IR 05000306/2010012

IR 05000306-10-012; 11/01/10 - 11/05/10; Prairie Island Nuclear Generating Plant, Unit 2; Followup Inspection to Supplemental Inspection Report 05000306/2010009

ML103280151
Person / Time
Site:	Prairie Island
Issue date:	11/24/2010
From:	Jack Giessner Reactor Projects Region 3 Branch 4
To:	Schimmel M Northern States Power Co
References
EA-09-167 IR-10-012
Download: ML103280151 (16)

Text

ber 24, 2010

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 NRC SUPPLEMENTAL (95001) FOLLOWUP INSPECTION REPORT 05000306/2010012

Dear Mr. Schimmel:

On November 5, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed a followup inspection at your Prairie Island Nuclear Generating Plant, Unit 2. The enclosed report documents the inspection results which were discussed on November 5, 2010, with you and members of your staff.

Earlier this year, the NRC performed a supplemental inspection consistent with the NRC Action Matrix due to a White finding in the Mitigating System Cornerstone. Specifically, on September 3, 2009, the NRC issued its Final Significance Determination and a Notice of Violation (NRC Inspection Report 05000306/2009013) for a White finding that involved failures by your staff to design the component cooling water system such that it would be protected from the impact of high energy line break, seismic, or tornado events. Your staff informed the NRC on April 26, 2010, of your readiness for this inspection. During the time period of June 14 through September 7, 2010, the staff conducted inspection activities associated with this issue, but were unable to complete the inspection because the extent of condition review performed by your staff did not have sufficient breadth to identify conditions similar to the condition which resulted in the White Finding. After performing additional activities, your staff informed the NRC on October 6, 2010, of your readiness for this inspection.

This followup inspection utilized applicable sections of NRC Inspection Procedure 95001, Inspection for One or Two White Inputs in a Strategic Performance Area, and was conducted to evaluate actions you have taken to assure that that the root cause evaluation associated with the White finding appropriately addressed the extent of condition and extent of cause of the problem. Based on the results of this inspection, the NRC concluded that the extent of condition review performed by your staff had sufficient breadth to indentify conditions similar to the condition that led to the White finding. Based on the inspection activities documented in this Inspection Report and Inspection Report 05000306/2010009, the NRC will close the White Finding. Therefore, consistent with NRC Inspection Manual Chapter (IMC) 0305, Operating Reactor Assessment Program, based on successful completion of the supplemental inspection, Prairie Island Unit 2 has transitioned to the licensee response band (Column I). However, also consistent with IMC 0305, the finding will still be considered (counted toward future determination) in the Action Matrix for the remainder of the quarter (fourth quarter 2010).

Based on the results of this inspection, one NRC-identified Severity Level IV violation and associated Green finding was identified. However, because the finding was of very low safety significance, and entered into your corrective action program, the NRC is treating the issue as a non-cited violation (NCV) in accordance with Section 2.3.2 of the NRC Enforcement Policy, dated September 30, 2010. If you contest the subject or severity of the 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 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 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/

John B. Giessner, Chief Branch 4 Division of Reactor Projects Docket Nos. 50-306 License Nos. DPR-60

Enclosure:

Inspection Report 05000306/2010012 w/Attachment: Supplemental Information

REGION III==

Docket Nos: 50-306 License Nos: DPR-60 Report No: 05000306/2010012 Licensee: Northern States Power Company, Minnesota Facility: Prairie Island Nuclear Generating Plant, Unit 2 Location: Welch, MN Dates: November 1 to November 5, 2010 Inspector: C. Thomas, Senior Resident Inspector, Monticello Nuclear Generating Plant Approved by: J. Giessner, Branch Chief Branch 4 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

Inspection Report 05000306/2010012; 11/01/10 - 11/05/10; Prairie Island Nuclear Generating

Plant, Unit 2; Followup Inspection to Supplemental Inspection Report 05000306/2010009.

The report covers a follow-up inspection performed by the Monticello Nuclear Generating Plant Senior Resident Inspector. This inspection completes the inspection activities documented in Inspection Report 05000306/2010009 and closes the White finding which was documented in Inspection Report 05000306/2009013. The inspector identified one Severity IV violation and associated Finding of very low safety significance (Green). 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 are determined using IMC 0305, Operating Reactor Assessment Program. Findings for which the SDP does not apply may be Green or 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.

Cornerstone: Mitigating Systems

Green, SLIV A Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the inspector for the licensees failure to provide an evaluation that adequately documented why implementing new manual operator actions during periods of adverse weather, which isolated portions of the component cooling water system susceptible to hazards associated with tornado-generated missiles, did not present a more than minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component (SSC) important to safety previously evaluated in the updated safety analysis report (USAR). The licensee initiated CAP 1257118, 50.59 Screening Not Sufficient - 122 Spent Fuel Pool Heat Exchanger Component Cooling Loss, and, at the end of the inspection, was in the process of correcting the deficiency.

The violation was determined to be more than minor because the inspector could not reasonably determine that the changes would not have ultimately required prior NRC approval. Violations of 10 CFR 50.59 are dispositioned using Traditional Enforcement process instead of the SDP because they are considered to be violations that could potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspector determined that the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04,

Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspector answered Yes to Question 5 under the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspector concluded that the finding screened as potentially risk significant due to a severe weather initiating event. Based upon Phase 3 SDP evaluation performed by a NRC Region III Senior Risk Analyst (SRA), the inspector concluded that the issue was of very low safety significance (Green). The inspectors concluded that this finding was cross-cutting in the Problem Identification and Resolution area, corrective action component, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of conditions as necessary P.1(c).

(Section 3.02)

REPORT DETAILS

4OA4 Followup Inspection

.1 Background Information

The licensee entered the Regulatory Response Column of the NRCs Action Matrix in the third quarter of 2009 as a result of one inspection finding of low-to-moderate safety-significance (White). As of July 31, 2008, the licensee had failed to ensure that the design of the Unit 2 component cooling (CC) system would mitigate licensing basis events such as a High Energy Line Break (HELB), seismic, and tornado.

Specifically, a CC system pipe was in close proximity to the 15A and 15B feedwater heater high energy piping. Due to this proximity, a HELB could impact and cause failure of the CC system pipe which would cause a loss of CC system function in 6 minutes due to loss of CC system inventory. The NRC determined the delta core damage frequency (CDF) for this condition to be 3.2E-6/year and, therefore, the finding was of low-to-moderate safety-significance (White). On August 5, 2009, the NRC issued Inspection Report (IR) 05000282/2009010; 05000306/2009010, which documented a preliminary White finding associated with this issue. Subsequent to determining the Findings risk significance, on September 3, 2009, the NRC issued IR 05000306/2010013, which documented a Notice of Violation (NOV) and associated White finding.

The licensee informed the NRC staff that they were ready for the supplemental inspection on April 26, 2010. The licensee performed root cause evaluation (RCE) 01145695 to identify the direct causes, contributing causes, and other causal factors which allowed for the White finding. The RCE also reviewed organizational attributes that resulted in the White finding, including a review of safety culture.

On September 7, 2010, the NRC staff completed the initial supplemental inspection in accordance with Inspection Procedure (IP) 95001. This inspection assessed the licensees evaluation of the White finding which impacted the Mitigating Systems Cornerstone in the Reactor Safety strategic performance area. The inspection objectives were to:

• provide assurance that the root and contributing causes of risk-significant issues were understood;

• provide assurance that the extent of condition and extent of cause of risk-significant issues were identified; and

• provide assurance that the licensees corrective actions for risk-significant issues were sufficient or would be sufficient to address the root and contributing causes to preclude repetition.

The inspectors reviewed the licensees RCE, as well as other evaluations conducted in support and as a result of the RCE. The inspectors reviewed corrective actions that were taken or planned to address the identified causes. The inspectors also held discussions with licensee personnel to ensure that the root and contributing causes and the contribution of safety culture components were understood and corrective actions taken or planned were appropriate to address the causes and preclude repetition.

The inspection team was unable to complete the Supplemental Inspection because they concluded that the licensees extent of condition review did not have sufficient breadth to identify conditions similar to the condition which resulted in the White Finding. As a result of this condition, the NRC did not close the White Finding. Instead, the NRC informed the licensee that the Agency would re-perform those portions of the supplemental inspection subsequent to the completion of corrective actions to address concerns that were identified in the supplemental IR (IR 05000306/2010009).

The licensee informed the NRC staff that they were ready for the followup to the supplemental inspection on October 6, 2010.

2. Inspection Scope

Inspection Procedure 95001 requires that the inspection staff determine that the licensees RCE addresses the extent of condition and extent of cause of the issue (Section 02.02d of IP95001). Since the inspection team performing the supplemental inspection was unable to complete their inspection activities in this section due to the licensees extent of condition review not having sufficient breadth to identify conditions similar to the condition which resulted in the White Finding, the primary focus of the followup inspection was to assess the corrective actions implemented by the licensee to address this deficiency. To assess the issues identified in IR 05000306/2010009, the inspector reviewed information associated with the following five areas:

• extent of condition walkdowns and associated engineering evaluations of interactions between the CC system, auxiliary feedwater system, and high energy lines located in the auxiliary building;

• operability assessments and corrective action documents related to previously unidentified or unevaluated interactions that were identified as part of the extent of condition walkdowns;

• how HELB induced pressure and temperature effects were evaluated for the systems that were evaluated as part of the extent of condition;

• scope and timeline for the licensees broader HELB analysis reconstitution project; and

• Unresolved Item 05000282/2010009-02, Classification of CC Piping.

Specific documents reviewed as part of this inspection are listed Attachment 1 of this report.

3. Observations and Findings

3.01 Closure of VIO 05000306/2009013-01; Failure to Ensure Design Measures Were Appropriately Established for the Unit 2 Component Cooling Water System The supplemental inspection was completed on July 9, 2010, as documented in IR 05000306/2010009, and concluded that the corrective actions were adequate to address the causes that were identified in the evaluation so as to prevent recurrence.

However, the extent of condition performed to understand the breadth of potential interactions of the CC system and a HELB failed to validate that CC system piping in proximity to high energy lines would remain operable following a HELB.

The followup inspection was completed on November 5, 2010, and the results of that inspection are documented in this report. The primary purpose of the follow-up inspection was to determine if the licensee had addressed the deficiencies associated with their original extent of condition evaluation. The licensee implemented a systematic process to walk down, identify and evaluate CC piping that was in the vicinity of HELB piping to ensure the CC safety function was maintained following HELB events. In addition, the licensee evaluated another high risk system (auxiliary feedwater) to determine potential impacts from HELBs. The inspector concluded that the corrective actions taken by the licensee to strengthen their extent of condition evaluation were adequate to address the deficiencies that were identified in the supplemental inspection.

Together, these two inspections completed the Agencys assessment of the licensees evaluation of the White finding associated with VIO 05000306/2009013-01. The inspector concluded that the White finding could be closed. Therefore, consistent with NRC Inspection Manual Chapter (IMC) 0305, Operating Reactor Assessment Program, based on successful completion of the supplemental inspection, Prairie Island Unit 2 has transitioned to the licensee response band (Column I). However, also consistent with IMC 0305, the finding will still be considered (counted toward future determination) in the Action matrix for the remainder of the quarter (fourth quarter 2010).

3.02 Closure of Unresolved Item 05000282/2010009-02; Classification of Component Cooling Piping

Introduction:

A Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the inspector for the licensees failure to provide an evaluation that adequately documented why implementing new manual operator actions during periods of adverse weather, which isolated portions of the component cooling water system susceptible to hazards associated with tornado-generated missiles, did not present a more than minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the updated safety analysis report (USAR).

Description:

On May 22, 2009, the licensee approved an operability recommendation that addressed a condition where CC system piping to the 122 spent fuel pool (SFP)heat exchanger, aerated drain tank (ADT) evaporator, and 121/122 waste gas recombiners was routed through an area of the fuel handling building where the piping was not adequately protected against tornado-generated missile hazards. The licensee determined that the absence of missile protection constituted an unanalyzed condition that had the potential to result in the loss of CC system function. The operability recommendation established compensatory measures to isolate the 121/122 waste gas recombiner piping from the rest of the CC system, but to leave CC system water supply to the 122 SFP heat exchanger and ADT evaporator in service. In the event of severe weather, the operability recommendation identified a compensatory measure that relied on manual operator actions to isolate the remainder of the in-service CC system piping from the piping that was susceptible to tornado-generated missile hazards. The licensee originally implemented the compensatory actions as a temporary change to Procedure AB-2, Tornado/Severe Thunderstorm/High Winds. The licensee would later make the temporary procedure change permanent. On both occasions, the licensee performed an applicability screening in accordance with their 50.59 Resource Manual, FG-E-SE-03, but incorrectly concluded that an evaluation was not required for a change that involved a new or modified operator action that supports a design function that is credited in the USAR.

With respect to adverse effect, NEI 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1 states, in part, that the screening process is not concerned with the magnitude of adverse effects, any adverse effect would screen in, and its magnitude would be evaluated using the10 CFR 50.59 evaluation process.

NEI 96-07 was endorsed by the NRC in Regulatory Guide 1.187. The inspector noted that the need to isolate portions of the piping during certain weather conditions represented an adverse effect. Specifically, the need to isolate component cooling water piping during adverse weather to protect it from tornado-generated missiles is adverse compared to piping that is designed to survive impact by tornado-generated missiles.

Analysis:

The inspector determined that the licensees failure to adequately evaluate implementing new manual operator actions designed to, during periods of adverse weather, isolate portions of the CC system susceptible to hazards associated with tornado-generated missiles was contrary to 10 CFR 50.59(d)(1) and was a performance deficiency warranting a significance evaluation. The violation was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"

Appendix B, "Issue Screening," because the inspector could not reasonably determine that the changes would not have ultimately required prior NRC approval.

Violations of 10 CFR 50.59 are dispositioned using the Traditional Enforcement process instead of the SDP because they are considered to be violations that could potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspector determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 -

Initial Screening and Characterization of Findings, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspector answered "Yes" to Question 5 under the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspector concluded that the finding screened as potentially risk significant due to a severe weather initiating event. A NRC Region III SRA performed a Phase 3 SDP evaluation using information from NUREG/CR-4461, Tornado Climatology of the Contiguous US, Revision 2, and the standardized plant analysis risk (SPAR) model, Revision 8.15, for Prairie Island. The probability of a damaging tornado at the site was estimated to be less than 1E-4/year. Assuming that the tornado could cause a total loss of the CC system, the SPAR model was used to estimate a conditional core damage probability (CCDP). Combining the probability of a tornado event with the CCDP for a loss of CC system event resulted in a delta CDF estimate for the finding of less than 1.0E-6/year.,

which is a finding of very low safety significance (Green).

In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance (Green).

The inspector determined that the contributing cause that provided the most insight into the performance deficiency affected the cross-cutting area of Problem Identification and Resolution, having corrective action components, and involving aspects associated with the licensee thoroughly evaluating problems such that the resolutions address causes and extent of conditions as necessary. This included properly classifying, prioritizing, and evaluating for operability and reportability, conditions adverse to quality. P.1(c)

Enforcement:

Title 10 CFR 50.59, Changes, Tests, and Experiments, Section (d)(1)states, in part, that the licensee shall maintain records of changes in the facility or procedures, and that the records must include a written evaluation that provides the bases for the determination that the change does not require a license amendment pursuant to paragraph 10 CFR 50.59(c)(2). Contrary to this requirement, on May 21, 2009, and September 1, 2009, the licensee failed to provide an evaluation that adequately documented why implementing new manual operator actions during periods of adverse weather, which isolated portions of the component cooling water system susceptible to hazards associated with tornado-generated missiles, did not present a more than minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the USAR. In both cases, the licensee failed to include in the written evaluation a basis as to why the newly introduced manual actions would not increase the likelihood of a malfunction of equipment important to safety. In accordance with NRC Enforcement Policy, the violation was classified as a Severity Level IV violation because the underlying technical issue was of very low risk significance. Because this violation was of a very low safety-significance, was not repetitive or willful, and was entered into the licensees corrective action program as CAP 1257118, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000306/2010012-01; Failure to Adequately Evaluate New Manual Operator Actions)

In addition, the ROP finding of very low safety significance, Green, is dispostioned separately from the Traditional Enforcement violation and, therefore, the finding is being assigned a separate tracking number. Although there is an additional tracking number, the cross-cutting aspect is assigned only once. (FIN 05000306/2010012-02; Failure to Adequately Evaluate New Manual Operator Actions)

This Finding and NCV closes URI 05000282/2010009-02; Classification of CC Piping.

.1 Exit Meeting Summary

On November 5, 2010, the inspector presented the inspection results to Mr. Mark Schimmel, Site Vice President, and other members of the staff who acknowledged the results of the inspection and the violation of applicable regulatory requirements. The inspector confirmed that proprietary information was not provided or examined during this inspection.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

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M. Schimmel, Site Vice President

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K. Davidson, Plant Manager

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D. Roddey, Engineering Director

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D. Potter, Engineering Supervisor

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J. Anderson, Regulatory Affairs Manager

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B. Sawatzke, Site Director of Operations

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S. Dipasquale, Licensing

NRC

J. Giessner, Chief, Reactor Projects Branch 4

LIST OF ITEMS

Opened and Closed

05000306/2010012-01 NCV Inadequate 50.59 Evaluation for New Manual Operator Actions

05000306/2010012-02 FIN Inadequate 50.59 Evaluation for New Manual Operator Actions

Closed

05000306/2010013-01 VIO Failure to Ensure Design Measures Were Appropriately Established for the Unit 2 Component Cooling Water System

05000282/2010009-02 URI Classification of CC Piping Attachment