IR 05000263/2010002

IR 05000263-10-002, on 01/01/2010 - 03/31/2010, Monticello Nuclear Generating Plant, Surveillance Testing, Identification and Resolution of Problems

ML101160511
Person / Time
Site:	Monticello
Issue date:	04/26/2010
From:	Kenneth Riemer NRC/RGN-III/DRP/B2
To:	O'Connor T Northern States Power Co
References
IR-10-002
Download: ML101160511 (42)

Text

April 26, 2010

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANT NRC INTEGRATED AND POWER UPRATE REVIEW INSPECTION REPORT 05000263/2010002

Dear Mr. OConnor:

On March 31, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Monticello Nuclear Generating Plant. The enclosed report documents the inspection findings, which were discussed on April 8, 2010, 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 the issues as non-cited violations (NCVs) in accordance with Section VI.A.1 of the NRC Enforcement Policy. Additionally, a licensee-identified violation is listed in Section 4OA7 of this report.

If you contest the subject or severity of these NCVs, 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 Monticello Nuclear Generating Plant. 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 Monticello Nuclear Generating Plant. 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 and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Kenneth Riemer, Chief

Branch 2

Division of Reactor Projects

Docket No. 50-263; 072-058 License No. DPR-22

Enclosure:

Inspection Report 05000263/2010002 w/Attachment: Supplemental Information

REGION III==

Docket No:

50-263; 072-058 License No:

DPR-22 Report No:

05000263/2010002 Licensee:

Northern States Power Company, Minnesota Facility:

Monticello Nuclear Generating Plant Location:

Monticello, MN Dates:

January 1 through March 31, 2010 Inspectors:

S. Thomas, Senior Resident Inspector

L. Haeg, Resident Inspector

J. Bozga, Reactor Inspector S. Bakhsh, Reactor Inspector, DNMS M. Learn, Reactor Engineer, DNMS

Approved by:

Kenneth Riemer, Chief Branch 2 Division of Reactor Projects

Enclosure

SUMMARY OF FINDINGS

IR 05000263/2010002; 01/01/2010 - 03/31/2010; Monticello Nuclear Generating Plant;

Surveillance Testing; Identification and Resolution of Problems.

This report covers a three-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. The report covers a review of documentation and interviews with licensee staff regarding Unresolved Item (URI) 05000263/2008-005-01,

Non-Destructive Examination of Weld on the Outer Lid of Casks Performed Outside the Temperature Range Specified by the Applicable Welding Procedure, associated with Monticellos 2008 initial dry cask loading campaign. Two Green 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).

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: Barrier Integrity

Green.

The inspectors identified a finding of very low safety significance and NCV of Technical Specification 5.4.1 for the licensee failing to appropriately implement an applicable procedure recommended in Regulatory Guide 1.33, Revision 2, Appendix A,

February 1978. Specifically, when unexpected local alarms were received during the performance of the safety relief valve (SRV) low low set system quarterly test,

Instrument and Control (I&C) personnel elected to attempt to clear the alarms prior to notifying operations and without fully understanding which alarms were present. The surveillance procedure provided no guidance on how to clear the unexpected module trip alarms and relay energized lights. The licensee entered this issue into their corrective action program. The inspectors determined that the performance deficiency affected the cross-cutting area of Human Performance, having decision-making components, and involving aspects associated with using conservative assumptions in decision making. H.1(a)

The inspectors determined that the performance deficiency was more than minor and a finding because it was associated with the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations, using the Phase 1 Worksheet for the Barrier Integrity Cornerstone. Since the inspectors answered no to all four questions in the Containment Barrier column of the Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance.

(Section 1R22)

Green.

The inspectors identified a finding of very low safety significance and NCV of 10 CFR 50, Appendix B, Criterion XVI, for the licensees failure to adequately evaluate and take corrective actions for a condition adverse to quality. Specifically, the licensee failed to appropriately evaluate the implications of the unexpected trips of high/low pressure switches, PSHL-4065A and PSHL-4066A, during the January 28, 2009, performance of the SRV low low set system quarterly tests and implement appropriate corrective actions. The failure to adequately evaluate the unexpected trips and correct the condition adverse to quality directly contributed to a repeat occurrence and subsequent unplanned Technical Specification Action entry during the January 27, 2010, performance of the same surveillance test. The licensee entered the issue into their corrective action program. The inspectors determined that the performance deficiency affected the cross-cutting aspect in the area of Problem Identification and Resolution, having corrective action program components, and involving aspects associated with the licensee thoroughly evaluating problems such that the resolutions address causes and extent of conditions, as necessary. P.1(c)

The inspectors determined that the performance deficiency was more than minor and a finding because it was associated with the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations, using the Phase 1 Worksheet for the Barrier Integrity Cornerstone. Since the inspectors answered no to all four questions in the Containment Barrier column of the Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance.

(Section 4OA2)

Licensee-Identified Violations

Violations of very low safety significance that were identified by the licensee have been reviewed by inspectors. Corrective actions planned or taken by the licensee have been entered into the licensees corrective action program. These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Monticello operated at full power for most of the assessment period with the exception of brief downpower maneuvers to accomplish rod pattern adjustments and to conduct planned surveillance testing activities.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

.1 Readiness for Impending Adverse Weather Condition - Extreme Cold Conditions

a. Inspection Scope

Since extreme cold conditions were forecast in the vicinity of the facility for January 28 and 29, 2010, the inspectors reviewed the licensees overall preparations/protection for the expected weather conditions. On January 28, 2010, the inspectors walked down the plant heating boiler and safety-related systems in the intake structure because their functions could be affected or required as a result of the extreme cold conditions forecast for the facility. The inspectors observed insulation, heat trace circuits, space heater operation, and weatherized enclosures to ensure operability of affected systems. The inspectors reviewed licensee procedures and discussed potential compensatory measures with control room personnel. The inspectors focused on plant managements actions for implementing the stations procedures for ensuring adequate personnel for safe plant operation and emergency response would be available. Specific documents reviewed during this inspection are listed in the Attachment to this report.

This inspection constituted one readiness for impending adverse weather condition sample as defined in Inspection Procedure (IP) 71111.01-05.

b. Findings

No findings of significance were identified.

.2 External Flooding

a. Inspection Scope

The inspectors evaluated the design, material condition, and procedures for coping with the design basis probable maximum flood. The evaluation included a review to check for deviations from the descriptions provided in the Updated Safety Analysis Report (USAR) for features intended to mitigate the potential for flooding from external factors.

As part of this evaluation, the inspectors checked for obstructions that could prevent draining and determined whether barriers required to mitigate the flood were available to be utilized in accordance with the licensees implementing procedures should flooding conditions exist. Additionally, the inspectors performed a walkdown of the protected area to identify any modification to the site which would inhibit site drainage during a probable maximum precipitation event or allow water ingress past a barrier. The inspectors also reviewed the abnormal operating procedure (AOP) for mitigating the design basis flood to ensure it could be implemented as written. Specific documents reviewed during this inspection are listed in the Attachment to this report.

This inspection constituted one external flooding sample as defined in IP 71111.01-05.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant systems:

• A residual heat removal service water (RHRSW) system with B RHRSW out-of-service (OOS) for planned maintenance (PM);

• A control room ventilation (CRV)/control room emergency filtration (CREF)system with B CRV/CREF OOS for PM; and

• B residual heat removal (RHR) system during A RHR system surveillance testing.

The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system and; therefore, potentially increase risk. The inspectors reviewed applicable operating procedures; system diagrams; the USAR; Technical Specification (TS) requirements; outstanding work orders (WOs); condition reports; and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program (CAP) with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

These activities constituted three partial system walkdown samples as defined in IP 71111.04-05.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Routine Resident Inspector Tours

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

• Fire Zone 5A/5B (reactor building 1001 elevation south and north);

• Fire Zone 4-B (reactor building closed cooling water heat exchanger area);

• Fire Zone 1-C (reactor core isolation cooling (RCIC) room);

• Fire Zone 1-B (11 RHR and core spray pump room); and

• Fire Zone 26 (off-gas stack).

The inspectors reviewed areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and implemented adequate compensatory measures for OOS, degraded or inoperable fire protection equipment; systems; or features in accordance with the licensees fire plan. The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights; their potential to impact equipment which could initiate or mitigate a plant transient; or their impact on the plants ability to respond to a security event. Using the documents listed in the Attachment to this report, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees CAP.

These activities constituted five quarterly fire protection inspection samples as defined in IP 71111.05-05.

b. Findings

No findings of significance were identified.

.2 Annual Fire Protection Drill Observation

a. Inspection Scope

On March 7, 2010, the inspectors observed a fire brigade activation for a simulated fire in the cable spreading room. Based on this observation, the inspectors evaluated the readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee staff identified deficiencies; openly discussed them in a self-critical manner at the drill debrief; and took appropriate corrective actions. Specific attributes evaluated were:

(1) proper wearing of turnout gear and self-contained breathing apparatus;

(2) proper use and layout of fire hoses;

(3) employment of appropriate fire fighting techniques;

(4) sufficient firefighting equipment brought to the scene;

(5) effectiveness of fire brigade leader communications, command, and control;

(6) search for victims and for propagation of the fire into other plant areas;

(7) smoke removal operations;

(8) utilization of preplanned strategies;

(9) adherence to the preplanned drill scenario; and

(10) drill objectives. Documents reviewed are listed in the Attachment to this report.

These activities constituted one annual fire protection inspection sample as defined in IP 71111.05-05.

b. Findings

No findings of significance were identified.

1R06 Flooding

.1 Internal Flooding

a. Inspection Scope

The inspectors reviewed selected risk important plant design features and licensee procedures intended to protect the plant and its safety-related equipment from internal flooding events. The inspectors reviewed flood analyses and design documents, including the USAR, engineering calculations, and AOPs to identify licensee commitments. The specific documents reviewed are listed in the Attachment to this report. In addition, the inspectors reviewed licensee drawings to identify areas and equipment that may be affected by internal flooding caused by the failure or misalignment of nearby sources of water, such as the fire suppression or the circulating water systems. The inspectors also reviewed the licensees corrective action documents with respect to past flood-related items identified in the CAP to verify the adequacy of the corrective actions. The inspectors performed a walkdown of the following plant area to assess the adequacy of watertight doors and verify drains and sumps were clear of debris and were operable, and that the licensee complied with its commitments:

• 911 elevation of turbine building including Division I 4160 V switchgear area.

This inspection constituted one internal flooding sample as defined in IP 71111.06-05.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

.1 Resident Inspector Quarterly Review

a. Inspection Scope

On March 8, 2010, the inspectors observed a crew of licensed operators in the plants simulator during licensed operator requalification examinations to verify that operator performance was adequate; evaluators were identifying and documenting crew performance problems; and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• licensed operator performance;

• crews clarity and formality of communications;

• ability to take timely actions in the conservative direction;

• prioritization, interpretation, and verification of annunciator alarms;

• correct use and implementation of abnormal and emergency procedures;

• control board manipulations;

• oversight and direction from supervisors; and

• ability to identify and implement appropriate TS actions and emergency plan actions and notifications.

The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one quarterly licensed operator requalification program sample as defined in IP 71111.11.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

.1 Routine Quarterly Evaluations

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk-significant system:

• 1AR transformer exceeded unavailability goals in (a)(1) action plan.

The inspectors reviewed specific events, such as where ineffective equipment maintenance had resulted in valid or invalid automatic actuations of engineered safeguards systems, and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• implementing appropriate work practices;

• identifying and addressing common cause failures;

• scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;

• characterizing system reliability issues for performance;

• charging unavailability for performance;

• trending key parameters for condition monitoring;

• ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and

• verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2) or appropriate and adequate goals and corrective actions for systems classified as (a)(1).

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

In addition to the sample listed above, the inspectors performed a review of the licensees (a)(3) Periodic Evaluation (April 2008 to March 2009). The inspectors reviewed the Periodic Evaluation to ensure that the licensee had performed the following:

• a Periodic Evaluation within the time restraints of the Maintenance Rule;

• reviewed (a)(1) goals, (a)(2) performance criteria, monitoring, and PM activities, and effectiveness of corrective actions;

• included industry operating experience (OE) where practicable; and

• made appropriate adjustments as of result of Periodic Evaluation.

These inspections constituted two quarterly maintenance effectiveness samples as defined in IP 71111.12-05.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

.1 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the licensee's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

• stator water cooling pressure control valve adjustment;

• V-AC-1 cooling water leak in condenser room;

• D moisture separator drain tank drain valve not controlling level;

• recirculation riser differential pressure switch (DPIS-2-129D) unable to reset; and

• standby liquid control (SBLC) system tank flaws.

These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work; discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor; and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems; when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Documents reviewed are listed in the Attachment to this report.

These maintenance risk assessments and emergent work control activities constituted five samples as defined in IP 71111.13-05.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

.1 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• non-safety-related pressure gauges installed in safety-related RHRSW system;

• refurbished fuel pool and plenum exhaust radiation monitors not like-for-like;

• isolation of feedwater crack at feedwater pumps may be too quick;

• flaws identified on base of SBLC system tank;

• degraded 13 RHRSW pump performance; and

• high pressure coolant injection (HPCI) discharge line air void.

The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available, such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and USAR to the licensees evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined; where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the to this report.

These operability inspections constituted six samples as defined in IP 71111.15-05.

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

.1 Permanent Plant Modifications

a. Inspection Scope

The following engineering design packages were reviewed and selected aspects were discussed with engineering personnel:

• Engineering Change (EC) 9561; Instrument/Service Air System Upgrade, Revision 2; and

• EC 13078; GL [Generic Letter] 2008-01 Generic Vent Line Modification.

These documents and related documentation were reviewed for adequacy of the associated 10 CFR 50.59 safety evaluation screening; consideration of design parameters; implementation of the modification; post-modification testing; and relevant procedures, design, and licensing documents were properly updated. The inspectors observed ongoing and completed work activities; as applicable, to verify that installation was consistent with the design control documents. These modifications replaced the existing station air compressors, receivers, and dryers to resolve longstanding capacity and reliability issues and developed a modification to be used for the installation of high point vents to address issues identified during the licensees evaluation of GL 2008-01.

Documents reviewed in the course of these inspections are listed in the Attachment to this report.

This inspection constituted two permanent plant modification samples as defined in IP 71111.18-05.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

.1 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the following post-maintenance (PM) activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• paralleling of 1AR transformer to buses 15 and 16 following resolution of load-tap changer instrumentation issues;

• testing of Division II 250 Vdc battery room exhaust fan following inspection and maintenance;

• testing of RCIC following valve maintenance;

• testing of 11 control rod drive (CRD) pump following bearing replacement;

• testing of MO-2009 following actuator repair; and

• testing of AO-2945 following maintenance and adjustment.

These activities were selected based upon the SSCs ability to impact risk. The inspectors evaluated these activities for the following (as applicable): the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion); and test documentation was properly evaluated. The inspectors evaluated the activities against TS, the USAR, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with PM tests to determine whether the licensee was identifying problems and entering them in the CAP and that the problems were being corrected commensurate with their importance to safety.

Documents reviewed are listed in the Attachment to this report.

These inspections constituted six PM testing samples as defined in IP 71111.19-05.

b. Findings

No findings of significance were identified.

1R20 Outage Activities

.1 Refueling Outage Activities

a. Inspection Scope

From January 25, 2010, through March 25, 2010, the inspectors reviewed the licensees control of heavy loads program in conjunction with the NRCs Operating Experience Smart Sample (OpESS) FY2007-03, Revision 2, Crane and Heavy Lift Inspection, Supplemental Guidance for IP-71111.20, specifically related to the removal and installation of the reactor pressure vessel (RPV) head during refueling outages.

Documents reviewed during the inspection are listed in the Attachment to this report.

This review did not constitute a baseline inspection sample.

The inspectors performed the following activities during the inspection:

• Reviewed licensees submittals and commitments related to GLs80-113 and 81-07, Control of Heavy Loads.

• Reviewed licensees design calculations for reactor building crane and crane support structure.

• Reviewed licensees preventative maintenance program procedures of rigging and special lifting devices used to remove and install the RPV head during refueling operations.

• Reviewed licensees procedures that control RPV head safe load path to remove and install the RPV head during refueling operations.

• Reviewed licensees procedures that provide training and qualification of reactor building crane operators.

• Reviewed a licensees sample of records of RPV Head Strongback and Dryer and Steam Separator Sling Lifting Device inspections completed prior to reactor disassembly and RPV head lift.

• Reviewed licensees procedures that control the total weight lifted by the reactor building crane to remove and install the RPV head during refueling operations and the reactor building crane rated lift capacity.

• Reviewed licensees reactor building crane inspection, maintenance and testing procedure.

b. Findings

(1) Reactor Building Crane Design and Licensing Basis Issues Introduction The inspectors determined that an unresolved item (URI) existed concerning the licensing and design basis of the reactor building crane and reactor building crane support structure.

Description The inspectors reviewed the following licensing documents for the reactor building crane:

• NRC letter to Northern States Power (NSP), Safety Evaluation by the Office of Nuclear Reactor Regulation [NRR] Supporting Approval of Crane Modification and Use of 70 Ton Spent Fuel Shipping Cask IF-300, dated May 19, 1977;

• NSP letter to NRC, Response to Request for Additional Information, dated February 28, 1977; and

• USAR, Section 10.2, page 4 of 24, and Section 12.2 page 28 of 49, and page 29 of 49, Revision 23 and Revision 25.

The NRC letter to NSP, Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Approval of Crane Modification and Use of 70 Ton Spent Fuel Shipping Cask IF-300, dated May 19, 1977, established the reactor building overhead crane capacity as a maximum of 85 tons and the crane seismic analysis did not analyze for a maximum 85 ton lifted load concurrent with a seismic event based on extremely low probability of both events occurring simultaneously. The licensee subsequently changed the reactor building crane capacity from 85 tons in USAR, Section 10.2, page 4 of 24, and Section 12.2, page 28 of 49, and page 29 of 49, Revision 23 to a crane capacity of 105 tons in USAR Section 10.2, page 4 of 24 and Section 12.2, page 28 of 49 and page 29 of 49, Revision 25.

The inspectors noted that the licensee did not perform a written 10 CFR 50.59 evaluation to assess the following: 1) whether the change of increasing design loads on the crane and the crane support structure required a license amendment and, 2) probabilistic analysis with consideration for a new maximum crane lifted load of 105 tons that evaluates whether or not a lifted load must be considered during a seismic event for the design of the reactor building crane and crane support structure.

The inspectors reviewed Calculation Nos. CA 76 138, Structural Requalification for New 85 Ton Crane, Revision 0; CA-05-103, Reactor Building Superstructure Seismic Response Analysis with 105 Ton Crane, Revision 0A; and CA-05-107, Structural Seismic Qualification Reactor Building Crane Upgrade for ISFSI, Revision 0B. The inspectors were concerned that the reactor building crane and reactor building crane support structure had been evaluated using friction in a linear elastic analysis to reduce seismic load effects applied to the reactor building crane and crane support structure. The licensee used much smaller seismic loads limited by the friction force and this resulted in a significant load reduction for qualification of the reactor building crane and reactor building crane support structure. In addition, the non-linear effects of friction have not been addressed in the aforementioned calculations.

The licensee was unable to provide evidence that the NRR staff had approved friction in a linear elastic analysis as a method of evaluation for this application. The use of friction to reduce seismic load effects on the reactor building crane and reactor building crane support structure was not discussed in the USAR.

The inspectors reviewed Calculation No. CA-05-101, Evaluation of Reactor Steel Superstructure for 105 Ton Reactor Building Crane, Revision 3A. The inspectors were concerned with the following: 1) The minimum yield strength for the American Society of Testing and Materials (ASTM) A1 trolley and the bridge rail has not been established in accordance with the American Institute of Steel Construction (AISC) code and, 2) The restraint mechanism in the longitudinal direction of the trolley rail and bridge rail connection was based on the use of friction resistance between the bottom of the rail and the supporting beam to resist the sliding of the rail during a design and licensing basis event.

In response to the concern, the licensee initiated corrective action program documents CAP 01214808, RB Crane Seismic Calc may not be Consistent W/License Basis, dated January 22, 2010 and CAP 01222530, Crane Heavy Lift Inspection URI:

10 CFR 50.59 for Crane Upgrade, dated March 13, 2010. Near the end of the inspection period, the licensee provided the inspectors additional information relevant to the design basis and licensing basis of the reactor building crane and reactor building crane support structure which will require additional review. Therefore, this issue is considered an unresolved item (URI 05000263/2010002-01, Reactor Building Crane Design and Licensing Basis Issues) pending additional inspector review to determine design and licensing basis requirements.

1R22 Surveillance Testing

.1 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements:

• 0187-02; 12 emergency diesel generator(EDG)/12 emergency service water (ESW) quarterly pump and valve tests; Revision 70 (routine);

• 0397; safety relief valve (SRV) low-low set system quarterly tests; Revision 15 (routine);

• 0051; main steam line high flow Group I instrument test; Revision 26 (routine);

• 0255-03-IA-1-2; core spray loop B quarterly pump and valve tests (in-service test (IST));

• 0391; RHR shutdown cooling supply; Revision 19 (containment isolation valve (CIV)); and

• 0533; containment sump flow measurement instrumentation; Revision 16 (reactor coolant system (RCS) leakage detection).

The inspectors observed in-plant activities and reviewed procedures and associated records to determine the following:

• did preconditioning occur;

• were the effects of the testing adequately addressed by control room personnel or engineers prior to the commencement of the testing;

• were acceptance criteria clearly stated, demonstrated operational readiness, and consistent with the system design basis;

• plant equipment calibration was correct, accurate, and properly documented;

• as-left setpoints were within required ranges; and the calibration frequency were in accordance with TSs, the USAR, procedures, and applicable commitments;

• measuring and test equipment calibration was current;

• test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied;

• test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures, jumpers and lifted leads were controlled and restored where used;

• test data and results were accurate, complete, within limits, and valid;

• test equipment was removed after testing;

• where applicable for IST activities, testing was performed in accordance with the applicable version of Section XI, American Society of Mechanical Engineers (ASME) code, and reference values were consistent with the system design basis;

• where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable;

• where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure;

• where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished;

• prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test;

• equipment was returned to a position or status required to support the performance of its safety functions; and

• all problems identified during the testing were appropriately documented and dispositioned in the CAP.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted three routine surveillance testing samples; one IST sample; one RCS leak detection inspection sample; and one CIV sample as defined in IP 71111.22, Sections -02 and -05.

b. Findings

Introduction A finding of very low safety significance and non-cited violation (NCV) of TS 5.4.1 was identified by the inspectors for the licensee failing to appropriately implement an applicable procedure recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Specifically, when unexpected local alarms were received during the performance of the SRV low low set system quarterly test, Instrument and Control (I&C)personnel elected to attempt to clear the alarms prior to notifying operations and without fully understanding which alarms were present. The surveillance procedure provided no guidance on how to clear the unexpected module trip alarms and relay energized lights.

Description On January 27, 2010, the inspectors observed the performance of 0397-A, SRV Low-Low Set System Quarterly Tests, Revision 15. The first half of the surveillance is performed at the SRV low low set Division I control panel located in the cable spreading room. The purpose of this surveillance test was to perform a channel functional test of the low-low set pressure switches; tailpipe pressure switch; and inhibit timers that were associated with the low-low set instrumentation. Step 124 of Procedure 0397-A states, IF reactor vessel is pressurized, THEN verify differential voltage between jack J-1 on PSHL-4064A and jack J-1 on PSHL-4064C is between -0.066 to 0.066 Vdc. Using a digital voltmeter, the I&C technician performed Step 124. During the performance of Step 124, the inspectors observed the PSHL-4065A (opens SRV G at 1062 psig if a scram has occurred) and PSHL-4066A (opens SRV H at 1052 psig if a scram has occurred) red trip LEDs and the associated amber K11A and K17A relay energized lights were illuminated (on the C-253A, SRV low low set Division I control panel). Since it was not apparent to the inspectors that the I&C technicians had observed the illumination of the red LEDs and amber lights, the inspectors asked the I&C technicians if the indications were expected. Since the technicians did not understand why the lights illuminated and the Procedure 0397-A did not provide guidance on how to proceed, the I&C technicians contacted their supervisor.

The inspectors observed the I&C supervisor enter the cable spreading room. Without verifying the actual indications present on the SRV low low set Division I control panel, he informed the I&C technicians that he had seen this happen before and that it was due to static discharge from the digital voltmeter. The inspectors asked the I&C supervisor if this was a known phenomena, why the 0397-A procedure did not address it.

Additionally, the inspectors asked if the issue had previously been entered into the CAP.

The supervisor could not answer either of the inspectors questions. Without first verifying which alarm lights were illuminated or notifying Operations of the unexpected indications, and without procedural guidance to do so, the I&C supervisor directed the technician to attempt to clear the alarms by depressing the gross fail trip reset buttons on the PSHL-4065A and PSHL-4066A trip modules. After this action did not clear the existing alarms, and subsequently realizing the gross fail red LEDs were not illuminated even before their reset buttons were depressed, the I&C supervisor notified the duty shift manager of the issue.

Once the duty shift manager was apprised of the situation, performance of the 0397-A procedure was stopped pending further evaluation of the event and until the existing configuration of the Division I SRV low low set trip logic was evaluated.

The trips were eventually reset utilizing a calibration unit in accordance with Procedure B.03.03-05, Resetting SRV Low Low Set System High/Low Pressure Switch (PSHL) Trip Units. Since turning on the power to any calibration unit on Panel C-253A or C-253B and selecting a master trip unit will block out the pressure transmitter input to the selected master trip unit and its associated slave trip units, this process of resetting the trips resulted in an unplanned entry in to TS 3.3.6.3.A.

The licensee entered this issue into their corrective action as CAP 1215801 (Inappropriate Direction Provided for Unexpected Indication).

Analysis The inspectors determined that the failure to operate TS-related equipment in accordance with approved procedures was a performance deficiency warranting significance evaluation. The inspectors concluded that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued December 24, 2009. The finding was more that minor because it was associated with the procedure adherence attribute of the Barrier Integrity (BI) Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that the performance deficiency affected the cross-cutting area of Human Performance, having decision-making components and involving aspects associated with making safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain plant conditions, to ensure safety is maintained. H.1(a)

The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations, using the Phase 1 Worksheet for the Barrier Integrity Cornerstone. Since the inspectors answered 'no' to all four questions in the Containment Barrier column of the Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance.

Enforcement Technical Specification 5.4.1 requires, in part, that written procedures be implemented covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Contrary to this requirement, Procedure 0397-A, SRV Low-Low Set System Quarterly Tests, Revision 15, was not appropriately implemented. Specifically, when unexpected local alarms were received during the performance of the 0397-A procedure, I&C personnel elected to attempt to clear the alarms prior to notifying operations and without fully understanding which alarms were present. The surveillance procedure provided no guidance on how to clear the unexpected module trip alarms and relay energized lights. Because the event was of very low safety significance and because the issue was entered into the licensees corrective action program (CAP 1215801), this violation is being treated as an NCV, consistent with Section VI.A.1 of the Enforcement Policy. (NCV 05000263/2010002-02, SRV Low Low Set Surveillance Procedure Implementation)

1EP6 Drill Evaluation

.1 Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of a routine licensee emergency drill on February 10, 2010, to identify any weaknesses and deficiencies in classification; notification; and protective action recommendation development activities.

The inspectors observed emergency response operations in the simulator control room; technical support center; and emergency offsite facility to determine whether the event classification; notifications; and protective action recommendations were performed in accordance with procedures. The inspectors also attended the licensee drill critique to compare any inspector-observed weakness with those identified by the licensee staff in order to evaluate the critique and to verify whether the licensee staff was properly identifying weaknesses and entering them into the CAP. As part of the inspection, the inspectors reviewed the drill package and other documents listed in the Attachment to this report.

This emergency preparedness drill inspection constituted one sample as defined in IP 71114.06-05.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

.1 Unplanned Scrams per 7000 Critical Hours

a. Inspection Scope

The inspectors sampled licensee submittals for the Unplanned Scrams per 7000 Critical Hours performance indicator (PI) for the period from the 1st Quarter 2009 through the 4th Quarter 2009. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the Nuclear Energy Institute (NEI)

Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, were used. The inspectors reviewed the licensees operator narrative logs, issue reports, event reports and NRC Inspection Reports for this period to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified.

This inspection constituted one unplanned scrams per 7000 critical hours sample as defined in IP 71151-05.

b. Findings

No findings of significance were identified.

.2 Unplanned Scrams with Complications

a. Inspection Scope

The inspectors sampled licensee submittals for the Unplanned Scrams with Complications PI for the period from the 1st Quarter 2009 through the 4th Quarter 2009.

To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, were used. The inspectors reviewed the licensees operator narrative logs, issue reports, event reports and NRC Integrated Inspection Reports for this period to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified.

This inspection constituted one unplanned scrams with complications sample as defined in IP 71151-05.

b. Findings

No findings of significance were identified.

.3 Unplanned Transients per 7000 Critical Hours

a. Inspection Scope

The inspectors sampled licensee submittals for the Unplanned Transients per 7000 Critical Hours PI for the period from the 1st Quarter 2009 through the 4th Quarter 2009. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, were used.

The inspectors reviewed the licensees operator narrative logs, issue reports, maintenance rule records, event reports and NRC Integrated Inspection Reports for this period to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified.

This inspection constituted one unplanned transients per 7000 critical hours sample as defined in IP 71151-05.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical Protection

.1 Routine Review of Items Entered into the Corrective Action Program

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees CAP at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Attributes reviewed included: the complete and accurate identification of the problem; that timeliness was commensurate with the safety significance; that evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent-of-condition reviews, and previous occurrences reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue.

Minor issues entered into the licensees CAP as a result of the inspectors observations are included in the Attachment to this report.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

No findings of significance were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees CAP. This review was accomplished through inspection of the stations daily condition report packages.

These daily reviews were performed by procedure as part of the inspectors daily plant status monitoring activities and; as such, did not constitute any separate inspection samples.

b. Findings

No findings of significance were identified.

.3 Annual Sample:

Review of Operator Workarounds

a. Inspection Scope

The inspectors evaluated the licensees implementation of their process used to identify, document, track, and resolve operational challenges. Inspection activities included, but were not limited to, a review of the cumulative effects of the operator workarounds (OWAs) on system availability and the potential for improper operation of the system, for potential impacts on multiple systems, and on the ability of operators to respond to plant transients or accidents.

The inspectors performed a review of the cumulative effects of OWAs. The documents listed in the Attachment to this report were reviewed to accomplish the objectives of the inspection procedure. The inspectors reviewed both current and historical operational challenge records to determine whether the licensee was identifying operator challenges at an appropriate threshold, had entered them into their CAP and proposed or implemented appropriate and timely corrective actions which addressed each issue.

Reviews were conducted to determine if any operator challenge could increase the possibility of an Initiating Event, if the challenge was contrary to training, required a change from long-standing operational practices, or created the potential for inappropriate compensatory actions. Additionally, all temporary modifications were reviewed to identify any potential effect on the functionality of Mitigating Systems, impaired access to equipment, or required equipment uses for which the equipment was not designed. Daily plant and equipment status logs, degraded instrument logs, and operator aids or tools being used to compensate for material deficiencies were also assessed to identify any potential sources of unidentified OWAs.

This review constituted one OWA annual inspection sample as defined in IP 71152-05.

b. Findings

No findings of significance were identified.

.4 Selected Issue Follow-Up Inspection:

Inadequate Corrective Actions for Unexpected SRV Low Low Set Trips Encountered during Surveillance Testing

a. Inspection Scope

During a review of a current issue associated with unexpected Rosemount trip unit trips encountered during SRV low low system set quarterly surveillance testing (see Section 1R22 of this report), the inspectors discovered that an almost identical testing anomaly had occurred during the performance of the same surveillance procedure in January 2009. The inspectors evaluated the licensees response to the January 2009 event.

This review constituted one in-depth problem identification and resolution sample as defined in IP 71152-05.

b. Findings

Introduction The inspectors identified a finding of very low significance and NCV of 10 CFR 50, Appendix B, Criterion XVI, for the licensees failure to adequately evaluate and take corrective actions for a condition adverse to quality (CAQ). Specifically, the licensee failed to appropriately evaluate the implications of the unexpected trips of PSHL-4065A and PSHL-4066A during the January 28, 2009, performance of the SRV Low Low Set System Quarterly Tests and implement appropriate corrective actions. The failure to adequately evaluate the unexpected trips and correct the CAQ directly contributed to a repeat occurrence and subsequent unplanned TS action entry during the January 27, 2010, performance of the same surveillance test.

Description On January 28, 2009, during the performance of Procedure 0397-A, SRV Low-Low Set System Quarterly Test, Revision 15, Step 124, IF reactor vessel is pressurized, THEN verify differential voltage between jack J-1 on PSHL-4064A and jack J-1 on PSHL-4064C is between -0.066 to 0.066 Vdc, the licensee received unexpected trips of PSHL-4065A and PSHL-4064A when a digital voltage meter (DVM) lead was inserted in the J-1 jack on PSHL-4064A. Since Procedure 0397-A has no procedural guidance for clearing these trips, Operations Manual Procedure B.03.03-05.G.3, Resetting SRV Low Low Set System High/Low Pressure Switch (PSHL) Trip Units, was utilized to clear the two trips. The 0397-A procedure was recommenced and completed with no further issues. The licensee wrote CAP 01167144, Unexpected Trip of SRV Lo-Lo Slave Units for PSHL-4064 to document the issue. The following was documented in the Notes section of the CAP:

Immediate Action Taken:

Stopped and informed I&C and Operations supervision. Reset tripped units per operations procedure.

Recommendations:

Close to actions taken and for trending purposes.

Why Did This Occur:

Unknown, potentially a static charge on the DVM or the technician inserting the probe caused a momentary spike. Once the units were reset, Step 124 was completed with no problems noted.

The technicians did discharge any static electricity prior to inserting the probes on the second attempt at Step 124.

General Notes:

Close to actions taken.

On January 27, 2010, during the performance of the same (0397-A) procedure approximately one year later, PSHL-4064A and PSHL-4065A unexpectedly tripped during the performance of Step 124.

Criterion XVI of 10 CFR 50, Appendix B, states, in part, that measures shall be established to assure that CAQ, such as malfunctions, are promptly identified and corrected. The inspectors identified that, although the licensee documented the CAQ in their CAP (that the unexpected trips occurred), the licensee failed to appropriately evaluate the cause of the issue and implement appropriate corrective actions. The inspectors noted the following deficiencies in the licensees action to address this issue:

• The Licensee did not Adequately Evaluate the Technical Specification Implications of the Unexpected Trips of PSHL-4065A and PSHL-4066A during the January 28, 2009, Performance of the SRV Low Low Set System Quarterly Tests.

Technical Specification 3.3.6.3, Low Low Set Instrumentation, Note 2, states:

When a channel is placed in an inoperable status solely for the performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains LLS initiation capability. Surveillance Procedure 0397-A provides no guidance on how to clear/reset unexpected Rosemount trip module trips.

This necessitates using a different procedure, Operations Manual Procedure B.03.03-05.G.3, Resetting SRV Low Low Set System High/Low Pressure Switch (PSHL) Trip Units, to clear the unexpected trips.

Precaution and Limitation 1 of this procedure states, in part, that use of the calibration unit causes one channel of one SRV low low set trip system to be inoperable and that per TS 3.3.6.3, loss of one channel makes the trip system inoperable. By procedure, TS 3.3.6.3.A should have been entered, and was not, when the performance of B.03.03-05.G.3 was required to reset the unexpected trips on PSHL-4065A and PSHL-4064A.

• The Licensee did not Adequately Identify the CAQ Documented in CAP 1167144 (Unexpected Trip of SRV Lo-Lo Slave Units for PSHL-4064).

The licensee utilizes, in part, Procedure FP-PA-ARP-01, CAP Action Request Process, Revision 24, to implement their CAP. This procedure defines a CAQ, in part, as failures; malfunctions; deficiencies; deviations; defective material and equipment and non-conformances that affect or have a reasonable potential to affect the operability or functionality of critical SSCs. Also, each CAQ is addressed in the CAP as a level A, B, or C issue.

1 to FP-PA-ARP-01 is used by the licensee as a screening tool for evaluating the severity level of issues entered into the CAP. Category 14 (Plant Operation & Equipment Related) of Attachment 1 lists unexpected equipment response as a Severity Level B issue.

7 of FP-PA-ARP-01 documents the CAP Screening Charter.

Listed in this charter are the functions of the CAP Screening Team.

Listed among these duties are determining if the issue is a condition adverse to quality (SCAQ, CAQ, CNAQ), determining the issues Severity Level (A, B, C),and to assign default evaluations (RCE, ACE, or CE) that are required for CAPs.

The Charter also directs that for CAPs closed to trend or actions taken, the Screening Team will ensure that sufficient information exists in the CAP to serve as an adequate historical record (e.g., CAP clearly describes the actions that corrected the condition).

The licensees CAP screening team failed to identify that the CAQ documented in CAP 1167144 was associated with what caused the unexpected trips to occur and that simply resetting the trips and closing the CAP to actions taken was insufficient to correct the documented CAQ.

• The Licensee did not correct the Condition Adverse to Quality Identified in CAP 1167144 prior to Closing the CAP.

The inspectors determined that the CAQ was that unexpected trips occurred on equipment required by TSs during the performance of required surveillance testing. The inspectors noted that the focus of the licensee, as documented in CAP 1167144, was to clear the unexpected trips, not to understand what precipitated them. Even though the CAP notes stated that the answer to the question, Why did this occur? was Unknown, the Operations (OPS) Status entry under the CAP Attribute notes stated, condition resolved using operations manual procedure and surveillance testing completed satisfactory. Without any documented formal evaluation of why the trips occurred, the licensee closed CAP 1167144 to Actions Taken.

The licensee entered this issue into their corrective action program as CAP 1215539.

Analysis The inspectors determined that the failure to adequately evaluate the implications of the unexpected trips of PSHL-4065A and PSHL-4066A during the January 28, 2009, performance of the SRV low low set system quarterly tests and implement appropriate corrective actions was a performance deficiency warranting significance evaluation.

The inspectors concluded that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued December 12, 2009. The finding was more that minor because it was associated with configuration control attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. The inspectors determined that the performance deficiency affected the cross-cutting aspect area of Problem Identification and Resolution, having corrective action program components, and involving aspects associated with the licensee thoroughly evaluating problems such that the resolutions address causes and extent of conditions, as necessary. P.1(c)

The inspectors evaluated the finding using IMC 0609, Appendix A, Attachment 1, Significance Determination of Reactor Inspection Findings for At-Power Situations, using the Phase 1 Worksheet for the Barrier Integrity Cornerstone. Since the inspectors answered no to all four questions in the Containment Barrier column of the Characterization Worksheet for Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones, the inspectors concluded that the finding was of very low safety significance.

Enforcement Title 10 CFR 50, Appendix B, Criteria XVI, Corrective Action, requires, in part, that measures be established to assure that conditions adverse to quality, such as malfunctions, are promptly identified and corrected. Contrary to the above, the licensee failed to appropriately evaluate the implications of the unexpected trips of PSHL-4065A and PSHL-4066A which occurred during the January 28, 2009, performance of the SRV low low set system quarterly tests and implement appropriate corrective actions.

The failure to adequately evaluate the unexpected trips and correct the CAQ directly contributed to a repeat occurrence and subsequent unplanned TS Action entry during the January 27, 2010, performance of the same surveillance test. Because the event was of very low safety significance and because the issue was entered into the licensees corrective action program (CAP 1215539), this violation is being treated as an NCV, consistent with Section VI.A.1 of the Enforcement Policy.

(NCV 05000263/2010002-03, Inadequate Corrective Actions for Unexpected SRV Low Low Set Trips Encountered during Surveillance Testing)

4OA3 Follow-Up of Events and Notices of Enforcement Discretion

.1 Infrequently Performed Test and Evolution:

SHERCO 3 Generator Maximum MVA Test Plan

a. Inspection Scope

The inspectors evaluated the licensees preparation for and performance during the SHERCO 3 Generator Max MVA test. This test was of specific interest to the inspectors because, although the test was being conducted at a large coal fueled generating facility near the Monticello plant, it had the potential to result in significant grid voltage perturbations. Additionally, because the test required frequent manipulation of the MVARS each of the three SHERCO Units and the Monticello plant was sending/receiving, close coordination was required between the SHERCO and Monticello control rooms was required. The inspectors reviewed the associated test plan, the licensees evaluation of the tests potential risk to the Monticello plant, and observed in the control room licensee performance during the portion of the test which established generator/grid conditions that supported the SHERCO 3 generator maximum MVA test.

This event follow-up review constituted one sample as defined in IP 71153-05.

b. Findings

No findings of significance were identified.

4OA5 Other Activities

.1 Preoperational Testing of an Independent Spent Fuel Storage Installation

a. Inspection Scope

(1) Training The inspectors reviewed training for the staff involved in Independent Spent Fuel Storage Installation (ISFSI) activities. This included a review of the licensees Systematic Approach to the Training program for this area which consisted of classroom and on-the-job training. The inspectors reviewed the training material, including the content of the manuals, visual aids, and techniques used to perform on-the-job training.

The inspectors independently verified satisfactory completion of training by applicable staff by comparing training documentation to the licensees personnel qualifications list.

The inspectors interviewed training instructors and select individuals who were responsible for performance of specific tasks during loading to evaluate their knowledge regarding the cask loading process and use of the equipment. The inspectors reviewed training records of welders and other personnel who the licensee authorized to perform the non-destructive examination (NDE) inspections to ensure that the individuals training was current.

b. Findings

No findings of significance were identified.

.2 Operation of an Independent Spent Fuel Storage at Operating Plants (60855.1)

a. Inspection Scope

The inspection was an examination of the dry fuel storage activities as they related to safety and compliance with the Commissions rules and regulations, certificate of compliance and technical specifications. Specifically, the inspectors reviewed select documentation and interviewed staff regarding the 2008 ISFSI loading campaign to address unresolved item (URI 05000263/2008-005-01, Non-Destructive Examination of Weld on the Outer Lid of Casks Performed Outside the Temperature Range Specified by the Applicable Welding Procedure). Non-destructive examination of weld on outer lid of casks performed outside the temperature range specified in the applicable welding procedure. One licensee-identified violation pertaining to URI 05000263/2008-005-01, Non-Destructive Examination of Weld on the Outer Lid of Casks Performed outside the Temperature Range Specified by the Applicable Welding Procedure, is discussed in Section 40A7. Based on the results of the review, this URI is considered closed.

b. Findings

A licensee-identified violation pertaining to URI 05000263/2008-005-01, Non-Destructive Examination of Weld on the Outer Lid of Casks Performed outside the Temperature Range Specified by the Applicable Welding Procedure, was identified and is discussed in Section 40A7.

4OA6 Management Meetings

Exit Meeting Summary

On April 8, 2010, the inspectors presented the inspection results to Mr. OConnor, and other members of the licensee staff. The licensee acknowledged the issues presented.

The inspectors confirmed that none of the potential report input discussed was considered proprietary.

.2 Interim Exit Meetings

Interim exits were conducted for:

• An exit teleconference for ISFSI IP 60855.1 was held on February 25, 2010.

The inspectors presented the inspection results to members of the licensee management and staff. Licensee personnel acknowledged the information presented.

• The inspector presented the results of the inspection review of licensee corrective actions pertaining to URI 05000263/2010002-01, Reactor Building Crane Design and Licensing Basis Issues, to Mr. N. Haskell, Site Engineering Director, and other members of the licensees staff via telephone on March 25, 2010. Licensee personnel acknowledged the inspection results presented.

The inspectors confirmed that none of the potential report input discussed was considered proprietary. Proprietary material received during the inspection was returned to the licensee.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Severity Level IV) was identified by the licensee and is a violation of NRC requirements, which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.

• The licensee identified that during the Monticello 2008 ISFSI loading campaign a contractor performing non-destructive examinations (NDE) on two dry shielded canisters (DSCs) failed to follow a step of TriVis Procedure QP-9.202, Revision 3, Color Contrast Liquid Penetrant (PT) Examination Using the Solvent-Removable Method, which required the base metal temperature of the surface for the NDE to be below 250 degrees F. Specifically, the base metal temperature for DSC 002 was 270 degrees F and 253 degrees F for DSC 005 when the NDE was performed. The licensee immediately evaluated the situation and held a stand down to emphasize the importance of procedural adherence. Since the loaded and welded canisters were already stored in the Horizontal Storage Modules on the ISFSI pad at the time of discovery of the discrepancies, the licensee performed qualification PT examinations using two comparator blocks at a procedural non-standard temperature of 325 degrees F.

The NDE products adequately identified flaws on the comparator blocks even beyond the qualified procedure range for the base metal temperature.

Title 10 CFR 72.150, Instructions, Procedures, and Drawings, requires, in part, that the licensee prescribe activities affecting quality by documented instructions, procedures, or drawings of a type appropriate to the circumstances and requires that these instructions, procedures, and drawings be followed. The violation was addressed by traditional enforcement since 10 CFR Part 72 is not risk based and is not covered under the Reactor Oversight Process. The inspectors reviewed the examples in the Enforcement Policy, Supplement I and determined that the failure to follow the procedure was a violation that had more than minor safety or environmental significance but did not rise to a Severity Level I, II, or III violation due to the above mentioned results of the comparator blocks and the fact that the test port plug (weld no. 5) did not serve a structural or pressure retaining function. The inspectors determined that the violation had more than minor safety significance because failure to follow procedures and keep the base metal temperature below the temperatures at which the PT examination materials function properly could have lead to errors in identifying potential flaws in more critical welds that serve as pressure boundaries. The licensee entered these issues into its corrective action program as AR 01156986 and AR 01155771.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. OConnor, Site Vice President

J. Grubb, Plant Manager

W. Paulhardt, Assistant Plant Manager

N. Haskell, Site Engineering Director

K. Jepson, Business Support Manager

S. Radebaugh, Maintenance Manager

M. Holmes, Radiation Protection/Chemistry Manager

S. Speight, Regulatory Affairs Manager

K. Shriver, Dry Fuel Storage Project Support

R. Anderson, System Engineering Supervisor

R. Loeffler, Senior Licensing Engineer

T. Crippes, Refueling Project Supervisor

M. Marashi, Senior Plant Engineer

Nuclear Regulatory Commission

K. Riemer, Chief, Reactor Projects Branch 2

LIST OF ITEMS

Opened

05000263/2010002-01 URI Reactor Building Crane Design and Licensing Basis Issues (Section 1R20)

05000263/2010002-02 NCV SRV Low Low Set Surveillance Procedure Implementation (Section 1R22)

05000263/2010002-03 NCV Inadequate Corrective Actions for Unexpected SRV Low Low Set Trips Encountered During Surveillance Testing (Section 4OA2)

Closed

05000263/2010002-02 NCV SRV Low Low Set Surveillance Procedure Implementation (Section 1R22)

05000263/2010002-03 NCV Inadequate Corrective Actions for Unexpected SRV Low Low Set Trips Encountered During Surveillance Testing (Section 4OA2)

05000263/2008005-01 URI Non-Destructive Examination of Weld on the Outer Lid of Casks Performed Outside the Temperature Range Specified by the Applicable Welding Procedure (Section 4OA5)