IR 05000255/2006006
| ML063040322 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 10/27/2006 |
| From: | Christine Lipa NRC/RGN-III/DRP/RPB4 |
| To: | Harden P Nuclear Management Co |
| References | |
| IR-06-006, IR-06-012 | |
| Download: ML063040322 (30) | |
Text
ber 27, 2006
SUBJECT:
PALISADES NUCLEAR PLANT NRC INTEGRATED INSPECTION REPORT 05000255/2006006 and 05000255/2006012
Dear Mr. Harden:
On September 30, 2006, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Palisades Nuclear Plant. The enclosed report documents the inspection findings which were discussed on September 26, 2006, with you and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
Based on the results of this inspection, two NRC-identified findings of very low safety significance (Green) were identified. Both of these findings were determined to involve a violation of NRC requirements. However, because the violations were of very low safety significance and because the issues have been entered into your corrective action program, the NRC is treating these findings as a non-cited violations (NCVs) consistent with Section VI.A.1 of the Enforcement Policy. If you contest the subject or severity of a NCV, you should provide a response with a basis for your denial, within 30 days of the date of this inspection report, 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 at the Palisades facility. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure 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). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Christine A. Lipa, Chief Branch 4 Division of Reactor Projects Docket No. 50-255 License No. DPR-20 Enclosure:
Inspection Report 05000255/2006006 and 0500025/2006012 w/Attachment: Supplemental Information cc w/encl: J. Cowan, Executive Vice President and Chief Nuclear Officer R. Fenech, Senior Vice President, Nuclear Fossil and Hydro Operations D. Cooper, Senior Vice President - Group Operations L. Lahti, Manager, Regulatory Affairs J. Rogoff, Vice President, Counsel and Secretary A. Udrys, Esquire, Consumers Energy Company S. Wawro, Director of Nuclear Assets, Consumers Energy Company Supervisor, Covert Township Office of the Governor State Liaison Office, State of Michigan L. Brandon, Michigan Department of Environmental Quality -
Waste and Hazardous Materials Division
SUMMARY OF FINDINGS
IR 05000255/2006006 and 05000255/2006012; 07/01/2006 - 09/30/2006; Palisades Nuclear
Plant; Operator Performance During Non-routine Evolutions and Event Follow-up.
This report covers a 3-month period of baseline inspections. The inspections were conducted by Region III inspectors and resident inspectors. This report includes two green findings, both of which were NCVs. 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.
NRC-Identified
and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation (NCV) of 10 CFR 50.65 "Requirements for monitoring effectiveness of maintenance at nuclear power plants." Specifically, contrary to 50.65(a)(2), the licensee failed to demonstrate that the performance of condition of the HPSI System had been effectively controlled through performance of appropriate maintenance, and did not place the system in 50.65(a)(1) status when system performance deteriorated. The licensee subsequently placed the HPSI system in 50.65(a)(1) status and entered the finding into their corrective action program.
The inspectors determined that not placing the system in (a)(1) status when performance deteriorated is more than minor because it matched an example in IMC 0612, Appendix E, "Examples of Minor Violations," as being more than minor.
The finding is of very low safety significance because the finding did not result in loss of a safety function. (Section 1R12)
- Green.
A Green NCV was self-revealed on March 29, 2006, when control valve CV-3070, left train HPSI sub-cooling valve for HPSI pump P-66B, failed to open during preventive maintenance. Subsequent investigation by the licensee identified that a design change had removed a support for the valve. The removal of this support caused the valve to bind. The finding is a violation of 10 CFR 50, Appendix B,
Criterion III. The licensee entered the finding into the corrective action program, repaired the valve and added additional support to prevent recurrence.
The inspectors concluded that the issue is more than minor because it affected the operability, reliability, and availability of a mitigating system. The inspectors concluded a phase 3 assessment was required based on the results of phase 1 and 2 assessments.
Following a phase 3 assessment, the Senior Reactor Analyst concluded that the finding is of very low safety significance. (Section 4OA5.2)
Licensee-Identified Violations
A violation of very low safety significance, which was identified by the licensee has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and corrective actions are listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant Status
The plant operated at or near full Rated Thermal Power during the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness
1R04 Equipment Alignment
.1 Partial Walkdowns
a. Inspection Scope
The inspectors completed five equipment alignment inspection samples by performing partial walkdowns of the following risk-significant plant equipment:
- Right train diesel 1-2 alignments including service water (SW) and fuel support system while left train emergency diesel generator (EDG) was out for testing
- EDG 1-2 during elevated risk due to adverse weather;
- Steam driven auxiliary feedwater pump (AFWP) during inoperability of motor driven AFWP; and
- "B" train high pressure injection system while "A" train was out of service for maintenance.
During the walkdowns, the inspectors verified that power was available, that accessible equipment and components were appropriately aligned, and that no open work orders for known equipment deficiencies existed which would impact system availability.
The inspectors also reviewed selected condition reports (CRs) related to equipment alignment problems and verified that identified problems were entered into the corrective action program. The documents reviewed during this inspection are listed in the attachment.
b. Findings
No findings of significance were identified.
.2 Complete Walkdown
The inspectors completed one semi-annual equipment alignment inspection sample by performing a complete walkdown of the SW system. Utilizing piping and instrumentation diagrams and system checklists, the inspectors verified that accessible system components were correctly aligned. The inspectors also reviewed open maintenance work orders to verify that the equipments safety function was not adversely impacted by pending work.
The inspectors reviewed selected CRs associated with the SW system to verify that identified problems were entered into the corrective action program with the appropriate significance characterization. The inspectors also verified that planned and completed corrective actions were appropriate.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
.1 Fire Area Walkdowns
a. Inspection Scope
The inspectors completed six fire protection inspection samples by touring the following areas in which a fire could affect safety-related equipment:
C AFWP room (Fire Area 24)
C Screenhouse - intake structure (Fire Area 9)
C Outside areas and post indicating valve positions C 1-C switchgear room (Fire Area 4)
C Control room (Fire Area 1)
C Turbine building 625' elevation (Fire Area 23D)
The inspectors verified that transient combustibles and ignition sources were appropriately controlled, and that the installed fire protection equipment in the fire areas corresponded with the equipment which was referenced in the Updated Final Safety Analysis Report (UFSAR), Section 9.6, "Fire Protection." The inspectors also assessed the material condition of fire suppression systems, manual fire fighting equipment, smoke detection systems, fire barriers and emergency lighting units. For selected areas, the inspectors reviewed documentation for completed surveillances to verify that fire protection equipment and fire barriers were tested as required to ensure availability.
The inspectors reviewed selected CRs associated with fire protection to verify that identified problems were entered into the corrective action program with the appropriate significance characterization. The inspectors also verified that planned and completed corrective actions were appropriate. The documents reviewed during this inspection are listed in the attachment.
b. Findings
No findings of significance were identified.
1R06 Flood Protection
a. Inspection Scope
The inspectors completed one inspection sample pertaining to flood protection measures for internal flooding events. The inspectors performed a walkdown of the component cooling water (CCW) room and its flood barriers to verify the flood barriers were in acceptable condition. The inspectors reviewed the licensees flood analysis for the CCW room and determined the analysis was consistent with configuration of the room.
Further, the inspectors reviewed CRs to verify that corrective actions for previously identified flood protection problems were appropriate and had been properly implemented.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance
a. Inspection Scope
The inspectors reviewed heat exchanger performance for the component cooling water system. The inspectors checked the licensee's data for heat exchanger flow to verify no evidence of fouling existed. The inspectors also reviewed the licensee's program for monitoring and ensuring the operability of the plant's heat exchangers. In addition, the inspectors verified that the heat exchangers were correctly categorized under the Maintenance Rule and verified that they were receiving the required maintenance. This represents one sample.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification
a. Inspection Scope
The inspectors completed one inspection sample of licensed operator requalification training by observing a crew of licensed operators during simulator training on August 30, 2006. The inspectors assessed the operators response to the simulated events which included a loss of pressurizer control and a loss of containment integrity.
The inspectors verified that the operators were able to effectively mitigate the events through accurate and timely implementation of applicable alarm response procedures; Off-Normal Procedure 4.2, "Loss of Containment Integrity," Off-Normal Procedure 18, "Pressurizer Pressure Control Malfunctions," and Emergency Operating Procedure 2.0, "Reactor Trip Recovery." The inspectors also observed the post-training critique to assess the licensee evaluators and the crews ability to self-identify performance deficiencies.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors completed two inspection samples pertaining to maintenance effectiveness by reviewing maintenance rule implementation activities for the following system and components:
- High Pressure Safety Injection System
- 125 Volt Vital DC Power The inspectors reviewed the licensee's implementation of the maintenance rule requirements to verify that component and equipment failures were evaluated and appropriately dispositioned. The inspectors also verified that the selected systems and components were scoped into the maintenance rule and properly categorized as (a)(1) or (a)(2) in accordance with 10 CFR 50.65.
b. Findings
Introduction The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50.65 "Requirements for monitoring effectiveness of maintenance at nuclear power plants." Specifically, contrary to 50.65(a)(2), the licensee failed to demonstrate effective control of performance of the HPSI System and did not place the system in (a)(1) status, establish goals and commence monitoring performance or condition when system performance deteriorated.
Description While performing a Maintenance Effectiveness inspection of the HPSI system in accordance with Inspection Procedure (IP) 71111.12, the inspectors noted that from May 10, 2005, to March 29, 2006, the system had experienced four Maintenance Preventable Functional Failures (MPFFs). As specified by the licensees Maintenance Rule Scoping Document, the licensee established performance criterion as less than two failures in a 24 month period. In addition, during the review of the HPSI system Maintenance Rule documents, the inspectors noted that the Mitigating Systems Performance Index (MSPI) evaluation resulted in White.
Following the fourth MPFF on March 29, 2006, the licensee wrote an Action Request (AR) to evaluate the HPSI system for placement in 10 CFR 50.65(a)(1). The licensee concluded that because the failures were not related to each other and corrective actions had been taken for failures, the HPSI system did not warrant placement in 10 CFR 50.65(a)(1). The inspectors questioned the licensees conclusion that the system should remain in 10 CFR 50.65(a)(2) given the number of failures and the licensee reiterated their conclusion. On August 24, 2006, the Maintenance Rule expert panel met and recommended to the engineering systems manager that the system be placed in 10 CFR 50.65(a)(1). Subsequently, the engineering systems manager approved the decision and placed the system in 10 CFR 50.65(a)(1) status on August 31, 2006 and commenced monitoring performance against licensee-established goals.
The inspectors concluded that effective control of the HPSI system performance had not been demonstrated as evidenced by the four MPFFs in less than 12 months, compounded with a white MSPI. In addition, the inspectors concluded that the licensee had ample time and opportunities prior to the start of the inspection to appropriately evaluate the effectiveness of the maintenance of the system. Regardless of the causes for the four failures, the licensee should have established goals and commenced monitoring the performance or condition of the system under 10 CFR 50.65(a)(1) when the multiple MPFFs indicated that system performance was not being effectively controlled through appropriate maintenance, to ensure that the system remained capable of performing its intended safety function. The licensees failure to place the system in 10 CFR 50.65(a)(1) status when performance deteriorated resulted in the system not being monitored against licensee-established goals.
Analysis The inspectors determined that the failure to place the HPSI System into 10 CFR 50.65(a)(1) when performance was not being effectively controlled through preventive maintenance was a licensee performance deficiency that warranted review in accordance with the Significant Determination Process. Using the IMC 0612, Appendix E, "Examples of Minor Violations," the inspectors determined that the finding was more than minor. Specifically the finding matches example 7.b.,
in that violations of 10 CFR 50.65(a)(2), are not minor because they necessarily involve degraded system performance.
To assess the significance of the finding, the inspectors used Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations" of IMC 0609, "Significance Determination Process," dated November 22, 2005. The inspectors determined that the finding was of very low safety significance, in accordance with the Phase 1 screening worksheet, because:
- (1) it did not represent an actual loss of safety function of a system;
- (2) it did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time;
- (3) it did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk-significant per 10 CFR 50.65 for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and
- (4) it did not screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.
Enforcement 10 CFR 50.65(a)(1), requires, in part, that the holders of an operating licensee shall monitor the performance or condition of structures, systems, or components (SSCs)within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems and components, are capable of fulfilling their intended functions.
10 CFR 50.65(a)(2) states, in part, that monitoring as specified in 10 CFR 50.65(a)(1) is not required where it has been demonstrated that the performance of condition of an SSC is being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remains capable of performing its intended function.
Contrary to the above, the licensee failed to demonstrate that the performance or condition of HPSI System had been effectively controlled through performance of appropriate maintenance, yet, did not monitor against licensee-established goals.
Specifically, from May 10, 2005, to March 29, 2006, the HPSI system experienced four MPFFs, which showed that performance was not being effectively controlled since the performance criterion specified for the system was less than two MPFFs in a 24 month period. Subsequently, on August 31, 2006, the licensee placed the system in 10 CFR 50.65(a)(1). However, because of the very low safety significance and because the issue has been entered into the licensees corrective action program (AR 01052257), the issue is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000255/2006006-01)
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a. Inspection Scope
The inspectors completed four inspection samples. The inspectors reviewed the following four activities to verify that the appropriate risk assessments were performed prior to removing equipment for work. The inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4), and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors verified the appropriate use of the licensees risk assessment tool and risk categories in accordance with Administrative Procedure 4.02, Control of Equipment, Revision 29, and Fleet Procedure FP-OP-RSK-01, Risk Monitoring and Risk Management, Revision 0.
Documents reviewed are listed in the attachment.
- Yellow risk path due to emergent adverse weather and EDG maintenance activities
- Yellow risk due to scheduled EDG maintenance activities
- Emergent work on load center 13
- Yellow risk due to EDG testing and main transformer work The inspectors also verified that CRs related to emergent equipment problems were entered into the corrective action program with the appropriate significance characterization. The inspectors reviewed selected CRs related to risk management during maintenance activities to verify that planned corrective actions were appropriate and had been implemented as scheduled.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations
a. Inspection Scope
For the four operability evaluations described in the Operability Recommendations (OPRs) listed below, the inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors reviewed the UFSAR to verify that the system or component remained available to perform its intended function. In addition, the inspectors reviewed compensatory measures implemented to verify that the compensatory measures worked as stated and the measures were adequately controlled. In addition, the inspectors verified that the CRs generated for equipment operability issues were entered into the licensees corrective action program with the appropriate significance characterization. Documents reviewed are listed in the attachment.
- Operability of EDG 1-2 from October 28 to November 20, 2005
- Operability of SW System with pin hole leak
- Operability of AFWP with questionable oil sample results
- Operability of EDG 1-2 with erratic fuel oil pressure profile
b. Findings
No findings of significance were identified.
1R17 Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed one permanent plant modification package that involved isolating instrument air to two SW supply valves. The inspectors reviewed the design change information, related design basis documents and the 10 CFR 50.59 screening evaluation to verify that the design bases, licensing bases and performance capability of the involved diesel generator system were not degraded by this modification. In addition, the inspectors reviewed applicable plant documents to verify that any appropriate changes were made. Documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors completed one inspection sample pertaining to post maintenance testing by assessing testing activities that were conducted for the following maintenance activities:
- Testing of EDG 1-2 after replacement of a temperature switch The inspectors observed portions of the post maintenance testing and reviewed documentation to verify that the tests were performed as prescribed by the work orders and test procedures; that applicable testing prerequisites were met prior to the start of the tests; and the effect of testing on plant conditions was adequately addressed by the control room operators. The inspectors reviewed documentation to verify the test criteria and acceptance criteria were appropriate for the scope of work performed; reviewed test procedures to verify the tests adequately verified system operability; and reviewed documented test data to verify the data was complete, and that the equipment met the prescribed acceptance criteria. Further, the inspectors reviewed CRs to verify that post maintenance testing problems were entered into the corrective action program with the appropriate significance characterization. For select CRs, the inspectors verified that the corrective actions were appropriate and implemented as scheduled.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors witnessed four surveillance tests and/or reviewed test data of selected risk-significant SSCs, listed below, to assess, as appropriate, whether the SSCs met the requirements of the TS; the USAR; Palisades Administrative Procedure 9.20, TS Surveillance and Special Testing Program; Engineering Manual EM-09-02 and EM-09-04, Inservice Testing of Plant Valves and Inservice Testing of selected safety-related pumps. One of the samples was an inservice test and one was an isolation valve. The inspectors also determined whether the testing effectively demonstrated that the SSCs were operationally ready and capable of performing their intended safety functions.
Further, the inspectors reviewed selected CRs regarding surveillance testing activities.
The inspectors verified that the identified problems were entered into the licensees corrective action program with the appropriate significance characterization and that the planned and completed corrective actions were appropriate. Additional documents reviewed are listed in the attachment.
C COP-22A, Diesel Fuel Oil Testing Program C QO-5, Valve Test Procedure - Includes Containment Isolation Valves (isolation valve)
C QO-21, Inservice Test for AFWPs (IST)
C RT-202, Control Room HVAC Heat Removal Capability
b. Findings
No findings of significance were identified.
1R23 Temporary Plant Modifications
a. Inspection Scope
The inspectors completed one baseline inspection sample by reviewing the following temporary modification:
- Installation of new non safety-related Diesel Generator 1-3 The inspectors reviewed the design documents and 10 CFR 50.59 safety screening to verify that the temporary modification did not affect the operability of the related systems and other interfacing systems. The inspectors reviewed documentation to verify that the modification was implemented as designed. Post modification testing results were reviewed to verify that the system functioned as intended after the modification was implemented.
b. Findings
No findings of significance were identified.
1EP6 Emergency Preparedness Drill Evaluation
a. Inspection Scope
The inspectors observed emergency response personnel in the Technical Support Center during an emergency planning drill on August 15, 2006. Emergency Preparedness personnel had pre-designated that the opportunities to classify the event and make protective action recommendations would be evaluated and included in the performance indicator data regarding drill and exercise performance.
The inspectors verified the emergency classifications, notifications to offsite agencies, and the development of protective action recommendations were completed in an accurate and timely manner as required by the emergency plan implementing procedures The inspectors also verified that the drill was conducted in accordance with the prescribed sequence of events and that the drill objectives were met.
The inspectors observed the post-drill critique in the Technical Support Center to verify that emergency response personnel and drill evaluators adequately self-identified performance problems. The inspectors reviewed the post-drill critique report to verify that the data regarding the indicator for drill and exercise performance was accurate.
Condition reports generated for identified drill performance problems were reviewed to verify that the problems were entered into the corrective action program with the appropriate significance characterization. This represents one sample.
b. Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1 Plant Walkdowns/Boundary Verifications and Radiation Work Permit Reviews
a. Inspection Scope
The inspectors reviewed the licensees physical and programmatic controls for highly activated and/or contaminated materials (non-fuel) that could be stored within the spent fuel pool. Specifically, applicable radiation protection (RP) procedures were reviewed, RP staff were interviewed, and a walkdown of the refuel floor was conducted. Although highly activated/contaminated materials are not stored in the spent fuel pool in a manner that readily allowed their inadvertent movement, the radiological controls for the storage of such materials was discussed with RP staff to ensure adequate barriers would be established should the licensee change its practices. This review represented one sample.
b. Findings
No findings of significance were identified.
.2 Problem Identification and Resolution
a. Inspection Scope
The inspectors reviewed the licensees documentation for all potential Performance Indicator (PI) events occurring since the last radiological access control inspection in March 2004 to determine if any of these events involved dose rates greater than 25 Rem/hour at 30 centimeters or greater than 500 Rem/hour at 1 meter or involved unintended exposures greater than 100 millirem total effective dose equivalent (or greater than 5 Rem shallow dose equivalent or greater than 1.5 Rem lens dose equivalent). None were identified. This review represented one sample.
b. Findings
No findings of significance were identified.
.3 High Risk Significant, LHRA and Very High Radiation Area (VHRA) Access Controls
a. Inspection Scope
The inspectors discussed with RP staff the controls that were in place for areas that had the potential to become high or locked high radiation areas during certain plant operations to determine if these plant operations required communication before hand with the RP group, so as to allow corresponding timely actions to properly post and control the radiation hazards. In particular, reactor operations procedures and RP procedures/job files developed to identify vulnerable areas subject to changing radiological conditions were reviewed and their implementation was discussed with RP supervisory staff. This review represented one sample.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES (OA)
4OA1 Performance Indicator Verification
Cornerstones: Occupational and Public Radiation Safety
.1 Radiation Safety Strategic Area
a. Inspection Scope
The inspectors sampled the licensees PI submittals for the periods listed below. The inspectors used PI definitions and guidance contained in Revision 3 of Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, to verify the accuracy of the PI data. The following PIs were reviewed:
- Occupational Exposure Control Effectiveness:
The inspectors reviewed the licensees assessment of the PI for occupational radiation safety, to determine if indicator related data was adequately assessed and reported during the previous four quarters. The inspectors compared the licensees PI data with the CR database, reviewed radiological restricted area exit electronic dosimetry transaction records, and conducted walkdowns of accessible locked high radiation area entrances to verify the adequacy of controls in place for these areas. Data collection and analysis methods for PIs were discussed with licensee representatives to determine if there were any unaccounted for occurrences in the Occupational Radiation Safety PI as defined in Revision 3 of Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline." This review represented one sample.
- Radiological Environmental TS/Off-site Dose Calculation Manual (ODCM)
Radiological Effluent Occurrences:
The inspectors reviewed data associated with the RETS/ODCM PI to determine if the indicator was accurately assessed and reported. This review included the licensees CR database for the previous four quarters, to identify any potential occurrences such as unmonitored, uncontrolled or improperly calculated effluent releases that may have impacted offsite dose. The inspectors also selectively reviewed gaseous and liquid effluent release data and the results of associated offsite dose calculations and quarterly PI verification records generated over the previous four quarters. Data collection and analyses methods for PIs were discussed with licensee representatives to determine if the process was implemented consistent with industry guidance in Revision 3 of Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline." This review represented one sample.
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 CRs were being generated and entered into the corrective action program with the appropriate significance characterization. For select CRs, the inspectors also verified that identified corrective actions were appropriate and had been implemented or were scheduled to be implemented in a timely manner commensurate with the significance of the identified problem.
b. Findings
No findings of significance were identified.
.2 Semi-Annual Trend Review
a. Inspection Scope
As required by IP 71152, "Identification and Resolution of Problems," the inspectors performed a review of the licensees corrective action program (CAP) action requests to identify trends that could indicate the existence of a more significant safety issue. The inspectors also reviewed the Operations Department Monthly Performance Report dated March 2006, the Site DRUM Report for the 1st Quarter of 2006 and the Corrective Action Program Performance Indicator Summary, dated May 2006. The inspectors review for potential trends included the results from the daily inspector CAP item screening discussed in Section 4OA2.1. The plant CAP action request screening meetings were observed to review the licensees level of effort in identifying potential trends, and any associated corrective actions that were planned or implemented. In addition, the inspectors reviewed issues documented outside the normal CAP that included, maintenance work orders, component status reports, performance indicators and Operations control room logs. The inspectors review nominally considered the 6 month period of January through June 2006. The inspectors compared and contrasted their results with the results obtained by the licensee during previous internal reviews.
b. Findings
No findings of significance were identified.
.3 Annual Sample for Follow-up Inspection
a. Inspection Scope
The inspectors completed one inspection sample for annual reviews of selected follow-up by reviewing the cumulative effects of operator work-arounds. The inspectors reviewed licensee practices for the identification, review and assessment of operator work-arounds including any cumulative impact. This inspection included review of the licensees Administrative Procedure 4.12, "Operator Work-around Program" and biweekly "Palisades OPS Burden Status Report," which also addressed other "operator challenges," and discussions with operations supervisors about program implementation.
b. Findings
No findings of significance were identified.
4OA3 Event Follow-up
.1 The inspectors reviewed Licensees Event Report (LER) 2005-07-00, "Inoperable
Emergency Diesel Generator for a Time Longer than Permitted by TSs" and subsequent correspondence that retracted the LER. The LER reported a condition where a failed snubber valve resulted in both diverting fuel from a single cylinder as well as spreading combustible fuel. In order to understand the licensees rational for the LER retraction, the inspectors reviewed the basis documents prepared to support the retraction. Based on this review, the inspectors concluded additional information was needed to determined the acceptability of retracting the LER. Unresolved Item (URI)05000255/2006004-06 addressing assessment of the diesel operability remains open.
This LER remains open.
4OA5 Other Activities
.1 (Closed) URI 05000255/2005202-01, "Verification of contents of two storage tubes
which contain rod fragments from assembly 1-024 "
The inspectors reviewed the licensees response to the subject URI and discussed this issue with the NRC Senior MC&A Physical Scientist who identified the issue originally.
The inspectors reviewed records developed in conjunction with the physical inventory of a failed fuel rod. The inspectors compared the results with documentation developed when the failed fuel was initially loaded into tubes. Based on these records, the inspectors concluded the licensee completed the physical inventories to verify the contents of storage tubes containing fuel pin fragments. These inventories confirmed the location of the fuel pins. This URI is closed.
.2 (Closed) URI 05000255/2006004-07,"Control Valve CV-3070 Fails to Stoke"
a. Inspection Scope
In NRC Inspection Report 05000255/2006004, the inspectors documented URI 05000255/2006004-07 regarding the failure of valve CV-3070, left train HPSI sub-cooling control valve, to stroke during maintenance. The licensee has performed and the NRC has reviewed the analysis of the valves capability needed to close this URI. The licensees analysis concluded the valve was not operable. This item is closed.
b. Findings
Introduction A Green NCV was self-revealed on March 29, 2006 when control valve CV-3070, left train HPSI sub-cooling valve for HPSI pump P-66B, failed to open during preventive maintenance. Subsequent investigation by the licensee identified that a design change had removed a support for the valve. The removal of this support caused the valve to bind.
Description While performing maintenance to inspect and repair the valves oiler, the licensee attempted to test stroke the valve. Once the valve failed to stroke, the plant entered the applicable 72-hour limiting condition for operation action 3.5.2B.1, that requires the train to be restored to operable. After troubleshooting the failure, the licensee determined the valve could not be repaired within the Limiting Condition for Operation (LCO) action time and shut down the plant as required.
Troubleshooting of the Control valve revealed the valve could open once it was mechanically agitated. In addition, diagnostic testing of CV-3070 showed the stem movement of the valve was not smooth. These test results indicated internal valve component interference. Subsequent disassembly of the valve indicated the valves stem was interacting with the back seat surface, increasing the force needed to open the valve.
The licensees causal analysis identified that in 1994 a support for the valve had been removed. A document search did not locate the design basis for the hanger. In addition, the licensee identified that the valve had also failed to stroke on 2003. The licensee concluded that inadequate support for the valve and its operator caused the internal components to bind and the valve to fail to stroke. Prior to restarting the reactor, the licensee repaired the valve and added additional support to prevent recurrence.
Analysis The inspectors determined that the failure to adequately design the valve support was a performance deficiency. The inspectors assessed this finding using the SDP. The inspectors reviewed the samples of minor issues in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," and determined that there were no examples related to this issue. Consistent with the guidance in IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening, "the inspectors determined that the finding was of more than minor significance because the failure of the valve to stroke is associated with the operability, reliability and availability of a mitigating system. Specifically, the sub-cooling valve is credited with maintaining subcooling of water to the HPSI pump when the plant is on sump recirculation.
Phase 1 Assessment The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in accordance with IMC 0609, Significance Determination Process, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-power Situations," Attachment 1. The inspectors determined that the finding represented an actual loss of safety function of a single train of safety-related equipment for greater than its TS allowed outage time and a Phase 2 SDP evaluation was warranted.
Phase 2 Assessment The inspectors utilized all of the initiating event worksheets except the "Anticipated Transient without Scram" and "Loss of Service Water " worksheets and solved only those sequences that involved the high pressure recirculation function with a duration of 3-30 days. Based on the results of the SDP worksheets, the inspectors determined that the finding was potentially of low to moderate safety significance (White). The regional senior reactor analyst (SRA) reviewed these results and determined that an SDP Phase 3 assessment was necessary to refine the risk characterization.
Phase 3 Assessment Internal Events - The SRA performed the risk evaluation using the Palisades Standardized Plant Analysis Risk (SPAR) Model, Level 1, Revision 3P, Change 3.21, created October 2005. The SRA ran the SPAR model assuming failure of Valve CV-3070 to open throughout an exposure time of 672 hours0.00778 days <br />0.187 hours <br />0.00111 weeks <br />2.55696e-4 months <br /> (28 days). The SRA obtained a change in core damage frequency (CDF) of 4.9E-7 (Green) for internal events. The dominant sequences involved loss of coolant accident scenarios with failure of high pressure recirculation.
External Events - The SRA reviewed the Individual Plant Examination of External Events" report for Palisades dated February 1998 and the Palisades Fire Hazards Analysis Report, Revision 6, for insights on risk contribution from external events. In reviewing these documents, Valve CV-3070 was not shown to be needed or relied upon to support post-fire safe shutdown. The SRA determined that the risk contribution of this finding due to fire is insignificant. Regarding seismic and flooding contributors to risk, none of the dominant contributors involved Valve CV-3070. The SRA determined that the risk contribution of this finding due to seismic and flooding events was likewise insignificant.
Large Early Release Frequency (LERF) - Using IMC 0609, Appendix H, "Containment Integrity Significance Determination Process," the SRA determined that this was a Type "A" finding for a pressurized water reactor with large dry containment. Using Table 5.1, the SRA concluded that none of the accident sequences were contributors to LERF. The attributes considered in Table 5.1 were inter-system loss of coolant accidents and steam generator tube ruptures. None of these scenarios was impacted by this finding. The SRA concluded that the LERF was negligible and did not contribute to the risk associated with this finding.
Significance Determination Conclusion- -The SRA concluded that the total CDF for this issue, considering internal events, external events, and LERF was estimated to be 4.9 E-7 (Green).
Enforcement 10 CFR 50 Appendix B Criterion III, requires, in part, that "Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the licensee application , for those structures systems and components to which this appendix applies are correctly translated into specifications , drawings, procedures, and instructions...." Contrary to these requirements, the licensee failed to include the hanger in design documentation. In addition, analysis to support removal of the hanger failed to recognize the potential for internal valve binding. As a result, the licensee removed the hanger, thus permitting the valve to bind. However, because this violation was of very low safety significance and because the issue was entered into the licensee's corrective action program as CAP01021152, this violation is being treated as an NCV, consistent with Section VI.A.1 of the Enforcement Policy (NCV 05000255/2006006-02). The licensee's initial corrective action included performing a plant shutdown.
.3 Temporary Instruction (TI) 2515/169 Mitigating System Performance Index Verification
a. Inspection Scope
The inspectors reviewed the licensees implementation of MSPI reporting guidance in accordance with TI 2515/169. The inspectors reviewed a sample of procedures where unavailability was excluded due to either the short duration of the unavailability or where the licensee took credit for operator action to recover. For the MSPI systems, the inspectors verified the baseline planned unavailability hours and confirmed the hours were correctly translated into the basis document. In addition, the inspectors selected a sample of planned and unplanned unavailable hours and confirmed through operating logs, maintenance records or CRs that the data was accurate. Finally, the inspectors confirmed the accuracy of failure data on a sampling basis.
The inspectors determined that the licensee accurately documented baseline planned unavailability data; and accurately documented actual unavailability hours. The inspectors determined that one event related to the 1-2 EDG may result in an additional failure. If included, the failure would result in the Emergency AC Power MSPI index crossing the Green/White threshold. This issue is already the subject of URI 05000255/2006004-06. No additional errors were identified in submitted data.
No significant discrepancies were identified in the basis document.
b. Findings
No findings of significance were identified.
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Mr. P. Harden and other members of licensee management on September 26, 2006. Licensee personnel acknowledged the findings presented. The inspectors asked licensee personnel whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
.2 Interim Exit Meetings
An interim exit meeting was conducted for:
- Occupational radiation safety program for access control and Performance Indicator verification with Mr. B. Patrick and Ms. B. Dotson on October 3, 2006.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance was identified by the licensee and is a violation of NRC requirements that meets the criteria of Section VI of the NRC Enforcement Policy for being dispositioned as an NCV.
Cornerstone: Mitigating Systems
.1 On June 9, 2006, a Nuclear Control Operator (NCO) identified that LCO 3.0.3 had been
inadvertently entered due to two Safety Injection Tank (SIT) bottles being rendered inoperable at the same time for approximately 40 seconds. Specifically, while performing RI-15C, SIT Level Channel Calibration on T-82B, the NCO vented T-82D due to a pressure rise. The Control Room Supervisor was subsequently made aware of the evolution. Technical Specification 5.4, requires procedures to be implemented and maintained for safety related equipment. Contrary to this requirement, the licensee failed to implement procedures to control operability of the SITs and unknowingly entered LCO 3.0.3. The inspectors reviewed the finding in accordance with IMC 0609 and concluded that since the two SITs were inoperable simultaneously for only 40 seconds and it was recognized promptly by the operating crew, the finding was of very low safety significance. The licensee entered this issue into their corrective action program as AR 01034847.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- P. Harden, Site Vice President
- G. Baustian, Training Manager
- T. Blake, Nuclear Safety Assurance Manager
- M. Carlson, Engineering Director
- B. Dotson, Regulatory Compliance
- G. Baustian, Training Manager
- G. Hettel, Plant Manager
- L. Lahti, Licensing Manager
- D. Malone, Regulatory Affairs
- C. Moeller, Radiation Protection General Supervisor - Ops
- D. Nestle, Radiation Protection General Supervisor - Technical
- B. Patrick, Radiation Protection & Chemistry Manager
- K. Yeager, Assistant Operations Manager
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
- 05000255/2006006-01 NCV The licensee failed to demonstrate that the performance or condition of High Pressure Injection System had been effectively controlled per 10 CFR 50.65 (Section R12 )
- 05000255/2006006-02 NCV Control Valve CV-3070 Failed to Stroke (Section 4OA5.2)
Closed
- 05000255/2006006-01 NCV The licensee failed to demonstrate that the performance or condition of High Pressure Injection System had been effectively controlled (Section R12)
- 05000255/2006006-02 NCV Control Valve CV-3070 Failed to Stroke (Section 4OA5.2)
- 05000255/2006004-07 URI Control Valve CV-3070 Fails to Stroke (Section 4OA5.2)
- 05000255/2005202-01 URI Verification of contents of two storage tubes which contain rod fragments from assembly 1-024 (Section 4OA5.1)
Discussed
- 05000255/2006004-06 URI Failure of Component on 1-2 EDG Causes Surveillance Failure (Section 4OA5.3)
Attachment