspector reviewed the conduct of engineering activities this period that supported shutdown operations, decommissioning planning, and spent fuel safety, Observations and Findinas The inspector reviewed installation of equipment for the RCS decontamination and
.,for the phase 1 of the nuclear island modifications, which were implemented per the i following documents:
e DCR-CY97029, Temporary Power for Decontamination Equipment e DCR-CY97015,480 Volt Motor Control Centers 2A and 2B o DCR-CY97016, Engineered Supports and Piping e DCR-CY97017, Mechanical Equipment Installation e DCR-CY97018, Instrumentation and Radiation Monitoring e DCR-CY97010, Spent Fuel Island Tie-ins and Testing e DCR-CY97015, Reclassification of QA Categories e AWOs 97-3797,97-3798,98-395 and 98-0020 The licensee completed the installation or the major portions of the intermediate and spray cooling systems during this inspection, excluding tie-in with the existing spent 1
' fuel heat exchangers. Continued work on the installations and final system testing was suspended until after the completion of the RCS decontamination. The inspector reviewed the modification activities in progress to verify that the installation conformed with the design and work packages. The inspector noted based on walkdown and sampling review of the field installations, that the as-built conditions conformed with the design details. Equipment conditions, including welding, supports, anchorages and material conditions were generally good. No
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The licensee design for spent fuel cooling includes the use of two fan cooling units, I which were mounted outdoors on the roof of the SFB. The licensee ran the fan l
[ coolers and conducted noise measurements at various locations on site and offsit The measurements showed that the noise from the fan coolers was acceptable onsite, and could not be dstected above general background levels at locations beyond the site boundar ;
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Completion of the nuclear island modifications and NRC review of the process continued at the end of the inspection perio . Conclusions Engineering supported plant decommissioning through modifications for the RCS decontamination and the nuclear island. Administrative controls were adequate to assure design requirements were reflected in the as-built conditio ,
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E2 Engineering Support of Decommissioning Activities E2.1 Plannina and Preparation for Chemical Decontamination ! Insoection Scone (42700,37801)
A review was performed of licensee efforts to plan the chemical decontamination of the primary system. Specific areas reviewed included activities associated with i ALARA planning, process and configuration control, and procedural control , Observations and Findinas
Chemical Decontamination Overview Connecticut Yankee continued with preparations for decontaminating the primary system in order to reduce radiation exposure during decommissioning activitie This licensee chose the Siemen's process called HP/ CORD D UV. These letters describe the decontamination process and represent the following: HP =
permanganic acid, CORD = chemical oxidation reduction decontamination, D =
decommissioning, and UV = ultraviolet light. This will be the first application of <
this process in the United States. However, this process, or a variation of it, has been used worldwide on seven primary systems since 199 The licensee will utilize existing plant equipment to the maximum extent possibl However, some modifications were required to incorporate necessary Siemen's i equipment and to increase flow through the plant domineralizers. The Siemen's equipment includes the installation of a letdown booster pump which will provide 200 gpm flow through the domineralizers. The use of the installed purification pump will provide an additional 100 gpm flow for a total of 300 gpm. This flow rate will help prevent the precipitation of particulate (e.g., nickel from the steam generator tubes) and remove process fluid radioactivity more quickly. The HP/
CORD D UV process will be controlled using a heater skid, two chemical addition skids, and three ultraviolet light skids. The flow path will be from the residual heat removal (RHR) discharge through the core deluge piping through the heater and
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ultraviolet light skids and finally through the pressurizer via an eight-inch header installed on top of the pressurizer manway, i-( Preparations The inspector reviewed the status of the preparations during the week of June 29 -
July 3,1998. This review included a discussion with the decontamination manager l
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to determine the scope of the project. The inspector conducted a field walkdown of
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the Siemen's equipment located in containment. The contractor personnel who were assembling the equipment were questioned to ascertain their level of understanding of the process and potential problems including material compatibility ;
issues associated with carbon steel. The contractors displayed an excellent i knowledge of the licensee's systems and decontamination process. In addition, the l contractor and licensee had determined that the decontamination process would !
only be performed on stainless steel components including the contractor's i aquipment and any sample sinks installed by the license Carbon steel is a concern during chemical decontamination because the chemicals attack this material more aggressively and could cause component failure in these areas resulting in leakage of the processing fluid. The licensee conducted a material review of the main coolant pump (MCP) seals and determined that they were j compatible with the processing fluid. Leakage is not expected from the MCP
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leakoff lines because of the relatively low flow and pressure in the syste During the walkdown, the inspector observed a contractor with both anti-c gloves off and his anti-c suit partially unzipped. The inspector considered this a poor work practice and questioned the contractor. The contractor responded that this was an
. acceptable practice when retrieving items inside the anti-c suit (e.g., pens). The .
Inspector later learned that this was not the accepted practice and this issue will be briefed to workers to ensure they have a clear understanding of radiological controls l inside the containment which is a posted contamination area, j The licensee controlled the chemical decontamination activities through Master {
Decontamination Procedure SPL-10.11.1. Before any employee is allowed to perform work associated with this procedure, they must be qualified as "decon personnel". As a minimum, this requires attending one of the four hour RCS decon general overview training sessions provided by the licensee. The inspector sat in on portions of this training and reviewed the lesson plan. This training included an :
overview of the chemical decontamination process, the plant modifications made, l the project organization, and chemical and radiological concerns. The inspector i found this training to be thorough and easy to understand and considered this a good work practice before performing a process of this nature, in addition to this training, the radiological protection staff involved with this process was given a separate 45-minute training session on the radiological aspects of the process. This training included greater detail on the radiological concerns and controls that will be used during this activit The inspector reviewed the licensee's procedure for handling spills of chemicals and precess leaks (SPL-10.11-21). This procedure is based on three levels of leak rate For leak rates less than 0.5 gpm and if the leak cannot be repaired, decontamination operations may continue as long as the leak can be contained and the liquid is i- returned to the system during the cleanup phase. For leak rates greater than !
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- gpm, the decon process will be' secured and the leak contained and repaired. The third level is for intermediate size Mts between 0.5 and 1.5 gpm. The procedure '
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states that these leaks should be repaired if possible during the decontamination l operations. It also states that after a cleanup phase of operations the leak shall be isobted and repaired. The inspector questioned the licensee how a leak of this
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' magnitude will be contained considering that a fuli ycle lasts approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />. The licensee responded that it is their inteia to stop the decon process if a
. leak of this nature cannot be stopped. They will review this procedure and make
. any necessary changes to ensure that it reflects their expectation The inspector reviewed the licensee controls in the Master Decontamination l Procedure SPL-10.11-1 and AOP 3.2-70 to verify adequate controls were in place to control the plant operating system during the decontamination process. The licensee used a combination of plant and vendor personnel to monitor and control the plant and vendor equipment. The licensee estabi!shed responsibilities and lines ;
of communications, and installed and calibrated necessary instrumentation and established requirements for monitoring and recording key parameters periodicall The inspector verified that the instrumentation was installed and operating in accordance with SPL 10.11- Decontamination Boundarv and Configuration Control i
The licensee planned to decontaminate the reactor coolant system (RCS) (excluding !
'the reactor vessel) and the attached piping, including portions of the pressurizer, steam generator water boxes, the residual heat removal system, the charging and
. volume control system, the RCS drain headers, and portions of the safety injectio and purification systems. The system lineups and valve configurations needed to -
accomplish the decontamination were dafined by procedures SPL 10.11-1 and SPL 10.11-3. The plant residual heat removal and purification systems were used ,
in conjunction with the vendor equipment to flush the piping systems and remove l the radioactive contaminants. The licensee modified plant systems and performed a hydraulic analysis of the flow boundary to assure adequate flow throughout the system. The inspector reviewed the modification activities in progress and verified the new plant configurations supported the decontamination boundary. Plant modifications to support the decontamination included the following:
DCR CY 97009, Artifact Removal MMOD 98507, RPV Head Deluge Jumper to Pressurizer MMOD 98509, Letdown Booster Pump i MMOD 98508, Letdown Return Pump i MMOD 98511, SFP IX Return Jumper MMOD 98512, Letdown Orifice Removal and Piping Replacement DCR 97029, Temporary Power to Decontamination Equipment i
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The licensee established a system valve lineup that used red Og controls to define the decontamination boundary per Section 7.1 of SPL 10.11-3. The initial valve lineup for systems within the boundary were defined in Section 7.2 of SPL 10.11-3, and a yellow DECON tag was created to mark the flowpath valves within the boundary. Double valve isolation, pipe caps and modified piping configurations were used to minimize the potential leakage of radioactive fluids outside the
! decontamination boundary. Through a sampling review, the inspector verified that the boundary was correct to accomplish the RCS decon and that the valve lineup was implemented as stipulated in the licensee procedures and administrative controls. Following events that indicated a weakness in the tagging controls (reference Section 01.4 above), the licensee conducted additional verifications of
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~the decontamination boundary and valve lineup during the period of July 17-1 The inspector monitored the implementation of the redundant and independent verifications, and noted that the licensee measures provided assurance that the plant was in the desired configuration for the decontamination. Besed on the above, the NRC concurred with the licensee's plan to proceed with the RCS decontamination, which occurred on July 20,199 i ALARA olannina Licensee memoranda HP-97-0155, dated July 29,1997, and HP-97-02112, Rev.1, dated August 15,1997, provide the radiation exposure estimates for decommissioning the Haddam Neck plant. It is estimated that decommissioning th s i plant without decontamination 0 the primary system would result in 1970.8 person-rems. A chemical decontamination of the primary system using the Siemen's method with a dose reduction factor of 15 would lower the exposure to an estimated 934.4 person-rem The inspector met with the radiological engineering supervisor to discuss issues !
related to radiation exposure control. The inspector reviewed the-licensee's method for tracking exposure during chemical decontamination preparations. The inspector
- -noted that the process is broken down into specific work activities (e.g., core barrel removal / replacement) and the exposure for each activity is tracked against estimated exposure goals. The exposure goal for the entire chemical decontamination process is 39 rems. One activity of particular interest is the removal of the core barrel. The licensee changed the method for removing the core barrel by lowering water shields in place before lifting the core barrel. This resulted in an exposure of 2.346 rems compared to approximately 12 rems for the previous core barrel remova The licensee had sent several employees to the chemical decontamination efforts at both Big Rock Point and Maine Yankee power plants. The inspector questioned the licensee as to what they had learned and how they had incorporated this knowledge into their planning efforts. In response to high radiation levels on processing filters experienced at other plants, the licensee has removed the filters from the filter j housings and will rely solely on the resin to remove particulate contamination. In !
response to particulate settling out on plant piping and creating high radiation areas and additional exposure, the licensee has maximized flow through their system as described earlier. In response to failed batteries in teledosimetry units, the licensee is using hard wired units. This will allow the licensee to monitor the progress of the decontamination without incurring dose to change batteries. To help minimize waste, the licensee will perform their initial crud burst using older resin. In addition, a management organization was created to accomplish this project with clear lines of responsibilities established. The licensee's efforts to incorporate industry lessons learned were well planned and thorough.
i Implementation of the RCS Decontamination I
After completing the required prerequisites and system leakage checks, the RCS was heated up to 200 degrees F to begin the chemical decontamination proces The start of the decontamination was delayed as the licensee conducted required
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maintenance on domineralizer system filters and valves which failed on demand prior to the start of chemicalinjection. After injecting chemical for the first chemical application of July 27,1998, the licensee responded to severe piping vibrations and the leakage of about 1000 gallons of RCS fluid. The decontamination process was delayed pending the resolution of that event. NRC 4 review of the decontamination process and the response to the leak is being provided in a special inspection (reference inspection 98-04). Conclusions Licensee preparations to complete the RCS decontamination were evident in the development of operating procedures, ALARA planning, training and management oversight and involvement. Licensee ar;tions were good to assure the decontamination boundary was properly established after identifying a weakness in the implementation of the tagging program. Further NRC evaluation continued following a 1000 gallon leak of decontamination flui :
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R1 Radiological Surveys and utamination Control R1.1 Offsite Radiation Survevs and Remediation Plannino Inspection Scone (83726)
The purpose of this inspection was to continue the review of licensee radiation and contamination surveys at private residences and properties located offsite that had received cement blocks or other material from Haddam Neck plant in the past. The inspector reviewed the planning and implementation of remediation plans at offsite location Observations and Findinos inspection item 97-09-04 concerned NRC review of the licensee actions to survey offsite areas that received soils, concrete blocks and other potentially contaminated !
equipment from the plant, and to recover contaminated materials for proper disposal as radwaste. Past NRC reviews in this area were described in Inspections 97-09, 97-10 and 98-01. The inspection methodologies described in the referenced reports continued during this period. Soil samples obtained from offsite properties were analyzed were split with the licensee and analyzed at the NRC Radiation Laboratory !
in King of Prussia, Pennsylvania. NRC review of the licensee and NRC data continued at the end of this inspectio ' The following activities occurred during this inspection period to verify that licensee i effectively controlled radioactive materials, and performed adequate survey (1) On March 24,1998, the NRC issued a letter to CYAPCo regarding the offsite remediation plans. The licensee responded with a generic final status survey plan. Site specific final plans are submitted at least 3 days prior to offsite remediation. The schedule for offsite work was updated as progress was i
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made. The offsite surveys and remediation efforts are expected to continue through the Fall of 1998. As of August 3,1998, the licensee was tracking 110 offsite survey locat;ons, and had accounted for 5031 of the concrete blocks released in 1975. The licensee had completed surveys and remediation at 56 locations, and was processing the survey areas and files for closure. The remaining 54 locations were in various stages of activity, including investigation, scoping surveys, remediation and restoratio Licensee, NRC and Connecticut Department of Environmental Protection (DEP) representatives began weekly status meetings to review the status of the offsite remediation efforts. The meetings were held either at the plant site or at DEP Offices in Hartford, CT. NRC Region I and NRR personnel participated by teleconference as necessary. The meetings were beneficial to assure better communications and coordination of efforts, and provide joint oversight of progress on the offsite efforts and the closure of icurvey l area (2) - The licensee prepared specific plans to survey and remediate affected offsite locations. The plans were provided for review by the NRC and the DEP. The plans were generally acceptable and reflected the criteria in the generic plans. The inspector verified during observations at the offsite job sites that the remediation plans were available at the job site, in use and followed by !
the licensee technician In general, licensee performance was very good to conduct the offsite surveys, recover blocks and implement the remediation plans to assure no plant related materials remained offsite. One area for improvement was the r eed for the licensee survey teams to identify all the blocks at the survey locations. While the licensee technicians were generally thorough to identify all target blocks, some blocks were identified after remediation activities had begun. Additional blocks were identified by various oversight organizations i (NRC, DEP, Quality Assurance (QA), ORISE) at Locations 9624,9632,9642, and 9657. This matter was discussed with the licensee, who stated the lessons learned would be used to enhance the proces Another area for improvement was for the licensee to ensure that scoping and remediation efforts were complete prior to NRC and DEP involvement to conduct confirmatory final status surveys. The licensee stated that lessons learned from Locations 9624 and 9632 would be used to enhance the proces (3) Although the vnt majority of materials checked did not show any radiation levels above background, the surveys continued to identify concrete blocks
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and other materials with very low levels of plant related ruoactivity. None
of the contaminated items had significant radiation dose retes. The blocks
! and materials were returned to Haddam Neck for disposal. Areas where the
~ contaminated materials were stored were checked to assure no residual contamination was present. In some cases, radioactivity from the contaminated blocks had washed into the soil; the soil was excavated as
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' needed to assure there was no detectable plant related activity left at the locatio During a survey at Location 9639 on July 6,1998, the licensee identified fixed and removable contamination on a we! ding unit. Subsequent isotopic analysis of the contamination removed from the unit identified transuranic radionuclides (ACR 98-609). Tnis was the first time transuranic radionuclides were identified offsite since the start of the Haddam Neck material recevery program. The welding unit was returned to the plant, and the licensee had recovered the transuranic activity identified at Location 9639. Licensee action in response to this finding continued at the end of the inspection, including additional surveys at the affected location and the followup on other welding units that had been released from the plant. The licensee plans to update LER 97-21 to reflect this findin l The inspector identified a need for improved communications from the licensee regarding this event. Following the July 6 survey, the licensee had preliminary analytical results of transuranic activity on July 8, but failed to disclose this significant finding to the NRC until the nspector ide.itified the !
information during routine rmews on July 15. NRC concerns regarding the need for timely communication of potentially significant findings offsite were ]
j discussed with in a meetiqg with the licensee and DEP representatives on July 16 (ACR 98-543). Past NRC concerns on the timely communication of ]
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Information was described in inspection 97-05. The licensee noted the inspector's comment (4) The licensee continued to survey properties that had received soils from Haddam Neck in the past, including Location 9606. Based on the direct radiation measurements, there were no anomalous dose rates identified at the site. However, soil sample results showed very low levels of detectable plant related radioactivity. NRC, DEP and CYAPCo soil analyses identified plant related radioactivity at six of the survey units at this location. The NRC results were comparable with those by CYAPCo and DEP; the NRC results are shown below:
Samole N Co-60 (oCi/a) ,
I 9606SSOO5 0.014tO.OO7 i 9606SSOO7 0.039*O.009 ;
9606SSO22 0.032 i O.015 The soil concentrations of Co-60 detected were below the Radiological Environmental Monitoring Program (REMP) environmental lower limits of 1 detection (LLDs). The licensee followup and assessment of ibis location was in progress at the conclusion of this inspection. The licensee planned to develop a full scope characterization survey for this sit The soil characterization results were comparable ,to previous sites where low level radioactivity was found, and for which dose assessments had been completed using the calculational methodologies of RESRAD Version 5.0.
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- The inspector's independent asses'3 ment for Location 9606 and found comparable results (maximum do',e a small fraction of that from natural background in the area). NRC raview of the site was in progress at the !
conclusion of the inspectio (5) The licensee developed a Technical Basis Document regarding the te@nique to distinguish between plant related Cesium-137 (Cs-137) and those l concentrations now found in the background in the environment. The l licensee adopted a' plan to differentiate plant related Cs-137 using a nonparametric statistical sampling methodology, which was discussed during a meeting on July 9,1998. The Technical Basis Document was sent to the NRC for review (reference copy of letter from R. Mellor to K. McCarthy, j Connecticut Department of Environmental Protection, dated June 12,1998).
The NRC staff determined that the method proposed by the licensee was appropriate for this application. The licensee established criteria for Cs-137 concentrations as follows: levels below 1.68 picocurie / gram (pCi/g) require !
no further remediation; levels between 1.68 and 2.69 pCi/g would receive j further evaluation; and levels greater than 2.69 pCl/g would require further remediation. The NRC staff provided comments on the plan, including the number of iterations that could be taken while applying a Wilcoxon Rank Sum screening test to the sample results. The licensee noted the NRC comments and stated the plan would be changed to include an administrative limit on the resample schem The licensee began development of a Technical Basis Document to perform dose assessments of plant related radioactive materialidentified offsite. The
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computer calculation will use a model to back calculate dose rates to the time of release from the plant. The licensee was developing a method to approximate the mix of radioisotopes on the blocks at the time of release, using plant records and measurements of the radioactivity currently on a representative sample of recovered blocks. The licensee will present the TBD for NRC and DEP review. Licensee efforts in this area continued at the end of this inspection perio (6) The Quality Services organization reviewed the offsite remediation activities i
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and was effective in evaluating the program controls and identifying areas for improvement (reference QSR 98-42 and ACR 98-504).
Preliminary Dose Assessments NRC findings continue to show that no significant radiological safety concerns exists from any of the Haddam Neck materials recovered to date from offsite locations. This conclusion was based on NRC and licensee data from direct gamma exposure rates and measured soil concentrations, and preliminary dose l- assessments. The dose assessments considered the potential whole body dose and I
accounted for occupancy factors. Consideration of potential exposure from transuranic will be included in the final dose assessments. Since the above results correspond to dose rates after 23 years of decay, the licensee and NRC followup asseesment were in progress at the conclusion of this inspection.
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NRC review of all sites ivas in progress at the conclusior' of the inspection, item 97-09-04 remains unresolved pending the completion of the licensee actions to identify and recover con: rete blocks, and evaluate other plant related material conclusions Licensee activities this peciod were good to conduct followup survey and assessments at offsite locations that had received potentially contaminated blocks and miscellaneous materials. Licensee surveys of offsite areas were thorough to assess the present radiological conditions. Areas for improved performance were identified in the need to thoroughly evaluate offsite properties for the presence of all materials to be surveyed, assuring areas are completely remediated prior to NRC and State final status surveys, and in assuring timely communication with the NRC and State of Connecticut regarding significant findings. NRC review of licensee actions to assess and remediate the offsite areas continued at the end of the inspectio R2 Radiological Protection and Chemistry (RP&C) Facilities and Equipment R2.1 Stack Flow and Radiation Monitors (VIO 98-03-06. URI 98-03-07) Insoection Scooe (84750)
The inspection consisted of a review of licensee actions to address discrepancies in the measurement of plant stack effluent Observations and Findinas Summarv of issues During engineering reviews in October,1997, to validate the design basis for the plant, licensee testing of the spent fuel building ventilation system identified stack and ventilation flow rates different than assumed (LER 97-18). The recognition that different ventilation system flows would impact the measurement of stack effluents led to further investigation of the stack flow and rahtion monitoring instrumentation. During this period, the licensee identifd several issues regarding the ability to monitor effluents via the main stack. The issues included discrepancies with isokinetic flow m6asurement, sample line deposition, total stack flow, and containment purge flow. The licensee identified that:
e The sample for RMS-14B was not drawn from the process stream in an isokinetic fashion. The cause for this condition was flow from the SFB that was not accounted for, inadequate functional testing of the original monitor design; and incorrect data used for stack radiation measurement (LER GOM l
- The stack total flow channel F-1101 did not accurately measure stack flow, nor was it capable of covering the entire range of flow through the stac l This was a condition outside the design basis of the plant, and a condition contrary to TS 3.3.3.8. The cause for this condition was the failure to 24 . q
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properly implement a 1974 design change of the primary auxiliary building ;
ventilation system. (LER 98-05) )
i e The RMS-14B sample lines were designed such that losses of particulate and iodine in the sample path were not properly accounted for. The cause-for this condition was inadequate original designs and inadequate material conditions. (LER 98-06)
The licensee committed to resolve these issues prior to the RCS decontaminatio l The licensee developed an action plan, Corrective Action Plan for RMS-14B dated June 15,1998, that addressed sample line deposition (ST 11.7-217); ventilation {
flow rate testing, installation of temporary flow monitoring instrumentation with '
associated procedures, and adjustments to the RMS-148 sampling system (SUR 5.2-69). Licensee responses to each discrepancy were appropriate to implement !
compensatory measures, institute short term corrective actions, conduct operability reviews and notify the NRC. The status of corrective actions were summarized in letter CY-98-121 dated July 16,199 NRC inspections during this period included verification the licensee actions, including: (1) mview of corrective actions, including compensatory measures; (2)
verifying the current methods to measure the main stack flow and containment purge flow; (3) a review of calibration results of the main stack flow rate monito i and the wids range gas monitor, RMS-148;(4) revicw of licensee engineering )
evaluations; (5) review of actions to meet commitments and Technical Specification requirements; (6) reviews of system geometry contributing to line loss and observation of system testing; (7) review of the projected public dose assessment; and, (8) confirmation of system status to support RCS decoritamination, as described in the July 16,1998 letter, inadeouate Calibration Method Prior NRC review of the isokinetic /anisokinetic sampling line for Rd4A/B was documented in inspection Report 97-12. The line loss for radioactive particulate and radiolodines was identified by the NRC and documented in NRC Inspection ,
83-12. The stack flow monitoring, RMS-14B design discrepancies and SFB ventilation system design deficiencies were additional historical example of the types of issue cited in the May 1997 escalated enforcement action. The 1974 design inadequacies were addressed in a prior NRC inspection (reference inspection item VIO 97-09-02).
The inspector discussed with responsible individuals relative to the line loss study performed by a vendor. The licensee stated that the vendor used reasonable assumptions to calculate the line loss. However, the length of the sampling line was about 250 ft with many bends and turns. Therefore, some assumptions may l not be valid in this case. The licensee performed line loss testing using DOP to
! avoid assumptions and associated uncertainties (AWO 98-1828). The licensee identified the source of the losses as a loose fitting in the sample tubing. The affected sections of the sample line were replaced.
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The main stack flow monitor channel (FT-1101) read about 51,000 cfm at the time of this inspection. Annual calibration results (1994,1995, and 1996) of the main stack flow rate monitor, as per Procedure SUR 5.2-82, Total Flow to Stack Channel Calibration, were within the licensee's acceptance criteria. However, this calibration procedure did not include a sensor (pitot tube). Section 1.4 of the TS defined the channel calibration, in part, "The Channel Calibration shall encompass the entire channelincluding the_ sensors and alarm,...." This was a violation of TS requirement (VIO 98-03 06).
This violation is one of several examples identified during this period involving the failure to meet TS requirements for surveillance testing and calibrations. Other examples are described in Section E8.1 above, which were attributed to the failure i to adequately implement License Amendment 125 by assuring surveillance procedures reflected the license commitments. Licensee actions continued at the end of the inspection to review License Amendment 125 to ensure other license commitments applicable to the present plant conditions were incorporated in surveillance procedures. The NRC identified concerns in the past regarding the i adequacy of surveillance tests relative to License Amendment 125 (VIO 92-04-01). l The inspection record-also reflects the past failure of plant procedures to implement the TSs (reference LERs 98-02,98-01,97-15,96-04,95-12,94-14,92-22,92-06, 91-09 and 91-07.) The LER history indicates the licensee's past actions were ineffective to assure technical specification requirements are adequately reflected in plant procedure Acceptance criteria of Procedure SUR 5.2-82 were established based on the vendor's calibration curve. The calibration curve was constructed based on the stack flow (feet / minute), and associated pressures (inches of water), which measured by a pitot tube at the design fan capacity (cfm). When the total fan capacity was upgraded in 1974, from about 70,000 cfm to about 113,000 cfm, the calibration curve should have been re-evaluated and re-validated, i.e., bacause the pressures measured by the pitot tube might have changed due to higher stack flow rate (113,000 cfm). The calibration range of the instrumental channel
- including readout device should siso be evaluated. The calibration range of the instrument channel is zero to about 87,100 cfm, which is lower than the actual total main stack flow rat !
Lona Term Actions Licensee actions in the area continued at the end d Me inspection period to address long term improvements to the stack effluent montw.ing. The licensee planned to remove, inspect, and replace as necessary, the main stack flow element; evaluate relocating the sample skid portions of RMS-148 to the upper level of the PAB, and connect the isokinetic probe to the RMS-14A; modify and relocate the containment purge flow element such that containment purge is indicated for all ventilation configurations; and, evaluate the requirement for two PAB fan operation and the methodology for measuring the maximum containment purge flow rate. The licensee evaluated the projected dose impact to the public due to using incorrect stack flow. However, the licensee recognized that this evaluation had a limited scope. The licensee will perform further evaluation for the projected dose impact to the public incorporating the: (1) correction of the line loss; (2) compensation of
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anisokinetic sampling line; and (3) correction. of the main stack flow rate. The licensee committed to update LER 98-05. The completion of licensee long term actions described above, and the evaluation of the dose impact from the stack effluent pathway is an unresolved item (URI 98-03-07).
c. Conclusion Licensee actions this period to identify, investigate and address the stack flow and RMS-14Bissues were thorough and appropriate. The stack flow calibration deficiencies were caused by inadequate engineering evaluations when PAB fans were upgraded in 1974. Critical engineering evaluation items were missed, such as re-validation of the flow calibration curve, flow channel calibration range including readout device, and the calibration methodology. The multitude of discrepancies in the stack effluent pathway indicate a past lack of understanding of design and operational requirements. Chemistry, Health Physics, l&C, Operations, Engineering, end Corporate personnel had responsibilities and were involved with the effluent control programs. However, there was a lack of integration in the review and oversight of system design and operating practices of the effluent control program by a single designated grou R2.2 Plant Ventilation System Air Balanga insoection Scone (84750-03) j The inspection consisted of reviews of the ali balance relative to the containment building, SFB, PAB, and waste disposal building (WDB). Supply and exhaust air {
capacities (cfm), and safety evaluations are described in Section 9.4.3 of Decommissioning Updated Final Safety Analysis iteport (DUFSAR). Observations and Findinas A wmplified ventilation systems diagram was prepared by the licensee during this inspection, as shown in Figure 1. Buildings shown in Figure 1 must be maintained at a negative pressure in order to prevent unmonitored releases. The negative pressure can be rea'd directly at the installed pressure gauges or calculated based on supply and exhaust air capacity. For example, there are three gauges at the SF All of these gauges displayed (indicated) negative pressures on May 28,199 There was no installed pressure gauge for the WDB. As shown in Figure 1, supply air capacity is 11,250 cfm and exhaust air capacity is 12,350 cfm. Therefore, the WDB maintained at a negative pressur Exhaust air from containment building, SFB, PAB, and WDB flow through the air cleaning systems, denoted as HEPA/HECA in Figure 1. HEPA and HECA filters are designed to remove radioactive particulate and radioiodines, respectively. If the licensee maintained negative pressures for these buildings shown in Figure 1 and no by-pass the air cleaning systems, then any radioactive particulate from these
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buildings will be removed at the 99% removai aate (e.g., Section 3/4.7.11.a.1 of the TS for the PAB).
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Prior to the major decommissioning activities in the containment building, SFB, PAB, (
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and WDB, the air cleaning systems must be tested and potential by-pass releases must be inspected to prevent unmonitored releases to the environmen Conclusion Licensee controls were appropriate to require that the containment building, SFB, PAB, and WDB must be maintained at negative pressures during decommissioning; the air cleaning systems must be tested at the required frequency; and, the potential air by-passes of air cleaning systems must be inspected periodicall V. Manecement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management periodically during the inspections, and at the conclusion of the inspection on August 13, 1998. The licensee acknowledged the findings presente The inspectors asked the licensee whether any materials exemined during the inspection should be considered proprietary. No proprietary information was identifie i f
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e PARTIAL LIST OF PERSONS CONTACTED Licensee Russell Mellor, Vice President Operations and Decommissioning
= Gary Bouchard, Unit Director Kerry Harner, Chemistry Manager Doug Heffernan, Maintenance Manager Gerry Waig, Operations Manager James Pandolfo, Security Manager ,
Richard Sexton, Radiation Protection Manager I Gerry van Noordennen, Nuclear Licensing j Pete Hollenbeck, Site Characterization Supervisor i Keith Sickles,-Design Engineer Edward Bingham, Engineering John Haseltine, Engineering Director t Jay Tarzia, HP/ Chemistry Technical Support INSPECTION PROCEDURES USED IP 36801: Organization, Management & Cost Control IP 37001: Facility Modifications IP 37801: Safety Reviews, Design Changes, and Modifications at PSRs IP 42700: Plant Procedures IP 61726: Surveillance Observations IP 62707: Maintenance Observations IP 62801: Maintenance and Surveillance at Shutdown Reactors IP 71707: Operational Safety Verification
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IP 71801: Decommissioning Performance and Status Review at PSRs IP 83726: Control of Radioactive Materials and Contamination, Surveys and Monitoring IP 84750: RadWaste Treatment, and Effluent & Environmental Monitoring IP 90712: Inoffice Review of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92700: Onsite Followup of Written Reports of Nonroutine Events IP 93702: Response to Operational Events ITEMS OPEN, CLOSED, AND DISCUSSED
.QQ.9B l- 98-03-01 VIO Failure to classify unusual event
! 98-03-02 VIO Failure to follow procedures for configuration control l
98-03-03 URI Comparative evaluation of SFP seismicity 98-03-04 URI Evaluote process for commitment tracking 98-03-05 VIO Failure to provide complete and accurate information 98-03-06 VIO Stack flow monitoring system not calibrated as required by TS 98-03-07 URI Dose impact to the public due to use of incorrect main stack flow
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Closed 96-11-07 URI SFPCS design vulnerability-electrical separation 97 03-05 URI Trend E10-1B Performance for Fouling See Attachment i Discussed 97-09-04 URI Survey and Remediation of Offsite Lo:ations
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LIST OF ACRONYMS USED ACR Adverse Condition Report ALARA As low As is Reasonably Achievable AOP Abnormal Operating Procedure CDAC Community Decommissioning Advisory Committee CFR Code of Federal Regulations cfm cubic feet per minute CYAPCo Connecticut Yankee Atomic Power Company DCR Design Change Request DEP Department of Environmental Protection DUFSAR Decommissioning Updated Final Safety Analysis Report DWST Demineralized Water Storage Tank EAL Emergency Action Level EPIP Emergency Plan Implementing Procedure F Fahrenheit gpm gallons per minute HECA High Efficiency Charcoal Air (filter)-
HEPA High Efficiency Particulate Air (filter)
I&C Instrumentation and Calibration IR inspection Report LER Licensee Event Report LLD Lower Limit of Detection MCC Motor Control Center MCP Main Coolant Pump NOP Normal Operating Procedure NOV Notice of Violation NRC Nuclear Regulatory Commission NRR (Office of) Nuclear Reactor Regulation PAB Primary Auxiliary Building PDR Public Document Room QA Quality Assurance QSR Quality Surveillance Report RCS Reactor Coolant System REMP Radiological Environmental Monitoring Program RHR Residual Heat Removal RP&C Radiological Protection and Chemistry RPM Radiation Protection Manager ,
RTT Recycle Test Tank l RWST Refueling Water Storage Tank l
SFB Spent Fuel Building SFP Spent Fuel Pool l SFPCS Spent Fuel Pool Cooling System SUR Surveillance Procedure l SW Service Water i TS Technical Specification WCM Work Control Manual l WDB Waste Disposal Building WTT Waste Test Tank
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ATTACHMENT 1 FOLLOWUP OF ESCALATED ENFORCEMENT ITEMS The NRC issued a Notice of Violation and Proposed imposition of Civil Penalty based on numerous inspections to review several facets of Haddam Neck performance. Some of the 1997 enforcement action issues concerned aspects of plant cperation that were important only for continued reactor operation, or plant operation at power. Following the certification to the NRC on December 5,1996 of the decision to permanently shutdown the plant and cease operations, the corrective actions (plaat modifications) for these violations were no longer appropriate. For each item listed, the violation (VIO) number referenced in the May 12,1997 Notice is listed, along with the original inspection report item number (if applicable) in parentheses. The items listed below are close Violation issues No Longer Applicable to Decommissioning URI 96-08-12, Containment Isolation Valve Design URI 96-08-13, Main Steam and Auxiliary Feedwater Structural Deficiencies VIO 9606-07 (01282), Feedwater Isolation During Main Steamline Break VIO 96-08-10 (01432), Containment Sump Screen Design VIO 96-201 (01022), NPSH Calculation for HPSI VIO 96-201 (01102), Feedwater Valve Leak Sealant VIO 96-201 (01162), Degraded Voltage Protection Setpoints VIO 96-201 (01382), RWST Level Instrument Classification VIO 96-201 (01392), Timeliness of PIR closure VIO 96-201 (01402), Correction of Station Blackout issues VIO 96-201 (01082), LPSI System Flowrate Nonconservative VIO 96-201 (01422), Degraded Grid Voltage Setpoints VIO 96-11-10 (07014), Containment Penetration Leak Rate Testing VIO 96-11-09 (09014), Operability of Boric Acid Flow Path VIO 9611-04 (10014), Nuclear instrument Setpoint Calculation VIO 96-11 (12014), inoperable Boric Acid Flow Path VIO 96-11-01 (13014), Inadequate Refueling Procedures l
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