IR 05000282/2007002
| ML071340358 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/14/2007 |
| From: | Richard Skokowski NRC/RGN-III/DRP/RPB3 |
| To: | Thomas J. Palmisano Nuclear Management Co |
| References | |
| FOIA/PA-2010-0209 IR-07-002 | |
| Download: ML071340358 (39) | |
Text
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC INTEGRATED INSPECTION REPORT 05000282/2007002 AND 05000306/2007002
Dear Mr. Palmisano:
On March 31, 2007, the U. S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Prairie Island Nuclear Generating Plant, Units 1 and 2. The enclosed report documents the inspection findings which were discussed on April 4, 2007, with you and other members of your staff.
This inspection examined activities conducted under your license as they relate to safety and to compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
Based on the results of this inspection, one NRC identified finding of very low safety significance was identified. This finding was determined to involve a violation of NRC requirements. However, because of its very low safety significance, and because the issue was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation (NCV) in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
If you contest any finding or the subject/severity of any Non-Cited Violation in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector Office at the Prairie Island facility. 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 available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Richard A. Skokowski, Chief Branch 3 Division of Reactor Projects Docket Nos. 50-282; 50-306 License Nos. DPR-42; DPR-60 Enclosure: Inspection Report 05000282/2007002 and 05000306/2007002 w/Attachment: Supplemental Information cc w/encl: D. Cooper, Senior Vice President and Chief Nuclear Officer M. Sellman, President and Chief Executive Officer Regulatory Affairs Manager J. Rogoff, Vice President, Counsel & Secretary Nuclear Asset Manager State Liaison Officer, Minnesota Department of Health Tribal Council, Prairie Island Indian Community Administrator, Goodhue County Courthouse Commissioner, Minnesota Department of Commerce Manager, Environmental Protection Division Office of the Attorney General of Minnesota
SUMMARY OF FINDINGS
IR 05000282/2007002, 05000306/2007002; 01/01/07 - 03/31/07; Prairie Island Nuclear
Generating Plant, Units 1 and 2; Occupational Radiation Safety.
This report covers a 3-month period of baseline resident inspection and announced baseline inspection of the operator requalification program and occupational radiation protection. The inspection was conducted by the resident inspectors and inspectors from the Region III office.
The inspectors identified one finding and associated Non-Cited Violation. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review.
The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
A. Inspector-Identified and Self-Revealed Findings
Cornerstone: Occupational Radiation Safety
- Green.
A finding of very low safety significance and associated Non-Cited Violation was inspector-identified during review of an issue where a station operator entered into a Locked High Radiation Area (LHRA) without authorization while a high integrity container was being moved to the radioactive waste barrel yard. The licensee has entered this finding into the corrective action program.
The finding was more than minor because it was associated with the Program/Process attribute of the Occupational Radiation Safety cornerstone, and potentially affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. The finding was determined to be of very low safety significance because the finding did not involve As-Low-As-Reasonably-Achievable planning, collective dose was not a factor, it did not involve an overexposure, and the individual involved received very low dose. Additionally, there was not a substantial potential for a worker overexposure, and the licensees ability to assess worker dose was not compromised.
The initial licensee evaluation of this issue was inadequate because it failed to address this event in relationship to previous similar events concerning the performance and effectiveness of LHRA guards. Specifically, Prairie Island had a similar event involving the performance of LHRA guards controlling access to radiologically significant areas during its April 2006 refueling outage. Had the previous event been properly identified, entered into the licensees corrective action program, and evaluated adequately and in a timely manner, this December 2006 event may not have occurred. Consequently, this finding also related to the cross-cutting area of problem identification and resolution dealing with the corrective action program component to ensure issues are promptly identified and fully evaluated to allow timely corrective actions. (Section 2OS1)
Licensee-Identified Violations
None.
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at or near full power throughout the inspection period.
Unit 2 operated at or near full power until January 4, 2007, when power was reduced to about 98 percent to perform maintenance on the feed water control system. Full power was restored on January 5, 2007. The unit remained at full power until February 28, 2007, when the unit was shut down for a maintenance outage (2F2401) to repair the 22 main steam isolation valve. The reactor was restarted and the generator placed on-line on March 8, 2007. Unit 2 achieved full power on March 13, 2007, and operated at or near full power for the remainder of the inspection period.
REACTOR SAFETY
Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity
1R04 Equipment Alignment (71111.04Q and S)
.1 Partial Walkdowns
a. Inspection Scope
The inspectors performed three partial system equipment alignment inspection samples comprised of in-plant walkdowns of accessible portions of trains of risk-significant equipment associated with the mitigating systems cornerstone. The inspectors conducted the inspections during times when the trains were of increased importance due to the redundant trains or other related equipment being unavailable. The inspectors also reviewed documents entering deficient conditions associated with equipment alignment issues into the corrective action program verifying that the licensee was identifying issues at an appropriate threshold and entering those issues into their corrective action program in accordance with the licensees corrective action procedures.
The inspectors utilized the valve and electric breaker checklists, where applicable, to verify that the components were properly positioned and that support systems were lined up as needed. The inspectors also examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious performance deficiencies. The inspectors reviewed outstanding work orders (WOs) and corrective action program documents (CAPs) associated with the operable trains to verify that those documents did not reveal issues that could affect the completion of the available trains safety functions. The inspectors used the information in the appropriate sections of the Technical Specifications (TS) and the Update Safety Analysis Report (USAR) to determine the functional requirements of the systems.
The inspectors verified the alignment of the following trains:
- diesel generator D2 while D1 was out of service for planned maintenance on January 23, 2007;
- 12 component cooling pump while the 11 component cooling pump was out of service for planned maintenance on January 31, 2007; and
- diesel generator D6 while D5 was out of service for planned maintenance on February 21, 2007.
Key documents used by the inspectors in conducting this inspection are listed in the to this report.
b. Findings
No findings of significance were identified.
.2 Complete System Alignment Walkdown of Unit 2 Safety Injection System
a. Inspection Scope
During the week of March 5, 2007, the inspectors performed a detailed in-plant walkdown of the alignment and condition of the Unit 2 safety injection system. The safety injection system is a risk-significant and safety-related mitigating system that provides water inventory to the reactor coolant system during off-normal and accident conditions. The inspectors also reviewed CAPs associated with equipment alignment issues to verify that the licensee was identifying issues at an appropriate threshold and entering them into their corrective action program in accordance with the licensees corrective action procedures.
The inspectors used applicable alignment checklists and plant drawings to verify that system components were properly positioned to support the completion of system safety functions and to verify that the as-found system configuration matched the configuration specified in the system alignment checklists and plant drawings. The inspectors examined the material condition of the components, such as pumps, supports and snubbers, motors, valves, instrumentation, controls, bus relays, and electrical panels.
Where applicable, the inspectors examined outstanding design issues, temporary modifications, and operator workarounds. Where applicable, the inspectors verified that tagging clearances were appropriate and attached to the specified equipment. The inspectors reviewed outstanding WOs, Work Requests, and CAPs associated with the trains to determine if any degraded conditions existed that could affect the accomplishment of the systems safety functions. The inspectors referred to the TS, USAR, and other design basis documents to determine the functional requirements of the systems and verified those functions could be performed if needed. Key documents used by the inspectors in conducting this inspection are listed in the Attachment to this inspection report.
This inspection effort constituted one complete system alignment inspection sample for a system associated with the mitigating systems cornerstone.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
.1 Quarterly Fire Protection Area Walkdowns
a. Inspection Scope
The inspectors conducted in-office and in-plant reviews of portions of the licensees Fire Hazards Analysis and Fire Strategies to verify consistency between those documents and the as-found configuration of the installed fire protection equipment and features in the fire protection areas listed below. The inspectors selected fire areas for inspection based on their overall contribution to internal fire risk as documented in the Individual Plant Examination of External Events (IPEEE), potential to impact equipment which could initiate a plant transient, or impact on the plants ability to respond to a security event. The inspectors assessed the control of transient combustibles and ignition sources, the material and operational condition of fire protection systems and equipment, and the status of fire barriers. In addition, the inspectors reviewed CAPs associated with fire protection issues to verify that the licensee was identifying issues at an appropriate threshold and entering them into their corrective action program in accordance with the licensees corrective action procedures.
The following nine fire areas were inspected by in-plant walkdowns supporting the completion of nine fire protection zone walkdown samples:
- Fire Area 25, diesel generator D1 room on January 16, 2007;
- Fire Area 31, east auxiliary feedwater pump room on January 16, 2007;
- Fire Area 32, west auxiliary feedwater pump room on January 16, 2007;
- Fire Area 81, bus 16 switchgear room on January 17, 2007;
- Fire Area 113, diesel generator D5 day tank room on January 17, 2007;
- Fire Area 115, diesel generator D5 lubricating oil make-up tank room on January 17, 2007;
- Fire Area 117, bus 25 switchgear room on January 17, 2007;
- Fire Area 41A, plant screenhouse 670-foot elevation on January 18, 2007; and
- Fire Area 41B, cooling water pump and safeguard traveling screen rooms on January 18, 2007.
Key documents used by the inspectors in conducting this inspection are listed in the to this inspection report.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures
a. Inspection Scope
The inspectors performed an in-office review of the most recently completed surveillance procedure (SP) for the inspection of plant flooding barriers and the abnormal operating procedure for flooding. The contents of these documents were compared to the plant flood protection design sections in the USAR and the assumption contained in the IPEEE associated with an external flooding event. This inspection effort completed the annual external flood protection inspection sample under the initiating events cornerstone.
The inspectors performed an in-plant inspection of flood protection barriers in the auxiliary building, turbine building, D5/D6 diesel generator building, and the old screenhouse comparing the as-found conditions of the flood protection panels against the acceptance criteria in the SP. The inspectors also verified that the actions specified in the abnormal operating procedure for flooding could be performed in a timely manner (three days), if required, and the necessary hardware and consumable materials were available and still within the usable shelf life.
The inspectors reviewed several CAPs and Work Requests to verify that minor deficiencies identified during this inspection were entered into the licensees corrective action program; that problems associated with plant equipment relied upon to prevent or minimize flooding were identified at an appropriate threshold; and that corrective actions commensurate with the significance of the issue were identified and implemented. As part of this inspection, the inspectors reviewed the documents listed in the Attachment.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance
a. Inspection Scope
On February 26, 2007, the inspectors observed the licensees inspection of the following safety-related heat exchangers:
- 12 DDCLP right angle gear drive lubricating oil cooler.
These heat exchangers were selected for review because the cooling water system was ranked high in the plant specific risk assessment and functions to support the proper operation of nearly all safety-related mitigating systems and provided the plants connection to the ultimate heat sink. This inspection effort completed one heat sink inspection procedure sample under the mitigating systems cornerstone.
The inspectors performed an independent as-found inspection of the HXs associated with the 12 DDCLP immediately after their opening and discussed the as-found condition of the HXs with the system engineer and the Generic Letter 89-13 program engineer. The inspectors reviewed the completed work package for the inspection and cleaning of the 12 DDCLP jacket water and right angle gear drive lubricating oil HXs comparing the as-found condition to the applicable tube plugging calculation acceptance criteria. The inspectors reviewed Procedure H21, Generic Letter 89-13 Implementing Program, Revision 10, governing Generic Letter 89-13 heat exchanger inspections in order to verify that the licensee was properly implementing their program.
The inspectors also reviewed CAPs to verify that the licensee was identifying issues at an appropriate threshold and entering them into their corrective action program in accordance with licensees corrective action procedures. Key documents used by the inspectors in conducting this inspection are listed in the Attachment to this report.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification
.1 Operating Test Results
a. Inspection Scope
The inspectors reviewed the overall pass/fail results of the comprehensive annual job performance measure operating tests and the annual simulator operating tests (required to be given per 10 CFR 55.59(a)(2)). The operating tests were administered by the licensee from August 2006 through October 2006. The overall results were compared with the Significance Determination Process (SDP) in accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix I, Operator Requalification Human Performance Significance Determination Process. Key documents used by the inspectors in conducting this inspection are listed in the Attachment to this inspection report.
b. Findings
No findings of significance were identified.
.2 Quarterly Review of Licensed Operators Requalification Training
a. Inspection Scope
On January 22, 2007, the inspectors performed a quarterly review of licensed operator requalification training in the simulator, which constituted one licensed operator requalification inspection sample. The inspectors observed a crew during an evaluated exercise in the plants simulator facility. The inspectors compared crew performance to licensee management expectations. The inspectors verified that the crew completed all of the critical tasks for each exercise scenario. For any weaknesses identified, the inspectors observed that the licensees evaluators noted the weaknesses and discussed them in the critique at the end of the session.
The inspectors assessed the licensees effectiveness in evaluating the requalification program ensuring that licensed individuals would operate the facility safely and within the conditions of their licenses, and evaluated licensed operator mastery of high-risk operator actions. The inspection activities included, but were not limited to, a review of high-risk activities, emergency plan performance, incorporation of lessons learned, clarity and formality of communications, task prioritization, timeliness of actions, alarm response actions, control board operations, procedural adequacy and implementation, supervisory oversight, group dynamics, interpretations of TS, simulator fidelity, and licensee critique of performance.
Key documents used by the inspectors in conducting this inspection are listed in the to this inspection report.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
.1 Routine Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed repetitive maintenance activities to assess maintenance effectiveness, including Maintenance Rule (10 CFR 50.65) activities, work practices, and common cause issues. The inspectors performed two issue/problem-oriented maintenance effectiveness samples under the mitigating systems and barrier integrity cornerstones. The inspectors assessed the licensees maintenance effectiveness associated with problems on:
- nuclear fuel cladding failures; and
- 11 turbine-driven auxiliary feedwater pump turbine bearing failure and overspeed trip throttle latch mechanism engagement on March 8, 2007.
The inspectors conducted in-office reviews of the licensees maintenance rule evaluations of equipment failures for maintenance preventable functional failures and equipment unavailability time calculations, comparing the licensees evaluation conclusions to applicable Maintenance Rule (a)(1) performance criteria. Additionally, the inspectors reviewed scoping, goal-setting (where applicable), performance monitoring, short-term and long-term corrective actions, functional failure definitions, and current equipment performance status.
The inspectors reviewed CAPs for significant equipment failures associated with risk-significant and safety-related mitigating equipment to ensure that those failures were properly identified, classified, and corrected. The inspectors reviewed other CAPs to assess the licensees problem identification threshold for degraded conditions, the appropriateness of specified corrective actions, and that the timeliness of the implementation of corrective actions were commensurate with the safety significance of the identified issues. Key documents used by the inspectors in conducting this inspection are listed in the Attachment to this report.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed the licensees management of plant risk during emergent maintenance activities or during activities where more than one significant system or train was unavailable. The activities were chosen based on their potential impact on increasing the probability of an initiating event or impacting the operation of safety-significant equipment. The inspections were conducted to determine whether evaluation, planning, control, and performance of the work were done in a manner to reduce the risk and minimize the duration where practical, and that contingency plans were in place where appropriate.
The licensees daily configuration risk assessment records and observations of work in progress were used by the inspectors to verify that the equipment configurations were properly listed, protected equipment were identified and were being controlled where appropriate, work was conducted properly, and significant aspects of plant risk were communicated to the necessary personnel. The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program.
In addition, the inspectors reviewed selected issues, listed in the Attachment, that the licensee encountered during the activities, to determine whether problems were entered into the corrective action program with the appropriate characterization and significance.
The inspectors completed seven samples under the initiating events and mitigating systems cornerstones by reviewing the following activities:
- the emergent failure of bus 26 load sequencer with the planned unavailability of the 121 instrument air compressor, the 12 condensate pump, 13 charging pump, and D1 diesel generator on January 9, 2007;
- the planned unavailability of diesel generator D1, 21 component cooling pump, and the 121 instrument air compressor on January 25, 2007;
- the planned unavailability of the 121 intake bypass gate, the 121 safeguards traveling screen, and the 11 component cooling pump on January 31, 2007;
- the planned unavailability of volume control tank level transmitter LT-112, 122 intake bypass gate, and the 21 charging pump on February 7, 2007;
- the planned unavailability of the 122 intake bypass gate, D2 diesel generator, 121 motor-driven cooling water pump, bus 27, and the Red Rock 2 transmission line;
- the planned unavailability of the 122 intake bypass gate, the Red Rock 2 transmission line, and breaker 2RSY on February 13, 2007; and
- the planned unavailability of the 122 instrument air compressor, the 13 charging pump, the 12 component cooling pump, the Byron transmission line, and the Red Rock 2 transmission line on March 14, 2007.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed the technical adequacy of six operability evaluations completing six operability evaluation inspection samples associated with equipment performance under the initiating events, mitigating systems, and barrier integrity cornerstones. The inspectors conducted these inspections by in-office review of associated documents and in-plant walkdowns of affected areas and plant equipment.
The inspectors compared degraded or nonconforming conditions of risk-significant structures, systems, and components associated with barrier and mitigating systems and against the functional requirements described in the TS, USAR, and other design basis documents; determined whether compensatory measures, if needed, were implemented; and determined whether the evaluation was consistent with the requirements of Administrative Work Instruction 5AWI 3.15.5, Operability Determinations. The following operability evaluations were reviewed by inspectors:
- Operability Recommendation (OPR) 01070125 that documented the operability of the reactor coolant pumps with impellers that had different serial numbers than those used in the safety analysis for core design, loss of coolant accident, loss of reactor coolant system flow, steam line break, and fuel assembly hold down;
- OPR 01070049 that documented the operability of emergency diesel generator D6 with generator axial vibration greater than the manufacturers recommendation;
- OPR 01070752-01 that documented the operability of the bus 26 load sequencer following receipt of an unexpected error code during testing;
- OPR 01073261 that documented the operability of emergency diesel generator D2 with a standby jacket coolant pump impeller that was a different size than was used for seismic qualification;
- CAP 01069591 prompt operability determination for the abnormal noise and vibration exhibited by the 22 main steam isolation valve; and
- OPR 01076278 that documented the operability of emergency diesel generator D6 with generator bearing vibration exceeding vendor limits on February 12, 2007.
Key documents used by the inspectors in conducting this inspection are listed in the to this inspection report.
b. Findings
No findings of significance were identified.
1R17 Permanent Plant Modifications (Annual)
a. Inspection Scope
The inspectors evaluated Design Change EC 652, Cooling Water In-Service Test Modification. The modification added flow meters/transmitters in the piping that supplies cooling water to the diesel-driven cooling water pump diesel jacket and pressure transmitters on the pump discharge piping. These instruments were added to address a concern regarding unmonitored flow paths which did not comply with American Society of Mechanical Engineers (ASME) Code requirements for in-service testing. The inspectors effort completed one permanent plant modification inspection sample.
The inspectors reviewed the modification installed in March 2007 to verify that the design basis, licensing basis, and performance capability of risk-significant systems were not degraded by the installation of the modification. The inspectors considered the design adequacy of the modification by performing a review of the modifications impact on licensing basis (10 CFR 50.59), flow paths, plant electrical requirements, equipment protection, operation, failure modes, and other related process requirements. The inspectors conducted this inspection by review of documents and in-plant walkdowns of associated plant equipment.
Key documents used by the inspectors in conducting this inspection are listed in the to this inspection report.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors assessed post-maintenance testing completing six post-maintenance test inspection samples. The inspectors selected post-maintenance tests associated with important mitigating and barrier integrity systems to ensure that the testing was performed adequately, demonstrated that the maintenance was successful, and that operability of associated equipment and/or systems was restored. The inspectors conducted these inspections by in-office review of documents, in-plant walkdowns of associated plant equipment, and interviews with responsible personnel. The inspectors observed and assessed the post-maintenance testing activities for the following maintenance activities:
- D1 diesel generator 18-month inspection on January 25, 2007;
- 121 safeguard traveling screen following preventive maintenance on January 30, 2007;
- 121 motor-driven cooling water pump inspection on February 13, 2007;
- D5 diesel generator 18 month inspection on February 20, 2007;
- 12 diesel cooling water pump following preventive maintenance on February 28, 2007; and
- 22 main steam isolation valve packing replacement and actuator adjustment on March 8, 2007.
The inspectors reviewed the appropriate sections of the TS, USAR, and maintenance documents to determine the systems safety functions and the scope of the maintenance. The inspectors also reviewed CAPs to verify that the licensee was identifying issues at an appropriate threshold and entering them into their corrective action program in accordance with the licensees corrective action procedures. Key documents used by the inspectors in conducting this inspection are listed in the to this report.
b. Findings
No findings of significance were identified.
1R20 Refueling and Other Outage Activities
a. Inspection Scope
The inspectors observed the licensees performance during a planned Unit 2 maintenance outage (2F2401) conducted between February 28, 2007, and March 8, 2007. The purpose of the outage was to repair the 22 main steam isolation valve. These inspection activities represent one outage inspection sample.
This inspection consisted of an in-office review of the licensees outage schedule, safe shutdown plan, and procedures governing the outage. Specifically, the inspectors assessed whether the licensee planned to effectively manage elements of shutdown risk pertaining to reactivity control, decay heat removal, inventory control, electrical power availability, and containment integrity. Key documents used by the inspectors in conducting this inspection are listed in the Attachment to this report.
The inspectors conducted in-plant observations of the following outage activities daily:
- attended outage management turnover meetings to verify that the current shutdown risk status was accurate, well understood, and adequately communicated;
- performed walkdowns of the main control room to observe the alignment of systems important to shutdown risk; and
- performed walkdowns to observe ongoing work activities and foreign material exclusion control.
The inspectors performed in-plant observations of the following specific activities:
- shutdown safety assessment;
- plant cooldown;
- the implementation of foreign material exclusion controls at the containment airlock;
- containment walkdown; and
- plant start up and power ascension.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
During this inspection period, the inspectors completed six surveillance inspection samples. Observation of surveillance procedures SP 1089A and SP 2371 completed the quarterly inservice testing inspection sample requirement of a risk-significant pump or valve surveillance test. The observation of SP 2405 completed the requirement inspection sample to observe a reactor coolant system leakage integrity test sample.
The inspectors selected the following surveillance testing activities as samples:
- SP 2307, D6 Diesel Generator 6-Month Fast Start Test on January 4, 2007;
- SP 1334; D1 Diesel Generator 18-Month 24-Hour Load Test on January 8, 2007;
- SP 1089A; Train A Residual Heat Removal (RHR) Pump and Suction Valve from Refueling Water Storage Tank Quarterly Test on February 1, 2007;
- SP 1106B; 22 Diesel Cooling Water Pump Monthly Test on February 15, 2007;
- SP 2405; Unit 2 Mid-Cycle and Refueling Outage Boric Acid Corrosion Examinations Inside Containment on February 28, 2007; and
- SP 2371; Cold Shutdown Test of RHR Pumps and Check Valves on March 4, 2007.
During completion of the inspection samples, the inspectors observed in-plant activities and reviewed procedures and associated records to verify, when applicable, that:
- preconditioning did not occur;
- effects of the testing had been adequately addressed by control room personnel or engineers prior to the commencement of the testing;
- acceptance criteria was clearly stated, demonstrated operational readiness, and was consistent with the system design basis;
- plant equipment calibration was correct, accurate, properly documented, and the calibration frequency was in accordance with TS, USAR, procedures, and applicable commitments;
- measuring and test equipment calibration was current;
- test equipment was used within the required range and accuracy;
- applicable prerequisites described in the test procedures were satisfied;
- test frequency met TS requirements to demonstrate operability and reliability;
- the tests were performed in accordance with the test procedures and other applicable procedures;
- jumpers and lifted leads were controlled and restored where used;
- test data/results were accurate, complete, and valid;
- test equipment was removed after testing;
- where applicable for inservice testing activities, testing was performed in accordance with the applicable version of ASME Code,Section XI, and reference values were consistent with the system design basis;
- where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or declared inoperable;
- where applicable for safety-related instrument control surveillance tests, reference setting data have been accurately incorporated in the test procedure;
- equipment was returned to a position or status required to support the performance of its safety functions; and
- all problems identified during the testing were appropriately documented in the corrective action program.
Key documents used by the inspectors in conducting this inspection are listed in the to this report.
b. Findings
No findings of significance were identified.
1R23 Temporary Plant Modifications
a. Inspection Scope
The inspectors conducted in-plant observations of the physical changes to the equipment and an in-office review of documentation associated with one temporary modification. This constituted one temporary modification inspection sample. The inspectors reviewed Temporary Modification EC 9802, which was implemented to install a seal on the snubber for the feedwater control valve to the 21 steam generator, CV-31135, as a temporary repair for an oil leak.
The inspection activities included a review of design documents, safety screening documents, and the USAR to determine that the temporary modification was consistent with modification documents, drawings, and procedures. The inspectors also reviewed the post-installation test results to confirm that tests were satisfactory and the actual impact of the temporary modification on the permanent system and interfacing systems were adequately verified. Additionally, the inspectors reviewed the corrective action documentation associated with an identified problem with the air supply to the power operated relief valves to verify that the licensee was identifying issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action. The key documents reviewed by the inspectors are listed in the Attachment to this report.
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
The inspectors observed the licensee perform an emergency preparedness drill on February 6, 2007. This inspection effort completed one emergency planning drill evaluation sample.
The inspectors observed activities in the Technical Support Center and Operations Support Center and attended the post-drill critique on February 6, 2007. The focus of the inspectors activities was to note any weaknesses and deficiencies in the drill performance and ensure that the licensee evaluators noted the same weaknesses and deficiencies and entered them into the corrective action program. The inspectors placed emphasis on observations regarding event classification, notifications, protective action recommendations, and site evacuation and accountability activities. Key documents used by the inspectors in conducting this inspection are listed in the Attachment to this report.
b. Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1 Review of Licensee Performance Indicators for the Occupational Exposure Cornerstone
a. Inspection Scope
The inspectors reviewed the licensees occupational exposure control cornerstone performance indicators (PIs) to determine whether or not the conditions surrounding the PIs had been evaluated and to determine if identified problems had been entered into the corrective action program for resolution.
This review represented one sample.
b. Findings
No findings of significance were identified.
.2 Plant Walkdowns and Radiation Work Permit Reviews
a. Inspection Scope
The inspectors identified radiologically significant work areas within radiation areas, high radiation areas (HRAs), locked high radiation areas (LHRAs), and airborne areas in the auxiliary and containment buildings. Selected work packages and radiation work permits (RWPs) were reviewed to determine if radiological controls, including surveys, postings, air sampling data and barricades, were acceptable. Work areas included, but were not limited to:
- Wooden Door with Metal Door at Spent Resin Tank Replacement;
- 2R24 Incore Drives;
- 2R24 Containment Inspections;
- 2R24 Insulation Removal; and
- Resin Liner No. 129 to the Shipping Cask Transfer.
This review represented one sample.
The inspectors reviewed selected RWPs and associated radiological controls used to access these and other radiologically significant areas and evaluated the work control instructions and control barriers that were specified in order to determine if the controls and requirements provided adequate worker protection. The Site Technical Specification requirements for HRAs and locked high radiation areas were used as standards for the necessary barriers. Electronic dosimeter alarm set points for both integrated dose and dose rate were evaluated for conformity with survey indications and plant policy.
This review represented one sample.
b. Findings
No findings of significance were identified.
.3 Problem Identification and Resolution
a. Inspection Scope
The inspectors reviewed the licensees self-assessments, audits, and condition reports related to the access control program to determined if identified problems were entered into the corrective action program for resolution.
This review represented one sample.
Corrective action reports related to access controls and high radiation area radiological incidents (non-performance indicator occurrences identified by the licensee in HRAs of less than 1Rem/hr) were reviewed. Selected staff members were interviewed and a selected sample of corrective action documents were reviewed to determine if follow-up activities were conducted in an effective and timely manner commensurate with their importance to safety and risk based on the following:
- initial problem identification, characterization, and tracking;
- disposition of operability/reportability issues;
- evaluation of safety significance/risk and priority for resolution;
- identification of repetitive problems;
- identification of contributing causes;
- identification and implementation of effective corrective actions;
- resolution of Non-Cited Violations (NCV) tracked in the corrective action system; and
- implementation/consideration of risk significant operational experience feedback.
This review represented one sample.
The inspectors determined if the licensees self-assessment and audit activities completed for the year that preceded the inspection were identifying and addressing repetitive deficiencies or significant individual deficiencies in problem identification and resolution, as applicable.
This review represented one sample.
The inspectors discussed performance indicators with the radiation protection staff and reviewed data from the licensee's corrective action program to determine if there were any performance indicators for the occupational exposure cornerstone that had not been reviewed. There were none to evaluate.
This review represented one sample.
b. Findings
No findings of significance were identified.
.4 High Risk Significant, High Dose Rate High Radiation Area, and Very High Radiation
Area Controls
a. Inspection Scope
The inspectors reviewed the licensees performance indicators for high risk, high dose rate HRAs, and for very high radiation areas to determine if workers were adequately protected from radiological overexposure. Discussions were held with selected radiation protection management and technicians concerning high dose rate HRA and very high radiation area controls and procedures, including procedural changes that had occurred since the last inspection. This review was completed to determine if procedure modifications had substantially reduced the effectiveness and level of worker protection.
This review represented one sample.
The inspectors interviewed radiation protection (RP) supervisors to determine if plant evolutions that could impact radiological conditions were communicated between the RP group and other involved groups beforehand in order to allow corresponding timely actions to properly post and control the radiation hazards. This review represented one sample.
During plant walkdowns, the posting and locking of entrances to high dose rate HRAs, and very high radiation areas were reviewed for adequacy. This review represented one sample.
b. Findings
One finding of very low safety significance was identified.
Introduction:
A Green finding and associated NCV was identified when an NRC inspector reviewed an issue where a station operator entered without authorization into an LHRA while personnel were moving a high integrity container (HIC) to the radioactive waste barrel yard.
Description:
On December 4, 2006, Prairie Island station personnel were transferring HIC No. 129 containing radioactive resin from the back of a flat bed trailer into a HIC storage area in the radioactive waste barrel yard. The dose rates on the HIC were 18,000 millirem per hour (mRem/hr) at contact; 8,000 mRem/hr at a distance of one foot; and 3,000 mRem/hr at a distance of one meter. Given these radiological conditions, the HIC transfer work area was posted and controlled in accordance with station procedures and Technical Specification 5.7.2 as a locked high radiation area.
Station procedures and 10 CFR 20.1903 permit certain exceptions to radiological posting requirements for short periods of time so long as specified alternate measures are in place. At Prairie Island the alternate actions for LHRA posting and access control was achieved through the use of LHRA guards at various access points. Locked high radiation area guards were personnel specifically assigned to control access to radiologically significant areas of the plant. At Prairie Island, the specific responsibilities of the LHRA guards were defined in station procedure PINGP 1470 and included the requirement that LHRA guards not allow unauthorized or inadvertent access into the LHRA.
On December 4, 2006, at approximately 10:30 a.m., a station operator, while responding to a high level tank alarm on the radioactive waste liquid processing panel in the radioactive waste facility, entered a LHRA that was controlled by a LHRA guard.
The operator was not on an RWP that permitted access to locked high radiation areas and was not pre-briefed on the radiological conditions specific to the work area being entered nor the potential impact of the transient radiological conditions being created by the HIC transfer activities. The operator was advised by the LHRA guard not to enter the radioactive waste processing area, however, the operator was not specifically informed that the area was being controlled for radiological purposes as a locked high radiation area. Consequently, the operator entered the LHRA because the radiological status of the area was not clearly communicated by the LHRA guard.
Analysis:
The inspectors determined that the LHRA guard allowed unauthorized entry into the area contrary to procedural requirements which represents a performance deficiency and a finding as described in IMC 0612, Appendix B, Issue Screening. The issue was more than minor because it was associated with the Program/Process attribute of the Occupational Radiation Safety cornerstone and potentially affected the cornerstone objective to ensure worker health and safety from exposure to radiation.
The finding does not involve the application of traditional enforcement, because it did not result in actual safety consequences or the potential to impact the NRCs regulatory function, and was not the result of willful actions. The finding was evaluated using the SDP in accordance with IMC 0609, Appendix C, for the Occupational Radiation Safety cornerstone. The finding was determined to be of very low safety significance (Green)because it did not involve As-Low-As-Reasonably-Achievable (ALARA) planning, was not associated with an overexposure given the actual radiological conditions in the area, and there were other workers present in the general area of the HIC transfer that could have interceded to minimize any actual radiological exposure. Consequently, there was not a substantial potential for a worker overexposure and the licensees ability to assess worker dose was not compromised.
However, this issue was not entered into the licensees corrective action program in a timely manner and the evaluation of the issue was not comprehensive or thorough relative to regulatory impact on station technical specifications or 10 CFR Part 20 compliance. Additionally, the licensees evaluation did not fully develop the cause of the event nor evaluate this event in relationship to previous events concerning the performance and effectiveness of LHRA guards. Specifically, Prairie Island had a similar event involving the performance of LHRA guards controlling access to radiologically significant areas during the Unit 1 April 2006 refueling outage. Had that previous event been properly identified, entered into the licensees corrective action program, and evaluated adequately and in a timely manner, the December 2006 LHRA event may not have occurred.
On the night shift of April 28, 2006, a work crew was in the Unit 1 containment airlock (a posted locked high radiation area) without a RP specialist escort. This earlier event, although known to members of the licensees staff was not entered into the corrective action program until it was brought to the licensees attention by the NRC approximately nine months later (IR 01075188; dated February 01, 2007). Consequently, this finding also relates to the cross-cutting area of problem identification and resolution dealing with the corrective action program component intended to ensure issues are promptly identified and thoroughly evaluated to allow timely corrective actions.
Licensee corrective actions included revising the process for LHRA guards such that the guard obtains positive verification that all radiological requirements are met prior to authorizing entry to an area and training for Nuclear Plant Service Attendants on Technical Specification requirements, area guarding, and command and control techniques.
Enforcement:
Technical Specification 5.4.1.a. requires the licensee to establish, implement, and maintain procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Procedures specified in Regulatory Guide 1.33 include RP procedures for access control to radiological areas, which are provided by licensee procedure PINGP 1470, High Radiation Area/Locked High Radiation Area Entry and Control Briefing Sheet (Revision 5). That procedure states that if an LHRA guard is assigned, then that persons only responsibilities are to not allow unauthorized or inadvertent entries into the LHRA.
Contrary to the above, on December 4, 2006, a station operator entered an LHRA that was being controlled by an LHRA guard and into the radioactive waste facility without proper authorization. Since the finding is of very low safety significance and had been entered into the corrective action system as CAP 1070811, the associated violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000282/2007002-01; 05000306/2007002-01).
2OS2 As-Low-As-Reasonably-Achievable (ALARA) Planning and Controls (71121.02)
.1 Inspection Planning
a. Inspection Scope
The inspectors reviewed plant collective outage exposure history, current exposure trends, and ongoing outage activities in order to assess current performance and exposure challenges. This review included determining the plants current 3-year rolling average for collective exposure in order to help establish resource allocations and to provide a perspective of significance for any resulting inspection finding assessment.
The inspectors reviewed the outage work scheduled during the inspection period and associated work activity exposure and time/labor estimates for the following five-work activities which resulted in the highest personnel collective exposures or were otherwise activities that were conducted in radiologically significant areas:
- Wooden Door with Metal Door at Spent Resin Tank Replacement;
- 2R24 Incore Drives;
- 2R24 Containment Inspections;
- 2R24 Insulation Removal; and
- Resin Liner No. 129 to the Shipping Cask Transfer.
The inspectors determined site specific trends in collective exposures based on plant historical exposure and source term data. The inspectors reviewed procedures associated with maintaining occupational exposures ALARA and assessed those processes used to estimate and track work activity exposures.
These reviews represented three inspection samples.
b. Findings
No findings of significance were identified.
.2 Radiological Work Planning
a. Inspection Scope
The inspectors evaluated the licensees list of work activities ranked by estimated exposure that were completed during the outage and reviewed the following work activities of highest exposure significance:
- Wooden Door with Metal Door at Spent Resin Tank Replacement;
- 2R24 Incore Drives;
- 2R24 Containment Inspections;
- 2R24 Insulation Removal; and
- Resin Liner No. 129 to the Shipping Cask Transfer.
For the activities listed above, the inspectors reviewed the RWP packages, work orders, exposure estimates, and exposure mitigation requirements in order to verify that the licensee had established radiological engineering controls that were based on sound radiation protection principles in order to achieve occupational exposures that were ALARA. This review also involved determining that the licensee had reasonably grouped the radiological work into work activities, based on historical precedence, industry norms, and/or special circumstances.
These reviews represented two inspection samples.
b. Findings
No findings of significance were identified.
.3 Verification of Dose Estimates and Exposure Tracking Systems
a. Inspection Scope
The inspectors reviewed the licensees assumptions and basis for its collective outage exposure estimate and evaluated the methodology and practices for projecting work activity specific exposures. This review included evaluating both dose rate and time/labor estimates for adequacy compared to historical station specific or industry data.
This review represented one inspection sample.
b. Findings
No findings of significance were identified.
.4 Problem Identification and Resolution
a. Inspection Scope
The inspectors reviewed the licensees self-assessments, audits, and Special Reports related to the ALARA program since the last inspection to determine if the licensees overall audit programs scope and frequency for all applicable areas under the occupational cornerstone met the requirements of 10 CFR 20.1101(c). The licensees corrective action program was also reviewed to determine if repetitive deficiencies and/or significant individual deficiencies in problem identification and resolution had been addressed.
These reviews represented two inspection samples.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator Verification
Cornerstone: Barrier Integrity
a. Inspection Scope
The inspectors reviewed the licensees submittals for three performance indicators for Prairie Island Units 1 and 2, completing six PI verification inspection procedure samples.
The inspectors used PI guidance and definitions contained in National Energy Institute Document 99-02, Revision 4, Regulatory Assessment Performance Indicator Guideline, to verify the accuracy of the PI data. The inspectors review included conditions and data from logs, condition reports, and calculations for each PI specified.
The inspectors also reviewed the CAPs listed in the Attachment to this report to verify that the licensee was identifying issues at an appropriate threshold and entering them into their corrective action program in accordance with corrective action procedures.
The licensees reporting of the following PIs were verified:
Unit 1
- Unplanned Scrams per 7000 Critical Hours for the 1st Quarter 2006 through the 4th Quarter 2006;
- Unplanned Scrams with the Loss of Normal Heat Removal 1st Quarter 2006 through the 4th Quarter 2006; and
- Unplanned Power Changes per 7000 Critical Hours for the 1st Quarter 2006 through the 4th Quarter 2006.
Unit 2
- Unplanned Scrams per 7000 Critical Hours for the 1st Quarter 2006 through the 4th Quarter 2006;
- Unplanned Scrams with the Loss of Normal Heat Removal 1st Quarter 2006 through the 4th Quarter 2006; and
- Unplanned Power Changes per 7000 Critical Hours for the 1st Quarter 2006 through the 4th Quarter 2006.
Key documents used by the inspectors in conducting this inspection are listed in the to this report.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
.1 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees corrective action program at an appropriate threshold, that adequate attention was given to ensure timely corrective actions, and that adverse trends were identified and addressed. This review did not count as an annual sample.
b. Findings
No findings of significance were identified.
.2 Selected Issue Follow-up Inspection
a. Inspection Scope
The inspectors selected an issue associated with diesel generator D6 generator bearing vibration measurements exceeding the manufacturers recommended limits for a more in-depth review in accordance with Inspection Procedure 71152, Identification and Resolution of Problems. This effort completed one in-depth review of the Problem Identification and Resolution inspection sample to review the corrective action aspects associated with this event. The inspectors reviewed the evaluation and corrective actions. The key documents reviewed by the inspectors associated with this inspection are listed in the Attachment to this report.
b. Findings and Observations
No findings of significance were identified. The licensee evaluated the impact of the vibration on the operability of the diesel generator and concluded that D6 was fully capable of performing its specified safety functions. The licensee has developed a corrective action plan to address the elevated vibration during the next scheduled maintenance period.
4OA3 Event Followup
(Closed) Licensee Event Report (LER) 05000306/2006-002-00: Unit 2 Event Monitoring Instrument Inoperable Longer Than Allowed by Technical Specifications.
On May 5, 2006, during a refueling outage on Unit 1, neutron flux monitor 1N51 and 1N52 displayed erratic indications. Troubleshooting and investigation involved purging moisture from the cables and performing a pressure test. The pressure test was not successful for 1N51 and subsequent inspection of the cables revealed that a cable splice connection sleeve did not include a shim that was required for the gap between the outside diameter of the cable and the inside diameter of the sleeve.
During the Unit 2 refueling outage, licensees staff inspected the cables for the Unit 2 neutron flux monitor channels 2N51 and 2N52. The licensees staff found that the respective cables for 2N51 and 2N52 did not have the required shim in the sleeves for the splice connections. The inspectors reviewed the LER, CAPs, evaluation, and corrective actions, and no findings of significance were identified. The issue was considered minor because none of the minor questions from IMC 0612, Appendix B, dated November 2, 2006, were answered in the affirmative. Specifically, the performance deficiency did not result in a loss of system safety function, and the inspectors did not identify any earlier opportunities for identification of the problem by the licensee. This LER is closed.
4OA5 Other Activities
.1 (Closed) Unresolved Item (URI) 050000282/2006002-02; 050000306/2006002-02:
Licensee Continuing On-site Tritium Well Sample Results Assessment The inspectors reviewed the licensees progress in investigating the cause of the seasonally elevated tritium levels in the water in a singular on-site monitoring well (P-10). The licensee conducted additional analysis of their on-site water monitoring program including a self-assessment and a hydrological review. Although there was not sufficient data to confirm a link or definitively exclude other potential sources, the licensee determined that the most likely contributor to the fluctuating tritium levels in the (P-10) well samples was radionuclide migration of discharges from the turbine building sump. The turbine building sump was a monitored effluent discharge pathway that, by plant design, may contain small but measurable amounts of tritiated water. The contents of the turbine building sump were routinely analyzed, characterized, and the radiological impact of any discharge to the land lock was evaluated in accordance with the Off-Site Dose Calculation Manual. The licensee continued to monitor and evaluate the analytical results from its on-site well water program relative to their groundwater protection initiative program. Anomalous sample results were assessed for radiological impact and identified findings were reported in the Annual Effluent Report. Based on the licensees evaluation, the continued onsite well monitoring program was in compliance with the Off-Site Dose Calculation Manual, this item is closed.
4OA6 Meeting(s)
.1 Exit Meeting
On April 4, 2007, the resident inspectors presented the inspection results to Mr. T. Palmisano and other members of his staff, who acknowledged the finding. The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.
.2 Interim Exit Meetings
Interim exits were conducted for:
- Biennial Operator Requalification Program inspection with Mr. J. Lash, Training Manager, on January 23, 2007; and
- Access control to radiologically significant areas and ALARA program inspection with Mr. P. Huffman, Plant Manager, on February 02, 2007, with a followup teleconference to discuss the final outcome of the inspection with Mr. J. Kivi, Compliance Engineer, on March 21, 2007.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- T. Palmisano, Site Vice President
- J. Sorensen, Site Director
- P. Huffman, Plant Manager
M.Carlson, Engineering Director
- J. Anderson, Radiation Protection Manager
- F. Forrest, Operations Manager
- J. Lash, Training Manager
- S. Northard, Nuclear Safety Assurance Manager
- R. Womack, Production Planning Manager
- J. Kivi, Regulatory Compliance Engineer
Nuclear Regulatory Commission
- R. Skokowski, Chief, Reactor Projects Branch 3
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
- 05000282/2007002-01 NCV Failure to Properly Control Access to a Locked High Radiation Area
Closed
- 05000282/2007002-01 NCV Failure to Properly Control Access to a Locked High Radiation Area
- 05000306/2006-002-00 LER Unit 2 Event Monitoring Instrument Inoperable Longer Than Allowed by Technical Specifications
- 05000282/2006002-02 URI Licensee Continuing On-site Tritium Well Sample
- 05000306/2006002-02 Results Assessment
Discussed
None Attachment