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IR 05000285/2015012; 10/19/2015 - 11/06/2015; Fort Calhoun Station; Triennial Fire Protection Team Inspection
	ML15351A398
Person / Time
Site:	Fort Calhoun
Issue date:	12/17/2015
From:	Werner G E; NRC/RGN-IV/DRS/EB-2
To:	Cortopassi L; Omaha Public Power District
References
	IR 2015012
	Download: ML15351A398 (46)
	v • d • e

December 17, 2015

Mr. Louis Cortopassi Site Vice President and Chief Nuclear Officer Omaha Public Power District

Fort Calhoun Station FC-2-4 P.O. Box 550

Blair, NE 68023-0550

SUBJECT: FORT CALHOUN STATION -

NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000285/2015012

Dear Mr. Cortopassi:

On November 6, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Fort Calhoun Station and discussed the results of this inspection with you and other members of your staff. The inspectors documented the results of this inspection in the enclosed inspection report.

The inspectors documented three findings of very low safety significance (Green) in this report. All three of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violations or significance of the violations in this report, you should provide a written 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 copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Fort Calhoun Station.

If you disagree with a cross-cutting aspect assignment 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 IV, and the NRC resident inspector at the Fort Calhoun Station.

In accordance with Title 10 of the Code of Federal Regulations 2.390, "Public Inspections, Exemptions, Requests for Withholding," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC's Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/

Gregory E. Werner, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-285 License No. DPR-40

Enclosure:

Inspection Report 05000285/2015012

w/Attachment:

Supplemental Information

cc w/encl: Electronic Distribution

ML15351A398 SUNSI Review By: JMateychick ADAMS Yes No Publicly Available Non-Publicly Available Non-Sensitive Sensitive Keyword: NRC-002 SUNSI Review Complete OFFICE DRS/EB2 DRS/EB2 DRS/EB2 RI/DRS/EB3 DRS/PSB2 DRS/EB2 NAME S. Alferink N. Okonkwo E. Uribe S. Galbreath D. Loveless J. Mateychick SIGNATURE /RA/ Email Email Email Email /RA/ DATE 12/10/15 12/10/15 12/7/15 12/14/15 12/14/15 12/10/15 OFFICE BC:DRP/D BC:DRS/EB2 NAME G. Miller G. Werner SIGNATURE /RA/ /RA/ DATE 12/14/15 12/16/15 Letter to Louis Cortopassi from Gregory E. Werner, dated December 17, 2015

SUBJECT: FORT CALHOUN STATION - NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000285/2015012

Electronic distribution by RIV

Regional Administrator (Marc.Dapas@nrc.gov)

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ACES (R4Enforcement.Resource@nrc.gov) Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket: 05000285 License: DPR-40 Report Nos.: 05000285/2015012 Licensee: Omaha Public Power District Facility: Fort Calhoun Station Location: 9610 Power Lane, Blair, NE 68008 Dates: October 19 through November 6, 2015 Team Leader: J. Mateychick, Senior Reactor Inspector, RIV, Engineering Branch 2 Inspectors: S. Alferink, Reactor Inspector, Region IV, Engineering Branch 2 N. Okonkwo, Reactor Inspector, Region IV, Engineering Branch 2 E. Uribe, Reactor Inspector, Region IV, Technical Support Services S. Galbreath, Reactor Inspector, Region I, Engineering Branch 3

D. Loveless, Region IV, Senior Reactor Analyst Accompanying Personnel: H. Barrett, Senior Fire Protection Engineer, Office of Nuclear Reactor Regulation, Division of Risk Assessment Approved By: Gregory E. Werner Chief, Engineering Branch 2

Division of Reactor Safety

SUMMARY

IR 05000285/2015012; 10/19/2015 - 11/06/2015; Fort Calhoun Station; Triennial Fire Protection Team Inspection.

The report covered a two week triennial fire protection team inspection by specialist inspectors from Region I and Region IV. Three findings of very low safety significance (Green) are documented in this report. These findings involved violations of NRC requirements. The significance of inspection findings is indicated by their color (i.e., Green, White, Yellow, or Red) and determined using Inspection Manual Chapter 0609, "Significance Determination Process," dated April 29, 2015. Cross-cutting aspects are determined using Inspection Manual Chapter 0310, "Aspects within the Cross-Cutting Areas," dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy, dated February 4, 2015. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 5.

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified two examples of a non-cited violation of License Condition 3.D, "Fire Protection Program," for the failure to adequately implement required National Fire Protection Association Standard 805 implementation items in accordance with the approved fire protection program. Specifically, the licensee did not implement two items listed in Table S-3, "Implementation Items," of Omaha Public Power District letter LIC-14-0042 by June 15, 2015. There was no immediate safety concern with either example and the licensee entered this violation into the corrective action program as Condition Reports 2015-2620 and 2015-2683.

The failure to implement a requirement of a license condition within the allowed implementation period was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," and determined that the issue was of very low safety significance (Green). These findings had a cross-cutting aspect associated with change management within the human performance area since the leaders failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the inspectors determined that the licensee did not have a process in place to ensure system level design basis documents were updated within the period required by a license condition and to assure plant-specific requirements were incorporated into the appropriate procedures (H.3). (Section 1R05.01.b)

Green.

The inspectors identified a non-cited violation of License Condition 3.D, "Fire Protection Program," for the failure to ensure one success path necessary to achieve and maintain the nuclear safety performance criteria was maintained free of fire damage for all single fires. Specifically, the licensee failed to provide adequate isolation for the pressurizer heaters credited for achieving safe and stable plant conditions for fires that require shutdown from outside the control room. The licensee entered this issue into their corrective action program as Condition Report 2015-12195 and added this issue to their compensatory measures for the control room and cable spreading room.

The failure to provide adequate isolation for equipment relied upon to achieve safe and stable plant conditions for a shutdown from outside of the control room was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." Because the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire requiring evacuation of the control room, a senior reactor analyst performed a Phase 3 evaluation and determined that the issue was of very low safety significance (Green). This finding did not have a cross-cutting aspect since it was not indicative of present performance in that the performance deficiency occurred more than three years ago. (Section 1R05.05.b)

Green.

The inspectors identified a non-cited violation of License Condition 3.D, "Fire Protection Program," for the failure to establish an appropriate monitoring program in accordance with National Fire Protection Association Standard 805, Section 2.6.

Specifically, the licensee failed to set the action level for the availability of the raw water system pumps to ensure that the assumptions in the engineering analysis remained valid. There was no immediate safety concern since the raw water pumps availability remained above the value assumed in the analysis and the licensee entered this violation into the corrective action program as Condition Report 2015-12612.

The failure to set the action level for the ava ilability of the raw water system pumps to ensure that the assumptions in the engineering analysis remained valid was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," and determined that the issue was of very low safety significance (Green). This finding had a cross-cutting aspect associated with change management within the human performance area since the leaders failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the inspectors determined that the licensee did not use the process that was in place to ensure that the appropriate fire risk assessment monitoring action levels were incorporated into the maintenance rule program and monitored (H.3). (Section 1R05.13.b)

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R05 Fire Protection

Main article: IP 71111.05XT

This report presents the results of a triennial fire protection inspection conducted in accordance with NRC Inspection Procedure 71111.05XT, "Fire Protection - NFPA 805 (Triennial)," issued January 31, 2013. The inspectors reviewed the licensee's fire protection program against the requirements of NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, as incorporated by Title 10 of the Code of Federal Regulation (CFR) 50.48(c). The NFPA 805 standard establishes a comprehensive set of requirements for fire protection programs at nuclear power plants. The standard incorporates both deterministic and risk-informed performance-based concepts. The inspectors evaluated the implementation of the approved fire protection program in selected risk-significant areas with an emphasis on the procedures, equipment, fire barriers, and systems that ensure the post-fire capability to safely shutdown the plant.

Inspection Procedure 71111.05XT requires the selection of three to five fire areas and one or more mitigating strategies for review. The inspectors used the Fort Calhoun Station NFPA 805 Fire Probabilistic Risk Assessm ent to select the following four fire areas (inspection samples) for review:

Fire Area Description Category 34B-1 Electrical Penetration Area Ground and Intermediate Levels (Room 57) Performance-Based 36B West Switchgear Area (Room 56W) Performance-Based 42 Control Room Complex Performance-Based 43 Emergency Feedwater Tank Area (Room 81) Performance-Based Since this was the first triennial inspection following NRC approval of the risk-informed, performance-based fire protection program, the inspectors reviewed samples of the implementation items required to have been completed in accordance with Operating License Condition 3.D. The inspectors also reviewed samples of the plant modifications credited to support the approved fire protection program.

The inspectors evaluated the licensee's fire protection program using the applicable requirements, which included plant Technical Specifications, Operating License Condition 3.D, NRC safety evaluations, 10 CFR 50.48, and NFPA 805. The inspectors also reviewed related documents that included the Updated Safety Analysis Report Section 9.5, the nuclear safety capability assessment, and the fire safety analyses.

Specific documents reviewed by the inspectors are listed in the attachment.

Four fire area inspection samples and one mitigating strategy sample were completed.

.01 Protection of Safe Shutdown Capabilities

a. Inspection Scope

The inspectors reviewed the nuclear safety capability assessment, piping and instrumentation diagrams, and fire response procedures to verify that a safe shutdown success path, free of fire damage, would be available to meet the nuclear safety goals, objectives, and performance criteria in the event of a fire under any plant operational mode or configuration.

The inspectors reviewed applicable sections of the fire response procedures for the selected fire areas and their associated fire scenarios to verify that the shutdown methodology properly identified the component s and systems necessary to achieve and maintain safe and stable plant conditions. The inspectors performed walk-throughs of procedure steps to ensure the implementation and human factors adequacy of the procedures. The inspectors verified that licensee personnel credited for procedure implementation had procedures available, were trained on implementation, and were available in the event a fire occurred. The inspectors verified that the operators could reasonably be expected to perform the specific actions within the time required to maintain plant parameters within specified limits.

b. Findings

Introduction.

The inspectors identified two examples of a Green non-cited (NCV) violation of License Condition 3.D, "Fire Protection Program," for the failure to adequately implement required NFPA 805 implementation items in accordance with the approved fire protection program.

Description.

The licensee adopted a performance-based, risk-informed fire protection program meeting the requirements of NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, as allowed per 10 CFR 50.48(c).

The fire protection program was approved by the NRC in a revised License Condition 3.D, "Fire Protection Program," issued in License Amendment Number 275 on June 16, 2014. License Condition 3.D, Section (3), "Transition License Condition," Sub-Section

(c) states:

The licensee shall implement the items listed in Enclosure 1, Attachment S, Table S-3, "Implementation Items," of OPPD letter LIC-14-0042, dated April 10, 2014, no later than 12 months after issuance of the license amendment.

The inspectors identified two examples of the licensee failing to adequately implement

items listed in Table S-3.

Example 1: Failure to Complete a Required Implementation Item within the Allowable Time Period Implementation item REC-104 states, "Update system level design basis documents to reflect the NFPA 805 NSPC [Nuclear Safety Performance Criteria] role that the system components now play (as described in Section 2.7.2 of NFPA 805)."

The licensee failed to complete this implementation item within the required 12-month period. The applicable NFPA 805 information was not incorporated into the system design basis documents currently available for use by plant personnel. An engineering change was created to update all the impacted system design basis documents as part of the Design Basis Reconstitution Project and marked-up system design basis documents were provided to the project, but revised documents have not been issued.

Analysis.

The failure to implement a requirement of a license condition within the allowed implementation period was a per formance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency could adversely affect the capability for the systems that are used to respond to fire initiating events.

The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," Attachment 4, "Initial Characterization of Findings," dated June 19, 2012. The inspectors determined that an IMC 0609, Appendix F, "Fire Protection Significance Determination Process," dated September 20, 2013, review was required as the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire. Using IMC 0609, Appendix F, Attachment 1, "Fire Protection Significance Determination Process Worksheet," dated September 20, 2013, the finding was screened as a Green finding of very low safety significance in accordance with Step 1.3, "Ability to Achieve Safe Shutdown," Question A. The inspectors concluded that the failure to update the system design basis documents with the NFPA 805 fire protection program in a timely manner did not directly impact the plant's ability to achieve and maintain hot shutdown conditions as required by the approved fire protection program.

Example 2: Failure to Adequately Maintain a Required Implementation Item Implementation item REC-136 stated, in part, that revisions will be made to Standing Order SO-O-21, "Shutdown Operations Protection Plan," Standing Order SO-O-25, "Temporary Modification," and General Form FC-66, "Temporary Modification Control Form," to caution against the use of temporary electrical installations during higher risk evolutions that take place in planned outages.

Section 4.1 of Standing Order SO-O-21, Revision 55, contained the following note:

NOTE: The performance of, or use of temporary modifications during "HIGHER RISK EVOLUTIONS" (as defined for NFPA 805, see definition 2.11 in Section 2.0) must be evaluated for impact to identify where the station departs from the assumptions of the non-power operations separation analysis, and that heightened administrative controls or other compensatory measures be applied, as applicable, to address the change(s) from the perspective of fire risk.

The inspectors identified that this note did not adequately implement the specific requirement of implementation item REC-136.

Section 4.4.12.A of Standing Order SO-O-25, Revision 86, and General Form FC-66, Revision 69, were reviewed and the inspectors determined that the requirements of implementation item REC-136 had been incorporated. However, Standing Order SO-O-25, Section 1.1.3.A, also stated, "All new Temporary Configuration Changes (previously called Temporary Modifications) will be processed using CC-AA-112." Procedure CC-AA-112, "Temporary Configuration Changes," Revision 21, did not list any Fort Calhoun Station specific commitments. Section 4.5.1.4 stated, "Determine applicable design consideration impacts using the Design Impact Screening Procedure CC-AA-102 and the Design Attribute Review (DAR), Attachment 1 in CC-AA-102."

The inspectors reviewed Procedure CC-AA-102, "Design Input and Configuration Change Impact Screening," Revision 1, and the additional procedures referenced for reviewing the impact on the fire protection program, Procedure CC-AA-209, "Fire Protection Program Configuration Change Review," Revision 5, and Procedure CC-AA-209-1001, "Guidelines for Performing Fire Protection Program Configuration Change Review," Revision 1. The inspectors determined that all three procedures had a general requirement to review the change for impact on the fire protection program, but did not have the specific information required by implementation item REC-136.

Analysis.

The failure to implement a requirement of a license condition within the allowed implementation period was a per formance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency could adversely affect the availability of systems that are used to respond to fire initiating events during higher risk evolutions that take place in planned outages.

The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," Attachment 4, "Initial Characterization of Findings," dated June 19, 2012. The inspectors determined that an IMC 0609, Appendix F, "Fire Protection Significance Determination Process," dated September 20, 2013, review was required as the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire. Using IMC 0609, Appendix F, Attachment 1, "Fire Protection Significance Determination Process Worksheet," dated September 20, 2013, the finding was screened as a Green finding of very low safety significance in accordance with Step 1.3, "Ability to Achieve Safe Shutdown," B Question. Based on the criteria in Appendix F, Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," dated February 28, 2005, the inspectors assigned the finding a "Low" degradation rating. Using Table A2.3, the inspectors considered a low degradation rating because the procedural deficiencies were not significant since there was a general requirement to review the impact of changes on fire protection program, and other licensee procedures require risk impacts to be evaluated during outages.

These findings had a cross-cutting aspect associated with change management within the human performance area since the leaders failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the inspectors determined that the licensee did not have a process in place to ensure system level design basis documents were updated within the period required by a license condition and to assure plant-specific requirements were incorporated into the appropriate procedures (H.3).

Enforcement.

License Condition 3.D, "Fire Protection Program," states, in part, "Omaha Public Power District shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c)." The fire protection program was approved by the NRC in the revised License Condition 3.D, issued in License Amendment Number 275 on June 16, 2014.

License Condition 3.D, Section (3), "Transition License Condition," Sub-Section

(c) states: The licensee shall implement the items listed in Enclosure 1, Attachment S, Table S-3, "Implementation Items," of OPPD letter LIC-14-0042, dated April 10, 2014, no later than 12 months after issuance of the license amendment.

Implementation item REC-104, from Table S-3 states, "Update system level design basis documents to reflect the NFPA 805 NSPC role that the system components now play (as described in Section 2.7.2 of NFPA 805)."

Implementation item REC-136, From Table S-3 states, in part, that, "Revisions will be made to Standing Order SO-O-21, "Shutdown Operations Protection Plan," Standing Order SO-O-25, "Temporary Modification," and General Form FC-66, "Temporary Modification Control Form," to caution against the use of temporary electrical installations during higher risk evolutions that take place in planned outages." Contrary to the above, from June 16, 2015, to November 6, 2015, the licensee failed to implement two items listed in Table S-3 of OPPD letter LIC-14-0042 within the required 12-month period that ended June 15, 2015. Specifically, the license did not implement items:

(1) REC-104, update of system level design basis documents to reflect NFPA 805 nuclear safety performance criteria, and

(2) REC-136, which cautioned against the use of temporary modifications during higher risk evolutions while in planned outages.

The licensee entered this violation into the corrective action program as Condition Reports 2015-12620 and 2015-12683 and were developing corrective actions. Because this violation was of very low safety significance and no immediate safety concerns existed, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000285/2015012-01, "Failure to Adequately Implement and Maintain Required NFPA 805 Implementation Items."

.02 Passive Fire Protection

a. Inspection Scope

The inspectors walked down accessible portions of the selected fire areas to observe the material condition and configuration of the installed fire area boundaries (including walls, fire doors, and fire dampers) and verify that the electrical raceway fire barriers were appropriate for the fire hazards in the area. The inspectors compared the installed configurations to the approved construction details, supporting fire tests, and applicable license commitments.

The inspectors reviewed installation, repair, and qualification records for a sample of penetration seals to ensure the fill material possessed an appropriate fire rating and that the installation met the engineering design. The inspectors also reviewed similar records for the rated fire wraps to ensure the material possessed an appropriate fire rating and that the installation met the engineering design.

The inspectors also reviewed recent inspection and functional test records for fire dampers to verify whether the inspection and testing was adequately conducted, the acceptance criteria were met, and any potential performance degradation was identified.

b. Findings

No findings were identified.

.03 Active Fire Protection

a. Inspection Scope

The inspectors reviewed the design, maintenance, testing, and operation of the fire detection and suppression systems in the selected fire areas. The inspectors verified the automatic detection systems and the manual and automatic suppression systems were installed, tested, and maintained in accordance with the NFPA code of record or approved deviations, and that each suppression system was appropriate for the hazards in the selected fire areas.

The inspectors performed a walkdown of accessible portions of the detection and suppression systems in the selected fire areas. The inspectors also performed a walkdown of major system support equipment in other areas (e.g., fire pumps and Halon supply systems) to assess the material condition of these systems and components.

The inspectors reviewed the electric and diesel fire pumps' flow and pressure tests to verify that the pumps met their design requirements. The inspectors reviewed the fire main flow tests to verify that the flow loops met their design requirements. The inspectors also reviewed the Halon suppression functional tests to verify that the system capability met the design requirements.

The inspectors assessed the fire brigade capabilities by reviewing training, qualification, and drill critique records. The inspectors also reviewed pre-fire plans and smoke removal plans for the selected fire areas to determine if appropriate information was provided to fire brigade members and plant operators to identify safe shutdown equipment and instrumentation, and to facilitate suppression of a fire that could impact post-fire safe shutdown capability. In addition, the inspectors inspected fire brigade equipment to determine operational readiness for firefighting.

The inspectors observed an unannounced fire drill and subsequent drill critique on November 3, 2015, using the guidance contained in Inspection Procedure 71111.05AQ, "Fire Protection Annual/Quarterly." The inspectors observed fire brigade members fight a simulated fire in the compressor area. The inspectors verified that the licensee identified problems, openly discussed them in a self-critical manner at the drill debrief, and identified 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 firefighting techniques,

(4) sufficient firefighting equipment brought to the scene,

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

(6) utilization of pre-planned strategies,

(7) adherence to the pre-planned drill scenario, and

(8) drill objectives.

b. Findings

No findings were identified.

.04 Protection from Damage from Fire Suppression Activities

a. Inspection Scope

The inspectors performed plant walkdowns and reviewed documents to verify that redundant success paths for achieving safe and stable plant conditions, which are located in the same fire area, would not be subject to damage from fire suppression activities, or from the rupture or inadvertent operation of fire suppression systems. Specifically, the inspectors verified:

A fire in one of the selected fire areas would not directly, through production of smoke, heat, or hot gases, cause activation of suppression systems that could potentially damage all redundant success paths.

A fire in one of the selected fire areas or the inadvertent actuation or rupture of a fire suppression system would not directly cause damage to all redundant success paths (e.g., sprinkler-caused flooding of other than the locally affected

train).

Adequate drainage was provided in areas protected by water suppression systems and for manual use of fire hoses by the fire brigade.

b. Findings

No findings were identified.

.05 Shutdown from a Primary Control Station

a. Inspection Scope

Review of Methodology The inspectors reviewed the nuclear safety capability assessment, procedures, piping and instrumentation drawings, electrical drawings, and other supporting documents to verify that the licensee can achieve and maintain safe and stable plant conditions from the primary control station in the event a fire required evacuation of the control room.

The inspectors verified that the nuclear safety capability assessment properly identified the components and systems necessary to meet the nuclear safety performance criteria for the fire area selected. Specifically, the inspectors determined the adequacy of the

systems selected to meet the criteria for reactivity control, inventory and pressure control, decay heat removal, vital auxiliaries, and process monitoring. For the primary control station, which was analyzed using a performance-based approach, the inspectors verified that the analysis included a consideration of all the necessary cables and equipment associated with operation and control of off-site power.

The inspectors verified that the transfer of command and control from the control room to the primary control station would be unaffected by fire-induced circuit faults (e.g., by the provision of separate fuses and power supplies for shutdown control circuits).

Review of Operational Implementation

The inspectors verified that the training program for licensed and non-licensed operators included the procedures for achieving and maintaining safe and stable plant conditions, including any necessary recovery actions. The inspectors also verified that sufficient personnel required to achieve and maintain safe and stable plant conditions were properly trained and were available at all times among the normal on-site staff, exclusive of the fire brigade.

The inspectors performed a timed walkdown of Procedure AOP-06, "Fire Emergency," Revision 29, with licensed and non-licensed operators to determine the adequacy of the procedure. The inspectors verified that the recovery actions taken were feasible and that operators could reasonably be expected to implement the procedure within the applicable time requirements to achieve the nuclear safety performance criteria. The inspectors evaluated the feasibility of the recovery actions using the criteria established in the licensee's approved fire protection program.

The inspectors also verified that the licensee conducted periodic operational tests of the transfer and isolation capability and instrumentation and control functions used for transferring control from the main control room to the primary control station and other locations where recovery actions would be performed. The inspectors verified that the tests were adequate to prove the functionality of the primary control stations' capability to meet performance criteria and achieve and maintain safe and stable plant conditions.

b. Findings

Introduction.

The inspectors identified a Green non-cited violation of License Condition 3.D, "Fire Protection Program," for the failure to ensure one success path necessary to achieve and maintain the nuclear safety performance criteria was maintained free of fire damage for all single fires. Specifically, the licensee failed to provide adequate isolation for the pressurizer heaters credited for achieving safe and stable plant conditions for fires that require shutdown from outside the control room.

Description.

For fires in the control room or cable spreading room that require control room evacuation and shutdown from outside the control room, the licensee was required to achieve and maintain safe and stable plant conditions by entering Hot Shutdown (Mode 3) and maintaining that condition for an initial 24-hour coping time. The licensee documented their compliance strategy for each fire area in the plant in Attachment 8 of NFPA 805 analysis, EA10-036, "Fort Calhoun Station Automation and Update of Safe Shutdown

Analysis.

" For the control room and cable spreading room fire areas, the licensee documented that the inventory and pressure control nuclear safety performance criterion would be satisfied using the pressurizer backup heater bank 4 (RC4-4) for fires that require shutdown from outside the control room.

C, "NEI 04-02 Table B-3 - Fire Area Transition," of the license amendment request, dated September 28, 2011, provided the method of accomplishment for the performance goals for each fire area. This table also specified that operators would initially maintain reactor coolant system pressure by operating the pressurizer backup heater bank 4 locally at the breaker for the reactor coolant system pressure control performance goal for fires that require shutdown from outside the control room.

The licensee used Procedure AOP-06, "Fire Emergency," Revision 29, to shut down the reactor in the event a control room or cable spreading room fire required control room evacuation and shutdown from outside the control room. During a walkdown of this procedure, the inspectors inquired about the isolation of pressurizer backup heater bank

4 from the control room.

In response to the inspectors' questions, the licensee reviewed the drawings for pressurizer backup heater bank 4 and identified a circuit vulnerability that could result in the loss of control for the pressurizer heater bank when performing a shutdown from outside the control room. This vulnerability involved an external hot short affecting the conductor connecting to the control room switch which could keep a relay energized and defeat control of the heater from the motor control center. If this condition occurred, operators would be unable to control pressurizer backup heater bank 4, which was relied upon to achieve safe and stable plant conditions for fires that require shutdown from outside the control room.

The licensee reported this issue as an unanalyzed condition in Event Notification 51487 on October 21, 2015.

Analysis.

The failure to provide adequate isolation for equipment relied upon to achieve safe and stable plant conditions for a shutdown from outside of the control room was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency could adversely affect the availability, reliability, and capability of pressurizer heater bank 4, which is used to respond to fire initiating events. The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," Attachment 4, "Initial Characterization of Findings," dated June 19, 2012. The inspectors determined that an IMC 0609, Appendix F, "Fire Protection Significance Determination Process," dated September 20, 2013, review was required as the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire. A senior reactor analyst performed a Phase 3 evaluation to determine the risk significance of this finding since it involved a postulated control room fire that led to

control room evacuation.

For the control room, the senior reactor analyst assigned a plant-specific fire ignition frequency for the main control board (FIF MCB) from Calculation CN-RAM-09-021, "Fire Ignition Frequencies for Fort Calhoun Station," Revision 2. The analyst multiplied the fire ignition frequency by a severity factor (SF) and a non-suppression probability indicating that operators failed to extinguish the fire within 20 minutes, assuming a 2-minute detection that required a control room evacuation (NP CRE). The resulting control room evacuation frequency (F CR) was: F CR = FIF MCB

SF

NP CRE

= 1.63E-3/year

0.1

1.3E-2

= 2.12E-6/year

The main control board had a total of 7 panels. The analyst determined that only a fire in one of these panels (CB-1) could lead to the loss of the credited pressurizer heater.

Therefore, a bounding change in core damage frequency for a control room fire that leads to evacuation and the loss of the pressurizer heater (F CR+PZR) was determined to be:

F CR+PZR = F CR * (1 / 7)

= 2.12E-6/year * (1 / 7)

= 3.03E-7/year

For the cable spreading room, the senior reactor analyst limited the risk determination to transient and hot work fires since there were no fixed ignition sources in the cable spreading room. The senior reactor analyst assigned a plant-specific fire ignition frequency for transients (FIF CSR-TR) and hot work activities (FIF CSR-HW) from Calculation CN-RAM-10-013, "Fire Scenario Selection and Characterization for Fort Calhoun Station," Revision 1.

Based on a walkdown of the cable spreading room, the senior reactor analyst assigned a screening weighting factor (W) of 0.1 for transient and hot work fires. The senior reactor analyst noted that the cable spreading room was protected by a Halon system. The senior reactor analyst assigned a nominal failure probability of the Halon system (NP Halon) from Calculation CN-RAM-10-013.

Therefore, a bounding change in core damage frequency for a cable spreading room fire that leads to evacuation and the loss of the credited pressurizer heater (F CSR+PZR) was determined to be:

F CSR+PZR = (FIF CSR-TR + FIFCSR-HW)

W

NP Halon = (2.3E-5/year + 7.52E-6/year)

0.1

0.069

= 2.11E-7/year

Since fires in the control room are independent of fires in the cable spreading room, the senior reactor analyst calculated a bounding total change in core damage frequency

(CDF TOT) for the performance deficiency by adding the change in core damage frequencies for the control room and cable spreading room calculated above. The senior reactor analyst calculated a bounding total change in core damage frequency of:

CDF TOT = F CR+PZR + FCSR+PZR

= 3.03E-7/year + 2.11E-7/year

= 5.1E-7/year This change in core damage frequency was considered to be bounding since it

assumed:

A fire in any of the applicable main control board panels or the cable spreading room areas would cause a loss of the credited pressurizer heater,

The conditional core damage probability given a control room fire with evacuation and the loss of the credited pressurizer heater was equal to one, and

The performance deficiency accounted for the entire change in core damage frequency (i.e., the baseline core damage frequency for this event was zero).

In accordance with the guidance in Inspection Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process," dated May 6, 2004, the senior reactor analyst screened the performance deficiency for its potential risk contribution to large early release frequency since the bounding change in core damage frequency provided a risk significance estimate greater than 1E-7/yr. Given that Fort Calhoun Station has a large dry containment and that control room evacuation sequences do not include steam generator tube ruptures or intersystem loss of coolant accidents, the analyst determined that this example was not significant with respect to large early release frequency. The analyst determined this example was of very low risk significance (Green).

This finding did not have a cross-cutting aspect since it was not indicative of present performance in that the performance deficiency occurred more than three years ago.

Enforcement.

License Condition 3.D, "Fire Protection Program," states, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request, dated September 28, 2011 (and supplements, dated December 19, 2011; December 22, 2011; March 20, 2012; July 24, 2012; August 24, 2012; September 27, 2012; April 23, 2013; May 21, 2013; July 29, 2013; September 12, 2013; October 11, 2013; November 4, 2013; November 11, 2013; December 18, 2013; January 24, 2014; February 28, 2014; April10, 2014; and June 11, 2014) and as approved in the safety evaluation, dated June 16, 2014. Table B-3, "Fire Area Transition," of the license amendment request, dated September 28, 2011, states that the licensee would initially maintain reactor coolant system pressure by operating pressurizer heater RC4-4 locally at the breaker for fires in the control room or cable spreading room.

Contrary to the above, from September 28, 2011, to November 6, 2015, the licensee failed to implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c). Specifically, a fire in the control room or cable spreading room could disable the pressurizer heater RC4-4 such that the heater could not maintain reactor coolant system pressure.

The licensee entered this issue into their corrective action program as Condition Report 2015-12195 and added this issue to their compensatory measures for the control room and cable spreading room. Because this violation was of very low safety significance and has been entered into the corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000285/2015012-02, "Failure to Provide Adequate Isolation

for Pressurizer Heaters."

.06 Circuit Analyses

a. Inspection Scope

The inspectors reviewed the nuclear safety capability assessment to verify that the licensee identified the circuits that may impact the ability to achieve and maintain safe and stable conditions. The inspectors verified, on a sample basis, that the licensee properly identified the cables for equipment required to achieve and maintain safe and stable conditions in the event of a fire in the selected fire areas. The inspectors verified that these cables were either adequately protected from the potentially adverse effects of fire damage or were analyzed to show that fire-induced faults (e.g., hot shorts, open circuits, and shorts to ground) would not prevent achieving safe and stable conditions.

a. Inspection Scope

The inspectors verified that the licensee established a monitoring program to ensure that the availability and reliability of the fire protection systems, structures, and components credited in the performance-based analyses are maintained and to assess the performance of the fire protection program in meeting the performance criteria specified in NFPA 805. The inspectors verified that the monitoring program ensured the assumptions in the engineering analysis remain valid. The inspectors also verified that the licensee was maintaining acceptable levels of availability, reliability, and performance per its license condition. When the established levels of availability, reliability, and performance were not met, the inspectors verified that the licensee took appropriate corrective actions to return to the established levels.

b. Findings

Introduction.

The inspectors identified a Green non-cited violation of License Condition 3.D, "Fire Protection Program," for the failure to establish an appropriate monitoring program in accordance with NFPA 805, Section 2.6. Specifically, the licensee failed to set the action level for the availability of the raw water system pumps to ensure that the assumptions in the engineering analysis remained valid.

Description.

The inspectors reviewed selected samples of equipment monitored by the licensee utilizing Procedure ER-AA-610-1003, "NFPA 805 Monitoring Program,"

Revision 0, to ensure that the licensee's program properly implemented the requirements of NFPA 805, Section 2.6.

Licensee letter LIC-11-0099, Section 4.6.2, "Overview of Post-Transition NFPA 805 Monitoring Program," states that, "Monitoring will ensure that assumptions in engineering analysis remain valid. The monitoring program will be documented in an administrative process (i.e., program manual or directive) that provides the process and sets clear guidelines to consistently measure the performance of the fire protection program. Table S-3, "Implementation Items", item REC-101, states: "Development of the NFPA 805 monitoring program for FCS utilizing guidance from FAQ [Frequently Asked Question] 10-0059 (as described in Section 2.6 of NFPA 805)." Frequently Asked Question 10-0059, "NFPA 805 Monitoring," Revision 5, states in the Details section, Item 3, "Action level threshold - When establishing the action level threshold for reliability and availability, the action level should be no lower than the fire PRA [probabilistic risk assessment] assumptions."

For the raw water system pumps, the licensee opted to use their maintenance rule program as the monitoring program. The inspectors noted that the maintenance rule program allowable availability was 95 percent. However, the fire probabilistic risk assessment basic event for test and maintenance unavailability was 0.0134, which relates to an availability of 98.66 percent. Therefore, the action level for availability in the monitoring program was lower than the assumptions in the fire probabilistic risk assessment.

Equipment Reliability Procedure ER-AA-610-1003, "NFPA 805 Monitoring Program," Section 2.1, defines Action Level, "For those SSCs monitored by the NFPA 805 Monitoring Program, the action level is the threshold at which mitigating actions are required to keep assumptions made in the engineering analysis valid." Section 4.4 "Risk Target Value Determination," Item 4.4.1, states, "Using the Fire PRA, or other processes as appropriate, to determine target values of reliability and availability for the HSS fire protection/NSCA SSCs and programmatic elements established in Section 4.3 as requiring additional monitoring beyond inspection and test programs and system/ program health programs." Item 4.4.4 states, "If HSS SSCs have been identified using the maintenance rule guidelines, then the associated SSC specific performance criteria may be ESTABLISHED as in the Maintenance Rule, provided the criteria are consistent with Fire PRA assumptions." The fourth criteria provided states, "If the maintenance rule criteria does not bound the Fire PRA assumptions, then SUBMIT a Maintenance Rule Change request per reference 6.2.5." The inspectors noted that the licensee had not submitted a maintenance rule change request despite the maintenance rule criteria failing to bound the fire probabilistic risk assessment assumptions.

Analysis.

The failure to set the action level for the availability of the raw water system pumps to ensure that the assumptions in the engineering analysis remained valid was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency could adversely affect the acceptable level of availability of the raw water pumps, which are used to respond to fire initiating events.

The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," Attachment 4, "Initial Characterization of Findings," dated June 19, 2012. The inspectors determined that an IMC 0609, Appendix F, "Fire Protection Significance Determination Process," dated September 20, 2013, review was required as the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire. Using IMC 0609, Appendix F, Attachment 1, "Fire Protection Significance Determination Process Worksheet," dated September 20, 2013, the finding was screened as a Green finding of very low safety significance in accordance with Step 1.3, "Ability to Achieve Safe Shutdown," B Question. Based on the criteria in Appendix F, Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," dated February 28, 2005, the finding was assigned a "Low" degradation rating. Using Table A2.3, the inspectors considered a low degradation rating because the issue involved monitoring of components that did not degrade below acceptable levels during the exposure period.

This finding had a cross-cutting aspect associated with change management within the human performance area since the leaders failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the inspectors determined that the licensee did not use the process that was in place to ensure that the appropriate NFPA 805 fire risk assessment monitoring action levels were incorporated into the maintenance rule program and

monitored (H.3).

Enforcement.

License Condition 3.D, "Fire Protection Program," states, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 28, 2011 (and supplements, dated December 19, 2011, December 22, 2011, March 20, 2012, July 24, 2012, August 24, 2012, September 27, 2012, April 23, 2013, May 21, 2013, July 29, 2013, September 12, 2013, October 11, 2013, November 4, 2013, November 11, 2013, December 18, 2013, January 24, 2014, February 28, 2014, April10, and June 11, 2014), and as approved in the safety evaluation, dated June 16, 2014.

The approved fire protection program is a risk-informed performance-based program in accordance with NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition" (NFPA 805), incorporated by reference into 10 CFR 50.48(c). NFPA 805, Section 2.6, "Monitoring," states that, "Monitoring shall ensure that the assumptions in the engineering analysis remain valid." Contrary to the above, from June 16, 2015, to November 6, 2015, the licensee failed to appropriately monitor plant equipment to ensure that the assumptions in the NFPA 805 engineering analysis remained valid. Specifically, the licensee's monitoring program did not monitor the availability of the raw water system pumps to ensure that it was no lower than the fire probabilistic risk assessment assumptions.

Because the finding was of very low safety significance (Green), was entered into the licensee's corrective action program as Condition Report 2015-12612, the availability of the pumps was greater than that assumed in the fire probabilistic risk assessment, and the licensee was in the process of developing corrective actions to address the monitoring of the pumps, this finding is being treated as a non-cited violation consistent with Section 2.3.2.a of the NRC Enforcement policy. This finding is identified as NCV 05000285/2015012-03, "Failure to Set Action Levels to Ensure that the

Assumptions in the Engineering Analysis Remain Valid."

.14 Plant Change Evaluation

a. Inspection Scope

The inspectors reviewed plant change evaluations to verify that where performance-based methods were applied, the methods adequately represented plant design and conditions in the fire area, were performed by qualified persons, were acceptable for the application and met the requirements of the fire protection license condition for self-approved changes to the fire protection program. The licensee had not yet implemented the plant changes. Therefore, the inspectors reviewed a sample of engineering changes planned by the licensee. Additionally, the inspectors reviewed the governing procedures related to engineering changes and the requirements for performing plant change evaluations.

The inspectors sampled the following changes:

REC-111, high energy arc fault shielding to limit damage.

REC-114, provide protection for SOV HCV-247 control cables to prevent spurious isolation of CVCS injection flow path.

REC-116, modification to ensure that diesel generator 1 supply breaker cannot close unless breakers 1A3-1 and 1A3-3 are open. Offsite power can potentially be lost with fire in DG1 area and diesel connects to grid.

REC-131, provide nitrogen backup to valve YCV-1045 so that pump FW-10 can operate for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. Findings

No findings were identified.

.15 Implementation of Risk-Related Implementation Items

a. Inspection Scope

The inspectors verified that the licensee appropriately implemented risk-related items in the establishment and early operation of their NFPA 805 program. This review included the risk management policy for fire brigade emergency response, upgraded fire human reliability analysis methods, alternate steam generator level monitoring methods, and incorporation of revised fire risk methods in the fire probabilistic risk assessment. The inspectors also reviewed the licensee's integration of multi-compartment scenarios into the base fire probabilistic risk assessment. Finally, the inspectors reviewed the licensee's probabilistic risk assessment to ensure the basic event data in the fire probabilistic risk assessment matched the internal events probabilistic risk assessment basic events, which included a verification of multiple spurious operation and human reliability analysis treatment.

The inspectors sampled the following alternate methods to ensure that the licensee appropriately replaced them with approved methods:

Fire Pump Frequency Apportioning The licensee based apportioning on pump runtime to account for lower usage with standby pumps. Standard NFPA 805 suggested apportioning on pump count alone. The licensee had performed sensitivity studies using pump count apportioning. Acceptable results were achieved.

Diesel Generator Fire Treatment The licensee assumed that most fires occur during surveillance testing. Therefore, it was assumed that an operator would be physically stationed in the room. This artificially improved manual suppression credit. Standard NFPA 805 suggested that surveillance testing is indicative of demand runs and likely will not have manual suppression. The licensee had performed sensitivity indicated studies indicating that the acceptance criteria were still met.

Use of Draft Frequently Asked Question 08-0050 The licensee indicated there was no change to their conclusions given implementation of the final frequently asked question.

Electrical Cabinet Generic Severity Factor The licensee assumed that 10 percent of fires will damage targets external to the cabinet. However, the acceptance criteria were still met when the severity factor was set to 1.0.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

K. Ahrens, Fire Protection System Engineer

C. Averett, Site Communicator

D. Bacalan, Site Security Manager

R. Beck, Chemistry Manager

B. Bittle, Design Engineer

B. Blame, Site Regulatory Assurance Manager

D. Brehm, Radiological Protection Supervisor

C. Cameron, Regulatory Assurance

H. Childs, Site Security Manager

G. Coleman, Training

L. Cortopassi, Site Vice President and Chief Nuclear Officer

S. Dean, Plant Manager

S. Fatora, Work Management Director

J. Geschwender, Probabilistic Risk Assessment Engineer

H. Goodman, Site Engineering Director

H. Hackeroft, Probabilistic Risk Assessment Engineer

B. Heimes, Security

R. Hugenroth, Nuclear Oversight Manager

T. Kapuan, Maintenance Director

J. LeMaire, Fire Protection Engineer (EPM Contractor)

S. Lindquist, Unit Supervisor

E. Matzke, Regulatory Assurance

J. McManuis, Project Engineering Manager

H. Minassian, Program Engineer

T. Muff, Maintenance

F. Pellizzori, Safe Shutdown Analyst (EPM Contractor)

C. Pragman, Exelon Corporate Fire Protection

M. Prospero, Executive Outage Manager

A. Rochon, Westinghouse (Fire PRA Support)

T. Swanson, Fire Marshall

M. Seip, Fire Brigade Instructor

A. Sublett, NFPA 805 Safe Shutdown Engineer

S. Swanson, Operations Director

R. Swerczek, Fire Protection Engineer

J. Tarpinian, NFPA 805 Fire Protection Engineer (EPM Contractor)

T. Uehling, Training

J. Wiegand, Operations Support and Services Manager

J. Wilson, Design Engineer

Z. Wineinger, Design Engineer

NRC Personnel

S. Schneider, Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000285/2015012-01 NCV

Failure to Adequately Implement and Maintain Required NFPA 805 Implementation Items (Section 1R05.01.b)

05000285/2015012-02 NCV

Failure to Provide Adequate Isolation for Pressurizer Heaters (Section 1R05.05.b)

05000285/2015012-03 NCV

Failure to Set Action Levels to Ensure that the

Assumptions in the Engineering Analysis Remain Valid

(Section 1R05.13.b) `

LIST OF DOCUMENTS REVIEWED

Cable Routing Data Components

EC399

ED363A

ED362

EC391A

EC391

EC392

ED363

EB365

EA396

ED985

Calculations

Number Title Revision

CN-RAM-09-021 Fire Ignition Frequencies for Fort Calhoun Station 2

CN-RAM-09-026 Fire Human Reliability Analysis for Fort Calhoun Station 2

CN-RAM-09-048 Fire PRA Plant Response Model for Fort Calhoun Station 2

CN-RAM-10-013 Fire Scenario Selection and Characterization for Fort Calhoun Station and 2

CN-RAM-10-015 Fire PRA Qualitative Screening, Quantitative Screening, Quantification, and Uncertainty Analysis for

Fort Calhoun Station and 4

CN-RAM-10-021 Fire Ignition Frequencies for Fort Calhoun Station 3

CN-RAM-10-025, Attachment 9 Fire Risk Assessment of NFPA 805 Variances from the Deterministic Requirements in Fire Compartment 36B

2C

CN-RAM-10-025, Attachment 11 Fire Risk Assessment of NFPA 805 Variances from the Deterministic Requirements in Fire Compartment 42

2C

CN-RAM-10-025, Attachment 12 Fire Risk Assessment of NFPA 805 Variances from the Deterministic Requirements in Fire Compartment 43

2C FC05690 Battery Load Profile and Voltage Drop Calculation 11 FC6506 Hydraulic Model of Fire Protection Water Supply System 8 FC07869 NFPA 805 Recovery Actions Evaluation at FCS for

EPU 0

Condition Reports

2014-14866

2015-01736

2015-07943

2015-08746

2015-08835

2015-09850

2015-11044

2015-11076

2015-11079

2015-11080

2015-11088

2015-11901

2015-11345

2015-11346

2015-11347

2015-11519

2015-11525*

2015-11566*

2015-11702

2015-11750*

2015-11755*

2015-11758*

2015-11821*

2015-11819*

2015-11904*

2015-11905*

2015-11935*

2015-12107*

2015-12057*

2015-12120*

2015-12173*

2015-12144

2015-12156*

2015-12158*

2015-12160*

2015-12173

2015-12190*

2015-12195*

2015-12204*

2015-12208*

2015-12221*

2015-12223

2015-12436

2015-12254

2015-12444

2015-12465

2015-12505

2015-12535

2015-12590*

2015-12599*

2015-12604*

2015-12611*

2015-12612*

2015-12620*

2015-12631*

2015-12623*

2015-12634*

2015-12653*

2015-12657*

2015-12662*

2015-12665*

2015-12730

2015-12789

2015-13041

Issued as a result of inspection activities

Drawings

Number Title Revision B-4250,

Sheet 17 Cable Block Diagram

FCV-1368 3

B-4250,

Sheet 260 Cable Block Diagram

YCV-1045 2

B-4250,

Sheet 261 Cable Block Diagram,

YCV-1045A 3

B120F11503, Sheet 2 Emergency Generators Schematic Diagram, A1-132A

& A1-133B

B120F14501, Sheet 1 Schematic Engine Control 15

B120F15503, Sheet 1 Schematic 480 VAC Auxiliary System 18

B120F15503, Sheet 2 Emergency Generators 480 VAC Auxiliary System Schematic

Drawings

Number Title Revision B-4347,

Sheet 1 Penetration Typical Drawing Conduit/Tubing/Piping thru

Sleeve or Corebore Mechanical or Electrical

Seal-Floor/Wall

B-4347,

Sheet 2 Penetration Typical Drawing Conduit/Tubing/Piping thru Sleeve or Corebore Mechanical or Electrical Seal-

Floor/Wall

B-4339,

Sheet 1 Penetration Typical Drawing Single Conduit - Sleeve or Corebore-9/16" or Less Average Annulus Electrical Seal Floor/Wall

B-4342,

Sheet 1 Penetration Typical Drawing Cable Trays Thru Walls of 11" or Greater Thickness Electrical Seal-Wall

CC-FC-211 Fire Protection Diagram 0

D-4098,

Sheet 1 Auxiliary Building Fire Areas Operating Level 3

D-4098,

Sheet 2 Auxiliary Building Fire Areas Ground Level 6

D-4098,

Sheet 3 Auxiliary Building Fire Areas Basement Level 1

D-4098,

Sheet 4 Auxiliary Building Fire Areas Intermediate Level 3 D-4665

DG-1 Diesel Generator one-line Diagram P&ID 6 D-4666

DG-2 Diesel Generator one-line Diagram P&ID 6 E-4113 Auxiliary Bldg. Electrical Partial Plan,

EL 1036-0"

3

EM-001/010, Sheet 4 Instrument and Control Equipment List 8

GHDR Inc., Fig. 7.3-2, Sheet 2

Simplified Master Diagram Functional Circuit Logic Engineered Safeguard Signals Spec. No. 17.2, Sheet Control Valves 20

USAR-FIG 8.1-1 Simplified One Line Diagram Plant Electrical System 16

23R0454, Sheet 24 Bus No. 1A2, Power Control Circuit Unit 1A2-9 Heater Drain Pump No.

FW-58 FDR

0108D8660 Metal Clad Switchgear Interconnection Diagram 7

Drawings

Number Title Revision 0108D8674, Sheet 3 Metal Clad Switchgear Interconnection Diagram 10

23R0456, Sheet 19 Bus No. 1A4 Power & Control Circuit Unit 1A4-10 L.P. Safety Injection Pump

SI-1B

0223R0456, Sheet 24 Bus No. 1A4 Power & Control Circuit Unit 1A4-10 Transfer T1B-4A Feeder

09313031-WD-1 Wiring Diagram for 5GSB2 1200 and 2000A

Replacement Breakers

11405-E-3 4.16KV. Auxiliary Power, One Line Diagram P &

ID 13

136B22432, Sheet 3 Elementary Diagram Switch Developments 16

136B2341, Sheet 35 Elementary Diagram Electrical Control 480V System 22

136B2432, Sheet 1 Elementary Diagram Switch Developments 26

136B2432, Sheet 12 Elementary Diagram Switch Developments 4

136B2432, Sheet 14 Elementary Diagram Switch Developments 7

136B2432, Sheet 34 Elementary Diagram Switch Developments 17

136B2432, Sheet 40 Elementary Diagram Switch Developments 16

136B2432, Sheet 44 Elementary Diagram Switch Developments 23

136B2432, Sheet 44A Elementary Diagram Switch Developments 3

136B2432, Sheet 7 Elementary Diagram Switch Developments 17

136B2492, Sheet 28 Elementary Control Valves and Pumps Elementary Diagram 11 136B2492, Sheet 110 Elementary Control Valves and Pumps Elementary Diagram 6

Drawings

Number Title Revision 136B2492, Sheet 111 Elementary Control Valves and Pumps Elementary Diagram 5 136B2492, Sheet 31 Elementary Diagram Electrical Control Valves and Pumps 6 136B2523, Sheet 17 Electrical Control Auto Start Circuit Elementary Diagram 4 136B2736, Sheet 9 Elementary Control Valves and Pumps Elementary Diagram 14 136B2736, Sheet 12 Elementary Control Valves and Pumps Elementary Diagram 12 136B2736, Sheet 13 Elementary Control Valves and Pumps Elementary Diagram 10 136B2736, Sheet 15 Elementary Diagram Emergency Manual Transfer Switch 3 136B2736, Sheet 2 Elementary Diagram Electrical Control Valves and Pumps 8 136B2736, Sheet 7 Elementary Diagram Electrical Control Valves and Pumps 7 161F531, Sheet 6 Three Line Diagram Main Three Line Diagram 37

161F532, Sheet 1 Elementary Diagram Main Breaker Control - 4.16 kV 37

161F532, Sheet 9 Elementary Diagram Main Breaker Control - 4.16 kV 19

161F561, Sheet 27A Interconnection Diagram 10

161F561, Sheet 75A Interconnection Diagram Wide Range Excore Neutron Detectors

161F592, Sheet 1 Connection Diagram

AI-179 41

161F593, Sheet 2 Connection Diagram

AI-179 20

11405-E-7, Sheet 1 480 Volt Primary Plant Motor Control Center One Line Diagram, Sheet # 2A P & ID

11405-E-8, Sheet 2 125V DC Misc. Power, Distribution Diagram, P&ID 13

Drawings

Number Title Revision 11405-E-19, Sheet 7 Schematic Diagram 480V But Tie BKR 1B4C 2

11405-E-32, Sheet 1 Pressurizer Heaters Wiring Diagram: MCC, Cables, Terminal Blocks, Heater Arrangement

11405-E-45, Sheet 2 MCC Auto Load Shed Channel "B" Schematic Diagram 41 11405-E-45, Sheet 4 Aux. Feedwater Control Valves to

RC-2A &

RC-2B 35

11405-E-52, Sheet 6

RC-4-HTRS Pressurizer Backup Heaters Bank #4 23

11405-E-52, Sheet 6

RC-4-HTRS Pressurizer Backup Heaters Bank #4 23

11405-E-67, Sheet 31 Cable Tray Section 0

11405-E-67, Sheet 81 Cable Tray Section 1

11405-E-74, Sheet 1 Main Steam Aux. Piping & Control Room Area-Plan,

El. 1036'-0"

11405-E-78, Sheet 1 Cable Room Tray and Conduit Layout Plan, EL. 1025'-

0" & Sections

11405-E-199 Alternate Shutdown Panel

AI-185 Relay Schematic 10

11405-E-137, Sheet 1 Schematic, Wiring Diagram & Switch Developments for Control Valve

YCV-1045, to Steam Driven Aux.

Feed Water Pump FW-10.

11405-E-143, Sheet 2 480V SWGR 1B4B, BKR 1B4B-1 (Unit 403B)

Schematic

SI-3B

11405-E-143, Sheet 5 480V SWGR 1B4C, BKR 1B4C-6 Schematic

CH-1B 7

11405-E-330, Sheet 4 Wiring Diagram, Terminal Box 56T thru 57T Heater

Dr Condensate Pump Push Button Station

11405-E-360, Sheet 2

VA-52A-MS &

VA-528-MS, Internal Connection Diagram 10 11405-E-360, Sheet 3

ATA-D1 &

ATA-D2 480 V AC Automatic Transfer Switch Schematic

Drawings

Number Title Revision 11405-E-360, Sheet 4 Diesel Air Compressor

SA-1-1,

SA-1-2 Aux. Diesel Schematic

11405-E-360, Sheet

6 A1-41A-MTS, A1-41B-MTS, 1A1-1A3-MTS &

1A2-1A4-MTS Transfer Switch Control Schematics

11405-E-360, Sheet

7

ATD-D1-MTS, 1B3A-4A-MTS, 1B3B-4B-MTS &

1A2-B3C-4C-MTS Transfer Switch Control Schematics

11405-M-133 Reactor Coolant System Flow Diagram P&ID 84 11405-M-136 Chemical and Volume Control System 56 P&ID 78

11405-M-137 Chemical and Volume Control System P&ID 26

11405-M-142 Chemical and Volume Control System P&ID 15

11405-M-144 Safety Injection and Containment Spray System Flow Diagram P&ID

118 11405-M-252 Composite Flow Diagram, Main Steam, P&ID 61 11405-M-253, Sheet 1 Flow Diagram Steam Generator Feedwater and Blowdown P&ID

11405-M-253, Sheet 2 Flow Diagram Steam Generator Feedwater and Blowdown P&ID

11405-M-253, Sheet 3 Flow Diagram Steam Generator Feedwater and Blowdown P&ID

11405-M-253, Sheet 4 Flow Diagram Steam Generator Feedwater and Blowdown P&ID

11405-M-253, Sheet Cover Composite Flow Diagram Steam Generator Feedwater and Blowdown P&ID

1244-01 Radio System Block Diagram 2 13007.42-EE-42B Wiring Diagram Alternate Shutdown PNL.

AI-185 24

13007.42-ESK-4A, Sheet 3 Equipment List 11

13007.42-ESK-4A, Sheet 4

11 13007.42-ESK-4A, Sheet 5 Nameplate Legend for

AI-185 9

Engineering Information Records Number Title Revision

EC-95-22 NFPA Code Compliance 4

EC-43758 Loop 106 Upgrade, Replace Transmitter, Manifold & Indicator

EC-63269 Replace Cable EC68 for the

AC-1

OC-M Raw Water Pump Motor

Miscellaneous Documents

Number Title Revision/Date Design Analysis

EA 10-036

Fort Calhoun Station Automation and Update of Safe Shutdown Analysis Design Analysis

EA 10-037

Fort Calhoun Station

NFPA-805 NSPC and Fire PRA Circuit Analysis, Cable Selection and Cable Location 0 Design Analysis

EA 10-041 NFPA 805 recovery Action Feasibility Assessment Design Analysis

EA 10-042 Non-Power Operations Mode Transition Report 0

Design Analysis

EA 10-044 Fort Calhoun Station Fire Area Review / Licensing Action Review Design Analysis

EA 11-013 NFPA 805 Risk Informed Performance Based Fire Risk Evaluation for Fire Area 34B-1

Design Analysis

EA 11-015 NFPA 805 Risk Informed Performance Based Fire Risk Evaluation for Fire Area 36B

Design Analysis

EA 11-017 NFPA 805 Risk Informed Performance Based Fire Risk Evaluation for Fire Area 42

Design Analysis

EA 11-018 NFPA 805 Risk Informed Performance Based Fire Risk Evaluation for Fire Area 43

Design Analysis

EA 15-015 NFPA 805 Fire Protection Monitoring Program 0

Design Analysis

EA 97-001 Updated Fire Hazard Analysis (UFHA) 19

Miscellaneous Documents

Design Analysis

EA 99-023 Fire Suppression Effects Analysis 1

Email From L. Wilkins to

J. Sebrosky Fort Calhoun Station RAIs Re:

NFPA-805 (ME7244) December 23, 2013 EPM Report R2407-

006-001 Task 6 Review of Technical Specifications (Implementation Item From FCS

LAR 10-07 Attachment S-3, Item #

REC-138)

Event Notification

51487

October 21, 2015

FAQ 07-0035 Bus Duct Counting Guidance for High Energy Arcing Faults 2

FAQ 13-0005 Cable Fires Special Cases: Self Ignited and Caused by Welding and Cutting General Form

FC-66 Temporary Modification Control Form 69 NRC Letter ML14098A092 Fort Calhoun Station - Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance With 10

CFR 50.48(c)

(TAC No. ME7244) June 16, 2014

OPPD Letter

LIC-13-0060 Remaining Responses to Second Request for Additional Information Re: License Amendment Request to Adopt NFPA 805 at Fort Calhoun Station (TAC No. ME7244) May 21, 2013

OPPD Letter

LIC-14-0004 Probabilistic Risk Assessment (PRA) RAO Responses 26.01 and 24.01 -

NFPA-805 Transition

(ME7244) January 24, 2014 Project Document

649-30-06 Initiator Frequency Data Notebook 8

Regulatory Guide

1.174 An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-

Specific Changes to the Licensing Basis

RERP-Appendix A

Blair Fire Department Aid Agreement June 6, 2006

Miscellaneous Documents

SDBD-DG-112 Emergency Diesel Generators 31

SDBD-FP-115 Fire Protection 32

SDBD-FW-AFW-117 Auxiliary Feedwater 45

SDBD-EE-200 120 VAC Vital Distribution 18

SDBD-EE-201 AC Distribution 24

SDBD-EE-202 DC Distribution 18

USAR-8.5 Electrical Systems, Initial Cable Installation Design Criteria 14

USAR-9.11 Auxiliary Systems, Fire Protection System 28 Westinghouse Letter

LTR-RAM-II-15-006 Focused Scope Fire PRA Peer Review 0

Fort Calhoun Fire & Rescue Agreement January 1, 2015

Fire Protection System Health Report 1

st Quarter 2015 September 10, 2015

Fire Protection System Health Report 4

th Quarter 2014 September 10, 2015

NFPA 805 Implementation Items

REC-001

REC-002

REC-004

REC-006

REC-008

REC-011

REC-012

REC-021

REC-027

REC-028

REC-031

REC-032

REC-034

REC-035

REC-037

REC-039

REC-050

REC-055

REC-059

REC-067

REC-068

REC-086

REC-094

REC-095

REC-097

REC-100

REC-101

REC-102

REC-103

REC-104

REC-105

REC-106

REC-107

REC-108

REC-109

REC-110

REC-134

REC-136

REC-139

REC-140

REC-141

REC-142

REC-144

REC-145

REC-147

REC-149

REC-150

REC-151

REC-152

REC-153

REC-154

REC-155

REC-156

REC-157

REC-158

REC-159

Procedures

Number Title Revision

AOP-06 Fire Emergency

29

AOP-06-02 Fire Emergency Uncontrolled Areas of Auxiliary Building

6

AOP-17 Loss of Instrument Air 16

AOP-30 Emergency Fill of Emergency Feedwater Storage Tank 13

CC-AA-102 Design Input and Configuration Change Impact Screening

CC-AA-112 Temporary Configuration Changes 21

CC-AA-209 Fire Protection Program Configuration Change Review 5

CC-AA-209-1001 Guidelines for Performing Fire Protection Program Configuration Change Review

CC-FC-102 Design Input and Configuration Change Impact Screening

CC-FC-211 Fire Protection Program 0

EN-FC-402-0002 Storm Water Pollution Prevention Plant 0

EM-ST-060FDZH-QA Calibration and Functional Test of QA Vault Halon Fire Detection and Protection System

EM-ST-06-FDZ-TG Disassembly, Reassembly, and Testing of Turbine Generator Fire Detector System

EM-ST-EE-0011 Annual Testing of Emergency Lighting Panels 9

EP-FC-124-AD-F-01 Control Room Inventory 1

ER-AA-310-1003 Maintenance Rule - Performance Criteria Selection 4

ER-AA-600-1068 Fire PRA Risk Evaluation Support for NFPA 805

Program 1

ER-AA-610 NFPA 805 Monitoring Program 0

ER-AA-610-1003 NFPA 805 Monitoring Program 0

GM-ST-FP-0002 Fire Door Eighteen Month Inspection 13

Procedures

GM-ST-FP-0004 Semi-Annual Control Room Halon Cylinders Weight and Pressure Test

GM-ST-FP-0005 Semi-Annual Switchgear Rooms Halon Cylinders

Weight and Pressure Test

24a

GM-ST-FP-0006 Fire Damper Eighteen Month Inspection 7

IC-ST-FP-0001 Calibration and Functional Test of Auxiliary Building, Elevation 1036' (Room 81) Fire Detection System

(AI-230 and

AI-231)

LA-AA-128-101 Regulatory Review of Proposed Changes to the Approved NFPA 805 Fire Protection Program

MM-ST-FP-0001 Inspection of Diesel Fire Pump Engine 18

MM-ST-FP-0002 Inspection of Fire Pump Strainer

FP-6A 17

MM-ST-FP-0003 Inspection of Fire Pump Strainer

FP-6B 5 N-AN-ENG-CERT-RM09 Engineering Training Certification Guide for

N-AN-ENG-CERT-RM09 Risk Management Engineer -

Fire External Events

NO-FC-10 Quality Assurance Topical Report (QATR) 3

OCAG-1 Operational Contingency Action Guideline 28

OP-AA-201-001 Fire Marshal Tours 6

OP-AA-201-002 Fire Event Reports 5

OP-AA-201-003 Fire Drill Performance 14

OP-AA-201-004 Fire Prevention For Hot Work 12

OP-AA-201-005 Fire Brigade Qualification 8

OP-AA-201-006 Control of Temporary Heat Generating Equipment 8

OP-AA-201-009 Control of Transient Combustible Material 17

OP-AA-201-010-1001 B.5.B Mitigating Strategies Equipment Expectations 3

OP-AA-201-011-1001 Fire Marshal Certification Process 2

OP-ST-ASP-0002 Surveillance Test Alternate Shutdown Capability Control Circuitry Verification

Procedures

OP-FT-DG-0001 Master Electrical Switch 183-MES/D2 Function Test 12

OP-ST-FP-0001A Fire Protection System Inspection and Test 19

OP-ST-FP-0001C Fire Protection System Inspection and Test 25

OP-ST-FP-0001D Fire Protection System Inspection and Test 31

OP-ST-FP-0002 Fire Protection Water Suppression System Valve Cycling Test

OP-ST-FP-0003 Fire Protection System Diesel Generator Rooms Sprinkler Functional Test

OP-ST-FP-0004 Battery Powered Smoke Detector Functional Test 8

OP-ST-FP-0008 Fire Protection System Auxiliary Building Sprinkler Functional Test

OP-ST-FP-0009 Fire Protection System Fire Hose Inspection 10

OP-ST-FP-0010 Fire Protection System Sprinkler System Air Flow Test 10

OP-ST-FP-0011 Fire Protection System Hose Station Operability Test 10

OP-ST-FP-0013 Fire Protection System Turbine Building Sprinkler Systems Monthly Inspection

OP-ST-FP-0019 Fire Protection System Motor Driven Fire Pump Full Flow Test

OP-ST-FP-1000 Quarterly Fire Protection Drain Valve Flush and Alarm Test 0

OP-ST-HG-0001 Hydrogen Bottle Dock Inspection and Maintenance Test 0

OP-PM-FP-1001A Monthly Fire Protection System Inspection (Week 1) 34

OU-AA-103 Shutdown Safety Management Program 15

PE-RR-AE-1001 Flood Barrier and Sandbag Staging and Installation 22

PRA-PR-5 Systems Analysis Procedure 4

SA-AA-129-2118 Management and Control of Temporary Power 8

SE-ST-FP-0005 Fire Barrier and Penetration Seals Eighteen Month Inspection

Procedures

SO-G-28 Station Fire Plan 89

SO-G-74 Fort Calhoun Station EOP/AOP Generation Program 22

SO-G-96 Planned LCO Entry Criteria and Equipment Reliability Control 15

SO-O-1 Conduct of Operations 107

SO-O-41 Control of Operator Aids and Emergency Equipment 141

SO-G-102 Fire Protection Program Plan 20

SO-G-103 Fire Protection Functionality and Surveillance Requirements

SO-O-21 Shutdown Operations Protection Plan 55

SO-O-25 Temporary Modification Control 86

Initial and Annual Re-Qual Fire Brigade Training 0

Vendor Documents Number Title Revision TD E155.0020 Series 24 Instrument & Control Switches for Power Industry & Heavy Duty Industrial Applications

IM-09313031-1 Instruction Manual for Square D Magnum Circuit Breakers 4 TD A499.0030 3 and 4 Way Solenoid Valves for Pilot Control of Diaphragm & Cylinder Operated Valves in Nuclear

Power Plants

TD E155.0030 High Speed Multi-Contact Lockout Relays for Power Industry Application

Work Orders

00528260

00458293

00485992

00547494

00526289

00495997

00459107

00520797

00493412

00543564

00136878

00481787

00490734

00545104

00423873

00468421

00474319

00489780

00502848

00517076

00521343

00524835

00532524

00532526

00535870

00542289

00544555

00423873

00433574

00454813

00474319

00489480

00502729

00515670

00517076

00521106

00521344

00523555

00532522

00535833

00536697

00511394

00513561

00519996

00532520

00553033

00427081

00495989

00428827

00465988

00541251

00542788

00540285

00545024

00547850

00427634

00527255

00435313

00465191

00495803

00499992

00522079

00470554

00538685

00530856

00559496

00545026

00540965

00543971

00544556

00305165

IR 05000285/2015012: Difference between revisions

Revision as of 20:44, 16 September 2018

Text

Dear Mr. Cortopassi:

Enclosure:

w/Attachment:

SUMMARY

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

REPORT DETAILS

REACTOR SAFETY

1R05 Fire Protection

.01 Protection of Safe Shutdown Capabilities

a. Inspection Scope

b. Findings

Introduction.

Description.

Analysis.

Analysis.

Enforcement.

.02 Passive Fire Protection

a. Inspection Scope

b. Findings

.03 Active Fire Protection

a. Inspection Scope

b. Findings

.04 Protection from Damage from Fire Suppression Activities

a. Inspection Scope

b. Findings

.05 Shutdown from a Primary Control Station

a. Inspection Scope

b. Findings

Introduction.

Description.

Analysis.

Analysis.

Enforcement.

.06 Circuit Analyses

a. Inspection Scope

b. Findings

.07 Communications

a. Inspection Scope

b. Findings

.08 Emergency Lighting

a. Inspection Scope

b. Findings

.09 Cold Shutdown Repairs

a. Inspection Scope

b. Findings

.10 Compensatory Measures

a. Inspection Scope

b. Findings

.11 Radiological Release

a. Inspection Scope

b. Findings

.12 Non-Power Operations

a. Inspection Scope

b. Findings

.13 Monitoring Program

a. Inspection Scope

b. Findings

Introduction.

Description.

Analysis.

Enforcement.

.14 Plant Change Evaluation

a. Inspection Scope

b. Findings

.15 Implementation of Risk-Related Implementation Items

a. Inspection Scope

b. Findings

.16 Alternative Mitigation Strategy Inspection Activities

a. Inspection Scope

b. Findings

OTHER ACTIVITIES (OA)

4OA2 Identification and Resolution of Problems

a. Inspection Scope

b. Findings

4OA6 Meetings, Including Exit

Exit Meeting Summary