IR 05000285/2014008

IR 05000285-14-008; on 05/12/2014 - 05/16/2014; Fort Calhoun Station, Post-Approval Site Inspection for License Renewal

ML14164A638
Person / Time
Site:	Fort Calhoun
Issue date:	06/13/2014
From:	John Dixon NRC/RGN-IV/DRS/EB-2
To:	Cortopassi L Omaha Public Power District
G. Pick
References
IR-14-008
Download: ML14164A638 (16)

Text

une 13, 2014

SUBJECT:

FORT CALHOUN STATION - NRC POST-APPROVAL LICENSE RENEWAL INSPECTION REPORT 05000285/2014008

Dear Mr. Cortopassi:

On May 16, 2014, U.S. Nuclear Regulatory Commission (NRC) inspectors completed a Post-Approval Site Inspection for License Renewal at your Fort Calhoun Station. The enclosed report documents the inspection findings, which were discussed with Mr. Scot Swanson, Acting Plant Manager, and other members of your staff.

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

Based upon the results of this inspection, no findings of significance were identified.

In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRCs Agencywide Documents Access and Management 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/

John L. Dixon, Jr. Acting Chief Engineering Branch 2 Division of Reactor Safety Docket: 50-285 License: DPR-40 cc w/enclosure:

Electronic Distribution to Fort Calhoun Station

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 05000285 License: DPR-40 Report: 05000285/2014008 Applicant: Omaha Public Power District Facility: Fort Calhoun Station Location: P.O. Box 310 Fort Calhoun, NE 68023 Dates: May 12 - 16, 2014 Inspector: G. Pick, Senior Reactor Inspector Approved By: John L. Dixon, Jr, Acting Chief Engineering Branch 2 Division of Reactor Safety-1- Enclosure

SUMMARY

IR 05000285/2014008; 05/12 - 16/2014; Fort Calhoun Station, Post-Approval Site Inspection for

License Renewal The report covers an inspection conducted by a regional inspector in accordance with the Nuclear Regulatory Commission (NRC) Manual Chapter 2515 and the NRC Inspection Procedure 71003.

The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red),

which is determined using Inspection Manual Chapter 0609, Significance Determination Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310,

Components Within the Cross-Cutting Areas. Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1646, Reactor Oversight Process.

NRC-Identified Findings and Self-Revealing Findings

None

Licensee-Identified Violations

None

REPORT DETAILS

OTHER ACTIVITIES (OA)

Phase 3 Inspection Activities The Phase 3 Inspection activities are performed after the licensee enters the period of extended operation. The period of extended operation is the additional 20 years beyond the original 40-year licensed term. Fort Calhoun Station began their period of extended operation after midnight on August 9, 2013.

The inspector performed this inspection to evaluate whether the licensee completed outstanding actions required to comply with the license renewal license condition and commitments and effectively implemented outstanding actions related to select aging management programs.

The inspector closed Commitments 23, 24, 34, and 39 during this inspection.

4OA2 Problem Identification and Resolution (71003 and 71152)

a. Inspection Scope

During this inspection, the inspector reviewed the licensees actions related to the condition reports listed in the attachment, which were identified during the Phase 2 license renewal inspection (Inspection Report 05000285/2013009, accessible from ADAMS as ML13206A346). The inspector evaluated whether the licensee had taken appropriate short term corrective actions and verified that the licensee developed corrective actions commensurate with the significance of the identified problems.

b. Observations Following the initial screening, the inspector selected 14 condition reports for detailed followup and evaluation of the implementation of corrective actions. From this review the inspector had the following observations:

• The inspector determined that Procedure SO-M-100, Conduct of Maintenance, Revision 57b, required revising to specifically discuss destructively testing components subject to selective leaching. Procedure steps refer to contacting the buried pipe program engineer; however, it was dependent upon the knowledge and awareness of the program engineer to recognize when the personnel removed elbows and valves from the fire protection piping that had the potential for selective leaching requiring destructive testing. The licensee initiated Requested Action 2014-0963 to revise Procedure SO-M-100 to ensure that the maintenance process would have maintenance personnel contact engineering when removing fire protection fittings to ensure destructive testing could be performed to look for selective leaching.

• While reviewing the corrective actions for Violation 05000285/2011006-02 (discussed below), the inspector determined that, although the licensee had installed the required level monitoring system, the licensee had not initiated all the administrative controls necessary for the system. Specifically, the licensee had not identified an owner for the monitoring system; had not included this requirement in the license renewal Program Basis Document 30, Cables and Connections Program, Revision 2; had not developed training for operation of the monitoring system; and had not established any administrative monitoring limits for rising levels in the manholes. The inspector determined that the licensee continued to rely on weekly visual inspections to determine the water level in Manholes 5 and 31 rather than fully utilize the level monitoring system.

c. Findings

No findings were identified.

4OA5 Other Activities Post-Approval Site Inspection for License Renewal

.1 (Closed) Violation 05000285/2011006-02: Inadequate Corrective Actions to Ensure the

Reliability of the Raw Water Pump Power Cables This violation documented the inability to ensure that water intrusion into Manholes 5 and 31, which contained the power cables for the raw water pumps, would not cover the power cables. This violation also tracked resolution of the system installed to ensure that the cables remained dry as specified in the license renewal commitments.

As documented in Inspection Report 05000285/2013009 the monitoring system had insufficient operating history and personnel were not familiar with the operation of the system. Similarly, as documented in Inspection Report 05000285/2013019 (ML14042A238), NRC reviewed the interim actions to monitor for water intrusion into Manholes 5 and 31 and found them acceptable but did not close the violation.

During this inspection, the inspector verified that the licensee continued to implement weekly preventive maintenance inspections of the manholes and to pump out the manholes when water accumulated above 4 inches. After comparing the preventive maintenance activities that removed water from Manholes 5 and 31 with water level graphs generated from the level monitoring system, the inspector determined that good correlation existed that demonstrated the level monitoring system accurately monitored level. The inspector determined that the level monitoring system identified increasing water levels and would provide sufficient warning to keep the inaccessible medium voltage cables dry.

While reviewing implementation of the level monitoring system, the inspector identified several additional concerns with the implementation of the level monitoring program.

Specifically, the licensee had not:

(1) updated Program Basis Document 30 to describe the actions being taken to keep the inaccessible medium voltage cables dry;

(2) established a program owner for the system;

(3) developed/conducted training for operators related to system operation; and

(4) established procedures for verifying proper system operation, responding to administrative limits, and operation of the system. The licensee initiated Condition Report 2014-05955 documenting the need to correct these deficiencies.

The licensee expected to have their corrective actions completed prior to the end of 2014 and indicated they would notify the regional office when they considered the system fully implemented. Further, the inspector did not identify a safety concern since the licensee had continued to implement weekly preventive maintenance activities to monitor for water in the manholes.

The licensee implemented the required actions to install a level monitoring system that will alarm when water is accumulating that could affect the safety-related cables. The inspector closed this violation.

.2 (Closed) Unresolved Item 05000285/2013009-01: Evaluation of Environmentally-

Assisted Fatigue for Charging Line Nozzle Commitment 23 specified: Add the following to the scope of components subject to the FCS Fatigue Monitoring Program:

• Pressurizer Surge Line bounding locations, and elbow;

• Class 2 and 3 components not included in the NUREG-1801 program which are subject to fatigue as an aging effect requiring management; and

• The number of cycles assumed for the evaluation of the charging line nozzle will be included in the Fatigue Monitoring Program Basis Document, when it is generated, to assure that a CUF of 1.0 is not exceeded.

The licensee had not completed actions related to evaluating environmentally-assisted fatigue of the charging line nozzle at the time of the Phase 2 inspection. A preliminary environmentally-assisted fatigue evaluation determined that the charging line nozzle could exceed the design limit prior to the end of the period of extended operation. The licensee initiated Condition Report 2011-10000 to track this commitment. The licensee expected to complete the fatigue analysis for the charging line nozzle prior to entering the period of extended operation.

During this inspection, the inspector reviewed the fatigue analysis for the charging line nozzle. The licensee completed Calculation CN-PAFM-10-15, Fort Calhoun: Transfer Function Database Development for CVCS Charging Nozzles, Revision 1, on August 6, 2013. The calculation described the method used to monitor fatigue and environmentally-assisted fatigue using a finite element model, which included the critical point on the charging inlet nozzle to the reactor coolant system cold leg. The licensee modeled the charging inlet nozzle using the design configuration contained in Calculation CN-MRCDA-10-56, Elastic-Plastic Fatigue Evaluation of the Fort Calhoun, Unit 1 Charging Inlet Nozzle for Extended Power Uprate, Revision 4. The inspector identified no concerns with the environmentally-assisted fatigue analysis actions taken by the licensee.

Based upon review of the actions taken to monitor for fatigue cycles of the charging line nozzle, the inspector concluded the licensee implemented actions to effectively manage the effects of aging during the period of extended operation. The inspector concluded the licensee met Commitment 23 prior to the period of extended operation.

The inspector closed this unresolved item.

.3 (Closed) Unresolved Item 05000285/2013009-02: Evaluation of Operating Cycles for

Fatigue Monitoring Program Commitment 24 specified, Cycles which involve power changes, operating pressure and temperature variations, and feedwater additions with the plant in hot standby conditions will be conservatively estimated from a review of plant operating records to predict current cycles under the fatigue monitoring program. Once current number of cycles has been established, a review will be performed to determine if there is a potential for exceeding the allowable cycles and should be managed. If so, they'll be counted and managed by the fatigue monitoring program.

The licensee had not conservatively estimated the number of cycles nor performed an evaluation of whether the potential for exceeding the allowable cycles in the period of extended operation existed as required by Commitment 24. Specifically, the licensee did not count the cycles related to the following transients:

• plant loading/unloading at 10 percent of full power per minute,

• 10 percent step load increase/decrease,

• operating variations of +100 pounds per square inch and +6 degrees Fahrenheit from normal operating pressure and temperature, and

• feedwater additions of 300 gallons per minute at 32 degrees Fahrenheit with the plant in HOT STANDBY.

The licensee documented this failure to count the cycles in Condition Report 2013-10756.

During this inspection, the inspector reviewed the actions taken by the licensee to count cycles related to the above transients. The inspector verified that the licensee had completed their actions to estimate the number of cycles on July 11, 2013, prior to entering the period of extended operation. The licensee had not had any unanticipated feedwater additions that met the criteria since they had replaced their steam generators in 2006.

For the loading/unloading transients, the licensee reviewed 10 of their 26 operating periods. The licensee determined they had 690 transients in this category for the reviewed operating periods and the maximum number of transients in any one period was 144 cycles. The licensee multiplied this maximum number by the periods not reviewed for an estimate of 2304 cycles. The total estimated cycles that resulted were 2994 cycles. This number of cycles was well below the 15,000 allowed cycles.

For the 10 percent step load increase or decrease, the licensee counted cycles for the same 10 operating periods. The licensee data demonstrated that they had 62 cycles for these 10 periods with a maximum cycle count of 16 cycles in any one period. The licensee estimated the 16 unanalyzed periods had a total of 256 cycles. The total estimated cycles was 318 cycles for the life of the plant. This number of cycles was well below the 2000 allowed cycles.

The licensee only had pressure and temperature data for operating period 26. The licensee determined from review of one hot leg thermocouple that there were 43 instances of exceeding +6 degrees Fahrenheit from the normal operating temperature.

Multiplying this value by 26 resulted in 1118 cycles. The licensee determined that there were 27 instances when the monitoring point had plus or minus 100 psi change from the nominal pressure. Multiplying this value resulted in 702 cycles for the life of the plant.

To be conservative the licensee combined the pressure and temperature and pressure transients and multiplied by a factor of ten ((702 + 1118)

• 10) to get an estimated 18,200 cycles. This is well below the 1,000,000 cycles.

Based upon review of the actions taken to identify the number of cycles for each type of transient listed in the commitment prior to entering the period of extended operation, the inspector concluded the licensee implemented actions to effectively manage the effects of aging during the period of extended operation. The inspector concluded the licensee met Commitment 24 prior to the period of extended operation.

The inspector closed this unresolved item.

.4 (Closed) Unresolved Item 05000285/2013009-03: Flaw Tolerance Evaluation for

Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components Commitment 34 specified, Develop the thermal aging embrittlement of cast austinetic stainless steel program that reflects the program elements of Generic Aging Lessons Learned (GALL) Report, GALL Aging Management Program (AMP) XI.M12, and other commitments in response to the NRC staff's review, as documented in the responses to staff requests for additional information and potential open items.

The licensee opted to manage thermal aging embrittlement of cast austinetic stainless steel components in the reactor coolant system through a flaw tolerance evaluation of reactor coolant system cast austinetic stainless steel piping. However, the licensee made two errors related to the assumptions and conclusions in the flaw tolerance evaluation. Specifically, the licensee:

• Inappropriately ruled out stress corrosion cracking effects because they concluded it was not credible since they maintained an effective Water Chemistry Program aging management program. However, the safety evaluation report and license renewal application credited this program for managing crack initiation and growth caused by stress corrosion cracking.

• Did not provide any correlation between the postulated flaw sizes and the detection capability of the non-destructive in-service inspection examination technique currently used or expected to be used.

The licensee documented this deficiency in Condition Report 2013-11991.

During this inspection, the inspector determined the licensee completed Calculation LTR-PAFM-11-119, Flaw Tolerance Evaluation for Susceptible CASS (cast austenitic stainless steel) Reactor Coolant Piping Components Including Auxiliary Nozzles in Fort Calhoun, Revision 1, in August 2013, prior to entering the period of extended operation. The inspector compared this revision of the calculation to the prior revision. The inspector verified that the calculation evaluated the impact of stress corrosion cracking on reactor coolant system piping. The calculation acknowledged that the cast austenitic stainless steel retained residual stresses but discussed that the environmental factor was strictly controlled with water chemistry; therefore, the calculation concluded that cracking and flaw growth caused by stress corrosion cracking was highly unlikely. The calculation further acknowledged that the state-of-the-art nondestructive examination did not have a reliable method to examine cast austenitic stainless steels. The calculation described that it would take a very large flaw (greater than 20 percent thru wall) to be seen. However, the analysis demonstrated that the material was unlikely to develop a flaw.

Based upon review of the actions taken to evaluate the effects of stress corrosion cracking and demonstration that nondestructive examinations would not be required, the inspector concluded the licensee implemented actions to effectively manage the effects of aging during the period of extended operation. The inspector concluded the licensee met this aspect of Commitment 34 prior to the period of extended operation.

The inspector closed this unresolved item.

.5 (Closed) Unresolved Item 05000285/2013009-04: Thermal Aging Embrittlement of Cast

Austenitic Stainless Steel Nozzles in the Reactor Coolant System Commitment 34 specified, Develop the thermal aging embrittlement of cast austinetic stainless steel program that reflects the program elements of GALL AMP XI.M12, and other commitments in response to the NRC staff's review, as documented in the responses to staff requests for additional information and potential open items.

The licensee did not evaluate all reactor coolant system components for susceptibility to thermal aging embrittlement in accordance with the NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 0. Specifically, the licensee had not included the cast austinetic stainless steel components (as listed below) in a flaw tolerance evaluation for their reactor coolant system piping. The affected reactor coolant system nozzles included:

• RC-PIPE-2501Q-CHARGE-NOZZ

• RC-PIPE-2501Q-DRAIN-NOZZ

• RC-PIPE-25011-PM-NOZZ

• RC-PIPE-2501Q-PMS-NOZZ

• RC-PIPE-2501Q-SDC-INLET-NOZZ

• RC-PIPE-2501Q-SDC-OUT-NOZZ The licensee documented this deficiency in Condition Report 2013-11991.

During this inspection, the inspector verified that Calculation LTR-PAFM-11-119, Revision 1, specifically addressed the affected reactor coolant system nozzles and determined that the charging nozzle experienced the highest of the stresses and would need to be monitored for fatigue.

Based upon review of the actions taken to include these components in the flaw tolerance evaluation, the inspector concluded the licensee implemented actions to effectively manage the effects of aging during the period of extended operation. The inspector concluded the licensee met this aspect of Commitment 34 prior to the period of extended operation.

The inspector closed this unresolved item.

.6 (Closed) Unresolved Item 05000285/2013009-05: Submittal of Leak-Before-Break

Analysis Commitment 39 specified, OPPD will complete a plant-specific leak before break (LBB)analysis using the latest LBB criteria. OPPD will submit to the NRC a license amendment request containing the plant-specific LBB evaluation.

During the last inspection, the inspectors determined that the licensee had identified that some stress locations on the hot and cold leg reactor vessel nozzles at the extended power uprate were lower than those calculated for their current license basis locations at the reduced power levels. Since the licensee had not implemented the power uprate, the licensee initiated actions to perform the leak-before-break calculation to 60 years at the current operating conditions. Condition Report 2013-10952 tracked completion of the leak-before-break analysis for current power levels.

During this inspection, the inspector determined that the licensee had submitted Letter LIC-13-0100, License Amendment Request: Plant-Specific Leak-before-Break Analysis, dated August 5, 2013, which attached the revised leak-before-break evaluation prior to entering the period of extended operation. The analysis demonstrated that the dynamic loading effects over the postulated reactor coolant system pipe breaks need not be included in the structural design basis for the additional 20-year license renewal period.

Since the licensee submitted their leak before break re-analysis prior to entering the period of extended operation; the inspector concluded the licensee met Commitment 39 prior to the period of extended operation.

The inspector closed this unresolved item.

4OA6 Meetings, Including Exit

The inspector presented the initial inspection results to Mr. Scot Swanson, Acting Plant Manager, and other members of the licensee staff during an exit meeting conducted on May 16, 2014. The licensee acknowledged the NRC inspection observations. The inspector retained no proprietary information and verified that no proprietary information was documented in this report.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

L. Cortopassi, Vice President and Chief Nuclear Officer

K. Erdman, Engineering Programs Supervisor

C. Halley, Electrical Systems Engineer

J. Johnston, Program Engineer

P. Koneck-Wilwerding, Mechanical Design Engineer

K. Maassen, License Renewal Program Engineer

E. Matzke, Compliance Engineer

J. Smidt, Electrical Cables Program Engineer

R. Swerczek, Fire Protection Program Engineer

P. Turner, Structural Program Engineer

Westinghouse Personnel

C. Burton, Site Representative

T. Meikle, Structural Analysis Design Engineer

M. Norge, Piping Analysis Design Engineer

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Closed

Item Description

05000285/2011006-01 VIO Inadequate Corrective Actions to Ensure the Reliability of the Raw Water Pump Power Cables (Section 4OA5.1)

05000285/2013009-01 URI Evaluation of Environmentally-Assisted Fatigue for Charging Line Nozzle (Section 4OA5.2)

05000285/2013009-02 URI Evaluation of Operating Cycles for Fatigue Monitoring Program (Section 4OA5.3)

05000285/2013009-03 URI Flaw Tolerance Evaluation for Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components (Section 4OA5.4)

05000285/2013009-04 URI Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Nozzles in the Reactor Coolant System (Section 4OA5.5)

05000285/2013009-05 URI Submittal of Leak-Before-Break Analysis (Section 4OA5.6)

-1- Attachment

COMMITMENTS The inspectors closed the following commitments during this inspection: 23, 24, 34, and 39.

NRC closed Commitments 1, 2, 3, 4, 5, 17, and 40 in Inspection Report : 05000285/2012009.

NRC closed Commitments 6, 7, 8, 9, 10, 11, 12, 13, 14, 15, 16, 18, 19, 20, 21, 22, 25, 26, 27, 28, 29, 30, 31, 32, 33, 35, 36, 37, 38, and 41 in Inspection Report : 05000285/2013009.