IR 05000206/2007011
| ML071440275 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/24/2007 |
| From: | Spitzberg D NRC/RGN-IV/DNMS/FCDB |
| To: | Rosenblum R Southern California Edison Co |
| References | |
| IR-07-011 | |
| Download: ML071440275 (13) | |
Text
SUBJECT:
NRC INSPECTION REPORT 050-00206/07-011
Dear Mr. Rosenblum:
This refers to the inspection conducted on April 16-19, 2007, at Southern California Edison Companys San Onofre Nuclear Generating Station, Unit 1 facility. This inspection was an examination of decommissioning activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license. The inspection included an examination of selected procedures and representative records, observations of activities, and interviews with personnel.
A preliminary exit briefing was presented to your staff at the conclusion of the onsite inspection, and a final briefing was presented telephonically to members of your staff on May 11, 2007, following the receipt of the NRCs soil sample results on the same day. The soil samples were collected from the Unit 1 site during the inspection and were analyzed by Oak Ridge Institute for Science and Education (ORISE) on behalf of the NRC. The enclosed report presents the results of the samples and the overall results of the inspection. In summary, the inspection determined that you were conducting decommissioning activities in compliance with regulatory and license requirements.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be made available electronically for public inspection in the NRC Public Document Room or from the NRCs document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction.
Should you have any questions concerning this inspection, please contact the undersigned at (817) 860-8191 or Mr. Robert Evans, Senior Health Physicist, at (817) 860-8234.
Sincerely,
/RA/
D. Blair Spitzberg, Ph.D., Chief Fuel Cycle and Decommissioning Branch
Southern California Edison Co. -2-Docket No.: 050-00206 License No.: DPR-13
Enclosure:
NRC Inspection Report 050-00206/07-011
REGION IV==
Docket No: 050-00206 License No: DPR-13 Report No: 050-00206/07-011 Licensee: Southern California Edison Co.
P.O. Box 128 San Clemente, California 92674 Facility: San Onofre Nuclear Generating Station, Unit 1 Location: San Clemente, California Dates: April 16-19, 2007 Inspectors: Robert J. Evans, P.E., C.H.P., Senior Health Physicist Fuel Cycle & Decommissioning Branch Janine F. Katanic, Ph.D., Health Physicist Nuclear Materials Inspection Branch Approved By: D. Blair Spitzberg, Ph.D., Chief Fuel Cycle & Decommissioning Branch Attachment: Supplemental Inspection Information ADAMS Entry: IR05000206-07-011 on 04/16/2007 - 04/19/2007; Southern California Edison Co., San Onofre Nuclear Generating Station; Unit 1. Decommissioning Report. No VIOs.
-2-EXECUTIVE SUMMARY San Onofre Nuclear Generating Station, Unit 1 NRC Inspection Report 050-00206/07-011 This inspection was a routine, announced inspection of decommissioning activities being conducted at the San Onofre Nuclear Generating Station, Unit 1 facility. Areas inspected include safety reviews, design changes and modifications; self-assessment, auditing, and corrective action; decommissioning performance and status review; occupational radiation exposure; and inspection of final surveys. The inspection determined that the licensee was conducting decommissioning activities in compliance with regulatory and license requirements.
Safety Reviews, Design Changes and Modifications at Permanently Shutdown Reactors
- The licensees safety review and design change program was in compliance with 10 CFR 50.59 requirements (Section 1).
Self Assessment, Auditing, and Corrective Action at Permanently Shutdown Reactors
- The licensee conducted self-assessments and audits in accordance with quality assurance program requirements (Section 2).
Decommissioning Performance and Status Review at Permanently Shutdown Reactors
- The licensee was conducting demolition work with an emphasis on industrial and radiological safety. Radiation protection controls had been implemented including postings, boundaries, and labels (Section 3).
- A spill of approximately 2000 gallons of water consisting of groundwater and rainwater collected from within the Unit 1 site had occurred. The inspectors determined that the spill was not reportable to NRC and did not result in a release to the environment in excess of the licensees Offsite Dose Calculation Manual limits (Section 3).
Occupational Radiation Exposure
- No individual exceeded the regulatory limit for total effective dose equivalent during 2006. The licensees As Low As Reasonably Achievable program was determined to be effective (Section 4).
Inspection of Final Surveys at Permanently Shutdown Reactors
- The inspectors conducted confirmatory surveys of the former Building A42 pit and the containment sphere concrete foundation. The surveys included measurement of ambient gamma exposure rates, collection of two soil samples, and measurement of surface contamination levels. Radioactive material was not identified in concentrations in excess of the NRCs screening values (Section 5).
-3-Report Details Summary of Plant Status San Onofre Nuclear Generating Station, Unit 1 was permanently shut down during November 1992 and was permanently defueled by March 1993. The unit remained in SAFSTOR until June 1999 when decommissioning was initiated. At the time of this inspection, the licensee was conducting decommissioning activities under the DECON option as stated in its Post Shutdown Decommissioning Activities Report dated December 15, 1998. DECON is defined as the immediate removal and disposal of all radioactivity in excess of levels which would permit the release of the facility for unrestricted use.
Since the previous inspection, the licensee removed the remainder of the steel containment sphere wall during late-March 2007 and shipped the material to an offsite disposal site. During the inspection, the licensee was demolishing the former radwaste building concrete walls. Final status surveys were in progress in the containment sphere foundation area. The licensee plans to complete the above-ground phase of decommissioning by late 2008.
1 Safety Reviews, Design Changes, and Modifications at Permanently Shutdown Reactors (37801)
1.1 Inspection Scope The purpose of this portion of the inspection was to ascertain whether design changes, tests, experiments, and modifications were effectively reviewed, conducted, managed, and controlled during plant decommissioning in accordance with 10 CFR 50.59.
1.2 Observations and Findings Regulation 10 CFR 50.59 addresses the change control process, a process used by the licensee to determine if a proposed change to the facility, procedures, tests, or experiments is subject to a license amendment and NRC approval. The process is implemented through site procedure SO123-XV-44, "10 CFR 50.59 and 72.48 Program." This procedure provided instructions for both initial screening and subsequent full evaluation, if necessary, of facility or procedure changes to confirm if the licensee can implement these changes without NRC approval. This was a common program for the two operating units and the decommissioning unit.
The inspectors reviewed 10 CFR 50.59 screens of various facility changes and found that all screens had been completed in accordance with procedural requirements. No full 10 CFR 50.59 evaluations had been performed for Unit 1 changes since the last inspection, and the inspectors did not identify any 10 CFR 50.59 screens that should have been processed as full evaluations.
-4-1.3 Conclusions The licensees safety review and design change program was in compliance with 10 CFR 50.59 requirements.
2 Self-Assessment, Auditing, and Corrective Action at Permanently Shutdown Reactors (40801)
2.1 Inspection Scope The objective of this portion of the inspection was to evaluate the effectiveness of licensee controls for identifying, resolving, and preventing issues that degrade safety or the quality of decommissioning.
2.2 Observations and Findings The inspectors conducted a review of the licensees implementation of the quality assurance (QA) program for Unit 1 to ensure compliance with the QA Program Topical Report SCE-1-A. The requirements for QA audits are provided in Section 17.2.18 of the Topical Report. This section provides the areas to be audited and the audit frequency.
The licensee maintained a master audit schedule to track all Unit 1 audits. The inspector confirmed that all audits were up-to-date.
The inspectors reviewed the audits of the environmental, fire protection, and maintenance programs. The licensee conducted these reviews in accordance with QA program requirements, and these reviews provided useful information to licensee management.
The licensee also conducted surveillances of the Unit 1 final status survey program since the last inspection. Further, a QA review of field activities was in progress during the inspection. The auditor concluded that the Unit 1 decommissioning survey program met the intent of the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), NUREG-1575, Revision 1. The audit resulted in no findings and one recommendation. The inspectors concluded that the licensees QA program was providing sufficient oversight of the decommissioning process, including final surveys.
Appendix B to 10 CFR Part 20, Criterion XVI states that measures shall be established to assure that conditions adverse to quality and nonconformances are promptly identified and corrected. The inspectors reviewed the licensees self-assessment reports for events that have been identified since the last inspection of this program area. The reports included one root cause evaluation, four apparent cause evaluations, and one common cause evaluation.
All evaluations were related to industrial near-misses or minor decommissioning incidents. The events included an overloaded intermodal, dropped metal beam, smoldering fire incidents, and accidental cutting of an electrical circuit. The licensee conducted detailed evaluations of each incident and formulated corrective actions. In
-5-summary, the licensee continued to conduct self assessments to identify and to correct incidents that may impact safety or decommissioning.
2.3 Conclusions The licensee conducted self-assessments and audits in accordance with QA program requirements.
3 Decommissioning Performance and Status Review at Permanently Shutdown Reactors (71801)
3.1 Inspection Scope The inspectors evaluated whether the licensee and its contracted workforce were conducting decommissioning activities in accordance with license and regulatory requirements. The inspectors also reviewed the circumstances surrounding the April 13, 2007, spill of water that contained low levels of radioactivity.
3.2 Observations and Findings During site tours, the inspectors observed decommissioning work in progress. The decommissioning work included demolition of the former radwaste building walls.
Industrial safety and radiation protection controls were evident in all areas of Unit 1.
Safety representatives were continuously present during work activities. The inspectors observed the movement and handling of radioactive material and concluded that radiological controls and postings met regulatory requirements.
On April 13, 2007, the licensee informed the NRC of a spill of approximately 2000 gallons of water within the Unit 1 site boundary. The water was a combination of both groundwater and rainwater collected from the Unit 1 radiologically restricted area.
The water was being stored in a tank located in the Unit 1 restricted area. The licensee determined that it no longer needed the tank and decided to discharge the contents in preparation for release of the tank from the restricted area. Prior to the discharge of the tank contents, licensee personnel obtained a sample of the water for analysis. The sample results indicated that the tank water contained 1.3 E-7 microcuries per milliliter of cesium-137. This sample result was below the effluent concentration limit of 1 E-6 microcuries per milliliter specified in 10 CFR Part 20, Appendix B, Table 2. The tritium concentration was below detection levels.
The licensee decided to discharge the contents of the tank into the Unit 1 yard sump.
The tank contents were discharged into catch basin #9 which should have routed the water through a series of underground pipes to the yard sump. The water then would have been released through the monitored effluent release point to the environment via the Units 2/3 outfalls. However, licensee personnel performing the tank discharge were unaware that a large section of the 6-inch cast iron pipe that led from catch basin #9 to the yard sump had been removed. Accordingly, when the contents of the tank were discharged into the catch basin, the contents poured out of the broken pipe and onto a sandy area within the Unit 1 site. Water was observed by licensee personnel and
-6-contractors to be coming out of the broken pipe, but the source of the water was not identified until almost 2,000 gallons had been discharged, roughly the entire contents of the tank. The licensee suspects that the spilled fluid was captured by an adjacent dewatering well and was subsequently transferred to the yard sump. The inspectors agreed with the licensees assessment, because the dewatering well was situated about 10 feet from the spill location.
Based on the water sample results, the licensee determined that the spill was not reportable to NRC under 10 CFR 20.2203, 10 CFR 50.72 or 10 CFR 50.73. However, pursuant to the licensees Groundwater Protection Initiative, the licensee voluntarily notified NRC and certain State and local officials regarding the spill. The licensee concluded that the discharge of the water did not result in a release to the environment in excess of Offsite Dose Calculation Manual limits. The licensee initiated an apparent cause evaluation of the event for management review and for determination of corrective actions to prevent recurrence of the event.
3.3 Conclusions The licensee was conducting demolition work with an emphasis on industrial and radiological safety. Radiation protection controls had been implemented including postings, boundaries, and labels. A spill of approximately 2000 gallons of water consisting of groundwater and rainwater collected from within the Unit 1 site had occurred. The inspectors determined that the spill was not reportable to NRC and did not result in a release to the environment in excess of the licensees Offsite Dose Calculation Manual limits.
4 Occupational Radiation Exposure (83750)
4.1 Inspection Scope The inspectors reviewed Unit 1 occupational radiation exposures for calendar year 2006 to verify whether they met the limits specified in 10 CFR Part 20. A review of the licensees As Low As Reasonably Achievable (ALARA) program for Unit 1 was also performed.
4.2 Observations and Findings The inspectors found that no individual at Unit 1 exceeded the regulatory limit for total effective dose equivalent during 2006. External radiation exposures were obtained for personnel entering a radiologically controlled area through the use of quarterly thermoluminescent dosimeters (TLD) and direct reading electronic dosimeters. For calendar year 2006, 1470 individuals at Unit 1 were monitored for external radiation exposure with a TLD and 183 individuals of those monitored received a measurable dose. The highest individual dose for Unit 1 personnel for 2006 was 1.024 rem.
Licensee personnel routinely compared the quarterly TLD data to the readings obtained from the electronic dosimeters. Based on industry guidance, the licensee implemented
-7-a review for cases over 100 millirems when the TLD differed from the electronic dosimeter by +/- 25 percent.
Regarding internal radiation exposures and skin contamination, a Personnel Contamination Report (PCR) was generated for each instance when personnel alarmed a whole body counter when leaving a radiologically controlled area. The licensee had established procedures to assess the level of contamination and perform decontamination. During May-December 2006, 96 PCRs were generated for Unit 1.
None of these PCRs was the result of a contamination that exceeded trigger levels that would result in assigning dose to the skin. For PCRs with suspected contamination on the face, neck, or about the shoulders, a whole body count was performed. Of the Unit 1 whole body counts performed in 2006, none resulted in the assignment of internal dose.
The inspectors reviewed the licensees ALARA program for Unit 1. Performance indicators and ALARA goals were tracked and evaluated by the licensee. The ALARA goal for Unit 1 for 2006 was 25.470 person-rem and was based on work activities planned for the calendar year. The actual collective dose for 2006 was 17.396 person-rem. The activities that were major contributors to collective dose included containment demolition, radioactive waste system removal and remediation, spent fuel pool liner removal, and routine health physics support activities. The licensee implemented several effective dose reduction techniques, such as the use of a suspended work platform instead of scaffolding during the spent fuel pool liner project.
4.3 Conclusions The licensee had an occupational exposure monitoring program that effectively monitored internal and external doses to radiation. No individual exceeded the regulatory limit for total effective dose equivalent during 2006. The licensees ALARA program was determined to be effective.
5 Inspection of Final Surveys at Permanently Shutdown Reactors (83801)
5.1 Inspection Scope Confirmatory surveys were conducted in the former containment sphere foundation area and the Building 42A reactor examination detensioning pit to independently ascertain the radiological conditions of these areas.
5.2 Observations and Findings The inspectors conducted a confirmatory radiological survey of the former Building 42A reactor examination detensioning pit. This area had been recently decommissioned by the licensee including removal of a metal liner from the pit. The survey consisted of measurement of ambient gamma exposure rates. In addition, two soil samples were collected, one from inside the pit and one from the backfill material subsequently used to refill the pit.
-8-The ambient gamma exposure rates in and around the pit were measured with a Ludlum Model 2401-P survey meter (NRC No. 21189G with calibration due date of 11/06/07).
Although the licensee had not established an acceptance criteria for exposure rate, the inspectors surveyed the area to identify any areas of elevated radioactivity for further sampling. The exposure rate measurements in the pit were consistent with background measurements (10-15 microRoentgens per hour).
The inspectors also collected two soil samples from the pit area for offsite analysis. The samples were shipped to Oak Ridge Institute for Science and Education (ORISE) for analysis by gamma spectroscopy. The inspectors allowed the licensee to analyze the two samples for gamma-emitting radionuclides. The sample results are provided in the Table below, in units of picocuries per gram (pCi/g):
Table: Split Sampling Results Sample Description NRCs Results (pCi/g) Licensees Results (pCi/g)
NRC-1, Building 42A cesium-137 0.01 +/- 0.02 cesium-137 Not Detected Pit Sample cobalt-60 0.03 +/- 0.04 cobalt-60 Not Detected uranium-235 0.06 +/- 0.09 uranium-235 Not Detected NRC-2, Excavated cesium-137 0.01 +/- 0.02 cesium-137 Not Detected Backfill Material adjacent to Building cobalt-60 0.01 +/- 0.02 cobalt-60 Not Detected 42A Pit uranium-235 0.12 +/- 0.10 uranium-235 Not Detected The sample results indicate that the radionuclides of concern were either not detected or were essentially zero in all of the samples. Further, the sample results indicate agreement between ORISE and the licensees laboratories. Accordingly, the licensees laboratory was determined to be technically capable of accurately detecting and quantifying radioactive material in site samples.
The inspectors also conducted confirmatory surveys in the containment sphere foundation area. At the time of the inspection, the licensee had completed the characterization survey and was commencing with the final status survey. The inspectors elected to conduct the confirmatory survey during the inspection because of the limited window of opportunity. Once the final status survey is complete, the licensee plans to partially demolish and backfill the area.
The survey consisted of measurement of beta-gamma activity on concrete surfaces.
The surveys were conducted using two Eberline E600 survey meters (NRC No. 079977, calibration due date of 09/15/07 and NRC No. 063473, calibration due date of 11/21/07)
with SHP380AB alpha-beta probes. The surveys consisted of limited scan surveys and 33 1-minute, fixed point measurements.
The inspectors also conducted limited gamma exposure rate measurements in the foundation area. The ambient gamma exposure rates ranged from 10-20
-9-microRoentgens per hour. These exposure rates were comparable to background levels for concrete surfaces.
As part of the survey, the inspectors collected background measurement from concrete located in the Unit 1 industrial area that was not situated in the radiologically restricted area. Using this background data, the inspectors calculated lower limits of detection for the survey meters. Two survey results from one area slightly exceeded the lower limit of detection. The area with the two elevated measurements was labeled as Grid 24 by the licensee.
The two sample results above the lower limit of detection of the survey meter were compared to the screening criteria for concrete surfaces. NRC guidance document NUREG-1757, Consolidated Decommissioning Guidance, Volume 2, Table H.1 provides the acceptable license termination screening values for building surface contamination.
The table includes acceptable screening levels for cobalt-60 and cesium-137. The most restrictive screening level is cobalt-60 at 7100 dpm/100 cm2. The highest fixed point measurement in the foundation area was less than half this screening level. Therefore, the NRCs confirmatory sample results were less than the NRCs screening criteria.
5.3 Conclusions The inspectors conducted confirmatory surveys of the former Building A42 pit and the containment sphere concrete foundation. The surveys included measurement of ambient gamma exposure rates, collection of two soil samples, and measurement of surface contamination levels. Radioactive material was not identified in concentrations in excess of the NRCs screening values.
6 Exit Meeting Summary The inspectors presented the preliminary inspection results to members of licensee management at the exit meeting on April 19, 2007. A final exit briefing was held telephonically with the licensee on May 11, 2007, following receipt of the soil sample results on the same day. The licensee did not identify as proprietary any information provided to, or reviewed by, the inspectors.
ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee T. Clepper, Project Manager, Unit 1 Decommissioning S. Enright, Project Manager, Unit 1 Health Physics B. Katz, Vice President, Nuclear Oversight & Regulatory Affairs J. Morales, Manager, Unit 1 Decommissioning J. Reilly, Vice President, Engineering & Technical Services A. Scherer, Manager, Nuclear Regulatory Affairs R. Waldo, Vice President, Nuclear Generation INSPECTION PROCEDURES USED IP 37801 Safety Reviews, Design Changes, and Modifications at Permanently Shutdown Reactors IP 40801 Self Assessment, Auditing and Corrective Action at Permanently Shutdown Reactors IP 71801 Decommissioning Performance and Status Review at Permanently Shutdown Reactors IP 83750 Occupational Radiation Exposure IP 83801 Inspection of Final Surveys at Permanently Shutdown Reactors ITEMS OPENED AND CLOSED Opened None Closed None Discussed None LIST OF ACRONYMS ALARA As Low As Reasonably Achievable IP NRC Inspection Procedure MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual (NUREG-1575)
ODCM Offsite Dose Calculation Manual ORISE Oak Ridge Institute for Science and Education PCR Personnel Contamination Report QA quality assurance TLD thermoluminescent dosimeters