Text

IR 05000263-11-008 (DRS), on 02/07/2011 - 02/11/2011 and 06/15/2011 - 07/15/2011, Monticello Nuclear Generating Plant (MNGP); Triennial Fire Protection Baseline Inspection
	ML11235A897
Person / Time
Site:	Monticello
Issue date:	08/23/2011
From:	Daley R C; Engineering Branch 3
To:	O'Connor T J; Northern States Power Co
References
	IR-11-008
	Download: ML11235A897 (26)
	v • d • e

August 23, 2011

Mr. Timothy Site Vice President Monticello Nuclear Generating Plant Northern States Power Company, Minnesota 2807 West County Road 75 Monticello, MN 55362-9637

SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000263/2011008(DRS)

Dear Mr. O'Connor:

On July 15, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed a triennial fire protection inspection at your Monticello Nuclear Generating Plant. The enclosed inspection report documents the inspection results, which were discussed on July 15, 2011, with you and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's 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 these issues were entered into your corrective action program, the NRC is treating these issues as Non-Cited Violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy. If you contest the subject or severity of 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 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 Monticello Nuclear Generating Plant. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter 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 Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). However, because of the security-related concerns contained in the enclosure, and in accordance with 10 CFR 2.390, a copy of this letter's enclosure will not be available for public inspection.

Sincerely,/RA/ Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket No. 50-263 License No. DPR-22 Nonpublic

Enclosure:

Inspection Report 05000263/2011008(DRS)

w/Attachment:

Supplemental Information cc w/encl: S. Coker, NSIR/DSO/DDSO/ST C. Johnson, NSIR J. Trapp, Region I M. Ernstes, RII M. Hay, RIV cc w/o encl: Distribution via ListServ Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION III Docket No: 050000263 License No: DPR-22 Report No: 05000263/2011008(DRS) Licensee: Northern States Power Company, Minnesota Facility: Monticello Nuclear Generating Plant (MNGP) Location: Monticello, MN Dates: February 7 - 11, 2011 June 15, 2011, through July 15, 2011 Inspectors: George M. Hausman, Senior Reactor Inspector, Lead Alan K. Dahbur, Senior Reactor Inspector Benny Jose, Senior Reactor Inspector Dariusz Szwarc, Reactor Inspector Nicholas A. Valos, Senior Reactor Analyst Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety 1 Enclosure

SUMMARY OF FINDINGS

IR 05000263/2011008 (DRS); 02/07/2011 - 02/11/2011 and 06/15/2011 - 07/15/2011; Monticello Nuclear Generating Plant (MNGP); Triennial Fire Protection Baseline Inspection. This report covers an announced triennial fire protection baseline inspection. The inspection was conducted by Region III inspectors. Two (Green) findings were identified by the inspectors. The findings were considered Non-Cited Violations (NCVs) of Nuclear Regulatory Commission (NRC) regulations. The significance of most findings is indicated by their color (Green, White, Yellow, or 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 cross-cutting aspects were determined using IMC 0310, "Components Within the Cross-Cutting Areas." The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006. A.

===Cornerstone: Initiating Events

NRC-Identified

and Self-Revealed Findings===

GreenThe inspectors determined that the finding was more than minor because the finding was associated with the Initiating Events cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone's objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance due to the low fire initiating frequency and the availability of remaining mitigating systems. This finding did not have a cross-cutting aspect because the finding was not representative of current performance. (Section 1R05.1b) . The inspectors identified a finding of very low safety significance and associated NCV of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, "Design Control," for the failure to evaluate the impact of the installation of the hydrogen/oxygen analyzer system on safety-related residual heat removal (RHR) system cables. Specifically, the licensee failed to evaluate how a failure of the hydrogen bottles and the resulting fire or explosion could impact RHR cables located directly above the hydrogen bottles. The licensee entered this issue into their corrective action program to review the placement of the hydrogen bottles.

Cornerstone: Mitigating Systems

GreenThe inspectors determined that the finding was more than minor because if left uncorrected, the failure to inspect and test the MCR air intake smoke detector would become a more significant safety concern. Specifically, if the licensee continued . The inspectors identified a finding of very low safety significance and associated NCV of License Condition 2.C.4 for the licensee's failure to inspect and test the main control room (MCR) air intake smoke detector. Specifically, the licensee failed to inspect and test the smoke detector between 2006 and 2011 as required by the preventative maintenance program. The licensee successfully tested the detector once the performance deficiency was identified and entered this issue into their corrective action program to evaluate the status of the detector.

not testing and maintaining the detector it would eventually fail to respond properly and result in a delayed notification to control room operators of a fire that could result in smoke entering the control room. This finding was of very low safety significance because the licensee successfully tested the detector. This finding did not have a cross-cutting aspect because the finding was not representative of current performance. (Section 1R05.3b) B. No violations of significance were identified.

Licensee-Identified Violations

1. REACTOR SAFETY

REPORT DETAILS

Cornerstones: Initiating Events and Mitigating Systems

1R05 Fire ProtectionThe purpose of the fire protection triennial baseline inspection was to conduct a design based, plant specific, risk-informed, onsite inspection of the licensee's fire protection program's defense-in-depth elements used to mitigate the consequences of a fire. The fire protection program shall extend the concept of defense-in-depth to fire protection in plant areas important to safety by:

Main article: IP 71111.05T

preventing fires from starting; rapidly detecting, controlling and extinguishing fires that do occur; providing protection for structures, systems, and components important to safety so that a fire that is not promptly extinguished by fire suppression activities will not prevent the safe shutdown of the reactor plant; and taking reasonable actions to mitigate postulated events that could potentially cause loss of large areas of power reactor facilities due to explosions or fires. The inspectors' evaluation focused on the design, operational status, and material condition of the reactor plant's fire protection program, post-fire safe shutdown systems and B.5.b mitigating strategies. The objectives of the inspection were to assess whether the licensee had implemented a fire protection program that: (1) provided adequate controls for combustibles and ignition sources inside the plant; (2) provided adequate fire detection and suppression capability; (3) maintained passive fire protection features in good material condition; (4) established adequate compensatory measures for out-of-service, degraded, or inoperable fire protection equipment, systems or features; (5) ensured that procedures, equipment, fire barriers, and systems exist so that the post-fire capability to safely shut down the plant was ensured; (6) included feasible and reliable operator manual actions when appropriate to achieve safe shutdown; and (7) identified fire protection issues at an appropriate threshold and ensured these issues were entered into the licensee's problem identification and resolution program. In addition, the inspectors' review and assessment focused on the licensee's post-fire safe shutdown systems for selected risk-significant fire areas. Inspector emphasis was placed on determining that the post-fire safe shutdown capability and the fire protection features were maintained free of fire damage to ensure that at least one post-fire safe shutdown success path was available. The inspectors' review and assessment also focused on the licensee's B.5.b related license conditions and the requirements of Title 10, Code of Federal Regulations (CFR) Part 50.54(hh)(2). Inspector emphasis was to ensure that the licensee could maintain or restore core cooling, containment, and spent fuel pool cooling capabilities utilizing the B.5.b mitigating strategies following a loss of large areas of power reactor facilities due to explosions or fires. Documents reviewed are listed in the Attachment to this report.

4 Enclosure The fire zones and B.5.b mitigating strategies selected for review during this inspection are listed below and in Section 1R05.11. The fire zones selected constitute four samples and the B.5.b mitigating strategies selected constitute four samples each, respectively, as defined in Inspection Procedure (IP) 71111.05T. Fire Area Fire Zone II Description 2-C Control Rod Drive Hydraulic Control Unit (HCU) Area and Heating, Ventilation and Air Conditioning Areas 3-C Motor Control Center Area 3-D Cooling Water Pump and Chiller Area VI 8 Cable Spreading Room

.1 a. Protection of Safe Shutdown Capabilities For each of the selected fire areas, the inspectors reviewed the fire hazards analysis, safe shutdown analysis, and supporting drawings and documentation to verify that safe shutdown capabilities were properly protected. Inspection Scope The inspectors reviewed the licensee procedures and programs for the control of ignition

sources and transient combustibles to assess their effectiveness in preventing fires and in controlling combustible loading within limits established in the fire hazards analysis. The inspectors performed plant walkdowns to verify that protective features were being properly maintained and administrative controls were being implemented. The inspectors also reviewed the licensee's design control procedures to ensure that the process included appropriate reviews and controls to assess plant changes for any potential adverse impact on the fire protection program and/or post-fire safe shutdown analysis and procedures.

b. Findings

Hydrogen Bottles Located Below Residual Heat Removal (RHR) System Cables

Introduction:

The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to evaluate the impact of the installation of the hydrogen/oxygen analyzer system on safety-related RHR system cables. Specifically, the licensee failed to evaluate how a failure of the hydrogen bottles and the resulting fire or explosion could impact RHR cables located directly above the hydrogen bottles.

Description:

In response to NUREG-0737, "Clarification of TMI Action Plan Requirements," dated October 31, 1980, the licensee made modifications to the post accident containment air monitoring system in 1982. On January 6, 1982, the licensee completed a safety evaluation for Modification 81Z067, "Post Accident Containment Air Monitoring System," dated October 8, 1981, that determined that the probability of an occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report (SAR) had not been increased. As a result, the licensee installed a hydrogen/oxygen analyzer system that 5 Enclosure consisted of three compressed hydrogen, three compressed oxygen bottles, and associated equipment in the west HCU area on the 935 foot elevation of the reactor building. This was located in Fire Zone 2-C in Fire Area II. During a walkdown on June 14, 2011, the inspectors noticed that the hydrogen bottles were located approximately 10 feet below a vertical cable riser. The licensee determined that the cable riser contained Division II RHR system cables after questioning by the inspectors. The RHR system is a safety-related system that was evaluated in the SAR. Division II of RHR contains 'A' and 'B' RHR pumps. The inspectors raised concerns about the location of the hydrogen bottles with respect to the safety-related cables. The hydrogen bottles present a fire and an explosion hazard. According to Table 2-7.1 of the Society of Fire Protection Engineers (SFPE) Handbook of Fire Protection Engineering (Third Edition) hydrogen has a lower flammability limit of 4 and an upper flammability limit of 75 percent. This means that a hydrogen mixture of between 4 and 75 percent will burn. Two of the hydrogen bottles had a mixture of 50 percent hydrogen and 50 percent nitrogen and the third bottle contained 18 percent hydrogen and 82 percent nitrogen. Each of the hydrogen bottles had a regulator attached to the discharge. However, if a piece of equipment or some object were to hit the regulator it could fail, cause a spark, and ignite the hydrogen. A fire from one of the hydrogen bottles could damage the RHR system cables and an explosion could additionally damage other nearby equipment. Further, the oxygen located in the three oxygen bottles could enrich the fire and increase its severity. With the exception of the RHR system cables there was no other safety-related equipment identified in the area or on the other side of the wall where the bottles were located. During the inspection the inspectors identified a battery operated man lift located approximately 15 feet away from the bottles that was not properly secured. If such a piece of equipment were to fall over and hit the bottles it could create a scenario where a fire or an explosion could occur. The licensee secured the man lift and issued A/R 01292836, "NRC Identified Unsecured Wheel Chock on Battery Man Lift," dated June 30, 2011. The licensee did not specifically mention the installation of the hydrogen bottles in the documentation for Modification 81Z067 and did not address the impact of that modification on safety-related structures, systems, or components. The inspectors determined that the licensee therefore failed to check the adequacy of the design of the hydrogen/oxygen analyzer system and its impact on the RHR system cables during the original installation. The licensee determined during the fire protection review completed for Modification 81Z067 that the change did not reduce the fire protection provisions for safety-related or safe shutdown equipment and that no fire hazards analysis was necessary. The licensee entered this issue into their corrective action program as A/R 01292906, "FPT 11 - Review Placement of the H2O2 Analyzer System," dated July 1, 2011, to review the placement of the hydrogen bottles.

Analysis:

The inspectors determined that the failure to evaluate the impact of the installation of the hydrogen/oxygen analyzer system on safety-related RHR system cables was contrary to 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and was a performance deficiency. The inspectors determined that the finding was more than minor because the finding was associated with the Initiating Events cornerstone 6 Enclosure attribute of Protection Against External Factors (Fire) and affected the cornerstone's objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the installation of the hydrogen bottles as part of the hydrogen/oxygen analyzer system directly below safety-related RHR system cables could have resulted in damage to the cables if the hydrogen in the bottles were to ignite or explode. In accordance with Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," Table 3b the inspectors determined the finding degraded the fire protection defense-in-depth strategies. Therefore, screening under IMC 0609, Appendix F, "Fire Protection Significance Determination Process," was required. The inspectors determined that the finding impacted the Fire Prevention and Administrative Controls category. Based on review of IMC 0609, Appendix F, Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," the inspectors determined the degradation rating to be high because of the hydrogen gas being more flammable than low flashpoint combustibles and there being a significant fire hazard associated with release of the gas. The Duration Factor was 1.0 based on the duration of the degradation being greater than 30 days per Table 1.4.1, "Duration Factors." An overall fire frequency of 2E-3 per year was calculated for the three hydrogen bottles based on information from IMC 0609, Appendix F, Attachment 4, "Fire Ignition Source Mapping Information: Fire Frequency, Counting Instructions, Applicable Fire Severity Characteristics, and Applicable Manual Fire Suppression Curves." The conditional core damage probability (CCDP) from Table 2.1.1, "Total Unavailability Values for SSD [Safe Shutdown] Path Based Screening CCDP," included a screening value of 1E-2, which was conservative. The Region III Senior Reactor Analyst used the Monticello Standard Plant Analysis Risk (SPAR) Model Version 8.15, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8.0.7.17, to calculate a revised CCDP of 4.2E-8. The Senior Reactor Analyst assumed that a fire due to failure of the bottles would result in a plant scram complicated by failure of Division II RHR pumps (i.e., "B" and "D" pumps). Based on the above CCDP and frequency values the risk associated with this finding is very low (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. EnforcementContrary to the above, from November 11, 1981, through July 15, 2011, the licensee failed to check the adequacy of design of the hydrogen/oxygen analyzer system and its impact on safety-related cables. Specifically, the licensee failed to evaluate how a failure of the hydrogen bottles and resulting fire or explosion could impact RHR cables located directly above the hydrogen bottles. Because this violation was of very low safety significance and it was entered into the licensee's corrective action program as A/R 01292906, this violation is being treated as an NCV, consistent with Section 2.3.2 of : Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

7 Enclosure the NRC Enforcement Policy (NCV 05000263/2011-008-01(DRS), Hydrogen Bottles Located Below RHR System Cables).

.2 a. Passive Fire Protection For the selected fire areas, the inspectors evaluated the adequacy of fire area barriers, penetration seals, fire doors, electrical raceway fire barriers, and fire rated electrical cables.

The inspectors observed the material condition and configuration of the installed barriers, seals, doors, and cables. The inspectors reviewed approved construction details. In addition, the inspectors reviewed license documentation, such as NRC safety evaluation reports (SERs), and deviations from NRC regulations and the National Fire Protection Association (NFPA) standards to verify that fire protection features met license commitments. Inspection Scope The inspectors walked down accessible portions of the selected fire areas to observe material condition and the adequacy of design of fire area boundaries (including walls, fire doors, and fire dampers) to ensure they were appropriate for the fire hazards in the area. b. No findings of significance were identified. Findings

.3 a. Active Fire Protection For the selected fire areas, the inspectors evaluated the adequacy of fire suppression and detection systems.

The inspectors observed the material condition and configuration of the installed fire detection and suppression systems. The inspectors reviewed design documents and supporting calculations. In addition, the inspectors reviewed license basis documentation, such as, NRC SERs, deviations from NRC regulations, and NFPA standards to verify that fire suppression and detection systems met license commitments. Inspection Scope

b. Findings

Failure to Inspect and Test Main Control Room (MCR) Air Intake Smoke Detector

Introduction:

The inspectors identified a finding of very low safety significance (Green) and associated NCV of License Condition 2.C.4 for the failure to inspect and test the MCR air intake smoke detector. Specifically, the licensee failed to inspect and test the smoke detector as required by the preventative maintenance program between 2006 and 2011.

Description:

The purpose of the MCR air intake smoke detector was to alert control room operators of smoke entering the MCR through the ventilation system. Under normal operating conditions the MCR ventilation system draws a portion of the air supplied to the room from the outside. Normally, the functioning smoke detector would 8 Enclosure alarm in the control room and Alarm Response Procedure C-020-A-22, "Smoke Detection EFT [Emergency Filtration Train] Air Intake," Revision 2 would direct operators to isolate the MCR ventilation system and place the system into the recirculation mode to prevent smoke from entering the MCR. If the detector was not functional smoke could enter the MCR before the operators would be aware of it. On December 10, 1976, in a letter to the NRC, the licensee committed to comply with the recommendations contained in Standard Review Plan 9.5.1 and Appendix A to Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1. One of the recommendations in APCSB 9.5-1, Appendix A was to provide the control room ventilation intake with smoke detection capability. The licensee stated in the December 10, 1976, correspondence that a modification would be made to meet the detection and alarm requirements. The NRC in Section 4.2 of fire protection SER dated August 29, 1979, stated that the, "Monticello Nuclear Generating Plant is provided with a fire detection system which monitors several general areas of the plant and a ventilation duct smoke detector in the control room air handling system." Therefore, the MCR air intake detector was required by the licensee's fire protection program. Further, the Updated Safety Analysis Report (USAR), Section 6.7.2 stated that a, "non-redundant, non-Class IE smoke detector, which meets NFPA requirements, is provided in the outside air intake." The licensee performed an inspection and test of the smoke detector provided in the emergency filtration air intake on August 16, 2006, under Work Order (WO) 156483, as required by preventative maintenance program PMRQ 10635-01, "Inspect, Clean, and Test SD-9070," dated August 17, 2006. That preventative maintenance program required that the detector pass a smoke and a sensitivity test on an annual basis. The detector passed the smoke test but failed the sensitivity test and the licensee initiated A/R 01044835, "SD-9070 EFT Air Intake Smoke Detector Fails Sensitivity Test," dated August 16, 2006, to document the failure. The licensee then initiated Work Request 00012355, "SD-9070 EFT Air Intake Smoke Detector Fails Sensitivity Test," dated August 16, 2006, to repair the detector. That work request was subsequently cancelled as the licensee completed a fire protection engineering evaluation FPEE-06-001, "Deletion of Requirement for EFT Air Intake Smoke Detector (SD-9070)," on October 24, 2006, that provided justification for abandoning the smoke detector. The licensee changed the status of the detector to inactive in their "Passport" computer database. As a result, further maintenance on the detector was suspended. However, after questioning by the inspectors the licensee determined that no modification was performed on the system and the detector was not removed from service. The detector was connected to the fire detection system and to an alarm in the control room. As a result, the licensee performed a smoke test on July 12, 2011, and determined that the detector passed the test. On July 15, 2011, the licensee successfully conducted a sensitivity test on the detector and determined that the detector was still functional. The licensee entered this issue into their corrective action program as A/R 01294094, "FPT11 - Installed Instrument SD-9070 Abandoned in Place," and reinstituted the preventative maintenance program for the detector. The inspectors determined that had it not been for their pursuit of the issue the licensee would not have realized that the detector remained in service and was not being maintained. The lack of maintenance and testing would have affected the functionality of the smoke detector over time and could have resulted in a delayed response by the control room operators to smoke entering the room.

9 Enclosure AnalysisThe finding was determined to be more than minor because if left uncorrected, the failure to inspect and test the MCR air intake smoke detector would become a more significant safety concern. Specifically, if the licensee continued not testing and maintaining the detector it would eventually fail to respond properly and result in a delayed or no notification to control room operators of a fire that could result in smoke entering the control room. The inspectors concluded this finding was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and affected the cornerstone's objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to perform the required annual surveillances on the MCR air intake smoke detector that was required to notify the MCR operators to isolate and place the MCR ventilation system into the recirculation mode of operation. : The inspectors determined that failure to inspect and test the MCR air intake smoke detector was contrary to the preventative maintenance program and was a performance deficiency. In accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," Table 3b the inspectors determined the finding degraded the fire protection defense-in-depth strategies. Therefore, screening under IMC 0609, Appendix F, "Fire Protection Significance Determination Process," was required. The inspectors assigned a low degradation to the finding because the smoke detector was likely functional during the five years that it was not tested because it passed its smoke and sensitivity test in July 2011. Therefore, the inspectors determined that the finding screened as having very low safety significance (Green) in Task 1.3.1 of IMC 0609, Appendix F. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance. EnforcementSection 10.3.1.2.4 of USAR stated that administrative controls for the Fire Protection Program include Administrative Work Instruction 4 AWI-08.01.00, "Fire Protection Program Plan," Revision 7 and are, therefore, part of the USAR by reference. Section 4.9 of 4 AWI-08.01.00 required that a surveillance program be established and maintained for fire protection systems and equipment. Preventative maintenance program PMRQ 10635-01, "Inspect, Clean, and Test SD-9070," dated August 17, 2006, required that a smoke and sensitivity test be performed annually for the MCR air intake smoke detector. : License Condition 2.C.4 required the licensee to implement and maintain in effect all provisions of the approved fire protection program as described in the USAR and as approved through SER dated August 29, 1979, and supplements dated February 12, 1981, and October 2, 1985. Contrary to the above, from August 17, 2006, until July 15, 2011, the licensee failed to implement and maintain all provisions of the approved fire protection program. Specifically, the licensee failed to inspect and test the MCR air intake smoke detector as required by the preventative maintenance program. Because this violation was of very low safety significance and it was entered into the licensee's corrective action program as A/R 01294094, this violation is being treated as an NCV, consistent with Section 2.3.2 10 Enclosure of the NRC Enforcement Policy (NCV 05000263/2011-008-02(DRS), Failure to Inspect and Test MCR Air Intake Smoke Detector).

.4 a. Protection from Damage from Fire Suppression Activities For the selected fire areas, the inspectors verified that redundant trains of systems required for hot shutdown would not be subject to damage from fire suppression activities or from the rupture or inadvertent operation of fire suppression systems including the effects of flooding.

The inspectors conducted walkdowns of each of the selected fire areas to assess conditions such as the adequacy and condition of floor drains, equipment elevations, and spray protection. Inspection Scope b. No findings of significance were identified. Findings

.5 a. Alternative Shutdown Capability The inspectors reviewed the licensee's systems required to achieve alternative safe shutdown to determine if the licensee had properly identified the components and systems necessary to achieve and maintain safe shutdown conditions.

The inspectors also focused on the adequacy of the systems to perform reactor pressure control, reactivity control, reactor coolant makeup, decay heat removal, process monitoring, and support system functions. Inspection Scope The inspectors conducted selected area walkdowns to determine if operators could reasonably be expected to perform the alternate safe shutdown procedure actions and that equipment labeling was consistent with the alternate safe shutdown procedure. The review also looked at operator training, as well as consistency between the operations shutdown procedures and any associated administrative controls. b. No findings of significance were identified. Findings

.6 As a result of the licensee's decision to withdraw their intention to comply with 10 CFR Part 50.48(c), this section of the inspection was completed during the week of February 7, 2011.

The results of that inspection are documented in the NRC Integrated Inspection Report (IR) 05000263/2011-002.Circuit Analyses 11 Enclosure

.7 a. Communications The inspectors reviewed, on a sample basis, the adequacy of the communication system to support plant personnel in the performance of alternative safe shutdown functions and fire brigade duties.

The inspectors verified that plant telephones, page systems, sound powered phones, and radios were available for use and maintained in working order. The inspectors reviewed the electrical power supplies for these systems to verify that either the telephones or the radios would remain functional following a fire. Inspection Scope b. No findings of significance were identified. Findings

.8 a. Emergency Lighting The inspectors performed a plant walkdown of selected areas in which a sample of operator actions would be performed in the performance of alternative safe shutdown functions.

As part of the walkdown, the inspectors focused on the existence of sufficient emergency lighting for access and egress to areas and for performing necessary equipment operations. The locations and positioning of the emergency lights were observed during the walkdown and during review of manual actions implemented for the selected fire areas. Inspection Scope b. No findings of significance were identified. Findings

.9 a. Cold Shutdown Repairs The inspectors reviewed the licensee's procedures to determine whether repairs were required to achieve cold shutdown and to verify that dedicated repair procedures, equipment, and material to accomplish those repairs were available onsite.

The inspectors also evaluated whether cold shutdown could be achieved within the required time using the licensee's procedures and repair methods. The inspectors also verified that equipment necessary to perform cold shutdown repairs was available onsite and properly staged. Inspection Scope b. No findings of significance were identified. Findings 12 Enclosure

.10 a. Compensatory Measures The inspectors conducted a review to verify that compensatory measures were in place for out-of-service, degraded or inoperable fire protection and post-fire safe shutdown equipment, systems, or features (e.g., detection and suppression systems, and equipment, passive fire barriers, pumps, valves or electrical devices providing safe shutdown functions or capabilities).

The inspectors also conducted a review on the adequacy of short term compensatory measures to compensate for a degraded function or feature until appropriate corrective actions were taken. Inspection Scope b. No findings of significance were identified. Findings

.11 a. B.5.b Inspection Activities The inspectors reviewed the licensee's preparedness to handle large fires or explosions by reviewing mitigating strategies identified below.

This review ensured that the licensee continued to meet the requirements of their B.5.b related license conditions and 10 CFR Part 50.54(hh)(2) by determining that: Inspection Scope Procedures were being maintained and adequate; Equipment was properly staged, maintained, and tested; Station personnel were knowledgeable and could implement the procedures; and Additionally, inspectors reviewed the storage, maintenance, and testing of B.5.b related equipment. The inspectors reviewed the licensee's B.5.b related license conditions and evaluated selected mitigating strategies to ensure they remain feasible in light of operator training, maintenance/testing of necessary equipment and any plant modifications. In addition, the inspectors reviewed previous inspection reports for commitments made by the licensee to correct deficiencies identified during performance of Temporary Instruction (TI) 2515/171 or subsequent performances of these inspections. The B.5.b mitigating strategies selected for review during this inspection are listed below. The offsite and onsite communications, notifications/emergency response organization activation, initial operational response actions, and damage assessment activities identified in Table A.3-1 of Nuclear Energy Institute (NEI) 06-12, "B.5.b Phase II and III Submittal Guidance," Revision 2 are evaluated each time due to the mitigation strategies' scenario selected.

13 Enclosure NEI 06-12, Revision 2, Section Licensee Strategy Table 2.2 Spent Fuel Pool (SFP) Makeup - Internal Strategy (Table A.2 1) 2.3.1 SFP Makeup - External Strategy (Table A.2-2) 2.3.2 SFP Spray - External Strategy (Table A.2-3) 3.4.2 DC [Direct Current] Power Supplies to Allow Depressurization of RPV [Reactor Pressure Vessel] and Injection with Portable Pump (Table A.5-2) b. One finding was identified which is discussed in IR 05000263/2011-408(DRS). Findings

OTHER ACTIVITIES (OA)

4OA2 Problem Identification and Resolutiona.

Main article: IP 71152

The inspectors reviewed the licensee's corrective action program procedures and samples of corrective action documents to verify that the licensee was identifying issues related to the fire protection program at an appropriate threshold and entering them in the corrective action program. The inspectors reviewed selected samples of condition reports, design packages, and fire protection system non-conformance documents. Inspection Scope b. No findings of significance were identified. Findings

4OA3 Follow-Up of Events and Notices of Enforcement Discretion.1

Main article: IP 71153

The licensee issued LER 05000263/2007-005-00 regarding the use of operator manual actions in an Appendix R Section III.G.1/G.2 fire area. The LER was closed in IR 05000263/2007-004 and documented as a licensee-identified violation of NRC requirements. Because the licensee was transitioning to NFPA 805 and the violation met the criteria established by the NRC's Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR Part 50.48(c)) for a licensee in NFPA 805 transition, the NRC exercised enforcement discretion to not cite the violation in accordance with the NRC's Enforcement Policy. On December 22, 2010, the licensee provided an update to LER 05000263/2007-005 to reflect their withdrawal of their letter of intent to voluntarily implement 10 CFR Part 50.48(c) at MNGP. (Closed) Licensee Event Report (LER) 05000263/2007-005-01: Discovery of Appendix R - Non Compliant Manual Actions During Review of NFPA 805 On May 14, 2009, the NRC issued Enforcement Guidance Memorandum (EGM) 09-002, "Enforcement Discretion for Fire Induced Circuit Faults," dated May 14, 2009, which specified that Regulatory Guide (RG) 1.189, "Fire Protection for Nuclear Power Plants," Revision 2 (issued October 2009) would include a method of compliance for licensees to resolve multiple fire induced circuit faults for licensees that chose not to adopt 14 Enclosure NFPA 805. Regulatory Guide 1.189, Revision 2 provided methods acceptable to the NRC to protect components important to safe shutdown. One of the methods specified in Section 5.3.1.2 of RG 1.189, Revision 2 was operator manual actions. The licensee evaluated the operator manual actions to reestablish ventilation in Fire Area IX, Fire Zone 12-A (Lower 4kV Room) and Fire Area XX, Fire Zone 14-A (Upper 4kV Room) and determined that those manual actions would be classified as important to safe shutdown in Evaluation FPEE-11-008, "Determination of Ventilation Manual Action Classification," Revision 2. The licensee is in the process of revising Procedure C.4-B.08.07A, "Ventilation System Failure" to include manual actions to restore ventilation through the use of a portable generator. The licensee has three years from the issuance of RG 1.189, Revision 2 (i.e., until November 2012) to complete corrective actions. The inspectors determined that the establishment of appropriate manual actions through their corrective action program will achieve compliance in accordance with EGM 09-002. Since the inspectors did not identify any significant issues during the review of this LER, and the underlying issue was given enforcement discretion during the closure of LER 05000263/2007-005, this LER is considered closed. Documents reviewed as part of this inspection are listed in the Attachment.

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

.2 On November 12, 2010, during a fire protection program assessment, the licensee discovered that the fire protection safe shutdown analysis did not address a postulated reactor vessel overfill event.

During a potential fire, concurrent with a loss of offsite power, requiring the evacuation of the control room, both the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) pumps would start if a low reactor water level setpoint was reached. Potential fire damage could result in the failure of the safety-related HPCI and RCIC high reactor water level trip circuitry leading to a reactor vessel overfill event. Specifically, level switches LS-2-3-672E(LIS) and LS-2-3-672F provide the high reactor water level trip signal. Fire damage to the switch circuitry could prevent the HPCI and RCIC high reactor water level trip, which would result in the reactor vessel overfill event. The event would continue until sufficient water filled the HPCI and RCIC steam lines to stall the HPCI and RCIC pumps. As a result of the water filled steam lines, when Operations personnel manually open (i.e., lifts) the safety relief valves (SRVs) to allow for low pressure reactor water inventory makeup and decay heat removal, the SRVs may be subjected to high pressure steam/water flow. (Closed) LER 05000263/2011-001-01: Reactor Vessel Overfill in Appendix R Scenario Upon discovery of this postulated event, the licensee determined that for a fire in the control room or cable spreading room, that one train of RHR and one train of core spray, the SRVs and associated auxiliary and support systems were required (i.e., credited) for safe shutdown. The HPCI and RCIC systems were not credited for this Appendix R scenario. In addition, industry operating experience and the licensee's plant specific analysis concluded that the SRV tail pipes would remain intact following the SRV lifts while subjected to liquid and/or two phase flow. Also, it was confirmed that Operations personnel were directed by procedures to take manual control of HPCI and RCIC for this scenario, which further reduced the probability of a reactor vessel overfill event.

15 Enclosure The licensee concluded that the cause of this event was due to an inadequate Appendix R analysis. Specifically, the previous analysis failed to consider a HPCI and RCIC automatic initiation with a subsequent failure to trip on high reactor water level. The licensee entered this issue into their corrective action program as A/R 01290369, "GAR to Update Piping Database," dated June 20, 2011. The inspectors did not identify any significant issues during the review of this LER. Documents reviewed as part of this inspection are listed in the Attachment. This LER is closed. This event follow-up review constituted one sample as defined in IP 71153-05.

4OA6 Management Meetings The inspectors presented the inspection results to Mr. Timothy O'Connor and to other members of the licensee staff on July 15, 2011.

The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary. Exit Meeting Summary ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

P. Albares, Operations Support Manager Licensee

P. Anderson, Regulatory Affairs Director

J. Antignano, Site Fire Protection Program Owner

B. Callstrom, Shift Captain - Security

D. Crane, Appendix R Engineer

D. Crofoot, Nuclear Oversight Supervisor

J. Earl, Emergency Preparedness Manager

S. Einbinder, Fleet Fire Protection Lead

J. Grubb, Plant Manager

N. Haskell, Engineering Director

D. Hazard, Multiple Spurious Operation Modification Design Engineer

T. Hurrle, Control Room Supervisor

M. Huting, Fleet Engineering Director

K. Jepson, Business Support Manager

T. Johnson, Fire Marshall, Operations

M. Kelly, Fleet Engineering Supervisor

S. Kibler, Engineering Programs Supervisor

J. Molden, Vice President Engineering

D. Neve, Regulatory Affairs

T. O'Connor, Site Vice President

S. Oswald, Regulatory Affairs

L. Schulze, Operations Support

G. Sherwood, Program Engineering Manager

L. Sueper, Regulatory Affairs

J. Zimmerman, Electrical Design Engineer

S. Thomas, Senior Resident Inspector Nuclear Regulatory Commission

P. Voss, Resident Inspector

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

05000263/2011-008-01(DRS) Opened NCV Hydrogen Bottles Located Below RHR System Cables (Section 1R05.1b)

05000263/2011-008-02(DRS) NCV Failure to Inspect and Test the MCR Air Intake Smoke Detector (Section 1R05.3b)

05000263/2011-008-01(DRS) Closed NCV Hydrogen Bottles Located Below RHR System Cables (Section 1R05.1b)

05000263/2011-008-02(DRS) NCV Failure to Inspect and Test the MCR Air Intake Smoke Detector (Section 1R05.3b)

05000263/2007-005-01 LER Discovery of Appendix R - Non Compliant Manual Actions during Review of NFPA 805 (Section 4OA3.1)

05000263/2011-001-01 LER Reactor Vessel Overfill in Appendix R Scenario (Section 4OA3.2) None.

Discussed

LIST OF DOCUMENTS REVIEWED

The following is a list of documents reviewed during the inspection.

Inclusion on this list does not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that selected sections of portions of the documents were evaluated as part of the overall inspection effort.

Inclusion of a document on this list does not imply NRC acceptance of the document or any part of it, unless this is stated in the body of the inspection report. CALCULATIONS Number Description or Title 11-153 Date or Revision Development of MS System Transient Response to Relief Valve Water Actuation 0 11-154 SRV Piping Analysis for Postulated Hot Short Condition 0

CORRECTIVE ACTION PROGRAM DOCUMENTS (A/Rs) ISSUED DURING INSPECTION Number Description or Title

01290369 Date or Revision GAR to Update Piping Database June 20, 2011

01290573 Unsecured Ladder Stored to Close to Sensitive Equipment June 14, 2011

01290835

FPT 11 - USAR App J.03 Document Error June 16, 2011

01290863 Work Task

402903-01 Cancelled Without Work Complete June 16, 2011

01292579

FPT 11 - Under Plate Fire Penetration Seals Not Inspected June 30, 2011

01292624 Motor Oil Not Staged for P-229 Portable Pump per Ops Procedure June 29, 2011

01292836 NRC Identified Unsecured Wheel Chock on Battery Man Lift June 30, 2011

01292906

FPT 11 - Review Placement of the H2O2 Analyzer System July 1, 2011

01294094

FPT 11 - Installed Instrument

SD-9070 Abandoned In Place July 11, 2011

01294208 PMCR for New PMID for 8A and 8B Generators July 12, 2011

01294395

FPT 11 - Improvements to Communication Equipment Documentation July 13, 2011

CORRECTIVE ACTION PROGRAM DOCUMENTS (A/Rs) REVIEWED Number Description or Title

01002810 Date or Revision Potential NRC Violation for Sprinkler Obstruction in EDG Rooms November 5, 2005

01044835

SD-9070 EFT Air Intake Smoke Detector Fails Sensitivity Test August 16, 2006

01223611 GAR to Perform TCOA Periodic Assessment March 22, 2010

01225914 GAR to Track TCOA Validations Performed for

OWI-03.07, Rev 1 May 25, 2010

DRAWINGS Number Description or Title

NX-16518-2 Date or Revision Fire Detector Location and Conduit Layout Reactor Building D

NX-16518-3 Fire Detector Location and Conduit Layout Reactor Building D

NX-20598-2 C-108A Cable Spreading Halon System

B

NF-36027 Reactor Building Floor Framing Plan - El. 962' - 6" B

NF-36282 Reactor Building - South Conduit & Trays Above EL. 935' - 0" 76

NE-36375-40 Turbine Plant Vertical Board C-20 Annunciator 77

NF-36496 Reactor Building Beam & Slab Details - El. 962' - 6" 2

NX-9215-2 C-108 Cable Spreading Smoke Detection System A

EVALUATIONS Number Description or Title

FPEE-06-001 Date or Revision Deletion of Requirement for EFT Air Intake Smoke Detector (SD-9070) 0

FPEE-10-003 Lack of Sequential Timer on Diesel Fire Pump 0

FPEE-11-008 Determination of Ventilation Manual Action Classification

0

MODIFICATIONS Number Description or Title 81Z067 Date or Revision Post Accident Containment Air Monitoring System October 8, 1981 EC9419 EDG Rooms Sprinkler Modification 0

PROCEDURES Number Description or Title 0274 Date or Revision Fire Hose Hydrostatic Test - Interior Hose Stations 27 1224 Fire Brigade Equipment Inventory 28 4

AWI-08.01.00 Fire Protection Program Plan 7 A.3-02-C West HCU Area Strategy A.3-02-C 8 A.3-03-C RX Vessel Instrument Rack Area (962' Elevation) Strategy A.3-03-C 7 A.3-03-D Rx Building RBCCW Pump Area Strategy A.3-03-D 8 A.3-08 Cable Spreading Room Strategy A.3-08 12 B.08.04.03-05 Alternate Nitrogen System 15 B.08.11-05 Diesel Oil System 22 C.4-J Operations During a Toxic Gas Event 5 C-020-A-22 Smoke Detection EFT Air Intake 2

EC-0441 Multiple Spurious Operations Review February 4, 2011

FG-WM-PMA-01 Preventive Maintenance and Surveillance Administration 7

PROCEDURES Number Description or Title

FP-OP-CTC-01 Date or Revision Control of Time Critical Actions 2

MWI-3-M-2.06 Fuse/Breaker Coordination Study and Electrical Coordination 9 REFERENCES Number Description or Title 536-1330 Date or Revision Halon Acceptance Test - Cable Spreading Room October 24, 1981

MWI-3-M-2.06 Fuse/Breaker Coordination Study and Electrical Coordination 9

NEDE-24988-P; Class III Analysis of Generic BWR Safety/Relief Valve Operability Test October 1981 NFPA

HFPE-02 The SFPE Handbook of Fire Protection Engineering 3rd Edition NRC Letter NRC (D.B. Vassallo) to NSP (L.O. Mayer),

NUREG-Items

II.F.1.4,

II.F.1.5, and

II.F.1.6 May 14, 1982 NRC Letter NRC (D.B. Vassallo) to (NSP Company ), "Completed Review of

NUREG-0737, Items

II.F.1.4,

II.F.1.5 and

II.F.1.6" July 28, 1982 NRC Letter NRC (W.D. Shafer) to NSP Company (C.E. Larson) April 5, 1984 NSP Letter NSP (L.O. Mayer) to NRC (V. Stello), Comparison of Existing Fire Protection Provisions to the Guidelines Contained December 10, 1976

NUREG-0737 Clarification of TMI Action Plan Requirements October 31, 1980 PMRQ 10635 Inspect, Clean, and Test

SD-9070 August 17, 2006 VENDOR DOCUMENTS Number Description or Title

NX-17048 Date or Revision Pyrotronics Training Manual 2 WORK ORDERS (WOs) Number Description or Title

00156483 Date or Revision Inspect, Clean, and Test

SD-9070 August 18, 2006

00295615

SD-9070 EFT Air Intake Smoke Detector Fails Sensitivity Test September 11, 2006

00344033 0275-01 Fire Barrier Penetration on Seal Visual Inspection April 8, 2009

00389802 0275-03 Fire Door Inspection April 1, 2010

00401499 1216-02 Fire Door Inspections - Semiannual September 21, 2010

00404995 Fire Detection Instrumentation Check December 8, 2010

00414101 0275-01 Fire Barrier Penetration on Seal Visual Inspection April 9, 2011