IR 05000397/2006004

IR 05000397-06-004; Energy Northwest; 7/1/2006 - 9/30/2006; Columbia Generating Station; Operability Evaluations; Other Activities

ML063180298
Person / Time
Site:	Columbia
Issue date:	11/14/2006
From:	Clay Johnson NRC/RGN-IV/DRP/RPB-A
To:	Parrish J Energy Northwest
References
IR-06-004
Download: ML063180298 (37)

Text

ber 14, 2006

SUBJECT:

COLUMBIA GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000397/2006004

Dear Mr. Parrish:

On September 30, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Columbia Generating Station. The enclosed inspection report documents the inspection findings which were discussed on October 5, 2006, with Mr. W. Oxenford and other members of your staff, and re-exit with Mr. T. Lynch, Plant Manager on October 10, 2006.

The inspections examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents one NRC-identified finding and one licensee identified finding of very low risk significance. One of these findings was determined to involve a violation of NRC requirements. However, because of the very low safety significance and because it is entered into your corrective action program, the NRC is treating this finding as a noncited violation (NCV) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011-4005; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Columbia Generating Station.

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

Energy Northwest -2-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.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Claude E. Johnson, Chief Project Branch A Division of Reactor Projects Docket: 50-397 License: NPF-21

Enclosure:

NRC Inspection Report 05000397/2006003

REGION IV==

Docket: 50-397 License: NPF-21 Report: 05000397/2006004 Licensee: Energy Northwest Facility: Columbia Generating Station Location: Richland, Washington Dates: July 1 through September 30, 2006 Inspectors: Z. Dunham, Senior Resident Inspector, Project Branch A, DRP R. Cohen, Acting Senior Resident Inspector and Resident Inspector, Project Branch A, DRP J. Drake, Operations Engineer, Operations Branch and Acting Resident Inspector Project Branch A, DRP R. Lantz, Emergency Preparedness Inspector, Operations Branch B. Larson, Operations Engineer, Operations Branch L. Ricketson, Health Physicist, Plant Support Branch Approved By: C. E. Johnson, Chief, Project Branch A, Division of Reactor Projects ATTACHMENT: SUPPLEMENTAL INFORMATION Enclosure

SUMMARY OF FINDINGS

IR 05000397/2006004; 7/1/2006 - 9/30/2006; Columbia Generating Station; Operability

Evaluations; Other Activities The report covered a 13-week period of inspection by resident inspectors, a senior health physicist, and emergency preparedness inspectors. One Green noncited violation was identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, Significance Determination Process. Findings for which the significance determination process 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.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, when Energy Northwest failed to perform adequate design reviews to maintain appropriate control of the design and qualification of the stations safety-related batteries. Specifically, the repetitive failure to provide adequate engineering analysis supporting the temporary installation of a non-Class 1E battery rail charger on a safety-related battery was not commensurate with ensuring the reliability of the stations safety-related batteries.

This finding was more than minor because the finding was a design control issue which affected the mitigating systems cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Utilizing Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the inspectors determined that the finding was of very low risk significance because it was a qualification issue confirmed not to result in loss of operability. Additionally, the finding did not represent a loss of safety function for a single train or for the system, and did not screen as potentially risk significant due to external events. This finding had crosscutting aspects in the area of problem identification and resolution associated with the corrective action program component in that the licensee did not thoroughly evaluate design issues with the nonqualified rail charger, as documented in Condition Report 2-05-01894. This resulted in additional examples of the failure to maintain adequate design control of the batteries. (Section 1R15)

Licensee Identified Violations

One violation of very low significance was identified by the licensee and reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared to be reasonable. This violation is listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status:

The inspection period began with Columbia Generating Station at 100 percent power. The plant was maintained at essentially 100 percent power for the entire inspection period except for three unscheduled reductions in power due to two reactor recirculation pump power supply cooling system failures and a hydraulic leak on a low pressure turbine intercept valve.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather

a. Inspection Scope

The inspectors completed a review of the licensee's readiness of seasonal susceptibilities involving extreme high temperatures and high winds. The inspectors:

(1) reviewed plant procedures, the Updated Safety Analysis Report, and Technical Specifications to ensure that operator actions defined in adverse weather procedures maintained the readiness of essential systems;

(2) walked down portions of the system listed below to ensure that adverse weather protection features were sufficient to support operability, including the ability to perform safe shutdown functions;

(3) evaluated operator staffing levels to ensure the licensee could maintain the readiness of essential systems required by plant procedures; and

(4) reviewed the corrective action program to determine if the licensee identified and corrected problems related to adverse weather conditions.

• Reactor Recirculation Pump 1A and 1B Adjustable Speed Drives; August 22, 2006 The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignments

.1 Partial Walkdown

a. Inspection Scope

The inspectors:

(1) walked down portions of the risk important systems listed below and reviewed plant procedures and documents to verify that critical portions of the selected systems were correctly aligned; and

(2) compared deficiencies identified during the walk down to the licensee's corrective action program to ensure problems were being identified and corrected.

• Fire Protection System; July 13, 2006

• Condensate System as an Emergency Source of Water to the Reactor; July 18, 2006

• Hydraulic Control Units; August 3, 2006 The inspectors completed three samples.

b. Findings

No findings of significance were identified.

.2 Complete Walkdown

a. Inspection Scope

The inspectors:

(1) reviewed plant procedures, drawings, the updated safety analysis report, technical specifications, and vendor manuals to determine the correct alignment of the system;

(2) reviewed outstanding design issues, operator work arounds, and corrective action program documents to determine if open issues affected the functionality of the system; and

(3) verified that the licensee was identifying and resolving equipment alignment problems.

• Standby Liquid Control (SLC); July through August, 2006 The inspectors completed one sample.

b. Findings

Introduction.

An unresolved item (URI) was identified pending the NRCs review of Energy Northwests evaluation to determine the impact of not venting the temporary hose during alternate boron injection and a potential performance issue associated with the use of alternate boron injection via the Reactor Core Isolation Cooling (RCIC)

System.

Description.

The inspectors reviewed and walked down the implementation of Emergency Support Procedure, ESP 5.5.8, Alternate Boron Injection, Revision 8, to determine the adequacy of the procedure. Procedure ESP 5.5.8 provides for connecting a pre-staged hose between a SLC pump relief valve flange (approximate 548 foot elevation) and a suction drain (approximate 422 foot elevation) on the RCIC pump. This was done to allow the RCIC system to inject the contents of the SLC boron tank to the reactor coolant system to provide a diverse method of shutting down the reactor in the event that AC power was not available to operate the SLC pumps. The inspectors noted that, after the hose is connected, that the procedure does not provide for venting and filling the hose prior to opening valves in the RCIC and SLC systems to establish the alternate flow path. The inspectors were concerned that because of a high point in the configuration of the alternate flow path that the temporary hose may not vent completely prior to operators starting the RCIC pump to inject the SLC tank contents.

Without venting, the inspectors postulated that the RCIC pump may become air bound due to air in the temporary hose being pumped into the RCIC pump or may challenge RCIC pump operation in other unforseen ways. The licensee was unable to provide any prior analysis or calculations which demonstrated that the temporary hose would self vent or if not vented that RCIC pump operations would not be challenged. The licensee documented the concerns in the corrective action program as CR 2-06-06510 and initiated an action request, AR 249460, to evaluate the condition. The licensee had not completed their evaluation of the concern at the end of the inspection period. An URI was opened pending a completion of the NRCs review of Energy Northwests evaluation to determine the adequacy of not venting the temporary hose during alternate boron injection and the impact on RCIC pump operation (URI 05000397/2006004-01; Non-Venting of Alternate Boron Injection Hose).

Analysis.

A determination of the safety significance associated with any performance deficiencies will be addressed in the resolution to the URI.

Enforcement.

A determination of the enforcement aspects associated with any performance deficiencies will be addressed in the resolution to the URI.

1R05 Fire Protection

.1 Quarterly Inspection

a. Inspection Scope

The inspectors walked down the plant areas listed below to assess the material condition of active and passive fire protection features and their operational lineup and readiness. The inspectors:

(1) verified when applicable that transient combustibles and hot work activities were controlled in accordance with plant procedures;

(2) observed the condition of fire detection devices to verify they remained functional;

(3) observed fire suppression systems to verify they remained functional;

(4) verified that fire extinguishers and hose stations were provided at their designated locations and that they were in a satisfactory condition;

(5) verified that passive fire protection features (electrical raceway barriers, fire doors, fire dampers, steel fire proofing, penetration seals, and oil collection systems) were in a satisfactory material condition;

(6) verified when applicable that adequate compensatory measures were established for degraded or inoperable fire protection features; and

(7) reviewed the corrective action program to determine if the licensee identified and corrected fire protection problems.

• Fire Area R-3; High Pressure Core Spray Pump Room; August 29, 2006

• Fire Area RC-2; Cable Spreading Room; August 29, 2006

• Fire Area R-4; Residual Heat Removal Pump 2B Room; August 28, 2006

• Fire Area ASD; Adjustable Speed Drive Building; August 21, 2006

• Fire Area TG-1; Main Transformer Yard, July 20, 2006

• Fire Area RC-18; Motor Control Center Room, July 19, 2006

• Fire Area R-5; Residual Heat Removal Pump 2A Room; July 16, 2006 The inspectors completed seven samples.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

.1 Annual External Flood Protection

a. Inspection Scope

The inspectors reviewed the Columbia Generating Station Final Safety Analysis Report (FSAR), Technical Specifications, and corrective action database to identify any external flood threats to the facility. Final Safety Analysis Report Sections 2.4.2 and 3.4.1.5.1, document that there are no external flood threats, either from ground water, local precipitation, or from the nearby Columbia River. The inspectors toured the external areas for any credible flood sources.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.2 Internal Flood Protection

a. Inspection Scope

The inspectors performed the following:

(1) reviewed the Updated Safety Analysis Report, the flooding analysis, and plant procedures to assess seasonal susceptibilities involving internal flooding;

(2) reviewed the corrective action program to determine if the licensee identified and corrected flooding problems;

(3) inspected underground bunkers/manholes to verify the adequacy of

(a) sump pumps,

(b) level alarm circuits,

(c) cable splices subject to submergence, and

(d) drainage for bunkers/manholes;

(4) verified that operator actions for coping with flooding can reasonably achieve the desired outcomes; and

(5) walked down the areas listed below to verify the adequacy of:

(a) equipment seals located below the floodline,

(b) floor and wall penetration seals,

( c) watertight door seals,

(d) common drain lines and sumps,

(e) sump pumps, level alarms, and control circuits, and

(f) temporary or removable flood barriers.

• Reactor Building Elevation 422 ft level which consisted of all the emergency core cooling systems pump rooms; September 25, 2006

• Division 1 Emergency Diesel Generator Room; September 26, 2006;

The inspectors completed two samples.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance

a. Inspection Scope

During July and August, 2006, the inspectors reviewed the licensees controls and methods for ensuring that the ultimate heat sink (service water spray ponds) was maintained in a condition to support the safety function of the ultimate heat sink. The inspectors assessed:

(1) inspection results;

(2) verified the adequacy of the licensees bio-fouling control program; and

(3) implementation of the stations in-service inspection plan.

The inspectors completed one sample.

Introduction.

An URI was identified pending the NRCs determination of the regulatory aspects and evaluation of the safety significance of potential performance issues associated with inspection of the service water (SW) siphon line, an ASME Code Class 3 component, and the facilitys Inservice Inspection Plan to perform required periodic testing and examination of the submerged SW siphon line in accordance with ASME Section XI.

Description.

The stations SW spray ponds comprise the facilitys ultimate heat sink.

The design basis function of the ultimate heat sink is to provide a source of water for the SW system for 30 days without relying on makeup water, and to absorb the heat transferred to it from the plant via the SW system during that time period without exceeding its design temperature. The design of the spray ponds includes a non-isolable 30 inch siphon line connecting both spray ponds to automatically transfer water from one pond to the other. A majority of the siphon line is buried, however, sections of the line are not buried and are submerged in the spray pond after penetrating a spray pond wall. In each spray pond, the submerged siphon line includes a ninety degree elbow which then directs the siphon line downward terminating 18 inches from the bottom of the spray pond floor. The siphon line assures that the entire inventory of water in the ponds is available to either SW train and that makeup water added to one pond is available to the other. The siphon allows a single train of SW access to the full 30 day water supply provided by the two SW ponds as prescribed by NRC Regulatory Guide 1.27, Ultimate Heat Sink for Nuclear Power Plants, Revision 2, and as required by 10 CFR 50, Appendix A, Criterion 44, Cooling Water.

On July 20, 2006, the inspectors identified that Energy Northwest had not included the siphon line in the facilitys Inservice Inspection Program Plan and therefore had not conducted specific periodic inspections of the siphon line to confirm the integrity of the line. The inspectors noted that the applicable edition of Section XI of the ASME Boiler and Pressure Vessel Code, for the third inservice inspection interval, as provided in the

facilitys Inservice Inspection Program Plan, Interval 3, dated December 9, 2005, was the 2001 Edition, 2003 Addenda of Section XI. Article IWA-5244, Buried Components, Section (b)(2), provided that system pressure test for nonisolable buried components shall consist of a test to confirm that flow during operation is not impaired. Contrary to this requirement, Energy Northwest had not included the spray pond siphon line in the inservice inspection plan and therefore never conducted an inspection or examination of the line specifically for ASME code testing requirements. Energy Northwest documented the issue in the corrective action program in CR 2-06-05951. The inspectors reviewed Energy Northwests evaluation of CR 2-06-05951 and noted that the licensee determined that although the SW siphon line had not been examined historically for the purpose of ASME code inspection requirements, that past quarterly surveillance testing of the SW pumps demonstrated that the siphon line flow capability was not impaired. This demonstrated that flow through the buried portion of the siphon line was not impaired consistent with the testing requirements of IWA-5244, Section (b)(2).

Energy Northwest also stated that based on a Code interpretation that the siphon line was excluded from Section XI requirements because it was considered an open ended pipe. The inspectors disagree with the licensees interpretation. Energy Northwest also stated that they were submitting this issue to ASME Code committee for review and determination of the Code requirements for this siphon piping.

The inspectors also noted that Energy Northwest had not conducted any specific examination or inspection of the submerged sections of the SW siphon line, nor had they performed any visual exams of the portion (siphon) that penetrates into the two spray ponds. A through wall flaw in the exposed sections of the siphon line that penetrates the spray pond, would jeopardize the ability of the line to complete its siphon function and would therefore impact the ability of the ultimate heat sink to complete its design safety function as discussed above. The inspectors noted that development of a through wall flaw through these portions of the siphon line would not reveal themselves during periodic quarterly SW system testing as discussed above for the buried and exposed portions of the siphon line. Energy Northwest documented this additional concern in CR 2-06-06306. An URI was opened pending the NRCs evaluation of the resolution to CR 2-06-06306 to determine if a violation of ASME code testing requirements occurred with Energy Northwest not examining or testing the submerged and exposed portions of the siphon line (URI 05000397/2006004-02; ASME Code Testing of Service Water Siphon Line).

Analysis.

A determination of the safety significance associated with any performance deficiencies will be addressed in the resolution to the URI.

Enforcement.

A determination of the enforcement aspects associated with any performance deficiencies will be addressed in the resolution to the URI.

1R11 Licensed Operator Requalification

a. Inspection Scope

On September 19, 2006, the inspectors observed testing and training of senior reactor operators and reactor operators to identify deficiencies and discrepancies in the training, to assess operator performance, and to assess the evaluator's critique. The inspectors also observed the ability of the operators to respond to events and verified that the licensee configured the simulator consistent with the control room and plant.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the maintenance activities listed below to verify:

(1) the appropriate handling of structure, system, and component (SSC) performance or condition problems;

(2) the appropriate handling of degraded SSC functional performance;

(3) evaluate the role of work practices and common cause problems; and

(4) evaluate the handling of SSC issues reviewed under the requirements of the Maintenance Rule, 10 CFR Part 50 Appendix B, and the Technical Specifications.

• Adjustable Speed Drive failure; July 27, 2006

• Digital Electrical Hydraulic leak and closure of MS-V-164B; August 15, 2006

• Primary Containment Isolation Valve PI-VX-253 failed Surveillance; September 18, 2006

• Reactor Protection Motor Generator Set Bearing Lubrication; September 20, 2006 The inspectors completed four samples.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

.1 Risk Assessment and Management of Risk

a. Inspection Scope

The inspectors reviewed the risk assessment activities listed below to verify:

(1) performance of risk assessments when required by 10 CFR 50.65 (a)(4) and licensee procedures prior to changes in plant configuration for maintenance activities and plant operations;

(2) the accuracy, adequacy, and completeness of the information considered in the risk assessment;

(3) that the licensee recognizes, and/or enters as applicable, the appropriate licensee-established risk category according to the risk assessment results and licensee procedures; and

(4) the licensee identified and corrected problems related to maintenance risk assessments.

• Planned maintenance on RHR-V-6A, RHR-V-68A, RHR-V-27A; July 10, 2006

• Battery E-B2-1 250 VDC Battery On-line Replacement; August 17, 2006

• Planned maintenance on WMA-FN-53A; September 6, 2006

• Planned maintenance on WMA-FN-52A; September 5, 20006 The inspectors completed four samples.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors:

(1) reviewed plant status documents such as operator shift logs, emergent work documentation, deferred modifications, and standing orders to determine if an operability evaluation was warranted for degraded components;

(2) referred to the Updated Safety Analysis Report and design basis documents to review the technical adequacy of licensee operability evaluations;

(3) evaluated compensatory measures associated with operability evaluations;

(4) determined degraded component impact on any Technical Specifications;

(5) used the significance determination process to evaluate the risk significance of degraded or inoperable equipment; and

(6) verified that the licensee has identified and implemented appropriate corrective actions associated with degraded components.

• CR 2-06-06581; LPRM detector 40/17B has reached its end-of-life per Revision 1 to GE SIL 501; September 5, 2006

• CR 2-06-06305; The qualification of equipment components used in the battery rail charger maintenance activity has been questioned; August 23, 2006

• CR 2-06-03036; While manipulating the valve to the closed position, the stem bushing pushed up and out of the yoke assembly; July 28, 2006

• CR 2-06-05506; Breaker E-CB-8/DG2 did not show adequate gap between the trip latch and the trip shaft during inspection under Work Order 01121474; July 25/2006 The inspectors completed four samples.

b. Findings

Introduction.

The inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control for failure to implement adequate design control measures for the stations safety-related batteries in that Energy Northwest used a non-qualified battery charger. The finding had cross-cutting aspects in the area of problem identification and resolution due to a lack of engineering rigor associated with the resolution of a previously identified concern with qualification of a non safety-related battery charger used in safety-related applications.

Description.

On August 16, 2006, during observation of the on-line replacement of the safety-related 250 VDC battery (E-B2-1), per procedure PPM 10.25.206, Online Battery Replacement of 250 VDC E-B2-1, Revision 0, the inspectors noted that a non-class 1E qualified battery charger was installed on an alternate battery rack. The alternate battery rack was installed and bypassing a tier of battery cells on E-B2-1 which were being replaced. The alternate battery rack was therefore necessary to ensure that the safety-related function of E-B2-1 would be maintained. The inspectors were concerned that use of a non-qualified charger in this application may result in failure to ensure that the charger and E-B2-1 remain electrically separated and therefore could impact operability of E-B2-1. Energy Northwest documented the inspectors concern in CR 2-06-06165. A followup assessment by Energy Northwest concluded that the battery was operable but non-conforming in the as-found condition. Energy Northwest also determined that the non-qualified battery charger had been installed previously on E-B2-1 during a replacement of a different tier of battery cells on August 15, 2006.

The inspectors reviewed Energy Northwests corrective action program and noted the following condition reports which documented NRC identified examples of Energy Northwests failure to maintain configuration control of the stations safety-related batteries. Examples include:

• CR 2-06-06165; Alternate battery charger connected to E-B2-1; August 17, 2006

• CR 2-06-05937; Rail charger installed on cell 230 of E-B2-1 battery had leads incorrectly tied to safety-related battery cables; August 8, 2006

• CR 2-05-08510; Rail (single cell) charger was placed on E-B2-1 without the concerns of CR 2-05-08194 being addressed; November 2, 2005

• CR 2-05-08194; Design engineering position needed to identify any formalized design requirements are applicable to the battery rail charger; October 24, 2005 The safety-related function of the associated battery in each example noted above was maintained. However, in each case an adequate engineering evaluation justifying installation of a non-Class 1E battery charger in a safety-related application was not performed prior to installing the charger. The inspectors considered these examples to represent a failure to implement adequate design controls for the stations safety-related batteries.

The inspectors also reviewed the qualification requirements of the installed electrical isolation components for the single cell rail charger that Energy Northwest had evaluated in CR 2-05-08194 as noted above. The inspectors concluded that Energy Northwest failed to identify that the electrical isolation components (fuses, diodes) which had been installed to ensure electrical separation between the charger and the battery were not Class 1E qualified. The inspectors noted that FSAR, Section 8.3.2.1, provided that all non-Class 1E loads supplied by the Class 1E DC power systems are connected to Class 1E DC power supplies through Class 1E isolation devices. Although a battery charger, when operating normally would not be considered a load on a DC battery, under faulted conditions when electrical isolation devices would be needed, the charger could become a load. Energy Northwest failed to identify that the electrical isolation components between the charger and the battery were not Class 1E qualified as provided by the FSAR. Energy Northwest documented the inspectors observation in CR 2-06-06305 on August 23, 2006.

Analysis.

The performance deficiency associated with this finding is Energy Northwests failure to implement adequate design controls regarding use of temporarily installed non-Class 1E battery rail chargers on the stations safety-related batteries. This finding was more than minor because the finding was a design control issue which affected the mitigating systems cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the repetitive failure to provide adequate engineering analysis supporting the temporary installation of a non-qualified battery rail charger on a safety-related battery was not commensurate with ensuring the reliability of the stations safety-related batteries. Utilizing MC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the inspectors determined that the finding was of very low risk significance (Green) because it was a qualification issue confirmed not to result in loss of operability. Additionally, the finding did not represent a loss of safety function for a single train or for the system, and did not screen as potentially risk significant due to external events. This finding had cross-cutting aspects in the area of problem identification and resolution associated with the corrective action program component in that the licensee did not thoroughly evaluate design issues with the nonqualified rail charger, as documented in CR 2-05-08194.

Enforcement.

10 CFR Part 50, Appendix B, Criterion III, Design Control, requires in part that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews. Contrary to this requirement, since September 5, 2002, Energy Northwest failed to perform adequate design reviews to maintain appropriate control of the design and qualification of the stations

safety-related batteries with respect to temporary installation of a non-qualified battery charger as provided in PPM 10.25.181, Single Cell Charging of Batteries, Revision 0.

Because this finding was of very low safety significance and was entered into the licensees corrective action program as PER 206-0522, this violation is being treated as an NCV, consistent with Section VI.A of the Enforcement Policy (NCV 05000397/2006004-03, Inadequate Design Control of Safety-Related Batteries). Energy Northwest took immediate action to stop use of the single cell rail charger until proper qualification of the electrical isolation components could be provided.

1R19 Postmaintenance Testing

a. Inspection Scope

The inspectors selected the postmaintenance test activities of risk significant systems or components listed below for review. For each item, the inspectors:

(1) reviewed the applicable licensing basis and/or design-basis documents to determine the safety functions;

(2) evaluated the safety functions that may have been affected by the maintenance activity; and

(3) reviewed the test procedure to ensure it adequately tested the safety function that may have been affected. The inspectors either witnessed or reviewed test data to verify that acceptance criteria were met, plant impacts were evaluated, test equipment was calibrated, procedures were followed, jumpers were properly controlled, the test data results were complete and accurate, the test equipment was removed, the system was properly re-aligned, and deficiencies during testing were documented. The inspectors also reviewed the corrective action program to determine if the licensee identified and corrected problems related to postmaintenance testing.

• WO 01097263; Replace Kapton Diaphragm on MS-LS-61C; August 22, 2006

• WO 01116517; Replace DSA-V-34A; August 24, 2006

• WO 01121277; Diesel Cooling Water Leak Engine 1B2; July 1, 2006

• WO 01122403; GY-P-10A2 Discharge Hose Leak; July 27, 2006

• WO 01123082; Repair MS-V-164B Hydraulic Operator; Sept 21, 2006 The inspectors completed five samples.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report, procedure requirements, and Technical Specifications to ensure that the surveillance activities listed below demonstrated that the SSCs tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the following significant surveillance test attributes were adequate:

(1) preconditioning;

(2) evaluation of testing impact on the plant;

(3) acceptance criteria;

(4) test equipment;

(5) procedures;

(6) jumper/lifted lead controls;

(7) test data;

(8) testing frequency and method demonstrated Technical Specification operability;

(9) test equipment removal;

(10) restoration of plant systems;

(11) fulfillment of ASME Code requirements;

(12) updating of performance indicator data;

(13) engineering evaluations, root causes, and bases for returning tested SSCs not meeting the test acceptance criteria were correct;

(14) reference setting data; and

(15) annunciators and alarms setpoints. The inspectors also verified that the licensee identified and implemented any needed corrective actions associated with the surveillance testing.

• PPM 8.3.449; Control Rod Settle Time Test; Revision 0; August 28, 2006

• ISP-MS-Q902; RPS and Isolation Reactor Vessel Level Low, Level 3; RCIC Isolation; Revision 4; July 26, 2006

• PPM 15.2.1; Monthly Fire Pump Battery Testing; Revision 8; August 17, 2006

• WO 01113082; Open and Inspect External Condition of HPCS-P-2; July 18, 2006

• PPM 2.11.3; Equipment Drain System; Revision 24; August 14, 2006

• OSP-RHR/IST-Q704; RHR Loop A Operability Test; Revision 17; September 7, 2006

• ESP-B21-Q101 Quarterly Battery Testing 250 VDC E-B2-1; Revision 7; September 24, 2006 The inspectors completed seven samples which included a review of an in-service pump and valve test.

1R23 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report, plant drawings, procedure requirements, and Technical Specifications to ensure that the below listed temporary modification were properly implemented. The inspectors:

(1) verified that the modification did not have an affect on system operability/availability;

(2) verified that the installation was consistent with the modification documents;

(3) ensured that the post-installation test results were satisfactory and that the impact of the temporary

modification on permanently installed SSCs were supported by the test;

(4) verified that the modifications were identified on control room drawings and that appropriate identification tags were placed on the affected drawings; and

(5) verified that appropriate safety evaluations were completed. The inspectors verified that licensee identified and implemented any needed corrective actions associated with temporary modifications.

The inspectors completed one sample.

• TMR 05-026;MS-V-120A Level Switch MS-LS-24A Stuck in High Position; September 18, 2006

b. Findings

No findings of significance were identified.

1EP1 Exercise Evaluation

a. Inspection Scope

The inspectors reviewed the objectives and scenario for the 2006 biennial emergency plan exercise to determine if the exercise would acceptably test major elements of the emergency plan. The scenario simulated a large condensor tube leak, with a subsequent failure of the reactor protection system to complete a reactor scram.

Multiple main steam isolation valve failures then resulted in initiation of a small release of radioactivity to the environment. An initially small reactor coolant leak in containment greatly increased, ultimately resulting in a loss of reactor vessel level, uncovery and damage of reactor fuel, with a rapid increase in the offsite release of radioactivity to the environment. These simulated events enabled the licensee to demonstrate their capability to implement the emergency plan.

The inspectors evaluated exercise performance by focusing on the risk-significant activities of event classification, offsite notification, recognition of offsite dose consequences, and development of protective action recommendations, in the simulator control room and the following dedicated emergency response facilities:

• Technical Support Center

• Operations Support Center

• Emergency Operations Facility The inspectors also assessed recognition of and response to abnormal and emergency plant conditions, the transfer of decision making authority and emergency function responsibilities between facilities, onsite and offsite communications, protection of emergency workers, emergency repair evaluation and capability, and the overall implementation of the emergency plan to protect public health and safety and the

environment. The inspectors reviewed the current revision of the facility emergency plan, and emergency plan implementing procedures associated with operation of the above facilities and performance of the associated emergency functions. These procedures are listed in the Attachment to this report.

The inspectors compared the observed exercise performance to the requirements in the facility emergency plan, 10 CFR 50.47(b), 10 CFR Part 50, Appendix E, and to the guidance in the emergency plan implementing procedures and other federal guidance.

The inspectors attended the post-exercise critiques in each of the above facilities to evaluate the initial licensee self-assessment of exercise performance. The inspectors also attended a subsequent formal presentation of critique items to plant management.

The inspectors completed one sample during this inspection.

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation

a. Inspection Scope

The inspectors evaluated the conduct of a routine licensee emergency drill on August 1, 2006, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation (PAR) development activities. The inspectors observed emergency response operations in the simulated control room to verify that event classification and notifications were done in accordance with Procedure PPM 13.1.1, Classifying the Emergency, Revision 34. The inspectors also reviewed the licensees evaluation of the drill to compare any inspector-observed weakness with those identified by the licensee in order to verify whether the licensee was properly identifying failures.

The inspectors completed 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)

a. Inspection Scope

This area was inspected to assess the licensees performance in implementing physical and administrative controls for airborne radioactivity areas, radiation areas, high

radiation areas (HRAs), and worker adherence to these controls. The inspectors used the requirements in 10 CFR Part 20, the technical specifications, and the licensees procedures required by technical specifications as criteria for determining compliance.

During the inspection, the inspector interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors performed independent radiation dose rate measurements and reviewed the following items:

• Controls (surveys, posting, and barricades) of three radiations, high radiation, or airborne radioactivity areas

• Radiation work permits, procedures, engineering controls, and air sampler locations

• Barrier integrity and performance of engineering controls in an airborne radioactivity area

• Corrective action documents related to access controls

• Radiation work permit briefings and worker instructions

• Adequacy of radiological controls such as, required surveys, radiation protection job coverage, and contamination control during job performance

• Dosimetry placement in high radiation work areas with significant dose rate gradients

• Changes in licensee procedural controls of high dose rates - high radiation areas and very high radiation areas

• Posting and locking of entrances to all accessible high dose rates - high radiation areas and very high radiation areas

• Radiation worker and radiation protection technician performance with respect to radiation protection work requirements Either because the conditions did not exist or an event had not occurred, no opportunities were available to review the following items:

• Adequacy of the licensees internal dose assessment for any actual internal exposure greater than 50 millirem Committed Effective Dose Equivalent The inspectors completed 14 of the required 21 samples.

b. Findings

No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspectors assessed licensee performance with respect to maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). The inspectors used the requirements in 10 CFR Part 20 and the licensees procedures required by technical specifications as criteria for determining compliance. The inspectors interviewed licensee personnel and reviewed:

• Current 3-year rolling average collective exposure

• Site specific ALARA procedures

• Integration of ALARA requirements into work procedure and radiation work permit documents

• Shielding requests and dose/benefit analyses

• Dose rate reduction activities in work planning

• Use of engineering controls to achieve dose reductions and dose reduction benefits afforded by shielding

• Workers use of the low dose waiting areas

• First-line job supervisors contribution to ensuring work activities are conducted in a dose efficient manner

• Radiation worker and radiation protection technician performance during work activities in radiation areas, airborne radioactivity areas, or high radiation areas

• Self-assessments, audits, and special reports related to the ALARA program since the last inspection

• Corrective action documents related to the ALARA program and follow-up activities such as initial problem identification, characterization, and tracking The inspectors completed 5 of the required 15 samples and 6 of the optional samples.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

.1 Cornerstone: Occupational Radiation Safety

a. Inspection Scope

The inspectors reviewed licensee documents from April 1 through June 30, 2006. The review included corrective action documentation that identified occurrences in locked high radiation areas (as defined in the licensees technical specifications), very high radiation areas (as defined in 10 CFR Part 20.1003), and unplanned personnel exposures (as defined in NEI 99-02). Additional records reviewed included as low as reasonably achievable (ALARA) records and whole body counts of selected individual exposures. The inspectors interviewed licensee personnel that were accountable for collecting and evaluating the performance indicator (PI) data. In addition, the inspectors toured plant areas to verify that high radiation, locked high radiation, and very high radiation areas were properly controlled. PI definitions and guidance contained in NEI 99-02, "Regulatory Assessment Indicator Guideline," Revision 3, were used to verify the basis in reporting for each data element.

• Occupational Exposure Control Effectiveness The inspector completed the required sample

(1) in this cornerstone.

.2 Cornerstone: Public Radiation Safety

a. Inspection Scope

The inspectors reviewed licensee documents from April 1 through June 30, 2006.

Licensee records reviewed included corrective action documentation that identified occurrences for liquid or gaseous effluent releases that exceeded PI thresholds and those reported to the NRC. The inspectors interviewed licensee personnel that were accountable for collecting and evaluating the PI data. PI definitions and guidance contained in NEI 99-02, "Regulatory Assessment Indicator Guideline," Revision 3, were used to verify the basis in reporting for each data element.

• Radiological Effluent Technical Specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences The inspectors completed the required sample

(1) in this cornerstone.

b. Findings

No findings of significance were identified.

.3 Cornerstone: Emergency Preparedness

a. Inspection Scope

The inspectors reviewed licensee evaluations for the three emergency preparedness cornerstone performance indicators of Drill and Exercise Performance, Emergency Response Organization Participation, and Alert and Notification System Reliability, for the period October 1, 2005, through June 30, 2006. The definitions and guidance of NEI 99-02, "Regulatory Assessment Indicator Guideline," Revisions 3 and 4, and the licensee Emergency Plan Instruction 18, "Emergency Preparedness NRC Performance Indicators," 6/15/2006, were used to verify the accuracy of the licensees evaluations for each performance indicator reported during the assessment period.

• Drill and exercise scenarios and licensed operator simulator training sessions, notification forms, and attendance and critique records associated with training sessions, drills, and exercises conducted during the verification period.

• Emergency responder qualification, training, and drill participation records.

• Alert and notification system testing procedures, maintenance records, and a 100 percent sample of siren test records. The inspectors also reviewed other documents listed in the Attachment to this report.

The inspectors completed three samples during the inspection.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Cross-References to PI&R Findings Documented Elsewhere

Section 1R04 describes a finding for the failure to perform adequate design reviews to maintain appropriate control of the design and qualification of the stations safety-related batteries with respect to temporary installation of a non-qualified charger.

.2 Daily Corrective Action Document Review

a. Inspection Scope

The inspectors performed a review of all documented condition reports and problem evaluation reports to help identify repetitive equipment failures or specific human performance issues for followup inspection using other baseline inspection procedures.

The review was accomplished by evaluating Energy Northwests electronic condition report and problem evaluation report databases and attending periodic plant status meetings.

b. Findings

No findings of significance were identified.

.3 Review of Identification and Resolution of Problems Associated with Radiation Protection

a. Inspection Scope

The inspectors evaluated the effectiveness of the licensees problem identification and resolution process with respect to the following inspection areas:

• Access Control to Radiologically Significant Areas (Section 2OS1)

• ALARA Planning and Controls (Section 2OS2)

The inspectors completed two samples.

b. Findings

No findings of significance were identified.

4OA3 Event Followup

.1 Adjustable Speed Drive Cooling Failure

On July 20, 2006, a leak in a adjustable speed drive (ASD) flexible coolant discharge hose forced one of two channels of ASD supplying a reactor recirculation pump (RRC) to be removed from service. Prior to the removal of the ASD channel 1A/2, reactor power was lowered to 90 percent to allow removal of the ASD channel. After the hose was replaced, a satisfactory test of the ASD channel was performed. Reactor power was lowered to 70 percent power to allow recovery of the ASD system. Reactor power was restored to 100 percent on July 21, 2006. The inspectors reviewed plant parameters, operator logs and operator response to the event including adherence to and quality of plant procedures used to address the failure.

b. Findings

No findings of significance were identified.

.2 Loss of Reactor Recirculation Pump

a. Inspection Scope

On July 27, 2006, one week after replacement, the discharge coolant hose on ASD channel 1A/2 ruptured during full power operations. As a result, electrical components of the ASD system channel 1A/2 supplying RRC-P-1A were sprayed with coolant. The electrical perturbation caused the other channel of ASD channel 1A/1 supplying RRC-P-1A to trip, resulting in the loss of RRC-P-1A and entering into single loop operation.

Reactor power was stabilized at approximately 67 percent power. ASD channel 1A/1 was tested satisfactorily and RRC-P-1A was started. The plant was returned to two loop operation. Wetted electrical components and all flexible cooling hoses in the ASD system were replaced and tested satisfactorily. The replacement hose was of a different material type and was procured from a different manufacturer. The plant was returned to 100 percent power on July 29, 2006. An Energy Northwest analysis performed on both

hose failures indicated that a similar failure mechanism in both cases. Common to these failures was a defect found on the inside diameter surface of the hose which resulted in breaching the inner rubber sleeve and pressurizing the outer cover to failure. The inspectors reviewed plant parameters, operator logs and operator response to the event including adherence to and quality of plant procedures used to address the leak.

b. Findings

No findings of significance were identified.

.3 Hydraulic Leak on Low Pressure No. 2 Intercept Valve

a. Inspection Scope

On August 14, 2006, Energy Northwest reduced power to 60 percent in response to a hydraulic leak on an intercept valve on low pressure turbine No. 2. The inspectors reviewed plant parameters, operator logs and operator response to the event including adherence to and quality of plant procedures used to address the leak.

b. Findings

No findings of significance were identified.

.4 Battery Acid Spill in Vital Area

a. Inspection Scope

On August 29, 2006, a battery jar, which had been previously installed in a station battery, was inadvertently tipped over during movement of the jar onto a pallet located in the 467 foot vital island area of the radwaste building. As a result, approximately 12 gallons of sulfuric acid spilled on the floor. Electrical workers promptly contacted the control room and contained the spilled acid. The inspectors assessed the situation locally to determine the extent of the spill and any impact on operators to operate safety-related equipment located in the 467 foot vital area.

b. Findings

No findings of significance were identified.

4OA6 Meetings, Including Exit

On August 24, 2006, the inspector presented the occupational radiation safety inspection results to Mr. R. Hogue, Acting Vice President, Nuclear Generation, and other members of his staff who acknowledged the findings. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

On August 24, 2006, the inspector conducted an exit meeting at the NRC Region IV offices to present the inspection results to Mr. M. Reis, Supervisor, Emergency Preparedness, and other members of his staff, who acknowledged the findings. The inspector confirmed that proprietary information was not provided or examined during the inspection.

On September 15, 2006, the lead inspector presented the results of the biennial emergency preparedness exercise inspection to Mr. D. Atkinson, Vice President, Nuclear Generation, and other members of his staff, who acknowledged the findings. The inspector confirmed that proprietary information was not provided or examined during the inspection.

On October 5, 2006, the resident inspectors presented the inspection results to Mr. Scott Oxenford, Vice President Technical Services, and other members of his staff who acknowledged the findings. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

On October 10, 2006, the resident inspectors re-exited with Mr. T. Lynch, Plant General Manager.

4OA7 Licensee Identified Violations

The following finding of very low significance was identified by the licensee and is a violation of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600 for being dispositioned as Non-Cited Violations (NCV).

Technical Specification 5.7.1 requires high radiation areas with dose rates not exceeding 1.0 rem per hour be barricaded and conspicuously posted. However, on August 3, 2006, the licensee identified that the septa decon tank on the 507-foot elevation of the radwaste building was not barricaded and conspicuously posted even though it contained areas accessible to a portion of the whole body with dose rates as high as 120 millirems per hour. The finding was documented in the licensee's corrective action program in Condition Report 2-06-05784 and 2-06-06339. This finding is of very low significance because it did not involve:

(1) an ALARA finding,

(2) an overexposure,

(3) a substantial potential for overexposure, or

(4) an impaired ability to assess doses.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Energy Northwest

D. Atkinson, Vice President, Nuclear Generation

S. Belcher, Manager, Operations

I. Borland, Manager, Radiation Protection

D. Coleman, Manager, Performance Assessment and Regulatory Programs

G. Cullen, Licensing Supervisor, Regulatory Programs

A. Khanpour, General Manager, Engineering

T. Lynch, Plant General Manager

W. Oxenford, Vice President, Technical Services

J. Parrish, Chief Executive Officer

R. Torres, Manager, Quality Assurance

C. Whitcomb, Vice President, Organizational Performance and Staffing

S. Boynton, Systems Engineering Manager

D. Holmes, Emergency Planner

R. Jorgensen, Emergency Planner

A. Mouncer, Vice President, Corporate Services

T. Powell, Emergency Planner

M. Reis, Supervisor, Emergency Preparedness

W. Sawyer, Emergency Planner

F. Schill, Licensing

NRC Personnel

R. Cohen, Resident Inspector and Acting Senior Resident Inspector

J. Drake, Acting Resident Inspector

Z. Dunham, Senior Resident Inspector

ITEMS OPENED AND CLOSED

Items Opened, Closed, and Discussed During this Inspection

Opened

05000397/2006004-01 URI Potential performance issue associated with the use of

alternate boron injection via the Reactor Core Isolation

Cooling (RCIC) System (Section 1R04)05000397/2006004-02 URI Potential performance issues associated with inspection of

the service water (SW) siphon line, an ASME Code Class 3

component, and the facilitys Inservice Inspection Plan to

perform required periodic testing and examination of the

submerged SW siphon line in accordance with ASME

Section XI. (Section 1R07)

Enclosure

Opened and Closed

05000397/2006004-03 NCV Failure to implement adequate design control measures for

the stations safety-related batteries in that Energy

Northwest used a non-qualified battery charger

(Section 1R15)

Closed

None.

Discussed

None.

PARTIAL