IR 05000282/2011012

IR 05000282-11-012(DRS); 05000306-11-012(DRS); 10/31/2011 - 11/18/2011; Prairie Island Nuclear Generating Plant; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications

ML11349A413
Person / Time
Site:	Prairie Island
Issue date:	12/15/2011
From:	Robert Daley Engineering Branch 3
To:	Schimmel M Northern States Power Co
George Hausman
References
IR-11-012
Download: ML11349A413 (22)

Text

ber 15, 2011

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT EVALUATIONS OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000282/2011012(DRS); 05000306/2011012(DRS)

Dear Mr. Schimmel:

On November 18, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications Inspection at your Prairie Island Nuclear Generating Plant. The enclosed inspection report documents the inspection results which were discussed on November 18, 2011, with you 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, two NRC-identified findings of very low safety significance were identified. The findings involved violations of NRC requirements. However, because of their very low safety significance and because the issues were entered into your corrective action program, the NRC is treating these issues as Non-Cited Violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the subject or severity of any NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with 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 Prairie Island Nuclear Generating Plant. In addition, if you disagree with the cross-cutting aspect assigned to 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 Prairie Island Nuclear Generating Plant.

In accordance with Title 10, Code of Federal Regulations (CFR), Part 50, Section 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), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA by A. Dahbur for/

Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket Nos.: 50-282; 50-306 License Nos.: DPR-42; DPR-60

Enclosure:

Inspection Report 05000282/2011012(DRS); 05000306/2011012(DRS)

w/Attachment: Supplemental Information

REGION III==

Docket Nos.: 50-282; 50-306 License Nos.: DPR-42; DPR-60 Report No: 05000282/2011012(DRS); 05000306/2011012(DRS)

Licensee: Northern States Power Company, Minnesota Facility: Prairie Island Nuclear Generating Plant Location: Welch, MN Dates: October 31, 2011, through November 18, 2011 Inspectors: George M. Hausman, Senior Reactor Inspector (Lead)

Jasmine A Gilliam, Reactor Inspector Dariusz Szwarc, Reactor Inspector Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000282/2011012(DRS); 05000306/2011012(DRS); 10/31/2011 - 11/18/2011; Prairie

Island Nuclear Generating Plant; 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. Two findings were identified by the inspectors. The findings were considered Non-Cited Violations (NCVs) 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). Cross-cutting aspects were determined using IMC 0310, Components Within the Cross-Cutting Areas. 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: Initiating Events

Green.

The inspectors identified a finding of very low safety significance and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B,

Criterion III, Design Control, for the licensees failure to check the adequacy of design for flammable gas bottles installed in areas located within the auxiliary building and their impact on safety-related cables and equipment. Specifically, the licensee failed to evaluate how a failure of the flammable gas bottles and a resulting fire or explosion at the installed locations could impact nearby safety-related structures, systems, or components. The licensee entered this issue into their corrective action program to review the placement of the flammable gas bottles.

The inspectors determined that the finding was more than minor because the finding was associated with the Initiating Events cornerstones attribute of Protection against External Factors (Fire) and affected the cornerstones objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance due to the low fire initiating frequency and the availability of remaining mitigating systems. This finding did not have a cross-cutting aspect because the finding was not representative of current performance. (Section 1R17.2b)

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a condition adverse to quality.

Specifically, the licensee failed to submit a license amendment request (LAR) to correct the non-conservative Technical Specification (TS) surveillance requirements in Section 3.8.1 for the emergency diesel generators (EDGs) allowable steady state frequency. The issue was originally identified and entered into the licensees corrective action program on September 8, 2006. During this inspection, the licensee entered the finding into their corrective action program to evaluate how to resolve the issue.

The inspectors determined that the finding was more than minor because the finding was associated with the Mitigating Systems cornerstones attribute of Equipment Performance 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 licensee could not be assured that the design requirements for the EDGs system loads would operate within the appropriate design specifications if the EDGs were allowed to operate within the non-conservative TS allowable steady state frequency of 58.8 Hertz (Hz) and 61.2 Hz. As a result, the licensee established an administrative limit to limit operation of the EDGs to a frequency between 59.5 Hz and 60.5 Hz. The finding was of very low safety significance because it did not result in a loss of operability. The finding had a cross-cutting aspect in the area of human performance, decision-making because the licensee repeatedly delayed submitting the license amendment until a resolution was developed by an industry working group. [H.1 (a)] (Section 4OA2.1b)

Licensee-Identified Violations

No violations of significance were identified.

REPORT DETAILS

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

.1 Evaluation of Changes, Tests, or Experiments

a. Inspection Scope

From October 31, 2011, through November 18, 2011, the inspectors reviewed seven safety evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR),

Part 50, Section 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 seven 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 October 31, 2011, through November 18, 2011, the inspectors reviewed seven permanent plant modifications that had been installed in the plant during the last three years. This review included in-plant walkdowns for portions of the modified 11 turbine driven auxiliary feedwater (AFW) pump, the D1/D2 emergency diesel generators (EDGs)and the diesel driven cooling water pump (DDCLP) fuel oil storage tank (FOST) pump motor starter equipment. 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 seven permanent plant modification samples as defined in IP 71111.17-04.

b. Findings

Flammable Gas Bottles Located in the Auxiliary Building

Introduction:

The inspectors identified a finding of very low safety significance (Green)and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to check the adequacy of design for flammable gas bottles installed in areas located within the auxiliary building and their impact on safety-related cables and equipment. Specifically, the licensee failed to evaluate how a failure of the flammable gas bottles and a resulting fire or explosion at the installed locations could impact nearby safety-related structures, systems, or components.

Description:

The inspectors identified two instances of flammable gas bottles installed on the 715 foot elevation of the auxiliary building near safety-related cables and equipment. A bottle of acetylene was located in the chemistry hot sample lab. That acetylene was used to provide acetylene gas to the internal burner of an atomic absorption analyzer located in the lab. A bottle of 80 percent hydrogen was also located behind a stairwell outside the chemistry hot sample lab. The hydrogen was used in support of a gas analyzer system and was installed prior to March 13, 1980, as documented in a licensees letter to the Office of Nuclear Reactor Regulation (NRR).

During the inspectors walkdowns, the inspectors noticed that the flammable gas bottles were within close proximity of safety-related cables and equipment (e.g., Technical Specification (TS) specified equipment) at each location. The inspectors raised concerns about the locations of the flammable gas bottles with respect to the safety-related cables and equipment located nearby.

The flammable gas bottles present a fire and an explosion hazard. According to Table 2-7.1 of the Society of Fire Protection Engineers (SFPE) Handbook of Fire Protection Engineering (Fourth Edition) acetylene and hydrogen have a lower flammability limit of 2.5 and 4 percent, and an upper flammability limit of 100 and 75 percent, respectively. This means that an acetylene mixture of between 2.5 and 100 percent and a hydrogen mixture of between 4 and 75 percent will burn. The hydrogen bottle had a mixture of 80 percent hydrogen. If the hydrogen were to escape from the bottle, it would dilute with the surrounding atmosphere and fall into the flammable range of between 4 and 75 percent.

The hydrogen and acetylene bottles each had a regulator attached to the discharge.

However, if a piece of equipment or some other object were to hit the regulator it could fail, cause a spark, and ignite the flammable gas. A fire from one or more of the flammable gas bottles could damage safety-related cables and an explosion could additionally damage other nearby safety-related equipment.

The licensee could not locate a safety or design evaluation that adequately analyzed the hazards associated with the hydrogen or the acetylene gas bottles. As a result, the inspectors concluded that the licensee failed to check the adequacy of the design for the flammable gas bottles installed and their impact on safety-related cables and equipment.

Specifically, the licensee failed to evaluate how a failure of the flammable gas bottles and a resulting fire or explosion at the installed locations could impact nearby safety-related structures, systems, or components.

The licensee entered this issue into their corrective action program (CAP) as CAP 01313606, Combustible Gas Cylinders May Not be Evaluated Adequately, dated November 17, 2011, to review the placement of the flammable gas bottles.

Subsequently, the licensee also issued CAP 01314188, Compressed Gas Cylinder Storage without 50.59 Process, on November 21, 2011, to document the failure to evaluate the placement of the flammable gas bottles.

Analysis:

The inspectors determined that the failure to evaluate the impact of the flammable gas bottles installed locations near safety-related cables and equipment was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. The inspectors determined that the finding was more than minor because the finding was associated with the Initiating Events cornerstones attribute of Protection against External Factors (Fire) and affected the cornerstones objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations.

Specifically, the installed locations of the flammable gas bottles could have resulted in damage to safety-related cables and equipment if the gas bottles were to ignite or explode.

In accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Phase I - Initial Screening and Characterization of Findings, Table 3b, the inspectors determined the finding degraded the fire protection defense-in-depth strategies. Therefore, screening under IMC 0609, Appendix F, Fire Protection Significance Determination Process, was required. The inspectors determined that the finding impacted the Fire Prevention and Administrative Controls category.

Based on review of IMC 0609, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, the inspectors determined the degradation rating to be high because of the flammable gases being more flammable than low flashpoint combustibles and there being a significant fire hazard associated with release of the gases. The Duration Factor was 1.0 based on the duration of the degradation being greater than 30 days per Table 1.4.1, Duration Factors. An overall fire frequency of 1.3E-3 per year was assigned for the flammable gas bottles based on information from IMC 0609, Appendix F, Attachment 4, Fire Ignition Source Mapping Information: Fire Frequency, Counting Instructions, Applicable Fire Severity Characteristics, and Applicable Manual Fire Suppression Curves.

The Region III Senior Reactor Analyst used the Prairie Island Standard Plant Analysis Risk (SPAR) Model, Version 8.15, and Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), Version 8.0.7.17, to calculate a conditional core damage probability (CCDP) less than 1E-6 conservatively assuming a fire due to failure of the flammable gas bottles that resulted in a plant trip and damage to a safe shutdown division. Based on the above CCDP and frequency values, the risk associated with this finding is very low (Green).

The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.

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 March 13, 1980, through November 18, 2011, the licensee failed to check the adequacy of design for flammable gas bottles installed within the auxiliary building and their impact on safety-related cables and equipment. Specifically, the licensee failed to evaluate how a failure of the flammable gas bottles and a resulting fire or explosion at the installed locations could impact nearby safety-related structures, systems, or components.

Because this violation was of very low safety significance and it was entered into the licensees corrective action program as CAP 01313606 and CAP 01314188, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000282/2011012-01(DRS);05000306/2011012-01(DRS),

Flammable Gas Bottles Installed in the Auxiliary Building).

4OA2 Problem Identification and Resolution

.1 Routine Review of Condition Reports

a. Inspection Scope

From October 31, 2011, through November 18, 2011, the inspectors reviewed fifteen 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 evaluations of changes, tests, or experiments and permanent plant modifications. 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

Failure to Correct a Condition Adverse to Quality

Introduction:

The inspectors identified a finding of very low safety significance (Green)and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a condition adverse to quality.

Specifically, the licensee failed to submit a license amendment request (LAR) to correct the non-conservative Technical Specification (TS) surveillance requirements in Section 3.8.1 for the emergency diesel generators (EDGs) allowable steady state frequency. The issue was identified and entered into the licensees corrective action program on September 8, 2006.

Description:

The licensees TS surveillance requirements in Section 3.8.1 stated, Verify each DG [Diesel Generator] starts from standby conditions and achieves steady state voltage 3740 V [Volts] and 4580 V, and frequency 58.8 Hz [Hertz] and 61.2 Hz.

The inspectors found that the licensees EDG loading calculations ENG-EE-018, Diesel Generator Sequence Loading for an SI Event Concurrent with Loss of Offsite Power (LOOP) for D1, D2, D5, D6, ENG-EE-021, Diesel Generator Steady State Loading for an SI Event Concurrent with Loss of Offsite Power (LOOP) for D1, D2, D5, D6, and ENG-EE-045, Diesel Generator Steady State Loading for a LOOP Coincident with an SBO [Station Blackout], assumed/established an administrative limit for the EDGs steady state frequency with a range of 59.5 Hz to 60.5 Hz. The administrative limit was established because if the EDGs were allowed to operate within the non-conservative TS allowable steady state frequency range of 58.8 Hz and 61.2 Hz, the design requirements for the EDGs system loads could not be assured.

On September 8, 2006, the licensee identified the non-conservative TS and entered the issue into their corrective action program as CAP 01049042, Evaluation of Impacts from EDG Frequency Variation. The licensee performed an operability recommendation as part of that CAP and implemented the EDG frequency compensatory measure to limit operation of the EDGs to between 59.5 Hz and 60.5 Hz. In that CAP, the licensee also determined that an amendment would have to be submitted to the NRC in order to correct the non-conservative TS.

After identifying this issue in 2006, the licensee observed that resolution of the non-conservative TS values for the EDGs steady state frequency and voltage were being pursued by other utilities due to receiving NCVs from the NRC. As a result, the licensee subsequently initiated discussions with the Pressurized Water Reactor Owners Group (PWROG) and decided to postpone submitting an LAR for the EDGs steady state frequency range until a resolution was developed by the industry. The licensee stated that they considered the non-conservative TS value for the EDGs steady state frequency range an industry-wide issue for which the PWROG would pursue a generic resolution for the PWR plants. However, after discussions with NRR, the inspectors concluded that the non-conservative TS value for the EDGs steady state frequency was to be treated on a plant specific basis, since no approved generic resolution presently exists.

The inspectors review of CAP 01090396, Inadequate EDG Surveillance Test Procedures, Revision 0, dated May 1, 2007, which was issued to address the non-conservative TS steady state voltage and initiate action to submit an LAR to revise the EDGs steady state voltage range, identified the due date to complete this action as June 5, 2013. The licensee stated that the PWROG met with the NRC several times to discuss a proposed resolution for the non-conservative TS value for the EDGs steady state voltage. During this inspection, the licensee stated they were in the final stages of submitting an LAR to correct the non-conservative TS value for the EDGs steady state voltage.

The licensee stated that they did not initiate an LAR to correct the non-conservative TS value for the EDGs steady state frequency because the resolution currently proposed by the PWROG did not specify a TS steady state frequency range. The inspectors concluded that the licensee was well aware that a condition adverse to quality, such as the non-conservative TS value for the EDGs steady state frequency, must be promptly corrected. Therefore, given the time period following discovery (i.e., September 8, 2006, until November 18, 2011) the licensee should have submitted an LAR to address the non-conservative TS value for the EDGs steady state frequency.

The NRC provided guidance for improper or inadequate TS in Administrative Letter 98-10. Imposing administrative controls in response to improper or inadequate TS was an acceptable short-term corrective action. However, the NRC expected that, following the imposition of administrative controls, an amendment to the TS would be submitted in a timely fashion. Therefore, the licensee should not have relied on the administrative controls limiting EDG operation to between 59.5 Hz and 60.5 Hz for a period of five years to address the non-conservative TS surveillance requirements in Section 3.8.1.

The licensee entered this issue into their corrective action program as CAP 01313783, EDG Loading Calculations Do Not Consider EDG Operation at 61.2 Hz, dated November 18, 2011, to document the inadequate calculations and CAP 01314190, Lack of Timely Response to Non-Conservative Technical Specification, dated November 21, 2011, in order to evaluate how to resolve the issue.

Analysis:

The inspectors determined that failure to correct the non-conservative TS surveillance requirements in Section 3.8.1 in a timely fashion was contrary to 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and was a performance deficiency. Specifically, the licensee failed to submit a license amendment to correct the non-conservative TS value for the EDGs allowable steady state frequency since entering the issue into their corrective action program on September 8, 2006.

The finding was more than minor because the finding was associated with the Mitigating Systems cornerstones attribute of Equipment Performance 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 licensee could not be assured that the design requirements for the EDGs system loads would operate within the appropriate design specifications if the EDGs were allowed to operate within the non-conservative TS allowable steady state frequency of 58.8 Hz and 61.2 Hz. As a result, the licensee established an administrative limit to limit operation of the EDGs to a frequency between 59.5 Hz and 60.5 Hz.

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, Characterization Worksheet for IE, MS, and BI Cornerstones. The inspectors determined that the cornerstone best reflecting the dominant risk was the Mitigating Systems cornerstone.

The inspectors confirmed that the finding did not result in a loss of operability or functionality per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment, because of the administrative procedures already in place (i.e., limiting operation of the EDGs between 59.5 Hz and 60.5 Hz).

Therefore, this finding was of very low safety significance (Green).

This finding has a cross-cutting aspect in the area of human performance, decision making because the licensee did not formally define the authority and roles for decisions affecting nuclear safety and as a result did not take the necessary steps to resolve a non-conservative TS in a time manner. Specifically, the licensee made a decision to delay resolving the non-conservative TS until the PWROG came up with a solution.

[H.1 (a)]

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that 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.

Contrary to the above, from September 8, 2006, until November 18, 2011, the licensee failed to promptly correct a condition adverse to quality. Specifically, the licensee failed to submit an LAR to correct the non-conservative TS surveillance requirements in Section 3.8.1 for the EDGs allowable steady state frequency.

Because this violation was of very low safety significance and it was entered into the licensees corrective action program as CAP 01313783 and CAP 01314190, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000282/2011012-02(DRS);05000306/2011012-02(DRS),

Failure to Correct a Condition Adverse to Quality).

4OA6 Meetings

.1 Exit Meeting Summary

On November 18, 2011, the inspectors presented the inspection results to Mr. Mark A. Schimmel 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

J. Anderson, Regulatory Affairs Manager

P. Anderson, Fleet Director of Licensing/Regulatory Affairs

M. Birkel, Compliance Engineer

J. Boesch, Plant Engineering Supervisor

J. Connors, Fleet Design Engineering Supervisor

S. DiPasquale, Licensing Engineer

S. Ford, Mechanical Design/Civil Supervisor

B. Horner, Reactor System Engineer

P. Huffman, Site Engineering Director

J. Lash, Nuclear Oversight Manager

S. McCall, Engineering Design Manager

S. Northhard, Plant Manager

L. Pfingsten, Administrative Support

M. Schimmel, Site Vice President

D. Vincent, Senior Project Manager

K. Vincent, Regulatory Programs Supervisor

B. Wegner, Mechanical Design/Civil Engineer

H. Wike, Electrical/I&C Design Supervisor

Nuclear Regulatory Commission

K. Stoedter, Senior Resident Inspector

P. Zurawski, Resident Inspector

Attachment

LIST OF ITEMS

Opened

05000282/2011012-01(DRS); NCV Flammable Gas Bottles Installed and/or Stored in

05000306/2011012-01(DRS) the Auxiliary Building (Section 1R17.2b)

05000282/2011012-02(DRS); NCV Failure to Correct a Condition Adverse to Quality

05000306/2011012-02(DRS) (Section 4OA2.1b)

Closed

05000282/2011012-01(DRS); NCV Flammable Gas Bottles Installed and/or Stored

05000306/2011012-01(DRS) in the Auxiliary Building (Section 1R17.2b)

05000282/2011012-02(DRS); NCV Failure to Correct a Condition Adverse to

05000306/2011012-02(DRS) Quality (Section 4OA2.1b)

Discussed

None.

Attachment