IR 05000312/2008001
| ML081580536 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/06/2008 |
| From: | Whitten J Division of Nuclear Materials Safety IV |
| To: | Shetler J Sacramento Municipal Utility District (SMUD) |
| References | |
| IR-08-001 | |
| Download: ML081580536 (47) | |
Text
UNITED STATES
SUBJECT:
NRC INSPECTION REPORT 050-00312/08-001
Dear Mr. Shetler:
A Nuclear Regulatory Commission (NRC) inspection was conducted from April 14 through May 16, 2008, at your Rancho Seco Nuclear Generating Station. At the end of the site visits on April 17, 2008, and April 26, 2008, the inspectors briefed the Plant Closure and Decommissioning (Plant Manager) regarding the preliminary inspection findings. A telephonic exit briefing was conducted with the Plant Manager on May 19, 2008. The enclosed inspection report presents the scope and results of the inspection.
The inspection was an examination of activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. Within these areas, the inspection included reviews of your self-assessment, auditing, and corrective action programs, maintenance and surveillance activities, decommissioning performance, occupational radiation exposure, and final status surveys.
Based on the results of this inspection, the NRC has determined that a Severity Level IV violation of NRC requirements occurred involving failure to implement the Emergency Plan. This non-repetitive, licensee identified violation is being treated as a Non-Cited Violation (NCV), consistent with Section VI. A of the Enforcement Policy. This violation is described in Section 4.3 of the inspection report. If you contest the violation or significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to: (1) the Regional Administrator, Region IV; (2) the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.
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 NRC's document system (ADAMS), accessible from the NRC Web site at http://www,nrc.qov/readinq-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.
Sacramento Municipal Utility District-2-I k.,
Docket No.: 050-0031 2 License No.: DPR-54
Enclosures:
I.
NRC Inspection Report 050-00312/08-001 2. ORISE Survey Report DCN 1695-SR-02-0 (w/Attachments 1 & 2)
Sacramento Municipal Utility District-3-
REGION IV==
050-0031 2 DPR-54 050-0031 2/08-001 Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station 14440 Twin Cities Road Herald, California April 14 through May 16, 2008 Emilio M. Garcia, Health Physicist James L. Montgomery, Health Physicist Thomas H. Youngblood, Jr., Health Physicist Reactor Decommissioning Branch, Office of Federal and State Materials and Environmental Management Programs Jack E. Whitten, Chief Nuclear Materials Safety Branch B Supplemental Information Partial List of Documents Reviewed Enclosure 1
EXECUTIVE SUMMARY Rancho Seco Nuclear Generating Station NRC Inspection Report 050-0031 2/08-001 This inspection was a routine, announced examination of the decommissioning activities being conducted by the licensee at the Rancho Seco Nuclear Generating Station. Areas reviewed by the NRC inspectors included self-assessment, auditing, and corrective action programs, maintenance and surveillance activities, decommissioning performance, occupational radiation exposure, and final status surveys.
Self-assessment, Auditina and Corrective Action The licensee had effectively maintained its Corrective Action Program (CAP) used to control the identification, evaluation, and resolution of problems (Section 1.I).
- Audits completed by the licensee in calendar year (CY) 2007, and those scheduled for completion in CY 2008, when completed will collectively addressed all facility activities that are required to be audited. All auditors satisfied the qualification requirements (Section 1.2).
Maintenance and Su rvei I lance
The licensee had no safety-related structures, systems or components, and the requirements of the Maintenance Rule were no longer required to be implemented at this site. The licensee had continued to maintain their schedule on other surveillances and routine tests (Section 2).
Decommissionins Performance and Status Review The licensee continued efforts to dismantle and remove contaminated components and to remediate contaminated surfaces in a safe manner. The licensee had completed final status surveys on 21 7 of the projected 295 survey units (Section 3).
Occupational Radiation Exposure
A licensee audit of the occupational radiation exposure program was conducted by the licensee in accordance with quality assurance program requirements (Section 4.1).
- The licensee maintained an effective program for monitoring occupational radiation exposures. Occupational exposures for CY 2007 were determined by the inspectors to be below regulatory limits (Section 4.2).
A non-cited violation related to the licensees failure to follow the Emergency Plan was identified (Section 4.3).
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Enclosure 1
Inspection of Final Status Survevs The ORISE staff conducted confirmatory measurements on selected surfaces of the auxiliary building, including embedded piping. The results of the ORISE surveys conducted during this inspection will be reported at a later date (Section 5).
The results of survey activities conducted by ORISE during the December 10 through 13, 2007, site visit were documented on a report issued on March 12, 2008.
A copy of this ORISE report is attached to the inspection report as Enclosure 2. With the exceptions of the elevated direct reading measurement identified by ORISE in two locations, the ORISE surveys confirmed the accuracy of the licensees final status survey results for the locations surveyed (Section 5).
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Enclosure 1
Report Details Summary of Facility Status The Rancho Seco Nuclear Generating Station was permanently shut down in June 1989. All spent reactor fuel has been moved to an onsite Independent Spent Fuel Storage Installation (ISFSI). At the time of this inspection, the licensee was conducting decommissioning activities under the provisions of the incremental decommissioning option of Rancho Secos Post Shutdown Decommissioning Activities Report (PSDAR)
dated March 20, 1997.
Decommissioning work activities conducted by the licensee under the PSDAR included the auxiliary building, reactor building, spent fuel building, and exterior areas. All major components located in these buildings and areas had been removed, packaged, and shipped offsite for disposal. In the auxiliary building, remediation and final status survey activities were being conducted. In the reactor building, the concrete and steel removal project continued with approximately 28.6 million pounds of the concrete and steel being removed and shipped offsite. As part of planned decommissioning activities, on April 26, 2008, the reactor building polar crane was brought down by the licensees contractor using explosives. In the fuel handling building, remediation of the surfaces continued.
At the time of the inspection, the licensee had completed final status surveys of approximately 73 percent of all survey units.
I 1.1 a.
b.
Self-assessment, Auditing, and Corrective Action (IP 40801)
Identification, Evaluation, and Resolution of Problems Inspection Scope The inspectors reviewed the licensees administrative procedures that control the identification, evaluation, and resolution of problems.
Observations and Findings The licensees program for assessing the resolution of non-conformances, material or programmatic~deficiencies, and conditions adverse to quality or safety remained as described in Section 1.2 of Inspection Report 050-00312/2006-002 and Section 1.2 of Inspection Report 050-00312/2007-002. There were no changes to the procedures of the Rancho Seco CAP since this program area was last inspected by the NRC in April 2007.
During CY 2007, 20 Potential Deviation from Quality (PDQ) reports were initiated and 11 were determined to be a Deviation from Quality (DQ). As of April 16, 2008, the inspectors noted that 13 PDQs had been initiated by the licensee during the calendar year with 6 determined to be DQs. No Corrective Action Requests or Stop Work Orders have been initiated in CY 2007 or as of the date of this inspection in CY 2008. The inspectors reviewed the list of PDQs and DQs
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Enclosure 1
that remained open. There were 15 items that remained open, including 8 opened in CY 2008. The oldest item remaining open was from CY 2005. The inspectors concluded that the licensee was appropriately addressing the timely resolution of these open items.
The inspectors reviewed agendas and minutes of Commitment Management Review Group (CMRG) meetings and noted that the CMRG had met, at a minimum, monthly as required by Rancho Seco Administrative Procedure (RSAP) RSAP-0260, Commitment Management Review Group & Compliance Management Tracking Systems. The CMRG membership was composed of the individuals described in RSAP-0260. The meeting records reviewed by the inspectors indicate that the CMRG was conducting initial reviews and providing characterizations of the new PDQs. Additionally, the CMRG was assigning tasks, establishing priorities, and reviewing proposed resolutions for PDQs and other identified problems. The inspectors concluded that the licensee was effectively maintaining the Rancho Seco CAP that was established by the licensee to control the identification, evaluation, and resolution of problems.
c.
Conclusion The licensee was effectively maintaining the CAP that was established by the licensee to control the identification, evaluation, and resolution of problems.
1.2 Quality Assurance Audit Organization, Staffing, and Qualifications a.
Inspection Scope The inspectors reviewed the licensees process of conducting Quality Assurance (QA) audits and surveillances. Additionally, the inspectors examined the status and composition of the QA audit organization, including the staffing and qualifications of individual members.
b.
Observations and Findinqs The inspectors noted that the composition of the licensees QA audit organization had not changed since the last inspection, conducted in April 2007, but some of the individual staff member in specific positions had changed. Of particular note, one of the lead auditors had left the organization and had been replaced by another full time auditor. The records reviewed by the inspectors indicated that additional auditors had been brought in as needed to conduct special audits.
Records also indicated that auditors with current lead auditor certifications had led all audits.
The inspectors reviewed records of audits completed since the last inspection of this area conducted by the NRC in April 2007. The licensee had conducted 15 audits in CY 2007 and 2 audits thus far in CY 2008. One additional audit not previously scheduled was in progress during the inspection. The licensee had a total of 10 audits scheduled for CY 2008. Audits conducted in CY 2006 - 2007, and those planned for CY 2008, addressed all 34 facility activities listed in Rancho Seco Quality Manual (RSQM),Section XVIII.
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Enclosure 1
In CY 2007, the licensee conducted a total of 32 audit surveillances. As of April 16, 2008, the licensee had conducted 15 surveillances thus far in CY 2008.
The inspectors selected audit reports 07-A-004, 07-A-01 4, 08-A-007, and surveillance reports 07-S-020, 07-S-027, 07-S-029, 08-S-05, and 08-S-11 for review. The inspectors confirmed that the audits and surveillances were conducted in accordance with the RSQM. Individuals that conducted the audits and surveillances were independent of the areas being audited. The auditors when conducting audits used approved checklists. The inspectors determined that the audit team personnel were qualified, were authorized by the licensee to perform the audits or surveillances in the areas audited, and the audits and surveillances conducted by the audit team had been conducted in a timely manner.
c.
Conclusion Audits conducted in CY 2007 and those scheduled for CY 2008 addressed all facility activities required to be audited. All auditors satisfied the qualification requirements.
Maintenance and Surveillance (IP 62801)
2.1 Inspection Scope The inspectors reviewed the status of required maintenance, surveillance, and testing. Inspectors interviewed the Maintenance Superintendent and examined selected records.
2.2 Observations and Findinas With the relocation of the spent fuel to the ISFSI, the licensee no longer had any safety-related structures, systems or components (SSC) as defined in 10 CFR 50.65(b)(1), nor any non-safety-related SSC as defined in 10 CFR 50.65(b)(2). The licensee had previously reviewed its Maintenance Rule procedure and concluded that it no longer was required to implement these activities at the site.
The inspectors reviewed a printout of the Rancho Seco Computerized Surveillance Schedule dated April 15, 2008, that listed the status of surveillances and routine tests. Ten routine tests or surveillances were beyond their due dates, but none were beyond the licensee's established grace period. When questioned about the overdue tests or surveillances, the Maintenance Superintendent indicated that he did not expect any of the surveillances or tests would exceed their grace period.
2.3 Conclusion The licensee no longer had any SSCs, and the requirements of the Maintenance Rule were no longer required to be implemented at this site. The licensee was maintaining the schedules on other surveillance and routine testing activities as required.
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Enclosure 1
3.1 3.2 Decommissioning Performance and Status Review (IP 71 801)
Inspection Scope The inspectors interviewed personnel, reviewed selected documents, and toured portions of the site to observe work activities including housekeeping, safety practices, fire protection practices, and radiological controls. An NRC inspector observed the preparations and was present during the demolition of the reactor building polar crane.
Observations and Findinqs The inspectors conducted tours of the reactor, auxiliary, fuel handling, and turbine buildings and observed dismantlement and decommissioning activities in progress. Decommissioning work observed by the inspectors during the tour was conducted in a safe and orderly manner. The inspectors conducted independent radiological surveys using a Ludlum Model 2401 -EC survey meter (NRC No. 21 175G, calibration due date July IO, 2008). Radiological controls, including postings and barriers, were observed by the inspectors to be in place.
The reactor building concrete and steel removal project was in progress. As of April 5, 2008, the contractor had completed approximately 97 percent of the demolition. Approximately 28.7 million pounds of concrete and steel had been packaged for shipment with approximately 28.6 million pounds shipped. These shipments had been made in 146 railcars assembled into 26 rail pickups. The decommissioning project continues with the licensees and contractors staff working two ten-hour shifts Monday through Friday.
On Saturday, April 26, 2008, the reactor building polar crane was brought down using explosives. An NRC inspector was dispatched and was onsite during the final preparations and the detonation of the explosives. Following completion of the roll call, a count down was begun and the detonation began at about 2:20 PM P.D.T. Following the detonation, radiation monitoring teams were dispatched to various locations and after about 20 minutes an all clear signal was broadcasted by the licensee indicating that no radioactivity above background had been detected in either the air samples or ambient portable radiation surveys. The plant manager confirmed this information during subsequent discussions with the NRC inspector onsite during the detonation.
When questioned by the inspectors about the status of the reactor building concrete and steel removal project, the licensee indicated that the project was expected to be completed by the end of May 2008. The licensees schedule anticipates the last of the final status surveys will be completed by the end of September 2008 and the last final status survey report will be submitted to the NRC by the end of October 2008.
In the fuel handling building, remediation continued on the building walls. When questioned about the schedule, the licensee projected that it would complete remediation of the fuel handling building by April 30, 2008.
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Enclosure 1
3.3
4.1 a.
b.
C.
4.2 a.
As of April 18, 2008, the licensee had completed final status surveys on 217 of a projected 295 survey units. The number of projected survey units had changed since the last inspection as some units were split into multiple units. Based on discussions with the licensee, the 21 7 completed survey units constituted approximately 73.6% of the total projected survey units.
Conclusion The licensee continued to dismantle and remove contaminated components and to remediate contaminated surfaces in a safe manner. Final status surveys had been completed on 217 of a projected 295 survey units.
Occupational Radiation Exposure (IP 83750)
Audits and Surveillances Inspection Scope The inspectors reviewed the recently conducted radiation safety audit report to verify implementation of the commitments made in Section XVIII, Audits, of the RSQM, as it relates to occupational radiation safety. The inspectors reviewed the qualification records for the individuals involved in conducting the radiation safety audit.
Observations and Findings The inspectors reviewed the licensees audit Report 07-A-01 4, Radiological Safety and Control and ALARA Program. The audit was conducted from November 21 through December 20,2007, and the report was issued on January 9, 2008. The inspectors reviewed the audit report, confirmed that the audit was conducted according to the commitments in the RSQM, determined that the individual who conducted the audit was independent of the function being audited, and concluded that the lead auditor was qualified and authorized to perform the audit in the areas audited. The auditors used an approved checklist when conducting the audit. The inspectors observed that the audit was conducted in a timely manner.
Conclusions A licensee audit of the occupational radiation exposure program was conducted in accordance with QA program requirements.
External and Internal Exposure Control and Other Radiation Protection Inspection Areas Inspection Scope The licensees personnel radiation monitoring program was inspected for compliance with applicable requirements and commitments.
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Enclosure 1
b.
Observations and Findings The licensee was continuing its use of Optically Stimulated Luminescent (OSL)
dosimeters for evaluating beta/gamma external doses, and neutron dosimeters for neutron dose. The OSL dosimeters were provided by a vendor that was accredited under the National Voluntary Laboratory Accreditation Program for the type of dosimeters used. In addition, the licensee used Electronic Dosimeter (EDs) for controlling the day-to-day personnel exposures. The licensee continued to use a vendor-supplied computerized dose tracking system for reading the EDs and for automatically assigning the estimated dose to the individual.
The Radiological Health Supervisor stated that during CY 2007 no individual had received an internal exposure that required a committed effective dose equivalent be assigned.
During CYs 2007 and to the date of this inspection in CY2008, no individual had been classified as a declared pregnant worker, and no planned special exposures had been conducted.
The annual report for CY 2007 radiation exposures required by 10 CFR 20.2206(b) was in final preparation during the site visit. The Radiological Health Supervisor stated that the annual report will indicate, when completed, that the Total Effective Dose Equivalent (TEDE) received by each occupationally exposed individual would be below the regulatory limit of 5 rem. A review of occupational dosimetry records indicated that the highest reported TEDE in the summary report was 1.620 rem. The doses to lens of the eye, skin of the whole body, internal, and skin of the maximally exposed extremity were all below applicable 10 CFR Part 20 limits.
The required Regulatory Guide 1.16 Annual Exposure Report for CY 2007 was submitted on March 25, 2008. The report was submitted to the NRC in a timely manner before the April 30, 2008, deadline.
c.
Conclusions The licensee maintained an effective program for monitoring occupational radiation exposures. Occupational exposures for CY 2007 were noted by the inspectors to be below regulatory limits specified in 10 CFR Part 20.
4.3 Emergency Plan a.
Inspection Scope The inspectors reviewed the circumstance related to a recent event when a Notification of an Unusual Event (NOUE) was declared by the licensee.
b.
Observations and Findings On Saturday April 5, 2008, the demolition subcontractor was using a cutting torch to remove a non-structural beam in the reactor building. Slag from the cutting
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Enclosure 1
operation started a fire in combustible material concealed below concrete rubble.
The fire watch detected the fire at approximately 10:20 AM PDT. At 10:30 AM PDT the on-shift ISFSI technician (the senior licensee person on site during weekends and backshift) was informed of the fire and promptly responded to the scene. The ISFSI technician took immediate action to evacuate the area, improve efforts to extinguish the fire, turned off the reactor building ventilation, and ensured that air samples were being taken to assess potential airborne radioactivity.
As a result of the fire, smoke escaped from the reactor building access hatch.
The Radiation Protection Technician on-duty monitored the continuous air sampler located at the reactor building hatch for airborne radioactivity. The air samples taken at the access hatch and in other locations in the reactor building revealed no contamination above background.
Due to the difficulty encountered by ISFSI technician in observing the fire that was concealed by concrete rubble, it was not possible for the Radiation Protection Technician to determine precisely when the fire was extinguished.
The licensee estimated that the fire was extinguished sometime between 1 I :30 AM and 12:30 PM PDT.
Due to the inability to contact offsite management, and as a result of the continuing smoke, at approximately 11:40 AM PDT, the ISFSI technician declared a NOUE in accordance with Emergency Plan Implementing Procedure (EPIP), EPIP-01 Emergency Action. The ISFSI technician took this action on the provision included in the procedure that allows discretionary declaration of an event if the event does not fall into one of the predefined conditions.
When the NOUE was declared by the ISFl technician, communications errors occurred due the licensees failure to use the EPlP checklist. Not using the EPlP checklist resulted in a failure of the licensee to make notifications to the County, State, and NRC within one hour of declaring a NOUE; use the licensee pager callout notification of the Rancho Seco on-call Emergency Response Organization (ERO); and make proper notification to the licensees management.
Following the declaration of the emergency, the ISFSI technician was able to contact members of the Rancho Secos management team and at that time a decision was made to conduct a conference call. Individuals participating in the conference call included the ISFSI Technician and Rancho Secos on-call management and corporate safety, fire protection, security, and emergency planning staff. During the conference call, the plant status was reviewed, the need for additional actions was evaluated, weekend work in the reactor building was stopped, and a continuous fire watch posted until Monday (April 7, 2008)
morning. Following the conference call, the ISFSI technician closed out the NOUE but did not use the requisite EPlP closeout checklist.
On Monday, April 7,2008, Rancho Seco management met and determined that a NOUE had been both declared and closed out on April 5, 2008. Accordingly, on April 7, 2008, the Rancho Seco personnel notified the NRC Operations Center, as well as the State and County.
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Enclosure 1
In accordance with the licensees corrective action program two PDQs, 08-01 0 and 08-01 1, were initiated and subsequently classified as DQs. A post event critique and investigation interviews were conducted and a detailed investigation report was prepared by the licensee. On May 1, 2008, the licensee submitted a 30-day follow-up report to the NRC for the NOUE that occurred on April 5, 2008.
This report identified three fundamental issues and associated causes.
e The fire in the reactor building occurred due to inadequate housekeeping practices that allowed combustible materials to accumulate and be covered up by concrete rubble thus setting the stage for a slag-induced fire.
e Training and drills had not adequately emphasized backshift implementation of the Emergency Plan, thus once the NOUE was declared the Emergency Plan was not properly implemented.
e No formal process existed to rapidly contact management during backshifts and weekends in a non-Emergency Plan situation. The inability of the ISFSI technician to contact offsite management in a timely manner lead the individual to declare a NOUE as a discretionary measure.
The licensees report described the corrective actions taken and those that the licensee had scheduled to be completed. Completed actions included:
1.
Removing the combustible debris from hot work areas having the potential for a fire starting; 2.
Reviewing the procedures, training, preventive measures and the specific documentation related to housekeeping and control of combustibles near hot work areas; 3.
Providing additional training to ISFSI Technicians, emergency response organization members, and Security Shift Supervisors on the Emergency Plan, EPIPs, emergency classification, importance of clear and explicit communications, proper use of required checklists, and mandating that these individuals pass a proficiency test on the emergency plan requirements; 4.
Providing additional training to ISFSI technicians on procedure compliance and the importance of following procedures; and 5.
Revising the applicable procedure to direct the ISFSI technician to call for outside agency assistance for any fire that cannot be extinguished with a portable fire extinguisher. Additionally, when the procedure is revised provide training of the ISFSI technicians.
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Enclosure 1
Actions to be completed included:
1. Develop backshift Emergency Plan implementation checklists, train ISFSI technicians on the proper use of the checklist, and perform an unannounced backshift drill that requires the use of the checklist; 2. Emphasize backshift Emergency Plan implementation in training and drills; and 3. Develop a formal process to notify management in a non-Emergency Plan situation and provide the requisite training to the staff.
Section 50.54(q) of Title 10 Code of Federal Regulations (CFR) requires in part that a holder of a nuclear power reactor operating license shall follow and maintain in effect emergency plans that meet the standards in Section 50.47(b)
of this Title. Contrary to the above on April 5, 2008, the licensee failed to follow their Emergency Plan in that an Emergency Classification (NOUE) was declared and the offsite notifications to the County, State and NRC where not made and the licensees on-call emergency organization was not activated. This non-repetitive, licensee identified and corrected failure satisfies the criteria as a non-cited violation (NCV 050-0031 2/0801-01).
c.
Conclusions A non-cited violation related to the licensee failure to follow the Emergency Plan was identified.
Inspection of Final Surveys (IP 83801)
5. I Inspection Scope Independent confirmatory radiological measurements were performed by ORISE on surfaces of the auxiliary building.
5.2 Observations and Findings On April 12, 2006, the licensee submitted their License Termination Plan (LTP) to the NRC. This LTP included proposed Derived Concentration Guide Lines (DCGLs) for meeting the public dose limits after license termination. On November 27, 2007, the NRC issued License Amendment Number 133 that approved the licensees LTP and the respective DCGLs.
Representatives from ORISE, working as the NRCs contractor, reviewed records of final status surveys taken in embedded piping and on surfaces in the auxiliary building. At the NRCs request, ORISE personnel conducted independent confirmatory radiological measurements of selected locations including embedded piping and compared their survey results with those of the licensee.
The results of these surveys will be reported to the licensee at a later date under separate correspondence. The ORISE staff also collected four soil samples in the area of the basement of the auxiliary building called pump alley. Two of the soil samples were counted by the licensee during the inspection and base on
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Enclosure 1
these results the licensee decided that this area required additional remediation.
After analysis of the soil samples by ORISE, the soil samples were returned to the licensee.
The results of survey activities conducted by ORlSE staff during December 10 through 13, 2007, inspection were documented in a report issued on March 12, 2008. A copy of that report is attached. With the exceptions of the elevated direct reading measurement that occurred in Room 18 of the Auxiliary Building and also in a small area near where the Regenerate Holdup Tank, the surveys taken by ORISE confirmed the accuracy of the licensees final status surveys for the locations surveyed. These elevated readings were discussed in Section 5 of Inspection Report number 050-0031 2107-006.
5.3 Conclusion The ORlSE staff conducted confirmatory measurements on selected surfaces of the auxiliary building, including embedded piping. The results of the ORISE surveys conducted during this inspection will be reported at a later date.
The results of survey activities conducted by ORISE during the December 10 through 13, 2007, site visit were documented on a report issued on March 12, 2008. A copy of this ORlSE report attached as Enclosure 2. With the exceptions of the elevated direct reading measurement identified by ORISE in two locations, the ORlSE surveys confirmed the accuracy of the licensees final status survey results for the locations surveyed.
Exit Meeting Summary At the end of the site visits on April 17, 2008 and April 26, 2008, the inspectors briefed the Plant Closure and Decommissioning Manager (Plant Manager) and other members of licensee staff regarding the preliminary inspection findings. A telephonic exit briefing was conducted with the Plant Manager on May 19, 2008.
The licensee did not identify any information provided to, or reviewed by, the inspectors as proprietary.
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PARTIAL LIST OF PERSONS CONTACTED Sacramento Municipal Utilitv District J Astronom, ISFSI Technician M. Bua, Radiation Protection/Chemistry Superintendent W. Hawley, Dismantlement Superintendent - Operations L. Hoist, Nuclear Document Control Supervisor R. Jones, Supervising Quality Engineer D. Koontz, ISFSI Supervisor S. Redeker, Manager, Plant Closure and Decommissioning (Plant Manager)
J. Roberts, Maintenance Superintendent E. Ronningen, Dismantlement Superintendent - Radiological D. Ward, Quality Assurance Auditor INSPECTION PROCEDURES USED IP 40801 IP 62801 Maintenance and Surveillance IP 71801 IP 83750 Occupational Radiation Exposure IP 83801 Self-assessment, Auditing, and Corrective Action Decommissioning Performance and Status Review Inspections of Final Surveys ITEMS OPENED, CLOSED, AND DISCUSSED Opened 050-00312/0801-01 NCV Failure to follow their Emergency Plan in that an Emergency Classification was declared and the offsite notifications to the County, State and NRC where not made and the on-call emergency organization was not activated.
Closed 050-00312/0801-01 NCV Failure to follow their Emergency Plan in that an Emergency Classification was declared and the offsite notifications to the County, State and NRC where not made and the on-call emergency organization was not activated.
Discussed None
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Attachment 1
LIST OF ACRONYMS CFR CMRG CY DCGLs DQ ED EPlP ERO IP ISFSI LTP NCV ORISE OSL PDQ PSDAR QA RSAP RSQM ssc TEDE Code of Federal Regulations Commitment Management Review Group Calendar Year Derived Concentration Guide Lines Deviation from Quality Electronic Dosimeter Emergency Plan Implementing Procedure Emergency Response Organization Inspection Procedure Independent Spent Fuel Storage Installation License Termination Plan Non-C ited Vi0 lat ion Oak Ridge Institute for Science Education Optically Stimulated Luminescent dosimeters Potential Deviation from Quality Post Shutdown Decommissioning Activities Report Qua I ity Assurance Rancho Seco Administrative Procedure Rancho Seco Quality Manual Structures, Systems Or Components Total Effective Dose Equivalent
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Attachment 1
PARTIAL LIST OF DOCUMENTS REVIEWED Audits and Surveillances
.
Audit Log, 2007.
.
Audit Log, 2008 as of April 16, 2008.
.
Audit File, Audit 07-A-004, Emergency Preparedness Program, performed May 2-22, 2007.
.
Audit File, Audit 07-A-01 4, Radiological Safety and Control; ALARA, performed November 21 through December 20, 2007.
.
Audit File, Audit 08-A-007, Process Control Program + Packaging &
Transportation of Radioactive Waste, performed March 11-20, 2008.
.
Surveillance Log, 2007.
.
Surveillance Log, 2008 through April 16, 2008.
.
Surveillance report 07-S-020, approved August 21,2007
.
Surveillance report 07-S-027, approved November 28, 2007
.
Surveillance report 07-5-029, approved December 10, 2007.
.
Surveillance report 08-S-005, approved February 14, 2008.
.
Surveillance report 08-S-011, approved March 13, 2008.
Correspondences and Memorandums
.
Correspondence dated May 1, 2008, from Steve Redeker, Manager, Plant Closure and Decommissioning to U. S. Nuclear Regulatory Commission, Attention John Hickman, subject Special Report Regarding a Fire in the Rancho Seco Reactor Building.
Reports
Master Surveillance Schedule by Due Date, dated April 15, 2008.
a Master Routine Tests Schedule by Department, dated April 15, 2008.
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Attachment 2
INTERIM LETTER REPORT CONFIRMATORY SURVEY RESULTS FOR ACTIVITIES PERFORMED IN DECEMBER 2007 RANCHO SECO NUCLEAR GENERATING STATION HERALD, CALIFORNIA INTRODUCTION The Sacramento Municipal Uallty District (SMUD) operated the Rancho Seco Nuclear Generating Station (RSNGS) from 1976 to 1989 under Atoinic Energy Coininission Docket Number 50-312 and License Number DPR-54. In August 1989, SMUD notified the U.S. Nuclear Regulatoiy Commission (NRC) that they shut down RSNGS permanently. In May 1991, SMUD submitted the Rancho Seco Decommissioning Plan wlvch was approved by the NRC in March 1995. SlClUD began decommissioning activities in Februaiy 1997 and completed transfer of dl the spent nuclear fuel in August 2002 (SMUD 2006a).
In April 2006, SMUD submitted a license termination plan (LTP) that was recently approved by the NRC on November 26,2007 (SMUD 2006a and NRC 2007). SibIUD currently is conducting decontamination efforts and performing final status surveys (FSS) on the remaining structural surfaces and in open land areas.
The NRC requested that the Independent Environmental Assessment and Verification (IEAV)
Program of the Oak Ridge Institute for Science and Education (ORISE) perform confirmatory surveys of structural surfaces and the Acid Waste System drains in the Auxhai-y Bddmg at the RSNGS (Figures 1 and 2). Wlde on site, the NRC site representative also requested that ORISE perform cursory gainma surface scans and collect soil samples at the Outfall Area and the Regenerant Hold-Up Tank Audai-y Boiler Land Area (RHUT Area). The confirmatory surveys were performed during the period of December 10 through 13,2007.
PROCEDURES Confumatory surveys were performed in accordance with a site-specific survey plan that was submitted to and approved by the NRC (ORISE 2007a). The site-specific survey plan follows the guidance provided in the IEAV Survey Procedures and Quality Program Manuals (ORISE 2007b and ORAU 2007).
In the AudaiTT Burlding, ORISE judgmentally selected four of the eight vaults (Figures 3 through 5), three rooms (Figures 6 through S), two Acid Waste System drains (Figures 9 and lo), and various pipe penetrations for confirinatoiy surveys based upon preliminary FSS results. At the request of the NRC site representative, ORISE also performed limited ra&ological surveys of the Outfall and RHUT Areas (Figures 11 and 12).
SURFACE SCANS Auxiliary Buildirw Structural Surfaces G a m a surface scans were performed using sodium iodide, thalhum-activated PaI(Tl)] gainma scindlation detectors coupled to ratemeters with audble indicators. Beta surface scans were performed using large area gas proportional, hand-held gas proportional, and Geiger-Muller (GY I<mcho Scco Nuclear Gcncrating Station 1695-SR-02-0
detectors coupled to ratemeter-scalers with audible indicators. Particular attention was given to cracks, joints, embedded piping openings and horizontal surfaces in the evaluated structural surfaces where material map have accumulated.
Drains and PiDe Penetrations Liinited qualitative garmna scans were performed in portions of two 2" inner diameter (ID) Acid Waste Systems drain lines on the -20 foot level elevation and in numerous pipe penetrations witlvn the vaults and rooms that were part of these suivey activities. ORISE recorded the gamma scan range for the Acid Waste System drains and the pipe penetrations. Gamma scans were performed using a cesium iolde, thahuin-activated [CsI(Tl)] gaimna scintillation detector coupled to a ratemeter with an audible indicator.
Outfall and RHUT Areas Gamna scans of the soils withn the excavated Outfall Area and immediately adjacent to the excavated area and within the RHUT Area were performed using a NaI(T1) gainma scinullation detector coupled to a ratemeter with an audble indicator.
SURFACE ACTIVITY MEASUREMENTS Based on beta and garmna surface scan results, direct measurements for beta activity were performed at 26 judgmentally-selected locations on the evaluated stivctural surfaces within the Auxiliaiy Budding which were available for confumatoiy suivey activities. Direct measurements were performed using hand-held gas proportional detectors coupled to ratemeters-scalers. A smear sample for determining removable gross alpha and gross beta activity levels was collected from each drect measurement location. Direct ineasurement and smear locations are indicated on Figures 4 through 8.
SOIL SAMPLING Based on gaimna scan results, ORISE judgmentally collected two soil samples from the Outfall Area; one sample to the east and one sample to the west, both immediately adjacent to the remedated portion of the Outfall Area. Also based on ORISE gamma scan results, ORISE asked SMUD personnel to collect a soil sample from an area of elevated gainma radiation (suspected to possibly be a discrete particle) in the RHUT Area and to provide analytical results to ORISE.
SAMPLE ANALYSIS AND DATA INTERPRETATION Ralological data and sample media were returned to the ORISE laboratoiy in Oak Ridge, Tennessee for analysis and interpretation. Radioassays were performed in accordance with the ORISE Laboratoiy Procedures Manual (ORISE 2007~). The soil samples were analyzed by gaimna spectroscopy for the priinaiy ralonuclides-of-concern (ROC), CO-60 and Cs-137. However, spectra were also reviewed for additional gamna-emitting fission and activation products associated with the RSNGS and other identifiable total absorption peaks. The soil sample results were reported in units of picocuries per grain (pCi/g). Smear samples were analyzed for gross alpha and gross beta activity using a low-background gas proportional counter. Sinear results and direct measurements for total surface activity were converted to units of disintegrations per ininute per 100 square centimeters (dpm/100 cm'). Since a pipe detector efficiency was not required,
Rancho Scco Nuclear Gcncrating Station 1695-SR-02-0
embedded piping scan data were reported in units of counts per minute (cpm) to compare with SMUDs gross gamma cpm results.
FINDINGS AND RESULTS AUXILIARY BUILDING STRUCTURAL SURFACES The scan percent coverage and room area classifications are provided in Table 1. Beta surface scans determined that localized areas of residual elevated beta-gamma surface activity were present on floor and lower wall surfaces witlin the evaluated suivey units (SUs). TVitli tlie exception of duect measurement Location 14 in Rooin 18, residual surface activity was hiited to small areas that were interspersed throughout the rooms. Due to the elevated beta and gamma surface activity levels determined at Location 14, SMUD and NRC personnel were notified and SMUD personnel remediated the location.
Beta measurements were performed at locations of residual elevated beta-gamma surface activity detected during surface scans. With one exception, total net beta activity measurements ranged from 130 to 16,000 dpm/100 cm2. The one exception was at Location 14 in Room 18, svlich had a total net beta activity of 110,000 dpm/100 cm2. It was deteiinined that the residual beta activity was due to a discrete particle (0.22 pCi/g of Cs-137 and 0.0008 pCi/g of CO-60); the discrete particle was reinoved and tlie post-remedial total net beta activity was 1,100 dpin/100 cm2. Removable gross alpha and gross beta activity ranged from 0 to 5 and -3 to 8 dpin/100 cm. Surface activity and removable activity level results are presented in Table 2.
DRAINS AND PIPE PENETRATIONS A comparison of ORISE and SMUD gamma scan results for the Acid Waste System Drain Line 4-2-1 5 indcated elevated gaimna radiation levels at approximately tlie same length/depth and levels as reported by SMUD personnel in the prehniiiaiy final status sui-vey (FSS) data packages.
However, gamma scans of Acid Waste System Drain Line 4-1-12, Segment 1 (originally iiumbered incorrectly by SMUD as Drain Line 4-1-13) indmted a large discrepancy between ORISE and SMUD data in the gamma scan range for the 0 to 8 foot portion of the pipe. SMUDs results, although more conseivative than ORISEs results, were thirteen times higher than the ORISE results. The reason for tlis discrepancy is unclear and ORISE recoinmends further evaluation of tlis drain line during a future confumatoi7 suivey trip. The ORISE confirmatoq and SMUD gaixtna scan ranges are provided in Table 3.
Gamma scans of numerous pipe penetrations within tlie evaluated SUs indicated that gamma rahation levels ranged from 200 to 1,200 cpin. For coinparison, the CsI(T1) detector background range for tlie conduits along the east side of the Turbine B d d n g at the +40 level elevation ranged from 200 to 800 cpin. The ORISE gaimna scan ranges for the pipe penetrations are also provided in Table 3.
OUTFALL AND RHUT AREAS Gamma scans of the remediated portion of the Outfall Area and the immediately adjacent areas detected residual elevated gamma radiation levels to the west and east of the remediated portion of Electronic mail from G. Pillsbury (SMVD) to W. h d m s (OIUSE); January 14,2005.
I
Ilnncho Seco Nuclear Generating Station
the Outfall Area. A soil sample was collected from each of these locations. The radionuclide concentrations for the two soil samples collected immediately adjacent to the remedated portion of the Outfall Area ranged from 0.32 to 0.49 pCi/g for co-60 and 34.9 to 47.1 pCi/g for Cs-137. The confirmatoqi radionuclide concentrations for tlie soil samples are provided in Table 4.
Gainma scans of the RHUT Area indicated that the vast majority of the surface soil was at background levels. The one exception was along the west perimeter of the area where a small location indicated elevated gamma radiation levels at approximately 40 times the average background. SMUD and NRC personnel were notified of tl3s finding and SMUD personnel remediated the location and collected a soil sample. SMUD analyzed the RF-IUT soil sample and determined that the sample contained 66 pCi/g of Co-GO. The sample was further divided and SMUD isolated a dmrete particle that was counted as a point source and reported the particle activity as 0.485 pCi of CO-60; the recount of the soil sample (minus the discrete particle) indicated background levels.
COMPARISON OF SURVEY RESULTS WITH GUIDELINES STRUCTURAL SURFACE ACTIVITY LEVELS The major containinants identified by SMUD at RSNGS are beta-gamma einitters-fission and activation products-resulting from reactor operation. Cesium-137 and CO-60 have been identified during characterization as the predominant radionuclides present on structural surfaces. SMUD developed site-specific derived concentration guideline levels (DCGLs), whch were recently approved by the NRC, based on a dose modeling to future occupants not to exceed 25 mrem/pear total effective dose equivalent (TEDE) as presented in Section 6 of the LTP (SMUD 2006a and NRC 2007). The DCGLs for surfaces were mockfied by SMUD to reflect the ratio of radionuclide concentrations (account for the presence of unmeasured containinants based on contaminant ratios)
in the specific SUs that were being evaluated. The applicable surface activity guidelines for the evaluated structural surfaces for these surveys are provided in Table 5. These DCGLs were provided in the preliminaiy FSS data packages for each evaluated SU and were derived from the LTP and decommissioning teclinical basis document (DTBD)-05-015 (SMUD 2006a and b).
Confuinatoiy survey data for Auxilmy Bullding structural surfaces were compared with the site-specific DCGL for the evaluated Auxiliary B d l n g SUs. One of the 26 direct beta activity measurement results on the concrete structural surfaces exceeded the Gross Beta DCGL of 43,000 dpin/l00 cin. Using the gross activity DCGL as determined in DTBD-05-015 (SMUD 2006b) and the area factor deterinined for each SU, SMUD calculated Design DCGL elevated measurement comparison (DCGLI,,,,) values which are also provided in Table 5. All confirmatory duect surface activity measurements on tlie Audtary Budding structural surfaces in the evaluated SUs were wirhm the site-specific SU DCGL,,,, as provided by SMUD in the prehninaiy FSS data packages. However, it was determined that the elevated beta surface activity at Location 14 in Room 18 was from a discrete Cs-137 and Co-60 particle; hence, the particle was remediated by SMUD personnel while ORISE was on site and a confirmatoq7, post-remedation direct measurement was performed with the results being well within the gross activity DCGL.
Electronic m d from E. Ronningen (SibIUD) to W. Adams (ORISE), RE: RHUT Area Contamination; December 30,2007.
Rancho Scco Nuclcar Gcncwting Station
DRAIN LINES AND PIPE PENETRATIONS Co-60 is the primaiy ROC witlin the embedded piping. SMUD has established a dose-based restriction for embedded piping not to exceed 25 mrem/year that assumes a budding occupancy scenario witlin rooms where embedded piping is present. The correspondmg modeled DCGL is 100,000 dpin/100 cin. SMUDs grouting action level for embedded piping is 21,000 dpm/100 cin (SMUD 2007).
ORISEs confirinatoiy drain line and pipe penetration results were not directly coinpared to the embedded piping DCGL; instead, since ORISE and SMUD used the same Ludlum Model 41-159 CsI(T1) detector, ORISE coinpared gross gamma scan readings with either SMUDs prehninai77 FSS data package ganxna scan results for each sunreyed pipe at various depths or with background levels as determined during a previous ORISE confirinatoi7r sumey (ORISE 2007d).
Confrrmatoiy suivey data for the Acid Waste System drain lines were coinpared with the preliminaiy FSS data package gross gainma cpm results. The confirmatoiy gaiuna scan results indcated that ORISE gross gamma radiation levels within the drain line pipes were consistent with the SMUD p r e h n a r y FSS data package results for Drain Line 4-2-15. SMUDs gross gainma cpin results for Drain Line 4-1-12, Segment 1 were thirteen times higher than the ORISE gross gainma cpm results; the results for this drain line need to be re-evaluated during the next ORISE confirmatoiy suivey.
Gamma scans of the other evaluated Auxlliaiy Budding pipe penetrations that were part of these suivey activities d d not detect gainma radiation levels in excess of the detector background.
SOIL SAMPLES Table 6-5 (Table 6) from the LTP provides the single nuclide DCGLs for soil at RSNGS. The DCGL,v is 12.6 pCi/g for Co-60 and 52.8 pCi/g for Cs-137 (SMUD 2006a). The Outfall Area soil sainple concentrations were below the respective single radionuclide DCGLs. The soil sainple collected by SMUD persoiinel from the location of elevated residual gamma radiation detected by ORISE personnel in the RHUT Area was analyzed by SMUD; SMUDs analyses indicated that the sample (containing 66 pCi/g of CO-60) exceeded the DCGL for Co-60. The sainple was further divided and a dscrete particle containing 0.485 pCi of CO-60 was identified by SMUD.
SUMMARY During the period of December 10 through 13,2007, ORISE performed confumatoiy radiological suivey activities which included beta and gainma surface scans, beta activity direct measurements, and removable gross alpha and gross beta activity ineasureinents on structural surfaces within the A~ixlliaiy B d d n g ; gainma scans withn A u d a i y Budding embedded piping; and, gainma scans and the collection of soil samples from the Outfall and RHUT Areas.
Beta and gamma surface scans identified several areas of elevated beta surface activity on the structural surfaces of the evaluated SUs with the Auxiltaqi Building. With one exception, additional investigation of these locations indicated that the majority of the elevated surface activity levels were attributable to localized areas of residual beta-gamma radiation within the matrix of the concrete inela. In general, the elevated surface activity was hnited to small areas that mere interspersed throughout the rooms. The one exception was a discrete particle of Cs-137 and co-60 that was found in Room 18. Direct measureinents were performed at 26 locations. As mehtioned above, Rancho Seco Nuclcnr Generating Station
1695-SI<-02-0
only one du-ect measurement esceeded the site-specific gross beta DCGL but all were witlvn the DCGL,,,, criteria. A review of the prelirnnai-p FSS data package for Room 18 indicated that SMUD personnel did not identify the elevated residual beta and gainma radation levels from Location 14 in Room 18. Therefore, the results of the confirmatory suivey activities for the evaluated structural surfaces of Rooin 18 in the Audiaqr Building I d not confirm the ralological status of the SU as presented in the licensees p r e h n a i y Room 18 FSS data package. SMUD personnel were notified and are investigating the confirmatoiy finding in Room 18. The confirmatoiy suivey results for Rooins 50 and 53 are in agreement with the ralological status of these SUs as presented in the licensees prehninaiy FSS data packages.
Gamma surface scans of the evaluated Auxiliaiy Buillng pipe penetrations I d not indicate any areas of elevated gamma radiation levels; the scan results within Acid Waste System Drain Line 4-2-15 were consistent with the results presented in the prehninaiy FSS data packages for that drain.
However, the gamma scan results for Acid Waste System Drain Line 4-1-12, Segment 1 were not in agreement with the prehinaiy FSS results. Further investigations by SMUD indicated that Drain Line 4-1-13 had been removed and the prehninaiy FSS results for Drain Line 4-1-12, Segment 1 were erroneously reported as the results for 4-1-13 (Refer to Figure 9). Due to this discrepancy, SMUD re-surveyed Segments 1 and 2 of Drain Line 4-1-12 and provided p r e h n a i y results to ORISE on February 12, 200i3 SMUDs gamma scan range for the 0 to 8 foot length of Segment 1 is thvteen times higher and not in agreement with ORISEs gamma scan results (Table 3). Based on the lscrepancy between the ORISE confirmatoiy and SMUD FSS gamma scan ranges and the confusion with the incorrect numbering of the drain line, ORISE recommends that a side-by-side instrument comparison be performed in tlvs drain line during a subsequent confirinatoiy suivey trip.
The soil sample results from the Outfall Area were below the individual ralonuclide DCGLs and ineet the soil release criteria. The SMUD radiological analyses of the soil sample from the RHUT area exceeded the soil DCGL for Co-60. SMUD personnel are performing further investigations witlvn the RHUT Area.
Electronic mail from E. Roiiningen (SMUD) to W. Adams (ORISE), RE: Survey of Acid Waste Line 4-1-12;
February 12,2008.
Ilancho Scco Nuclear Gencrating Scation
Rancho Scco Nuclear Generating Station FIGURES
1695-001 (2)
\\ \\
N f
NOT TO SCA1.E Figure 1: Location of Rancho Seco Nuclear Generating Station, Herald, California 16)5-S1L02-0
Rancho Seco N L I C ~ ~ X Gcnerating Station
169.5-002 (7)
I COOLING TOWERS TURBINE BUILDING FUEL STORAGE RUILDING REACTOR BUILDING
LTECHNICAT, CENTE WAl'LR 'I'UAYMENT PLAh'l'
i I
i ADMINIS TRAI'10N 1 BUILDING RLJILDINC T d R 131JILDINO I
NOT TO SCALE Figure 2: Plot Plan of the Industrial Area at Rancho Seco Nuclear Generating Station
1695-SR-02-0 Itancllo Scco Nuclear Gcncrating Station
Igure provided by SMUD.
N A
NOT TO SCALE Figure 3: Plot Plan of Auxiliary Building Vaults - Survey Units F8130401 and F8130411 1695-SR-02-0
Rancho Seco Nuclear Gcncrating Station
MEASUREMENT/SAMTLE LOCATION
@I Single Point - ~ o o r A
Single Point - Lower \\vau Figure 4: Survey Unit F8130401, Vaults 30 and 31 - Direct Measurement and Sample Locations
1695-SR-02-0 Rancho Seco Nuclcar Gcncrating Station N -2'L NOT T O SCALE 11- -
\\ ! - * \\
I
@re provided by SMUD.
MEASUEMENT/SAMPLE LO CAT1 0 N
@ Single Point - lo or A
Single Point - Lower Wall A
Single Point - Upper Surfaces Pipe Penetration Figure 5: Survey Unit F8130411, Vaults 34 and 35 - Direct Measurement and Sample Locations
1695-SR-02-0 Rancho Seco Nuclear Gcnerating Stntmn
I Figure Provided by SMUD MEASUREMENT/SAMPLE LOCATION
@ Single Point - ~1001
- E@: P
.......
.
. _...
.
I N ---+
--..IC NOT TO SCALE Figure 6: Survey Unit F8130201, Room 18 - Direct Measurement and Sample Locations 1695-SR-02-0
Rancho Seco Nuclear Generating Station
'igure provided by SMUD MEASUREMENT/SAMLE LOCATION
@ Single Point - Floor N --</b--4 NOT TO SCALE Figure 7: Survey Unit F8130681, Room 50 - Direct Measurement and Sample Locations 1695-SR-02-0
Rancho Scco Nuclear Gencrntmg Station
N n *
b W ( 0
'igire provided by SMUD.
TVIEASUREMENT/SAMPLE LOCATION
@ Single Point - Floor N -=a-1- -<
NOT TO SCALE Figure 8: Survey Unit F8130781, Room 53 - Direct Measurement and Sample Locations 1 m - s i ~ o 2 - r )
Rancho Scco Nuclear Generating Station
pure pro
- Surveyed Drain MEASUREMENT LOCATION NOT TO SCALE 1695-SR-02-0
Itancho Scco Nuclear Generating Station
'&ire provided by SMUD.
MEASUREMENT LOCATION
- Surveyed Drain NOT TO SCALE Figure 10: Survey Unit F8990521, Acid Waste Drains - Surveyed Drain 4-2-15 1695-SI{-02-0
Rancho SCCO Nuclcar Generating Station
'igure provided by SMUD.
SOIL SAMTLE LOCATION
- Soilsample-2LB4-N NOT TO SCALE Figure 11: Outfall Area - Soil Sample Locations
1635-SR-02-fl Ranclio Sccn Nuclear Gcncrmng Station
i i
Figure provided by SMUD.
SOIL SAMT'LE LOCATION Soil Sainple Figure 12: RHUT Auxiliary Boiler Land Area - Soil Sample Locations
Rancho Seco Nuclenr Gcnerating Station
TABLES Ilancho Scco Nuclcar Gcncrating Station 1695-SI<-02-0
TABLE 1 Auxiliary Building Survey Unit/Room" Class Vault 30
Vault 31
Vault 34
Vault 35
18 FL and LW
18 US
50 FL and LW
50 US
53 FL and LW
53 US
SURVEY UNIT CLASSIFICATION AND SCAN COVERAGE FOR SURVEYED ROOMS IN THE AUXILIARY BUILDING RANCHO SECO NUCLEAR GENERATING STATION HERALD, CALIFORNIA Percent Scan Coverage Beta Beta Beta Floor Lower Wall Upper Surfaces Gamma Floor/Lower Wall 100
5
___
100
5
___
100
___
100
100
50 b
___
--_
___
___
___
___
___
___
___
100
50
___
___
___
___
___
100
50
___
___
___
___
liancho Seco Nuclear Gcncrnting S t a t i m
1695-SR-02-0
TABLE 2
10
SURFACE ACTIVITY LEVELS AUXILIARY BUILDING STRUCTURAL SURFACES RANCHO SECO NUCLEAR GENERATING STATION HERALD, CALIFORNIA FL 16,000
8 YES FL 4,300
1 YES LW 2.600
1 YES Removable Activity Activity Meets Gross Beta DCGL~
Total Beta (dpm/100 ern')'
Room/
Location" Surfaceb Activity I
I I
I I
- -_
I/ Vault 34 I
I I
I I
I1
I L\\V I
740 I
I
I YES II I
7 LW 6,000
-1 YES FL 6,100
7 YES I
I I
I
~~
/[Room50
'
FL 1,200
6 YES I
FL 4.100
-1 YES I
LW 2,400
3 YES
us 130
2 YES
22
24
26
us 170
3 YES FL 12,000
5 YES FL 14,000
6 YES FL 5,700
3 YES FL 10,000
8 YES FL 15,000
6 YES FL 14,000
8 YES
___
NO'
14 After FL 1.100
2 YES f
14 Before'
FL 110,000
___
FL 3,400
3 YES
FL 4,200
6 YES
FL 12.000
2 YES 1695-SR-02-0
Rancho Seco Nuclear Generating Station
TABLE 2 (continued)
Room/
Location" Surfaceb SURFACE ACTIVITY LEVELS AUXILIARY BUILDING STRUCTURAL, SURFACES RANCHO SECO NUCLEAR GENERATING STATION HERALD, CALIFORNIA Removable Activity Activity Meets Gross Beta DCGL~
Total Beta Activity (dpm/100 cm2)'
19
I I
I I
I Rooin 53 I
I I
I I
FL 5,400
4 1'ES FL 3,700
1 YES FL 2,500
6 YES Rancho Scco Nuclear Gcnerating Station
1695-SR-02-0
TABLE 3 AUXILIARY BUILDING ACID WASTE SYSTEM DRAINS AND PIPE PENETRATIONS CONFIRMATORY GAMMA SCAN RANGES RANCHO SECO NUCLEAR GENERATING STATION HERALD, CALIFORNIA Gamma Scan Range (cpm)
SMUD Drain Line Diameter Scan Length Location (inches)
(feet)
ORISE
~
.-I-(
I I
Turbine Building Backgrounds" b
Conduit, East Side 1
1 300 to 600
___
Conduit, East Side 2
1 300 to 600
-_-
Conduit. East Side 3 I
I
I 200 to 600 I
II Conduit, East Side 4
1 300 to 600
__-
Penetration, East Side
1 300 to 600
_--
Exciter Pad East
12 200 to so0
__-
Exciter Pad West
12 200 to 800
__-
Background Range
___
I
-__
I 200 to800
-
-
Auxiliary Building Pipe Penetrations'
Vault 30 2 to 3
400 to 900 N A ~
Vault 34 2 to 3
600 to 1,200 NA NA Rooin 18 2 to 14 3 to 5 200 to 800 Rooin 53 2 to 4 2 to 3 200 to 600 I
I I
I Auxiliary Building Acid Waste Drainse I
510 to 19,000
4-1-12, Segment 1' I
0 to 8 I
400 to 1,400
[
4-2-1 5
900 990
600 520
800 660
3,200 7,700
5,000 16,000 4.5 20,000 3,000
- Turbine Building embedded piping backgrounds were dctcrmiticd within Turbine Building conduits. This data was collected during a prcvious OlZISE survey (OIZISE 2007d).
Wcasurcments not performed by SMUD within thc Turbine Bidding conduits.
CFigurc not provided. Each room had iiumerous pipe pcnetratioiis and a portion of those penetrations wcrc scaniicd.
dNot applicable. 01IISE compared the pipe pciietratioii scan results to background.
~licfcr to Figures 9 and 10. SMUD data was provided in to OIZISE in a pichinary I S
data package. OlIISE aiid SMUD results were rounded to hvo sigiificant digits.
fSb1UD originally numbered incorrectly Acid Wastc Drain Line 4-1-12, Scgncnt 1 as 4-1-13, Due to discrepancies in tlic survey data, OIZISE recommends further evaluation of this drain hie and the itistrumcntation used to collect the prcliniinary FSS data.
1695-SR-02-0
liancho Scco Nuclear Generating Station
TABLE 4 Auxiliary Building Survey Unit / Roomn Vault 30 Vault 31 Vault 34 Vault 35 18 FL and LW 18 us 50 FL and LIT 50 US 53 FL and LW 53 us RADIONUCLIDE CONCENTRATIONS I N SOIL SAMPLES RANCHO SECO NUCLEAR GENERATING STATION HERALD, CALIFORNIA Gross Beta DCGLb Design DCGL,,,'
(dpm/100 cm')
(dpm/100 cm')
Class
43,000 163,400
43,000 163,400
43,000 154,800
43,000 154,800
43,000 137,600
43,000 NAd
43,000 206,400
43,000 NA
43,000 172,000
43,000 NA
~
'Ilcfcr to Figurc 1 1.
I>Unccrtaititics represcat thc 95% confidence lcvcl bascd 011 totd propaptcd unccrtamties.
TABLE 5 Rancho Seco Nuclear Generating Station
1695-SR-02-0
TABLE G CO-60 Ni-63 Sr-90 CS-131 CS-137 DERIVED CONCENTRATION GUIDELINE LEVELS FOR SOIL SAMPLES RANCHO SECO NUCLEAR GENERATING STATION HERALD, CALIFORNIA
~~
1.93E-tOO 1.26E+01 1.6OE-06 1.52Et-07 3.7 6E-03 6.49B+03 1.09E+00 2.24Ei-01 4.62B-01 5.28Et-01 Single Nuclide DCGL, Values for Detectable Radionuclides" II Peak of the Mean Dose I
C-14 I
2.9 3E-0 6 I
8.33E+06 II
,Table 6-5 from the Licc~isc Termination Plan (SIvlUU 2006n).
Rancho Scco Nuclear Gcncrating Station
REFERENCES Oak Rtdge Associated Universities (OIWU). Quhty Prograin Manual for the Independent Environmental Assessment and Verification Program. Oak Ridge, Tennessee; P\\/larch 1,2007.
Oak Rtdge Institute for Science and Education (ORISE). Final Confirmatoi-y Survey Plan for the Remaining Structural Surfaces, Embedded Piping, StanQng Water and Open Land Area Suivey Units, Rancho Seco Nuclear Generating Station, Herald Cahfornia pocket No. 50-312; RFTA No.
06-0031. Oak Ridge, Tennessee; August 10,2007~
Oak Rtdge Institute for Science and Education. Suivey Procedures Manual for the Independent Environmental Assessment and Verification Prograin. Oak Ibdge, Tennessee; August 3,2007b.
Oak Rtdge Institute for Science and Education. Laboratoiy Procedures Manual for the Environmental Suivey and Site Assessment Prograin. Oak Ridge, Tennessee; June 15,2007~.
Oak Ibdge Instihite for Science and Education. Revised-Confirmatoiy Suivey Report for Portions of the A~i~iliaiy Building Structural Surfaces and Turbine Budding Embedded Piping, Rancho Seco Nuclear Generating Station, Herald, California. DCN 1695-SR-01-1. Oak Ibdge, Tennessee; December 21,2007d.
Sacramento Municipal Utlllty District (SMUD). License Termination Plan, Rancho Seco Nuclear Generating Station, Herald, California; April 2006a.
Sacramento Municipal Uulity District. Decommissioning Technical Basis Document: Structure Nuclide Fraction and DCGLs. DTBD-05-015, Revision 0. Rancho Seco Nuclear Generating Station, Herald, California; June 2, 2006b.
Sacramento Iclunicipal Uulity District. Decominissioning Technical Basis Document: Embedded Piping Scenario and DCGL Deterinination Basis. DTBD-05-009, Revision No. 1.O. Rancho Seco Nuclear Generating Station, Herald, California; June 1,2007.
U.S. Nuclear Regulatory Coinrnission (NRC). Letter froin J. Hiclman (NRC-HQ to S. Redeker (SMUD); SUBJECT: Rancho Seco Nuclear Generating Station - Issuance of Amendment RE:
License Termination Plan (TAC No. J52668). Washmgton, DC; November 27,2007.
Rancho Scco Nuclear Gcncrating Statlon
1695-SR-02-0