IR 05000456/2007004
| ML072070555 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 07/26/2007 |
| From: | Richard Skokowski NRC/RGN-III/DRP/RPB3 |
| To: | Crane C Exelon Generation Co, Exelon Nuclear |
| References | |
| FOIA/PA-2010-0209 IR-07-004 | |
| Download: ML072070555 (47) | |
Text
uly 26, 2007
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2 NRC INTEGRATED INSPECTION REPORT 05000456/2007004 AND 05000457/2007004
Dear Mr. Crane:
On June 30, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Braidwood Station, Units 1 and 2. The enclosed report documents the inspection results, which were discussed on June 29, 2007, with Mr. G. Boerschig 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 were identified. Both findings involved violations of NRC requirements. However, because of their very low safety significance, and because the issues were entered into your corrective action program, the NRC is treating the issues as Non-Cited Violations (NCVs) in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
If you contest the subject or severity of a finding, 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 Resident Inspector Office at the Braidwood Station. In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS)
component of NRCs 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. 50-456; 50-457 License Nos. NPF-72; NPF-77 Enclosure: Inspection Report 05000456/2007004 and 05000457/2007004 w/Attachment: Supplemental Information cc w/encl: Site Vice President - Braidwood Station Plant Manager - Braidwood Station Regulatory Assurance Manager - Braidwood Station Chief Operating Officer Senior Vice President - Nuclear Services Vice President - Operations Support Vice President - Licensing and Regulatory Affairs Director Licensing Manager Licensing - Braidwood and Byron Senior Counsel, Nuclear, Mid-West Regional Operating Group Document Control Desk - Licensing Assistant Attorney General Illinois Emergency Management Agency State Liaison Officer Chairman, Illinois Commerce Commission
SUMMARY OF FINDINGS
IR 05000456/2007004, 05000457/2007004; 04/01/2007 - 06/30/2007; Braidwood Station,
Units 1 & 2; Radiation Safety Event Followup Inspection.
This report covers a three-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. Two Green findings were identified by the inspectors. Both findings were considered non-cited violations (NCVs) of NRC regulations.
The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 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 4, dated December 2006.
NRC-Identified and Self-Revealing Findings
Cornerstone: Occupational and Public Radiation Safety
- Green.
A finding of very low safety significance and an associated NCV of 10 CFR 20.1101(a) was identified by the inspectors for the licensees failure to implement a radiation protection program commensurate with licensed activities and the ongoing radiological issues at the plant. Specifically, radiological controls were not effectively applied to secondary systems, which contained contaminated (tritium) fluids, to ensure that worker exposures and radiological effluents were fully monitored and controlled.
The finding is greater than minor because it was associated with the process and procedures attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective to ensure the adequate protection of worker health and safety from exposure to radiation from radioactive material during nuclear reactor operation. Specifically, the inspectors determined that the finding did not involve unintended collective dose resulting from a deficiency in As-Low-As-Reasonably-Achievable (ALARA) planning, work control, or exposure control. The inspectors also determined that the finding did not involve an overexposure, the substantial potential for an overexposure, and did not compromise the licensees ability to assess dose. Consequently, the inspectors concluded that the finding was of very low safety significance. Corrective actions taken by the licensee included characterizing secondary systems to determine tritium concentration and prescribing radiological coverage and contamination control requirements for each system based upon this characterization. The cause of the finding was related to a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not address radiological safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity (P.1(d)). (Section 4OA3.2)
- Green.
A finding of very low safety significance and an associated NCV of 10 CFR 20.1902(e) was identified by the inspectors for the failure to post areas in which licensed material is used or stored. Specifically, two waste water lagoons, located within the Protected Area, and the Turbine Building each contained greater than 10,000 uCi of tritium and were not posted in accordance with 10 CFR 20.1902(e).
The finding is greater than minor because it was associated with the process and procedures attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during nuclear reactor operation. Specifically, the inspectors determined that the finding did not concern unintended collective dose resulting from a deficiency in ALARA planning, work control or exposure control. The inspectors also determined that the finding did not involve an overexposure, the substantial potential for an overexposure, and did not compromise the licensees ability to assess dose. Consequently, the inspectors concluded that the Significance Determination Process (SDP) assessment for this finding was of very low safety significance. Corrective actions taken by the licensee included posting the lagoons and areas of the turbine building appropriately as "CAUTION, RADIOACTIVE MATERIAL(S). The cause of the finding was related to a cross-cutting aspect in the area of Human Performance because the licensee did not use conservative assumptions in decision making (H.1(b)). (Section 4OA3.2)
Licensee-Identified Violations
None.
REPORT DETAILS
Summary of Plant Status
Unit 1 was operated at or near full power until June 27, 2007, when a lightning strike on an offsite power line created a grid disturbance resulting in an overcurrent trip of the 1D reactor coolant pump and subsequent automatic reactor trip. Unit 1 was restarted and synchronized to the grid the next day and was returned to full power over the remainder of the inspection period.
Unit 2 was operated at or near full power for the entire inspection period except that power was reduced to about 85 percent on May 12 through 14, 2007, for turbine valve testing.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
Readiness for Seasonal Susceptibilities
a. Inspection Scope
The inspectors monitored and reviewed the licensees preparations for summer weather conditions. This effort was accomplished by verifying that the licensee had completed the requirements for summer readiness as required by Exelon Corporate Procedure WC-AA-107, Seasonal Readiness. The inspectors also reviewed the Updated Final Safety Analysis Report (UFSAR) and the Technical Specifications (TS) to identify systems and components potentially susceptible to failure in high temperature conditions. The inspectors verified that the licensee had addressed these components in preparation for summer conditions. In addition, the inspectors selected the following risk-significant systems/areas for a detailed walkdown and review:
- lake screen house;
- refueling water storage tanks; and
- rod control system.
The inspectors also reviewed the licensees procedures for coping with adverse electrical grid conditions, with emphasis on the licensees communications protocol with grid operators. Documents reviewed during this inspection are listed in the Attachment.
This inspection constituted one sample of the inspection requirement for site readiness prior to the onset of extreme seasonal weather.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment
Partial Walkdowns
a. Inspection Scope
The inspectors performed partial walkdowns of the accessible portions of risk-significant system trains during periods when the train was of increased importance due to redundant trains or other equipment being unavailable. The inspectors utilized the valve and electric breaker lists to determine whether the components were properly positioned and that support systems were aligned as needed. The inspectors also examined the material condition of the components and observed operating parameters of equipment to determine whether there were any obvious deficiencies. The inspectors reviewed Issue Reports (IRs) associated with the train to determine whether those documents identified issues affecting train function. The inspectors used the information in the appropriate sections of the TS and the UFSAR to determine whether the licensee had maintained the functional requirements of the system. The inspectors also reviewed the licensees identification of and controls over the redundant risk-related equipment required to remain in service. Documents reviewed during this inspection are listed in the Attachment.
The inspectors completed three samples of this requirement by walkdowns of the following trains:
- 1B residual heat removal (RH) train with the 1A RH train in a planned outage; and
- 2A auxiliary feedwater (AF) train with the 2B AF pump in a corrective maintenance outage.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
Quarterly Inspection
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of fire fighting equipment, the control of transient combustibles and ignition sources, and on the condition and operating status of installed fire barriers. 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. Also reviewed was the revised Individual Plant Examination of External Events, which contained additional insights on selected fire areas that impact equipment potentially causing plant transient or adversely affecting safe shutdown capability. The inspectors used the Fire Protection Report, Revision 22, to determine: that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and that fire doors, dampers, and penetration seals appeared to be in satisfactory condition.
The inspectors completed nine samples of this inspection requirement during the following walkdowns:
- Unit 1 steam tunnels and main steam isolation valve rooms (Zone 18.3-1);
- Unit 2 steam tunnels and main steam isolation valve rooms (Zone 18.3-2);
- Unit 1 containment mechanical penetration area, auxiliary building 364 elevation (Zone 11.3-1);
- Unit 2 containment mechanical penetration area, auxiliary building 364 elevation (Zone 11.3-2);
- Unit 1 auxiliary electrical equipment room (Zone 5.5-1);
- Unit 2 auxiliary electrical equipment room (Zone 5.5-2);
- 11 engineered safeguards features switchgear room (Zone 5.2-1);
- 12 engineered safeguards features switchgear room (Zone 5.1-1); and
- 21 engineered safeguards features switchgear room (Zone 5.2-2).
The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program. Documents reviewed during this inspection are listed in the Attachment.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures
Internal Flood Protection Features
a. Inspection Scope
The inspectors reviewed Braidwoods flood analysis and design basis documents to identify design features important to internal flood protection, and flood protection measures in place to prevent or mitigate effects of internal flooding. For this sample, the inspectors focused on the 330-foot elevation of the auxiliary building, the essential service water (SX) pump rooms. The inspectors examined the flood doors, leak detection sumps, normal sumps, and cross connections between the auxiliary building and turbine building floor drain systems. This review represented one annual inspection sample. Documents reviewed during this inspection are listed in the Attachment.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program
Quarterly Review of Testing/Training Activity
a. Inspection Scope
The inspectors observed operating crew performance during an evaluated simulator examination scenario involving a faulted steam generator and an anticipated transient without SCRAM following various instrument and controller failures.
The inspectors evaluated crew performance in the following areas:
- clarity and formality of communications;
- ability to take timely actions in the safe direction;
- prioritization, interpretation, and verification of alarms;
- procedure use;
- control board manipulations;
- oversight and direction from supervisors; and
- group dynamics.
Crew performance in these areas was compared to licensee management expectations and guidelines.
The inspectors verified that the crew completed the critical tasks listed in the simulator guide. The inspectors also compared simulator configurations with actual control board configurations. For any weaknesses identified, the inspectors observed the licensee evaluators to determine whether they also noted the issues and discussed them in the critique at the end of the session. Documents reviewed are listed in the Attachment.
This review constituted one sample of this inspection requirement.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
Routine Inspection
a. Inspection Scope
The inspectors reviewed the licensees overall maintenance effectiveness for selected plant systems. This evaluation consisted of the following specific activities:
- observing the conduct of planned and emergent maintenance activities where possible;
- reviewing selected IRs, open work orders, and control room log entries in order to identify system deficiencies;
- reviewing licensee system monitoring and trend reports;
- attending various meetings throughout the inspection period where the status of maintenance rule activities was discussed;
- conducting partial walkdowns of the selected system; and
- interviewing appropriate system engineers.
The inspectors also reviewed whether the licensee properly implemented Maintenance Rule, 10 CFR 50.65, for the affected systems. Specifically, the inspectors determined whether:
- the system was scoped in accordance with 10 CFR 50.65;
- performance problems constituted maintenance rule functional failures;
- the system had been assigned the proper safety significance classification;
- the system was properly classified as (a)(1) or (a)(2); and
- the goals and corrective actions for the system were appropriate.
The above aspects were evaluated using the maintenance rule program and other documents listed in the Attachment. The inspectors also verified that the licensee was appropriately tracking reliability and/or unavailability for the systems. The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program. Documents reviewed in this inspection are listed in the Attachment.
The inspectors completed three samples in this inspection requirement by reviewing the following systems and equipment performance issues:
- SX subsequent to a through wall leak in Code Class III Piping;
- process radiation monitoring subsequent to numerous system communication failures; and
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 inspectors verified that the evaluation, planning, control, and work in progress met the requirements of 10 CFR 50.65, Maintenance Rule. Specifically, the inspectors verified that the equipment was identified and controlled where appropriate, work was being conducted properly, and significant aspects of plant risk were being communicated to the necessary personnel.
The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program. This review determined whether those problems were being entered into the corrective action program with the appropriate characterization and significance; documents reviewed during this inspection are listed in the Attachment.
The inspectors completed five samples by reviewing the following activities:
- 1A DG work window with predicted severe weather period;
- emergent setpoint adjustment and regulator replacement in feedwater heater 16B normal level control valve circuit;
- offsite power cross-tied between Units during Unit 2 system auxiliary transformer outage;
- 1A DG emergent governor problem during Unit 2 system auxiliary transformer outage; and
- removal for maintenance of the G fuel storage rack from the spent fuel pool.
b. Findings
No findings of significance were identified..
1R15 Operability Evaluations
a. Inspection Scope
The inspectors evaluated plant conditions and selected IRs for risk-significant components and systems in which operability issues were questioned. These conditions were evaluated to determine whether the operability of components was justified. The inspectors compared the operability and design criteria in the appropriate section of the UFSAR to the licensees evaluations presented in the IRs and other documents to verify that the components or systems were operable. The inspectors also conducted interviews with the appropriate licensee system engineers and conducted plant walkdowns, as necessary, to obtain further information regarding operability questions. Documents reviewed as part of this inspection are listed in the
.
The inspectors completed five samples by reviewing the following operability evaluations and conditions:
- Unit 2 high range containment radiation monitors impacted by temperature changes;
- 480 volt breaker ratings lowered in newly manufactured breakers;
- 1A DG fuel oil day tank sample indicated water and sediment following diesel oil storage tank cleaning;
- pin hole leak in an American Society of Mechanical Engineers (ASME) Code Class III pipe in the SX system; and
- Unit 1 intermediate range nuclear instrument, N35, overcompensated during reactor shutdown and subsequent reactor start-up.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed post-maintenance testing activities associated with important mitigating systems, barrier integrity, and support systems to ensure that the testing adequately demonstrated system operability and functional capability. The inspectors used the appropriate sections of the TS and UFSAR, as well as work orders for work performed, to evaluate the scope of the maintenance and to determine whether the post-maintenance testing was performed adequately, demonstrated that the maintenance was successful, and that operability was restored. The inspectors determined whether the tests were conducted in accordance with their procedures, including establishing the proper plant conditions and prerequisites, that the test acceptance criteria were met; and that the results of the tests were properly reviewed and recorded. The activities were selected based on their importance in demonstrating mitigating systems capability and barrier integrity. The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program. Documents reviewed as part of this inspection are listed in the Attachment.
Six samples were completed by observing post-maintenance testing of the following components:
- 1A DG following work window;
- 1A DG following mechanical governor replacement;
- 2B AF pump following the repair of a fuel oil leak;
- BT 7-11 switchyard circuit breaker following control power switch replacement;
- 1B DG following planned work window; and
- security DG following a planned work window.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors reviewed surveillance testing activities associated with important mitigating systems, barrier integrity, and support systems to ensure that the testing adequately demonstrated system operability and functional capability. The inspectors used the appropriate sections of the TS and UFSAR to determine whether the surveillance testing was performed adequately and that operability was restored. The inspectors determined whether the testing met the frequency requirements; that the tests were conducted in accordance with the procedures, including establishing the proper plant conditions and prerequisites; that the test acceptance criteria were met; and that the results of the tests were properly reviewed and recorded. Activities were selected based on their importance in demonstrating mitigating systems capability, barrier integrity and the initiating events cornerstones. Documents reviewed as part of this inspection are listed in the Attachment.
Six samples were completed by observing and evaluating the following surveillance tests:
- Unit 1 K616B main steam isolation valve emergency closure slave relay test (Routine);
- 2B RH pump ASME test (Inservice Testing);
- 1A AF pump simulated undervoltage conditions start (Routine);
- 1B containment spray pump slave relay and quarterly ASME test (Inservice Testing); and
- Unit 1 reactor coolant system leakrate evaluation (Leakrate).
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
The inspector observed the licensee performance during an emergency preparedness duty team drill. The inspectors observed activities in the control room simulator and the technical support center. The inspectors also attended the post-drill critiques in both facilities. 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 issues and entered them into the corrective action program. The inspectors placed emphasis on observations regarding event classification, notifications, protective action recommendations, site evacuation, and accountability activities. As part of the inspection, the inspectors reviewed the drill package and other documents listed in the
.
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 identified problems had been entered into the corrective action program for resolution. These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
.2 Plant Walkdowns and Radiation Work Permit Reviews
a. Inspection Scope
The inspectors reviewed licensee controls and surveys in the following three radiologically significant work areas within radiation areas, high radiation areas and airborne radioactivity areas in the plant and reviewed work packages which included associated licensee controls and surveys of these areas to determine if radiological controls including surveys, postings and barricades were acceptable:
- spent fuel pool inventory area;
- emergency core cooling system piping walkdown areas; and
- Unit 1 containment at-power entry.
These reviews represented one inspection sample.
The inspectors reviewed the radiation work permits (RWPs) and work packages used to access these three areas and other high radiation work areas to identify the work control instructions and control barriers that had been specified. Electronic dosimeter alarm set points for both integrated dose and dose rate were evaluated for conformity with survey indications and plant policy. Workers were interviewed to verify that they were aware of the actions required when their electronic dosimeters noticeably malfunctioned or alarmed. These reviews represented one inspection sample.
The inspectors walked down and surveyed (using an NRC survey meter) one of the three areas to verify that the prescribed RWP, procedure, and engineering controls were in place, that licensee surveys and postings were complete and accurate, and that air samplers were properly located. These reviews represented one inspection sample.
The adequacy of the licensees internal dose assessment process for internal exposures greater than 50 millirem committed effective dose equivalent (CEDE) was assessed.
There were no internal exposures greater than 50 millirem CEDE. These reviews represented one inspection sample.
The inspectors also reviewed the licensees physical and programmatic controls for highly activated and/or contaminated materials (non-fuel) stored within spent fuel or other storage pools. These reviews represented one inspection 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, Licensee Event Reports, and special reports related to the access control program to verify that identified problems were entered into the corrective action program for resolution.
These reviews represented one inspection sample.
The inspectors reviewed nine corrective action reports related to access controls and high radiation area radiological incidents (non-PIs identified by the licensee in high radiation areas <1R/hr). Staff members were interviewed and corrective action documents were reviewed to verify that follow-up activities were being 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 (NCVs) tracked in the corrective action system; and
- implementation/consideration of risk significant operational experience feedback.
These reviews represented one inspection sample.
The inspectors evaluated the licensees process for problem identification, characterization, prioritization, and verified that problems were entered into the corrective action program and resolved. For repetitive deficiencies and/or significant individual deficiencies in problem identification and resolution, the inspectors verified that the licensees self-assessment activities were capable of identifying and addressing these deficiencies. These reviews represented one inspection sample.
The inspectors reviewed licensee documentation packages for all PI events occurring since the last inspection to determine if any of these PI events involved dose rates greater than 25 R/hr at 30 centimeters or greater than 500 R/hr at 1 meter. Barriers were evaluated for failure and to determine if there were any barriers left to prevent personnel access. 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), were evaluated to determine if there were any regulatory overexposures or if there was a substantial potential for an overexposure. There were no PI events since the last inspection. These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
.4 Job-In-Progress Reviews
a. Inspection Scope
The inspectors observed the following three jobs that were being performed in radiation areas and high radiation areas for observation of work activities that presented the greatest radiological risk to workers:
- transient in-core probe inventory in the spent fuel pool;
- Unit 1 at-power entry; and
- emergency component cooling system venting and surveillance.
The inspectors reviewed radiological job requirements for these three activities including RWP requirements and work procedure requirements, and attended As-Low-As-Reasonably-Achievable (ALARA) job briefings. These reviews represented one inspection sample.
Job performance was observed with respect to these requirements to verify that radiological conditions in the work area were adequately communicated to workers through pre-job briefings and postings. The inspectors also verified the adequacy of radiological controls including required radiation, contamination, and airborne surveys for system breaches; radiation protection job coverage which included audio and visual surveillance for remote job coverage; and contamination controls. These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
.5 High Risk Significant, High Dose Rate/High Radiation Area and Very High Radiation
Area Controls
a. Inspection Scope
The inspectors held discussions with the Radiation Protection Manager concerning high dose rate/high radiation area and very high radiation area controls and procedures, including procedural changes that had occurred since the last inspection, in order to verify that any procedure modifications did not substantially reduce the effectiveness and level of worker protection. These reviews represented one inspection sample.
The inspectors discussed, with Radiation Protection supervisors, the controls that were in place for special areas that had the potential to become very high radiation areas during certain plant operations. The purpose of these discussions was to determine if these plant operations required communication beforehand with the Radiation Protection group, so as to allow corresponding timely actions to properly post and control the radiation hazards. These reviews represented one inspection sample.
The inspectors conducted plant walkdowns to verify the posting and locking of entrances to high dose rate/high radiation areas, and very high radiation. These reviews represented one inspection sample.
b. Findings
No findings of significance were identified
.6 Radiation Worker Performance
a. Inspection Scope
During job performance observations, the inspectors evaluated radiation worker performance with respect to stated radiation protection work requirements and evaluated whether workers were aware of the significant radiological conditions in their workplace, the RWP controls and limits in place, and that their performance had accounted for the level of radiological hazards present. These reviews represented one inspection sample.
The inspectors reviewed radiological problem reports which found that the cause of the event was due to radiation worker errors to determine if there was an observable pattern traceable to a similar cause, and to determine if this perspective matched the corrective action approach taken by the licensee to resolve the reported problems. These problems, along with planned and taken corrective actions were discussed with the Radiation Protection Manager. These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
.7 Radiation Protection Technician Proficiency
a. Inspection Scope
During job performance observations, the inspectors evaluated Radiation Protection Technician performance with respect to radiation protection work requirements and evaluated whether they were aware of the radiological conditions in their workplace, the RWP controls and limits in place, and if their performance was consistent with their training and qualifications with respect to the radiological hazards and work activities.
These reviews represented one inspection sample.
The inspectors reviewed radiological problem reports which found that the cause of the event was radiation protection technician error to determine if there was an observable pattern traceable to a similar cause, and to determine if this perspective matched the corrective action approach taken by the licensee to resolve the reported problems.
These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
Cornerstone: Public Radiation Safety
2PS1 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems (71122.01)
.1 Review of Blowdown Line and Tritium Remediation System Integrity
a. Inspection Scope
The inspectors continued to monitor the licensees activities resulting from previous inadvertent leaks of tritiated liquid from the blowdown line to the Kankakee River.
The inspection activities included the following:
- routine liquid effluent discharges to the river;
- operation of the pond remediation system;
- operations of and repairs to the remediation system for areas near vacuum breaker one;
- response to increased tritium levels in the secondary plant, lagoons, and cooling lake;
- installation of a remediation system for areas near the oil separator;
- vault liner repairs of vacuum breaker six; and
- periodic inspections of all of the vacuum breaker pits and remediation pumps.
The inspectors verified that minor issued identified during the inspection were entered into the licensees corrective action program. Documents reviewed are listed in the
. This inspection did not constitute a complete sample.
b. Findings
No findings of significance were identified.
.2 Inspection Planning
a. Inspection Scope
The inspectors reviewed the most current Radiological Effluent Release Report to verify that the program was implemented as described in Radiological Effluent TSs/Offsite Dose Calculation Manual (RETS/ODCM) and to determine if ODCM changes were made in accordance with Regulatory Guide 1.109 and NUREG-0133. The inspectors determined if the modifications made to radioactive waste system design and operation changed the dose consequence to the public. The inspectors verified that technical and/or 10 CFR 50.59 reviews were performed when required and determined whether radioactive liquid and gaseous effluent radiation monitor setpoint calculation methodology changed since completion of the modifications. The inspectors determined if anomalous results reported in the current Radiological Effluent Release Report were adequately resolved.
The inspectors reviewed RETS/ODCM to identify the effluent radiation monitoring systems and accompanying flow measurement devices, effluent radiological occurrence performance indicator incidents in preparation for onsite follow-up, and the UFSAR description of all radioactive waste systems. The inspectors reviewed the licensees RETS/ODCM which provided the licensees program for identifying potential contaminated spills and leakage and the licensees process for control and assessment.
These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
.3 Onsite Inspection
a. Inspection Scope
The inspectors walked-down the major components of the gaseous and liquid release systems (e.g., radiation and flow monitors, demineralizers and filters, tanks, and vessels) to observe current system configuration with respect to the description in the UFSAR, ongoing activities, and equipment material condition. These reviews represented one inspection sample.
The inspectors observed the routine processing (including sample collection and analysis) and release of radioactive liquid waste to verify that appropriate treatment equipment was used and that radioactive liquid waste was processed and released in accordance with procedure requirements. The inspectors reviewed routine processing (including sample collection and analysis) and release of radioactive gaseous effluent to verify that appropriate treatment equipment was used and that the radioactive gaseous effluent was processed and released in accordance with RETS/ODCM requirements.
The inspectors reviewed several radioactive gaseous effluent release permits, including the projected doses to members of the public. These reviews represented one inspection sample.
The inspectors reviewed the records of abnormal releases or releases made with inoperable effluent radiation monitors and reviewed the licensees actions for these releases to ensure an adequate defense-in-depth was maintained against an unmonitored, unanticipated release of radioactive material to the environment.
These reviews represented one inspection sample.
For unmonitored releases, (i.e., via typical, routine effluent pathways, or via spills, leaks, abnormal, or unexpected liquid or gaseous discharge, or other unusual occurrences),the inspectors verified that the licensee did perform an evaluation of the type and amount of radioactive material that was released and the associated projected doses to members of the public. These reviews represented one inspection sample.
Additionally, for any areas where spills, leaks, or other unusual occurrences (i.e., involving the spread of licensed radioactive material in and around the facility, equipment, or site) took place, the inspectors verified that these areas have been properly documented in the sites decommissioning file, as required by 10 CFR 50.75(g). These reviews represented one inspection sample.
The inspectors assessed the licensees understanding of the location and construction of underground pipes and tanks, and storage pools (spent fuel pool) that could result in leakage of contaminated fluids to the groundwater as a result of degrading material conditions or aging of facilities. The inspectors appraised the licensees capabilities (such as monitoring wells) of detecting spills or leaks and of identifying groundwater radiological contamination both on-site and beyond the owner controlled area. The inspectors discussed with the licensee, its understanding of groundwater flow patterns for the site, and in the event of a spill or leak of radioactive material, if the licensees staff can estimate the pathway of a plume of contaminated fluid both onsite and beyond the owner controlled area. These reviews represented one inspection sample.
The inspectors reviewed the licensees technical justification for changes made by the licensee to the ODCM as well as to the liquid or gaseous radioactive waste system design, procedures, or operation since the last inspection to determine whether the changes affect the licensees ability to maintain effluents ALARA and whether changes made to monitoring instrumentation resulted in a non-representative monitoring of effluents. These reviews represented one inspection sample.
The inspectors reviewed a selection of monthly, quarterly, and annual dose calculations to ensure that the licensee properly calculated the offsite dose from radiological effluent releases and to determine if any annual RETS/ODCM (i.e., Appendix I to 10 CFR Part 50 values) were exceeded. These reviews represented one inspection sample.
The inspectors reviewed air cleaning system surveillance test results to ensure that the system was operating within the licensees acceptance criteria. The inspectors reviewed surveillance test results the licensee uses to determine the stack and vent flow rates. The inspectors verified that the flow rates were consistent with RETS/ODCM or UFSAR values. These reviews represented one inspection sample.
The inspectors reviewed records of instrument calibrations performed since the last inspection for each point of discharge effluent radiation monitor and flow measurement device and reviewed any completed system modifications and the current effluent radiation monitor alarm setpoint value for agreement with RETS/ODCM requirements.
The inspectors also reviewed calibration records of radiation measurement (i.e.,
counting room) instrumentation associated with effluent monitoring and release activities and the quality control records for the radiation measurement instruments. These reviews represented one inspection sample.
The inspectors reviewed the results of the interlaboratory comparison program to verify the quality of radioactive effluent sample analyses performed by the licensee. The inspectors reviewed the licensees quality control evaluation of the interlaboratory comparison test and associated corrective actions for any deficiencies identified. The inspectors reviewed the licensees assessment of any identified bias in the sample analysis results and the overall effect on calculated projected doses to members of the public. In addition, the inspectors reviewed the results from the licensees quality assurance audits to determine whether the licensee met the requirements of the RETS/ODCM. These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
.4 Identification and Resolution of Problems
a. Inspection Scope
The inspectors reviewed the licensees self-assessments, audits, Licensee Event Reports, and Special Reports related to the radioactive effluent treatment and monitoring program since the last inspection to determine if identified problems were entered into the corrective action program for resolution. The inspectors also verified that the licensee's self-assessment program was capable of identifying repetitive deficiencies or significant individual deficiencies in problem identification and resolution.
The inspectors also reviewed corrective action reports from the radioactive effluent treatment and monitoring program since the previous inspection, interviewed staff, and reviewed documents to determine if the following activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk:
- 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 tracked in the corrective action system; and
- implementation/consideration of risk significant operational experience feedback.
These reviews represented one inspection sample.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
.1 Initiating Event Performance Indicators
a. Inspection Scope
Cornerstone: Barrier Integrity
The inspectors sampled the licensees PI submittals for the periods listed below.
The inspectors used PI definitions and guidance contained in Nuclear Energy Institute Document 99-02; Regulatory Assessment Performance Indicator Guideline, Revision 4, to verify the accuracy of the PI data. The following PIs were reviewed for a total of four samples:
Unit 1
- reactor coolant system specific activity; and
- reactor coolant system leakage.
Unit 2
- reactor coolant system specific activity; and
- reactor coolant system leakage.
The inspectors reviewed licensee IRs, electronic logs, and other records for the period from April 1, 2006, through March 31, 2007, for each PI area specified above. The inspectors independently re-performed calculations where applicable. The inspectors compared the information acquired for each PI to the data reported by the licensee.
The inspectors verified that minor issues identified during the inspection were entered into the licensees corrective action program. Documents reviewed are listed in the
.
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 required by Inspection Procedure 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed screening of all items entered into the licensees corrective action program. This screening was accomplished by reviewing the description of each new Issue Report and attending selected daily management review committee meetings. Documents reviewed are listed in the Attachment to this report. Minor issues entered into the licensees corrective action program as a result of the inspectors observations are generally denoted in the Attachment. These activities were part of normal inspection activities and were not considered separate samples.
b. Findings
No findings of significance were identified.
.2 Semi-annual Review to Identify Trends
a. Inspection Scope
The inspectors conducted a review of licensee corrective action documents to identify trends that could indicate the existence of a more significant safety issue. For this review, the inspectors reviewed a sampling of equipment failures and other equipment problems for the period of January through May 2007 and then looked at the last five-year history to see if any of those failures were part of trend involving similar equipment in other trains or systems. This activity represented one inspection sample of the semiannual trend review. Documents selected for a more detailed review are listed in the Attachment.
b. Findings
No findings of significance were identified.
4OA3 Event Followup
.1 Unit 1 Reactor Trip - June 27, 2007
a. Inspection Scope
The inspectors responded to an automatic reactor trip of Unit 1 that occurred on June 27, 2007. The inspectors observed operator actions taken to verify that they were taken in accordance with licensee procedures, and reviewed unit and system indications to verify that system responses were as expected. The inspectors discussed the trip with operations, engineering, and licensee management personnel to gain an understanding of the event and assess followup actions. The inspectors later reviewed the initial investigation report and observed the Plant Operating Review Committee to assess the detail of review and adequacy of the licensees understanding of the apparent cause of the trip and proposed corrective actions prior to unit restart.
The licensees investigation identified that the apparent cause of the reactor trip was a lightning strike to a 345 kilovolt offsite power line which created a large current spike in the Unit 1 switchyard resulting in multiple switchyard breakers opening and an overcurrent trip of the 1D reactor coolant pump. Subsequent to the 1D reactor coolant pump trip, the reactor tripped on low reactor coolant flow in the D reactor coolant loop as expected. The inspectors also reviewed the initial licensee notification to verify that it met the requirements specified in NUREG 1022, Event Reporting Guidelines. The inspectors determined that the initial event notification contained errors regarding the response of the switchyard/main generator output breakers at the onset of the event and did not include details regarding the expected engineered safeguards feature actuation of the AF system. The licensee subsequently updated the event report to more accurately reflect the sequence of events.
The inspectors observations were compared to the requirements specified in the procedure listed in the Attachment. The above represents one inspection sample.
b. Findings
No findings of significance were identified.
.2 Spill from West Lagoon
Background On May 23, 2007, the licensee identified that approximately 1500 gallons of water from the West Lime Sludge Lagoon overflowed its banks due to high winds in the area. The lagoon is located within the restricted area and is a component of the licensees waste treatment system. Sampling of the lagoon indicated tritiated water, at a concentration of approximately 75,000 pico-curies per liter (pCi/L), which the licensee had not expected.
As a result, the licensee sampled the East Lime Sludge Lagoon, located adjacent to the West Lagoon, and also identified elevated levels of tritium at a concentration of approximately 2.6 million pCi/L. The licensee determined that the increased tritium level was due to leakage from one of the sites primary water storage tanks (via system valve leakage) into the turbine building equipment floor drains, which were subsequently routed to the East Lagoon. Upon identification of the leakage, the system was isolated from the lagoons. The licensee concluded that the West Lagoon became contaminated as a result of water transfers between the lagoons. The licensee discharged the contents of both lagoons to the Braidwood cooling lake, a recognized radioactive discharge path as provided in the ODCM. The licensees preliminary dose estimate concluded that the potential dose to the public from the lagoon releases was a small fraction of one millirem.
a. Usage of East and West Lagoons Inspection Scope The inspectors reviewed the material that the licensee processed and stored in the East and West Lime Sludge Lagoons. The East Lagoon provides a surge volume for processing waste from the turbine building equipment and turbine building floor drains.
Leakage from the primary water storage tank was collected and routed to the floor drain system and eventually to the East Lagoon. Residual water in the pumping systems cross contaminated the contents of the West Lagoon.
The inspectors reviewed the licensing basis for the lagoons, and assessed the activity of tritium contained within the lagoons, as provided in the UFSAR for compliance with the ODCM.
Findings No findings of significance were identified.
b. Impact on Public Radiation Safety Inspection Scope The inspectors collected split samples from the cooling lake and the waste treatment lagoons for analysis by an independent laboratory. Additionally, the inspectors independently calculated the dose to the public from lagoon discharges to the Braidwood cooling lake and compared the results to 10 CFR Part 50, Appendix I limits.
The inspectors determined that the dose to public was less than 1 millirem.
Findings No findings of significance were identified.
c. Impact on Occupational Radiation Safety Inspection Scope The inspectors evaluated the contributors to tritium accumulation in the primary water system, the cross contamination of the demineralized water system, and the subsequent use of demineralized water as make-up for other secondary systems to determine whether the licensee was adequately implementing the requirements contained in 10 CFR Part 20 and plant procedures.
The inspectors reviewed the work history of select components within the secondary side of the plant. The inspectors focused on job planning, radiological job coverage for system breaches, and monitoring for internal dose of workers involved with system breaches of the primary water system to determine if the licensee was adequately implementing the requirements contained in 10 CFR Part 20 and plant procedures.
Findings
Introduction:
A finding of very low safety significance and an associated Non-Cited Violation (NCV) of an NRC requirement was identified by the inspectors for the failure to implement the radiation protection program commensurate with the full extent of radiological issues at the plant.
Description:
In November 2005, the licensee identified leaks from the blowdown line vacuum breaker valves that resulted in violations of NRC requirements for the unauthorized release of radioactive material to the environment (Inspection Report 05000456/2006008; 05000457/2006008; ML061450522). That condition caused the licensee to halt all liquid radioactive waste releases until the cause(s) was identified, repaired, and evaluated. During this process, the licensee made a decision to further restrict the number of liquid radioactive waste releases and to reduce the amount of tritium released from the plant each year through enhanced water recycling. Recycling waste water minimizes impurities in the waste water with the exception of tritium which cannot be removed from the water using conventional technologies because of its chemical form. The recycled water was then stored for future use in the primary water storage tanks. By April 2006, tritium concentration in the primary water system exceeded 10 million picocuries/liter (pCi/L) and by April 2007, the concentration was several hundred million pCi/L.
Historically, the primary water system has been an ultra-pure source of water that was supplied by processing a non-radiological water source and removing chemical impurities. The water was then used in the reactor coolant system and served as make-up water to the component cooling water system. The primary water system and other systems that came into contact with primary water as make-up were outside the radiologically controlled area and not evaluated by the licensee to assess system radiological impact.
The decision to operate the plant in recycle mode was implemented only after significant changes were made to the facility, including installing new equipment, modification of existing equipment, and creating new procedures. However, the licensees radiation protection program failed to adequately monitor, assess, and control radiological hazards from the additional systems and areas of the plant that were impacted by tritium. Specifically, recycled plant water impacted plant systems and areas that had not previously been contaminated or had only trace levels of contamination. Additionally, the licensees radiological controls were not effectively applied to these systems to ensure that worker exposures and radiological effluents were fully monitored and controlled. Recent work activities that adversely impacted secondary side contamination included a spectacle flange installed on the primary water system; a pump that was replaced in the waste water treatment building; and drip catch containers installed on the primary water system which routed the water to the turbine equipment floor drains.
Analysis:
The failure to implement the radiation protection program commensurate with the full extent of radiological issues at the plant represents a performance deficiency as defined in NRC Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening. The inspectors determined that the issue was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Therefore, the issue was more than minor and represented a finding which was evaluated using the Significance Determination Process (SDP).
Since the finding primarily involved the ability to protect workers from exposure to radiation, the inspectors utilized IMC 0609, Appendix C, Occupational Radiation Safety SDP, to assess its significance. The inspectors determined that the finding did not involve unintended collective dose resulting from a deficiency in ALARA planning, work control or exposure control. The inspectors also determined that the finding did not involve an overexposure, the substantial potential for an overexposure, and did not compromise the licensees ability to assess dose. Consequently, the inspectors concluded that the SDP assessment for this finding was of very low safety significance (Green).
As described above, the corrective actions taken in response to vacuum breaker valve leakage and resulting violations of NRC requirements included a new operating philosophy (recycle). The licensee implemented these corrective actions without assessing the impact on the radiation protection program. Consequently, the cause of this deficiency had a cross cutting aspect in the area of Problem Identification and Resolution. Specifically, the corrective actions taken by the licensee did not adequately address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity (P.1(d)).
Enforcement:
Title 10 CFR 20.1101(a) requires each licensee to develop, document, and implement a radiation protection program commensurate with the scope and extent of licensed activities and sufficient to ensure compliance with the provisions of this part.
Contrary to the above, as of May 24, 2007, the work control process at the plant failed to evaluate the radiological impact of radioactive waste reduction activities for work on secondary systems that were contaminated with tritium.
Corrective actions taken by the licensee included characterizing secondary systems to determine tritium concentration and prescribing radiological coverage and contamination control requirements for each system based upon this characterization. Since the licensee documented this issue in its corrective action program (AR 00636122) and because the violation is of very low safety significance, it is being treated as an NCV (05000456/2007004-01; 05000457/2007004-01).
d. Notifying Workers of Hazards in the Workplace Inspection Scope The inspectors walked down areas of the turbine building and waste processing areas to identify whether the impact of tritium contamination was identified, evaluated, and communicated to the workers and to determine whether the licensee satisfied the requirements of 10 CFR Part 20.
Findings
Introduction:
A finding of very low safety significance and an associated NCV of an NRC requirement was identified by the inspectors for the failure to post areas that contain radioactive material in accordance with 10 CFR 20.1902(e).
Description:
The licensee notified the NRC of the increased concentration of tritium in the East Lagoon on May 24, 2007. The NRC assessed the concentration and volume and questioned whether the lagoons were posted in accordance with NRC requirements. The licensee responded by posting the access point to the lagoons and by reviewing the posting of other areas of the plant including the turbine building. A walkdown of the turbine building was conducted by the NRC inspectors and members of the licensees staff during the afternoon of May 29, 2007. This walkdown identified that some tanks within the building were posted as Radioactive Material but other areas within the turbine building that contained tritium in excess of 10,000 micro-curies (uCi) in secondary-side systems were not posted. The inspectors concluded that the licensee had not adequately evaluated the radiological conditions to determine if the requirements of 10 CFR 20.1902(e) were met.
Analysis:
The failure to post areas that contain radioactive materials in quantities that exceed specified values represents a performance deficiency as defined in NRC Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening. The inspectors determined that the issue was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Therefore, the issue was more than minor and represented a finding which was evaluated using the SDP.
Since the finding involved the ability to protect workers from exposure to radiation, the inspectors utilized IMC 0609, Appendix C, Occupational Radiation Safety SDP, to assess its significance. The inspectors determined that the finding did not involve unintended collective dose resulting from a deficiency in ALARA planning, work control or exposure control. The inspectors also determined that the finding did not involve an overexposure, the substantial potential for an overexposure and did not compromise the licensees ability to assess dose. Consequently, the inspectors concluded that the SDP assessment for this finding was of very low safety significance (Green).
Since the licensee had previously reviewed the conditions of the turbine building and concluded that posting the area would not benefit the workforce, this deficiency has a cross cutting aspect in Human Performance. Specifically, the licensee did not use conservative assumptions in decision making (H.1(b)).
Enforcement:
Title 10 CFR 20.1902(e) requires licensees to post each area or room in which there is used or stored an amount of licensed material exceeding 10 times the quantity of such material specified in Appendix C to Part 20 with a conspicuous sign or signs bearing the radiation symbol and the words "CAUTION, RADIOACTIVE MATERIAL(S)" or "DANGER, RADIOACTIVE MATERIAL(S)." Contrary to the above, as of May 24, 2007, the lagoons contained greater than 10,000 uCi of tritium and were not posted in accordance with 10 CFR 20.1902(e). Additionally, as of May 29, 2007, areas of the turbine building contained greater than 10,000 uCi of tritium and were not posted in accordance with 10 CFR 20.1902(e).
Corrective actions taken by the licensee included posting the lagoons and the turbine building with "CAUTION, RADIOACTIVE MATERIAL(S)." Since the licensee documented this issue in its corrective action program (AR 00636141)and because the violation is of very low safety significance, it is being treated as an NCV (05000456/2007004-02; 05000457/2007004-02).
4OA5 Other Activities
(Closed) Unresolved Item (URI) (05000456/2005007-06; 05000457/2005007-06):
Review of Seismic/Safety Classification for the Essential Service Water Strainer Backwash System During the 2005 Safety System Design and Performance Capability Inspection, the inspectors identified an unresolved item concerning SX strainer backwash system. The SX strainer backwash motor and isolation drain valve power supplies and control circuits were not safety-related or seismically qualified and following a seismic event these components could lose power. Without corrective actions to restore the SX strainer backwash function, the accumulation of sediment/debris present in the SX system would build up on the strainers and cause a loss of SX flow to safety-related equipment.
The licensee provided the inspectors with four points considered by the original Sargent and Lundy Engineers involved in Classification of the SX backwash system:
C Power can be restored to strainer backwash in the event of a loss of offsite power; C Strainer backwash can be operated manually; C There will be sufficient time to accomplish strainer backwash before adverse effects on the SX system; and C Strainer backwash capability is expected to function after a seismic event.
Since the SX strainer backwash system was required to maintain the safety-related SX function following a seismic event, the inspectors were concerned that it should have been provided with seismically qualified control circuits and power supplies in accordance with 10 CFR Part 50 Appendix A, General Design Criteria No. 2 and No. 17.
This issue was considered an URI pending NRC review of the plant licensing basis for the SX system backwash strainer function.
NRC Review and Conclusion:
Inspectors from the NRC Region III office, with support from the Office of Nuclear Reactor Regulation (NRR) staff, reviewed this issue. After a review of the plants licensing basis, the NRR staff concluded that the SX strainer backwash system was operating within its design and licensing basis. As such, the classification of the SX strainer backwash system as nonsafety-related or non-seismically qualified was acceptable. However, if the SX strainer backwash system was needed under a loss of power/seismic event, either its failure would not prevent the SX system from performing its safety-related function or proceduralized actions could be performed to operate the SX strainer backwash system under this scenario. This loss of SX function would be caused by allowing the strainer to become plugged without any means of cleaning the debris from the strainer without power to the backwash arm or the backwash motor-operated valves. In order to verify SX strainer backwash system would be able to be operated manually, the inspectors reviewed the licensees corrective action (AR00367473) for Finding 05000456/2005007-05; 05000457/2005007-05, which concerned the failure to provide operators with equipment, procedures, and training to manually operate the SX system strainers to recover the loss of the SX automatic backwash capability.
The inspectors review of the corrective actions identified several concerns with the work orders developed to address the issue for each of the strainers. These concerns included: there was no emergency, abnormal, or alarm response procedure to direct the operators to perform the work orders in the event of high strainer differential pressure, and a loss of non-safety power to the SX strainer backwash system could occur. As such, operators may not be aware of the work orders to address this scenario. The four work orders, each contained Steps 1.I and 1.K that referenced re-performing Step No. 8"; however, this step did not exist in the work orders. The work orders instructions were copied from the strainer vendor manual, but step numbers were changed that were not identified during the review process. In addition, the work package did not identify the tools required to perform the intended actions. An initial response from the licensee indicated that the only tool required would be an appropriately sized wrench to remove the two nuts on top of the strainer backwash shaft. The inspectors questioned the need for ladders to reach the strainer, tools to remove the set screws on the dust cover and the thrust bearing attached to the backwash shaft.
Based on the inspectors concerns with the resolution of the finding, the licensee initiated several actions. First, a new procedure was issued, BwMP 3300-103, SX Strainer Manual Backwash Operation on Loss of Power, which provided directions to operate the strainer backwash system without power. The procedure was based on the strainer vendor manual and provided sufficient guidance to perform the required operation. The only concern with the procedure was the lack of information on what tools would be necessary to perform the required actions. As previously discussed, scaffolding or ladders were needed to reach the strainer work area, and the associated tools necessary to perform the work were not identified such that the operation could be conducted in a timely manner. The licensee initiated Revision 1 to the procedure to include the necessary tools to manually operate the strainer to resolve this issue.
Secondly, BwOP SX-6, Essential Service Water Strainer Manual Operation, was revised to include the following statement under Limitations and Action, In the event the SX strainer cannot be backwashed due to a loss of power, mechanical maintenance department should be contacted to perform a manual backwash of the strainer per BwMP 3300-103. This change provided the operators with an acceptable link to the maintenance action needed to perform these manual actions.
Based on the above qualitative assessment and the facts the inspectors determined no performance deficiencies or violations of regulatory requirements exist, no additional enforcement action was warranted. The inspectors had no further concerns in this area.
This unresolved item is closed.
Because the inspection was counted in another inspection report, these inspection activities do not represent an inspection sample for this report.
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Mr. G Boerschig, and other members of licensee management at the conclusion of the inspection on June 29, 2007. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
.2 Interim Exit Meetings
Interim exit meetings were conducted for:
- Radioactive Gaseous And Liquid Effluent Treatment And Monitoring Systems program with Mr. G. Boerschig on May 4, 2007;
- Access Control to Radiologically Significant Areas with Mr. T. Couto on May 15, 2007;
- Event Follow-up for the spill of tritium from the West Lagoon and increase of tritium in the East Lagoon with Mr. T. Coutu on June 1, 2007; and
- Closure of URI 05000456/2005007-06; 05000457/2005007-06 with Mr. J. Perry on June 22, 2007.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- T. Coutu, Site Vice President
- L. Coyle, Plant Manager
- K. Aleshire, Emergency Preparedness Manager
- M. Cichon, Licensing Engineer
- G. Dudek, Site Training Director
- J. Gosnell, Project Manager
- D. Gullott, Regulatory Assurance Manager
- J. Knight, Nuclear Oversight Manager
- R. Leasure, Radiation Protection Supervisor
- T. McCool, Operations Director
- T. Meents, RETS/REMP Analyst
- J. Moser, Radiation Protection Manager
- D. Myers, Acting Maintenance Director
- J. Perry, Acting Manager, Regulatory Assurance
- M. Smith, Engineering Director
- T. Tierney, Chemistry, Environmental, and Radioactive Waste Manager
Nuclear Regulatory Commission
- R. Skokowski, Chief, Reactor Projects Branch 3
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened
456/2007004-01; NCV Failure to implement a radiation protection program 457/2007004-01 commensurate with the extent of plant radiological hazards.
(Section 4OA3.2)
456/2007004-02; NCV Failure to post areas that contain radioactive material.
457/2007004-02 (Section 4OA3.2)
Closed
456/2007004-01; NCV Failure to implement a radiation protection program 457/2007004-01 commensurate with the extent of plant radiological hazards.
(Section 4OA3.2)
456/2007004-02; NCV Failure to post areas that contain radioactive material.
457/2007004-02 (Section 4OA3.2)
456/2005007-06; URI Review of Seismic/Safety Classification for the SX Strainer 457/2005007-06 Backwash System Attachment
Discussed
None Attachment