IR 05000213/1997006
| ML20210P025 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 08/12/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20210P006 | List: |
| References | |
| 50-213-97-06, 50-213-97-6, NUDOCS 9708260242 | |
| Download: ML20210P025 (50) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
50 213; License No:
DPR 61 Report No:
50 213/97 06 Licensee:
Connecticut Yankee Atomic Power Company l
P. O. Box 270 Hartford, CT 06141-0270 Facility:
Haddam Neck Station Location:
Haddam, Connecticut
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Dates:
June 16 - July 18,1997 Inspectors:
Ronald L. Nimitz, CHP, Senior Radiation Specialist William J. Raymond, Senior Resident inspector Approved by:
John R. White, Chief, Radiation Safety Branch Division of Reactor Safety -
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9708260242 970812 PDR ADOCK 05000213
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EXECUTIVE SUMMARY Haddam Neck Station NRC Inspection Report No. 50 213/97 06 This inspsetion was an inspection of the radiological controls program at the Connecticut-Yankee Atomic Power Station, Haddam, Connecticut. The areas reviewed during the inspection included planning and preparation for shipment of non Irradiated fuel, applied radiological controls including radiological controls of diving activities, contamination controls, and r6dierogical controls program enhancements.
-The inspection revealed that the licenses continues to experience radiation protection procedure violations attributable to personnel non adherence to procedures and ineffective superviser review of work activities. The licensee did not' effectively plan and prepare for shipment of non Irradiated fueli;The licenses was developing Phase 1 of the radiation
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protection improvement plan. At the time of the inspection, it was not clear that progrrmmatic controls were adequately established to ensure that all hazmat employces wert antified and properly trained and tested as required by 49 CFR 172 Subpart H.
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I REPORT DETAILS PLANT SUPPORT l
R1 Radiological Protection and Chemistry (RP&C) Controls fLL1 External and Internal Exoosure Controls a.
Inspec.tlon Scooe (83750)
The insp3ctor toured the radiological controls area and reviewed radiological controli including posting, barricading, and access control, as appropriate, to radiation and high radiation areas and contaminated areas. The inspector also reviewed previous and planned radiological work activities, b.
Observations and Findinal b.1.
Divina Activities in the future, the licensee will need to perform diving operations within the reactor cavity fuel transfer area to complete botting of a blank flange to the fuel transfer tube, in preparation for such activity, the inspector reviewed applicable diving program procedures and reviewed the implementation of radiological controls and applicable procedure requirements for diving activities conducted by the licensee on November 8,1996 (reactor cavny dive) and December 10,1996 (fuel pool dive).
The inspector interviewed the radiation protection technicians who provided radiological coverage for the diving activities and also interviewed radiation protection supervisors who provided oversight, and reviewed and approved the operation. Tho inspector reviewed the diving activities relative to guidance provided in Technical Specifications,10 CFR 20, and the following NRC information and guidance documents.
NRC Information Notice No. 82 31, " Overexposure of Diver During Work in
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Fuel Storage Pool," dated July 28,1982 NRC Information Notice No. 84-61, " Overexposure of Diver in Pressurized
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Water Reactor (PWR) Refueling Cavity," dated August 8,1984 NRC Regulatory Guide 8.38, " Control of Access to High and Very High
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Radiation Areas in Nuclear Power Plants," dated June 1993, Appendix A, Procedure for Diving Operations in High and Very High Radiation Areas The inspector's reviow of the November 8, and December 10,1996, dives in the reactor cavity and spent fuel pool indicated that no significant exposure was sustained by the divers based on review of external monitoring data. The dive work area was surveyed prior to ently. The divers were provided with survey meters for personnel monitoring and were monitored with multiple radiation dosimeters. The dive activities were completed with no unplanned exposures. As required by procedures, bloassiys were initiated for the divers. The inspector's review of calibration data for the survey meters used indicated the underwater survey meters
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were calibrated in accordance with procedure requirements prior to use. The licensee controlled the diver's movement by viewing the diver through a viewing glass and controlling his movement with tether lines.
The following negative observations were made relative to the November 8,1996, dive in the reactor cavity.
There was no documented data to demonstrate that the under water
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radiation survey instrumentation, used to perform the underwater radiation surveys, was source checked for operability prior to use. One technician involved with the surveys indicated the source checks were performed and radiation protection services personnel recalled performing source checks of the instruments. The program for source checking instruments did not require documentation of the checks.
Since the dive activity, the license initiated a program to place source check stickers on instruments. The stickers were initialed when an instrument was source checked. The inspector observed instruments in use to have such stickers and be initialed. The inspector noted the use of the stickers was informal and not required by procedure.
The inspector was unable to determine what specific realtime electronic
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teledosimetry was worn by the diver as he worked on November 8,1996.
The licensee did not have any data that identified the specific devices worn at each location. Further, it was not s 1ar (due to a lack of documentation) if the electronic teledosimetry worn by the diver on November 8,1996, was source checked prior to use. However, the licensee did provide calibration data for what was believed to have been the electronic dosimeters used during the diving activities.
The inspector noted that radiological controls procedures for diving activities
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contained specific provisions for collection and analysis of bioassay samples from divers following their dive activities. The inspector noted that the post dive diver tritium bloassay sample (urine sample) for the November 8,1996, dive was not sent for analysis. The inspector observed that as of June 20, 1997, the sample remained in an onsite refrigerator. The licensee initiated action to send out the sample and evaluate the adequacy of the sample relative to any potential plateout of radioactivity on the inside of the sample container walls.
The licenseo subsequently sent out the sample and intercompared the results and concluded there was no statistical difference between the pre, and
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post-dive in vitro bioassay sample results, in addition, the inspector noted that a pre-and post in-vitro bloassay was conducted for the diver who dove on Cecember 10,1996. The inspector noted that the post dive sample result indicated a positive tritium indication (relative to the pre-dive result). The licensee initiated a review of the matter and indicated a dose assessment would be performed.
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The following negative observations were made relative to the Decemoer 10,1996, dive in the reactor cavity.
As discussed above, there was no documented data to demonstrate that the
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under water radiation survey instrumentation, used to perform the underwater radiation surveys, were source checked prior to use. One technician indicated the source checks were performed and radiation protection services personnel recalled performing source checks of the instruments.
Also as discussed above, the inspector was unable to determine what
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realtime electronic teledosimetry was worn by the diver as he worked on December 10,1996. Further, it was not clear {due to a lack of documentation) if the teledosimetry, worn by the diver on December 10, 1996, was source checked prior to use. One technician indicatad the source checks were performed and radiation protection services personnel recalled performing source checks of the instruments.
The licensee's procedures for radiological controls for diving contained
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specific requirements for performance and documentation of radiation surveys of underwater work locations. The inspector observed that followup radiation surveys, performed December 10,1996, after an initial survey on December 9,1996, appeared to be a photocopy of the initial survey superimposed on a new survey form with new times and date, and portrayed as survey data for both verification and diver surveys. This was considered a questionable prtetice, and was not consistent with procedure requirements as discussed below. Further, it was unclear how the supervisors who reviewed the survey data fulfilled their responsibilities to independently review and evaluate data to ensure that the survays were within 20% of the original survey data as required by procedure, prior to authorizing diving.
The technician who documented the surveys indicated the surveys were documented in the aforementioned manner to allow for easo in intercomparison with original data, the actual resurvey data was similar, and consequently no additional data was recorded. The supervisor who authorized the dive indicated he received verbalindication from the technician involved in the diving that the survey results were similar.
The survey maps for December 10,1996, did not clearly indicate that the
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dive area had been comprehensively surveyed. Subsequently, the inspector discussed survey performance with the responsible technician to establish the adequacy of the survey.
The following additional negative observations were noted.
Landmarks and restrictions were not indicated on underwater survey maps to
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clearly indicate to the diver the allowable extent of travel.
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All personnel dosimetry devices used for radiological controls were not
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identified (e.g., teledosimetry).
Radiation protection procedures for diving referenced a draft NRC Regulatory
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Guide on High and Very High Radiation Area access control. The procedures did not reference the actual approved NRC Regulatory Guide (Regulatory Guide 8.38) nor was it apparent that the approved guide had been reviewed for any changes of importance for inclusion in applicable procedures and implemented as appropriate.
The following areas for enhancement was identified.
There were no apparent checklists or other administrative tools to ensure all
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applicable procedure requirements or signoffs, for diving, were implemented.
The licensee's procedures contained specific requirements for confirmatory
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surveys of the underwater work area to be performed by the diver during his work. The inspector noted that there was no specific training program, as appropriate, that included survey mothodology or limitations and characteristics of the underwater survey instrumentation to be used.
Further, the inspector noted that divers could encounter significant radiation dose rate gradients and communications may involve relay of communications from a technician through a dive tender to the diver. The licensee indicated divers were provided pre job briefings which discussed survey techniques. The inspector questioned the consistency and adequacy of such training in that no lesson plan or standardized guidance was developed for personnel providing such training.
Based on the above reviews, the following apparent violation was identified.
(VIO 50 213/97 06 01)
Technical Specification 6.11, radiation protection program, requires that procedures for personnel radiation protection be prepared consistent with the requirements of 10 CFR Part 20 and shall be adhered to for all operations involving personnel radiation exposure.
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Radiation protection procedure RPM 1.4-5, Revision 5, Bioassay Shipment, requires that bloassays be shipped as expeditiously as possible, otherwise a preservative must be added.
The inspector noted that the diver tritium bloassay sample (urine sample) for a November 8,1996, dive was not shipped as expeditiously as possible for analysis and no preservative was added. The inspector noted that as of June 20,1997, the sample remained in an onsite refrigerator.
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Radiation protection procedure RPM 2.213, Revision 2, Underwater Surveys, requires in Section 3.3 that pre-dive underwater radiation surveys be performed with two independent underwater radiation survey meters and that the results be recorded on survey forms for comparison purposes.
The inspector noted that the pre-dive underwater radiation survey results, for the two independent "diation survey meters, were not recorded for pre-dive surveys performou ! n hMmber 8 and December 9,1996.
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Radiation protection pioca.ro RPM 2.213, Revision 2, Underwater Surveys, requires in Section 3.3 that a health physics manager / designee review and compare the recorued pre dive radiation survey results made with two independent radiation survey meters to verify that the results are within plus or minus 20%.
The inspector noted that no such review and comparison of recorded pre.
dive radiation survey results was made for pre dive radiation surveys made on November 8,1996 and December 9,1996, in that the survey results for the two independent radiation survey meters were not recorded for review purposes, iv.
Radiation protection procedure RPM 2.213, Revislon 2, Underwat6r Surveys, requires in Section 3.4 that a diver perform a dive work area survey to confirm pre dive underwater radiation survey results including head level, waist level, floor level, arms length in the direction of the largest radiation source, and that results be recorded on a survey map for comparison purposes.
The inspector noted that the dive work area surveys for a December 10, 1996, dive were not recorded. Rather a photocopy of a December 9,1996, pre dive survey was signed and dated to provide the results.
The inspector considered the above matters to be the result of failure to develop and maintain irnportant documentation relative to radiological surveys and evaluations. The inspector's interviews indicated there was alack of radiatior>
protection technician knowledge of procedure requirements in the area of documentation of surveys for diving activities. As an example, one technician involved with diving activities indicated there was no requirement to document surveys made by divers to confirm pre dive surveys. As indicated above, such a requirement clearly existed. In addition there was a lack of supervisor knowledge of procedure requirements. For example, one supervisor interviewed indicated he was unaware of requirements to intercompare pre-dive and diver work area surveys, b2 Entrv Into the PAB Pine Trench The inspector reviewed a radiation protection technician's entry into the Primary Auxiliary Building pipe trench and the radiation, contamination, and airborno radioactivity surveys made therein. The review was with respect to criteria contained within 10 CFR 20 and applicable licensee procedure *
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The entry was made on April 18,1997, under Radiation Work Permit (RWP) No.
97 09, Revision 1, PAB Pipe Trench Health Physics Surveys (Job Step 1). In preparation for performance of a valve lineup surveillance activity within the PAB pipe trench, a radiation protection technician entered the area to perform radiological surveys. The licensee used previous radiological surveys as a basis for selection of radiological controls for the technician making the entry.
The inspector's review indicated the licensee implemented the compensatory measures for entry into high risk areas (pipa trench) outlined in the licensee's responsa to the Confirmatory Action Letter 1 97 007, dated March 4,1997.
However, the following negative observations were made relative to the personnel entry.
The airborne radioni 4-samples collected in the trench were collected
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from 1:00 p.m. to 4:t o.m. However, the sample data sheets indicated the sample was analyzed e about 3:00 p.m. This discrepancy was not identified during a review of sample results by the responsible supervisor.
Subsequent licensee review indicated the clock time on the sample counting equipment was not correct.
The licensee relied on two general area air samples to provide an indication
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of airborne radioactivity encountered. However, the inspector noted that a technician entered areas with elevated contamination and it was not apparent that the general area air sample was representative of airborne radioactivity concentrations within those areas. The licensee initiated a review of the adequacy of sampling.
The following apparent violation was identified. (VIO 50 213/97 06 01).
Technical Specification 6.11, radiation protection program, requires that procedures for personnel radiation protection be prepared consistent with the requirements of 10 CFR Part 20 and be adhered to for all operations involving personnel radiation exposure, i.
Radiation protection procedure RPM 2.31, Revision 9, Posting of Radiological Controlled Areas, requires in Section 2.8, that any area in which airborne radioactivity concentrations are greater than or equal to 30% of one DAC shall be posted as Airborne Radioactivity Area.
The inspector noted that on April 15,1997, two altborne radioactivity samples collected in the PAB pipe trench at 1:00 p.m. Indicated airborne radioactivity concentrations greater than 30% of one DAC and the area was not posted as Airborne Radioactivity Area, ii.
Radiation protection procedure RPM 2.2 7, Revision 10, Air Sample Counting, requires in Section 2.12 that air samples with an alpha activity greater than or equal to 1.0 E-12 uCi/mi be recounted af ter 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> _
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The inspector noted that an air sample collected in the PAB pipe trench on April 15,1997, between 1:00 p.m. and 4:00 p.m. Indicated an alpha activity of 2.4 E 12 uCi/ml and was not recounted after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The licensee became aware of the above matters when a technician informed a supervisor of the concerns.
b.3 Hlah Radiation Area Controls The inspector observed that locked High Radiation Area gates on the reactor containment charging floor appeared to be easily opened by reaching through a fence with a tool. The licensee attempted to open the gates and noted the gates could not be opened, c.
Conclusion A violation of radiation protection procedures associated with documentation of survey results and initiation of actions as a result of review of survey results was identified. It was not evident that the technician and supervisor, associated with this work activity, were fully knowledgeable of the pertinent procedural requirements. The above observations, in conjunction with findings associated with diving activities discussed in Section R1.1.b.1, above, and the findings associated with the November 2,1996, fuel transfer canal event indicate that the licensee has not yet been successful in improving performance relative to procedural adherence.
(Note: As a result of these concerns, a management meeting was held at NRC Region I to discuss the licensee's actions on these matters. A discussion _of the _
meeting is contained in Section X1 of this report.)-
R3 Radiation Protection and Chemistry Procedures and Documentation R3.1 Radioactive Material Shionina and Preparation and Plannina for Shioment of Unirradiated New Fuel a.
Scope The inspector reviewed the licensee's planning and preparation for shipment of non-irradiated fuel from the facility. The inspector reviewed development of procedures, training of personnel, conformance with applicable NRC and DOT shipping requirements including NRC Certificate of Compliance, and conformance with applicable station procedures and License Amendment #188. As part of the review, the inspector reviewed fuel receipt activities conducted in July and August of 1996 and fuel storage activities including implementation of applicable radiological controls.- The inspector reviewed establishment of applicable radiation work permits, calibration of instrumentation, and use of trained and qualified personnel.
The inspection included a review of the process as described in procedure SNM 1,4 20, observations of the licensee's implementation of procedure SNM 1.4 20, and a review of the licensee activities for conformance with the requirements of 10 CFR 71.12.
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Observations and Findinas New fuel was received on site in July 1996 in preparation for use in core cycle 20.
The fifty two zircaloy clad fuel bundles were stored in the new fuel storage vault and never became externally contaminated. The inspector verified that the licensee's storage practices met the requirements in License Arnendment #188 for the criticality control of new fuelin dry storage. The fuel bundles in the new fuel storage vaults were topped with cloth dust covers, and did not contain plastic sleeves or other wrappings that could retain water from a postulated flood.
Licensee engineering personnel were familiar with the criticality controls and stated that, when plastic wrappings had been used, the bottoms were open to avoid the accumulation of water.
The licensee implemented plans to send all the new fuel bundles and 3 rod control cluster assemblies to Siemens Corporation facilities in Seattle, Washington. The licensee planned to send the fuelin 6 shipments using the Advanced Nuclear Fuels Corporation Model 51032 1 new fuel shipping canisters supplied by Siemens, which had received a specific NRC license on the containers. The NRC issued Certificate of Compliance No 6851, Revision 23, dated March 25,1996. The licensee proposed to use the canisters under the conditions of a generallicense per 10 CFR 71.12, and submitted a letter to the NRC dated May 28,1997 to register as a user of the canisters. Siemens provided a vendor representative to assist licensee personnel in the preparation of the shipment including providing Des;gn Specification EMF S35152 dated May 13,1997 to assist the licensee in developing a procedure to use the canisters in accordance with the conditions of the certificate of compliance.
The licensee developed procedure SNM 1.4 20, "New Fuel Assembly and RCCA Packaging and Shipping," Revision 10, to describe the process used at Haddam
Neck load the fuelinto the canisters and to prepare the fuel for shipment. The new procedure was reviewed and approved for use on June 10,1997. The licensee used training videos to familiarize plant personnel on the use of the canisters. The video was also used by the plant operations review committee (PORC) as part of its review of the process.
The licensee assigned engineering, operations, maintenance, health physics, and security personnel to prepare the fuel for shipment. The licensee used security procedures to inspect the canisters prior to allowing entry into the protected area.
The measures were reviewed and revised based on a plant worker's suggestion that the security checks be conducted outside the spent fuel building (reference Adverse Condition Report ACR 97 290). The inspector reviewed the old and revised security measures and identified no inadequacies with regulatory requirements.
The licensee began loading the first two new fuel bundles into a canister on June 16. The inspector reviewed licensee actions to complete cask loading operations per SNM 1.4 20, and to meet the requirements for the control of heavy loads as c.. scribed in WCM 2.2 8. Plant workers followed the procedures as written. The licensee assembled the canister in accordance with the vendor's
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directions on the number of separators, full clamps and non contacting clamps, but found that the design specification and procedure were inadequate because the full number of clamps could not be installed due to interferences with the separator blocks (ACR 97 305). The licensee suspended fuelloading activities on June 17, 1997, pending further review with the vendor. The problem was resolved when the vendor revised the design specification (Revision 1 dated June 18) to increase the tolerance for separator block positioning from 0.25 inches to 0.5 inches.
The inspector reviewed Certificate 6581 to verify the licensee had met each line item. The inspector also reviewed the licensee's process as described in the associated licensee application accompanying the certificate. This reviewed confirmed the licensee met the requirements that assured mechanical stability of the loaded canisters and nuclear criticality safety. During reviews on June 18, the inspector noted that the licensee had the certificate of compliance and planned to review it prior to shipment of the loaded canisters. However, it was not apparent that the licensee had considered the certificate and associated drawings during the development of SNM 1.4 20.
The licensee held a meeting on June 19 involving management and shipping personnel to identify the lessons learned from the experiences to date. This meeting resulted in the identification of 13 items to be resolved, including issues with the spacer / shim configuration, the general conformance with the certificate of compliance, foreign material controls for clear plastic, the adequacy of tre3ning in the requirements for the handling of hazardous materials, and severalissues on industrial safety (safety shoes, ladders, fork lift use).
The licensee initiated a review of SNM 1.4 20 and made changes to the procedure based on a review of the certificate. The procedures was rewritten and reissued in Revision 11 as "New Fuel Assembly and HCCA Packaging and Shipping Using the Model 510321 Shipping Container (Certificate of Compliance USA /6581/AF)".
One particular question identified by the PORC during a meeting on June 20 was whether the use of wood shims to make up the distance from the top of the fuel bundle to the upper thrust plate was in accordance with Certificate 6581. This question was referred to Siemens, who responded to the licensee in a letter dated June 23 which stated that the use of wood complied with the certificate of compliance. The inspector noted the use of the wood shims was not specified in the certificate or in Section 3.0 in the associated application on file with the NRC.
Upon review, the inspector determined that the use of wood shims inside the canister was not described in either the certificate or the supporting drawings and documentation. The inspector informed the licensee on June 26 that use of the wood was contrary to the use of the canister approved by the NRC (ACR 97 359).
The licensee had to either change the loading process to make the new fuel installation in the canisters to agree with the certificate, or revise the certificate to recognize the use of wood shims. The licensee stated that this issue would be resolved with NRC, Office of Nuclear Materials Safety (NMSS) prior to shipping the new fuel.
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The NRC issued Revision 24 of Certificate 6581 on July 2,1997. The record of review for Revision 24 stated that the use of wood inside the canisters was acceptable because it was bounded by the criticality and fire hazards analysis. The licensee stated that a review will be completed per 10 CFR Part 21 to determine whether a significant safety hazard existed relative to past use of the Siemens cask.
The licensee issued Revision 12 to procedure SNM 1.4 20 on July 9, following additional reviews, to assure it fully addressed the item in the Certificate of Compliance. Loading of new fuelinto the Siemens canisters resumed on July 10, and the first shipment of 8 new fuel bundles loaded into four canisters was made on July 14,1997. The inspector reviewed the licensee's final preparations for the shipment on July 14, including: the completion the shipping manifest, the radiation limits, marking and placarding to assure compliance with 49 CFR 172, and the completion of OC surveillances. No inadequacies were identified.
The following positive observation was made relative to the August 1996 fuel receipt activities.
The inspector's review indicated applicable radiological controls program
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procedures were implemented for receipt of the fuel, transfer of the fuel to the spent fuel storage building, opening of the fuel canisters, and transfer of the fuel to dry storage. Required radiological control documentation was available for each fuel element received. Selected review of instrument calibration records indicated instruments used for radiological surveys were calibrated as required.
The following positive observations were made relative to the June 1997 fuel shipment activities.
A program was established and implemented to train personnel on the
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requirements of 49 CFR 172 Subpart H for the planned upcoming fuel shipments. Subsequently, personnel had been trained and tested, relative to
49 CFR 172 Subpart H requirements. There appeared to be an adequate number of properly trained personnel to support this activity.
A new fuel shipment industrial safety job safety analysis was performed.
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Control measures, relative to industrial safety practices for the planned shipments were implemented.
A surveillances program for shipping activities was implemented. While
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some administrative errors were noted, effective action was taken to correct issues.
The licensee obtained a contractor engineer to support radwaste activities.
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in addition, the inspector noted that the licensee was in the process of obtaining standard industry software for performing shipping calculations and shipping paper generation, further, shipping procedures were being revised to provided enhanced detail and useability.
The following negative observations in the area of fuel receipt, and handling were noted.
Training documentation could not be located for one radiological controls
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technician who performed surveys of new fuel elements received in August 1996 to indicate that the individual had been properly trained and qualified as required by 49 CFR 172, Subpart H.
No clearly defined program could be identified, as required by 49 CFR 172
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Subpart H, for personnel (non radiological controls) who supported new fuel receipt in 1996. It was not apparent that the individuals who supervised and performed the fuel receipt and inspection activities (i.e., maintenance personnel, reactor engineering personnel) had received appropriate training.
(Inspector Note: The above negative observations were considered an unresolved item (UNR 50 213/97 06 02). An ACR was written by the licenseo pending licensee effort to provide evidence of appropriately established and implemented training and testing programs as required to 10 CFR 49.172, Subpart H for all hazardous materials handlers.)
Personnel managing the receipt of new fuelin 1996 did not appear to
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understand the requirements of 49 CFR 172, Subpart H relative to the need to utilize properly trained personnel, for fuel receipt activities.
The following negative observations were made relative to the upcoming planned shipment of new fuel elements, by the licensee, to its vendor.
There was no comprehensive evaluation of licensee conformance to the NRC
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Certificate of Compliance (C of C) for the fuel shipping containers as well as asstrances that detailed procedures had been developed for fuel shipment, consistent with the C of C. NRC review observed that fuel could not be secured in the shipping container using existing procedures. There was no verificatiori, relative to 10 CFR 71,12, that all applicable documents and drawings were present and they were implemented. The inspector observed that the licensee was planning to use wood pieces (shims) for support of the fuel. It was not apparent that the wood was authorized by the C of C.
(Inspector note: As discussed above, the licensee suspended fuel shipping activities and initiated actions to perform a comprehensive evaluation of the C of C and the establishment and implementation of applicable icel container loading, closure and shipping procedures. The licensee also obtained a revised C of C ( C of C No. 6581, Revision 24, July 2,1997) to allow use of wood shims for the fuel shipping container.)
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The coordinators for the current fuel shipment campaign did not know if all
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applicable personnel had been tested relative to Subpart H requirements.
The inspector observed that some individuals did not appear to have been tested based on training records provided by a fuel shipping coordinator.
The licensee subsequently provided additional documentation indicating that all personnel reviewed had received training and testing.
Radiological Controls supervision were unaware of 49 CFR 172 Subpart i or
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the status of its implementation (as appropriate).
It was not apparent that appropriate controls were established to ensure
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performance of contamination surveys as outlined in 49 CFR 173.443 (i.e.,
survey to 300 cm2). No apparent concern was noted reletive to conformance with applicable limits based on discussions with cognizant personnel.
However, the licensee subsequently reviewed this matter and clarified procedurds, in addition, the following programmatic concerns were identified.
Current radioactive material shipping procedures lacked oetail and required
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substantial knowledge of applicable regulations to properly implement. Only one individual was currently qualified to sign shipping papers and authorize a shipment.
The inspector noted that although the licensee established appropriate
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training and testing for individualt, involved in handling, packaging and shipment of the unirradiated fuel to the vendor, and there appeared to be an adequate number of personnel trained to support this activity, programmatic controls appeared to need to be established to ensure the hazardous material training and testing requirements of 49 CFR 172 Subpart H (i.e.,
familiarization training, emergency training, function specific training) were implemented, as appropriate, for all employees involved in hazmat activities.
The inspector noted that during decommissioning activities, extensive shipping of radioactive material would occur. Consequently, an appropriately detailed shipping program would be required to support decommist,loning activities, including assurance that all personnel involved in hazmat activities were properly trained.
The licensee initiated a review of the above matters.
In addition to the above, the inspector observed that personnel were performing dry runs of fuelloading activities in radiation fields emanating from contaminated fuel handling tools left in close proximity to fuel containers. Although endiation dose rates were low, the observation indicated an apparent lack of sensitivity to unnecessary personnel radiation exposur.
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Further, the inspector noted that the licensee identified and documented, in an Adverse Condition Report (ACR No. 97 00-42, dated January 21,1997), that a 5 gallon pail of contaminated fuelinspection lights, shipped to another nuclear power station was left unattended on January 17,1997, for a short period of time at the other station's protected area access point. Although the pail contained a relatively small quantity of radior.ctive material (Limited Guantity), the ACR indicated that the individual who delivered the pall, did not wait with the shipment to physically turnover the licensed material to the authorized representative. The individual did, however, contact a waste services representative at the other station and inform that individual that the pail had been delivered and subsequently left the licensed material under the temporary surveillance of a clerical worker. The inspector's review determined that, in this circumstance, sufficient control was maintained to technically meet the requirements of 10 CFR 20.1802.
The licensee completed a causal factor and corrective action plan relative to this matter and concluded that the individual did not understand that the package was to be physically received and not lef t unattended. Also, the licensee's procedure for shipment of non waste radioactive material packages did not include requirements for receipt of packages. The licensee subsequently initiated a revision to the applicable procedure (RPM 3.6 3) to include guidance in this area. In addition, the licensee suspended use of building services personnel to transport radioactive material. The licensee has elected to use commercial carriers to transport such material, it was not apparent that this individual had received proper training. This matter will be reviewed relative to Unresolved item No. 50 213/97 06 02 discussed above, c.
Conclusions Varied performance was noted in the development of the process to use the Siemens cask for the first time at Haddam Neck. Good performance was noted to implement the License Amendment #188 requirements for criticality control of new fuel while in dry storage; there was good coordination between engineering and other plant groups to support the shipment; there was good interface with the vendor and security to resolve emerging issues (separator block / clamp interference and location to conduct security checks); and, the process complied with the Siemens design specification and the fuel nuclear criticality restrictions stipulated in the certificete of compliance. Once initial problems with the loading process were identified, there was good management involvement in the staff initiatives to develop lessons learned and resolve problems following the voluntary work stoppage on June 17.
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Despite CY staff efforts to prepare the fuel for shipment in accordance with the vendor's instructions, problems were identified by the NRC in the failure to assure full complianco with Certificate 0851. Further CY review of the certificate could have been completed earlier in the process and for the development of procedure SNM 1,4 20. Support from the vendor was not adequate to assure that the use of wood shims explicitly conformed with either the certificate or the supporting documentation and drawings associated with the application. No violations were identified rotativo to use of the Siemens canisters at Haddam Neck due to the issuance of Revision 24 of the Certificate prior to shipment.
An unresolved item relative to establishment of a training program as required by 49 CFR 172, Subpart H, requirements was identified.
R3.2 Contamination Controls a.
Insoection Scone The inspector reviewed selected aspects of the licensee's contamination control program. The inspector toured the radiological controlled area and observed contamination control practices. 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 10,1997, was identified by the vendor to exhibit radioactive contamination. This matter had been previously reviewed during NRC inspection No. 50 213/97 01, dated May 8,1997.
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As a result of this event and other concerns involving radiological controls, the NRC issued a Confirmatory Action Letter (CAL No. 1 97-007) on March 4,1997, specifying actions to be taken to effect overall pro 9 ram improvament. (See Section R8.1 of this inspection report.)
b.
Observations and Findinas b.1 Releate of Contaminated Material from Station (URI 97 01-11]
As discussed in NRC Inspection Report No. 50 213/97 01, interim actions were taken (February 27,1997) to restrict the release of material from the radiological controlled and protected areas pending evaluation of the event and development and implementation of long term corrective actions.
During this inspecuon, the inspector determined that in March 1997, the licensee performed a review of the radiation protection instrumentation program with a focus on instrumentation used for monitoring of material to be released to the unrestricted area. The calibration and source check practices of other survey and monitoring equipment used to perform in plant surveys was also evaluated. The review identified the following discrepancies:
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use of calibration sources whose radiation energies were not consistent with
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those encountered in the station; lack of validation of instrument operability by the source check program and
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a need for improved training of personnel on the program; lack of procedure guidance for source checking instrumentation including
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lack of analytical guidance for interpretation of instrument response.
The licensee initiated severalimmediate and long term corrective actions to effect program improvement. These included the following.
A comprehensive evaluation of the stations radioactive material and radiation
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energies was conducted.
Sources were selected to calibrate instruments for optimum response to the
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radionuclide mix present at the station.
The training program for personnel use of contamination monitors was
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enhanced and implemented.
Quantitative instrument source check criteria was developed for small article
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monitors and handheld friskers. The methodology was implemented.
Source check Jigs were purchased to provide consistent instrument source
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check geometries and quantitativo dose rates.
Use of micro R meters was implemented for performing surveys of aggregate
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checks of bulk material.
A review and revision of the calibration and source check program was
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initiated, in addition to the above, and in accordance with commitments outlined in the licensee's action plan to determine the extent of radioactive material that may have been inappropriately released form the site (contained in the licensee's March 20, 1997 letter CY 97 017, Supplemental Response to Confirmatory Action i.etter), the licensee continued to survey materials arid equipment located at the site both inside and outside the protected area. As of June 19,1997, the licensee had surveyed 3950 items outside the protected area within various warehouses and identified 28 items with confirmed low level non removable contamination. These items were returned to the RCA. The licensee also surveyed 11,057 items inside the protected area and identified 56 items with low level (but detectable) fixed activity.
The following negative observation was made:
During a previous inspection (Reference NRC Inspection No. 50 213/97-01)
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the inspector observed that the licensee had designated outdoor yard areas as portions of the radiological controlled area (RCA) and that personnel routinely egressed main plant buildings (e.g., auxiliary building) to access the outdoor areas. The licensee did not require personnel to frisk for contamination prior to exiting the plant buildings. This was considered a program weakness in that personnel may inadvertently track contamination outdoors from the main plant building..
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During th$ current inspection the inspector entered and toured potions of the reactor containment. The inspector observed that the access ways to the containment were not well demarcated for contaminatio. control purposes.
The inspector noted that although the containment walk areas were not considered contaminated areas, the potential for inadvertent contamination was present (e.g., overhet.d contamination) and containment access / egress points did not have a clearly demarcated boundary for personnel frisking and contamination control purposes. The licensee initiated a review of this matter.
b.2.
Chemistrv Lab Contamination Event The inspector reviewed the licensee's followup of the contamination of the chemistry lab on June 2,1997, due to inadvertent breakage of a source vial. The vial (glass ampule) inadvertently broke at a different location on the vial during its opening rather than the expected location. The inspector noted use of such vials and the practice of " snapping off" of the glass top to be a common industry practice. Personnel sustained Cs 137 shoe contamination and a chemist sustained contamination of the hand.
The inspector's review indicated that the licensee performed an appropriate followup of this matter including decontamination of the area, initiation of an adverse condition report, and discussion with the source vendor. The licensee also performed an assessment of the shallow-dose equivalent to the extremity of the chemist who sustained contamination of the hand and performed a calculation of expected committed effectivo dose equivalent. The calculations indicated no significant exposure. The inspector selectively reviewed the calculations and had no questions, c.
Conclusions The licensee performed a comprehensive review of its contamination controls for egress of material from the radiological controlled area and took a number of actions to enhance the program Contamination control practices appear weak relative to control of inadvertent tracking of contamination to outdoor areas from process building within the radiological controlled area.
The unresolved item associated with the effectiveness of contamination controls program remains open (URI 97 01 11)
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R8 Miscel:aneous Matters R8.1 Confirmatorv Action 1gugt a.
kpng The NRC issuso * Confirmatory Action Letter (CAL 197 007) dated March 4,1997, to confirm the licensee's actions and commitments to identify and effectively resolve weaknesses and deficioneles in tha implementation of its radiological controls prograrn. These included the November 2,1996, reactor cavity airborne radioactivity event (Reference NRC Inspection Report No. 50 213/9612), the programmatic deficiencies associated with calibration of effluent monitoring systems (Reference NRC Inspection Peport No. 50 213/97 02), and the problems associated with release of contaminated material to an unrestricted area.
The inspector reviewed the licensee's implementation of the CAL and associated corrective actions and program improvement initiatives, b.
Findinas and Observations The following observations were made relative to implernentation of CAL No.
1 97 007 dated March 4,1997.
CAL ltem 1.
By March 7,1997, identify, in writing, specific compensatory measures that will be put in place to assure sufficient management, control, and oversight of ongoing or planned activities that require radiological controls.
NRC Findinas (Item 11 The licensee provided a letter (CY 97 011) dated March 7,1997, that identified, in writing, the compensatory measures put in place, that, according to the licensee would assure sufficient management, control, and oversight of ongoing or planned activities that required radiological controls.
The inspector discussed implementation of the compensatory measures and verified that the licensee accomplished this item.
CAL ltem 2.
By March 31,1997, engage the services of an independent assessor to assess the quality and performance of the radiological control program and its implementation.
NRC Findinas (item 2)
The licensee provided a letter (CY 97 011) dated March 7,1997, that identified, that the services of an independent assessor would be engaged by March 31, 1997.
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The licenseo provided a March 27,1997, supplemental response (CY 97 033) to the CAL which Indicated that the independent assessor was engaged to assess the quality and performance of the radiolagical control program and its implementation at the Haddar.# Neck Plant. This item has been accomplished.
CAL ltem 3.
By May 30,1997, based on input from the independent assessment, (1) Identify problems, determine root causes, and develop broad based and specific corrective actions; (2) identify performance measures that may be used to determine the effectiveness of radiological control programs, and (3) submit a plan and schedule to the Regional Administrator, NRC Region I, for the implementation of improvements in the radiological control programs.
NRC Findinns (Item 3)
The licenseo submitted a letter (CY 97-011) dated March 7,1997, that identified, that the information identified in CAL ltem 3 would be submitted by May 30,1997, The licensee subsequently submitted a letter (CY 97 049) dated May 30,1997, which provide a radiation protection Improvement Plan. The letter provided a description of root causes and a plan to develop corrective actions including improvement initiatives and effectiveness measurements. The plan was established to provide overall program improvement and effectiveness reviews using a three phase approach. Phase Iincluded identification of deficiencies which have a potential for affecting worker or public safety and or compliance with regulations within the near torn and are scheduled to be identified and corrected (with appropriate training of personnel by July 31,1997. The Phase ll program was designed to enhance the radiological controls program to meet or exceed standard industry radiation protection good practices. Items associated with Phase il are scheduled to be completed by September 30,1997. Phase 111 of the program is expected to provide further revaluation of effectiveness reviews and i nplementation of additional actions, as appropriate to support reactor coolant decontamination.
The latter phase is expected to be completed by December 31,1997.
The inspector met with the licensee and contractor personnel developing the plan and discussed the Phase I program development and current status. The inspector noted a consolidation of all program area review findings in the radiological controls area (e.g., audit findings, independent assessment findings) for purposes of corrective action matrix development. The licensee appeared to be developing a comprehensive matrix of allitems for review and resolution as appropriate.
The licensee cubsequently submitted a letter (CY 97-067) dated June 13,1997, which provided the contractor's complete Independent Assessment of the radiological controls progra !
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The inspector indicated the following areas did not appear to be clearly identified within the matrix for Phase I review, as appropriate.
Evaluation of the adequacy and effectiveness of radiological contamination
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control practices within the radiological controlled area. Such an evaluation should examine the appropriateness of egress of personnel from process building to outdoor areas without performing personnel contamination monitoring.
Evaluation of the adequacy and effectiveness of radiography practices.
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The licensee indicated these matters would be reviewed for inclusion as Phase i review items, as appropriate.
This item has been accomplished.
_ CAL ltem 4.
Prior to eliminating any interim compensatory measures (as committed in response to item 1, above), the licensee is to meet with the Regional Administrator, NRC Region I for the purpose of describing program implemen'.ation and performance improvements achieved or planned.
NRC Findinas (Item 41 The licensee submitted a letter (CY 97 011) dated March 7,1997, that identified, that the ' licensee will comply with the commitment. While actively engaged in program improvement efforts, the licensee has not yet indicated that they are prepared for such meeting, c.
Conclusion The licenseo had effectively completed all scheduled commitments specified by the Confirmatory Action Letter. Notwithstanding a meeting with the Regional Administrator remains to be scheduled when the licensee is prepared.
R8.2 Use of Electronic Dosimetrv a.
Scope During a previous inspection, (Reference NRC Inspection No. 50 213/9612, dated December 19,1996), the inspector noted that an electronic dosimetry (ED) device, based on information provided by its wearer, did not alarm upon encountering radiation dose rates in excess of its dose rate alarm setpoint. The device's integrated dose alarm did however alarm properly when its integrated dose rate alarm was exceeded. The inspector reviewed the adequacy of the device and attempted to ascertain the reason for its not alarming as anticipated.
The inspector reviewed the performance characteristics of the devico including calibration records and discussed the use of the 'evice with licensee personne _
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b.
Findinas and Observations The licensee was able to identify the electronic dosimetry worn by the individuals who entered the fuel transfer canal on November 2,1996. The licensee subjected the dosimeters to a series of tests including irradiation of the dosimeters to known radiation dose rates to validate the operability of the devices dose rate alarm
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feature. The devices had not been serviced or repaired since November 2,1996.
The licensee concluded that the uvices were working properly, in addition, the licensee selected 10 other similar dosimeters, tested their operability, and concluded that those devices also were working properly, The inspector reviewed the relevant calibration data for the devices in effect prior to their use on November 2,1996.
The data indicated the devices were calibrated. The licensee concluded that the devices were operational at the time of the entry based on the following facts.
The devict.s were calibrated prior to their use.
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Periodic checks of the devices were completed.
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The device reader performed a self check of the device as one logs into the
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RCA, No failures had been detected during performance of additional testing of
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10 dosimeters. Each dosimeter responded properly to all alarms (both audibly and visibly).
Doserate alarms exhibited an elevated sound (decibel) level.
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Doserate alarms wcu!d cease alarm tone when removed from the source.
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The alarm circuitry was operational in that one of the devices was noted to
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be in alarm upon exit of the worker from the reactor cavity since it had exceeded its integrate-I dose alarm set point.
The licensee contacted the manufacture of the devices and discussed the
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performance of the device and its predecessors. The manufacture indicated no problems were noted with the alarm feature.
The licensee noted the manufacture's alarm response time of the device and
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concluded that the device exhibited an appropriate response time to alarm.
The licenseo could not explain the workers' statements that no alarm was heard wl hin the fuel transfer canal. The license indicated that, based on practices during the event, workers would back out of an elevated dose rate area when the dose rate alarm would alarm which would cause the alarm to reset. The inspector noted however, that licensee procedures specifically required that personnel exit the area upon any alarm The inspector noted that the worker's general employee training program provided guidance to workers to exit the area on a continuous alarm (integrated dose) and back out on an intermittent alarm (dose rate alarm), if the latter dose alarm does not reset, workers were to exit the area. The licensee initiated a procedure change to change the procedure to reflect current practices.
The licensee indicated that a recent review of the training program had been performed to determine if the program was consistent with expectations, the program was consistent with expectations but this issue was not noted.
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The inspector made the following observation.
The licensee did not specifically test the dose rate alarm feature of the
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dosimeter during its calibration and checks. Rather, the integrated doss alarm was tested. The inspector noted that the integrated dose alarm was an option to be used (based on requirements within the licensee's Technical Specifications (TS)) for High Radiation Area entries. There was no specific requirement within the TS for use of a dose rate alarm feature. The licensee indicated this matter would be reviewed.
The licensee indicated that technicians were instructed to perform a check of the doserate and doso alarm functions during pre calibration of dosimeters and that the current calibration procedure was to be evaluated, c.
Conclusion The electronic dosimeters worn by personnel who entered the fuel transfer canal on November 2,1996, appeared to be working properly, R8.3 NRC Information Notice 96-47 a.
Sup2n NRC Information Notice 96-47, Recordkeeping, Decommissioning Notifications for Disposals of Radioactive Waste By Land Burial Authorized Under Former 10 CFR 20.3043, 20.302, and Current 20.2002, dated August 19,1996, provides guidance to licensees relative to recordkeeping requirements for previous onsite disposal of radioactive material, b.
Findinas and Observations The licensee was aware of the Information Notice. The licensee was reviewing its records and contacting current and previous employees to determine if any radioactive material had been previously buried at the station, c.
Conclusion The licensee was reviewing the history of the station to determine if any radioactive material had been buried at the station.
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R8.4 40 CFR 190 Dose Limits (IFl 50 213/97-01 10)
a.
kqgan During the referenued inspection, the inspector discussed conformance with the Environmental Protection Agency (EPA) regulation 40 CFR 190 with the licensee.
The review was prompted, in part by elevated radiation dose rates in the back yard areas of the station and posted "ALARA walk ways." The inspectos determined that the environmental TLDs posted on and near the protected area fence were A
considered as extra environmental TLDs and were not intended to measure direct radiation from the site to demonstrate compliance with 40 CFR 190 but were only considered for on site measurements to ensure compliance with certain 10 CFR 20 requirements. The inspection was incomplete pending further discussion to determine how the licensee ensured compliance with 40 CFR 190 limits, b.
Findinas and Observations The inspector met with cognizant personnel to discuss data used to confirm conformance with the requirements of 40 CFR 190. While there is no evidence to suggest that the licensee is not controlling exposures to the public as required by 40 CFR 190, documentation and basis was waak or not available for review.
Subsequently, the licensee initiated a review o' this matter, c.
Conclusion Licensee personnel were not able to provide appropriate data to demonstrate conformance with 40 CFR 190 dose limits. The fo.lowup item remains open.
(IFl 50 213/97 01 10)
J X1 - Exit Meeting Summary The inspector presented the inspection results to members of licensee management at the conclusion of the onsite portion of the inspection on June 20,1997. The licensee acknowledged the findings presented, in addition, a telephone discussion was held with the licensee on June 27,1997, in addition, a management meeting was held at the NRC Region i Office, King of Prussia, Pennsylvania on July 18,1997. The meeting was open to the public and members of the
- public did attend. The purpose of the meeting was to allow the licensee to update the NRC on the status of various NRC areas of interest including actions to improve personnel adherence to procedures, training of personnel, and contamination monitoring._ The attachment to this report provides the licensee's presentation.
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PARTIAL LIST OF PERSONS CONTACTED Licensee g
1. Felgenbaum, Executive Vice President and Chief Nuclear Officer R. Mellor, Director Site Operations and Decommissioning G. Bouchard, Work Services Director R. Gault, Radiation Protection Supervisor J. Goergen, Assistant Health Physics Manager G. Goncarovs, Chemistry Manager M. Sweeney, Radiation Protection Services Supervisor G. van Noordennen, Licensing Manager J. Warnock, Quality Assurance Manager NBL W. Raymond, Senior Resident inspector Haddam Neck Project Manager INSPECTION PROCEDURES USED IP 71707:
Plant Operations IP 92904:
Followup - Plant Support ITEMS OPEN, CLOSED, AND DISCUSSED
.QItita 97-06 01 VIO Failure to Follow Procedures (Multiple)
PJs.stli.sai 97 01 10 IFl Compliance with 40 CFR 190 not verified 97-01 11 URI Review Release of Contaminated Material Closed None
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LIST OF ACRONYMS USED ACR Adverse Condition Report CAL Confirmatory Action Letter
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CFR Code of Federal Regulations CYAPCo Connecticut Yankee Atomic Power Company FSAR Fina) Safety Analysis Report GL Go.eric Letter HP Health Physics IFl inspection Followup item IR inspection Report NMSS Nuclear Materials Safety and Safeguards NOV Notice of Violation NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation PAB Primary Auxiliary Building QA Quality Assurance OAS Quality Assurance Survaillance QC Quality Control UFSAR Updated Final Safety Analysis Report URI Unresolved item VIO Violation WCM Work Control Manual
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Connecticut Yankee Atomic Power Company MANAGEMENT MEETING i
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Ju!y 18,1997 l
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Agenda
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Opening Remarks NRC/T. Feigenbaum
NRC Exit Findings R. Nimitz Corrective Action Program J. Haseltine Human Performance R. Mellor
Procedure Compliance R. Mellor NRC Inspection Response R. Sexton Fuel Shipping J. Haseltine h
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Agenda { cont.}
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Confirmatory Survey R. Sexton
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Site Contamination /
R. Mellor
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Characterization Confirmatory Action Letter R. Mellor
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Summary T. Feigenbaum
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Corrective Action Program New Software and Process implemented
-Reduced Time for Processing ACRs (1/4 Time)
-Much Easier to Use Computer to Determine ACR and AR Status-More Management Control (Extensions and Closures)
-Categorize Important Assignments (QA, NRC, ACR, PORC)
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Corrective Action Program <'< Continued)
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Training Complete on ACR/ATS
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- High Percentage of Site Personnel Attended
- Low Threshold and Finding Your Own Problems Emphasized
- Excellent Response 141 Of 275 Personnel Had initiated an ACR n Understanding of ACR Initiation Program (5.521 To 8.354)
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Value of ACR Program (6.436 To 6.867)
Opinion of New ACR Program (7.543)
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- Excellent Feedback Feedback to initiators (12)
n Communicate Trends and Sta'.as (12)
n Faster Process (11)
n Did Not Agree With ACR Program (5)
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Corrective Action Program (Continued}
Results
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-Large increase in ACRs initiated (116 in June With 53 Per Month for First 5 Months)
-Better Quality of ACRs
- ACR Processing is Much Easier and Faster-Receiving Comments on Software and Resolving
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I First. Quarter Baseline Report (162_ ACRs)
- Personnel Error ACRs (21)
-Event ACRs (25)
-Procedure ACRs (9)
-High Percentage of ACRs Found by Oversight (39 or 24%)
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Corrective Action Trends l
Second Quarter Trend Report (221 ACRs)
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-Personnel Error ACRs (24)
-Event ACRs (24)
-Procedure ACRs (20)
-Lower Percentage of ACRs Found by Oversight (26 or 11.7%)
Unacceptable Trend for Personnel Errors and Procedures
-Level B ACRs initiated
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Human Performance Trend Analysis Indicates Personnel Errors Continue
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Actions to improve Performance
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-High Standards and Expectations Have Been Set Site Wide Department Level
- Assimilation of New Management Team Comaleted at All Levels of Organization Core Values imparted
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Human Performance (Continued)
Actions to improve Performance (Cont'd)
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-Work Observation Program Training Complete / Procedure issued Field implementation in Progress
- Director Level Involvement in New Tasks
- ACR Initiated to Evaluate Adverse Trend
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Human Performance (Continued)
Measuring Performance
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-Work Observation Trending-Corrective Action Program Trending-Culture Survey-Leadership Survey
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Procedure Compliance
1 Trending Indicates Non-Compliance Continues
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- Adverse Condition Report initiated Immediate Actions
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-Re-Inforce Standards Through CY Today-Revise Site Procedure to More Clearly Define Management Expectations-Train All Site Personnel on Expectations and Use of Procedures l
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Procedure Compliance l
(Continued}
Assuring Knowledge of Procedures
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-Oversight Emphasis-Pre-Job Briefings-Work Observation Program-Training Programs
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Immediate Corrective Actions Following NRC HP Inspection
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l Senior Management Meetings With Technicians
HP Technician Rotation to Seabrook Scheduled
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Evaluation of Technician Staff Strengths and Weaknesses
Procedural Adherence Enhancements
-Round Table Meetings Conducted to Enhance Procedure Knowledge, Identify Improvements and Set Expectations for Compliance-Use of Procedures "In Hand" l
Mentor for Radiological Operations Supervisor
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Survey Documentation / Supervisors Review
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Detailed Expectations Developed
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Training Provided
Redundant Supervisory Review & Feedback
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Increased Self Assessment
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Status of Improvement Plan Phase I Actions Major Actions Establish Organization and Expectations
Improve Instrumentation Program
Improve RWP/ Work Process
Assess Corrective Actions From 11/2/96 and l
2/26/97 Events l
Improve Alpha Monitoring Capabilities, Airborne l
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Status of Improvement Plan
Phase I Actions (Cont.)
J 228 Actions (Identified to Date)
-101 Working
- i27 Completed of Which 44 Are Closed
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Phase i Actions Expected to Be Complete By
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July 31,1997 Training & Final Closure Review by Mid-August
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Procedure, Documentation, Calibration, Training, Job
Hazard Analysis identified and Performed Initial Housekeeping issues identified by Oversight and
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Corrected Lessons Learned Meeting Held - 13 Items identified and i
Corrected
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Certificate of Compliance Validation Requirements Not
Fully Understood.
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Outside Protected Area Sample Protocol Identified Potential Through:
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-Warehouse Personnel Input-VisualInspection of AllItems 100% Survey of items With Potential
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Results
-0.7% Contamination Rate (27 of 4000)
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Protected Area Protocol
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100% Of Tools and Equipment
Started at 10% of Scaffolding (Increased to
100%.Due to Identified Knuckles)
i Sampling of Nuts, Bolts, Washers, Etc.
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Sample 10% of Clean Area Reusable Lumber
Supply (Just Beginning Survey)
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Status l
Outside Protected Area Complete
Protected Area Approximately 50% Complete
- About 63,000 items Surveyed
- About 0.15% Contamination Rate (95 Items)
No item Greater Than 1500ccpm
- About 75% of items 100ccpm or Less h
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Conclusions Items Being Fov>." P marily Due To Lower Backgrounds
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and New Techr:
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There is A Low Probability Of Material With Significant
Levels of Contamination Having Been Released From Site.
Minimal Risk
-Generally Less Than 5000dpm/100cm-No Removable Contamination Identified
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Site Contamination Perspective FERC Estimate - Best Effort in Nov.1996
Two Events Dominate
-1979 - Rupture Disc
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-Both Events Reported and Documented-Soil Remediated For Both Events-Routine Monitoring Programs Historical Review
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Site Contamination Perspective (Continued)
Preliminary Scoping Survey Underway in Targeted Areas
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Outside Protected Area Detailed Scoping Survey for Systems and Structures
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Anticipated by November 1997 Detailed Scoping Survey for Soil Anticipated by April
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1998.
.