IR 05000259/2009008

0500259-09-008, 05000260-09-008, and 05000296-09-008; on 10/19/09 - 12/28/09; Browns Ferry Nuclear Plant, Units 1, 2, and 3; Component Design Basis Inspection

ML100420056
Person / Time
Site:	Browns Ferry
Issue date:	02/09/2010
From:	Binoy Desai NRC/RGN-II/DRS/EB1
To:	Krich R Tennessee Valley Authority
References
IR-09-008
Download: ML100420056 (37)

Text

February 9, 2010

SUBJECT:

BROWNS FERRY NUCLEAR PLANT - NRC COMPONENT DESIGN BASIS INSPECTION REPORT 05000259/2009008, 05000260/2009008, AND 05000296/2009008

Dear Mr. Krich:

On December 28, 2009, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Browns Ferry Nuclear Plant, Units 1, 2, and 3. The enclosed inspection report documents the inspection findings which were discussed on November 20, 2009 with Mr. Rusty West, Site Vice President, and on December 28, 2009, via telephone, with Mr. Jim Randich, Plant Manager as well as other members of your staff.

The inspection examined activities conducted under your licenses as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your licenses. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, the inspectors identified two findings of very low safety significance (Green), which involved violations of NRC requirements. Additionally, one licensee-identified violation which was determined to be of very low safety significance is listed in this report. However, because of their very low safety significance and because they are entered into your corrective action program, the NRC is treating these findings as Non-Cited Violations (NCVs) consistent with Section VI.A.1 of the NRCs Enforcement Policy. If you contest any of these NCVs you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the United States Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Browns Ferry Nuclear Plant. In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at Browns Ferry Nuclear Plant. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

TVA

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Binoy B. Desai, Chief

Engineering Branch 1

Division of Reactor Safety

Docket Nos.: 50-259, 50-260, 50-296 License Nos.: DPR-33, DPR-52, DPR-68

Enclosure:

Inspection Report 05000259/2009008, 05000260/2009008, and

05000296/2009008 w/Attachment: Supplemental Information

REGION II==

Docket Nos.:

50-259, 50-260, 50-296

License Nos.:

DPR-33, DPR-52, DPR-68

Report No.:

05000259/2009008, 05000260/2009008, and 05000296/2009008

Licensee:

Tennessee Valley Authority (TVA)

Facility:

Browns Ferry Nuclear Plant, Units 1, 2, and 3

Location:

Corner of Shaw and Nuclear Plant Roads

Athens, AL 35611

Dates:

October 19, 2009 through December 28, 2009

Inspectors:

R. Berryman, P.E., Senior Reactor Inspector (Lead)

R. Aiello, Senior Operations Engineer

C. Even, Reactor Inspector

R. Patterson, Reactor Inspector

J. Leivo, Contractor

H. Campbell, Contractor

Accompanying Personnel:

J. Kent, Construction Inspector (trainee)

Approved by:

Binoy B. Desai, Chief

Engineering Branch 1

Division of Reactor Safety

SUMMARY OF FINDINGS

IR 05000259/2009008, 05000260/2009008, and 05000296/2009008; 10/19/09 - 12/28/09;

Browns Ferry Nuclear Plant, Units 1, 2, and 3; Component Design Basis Inspection.

This inspection was conducted by a team of four NRC inspectors, one NRC inspector who was in training, and two NRC contract inspectors. Two Green findings, both of which were non-cited violations (NCVs) were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP). Finding for which the SDP does not apply may be Green or is assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, (ROP) Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Green non-cited violation (NCV) of TS 5.4.1 for the failure to have an adequate procedure that would ensure tornado depressurization protection of the emergency diesel generators (EDGs).

Abnormal Operating Instruction, 0-AOI-100-7, Severe Weather, did not provide guidance on how to provide pressure equalization of the EDG building for mitigating atmospheric depressurization associated with tornado conditions that could impact the EDG building ventilation system. The design of the EDG ventilation system intake and exhaust dampers requires the dampers to be manually opened prior to a tornado depressurization event to ensure the EDG building and ventilation system remain intact and operable during and after a tornado. This finding was entered into the licensees corrective action program as problem evaluation report (PER) 206919. As an immediate corrective action, the licensee added steps to procedure 0-AOI-100-7 to station an operator in the EDG building to perform required manual actions in the event of a tornado warning in the area.

This finding is more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and operability of the EDGs to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e. Operating (Post Event)

Procedures (AOPs). The inspectors assessed the finding using a Phase I Significance Determination Process (SDP) screening which determined a Phase III SDP evaluation was required due to the fact that the finding involved the loss or degradation of equipment specifically designed to mitigate a severe weather initiating event (e.g., tornado doors). The loss of this equipment by itself, during the external initiating event it was intended to mitigate, would degrade two or more trains of a multi-train safety system. A Phase III SDP evaluation was performed in accordance with NRC Inspection Manual Chapter 0609 Appendix A by a regional Senior Reactor Analyst using the NRC Standardized Plant Analysis

Risk (SPAR) model. The analysis determined that the performance deficiency resulted in a core damage frequency (CDF) risk increase of less than 1E-6.

Therefore, the finding was characterized as having very low safety significance (Green). The large early release frequency (LERF) result was less than 1E-7 which would not override the CDF risk characterization. The dominant sequence was a tornado which caused damage to the EDG dampers for all EDGs and resulted in a loss of offsite power (LOOP). The risk was mitigated by the low initiating event likelihood for tornadoes and the probability for EDG recovery which was proceduralized.

The inspectors determined that the use of operating experience (OE) information was a significant cause of this performance deficiency. Regulatory Information Summary 2006-03, Post Tornado Operability of Ventilation and Air Conditioning Systems as well as an internal licensee OE had raised a similar concern. The licensee was unaware of the vulnerability of the EDG ventilation system to tornado depressurization events until it was brought to their attention by the inspectors. The licensees failure to use available OE is directly related to the OE component of the cross-cutting area of Problem Identification and Resolution and the aspect of implementing OE through changes to procedures (P.2.(b)).

(Section 1R21.2.12)

Green.

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to provide adequate guidance in existing procedures utilized for flow balancing of the emergency equipment cooling water (EECW) system. The EECW system provided the heat sink for station safety-related heat loads including cooling for the residual heat removal (RHR) and core spray (CS) room coolers. The installed strainers on the EECW system are capable of filtering debris greater than 1/8 inch (.125 inches), potentially allowing debris less than 1/8 inch to pass through and clog downstream throttle valves. A clog in the throttle valves would prevent adequate flow from reaching safety-related heat exchangers unless procedural guidance or limitations prevented throttling valves to disk-to-seat clearances of less then 1/8 inch. The existing EECW flow balance procedure was inadequate in that it made no provision in the acceptance criteria to limit or evaluate minimum throttle valve seat/disc clearance, and the subsequent potential for increased flow obstruction, resulting from system flow balancing. This finding was entered into the licensees corrective action program as problem evaluation reports (PERs) 208374 and 208636. Planned corrective actions included a revision to EECW flow balancing procedures. The inspectors verified and discussed with the licensee existing indications that are available to alert the operator of potential clogging.

This finding is more than minor because it affects the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and operability of the RHR and CS pump room coolers to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e.

Maintenance and Testing (Pre-event) Procedures. The team assessed this finding using the Significance Determination Process (SDP) and determined that the finding was of very low safety significance (Green) because the inspectors found no documented history of an actual loss of safety system function. This finding was reviewed for cross-cutting aspects and none were identified. (Section 1R21.2.15)

Licensee-Identified Violations

• One violation of very low safety significance, which was identified by the licensee, has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and corrective action tracking number are listed in Section 4OA7 of this report.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R21 Component Design Bases Inspection

.1 Inspection Sample Selection Process

The team selected risk-significant components and operator actions for review using information contained in the licensees Probabilistic Risk Assessment (PRA). In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum value greater than 1 X10-6. The components selected were located within the residual heat removal (RHR) system, plant and drywell control air systems, main steam isolation valves (MSIVs), emergency diesel generator (EDG)support and electrical subsystems, 4160 VAC electrical system, common station service transformers (CSSTs), 125/250 VDC battery system, and main steam line radiation monitors. The sample selection included 17 components, five operator actions, and five operating experience items. Additionally, the team reviewed one modification related to extended power uprate activities documented in Section 4OA5 of this report.

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases had been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modification, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed performance test results, significant corrective action, repeated maintenance, maintenance rule (a)1 status, RIS 05-020 (formerly GL 91-18) conditions, NRC resident inspector input of problem equipment, system health reports, industry operating experience and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. An overall summary of the reviews performed and the specific inspection findings identified are included in the following sections of the report.

.2.1 RHR Pump B and D Minimum Flow Valve (MOV FCV-74-30)

a. Inspection Scope

The team reviewed design and licensing documentation for motor-operated valve (MOV)

FCV-74-30, including the Updated Final Safety Analysis Report (UFSAR), Design Basis Documents (DBDs), and drawings to verify the capability of the valves to meet design basis requirements. The design function of these valves is to protect the RHR Loop II pumps from possible damage during low flow conditions; at high pump flow the valve closes while low pump flow results in a signal to open the valve thereby providing a minimum flow path. Orifice sizing, setpoint, and related flow calculations were reviewed to verify that they were appropriate. Additionally, discussions with design engineering personnel were conducted and completed surveillance test results were reviewed to verify that the recirculation flow path was adequate.

b. Findings

No findings of significance were identified.

.2.2 RHRSW to A Train RHR Heat Exchanger Check Valve (1-CKV-23-510)

a. Inspection Scope

The team reviewed portions of the RHR service water (RHRSW) and RHR system DBDs and drawings to determine the functional requirements for valve 1-CKV-23-510. This valve allows forward flow through RHRSW coolant heat exchanger and prevents potential reverse flow from the A train RHR heat exchanger. The inspectors also conducted a walkdown of the RHRSW pumps, piping and valves, with particular focus on CKV-23-510. Additionally, the inspectors verified that a program was in place to appropriately monitor and address any valve degradation.

b. Findings

No findings of significance were identified.

.2.3 A and B Train RHR Supply to Torus Cooling (MOV FCV-74-57 and -59)

a. Inspection Scope

The team reviewed applicable DBDs, drawings, and calculations to verify the capability of these valves to perform the design basis functions. American Society of Mechanical Engineers (ASME)Section XI stroke-timing test data was reviewed for the valves to verify that requirements were satisfied and potential adverse trends were monitored.

Test procedures were also reviewed to evaluate the valve throttling capability during RHR pump flow testing to verify it was adequate. Additionally, calculations documenting the design requirements of maximum differential pressures, actuator requirements and weak-link evaluations were reviewed to verify that they were appropriate.

b. Findings

No findings of significance were identified.

.2.4 RHR Pumps A, B, C and D

a. Inspection Scope

The team reviewed portions of the UFSAR, Technical Specifications (TS), and the RHR DBD to verify the capability of the pumps to perform the required function during postulated design basis events. Calculations establishing pump performance requirements during design basis events were reviewed to verify that the conclusions were appropriate. In-service testing (IST) and comprehensive test results were reviewed to assess potential pump degradation and to verify that pump performance was sufficient to satisfy design basis accident requirements. In addition, the team reviewed the system health reports, corrective action documentation, and the station response to Bulletin 88-04, Potential Safety-Related Pump Loss, to verify that any potential degradation was being tracked and addressed appropriately. Additionally, the inspectors interviewed the RHR system engineer to discuss the overall health and performance history of the RHR system. The inspectors also conducted a walkdown of the Unit 1 loop II pumps and associated piping and valves to verify that any adverse material conditions were being appropriately addressed.

b. Findings

No findings of significance were identified.

.2.5 EDG-Mechanical (Unit 1-2 EDGs A, B, C, D)

a. Inspection Scope

The team reviewed portions of the UFSAR, EDG DBD, calculations, System health reports, related problem evaluation reports (PERs), operating procedures, and a selection of completed TS surveillances to verify the capability of the EDGs to supply onsite AC electrical power during design basis events. A walkdown of the B EDG was conducted to verify that adverse material conditions, if any, were being appropriately addressed. The inspectors also reviewed the EDG fuel oil transfer pump test procedures and completed test results to verify that the fuel oil transfer pumps were capable of performing their required function during design basis events.

b. Findings

No findings of significance were identified.

.2.6 Unit 1 Inboard Main Steam Isolation Valves (MSIVs), (1-FCV-1-14, -26, -37, and -51)

a. Inspection Scope

The team reviewed applicable portions of the UFSAR, TS, calculations, and drawings to verify the capability of the MSIVs to perform the required functions during design basis events. The inspectors interviewed the system engineer to discuss the overall health of the MSIVs and associated PERs to verify that adverse material conditions, if any, were being appropriately addressed. The inspectors also reviewed design change notice (DCN) 51143, Main Steam and Standby Liquid Control System. This DCN was reviewed by the team, with focus on the technical evaluation of the major changes to the MSIVs required for a planned power uprate. The associated 10 CFR 50.59 Evaluation and related thermal hydraulic calculations for this DCN were also reviewed by the team to verify that the results were appropriate. MSIV stroke-timing test results were also reviewed to verify that any degradation was being appropriately monitored and addressed.

b. Findings

No findings of significance were identified.

.2.7 250 VDC Battery Board 1

a. Inspection Scope

The team selectively reviewed electrical one-line diagrams, loading calculations, voltage drop calculations, short-circuit calculations, and associated electrical protection to determine the capability of the battery board to serve the required power to 250 VDC loads in accordance with the design and licensing basis as well as for station blackout (SBO) events. For a sample of limiting conditions identified for the battery boards in the design calculations, the team reviewed operating procedures to confirm that the procedures were consistent with the actions prescribed by the calculations for selected out-of-service configurations. The team reviewed recent system health reports, a sample of available preventive maintenance and surveillance records, and associated corrective action history. The team also performed a non-intrusive visual inspection of the battery board to assess the installation configuration, material condition, and potential vulnerability to external hazards to verify that any degraded conditions were being appropriately addressed.

b. Findings

Introduction.

The team indentified an Unresolved Item (URI) regarding safety-related molded-case circuit breakers (MCCBs) in safety-related applications. TVA had not implemented a test program to detect potential deterioration or to assure that all installed safety-related MCCBs would perform satisfactorily in service.

Description.

TVA procedure 0-TI-395, Breaker Testing and Maintenance Program, required that critical molded case circuit breakers be subject to preventative maintenance (PM) activities. This included, in part, inspection for overheating, mechanical operation, enclosure inspection, overload trip testing, and instantaneous trip testing. The TVA program required that the testing and PM activities be performed every four to six years. UFSAR 8.6.4.1.1 stated, in part, that zero-resistance short circuits at the battery board or any point downstream can be cleared by the breakers operating within their ratings. To ensure this was satisfied by the installed equipment, degradation of breaker performance should have been detectable and acceptably controlled by periodic testing and preventive maintenance. Additionally, UFSAR 8.5.2.11 stated, in part, that the standby AC power system will meet or exceed the requirements of IEEE-308, Criteria for Class 1E Power Systems at Nuclear Generating Stations. This standard recommended that periodic tests be performed at scheduled intervals to detect deterioration of equipment and to demonstrate operability of components that are not exercised during normal operation.

Four MCCBs on 250 VDC Battery Board 1 were selected by the team (breakers 607, 705, 712, 715) as part of the inspection sample. No records of PM activities were found on these four MCCBs. While addressing the extent of this condition, TVA identified a total of 622 safety-related MCCBs in both AC and DC applications for which no testing was being performed.

As a result of the inspectors observations, TVA initiated PER 209095 and scheduled testing of the 622 safety-related MCCBs that were not being tested in accordance with the TVA PM program. As of December 28, 2009, TVA had successfully tested 30 of the 622 breakers. This represented a small sample of the total breaker population. The inspectors plan to review additional PM test results to determine if this performance deficiency is more than minor.

Summary. This issue is unresolved pending further inspection to determine the extent of condition and impact of not implementing a test program to assure that all installed safety-related MCCBs would perform satisfactorily. (URI 05000259, 260, 296/2009008-01, Safety-Related Molded Case Circuit Breakers)

.2.8 4160 VAC Shutdown Board C

a. Inspection Scope

The team selectively reviewed electrical one-line diagrams, loading calculations, short circuit calculations, and electrical protection, to determine the capability of shutdown board C to serve the required power to the associated 4160 VAC loads in accordance with the design and licensing basis. The team also selectively reviewed the schematic diagrams for the shutdown board trip and transfer circuits to confirm that the design basis for shutdown board interlocks and transfer to the diesel generators was satisfied and that no credible single failure vulnerabilities existed in the logic circuits. The team reviewed selected results of preventive maintenance and protective relay surveillances and functional tests to verify that any equipment performance degradation was being adequately addressed. The team reviewed system health reports and associated corrective action history to verify that any degraded conditions were being appropriately addressed. The team also performed a non-intrusive visual inspection of the shutdown board and logic panel to assess the installation configuration, material condition, and potential vulnerability to external hazards to verify that any degraded material conditions were being appropriately addressed.

b. Findings

No findings of significance were identified.

.2.9 Transformer 161 KV/4160 VAC CSST-A

a. Inspection Scope

The team reviewed the electrical one-line diagrams, loading calculations, electrical protection, tap settings, and nameplate data, to determine the adequacy of transformer 161KV/4160 VAC CSST-A to supply required power as a qualified alternate source under design and licensing basis conditions. The team reviewed recent results of transformer preventive maintenance including dissolved gas analysis, temperature trending, and protective relay calibrations to verify that equipment performance degradation was being appropriately addressed. The team reviewed recent system health reports and associated corrective action history to verify that any degraded conditions were being appropriately addressed. The team also interviewed the transformer system engineer and performed a non-intrusive visual inspection of the transformer to assess the installation configuration, material condition, and potential vulnerability of the transformer to external hazards to verify that any degraded material conditions were being appropriately addressed.

b. Findings

No findings of significance were identified.

.2.10 EDG (Electrical)

a. Inspection Scope

The team reviewed related design basis documentation, drawings, TS, and the UFSAR to identify design, maintenance, and operational requirements for the EDGs. EDG start, stop, and shutdown circuits were reviewed to verify that the logic of operation was consistent with the accident analysis and that the energy sources used for control functions would be available and unimpeded during accident conditions. The team reviewed EDG loading calculations including voltage, frequency and loading sequences to verify the capability of the EDGs to perform their intended safety function. The team also reviewed diesel batteries including voltage calculations, maintenance requirements, and operating procedures to verify that the terminal voltage of the battery would remain at or greater than design basis minimum voltage during the worst case design basis accident. Surveillance test procedures were reviewed to verify that applicable test acceptance criteria and test frequency requirements for the EDGs were satisfied. Field walkdowns were performed to assess observable material conditions and verify that system alignment was consistent with current drawings and operating procedures.

Selected PERs and work orders were reviewed by the team to verify that any potential component degradation was monitored and appropriately addressed.

b. Findings

No findings of significance were identified.

.2.11 RHR Pump Motors

a. Inspection Scope

The team reviewed the TS, UFSAR, and design basis documentation to identify the RHR pump motors design basis functions and related accident analysis assumptions. The team reviewed the adequacy, reliability, and availability of the power supply to the RHR pumps in normal and degraded voltage conditions. This review included review of motor specifications, vendor manual requirements, and breaker specifications to verify that they were appropriate. A walk-down of the motors and pumps was conducted to assess the physical condition of the motor and verify that the system configuration was consistent with the design basis assumptions, system operating procedures, and plant drawings.

b. Findings

No findings of significance were identified.

.2.12 EDG Building Ventilation Dampers

a. Inspection Scope

The team reviewed the UFSAR, system description, abnormal operating instructions (AOIs), control logic, drawings and operator actions to verify that the EDG building intake and exhaust ventilation dampers could perform their function of providing cooling to components in the EDG building. A system walkdown was performed to verify physical damper dimensions and that linkages could operate as designed. The team also reviewed selected operating experience issues dealing with EDG building ventilation dampers to verify that issues at other facilities were being appropriately identified and addressed.

b. Findings

Introduction.

The inspectors identified a Green non-cited violation (NCV) of TS 5.4.1 for the failure to have an adequate procedure that would ensure tornado depressurization protection of the EDGs. Abnormal Operating Instruction, 0-AOI-100-7, Severe Weather, did not provide guidance on how to provide pressure equalization of the EDG building for mitigating atmospheric depressurization associated with tornado conditions that could impact the EDG building ventilation system. The significance of this violation was determined using Phase III of the Significance Determination Process (SDP).

Description.

The EDGs are part of the standby AC power supply and distribution system. BFN UFSAR Section 8.5.2 states, in part, that the standby AC power system will meet or exceed the requirements of IEEE-308. IEEE-308-1971, Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations Section 4.1 states, in part, that Class 1E electric systems shall be designed to assure that a tornado will not cause a loss of electric power to a number of engineered safety features sufficient to jeopardize the safety of the station. BFN UFSAR Section 12.2.1 states, in part, that loading considerations for structures should include tornado depressurization. However, there was not an explicit requirement for the protection of the Class 1E power systems from environmental events to be accomplished by a specific fixed design feature or by procedural guidance to provide protection using system configuration.

The EDG ventilation cooling system intake and exhaust dampers are interlocked with the ventilation fans that take outside air to cool the components inside the EDG room. The intake and exhaust ventilation dampers are normally closed, and open, when the EDG starts or when the ventilation fans are manually started. Manual operation of the ventilation fans is performed locally. When the EDGs are not running, the EDG ventilation system intake and exhaust dampers would have to be manually opened prior to a tornado depressurization event to ensure the EDG building and ventilation system remain intact and operable during and after a tornado. Procedure 0-AOI-100-7 did not provide guidance during a tornado warning to verify that the ventilation dampers in the EDG building are opened. In a tornado event where the EDGs are not running, this condition would result in the inability to mitigate atmospheric depressurization of the EDG building associated with a tornado. This represented a potential common-cause failure of all eight EDG ventilation systems and subsequent potential inoperability of the EDGs. As an immediate corrective action, the licensee added steps to 0-AOI-100-7 procedure to station an operator in the EDG building to perform required manual actions in the event of a tornado warning in the area.

Analysis.

The failure to have an adequate procedure to ensure tornado depressurization protection of the EDG buildings is a performance deficiency. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and operability of the EDGs to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e.

Operating (Post Event) Procedures (AOPs). The inspectors assessed the finding using a Phase I SDP screening which determined a Phase III SDP evaluation was required due to the fact that the finding involved the loss or degradation of equipment specifically designed to mitigate a severe weather initiating event (e.g., tornado doors). The loss of this equipment by itself, during the external initiating event it was intended to mitigate would degrade two or more trains of a multi-train safety system. A Phase III SDP evaluation was performed in accordance with NRC Inspection Manual Chapter 0609 Appendix A by a regional SRA using the NRC SPAR model. The analysis determined that the performance deficiency resulted in a core damage frequency (CDF) risk increase less than 1E-6. Therefore, the finding was characterized as having very low safety significance (Green). The large early release frequency (LERF) result was less than 1E-7 which would not override the CDF risk characterization. The dominant sequence was a tornado which caused damage to the EDG dampers for all EDGs and resulted in a loss of offsite power (LOOP). The risk was mitigated by the low initiating event likelihood for tornadoes and the probability for EDG recovery which was proceduralized.

The inspectors determined that the use of operating experience (OE) information was a significant cause of this performance deficiency. Regulatory Information Summary 2006-03, Post Tornado Operability of Ventilation and Air Conditioning Systems as well as an internal licensee OE had raised a similar concern. The licensee was unaware of the vulnerability of the EDG ventilation system to tornado depressurization events until it was brought to their attention by the inspectors. The licensees failure to use available OE is directly related to the OE component of the cross-cutting area of Problem Identification and Resolution and the aspect of implementing OE through changes to procedures (P.2.(b)).

Enforcement.

Technical Specification 5.4.1, Procedures, requires in part that procedures shall be established, implemented, and maintained covering the activities in Regulatory Guide (RG) 1.33, Rev. 2, Quality Assurance Program Requirements.

Appendix A of RG 1.33 states, in part, that typical safety-related activities such as combating emergencies and other significant events including acts of nature, e.g.,

tornado, shall be covered by written procedures. Contrary to the above, TVA did not have adequate procedures to ensure tornado depressurization protection of the EDG buildings. This condition has existed since plant initial operation. Because this finding is of very low safety significance (Green) and is entered into the TVA corrective action program as PER 206919, this finding is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000259, 260, 296/2009008-01, Violation of Technical Specification 5.4.1 for Failure to Develop Adequate Procedures to Ensure Tornado Depressurization Protection of the Emergency Diesel Generators)

.2.13 Main Steam Line Radiation Monitors

a. Inspection Scope

The team reviewed the TS, UFSAR, and DBDs for the main steam line radiation monitors to identify the component design basis functions and related accident analysis assumptions. The team reviewed the adequacy, reliability, and availability of the power supply to the main steam line radiation monitors. This review included review of vendor manual requirements and breaker specifications. The team also reviewed the calibration settings, calculations and procedures to verify that the main steam line radiation monitors could operate consistent with the design basis assumptions, and system operating procedures. The inspectors reviewed system health, maintenance, and corrective action documents to determine whether the equipment had exhibited adverse performance trends.

b. Findings

No findings of significance were identified.

.2.14 RHR Pump A and C Minimum Flow Shutoff Return Valve to Suppression Pool (MOV

FCV-74-7)

a. Inspection Scope

The inspectors reviewed applicable portions of the TS, UFSAR, DBDs for the RHR system to identify the component design basis function and related accident analysis assumptions. The team conducted a visual inspection of MOV FCV-74-7 to verify that any degraded material conditions were being appropriately addressed. In addition, the team verified that the power demand requirements for the valves were captured in electrical load and degraded voltage calculations. The team also verified that the worst case/highest differential pressure (dP) was used to determine the maximum valve opening and/or closing requirements to ensure that the valve would perform its intended safety-related design basis function. A review was conducted of the licensees testing procedures and results from actual diagnostic valve testing that was performed to verify the MOV was tested in a manner that would detect a malfunctioning valve and verify compliance with GL 89-10 program plan requirements. A review of operating and maintenance procedures was conducted to verify that a malfunctioning MOV could be identified by operators, and that the valves were being properly maintained. The team reviewed historical maintenance work orders and modification packages associated with these valves to verify that they were being properly maintained to ensure reliability and margin for safe operation.

b. Findings

No findings of significance were identified.

.2.15 EECW Throttle Valve to RHR Pump Room Cooler 1C (HOV 67-565)

a. Inspection Scope

The team reviewed the system DBD, related design basis support documentation, drawings, TS, and the UFSAR to identify design, maintenance, and operational requirements for the RHR and core spray (CS) room cooler throttle valves. The team reviewed the cooling specifications, design bases information and supporting calculations to identify system flow requirements. The team reviewed the procedures and results of room cooler and strainer inspections and cleanings, flow balancing and dP testing and trending to verify that degraded conditions were being appropriately addressed. The team reviewed the adequacy of the emergency equipment cooling water (EECW) flow to the room cooler that is controlled by the manually operated flow control valve. Component related PERs, corrective maintenance activities, and system health reports were reviewed to evaluate TVAs capability for detection, monitoring, and correcting potential degradation. A field walkdown was performed with the system engineer to assess observable material conditions and verify that the system configuration was consistent with the design basis assumptions, system operating procedures, and plant drawings.

b. Findings

Introduction.

The team identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to provide adequate procedures for flow balancing of the EECW system. The EECW flow balance procedure was inadequate in that it made no provision in the acceptance criteria to limit or evaluate minimum throttle valve seat/disc clearance, and the subsequent potential for increased flow obstruction, resulting from system flow balancing.

Description.

Preventative Maintenance Test Procedure 3-PMT-BF-067.042,EECW Flow Balance, provides the guidance for performing a flow balance of the EECW system.

This flow balancing procedure was developed in 1995. The EECW system provided the heat sink for station safety-related heat loads including cooling for the RHR and CS room coolers. The flow balance procedure acceptance criteria addressed maintaining minimum flows to the safety-related equipment. The installed strainers on the EECW system are capable of filtering debris greater than 1/8 inch (.125 inches), potentially allowing debris less than 1/8 inch to pass through and clog downstream throttle valves.

A clog in the throttle valves would prevent adequate flow from reaching safety-related heat exchangers unless procedural guidance or limitations prevented throttling valves to disk-to-seat clearances of less then 1/8 inch.

The inspectors evaluated the seat and disc geometry of the installed flow control throttle valves and determined that valves 2-THV-067-0551 and 2THV-067-0594 had disk-to-seat clearances less than the 1/8 inch debris size. These valves on the EECW system provide cooling water to the RHR and CS room coolers respectively. The inspectors concluded that there was a potential for flow obstruction in the throttle valves due to the throttled valve positions resulting in disk-to-seat clearances less than the design size of the EECW strainers. The team noted that the EECW flow balancing procedure contained no provision in the acceptance criteria to limit or evaluate minimum disk-to-seat clearance with respect to content of debris in the system or EECW strainer gap design when changing position of flow control valves. Additionally, there was no reference in the limits and precautions section of the procedure related to the increased potential for flow obstruction due to throttling the flow control valves. The licensee entered this issue into their corrective action program with actions to further evaluate the EECW flow balancing procedure and to correct the throttle positions of the 2-THV-067-0551 and 2THV-067-0594 valves. The inspectors verified and discussed with the licensee existing indications that are available to alert the operator of potential clogging.

Analysis.

The failure to provide adequate procedures for flow balancing of the EECW system is a performance deficiency. This finding is more than minor because it affects the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and operability of the RHR and CS pump room coolers to perform the intended safety function due to potential clogging from debris of the throttle valves that supply cooling water during a design basis event. This affected the cornerstone attribute of Procedure Quality, i.e. Maintenance and Testing (Pre-event) Procedures. The team assessed this finding using the SDP and determined that the finding was of very low safety significance (Green) because the inspectors found no documented history of an actual loss of safety system function. This finding was reviewed for cross-cutting aspects and none were identified.

Enforcement.

10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, that activities affecting quality shall be prescribed by documented procedures include appropriate qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, the EECW system flow balance procedure did not include controls to preclude flow obstruction caused by reduced valve disk-to-seat clearance during valve throttling below 1/8 of and inch. This condition had existed since at least 1995. Because this finding was of very low safety significance and was entered into the TVA corrective action program as PERs 208374 and 208636, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy. (NCV 05000259, 260, 296/2009008-02, Violation of 10CFR50, Appendix B, Criterion V for Inadequate Procedure for Emergency Equipment Cooling Water System Flow Balancing)

.2.16 Plant Control Air Receiver Relief Valve (0-32-556)

a. Inspection Scope

The team reviewed the system DBD, related design basis support documentation, drawings, TS, and the UFSAR to identify design, maintenance, and operational requirements for the plant control air safety relief valve. Maintenance history and associated PERs were reviewed to verify that potential degradation was being monitored and appropriately addressed. Setpoint procedures were reviewed to verify that appropriate design inputs and vendor tolerances were appropriately incorporated into testing acceptance criteria. The team conducted a field walkdown of the relief valve to verify that the installed configuration was consistent with the design basis and plant drawings and to assess observable material conditions to verify that any degraded conditions were being appropriately addressed.

b. Findings

No findings of significance were identified.

.2.17 Drywell Control Air Check Valves (32-2516, 2521, 2163 and 0336)

a. Inspection Scope

The team reviewed the DBD, related design basis support documentation, drawings, TS, and the UFSAR to identify design, maintenance, and operational requirements for selected drywell control air system check valves. Maintenance history, as demonstrated by system health reports, preventive and corrective maintenance, and PERs, were reviewed to verify that any degradation was being appropriately monitored and addressed. The team conducted interviews with the control air (CA) system engineer to obtain additional information and verify that the stations implementation and analysis of industry operating experience related to check valves was appropriate.

b. Findings

No findings of significance were identified.

.2.18 Plant Control Air Check Valves (32-2171, 0243)

a. Inspection Scope

The team reviewed the DBD, related design basis support documentation, drawings, TS, and the UFSAR to identify design, maintenance, and operational requirements for selected drywell control air system check valves. Maintenance history, as demonstrated by system health reports, preventive and corrective maintenance, and PERs, were reviewed to verify that any degradation was being monitored and addressed. The team conducted interviews with the CA system engineer to obtain additional information and verify that the stations implementation and analysis of industry operating experience related to check valves was appropriate. The team also conducted a field walkdown of these check valves to verify that the installed configuration was consistent with the design basis and plant drawings and to assess observable material conditions to verify that any degraded conditions were being appropriately addressed.

b. Findings

No findings of significance were identified.

.3 Review of Low Margin Operator Actions

a. Inspection Scope

The team performed a margin assessment and detailed review of five risk-significant and time-critical operator actions. Where possible, margins were determined by the review of the assumed design basis and UFSAR response times and performance times documented by job performance measures (JPMs). For the selected components and operator actions, the team performed an assessment of the Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs), Annunciator Response Procedures (APPs), and other operations procedures to determine the adequacy of the procedures and availability of equipment required to complete the actions. Operator actions were observed on the plant simulator and during plant walk downs.

The following operator actions were observed on the licensees operator training simulator:

• Actions to align wetwell vent path after loss of suppression pool cooling per EOP procedure 3-EOI-Appendix 13, Emergency Venting Primary Containment

• Actions to align the diesel board for diesel C per 0-AOI-57-1A, Loss of Offsite Power (161 and 500 KV) Station Blackout

• Actions to inhibit an automatic depressurization system (ADS) during an Anticipated Transient Without Scram (ATWS) per 3-EOI-1, Reactor Pressure Vessel Control

• Actions to initiate emergency depressurization (ED) after loss of high pressure coolant injection (HPCI) per EOP procedure 3-EOI-C-1, Alternate Level Control

• Actions to lower and control reactor pressure vessel (RPV) level following an ATWS with isolated RPV per 3-EOI-C-5, Level / Power Control

b. Findings

No findings of significance were identified.

.4 Review of Industry Operating Experience

a. Inspection Scope

The team reviewed selected operating experience issues that had occurred at domestic and foreign nuclear facilities for applicability at the Browns Ferry Nuclear Plant. The issues that received a detailed review by the team included:

• BL 88-04, Potential Safety-Related Pump Loss

• IN 2006-029, Potential Common Cause Failure of Motor-operated Valves as a result of Stem Nut Wear

• RIS-06-023, Post-tornado Operability of Ventilating and Air-conditioning Systems Housed in Emergency Diesel Generator Rooms

• IN 2006-026, Failure of Magnesium Rotors in Motor-operated Valve Actuators

• IN 2000-21, Detached Check Valve Disk not detected by use of Acoustic and Magnetic Non-intrusive Test Techniques

b. Findings

No findings of significance were identified.

.5 Review of Permanent Plant Modifications

a. Inspection Scope

The team reviewed one risk-significant modification noted below to verify that the design bases, licensing bases, and performance capability of the component had not been degraded due to the modification. The adequacy of design and post-modification testing of the modification was reviewed by performing activities identified in IP 71111.17, Evaluations of Changes, Test, or Experiments and Permanent Plant Modifications, Section 02.02.a. Additionally, the team reviewed the modification in accordance with IP 71111.17 to verify that the licensee had appropriately evaluated it for 10 CFR 50.59 applicability.

• DCN 69118, Turbine First Stage Pressure Scram Bypass Setpoint Change

b. Findings

No findings of significance were identified.

4OA5 Other

.1 Power Uprate (IP 71104)

The inspectors reviewed the following modification in accordance with IP 71104, Power Uprate. This is documented in Section 1R21.5 of this report.

• DCN 69118, Turbine First Stage Pressure Scram Bypass Setpoint Change

4OA6 Meetings, Including Exit

Exit Meeting Summary

On November 20, 2009, the team presented the inspection results to Mr. Rusty West, Browns Ferry Nuclear Plant Site Vice President, and other members of the licensee staff.

The team returned all proprietary information examined to the licensee. No proprietary information is documented in the report.

On December 28, 2009, a telephone exit was conducted to disposition one item related to safety-related MCCBs as a URI (Section 1R21.2.7) to Mr. Jim Randich, Browns Ferry General Manager - Site Operations, and other members of the licensee staff.

4OA7 Licensee-Identified Violations (LIV)

The following violation of very low safety significance (Green) was identified by the licensee. This violation of NRC requirements met the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a NCV.

• 10 CFR 50, Appendix B, Criterion III, Design Control, requires that that measures shall be established to assure that applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. During a design basis loss of coolant accident (LOCA) on Unit 1

(2) concurrent with a LOOP on all three units, given a single active failure affecting a Unit 1/2 EDG or 4160V shutdown board, there could be insufficient power for two RHR pumps to provide suppression pool cooling on Unit 2 (1). Suppression pool cooling on the non-LOCA units will be required to maintain suppression pool temperature within acceptable limits during RCIC and/or HPCI operation during a LOOP. Contrary to this, the licensee had inadequate design basis documentation (calculations) to support parallel EDG operations. This was identified in the licensees corrective action program as PER 178142. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green)because of the low initiating event likelihood frequency of LOCA on one unit concurrent with a LOOP on all three units.

ATTACHMENTS:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

D. Binkley, Operator Training

S. Bono, Site Engineering Director

S. Carter, System Engineer, Transformers

J. Davenport, Licensing Engineer

K. Dollar, Electrical Design Engineering

M. Epting, System Engineer, I&C

R. Godwin, Site Licensing and Industry Affairs Manager

G. Harrison, Electrical Design Engineering

K. Harvey, System Engineer, RHR service water

J. Jackson, Civil Design Engineering

E. Kirby, Balance of Plant Engineering

R. Knight, Operator Training

R. Krich, Vice President Nuclear Licensing, TVA Nuclear

J. McCarthy, BFN Director, Safety and Licensing

R. Moxley, Electrical / I&C System Engineer

J. Randich, General Manager - Site Operations

NRC

T. Ross, Senior Resident Inspector, Browns Ferry Nuclear Plant

LIST OF ITEMS

OPENED, CLOSED, AND REVIEWED

Opened

05000259, 260, 296/2009008-01 NCV Violation of Technical Specification 5.4.1 for Failure to Develop Adequate Procedures to Ensure Tornado Depressurization Protection of the Emergency Diesel Generators (Section 1R21.2.12)

05000259, 260, 296/2009008-02 NCV Violation of 10CFR50, Appendix B, Criterion V for Inadequate Procedure for Emergency Equipment Cooling Water System Flow Balancing (Section 1R21.2.15)

05000259, 260, 296/2009008-01 URI Safety-Related Molded Case Circuit Breakers (Section 1R21.2.7)