IR 05000213/1998005
| ML20198R153 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 12/21/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20198R138 | List: |
| References | |
| 50-213-98-05, 50-213-98-5, NUDOCS 9901080110 | |
| Download: ML20198R153 (45) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION 1
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I Docket No.:
50-213 License No.:
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Report No.:
. 50-213/98-05 i
' Licensee:
Connecticut Yankee Atomic Power Company 362 injun Hollow Road East Hampton, CT 06424-3099 Facility:
Haddam Neck Station i
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Location:
Haddam, Connecticut
Dates:
July 20 - November 2,1998 j
inspectors:
Dr. Jason C. Jang, Senior Radiation Specialist j
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' Marie Miller, Senior Health Physicist Daniel Barss, EP Program Reviewer, NRR Thomas Fredrichsi Project Manager, NRR John Wray, Decommissioning Health Physicist Joseph Nick, Decommissioning Health Physicist -
Ronald L. Nimitz, Senior Radiation Specialist
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William J. Raymond, Senior Resident inspector
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Approved by:
Ronald Bellamy, Chief, Decommissioning and Laboratory Branch Division of Nuclear Materials Safety i
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9901000110 981221 I
PDR ADOCK 05000213
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EXECUTIVE SUMMARY
l NRC Inspection Report No. 50-213/98-05 (-
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l This integrated inspection included aspects of licensee operations, engineenng, maintenance, and plant support. The report covers a three-month period of resident
inspection. It also includes the results of announced inspections by regional specialists to i review the unplanned exposure of workers to radiation, and the implementation of the l gaseous effluent control program during the time period in which the Reactor Coolant
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System (RCS) decontamination was in progress and the stack monitoring function was degraded. Further, it includes the results of the inspection of the contaminated cables that were released to the vendor on February 19,1997.
j j Decomm!se!anina Operations and Maintenance:
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Operators' performance was good to monitor the status of operating plant equipment and to monitor the spent fuel pool (SFP). Operators showed good regard for plant procedures I and responded well to off-normal conditions. Maintenance personnel performed well to
, test and maintain equipment, and showed a good regard for spent fuel safety by providing
, reliable and redundant heat removal capability. Licensee actions were slow to address degraded electrical equipment and restore full functionality of a backup diesel generator.
Spent Fuel Safety:
i Licensee performance was very good to implement the nuclear island modifications per commitments and administrative controls. Engineering support was very good, as evidenced in the quality of the calculations and technical evaluations that supported the modifications. Engineering and technical support activities this period were good to
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support decommissioning planning and spent fuel safety.
i NRC review confirmed the licensee had taken the committed actions in the response to the l l May 12,1997 enforcement action and the 10 CFR 50.54(f) letter. The licensee co:,pleted l
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actions to improve performance in the areas of engineering, operational practices,
I emergency preparedness, the licensing end design basis, and the corrective action
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program. The corrective actions were generally effective, but some weaknesses remain
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that warrant continued licensee attention to assure station practices and programs remain effective for decommissioning. NRC reviews continued to confirm the effectiveness of i actions to address performance weaknesses.
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l Plant Support and Radioloalcal Controls:
l l The NRC reviewed the implementation of the gaseous effluent control program for the time periods when the RCS decontamination was in progress, and when the stack monitoring
, function was uncertain. The licensee maintaited good air balance for the primary auxiliary
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building (PAB) and the containment buildings; therefore, all air from these building was
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released to the environment through the air cleaning system (HEPA). Good laboratory i measurement techniques for in-plant and main stack samples were implemented. The L projected doses to the public from a potential unplanned release during the RCS decontamination period were calculated and the results were insignificant.
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An apparent violation of 10 CFR 20 was identified regarding the f ailure to survey which resulted in a release of contaminated equipment to an offsite vendor on February 19,1997.
The licensee's f ailure to perform adequate r, view of existing survey data for the Residual Heat Removal (RHR) room resulted in the intake of radioactive material by two individuals.
Although the estimated occupatianal exposures did not exceed 10 CFR 20.1204 limits, the NRC staff will continue to evaluate this item in future inspections. The licensee conducted adequate follow-up actions including taking of sufficient bioassay samples to assess the I dose to the whole body and body organs. An unresolved item will track the final dose assessments following additional sample analyses.
The licensee conducted security activities in a manner that protected public health and safety. The response to off-normal conditions was good to assure security functions remained effective. Compensatory measures were taken per procedures and instructions.
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One discrepancy was noted in the f ailure to assure emergency power for security equipment remained functional. An unresolved item will track the review to determine whether the availability of emergency power met the requirements of the security plan.
This insrection evaluated the licensee's performance during its first graded defueled emergency preparedness exercise. The inspectors observed facility staffing at the two emergency f acilities, control room (CR) and technical support center (TSC), command and control, and turnover from the CR to the TSC. The inspectors observed demonstration of procedure implementation including: event classification, notifications, on-site protective actions, and search and rescue. Communications, rumor control and drillmanship were also evaluated. The licensee conducted an adequate exercise in accordance with the defueled emergency plan and defueled emergency plan implementing procedures. All NRC and licensee exercise objectives were met.
The CR and TSC were staffed and activated in a prompt manner. The Emergency Director and the response staff demonstrated the following: adequate assessment, classification, notification, command and control, and onsite protective actions. The search and rescue effort was delayed because of cumbersome procedures. The licensee's self-critique was a noted strength. The licensee identified numerous issues and categorized those issues according to significance. The NRC inspectors agreed with the licensee's findings and re-emphasized a few of the licensee's self-identified improvements items.
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REPORT DETAILS
Summarv of Plant Status The Haddam Neck plant conditions remained stable with the spent fuel safely stored in the SFP. The licensee completed the RCS decontamination and finished installation and testing of design modifications for phase 1 of the nuclear island.
On July 27,1998, NRC Commissioner Nils Diaz and Nuclear Reactor Regulation (NRR)
Project Manager Thomas Fredrichs toured the site and met with licensee management to discuss the status of decommissioning activities.
On July 28,1998, Mr. W. Raymond and Ms. M. Miller attended a public meeting hosted by Mayor Bette Giesing of Groton to discuss public questions regarding the activities to complete radiation surveys and remove licensed materials from offsite properties. The
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inspectors discussed the NRC independent reviews and radiological measurements at the offsa locations, along with the basis for the NRC conclusions that there was no threat to pubuc health and safety based on the findings to date. The inspectors attended meetings of the Community Decommissioning Advisory Committee on July 29, September 15, September 29, October 13, and Oct 27,1998.
On September 4,1998, the licensae announced the selection of Mr. K. Heider to the position of Decommissioning Director, reporting to the Vice President of Operations and Deccmmissioning. The creation of the new position was part of a reorganization of Haddam Neck staff that would better integrate operations, maintenance and engineering organizations with the decommissioning staff.
The "B" loop intermediate and spray cooling systems were tied in to the SFP heat exchangers on September 27,1998. The inspectors performed a special inspection of an internal exposure that occurred when two her th physics technicians entered the RHR pit on September 30,1998.
1. Decommissionina Operations and Maintenance
Conduct of Operations 01.1 Operatina Activities and Status of Ooeratina Systems Insoection Scooe (71707. 62801)a.
Using Inspection Procedure 71707 and 62801,the inspector reviewed plant status and licensee activities to maintain the plant in the defueled condition.
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Observations and Findinas The operators maintained stable plant conditions, and adequate level and cooling for the SFP. The licensee operated plant equipment necessary to support the SFP, and assured the operability of support systems. The inspector verified compliance with Technical Specifications (TS) TS 3.9.11, SFP water level; TS 3.9.12, fuel storage air cleanup systems; and, TS 3.9.16, pool temperature below 150*F. The spent fuel pool cooling system (SFPCS) remained operating per Normal Operating Procedure (NOP) 2.10-1, with the "B" SFP heat exchanger in service. The
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operating procedure was revised to reflect the status of the new cooling systems for the SFP. On September 24,1998, following the revision of procedures and the completion of training, the licensee implemented License Amendment 193, which formally made the Defueled TSs effective (NL/CY-98-112).
The secondary side of the SFPCS was initially cooled using the service water (SW)system. The licensee completed installation and initial test activities on September 27,1998, to place the "B" loop of intermediate and spray cooling
-systems into service. The systems remained in a test and evaluation mode of operation under Surveillance Test (ST) 11.7-212B until October 8, when the "B" loop was declared operable. The licensee conducted routine surveillance of the SFP and building. Operator actions observed during periodic plant tours were consistent with the procedures.
Operators' activities to monitor the spent fuel from July to November were good to operate plant systems as necessary to maintain spent fuel storage, and to assist
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maintenance, testing and decommissioning activities. Operators performed well to follow procedure NOP 2.10-1 during the transition to the new cooling system. Plant personnel responded well to off-normal conditions, as described below.
Storm Preparations The plant staff prepared the site and tracked the progress of Hurricane Bonnie during the period of August 25 - 29,1998. The operators followed the guidance of abnormal operating procedure (AOP) 3.2-5, " Natural Disasters" on August 25 as the storm came within 800 miles of the site. The licensee inspected the site and secured loose material which could impact personnel, structures or power supplies.
Site inventories were replenished and emergency diesels were confirmed to be in a standby / ready condition. The inspector toured the site periodically during August 25 - 28, and verified licensee preparations. Hurricane Bonnie and subsequent storms tracked south of New England and had no impact on the site. Licensee preparations for adverse weather were generally thorough.
Leakaae From Dike Plant staff discovered on September 3,1998, the loss of water from the dike around the aerated drain holdup tank ed the recycle test tanks (Adverse Condition Report (ACR)98-851). The licensee normally processed rain water collected in dike areas as radwaste. That practice was stopped on August 26,1998 when radwaste tank inventory margins became low. Licensee follow-up actions were good to develop a temporary modification to allow the alternate processing of rain water from the dike areas to the aerated drain tank and to the waste test tank for processing and release.
01.2 Maintenance and Surveillance Activities a.
Insoection Scoce (62801. 61726 and 62707)
Using inspection Procedure 62801,61726 and 62707, the inspector conducted 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 plant equipment. The licensee maintained plant equipment necessary to support the SFP. The licensee
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^ implemented compensatory measures and the action statements in the TS and the technical requirements manual during periods of degraded equipment performance (e.g., for work on Radiation Monitor (RM) 148, the seismic monitor and the stack flow instrument). Additionally, the inspector reviewed on a sampling basis the planning, work controls and radiological controls for the following activities:
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pressurizer surge line cutting and capping
- tie-in of the "B" loop cooling to the SFP heat exchanger
"A" spent fuel heat exchanger cleaning and repair
- uninterruptible power supply UPS-2 troubleshooting and repair
- preparations to install the core barrel into the reactor vessel e
" A" Soent Fuel Heat Exchanaer (AWO 98-3177)
After initiating spent fuel cooling using the "B" loop intermediate cooling (IC)system, the licensee removed the " A" heat exchanger from service for cleaning prior to tie-in of the "A" IC loop. The "A" heat exchanger tubes were also cleaned
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i and leak tested, and the bonnet (end bell) was inspected. The licensee identified corrosion induced wear on the bonnet divider plate, which had the potential to create a bypass flow path. The "A" heat exchanger performance to cool the pool was acceptable during the installation of the "B" IC system. The "A" heat exchanger bonnet was dispositioned under Non-Conformance Report (NCR)98-092, Replacement item Evaluation (RIE) 96-5037 and Automated Work Order (AWO)98-3177. A new bonnet was available onsite and was installed. The new bonnet had superior mechanical and corrosion resistance characteristics compared with the original item. The licensee coated the internal carbon steel surfaces of the new bonnet to further minimize corrosion. The licensee showed a conservative approach to spent fuel safety through a good regard for maintaining redundancy in SFP heat removal capability and by expeditiously restoring the " A" heat exchanger to service in the best possible condition.
UPS-2 Failure (AWO 98-3249)
The licensee performed troubleshooting and repairs per AWO 98-3249 after UPS-2
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failed to carry supplied loads. The Uninterruptible Power Supply (UPS) failed when a voltage conditioning transformer on the input supply from a motor control center (MCC) failed in-service on October 25,1998. Temporary power was restored by bypassing the failed transformer. The in-service MCC was the attemate supply for the UPS-2 loads, which had been on-line since June 22,1998 (AWO 95-15493)when a failure of the UPS-2 batteries caused the loss of the rectifier / battery charger feed from the preferred MCC and a backup diesel generator. With the ', referred MCC feed inoperable, the backup diesel generator was unavailable to sply loads downstream of UPS-2.
Despite actions to replace the batteries and fuse FU-2 (AWO 98-2225).
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rectifier / charger / inverter portion of UPS-2 remained inoperable due to the tripping of
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_7 the battery supply breaker. The cause for the tripping remained unknown and the repair of UPS-2 was hampered by the lack of complete information at the site regarding the UPS-2 design. Licensee actions continued at the end of the inspection period to obtain vendor assistance to investigate and repair UPS-2.
While maintenance support was good in response to the loss of power on October 25, the licensee did not fully appreciate the vulnerability in the UPS-2 that had existed since June 22,1998. Also, the vulnerability was not fully known by all affected plant groups. Licensee actions were not timely to resolve degraded equipment conditions prior to the f ailure on October 25,1998, and to restore the functionality of a backup diesel generator.
01.3 Conclusions for Decommissionina Operations and Maintenance r
Operators' performance was good to monitor the status of operating plant equipment, and to monitor the SFP. Operators showed good regard for plant procedures and responded well to off-normal conditions. Maintenance personnel performed well to test and maintain equipment, and showed a good regard for spent fuel safety by providing reliable and redundant heat removal capability. Licensee actions were slow to address degraded conditions and restore full functionality of a backup diesel power supply.
II. Spent Fuel Safety E1 Engineering Support for Decommissioning E1.1 Soent Fuel Coolina System Modifications a.
Inspection Sccce (37801,60851,60P54,37700)
The insmetor reviewed modification activities that supported spent fuel safety.
Plant design change activities to establish the nuclear island were reviewed to verify that the modifications were implemented in accordance with licensee controls, b.
Observations and Findinas The modifications associated with design change record (DCR) CY-97010,015, 016,017,018 and 032 were reviewed, along with supporting engineering calculations and analyses. The list of references and materials reviewed is provided in Attachment 1. The DCRs detailed the installation of new equipment for the spent pool fuel island (SFPI) that eliminated reliance on the existing plant cooling water systems. Inspection Report Nos. 98-01 and 98-03 provide documentation of further NRC review in this area. The licensee classified the SFP and cooling systems under DCR-CY97015, and applied graded quality assurance (QA) to the installation per Administrative Control Procedure (ACP) 1.2-2.86 (the Spent Fue!
i Pool Quality Assurance Program). The modification and testing included:
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The use of the existing SFP heat exchangers (E-10-1 A and E-10-1B25) and foel peol cooling pumps (P-21-1 A and 21-1B) with the new IC and spray cooling (SC)systems. The SC system used fan spray coolers to disperse heat to the outside air.
The new IC system piping had inlet and outlet connections to allow for the j
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l emargency supply of cooling water from tha fire system in the event of the loss of the normal cooling system. Normal makeup of coolant is provided by the existing 100,000 gallon primary water storage tank (PWST) and the primary water transfer pumps. The primary water distribution header was modified to supply automatic make-up capability to the SC and IC surge tanks located in the truck bay of the
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spent fuel building.
l During this inspection, the licensee completed the modifications to install the IC and
.SC systems; performed component level testing; t ei d the "B" loop IC and SC systems into the SFPCS (functional on September 27,1998); and, completed l
thermal performance testing. The tie-in of the "A" loop systems continued at the
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end of the inspection period. The new cooling system operated smoothly, and l
performed well during the periods of high ambient temperature and humidity to cool the pool and maintain temperatures well below 100 F. The licensee noted that the spent fuel temperature would track ambient temperature and humidity conditions, as monitored by the temperature difference { delta-T) between the heat exchanger outlet (return of cooled water to the SFP) and the ambient wet bulb temperature.
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When the delta-T was greater than 26o, the pool would cool down; when the delta-T was less than 26*, the pool would heat up (reference ST 11.7-2128, Attachment 12.5). The maximum temperature for the design conditions,104*F at a wet bulb temperature of 76 F, was not reached.
New instrumentation was provided in the CR and locally in the spent fuel building, including: SFP temperature and levelindicators mounted in control panel F1;IC and SC surge tank low and high level alarms located on CR panels FF and EE: locally mounted control and indicators for makeup control valves, relative humidity and dry bulb temperatures, and wet bulb temperature; and newly installed radiation monitors RMS-19A and 19 B, used to monitor the IC loop for radioactivity. The existing CR alarms for SFP high and low level and high temperature remained operable on PanetE.
The IC/SC systems used two 100% capacity, water-to-air fan coolers. The IC/SC cooling loops were designed based on the decay heat load requirements for July 1, 1998, an assumed ambient wet bulb temperature of 76*, and the ability to cool the pool to below 110 F for at least 95% of operating hours per the American National Standards Institute (ANSI) 57.7. The f an coolers were located outdoors on top of the truck bay roof of the spent fuel building.
The SFPI electrical power was provided by a 480-volt power feed from the existing onsite power supplies, with a new feed provided from Bus 11 via a splice connection in the cable vault. The backup power supply remained the station emergency diesel generators, EG-2 A and EG-28. In the future, the licensee planned to use a mobile air-cooled diesel for the nuclear island instead of the station diesels.
The design control program was appropriately implemented and onsite fabrication and testing activities were conducted in accordance with the associated work package 3. The DCRs were developed and implemented per the administrative controls in the Design Change Manual. Inspector observations of as-built l
installations and activities in progress, based on a sampling review of work t
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packages, showed the modifications were well controlled and completed per the DCR packages. Plant operational procedures were changed to reflect the new j
systems, components and valve lineups.
Component and performance testing of the new systems was completed per DCR 97010 and ST 11.7-211,11.7-212A, and 11.7-2128. Test activities were controlled using the work order process. The testing included flushing of the IC and SC systems, and performance testing on spent fuel cooling after the final tie-in; functional testing of IC and SC surge tank high and low level alarms and interlocks; and, functional testing of new operator interface alarms and indicators that provide
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for monitoring of the SFP conditions in the CR.
The supporting calculations and technical evaluations reviewed for DCR-CY-97010 are listed in Attachment 1. The safety evaluations completed per 10 CFR 50.59 and supporting the DCR appeared to be completed per the administrative controls and were consistent with the Updated Final Safety Analysis Report (UFSAR) and design work to declassify plant systems. The inspector completed a sampling review of structural and thermal calculations that supported the development of the nuclear island. The calculations used acceptable methodologies, and conclusions were supported by the results. The analyses (Calculations 97-SFPI-1610 and 1611)and testing (SFP Heat up Rate Test per Procedure ST 11.7-2128) determined the pool thermal performance and provided the basis for the sizing and operational thermal requirements of the fan coolers.
NRC review focused on the mechanical and electrical portions of the SFPI cooling systems. Licensee planning for additional modifications continued at the end of the inspection, including the changes to the SFP purification system, installation of alternate normal and emergency power supplies; relocation of make-up equipment inside the spent fuel building; and installation of a new heating system.
c. Conclusions
Licensee performance was very good to implement the nuclear island modifications per commitments and administrative controls. Engineering support was very good, as evidenced in the quality of the calculations and technical evaluations that supported the modifications. Engineering and technical support activities this period were good to support decommissioning planning, and spent fuel safety.
E8 Miscellaneous Engineering issues E8.1 Review of Licensina Event Reports (LERs) and Telechonic Notifications
a. Inspection Scope
(92700,90712)
The purpose of this inspection was to review prompt reports and LERs to verify the requirements of 10 CFR 50.72 and 50.73 were met.
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Observations and Findinas LER 98-008, Stack Radiation Monitor RMS-14B Samples Not Analyzed Properly.
This LER described the discovery on September 1,1998 that some of the particulate and iodine samples taken from July 21 through August 28,1998 were not analyzed as required by the TSs and the Offsite Dose Calculation Manual. The causes included operation of the sample skid with a mispositioned vent valve, and the discovery that the sample system design allowed the formation of condensation in the sample lines under certain environmental conditions, with the consequent deposition of particulates and iodines at a higher fraction than previously evaluated.
The event was reported per 10 CFR 50.73(a)(2)(1) as a condition prohibited by the TSs. Licensee actions to correct the deficiency and NRC findings are described in Section R1.2 and R8.1. This LER is closed.
LER 98-009, Excessive Check Vaive Seat Leakage. This LER described the discovery on September 15,1998 that SW check valve SW-CV-963 had excessive seat leakage. The valve was declared inoperable, since it could not podorm the design function to eliminate the potential for water hammer event following a loss of norma! power and the subsequent restart of the SW pumps. The inoperable condition was discovered during a surveillence test that was being conducted at an increased frequency (monthly versus quarterly) after tasting in August identified a similar problem (LER 98-07). Licensee actions were appropriate to exercise and test the valve on September 16, which freed the disc from the stuck open position, and to increase the surveillance test frequency to weekly to assure the valve remained functional. The long term corrective actions were to complete modifications to install the new intermediate and spray cooling systems, and thereby eliminate reliance on the service water system to cool the SFP. The licensee made the "B" loop of the new cooling systems operable on October 8,1998. This LER is closed.
LER 97-21-01, Contaminated Material Found Offsite. The licensee provided a supplemental report to describe the discovery at an offsite location of a welding machine that showed radiological contamination containing transuranic radionuclides. The contamination was removed, and the welder was returned to the site. Licensee action to follow-up this finding were good to conduct additional surveys at the offsite property (no additional transuranic isotopes were found),complete detailed assessments of the internals of the welder (no additional transuranic isotopes were found), and to initiate plans to locate and conduct follow-
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up surveys of other welders from Haddam Neck and Millstone. The licensee reported that, as of October 8,1998, a total of 15 welders had been identified for l
further investigation. Inspection item 97-10-01 will track NRC follow up of licensee actions to identify and remediate radioactive materials at offsite locations.
l Inspection item 97-09-04 will track the potential exposure to members of the public l
from offsite radioactive materials. This LER is closed.
LER 96-29, PAB Internal Flood Protection. The licensee issued Revision 1 to this LER on July 7,1997 to reflect the decommissioned status of the plant. A licensee engineering evaluation concluded the plant could be maintained in safe shutdown following a postulated internal flooding event. Further corrective actions were no longer required. This LER is closed.
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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 10/1/98 (EN 34861), and 9/15/98 (EN 34783). No inadequacies were identified in meeting reporting requirements. The licensee reported to the NRC per 50.72(b)(2)(vi)on October 1, 1998, that about 6000 gallons of rain water collected in the dike around the recycle test tanks and the aerated drain tank may have seeped into the ground since August 26. The rain water became partially contaminated due to contact with contaminated surfaces, and contained trace levels of radioactivity, including 6.09E-9 Ci/ml of Cs-137, and 4.19E-5 Ci/ml of tritium. The concentrations were less that the TS 3.11 limits of 1E-6 pCi/ml for Cs-137 and 1E-3 Ci/ml for tritium, respectively. Potential dose to any offsite area will be reported in the annual effluent report to the NRC. Licensee follow-up actions were appropriate to reactivate the sump pump in the dike area, evaluate the status of the dike in the borated waste storage tank area, and to include the ground water leakage in the process to evaluate the site for remediation.
E8.2 Enforcement Follow-up - Manaaement, Desian and Corrective Actions
a. Inspection Scope
(92702)
The purpose of this inspection was to complete the review of the licensee corrective actions for the enforcement action issued by the NRC letter dated May 12,1997. This review also covered the licensee actions committed to in response to the NRC Request for Information under 10 CFR 50.54(f) dated October 9,1996, which addressed issues similar to those in the enforcement action, including the licensing basis and design basis (LB/DB) for the safe storage of spent fuel, the identification of commitments, configuration management of major processes, and updating the Final Safety Analysis Report (FSAR). Past NRC reviews in this area were provided in Inspection Report Nos. 97-03,97-05,98-01 and 98-03.
b.
Observations and Findinas The licensee took actions to address deficiencies in broad categories that included engineering and operating practices and the corrective action process. The licensee addressed deficiencies in engineering which included fai!ures to meet commitments, and the failure to establish and adequately maintain the LB/DB. The licensee provided a final response to the NRC Request for Information per 10 CFR 50.54(f)by letter dated September 30,1998 (CY-98-153).
Enaineerina Proarams and Practices in the licensee's response dated June 11,1997, the iicensee described the actions to address weaknesses in the processes for safety evaluations, UFSAR updates, safety classification, instrument set points and procurement engineering. The corrective actions addressed the common causes for the process deficiencies by upgrading the LB/DB information, the design control process, the configuration management pro:ess, resource management, training, and by setting new management staridards and expectations. The following areas were reviewed:
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I R*aulatorv Commitments l
A commitments database was used to verify plant design, procedures, programs and processes were as reflected in the license. The licensee identified over 13,950 licensing commitments, with 2080 commitments applicable to decommissioning, which were reviewed to determine which remained active commitments, or commitments that required further action. The review of outstanding commitments was tracked as a key performance indicator until the backlog was reduced. As of September 1998,130 items that required action to implement the commitment, such as procedure steps were being tracked by the licensee. The licensee demonstrated completion of this action through a review of the database, and the list of assignments in the action tracking system. NRC commitments due in 30 days are also tracked in the weekly performance indicators for the site. Procedure LDI 2.01, " Regulatory Commitments", will be used to identify, manage and track future regulatory commitments.
Inspection item 98-03-06 concerned the failure to meet TS commitments for the calibration of stack flow instruments. One of the corrective actions in response to this item was to verify that License Amendment No.125 was correctly translated into the surveillance procedures. The Quality Oversight organization completed a review to assure TS surveillance items were correctly covered in plant procedures.
Discrepancies were identified and addressed through ACRs.
Finally, inspection item 98-03-04 concerned the adequacy of review under LDI 2.02 to identify commitments, and specifically whether items received an independent review. The licensee completed a self-assessment in this area as described in NL/CY-98-086. The assessment found numerous discrepancies in the database in which the verification appeared lacking; however, the source documents showed the verification reviews were completed but the data entries for the reviewer names were not entered into the database. The assessment addressed the NRC concern in this area. The assessment was thorough and identified recommendations to improve the commitment tracking database. The licensee's action to address issues identified in the NOV response is complete.
Licensino Basis for Safe Storaae of Soent Fuel The UFSAR was revised to correct deficiencies applicable to Structures, Systerns and Components (SSCs) and programs required to support the defueled condition.
The revised UFSAR, Change No. 30, was submitted to the NRC by letter dated January 26,1998 (NL/CY-98-011). The inspector reviewed the UFSAR during routine inspections to verify the upgraded document accurately reflected plant conditions and the licensing commitment. The inspector reviewed, on a sampling basis, the following sections of the UFSAR for general accuracy, as well as for the adequacy of those systems that were important to the decommissioning status of the plant: Section 1.0, Plant Systems and Categorizations; Section 2.4, Hydrology and Ultimate Heat Sink; Section 2.5, Seismology; Section 3.2 Classification of SSCs; Section 3.1, Conformance with General Design Criteria (GDC); Section 3.4, Flood Protection; Section 3.10, Seismic Qualification Section 6.2, Containment Systems; Section 6.3, Emergency Core Cooling Systems: Section 6.4, Habitability Systems Section 7.0, instrumentation and Controls: Section 8.3, On-site Power; L
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Section 9.1, Fuel Storage and Handling; Section 9.2.1, Service Water System; Section 9.2, Water Systems Section 11.0, Radioactive Waste Management; Section 13.0, Conduct of Operations; and, Section 15.0, Accident Analyses. Within the scope of,this sampling review, the updated UFSAR was accurate in regard to
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significant plant design criteria, specifications and physical features.
The defueled condition LB/DB for spent fuel storage was issued in a new Design
- Basis Document, which was maintained as indicated in Revision 3 issued on October 8,1998. The licensee completed an independent review of the LB/DB, as j
described in the report, " Independent Third Party Review of the Defueled Licensing and Design Basis at Connecticut Yankee Atomic Power Company." The
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independent review was completed by two outside contractors (XWest Group, Inc.,
and Duke Engineering and Services) to assure that the LB/DB document was supported by the reference documents and acceptable for use. The open issues from both independent reviews were resolved, and verified complete by the auditing
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organizations (reference: DE&S memorandum CYR-98-0014 dated April 9,1998; and XWest memoranda 97-105 through 97-123 issued in December,1997). The L
inspector noted the licensee's method to track and resolve the open issues was thorough, and verified by a sampling review that the discrepancies were addressed.
The licensee's action to address issues identified in the Notice of Violation (NOV)
and 10 CFR 50.54(f) responses is complete.
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The licensee continued to review and categorize plant systems using the
- l methodology described in Section 1.1 of the UFSAR. Plant systems were abandoned when no longer needed to support the safe storage of spent fuel, or were no longer needed to support the RCS decontamination. The licensee achieved
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a significant milestone during this inspection period through the installation of new cooling systems for the SFP, which eliminated the reliance on the SW system and the Connecticut River as the ultimate heat' sink for decay heat removal. Further, the licensing requirements for several plant systems were eliminated by issuance of the Defueled TSs in License Amendment 193 on June 30,1998. Several NRC issues
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involving plant program or equipment problems were evaluated as no longer significant to the decommissioning status of the plant. A summary of resolved issues is provided in Attachment 2 of this report. While some discrepancies indicate some weaknesses in the actions to improve station practices and programs l
(e.g., the' DCR for the seismic monitor, and the initial DCR to declassify systems for the nuclear island), these items appeared as exceptions to the generally good effort
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to revise and maintain the LB/DB. The licensee's action to address issues identified
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in the NOV and 10 CFR 50.54(f) responses is complete.
Operations Procedures and Practices l
t The corrective actions to address the causes for violations involving inadequate procedures were previously inspected. Licensee actions to address procedure deficiencies for the core off-load and the nitrogen intrusion event were reviewed in NRC Inspection Report No. 96-11. The licensee took long term corrective actions to revise and improve operating procedures. The last major " operational" event prior t
to transitioning into the decommissioning and plant dismantlement mode was the l
conduct of the RCS decontamination during the Summer of 1998. NRC concerns l
regarding the adequacy of procedures for the RCS decontamination were addressed
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in Inspection Report No. 98 04. The licensee complsted the actions necessary to meet the commitments in response to the May 12,1997 enforcement item.
Additional NRC reviews of this areas are described in Attachment 2.
Corrective Action Proaram The actions to improve the corrective actions program (CAP) were described in Inspection Report Nos. 97-03 and 97-05. The licensee also completed an independent assessment of the CAP, and benchmarked the program against other nuclear power plants with a good CAP. The licensee implemented changes and completed actions to address the recommendations from the self-assessments, as described in memorandum CYCA 98-002 dated January 19,1998. The inspector reviewed the tracking and status of the follow-up actions in a meeting with the Corrective Actions Program Manager. The priority 1 and priority 2 items were completed in the third quarter of 1998. Based on the above, the licensee completed
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the actions necessary to meet the commitments in response to the May 12,1997 enforcement item.
Past NRC inspections verified the licensee actions to improve the CAP, and the generally good performance to implement the revised process. Good performance was noted in the following areas: definition, administration and implementation of the ACR process; low threshold for problem identification; generally thorough and appropriate reviews to establish apparent and root causes for identified discrepancies; use of appropriate indicators to track performance; generally j
thorough and appropriate corrective actions that address the causes for discrepancies; and, a strong performance by the management review team to assess individual issues and to identify general station performance trends to assure a viable CAP.
Notwithstanding the above findings indicating general improvement had been achieved, there were examples where licensee findings and plant events suggested the need for continued attention to assure effective corrective actions. Examples include the NRC findings describe in Inspection Report No. 98-04. Licensee actions continued at the end of this inspection period to review the CAP as implemented in the ACR process. The licensee completed the actions necessary to meet the commitments in response to the May 12,1997 enforcement item.
c. Conclusions
in summary, the NRC review confirmed that the licensee had taken the actions committed to in the response to the May 12,1997 enforcement action and the 10 CFR 50.54(f) letter. The licensee completed actions to improve performance in the areas of engineering, operational practices, emergency preparedness, the LB/DB, and the CAP. NRC review determined that the corrective actions were generally effective, but some weaknesses remained that warrant continued licensee attention to assure station practices and programs are effective throughout the j
decommissioning process. NRC reviews continued to confirm the effectiveness of actions to address performance weaknesses.
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E8.3 Status of Previous inspection items (Closed) Unresolved item 96-06-03: Corrective Actions for LERs 96-08 and 96-09.
This item was open pending the review of licensee actions for issues related to AOPs (96-09) and evaluation of flooding through an open floor block in the PAB (LER 96-08). The licensee reviewed the station AOPs and identified no other similar deficiencies. This result was reported in LER 96-09, Revision 1, dated June 10, 1998. In regard to PA8 flooding, a licensee engineering evaluation concluded the plant could be maintained in safe shutdown following a postulated internal flooding event. Further corrective actions to develop a barrier control program were no longer required following the decision to permanently shutdown the plant and the removal of all fuel from the reactor, as described in LER 98-08, Revision 1 dated
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January 28,1997. This item is closed.
(Closed) Unresolved item 96-10-01: Special Nuclear Material (SNM) Inventory l
Requirements. This item concerned the requirements for the conduct of inventories l
of SNMs per 10 CFR 50.71. The NRC determined that recording fuel assembly
serial numbers is not specifically required in order to meet the inventory i
requirements of Section 70.51(a)(8). Recording serial numbers remains a good practice as defined by industry standards such as ANSI Standard N15.8-1974, Nuclear Material Control Systems for Nuclear Power Plants, Section 6.4.2. The licensee is not committed to the requirements of ANSI N15.8; however, past inventories of SNM in the SFP per SNM 1.4-11 have included verifying serial numbers depending on whether fuel shuffle activities have been conducted.
Further, the inventory conducted in 1997 under AWO 96-9577 included the verification of serial numbers after cleaning the top nozzles of the fuel assemblies.
This item is closed.
(Closed) Violation 98-01-04: Seismic Monitoring instrumentation. This item concerned the failure to r 4se the TSs prior to modifying the seismic monitoring instrumentation. The lice,ee responded by letter CY-98-088 dated June 12, 1998. Modifications were completed under DCR 97-019 to place the new seismic instrumentation channel in service. The licensee proposed to revise the TSs to reflect the new system by letter to the NRC dated June 2,1998 (CY-98-084), and to relocate the requirements for the system from the TSs to the technical requirements manual. The system remained functional (except during periods of test or repair) but inoperable per TS 3.3.3.3 pending the issuance of a license amendment by the NRC. Licensee actions to address the causes of the violation included: revisions to the design change process: the conduct of a lessons-learned review by the oversight committees (Plant Operations Review Committee and Nuclear Safety Assessment Board): and, revising the procedure to conduct safety evaluations to better control changes to the TSs. This item is closed.
(Closed) Unresolved item 96-06-06: Battery Oscillations and DC Ground. Further licensee reviews and follow-up actions were described in memorandum CT-TS-97-053 dated February 12,1997. The current oscillations on the "A" battery charger decreased and were no longer observable after reducing loads on the battery following the plant shutdown. No subsequent oscillations were observed during inspector reviews of plant status. The batteries were no longer required to be operable as a safety system, and the TSs 3/4.8 operability and
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surveillance rsquiremants were dslated in Licensed Amendment 193. This item is closed.
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(Closed) Unresolved item 98-01-03: Quality Attributes for the Nuclear Island. The licensee completed the engineering evaluations to disposition the quality issues addressed in NCR 98-36 for the new spent fuel cooling systems. DCR CY-97023 classified the new systems as requiring augmented QA. NCR 98-036 addresst d the potential discrepancies for the initial actions to procure and install the systems as non-QA. The resolution included additional inspections to document the quality attributes of the mechanical, instrumentation and electrical modifications, the pressure testing and functional testing, and the development of drawings, manuals and project documentation. The licensee identified and corrected weld quality issues in the piping for the new pump skids. Items identified as adverse to quality
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were addressed using separate NCRs. The inspector conducted independent walkdowns of the new system components, piping and instrumentation. This review identified no discrepancies not already addressed by the licensee. This item i
is closed.
(Closed) Violation 96-201-08: Failure to Follow Procedure for Vendor Contacts.
The licensee used the VETIP process to obtain vendor information on component reliability problems. The licensee addressed this violation by reinstituting the vendor program per procedure PEG 6.05, " Vendor Interface for Key Safety Related Components." Implementation of this program was demonstrated in memoranda NMDM-98-OO7,98-005,97-005, and CY-TS-97-822 for 1997; and, PEG-MP-96-0350,399,456,457, and 504 for 1996. These actions satisfied the response to the violation.
Program changes included the implementation of the process at Haddam Neck, and revision of the Key Safety Related Equipment List to reflect the permanently shutdown status'of the plant (CY-TS-97-822 and 98-098). The licensee also planned to administer the vendor program in a revised Industry Experience Program
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for Haddam Neck, as described in procedure ACP 1.2-2.75," Regulatory Affairs Quality Procedure Industry Practices", Revision 1, which was in draft review on October 30,1998. This item is closed.
f (Closed) Violation 96-201-35: Failure to Disposition Non-conformances. The licensee revised ACP 1.2-15.2, Section 1.5.2 on 10/23/97 to address the
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scheduling and tracking issues identified by the NRC. The revised procedure guidance addressed the use of nonconforming items in plant systems. This item is closed.
(Closed) Unresolved item 98-03-04: Process for Commitment Tracking. Licensee actions to address this matter are described in Section E8.2. This item is closed.
(Closed) IJnresolved item 97-01-07: SW System Modifications. The licensee modified the service water system by installing check valve SW-CV-963 to eliminate the potential for water hammer during loss of power events. Past NRC inspections reviewed licensee actions to test the new check valve for continued operability.
The licensee completed modifications to the spent fuel cooling system during this
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inspection period to eliminate the reliance on the SW system for decay heat removal. This item is closed.
' Closed) Unresolved item 97-01-08: SW System Corrosion. The SW system operated through September 1998 without further indication of corrosion induced leakage. The licensee completed modifications to the SFPCS during this inspection period to eliminate the reliance on the SW system for decay heat removal. This item is closed.
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(Closed) Unresolved item 97-03-04: SFP Heat Exchanger Performance. Fast NRC inspections reviewed licensee actions to monitor the thermal performance in the "B" SFP heat exchanger, and to backwash the SW side of the heat exchanger to reduce fouling. The new SFPCS reduced the potential for future fouling by climinating the
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reliance on SW, and through the use of demineralized water in the IC system.
Further, the licensee addressed the operability of the heat exchangers in License Amendment No.193 issued on June 30,1998. This item is closed.
(Closed) Unresolved item 97-09-03: QA Classifications for SFPCS. NRC reviews in Inspection Report No. 98-01, and Section E1.1 of this report, describe the l
licensee's actions to apply graded QA to the SFPCS and the installation of the nuclear island. - This item is closed.
- LClosed) LEH 97-01-00: River Temperature Below UFSAR Analyzed Condition. The
-licensee revised Section 9.2.1.2 in the Defueled UFSAR to describe the temperature extremes at the SW inlet, and to specifically recognize that the minimum temperature was 28*F. The new SFPCSs eliminated the reliance on SW for decay heat removal. This item is closed.
Ill. Plant Support and Radioloalcal Controls
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R1 Radiological Surveys and Contamination Control R1.1 Contamination incident After the RCS Decontamination
a. Inspection Scope
(83750)
The purpose of this inspection was to evaluate the licensee's dose assessment of two individuals who received internal uptakes of radioactive material and to determine if either individual was exposed to radioactivity in excess of 10 CFR 20 limits.
b.
Observations and Findinas On September 30,1998, two health physics technicians were assigned to survey the RHR pit in preparation for entries later that week by maintenance personnel for plant repairs. Their supervisor reviewed the work scope with them and told them to perform the job on Radiation Work Permits (RWP) No. 0014. No previous RHR pit surveys were reviewed prior to their entry. In accordance with the RWP, the
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Upon exit of the Radiological Controlled Area (RCA) after completion of their task, they successfully passed through the plant's automated whole body friskers. These monitors did not alarm. At the end of the day, they were chosen to receive a whole i
body count as part of the licensee's random whole body count program. One individual was detected with internal contamination at a level below the licensee's immediate action level (less than %10 All Co-60). The other technician's whole body count indicated no internal uptake. This action level is based a beta / gamma to alpha ratio previously exhibited in the plant. The licensee analyzed the worker's j
l lapel air samples and determined that the beta / gamma to alpha ratio for
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contamination in this room was much less than expected (10 to 15:1 rather than historical 80:1). The licensee immediately realized that this change in ratio meant that the uptake of transuranic isotopes based on comparison to the measured Co-60
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uptake was greater than originally determined and required bioassay samples to j
accurately assess dose to the whole body and bone.
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Urine samples were obtained from both workers. Technician A did not show any detectable radioactive uptake. Technician B was asked to provide fecal samples
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and complied within the first 24-hours. A conservative dose assessment was
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performed which assumed that the entire uptake was inhaled. The inspector l-reviewed this assessment with the licensee and verified that the exposure did not exceed 5 rem whole body nor 50 rem bone dose which are the regulatory limits.
The exact exposures will be determined following further sample analyses.
However, based on the sarnple results to date, it appears from the elimination of the l
radionuclides from the body that the uptake was primarily ingestion. This would l
lower the dose assessments to the whole body and the bone. The final dose
assessments to the two exposed health physics technicians will be reviewed during a future inspection. (URI 98-05-01).
The inspector reviewed qualifications of the health physics supervisor and the two
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technicians and determined that they satisfied TS qualification requirements for their respective positions. The technicians were Senior Health Physics technicians.
i Previous surveys of the RHR room were reviewed and it was noted that the last entry into this area was on July 28,1998. This was a day after RCS
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decontamination liquid spilled into the room. Levels of contamination where the technician performed work were as high as 20 mrad smearable beta / gamma and total alpha. Results of the survey performed on September 24,600 dpm/100 cm 30,1998, indicated contamination levels remained high; a maximum of 204,000 dpm/100 cm beta / gamma and 8070 dpm/100 cm2 total alpha. Plant procedure RPM 2.1-10 Revision 1, RWPs, states in part that an area with total alpha contamination levels greater than 1000 dpm/100 cm2 is a high risk alpha area requiring a Specific-2 RWP, respiratory protection, and approval by the Health Physics Manager (or Designee). The two technicians entered the RHR pit on
' September 30,1998, on RWP-0014, a Specific-1 RWP which did not have approval of the Health Physics Manager (or Designee) nor did it require respiratory protection.
Previous surveys of the RHR room were not reviewed prior to establishing controls i
for the entry.
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A review of previous survey data on the RHR room should have been performed under the circumstances.10 CFR 20.1501 requires, in part, that a licensee make or cause to be made surveys that may be necessary to comply with the repuluions in 10 CFR Part 20 and are reasonable under the circumeances to evr! bate the extent j
of radiation levels and the potential hazards that could be present. 10 CFR 20.1201 provides occupational exposure limits.10 CFR 20.1204 requires determination of internal exposures. The licensee's failure to review previous surveys of the RHR room, knowing that a significant spill of RCS decontamination liquid had contaminated the room with high levels of alpha radioactivity, to aetermine potential internal exposures to two individuals as necessary to comply with the requirements of 10 CFR 20.1204, appears to be a violation of 10 CFR 20.1501. This item will be reviewed further in future NRC inspections. (URI 50-213/98-05-02)
The inspector reviewed the licensee's actions following the event to prevent recurrence. The disciplinary actions appeared to be adequate and thorough including requalification of the supervisor. The RWP process was strengthened requiring weekly supervisory reviews. All areas with high contamination following the RCS decontamination were controlled and evaluated for changing beta / gamma to alpha ratios. This event was reviewed with the entire department in small group meetings. The inspector had no further questions regarding the follow-up actions The inspector examined the respiratory protection issue log and compared it to entries into high contamination areas. Random names and areas were selected for review. In all cases reviewed respiratory protection was use as required. In addition, Specific-2 RWPs were used where necessary on these occasions. Based on this review, the inspector determined that this event was an isolated event and is not indicative of a programmatic breakdown.
c. Conclusions
The licensee's failure to perform adequate review of existing survey data for the RHR room resulted in the intake of radioactive material by two individuals.
Although the exposures did not exceed 10 CFR 20.1204 limits, this failure to perform an adequate survey is a vit.Jation of 10 CFR 20.1501. The licensee conducted adequate follow-up actions including taking sufficient bioassay samples to assess the dose to the whole body and the bone. An unresolvod item will track the final dose assessments following additional sample analyses. A review of similar radiological areas and conditions by the inspector resulted in the determination that this event was an isolated one and is not indicative of program deterioration.
R1.2 Imolementation of Radioactive Gaseous Effluent Control Proaram durina the Decontamination Period a.
Insoection Scoce (84750)
The RCS decontamination process was performed on July 21-August 22,1998.
Implementation of the gaseous effluent control program during the decontamination period was reviewed. The inspection included the review of:
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(1)air balances of ths conteinmsnt building and the PAB; (2)analytical results for the in-plant air samples; (3)analytical results for the main stack air; and (4)project dose calculation results for the public.
b.
Observations and Findinas According to the Operations Log, the PAB maintained negative pressure during the months of July and August 1998. The containment exhaust fan was operated during the decontamination period to maintain the air direction from containment building to the air cleaning system. Air from the PAB and the containment building passed through the air cleaning system (HEPA for particulates). After cleaning, the air was released to the environment through the main stack.
About 250 air samples were taken from the PAB and containment building and analyzed for gross ply during the period from July 26 to July 29,1998. The majority of the measurement results were either below the minimum detectable concentration or in the range from 1E-11 pCi/cc to 1E-12 pCi/cc.
Analytical results of the main stack samples, from July 12,1998 to August 23, 1998, were reviewed for radioactive particulates. All measurement results were reported as below the lower limits of detection (LLD) (TS required LLD is 1E-11 pCi/cc). For the samples taken between 8/9-16/98 and 8/16-18/98,the measured radionuclide was Cs-137 and the activities were 1.59(iO.35)E-12 Ci/cc and
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4.59(10.50)E-13 Ci/cc, respectively, which were also lower than the TS LLD value.
Projected doses to the public from a potential release of all airborne tritium were calculated by the licensee and by the NRC. Projected tritium dose (5.47E-3 mrem / July 1998) was well below the TS limit (5 mrem / year). The projected Cs-137 dose was calculated by the NRC and the result was 1.2E-3 mrem.
c. Conclusions
The NRC reviewed the implementation of the gaseous effluent control program for the time periods when the RCS decontamination was in progress, and when the stack monitoring function was uncertain. The licensee maintained good air balance for the PAB and the Containment buildings; therefore, all air from these building was released to the environment through the air cleaning system (HEPA). Good laboratory measurement techniques for in-plant and main stack samples were implemented. The projected doses to the public from a potential release of all airborne activity during the RCS decontamination period were calculated and the results were insignificant.
R1.3 Contamination Controls a.
insoection Scope For Inspection Report 97-01, dated May 8,1997, two unresolved items (URI 97-01-04 and URI 97-01-12)were identified regarding the effectiveness of
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contamination controls, and training of personnel on use of contamination monitoring equipment. Interim actions were taken to restrict the release of material from radiological controlled and protected areas. The details of the NRC review of this matter are included here and in Section R.S.
The inspector reviewed selected aspects of the licensee's contamination control program. The review was in response to the identification by a vendor on February 27,1997, that video equipment, picked up by the vendor at the Haddam Neck facility as clean equipment on February 19,1997, was identified by the vendor to exhibit radioactive contamination. The inspector's review was against criteria outlined in 10 CFR Part 20 and applicable licensee procedures, b.
Observations and Findinas identification of Contaminated Eauioment at an Offsite Vendor Facility On February 26,1997, a non-licensed representative of a video equipment vendor notified the licensee's radiation protection manager (RPM) that low level radioactive contamination had been detected on video cables that had been picked up by its personnel at the Haddam Neck facility on February 19,1997. The contamination was detected by the vendor's personnel who normally perform routine contamination surveys of video equipment that is returned to the vendor's facility from reactor facilities.
The video equipment was removed from the normally locked steam generator mock-up building (located outside the station protected area (PA)) and transported from the Haddam Neck facility in a closed van on February 19,1997. The van arrived at the vendor's facility, located in an unrestricted area, on February 19,1997. The van was locked. No other material was loaded or unloaded from the van during the trip. The material was unloaded from the van on February 20,1997, and placed on plastic sheets within the vendor facility at the lay down area. In preparation for survey of the material, an area was set up at a work table at the facility and individualitems were surveyed by the vendor's personnel who wore gloves. The video equipment remained untouched at the lay down area untilit was surveyed on February 26,1997, found contaminated, and placed in plascic bags.
When notified, on February 26,1997, of the contaminated equipment, the RPM requested that the material be placed in secure storage. The licensee subsequently dispatched a radiological survey team on February 27,1997, to the vendor's facility to survey and retrieve the material. The following table (Table 1) identifies the j
material surveyed and found to exhibit detectable contamination. The survey results indicated the contamination was principally non-removable (fixed)contamination.
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i Receiver 2144-4B 10,000
< 1000 (LAS)
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< 1000 (LAS)i Receiver 1338-2G 1,500
< 1000 (LAS)
Carof Cable 1 1,000
< 1000 (LAS)
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Carol Cable 2.
1,500 1000 (LAS)120 volt cable 5,000 1,000 DPM/100 cm8 I
Camera No.1 4,000 1000 (LAS)
Camera No. 2
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Crescent Wrench 2,500
< 1000 DPM/100 cm2
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l contamination present; LAS - targe area contamination survey; DPM/100 cm -
contamination survey results per 100 square centimeters l
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The inspector noted that the facility itself was surveyed including the transport van, work table, desks, bathroom facilities, shop vacuums and their contents, lay down area, equipment on shelving, and trash bags. Personnel who handled the material were also surveyed and found not contaminated. The vendor ;7ersonnel had previously identified a contaminated crescent wrench which us retrieved. Also, a blanket, upon which the video equipment was placed in the van was treated as contaminated. All material was packaged in a 55-gallon drum and returned to the Haddam Neck facility as a limited quantity shipment on February 28,1997.
L An NRC resident inspector observed the initial licensee surveys of the material, vendor f acility, and van on February 27,1997. Also, an NRC health physicist was dispatched to the vendor facility on February 28,1997. The specialist performed
' independent radiological surveys of the vendor's facility and van and did not detect l
any residual contamination.
As a result of the identification of contr.minated material released from the station, the licensee took a number of interim actions on February 27,1997, to control and restrict the release of material from station warehouses and the site protected area and RCA pending identification of the cause of failure to detect the contamination and implementation of appropriate long term corrective actions. The licensee took the following actions on February 27,1997.
A directive was implemented that any material being removed from the RCA
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be monitored by use of the small article monitors (SAMs). If the material to
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be released could not be surveyed by the SAMs, then material was to be i
direct frisked and specific approval was to be obtained from the Director,
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Work Services for the material to be removed from the RCA. Further, all
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material was to be surveyed for removable contamination. The licensee subsequently revised applicable procedures to address this matter.
A security guard was stationed at the station PA egress to control release of
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material from the PA and a directive was implemented that any material to be removed from the protected area was to be evaluated by radiation protection (RP) and surveyed, as appropriate. All such material was to be surveyed by a SAM or direct frisked. Further, vehicles leaving the protected area were to be surveyed for contamination.
A directive was issued that no material may be placed in or removed from
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the warehouse unless surveyed by RP.
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in addition, on February 28,1997, the steam generator mock-up building (from which the contaminated cameras were removed) was surveyed. The following
~ additional contaminated equipment was identified.
U A camera was identified with 4,000 dpm/100 cm removable contamination.
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Two camera control boxes were identified with 300 - 500 dpm fixed
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contamination. The boxes were placed in a SAM which caused the SAM to alarm.
j No other contaminated equipment was identified in the steam generator mock-up building. The facility itself (e.g., walls, floors, equipment) was found to be free of contamination. In addition, to assist in the survey of warehouse facilities for additional material that may exhibit contamination, the licensee obtained four additional contractor RP technicians to perform surveys.
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in addition to the above, as a result of the identification of contaminated material offsite, the licensee initiated and completed (April 1,1997) a root cause evaluation of the release of contaminated material. The licensee's analysis indicated that the principal cause of the release was the lack of sufficient control procedures. Further, the analysis indicated that previous examples of uncontrolled release of contaminated objects from the RCA had occurred in the latter part of 1996 including four indications of detection of contamination / contaminated material outside the RCA in 1995. The evaluation indicates that a September 19,1996, i
report stated that two SAMs would be purchased and used to monitor equipment leaving the RCA. Notwithstanding this decision, the evaluation indicates that
' ineffective corrective action was taken for the previously identified matters, and that the SAMs were not used to monitor the contaminated camera equipment.
The inspector noted that 10 CFP 20.1501 requires, in part, that licensees make or cause to be made survt t may be necessary for the licensee to comply with the regulations in 10 CF., Pan 20 and are reasonable under the circumstances to evaluate the extent of rL..,on levels and the concentrations of radioactive materials and the potential radiological hazards that could be present. 10 CFR 20.1003 defines a " survey" as an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive materials. When appropriate, such an evaluation includes a
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physical survey of the location of radioactive material and measurements or l
l calculations of levels of radiation, or concentrations or quantities of radioactive material present.
The inspector noted that 10 CFR 20.2001 provides general requirements for disposal of licensed material and 10 CFR 20.1301 provides specific dose limits for
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members of the public. The inspector noted that the contamination surveys of equipment and material released from the licensee's radiological controlled area were inadequate to ensure compliance with the requirements of 10 CFR 20.2001 in that as discussed above, the licensee unknowingly released radioactive contaminated material to in individual (camera vendor) not licensed to received radioactive material. Further, the surveys were inadequate to ensure compliance with the individual dose limits for members of the public in that the licensee unknowingly released contaminated material and consequently, an evaluation was not performed to evaluate the potential hazards incident to the release of the
radioactive material. The release of the contaminated materialis an apparent violation of 10 CFR 20.1501. This closes URI 97-01-01. (eel 50 213/98-05-03)
Review of Monitor Used to Survev Video Eauipment l
The inspector reviewed the monitoring equipment used to survey the camera equipment released from the RCA. The monitoring equipment used was a waste sorting table with 18 scintillation detectors. Each scintillation detector exhibited a j
surface area of 625 cm2 Material was placed on the detector array for monitoring.
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Based on calibration information (February 1997) provided by the licensee's health physics support staff, the 18 detectors appeared to have an average minimum detectable activity (MDA) of approximately 1.5 nanocuries (nCi) (i.e.,3,300 dpm).
The MDA ranged from approximately 1.1 nCi to 2.1 nCi.
The inspector noted the following and made the following observations.
The inspector noted in discussions with one technician that had surveyed the
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l camera equipment in late January 1997, that the technician did not
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l completely understand the procedure and would have unknowingly not followed the procedure for using the waste sorting table. Specifically, the
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technician indicated that had a detector alarmed, the material would be recounted and if a second count indicated clean, the material counted would be considered clean. However, the procedure for operation of the waste L
sorting table (procedure RPM 2.3-14, Revision 2, Operation of the NNC WST-18), specifically indicates in Section 3.3.7 that "If the WST-18 alarms, REMOVE the items over the alarming detector and dispose of them as contaminated wastes...". The technician indicated none of the surveyed camera equipment alarmed the detector.
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_On March 6,1997, the inspector observed material (retrieved from the offsite camera vendor's facility) to readily alarm the sorting table when placed on the table.
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The inspector noted that the MDA of the waste sorting tab le was highly
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background counting prior to each use of the table. The inspector noted that there were no procedure controls to ensure minimum background when performing background checks to ensure that the smallest possible MDA was achieved prior to use cf the device. The inspector's discussions with a j
technician who had surveyed the camera equipment indicated that the building that housed the table was filled with a significant amount of material to be surveyed when the camera equipment was surveyed for release.
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The detection capability of the waste sorting table was highly dependent on distance of material from the detector surface. Although guidance for material not to exceed two inches in height was contained within procedures, large objects were 71 aced on the table. Further, contaminated
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material was apparently found.nside equipment retrieved from the vendor's l
offsite facility.
The inspector reviewed operational check sheets for the sorting table and
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noted the sheets to indicate satisfactory. No control charts or other data sheets were available to provide indication of day-to-day instrument performance.
At the time of the initialinspection, the licensee had not taken action to suspend use of the waste sorting table pending determination of the root cause of the release of contaminated material and implementation of appropriate corrective actions. The licensee subsequeM!y suspended use of the waste sorting table.
c.
Conclusion The licensee did not adequately survey potentially contaminated material.
Contaminated material was detected outside the RCA, including being in the possession of a non-licensed individual. The failure to perform adequate surveys of potentially contaminated materialis an apparent violation of 10 CFR 20.1501.
Further, the procedures for use of the waste sorting table were considered poor.
R1.4 Radioloaical Survevs a.
Insoection Scoce The purpose of this inspection was to continue the review of licensee radiation and contaminatio.t surveys at private residences and properties located offsite that had received cement blocks or other material from the Haddam Neck plant in the past.
The inspector reviewed the planning and implementation of remediation plans at offsite locations.
b.
Observations and Findinas Unresolved item 97-09-04 concerned NRC raview of the licensee actions to survey offsite areas that received soils, concrete blocks and other potentially contaminated l
equipment from the plant, and to recover contaminated materials for proper disposal as radwaste. Past NRC reviews in this area were described in NRC Inspation
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Report Nos. 97-09,97-10,98-01 and 98-03. The inspection methodologies
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i dsscrib:d in the referenced reports continued during this period. Soil samples obtained from offsite properties were split with the licensee and analyzed at the NRC's Region I Radiation Measurement Laboratory in King of Prussia, Pennsylvania.
NRC review of the licensee and NRC data continued at the end of this inspection.
Preliminarv Dose Assessments NRC findings continue to show that no significant radiological safety concerns exist from any of the Haddam Neck materials recovered to date from offsite locations.
This conclusion was based on NRC and licensee data from direct garnma exposure rates and measured soil concentrations, and preliminary dose assessments. The
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dose assessments considered the potential whole body dose and accounted for occupancy factors. Consideration of potential exposure from transuranic isotopes 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 follow up assessment were in progress at the conclusion of this inspection, j
NRC review of all sites was in progress at the conclusion of the inspection. Item 97-09-04 remains unresolved pending the completion of the licensee actions to identify and recover concrete blocks, and evaluate other plant related materials.
c.
Conclusion Licensee activities this period continued to be good to conduct follow-up 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 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 inspection.
R1.5 Radioloaical Controls a.
Insoection Scoce (83750)
The inspector toured the RCA and discussed specific radiological controls with RP supervision and RP technicians. The inspectors also reviewed radiological controls implemented for work in the RCA including RWPs and associated radiological surveys.
b.
Observations and Findinas The inspectors toured various areas within the RCA including the PAB, the containment building, and outside areas. Appropriate controls were observed for Radiation Areas and High Radiation Areas. Postings and barriers were effectively l
l placed to notify workers regarding changes in radiation levels. Appropriate controls were noted to prevent the spread of radioactive contamination. The inspectors l
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- noted that workers were taking the proper precautions for radiation protection.
l During tours, good radiological housekeeping and good worker awareness of radiological hazards were noted.
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The inspectors reviewed the preparations for the core barrellift a..a other work in the containment building. The RP staff performed a good pre-job review including
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i the use of respirators to maintain total effective dose equivalents as low as is l
reasonably achievable (ALARA). A very good pre-job briefing was held with all
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workers to review the radiological conditions and radiation protection requiremerts.
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c. Conclusions
Radiological controls for radiological work were well-planned and RP personnel i
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l maa.tained close oversight of work. Radiation protection planning was effective to i
maintain workers' radiological exposures ALARA.
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R5 Staff Trainina and Qualification in Radiation Protection & Chemistry
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insoection Scope (83750)
l The inspectc,r selectively reviewed the training provided to the licensee's RP staff relative to the prcgrammatic weaknesses identified during the November 2,1996, I
fuel transfer canal airborne radioactivity event. The inspector also reviewed trainir,g
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of personnel to perform contamination monitoring.
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Observations and Findinas
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November 2.1996, Fuel Transfer Canal Event The licensee provided training to the radiological controls group on the
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circumstances surrounding the November 2,1996, fuel transfer canal airborne radioactivity event. The NRC inspection report, which detailed the NRC reviews of the event, was provided, including the licensee's internal lessons learned document.
Also, the licensee provided training on NRC Information Notice No. 92-75, l
" Unplanned intakes of Airborne Radioactive Material By Individuals at Nuclear Power
Plants," dated November 12,1992, which was referenced in the inspection report.
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Bglease of Contaminated Camera Eauioment to an Offsite Vendor
The inspector reviewed the training and qualifications of personnel who calibrated l
and operated the waste sorting table. As discussed above, camera equipment, that was apparently surveyed for contamination on the waste sorting table was identified to be contaminated and located at an offsite vendor facility. The inspector's review indicated the following.
l The inspector was not able to identify a formal training program or training
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The inspector was not able to identify formal training records to indicate that
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two of the three individuals who surveyed the camera equipment were in l
fact trained and qualified. The inspector was not able to identify a training and qualification program for operation of the waste sorting table.
The documentation of the training and qualification of personnel for calibration and operation of contamination monitoring equipment was considered poor. However,
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no violation of NRC requirements was identified. This matter was previously identified as an unresolved item (URI 97-01-12).
Subsequent to the NRC finding the following actions were taken. The waste sorting table was discontinued for free release surveys. Technician training was completed on the proper use of instrumentation for the free release of material. The inspector j
observed proper release of material by senior HP technicians. This item is closed
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l (URI 97-01-12).
c.
Conclugi9B The licensee provided training to the radiological controls group on the November 2, 1996, event. The documentation and records in the area of training and qualification of personnel who calibrate and operate contamination monitoring equipment war, poor. Subsequent actions taken to improve detection capability of instruments and train technicians for the release of material have been effective in j
contamination control.
R8 Miscellaneous Radiological Protection issues i
R8.1 Status of Previous inspection items I
(Closed) Insoector Follow-up item 9E-27-03: Contamination Event. This item was open pending the review of licensee actions to improve the controls of contaminated materials.
This area was previously reviewed in inspection Report Nos. 97-06,97-08,97-10 and 98-02, which provided NRC reviews of the licensee's program to control contamination, and the efforts to improve the radiological controls program. There have been no subsequent events regarding the loss of control of contaminated hoses. This item is closed.
(Closed) LER 97-03: Contrcl Room Habitability. This item is closed.
(Closed) Unresolved item 96-02-03: Crntrol Room Habitability. This item and LER 97-03 l
involved the adequacy of analyses and procedures to assure the CR remained habitable l
following design basis events. Licensee actions to assure procedure quality were addressed in inspection Report No. 96-11 and memorandum CY-TS-97-0316. Following
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the decision to permanently cease power operations, the licensee performed a calculation, with quality checks, for the new design basis accidents with Haddam Neck in a decommissioning status. Calculation CTCRDOSE-1-RAB evaluated the doses for operators in the CR for postulated accidents as defined in UFSAR Section 15.5.1 (waste systems accidents) and Section 15.5.2 (fuel handling accidents in the SFP). The results for both accidents showed results below the NRC guidelines # 5 rem whole body,30 rem skin and 30 rem thyroid. This item is closed.
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(Discussed) URI 50-213/98-03-07and Corrective Action Plan for RMS-14B:
The projected dose impact (due to gaseous effluent releases) to the public incorporating the:
- (1) correction of the line loss;
- (2) compensation of anisokinetic sampling line; and
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- (3) correction'of the main stack flow rate, would be evaluated.
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=The inspector reviewed ACRs related to the main stack effluent monitoring system
(including the URI) and determined that the total error analysis technique would be
suitable to resolve this item. The following areas, as illustrated in Figure 1, can be
- used for the error analysis of the main stack effluent monitoring system for historical data:
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I an error of air capacity, inlet to the main stack; a2:
error of air capacity, outlet from the main stack; a:
error of isokinetic nozzle and sampling line; s
as error of line loss for iodine and particulates; l
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error of measurement laboratory; i
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error of sample volume; and ag
. other errors, if any.
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The licensee had a similar method to quantify the uncertainties. Reviewing and discussing these errors with the licensee, the inspector noted that the licensee had
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the right approach to resolve this issue. Dose calculations and impact to the public will be complete by the licensee before the end of 1998 and/or beginning of 1999.
i S1 Security and Safeguards S1.1 Review of Security Activitigs
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inspection Scope (81700)i The purpose of this inspection was to determine whether the conduct of security activities met the licensee's commitments in the areas of access control and the
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control of personnel, packages / material, and vehicles.
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Observations and Findinas '
The inspector observed operations at the personnel access portal at various times during the inspection. Positive controls were in p! ace to ensure only authorized
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. individuals were granted access to the PA. The inspector observed security force i
operations et the vehicle entry point including verification of authorization and performance of vehicle searches prior to entry into the PA. All personnel, materials and hand carried items entering the PA were properly searched.
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The inspector observed the authorization and access control for plant vital areas, including the use of compensatory measures during periods of off-normal conditions. Positive controls were in place to ensure only authorized individuals were granted access to the vital area.
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Defueled Security Plaa On September 20,1998, the licensee implemented the defueled security plan. The revised security procedures that supported the new plan included changes to:
portions of the PA that became the industrial area (IA), with new IA access requirements; PA access controls; vehicle access control; and use of the secondary alarm station and the assessment aids. The licensee continued to make physical modifications to the facility in support of the plan changes. The licensee instituted compensatory measures as necessary pending completion of the modifications. The inspector verified the compensatory measures were in place during routine tours of the facility, and found that security officers were knowledgeable regarding post orders.
f Loss of Security Eouioment Power
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The inspector reviewed the response by the security force following a loss of power to security equipment at 5:15 p.m. on October 25,1998. Compensatory actions were taken in accordance with procedures SEC 1.3-32, SEC 1.3-41 and SCP 1.8-1.
The power to security equipment was restored by 5:47 p.m. on October 26,1998.
The loss of security power for about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> highlighted a degraded condition in which the security emergency diesel generator was not fully operable. The
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emergency power source was not capable of connecting to security loads since June,1998. The licensee's security plan (both for the defueled condition and the commitments in effect prior to September 20,1998) describe requirements for emergency power for security equipment. Other diesel generators at the site (EG-2A, EG-3B and EG-7) remained available during the period of interest. While actions continued to address degraded equipment problems, the licensee completed a temporary modification under Jumper 98-52 to restore full operability of the security diesel generator on November 5,1998. This matter is unresolved pending
further licensee and NRC review of the availability of emergency power for security loads (UNR 98-06-04).
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c. Conclusions
The licensee conducted security activities in a manner that protected public health and safety. The response to off-normal conditions was good to assure security functions remained effective. Compensatory measures were taken per procedures and instructions. One discrepancy was noted in the failure to assure emergency power for security equipment remained functional. An unresolved item will track l-the review to determine whether the availability of emergency power met the requirements of the security plan.
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P3 Emergency Preparedness Procedures and Documentation P3.1 Review of Exercise Obiectives and Scenario (82302)
The inspectors reviewed the 1998 exercise objectives and scenario and determined that they acceptably exercised major elements of the licensee's defueled emergency plan. The scenario included a declaration of an Unusual Event (UE) based on a fire
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in a RCA for greater than fifteen minutes; an Alert declaration based on damaged fuel with radioactive gas (primarily Krypton-85) released to the environment through
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the primary vent stack; and dispatching of a Search and Rescue Team for an unaccounted individualin the SFP Building. This scenario provided an adequately challenging framework to support demonstration of the licensee's emergency plan, procedures and defueled emergency response organization (DERO).
i P4 Staff Knowledge and Performance i
P4.1 1998 Evaluated Biennial Defueled Emeraency Preparedness Exercise
a. Inspection Scope
(82301)s
The purpose of this inspection was to observe and evaluate the licensee Defueled Emergency Plan (DEP) Exercise to assess the licensee's implementation of their
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emergency preparedness program and procedures. The inspectors observed the biennial emergency exercise conducted on September 23,1998, at the two emergency response facilities (CR and TSC). Findings are identified specific to NRC and licensee exercise objectives, rather than specific to the emergency facility.
Findings related to management of the exercise were based on the licensee's pre-exercise briefings, and the licensee's player, controller and management critiques.
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b.
Observations and Findinaq
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Shift Staffino and Command and Control The DEP describes the on-shift and augmented organizations that are intended to be part of the overall response organization in the event of an emergency at the Haddam Neck Facility. During this exercise, the licensee's normal on-shift
- decommissioning organization was seven individuals, including the Operations Shift Manager, who is the NRC Certified Fuel Handler. They completed the DEP
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responsibilities of emergency classification, command and control, offsite notification, authorization of onsite protective actions, and notification for activation of the DERO well within the response goal of one hour. Augmentation of the DERO I
(six additional positions) and continued implementation of the Defueled Emergency
Plan Implementing Procedures (DEPIP) were completed well within the response
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goal of two hours from the time of an ALERT classification.
The TSC was activated in accordance with the DEP response timeliness goals.
Responding personnel were familiar with their duties and responsibilities and utilized established procedural guidance in accomplishing assigned tasks in an effective manner. Back-up personnel, who were readily available due to the nature of the t!:ning of the scenario events, were also effectively utilized for several positions.
The licensee demonstrated that they had properly staffed the DERO at the TSC.
The inspector determined that the licensee had resolved the concern with over-staffing that was identified during the last biennial exercise (IFl 96-007-03). This item is closed.
The inspectors noted that there was a delay in the declaration by the Emergency Director (ED) regarding acceptance of the command-and-control responsibilities from i
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the Shift Manager (SM) and the TSC activation. Ths ED had assumed that the Health Physics Network (HPN) and the Emergency Notification System (ENS) had to i
be staffed, a procedural step in DEPIP 1.5-21, " Emergency Director", before announcing that the TSC was activated. The DEP response goal assumes that the TSC will be activated when the key response positions are staffed, as opposed to all positions.
Procedure Use The on-s'hift organization demonstrated good knowledge of and use of annunciator, AOPs and DEPIPs to confirm plant conditions and provide timely and accurate status information to the rest of the DERO. Augmented response personnel effectively used established procedural guidance to carry out assigned tasks.
I Duties, responsibilities, and turnover points for subsequent responders were clearly
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detailed and implemented in accordance with the established procedures.
Implementation of specific emergency response actions are detailed below.
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Accident Detection, Assessment and Classificatina
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l The SM, acting as ED, implemented the DEPlPs to promptly assess symptoms and classify events correctly. Key parameters appropriate for scenario events were i
determined and monitored by personnel in the TSC. Area radiation monitors and
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process monitors were utilized to assess plant radiological conditions and to estimate the magnitude of an offsite release. Monitor readings were obtained from
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the control room and recorded for display on a status board in the TSC. Reports from personnelin the field were utilized to confirm radiation monitor readings, and provide additional details to provide a clear picture of onsite radiological conditions.
Notification The SM/ED promptly notified the station and the DERO of the emergency classifications, and demonstrated good knowledge and use of the notification equipment and forms. The ED also demonstrated the ability to make timely notifications to the State of Connecticut and NRC. The DEP requires these notifications be completed within one hour of emergency classification. Because the exercise progressed from an UE to an ALERT quickly, the notification to the
NRC only addressed the ALERT conditions, and did not indicate that a fire had occurred which was the basis for an earlier UE declaration.
Emergency response personnel first received notification of event declarations by pagers. Soon after the pagers were activated, the plant public address (PA) system was utilized to notify onsite personnel of emergency conditions, classifications and the activation of the emergency response organization. PA announcements were clear, timely and informative. One ALERT announcement was inadvertently i
communicated as an UE; however, this was quickly corrected by a follow-up announcement.
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Communications Radio communications between the control room and the fire brigade were clear and concise. Request for actions such as de-energizing an electrical bus to assist the fire-fighting efforts were clearly communicated and acknowledged. However, when the TSC was first staffed, the status of the fire was not completely understood. The inspectors observed that the control room staff did not orderly log information of plant events. The need to log information was evident in the failure to note the actual time that the fire was reported by the fire brigade (9:30 am). The SM had recorded on a piece of paper this event at an estimated time (10:00 am). Inplace communication links and assigned responsibilities were followed to obtain the information need to resolve the question of the status of the fire-fighting effort.
During their critique, the licensee identified the need for maintaining logs by key DERO positions.
When establishing contact with the NRC over the HPN phone was a competing priority for the Radiological Assessment Coordinator, other available personnel were utilized to complete the task to ensure that appropriate communication links were established. Communications between the ED and the State Representative at the TSC, and telephone discussions between the ED and the CT Emergency Management Organization were timely and effective. The licensee identified tha communications by the ED as a strength during this exercise. The inspectors found that communications with the State had improved and the concern identified during the previous biennial exercise (IFl 96-007-04) had been adequately addressed. This item is closed.
Radic:oaical Exposure Control The SM/ED, Fire Brigade Leader and on-shift HP Technician performed well to initiate onsite actions in response to degraded conditions, including actions to respond to the fire and protect personnel from the gaseous radiological release.
Supervisory personnel in the TSC actively monitored radiological conditions and the impact of changing radiological conditions to personnel onsite. The Security Director considered the effect that both the plume of smoke, and later, the radiological release, could have on security guards stationed in the field. Known radiological monitoring information from reports of persons onsite, as well as installed radiation monitors, were used to determine radiological hazards and j
planning for the search and rescue team. Based on available information,
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appropriate protective measures were determined and the response team was briefed on expected conditions. The briefing included information on the best route to a work area (and return route) for assigned tusks based on known and expected radiological conditions. Dose limits were evaluated and extended to ensure
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sufficient margins to allow the team to accomplish their assigned tasks.
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The recently approved DEP discontinued offsite protective actions, thereby negating l
any objectives for offsite protection action assessment and recommendations.
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f Search and Rescue The need for the implementation of search and rescue measures was appropriately
l determined. The ED directed the Radiological Assessment Coordinator (RAC) and L
the Technical Response Coordinator (TRC) to coordinate efforts and dispatch the l
needed team. Established procedural guidance was followed to prepare, brief, and dispatch the search and rescue team. Team planning, preparation, and briefings were complcte and thorough. However, the inspectors noted that tha dispatching of this urgently needed team was delayed by administrative detail prescribed in the DEPIP and thoroughness of the briefings for the search and rescue team members.
Rumor Control l
The Public Information Coordinator called in a back-up person to assist with responding to numerous inquiries. Approved procedures and forms were utilized to respond to information requests. The inspectors identified that two press releases l
indicated that the county officials had been notified of the event, which was
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determined not to have occurred. The licensee attributed this mistake to familiarity
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with the former Site Emergemy Response Organization (SERO), which included offsite communications with the state and county governments.
Drillmanshio l
Both the players and controllers demonstrated good drillmanship in n,ponding to -
scenario events. Controllers were visible and available to provide r ided
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I information and control the exercise conduct, yet maintained an appropriate level of separation from players and not interfere with their actions. Players maintained a l
professional response attitude during the conduct of the exercise.
Conduct of Exercise Critiaue The licensee conducted several post-exercise critiques to solicit input independently from ths players, controllers and oversight observers. This was followed by a joint critique with drill personnel, and lastly a management critique. The inspectors observed several of these critiques, and noted that the licensee identified the same issues that were observed by the NRC. The licensee's self-assessment was thorough and appropriately critical, which the inspectors determined was an exercise strength.
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c. Conclusions
The licensee conducted an adequate exercise in accordance with the DEP and DEPIP. All exercise cbjectives were met. Two IFis were closed.
The CR and TSC were staffed ano. tivated in a prompt manner. The ED and the response staff demonstrated the following: adequate assessment, classification, l
notification, command and control, and onsite protection actions. The search and rescue effort was delayed because of cumbersome procedures. The licensee's self
critique was a noted strength. The licensee identified numerous issues and I
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categorized those issues according to significance. The NRC agreed with the licensee's findings IV. Manaaement Meetinas i
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Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management periodically during the inspections, and at the end of the inspection on October 29,1998.
The licensee acknowledged the findings presented.
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The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary, No proprietary information was identified.
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- PARTIAL LIST OF PERSONS CONTACTED
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. Russel1 Mellor, Vice President Operations and Decommissioning
' Ken Heider, Decommissioning Director
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Gary Bouchard, Unit Director
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Kerry Harner, Chemistry Manager-Doug Heffernan, Maintenance Manager
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Gerry Waig, Operations Manager l
James Pandolfo, Security a *anager l.
Richard Sexton, Radiation Pru action Manager Gerry van Noordonnen, Nuclear Licensing Pete Hollenbeck, Site Characterization Supervisor
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Keith Sickles, Design Engineer
. Edward Bingham, Engineering
- John Haseltine, Enginsering Director Jay Tarzia, HP/ Chemistry Technical Support
._ Kathleen Burgess, Emergency Planning Coordinator Lead
- Steve Hook, Manager, Emergency Planning for CY and MP '
Dan McDavitt, Emergency Planning Tom Blount, Emergency Planning Jim Lenois, Security Coordinator
-Joseph Bourassa, CY Oversight Manager
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Roy Brown, Staff Assistant to Unit Director
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l Ralph Reeves, Operations Dale Flick, NU, Radiological Engineering Services
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. Anthony Nericcio, Public Affairs -
. Gregory Wilson, Public Information Coordinator.
Robert Greenfield, Site Characterization Michael Faivus, Site Characterization State of CT Daa= tment of 0;Weasieatal Protection Dennis Galloway, Radiation Specialist
. Michael Firsick, Radiation Specialist
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William Raymond, Senior Resident inspector Marie Miller, Senior Health Physicist
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Dan Bares, Emergency Preparedness Specialis*, ' AR Dr. Jason Jang, Senior Radiation Specialist -
Thomas Fredrichs, Project Manager, NRR John Wray, Decommissioning Health Physicist Joseph Nick, Decommissioning Health Physicist
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INSPECTION PROCEDURES USED (P 37700:
Design Changes and Modifications IP 37801:
Safety Reviews, Design Changes, and Modifications at PSRs IP 60851:
Design Control of ISFSI Components IP 60854:
Preoperational Testing of an ISFSI iP 61726:
Surveillance Observations
. IP 62707:
Maintenance Observations IP 62801:
Maintenance and Surveillance at Permanently Shutdown Reactors IP 71707:
Operational Safety Verification IP 81700:
Physical Security Assessment PSRs IP 82302:
Review of Exercise Objectives and Scenarios for Power Reactors IP 82301:
Evaluation of Exercises for Power Reactors IP 83750:
Occupational Radiation Exposure IP 84750:
RadWaste Treatment, and Effluent & Environmental Monitoring IP 90712:
Inoffice Review of Written Reports of Nonroutine Events at Power Reactors Facilities IP 92700:
Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92702:
Follow-up on Corrective Actions for Violations and Deviations
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ITEMS OPEN, CLOSED, AND DISCUSSED Open 98-05-01 URI Availability of Security Emergency Power
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98-05-02 URI Dose Assessment for Worker in RHR Pit 98-05-03 URI Failure to Review Radiological Conditions Prior to Work in the RHR Pit 98-05-04 eel Failure to Survey / Release of Contamineted Equipment to an Offsite Vendor Closed See Attachment 2 95-27-03 IFl Contamination Event-96-02-03 URI Control Room Habitability
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96-06-03 URI Followup LER 96-08 and 96-09 96-06-06 URI Battery Oscillations
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96-07-03 IFl Overstaffing of the Site Emergency Response Organization 96-07 04 IFl Communication Problems with State of CT
96-10-01 URI SNM Inventory Requirements
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96-201-08 VIO Procedure for Vendor Contacts 96-201-35 VIO Disposition Non-conformances
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97-01-07 URI SW System Modifications
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97-01-08 URI SW System Corrosion 97-01-11 URI Effectiveness of Contamination Controls j
l 97-01-12 URI Training on Use of Contamination Monitoring Equipment 97-03-04 URI SFP Heat Exchanger Performance
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i 97-09-03 URI QA Classification for SFPCS 98-01-03 URI Quality Attributes for Nuclear Island 98-01-04 VIO Seismic Monitoring instrumentation 98-03-04 URI Process for Commitment Tracking 98-03-06 VIO Failure to Meet TS requirements for the Calibration of Stack Flow Instruments L
LER Internal Flood Protection 97-01 LER River Temperature Below UFSAR Limit l
97-03 LER
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97-21 LER Contaminated Material Found Offsite 98-008 LER Stack Radiation Monitor RMS-14B Samples Not Analyzed Properly 98-009 LER Excessing Check Valve Seat Leakage Discussed 97-09-04 URI Survey and Assessment of Offsite Areas i'
98-03-07 URI Projected Dose Impacts (RMS-14B)
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LIST OF
DOCUMENTS REVIEWED
The following documents were reviewed to evaluate the exercise objectives and scenarios
for a defueled power reactor:
Defueled Emergency Plan for Haddam Neck Plant - Section 13.3 of the Updated
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Final Safety Analysis Report, dated February 1998;
CYAPCO Letter to NRC dated March 25,1998 - Revised Requested Exemptions
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from 10 CFR 50.47(b) and 10 CFR 50 Appendix E; and
NRC Letter to CYAPCO dated August 28,1998 - Exemption from a Portion of 10
CFR Section 50.54(q) and Approval of Defueled Emergency Plan at HNP
The following Defueled Emergency Plan implementing Procedures' were used during
evaluation of the exercise:
DEPIP 1.5-1, Emergency Assessment Using Defueled EAL Tables
DEPIP 1.5-2, Notification and Communications
DEPIP 1.5-3, Notification of Unusual Event
DEPIP 1.5-14, Relocation and Assembly
DEPIP 1.5-19, Technical Support Center Activation
DEPIP 1.5-21, Emergency Director
DEPIP 1.5-22, Radiological Assessment Coordinator
DEPIP 1.5-23, Security Director
DEPIP 1.5 24, Communication Coordinator
DEPIP 1.5-25, Public Information Coordinator
DEPIP 1.5-26, Manager of Control Room Operations
DEPIP 1.5-28, Technical Response Coordinator
DEPIP 1.5-42. Emergency Teams
DEPIP 1.5-61, On-shift Dose Assessment
DEPIP 1.5-62, News Releases
DEPIP 1.5-63, Rumor and inquiry Control
DEPIP 1.5-65, Refined Dose Assessment
DEPIP 1.5-43, Personnel Radiation Exposure Control and Dosimetry issue
During Nuclear Emergencies
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The Defueled Procedures, Revision I were considered effective for the
purposes of the exercise. However, the effective date was scheduled for
October 4,
1998, to allow for implementation by the State of Connecticut.
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' LIST OF ACRONYMS USED
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'ACP
Administrative Control Procedure
ACR
Adverse Condition Report
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As Low As is Reasonably Achievable
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ANSI
American National Standards Institute
AOP.
Abnormal Operation Procedure
AWO
Automated Work Order
Corrective Action Program.
CR
Control Room
' Design Change Record
Defueled Emergency Plan
DEPIP -
Defueled Emergency Plan implementing Procedures
DERO
Defueled Emergency Response Organization
'dpm~
Disintegrations per Minute
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Emergency Director
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' ENS
. Emergency Network System
F.
Fahrenheit
Final Safety Analysis Report
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GDC.
General Design Criteria
HNF.
Haddam Neck Facility
Health Physics
Health Physics Netwcrk
IA -
Industrial Area
Intermediate Cooling System
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LD/LB
Licensing Basis / Design Basis
LERs
Licensing Event Reports
LLD-
Lower Limits of Detection
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Motor Control Center
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'MDA
Minimum Detectable Activity
Non-Conformance Report
. Normal Operating Procedures
Notice of Violation
NRC
Nuclear Regulatory Commission
Nuclear Reactor Regulation
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Office of Investigations
Protected Area
PAB:
Primary Auxiliary Building
- PWST
Primary Water Storage Tank
Quality Assurance
RAC
Radiological Assessment Coordinator
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Radiological Controlled Area
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RIE
Replacement item Evaluation
RM:
Radiation Monitor
Radiation Protection
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RPM-
Radiation Protection Manager
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.RWP
. Radiation Work Permits
Small Article Monitors
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SC
Spray Cooling
SERO
Site Emergency Response Organization
Spent Fuel Pool
SFPCS
Spent Fuel Pool Cooling System
FFPI
Spent Fuel Pool Island
SM.
Shift Manager
Structures, Systems and Components
Surveillance Test
Service Wate'r
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TRC--
. Technical Response Coordinator
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TS
Technical Specification
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TSC-
Technical Support Center-
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Unusual Event
Updated Final Safety Analysis Report
Uninterruptible Power Supply
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Attechment 1
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Spent Fuel Island Modifications
The inspection of the design, modification and testing of the nuclear island included
observation and/or review of design change records (DCRs), special tests (STs) and
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- associated safety evaluations, and work orders (WO) on a sampling basis, including the
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following references:
MODIFICATION PACKAGES
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DCR-97015, Spent Fuel Island Phase 1 - Electrical Equipment Installation
DCR-97016, Spent Fuel Island Phase 1 - Engineered Supports and Piping
DCR-97017, Spent Fuel Island Phase 1 - Mechanical Equipment Installation
DCR-97018, Spent Fuel Island Phase 1 - Instrumentation Installation -
DCR-97032, Spent Fuel Island - Interim Power Supply to MCC-2A and MCC-2B
TEST PACKAGES
ST 11.7-211, Phase i SFPCS IC and SC Loop Component Test and Flush
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ST 11.7-212B, Phase i SFPCS Testing - B Loop Thermal Performance Test
PROCEDURES
NOP 2.10-1, Spent Fuel Cooling System Operation, Revision 18
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NOP 2.10-5, Spent Fuel Pool Makeup, Revision O
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NOP 2.10-7, Adding Boric Acid to the SFP, Revision O
AOP 3.2-59, Loss of Spent Fuel Cooling, Revision 5
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ANN 4.17-4, B Spray surge Tank at Auto Makeup Level, Revision O
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ANN 4.17-8, B Spray surge Tank Low Level, Revision 0
ANN 4.17-12, B Spray surge Tank High Level, Revision O
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ANN 4.17-16, B Intermediate Surge Tank Low Level, Revision O
ANN 4.17-20, B Intermediate Surge Tank High Level, Revision O
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ANN 4.17-24, SFP Cooling Instrument Power supply Trouble, Revision O
CALCULATIONS
Calculation 97-SFPI-01610-MY,SC System Thermal Hydraulic Model, 2/19/98
Calculation 97-SFPI-01611-MY,1C system Thermal Hydraulic Model,3/13/98
Calculation 97-SFPI-01585-MY, Structural Qualification of IC/SC Skids,9/30/97
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Calculation 97-SFPI-1601-CY, Piping Analysis for SFP SC System, 10/23/97
Calculation 97-SFPI-1602-CY, Piping Analysis for SFP IC System, 10/29/97
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Calculation 97-SFPI-1605-MY,SC & IC Tanks Design, 12/18/97
Calculation 97-SFPI-1580-CY, Structural Qualification of Fan Anchorage,9/30/97
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Fallow-up of Escalated Enforcement items
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numerous inspections to review several facets of Haddam Neck performance. The
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violations were grouped into categories that included inadequacies in engineering programs
and practices; operational and procedural deficiencies; and, numerous deficiencies in the
corrective action program. The NRC review of licensce is described in Section E8.2 of this
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report. The NRC review of actions to address specific violations is summarized below. For
each item discussed, the violation (VIO) number referenced in the May 12,1997 Notice is
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listed, along with the original inspection report item number (if applicable) in parentheses,
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' Violation issues No Longer Applicable to Decommissioning
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Violations and Regulatory Concerns Related to Engineering
NRC reviews of licensee actions to correct deficiencies)lri the conduct of 10 CFR 50.59
safety evaluations, and to accurately maintain the UFSAR were reviewed previously, as
described in Inspection Report No. 97-03, and Section E8.2 of this report. Some of the
1997 enforcement action issues concerned aspects of plant operation 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 (program revisions and plant
modifications) for certain deficiencies were no longer appropriate. The licensee provided
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documentation to demonstrate the actions taken to address specific issues, and to assure
the correction of deficiencies for equipment or programs that remain important to
decommissioning. The items listed below are closed.
1.
URI 96-06-03: Regulatory Guide 1.97 Program for RHR and Plant Components
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URI 96-06-06: Battery Oscillations and DC Ground
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IFl 96-08-09: CMP Action to Translate Licensing / Design Basis
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URI 96-08-15): Startup Issues Related to MS Bridge and Station Batteries
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VIO 01362 (IR 96-201-18): Evaluation of EDG KW Meters
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VIO 01062 (IR 96-201-24): Seismic Qualification for EDG Air Start Piping
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VIO 01112 (IR 96-201-36): EDG Relay Classification
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VIO 01272 (IR 96-201-04): Evaluation for Battery Charger Replacement
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VIO 01023 (IR 96-201-28):EDG Testing an.d Load Calculations
10. VIO 01352 (IR 96-201-01): Failure to Update UFSAR per 10 CFR 50.71(e)
11. URI 96-201-24: Evaluation of Potential SW Waterhammer
.12. VIO 01042 (IR 96-201-22): Failure to Evaluate SW Two-Phase Flow
13. VIO 01202 (IR 96-201-05): Failure to Evaluate PAB Floor Blocks
14. VIO 01212 (IR 96-201-02): Failure to Evaluate UFSAR Change
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15. VIO 01222 (IR 96-201-02): Failure to Evaluate UFSAR Change
17. VIO 01232 (IR 96-201-02): Failure to Evaluate UFSAR Change
18. VIO 01242 (IR 96-201-02): Failure to Evaluate UFSAR Change
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19. VIO 01252 (IR 96-201-03): Failure to Evaluate Change per 10 CFR 50.59
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20. VIO 01262 (IR 96-201-03): Failure to Evaluate Change per 10 CFR 50.59
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21. VIO 01292 (IR 96-201-01): Failure to Update UFSAR per 10 CFR 50.71(e)
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2. VIO 01302 (IR 96-201-01): Failure to Update UFSAR per 10 CFR 50.71(e)
23. VIO 01312 (IR 96-201-01): Failure to Update UFSAR per 10 CFR 50.71(e)
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24. VIO 01322 (IR 96-201-01): Failure to Updata UFSAR p:r 10 CFR 50.71(e)
25. VIO 01332 (IR 96-201-01): Failure to Update UFSAR per 10 CFR 50.71(e)
26. VIO 01342 (IR 96-201-01): Failure to Update UFSAR per 10 CFR 50.71(e)
27. VIO 14014 (IR 96-11-03): SF Building Ventilation System
28. URI 96-201-13: Emergency Diesel Generator Full Load Testing
II.
Violations Associated with inadequate Corrective Actions
Some enforcement issues concerned inadequate corrective actions that applied to systems
not required in the defueled mode. Licensee actions to correct specific material
deficiencies were reviewed concurrent with the activity, as described in Inspection Report
Nos. 96 80,96-10 and 96-11. Following the certification to the NRC on December 5,
1996 of the decision to permanently shutdown the plant and cease operations, the
corrective actions for these violations were no longer appropriate. The items listed below
are closed.
1. VIO 01452 (IR 96-201-14): Evaluation of 8attery Charger Spiking
2. VIO 03112 (IR 96-80): Corrective Actions for Degraded Shutdown Conditions
3. VIO 03102 (IR 96-80): Corrective Actions for Degraded Shutdown Conditions
4. VIO 03092 (IR 96-80): Corrective Actions for Degraded Shutdown Conditions
5. VIO 03122 (IR 96-80): Ineffective Actions for RCS Inventory Diversions
6. VIO 11014 (96-11-05): Inadequate corrective Actions for Instrument Calibrations
7. VIO 08014 (96-80): Untimely Reportability for Nitrogen Intrusion Event
111.
Violations Associated with inadequate Procedures
The NRC reviewed licensee actions to address procedure weaknesses concurrent with the
activity, such as for the nitrogen intrusion event (Inspection Report No. 96-01) and the
final core off-load (Inspection Report No. 96-11). Following the certification on December
5,1996 of the decision to permanently cease operations, the corrective actions for these
violations were no longer appropriate. The items listed below are closed.
1. VIO 03082 (IR 96-80): Adequate Procedures Not Provided Per Criterion V
2. VIO 03072 (IR 96-80): Written Procedures Not Established / Implemented
3. VIO 03062 (IR 96-80): Written Procedures Not Established / implemented
4. VIO 03052 (IR 96-80): Written Procedures Not Established / Implemented
5. VIO 03042 (IR 96-80): Written Procedures Not Established / Implemented
6. VIO 03032 (IR 96-80): Written Procedures Not Established / Implemented
7. VIO 03022 (IR 96-80): Written Procedures Not Established / Implemented
8. VIO 03012 (IR 96-80): Written Procedures Not Established / Implemented
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