IR 05000133/1998003
| ML20236E777 | |
| Person / Time | |
|---|---|
| Site: | Humboldt Bay |
| Issue date: | 05/22/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20236E771 | List: |
| References | |
| 50-133-98-03, 50-133-98-3, NUDOCS 9805280043 | |
| Download: ML20236E777 (15) | |
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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.:
50-133 License No.:
DPR-7 Report No.:
50-133/98-03 Licensee:
Pacific Gas and Electric Company (PG&E)
Facility:
Humboldt Bay Power Plant Unit No. 3 Location:
1000 King Salmon Avenue Eureka, California 95503 Dates:
April 27 - 30,1998 Inspectors:
J. V. Everett, Senior Radiation Specialist L. L. Wheeler, NRR Project Manager Accompanied by:
D. Blair Spitzberg, Ph.D., Chief Nuclear Materials Safety Branch 2 Approved By:
D. Blair Spitzberg, Ph.D., Chief Nuclear Materials Safety Branch 2 Attachment:
Supplemental Information
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EXECUTIVE SUMMARY Humboldt Bay Power Plant, Unit No. 3 NRC Inspection Report 50-133/98-03 The Humboldt Bay Power Plant Unit No. 3 has been in a SAFSTOR decommissioning status since 1976. Based on observations made during the site tour, the facility was being maintained in an acceptable structural condition.
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The licensee had continued to actively monitor conditions related to Unit 3 while in SAFSTOR.
A review of the self-assessment and corrective action process at the site found numerous issues identified by the licensee indicating an aggressive approach to recognizing and identifying potential problem areas which were then brought to the attention of management fnr resolution and corrective action. The process appeared to be comprehensive and provided for tracking corrective actions to completion.
Decommissioning Performance and Status.
The licensee submitted a revision to their PSDAR on February 27,1998. On April 29,
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1998, a public meeting was held in Eureka, California, to inform the public of the NRC
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- regulations governing plant decommissioning and the licensee's plans for
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decommissioning activities, with particular attention to removal of the 250-foot stack and transfer of the spent fuel to dry cask storage (Section 1).
Observations during tours of the Unit 3 facility and radiologically controlled area indicated that the facility was being maintained in an acceptable SAFSTOR condition. Fire loading, radiological controls, structuralintegrity of the facility, and housekeeping were
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found to be adequate (Section 1).
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l Recent changes to emergency procedures were found to be acceptable and consistent with the current status of the facility (Section 1).
Organization. Mana9Cment and Cost Controls l
A recent change to the licensee's organization involved the replacement of the radiation
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protection manager. The new individual met all requirements of the technical specifications for this position (Section 2).
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' Radioactive Waste Treatment. Effluents. and Environmental Monitorina
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Offsite radiological doses due to gaseous and liquid effluents released from the
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Humboldt Bay facility in 1997 were reviewed and determined to be within regulatory limits (Section 3).
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-3-Self Assessment. Audits. and Corrective Actions The licensee had a formal self assessment and corrective action program that
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documented problems, brought the problems to the attention of management, and tracked the resolution and completion of corrective actions. Several specific events evaluated during the inspection verified the effectiveness of the program (Section 4).
Incoections of Final Survevs The licensee had completed a characterization survey and report to determine the levels
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of contamination in the soil and in the bay sediment adjacent to the Humboldt Bay facihty. Contamination levels were within expected ranges onsite with cesium (Cs)-137 being the predominant isotope. No radioactivity above natural background levels was found in the bay (Section 5).
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Report Details Summary of Plant Status The Humboldt Bay Power Plant, Unit No. 3 was placed in a SAFSTOR decommissioning status in 1976.' On February 27,1998, PG&E submitted a PSDAR update to the NRC to comply with 10 CFR 50.82(a)(7). A public meeting was held on Wednesday, April 29,1998, in Eureka,
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California, to discuss the PSDAR update and the licensee's future plans for Humboldt Bay. As desenbed in the PSDAR update, the licensee plans to remove the 250-foot ventilation stack, evaluate the option of moving the spent fuel from the spent fuel pool to dry cask storage, then initiate site dismantlement.
Ongoing maintenance and upkeep of the facility appeared to be adequate based on observations during the site tour. The facility was found to be in an acceptable structural cond; tion during this inspection. The in-leakage of ground water into the caisson basement, which was repaired in 1997, continued to remain very low at less than 15 gallons / day, indicating that the repair was successful. Work planning and development of the safety evaluation for removal of the 250-foot vent stack was underway with tentative plans for stack removal during the summer of 1998.
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Decommissioning Performance and Status Review (71801) and Organization, Management, and Cost Controls (36801)
1.1 Insoection Scooe Meetings were held with the licensee to review the schedule in the PSDAR update for
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future major decommissioning activities. A tour of the Unit 3 facility to evaluate structural integrity, housekeeping, fire loading, and radiological controls was completed. Recent changes to emergency plan procedures were also reviewed.
1.2 Observations and Findinos
On February 27,1998, PG&E submitted to the NRC a revised PSDAR describing planned changes to the status of the Humboldt Bay Unit 3 facihty. The planned changes -
were reviewed with the licensee. These discussions included three main areas:
1) removal of the Unit 3 ventilation stack,2) transfer of the spent fuel from the spent fuel pool to dry cask storage, and 3) eventual dismantlement of the site. The licensee was
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l evaluating several contractor bids to remove the 250-foot tall ventilation stack associated with Unit 3. A smaller stack to provide continued ventilation and filtering of the air from the Unit 3 structures would be constructed to replace the older stack. The 250-foot stack l
had been identified by the licensee as a potential hazard from a major earthquake. As a
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precautionary action, the licensee had determined that early removal of the stack was
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justified based on the potential consequences to the site structures should the stack not be able to withstand the recently postulated worse case earthquake from the nearby faults.
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The licensee was evaluating options for removing the 390 spent fuel assemblies from the spent fuel pool and storing the fuel in dry cask storage in an onsite ISFSI. The ISFSI
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would not be complete until the year 2002, at the earliest. The licensee also planned to evaluate the option to proceed with dismantlement of the facility once the fuel was removed from the pool.
NRC personnel from headquarters and Region IV attended a public maeting in Eureka, Cahfornia, to Mform local residents of the NRC regulations on plant decommissioning and to provide the licensee an opportunity to discuss its plans for vent stack removal and dry cask storage. Approximately 60 persons attended the meeting.
A tour of the Humboldt Bay facility was conducted as part of this inspection. The caisson in-leakage continued to remain low at less than 15 gallons / day. The facility appeared in good structural condition, fire loading and OSHA compliance appeared acceptable, good housekeeping practices were being implemented, and radiological posting and controls were observed and determined to be adequate. A tour of the radwaste storage building did not identify any problems with storage of material. The base of the 250-foot stack was examined and existing cracks in the stack were observed. Several of the cracks were being monitored by the licenses. There had been no significant changes in the width of the cracks being monitored.
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The licensee issued revisions to several emergency planning procedures in January and February 1998. This included the following procedures which were reviewed as part of j
this inspection:
EOP-2, Loss of 2.4 KV Bus, Rev. 73,2/26/98 EOP-7, Irradiated Fuel Damage, Rev. 22,1/29/98
EOP-8, Abnormal Conditions During a Planned Batch Release, Rev. 66,2/26/98 e
EOP-9, Unplanned Release of Liquid Radioactivity, Rev. 21,2/26/98
EPIP R-4, Personnel Instructions During Emergency Situations, Rev. 23,1/29/98 o
EPIP R-7, Establishment of the On-Site Emergency Organization and Notification l
e of Offsite Organizations, Rev. 25,2/26/98 LLOA, Local Letters of Agreement, Rev. 2,2/26/98
in general, changes to the procedures were minor. Procedures EOP-2, EOP-8 and EOP-9 revised the actions related to the caisson sump. This was necessary now that the in-leakage problem had been repaired. The letters of agreement were updated in November 1997. Letters of agreement included General Hospital, St. Joseph Hospital, City Ambulance, and Humboldt Bay Fire District.
1.3 Conclusions
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The licensee submitted a revision to their PSDAR on February 27,1998. On April 29, 1998, a public meeting was held in Eureka, Cahfornia, to inform the public of the NRC regulations governing plant decommissioning and the licensee's plans for l
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-6-decommissioning activities, with particular attention to removal of the 250-foot stack and transfer of the spent fuel to dry cask storage.
Observations during tours of the Unit 3 facility and radiologically cpntrolled area indicated that the facility was being maintained in an acceptable SAFSTOR condition. Fire loading, radiological controls, structural integrity of the facility, and housekeeping were found to be adequate.
Recent changes to emergency procedures were found to be acceptable and consistent with the current status of the facility.
- Organization, Management, and Cost Controls (36801)
2.1 Insoection Scooe Changes to the licensee's organization were reviewed for any management changes since the last inspection.
2.2 Observations and Findinas The licensee recently hired a new radiation protection manager for the site. The previous radiation protection manager was a long-term contractor, An overlap of several
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weeks had been scheduled to ensure that projects underway were not affected by the change in personnel. The new radiation protection manager had approximately 20 years experience at operating nuclear plants. He was a certified health physicist with a
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master's degree in Radiological Health Physics.
Technical Specification Vll.C.2.d required the radiation protection manager to have a bachelor's degree or equivalent,5 years experience in applied radiation votection, specialized knowledge in radiation and criticality requirements and practices, and knowledge of related regulatory requirements and practices. The new radiation protection manager met the requirements established in the technical specification.
All other personnel in the management organization remained the same as the last inspection.
~ 2.3 Conclusions A recent change to the licensee's organization involved the replacement of the radiation protection manager. The new individual met all requirements of the technical L
specifications for this position.
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Radioactive Waste Treatment, Effluents and Environmental Monitoring (84750)
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_Insoection Scoce On March 31,1998, PG&E submitted to the NRC the annual Radioactive Effluent Release Repcrt for 1997 as required by Technical Specification Vll.J.3. The data provided in the report was reviewed for any significant changes from previous years.
3.2 Observations and Findinos Technical Specification Vll.J.3 required the licensee to submit an annual effluent release report covering the previous calendar year to the NRC by April 1 of each year. The report must include a sumrnary of the quantities of radioactive liquids, gaseous effiuents and solid wastes released from the unit. The licensee submitted the required report on March 31,1998. A review of the 1997 data compared to the previous 2 years indicated that releases were comparable for fission and activation product gaseous and liquid effluents and lower in 1997 for particulate released in gaseous effluents. All releases were significantly below 1 percent of applicable regulatory limits. The licensee did report that higher than usual results were obtained for composite samples collected between July 11 and 22,1997, for cobalt (Co)-60. However, comparison with the liquid radwaste process monitor and with grab samples taken at the continuous discharge point did not support the higher Co-60 values. The licensee determined that contaminated sample tubing led to the unusual readings. The higher readings were used as the values reported for the period to be more conservative.
Two shipments of contaminated waste were sent to the Envirocare low level waste storage facility in Clive, Utah. The contaminated waste included material associated with the work performed in the caisson suppression chamber to repair the in-leakage problem. No spent resins or sludge were shipped in 1997.
The licensee calculated that the estimated radiological dose to the average adult-age public from the water pathway due to releases from Humboldt Bay was 0.003 mrem /yr total body and to the average child-age group was 0.004 mrem /yr to the bone. These exposures were significantly lower than the limits to the public established in 10 CFR j
Part 50, Appendix i of 3 mrem /yr to the total body and 10 mrem /yr to any organ. For i
gaseous effluents, the licensee calculated the maximum dose to an average individual was 0.001 mrem /yr. This value was also well below the limits in 10 CFR Part 50,
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l Appendix I of 10 mrem /yr gamma radiation and 20 mrad /yr beta radiation from noble l
gases, and 15 mrem /yr to any organ from tritium and radionuclides in particulate form.
i For the total-body dose to a member of the public at the site boundary, the estimated average dose was determined using environmental thermoluminescent dosimeter (TLD)
readings. The dose was estimated to be 0.016 mrem /yr. The 10 CFR 20.1302(b)(2)(ii)
l limit was 50 mrem /yr.
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-8-3.3 Conclusions Offsite radiological doses due to gaseous and liquid effluents released from the l
Humboldt Bay facility in 1997 were reviewed and determined to be within regulatory limits.
Self Assessment, Audits and Corrective Actions (40801) and Effectiveness of Licensee Controls in identifying, Resolving and Preventing Prob' ems (40500)
4.1 Insoection Scooe The licensee conducted periodic self-assessments and investigations into events that occurred during maintenance, operations and surveillance activities. These events were formally tracked and evaluated by management.. Several selected events were analyzed to determine whether the licensee's system of tracking, evaluating and dispositioning potential problems was being effectively implemented.
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l 4.2 Observations and Findinas The process used at Humboldt Bay for identifying and tracking potential quality related problems and nonconformances was documented in Procedure HBAP C-12,
" Identification and Resolution of Problems and Nonconformances," Revision 6A. When an issue was identified that could be a problem, the individual identifying the problem would complete a job order Procedure HBAP C-12 provided guidance criteria to determine if a job order was required. The job order was reviewed by the shift foreman, who then forwarded the job order to the plant engineer for evaluation. The plant engineer determined whether a quality problem or nonconformance was involved.
Criteria were provided in the procedure for determining if the problem was a nonconformance issue. A priority number was assigned ranging from high priority (requiring immediate attention or correction within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) to low priority (repair when practical). A computerized tracking system was used to monitor the progress of the l
repair and track the age of the job order. Goals were set to prevent job orders from becoming too old. To close a job order, the person assigned to the job order completed a section on the job order form describing the corrective action taken. Then quality control personnel performed an evaluation of the corrective action. If quality control personnel agreed with the action, the job order was closed.
Three events were selected from the list of job orders and nonconformance reports to
evaluate the process used by the licensee to recognize the significance of potential
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problems and perform self-assessments. The first self-identified event that was reviewed
involved the discovery in January 1998 that higher levels of krypton (Kr)-85 were present
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l in the spent fuel pool water. Monthly samples of spent fuel pool water had been collected and analyzed as required by Technical Specification 3.6.5. The analysis included a gamma scan of the sample. Typicalisotopes detected were Cs-137, neptunium (Np)-239, and Kr 85. The Kr-85 levels over the past year averaged l
l 1 x 10 microcuries/ml (uCi/mi). The sample taken in January 1998 indicated Kr-85 j
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levels of 1 x 10 '5 Ci/ml. This was 10 times higher than normallevels. The licensee performed additional water sampling which confirmed the higher readings. In addition, americium (Am)-241 and Cs-134 were detected slightly above their minimum detectable level. Normally these isotopes were below minimum detectable limits in the spent fuel pool water samples. The strip chart recorder for the stack gas monitoring system was evaluated and indicated no elevated radiation levels were released from the building.
Management was verbally informed and Job Order C17236 was initiated. Radiation I
protection personnel performed an analysis of the data. Contact was made with two other facilities with fuel that had been in long-term storage. Similar events had not been encountered at the other facilities. The data base maintained by the institute of Nuclear i
Power Operations was reviewed. No similar events had been noted. The licensee concluded that the Kr-85 was from a fuel pin cladding failure. Calculations by the radiation protection group indicated the levels of Kr-85 in the pool were representative of no more than one pin leaking. The levels of Kr-85 measured were also determined to be well within the boundaries of the SAFSTOR Plan accident analysis which assumed loss of all Kr-85 in all fuel assemblies. As of April 25,1998, the Kr-85 levels in the pool had shown a continued decrease and were down to twice normal levels for the pool water.
The increased Kr-85 in the spent fuel pool water had been recognized immediately by the individual performing the sample counting. Job Order C17236 was reviewed by the plant review committee. The issue was determined to be a quality problem and a technical review group meeting was conducted. Sampling was increased from monthly to weekly to track the trend for the isotopes in the spent fuel pool water. Management continued to track the weekly sam,nles to ensure that another Kr-85 levelincrease did not
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occur that would require additional evaluation of the condition of the spent fuel. Previous visualinspections of the fuel elements had not indicated deterioration of the fuel cladding. The inspector determined that for the Kr-85 event, the licensee took the appropriate actions.
The second event reviewed during this inspection involved the potential incorrect conclusion in the SAFSTOR decommissioning plan for the accident associated with the radwaste tank. During a review of the environmental report, a discrepancy was recognized between the environmental report, Section 6.3.2.4, and the decommissioning plan, Section 1.2.4, concerning the final concentration of four radioisotopes at the point of l
discharge after an uncontrolled release of the radwaste tank contents. There was a
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discrepancy as to whether the calculations showed the levels would be below the required 10 CFR 20, Appendix B, Table 2 limits at the discharge point. As a result of this i
concern, a job order was initiated and a review of the issue was completed by management. The isotopes involved were Cs-137, Cs-134, Co-60, and strontium (Sr)-
90. The issue was determined to be a potential quality problem and was evaluated to determine if a nonconformance existed. The licensee initiated a review of the 1984 calculations performed for the potential accident. There appeared to be a discrepancy between the environmental report and the decommissioning plan concerning the location of the final release point, which could be considered the point where the plant discharge line emptied into the canal or where the canal emptied into the bay. A review of NRC guidance in the Standard Review Plan, Section 15.7.3, determined that the evaluation of
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- 10-whether the concentration exceeded 10 CFR Part 20 limits should be made at the nearest potable water supply. The NRC, as part of the safety evaluation of the decommissioning plan, performed calculations for the radwaste tank discharge and found the releases to be well below regulatory limits. The plant technical review group concluded that the guidance in the NRC Standard Review Plan should have been used.
This confirmed the conclusion that the isotopes discharged would be significantly below regulatory limits. The licensee identified that the two documents should be consistent.
The inspector noted that the licensee had completed a very thorough review of the event and had adequately evaluated the issue.
The third event reviewed during this inspection involved an unplanned reduction in the water level of the spent fuel pool. On October 8,1997, a radiation protection monitor took a sample of the spent fuel pool water to conduct the monthly spent fuel pool water quality check required by technical specification 3.6.5. After taking the sample, the sample valve was left open. Later that day, operations personnel observed the chart recorder for the spent fuel pool water level and recognized that the water level had dropped slightly. It was noted that detecting this small change on the circular strip chart recorder by the operations personnel required attention to detail and could have been easily missed. The operations personnel initiated a walkdown of the spent fuel pool systems and located and closed the valve. Management was informed and a job order initiated. The issue was identified as a quality problem.
The technical review group evaluated the problem. A nonconformance report was issued and an evaluation completed of the deportability of the incident. The root cause was determined to be personnel error. It was noted that similar events had occurred over the past several years with other systems. Each case appeared to be an isolated event. The radiation protection individual involved with leaving the sample valve open was provided with a special training session by the plant r anager.
This event resulted in a small amount of water draining from the spent fuel pool. The floor of the fuel building was at elevation 12 feet. Technical Specification Ill.B.2.a required the water level to be maintained greater than elevation 10.5 feet, which would be 1.5 feet below the floor level. Typically, the water level was maintained at elevation 10.7 feet to 11 feet. The top of the spent fuel assemblies were at elevation -2 feet, which would provide for 12.5 feet of water over the fuel at the required technical specification level of elevation 10.5 feet. A low water alarm in the pool was set at elevation 10.7 feet.
The suction for the sample line was at elevation 10 feet. Therefore, leaving the sample line open could result in the low-water alarm being activated and the water level being reduced to 0.5 feet below the minimum technical specification level. This would leave 12 feet of water above the fuel.
l Alert operations personnel were credited with realizing that the pool had lost water and
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closed the sample valve before the technical specification limit was violated. Since the sample line intake was at the 10-foot elevation, water level could not drain through the open sample valve to reach a point of concern related to adequate coverage of water over the spent fuel. The licensee actions in evaluating this issue and their recognition of
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the potential seriousness of mispositioned valves were appropriate. The alertness of the operations personnel demonstrated their ability to quickly recognize an abnormal
condition at Unit 3.
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. Conclusions The licensee had a formal self assessment and corrective action program that documented problems, brought the problems to the at.ention of management, and tracked the resolution and completion of corrective actions. Several specific events evaluated verified the effectiveness of the program.
Inspections of Final Surveys (83801)
5.1 Insoection Scoce A characterization survey was completed by the licensee to assess the level of radiological contamination in the soil and the bay near the Humboldt Bay facility. A preliminary review was conducted of the results of this survey, 5.2 Observations and Findinas The licensee utilized the concepts prescribed in NUREG 1575, " Multi-Agency Radiation Survey and Site Investigation Manual," to develop a site characterization survey and report assessing the level of contamination near the Humboldt Bay facility and in the bay.
This report will be reviewed in detail by the NRC as part of the final site license termination process in the future when the licensee is in the final stages of decommissioning the site. The report provided a satisfactory overall summary of the current condition of the radiological environment around the site.
Numerous soil and sediment sa.1ples were taken in and near to the owner controlled area. Samples included the top soil, and were also taken from several feet below the
surface. Sampling was conducted at 281 locations. Over 700 core samples were collected. Computer codes were used to calculate the projected doses to the affected population for the various pathways. These results provided a basis for comparison with the NRC criteria of 25 mrem /yr for license termination specified in 10 CFR 20.1403. In addition, the licensee applied an ALARA concept to the criteria and evaluated the level of contaminated soil using a criteria of 15 mrem /yr. Soil in areas near the plant and the discharge canal were found to have sufficient contamination levels to require remediation. The predominant isotope found was Cs-137. Radiation levels in the bay sediment were found to be at natural background levels. The characterization report
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provided the licensee with a more accurate basis to evaluate the overall cost to j
decommission the site.
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12-5.3 Conclusions The licensee had completed a characterization survey and report to determine the levels of contamination in the soil and in the bay sediment adjacent to the Humboldt Bay I
facility. Contamination levels were within expected ranges onsite with Cs-137 being the predominant isotope. No radioactivity above natural background levels was found in the bay.
Follow-up of Open items (92701)
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6.1 (Discussed) IFl 50-133/98001-03 Tritium in Monitorina Wells MW-11 and MW-1:
The licensee was continuing to assess the potential source for the tritium detected in Monitoring Wells No.1 and No.11. Samples taken February 17,1998, found tritium levels of 274 pCi/lin Well No.11 and 447 pCi/lin Well No.1. This compared to samples taken November 18-20,1997, which indicated 312 pCi/l for Well No.11 and 401 pCi/l for Well No.1 Data back to 1987 had shown tritium in monitoring Well No.11. In 1987 levels were approximately 2,700 pCi/l. Humboldt Bay Technical Specification, Table V-1 established a reporting criteria for tritium of 30,000 pCi/l. For drinking water, the Environmental Protection Agency limit for tritium is 20,000 pCill.
The tritium has continued to decrease over the years at a rate faster than natural decay (the half life of tritium is 12.3 years). However, the tritium should have dissipated more quickly due to the transportability of tritium in ground water. This raised the concern that a source of tritium may be present near the monitoring wells. Tritium found in the natural environment is approximately 10 pCi/l and would not account for the levels found in the monitoring wells. The largest tritium source onsite was the spent fuel pool. Tritium levels in the pool were measured at 98,000 pCill on February 20,1997. A leak in the spent fuel pool stainless steel liner allows water to collect in the gap between the liner and the concrete wall. Tritium measured in the gap between the spent fuel pool liner and the building wall measured 12,000 pCi/l on February 20,1997. This indicated that there was also in-leakage into the gap from the ground water. The liner gap water was periodically drained to keep the gap water level below the groundwater level to ensure that the groundwater leaked into the gap and tritium was not leaking into the groundwater. The leak rate into the liner gap was approximately 1-1.5 gallons / day combined groundwater and spent fuel pool water. Based on the February 20,1997 tritium concentrations, the majority of the water in the liner gap was coming from groundwater. No tritium analysis had been conducted since February 1997 for the spent fuel pool or liner gap.
Another source of tritium onsite was a railroad drain pipe recently sampled and found to have tritium levels of 3,200 pCi/1.' This drain pipe runs near Monitoring Well No.11. The licensee planned to drain the tritiated water f;om the pipe and rnonitor the effect on the f
tritium in the wells. The results of this evaluation will be reviewed during a future inspection.
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Exit Meeting The inspectors presented the inspection results to members of the licensee management at the exit meeting on April 30,1998. The licensee acknowledged the findings presented. The licensee did not identify as proprietary any information provided to, or reviewed by, the inspector.
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ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee J. Albers, Senior Radiation Protection Engineer (Radiation Protection Manager)
L. Brown, Radiation Protection J. Crow, Training M. Grossman, Operations Supervisor E. Kahler, Decommissioning Project Manager P. Morris, Radiation Protection T. Moulia, Plant Manager B. Norton, Consultant R. Parker, Radiation Protection D. Sokolsky, Senior Licensing Engineer R. Willis, Plant Engineer INSPECTION PROCEDURES USED 36801 Organization, Management, and Cost Controls 40500 Effectiveness of Controls in Identifying, Resolving, and Preventing Problems 40801 Self Assessment, Audits, and Corrective Actions 71801 Decommissioning Performance and Status Review 84750 Radwaste Treatment and Effluent and Environmental Monitoring 92701 Follow-up on Open items i
ITEMS OPENED, CLOSED, AND DISCUSSED O_ened o
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None Discussed 50-133/98001-03 IFl Tritium in Monitoring Wells MW-11 and MW-1 l
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2-LIST OF ACRONYMS l
ALARA As Low As Reasonably Achievable
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CFR Code of Federal Regulations
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EOP Emergency Operating Procedure EPIP Emergency Plan Implementing Procedure IFl Inspection Follow-up item ISFSI Independent Spent Fuel Storage installation LLOA Local Letters of Agreement mR milliroentgen NRC Nuclear Regulatory Commission PG&E Pacific Gas & Electric PSDAR Post-Shutdown Decommissioning Activities Report TLD Thermoluminescent Dosimeter uCi microcurie VIO Violation
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