IR 05000305/2010007

IR 05000305-10-007 (DRS) on 01/25/2010 - 02/12/2010 for Kewaunee Power Station, Evaluations of Changes, Tests or Experiments and Permanent Plant Modifications

ML100710480
Person / Time
Site:	Kewaunee
Issue date:	03/12/2010
From:	Robert Daley Engineering Branch 3
To:	Heacock D Dominion Energy Kewaunee
References
IR-10-007
Download: ML100710480 (21)

Text

March 12, 2010

SUBJECT:

KEWAUNEE POWER STATION EVALUATIONS OF CHANGES, TESTS OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000305/2010007(DRS)

Dear Mr. Heacock:

On February 12, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an evaluation of changes, tests or experiments and permanent plant modifications inspection at your Kewaunee Power Station. The enclosed report documents the inspection findings, which were discussed on February 12, 2010, with Mr. Stephen Scace and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

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

Based on the results of this inspection, one NRC-identified finding of very low safety-significance was identified. The finding involved a violation of NRC requirements. However, because of its very low safety-significance, and because the issue was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation (NCV) in accordance with Section VI.A.1 of the NRC Enforcement Policy.

If you contest the subject or severity of a NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Kewaunee Power 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 Regional Administrator, Region III, and the NRC Resident Inspector at the Kewaunee Power Station. The information that you provide will be considered in accordance with Inspection Manual Chapter 0305.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS)

component of NRC's 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/

Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket No. 50-305 License No. DPR-43

Enclosure:

Inspection Report 05000305/2010007(DRS)

w/Attachment: Supplemental Information

REGION III==

Docket No:

50-305 License No:

DPR-43 Report No:

05000305/2010007(DRS)

Licensee:

Dominion Energy Kewaunee, Inc.

Facility:

Kewaunee Power Station Location:

Kewaunee, Wisconsin Dates:

January 25, 2010, through February 12, 2010 Inspectors:

G. Hausman, Senior Reactor Inspector (Lead)

N. Féliz Adorno, Reactor Inspector

M. Munir, Reactor Inspector Observer:

E. Sánchez Santiago, Reactor Engineer Approved by:

R. C. Daley, Chief

Engineering Branch 3

Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000305/2010007 (DRS); 01/25/2010 - 02/12/2010; Kewaunee Power Station; Evaluations of Changes, Tests or Experiments and Permanent Plant Modifications.

This report covers a two-week announced baseline inspection on evaluations of changes, tests or experiments and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors. One Green finding was identified by the inspectors. The finding was considered a Non-Cited Violation (NCV) of NRC regulations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

A finding of very low safety-significance and associated NCV of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. Specifically, the licensee failed to assure that the methodology used in calculation C11716, MCC [Motor Control Center] Control Circuit Voltage Drop, Revision 1, correctly represented the sequence of operation for the various devices contained within the plant equipments control circuitry, such that the minimum required MCC voltage was available for proper circuit operation. Upon discovery of this condition, the licensee performed a preliminary evaluation and entered the finding into their corrective action program (CR366627 and CR366865).

This finding was more than minor in accordance with IMC 0612, Appendix B because the finding was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate MCC voltages could render the safety-related loads required to mitigate the consequences of a design basis accident inoperable and not available. In addition, as a result of the calculation errors, the inspectors were concerned that unsubstantiated MCC voltage values could be used in future calculations and modifications to plant equipment. To resolve the inspectors concerns, the licensee completed an interim evaluation, which evaluated the calculations other circuit models and associated cases. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the control circuits modeled in the calculation. The finding was of very low safety-significance based on a Phase I screening in accordance with IMC 0609, Significance Determination Process,

Attachment 0609.04, Phase I - Initial Screening and Characterization of Findings,

Table 4a.

This finding has a cross-cutting aspect in the area of human performance, work practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported.

Specifically, the licensee failed to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. (H.4(c)) (Section 1R17.2b)

Licensee-Identified Violations

No violations of significance were identified.

REPORT DETAILS

1R17 Evaluations of Changes, Tests or Experiments and Permanent Plant Modifications

.1 Evaluations of Changes, Tests or Experiments

a. Inspection Scope

From January 25, 2010, through February 12, 2010, the inspectors reviewed nine safety evaluations (SEs) performed pursuant to 10 CFR 50.59 to determine if the evaluations were adequate and that prior NRC approval was obtained as appropriate. The inspectors also reviewed 12 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:

• the changes, tests or experiments performed were evaluated in accordance with 10 CFR 50.59 and that sufficient documentation existed to confirm that a license amendment was not required;

• the safety issue requiring the change, tests or experiment was resolved;

• the licensee conclusions for evaluations of changes, tests or experiments were correct and consistent with 10 CFR 50.59; and

• the design and licensing basis documentation was updated to reflect the change.

The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests and Experiments, dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for 10 CFR 50.59, Changes, Tests and Experiments.

This inspection constituted 9 samples of evaluations and 12 samples of changes as defined in IP 71111.17-04.

b. Findings

No findings of significance were identified.

.2 Permanent Plant Modifications

a. Inspection Scope

From January 25, 2010, through February 12, 2010, the inspectors reviewed 15 permanent plant modifications that had been installed in the plant during the last three years. This review included in-plant walkdowns of the train A emergency diesel generator room, fuel oil storage tank, the safety injection common header injection line gas accumulation chamber and the north penetration room scaffolding. The modifications were selected based upon risk-significance, safety-significance, and complexity. The inspectors reviewed the modifications selected to determine if:

• the supporting design and licensing basis documentation was updated;

• the changes were in accordance with the specified design requirements;

• the procedures and training plans affected by the modification have been adequately updated;

• the test documentation as required by the applicable test programs has been updated; and

• post-modification testing adequately verified system operability and/or functionality.

The inspectors also used applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors is included as an Attachment to this report.

This inspection constituted 15 permanent plant modification samples as defined in IP 71111.17-04.

b. Findings

Calculation Methodology Did Not Represent Actual Plant Equipment Configuration

Introduction:

A finding of very low safety-significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design.

Specifically, the licensee failed to assure that the methodology used in calculation C11716, MCC [Motor Control Center] Control Circuit Voltage Drop, Revision 1, correctly represented the sequence of operation for the various devices contained within the plant equipments control circuitry, such that the minimum required MCC voltage was available for proper circuit operation.

Description:

The inspectors reviewed safety-related calculation C11716, MCC Control Circuit Voltage Drop, Revision 1. The calculation determined the minimum required MCC voltage for starters/contactors, auxiliary relays, interposing relays and other devices used in the plant equipments control circuitry during inrush (pickup) and holding (sealing) conditions for all safety-related (QA1) and non-safety-related (QA2&3) circuits powered by control power transformers (CPTs) connected to MCCs. The calculation methodology involved the development of 18 control circuit models, which represented the control circuitry for each load connected to the MCCs. Each circuit model was then evaluated based on the sequence of operation of various devices reflected in the circuit model and the minimum MCC voltage required for proper circuit operation. The minimum required MCC voltages were then compared with the available MCC voltages derived from calculation C11450, Auxiliary Power System Modeling and Analysis, Revision 1, to assure that all the components of the control circuit had adequate voltage for proper circuit operation.

During the inspectors review, errors were identified with the charging pumps A and B control circuit models. The inspectors concluded that calculation C11716 did not accurately evaluate the actual control circuits operation for charging pumps A and B motors 1-067 and 1-106, respectively. Specifically, the inspectors questioned the circuit models operation and the licensees evaluation of relays 42IR and FSR in each circuit.

The calculation stated that relays 42IR and FSR picked-up simultaneously, however, the analysis evaluated them singularly. In addition, there were calculation errors made in modeling the series and parallel circuits. As a result, the formulae developed to calculate the required minimum MCC voltage did not accurately represent circuit operation.

To resolve the inspectors concerns the licensee issued condition report CR366627 and performed a preliminary evaluation. The licensee determined that because of the circuit operation modeling errors there was a decrease in the margin between the minimum MCC voltage required and the voltage available. However, the licensee concluded that enough margin remained such that the component was operable.

The inspectors review of calculation C11716 consisted of only one circuit model and the inspectors were concerned about the adequacy of the other 17 circuit models and the different cases associated with each circuit model. As a result, the licensee conducted a review of the remaining circuit models and different cases. The licensee discovered that calculation C11716 did not correctly evaluate the control circuitry for the radiation monitor channels R11/R12 for the containment particulate/containment gas pump motor 1-1227 and initiated condition report CR366865. The licensee found that the calculation methodology did not match the actual circuit operation but concluded that the error did not impact the operability or functionality of motor 1-1227.

Analysis:

The inspectors determined that the licensees failure to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design was contrary to the requirements of 10 CFR Part 50, Criterion III, Design Control and was a performance deficiency.

The finding was determined to be more than minor because the finding was associated with the mitigating systems cornerstone attribute of design control and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to assure that the methodology used in calculation C11716 correctly represented the actual plant equipment configuration and that adequate design reviews were performed. The inspectors considered this a significant calculational error. The inspectors were concerned that inadequate MCC voltages would have rendered the

safety-related loads required to mitigate the consequences of a design basis accident inoperable and not available. In addition, as a result of the calculation errors, the inspectors were concerned that unsubstantiated MCC voltage values could be used in future calculations and modifications to plant equipment. To resolve the inspectors concerns, the licensee completed an interim evaluation, which evaluated the calculations other circuit models and associated cases. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the control circuits modeled in the calculation.

The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase I - Initial Screening and Characterization of Findings, Table 4a for the mitigating systems cornerstone. The basis for selecting the mitigating systems cornerstone was that inadequate MCC voltages could render the safety-related loads required to mitigate the consequences of a design basis accident inoperable and not available. The finding screened as Green because it was a design deficiency that did not result in actual loss of safety function.

This finding has a cross-cutting aspect in the area of human performance, work practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported.

Specifically, the licensee failed to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. (H.4(c))

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion III, Design Control requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, from August 28, 2008, to February 12, 2010, the licensee failed to provide for calculation C11716, MCC Control Circuit Voltage Drop, Revision 1, design control measures that shall provide for verifying or checking the adequacy of design, such as by performance of design reviews or by the use of alternate or simplified calculational methods. Specifically, the licensee failed to assure that the methodology used in calculation C11716 correctly represented the sequence of operation for the various devices contained within the plant equipments control circuitry, such that the minimum required MCC voltage was available for proper circuit operation. Because this violation was of very low safety-significance and it was entered into the licensees corrective action program as CR366627 and CR366865, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000305/2010007-01 (DRS)).

4OA2 Identification and Resolution of Problems

.1 Routine Review of Condition Reports

a. Inspection Scope

From January 25, 2010, through February 12, 2010, the inspectors reviewed corrective action process documents that identified or were related to 10 CFR 50.59 evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent pant modifications and evaluations for changes, tests or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

4OA5 Other Activities

(Closed) Unresolved Item (URI)05000305/2009005-05 (DRP): Evaluation to Support Seismic Classification Downgrade to the Fuel Transfer Carriage The fuel transfer carriage is used to transport fuel assemblies between the reactor refueling cavity and the spent fuel pool. The carriage runs on tracks (i.e., transport rails)that extend from the refueling cavity through the transfer tube and into the fuel transfer canal. On September 19, 2009, and October 4, 2009, during testing and the transfer of irradiated fuel, the carriage stopped moving. The licensees assessment concluded that the cause for the carriage stopping was binding of the carriages seismic restraints. The seismic restraints were installed in 2008 to prevent carriage derailing. The seismic restraints were L-shaped clips that attached the carriage to the carriages transport rails.

To prevent the carriage from binding during core defueling activities, the licensee modified the carriage by removing all the seismic restraints per engineering change notice (ECN) 3784-003. The inspectors review of the ECN identified a contractors condition report (i.e., Number 2009-6596), which stated that the contractor recommended a minimum of three restraint angles to be maintained in-place on the carriage. Contrary to the contractors recommendation, the licensee performed an evaluation (i.e., ECN 3784-003, Attachment 2) and concluded that all of the seismic restraints could be removed. The inspectors noted that this evaluation was not a formal calculation. The licensee stated that the carriage was non-seismic and did not require a formal calculation. However, the inspectors questioned the licensees change to the carriages seismic qualification and the adequacy of the informal calculation that supported the removal of the carriages seismic restraints. As a result, this URI was issued in the stations 2009 fourth quarter integrated inspection report (ML100390005)pending completion of an NRC inspectors review.

During this inspection, the inspectors reviewed drawings, license basis documents, procedures, vendor documents, modification packages, and 50.59 screenings associated with the classification downgrade of the fuel transfer carriage and the removal of the seismic restraints. In addition, the inspectors conducted interviews of plant personnel and obtained clarification concerning the URI from the Office of Nuclear Reactor Regulations (NRR). The inspectors concluded to resolve this URI, the following two questions must be answered:

2. Was the calculation to support removal of the fuel transfer carriages seismic

restraints inadequate?

Question 1: The fuel transfer carriage was originally classified as quality assurance QA (1). This classification was documented in internal memos and communications by the licensee. The documents were not referenced in any correspondence with the NRC as part of the original licensing activities. In 2008, the licensee downgraded the carriage from QA

(1) to QA

(2) and classified the carriage as seismic class 3. As part of this change, the licensee added the carriages seismic classification to the final safety analysis report (FSAR). Prior to 2008, the carriage had no formal seismic classification.

The inspectors concluded that the licensees classification was consistent with the definition of seismic class 3 as described in the FSAR and with the licensees procedures applicable for QA classifications. The inspectors contacted NRR concerning this issue. Based on the discussions with NRR, it was considered reasonable to conclude that the NRC staff accepted a non-safety classification for the fuel transfer carriage or did not specifically review the classification. Since the original classification of the carriage was not previously described in the Kewaunee Power Stations licensing basis, the inspectors determined that a 50.59 evaluation was not required and that the seismic qualification of the fuel transfer carriage was not inappropriately changed.

Question 2: For the calculation that supported the physical modification of the fuel transfer carriage, the inspectors noted that in FSAR, Table B.6-1, a design basis earthquake load analysis was not applicable to seismic class III components. However, the licensee completed a calculation that concluded that the removal of the seismic restraints did not increase the likelihood of the fuel transfer carriage becoming derailed.

Specifically, the carriage would not derail or overturn during a design basis earthquake (DBE). In addition, the carriage was referenced in one of the safety analyses and the results of the analysis was not affected by the removal of the seismic restraints because the accident analysis already considered a scenario where the fuel transfer carriage became stuck in the transfer tube while carrying a fuel bundle. The safety analysis also concluded that if this condition were to occur the fuel bundle would not be damaged and would not lose cooling. The function of the seismic restraints was not referenced or taken credit for in the FSAR safety analysis. Therefore, the inspectors concluded that the calculation to support removal of the fuel transfer carriages seismic restraints was not a concern.

Based on the above assessment the inspectors determined that no performance deficiencies or violations of regulatory requirements exist. The inspectors had no further concerns in this area. The documents that were reviewed are included in the attachment to this report. This unresolved item is closed.

4OA6 Meetings

Exit Meeting Summary

On February 12, 2010, the inspectors presented the inspection results to Mr. Stephen Scace, and other members of the licensee staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material reviewed during the inspection was returned to the licensee staff.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

M. Aulik, Design Manager

T. Breene, Nuclear Licensing Manager

T. Evans, Maintenance Manager

S. Heironimus, ECP Specialist

W. Henry, NOD Manager

D. Laing, Training Manager

J. Marean, Rapid Response - Engineering

J. McNamara, Mechanical/Structural Design Engineering Supervisor

R. Repshas, Licensing _ Engineering

M. Rosseau, Electrical/I&C Design Engineering Supervisor

S. Scace, Site Vice President

R. Simmons, Plant Manager

M. Sortwell, Engineering Supervisor

D. Vorpaul, Balance of Plant/System Engineering Supervisor

M. Wilson, Safety and Licensing Director

S. Yuen, Engineering Director

Nuclear Regulatory Commission

J. Cassidy, Senior Health Physicist

M. Kunowski, Branch Chief

R. Ruiz, Senior Resident Inspector (acting)

LIST OF ITEMS

Opened

05000305/2010007(DRS)-01 NCV Calculation Methodology Did Not Represent Actual Plant Equipment Configuration

Closed

05000305/2010007(DRS)-01 NCV Calculation Methodology Did Not Represent Actual Plant Equipment Configuration

05000305/2009005(DRP)-05 URI Evaluation to Support Seismic Classification Downgrade to the Fuel Transfer Carriage

Discussed

None