IR 05000285/2009006

IR 05000285-09-006; August 24-28, 2009, and September 8-18, 2009; Fort Calhoun Station: Baseline Inspection, NRC Inspection Procedure 71111.21, Component Design Basis Inspection

ML093641134
Person / Time
Site:	Fort Calhoun
Issue date:	12/30/2009
From:	Thomas Farnholtz Region 4 Engineering Branch 1
To:	Bannister D Omaha Public Power District
References
FOIA/PA-2014-0210 IR-09-006
Download: ML093641134 (63)

Text

UNITED STATES NUC LE AR RE G UL AT O RY C O M M I S S I O N ber 30, 2009

SUBJECT:

FORT CALHOUN STATION- NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000285/2009006

Dear Mr. Bannister:

On September 17, 2009, the US Nuclear Regulatory Commission (NRC) completed a component design bases inspection at Fort Calhoun Station. The enclosed report documents our inspection findings. The preliminary findings were discussed on September 17, 2009, with M and other members of your staff. After additional in office inspection, a final telephonic exit meeting was conducted on December 28, 2009, with Mr. Reinhart, Site Vice President, and others of your staff.

The inspection 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 team reviewed selected procedures and records, observed activities, and interviewed cognizant plant personnel.

Based on the results of this inspection, the NRC has identified five findings that were evaluated under the risk significance determination process. Violations were associated with five of the findings. Five of the findings were found to have very low safety significance (Green) and the violations associated with these findings are being treated as noncited violations, consistent with Section VI.A.1 of the NRC Enforcement Policy.

If you contest any of the noncited violations, or the significance of the violations you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the US Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 East Lamar Blvd., Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, US Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Fort Calhoun Station. 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

Omaha Public Power District -2-Regional Administrator, Region IV, and the NRC Resident Inspector at Fort Calhoun Station.

The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

In accordance with Code of Federal Regulations, Title 10, Part 2.390 of the NRC's Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS)

component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Thomas Farnholtz, Chief Engineering Branch 1 Division of Reactor Safety Dockets: 50-285 License: DPR-40

Enclosure:

Inspection Report 05000285/2009006 w/Attachments: (1) Supplemental Information (2) Failure Probability of the Alternative Mitigation Strategy

REGION IV==

Docket: 50-285 License: DPR-40 Report Nos.: 05000285/2009006 Licensee: Omaha Public Power District Facility: Fort Calhoun Station Location: 9610 Power Lane Blair, NE 68008 Dates: August 24 - 28, 2009 Onsite September 8 - 18, 2009 Onsite September 18 - December 28, 2009 In-Office Team Leader: G. George, Senior Reactor Inspector Engineering Branch 1 Inspectors: P. Goldberg, Reactor Inspector, Plant Support Branch 2 S. Hedger, Operations Engineer, Operations Branch N. Okonkwo, Reactor Inspector, Engineering Branch 2 Accompanying G. Morris, Beckman and Associates Personnel: W. Sherbin, Beckman and Associates Others: S. Wong, Senior Reactor Analyst, Nuclear Reactor Regulation D. Loveless, Senior Reactor Analyst, Nuclear Reactor Regulation Approved By: Thomas Farnholtz, Branch Chief Engineering Branch 1-1- Enclosure

SUMMARY OF FINDINGS

IR 05000285/2009006; August 24-28, 2009, and September 8-18, 2009; Fort Calhoun Station:

baseline inspection, NRC Inspection Procedure 71111.21, Component Design Basis Inspection.

The report covers an announced inspection by a team of four regional inspectors and two contractors. Five findings were identified. Three of the findings were of very low safety significance (Green). Two of the findings were determined to be Severity Level IV violations.

The final 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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,

Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

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

Criterion III, Design Control. Specifically, from February 1992 to September 8, 2009, the licensee failed to adequately evaluate the seismic qualification of the raw water pumps to ensure that the pumps anchor bolts imbedded in the floor would meet Seismic Class I standards. The team determined that the February 1992 seismic analysis was not conservative for the following reasons:

• The weight distribution of the pump/motor assembly in the analysis did not correctly apply the center of gravity of the pump to the loading analysis.

• The stress analysis of the anchors did not include the weight of the water in the piping.

• The stress analysis did not include the nozzle loads applied to the pump due to the weight of the discharge piping.

The licensee entered the issue into their corrective action program as CR 2009-3977, and performed a preliminary operability evaluation of the support components which determined that the pumps would remain operable following a safe shutdown earthquake. The team reviewed the evaluation, and concurred with the operability evaluation.

The finding is more than minor because it adversely affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance (Green) because it was a design deficiency that did not result in actual loss of safety function. This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance (Section 1R21.2.15).

Green.

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

Criterion V, Instructions, Procedures, and Drawings. Specifically, from August 9, 1973, to September 8, 2009, the licensee failed to prescribe instructions into procedures that would ensure that the plant could be safely shutdown at the probable maximum flood elevation of 1009.3 feet mean sea level. The licensees updated safety analysis report, technical specifications, and station procedures state that protection of the raw water pumps against flooding up to the probable maximum flood height of 1009.3 feet mean sea level is accomplished by sandbag berms and flood gates. During an intake structure walkdown, the team observed two unsealed, 14-inch diameter fire protection piping penetrations in the outer wall, with the bottom of the penetration at elevation 1008.5 feet mean sea level. The penetrations had an air gap of about 1/2-inch between the wall and the pipe. After reviewing station procedures, the team determined that the unsealed penetrations would not be sealed during flooding conditions.

As a result of the teams concern, the licensee entered the issue into their corrective action program as CR 2009-4166 and CR 2009-6195, and verified that there are no other open penetrations in the building walls below the flood level of 1009.3 feet mean sea level. The licensee changed procedure GM-RR-AE-1002 to provide temporary sealing of the penetrations if predicted floods occurred before the permanent seals were installed. The licensee stated that the penetrations will be permanently sealed before the spring 2010 flood season.

This performance deficiency is more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of external events and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. The finding affected the Mitigating Systems Cornerstone because flood protection was degraded. The team determined that the finding resulted in the degradation of equipment and functions specifically designed to mitigate a flooding initiating event. In addition, during a flooding event, the loss would degrade two or more trains of a multi-train safety system. Therefore, the finding was potentially risk significant to flood initiators and a Phase 3 analysis was required. The final change in core damage frequency was calculated to be 8.2 x 10-7 indicating that the finding was of very low safety significance (Green). This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance (Section 1R21.2.15).

Green.

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

Criterion XVI, Corrective Action for failure to take adequate corrective action following the discovery of water intrusion in manholes MH-5 and MH-31 in 1998, 2005, and 2009.

Specifically, from 1998 to September 11, 2009, the licensee failed to take corrective action to establish an appropriate monitoring frequency that would mitigate potential common mode failure of raw water 5kV motor cables in underground ducts and manholes. The licensee entered the condition into the corrective action program as

CR 2009-4216. The corrective action changed the manhole inspection schedule from an 18-month schedule to a quarterly schedule.

The finding is more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of design control for ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance (Green) because it was a design deficiency that did not result in actual loss of safety function. This finding has a crosscutting aspect in the area of human performance, decision making, because the licensee failed to use conservative assumptions in decision-making and adopts a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action.

Specifically, since 2005, the licensee decided to postpone installation of proposed water level control corrective actions and failed to appropriately monitor water intrusion in MH-5 and MH-31 multiple times H.1(b)(Section 1R21.3.4).

• SL-IV. The team identified a Severity Level IV, noncited violation of 10 CFR Part 50,

Appendix B, Criterion XVII, Quality Assurance Records, for failure to maintain original records of the seismic and tornado analysis of the intake structure. Specifically, in 2005, the licensee could not retrieve the original design documentation of the seismic and tornado analysis of the intake structure. This condition was documented in CR 200504345. After the licensee determined the documentation was not retrievable, the licensee reconstituted the seismic and tornado analysis of the intake structure.

These analyses were available during the teams inspection.

This finding is assessed through traditional enforcement because the finding has the potential for impacting the NRCs ability to perform its regulatory function. Using Inspection Manual Chapter 0612, Appendix E, the finding is more than minor because the records were not retrievable. Using Supplement I of the NRC Enforcement Policy, this finding will be treated as a Severity Level IV violation. This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance (Section 4OA5.1).

• SL-IV. The team identified a Severity Level IV, noncited violation for failure to update the final (updated) safety analysis report in accordance with 10 CFR 50.71(e).

Specifically, the licensee failed to update Section 9.8, Raw Water Systems, of the Fort Calhoun Station Updated Safety Analysis Report after constructing a sheet pile alignment wall alongside the intake structure in 1982. Furthermore, this modification removed the slope from the river bottom. Additionally, recent sounding records indicate the river bottom near the intake structure is approximately the same depth as the center of the channel, thus, invalidating the updated safety analysis report statement. The licensee entered this condition into the corrective action program as CR 2009-3927.

The finding is more than minor because the finding is determined to have a material impact on safety. Specifically, with the new sheet pile alignment wall, it could lead to a barge strike that is different than described in the updated safety analysis report. Using Supplement I of the NRC Enforcement Policy, this finding will be treated as a Severity

Level IV violation. This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance (Section 4OA5.1).

Licensee-Identified Violations

.

None.

REPORT DETAILS

REACTOR SAFETY

Inspection of component design bases verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected components and operator actions to perform their design bases functions. As plants age, their design bases may be difficult to determine and important design features may be altered or disabled during modifications. The plant risk assessment model assumes the capability of safety systems and components to perform their intended safety function successfully.

This inspectable area verifies aspects of the Initiating Events, Mitigating Systems and Barrier Integrity Cornerstones for which there are no indicators to measure performance.

1R21 Component Design Bases Inspection

A team of NRC inspectors selected risk-significant components and operator actions for review using information contained in the licensees probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater than two or a Birnbaum value greater than 1E-6.

a. Inspection Scope

To verify that the selected components would function as required, the team reviewed design basis assumptions, calculations, and procedures. In some instances, the team performed calculations to independently verify the licensee's conclusions. The team also verified that the condition of the components was consistent with the design bases and that the tested capabilities met the required criteria.

The team reviewed maintenance work records, corrective action documents, and industry operating experience records to verify that licensee personnel considered degraded conditions and their impact on the components. For the review of operator actions, the team observed operators during simulator scenarios, as well as during simulated actions in the plant.

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions because of modifications, and 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 actions; repeated maintenance; 10 CFR 50.65(a)1 status; operable, but degraded 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.

The inspection procedure requires a review of 20 to 30 total samples, including 10 to 20 risk-significant and low design margin components, 3 to 5 relatively high-risk operator actions, and 4 to 6 operating experience issues. The sample selection for this inspection was 15 components, 5 operator actions, and 4 operating experience items.

.2.1 Voltage Regulator Emergency Diesel Generator (1 & 2)

a. Inspection Scope

The team inspected the electrical portions of the emergency diesel generator, including the voltage regulator, to verify the adequacy of the equipment to respond to design basis events. The team reviewed diesel generator load sequencing logic to verify the appropriate functionality was implemented. The team reviewed completed surveillances to verify that the technical specification requirements were met and to assess the response of the exciter and voltage regulator.

Additionally, the team reviewed protection/coordination and short-circuit calculations to verify the standby diesel generator was adequately protected by protective devices during emergency operation. The team reviewed calculations to verify that: 1) steady-state loading was within design capabilities; 2) adequate voltage would be present to start and operate connected loads; and, 3) operation at maximum allowed frequency would be within the design capabilities. The team reviewed the standby diesel generator load sequence time delays. The team reviewed the standby diesel generator voltage regulator maintenance and control voltage to verify that the components would function when required. Finally, the team reviewed the recorder printout from the latest diesel generator automatic loading test to ensure the voltage returned to nominal between automatic load steps.

b. Findings

No findings of significance were identified.

.2.2 Containment Fan Cooler 7C & 7D

a. Inspection Scope

The team reviewed the design basis loading, motor protective relaying, protective circuit, and device settings controls tied to the emergency diesel generator sequencing of the fan coolers. Additionally, the team reviewed selected calculations for electrical distribution system load flow/voltage drop, short-circuit, and electrical protection and coordination. The adequacy and appropriateness of design assumptions were reviewed to verify that bus capacity was not exceeded and bus voltages remained above minimum acceptable values under design basis conditions. The fan cooler motor protective device settings and breaker ratings were reviewed to ensure that selective coordination was adequate for protection of containment fan cooler and containment spray pump

during worst-case short-circuit conditions. The team also reviewed breaker calibration and preventive maintenance programs associated with the breakers.

Additionally, the team reviewed the Fort Calhoun Updated Safety Analysis Report (USAR), technical specifications, design basis documents, calculations, maintenance records, and modification documents for the water management modification. This modification removed the actuation of the containment spray pumps following a loss of coolant accident, and credited the containment fan coolers and filters for pressure and temperature control and removal of radioactive particulates. The team reviewed the analysis to ensure that the required mission time of the fan coolers was 30 days. The team reviewed the modifications to determine if the previously installed partial strainer was adequate by elimination of the containment spray function for a loss of coolant accident. Additionally, the team reviewed re-analyses of the containment pressure/temperature responses to ensure that the analyses provided the revised design basis for the containment fan cooler requirements.

b. Findings

No findings of significance were identified.

.2.3 125 Vdc Station Batteries 1 & 2

a. Inspection Scope

The team reviewed the 125 volt direct current (Vdc) system drawings, USAR, technical specifications, 125 Vdc system design basis documents, and the system health reports for the 125 Vdc system. The team reviewed the battery sizing calculations to ensure the replacement batteries were adequately sized. Additionally, the team reviewed the assumed correction factors for temperature and aging to establish design margin.

Finally, the team compared the calculated profiles to the results of last battery discharge tests to verify that the established margin was correct.

The team reviewed surveillance procedures for the batteries to ensure correct acceptance criteria, including minimum cell voltage and specific gravity and minimum battery terminal voltage to ensure correct action limits were set for those battery parameters. The team also compared those limits against the technical specifications, procurement specifications, and the manufacturers specifications and instruction manual. Additionally, the team reviewed the 125 Vdc voltage drop calculations to ensure that sufficient voltage would exist at the safety-related loads during a battery design discharge.

In addition, the team performed walkdowns of the battery rooms and 125 Vdc systems.

The team reviewed the battery room heating, ventilation, and air conditioning systems and heating load calculations to ensure the correct parameters were used for battery room temperature and air flow.

b. Findings

No findings of significance were identified.

.2.4 Motor-Driven Auxiliary Feedwater Pump FW6 Breaker 1A3-16

a. Inspection Scope

The team reviewed the 4 kilovolt (kV) switchgear breaker maintenance history and related condition reports to assess the reliability of the power supply for the motor-driven auxiliary feedwater pump. The team reviewed the control circuit for the circuit breakers controlling power to the power circuit to verify that it would operate during design basis events. The switchgears protective device settings and breaker ratings were reviewed to ensure that selective coordination was adequate for protection of connected equipment during worst-case, short-circuit conditions. Additionally, the team reviewed the degraded and loss of voltage relay protection schemes. The 125 Vdc voltage calculations were reviewed to determine if adequate voltage would be available for the breaker open/close coils and spring charging motors. Finally, the team reviewed the sizing of the 5 kV power cables for the auxiliary feedwater pump motors.

b. Findings

No findings of significance were identified.

.2.5 Raw Water Pump Breaker 1A3-9

a. Inspection Scope

The team reviewed the 4 kV switchgear breaker maintenance history and related condition reports to assess the reliability of the power supply for the raw water pumps.

The team reviewed the control circuit for the circuit breakers controlling power to the power circuit to verify that it would operate during design basis events. The switchgears protective device settings and breaker ratings were reviewed to ensure that selective coordination was adequate for protection of connected equipment during worst-case, short-circuit conditions. Additionally, the team reviewed the degraded and loss of voltage relay protection schemes. The 125 Vdc voltage calculations were reviewed to determine if adequate voltage would be available for the breaker open/close coils and spring charging motors. Finally, the team reviewed the sizing of the 5 kV power cables for the raw water pump motors.

b. Findings

No findings of significance were identified.

.2.6 House Service Power Transformer T1A-3 & -4 (161kV)

a. Inspection Scope

The team reviewed the system one-line diagrams, voltage tap settings, nameplate data, protective relay settings, and loading requirements to determine the adequacy of the transformer to supply required power to the associated 4160 volt alternating current (Vac) buses. In addition, the team reviewed the results of recently completed transformer protective relay preventive maintenance.

For the 161 kV switchyard, the team reviewed offsite power connections and the transmission operator notification protocols. To assess the potential vulnerabilities of the transformer to external hazards, the team performed a visual inspection of the 161 kV switchyard to assess the installation, configuration, and material condition.

The team reviewed selected calculations for electrical distribution system load flow/voltage drop, degraded voltage protection, short-circuit analysis, and electrical protection and coordination. This review was conducted to determine the adequacy and appropriateness of design assumptions. In addition, the team verified that transformer capacity was not exceeded and bus voltages remained above minimum acceptable values under design basis conditions. The team reviewed the operating procedures which would address potential minimum voltage at the 161 kV switchyard following a trip of the Fort Calhoun Station generator and resultant fast transfer of the loads from the station auxiliary transformers to the house service power transformers to ensure adequate interface with the offsite power supply organization.

Finally, the team evaluated selected portions of the licensee response to NRC Generic Letter (GL) 2006-02, Grid Reliability and the Impact on Plant Risk and the Operability of Offsite Power, dated February 1, 2006.

b. Findings

No findings of significance were identified.

.2.7 Containment Spray Pump Breaker 1B4B-1

a. Inspection Scope

The team reviewed actuation circuits, interlocks, and emergency diesel generator loading associated with containment spray pumps. The team reviewed control circuits, selected calculations for electrical distribution system load flow/voltage drop, short-circuit analysis, and electrical protection and coordination. The team reviewed the adequacy of design assumptions in calculations to verify that, upon starting and running of the containment spray pumps, bus stability is maintained within design basis conditions and requirements. The pump motor protective device settings and breaker ratings were reviewed to ensure that selective coordination was adequate for protection of the containment spray pump during worst case short-circuit conditions. The team also

reviewed breaker calibration and preventive maintenance programs and logic functional tests associated with the pumps.

b. Findings

No findings of significance were identified.

.2.8 Emergency Diesel Generators 1 & 2 Common Cause Failures

a. Inspection Scope

The team reviewed the USAR, technical specifications, design bases documents, calculations, corrective maintenance, and testing of the emergency diesel generator systems. The team performed several detailed walkdowns of the emergency diesel generator rooms and essential support systems to determine whether design or operational conditions existed that would compromise the performance of both emergency diesel generators. In particular, the team reviewed external flood and seismic evaluations of the fuel oil storage tank and emergency diesel generator day tanks to ensure that the selected equipment could withstand external flooding and seismic loads. Additionally, the team reviewed internal flooding studies to ensure that there was no potential to flood the building and cause common cause failure of the emergency diesel generators.

The team reviewed the fuel oil, lube oil, cooling, and starting air systems of the emergency diesel generators. For the fuel oil system, the team reviewed recent oil sample results and fuel oil consumption to ensure technical specification requirements were met. For the lube oil system, the team verified that sufficient lube oil supplies are onsite to support extended diesel generator runs, if required. The air-cooled diesel generator radiator system design requirements were reviewed to ensure adequate performance of the radiator under worst-case ambient conditions. The diesel generator room ventilation system design was reviewed, including failure positions of air-operated dampers. The team reviewed the emergency diesel generator air start system configuration which included a connection between two of the emergency diesel generator air start accumulators to ensure that any failure in the connecting air lines would not result in loss of air start capability for the emergency diesel generator. The team reviewed inservice testing of the safety-related portion of the starting air system.

b. Findings

No findings of significance were identified.

.2.9 Diesel-Driven Auxiliary Feedwater Pump FW-54

a. Inspection Scope

The diesel-driven auxiliary feedwater pump is independent of all plant support systems and has its own fuel oil storage tank. The team reviewed design documents, including drawings,

calculations, procedures, and the design basis document to determine the design requirements for the diesel-driven auxiliary feedwater pump. Hydraulic analyses were reviewed to verify adequacy of net positive suction head and verify adequacy of surveillance test acceptance criteria for pump minimum discharge pressure at required flow rate.

Maintenance testing results were reviewed to verify acceptance criteria were met and performance degradation would be identified. Fuel oil sample results were reviewed to ensure fuel oil quality is maintained. The team performed a walkdown of the diesel-driven auxiliary feedwater pump area and supporting equipment to determine whether the alignment was in accordance with design basis and procedural requirements, and to assess the material condition of the pump and diesel-engine driver. Preventive and corrective maintenance records were reviewed to ensure the auxiliary feedwater pump and diesel-engine driver were properly maintained. Finally, the team reviewed corrective action documents to ensure problems associated with the diesel-driven feedwater pump were appropriately identified and corrected.

b. Findings

No findings of significance were identified.

.2.10 Motor- Driven Auxiliary Feedwater Pump FW-6

a. Inspection Scope

The team reviewed design documents, including drawings, calculations, procedures, and the design basis document to determine the design requirements for the auxiliary feedwater motor driven pump, FW-6. Hydraulic analyses were reviewed to verify adequacy of net positive suction head and verify adequacy of surveillance test acceptance criteria for pump minimum discharge pressure at required flow rate. Inservice testing results were reviewed to verify acceptance criteria were met and performance degradation would be identified.

Pump actuation logic test results were reviewed to ensure the motor driven auxiliary feedwater pump would start in accidents and events as described in the USAR. In addition, the licensee responses and actions related to IE Bulletin 88-04, Potential Safety-Related Pump Loss, were reviewed to assess implementation of operating experience related to pump minimum flow requirements, and pump-to-pump interaction. The inspectors reviewed condensate storage tank design criteria, including seismic qualification and usable volume calculations, to ensure the motor-driven auxiliary feedwater pump had adequate safety-grade water supply.

The team performed a walkdown of the motor-driven auxiliary feedwater pump area and supporting equipment to determine whether the alignment was in accordance with design basis and procedural requirements, and to assess the material condition of the pump and motor. Preventive and corrective maintenance records were reviewed to ensure the auxiliary feedwater pump was properly maintained. Finally, the team reviewed corrective action documents to ensure problems associated with the pump were appropriately identified and corrected.

b. Findings

No findings of significance were identified.

.2.11 Power Operated Relief Valve PCV 102-1

a. Inspection Scope

The team reviewed design calculations that were performed to determine the lift settings and required stroke timing of the power operated relief valves while in the low temperature overpressure protection mode of operation. Low temperature overpressure protection mode for pressure relief consists of two power operated relief valves with variable lift settings that act as safety relief valves to limit the pressure buildup in the reactor coolant system. While operating in the low temperature overpressure protection mode, procedure restrictions ensure the integrity of the reactor coolant pressure boundary is not compromised by violating the pressure and temperature limits of 10 CFR Part 50, Appendix G, Fracture Toughness. The team also reviewed instrument setpoint and uncertainty calculations for pressurizer level, temperature, and pressure instruments that are relied upon during the low temperature overpressure protection operation. The team reviewed operational mode change procedures to ensure the low temperature overpressure protection controls on pump starts, and pressurizer level were in place when required. This review was done to ensure the assumptions of power operated relief valve stroke time, reactor coolant system pressurization rate due to the operation of safety injection pumps and reactor coolant pumps, and the low temperature overpressure protection procedure restrictions in the analysis were translated into plant operating procedures. The team reviewed power operated relief valve inservice tests to ensure valve opening timing is consistent with requirements from design documents. Corrective action documents were reviewed to ensure problems were identified and corrected in a timely manner.

b. Findings

No findings of significance were identified.

.2.12 Containment Spray Pump 3B

a. Inspection Scope

The team reviewed design documents, including drawings, calculations, procedures, and the design basis document to determine the design requirements for the containment spray pump 3B. Hydraulic analyses were reviewed to verify adequacy of net positive suction head and verify adequacy of surveillance test acceptance criteria for pump minimum discharge pressure at required flow rate. Inservice testing results were reviewed to verify acceptance criteria were met and performance degradation would be identified. Pump actuation logic test results were reviewed to ensure the containment spray pump would start in accidents as described in the USAR. In addition, the licensee responses and actions related to IE Bulletin 88-04, Potential Safety-Related Pump

Loss, were reviewed to assess implementation of operating experience related to pump minimum flow requirements, and pump to pump interaction.

The team performed a walkdown of the containment spray pump area and supporting equipment to determine whether the alignment was in accordance with design basis and procedural requirements. Preventive and corrective maintenance records were reviewed to ensure the pump was properly maintained. Finally, the team reviewed corrective action documents to ensure problems associated with the pump were appropriately identified and corrected.

b. Findings

No findings of significance were identified.

.2.13 Auxilliary Feedwater Steam Admission Valves, AOV 1045 and 1045 A&B

a. Inspection Scope

For steam admission valves AFW-1045 and 1045A/B, the team reviewed the USAR, technical specifications, design basis documents, drawings, and procedures to identify the design basis requirements for the valves. The valve testing procedures and valve specifications were reviewed to verify the design basis requirements, including worst case system and environmental conditions, were incorporated into the test acceptance criteria and equipment design. The adequacy and availability of the backup air source for valve operation was assessed by reviewing equipment history to verify that valves were adequately maintained and that identified equipment problems were resolved.

Recent performance tests were reviewed to assure that the equipment capability was monitored and maintained to meet their design basis function. Additionally, a walkdown was performed to assess the observable material condition and verify the installed configuration was consistent with the design bases and plant drawings.

b. Findings

No findings of significance were identified.

.2.14 Auxilliary Feedwater Containment Isolation Valve, AOV 1107B & 1108B

a. Inspection Scope

For containment isolation valves AFW-1107B and 1108B, the team reviewed the design basis documents, technical specifications, USAR, drawings, and procedures to identify the design basis requirements for the valves. The valve testing procedures and valve specifications were reviewed to verify the design basis requirements, including worst case system and environmental conditions, were incorporated into the test acceptance criteria and equipment design. The adequacy and availability of the backup air source for valve operation was assessed by reviewing equipment history to verify that valves were adequately maintained and that identified equipment problems were resolved.

Recent performance tests were reviewed to assure that the equipment capability was monitored and maintained to meet their design basis function. Additionally, a walkdown was performed to assess the observable material condition and verify the installed configuration was consistent with the design bases and plant drawings.

b. Findings

No findings of significance were identified.

.2.15 Intake Structure

a. Inspection Scope

The safety-related functions of the intake structure is to provide water from the Missouri River that is required for component cooling and fire fighting at Fort Calhoun Station and to provide the structural support and environmental protection necessary to ensure the functional integrity of the raw water pumps and fire protection pumps under all operational and environmental conditions. The team performed several detailed walkdowns of the intake structure and reviewed design and event analyses to ensure that the structure would remain intact during design bases events. The team reviewed the Fort Calhoun Station Individual Plant Examination for External Events (IPEEE) report to determine the probability of external events that the intake structure was able to sustain. The review focused on the potential for damage to the intake structure caused by barge impact, seismic, and external flooding events. The barge impact and seismic events were selected based on previous inspection unresolved items and the flooding event was chosen based on significance of core damage if flooding mitigation features do not work.

b. Findings

b.1 Inadequate Assessment of Seismic Qualification of Raw Water Pumps

Introduction.

The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee did not adequately evaluate the seismic qualification of the anchorage of raw water pumps in the intake structure, to ensure that the pumps anchor bolts imbedded in the floor would meet Seismic Class I standards.

Description.

The team reviewed the Fort Calhoun Station USAR, Appendix F, Section 1.3.b.2 which describes the raw water system as a Seismic Class I system, and all supports, hangers, etc., associated with Class I equipment are also Class I.

During a walkdown of the intake structure, the team observed the raw water pump anchor bolts and seismic bracing of the raw water system piping attached to the pumps.

The team noted that there were no longitudinal seismic restraints in the common raw water discharge header, and noted several disconnected seismic supports in the horizontal piping supports perpendicular to the piping header.

Following the walkdown, the team reviewed the seismic analysis of the raw water pumps contained in Screening Evaluation Worksheet, Motor-Driven Pump AC-10A, performed in February 1992. The seismic analysis for the raw water pumps included the analysis of the pump anchorage bolts embedded in the floor, and the piping nozzle loads applied to the pump anchorages. After reviewing the analysis, the team determined that the analysis was not conservative for the following reasons:

• The weight distribution of the pump/motor assembly in the analysis did not correctly apply the center of gravity of the pump to the loading analysis.

• The stress analysis of the anchors did not include the weight of the water in the piping.

• The stress analysis did not include the nozzle loads applied to the pump due to the weight of the discharge piping.

The team determined that failure of the anchors to remain in place would impact the capability of the safety-related raw water pumps to perform their design function following a seismic event. Failure of the raw water pumps to remain in place would result in a common cause failure of the raw water system to provide a cooling medium for the Seismic Class I component cooling water system.

As a result of the teams concern, the licensee entered the issue into their corrective action program as CR 2009-3977 and performed a preliminary operability evaluation of the support components which determined that the pumps would remain operable following a safe shutdown earthquake. The team reviewed the evaluation, and concurred with the operability evaluation.

Following the inspection, the licensee performed a final seismic evaluation to assess the integrity of the pumps in accordance with the methodology described in the USAR. The final seismic evaluation concluded that the pumps with the seismic brace are acceptable for the required loadings.

Analysis.

The team determined the failure to adequately evaluate the seismic qualification of the raw pumps was a performance deficiency because the evaluation did not meet Seismic Class I requirements. The finding is more than minor because it adversely affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, the finding was of very low safety significance (Green) because it was a design deficiency that did not result in actual loss of safety function. This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance.

Enforcement.

10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in §50.2 and as specified in the license application, for

those structures, systems, and components for which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to this, from February 1992 to August 28, 2009, the licensee did not correctly translate Seismic Class I requirements into seismic evaluations of the raw water pumps to ensure that they would function during a seismic event. Because this violation is of very low safety significance and it was entered into the corrective action program as CR 2009-3977, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000285/2009006-01, Inadequate Assessment of Seismic Qualification of Raw Water Pumps.

b.2 Inadequate Flood Protection for the Intake Structure

Introduction.

The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensee failed to prescribe instructions into procedures that would ensure that the plant could be safely shutdown at the probable maximum flood elevation of 1009.3 feet mean sea level.

Description.

Fort Calhoun Station is licensed and required to adhere to the 70 draft General Design Criteria published for comment in the Federal Register (32 FR 10213)on July 11,1967. General Design Criterion 2, Performance Standards, states, Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect public health and safety or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established shall reflect:

(a) Appropriate consideration for the most severe of these natural phenomena that have been recorded for the site and the surrounding area and

(b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.

The team determined license basis flood levels at Fort Calhoun Station by reviewing USAR Chapter 2.7, Hydrology, USAR Chapter 9.8, Raw Water System, and Technical Specifications 2.16, River Level.

USAR Section 2.7.1.2 states, in part:

The design flood elevation of 1,006 feet based on a 0.1 percent probability flood is considered conservative. Without special provisions, the plant can accommodate flood levels of up to 1,007 [feet mean sea level]. Steel flood gates are permanently mounted above and adjacent to openings in structures containing equipment required for a safe and orderly plant shutdown. In the event of high water levels, these flood gates can be installed to provide protection to a level of 1,009.5 [feet mean sea level]. In the Intake Structure, protection to 1009.5 [feet] msl is accomplished with flood gates and sandbagging. The plant can be protected by sandbags, temporary earth levees and other methods to allow a safe shutdown with a flood elevation of 1,013 [feet mean sea level].

USAR Section 9.8.6 states, in part:

Protection for the raw water pumps and their drives against floods is provided at three elevations as indicated on Figure 9.8-1. The pumps are permanently protected against any water level up to elevation 1007.5 feet by the Class I concrete substructure of the intake building. Protection is provided to elevation 1009.5 feet by sandbags around the traveling screen areas and by gasketed steel closures at exterior doorway openings in the intake structure reinforced concrete perimeter walls. Protection to elevation 1014.5 feet is provided by additional sandbags around the traveling screen areas, and by supplementing the intake structure perimeter walls with sandbags. The water level inside the intake cells can be controlled by positioning the exterior sluice gates to restrict the inflow into the cells.

Technical Specification 2.16 states If the Missouri River level exceeds 1009 feet, the reactor will be placed in a cold shutdown condition using normal operating procedures. When the river level reaches elevation 1004.2 feet and rising, the emergency plan to protect the plant will be instituted.

During an intake structure walkdown in September 2009, the team observed two 14-inch diameter fire protection piping penetrations in the outer wall, with the bottom of the penetration near an elevation of 1008 feet mean sea level. The penetrations were not sealed, and had an air gap of about 1/2-inch between the wall and the pipe. Following the walkdown, the team requested the licensee provide a flooding analysis for the intake structure including an analysis of water entering the structure through the penetrations.

The licensees preliminary evaluation determined that all raw water pumps would be subject to flood water damage in one to four hours with a probable maximum flood level of 1009.3 feet mean sea level.

To determine if the penetrations would be sealed during a flood, the team reviewed Fort Calhoun Station Abnormal Operating Procedure (AOP) 1, Acts of Nature, Revision 23 and procedure GM-RR-AE-1002, Flood Control Preparedness for Sandbagging, Revision 8. AOP-1 directs the facility to use procedure GM-RR-AE-1002 which provides instructions for the installation of sandbag berms and other flood control barriers at the intake structure, auxiliary building, and turbine building when the flood levels reach heights between 1000 and 1014 feet mean sea level. Between flood levels of 1004 and 1007 feet mean sea level, the licensee will begin to install steel flood gates and build sandbag berms at the doors of the plant structures to protect to 1009.3 feet mean sea level. At a flood level of 1009 feet mean sea level, the procedure states that additional sandbags would be draped across flood gates to protect the intake structure to 1014 feet mean sea level as stated in Attachment 9.8 of procedure GM-RR-AE-1002.

The team identified, although flood protection would be installed in and around the intake structure, procedure GM-RR-AE-1002 did not prescribe steps to mitigate external flooding through the open penetrations.

The team determined that failure to ensure appropriate protection of the intake structure at increased flood levels would result in a common cause failure of the safety-related raw water pumps to safely shutdown the plant following a flooding event at 1009 feet mean sea level as stated Technical Specification 2.16, River Level. Additionally, the team determined that the piping penetrations would need to be sealed and flooding protection strategies be revised to ensure the intake structure would be available for safe shutdown at 1009.5 feet mean sea level as stated in the USAR.

As a result of the teams concern, the licensee entered the issue into their corrective action program as CR 2009-4166 and CR 2009-6195, and verified that there are no other open penetrations in the building walls below the flood level of 1009.5 feet mean sea level. The licensee changed procedure GM-RR-AE-1002 to provide temporary sealing of the penetrations if predicted floods occurred before the permanent seals were installed. The licensee stated that the penetrations will be permanently sealed before the spring 2010 flood season.

Analysis.

The team determined the failure to prescribe steps to protect the intake structure and raw water pumps from external flooding up to a flood height of 1009.5 feet mean sea level is a performance deficiency. Specifically, the licensee failed to prescribe steps into procedure GM-RR-AE-1002, Flood Control Preparedness for Sandbagging, that would provide adequate protection to safely shutdown the plant at the flood level of 1009 feet mean sea level as stated in Technical Specification 2.16 and Abnormal Operating Procedure 1, Acts of Nature.

This performance deficiency is more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of external events and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. The team determined that the finding resulted in the degradation of equipment and functions specifically designed to mitigate a flooding initiating event. In addition, during a flooding event, the loss would degrade two or more trains of a multi-train safety system. Therefore, the finding was potentially risk significant to flood initiators and a Phase 3 analysis was required.

The senior reactor analyst completed a Phase 3 analysis using the plant-specific Standardized Plant Analysis Risk Model for Fort Calhoun, Revision 3.45; the Individual Plant Evaluation of External Events (IPEEE); and hand calculations. The exposure period of 1 year represented the maximum exposure time allowable in the significance determination process. The analysis covered the risk affected by the performance deficiency over a range of postulated external floods from 1008.4 to 1009.5 feet mean sea level. The flood hazard was developed using information provided by the U. S. Army Corp of Engineers and was confirmed by the licensees risk analysts.

Baseline failures at these flood elevations included an unrecoverable loss of offsite power, a loss of fuel transfer for Diesel Engine-Driven Auxiliary Feedwater Pump FW-54, and a loss of potable water. The performance deficiency resulted in a common cause failure of all Raw Water System pumps.

Given the postulated sequences and plant procedures, the analyst provided a screening factor of 0.1 for nonrecovery of the raw water pumps via diversion or pumping of the water coming into the intake structure. The analyst also provided a 0.1 nonrecovery screening value to account for alternative makeup sources of water to the emergency feedwater storage tank. Finally, the analyst estimated the failure rate for the minimum alternative mitigation equipment available to the licensee at 3.2 percent as provided in 2 of this inspection report.

The SPAR model provided that the primary accident sequences involve a loss of offsite power, loss of all auxiliary feedwater, and either a loss of containment cooling or a loss of once through cooling. The final change in core damage frequency was calculated to be 8.2 x 10-7 indicating that the finding was of very low safety significance (Green).

Enforcement.

10 CFR Part 50 Criterion V, Instructions, Procedures, and Drawings, states, in part, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Contrary to this, from August 9, 1973, to September 8, 2009, the licensee failed to prescribe documented instructions into procedures that would ensure that the plant could be safely shutdown at an external flood elevation of 1009.3 feet mean sea level.

Specifically, the licensee failed to ensure that station procedures would prescribe steps to mitigate external flooding through open wall penetrations in the intake structure, up to 1009.5 feet mean sea level. Because this violation was of very low safety significance and it was entered into the corrective action program as CR 2009-4166 and CR 2009-6195, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000285/2009006-02, Inadequate Flood Protection for the Intake Structure.

b.3 Failure to Update Flood Protection for Safety-Related Buildings

Introduction.

The team identified an unresolved item concerning external flood protection for plant areas considered vital to allow the reactor to achieve cold shutdown.

Specifically, the issue concerns:

(1) the ability of the licensee to protect the Fort Calhoun Station auxiliary building, intake structure, and turbine building basement from external floods up to flooding elevation 1013 feet mean sea level as stated in USAR and station procedures; and,

(2) upon receiving new flooding information in November 2003, the licensee was required to update the USAR.

Description.

The team determined flood elevations at Fort Calhoun Station by reviewing USAR Chapter 2.7, Hydrology, USAR Chapter 9.8, Raw Water System, and Technical Specifications 2.16, River Level.

USAR Section 2.7.1.2 states, in part:

The design flood elevation of 1,006 feet based on a 0.1 percent probability flood is considered conservative. Without special provisions, the plant can accommodate flood levels of up to 1,007 [feet mean sea level]. Steel flood gates are permanently

mounted above and adjacent to openings in structures containing equipment required for a safe and orderly plant shutdown. In the event of high water levels, these flood gates can be installed to provide protection to a level of 1,009.5 [feet mean sea level]. In the Intake Structure, protection to 1009.5 [feet] msl is accomplished with flood gates and sandbagging. The plant can be protected by sandbags, temporary earth levees and other methods to allow a safe shutdown with a flood elevation of 1,013 [feet mean sea level].

USAR Section 9.8.6 states, in part:

Protection for the raw water pumps and their drives against floods is provided at three elevations as indicated on Figure 9.8-1. The pumps are permanently protected against any water level up to elevation 1007.5 feet by the Class I concrete substructure of the intake building. Protection is provided to elevation 1009.5 feet by sandbags around the traveling screen areas and by gasketed steel closures at exterior doorway openings in the intake structure reinforced concrete perimeter walls. Protection to elevation 1014.5 feet is provided by additional sandbags around the traveling screen areas, and by supplementing the intake structure perimeter walls with sandbags. The water level inside the intake cells can be controlled by positioning the exterior sluice gates to restrict the inflow into the cells.

Technical Specification 2.16, Basis, states:

The maximum Missouri River level of 1009 feet is the level at which the installed flood gates will protect the plant. Any increase in river level will require sand bagging to repel the water to a maximum flood level of 1014 feet or greater.

When the licensee determines it is necessary to protect the plant at flood levels greater than 1009.3 feet, the licensee implements procedure GM-RR-AE-1002, Flood Control Preparedness for Sandbagging. Step 7.4 of the procedure states:

The primary focus for flood protection should be directed to those facilities which are considered vital with respect to nuclear safety and credited with flood protection in the IPEEE Flooding evaluation, Reference 2.3. These facilities shall be protected at the sacrifice of the other facilities if site conditions warrant. The vital facilities are:

Auxiliary Building, Intake Structure, and Turbine Building Basement.

Reference 2.3 of procedure GM-RR-AE-1002 is the IPEEE for Fort Calhoun Station Section 5.2, External Flooding, Table 5.2.3, which credits flood protection up to 1013.5 feet mean sea level by sandbagging for the switchyard, auxiliary building, intake structure, and turbine building. Table 5.2.3, Impact of Periodic Flood due to Rain and Snow, comments that severe core damage results if either intake or auxiliary building sandbagging fails.

9.5 of procedure GM-RR-AE-1002 gives specific instructions that plant operators would use to protect from flood crest above 1009 feet mean sea level. The attachment notes that sandbags would be tied and draped over the top of floodgates to

supplement the protection capability to the projected flood crest. Specifically, the attachment states, Place additional sandbags on top of the floodgates to raise the protection against the expected crest of the flood.

During the inspection, the team discussed, with the licensee, how protection of vital facilities against floods would occur at flood levels above the probable maximum flood level of 1009 feet mean sea level. As a result of this discussion, the team determined that stacking and draping sandbags at a height of four feet over the top of floodgates would be insufficient to protect the vital facilities. The sandbags would be insufficient because the cross section on the top of the floodgates is too small to support a stacked sandbag configuration that would retain four feet static head of water.

The team determined that the failure to protect the auxiliary building, intake structure, and turbine building to an external flood height of 1013 feet mean sea level is a performance deficiency. Specifically, the licensee failed to meet the self imposed standards of USAR Section 2.7.1.2 and USAR Section 9.8.6. In addition, the licensee failed to meet the self imposed standard of Step 7.4 in procedure GM-RR-AE-1002, Flood Control Preparedness for Sandbagging.

The licensee disagreed with the teams determination that this is a performance deficiency because Fort Calhoun Station is designed to protect to the design basis probable maximum flood height of 1009.3 feet mean sea level. The licensee believes that any flooding above the probable maximum flood level is incredible because it would involve a dam break upstream of the site and dam failure was considered incredible based on Army Corps of Engineers letter to Omaha Public Power District, dated December 12, 1967. However, the team believes, because the licensee stated flood protection would occur up to 1013 feet mean sea level in the USAR and station procedures, the licensee self-imposed a standard to provide flood protection up to 1013 feet mean sea level. Additionally, in November 2003, the licensee received new information from the Army Corps of Engineers that determined flood levels could be potentially higher than what was evaluated in 1967.

Because more inspection is necessary to resolve this issue, the issue is considered an unresolved item pending further NRC Region IV review. The NRC Region IV review will determine:

1. If the failure to meet the self imposed standard of flood protection up to 1013 feet is a performance deficiency in accordance with NRC Manual Chapter 0612.

2. If a violation of NRC requirements is associated with the performance deficiency because the licensee did not update the flooding design basis when new information was received in November 2003.

This unresolved item is identified as URI 05000285/2009006-03, Failure to Update Flood Protection for Safety Related Buildings.

.3.1 Inspection of Generic Letter 1988-14, Instrument Air Supply System Problems Affecting

Safety-Related Equipment

a. Inspection Scope

The team reviewed Generic Letter 1988-14 associated with instrument air problems causing adverse effects on safety-related equipment. The review was performed to ensure that the licensees actions were adequate to ensure that the safety-related equipment performed as required. The team reviewed the USAR, technical specifications, design basis documents, calculations, modification documents, maintenance documents, and surveillance testing to determine if the licensee had completed all of the items listed in their commitments to Generic Letter 88-14.

Additionally, the team reviewed preventive maintenance documents to ensure that the licensee had included desiccant replacement and changing air filters in the instrument air system. In addition, the team reviewed an abnormal operating procedure and training on the procedure to ensure that the procedure met all operator requirements in the event of a loss of instrument air.

b. Findings

No findings of significance were identified.

.3.2 Inspection of Generic Letter 1988-17, Loss of Decay Heat Removal

a. Inspection Scope

The team reviewed the Generic Letter 1988-17, which involved actions taken to address identified deficiencies in procedures, hardware, and training related to loss of decay heat removal when the reactor is not at power. In the letter, the NRC requested status of the licensees implementation of eight recommended expeditious actions. To verify that the licensee adhered to these actions, the team reviewed licensee response letters, NRC evaluation reports in response to the letters, NRC inspection reports, related operations procedures, instrument calibration procedures, and instrumentation vendor information on calibration practices. Additionally, the team reviewed the schedule and completion of six programmed enhancement recommendations.

b. Findings

No findings of significance were identified.

.3.3 Inspection of Information Notice 2005-30, Safe Shutdown Potentially Challenges by

Unanalyzed Internal Flooding Events and Inadequate Design

a. Inspection Scope

The team reviewed the licensees disposition of Information Notice 2005-30. This information notice discussed recent industry events where it was discovered that safe shutdown was potentially challenged by unanalyzed flooding events and inadequate design of safety-related systems. Specifically, the physical arrangement of safety-related systems essential to achieve safe shutdown made these systems vulnerable to flooding originating from failures of nonsafety-related systems located in the turbine building. The team reviewed the disposition of the Information Notice as documented by licensee in OE-2007-4005.

b. Findings

No findings of significance were identified.

.3.4 Inspection of Generic Letter 2007-01, Inaccessible or Underground Power Cable

Failures that Disable Accident Mitigation Systems of Cause Plant Transients

a. Inspection Scope

The team reviewed Information Notice 2002-12 and Generic Letter 2007-01, which documented the concern of inaccessible power cables. In addition, the team performed a walkdown of portions of the safety-related underground duct banks that contained the 5 kV cable for the raw water pump motors to assess the installation configuration, material condition, and potential vulnerability to hazards.

b. Findings

b.1 Inadequate Corrective Actions to Ensure the Reliability of the Raw Water Pump Power Cables

Introduction.

The team identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to take effective corrective action following the discovery of water intrusion in manholes MH-5 and MH-31 in 1998, 2005 and 2009. Specifically, the licensee failed to take corrective action to establish an appropriate monitoring frequency that would mitigate potential common mode failure of raw water 5kV motor cables in underground ducts and manholes.

Description.

The motors driving the raw water pumps, AC-10A through AC-10D, are powered through 5 kV cables routed through duct banks and manholes (MH-5 & MH-31)between the plant building and the intake structure.

USAR Section 8.5, Cable Installation, Section 8.5.1.f states:

The E prefixed cables inside the screenhouse [intake structure] and between the plant building and screenhouse are routed in separate conduits, tray sections, or in separate duct bank conduits.

There is one manhole between the pull boxes and the screen house. The cables are in cable trays and the routing is in conformance with the Cable and Conduit Schedule Notes (Figure 8.5-1). There is a 6 thick concrete wall separating cable trays with EA and EC cables from cable trays holding EB and ED cables.

On numerous occasions, between 1998 and 2009, the licensee discovered water in manholes MH-5 and MH-31. In response to the 1998 event, the licensee established a 5-year inspection schedule for MH-31, as documented in CR 199801719. The 5-year inspection schedule did not include MH-5 in the corrective action extent of condition.

In 2002, the licensee addressed Information Notice 2002-12, Submerged Safety-Related Electrical Cables, documented in CR 200200707, justifying the acceptability of lifetime submergence of the safety-related raw water 5 kV cables based on the cable manufacturers design basis accident qualification report. As a corrective action to address Information Notice 2002-12, the licensee established a 5-year inspection schedule for MH-5.

In 2005, the licensee discovered a sufficient water level in MH-5 to completely submerge all but the top two cable trays, as documented in CR 200503247. Following the discovery of water in 2005, the licensee established an 18-month inspection schedule for MH-5. As a corrective action, engineering suggested at that time adding either a level alarm or a sump pump to MH-5.

During August 2009, MH-5 was opened on three separate occasions and water was pumped from the vault each time. The inspectors determined, at the rate of water intrusion during these occasions, the amount of water would submerge the 5 kV cable trays in an 18-month inspection schedule. At the time of this inspection neither an alarm nor a sump pump had been approved for installation to ensure the safety-related cables in MH-5 would stay above the water level.

During the inspection, the team identified that the environmental qualification report used to address Information Notice 2002-12, that justified the submergence of safety-related cables, was not conservative because the cable ratings were based on a 30-day containment spray, not an uncontrolled lifetime of submergence. Additionally, the team identified that the licensee had initially failed to include MH-5 in the corrective action for the flooding found in MH-31 in 1998; subsequently failing to determine if the 5-year inspection cycle was appropriate for MH-5 in 2002. Since 2005, the licensee postponed installation of proposed level alarm or sump pump in MH-5 multiple times. The postponement was based on fire protection requirements and budget constraints.

Based on the teams finding, the licensee entered the issue into the corrective action program as CR 2009-4216. The corrective action changed the manhole inspection schedule from an 18-month schedule to a quarterly schedule.

Analysis.

The team determined that failure to take effective corrective action to ensure the reliability and capability of the safety-related cables powering the raw water pump motors is a performance deficiency. Furthermore, the finding was within the licensee's ability to foresee and correct because the licensee had multiple opportunities to correct the continuing challenge to the safety-related cables and raceway system for the raw water system. The finding is more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of design control for ensuring the availability, reliability, and capability of systems that respond to Initiating Events to prevent undesirable consequences. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, the finding was of very low safety significance (Green) because it was a design deficiency that did not result in actual loss of safety function.

This finding has a crosscutting aspect in the area of human performance, decision making, because the licensee failed to use conservative assumptions in decision making and adopts a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action. Specifically, since 2005, the licensee decided to postpone installation of proposed water level control corrective actions and failed to appropriately monitor water intrusion in MH-5 and MH-31 multiple times H.1(b).

Enforcement.

10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. Contrary to this, from 1998 to September 11, 2009, the licensee failed to establish appropriate monitoring frequency to correct or mitigate water intrusion in cable vaults containing safety-related cables. Because this violation was of very low safety significance and it was entered into the corrective action program as CR 2009-4216, this violation is being treated as a noncited violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy:

NCV 05000285/2009006-04, Inadequate Corrective Actions to Ensure the Reliability of the Raw Water Pump Power Cables.

.4 Results of Reviews for Operator Actions:

The team selected risk-significant components and operator actions for review using information contained in the licensees probabilistic risk assessment. This included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum value greater than 1E-6.

a. Inspection Scope

For the review of operator actions, the team observed operators during simulator scenarios associated with the selected components as well as observing simulated actions in the plant.

Inspection procedure 71111.21 requires a review of 3 to 5 relatively high-risk operator actions. The sample selection for this inspection was 5 operator actions.

The selected operator actions were:

• A bounding case major steam line break with feedwater isolation and containment air cooler failures (Scenario)

• A loss of offsite power with a fault of 125 Vdc Bus 2, the motor-driven auxiliary feedwater pump out of service, and a diesel drive auxiliary feedwater pump failure (Scenario)

• Establish emergency boration from outside the control room (Job Performance Measure)

• Cross-tie power supplies to open a power-operated valve during a loss of offsite power with an emergency diesel generator failure (Job Performance Measure)

• Align raw water system to shutdown cooling heat exchanger during a loss of coolant accident with a circulating cooling water system failure (Job Performance Measure)

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

The team reviewed a sample of problems that the licensee had identified previously and entered into the corrective action program. The team reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions. In addition, condition reports written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action system. The specific documents that were sampled and reviewed by the team are listed in the attachment.

4OA5 Other

.1 URI 05000285/2005011-05: Intake Structure Design

a. Inspection Scope

The team reviewed actions that the licensee completed to resolve concerns associated with the intake structure unresolved item which was reported in NRC Inspection Report 05000285/2005011. The licensees action was to provide original intake structure design documentation for the seismic, tornado, and barge impact analysis.

The licensee could not provide original design documentation, so licensee staff embarked on a reconstitution effort. The licensee has provided new calculations for the seismic design and tornado analysis; however, the barge impact was not completed.

To resolve this issue, the NRC inspectors will need to review the future barge impact analysis. From discussion, the licensee has agreed to complete the barge impact analysis by June 30, 2010. The unresolved item will remain open pending NRC review of this analysis.

b. Findings

b.1 Failure to Maintain Quality Records for the Intake Structure Design

Introduction.

The team identified a Severity Level IV, noncited violation of Criterion XVII, Quality Assurance Records, for failure to maintain original records as required by the Atomic Energy Commissions 70 draft General Design Criteria, July 1967. Specifically, the licensee failed to maintain the original records of the seismic and tornado analysis of the intake structure.

Description.

Fort Calhoun Station is licensed and required to adhere to the 70 draft General Design Criteria published for comment in the Federal Register (32 FR 10213)on July 11,1967. Criterion 5, Records Requirement, states, Records of the design, fabrication, and construction of essential components of the plant shall be maintained by the reactor operator or under its control throughout the life of the reactor.

During an NRC inspection in 2005 (NRC Inspection Report 05000285/2005011), NRC inspectors requested documentation of the seismic analysis for the Fort Calhoun Station intake structure. At the time of the inspection, the information was not readily available.

Since the information was not available, the licensee entered the concern into the corrective action program, as CR 200504345, to begin searching for the associated documents. The investigation concluded that the records were not retrievable.

After the licensee determined the documentation was not retrievable, the licensee reconstituted the seismic and tornado analysis of the intake structure. These analyses were available during the teams inspection.

Analysis.

The team determined that the failure to meet the 70 draft General Design Criteria is a performance deficiency. This finding is assessed through traditional

enforcement because the finding has the potential for impacting the NRCs ability to perform its regulatory function. Using Inspection Manual Chapter 0612, Appendix E, the finding is more than minor because the records were irretrievable. Using Supplement I of the NRC Enforcement Policy, this finding will be treated as a Severity Level IV violation. This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance.

Enforcement.

10 CFR Part 50, Appendix B, Criterion XVII, Quality Assurance Records, states, in part, Sufficient records shall be maintained to furnish evidence of activities affecting quality. Additionally, Records shall be identifiable and retrievable. Contrary to this, prior to 2005, the licensee failed to maintain the original records for the seismic and tornado analysis for the intake structure. Because this violation is of very low safety significance and was entered into the licensees corrective action program as CR 2009-3769, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000285/2009006-05, Failure to Maintain Quality Records of the Intake Structure Design.

b.2 Failure to Update Intake Structure Design in the FSAR

Introduction.

The team identified a Severity Level IV, noncited violation for failure to update the final (updated) safety analysis report in accordance with 10 CFR 50.71(e).

Specifically, the licensee failed to update Section 9.8, Raw Water Systems, of the Fort Calhoun Station USAR after constructing a sheet pile alignment wall alongside the intake structure in 1982.

Description.

USAR Section 9.8, Raw Water Systems, states, The intake structure is a massive concrete building set just back of the harbor line of the river. Furthermore, The river bottom slopes downward from the bank to the thread of the channel, thus keeping boats and barges away from the actual harbor line. Any blow that could be struck by such a vessel would be a glancing one at worst on the armored wall noses and any damage to the structure itself is considered unlikely.

In 1982, the licensee installed a sheet pile alignment wall on the upstream side of the intake structure, as documented in Modification Request FC-81-106. This modification removed the slope from the river bottom. Additionally, recent sounding records indicate the river bottom near the intake structure is approximately the same depth as the center of the channel, thus, invalidating the USAR statement.

Analysis.

The finding is a performance deficiency because the licensee failed to update USAR Section 9.8, as required by 10 CFR 50.71(e). The finding is assessed through traditional enforcement because the finding has the potential for impacting the NRCs ability to perform its regulatory function. The finding is more than minor because the finding is determined to have a material impact on safety. Specifically, it is not certain that a river vessel will strike the intake structure. Using Supplement I of the NRC Enforcement Policy, this finding will be treated as a Severity Level IV violation. This finding was not assigned a crosscutting aspect because the underlying cause was not indicative of current performance.

Enforcement.

10 CFR 50.71, Maintenance of records, making of reports, paragraph

(e) states, Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. Contrary to this, since September 30, 1982, the licensee did not update USAR Section 9.8. Because this violation is of very low safety significance and was entered into the licensees corrective action program as CR 2009-3927, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000285/2009006-06, Failure to Update Intake Structure Design.

4OA6 Meetings, Including Exit

On September 17, 2009, the team leader presented the preliminary inspection results to Mr. D. Bannister, Vice President and Chief Nuclear Officer, Fort Calhoun Station, and other members of the licensees staff.

On December 28, 2009, the team leader conducted a telephonic final exit meeting with Mr. J. Reinhart, Site Vice President, and other members of the licensees staff. The licensee acknowledged the findings during each meeting. While some proprietary information was reviewed during this inspection, no proprietary information was included in this report.

4OA7 Licensee Identified Violations

None.

s: Supplemental Information

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

J. Reinhart, Site Vice President

D. Bannister, Vice President, Nuclear Operations

T. Nellenbach, Plant Manager

D. Trausch, Assistant Plant Manager

R. Clemens, Manager, Nuclear Engineering

T. Mathews, Manager, Nuclear Licensing

J. Herman, Manager, Engineering Programs

M. Frans, Manager, System Engineering

J. Gasper, Manager, Design Engineering

K. Wells, Nuclear Design Engineer - Electrical/I&C

M. Anielak, Manager, Shift Operations

R. Westcott, Manager, Quality

R. Harig, Manager, Work Management

T. Miller, Superintendent, Mechanical

S. Baughn, Supervisor, Reactor Performance Analysis

D. Guinn, Supervisor, Regulatory Compliance

B. Blessie Supervisor, Operations Engineering

G. Cavanaugh, Supervisor, Corrective Actions

R. Mueller, Supervisor, Electrical/I&C Design Engineering Nuclear

S. Miller, Supervisor, System Engineering - I&C/Electrical

J. Adams, Principal Engineer, Electrical

S. Kalra, Senior System Engineer

M. Elzway, Senior Nuclear Design Engineer - Electrical

D. Gage, Senior Instructional Technician

A. Chladil, Senior Operations Engineer

D. Rollins, Reliability Engineer

K. Dworak, Working Crew Leader - Electrical Maintenance

T. Bottum, System Engineer

E. Linzer, Licensed Operator

J. Carlson, Design Engineer

A. Filips, Design Engineer

A. Koenig, System Engineer

E. Matzke, Regulatory Compliance

J. Bock, Maintenance Support Clerk, Work Management

K. Hyde, Design Engineer

P. Cronin, Operations

K. Landis, Member, SARC

Attachment 1

NRC Personnel

D. Loveless, Senior Reactor Analyst

J. Kirkland, Senior Resident Inspector

J. Wingebach, Resident Inspector

K. Manoly, Nuclear Reactor Regulation

R. Lantz, Chief, Operations Branch

W. Schaup, Acting Resident Inspector

L. Wilkins, Project Manager

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

Failure to Update Flood Protection for Safety

05000285/2009006-03 URI Related Buildings (Section 1R21.2.15)

Opened and Closed

Inadequate Assessment of Seismic Qualification of

05000285/2009006-01 NCV Raw Water Pumps (Section 1R21.2.15)

Inadequate Flood Protection for the Intake Structure

05000285/2009006-02 NCV (Section 1R21.2.15)

Inadequate Corrective Actions to Ensure the

05000285/2009006-04 NCV Reliability of the Raw Water Pump Power Cables (Section 1R21.3.4)

Failure to Maintain Quality Records of the Intake

05000285/2009006-05 NCV Structure Design (Section 4OA5.1)

Failure to Update Intake Structure Design

05000285/2009006-06 NCV (Section 4OA5.1)

Discussed

05000285/2005011-05 URI Intake Structure Design (Section 4OA5.1)

Attachment 1