ML13357A748: Difference between revisions

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{{Adams
#REDIRECT [[ULNRC-06060, Callaway, Unit 1, Enclosure 1 to ULNRC-06060 - Request for Additional Information Set #5 Response]]
| number = ML13357A748
| issue date = 12/19/2013
| title = Callaway, Unit 1, Enclosure 1 to ULNRC-06060 - Request for Additional Information Set #5 Response
| author name =
| author affiliation = Ameren Missouri, Union Electric Co
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000483
| license number = NPF-030
| contact person =
| case reference number = ULNRC-06060
| package number = ML13357A791
| document type = - No Document Type Applies
| page count = 777
}}
 
=Text=
{{#Wiki_filter:Enclosure 1 to ULNRC-06060 Page 1 of 2
 
REQUEST FOR ADDITIONAL INFORMATION (RAI) SET #5 RESPONSE
 
to ULNRC-06060  Page 2 of 2 Fire Protection Engineering RAI 20
 
In Attachment X of the License Amendment Request (LAR) dated August 29, 2011, (ADAMS Accession No. ML112420048), Approval Request 1 requests the elimination of the requirement to enter Technical Specification 3.0.3 for an inoperable fire suppression water system coupled with inability to establish a backup fire protection water system within 24 hours. The request indicates that the procedure that is currently used to address compensatory actions, does not address the National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants - 2001 Edition," requirements. The request states that the procedure "...will be used to establish the required compensatory actions and impairment durations following the transition to NFPA 805." However, there is no implementation item associated with the revision of this procedure to meet the requirements of NFPA 805 Section 3.2.3(2). Describe the actions needed to bring the procedure into compliance with these requirements and update LAR Attachment S, Table S-3 as necessary.
 
Callaway Response: 
 
The requirement to enter Technical Specification 3.0.3 will be removed from FSAR Table 9.5.1-2, Note 2. The NFPA 805 3.2.3(2) requirement to establish compensatory actions is currently included in procedure APA-ZZ-00703, Fire Protection Operability Criteria And Surveillance Requirements. The requirement to enter Technical Specification 3.0.3 will also be removed from this procedure. Implementation Item 13-805-008 will track these actions. LAR Attachment S, Table S-3 Implementation Items has been revised to include this new implementation item and the revised page is included in Enclosure 2 of this letter.
 
to ULNRC-06060 
 
CHANGES TO THE TRANSITION REPORT to ULNRC-06060 
 
Licensee Identified Changes To The Transition Report
 
LIC Description LAR Section LIC-1 Provided by ULNRC-05851 dated April 17, 2012  LIC-2 Provided by ULNRC-05851 dated April 17, 2012  LIC-3 Provided by ULNRC-05851 dated April 17, 2012  LIC-4 Provided by ULNRC-05851 dated April 17, 2012  LIC-5 Provided by ULNRC-05851 dated April 17, 2012  LIC-6 Provided by ULNRC-05851 dated April 17, 2012  LIC-7 Provided by ULNRC-05851 dated April 17, 2012  LIC-8 Provided by ULNRC-05851 dated April 17, 2012  LIC-9 Provided by ULNRC-05876 dated July 12, 2012  LIC-10 Provided by ULNRC-05876 dated July 12, 2012  LIC-11 Provided by ULNRC-05876 dated July 12, 2012  LIC-12 Provided by ULNRC-05876 dated July 12, 2012  LIC-13 Provided by ULNRC-05876 dated July 12, 2012  LIC-14 Provided by ULNRC-05876 dated July 12, 2012  LIC-15 Provided by ULNRC-05876 dated July 12, 2012  LIC-16 Provided by ULNRC-05876 dated July 12, 2012  LIC-17 Provided by ULNRC-05876 dated July 12, 2012  LIC-18 Provided by ULNRC-05876 dated July 12, 2012  LIC-19 Provided by ULNRC-05876 dated July 12, 2012  LIC-20 Provided by ULNRC-06011 dated August 5, 2013  LIC-21 Provided by ULNRC-06011 dated August 5, 2013  LIC-22 Based on a review of all previous RAI submittals, Ameren Missouri determined that select LAR changed pages identified in the responses to MP RAI 01, FM RAI 01c, FM RAI 01i  and Licensee Identified Changes LIC-10 and LIC-19, as transmitted in letter ULNRC-05876, were not included in the docketed response. The revised LAR pages for these previous RAI responses are included in Enclosure 2 to this letter.
LIC-23 Minor typographical errors with respect to equipment identification numbers were discovered in the LAR Transition Report Table G-1 and have been corrected in Attachment G to this letter. The errors include revising equipment ID ALV0087 to ABV0087; revising EGFHV0016 and EGFHV0054 to EGHV0016 and EGHV0054, respectively; revising the location of switch NBHS0014 from Switchgear NG02 to NB02; and revising the switch IDs GNHS009A and GNHS0017A to GNHIS0009A and GNHIS0017A. Attachment G, Table G-1 LIC-24 In Fire Area A-22, VFDR A-22-001 incorrectly states that Steam Generator A is not credited for Decay Heat Removal; however Table B-3 and Calculation KC-26 correctly state that Attachment C, Table B-3 to ULNRC-06060 Steam Generator A is credited. This is a typographical error only and is corrected as shown in this Enclosure. LIC-25 Summaries of EEEE and Licensing Actions in Table B-3 referred to fire zones as rooms (e.g., Room 1101); however, in the fire protection program, these rooms will be referred to as fire zones (e.g., Fire Zone 1101). The terminology has been revised and is included in Attachment C to this letter. There are no technical changes associated with this revision. In addition, the suppression effects discussion for select fire areas also contained the room vs. fire zone terminology. This has also been corrected. The revised pages are included in this Enclosure. Note that the entire LAR Attachment C Table B-3 is provided, not just the revised pages. Attachment C, Table B-3 LIC-26 In the original submittal, Ameren Missouri requested approval for 20-ft separation zones in Fire Areas A-1, A-16, A-27, C-1 and RB-1. It has since been determined that approval is not required for these fire areas. Fire Areas A-1, A-16 and C-1 meet the requirements for a 20-ft separation zone free of intervening combustibles. Fire Area A-27 has been revised to no longer credit a 20-ft separation zone; alternatively, VFDR A-27-014 has been developed to address separation concerns.
Fire Area RB-1 contains minor amounts of intervening combustibles within the 20-ft separation zones; these have been evaluated in VFDR RB-01/02/03/04/05-001. The FREs determined that there is no change in the total risk in Fire Area A-27 and the delta risk for this VFDR is zero. The FRE for Fire Area RB-1 has applied a bounding risk estimate; assuming delta risk is equivalent to the total risk of the area. Therefore, Approval Requests 3 and 4 of Attachment X have been rescinded. As a result, LAR Section 4.8.3, Table 4-3, Attachment C, Attachment W and Attachment X have been modified to reflect the revised methodologies for these fire areas. The revised pages are included in this Enclosure. Attachment C, Table B-3; Attachment X; Attachment W Table W-2; 4.8.3; Table 4-3 LIC-27 A drawing reference in LAR Table B-1 Section 3.10.2 was found to be in error. The drawing number has been revised from M-22KC04 to M-22KC07. The revised page is included in this Enclosure. Attachment A, Table B-1 LIC-28 Because cold shutdown is not required to achieve safe and stable or meet the performance goals of NFPA 805, Ameren Missouri has revised text to remove discussions regarding analysis to achieve cold shut down in LAR Attachment B, Table B-2. Additionally, the cold shutdown performance goals included in each fire area summary are removed from LAR Attachment C, Table B-3. Attachment B, Table B-2; Attachment C, Table B-3
 
to ULNRC-06060
 
ATTACHMENT 1: CHANGES TO THE TRANSITION REPORT MAIN BODY Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page 36  Plume Radius (Method of Heskestad)  Hot Gas Layer (Method of MQH)  Hot Gas Layer (Method of Beyler)  Hot Gas Layer (Method of Foote, Pagni, and Alvares [FPA])  Hot Gas Layer (Method of Deal and Beyler)  Ceiling Jet Temperature (Method of Alpert)  Hot Gas Layer Calculations using Fire Dynamics Simulator (Version 5)  Sprinkler Actuation Calculation using Fire Dynamics Simulator (Version 5)  Smoke Detection Actuation Correlation (Method of Heskestad and Delichatsios)  Sprinkler Activation Correlation  Control Room Abandonment Calculation using CFAST  Temperature Sensitive Equipment Hot Gas Layer Study  Temperature Sensitive Equipment Zone of Influence Study  Plume/Hot Gas Layer Interaction Study  Corner and Wall HRR  Correlation for Heat Release Rates of Cables (Method of Lee)  Correlation for Flame Spread over Horizontal Cable Trays (FLASH-CAT)  Smoke Detector Actuation using Fire Dynamics Simulator (Version 5) The acceptability of the use of these fire models is summarized in Attachment J.
For those models evaluated in NUREG 1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, April 2007, the models were evaluated to ensure they are used within the validated range. In cases where the models have been applied outside the validated range reported in NUREG 1824, the use has been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis. Technical details demonstrating the models are within range, as well as any justification of models outside the range, have been documented in R1984-001-002, "Callaway Plant Verification and Validation of Fire Modeling Tools and Approaches," Revision 1. For those models not included in the NUREG 1824 verification and validation, the bases for validation and the use of the models within the validated range, is fully documented in R1984-001-002.4.5.1.3 Results of Fire PRA Peer Review The Callaway Plant Fire PRA (Callaway Plant model of record 3Q09-FPRA) was peer reviewed against the requirements of ASME/ANS RA-Sa-2009, Part 4. The PWR Owner's Group (PWR OG) issued a report containing the results of the Callaway Plant Fire PRA Review on March 9, 2010 (LTR-RAM-II-10-019). The identification and resolution of the high level findings from the PWR OG Fire PRA Review are summarized in Attachment V.
FM  RAI 01-cFM RAI 01-iFM RAI 01-hLIC-10 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page 41 Results of Evaluation Process Disposition of VFDRs The Callaway Plant NSCA and the NFPA 805 transition project activities have identified a number of variances from the deterministic requirements of NFPA 805 Section 4.2.3. These variances were dispositioned using the fire risk evaluation process. Each variance dispositioned using a Fire Risk Evaluation was assessed against the Fire Risk Evaluation acceptance defense-in-depth and safety margin criteria from Section 5.3.5 of NEI 04-02 and RG 1.205. The results of these calculations are summarized in Attachment C. Following completion of transition activities and planned modifications and program changes, the plant will be compliant with 10 CFR 50.48(c). Risk Change Due to NFPA 805 Transition In accordance with the guidance in RG 1.205, Section C.2.2.4, Risk Evaluations, risk increases or decreases for each fire area using Fire Risk Evaluations and the overall plant should be provided. Note that the risk increase due to the use of recovery actions was included in the risk change for transition for each fire area. RG 1.205 Section C.2.2.4.2 states in part  "The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions). The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines. If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk neutral or represent a risk decrease." The risk increases and decreases are provided in Attachment W. 4.6 Monitoring Program 4.6.1 Overview of NFPA 805 Requirements for the NFPA 805 Monitoring Program Section 2.6 of NFPA 805 states: "A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid." 4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program The Monitoring program described in procedure EDP-ZZ-01101, "Fire Protection Monitoring Program Procedure," will be implemented after the safety evaluation issuance as part of the fire MP RAI 01 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page 42 protection program transition to NFPA 805  (see implementation item in Attachment S). The monitoring process is comprised of four phases. Phase 1 - Scoping  Phase 2 - Screening Using Risk Criteria  Phase 3 - Risk Target Value Determination  Phase 4 - Monitoring Implementation The evaluation conducted as described below which includes these 4 Phases will be documented in Callaway calculation KC-163,"Callaway Fire Protection Monitoring Program."
Phase 1 - Scoping The following categories of Systems, Structures and Components (
SSC's) and programmatic elements will be reviewed for inclusion in the NFPA 805 monitoring program: 1) Structures, Systems, and Components required to comply with NFPA 805, specifically:
Fire protection systems and features Required by the Nuclear Safety Capability Assessment Modeled in the Fire PRA  Required by Chapter 3 of NFPA 805  Nuclear Safety Capability Assessment equipment Nuclear safety equipment  Fire PRA equipment  Non Power Operations (NPO) equipment SSCs relied upon to meet radioactive release criteria 2) Fire Protection Programmatic Elements Phase 2 - Screening Using Risk Criteria Phase 2 of the process utilizes the risk significance of the SSC's and programmatic elements identified in Phase 1 to establish the appropriate monitoring process to be utilized. The categories of SSC's and programmatic elements from Phase 1 are each screened based on the ability to be able to explicitly establish their risk contribution.
Fire Protection Systems and Features The fire protection systems and features identified in Phase 1 as in scope are screened for risk significance. Risk significance is determined at the component, system, and/or functional level and evaluated on an individual fire area basis. Fire compartments smaller than fire areas may be used instead of fire areas provided the compartments are independent (i.e., share no fire protection SSC's) and sufficient basis is documented. The Fire PRA is used to establish the risk significance based on the following risk screening thresholds:
(AND) either Core Damage Frequency (CDF) x (RAW) E-7 per year MP RAI 01MP RAI 01MP RAI 01MP RAI 01 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page 43 (OR)  Large Early Release Frequency (LERF) x (RAW) er year CDF, LERF, and RAW (monitored parameter) are calculated for each fire area. The monitored parameter' will be established at a level commensurate with the amenability of the parameter to risk measurement (e.g., a fire barrier may be more conducive to risk measurement than an individual barrier penetration). All "required" fire protection systems and features identified as in scope in Phase 1 will be screened and categorized as either HSS or LSS and included within the appropriate monitoring process based on the risk significance. High Safety Significant (HSS) fire protection systems and features are those that meet or exceed the risk significant screening criteria. The HSS fire protection systems and features will be included in the monitoring program contained in the site Maintenance Rule Program described in procedure EDP-ZZ-01128, "Maintenance Rule Program."
Low Safety Significant (LSS) fire protection systems and features are those that do not meet the risk significant screening criteria and are monitored via the existing inspection and test programs and in the existing system/program health program described in EDP-ZZ-01131, "Plant Health and Performance Monitoring Program,". Nuclear Safety Capability Assessment Equipment For fires originating during non-power operational (NPO) modes, the qualitative use of fire prevention to manage fire risk during Higher Risk Evolutions does not lend itself to quantitative risk measurement. Therefore, for NSCA Equipment credited for NPO only, no screening is performed and fire risk management effectiveness is monitored programmatically similar to combustible material controls and other fire protection programmatic elements using the existing inspection and test programs and system/program health programs. All required NSCA equipment, except the NPO scope, identified as in scope in Phase 1 will be screened for safety significance using the fire PRA and the Maintenance Rule guidelines differentiating HSS equipment from LSS equipment. High Safety Significant (HSS) NSCA equipment are those that meet or exceed the risk significant screening criteria. The HSS NSCA equipment will be included in the monitoring program contained in the site Maintenance Rule Program. HSS NSCA equipment may already be appropriately monitored by the Maintenance Rule. A comparison of HSS NSCA equipment to the SSC's that are monitored in the Maintenance Rule program will be performed to determine what equipment will require additional NFPA 805 Monitoring. Also the review will ensure current Maintenance Rule functions are consistent with the required functions of the HSS NSCA equipment. All remaining NSCA equipment that is not screened HSS is considered to be LSS and is not included in the monitoring program. SSCs Relied upon for Radioactive Release Criteria The evaluations performed to meet the radioactive release performance criteria are qualitative in nature. The SSC's relied upon to meet the radioactive release performance criteria are not amenable to quantitative risk measurement. Additionally, since 10 CFR Part 20 limits (which are lower than releases due to core damage and containment breach) for radiological effluents are not being exceeded, equipment relied upon to meet the radioactive release performance criteria is considered inherently low risk. Therefore, monitoring is conducted using the existing inspection and test programs and system/program health programs.
MP RAI 01 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page 44 Fire Protection Programmatic Elements Monitoring of programmatic elements is required in order to "assess the performance of the fire protection program in meeting the performance criteria". These programs form the bases for many of the analytical assumptions used to evaluate compliance with NFPA 805 requirements. Programmatic aspects include:  Control of Combustible Materials; program compliance and effectiveness, transient exclusion zone effectiveness  Control of Ignition Sources; program compliance and effectiveness  Impairment and Compensatory Measures; program compliance and effectiveness  Industrial Fire Brigade; effectiveness  Monitoring of programmatic elements and program effectiveness is more qualitative in nature since they do not lend themselves to the numerical methods of reliability and availability. Therefore, monitoring is conducted using the existing system and program health programs. Fire protection health reports, self-assessments, regulator and insurance company reports provide inputs to this monitoring program.
Phase 3 - Risk Target Value Determination Phase 3 establishes target values for reliability and availability for the HSS fire protection systems and features and NSCA equipment. HSS SSC's Reliability and availability criteria are established by evaluation based on the HSS fire protection system or features or NSCA equipment's assumed level of reliability/availability in the supporting analyses. Action levels are established for the HSS fire protection system or features or NSCA equipment at the component level, program level, or functionally through the use of the pseudo system or the performance monitoring group' concept. The actual action level is determined based on the number of component, program or functional failures within a sufficiently bounding time period (~2-3 operating cycles). In addition, the EPRI Technical Report (TR) 1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features" may be used as input for establishing reliability targets, action levels, and monitoring frequency. When establishing the action level threshold for reliability and availability, the action level will be no lower than the fire PRA assumptions. Where HSS NSCA equipment is identified using the Maintenance Rule guidelines, the performance criteria may be established based on the Maintenance Rule, provided the criteria are consistent with Fire PRA assumptions. LSS SSC's LSS Fire Protection Systems and Features are included within the existing inspection and test programs and system and program health programs which ensure functionality and no reliability and availability criteria are assigned. LSS NS (Non-Safety) and FPRA equipment is not included in any monitoring process therefore reliability and availability criteria are not required. NPO equipment which is LSS is included within the existing inspection and test programs which ensure functionality and no reliability and availability criteria are assigned. Rad Release SSC's are LSS and are included within the existing inspection and test programs which ensure functionality and no reliability and availability criteria are assigned. Fire protection programmatic MP RAI 01 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page 45 elements are considered LSS and do not lend themselves to the numerical methods of reliability and availability so their effectiveness is based on the objective and anecdotal evidence evaluated by the engineers in charge of the programs.Phase 4 - Monitoring Implementation Phase 4 is the implementation of the monitoring program, once the monitoring scope and criteria are established.
For HSS fire protection systems and features and NSCA equipment that are monitored, theactual levels of availability, reliability, and performance will be reviewed against the established action levels. If an action level is triggered, the Corrective Action Program governed by APA-ZZ-00500, "Corrective Action Program," is used to identify the adverse condition. A corrective action plan will then be developed to ensure performance returns to the established level.When applicable, a sensitivity study can be performed to determine the margin below the action level that still provides acceptable fire PRA results to help prioritize corrective actions if the action level is reached. A periodic assessment of the Monitoring Program will be included within the scope of the Nuclear Oversight Department's routine Fire Protection Program assessment which is described inLARSection 4.7.3. The scope of the Monitoring Program assessment will include the following:- Review systems with performance criteria. Do performance criteria still effectively monitor the functions of the system? Do the criteria still monitor the effectiveness of the fire protection and nuclear safety capability assessment systems? - Have the supporting analyses been revised such that the performance criteria are no longer applicable or new fire protection and nuclear safety capability assessment SSCs, programmatic elements and/or functions need to be in scope? - Based on the assessment period, are there any trends in monitored elements that should be addressed that are not being addressed? - Has external Operating Experience and Internal Operating Experience when applicable been incorporated into the Monitoring program?
MP RAI 01MP RAI 01MP RAI 01MP RAI 01 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page 46 Function currently in Maintenance Rule?Component currently in FPRA?Fire Protection Systems and FeaturesNSEL ComponentsRad Release Engineered Systems and Features NoHigh Safety Significance of feature by compartment?NFPA 805 Specific Monitoring ProcessEstablish targets for reliability/unavailability in Phase 3Use Maintenance Rule for Monitoring YesYesNormal System & Program Health Monitoring  Process or Outage Risk Management for NPOInclude in Maintenance Rule?High Risk Significance?
YesNoYesNoFire Protection Programmatic ElementsYesNoNPO ComponentsFPRA Components NSCANoPhase 1 -ScopingPhase 2 -Screening*Fully describe process used*Figure 4 NFPA 805 Monitoring - Scoping and Screening MP RAI 01 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page 53 QA Program Utilized During Transition During the transition to 10 CFR 50.48(c), Callaway Plant performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805 and the existing FP QA Program described above. This included requirements that each analysis, calculation, or evaluation performed to support compliance with 10 CFR 50.48(c) be independently reviewed.
Post Transition QA ProgramCallaway Plant will utilize the existing Fire Protection Quality Assurance program with the following changes. For the post NFPA 805 Transition, the FP QA Program requirements will be consolidated within FSAR SP Sections 3.2.4 and 9.5.1 and the OQAM. In addition to editorial and administrative changes (i.e. replacing references to previous NRC guidelines with those associated with the NFPA 805 transition and ensuring the features required for a performance based program under NFPA 805 are addressed), the components and systems currently considered within the scope of the Fire Protection QA Program will be expanded to include those components and systems that are in the power block and are required by Chapter 4 of NFPA 805. This means that certain FP systems and features in some buildings not currently considered under the FP QA Program that are required by NFPA 805 Chapter 4 will now fall under the Fire Protection QA program. As such, any future modifications to these systems will be conducted under the design controls required by the FP QA program. The FP QA Program includes a requirement to conduct independent audits of the FP Program by the Nuclear Oversight Department to ensure that the requirements of the fire protection program are being effectively implemented.
OQAM Section 18 will be revised to change the audit frequency from 2 years to 3 years. Additionally, the details of the audit scope contained in OQAM Section 18.8.e will be relocated to FSAR SP Section 9.5.1. Also as noted in section 4.6.2 criteria for assessing the Monitoring Program will be added to the assessment scope (Ref. Implementation Item 11-805-073).Fire PRA Quality Configuration control of the Fire PRA model will be maintained by integrating the Fire PRA model into existing procedure APA-ZZ-00312, "Probabilistic Risk Assessment (PRA)", used to ensure configuration control of the internal events PRA model. This process complies with Section 5 of the ASME Standard for PRA Quality and ensures that Ameren Missouri maintains an as-built, as-operated PRA model of the plant. The process has been peer reviewed. Quality assurance of the Fire PRA is assured via the same processes applied to the internal events
 
model.This process follows the guidance outlined in RG 1.174 which requires the use of qualified individuals, procedures that require calculations be subject to independent review and verification, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when errors are discovered. Although the entire scope of the formal 10 CFR 50 Appendix B program is not applied to the PRA models or processes in general, often parts of the program are applied as a convenient method of complying with the requirements of RG 1.174. For instance, the procedure which addresses software controls for 10 CFR 50 Appendix B is applied to the PRA model software, as well. With respect to Quality Assurance Program requirements for independent reviews of calculations and evaluations, those existing requirements for Fire Protection Program LIC-19 CategoryIDRequired Fire Protection Feature and System DetailsRequired?
SLERDNFPA 805 Regulatory
 
BasisTable 4-3 Summary of NFPA 805 Compliance Basis an d Required Fire Protection Systems and FeaturesTypeFire AreaFire ZoneDescriptionAmeren MissouriCallaway Plant NFPA 805 Transition ReportA-24Containment Mechanical Piping Penetration Room A 4.2.4.21323117Ionization YNNYNPipe Penetration Room AA-24Detection1323NoneN/A-----Pipe Penetration Room AA-24Suppression1323NoneERFBSYNNYNPipe Penetration Room AA-24FeatureA-25Pipe Penetration Room B4.2.3.21322117Ionization NNNNNPipe Penetration Room BA-25Detection1322NoneN/A-----Pipe Penetration Room BA-25Suppression1322NoneN/A-----Pipe Penetration Room BA-25FeatureA-26Ops Storage/I&C Hot Shop4.2.3.21405118Ionization NNNNNOps Storage/I&C Hot ShopA-26Detection1405NoneN/A-----Ops Storage/I&C Hot ShopA-26Suppression1405NoneN/A-----Ops Storage/I&C Hot ShopA-26FeatureA-27Reactor Trip Switchgear Room 4.2.4.21403105Ionization YNNYNLoad Center and MG Sets RoomA-27Detection1403112Ionization YNNYNLoad Center and MG Sets RoomA-27Detection1403SKC03HalonYNNYNLoad Center and MG Sets RoomA-27Suppression1403NoneERFBSYNNYNLoad Center and MG Sets RoomA-27FeatureAugust 2011 Page 77LIC-26 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011  Page 143 4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase The Fire PRA model represents the as-built, as-operated and maintained plant as it will be configured at the completion of the transition to NFPA 805. The Fire PRA model includes credit for the planned implementation of the modifications listed in Attachment S. Following completion of the implementation items listed in Attachment S, such as further development of procedure changes and training, additional refinements may need to be incorporated into the FPRA. During the implementation phase there may also be refinements to the FPRA based on industry-initiatives. As the FPRA refinements are made, some adjustments to the list of Recovery Actions provided in Attachment G may be warranted prior to completion of implementation. Any changes to the list of Recovery Actions will be evaluated using the same process used in Attachments G and W of this submittal. Table S-1 summarizes plant modifications associated with the transition to NFPA 805 that have already been implemented. Table S-2 summarizes plant modifications that are committed for implementation. Table S-3 provides a list of those items (procedure changes, process updates, and training of affected plant personnel) that will be completed prior to the implementation of the new NFPA 805 FP program at Callaway Plant. 4.8.3 Supplemental Information - Other Callaway Plant Specific Issues 4.8.3.1 Request for Approval of Change to Technical Specification Requirement NRC Approval is being requested to eliminate the requirement to initiate a plant shutdown in accordance with Technical Specification LCO 3.0.3 in the event of a loss of the normal fire suppression water system and inability to establish a back-up water system in 24 hours. On February 19, 1987 Union Electric Company submitted a license amendment request via ULNRC-01447 to delete fire protection Technical Specifications and relocate those requirements to the FSAR under licensee control in accordance with Generic Letter 86-10, "Implementation of Fire Protection Requirements." On October 30, 1987 Union Electric Company responded to NRC questions related to this license amendment request via ULNRC-01667. Specifically, the following question and response is documented in ULNRC-01667.
NRC Question The shutdown requirement of Specification 3.7.10.1 ACTION b should be retained in an appropriate commitment document.
Response As part of implementing the proposed revisions to the Technical Specifications, the requirements of Specification 3.7.10.1 ACTION b will be retained and will not be modified without prior approval from the Nuclear Regulatory Commission (NRC). The requirements of Specification 3.7.10.1 ACTION b will be added' to FSAR (USAR for Wolf Creek) Table 9.5.1-2 with a statement that no modifications to these requirements will be made without prior approval of the NRC. On January 13, 1988 the NRC issued Amendment No. 30 to Facility Operating License No. NPF-30. In the accompanying safety evaluation the staff noted the following:  The licensee had originally proposed to delete the shutdown requirement of Specification 3.7.10.1 Action b. The staff's position is that the loss of the normal fire protection water supply and the inability to establish a back-up fire suppression water Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011  Page 144 system within 24 hours warrant plant shutdown. The licensee responded that the requirements of Specification 3.7.10.1 Action b. will be added to the FSAR with commitment that no modifications to these requirements will be made without prior approval from NRC. The staff considers this response to be acceptable. In response to the above, Callaway Plant has maintained the following statement in FSAR Table 9.5.1-2 for the Fire Suppression Water System, requirements a, b and c. With the Fire Suppression Water System in this condition, establish a backup Fire Suppression Water System within 24 hours. If this required action cannot be met, the requirements of Technical Specification 3.0.3 shall be initiated. Modifications to these requirements shall not be made without prior approval of the NRC. As part of the transition to NFPA 805, it is being requested that the NRC Staff review and approve the removal of the existing requirement in FSAR Table 9.5.1-2 to enter Technical Specification LCO 3.0.3 for an inoperable Fire Suppression Water System coupled with the inability to provide a backup fire suppression water system within 24 hours. NFPA 805 Section 3.2.3(2) requires compensatory actions to be implemented when fire protection systems and other systems credited by the fire protection program cannot perform their intended function. NFPA 805 Section 3.2.3(2) also requires that limits be established on the impairment duration. As stated in Attachment A, NEI 04-02 Table B-1, Callaway Plant procedure APA-ZZ-00703, "Fire Protection Operability Criteria and Surveillance Requirements," will be used to establish the required compensatory actions and impairment durations following the transition to NFPA 805. FSAR Table 9.5.1-2 will be eliminated following the transition to NFPA 805. Justification for this request is documented in Attachment X, Approval Request 1. 4.8.3.2 Request for Approval of Specific Current Transformer Configurations NRC Approval is being requested for a deviation from the common enclosure analysis requirements of NFPA 805 Section 2.4.2 for specific current transformer (CT) configurations where a fire induced open circuit failure could result in a secondary fire. A fire in plant fire area C-21, Lower Cable Spreading Room, or in plant fire area C-27, Main Control Room, could result in an open circuit failure for circuits associated with the Main
 
Generator CTs
. Due to the design of these CTs, a secondary fire due to overheating can be postulated to occur in plant fire area TB-1. Section 2.4.2 of NFPA 805 requires consideration of fire-induced open circuit failure modes and specifies that circuits which share a common enclosure with circuits required to achieve the nuclear safety performance criteria, be evaluated to ensure that such electrical faults will not cause the fire to extend beyond the immediate (initial) fire area. As discussed in NFPA 805 B.3.4.2 the evaluation of common enclosure issues should include consideration of CTs that are constructed such that an open secondary circuit could cause ignition of the transformer. As part of the transition to NFPA 805, it is being requested that the NRC Staff review and approve a deviation from the common enclosure requirements of NFPA 805 for the Main Generator CTs. Justification for this request is documented in Attachment X, Approval Request 2.
LIC-26 to ULNRC-06060 
 
ATTACHMENT A: CHANGES TO THE TRANSITION REPORT ATTACHMENT A NFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentAmeren Missouri Callaway Plant N FPA 805 Transition Report Attachment A . NEI 04-02 Table B Transition of Fundamental FP Program and Design Elements (NFPA 805 Chapter 3)Table B NFPA 805 Ch. 3 Transition%&!!'())&/!###"))!
August 2011 Page A-89 LIC?@A to ULNRC-06060 
 
ATTACHMENT B: CHANGES TO THE TRANSITION REPORT ATTACHMENT B Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection A comprehensive list of systems and equipment and their interrelationships to be ana lyzed for a fire event shall be developed.
The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety fu nctions and components whose fire-induced failure could prevent the operation or result in the maloperatio n of those compo nents needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.This section discusses a generic deterministic methodology and criteria that licensees can use to perform a post-fire safe shutdown analysis to address regulatory requirements. The plant-specific ana lysis approved by NRC is reflected in the plants licensing basis. The methodology described in this section is also an acceptable method of performing a post-fire safe shutdown analysis. This methodology is indicated in Figure 3-1. Other methods acceptable to NRC may also be used. Regardless of the method selected by an individual licensee, the criteria and assumptions provided in this guidance document may apply. The methodology described in Section 3 is based on a computer database oriented approach, which is utilized by several licensees to model Appendix R data relationships. This guidance document, however, does not require the use of a computer database oriented approach. The requirements of Appendix R Sections III.G.1, III.G.2 and III.G.3 apply to equipment and ca bles required for achieving and maintaining safe shutdown in any fire area. Although equipment and cables for fire detection and suppression systems, communications systems and 8-hour emergency lighting systems are important features, this guidance document d oes not address them.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.0Deterministic Methodology Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsA deterministic methodology was used to assess conformance with the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805 for the Callaway Plant.The Callaway Plant NFPA 805 Nuclear Safety Capability Assessment (NSCA) deterministic methodology has been reviewed in detail a gainst the guidance, criteria, and assumptions contained within NEI 00-01, Chapter 3, as documented in the subsequent sections of this table (i.e., Table B-2 from NEI 04-02). The results of this review conclude that the Callaway Plant NSCA has been performed consistent with (i.e., aligns with) the det erministic metho dology guidance, criteria, and assumptions from Chapter 3 of NEI 00-01.With the exception of a few specific paragraphs annotated with quotation marks, all other information contained herein from Callaway Plant Calculation KC-26 is paraphrased.
Page B-2August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-3August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection This section discusses the identification of systems available and necessary to perform the required safe shutdown functions. It also provides information on th e process for combining these systems into safe shutdown paths. Appendix R Section III.G.1.a requires that the capability to achieve and maintain hot shutdown be free of fi re damage. It is expected that the term free of fire damage will be further clarified in a forthcoming Regulato ry Issue Summary. Appendix R Section III.G.1.b requires that repairs to systems and equipment necessary to achiev e and maintain cold shutdown be completed within 72 hours. It is the intent of the NRC that requirements related to the use of manual operator actions will be addressed in a forthcoming rulemaking.The goal of post-fire safe shutdown is to assure that one train of shutdown systems, structures, and components remains free of fire damage for a single fi re in any single plant fire area. This goal is accompli shed by determining those functions important to achieve and maintain hot shutdown. Safe shut down systems are selected so that the capability to perform these required functions is a part of each safe shutdown path. The functions important to post-fire safe shutdown generally include, but are not limited to the following:- Reactivity Control- Pressure Control Systems- Inventory Control Systems- Decay Heat Removal Systems- Process Monitoring- Support Systems
* Electrical systems
* Cooling systems These functions are of importance because they have a dire ct bearing on the safe shutdown goal of being able to achieve and maintain hot shutdown which ensures the integrity of the fuel, the reactor pressure vessel, and the primary containment. If these functions are preserved, then the plant will be safe because the fuel, the reactor and the primary containment will not be damaged. By assuring that this equipment is not damaged and remains fu nctional, the protection of the health and safety of the public is assured.In addition to the above listed functions, Generic Letter 81
-12 specifies considera tion of associated circuits with the potential for spurious equipment operation and/or loss of power source, and the common enclosure failures. Spurious operations/actuations can affect the accomplishment of the post-fire safe shutdown functions listed above. Typical examples of the effects of the spurious operations of concern are the following:
- A loss of reactor pressu re vessel/reactor coolant inventory in excess of the safe shutdown makeup capability- A flow loss or blockage in the inventory makeup or decay heat remova l systems being used for the required safe NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1Safe Shutdown Systems and Path Development Page B-4August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection shutdown path.
Spurious operations are of concern because they have the potential to directly affect the ability to achieve and maintain hot shutdown, which could affect the fuel and cause damage to the reactor pressure vessel or the primary containment.
Common power source and common enclosure concerns could also affect these and must be addressed.
Applicable NoneApplicability CommentsAlignment StatementAlignment BasisAligns with IntentCallaway Plant systems / functions / components required to achieve and maintain "safe and stable" plant conditions post-fire per the Nuclear Safety Performance Criteria of NFPA 805 are identified in Callaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Co mponent Selection.The identification and analysis of these systems / functions / co mponents includes addressing associated circuit issues for spurious operations, high/low pressure interfaces, common power supplies, and common enclosures. These associated circuit issues are discussed in Section 8.7
, Associated Circuits -
Purpose and Scope
, of Callaway Plan t Calculation KC-26. The Callaway Plant definition for "safe and stable" plant operation post-fire per the Nuclear Safety Performance Criteria of NFPA 805 is provided in Section 5.6, Definition of "Safe and Stable" Plant Conditions for Callaway Plant, of Callaway Plant Calculation KC-26.A computer database tool, SAFE-PB, is utilized to demonstrate that the Nuclear Safety Performance Cri teria of NFPA 805 are met for each fire area of the plant. The computer database tool is identified as the "NSCA database" in the remainder of this table, and is described in Section 9.0, Description of SAFE-PB, of Callaway Plant Ca lculation KC-26.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-5August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection The following criteria and assump tions may be considered when identifying systems available and necessary to perform the required safe shutdown functions and combining these systems into safe shutdown paths.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.1.1Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-6August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection [BWR] GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths For The BWR" addresses the systems and equipment originally designed into the GE boiling water reactors (BWRs) in th e 1960s and 1970s, that can be used to achieve and maintain safe shutdown per Section III.G.1 of 10CFR 50, Appendix R. Any of the shutdown paths (methods) described in this repor t are considered to be acceptable methods for ach ieving redundant safe shutdown.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.1Criteria/AssumptionsNot Applicable Callaway Plant is PWR; BWR guidance not applicable.
Applicability CommentsAlignment StatementAlignment BasisNot ApplicableNot Applicable Reference DocumentsNot Applicable Page B-7August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection [BWR] GE Report GE-NE-T43-00002-00-03-R01 provides a discussion on the BWR Owners' Group (BWROG) position regarding the use of Safety Relief Valves (SRVs) and low pressure systems (LPCI/CS) for safe shutdown. The BWROG position is that the use of SRVs and low pressure systems is an acceptable methodology for ac hieving redundant safe shutdown in accordance with the requirements of 10CFR50 Appendix R Sections III.G.1 and III.G.2. The NRC has accepted the BWROG position and is sued an SER dated Dec. 12, 2000.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.2Criteria/AssumptionsNot Applicable Callaway Plant is PWR; BWR guidance not applicable.
Applicability CommentsAlignment StatementAlignment BasisNot ApplicableNot Applicable Reference DocumentsNot Applicable Page B-8August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection [PWR] Generic Letter 86-10,  , Section 5.3.5 s pecifies that hot shutdown can be maintained without the use of pressurizer heaters (i.e., pressure control is provided by controlling the makeup/charging pump s). Hot shutdown conditions can be maintained via natural circulation of the RCS through the steam generators. The cooldown rate must be controlled to prevent the formation of a bubble in the reactor head. Therefore, feedwater (either auxiliary or emergency) flow rates as well as steam release must be controlled.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.3Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredReactor Coolant System (RCS) pressure c ontrol capabilities required to achieve and maintain a "safe and stable" plant condition post-fire are identified in Callaway Plant Calcu lation KC-26, Section 7.0, NSCA Model Development and Component Selection.The NSCA model requires that pressurizer heater capability be available for RCS pressure control. The Pressurizer Backup Group heaters are analyzed to remain available from the Main Control Room (or Auxiliary Shutdown Panel) for RCS pressure control. The Pressurizer Control Gro up heaters ar e only analyzed for loss of Main Control Room trip capability.The NSCA model also allows for RCS pressure control to be achieved utilizing the Chemical and Volume Control Sy stem (CVCS) to add RCS inventory (and increase RCS pressure) and the Auxiliary Feedwater System (AFW) (t o remove decay heat, and decrease RCS pressure).RCS inventory is supplied with the CVCS utilizing either of two essential charging pumps, with pump suction taken from the borated Refueling Water Storage Tank (RWST), and pump discharge injected into the RCS through the RCP seals and/or the boron injection header. The non-credited charging pump(s) are analyzed for loss of Main Control Room trip capability.Feedwater for decay heat removal is supplied from either the Turbine Driven Auxiliary Feedwater (AFW) Pump (supplies all four S team Generators [SGs]) or the two Motor Driven Auxiliary Feedwate r Pumps (MDAFW). MDAFW-A supplies SGs B and C, MDAFW-B supplies SGs A and D. Atmospheric ste am dump valves (ASDs) are modeled to be operable as required for the credited SG (1 of 4). The non-credited AFW pump(s) are analyzed for loss of Main Control Room trip capability.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-9August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection The classification of shutdown capability as alternative shutdown is made independent of the selection of systems used for shutdown. Alternative shut down capability is determined based on an inability to assure the availability of a redundant safe shutdown path. Compliance to the separation requirements of Sections III.G.1 and III.G.2 may be supplemented by the use of manual actions to the extent allowed by the re gulations and the licensing basis of the plant, repairs (cold shutdown only), exemptions, deviations, GL 86-10 fire hazards analyses or fire protection design change evaluations, as appropriate. These may also be used in conjunction with alternative shutdown capability.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.4Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsUnlike 10 CFR 50 Appendix R, NFPA 805 makes no distinction for alternative / dedicated shutdown.Auxiliary Shutdown Panel RP118B is the primary control station for implementation of the 10 CFR 50 Appendix R Alternate Shutdown Strategy in the event of a fire that requires the evacuation of the Main Control Room. Based on the definition provided in RG 1
.205, and the additiona l guidance provided in FAQ 07-0030 Revision 5 (ML110070485), Aux iliary Shutdown Panel RP118B is also considered to be the Primary Control Station for NFPA 805, with the associated enabling, control, and indication functions as identified:* Enable RP118B with isolation transfer switches/control switches located at RP118B
* Steam Generator B (2) pr essure indication (ABPIC0002B)
* Steam Generator B (2) wide range level indicat ion (AELI0502A)
* Steam Generator B (2) AFW flow indication (ALFI0003B)
* Open control for steam s upply valve from Steam Generator B (2) to TDAFP (ABHV0005)* Open and close control for Steam Generator B (2) Atmospheric Steam Dump Valve (ABPV0002)* Open and close control for Steam Generator B (2) AFW flow control valve from TDAFP (ALHV0010)
Page B-10 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
* Open and close control for Essential Service Water to s uction of MDAFW Pump B (ALHV0030)* Open and close control for Condensate Storage Tank to suction of MDAFW Pump B (ALHV0034)* MDAFW Pump B suction pressure indication (ALPI0024B)
* Trip and close control for MDAFW Pump B breaker (NB0205)
* Steam Generator D (4) pressure indica tion (ABPIC0004B)* Steam Generator D (4) wide range level indicat ion (AELI0504A)
* Steam Generator D (4) AFW flow indication from MDAFW Pump B (ALFI0001B)* Open and close control for Steam Generator D (4) Atmospheric Steam Dump Valve (ABPV0004)* Open and close control for Steam Generator D (4) AFW flow control valve from MDAFW Pump B (ALHV0005)* Open and close control for Essential Serv ice Water to suction of TDAFP (ALHV0033)* TDAFP suction pressure indication (ALPI0026B)* Open and close control for TDAFP Governor Control valve (FCFV0313)* Open and close control for TDAFP Trip and Throttle valve (FCHV0312)* Pressurizer level ind ication (BBLI0460B)* Reactor Coolant System pressure indication (BBPI0406X)* Reactor Coolant System Loop 2 cold leg temperature indication (
BBTI0423X)* Reactor Coolant System Loop 4 hot leg temperature indication (
BBTI0443A)
* Intermediate and source range neutron monitoring indication (SENI0061X and SENI0061Y)
* Trip and close control for Pressurizer Backup Group B breaker (PG2201)NRC approval for the design of the Auxiliary Shutdown Panel, and for the overall Alternate Shutdown Strategy to meet the requirements of 10 CFR 50 Appendix Page B-11 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection R, Section III.G.3, was provided in NUREG-0830, SER Supplement No.
3, Docket No, STN 50
-483, May 1984, and in NUREG-0830, SER S upplement No. 4, Docket No, STN 50-483, October 1984. Clarification regarding this approval is requested in Attachment T of the Callaway Plant N FPA 805 License Amendment Request, LDCN 11-0012, Transition Report.Enabling of the Auxiliary Shutdown Panel involve s the transfer of control from the Main Control Room to RP118B through an opera tor action to ma nually position three isolation transfer switches and five control switches which are located on RP118B. Following activation of the Auxiliary Shutdown Panel, the plant operator is provided with the capability to control a nd monitor secondary side decay heat removal capability utilizing the Auxiliary Feedwater System, the capability to control Reactor Coolant System (RCS) pressure, and the capability to monitor critical RCS process parameters which are necessary to verify that natural circulation has been established in the RCS and that it is being successfully maintained thereafter.The Auxiliary Shutdown Panel has been transitioned to NFPA 805 as the Primary Control St ation for meeting the NSPC in the event of a fire that requires evacuation of the Main Control Room.Note:  NUREG-0830 Supplement 3 identifies the f ollowing for the Main Cont rol Room evacuation fire event: Some operations require cutting a control power cable at the equipment to ensure that a fault in the control room does not prevent certain equipme nt operation.
These operations have been superseded by NFPA 805 plant modifications which provide for the capability to isolate and transfer control of the fire affected component to the local control station, with redundant fusing. These NFPA 805 modifications are inc luded in Attachment S of the LAR. There are no NFPA 805 Recovery Actions that require cutting of control power cable. The NFPA 805 Recovery Actions associated with the capability to isolate and transfer control of the fire affected component to the local control station, with redundant fusing, are identified and evaluated as VFDRs since they do not occur at the Primary Control St ation, RP118B.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NUREG-0830,  SER Supplement No. 3, Docket No, STN 50-483, dated May 01, 1984NUREG-0830, SER Supplement No. 4, Docket No, STN 50-4 83, dated October 01, 1984 Page B-12 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection At the onset of the postu lated fire, all safe shutdown systems (including applicable redundant trains) are assumed operable and available for post-fire safe shutdown. Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions for Operation, etc.
in progress. The units are ass umed to be operating at full power under normal conditions and normal lineups.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.5Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.1.1.5 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-13 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, earthquake),
single failures or non-fire induced transients need be considered in conjunction with the fire.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.6Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.1.1.6 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-14 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection For the case of redun dant shutdown, offsite power may be credited if demonstrated to be free of fire damage. Offsite power should be assumed to remain available for those cases where its availability may adversely impact safety (i.e., reliance cannot be p laced on fire causing a loss of offsite power if the consequences of offsite power availability are more severe than its presumed loss). No credit should be taken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and alternative shutdown capability is necessary, shutdown must be demonstrated both where offsite power is available a nd where offsite power is not available for 72 hours.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.7Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.1.1.7 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection.
Callaway Plant Calcu lation KC-26, Section 7.7, Electrical Distribution Model Overview, describes how the NSCA models offsite po wer, including th e Alternate Emergency Power System (AEPS), as well as onsite power from the emergency diesel generators. As part of the NSCA model (through component-to-component logic success path in the NSCA database tool), offsite power is only credited in fi re areas where it can be demonstrated to be free of fire damage.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-15 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Post-fire safe shutdown systems and components are not required to be safety-related.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.8Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.1.1.8 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection.The Callaway Plant NSCA model does include non-safety re lated plant systems / functions / components. For example, non-safety related offsite power capability is included in the NSCA model as described in Callaway Plant Calculation KC-26, Section 7.7, Electrical Distribution Model Overview.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-16 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor scram/trip. Fire-induced impacts that provide no adverse consequences to hot shutdown within this 72-hour period need not be included in the post-fire safe shutdown analysis. At least one train can be repaired or made operable within 72 hours using onsite capability to achieve cold shutdown.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.9Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment BasisAligns with Intent Callaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.1.1.9 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection.From Section 3.0 of KC-26:"The NFPA 805 Nuclear Safety Performance Criteria (NSPC) requires the licensee to demonstrate that the plant can achieve and maintain a safe and stable condition, but it does not explicitly require the licensee to demonstrate that cold shutdown c an be achieved within 72 hours and maintained indefinitely thereafter. The Callaway NFPA 805 NSPC analysis has defined the safe and stable condition as being able to achieve and maintain Hot Stand by until such time as the plant can either transition to Cold Shutdown, or can safely return to power operation."
"Safe and stable" for Callawa y Plant is defined in S ection 5.6, Definition of Safe and Stable Plant Conditions for Callaway P lant, of Callaway Plant Calculation KC-26.From Section 5.6 of KC-26:
"The NFPA 805 Nuclear Safety Performance Criteria (NSPC) Analysis for Callaway Plant has been developed to ensure that the plan t can achieve and maintain the reactor fuel in a 'safe and stable' condition assuming that a fire event occurs during Callaway Plant Mode 1 (Power Operati on), Mode 2 (Startup
), Mode 3 (Hot Standby), and Mode 4 (Hot Shutdown), up to the point at which the MCC breakers for the Residual Heat Remova l Loop Suction Isola tion Valves, BBPV8702A, BBPV8702B, EJHV8701A, and EJHV8701B, are unlocked and closed. Refer to the Callaway Plant NFPA 805 L icense Amendm ent Request, L DCN 11-0012, Transition Report Attachment C (Table B-3) for the Systems and Components credited with supporting 'safe and stable' plant conditions by fire area.The NFPA 805 Nuclear Safety Capability Assessment (NSCA) has demonstrated that Callaway Plant can achieve and maintain 'safe and stable' conditions for at least 10 hours with the minimum shift operating staff before having to take action to recharge th e nitrogen accumulators. This initial 10 hours p rovides sufficient Page B-17 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection time for the Emergency Response Organizat ion (ERO) to respond and be available to support 'safe and stable' actions to extend Hot Standby conditions."From Section 7.0 of KC-26:
"The transition from Hot Standby to Cold Shutdown and plant operation in Cold Shutdown is not required to demonstrate that the NSPC safe and stable plant conditions defined for the Callaway Plant have been met. Operator manual ac tions and/or repair activiti es associated with these capabilities are not identified as Variances from the Deterministic Requirements of NFPA 805 (VFDRs), and are not implemented into the plant oper ations fire response procedures."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-18 August 2011 LIC-28 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Manual initiation from the main control room or emergency control stations of systems required to achieve and maintain safe shutdown is acceptable where permitted by current regulations or approved by NRC; automatic initiat ion of systems selected for safe shutdown is not required but may be included as an option.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.10Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.1.1.10 of NEI 00-01 is explicitly stated in the calculation
.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection.From Section 3.0 of KC-26:"The Callaway Plant NSCA credits Main Control Room operator action to align NSCA systems / functions / components. The Callaway Plant NSCA does not credit automatic initiation of NSCA systems / functions / components unless specifically modeled and analyzed. Automatic function of the Condensate Storage Tank (CST) Auxiliary Feedwater Low Suction Pre ssure (LSP) design feature and automatic function of th e Load Shed / Load Sequenc ing Panels are two automatic functions that are explicitly modeled and credited in the NSCA."
"The effects of fires on the Reactor Protection System (RPS), Reactor Trip Breakers
, Reactor Trip Bypass Breakers, and Control Rod Drive Mechanisms are not considered to preclude the init iation of an automatic or manual reactor trip and control rod insert ion due to the RPS fail-safe design. The RP S channels are designed fail in the trip condition on loss of power. The reactor trip breakers and reactor trip b ypass breakers are designed to trip on loss of DC control power.
Similarly, the control rod drive clutch mechanisms are designed to release the control rods on loss of power. This is based on Generic Letter 86-10 Enclosure 2, Section 3.8.4."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-19 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Where a single fire can impact more than one unit of a multi-unit plant, the ability to ac hieve and maintain safe shutdown for each affe cted unit must be demonstrated.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.1.11Criteria/AssumptionsNot Applicable Callaway Plant is a single-unit plant; multi-unit guidan ce not applicable.
Applicability CommentsAlignment StatementAlignment BasisNot ApplicableNot Applicable Reference DocumentsNot Applicable Page B-20 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for safe shutdown. For additional information on BWR system selection, refer to GE Report GENE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths for the BWR."
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.1.2Shutdown Functions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-21 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection [BWR] Control Rod Drive SystemThe safe shutdown performance and design requirements for th e reactivity control function can be met without automatic scram/trip capability. Manual scram/reactor trip is credited. The post-fire safe shutdown analysis must only provide the capability to manually scram/trip the reactor.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.1Reactivity ControlNot Applicable Callaway Plant is PWR; BWR guidance not applicable Applicability CommentsAlignment StatementAlignment BasisNot ApplicableNot Applicable Reference DocumentsNot Applicable Page B-22 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
[PWR] Makeup/ChargingThere must be a method for ensuring that adequate shutdo wn margin is maintained by ens uring borate d water is utilized for RCS makeup/charging.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.1Reactivity Control Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsReactivity control capabilities required to achieve and maintain a "
safe and stable" plant condition post-fire are identified in Callaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection.The Callaway Plant NSCA model requires short term reactivity control to be provided through automatic or manual reactor trip an d the subsequent insertion of the control rods. The effects of fires on the Reactor Protection System (RPS), Reactor Tr ip Breakers, Reactor Trip Bypass Break ers, and Control Rod Drive Mechanisms are not considered to preclude the init iation of an automatic or manual reactor trip and control rod insert ion due to the RPS fail-safe design. The RP S channels are designed fail in the trip condition on loss of power. The reactor trip breakers and reactor trip b ypass breakers are designed to trip on loss of DC control power.
Similarly, the control rod drive clutch mechanisms are designed to release the control rods on loss of power. This is based on Generic Letter 86-10 Enclosure 2, Section 3.8.4.The Callaway Plant NSCA model requires long term reactivity control to be provided utilizing the Chemical and Vo lume Control System (CVCS) to add borated Reactor Coolant System (RCS) inventory.Borated RCS inventory is supplied with the CVCS utilizing either of two essential charg ing pumps, with pump s uction taken from the borated Refueling Water Storage Tank (RWST), and pump discharge injected into the RCS through the RCP seals and/or the boron injection header.The credited source of borated water is the RWST, and the Volume Control Tank (VCT) is isolated to prevent dilution of borated water and inadvertent reduction of the shutdown margin.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-23 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
[BWR] Safety Relief Valves (SRVs)The SRVs are opened to maintain hot shutdown conditions or to depressurize the vessel to allow injection using low pressure systems. These are op erated manually. Automatic initiation of the Automatic Depressurization System is not a required function.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.2Pressure Control SystemsNot Applicable Callaway Plant is PWR; BWR guidance not applicable Applicability CommentsAlignment StatementAlignment BasisNot ApplicableNot Applicable Reference DocumentsNot Applicable Page B-24 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
[PWR] Makeup/Charging RCS pressure is contro lled by controlling the rate of charging/makeup to the RCS. Although utilization of the pressurizer heaters and/or auxiliary spray reduces operator burden, neither component is required to provide adequate pressure control. Pressure reduc tions are made by allowing the RCS to cool/shrink, thus reducing pressurizer level/pressure.
Pressure increases are made by initiating charging/makeup to maintain pressurizer level/
pressure. Manual control of the related pumps is acceptable.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.2Pressure Control Systems Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsReactor Coolant System (RCS) pressure c ontrol capabilities required to achieve and maintain a "safe and stable" plant condition post-fire are identified in Callaway Plant Calcu lation KC-26, Section 7.0, NSCA Model Development and Component Selection.The Callaway Plant NSCA model requires that pressurizer heater capability be utilized for RCS pressure control. The Pressurizer Backup Group heaters are analyzed to remain operable from the Main Control Room (or Auxiliary Shutdown Panel) for RCS pressure control. The Pressurizer Control Group heaters are only analyzed for loss of Main Control Room trip capability.The Callaway Plant NSCA mod el also allows for RCS pressure control to be achieved utilizing the Chemical and Volume Control System (CVCS) to add RCS inventory (and increase RCS pressure) and the Auxiliary Feedwater System (AFW) (to remove decay heat, and dec rease RCS pressure
).Addition of RCS inventory is accomplished with the CVCS utilizing either of two essential charging pumps, with pump suction taken from the borated Refueling Water Storage Tank (RWST), and pump discharge injected into the RCS through the RCP seals and/or the boron injection header. The non-credited charging pump(s) are analyzed for loss of Main Control Room trip capability.Feedwater is supplied from either the Turbine Driven Auxiliary Feedwater (AFW)
Pump (supplies all four Steam Generators [SGs]) or the two Motor Driven Auxiliary Feedwater Pumps (MDAFW). MDAFW-A supplies SGs B and C, MDAFW-B supplies SGs A and D. Atmospheric steam dump valves (ASDs) are modeled to be operable as required for the credited SG (1 of 4). The non-credited AFW pump(s) are analyzed for loss of Main Con trol Room trip capability.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-25 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection [BWR] Systems selected for the invento ry control function should be capable of supplying sufficient reactor coolant to achieve and maintain hot shutdown. Manual initiation of these systems is acceptable. Automatic initiation functions are not required.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.3Inventory ControlNot Applicable Callaway Plant is PWR; BWR guidance not applicable.
Applicability CommentsAlignment StatementAlignment BasisNot ApplicableNot Applicable Reference DocumentsNot Applicable Page B-26 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection [PWR] Systems selected for the invento ry control function s hould be capable of maintain ing level to achieve and maintain hot shutdown. Typically, the same components providing inventory control are capable of providing pressure control. Manual initiation of thes e systems is acceptable.
Automatic initiation functions are not required.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.3Inventory Control Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsReactor Coolant System (RCS) inventory contro l capabilities required to achieve and maintain a "safe and stable" plant condition post-fire are identified in Callaway Plant Calcu lation KC-26, Section 7.0, NSCA Model Development and Component Selection.The Callaway Plant NSCA model requires the Chemical and Volume Control System (CVCS) to be utilized for RCS inventory control. Addition of RCS inventory is accomplished with the CVCS utilizing either of two essential charging pumps, with pump suction taken from the borated Refueling Water Storage Tank (RWST), and pump discharge injected into the RCS through the RCP seals and/or the boron injection header. The non-credited charging pump(s) are analyzed for loss of Main Control Room trip capability.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-27 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection [BWR] Systems selected for the decay heat removal function(s) should be capable of:- Removing sufficient decay heat from primary containment, to prevent conta inment over-pressurization and failure.- Satisfying the net positive suction head requirements of any safe shutdown systems taking suction from the containment (suppression pool).- Removing sufficient decay heat from the reactor to ach ieve cold shutdown.
This does not restrict the use of other systems.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.4Decay Heat RemovalNot Applicable Callaway Plant is PWR; BWR guidance not applicable.
Applicability CommentsAlignment StatementAlignment BasisNot ApplicableNot Applicable Reference DocumentsNot Applicable Page B-28 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection [PWR] Systems selected for the decay heat removal function(s) should be capable of:- Removing sufficient decay heat from the reactor to reach hot shutdown conditions. Typically, this enta ils utilizing natural circulation in lieu of forced circulation via the reactor cool ant pumps and controlling steam release via the Atmospheric Dump valves.- Removing sufficient decay heat from the reactor to reach cold shutdo wn conditions.
This does not restrict the use of other systems.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.4Decay Heat Removal Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsReactor Coolant System (RCS) decay heat remova l capabilities required to achieve and maintain a "safe and st able" plant conditi on post-fire are identified in Callaway Plant Calcu lation KC-26, Section 7.0, NSCA Model Development and Component Selection.The Callaway Plant NSCA model requires the Auxiliary Feedwater System (AFW) to be utilized for RCS decay heat removal.Decay heat removal is achieved by initially tripping the Reactor Coolant Pumps (RCPs) and establishing natural circulation thro ugh at least 1 of the 4 Steam Generators (SGs). The SG code safety valves or the SG atmospheric steam dump valves (ASDs) are credited release steam.
"Safe and stable" for Callawa y Plant is defined in S ection 5.6, Definition of Safe and Stable Plant Conditions for Callaway P lant, of Callaway Plant Calculation KC-26.From Section 5.6 of KC-26:"The NFPA 805 Nuclear Safety Performance Criteria (NSPC) Analysis for Callaway Plant has been developed to ensure that the plan t can achieve and maintain the reactor fuel in a 'safe and stable' condition assuming that a fire event occurs during Callaway Plant Mode 1 (Power Operati on), Mode 2 (Startup
), Mode 3 (Hot Standby), and Mode 4 (Hot Shutdown), up to the point at which the MCC breakers for the Residual Heat Remova l Loop Suction Isola tion Valves, BBPV8702A, BBPV8702B, EJHV8701A, and EJHV8701B, are unlocked and closed. Refer to the Callaway Plant NFPA 805 L icense Amendm ent Request, L DCN 11-0012, Transition Report Attachment C (Table B-3) for the Systems and Components credited with supporting 'safe and stable' plant conditions by fire area.The NFPA 805 Nuclear Safety Capability Assessment (NSCA) has demonstrated that Callaway Plant can achieve and maintain 'safe and stable' conditions for at Page B-29 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection least 10 hours with the minimum shift operating staff before having to take action to recharge th e nitrogen accumulators. This initial 10 hours p rovides sufficient time for the Emergency Response Organizat ion (ERO) to respond and be available to support 'safe and stable' actions to extend Hot Standby conditions."From Section 7.0 of KC-26:
"The transition from Hot Standby to Cold Shutdown and plant operation in Cold Shutdown is not required to demonstrate that the NSPC safe and stable plant conditions defined for the Callaway Plant have been met. Operator manual ac tions and/or repair activiti es associated with these capabilities are not identified as Variances from the Deterministic Requirements of NFPA 805 (VFDRs), and are not implemented into the plant oper ations fire response procedures."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-30 August 2011 LIC-28 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection The process monitoring function is provided for all saf e shutdown paths. IN 84-09, Atta chment 1, Sect ion IX "Lessons Learned from NRC Inspections of F ire Protection Safe Shutdown Systems (10CFR50 Appendix R)"
provides guidance on the instrumentation acceptable to and preferred by the NRC for meeting the process monitoring function. This instrumentation is that which monitors the process variables necessa ry to perform and control the functions specified in Appendix R Section III.L.1. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 list of process monitoring is applied to alternative shutdown (III.G.3). IN 84-09 did not identify specific instruments for process monitoring to be applied to redundant shutdown (III.G.1 and III.G.2). In general, process monitoring instruments similar to those listed below are needed to successfully use existing operat ing procedures (in cluding Abnormal Operating Procedures).BWR:
-Reactor coolant level and pressure
-Suppression pool level and temperature-Emergency or isolation condenser level
-Diagnostic instrumentat ion for safe shutdown system
-Level indication for tanks needed for safe shutdownThe specific instruments r equired may be based on operator preference, safe shutdown procedural guidance strategy (symptomatic vs. presc riptive), and systems and pat hs selected for safe shutdown.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.5Process MonitoringNot Applicable Callaway Plant is PWR; BWR guidance not applicable Applicability CommentsAlignment StatementAlignment BasisNot ApplicableNot Applicable Reference DocumentsNot Applicable Page B-31 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection The process monitoring function is provided for all saf e shutdown paths. IN 84-09, Atta chment 1, Sect ion IX "Lessons Learned from NRC Inspections of F ire Protection Safe Shutdown Systems (10CFR50 Appendix R)"
provides guidance on the instrumentation acceptable to and preferred by the NRC for meeting the process monitoring function. This instrumentation is that which monitors the process variables necessa ry to perform and control the functions specified in Appendix R Section III.L.1. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 list of process monitoring is applied to alternative shutdown (III.G.3). IN 84-09 did not identify specific instruments for process monitoring to be applied to redundant shutdown (III.G.1 and III.G.2). In general, process monitoring instruments similar to those listed below are needed to successfully use existing operat ing procedures (in cluding Abnormal Operating Procedures).PWR:
-Reactor coolant temperature (hot leg / cold leg)-Pressurizer pressure and level-Neutron flux monitoring (source range)
-Level indication for tanks needed for safe shutdown
-Steam generator level and pressure
-Diagnostic instrumentat ion for safe shutdown systemsThe specific instruments r equired may be based on operator preference, safe shutdown procedural guidance strategy (symptomatic vs. presc riptive), and systems and pat hs selected for safe shutdown.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.5Process Monitoring Applicable NoneApplicability CommentsAlignment StatementAlignment BasisNot in Alignment, but Prior NCR ApprovalProcess monitoring instrumentation required to achieve and ma intain a "safe and stable
" plant condition post-fire is identified in Callaway Plan t Calculation KC-26, Section 7.0, NSCA Model Development and Compon ent Selection.The Callaway Plant NSCA model requires the following instrumentation to be utilized for process monitoring. This instrumentation is consistent with minimum process monitoring instrumentation expectations identified in USNRC Information Notice (IN) 84-09, and as previously approved by the USNRC in the 10 CFR 50 Appendix R licensing basis for the Callaway Plant.
* Reactor coolant temperatu re (T-hot / T-cold): The se instruments are modeled in support of the Decay Heat Removal Performance Goal.Page B-32 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
* Pressurizer pressure and level: These instruments are modeled in support of the Inventory and Pressure Contro l Performance Go al.* Neutron flux monitoring (source range)
: These instruments are modeled in support of the Reactivity Control Performance Goal.* Level indication for various tanks: These instr uments are include d in the system logics for wh ich the tank is required.
* Steam Generator (S G) level and pressure: These instru ments are modeled in support of the Decay Heat Re moval Performance Goal.
* Diagnostic instrumentation for safe shutdown systems: Diag nostic instrumentation such as pump suction pressure, flow, and temperature are generally provided by local indicators that require no electrical power. Where beneficial to reduce operator burden, instruments that read out in the Main Control Room have been included in the model and logically associated with the component being monitored. In addition, instruments which provide permissive or controlling signals to safe shutdown components are modeled in direct support of the component as part of the cable selection process.
Notes:* The RCS temperature instruments are mo deled with the steam generator level instruments. The required instrumentation is a sin gle T-hot instr ument and a single T-cold instrument, which can be on differ ent credited loops. This configuration was previously approved by the USNRC in the 10 CFR 50 Appendix R licensing basis for the Callaway Plant.* Reactor Coolant System (RCS) pres sure is assumed to be uniform throughout the RCS, including the pressurizer.* Neutron flux monitors have indication in the Main Control Room and at the Auxiliary Shutdown Panel.* The various tanks required for safe shutdown include the Condensate Storage Tank (CST), Refueling Water St orage Tank (RWST), and Emergency Diesel Generator (EDG) Fuel Oil Tanks.* SG level indication requires wide range level indication. When wide r ange level indication is unavailable, SG level is monito red using narrow range level. AFW flow indication is credited where available, but not required. This configuration was previously approved by the USNRC in the 10 CFR 50 Appendix R licensing basis for the Callaway Plant.* CST, RWST and EDG Fuel Oil Tank level indication is not provided at the Auxiliary Shutdown Panel RP118B. This does not align with Section 3.1.2.5 of NEI 00-01. However, this conf iguration was previously approved by the USNRC in the 10 CFR 50 Appendix R licensing basis for the Callaw ay Plant.NRC approval for the design of the Auxiliary Shutdown Panel, and for the overall Alternate Shutdown Strategy to meet the requirements of 10 CFR 50 Appendix R, Section III.G.3, was provided in NUREG-0830, SER Supplement No.
3, Docket No, STN 50
-483, May 1984, and in NUREG-0830, SER S upplement No. 4, Docket No, STN 50-483, October 1984. Clarification regarding this approval is requested in Attachment T of the Callaway Plant N FPA 805 License Amendment Request, LDCN 11-0012, Transition Report.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NUREG-0830,  SER Supplement No. 3, Docket No, STN 50-483, dated May 01, 1984NUREG-0830, SER Supplement No. 4, Docket No, STN 50-4 83, dated October 01, 1984 Page B-33 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
[Blank Heading - No specific guidance]
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.6Support Systems Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not Required Generic heading. Detailed alignment discussed in subsequ ent reference paragraphs.
Reference DocumentsNot Applicable Page B-34 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection AC Distribution SystemPower for the Appendix R safe shutdown equipment is typically provided by a medium voltage system such as 4.16 KV Class 1E busses either directly from the busses or through step down tr ansformers/ load centers
/ distribution panels for 600, 480 or 120 VAC loads.For redundant safe shutdown performed in accordance with the requirements of Appendix R Section III.G.1 and 2, power may be supplied from either offsite power sourc es or the emergency diesel generator d epending on which has been demonstrated to be free of fire damage. No credit should be taken for a fire causing a loss of offsite power.
Refer to Section 3.1.1.7.
DC Distribution System Typically, the 125VDC distribution system supplies DC control power to various 125VDC co ntrol panels including switchgear breaker controls. The 125VDC distribution panels may also supply power to the 120VAC distribution panels via static inverters. These distribution panels typicall y supply power for instrumentation necessary to complete the process monitoring functions.For fire events that result in an interruption of power to the AC electrical bus, the station batteries are necessary to supply any required control power during the interim time period required for the diesel genera tors to become operational.
Once the diesels are operational, the 125 VDC distribution system can be powered from the diesels through the battery chargers.The DC control centers may also supply power to various s mall horsepower Appendix R safe shutdown system valves and pumps. If the DC sys tem is relied upon to s upport safe shutdown without battery chargers being available, it must be verified that sufficient battery capacity exists to support the necessary loads for sufficient time (either until power is restored, or the loads are no longer required to operate).
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.6.1Electrical Systems Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsPage B-35 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Electrical distribution systems required to achieve and mainta in a "safe and stable" plant condition post-fire are identified in Callaway Plant Calculation KC-26, Section 7.7, Electrical Distribution Model Overview.The Callaway Plant NSCA model includes d iverse Class 1E and non-Class 1E electrical distribution syst em capabilities to provide the vital support function of electrical power for the mechanical systems / functions / components of the NSCA model.AC power may be supplied by offsite power through the switchyard or by offsite power through the Alternate Emergency Power System (AEPS), or AC power may be supplied by onsite power from the emergency diesel generators.The AC power supplies provide power to 13kV and 4kV buses, which in turn supply 480V AC buses. The 480V AC bus es are the normal supply for 120V AC buses.Battery backed 125V DC s upplies power to 120V AC distribu tion panels via inverters and static transfer switches. The static tra nsfer switches are used as a backup to automatically supply 120V AC from a 480V AC source thro ugh a constant voltage transformer.Section 8.7, Associated Circuits - Purpose and Scope, of Callaway Plant Calculation KC-26 identifies the electrical design calc ulations that demonstrate the capacity of the Class 1E and non-Class 1E station batteries as being adequate to supply power for station loads for a minimum of 240 minutes (4 hours) without charging.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-36 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Various cooling water systems may be required to support safe shutdown system operation, based on plant-specific considerations. Typical uses include:- RHR/SDC/DH Heat Exchanger cooling water- Safe shutdown pump co oling (seal coolers, oil coolers)- Diesel generator cooling- HVAC system cooling water.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.6.2Cooling Systems [Main Section]Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCooling water systems required to achieve and maintain a "safe a nd stable" plant condit ion post-fire are identified in Callaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Componen t Selection.The Callaway Plant NSCA model requires the Component Cooling Water System (CCW) and the Essential Service Water System (ESW) to provide the vital support function of cooling water for the other mechanical systems / functions / components of the NSCA model.The Component Cooling Water System (CCW) is cooled by the Essential Service Water System (ESW). There are two CCW trains; each provides cooling water to the following:* RHR Heat Exchangers* RHR Pump Seal Coolers
* CCP Pump Oil Cooler* Fuel Pool Cooling Heat Exchanger* Seal Return Heat Exchanger
* RCP Thermal Barr ier Heat Exchanger* Excess Letdown Heat Exchanger Page B-37 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection In addition to cooling the CCW heat exchangers, the two ESW trains are credited to cool the following:
* Diesel Generator Coolers
* AFW Pump Room Coolers* Control Room A/C Condensers* Class 1E Switchgear A/C Condensers* RHR Pump Room Coolers* CCW Pump Room Coolers
* CCP Room Coolers
* Penetration Room Coolers* Containment Air CoolersThe ESW system is also a backup water source for the AFW pumps and for CCW (manual makeup only).
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-38 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection HVAC Systems may be required to ass ure that safe shutdown equipment remains within its operating temperature range, as specified in manufactur ers literature or demonstrat ed by suitable test methods, and to assure protection for plant operations staff from the effects of fire (smoke, heat, toxic gases, and gaseous fire suppression agents). HVAC systems may be required to support safe shutdown system operation, based on plant-specific configurations. Typical uses include:
- Main control room, cable spreading room, relay room
- ECCS pump compartments- Diesel generator rooms
- Switchgear roomsPlant-specific evaluations are necessary to determine which HVAC systems are essential to safe shutdown equipment operation.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.2.6.2Cooling Systems [HVAC]
Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsHeating ventilation and air conditioning systems (HVAC) required to achieve and maintain a "saf e and stable" p lant condition po st-fire are identified in Callaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection.The NSCA model requires HVAC to provide the vital support function of air cooling for plant equipment operability (of other mechanical systems / functions / components in the NSCA mo del), and to maintain Main Control Room habitability for plant operations personnel.HVAC systems are required for the following plant areas:* Main Control Room* Containment* Class 1E Switchgear Rooms* Electrical Penetration Rooms Page B-39 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
* Diesel Generator Rooms* Class 1E Battery Rooms
* Vital Inverter Rooms
* Vital DC Switchgear Rooms* ESW Pump Rooms
* Ultimate Heat Sink (UHS) Electrical Equipment Rooms Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-40 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Refer to NEI-00-01 Rev 1 Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown systems and developing the shutdown paths.The following methodology may be used to define the safe shutdown systems and paths for an Appendix R analysis:
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.1.3Methodology for Shutdown System Selection Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-41 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Review available documentation to obtain an understanding of the available plant systems and th e functions required to achieve and maintain safe shutdown.Documents such as the following may be reviewed:
- Operating Procedures (Normal, Emergency, Abnormal)- System descriptions- Fire Hazard Analysis- Single-line electrical diagrams- Piping and Instrumenta tion Diagrams (P&IDs)- [BWR] GE Report GE-NE-T43-00002-00-01-R02 entitled "Original Shutdown Paths for the BWR" NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.3.1Identify Safe Shutdown Functions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsSection 4.0, NSCA Design Inputs, of Callaway Plant Calculation KC-26 identifies the following types of documents and databases as sources of design input utilized for the development of the NSCA model:* Callaway Plant FSAR
* Callaway Plant Technical Specifications* Callaway Plant De sign Basis Documents* Callaway Pla nt Design Calculations* Callaway Plant Piping and Instrumentation Diagrams* Callaway Plant Electrical One-line Diagrams (electrical drawings)* Callaway Plant Electrical Three-line Diagrams (electrical drawings)* Callaway Plant Electrical Schematic Diagrams (electrical drawings)
Page B-42 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
* Callaway Plant Instrument Loop Diagrams (electrical drawings)* Callaway Plant Operating Procedures* Callaway Plant Piping and Instrumentation Diagrams* Callaway Plant Electrical One-line Diagrams (electrical drawings)* Callaway Plant Electrical Three-line Diagrams (electrical drawings)* Callaway Plant Electrical Schematic Diagrams (electrical drawings)* Callaway Plant Instrument Loop Diagrams (electrical drawings)* Callaway Plant Raceway Plan Drawings* Callaway Plant Exposed Conduit Drawings* Callaway Plant USNRC Safety Evaluation Report s and Supplements* Plant fire zone boundary partitioning drawings from Callaway R eport R1843-004-001, Callaway Plant NFPA 805 Fire PRA Plant Boundary Definition and Partitioning* The NSCA database tool.*
*The NSCA database tool is a safety-related cable and raceway database that is populated with information from Callaways Cable and Raceway Tracking System (CARTS) and DIRECTOR databases.The NSCA database tool is described in Section 9.0, Description of SAFE-PB, of Callaway Plant Calculation KC-26.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-43 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Given the criteria/assumptions defined in Section 3.1.1, identify the available combin ations of systems capable of achieving the safe shutdown functions of reactivity control, pressure control, inventory control, decay heat removal, process monitoring and support systems such as electrical and cooling systems (refer to Section 3.1.2). This selection process does not restrict the use of other systems. In addition to achieving the required safe shutdown functions, consider spurious operations and power supply issues that could impact the required safe shutdown function.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.3.2Identify Combinations of Systems That Satisfy Each Safe Shutdown Function Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection, identifies the overall process utilized to identify the combinations of plant systems that sat isfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 80 5.From Section 7.0 of KC-26:
"The purpose of NSCA model development and component selection activity is to create an acc urate plant model that represents the Nuclear Safety Performance Criteria (NSPC) requirements from NFPA 805, Section 1.5.1. The NSCA model must identify and include plant systems /
functions / components that are required to actively function in order satisfy the NSPC requirements. The NSCA model also must identify include plant systems / functions / components that are not required to actively function, but whose mal-operation (i.e., spurious operation), alone or in combination, could be adverse to meeting the NSPC requirements. The plant model should, within constraints of complexity and cost, maximize the diversity and number of potential success paths t hat are available to satisfy the NSPC requirements."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-44 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Select combinations of systems with the capability of performing all of the required safe shutdown functions and designate this set of systems as a safe shutdown path. In many cases, paths may be defined on a divisional basis since the availability of electrical power and other support systems must be demonstrated for e ach path. During the equipment selection phase, identify any addit ional support systems and list them for the appropriate path.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.3.3Define Combination of Systems for Each Safe Shutdown Path Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection, identifies the overall process u tilized to determine the combinations of plant systems that sat isfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 80 5.Callaway Plant Calculation KC-26, Section 7.0, also identifies the overa ll process utilized to logically relate individual syst ems in support of each performance goal. Success paths for each performance goal are specified (i.e., performance goal-to-system logic success paths in the NSCA d atabase tool). Each success path represents the minimum system combinations required to achieve a specific performance goal.Certain support systems / functions, such as electrical power and cooling water, are modeled to dire ctly support specific components and systems rather than to directly support performance goals. These relationships are illus trated by component-to-component logic success paths in the Callaway Plant NSCA model.These relationships are maintained electronically in the NSCA database.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-45 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Assign a path designation to each combination of systems. The path will serve to document the combination of systems relied upon for safe shutdown in each fire area. Refer to Attachment 1 to this document for an example of a table illustrating how to document the various combinations of systems for selected shutdown paths.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.1.3.4Assign Shutdown Paths to Each Combination of SystemsApplicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection, identifies the overall process u tilized to determine the combinations of plant systems that sat isfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 80 5.Callaway Plant Calculation KC-26, Section 7.0, identifies that each performance goal may have multiple success paths representi ng a different combination of systems / functions (i.e., performance goal-to-system logic success paths in the NSCA database tool). Each combination of systems that represents a unique success path for a given performance goal is assigned a path number in the NSCA database tool. The path number is automatically assigned by the NSCA database tool.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-46 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection The previous section described the methodology for selecting the systems and paths necessary to achieve and maintain safe shutdown for an exposure fire event (see Section 5.0 DEFINITIONS for "Exposure Fire"). This section describes the criteria/assumptions and selection methodology for identifying the specific safe shutdown e quipment necessary for the systems to perform their Appendix R function.The selected equipment should be related back to the safe shutdown systems that they support and be assigned to the same safe shutdown path as that sy stem. The list of safe shutdown equipment will then form the basis for identifying the cables necessary for the operation or that can cause the maloperation of the safe shutdown systems.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2Safe Shutdo wn Equipment Selection Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-47 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Consider the following criteria and assumptions when identifying equipment necessary to perform the required safe shutdown functions:
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.2.1Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-48 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Safe shutdown equipment can be divided into two categories. Equipment may be categorized as (1) primary components or (2) secondary components. Typically, the following types of equipment are considered to be primary components:- Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc.- All necessary process indicators and recorders (i.e., flow indicator, temperature ind icator, turbine sp eed indicator, pressure indicator, level recorder)
- Power supplies or other electrical components that support oper ation of primary components (i.e., diesel generators, switchgear, motor control centers, load centers, power supplies, distribution panels, etc.).Secondary components are typically items found within the circuitry for a primary component. Th ese provide a supporting role to the overall circuit function. Some secondary components may provide an isolation function or a signal to a primary component via either an interlock or input signal processor. Examples of secondary components include flow switches, pressure switches, te mperature switches, level switches, temperature eleme nts, speed elements, transmitters, converters
, controllers, tr ansducers, signal conditioner s, hand switches, relays, fuses and various instrumentation devices.Determine which equipment should be included on the Safe Shutdown Equipment List (SSEL). As an option, include secondary components with a primary component(s) that would be affected by fire dam age to the secondary component.
By doing this, the SSEL can be kept to a manageable size and the equipment included on the SSEL can be readily related to required post-fire saf e shutdown systems and functions.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.1.1Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.2.1.1 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 8.0, Circuit Identification and Analysis (compo nent selection is also perf ormed during the circuit identification and analysis activity).Page B-49 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection There is no explicit distinction made in the Callaway Plant NSCA betw een primary and secondary equipment; however, a similar ap proach is maint ained through the system-to-component logic success paths and the component-to-component logic success paths in the NSCA database tool.* Mechanical and electrical system components such as pumps, air operated valves, motor operated valves, and solenoid operated valves, fans, heaters, electrically controlled circuit breakers, tr ansformers, switchgear
, motor control cent ers, batteries, battery cha rgers, inverte rs, distribution pan els, automatic transfer switches, diesel generators and engines, strainers, instrumentation, and dampers, etc. which have an active function i n achieving safe shutdown are included in the NSCA.* Mechanical and electrical system passive components such as pumps, air operated valves, motor operated valves, and solenoid o perated valves, fans, heaters, electrically controlled circuit brea kers, instrumentation, and dampers, etc. are included in the NSCA if they maintain a system boundary or if the spurious operation(s) of the passive component(s) has an adverse impact on NSCA capabilities.* Mechanical system passive components such as tanks, vessels, and heat exchangers which have no spurious failure mode are included in the NSCA for completeness.* For air operated valves, the convention used for the Callaway Plant NSCA is for the component-t o-cable logic success path to associate the required cable ID(s) to the pilot solenoid valve ID(s) or valve I/P ID (i.e., 4EMK04CA fails EMHY8843; 6BGI44CB fails BGHY0182), and for the component-to-component logic success path to associate the pilot solenoid valve ID(s) or valve I/P ID to the air operated valve ID (i.e., EMHY8843 fails EMHV8843; BGHY0182 fails BGHV0182).* For process monitoring instrumentation, the convention used for the Callaway Plant NSCA is for the component-to-cable logic success path to associate the required cable ID(s) to the transmitting component ID (i.e., 1BBI16KA fails BBPT0455), and for the component-to-component logic success path to associate the indicator ID to the transmitting component ID (i.e.,
BBPT0455 fails BBPI0455).* Control panels and discrete electrical and instrumentation components such as hand switches, relays, starters, fuses, indicat ing lights, molded c ase and other non-electrically operated circuit breakers, electrical disconnects, pull boxes, junction boxes, terminal boxes, signal converters, amplifiers, bistables, relay cards, instrument power supplies, etc. (excluding the tr ansmitting devices an d indicating devices) are not explicitly identified or included in the NFPA 805 NSPC Equipment List. These secondary components or sub-components are represented in the NSCA by virtue of the circuit conductors and cables that interconnect them to the prima ry component.
* Manual valves that are repositioned for credited NFPA 805 Recovery Actions are included in the NFPA 805 NSPC Equipment List.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-50 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Assume that exposure fire damage to manual valves and piping does not adversely impact their ability to perform their pressure boundary or safe shutdown function (heat sensitive piping materials, including tubing with brazed or soldered joints, are not included in th is assumption). Fire damage should be evaluated with respec t to the ability to manually open or close the valve should this be necessary as a part of the post-fire safe shutdown scenario.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.1.2Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.2.1.2 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 10.0, Deterministic Fire Area Assessment and Results.
Consideration of the fluid boundary isolation provided by a normally closed manual valve is applicable to, and included the dev elopment of th e NSCA model and component selection, and the deterministic fire area assessment.Consideration of the flowpath a lignment provided by a normally open manual valve is applicable to, and included the development of the NSCA model and component selection, and the deterministic fire area assessment.* Manual valves that are repositioned for credited NFPA 805 Recovery Actions are included in the NFPA 805 NSPC Equipment List, and are subject to assessment of feasibility per KC-26, Section 10.0, Deterministic Fire Area Assessment and Results.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-51 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Assume that manual valves are in their normal position as shown on P&IDs or in the plant operating procedures.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.1.3Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.2.1.3 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 10.0, Deterministic Fire Area Assessment and Results.
Consideration of the fluid boundary isolation provided by a normally closed manual valve is applicable to, and included the dev elopment of th e NSCA model and component selection, and the deterministic fire area assessment.Consideration of the flowpath a lignment provided by a normally open manual valve is applicable to, and included the development of the NSCA model and component selection, and the deterministic fire area assessment.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-52 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Assume that a check valve closes in the direction of potential flo w diversion and seats properly with sufficient leak tightness to prevent flow diversion. Therefore, check valves do not adversely affect the flow rate capability of the safe shutdown systems being used for inventory control, decay heat removal, equipment cooling or other related safe shutdown functions.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.1.4Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.2.1.4 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 10.0, Deterministic Fire Area Assessment and Results.
Consideration of the fluid boundary isolation provided by a chec k valve is applicable to, and included the development of the NSCA model and component selection, and the deterministic fire area assessment.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-53 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Instruments (e.g., re sistance temperature detectors, thermo couples, pressure transmitters, and flow transmitters) are assumed to fail upscale, midscale, or downscale as a result of fire damage, whichever is worse. An instrument performing a control function is assumed to provide an undesired s ignal to the control circuit.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.1.5Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not Required Callaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.2.1.5 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, Section 8.0, Circuit Identification and Analysis (component selection is also performed during the circuit identification and analysis activity), and Section 10.0, Deterministic Fire Area Assessment and Results.
Consideration of these instrumentat ion failure modes is applicable to, and included the developmen t of the NSCA model and component selection
, the circuit identification and analysis, and the deterministic fire area assessment.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-54 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Identify equipment that could spuriously operate or mal-ope rate and impact the performance of equipment on a required safe shutdown path during the equipment selection phase. Consider Bin 1 of RIS 2004-03 during the equipment identification process.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.1.6Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment BasisAligns with Intent Callaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. This criteria / assumption listed in Section 3.2.1.6 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 8.0, Circuit Identification and Analysis (compo nent selection is also perf ormed during the circuit identification and analysis activity).Identification of spurious equipment for the Callaway Plant NSCA does not include binning as described in RIS 2004-03.* Mechanical and electrical system passive components such as pumps, air operated valves, motor operated valves, and solenoid o perated valves, fans, heaters, electrically controlled circuit brea kers, instrumentation, and dampers, etc. are included in the NSCA if they maintain a system boundary or if the spurious operation(s) of the passive component(s) has an adverse impact on NSCA capabilities.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-55 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Identify instrument tubing that may c ause subsequent effects on ins trument readings or signals as a result of fire. Determine and consider the fire area location of the inst rument tubing when evaluating the effects of fire damage to circuits and equipment in the fire area.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.1.7Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection and fire area assessment. This criteria / assumption listed in Section 3.2.1.7 of NE I 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 10.0, Deterministic Fire Area Assessment and Results.From Section 7.8 of KC-26:
Consideration of the potential adverse NSCA impact resulting from the heating of instrument tubing sensing lines is applicable to, and included the development of the NSCA model and component selection, a nd the deterministic fire area assessment.
Instrument tubing sensing lines for NSCA instrumentation have been identified, located, and incorporated into the NSCA model as components with a "-SL" suffix. These instrument tubing sensing line components fail their associated transmitting device through the component-to-component logic success path relationship in the NSCA model database. The instrument tubing sensing lines components are evaluated on a fire area basis, as applicable.From Section 10.2 of KC-26:"* When resolving component and cable failures the analyst is required to consider
, and address as nece ssary, the potential impact resulting from the following:
* heating of instrument tubing sensing lines resulting in e rroneous or unreliable signals from NSCA analyz ed instrumentation (r efer to Section 7.
8 of [Calculation KC-26])"Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-56 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Refer to NEI-00-01 Rev 1 Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdown equipment.
Use the following meth odology to select the sa fe shutdown equipment for a post-fire safe shutdown analysis:
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.2.2Methodology for Equipment Selection Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-57 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Mark up and annotate a P&ID to highlight the specific flow paths for each system in support of each shutdown path.
Refer to Attachment 2 for an example of an annotated P&ID illustrating this concept.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.2.1Identify the System Flow Path for Each Shutdown PathApplicable NoneApplicability CommentsAlignment StatementAlignment BasisAligns with Intent Callaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection, identifies the overall process utilized to identify the combinations of plant components for each plant system that is identified as being required to sat isfy each of th e Nuclear Safety Performance Criteria (NSPC) from Section 1.5.
1 of NFPA 805.A review of P&IDs, electrical drawings, instrument loop diagrams, etc. is performed to identify the NSCA systems, and to identify and develop the NSCA system-to-component logic relationships (i.e., Boolean logic / success paths) and the NSCA component-to-component logic success path r elationships (i.e., success paths). The reviewed documentation (i.e., dr awing markups) is not required to be maintained as part of the NSCA record; however, the reviewed documentation (i.e., document numbers and revision levels) is recorded for configuration management.
Reference DocumentsNot Applicable Page B-58 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Review the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop diagrams) to assure that all equipment in each systems flow path has been identified. Assure that any equipment tha t could spuriously operate and adversely affect the desired system function(s) is also identified. If additional systems are identified which are necessary for the operation of the safe shutdown system under review, inclu de these as systems re quired for safe shutdown.
Designate these new systems with the same safe shutdown path as the primary safe shutdown system under review (Refer to Figure 3-1).
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.2.2Identify the Equipment in Each Safe Shutdown System Flow Path Including Equipment That May Spuriously Operate and Affect System Operation Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection, identifies the overall process utilized to identify the combinations of plant components for each plant system that is identified as being required to sat isfy each of th e Nuclear Safety Performance Criteria (NSPC) from Section 1.5.
1 of NFPA 805.A review of P&IDs, electrical drawings, instrument loop diagrams, etc. is performed to identify the NSCA systems, and to identify and develop the NSCA system-to-component logic relationships (i.e., Boolean logic / success paths) and the NSCA component-to-component logic success path r elationships (i.e., success paths). This is an iterative process.* Mechanical and electrical system components such as pumps, air operated valves, motor operated valves, and solenoid operated valves, fans, heaters, electrically controlled circuit breakers, tr ansformers, switchgear
, motor control cent ers, batteries, battery cha rgers, inverte rs, distribution pan els, automatic transfer switches, diesel generators and engines, strainers, instrumentation, and dampers, etc. which have an active function i n achieving safe shutdown are included in the NSCA.* Mechanical and electrical system passive components such as pumps, air operated valves, motor operated valves, and solenoid o perated valves, fans, heaters, electrically controlled circuit brea kers, instrumentation, and dampers, etc. are included in the NSCA if they maintain a system boundary or if the spurious operation(s) of the passive component(s) has an adverse impact on NSCA capabilities.* Mechanical system passive components such as tanks, vessels, and heat exchangers which have no spurious failure mode are included in the NSCA for completeness.
Page B-59 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
* Control panels and discrete electrical and instrumentation components such as hand switches, relays, starters, fuses, indicat ing lights, molded c ase and other non-electrically operated circuit breakers, electrical disconnects, pull boxes, junction boxes, terminal boxes, signal converters, amplifiers, bistables, relay cards, instrument power supplies, etc. (excluding the tr ansmitting devices an d indicating devices) are not explicitly identified or included in the NFPA 805 NSPC Equipment List. These secondary components or sub-components are represented in the NSCA by virtue of the circuit conductors and cables that interconnect them to the prima ry component.
* Manual valves that are repositioned for credited NFPA 805 Recovery Actions are included in the NFPA 805 NSPC Equipment List.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-60 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Prepare a table listing the equipment identified for each system and the shutdown path that it supports. Identify any valves or other equipment that could spu riously operate and impac t the operation of th at safe shutdown system.
Assign the safe shutdown path for the affected syst em to this equipment.
During the cable selection phase, identify additional equipment required to support the safe shutdo wn function of the path (e.g., electrical distribution system equipment). Include this additional equipment in the s afe shutdown equipment list. Atta chment 3 to this document provides an example of a (SSEL). The SSEL identifies the list of equipment within the plant considered for safe shutdown and it documents various equipment-related attributes used in the analysis.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.2.3Develop a List of Safe Shutdown Equipment and Assign the corresponding System and Safe Shutdown Path(s)
Designation to EachApplicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection, identifies the overall process utilized to identify the combinations of plant components for each plant system that is identified as being required to sat isfy each of th e Nuclear Safety Performance Criteria (NSPC) from Section 1.5.
1 of NFPA 805.The NFPA 805 NSPC Equipment List is maintained as Attachment 7-5 to Callaway Plant Calculation KC-26. Other attachments of KC-26 include data such as NSCA component fire zone and fire area location, NSCA component description, NSCA component association to system logic success path (i.e., system-to-component logic), NSCA system association to performance goal logic success path (i.e., performance goal-to-system logic), etc.
This data is also contained within the NSCA database. The NSCA database also contains NSCA component drawing references.Note: The NFPA 805 NSPC Equipment List also includes and specifically identifies Fire PRA and NPO equipment.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-61 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection Collect additional equipment-related information necessary for performing the post-fire safe shutdown analysis for the equipment. In order to facilitate the analysis, tabulate this data fo r each piece of equipment on the SSEL. Refer to Attachment 3 to this document for an example of a SSEL. Examples of related equipment data sh ould include the equipment type, equipment description, safe shutdown system, safe shutdown path, drawing reference, fire area, fire zone, and room location of equipment. Other information such as the following may be useful in performing the safe shutdown analysis: normal position, hot shutdown position, cold shutdown position, failed air position, failed electrical position, high/low pressure interface concern, a nd spurious operation concern.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.2.4Identify Equipment Information Re quired for the Safe Shutdown AnalysisApplicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection, identifies the overall process utilized to identify the combinations of plant components for each plant system that is identified as being required to sat isfy each of th e Nuclear Safety Performance Criteria (NSPC) from Section 1.5.
1 of NFPA 805.The NFPA 805 NSPC Equipment List is maintained as Attachment 7-5 to Callaway Plant Calculation KC-26. Atta chment 7-5 includes t he following data associated with each NSCA Component ID:
* System ID
* System Desig. (by EPM)* Component Type (by EPM) (e.g. MOV, AOV, Pilot Solenoid, etc. / failure position on loss of po wer and / or a ir, as applicable)* Normal Pos. (normal component position with the plant operating at power)* Hot Pos (SSD and/or PRA) (required hot shutdown position)
* Comments #1 (amplifying notes and co mments, including identification of high/low pressure interfaces)* Comments #2 (additional amplif ying notes and comments)* Support Equipment (e.g., associated support components required for the NSCA component to perform its required function - primarily electrical distribution Page B-62 August 2011 LIC-28 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection components; these associated support components are also identified as NSCA components)* Needs Power (e.g., if the NSCA component needs electrical power to perform its required function(s); Y- yes, N - no, or NA -
not applicable) Other attachments of KC-26 include data such as NSCA component fire zone and fire area location, NSCA component description, NS CA component association to system logic success path (i.e., system-to-component logic), NSCA system association to performance goal logic success path (i.e., performance goal-to-system logic), etc. This data is also contained within the NSCA database. The NSCA database also contains NSCA componen t drawing references.Note: The NFPA 805 NSPC Equipment List also includes and specifically identifies Fire PRA and NPO equipment.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-63 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection In the process of defining equipment and cables for safe shutdown, identify additional supporting equipment such as electrical power and interlocked equipment.
As an aid in assessing identified impacts to safe shutdown, consider modeling the dependency between equipment within each safe shutdown path either in a relational database or in the form of a Safe Shutdown Logic Diagram (SSLD). Attachment 4 provides an example of a SSLD that may be developed to document these relationships.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.2.2.5Identify Dependencies Between Equipment, Supporting Equipment, Safe Shutdown Systems and Safe Shutdown Paths Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calculation KC-26, Section 7.0, NSCA Model Development and Component Selection, identifies the overall process utilized to identify the combinations of plant components for each plant system that is identified as being required to sat isfy each of th e Nuclear Safety Performance Criteria (NSPC) from Section 1.5.
1 of NFPA 805.Supporting equipment is tied to supported components through the use of component-to-component logic success path relationships in the NSCA database tool. The NSCA includes these component-to-component logic success path relationships when determining the effect of the a postulated fire event.The NFPA 805 NSPC Equipment List is maintained as Attachment 7-5 to Callaway Plant Calculation KC-26. Atta chment 7-5 includes t he following data associated with each NSCA Component ID:
* System ID
* System Desig. (by EPM)* Component Type (by EPM) (e.g. MOV, AOV, Pilot Solenoid, etc. / failure position on loss of po wer and / or a ir, as applicable)* Normal Pos. (normal component position with the plant operating at power)* Hot Pos (SSD and/or PRA) (required hot shutdown position)
* Comments #1 (amplifying notes and co mments, including identification of high/low pressure interfaces)* Comments #2 (additional amplif ying notes and comments)
Page B-64 August 2011 LIC-28 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability Syste m and Equipment Selection
* Support Equipment (e.g., associated support components required for the NSCA component to perform its required function - primarily electrical distribution components; these associated support components are also identified as NSCA components)* Needs Power (e.g., if the NSCA component needs electrical power to perform its required function(s); Y- yes, N - no, or NA -
not applicable) Other attachments of KC-26 include data such as NSCA component fire zone and fire area location, NSCA component description, NS CA component association to system logic success path (i.e., system-to-component logic), NSCA system association to performance goal logic success path (i.e., performance goal-to-system logic), etc. This data is also contained within the NSCA database. The NSCA database also contains NSCA componen t drawing references.Note: The NFPA 805 NSPC Equipment List also includes and specifically identifies Fire PRA and NPO equipment.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-65 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety functions shall be identified. This includes circuits that are required for operation, that could prevent the oper ation, or that result in the maloperation of th e equipment identified in 2.4
.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to s upport the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Ot her circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Su pply Circuits. Those circuits whose fire-induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situ ation could occur if the upstream protection device (i.e., br eaker or fuse) is not properly co ordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits requ ired to achieve the nuclear safety perfor mance criteria and whose fire-induced failure could cause the loss of the required co mponents shall be identified. The concern is that the eff ects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately p rotected cables or via inadequately sealed fire area boundarie s.This section provides industry guidance on the recommended methodology and criteria for selecting safe shutdown cables and determining their potential impact on equipment required for achieving and maintaining safe shutdown of an operating nuclear power plant for the condition of an exposure fire. The Appendix R safe shutdown cable selection criteria are developed to ensure that all cables that could affect the proper operation or that could cause the maloperation of safe shutdown equipment are identified and that these cables are properly related to the safe shutdown equipment whose functionality th ey could affect. Through this cable-to-equipment relations hip, cables become part of the safe shutdown path assigned to the equipment affected by the cable.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3Safe Shutdown Cable Selection and Location Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-66 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis To identify an impact to safe sh utdown equipment based on cable routing, the equipment must have cables that affect it identified. Carefully consider how cables are related to safe shutdown equipment so that impacts from these cables can be properly assessed in terms of their ultimate impact on safe shutdown system equipment.
Consider the following cr iteria when selecting cables that impact safe shutdown equipment:
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.3.1Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-67 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis The list of cables whose failure could impact the operation of a piece of safe shutdown equ ipment includes more than those cables connected to the equipment. The relationship between cable and affected equipment is based on a review of the electrical or elementary wiring diagrams. To assure that all cables that could affect the operation of the safe shutdown equipment are identified, investigate the power, control, instrumentation, interlock, a nd equipment status indication cables related to the equipment. Consider review ing additional schematic diagrams to identify additional cables for interlocked circuits that also need to be considered for their impact on the ability of the equipment to operate as required in support of post fire safe shutdown. As an option, consider applying the screening criteria from Section 3.5 as a part of this section. For an example of this see Section 3.3.1.4.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.1.1Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. The criteria / assumptions listed in Section 3.3.1.1 of NEI 00-01 are explicitly stated in the calculation
.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 8.0, Circuit Identification and Analysis.
Callaway Plant Calculation KC-26, Section 8.0, identifies the overall process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.The circuit identification and analysis process includes consideration of "on-scheme" and "off-scheme" circuits (i.e., power, control, breaker protection current sensing loops, instru mentation, permissives, interlocks, etc.) with respect to the required function(s) for each NSCA component
.This is an iterative process which may result in the addition of new NSCA components to the NFPA 805 NSPC Equipment List, and the addition of new component-to-component logic success path relationships to the NSCA model. The new NSCA components added the NFPA 805 NSPC Equi pment List are also subject to the circuit identification and analysis process described above.The final set of NSCA cables identified to support the required function of each NSCA component are maintained utilizing a component-to-cable logic success path relationship in the NSCA database. The NSCA component-to-c able logic success path relationship is als o provided in Callaway Plant Calculation KC-26,  -1. Note: The Instrumen t Air System has not been credited or analyzed in the Callaway Plant NSCA and NPO. The initial circuit analysis and cable selection, and the subsequent deterministic fire area assessment for NFPA 805 NSCA and NPO components was performed utilizing the following cr iteria with respect to Page B-68 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis considerations for the availability of instrument air. Instrument air system pressure IS assumed to exist if it can have an ad verse consequence (i.
e., air pressure exists to keep an AOV in the undesired position absent operator action [from Main Control Room or credited Recovery Action] to ensure the pilot SOV is deenergized). Instrumen t air system pressure IS NOT assumed to exist if it can have a beneficial effect (i.e., air pressure ex ists to keep or p lace an AOV in the desired position).
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-69 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis In cases where the failur e (including spurious actuations) of a single cable could impact more than one piece of safe shutdown equipment, include the cable with each piece of safe shutdown equipment.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.1.2Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. The criteria / assumptions listed in Section 3.3.1.2 of NEI 00-01 are explicitly stated in the calculation
.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 8.0, Circuit Identification and Analysis.
Callaway Plant Calculation KC-26, Section 8.0, identifies the overall process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.The circuit identification and analysis process includes consideration of "on-scheme" and "off-scheme" circuits (i.e., power, control, breaker protection current sensing loops, instru mentation, permissives, interlocks, etc.) with respect to the required function(s) for each NSCA component
.This is an iterative process which may result in the addition of new NSCA components to the NFPA 805 NSPC Equipment List, and the addition of new component-to-component logic success path relationships to the NSCA model. The new NSCA components added the NFPA 805 NSPC Equi pment List are also subject to the circuit identification and analysis process described above.A single cable, if determined to be required fo r more than one NSCA c omponent, may be associated to each unique NSCA component ID utilizing the component-to-cable logic success path relationship in the NSCA database. Typically in the case of power cables, one cable may me associat ed to a single NSCA support component ID, and that NSCA support component ID may be associated to one or more other NSCA suppo rted component IDs utilizing the component-to-component logic success path relationship in the NSCA database.The final set of NSCA cables identified to support the required function of each NSCA component are maintained utilizing a component-to-cable logic success path relationship in the NSCA database. The NSCA component-to-c able logic success path relationship is als o provided in Callaway Plant Calculation KC-26,  -1.The final set of NSCA support components identified to support the required function of each NSCA supported co mponent are maint ained utilizing a component-to-component logic success path relationship in the NSCA database. The NSCA component-to-component logic success path relations hip is also provided in Callaway Plant Calcu lation KC-26, Attachment 7-4 and Attachment 7-5.
Page B-70 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-71 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Electrical devices such as relays, switches and signal resistor units are considered to be acceptable isolation devices. In the case of instrument loops, review the isolation capabilities of the devices in the loop to determine that an acceptable isolation device has been installed at each point where the loop must be isolated so that a fault would not impact the performance of the safe shutdown instrument function.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.1.3Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. The criteria / assumptions listed in Section 3.3.1.3 of NEI 00-01 are explicitly stated in the calculation
.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 8.0, Circuit Identification and Analysis.
Callaway Plant Calculation KC-26, Section 8.0, identifies the overall process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.
Section 8.2 of Callaway Plant Calculation KC-26 provides the following considerations with respect to iso lation devices and the ir application:* Electrical devices such as fuses, molded case circuit breakers, relays (coil-to-contact), switches, and signal resistor units are considered to be acceptable isolation devices. In the case of instrument loops, review the isolation capabilities of the devices in the loop to determine that an acceptable isolation device has been installed at each point where the loop must be isolated so that a fault would not impact the performance of the safe shutdown instrument function. Circuit breakers that require DC control power to perform the protective overcurrent trip function are acc eptable isolation devices provided that the DC control power and trip circuitry (inclusive of the current sensing loop) are demonstrated to be free of fire damage.* Circuit identification and analysis must consider the normal position of switch and relay contacts within control circuits, s imilarly the circuit identification and analysis must consider the potential impact from switch a nd relay contacts having been repositioned follow ing initiation of the fire event.* Circuit identification an d analysis for electrical distribution systems must consider the potential adverse impact from spuri ous closure, spurious t rip, failure to trip on demand, failure to trip on overcurrent, etc. for all electrically operated circ uit breakers, including lo ss of DC control po wer for close and trip function and including protective circuits (i.e.,
overcurrent protection, differential overcurrent protection, etc.).Note that selec tive coordination of br eakers / fuses is initially assumed during the circuit identification and analysis for ea ch unique NSCA component. Selective coordination of breake rs / fuses is subsequently confirmed for the electrical distribution systems, electrical distribution components, and electrical distribution Page B-72 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis alignments included in the NSCA, NPO, and Fire PRA as part of the associated circuits assessment provided in Callaway Plant Calculation KC-26, Section 8.7, Associated Circuits -
Purpose and Scope.Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-73 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Screen out cables for circuits that do not impact the safe shutdown function of a component (i.e., annunciator circuits, space heater circuits and computer input circuits) unless some reliance on these circuits is necessary. However, they must be isolated from the components control scheme in such a way that a cable fault would not impact the performance of the circuit.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.1.4Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. The criteria / assumptions listed in Section 3.3.1.4 of NEI 00-01 are explicitly stated in the calculation
.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 8.0, Circuit Identification and Analysis.
Callaway Plant Calculation KC-26, Section 8.0, identifies the overall process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.
Section 8.2 of Callaway Plant Calculation KC-26 provides the considerations with respect to the screening of cables and /or circuits that do not impact the required function for each NS CA component, as applicable.
Not required cables are generally only identified in the NSCA database for primary scheme (i.e., on-schem e) cables that are d etermined not to be required for the NSCA, NPO, or Fire PRA. These cables are identified in the NSCA database at the d iscretion of the preparer and reviewer of the circuit iden tification and analysis for each NSCA component to document that the prima ry scheme cables were indeed included and addressed in the circuit identification and analysis activity. The not required cables are typically assigned one of the following cable functions in the NSCA database:* RC: 1 - isolated annunciation input circuit* RC: 2 - isolated pe rmissive/interlock* RC: 3 - isolated space heater circuit* RC: 4 - power cable to passive component (i.e., non high/low pressure interface MOV)* RC: 5 - isolated co mputer input circuit Page B-74 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis
* RC: 6 - isolated or non-isolated indication/auto initiation circuit that cannot adversely impact required fu nction of analyze d componentThe final set of NSCA cables identified as not being required to support the required function of each NSCA component are maint ained utilizing a component-to-cable logic success path relationship in the NSCA database. The NSCA component-to-cable logic success path relationship is also provided in Callaway Plant Calculation KC-26,  -1.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-75 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis For each circuit requiring pow er to perform its safe shutdown function, identify the cable su pplying power to each safe shutdown and/or required interlock component. Initially, identify only the power cables from the immediate upstream power source for these interlocked circuit s and components (i.e., the closest power supply, load center or motor control center). Review further the electrical distribution system to capture the re maining equipment from the electrical power distribution system necessary to su pport delivery of power from either the offsite power source or the emergency diesel generators (i.e., onsite power sour ce) to the safe shutdown equipment. Add this equipme nt to the safe shutdown equipment list. Evaluate the power cables for this additional equipment for associated circuits concerns.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.1.5Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. The criteria / assumptions listed in Section 3.3.1.5 of NEI 00-01 are explicitly stated in the calculation
.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 8.0, Circuit Identification and Analysis (compo nent selection is also perf ormed during the circuit identification and analysis activity).Callaway Plant Calculation KC-26, Section 7.0, identifies the overall process utilized to develop the NSCA model and select com ponents required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.The NSCA model development and component selection process includes consideration for the following:* With respect to the convention for identifying and associating electrical distribution equipment to NSCA components, the NSCA model development and component selection should utilize a building block approach consistent with NEI 00-01 Criteria / Assumption 3.3.1.5. The bou ndary for NSCA model development and component selection (for each NSCA component) should include only, as applicable, the upstream electrical power source for each NSCA component.
Callaway Plant Calculation KC-26, Section 8.0, identifies the overall process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.The circuit identification and analysis process includes consideration for the following:* With respect to the convention for associating power cables to NSCA components, the NSCA circuit identification and analysis should utilize a building block Page B-76 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis approach consistent with NEI 00-01 Criteria / Assumption 3.3.1.5. The boundary for NSCA circuit identification and analysis (fo r each NSCA component) should include only, as ap plicable, the power cables from the NSCA component to the upstream electrical power source.
Application of NEI 00-01 Criteria / Assumption 3.3.1.5 for the Callaway Plant NSCA is reflected in the f inal set of NSCA support components identified to support the required function of each NSCA supported component. These relationships are maintained util izing a component-to-component l ogic success path relationship in the NSCA database. The NSCA component-to-component logic success path relationship is also provided in Callaway Plant Calculation KC-26,  -4 and Attachment 7-5.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-77 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis The automatic initiation log ics for the credited post-fire safe shutdown systems are not required to support safe shutdown. Each system can be controlled manually by o perator actuation in the main control room or emergency control station. If operator actions outside the MCR are necessary, those actions must conform to the regulatory requirements on manual actions. However, if not protected from the effects of fire, the fire-induced failure of automatic initiation logic circuits must not adversely affect any post-fire safe shutdown system function.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.1.6Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. The criteria / assumptions listed in Section 3.3.1.6 of NEI 00-01 are explicitly stated in the calculation
.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 8.0, Circuit Identification and Analysis (compo nent selection is also perf ormed during the circuit identification and analysis activity).Callaway Plant Calculation KC-26, Section 7.0, identifies the overall process utilized to develop the NSCA model and select com ponents required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.The NSCA model development and component selection process includes consideration for the following:
* Unless specifically included in the NFPA 8 05 NSPC Equipment List and modeled in the NSCA, credit is not taken in the NSCA model development and component selection for the actuation of any automatic safety features to assist in the operation of componen ts to achieve the NFPA 805 NSPC. With respect to automatic signals, the model deve lopment and component selection for the NSCA ensures that components functional requirements a re included in the NFPA 805 NSPC Equipment List and the NSCA database, as applicable, to identify NSCA components where the circuit identification and analysis will need to include the capability for the plant operator to reposition the NSCA component from the Main Control Room in the event that an ESFAS actuation had oc curred, following manual reset of the ESFAS signal.Section 7.10 of Callaway Plant Calculation KC-26 provides a discussion for the treatment of a valid and/or spurious actuation of the Engineered Safety Features Actuation System (ESFAS) in the Callaway Plant NSCA.
Callaway Plant Calculation KC-26, Section 8.0, identifies the overall process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.
Page B-78 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Section 8.2 of Callaway Plant Calculation KC-26 provides the considerations with respect to the treatment of automatic signals in the circuit identification and analysis activity, as applicable.Unless specifically inc luded in the NFPA 805 NSPC Equipment List and modeled in the NSCA, credit is not taken in the circuit identification and analysis for the actuation of any automatic safety features to assist in the operation of components to achieve the NFPA 805 NSPC. With respect to automatic signals, based on component functional requirements identified in the NFPA 805 NSPC Equipment List and the NSCA database, the circuit identificat ion and analysis for NSCA components ensures that circuits are included in the NSCA model, as applicable, that allow the plant operator to reposition the NSCA component from the Main Control Room in the event that an ESFAS actuation had occurred, following manual reset of the ESFAS actuation signal.The NFPA 805 NSPC Equipment List is maintained as Attachment 7-5 to Callaway Plant Calculation KC-26.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-79 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Cabling for the electrical distribution system is a concern for those breakers that feed associated circuits and are not fully coordinated with upstream breakers. With respect to electrical distribution cabling, two types of cable associations exist.
For safe shutdown considerations, the direct power feed to a primary safe shutdown componen t is associated with the primary component. For example, the power feed to a pump is necessary to support the pum
: p. Similarly, the power feed from the load center to an MCC supports the MCC. However, for cases where sufficient branch-circuit coordination is not provided, the same cables discussed above would also s upport the power supply. For example, the power feed to the pump discussed above would support the bus from which it is fed because, for the case of a common power source analysis, the concern is t he loss of the upstream power source and not the con nected load. Similarly, the cable feeding the MCC from the load center would also be necessary to support the load center.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.1.7Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection. The criteria / assumptions listed in Section 3.3.1.7 of NEI 00-01 are explicitly stated in the calculation
.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 8.0, Circuit Identification and Analysis.
Selective coordination of breakers / fuses has been confirmed for the electrical distribution syste ms, electrical distribution components, and electrical distribution alignments included in the NSCA, NPO, and Fire PRA as part of the associated circuits assessment provided in Callaway Plant Calculation KC-26, Section 8.7, Associated Circuits - Purpose and Scope. Selec tive coordination of breakers / fuses has been established through a review of the Callaway Plant Electrical Design Calculations identified in the associated circuits assessment.From Section 8.7 of KC-26:
"The calculations identified in the references section of this assessment (i.e.,
Section 8.7 of Callaway Plant Calculation KC-2
: 6) address breaker / fuse coordination for the overall plant electrical design. These calculations envelop the topic of Associated Circuits by Common Power Supply (i.e., breaker / fuse coordination) with res pect to 10 CFR 50 Appendix R and NFPA 805. The calculations address all of the electrical power supplies and electrical alignments being credited in the NFPA 805 NS PC Analysis and the Fire PRA. The calculations a re performed consistent with accepted industry practices, and demonstrate that selective coordination has been achieved through the proper application and sizing of circuit overcurrent protection devices.
Furthermore, plant circuit breakers that require an external source of control power to perform their protective overcurrent trip function have been identified and analyzed. The analysis was performed to ensure that all load breakers on the credited switchgear will remain functional to isolate potentially fire affected (non-Page B-80 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis credited) loads."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-81 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Associated Circuit Cables Appendix R, Section III.G.2, requires that separation features be provided for equipment and cables, including associated nonsafety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to grou nd, of redundant trains of sy stems necessary to achieve hot shutdown. The three types of associated circuits were identified in Reference 6.1.5 and further clarified in a NRC memorandum dated March 22, 1982 from R. Mattson to D. Eisenhu t, Reference 6.1.6. They are as follows:
- Spurious actuations- Common power source- Common enclosure NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.3.2Associated Circuit Cables Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-82 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Cables Whose Failure May Cause Spurious Actuations Safe shutdown system spurious actuation concerns can result from fire damage to a cable wh ose failure c ould cause the spurious actuation/mal-operation of equipment whose opera tion could affect safe shutdown. These cables are identified in Section 3.3.3 together with the remaining safe shutdown cables required to support control and operation of the equipment.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.3.2 [A]Associated Circuit Cables - Cable Whose Failure May Cause Spurious Actuations Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the over all process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Pe rformance Criteria (NSPC) from Section 1.5.1 of NFPA 805.
Section 8.2 of Callaway Plant Calculation KC-26 describes that the circuit identification and analysis for each NSCA component is defined from the functional requirements identified in the NFPA 805 NSPC Equipment List and the NSCA database. These functional requirements include: 1) no rmal position (at-power) and required position for hot standby.Plant components whose spurious operation alone, or in combina tion with other components, could adversely affect NSCA capabilities are included in the NFPA 805 NSPC Equipment List (as NSCA components).
One or more cable is identified by the circuit ident ification and analysis as being required for an NSCA component if its failu re alone (or their failure in combination) could adversely affect the desired position(s) / function(s) for the NSCA component, as a pplicable, based on consi deration of the ef fects of open circuits, short circ uits, and/or grounds.* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affect ing multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are p ostulated to occur excep t as modified by the f ollowing:
* Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valv es (as defined by the Fire PRA).
* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire PRA).Page B-83 August 2011 LIC-28 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis
* No credit is taken for self-healing of electrical failures.* The circuit identification a nd analysis does not screen out cables on the basis of ja cket material, insulation material, shie lding, and/or the cable being routed in a dedicated conduit.The final set of NSCA cables identified to support the required function of each NSCA component are maintained utilizing a component-to-cable logic success path relationship in the NSCA database. The NSCA component-to-c able logic success path relationship is als o provided in Callaway Plant Calculation KC-26, Attachment 8-1. The NSCA database and Attachment 8-1 identify the spurious cables together with other NSCA cables required to support control and operation of the NSCA components.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-84 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Common Power Source CablesThe concern for the common power source associated circuits is the loss of a safe shutdown power source due to inadequate breaker/fuse coordination. In the case of a fire-induced cable failure on a non-s afe shutdown load circuit supplied from the safe shutdown power source, a lack of coordination between the ups tream supply breaker/fuse feeding the safe shutdown power source and the load breaker/fuse supplying the non-safe shutdown faulted circuit can result in loss of the safe shutdown bus. This would result in the loss of power to the safe shutdown equipment supplied from that power source preventing the safe shutdown equipment from performing its required safe shutdown function. Identify these cables together with the remaining safe shutdown cables required to support control and operation of the equipment. Refer to Section 3.5.2.4 for an acceptable methodology for analy zing the impact of these cables on post-fire safe shutdown.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.3.2 [B]Associated Circuit Cables - Common Power Source CableApplicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsThe electrical power distribution systems at Callaway Plant are described in Chapter 8.0, Electric Power, of the Callaway - SP, Final Safety Analysis Report (FSAR), Revision OL-14, dated December, 2004. Chapter 8 of the FSAR includes discussions of the design bases for the electrical distribution system, and provides reference to the app licable codes and standa rds that were utilized in the design of the syst ems, inclusive of cable de sign and sizing, selection of circuit protection, electrical separation, etc. The FSAR documents that high level design criteria t hat Callaway must continue to meet as changes are made to the facility through design modifications.
Callaway Plant has per formed electrical design calculations which establish protective device se tpoints and coordination.Callaway Plant Calculation KC-26, Section 8.7, Associated Circuits - Purpose and Scope, confirms that selective coordination of breakers / f uses is maintained the electrical distribution systems, electrical distribution components, and electrical distribution alignments included in the NSCA, NPO, and Fire PRA. Selective coordination of breake rs / fuses has been established through review of the Callaway Plant Electrical Design Calcu lations identified in Section 8.7.From Section 8.7 of KC-26:"The calculations identified in the references section of this assessment (i.e.,
Section 8.7 of Callaway Plant Calculation KC-2
: 6) address breaker / fuse coordination for the overall plant electrical design. These calculations envelop the topic of Associated Circuits by Common Power Supply (i.e., breaker / fuse coordination) with res pect to 10 CFR 50 Appendix R and NFPA 805. The calculations address all of the electrical power supplies and electrical alignments being credited in the NFPA 805 NS PC Analysis and the Fire PRA. The calculations a re performed consistent with accepted industry practices, and demonstrate that selective coordination has been achieved through the proper application and sizing of circuit overcurrent protection devices.
Page B-85 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Furthermore, plant circuit breakers that require an external source of control power to perform their protective overcurrent trip function have been identified and analyzed. The analysis was performed to ensure that all load breakers on the credited switchgear will remain functional to isolate potentially fire affected (non-credited) loads."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-86 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Common Enclosure CablesThe concern with common enclosure associated circuits is fire damage to a cable whos e failure could propagate to other safe shutdown cables in the same enclosure either because the circuit is not properly protected by an isolation device (breaker/fuse) such that a fire-induced fault could result in ignition along its length, or by the fire propagating along the cable and into an adjacent fire area. This fire spread to an adjacent fire area could impact safe shutdown equipment in that fire area, thereby resulting in a condition that exceeds the criteria and assumptions of this methodology (i.e., multiple fires). Refer to Section 3.5.
2.5 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.3.2 [C]Associated Circuit Cables -
Common Enclosure Cables Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calculation KC-26, Section 8.7, Associated Circuits - Purpose and Scope, addresses common enclosure concerns, in clusive of an assessment for the open circuiting of current transformer (CT) secondaries.From Section 8.7 of KC-26:
For common enclosure:
"Chapter 8 of the FSAR includes discussions of the design bases fo r the electrical distribution sy stem, and provides reference to the applicable codes and standards that were utiliz ed in the design of the s ystems, inclusive of ca ble design and sizing, selection of circuit protection, electrical separat ion, etc. The FSAR documents the high level design criteria that Callaway must continue to meet as changes are made to the f acility through design modifications.At Callaway, circuits are provided with overcurrent protection devices that will trip prior to damage to the cable in areas away from the fire. The Callaway Plant electrical single line drawings are identified in the references section of this assessment (i.e., KC-26, Section 8.7.1). These drawings were reviewed with respect to the application of overcurrent protection devices at various voltage levels for both Class 1E and Non-Class 1E circuits. Thi s upper tier review in concert with proper cable sizing practices demonstrated by the available calculations discussed in the breaker / fuse coordination of this a ssessment (i.e., KC-26, Section 8.7.6) and by the general notes on the single line drawings which specify minimum cable sizes for standard breaker and fuse sizes establish a confidence level for cable protective device applications throughout the plant. These practices indicate that associated circuits of concern by common enclosure wi ll not impact the ability to achieve a safe shutdown."
For current transformers:
Page B-87 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis "In all cases but one, the Callaway plant electrical design has addressed the possibility for CT secondary ignition failures in fire areas remote from the CT through the application of isolation transducers to isolate CT secondary circuits that leave the enclosure (typically switchgear) where the CT is physically located. Where this is not the case, other design features of the CT (i.e., CT turns ratio or relay accuracy class) will ensure that the CT does not pose a secondary ignition fire hazard. Where this is not the case, a fire hazard assessment has been performed to address the potential impact to deterministic safe shutdown capability resulting from a postulated secondary fire occurring in all of the fire area through which the CT s econdary current loop is routed. These fire hazard assessments have identified no advers e impact to deterministic safe shutdown capability.In one case, the CT could not be screened out by any means. Callaway P lant has requested approval from the NRC for this non-screened CT in the NFPA 805 License Amendment Request, LDCN 11-0012, Transition Report Attachment X."
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-88 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Refer to Figure 3-4 for a flowchart illustrating the various steps involved in selecting th e cables necessary for performing a post-fire safe shutdown analysis. Use the following methodology to define the cables required for safe shutdown including cables that may cause associated circuits concerns for a post-fire safe shutdown analysis:
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.3.3Methodology for Cable Selection and Location Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-89 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical diagrams including the following documentation to identify the circuits (power, control, instrumentation) required for operation or whose failure may impact the operation of each piece of equipment:- Single-line electrical diagrams- Elementary wiring diagrams- Electrical connection diagrams
- Instrument loop diagrams.For electrical power distribution equipment such as power supplies, identify any circuits whose failure may cause a coordination concern for the bus under evaluation.If power is required for the equipment, include the closest upstream power distribution source on the safe shutdown equipment list. Through the iterative process described in Figures 3-2 and 3-3, include the additional upstream power sources up to either the offsite or the emergency power source.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.3.1Identify Circuits Required for the Operation of the Safe Shutdown Equipment Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the over all process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Pe rformance Criteria (NSPC) from Section 1.5.1 of NFPA 805.
Section 8.2 of Callaway Plant Calculation KC-26 describes the circuit identification and analysis for each NSCA component to ensure that the functional requirements identified in the NFPA 805 NSPC Equipment List and the NSCA database for the NSCA component are met.
These functional requirements include: 1) normal position (at-power) and required position for hot standby.
One or more cable is identified by the circuit ident ification and analysis as being required for an NSCA component if its failu re alone (or their failure in combination) could adversely affect the desired position(s) / function(s) for the NSCA component, as a pplicable, based on consi deration of the ef fects of open circuits, short circ uits, and/or grounds.* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affect ing multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are p ostulated to occur excep t as modified by the f ollowing:
Page B-90 August 2011 LIC-28 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis
* Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification and analysis for high/low pressure interface valves.
* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and ana lysis for high/low pressure interface valves.* No credit is taken for self-healing of electrical failures.* The circuit identification a nd analysis does not screen out cables on the basis of ja cket material, insulation material, shie lding, and/or the cable being routed in a dedicated conduit.
Section 8.2 of Calcu lation Plant Calculation KC-26 also identifies the process of iterative review, and the potential for new NSCA components to be identified through the circuit iden tification and analysis.The final set of NSCA cables identified to support the required function of each NSCA component are maintained utilizing a component-to-cable logic success path relationship in the NSCA database. The NSCA component-to-c able logic success path relationship is als o provided in Callaway Plant Calculation KC-26, Attachment 8-1. The NSCA database and Attachment 8-1 identify the NSCA cables required to support control and operation of the NSCA components. The NSCA database also includes drawing references with revision numbers for configuration management.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-91 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis In reviewing each control circuit, investigate interlock s that may lead to additional circuit schemes, cables and equipment. Assign to the equipment any cables for interlocked circuits that can affect the equipment.While investigating the interlocked circuits, additional equipment or power sources may be discovered. Include these interlocked equipment or power sources in the safe shutdown equ ipment list (refer to NEI-00-01 Rev 1 Figure 3-3) if they can impact the operation of the equipment under co nsideration.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.3.2Identify Interlocked Circuits and Cables Whose Spurious Operat ion or Mal-operation Could Affect ShutdownApplicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the over all process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Pe rformance Criteria (NSPC) from Section 1.5.1 of NFPA 805.
Section 8.2 of Callaway Plant Calculation KC-26 describes that the circuit identification and analysis for each NSCA component includes a review for secondary scheme (i.e., "off-scheme") circuits to ensure that the functional requirements identified in the NFPA 805 NSPC Equipment List and the NSCA database for the NSCA component are met. These functional requirements include: 1) normal position (at-power) and required position for hot standby.One or more cable is identified by the circuit ident ification and analysis as being required for an NSCA component if its failu re alone (or their failure in combination) could adversely affect the desired position(s) / function(s) for the NSCA component, as a pplicable, based on consi deration of the ef fects of open circuits, short circ uits, and/or grounds.* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affect ing multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are p ostulated to occur excep t as modified by the f ollowing:
* Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification and analysis for high/low pressure interface valves.
* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and ana lysis for high/low pressure interface valves.* No credit is taken for self-healing of electrical failures.* The circuit identification a nd analysis does not screen out cables on the basis of ja cket material, insulation material, shie lding, and/or the cable being routed in a Page B-92 August 2011 LIC-28 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis dedicated conduit.
Section 8.2 of Calcu lation Plant Calculation KC-26 also identifies the process of iterative review, and the potential for new NSCA components to be identified through the circuit iden tification and analysis.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-93 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Given the criteria/assumptions defined in Section 3.3.1, identify the cables re quired to operate or that may result in maloperation of each piece of safe shutdown equipment.Tabulate the list of cables potentially affecting each piece of equipment in a relational database includ ing the respective drawing numbers, their revision and any interlocks that are investigated to determine their impact on the operation of the equipment. In certain cases, the same cable may support multiple pieces of equipment. Relate the cables to each piece of equipment, but not necessarily to each supporting secondary component.If adequate coordination does not exist for a particular circuit, relate the power ca ble to the power source. This will ensure that the power source is identified as affected equipment in the fire areas where the cable may be damaged.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.3.3Assign Cables to the Safe Shutdown Equipment Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the over all process utilized to identify and analyze circuits for the NSCA components identified as being required to satisfy each of the Nuclear Safety Pe rformance Criteria (NSPC) from Section 1.5.1 of NFPA 805.
Section 8.2 of Callaway Plant Calculation KC-26 describes the circuit identification and analysis for each NSCA component to ensure that the functional requirements identified in the NFPA 805 NSPC Equipment List and the NSCA database for the NSCA component are met.
These functional requirements include: 1) normal position (at-power) and required position for hot standby.
One or more cable is identified by the circuit ident ification and analysis as being required for an NSCA component if its failu re alone (or their failure in combination) could adversely affect the desired position(s) / function(s) for the NSCA component, as a pplicable, based on consi deration of the ef fects of open circuits, short circ uits, and/or grounds.* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affect ing multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are p ostulated to occur excep t as modified by the f ollowing:
* Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification and analysis for high/low pressure interface valves.
* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and ana lysis for high/low pressure interface valves.
Page B-94 August 2011 LIC-28 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis
* No credit is taken for self-healing of electrical failures.* The circuit identification a nd analysis does not screen out cables on the basis of ja cket material, insulation material, shie lding, and/or the cable being routed in a dedicated conduit.The final set of NSCA cables identified to support the required function of each NSCA component are maintained utilizing a component-to-cable logic success path relationship in the NSCA database. The NSCA component-to-c able logic success path relationship is als o provided in Callaway Plant Calculation KC-26, Attachment 8-1. The NSCA database and Attachment 8-1 identify the NSCA cables required to support control and operation of the NSCA components. The NSCA database also includes drawing references with revision numbers for configuration management.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-95 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis This section on circuit analysis provides information on the potential impact of fire on circuits used to monitor, control and power safe shutdown equipment.
Applying the circuit analysis criteria will lead to an understanding of how fire damage to the cables may affect the ability to achieve and maintain post-fire safe shutdown in a particular fire area. This section should be used in conjunction with Section 3.4, to evaluate the potential fire-induced impacts that require mitigation. Appendix R Section III.G.2 identifies the fire-induced circuit failure types that are to be evaluated for impact from exposure fires on safe shutdown equipment. Section III.G.2 of Appendix R requires consi deration of hot shor ts, shorts-to-ground and open circuits.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5Circuit Analysis and Evaluation Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-96 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.5.1Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-97 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Consider the following circuit failure types on each conductor of each unprotected safe shutdown cable to determine the potential impact of a fire on the safe shutdown equipment associated with that conductor.- A hot short may result from a fire-induced insulation breakdown between conductors of the same cable, a different cable or from some other external source resulting in a compatible but undesired impress ed voltage or signal on a specific conductor.
A hot short may cause a spurious operation of safe shutdown equipment.- An open circuit may result fr om a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit may prevent the abili ty to control or power the affected equipment. An open circuit may also result in a change of state for normally energized equipment. (e.g. [for BWRs] loss of power to the Main Steam Isolation Valve (MSIV) solenoid valves due to an open circuit will result in the closur e of the MSIVs). Note that RIS 2004-03 indicates that open circuits, as an initial mode of cable failures, are considered to be of very low likelihood. The risk-informed inspection process will focus on failures with relatively high probabilities.
- A short-to-ground may re sult from a fire-induced breakdown of a cable insulation system, resulting in the potential on the conductor being applied to ground potential. A short-to-ground may have all of the sa me effects as an open circuit and, in addition, a short-to
-ground may also cause an impact to the control circuit or power train of which it is a part.
Consider the three types of ci rcuit failures identified above to occur individually on each conductor of each safe shutdown cable on the required safe shutdown path in the fire area.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.1.1Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. The criteria / assumptions listed in Section 3.5.1.1 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 10.2 of KC-26:* Circuit analysis may be utilized to assess and disposition specific circuit failures modes (as documented in the NSCA equipment resolutions)
. Circuit analysis conforms to the criteria / assumptions identified in Section 3. 0 and Section 8.0, Circuit Identification and Analysis, of Callaway Plant Calculation KC-26. The Page B-98 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis circuit analysis may discount sp urious operation based on a fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-99 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Assume that circuit contacts are positioned (i.e., open or closed) consistent with the normal mode/position of the safe shutdown equipment as shown on the schematic drawings. The analyst must consider the posi tion of the safe shutdown equipment for each specific shutdown scenario when determining the impact that fire damage to a particular circuit may have on the operation of the safe shutdown equipment.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.1.2Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. The criteria / assumptions listed in Section 3.5.1.2 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 10.2 of KC-26:* Circuit analysis may be utilized to assess and disposition specific circuit failures modes (as documented in the NSCA equipment resolutions)
. Circuit analysis conforms to the criteria / assumptions identified in Section 3. 0 and Section 8.0, Circuit Identification and Analysis, of Callaway Plant Calculation KC-26. The circuit analysis may discount sp urious operation based on a fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage.From Section 8.2 of KC-26:Circuit identification and analysis is performed in the NSCA database for each NSCA co mponent as applicable (circuit identifica tion and analysis is not required for mechanical equipment, etc.). The circuit identification and analysis process involves the following steps:a. identify and understand the normal position for the component at-power, or non-power, as applicableb. identify and understand the desired position(s) / function(s) for the component at-power and/or non-po wer, as applicablec. identify and understand the design function and response for the component under accident conditions, as applicableFrom Section 8.2 of KC-26:
Page B-100 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis "* Circuit identification and analysis must consider the normal position of switch and relay contacts within control circuits, similarly the circuit identification and analysis must consider the potential impact from switch a nd relay contacts having been repositioned follow ing initiation of the fire event."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-101 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Assume that circuit failure types resulting in spurious operations exist until action has been taken to isolate the given circuit from the fire area, or other actions have been taken to negate the effects of circuit failure that is causing the spurious actuation. The fire is not assumed to eventually clear the circuit fault. Note that RIS 2004-03 indica tes that fire-induced hot shorts typically self-mitigate after a limited period of time.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.1.3Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. The criteria / assumptions listed in Section 3.5.1.3 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 10.2 of KC-26:* Circuit analysis may be utilized to assess and disposition specific circuit failures modes (as documented in the NSCA equipment resolutions)
. Circuit analysis conforms to the criteria / assumptions identified in Section 3. 0 and Section 8.0, Circuit Identification and Analysis, of Callaway Plant Calculation KC-26. The circuit analysis may discount sp urious operation based on a fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage.From Section 8.2 of KC-26:"* No credit is taken for self-healing of electrical failures."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-102 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis When both trains are in the same fire area outside of primary containment, all cables that do not meet the separation requirements of Section III.G.2 are assumed to fail in their worst case configuration.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.1.4Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. The criteria / assumptions listed in Section 3.5.1.4 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 10.2 of KC-26:
* An initial deterministic analysis is run with the NSCA database tool for each fire area (or analysis area)
. The initial fire area analysis assu mes that all NSCA equipment and NSCA cables physically located in the fire area fail to their worst case condition or state.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-103 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis The following guidance provides the NRC inspection focus from Bin 1 of RIS 2004-03 in order to identify any potential combinations of spurious operations with higher risk significance. Bin 1 failures should also be the focus of the analysis; however, NRC has indicated that other types of failures required by the regulations fo r analysis should not be disregarded even if in Bin 2 or 3. If Bin 1 changes in subsequent revisions of RIS 2004-0 3, the guidelines in the revised RIS should be followed.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.1.5Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-104 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Cable Failure Modes.For multiconductor cables testing has demonstrated that co nductor-to-conductor shorting within the same cable is the most common mode of failure. This is often referred to as "intra-cable shorting." It is reasonable to assume that given damage, more than one conductor-to-conductor short will occur in a given cable. A second primary mode of cable failure is conductor-to-conductor shorting between separate cables, commonly referred to as "inter-cable shorting
." Inter-cable shorting is less likely than intra-cable shorting. Consistent with the current kn owledge of fire-induced cable failures, the following configurations should be considered:A. For any individual multic onductor cable (thermoset or thermoplastic), any and all potential spurious actuations that may result from intra-cable s horting, including any possible combination of conductors within the cable, may be postulated to occur concurrently regardless of number. However, as a practical matter, the number of combinations of potential hot shorts increases rapidly with the number of conductors within a given cable. For example, a multiconductor cable with three conductors (3C) has 3 possible combinations of two (including desired combinations), while a five conductor cable (5C) has 10 possible combinations of two (inc luding desired combinations
), and a seven conductor cable (7C) has 21 possible combinations of two (including desired combinations)
. To facilitate an inspection that considers most of the risk pr esented by postulated hot shorts within a multiconductor cable, inspectors should consider only a few (three or fo ur) of the most critical postulated combinations. B. For any thermoplastic cable, any and all potential spurious actuations that may result from intra-cable and inter-cable shorting with other thermoplastic cables, including any possible combination of conductors within or between the cables, may be postulated to occur concurrently regardless of number. (The consideration of thermoset cable inter-c able shorts is deferred pending additional research.)
C. For cases involving the potential damage of more than one multiconductor cable, a maximum of two cables should be assumed to be damaged concurrently. The spurious actuations should be evaluated as previously described. The consideration of more than two cables being damaged (and subsequent spurious actuations) is deferred pending additional research.D. For cases involving direct current (DC) circuits, the potential spurious operation due to failures of the associated control cables (even if the spurious operation requires two concurrent hot shorts of the proper polarity, e.g., plus-to-plus and minus-to-minus) should be considered when the required source and ta rget conductors are each located within the same multiconductor cable.E. Instrumentation Circuits. Required instrumentation circuits are beyond the scope of this associated circuit approach and must meet the same requirements as req uired power and control circuits. There is one cas e where an instrument circuit could potentially be con sidered an associated circu it. If fire-induced damage of an instrument circuit could prevent operation (e.g., lockout permissiv e signal) or cause ma loperation (e.g., unwanted start/stop/reposition signal) of systems NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.1.5Criteria/Assumptions Page B-105 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an associated circuit and handled accordingly.
Likelihood of Undes ired ConsequencesDetermination of the potential consequence of the damaged associated circuits is based on the examination of specific NPP piping and instrumentation diagrams (P&IDs) and review of components that could prevent operation or cause maloperation such as flow diversions, loss of coolant, or other scenarios that could significantly impair the NPPs ability to achieve and maintain hot shutdown. When considering the potential consequence of such failures, the [analyst] should also consider the time at which the prevented operation or maloperation occurs.
Failures that impede hot shutdown within the first hour of the fire tend to be most risk significant in a first-order evaluation. Consideration of cold-shutdown circuits is deferred pending additional research.
Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA circuit identification and analysis. The criteria / assumptions listed in Section 3.5.1.5 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 8.0, Circuit Identification and Analysis.From Section 8.2 of KC-26:
"e. Postulate the effects of open circuits, short circuits, and/
or grounds upon the desired position(s) / function(s) for the c omponent at-power and/or non-power, as applicable""* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affecting multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are p ostulated to occur excep t as modified by the f ollowing:
* Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valv es (as defined by the Fire PRA).
* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire Page B-106 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis PRA).* No credit is taken for self-healing of electrical failures.
* Multiple AC and DC grounds ar e postulated in the circuit ident ification and analysis. Multiple grounds in ungrounded AC or DC systems can result in clearing of fuses, or tripping of breakers.""* The circuit identification and analysis does not screen out cables on the basis of jacket material, insulation material, shi elding, and/or the cable being routed in a dedicated conduit. However, the deterministic NSCA area-by-area analyses may discount spurious operation based on the fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage."Circuit identification and analysis for the Callaway Plant NSCA does not include limiting assumptions as described in RIS 2004-03.The circuit analysis and cable selection performed for Callaway is consistent with the guidelines, criteria, and assumptions of NEI 00-01 Revision 2. However, it has become apparent that NEI 00-01 may be unclear to some individuals with respect to the guidance, criteria, and assumptions as pertaining to inter-cable hot shorts (i.e., direct inter-cable hot shorts - source cable to target cable, and indire ct inter-cable hot sh orts - source cable to target cable through a ground plane). As a consequence, Callaway is providing the following clarification in th e NFPA 805 LAR to describe the Callaway circuit analysis and cable selection treatment for inter-cable hot shorts, inclusive of direct and indirect inter-cable hot shorts:The Callaway circuit analysis and cable selection process includes that a positive DC or a negative DC inter-cable hot short can occur on the same target cable so as to result in the s purious operation of a non-high/low pressure interface component.The Callaway circuit analysis and cable selection process excludes that a positive DC and a negative DC inter-cable hot short can occur on t he same target cable so as to result in the sp urious operation of a non-high/
low pressure interface component.
Inter-cable hot shorts are c onsidered by Callaway to occur from direct source cable (s) to target cable int eractions or from indirect source cable(s) to target cable interactions through a gro und plane (i.e., the ground plane could be established through any fi re affected pla nt equipment, conduits, and/or raceways). No distinction is made by Callaway b etween direct and indirect inter-cable hot shorts. The mechanism for the externally applied voltage source (i.e., hot short) to contact the target cable is treated as a black box.Based on this treatment, a non-high/low pressure interface component cannot spuriously operate due to a single inter-cable hot short (positive DC or negative DC) so long as there are also no adequate sources of DC voltage originating within the target cable that could result in spurious operation of the non-high/low pressure interface component due to a combination of intra-cable short circuits and a single inter-cable hot short.The Callaway treatment is consistent with NRC Generic Letter 86-10, Question and Answer 5.3.1, and NEI 00-01 R evision 2, Figure 3.5.2-5. The Callaway treatment is also consistent with the test results from NUREG/CR-7100, SAND2012-0323P, Direct Current Electrical Shorting in R esponse to Exposure Fire (DESIREE-Fire): Test Results, specific to inter-cable hot shorts (Section 6.5.3). Note t he test configuration for the inter-cable hot shorts from NUREG/CR-7100, Comments:
Page B-107 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Reference DocumentsSAND2012-0323P, as depicte d in Figure A-54, was set up intentionally to obtain inter-cable hot shorts for the study, and is not representative of field typical installations which may fu rther reduce the likelihood of inter-cable hot shorts.Multiple grounds (in ungrounded circuits) are considered in the Callaway circuit analys is and cable selection process with respect to the potential for loss of required power for ungrounded circ uits. This approach is consistent with NEI 00-01 Revision 2, Figure 3.5.2-3.
Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-108 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Appendix R requires that nuclear power plants must be designed to prevent exposure fires from defeating the ability to achieve and maintain post-fire safe shutdown. Fire damage to circuits that provide control and power to equipment on the required safe shutdown path and any other equipment whose spurious operation/mal-operation could affect shutdown in each fire area must be evaluated for the effects of a fire in that fire area. Only one fire at a time is assumed to occur. The extent of fire damage is assumed to be limited by the b oundaries of the fire area. Given this set of conditions, it must be assured that one redundant train of equipmen t capable of achieving hot shutdown is free of fire damage for fires in every plan t location. To provide this assurance, Appendix R requires that equipment and circuits required for safe shutdown be free of fire damage and that these circuits be designed for the fire-induced effects of a hot short, short-to-ground, and open circuit. With respect to the electrical distribution sy stem, the issue of breaker coordination must a lso be addressed.This section will discuss specific examples of each of the following types of circuit failures:- Open circuit
- Short-to-ground- Hot short.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.5.2Types of Circuit Failures Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-109 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis This section provides guidance for addressing the effects of an open circuit for safe shutdown equipment. An open circuit is a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit will typically prevent the ability to control or power the affected equipment. An open circuit can a lso result in a change of state for normally energized equipment. For example, a loss of power to the main steam isolation valve (MSIV) solenoid valves [for BWRs] due to an open circuit will result in the closure of the MSIV.NOTE: The EPRI circuit failu re testing indicated that open circuits are not likely to be the initial fire-induced circuit failu re mode. Consideration of this may be helpful within the safe shutdown analysis. Consider the following consequences in the safe shutdown circuit analysis when determining the effects of open circuits:- Loss of electrical continuity may occur within a conductor resulting in deenergizing the circuit and causing a loss of power to, or control of, the required safe shutdown equipment.
- In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or other device. This loss of power may change the state of the equipment. Evalua te this to determine if equipment fails safe.- Open circuit on a high voltage (e.g., 4.16 kV) ammeter current transformer (CT) circuit may result in secondary damage.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.2.1Circuit Failures Due to an Open Circuit Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the over all process utilized to perform circuit identification and analysis for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.From Section 8.2 of KC-26:
"e. Postulate the effects of open circuits, short circuits, and/
or grounds upon the desired position(s) / function(s) for the c omponent at-power and/or non-power, as applicable""* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affecting multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are p ostulated to occur excep t as modified by the f ollowing:
Page B-110 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis
* Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valv es (as defined by the Fire PRA).
* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire PRA).""* Multiple AC and DC grounds are postulated in the circuit identification and analysis. Multiple grounds in ungrounded AC or D C systems can result in clearing of fuses, or tripping of breakers.""* The circuit identification and analysis does not screen out cables on the basis of jacket material, insulation material, shi elding, and/or the cable being routed in a dedicated conduit. However, the deterministic NSCA area-by-area analyses may discount spurious operation based on the fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage."Callaway Plant Calculation KC-26, Section 8.7, Associated Circuits - Purpose and Scope, addresses common enclosure concerns, in clusive of an assessment for the open circuiting of current transformer (CT) secondaries.From Section 8.7 of KC-26:
For current transformers:"In all cases but one, the Callaway plant electrical design has addressed the possibility for CT secondary ignition failures in fire areas remote from the CT through the application of isolation transducers to isolate CT secondary circuits that leave the enclosure (typically switchgear) where the CT is physically located. Where this is not the case, other design features of the CT (i.e., CT turns ratio or relay accuracy class) will ensure that the CT does not pose a secondary ignition fire hazard. Where this is not the case, a fire hazard assessment has been performed to address the potential impact to deterministic safe shutdown capability resulting from a postulated secondary fire occurring in all of the fire area through which the CT s econdary current loop is routed. These fire hazard assessments have identified no advers e impact to deterministic safe shutdown capability.In one case, the CT could not be screened out by any means. Callaway P lant has requested approval from the NRC for this non-screened CT in the NFPA 805 License Amendment Request, LDCN 11-0012, Transition Report Attachment X."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-111 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis This section provides guidance for addressing the effects of a short-to-ground on circuits for safe shutdown equipment. A short-to-ground is a f ire-induced breakdown of a cable insulation system resulting in the potential on the conductor being applied to ground potential. A short-to-ground can cause a loss of power to or control of required safe shutdown equipment. In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases where proper coordination does not exist.
Consider the following consequences in the post-fire safe shutdown analysis when determining the effects of circuit failures related to shorts-to-ground:
- A short to ground in a power or a control circuit may result in tripping one or more isolati on devices (i.e.
breaker/fuse) and causing a loss of power to or cont rol of required saf e shutdown equipment.- In the case of certain energized equipment such as HVAC dampers, a loss of control power may result in loss of power to an interlocked relay or other device that may cause one or more spurious operations.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.2.2Circuit Failures Due to a Short-to-Ground Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the over all process utilized to perform circuit identification and analysis for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.From Section 8.2 of KC-26:
"e. Postulate the effects of open circuits, short circuits, and/
or grounds upon the desired position(s) / function(s) for the c omponent at-power and/or non-power, as applicable""* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affecting multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are p ostulated to occur excep t as modified by the f ollowing:
* Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valv es (as defined by the Fire PRA).
* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire Page B-112 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis PRA).""* Multiple AC and DC grounds are postulated in the circuit identification and analysis. Multiple grounds in ungrounded AC or D C systems can result in clearing of fuses, or tripping of breakers.""* The circuit identification and analysis does not screen out cables on the basis of jacket material, insulation material, shi elding, and/or the cable being routed in a dedicated conduit. However, the deterministic NSCA area-by-area analyses may discount spurious operation based on the fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-113 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis This section provides guidance for analyzing the effects of a hot short on circuits for required safe shutdown equipment. A hot short is defined as a fi re induced insulation breakdown betw een conductors of the same cable, a different cable or some other external source resulting in an undesired impressed voltage on a specific conductor.
The potential effect of the undesired impressed volta ge would be to cause equipment to operat e or fail to operate in an undesired manner.Consider the following specific circuit failures related to hot shorts as part of the post-fire safe shutdown analysis:- A hot short between an energized conductor and a de-energized conductor within the same cable may cause a spurious actuation of equipment. The spuriously actuated device (e.g., relay) may be interlocked with another circuit that causes the spurious actuation of other equipment. This type of hot short is called a cond uctor-to-conductor hot short or an internal hot short.- A hot short between any external energiz ed source such as an energized conduct or from another cable (thermoplastic cables only) and a de-energized conductor may also cause a spu rious actuation of equipment. T his is called a cable-to-cable hot short or an external hot short. Cable-to-cable hot shorts between thermoset cables are not postulated to occur pending additional research.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.2.3Circuit Failures Due to a Hot Short Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the over all process utilized to perform circuit identification and analysis for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.From Section 8.2 of KC-26:
"e. Postulate the effects of open circuits, short circuits, and/
or grounds upon the desired position(s) / function(s) for the c omponent at-power and/or non-power, as applicable""* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affecting multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are p ostulated to occur excep t as modified by the f ollowing:
* Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification Page B-114 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis and analysis for high/low pressure interface valves and high consequence Fire PRA valv es (as defined by the Fire PRA).
* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire PRA).""* The circuit identification and analysis does not screen out cables on the basis of jacket material, insulation material, shi elding, and/or the cable being routed in a dedicated conduit. However, the deterministic NSCA area-by-area analyses may discount spurious operation based on the fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage."The circuit analysis and cable selection performed for Callaway is consistent with the guidelines, criteria, and assumptions of NEI 00-01 Revision 2. However, it has become apparent that NEI 00-01 may be unclear to some individuals with respect to the guidance, criteria, and assumptions as pertaining to inter-cable hot shorts (i.e., direct inter-cable hot shorts - source cable to target cable, and indire ct inter-cable hot sh orts - source cable to target cable through a ground plane). As a consequence, Callaway is providing the following clarification in th e NFPA 805 LAR to describe the Callaway circuit analysis and cable selection treatment for inter-cable hot shorts, inclusive of direct and indirect inter-cable hot shorts:The Callaway circuit analysis and cable selection process includes that a positive DC or a negative DC inter-cable hot short can occur on the same target cable so as to result in the s purious operation of a non-high/low pressure interface component.The Callaway circuit analysis and cable selection process excludes that a positive DC and a negative DC inter-cable hot short can occur on t he same target cable so as to result in the sp urious operation of a non-high/
low pressure interface component.
Inter-cable hot shorts are c onsidered by Callaway to occur from direct source cable (s) to target cable int eractions or from indirect source cable(s) to target cable interactions through a gro und plane (i.e., the ground plane could be established through any fi re affected pla nt equipment, conduits, and/or raceways). No distinction is made by Callaway b etween direct and indirect inter-cable hot shorts. The mechanism for the externally applied voltage source (i.e., hot short) to contact the target cable is treated as a black box.Based on this treatment, a non-high/low pressure interface component cannot spuriously operate due to a single inter-cable hot short (positive DC or negative DC) so long as there are also no adequate sources of DC voltage originating within the target cable that could result in spurious operation of the non-high/low pressure interface component due to a combination of intra-cable short circuits and a single inter-cable hot short.The Callaway treatment is consistent with NRC Generic Letter 86-10, Question and Answer 5.3.1, and NEI 00-01 R evision 2, Figure 3.5.2-5. The Callaway treatment is also consistent with the test results from NUREG/CR-7100, SAND2012-0323P, Direct Current Electrical Shorting in R esponse to Exposure Fire (DESIREE-Fire): Test Results, specific to inter-cable hot shorts (Section 6.5.3). Note t he test configuration for the inter-cable hot shorts from NUREG/CR-7100, SAND2012-0323P, as depicte d in Figure A-54, was set up intentionally to obtain inter-cable hot shorts for the study, and is not representative of field typical installations which may fu rther reduce the likelihood of inter-cable hot shorts.
Comments:
Page B-115 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Reference DocumentsMultiple grounds (in ungrounded circuits) are considered in the Callaway circuit analys is and cable selection process with respect to the potential for loss of required power for ungrounded circ uits. This approach is consistent with NEI 00-01 Revision 2, Figure 3.5.2-3.
Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-116 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location Physical location of equipment and cables shall be identified.Identify the routing for each cable including all raceway and cable endpoints. Typically, th is information is obtained from joining the list of safe shutdown cables with an existing ca ble and raceway database.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.3.4Identify Routing of Cables Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsSection 8.6 of Callaway Plant Calculation KC-26, describes the overall process for assigning fire zone locations to NSCA compon ents and NSCA cable via raceways. This includes assigning fire zone locations to NSCA cable to/from equipment.
Fire zone location is pe rformed through review of plant equipment and raceway layout drawings, and plant fire zone boundary partitioning drawings from Callaway Report R1843-004-001, Callaway Plant NFPA 805 Fire PRA Plant Boundar y Definition a nd Partitioning.The fire zone location for NSCA components, NSCA cable via r aceways, and NSCA cable to/from equipment, is identif ied in Callaway Plant Calculation KC-26,  -2 (NSCA components), Attachment 8-3 (NSCA ca ble to/from equipment), and Attachment 8-4 (NSCA cable raceways), and in the NSCA database.Fire areas are assigned to NSCA components, NSCA cable to/from equipment, and NSCA cable via raceways based on the fire zone-to
-fire area relationships maintained in the NSCA database. Callaway Plant Calculation KC-26, Section 6.0, Identification of Callaway Plant Fire Areas / Compartments, includes Attachment 6-1 which provides a table of the fire zone-to-fire area relationships. Cable fire areas are automatically assigned by the NSCA databa se through the relationship of a cable to its respective via raceways and their fire areas, and to its respective to/from equipment and their fire areas.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-117 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location Identify the fire area location of each raceway and cable endpoint identified in the previous step and join this information with the cable routing data. In addition, identify the location of field-routed cable by fire area. This produces a database containing all of the cables requiring fi re area analysis, their locations by fire area, and their raceway.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.3.3.5Identify Location of Raceway and Cables by Fire Area Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsSee Section 3.3.3.4 of this table.The existing methodology aligns with NEI 00-01 guidance.
Reference DocumentsNot Applicable Page B-118 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location The evaluation of associated circuits of a common power source consists of verifying proper coord ination between the supply breaker/fuse and the load breakers/fuses for power sources that are required for safe shutdown. The concern is that, for fire damage to a single power cab le, lack of coordination between the supply breaker/fuse and the load breakers/fuses can result in the loss of power to a safe shutdown power source that is required to provide power to safe shutdown equipment.
A coordination study should demonstrate the coordination status for each required common power source. For coordination to exist, the time-current curves for the breakers, fuses and/or protective relaying must demonstrate that a fault on the load circuits is isolated before tripping the upstream breaker that supplies the b us. Furthermore, the available short circuit current on the load circuit must be c onsidered to ensure that coordination is demonstrated at the maximum fault level.The methodology for identifying potential associated circuits of a common power source and evaluating circuit coordination cases of associated circuits on a single circuit fault basis is as follows:- Identify the power sources required to supply power to safe shutdown equipment.- For each power source, identify the breaker/fuse ratings, types, trip settings an d coordination char acteristics for the incoming source breaker supplying the bus and the breakers/fuses feeding the loads supplied by the bus.
- For each power source, demonstrate proper circuit coordination using acceptable industry methods.
- For power sources not properly coordinated, tabulate by fire area the routi ng of cables whose breaker/fuse is not properly coordinated with the supply breaker/fuse. Evaluate the potential for disabling power to the bus in each of the fire areas in which the associated circuit cables of concern are routed and the power source is required for safe shutdown.
Prepare a list of the following information for each fire area:- Cables of concern.
- Affected common power source and its path.- Raceway in which the cable is enclosed.
- Sequence of the raceway in the cable route.- Fire zone/area in which the raceway is located.
For fire zones/areas in which the power source is disabled, the effects are mit igated by appropriate methods. - Develop analyzed safe shutdown circuit dispositions for the associated circuit of concern cables routed in an area of the same path as required by the power source. Evaluate a dequate separation based upon the criteria in Appendix R, NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.2.4Circuit Failures Due to Inadequate Circuit Coordination Page B-119 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location NRC staff guidance, a nd plant licensing bases.
Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the over all process utilized to perform circuit identification and analysis for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.Selective coordination of breakers / fuses has been confirmed for the electrical distribution syste ms, electrical distribution components, and electrical distribution alignments included in the NSCA, NPO, and Fire PRA as part of the associated circuits assessment provided in Callaway Plant Calculation KC-26, Section 8.7, Associated Circuits - Purpose and Scope. Selec tive coordination of breakers / fuses has been established through a review of the Callaway Plant Electrical Design Calculations identified in the associated circuits assessment.From Section 8.7 of KC-26:"The calculations identified in the references section of this assessment (i.e.,
Section 8.7 of Callaway Plant Calculation KC-2
: 6) address breaker / fuse coordination for the overall plant electrical design. These calculations envelop the topic of Associated Circuits by Common Power Supply (i.e., breaker / fuse coordination) with res pect to 10 CFR 50 Appendix R and NFPA 805. The calculations address all of the electrical power supplies and electrical alignments being credited in the NFPA 805 NS PC Analysis and the Fire PRA. The calculations a re performed consistent with accepted industry practices, and demonstrate that selective coordination has been achieved through the proper application and sizing of circuit overcurrent protection devices.
Furthermore, plant circuit breakers that require an external source of control power to perform their protective overcurrent trip function have been identified and analyzed. The analysis was performed to ensure that all load breakers on the credited switchgear will remain functional to isolate potentially fire affected (non-credited) loads."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-120 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location The common enclosure associated circuit concern deals with the possibility of causing secondary failures due to fire damage to a circuit either whose isolation device fails to isolate t he cable fault or protect the faulted cable from reaching its ignition temperature, or the fire somehow propagates along the cable into adjoining fire areas.The electrical circuit design for most plants provides proper circuit protection in the form of ci rcuit breakers
, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable siz ing are included as part of the original plant electrical design maintained as part of the design change process. Proper protection can be verified by review of as-built drawings and change documentation. Review the fire rated barrier and penetration designs that preclude the propagation of fire from one fire area to the next to demonstrate that adequate measures are in plac e to alleviate fire propagation concerns.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.5.2.5Circuit Failures Due to Common Enclosure ConcernsApplicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the over all process utilized to perform circuit identification and analysis for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.Callaway Plant Calculation KC-26, Section 8.7, Associated Circuits - Purpose and Scope, addresses common enclosure concerns, in clusive of an assessment for the open circuiting of current transformer (CT) secondaries.From Section 8.7 of KC-26:
For common enclosure:
"Chapter 8 of the FSAR includes discussions of the design bases fo r the electrical distribution sy stem, and provides reference to the applicable codes and standards that were utiliz ed in the design of the s ystems, inclusive of ca ble design and sizing, selection of circuit protection, electrical separat ion, etc. The FSAR documents the high level design criteria that Callaway must continue to meet as changes are made to the f acility through design modifications.At Callaway, circuits are provided with overcurrent protection devices that will trip prior to damage to the cable in areas away from the fire. The Callaway Plant electrical single line drawings are identified in the references section of this assessment (i.e., KC-26, Section 8.7.1). These drawings were reviewed with respect to the application of overcurrent protection devices at various voltage levels for both Class 1E and Non-Class 1E circuits. Thi s upper tier review in concert with proper cable sizing practices demonstrated by the available calculations discussed in the breaker / fuse coordination of this a ssessment (i.e., KC-26, Section Page B-121 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology ReviewNFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location 8.7.6) and by the general notes on the single line drawings which specify minimum cable sizes for standard breaker and fuse sizes establish a confidence level for cable protective device applications throughout the plant. These practices indicate that associated circuits of concern by common enclosure wi ll not impact the ability to achieve a safe shutdown."
For current transformers:"In all cases but one, the Callaway plant electrical design has addressed the possibility for CT secondary ignition failures in fire areas remote from the CT through the application of isolation transducers to isolate CT secondary circuits that leave the enclosure (typically switchgear) where the CT is physically located. Where this is not the case, other design features of the CT (i.e., CT turns ratio or relay accuracy class) will ensure that the CT does not pose a secondary ignition fire hazard. Where this is not the case, a fire hazard assessment has been performed to address the potential impact to deterministic safe shutdown capability resulting from a postulated secondary fire occurring in all of the fire area through which the CT s econdary current loop is routed. These fire hazard assessments have identified no advers e impact to deterministic safe shutdown capability.In one case, the CT could not be screened out by any means. Callaway P lant has requested approval from the NRC for this non-screened CT in the NFPA 805 License Amendment Request, LDCN 11-0012, Transition Report Attachment X."
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-122 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment An engineering analysis shall be performed in accord ance with the requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. See Chapter 4 for meth ods of achieving these performance criteria (performance-ba sed or deterministic).By determining the location of each component and cable by fire area and using the cable to equipment relationships described above, the affected safe shutdown equipment in each fire area can be determined. Using the list of affected equipment in each fire area, t he impacts to safe shutdown systems, paths and functions can be determined. Based on an assessment of the number and types of these impacts, the required safe shutdown path for each fire area can be determined. The specific impacts to the selected safe shutdown path can be evaluated using the ci rcuit analysis andevaluation criteria contained in Section 3.5 of this document. Having identified all impacts to the required safe shutdown path in a particular fire area, this section provides guidance on the techniques available for individ ually mitigating the effects of each of the potential impacts.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4Fire Area Assessment and Compliance Strategies Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-123 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment The following criteria and assumptions apply when performing fire area compliance assessment to mitigate the consequences of the circuit failures identified in the previous sections for the required safe shutdown path in each fire area.NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.4.1Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-124 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Assume only one fire in any single fire area at a time.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.1.1Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. The criteria / assumptions listed in Section 3.4.1.1 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 10.2 of KC-26:* A deterministic NSCA is perf ormed for each fire area of the plant utilizing the NSCA database tool to determine the final NSC A compliance strategy assuming one all consuming fire in a single fire area at a time. The objective is to recover at least one NSCA success path for each NSC A performance goal-to-system logic based on identification of the least impacted train of plant equipment.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-125 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Assume that the fire may affect all unprotected cables and equipment within the fire area. This assu mes that neither the fire size nor the fire inten sity is known. This is co nservative and bounds the exposure fire that is required by the regulation.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.1.2Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. The criteria / assumptions listed in Section 3.4.1.2 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 10.2 of KC-26:
* An initial deterministic analysis is run with the NSCA database tool for each fire area (or analysis area)
. The initial fire area analysis as sumes that all unprotected NSCA equipment and NSCA cables physically located in the fire area fail to their worst case condition or state.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-126 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Address all cable and equipment impacts affecting the required safe shutdown path in the fire area. All potential impacts within the fire area must be addressed. The focus of this section is to determine and assess the potential impacts to the required safe shutdown path selected for achieving post-fire safe shutdown and to assure that the required safe shutdown path for a given fire area is properly protected.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.1.3Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. The criteria / assumptions listed in Section 3.4.1.3 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 10.2 of KC-26:* Where the initial NSCA analysis run identifies that there is no success path available to satisfy the NSPC in a given fire ar ea, then su bsequent iterations of the NSCA analysis is required, as necessary, to recover ("resolve") at least o ne success path to satisfy the NSPC in the given fire area. Separation requirements for the deterministic approach to demonstrate the NSPC are identified from NFPA 805, Section 4.2.3. These requirements must be satisfied for the fire area to be deterministically compliant.* The NSCA compliance strategy for these resolved fire areas in the NSCA database typically ends up crediting the least affected success path. This strategy for resolution generally involves the resolution of supporting functions first (i.e., electrical distribution, cooling water, etc.), followed by resolution of the supported front line systems / functions
/ components (i.e., Reactivity Control, Inventory Control, etc.).* The deterministic NSCA for each resolved fire area in the NSCA database may rely on NSCA database equipment and/or cable re solutions.
* NSCA equipment resolutions identify and provide a traceable link for each component failure on a fire area basis that requires further engineering justification to be determined acceptable as-is (i.e., not having any adverse impact to the NSCA), or that requires further engineering review to identify and PROPOSE a plant change such as an OPERATOR MANUAL ACTION, or a physical plant modification.
Each equipment res olution includes descriptive text fields in the NSCA database to document the engineering review basis.* NSCA cable resolutions identify and provide a traceable link for protected cables in the fire area (i.e., raceway protected by ERFBS, raceway embedded in concrete with evaluation, raceway routed in buried ductbank through one or more manhole). Each cable resolution includes descriptive text fields in the NSCA Page B-127 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment database to document the engineering review basis.* Circuit analysis may be utilized to assess and disposition specific circuit failures modes (as documented in the NSCA equipment resolutions)
. Circuit analysis conforms to the criteria / assumptions identified in Section 3. 0 and Section 8.0, Circuit Identification and Analysis, of Callaway Plant Calculation KC-26. The circuit analysis may discount sp urious operation based on a fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-128 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Use manual actions where appr opriate to achieve and maintain post fire safe shutdown conditions in accordance with NRC requirements.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.1.4Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. The criteria / assumptions listed in Section 3.4.1.4 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 3.0 of KC-26:"For NFPA 805, manual operator actions to achieve and maintain safe and stable plant conditions are not allowed by the separation requirements of NFPA 805 Section 4.2.3 with respect to demonstrating the Nuclear Safety Performance Criteria of NFPA 805 Section 1.5.1 is met. Recovery actions (RA) may be credited to achieve and maintain safe and stable plant conditions provided that they are evaluated per the risk-informed, performance bas ed (RIPB) requirements of NFPA 805 Section 4.2.4, and determined to be feasible and reliable (for Fire PRA credited RAs), or determined to be feasible and not adverse to plant risk (for defense-in-depth credited RAs)."From Section 10.2 of KC-26:* The deterministic NSCA for each resolved fire area in the NSCA database may rely on NSCA equipment and/or cable resolutions
.* NSCA equipment resolutions identify and provide a traceable link for each component failure on a fire area basis that requires further engineering justification to be determined acceptable as-is (i.e., not having any adverse impact to the NSCA), or that requires further engineering review to identify and PROPOSE a plant change such as an OPERATOR MANUAL ACTION, or a physical plant modification.
Each equipment res olution includes descriptive text fields in the NSCA database to document the engineering review basis.
* All proposed operator manual act ions are subject to a preliminary review for feasibility (inclusive of the failure mode for motor operated valves as described in USNRC Information Notice 92-18).
* NSCA equipment resolutions that propose operator manual actions are identified as separation issues, and Variations from the Deterministic Requirements (i.e., VFDR) of NFPA 805, Section 4.2.3.
Page B-129 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-130 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Where appropriate to achieve an d maintain cold shutdown within 72 hours, use repairs to e quipment required in support of post-fire shutdown.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.1.5Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment BasisAligns with Intent Callaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. This criteria / assumption listed in Section 3.4.1.5 of NEI 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 10.2 of KC-26:"The NFPA 805 Nuclear Safety Performance Criteria (NSPC) requires the licensee to demonstrate that the plant can achieve and maintain a safe and stable condition, but it does not explicitly require the licensee to demonstrate that cold shutdown c an be achieved within 72 hours and maintained indefinitely thereafter. The Callaway NFPA 805 NSPC analysis has defined the safe and stable condition as being able to achieve and maintain Hot Stand by until such time as the plant can either transition to Cold Shutdown, or can safely return to power operation."
"Safe and stable" for Callawa y Plant is defined in S ection 5.6, Definition of Safe and Stable Plant Conditions for Callaway P lant, of Callaway Plant Calculation KC-26.From Section 5.6 of KC-26:"The NFPA 805 Nuclear Safety Performance Criteria (NSPC) Analysis for Callaway Plant has been developed to ensure that the plan t can achieve and maintain the reactor fuel in a 'safe and stable' condition assuming that a fire event occurs during Callaway Plant Mode 1 (Power Operati on), Mode 2 (Startup
), Mode 3 (Hot Standby), and Mode 4 (Hot Shutdown), up to the point at which the MCC breakers for the Residual Heat Remova l Loop Suction Isola tion Valves, BBPV8702A, BBPV8702B, EJHV8701A, and EJHV8701B, are unlocked and closed. Refer to the Callaway Plant NFPA 805 L icense Amendm ent Request, L DCN 11-0012, Transition Report Attachment C (Table B-3) for the Systems and Components credited with supporting 'safe and stable' plant conditions by fire area.The NFPA 805 Nuclear Safety Capability Assessment (NSCA) has demonstrated that Callaway Plant can achieve and maintain 'safe and stable' conditions for at least 10 hours with the minimum shift operating staff before having to take action to recharge th e nitrogen accumulators. This initial 10 hours p rovides sufficient time for the Emergency Response Organizat ion (ERO) to respond and be available to support 'safe and stable' actions to extend Hot Standby conditions."
Page B-131 August 2011 LIC-28 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-132 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Appendix R compliance requires that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) is free of fire damage (III.G.1.a). Wh en cables or equipment, including associated circuits, are within the same fire area outside primary containment and separation does not already exist, provide one of the following means of separation for the required safe shutdown path(s):- Separation of cables and equipment and associated nonsafety circuits of redundant trains within the same fire area by a fire barrier having a 3-hour rating (III.G.2.a).- Separation of cables and equipment and associated nonsafety circuits of redundant trains within the same fire area by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards. In addition, fi re detectors and an automatic fire suppression system shall be installed in the fire area (III.G.2.b).- Enclosure of cable and equipment and associated non-safety circ uits of one redundant train within a fire area in a fire barrier having a one-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (III.G.2.c).
For fire areas inside noninerted c ontainments, the following additional options are also available:- Separation of cables and equipment and associated nonsafety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards (III.G.2.d);- Installation of fire detec tors and an automatic fire suppression system in the fire area (III.G.2.e); or- Separation of cables and equipment and associated non-safety circuits of redundant trains by a noncombustible radiant energy shield (III.G.2.f).
Use exemptions, deviations and licensing change processes to satisfy the requirements mentioned above and to demonstrate equivalency de pending upon the plant's license requirements.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.1.6Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA fire area assessment. The Page B-133 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment criteria / assumptions listed in Section 3.4.1.6 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 10.0, De terministic Fire Ar ea Assessment a nd Results.From Section 10.2 of KC-26:* Where the initial NSCA analysis run identifies that there is no success path available to satisfy the NSPC in a given fire ar ea, then su bsequent iterations of the NSCA analysis is required, as necessary, to recover ("resolve") at least o ne success path to satisfy the NSPC in the given fire area. Separation requirements for the deterministic approach to demonstrate the NSPC are identified from NFPA 805, Section 4.2.3. These requirements must be satisfied for the fire area to be deterministically compliant.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-134 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Consider selecting other equipment th at can perform the same safe shutdo wn function as the impa cted equipment. In addressing this situation
, each equipment impact, including spurious operations, is to be addressed in accordance with regulatory requirements and the NPPs current licensing basis.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.1.7Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection and fire area assessment. The criteria / assumptions listed in Section 3.4.1.7 of NEI 00-01 are explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 10.0, Deterministic Fire Area Assessment and Results.From Section 7.0 of KC-26:
"The purpose of NSCA model development and component selection activity is to create an acc urate plant model that represents the Nuclear Safety Performance Criteria (NSPC) requirements from NFPA 805, Section 1.5.1. The NSCA model must identify and include plant systems /
functions / components that are required to actively function in order satisfy the NSPC requirements. The NSCA model also must identify include plant systems / functions / components that are not required to actively function, but whose mal-operation (i.e., spurious operation), alone or in combination, could be adverse to meeting the NSPC requirements. The plant model should, within constraints of complexity and cost, maximize the diversity and number of potential success paths t hat are available to satisfy the NSPC requirements."From Section 10.2 of KC-26:* Where the initial NSCA analysis run identifies that there is no success path available to satisfy the NSPC in a given fire ar ea, then su bsequent iterations of the NSCA analysis is required, as necessary, to recover ("resolve") at least o ne success path to satisfy the NSPC in the given fire area. Separation requirements for the deterministic approach to demonstrate the NSPC are identified from NFPA 805, Section 4.2.3. These requirements must be satisfied for the fire area to be deterministically compliant.* The NSCA compliance strategy for these resolved fire areas in the NSCA database typically ends up crediting the least affected success path. This strategy for resolution generally involves the resolution of supporting functions first (i.e., electrical distribution, cooling water, etc.), followed by resolution of the supported front line systems / functions
/ components (i.e., Reactivity Control, Inventory Control, etc.).
Page B-135 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-136 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Consider the effects of the fire on the density of the fluid in instr ument tubing and any subsequent effects on instrument readings or signals associated wi th the protected safe shutdown path in evaluating post fire safe shutdown capability. This can be done systematically or via procedures such as Emer gency Operating Procedures.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.1.8Criteria/Assumptions Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 3.0, NSCA Criteria / Assumptions, lists criteria / assumptions pertaining to the NSCA model development and component selection and fire area assessment. This criteria / assumption listed in Section 3.4.1.8 of NE I 00-01 is explicitly stated in the calculation.KC-26, Section 3.0, identifies the criteria / assumptions utilized in KC-26, Section 7.0, NSCA Model Development and Component Selection, and Section 10.0, Deterministic Fire Area Assessment and Results.From Section 7.8 of KC-26:
Consideration of the potential adverse NSCA impact resulting from the heating of instrument tubing sensing lines is applicable to, and included the development of the NSCA model and component selection, a nd the deterministic fire area assessment.
Instrument tubing sensing lines for NSCA instrumentation have been identified, located, and incorporated into the NSCA model as components with a "-SL" suffix. These instrument tubing sensing line components fail their associated transmitting device through the component-to-component logic success path relationship in the NSCA model database. The instrument tubing sensing lines components are evaluated on a fire area basis, as applicable.From Section 10.2 of KC-26:"* When resolving component and cable failures the analyst is required to consider
, and address as nece ssary, the potential impact resulting from the following:
* heating of instrument tubing sensing lines resulting in e rroneous or unreliable signals from NSCA analyz ed instrumentation (r efer to Section 7.
8 of [Calculation KC-26])"Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-137 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Refer to Figure 3-5 for a flowchart illustrating the various steps involved in performing a fire area assessment.
Use the following methodology to assess the impact to safe shutdown and demonstrate Appendix R compliance:
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance 3.4.2Methodology for Fire Area Assessment Applicable NoneApplicability CommentsAlignment StatementAlignment Basis Not RequiredGeneric paragraph. Detailed alignment discussed in subsequent reference paragraphs.
Reference DocumentsNot Applicable Page B-138 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Identify the safe shutdown cables, equipment and systems located in each fire area that may be potentially damaged by the fire. Provide this information in a report format. The report may be sorted by fire area and by system in order to understand the impact to each safe shutdown path within each fi re area (see Attachment 5 for an example of an Affected Equipment Report).
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.2.1Identify the Affected Equipment by Fire Area Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 10.0, Dete rministic Fire Area Ass essment and Results, identifies the overall process utilized to perf orm deterministic fire area assessment and the fire area assessment results for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.From Section 10.2 of KC-26:
* An initial deterministic analysis is run with the NSCA database tool for each fire area (or analysis area)
. The initial fire area analysis assu mes that all NSCA equipment and NSCA cables physically located in the fire area fail to their worst case condition or state.The NSCA analysis structure a nd results for each fire area are displayed on the computer screen as a tree structure beginning with performance goals on the left leading to system logic. The system logic display shows all equipment in every success path. This tree structure diagram of the NSCA model shows the results of the analysis in the following way.
* Failed equipment, cables, sys tems, and performance goals (displayed in red text)* Resolved equipment, cables, systems (displayed in blue text, with check mark)
* Design Change (ACP) logics, if applicable (
displayed in bold text)
* All items that show with black text are unaffected by the initial failures used in this analysis.The following hardcopy reports are available from the NSCA database tool for each deterministic analysis area. T hese reports (f or the final "resolv ed" fire areas) are include in Attachmen t 10-2 of Callaway Plant Calculation KC-26:
* Status of Goals - a list of all NSCA performance goals, identifies eac h NSCA performance goal as useable or failed Page B-139 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment
* Status of Systems by Goals - identifies the NSCA systems associated with each NSCA performance goal as useable or failed, identifies each NSCA performance goal as useable or failed* Status of Systems
- a list of all NSCA systems, identifies each NSCA system as useable or failed
* Status of Safe Shutdown Equipment - a list of all NSCA components, identifies each component as useable, failed, or resolved; for each failed NSCA component, identifies the failed component-to-component logic success path (i.e., support component) and failed component-to-cable logic success path (i.e., required cable), as applicable; for each resolved NSCA component, identifies the equipment resolution (includes the equipment VFDR statement with the associated equipment VFDR cl osure statement
, as applicable)
* Status of Safe Shutdown Equipment by System - identifies the NSCA components associated with each NSCA system as useable, failed, or resolved, identifies each NSCA system as useable or failed; for each failed NSCA component, identifies the failed component-to-component logic success path (i.e., support component) and failed component-to-cable logic success path (i.e., required cable), as applicable; for each resolved NSCA component, identifies the equipment resolution (includes the equipment VFDR statement with the associated equipment VFDR closure statement, as applicable)* Failed Safe Shutdown Cables - a list of all NSCA cables in the fire area, identifies each cable as useable, failed, or recovered* Failed Safe Shutdown Cables, Fire Zone Location in this Analys is - a list of all NSCA cables in the fire area, identifies the fire zone(s) each cable is contained within, inside the fire area* Analysis Cables status for this analysis - identifies the NSCA cables in the fire area as useable, failed, or resolved; for each resolved NSCA cable, identifies the cable resolution (includes the cable VFDR statement with the associated cable VF DR closure stat ement, as applicable)* Resolutions used in th is analysis - a list of all resolved NSCA components followed by a list of all resolved NSCA cables (for cables in the fire area); for each resolved NSCA component, identif ies the equipment resolution (includes the equipment VFDR statement with the associated equipme nt VFDR closure statement, as applicable); for each resolved NSCA cable, identifies the cable resolution (includes the cable VFDR statement wit h the associated cable VFDR closure state ment, as applicable)
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-140 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Based on a review of th e systems, equipment and cables within each fire area, d etermine which s hutdown paths are either unaffected or lea st impacted by a postulated fire within the fire area. Typically, the safe shutdown path with the least number of cables and equipment in the fire area would be select ed as the required safe shutdown path. Consider the circuit failure c riteria and the possible mitigating strateg ies, however, in selecting the required safe shutdown path in a particular fire area. Review support systems as a part of this assessment since their availabil ity will be important to the ability to achieve and maintain safe shutdown. For example, impacts to the electric power distribution system for a particular safe shutdown path could present a major impediment to using a particular path for safe shutdown. By identifying this early in the assessment process, an unnecessary amount of time is not spent assessing impacts to the frontline systems that will require t his power to suppor t their operation.Based on an assessment as described above, designate the required safe shutdown path(s) for the fire area. Identify all equipment not in the safe shutdown path whose spurious op eration or mal-operat ion could affect the shutdown function. Include these cables in the s hutdown function list. Fo r each of the safe shutdown cables (located in the fi re area) that are part of the required safe shutdown path in the fire area, perform an evaluation to determine t he impact of a fire-induced cable failure on the corresponding safe shutdown equipment and, ultimately, on the required safe shutdown path. When evaluating the safe shutdown mode for a particular piece of equipment, it is important to consider the equipments position for the specific safe shutdown scenario for the full duration of the shutdown scenario. It is possible for a piece of equipment to be in two different states depending on the shutdown scenario or the stage of shutdown within a particular shutdown scenario. Document information related to the normal and shutdown positions of equipment on the safe shutdown equipment list.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.2.2Determine the Shutdown Paths Least Impacted By a Fire in Each Fire Area Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 10.0, Dete rministic Fire Area Ass essment and Results, identifies the overall process utilized to perf orm deterministic fire area assessment and the fire area assessment results for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.From Section 10.2 of KC-26:* Where the initial NSCA analysis run identifies that there is no success path available to satisfy the NSPC in a given fire ar ea, then su bsequent iterations of the NSCA analysis is required, as necessary, to recover ("resolve") at least o ne success path to satisfy the NSPC in the given fire area. Separation requirements for Page B-141 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment the deterministic approach to demonstrate the NSPC are identified from NFPA 805, Section 4.2.3. These requirements must be satisfied for the fire area to be deterministically compliant.* The NSCA compliance strategy for these resolved fire areas in the NSCA database typically ends up crediting the least affected success path. This strategy for resolution generally involves the resolution of supporting functions first (i.e., electrical distribution, cooling water, etc.), followed by resolution of the supported front line systems / functions
/ components (i.e., Reactivity Control, Inventory Control, etc.).The NSCA model development and component selection process (refer to Section 7.0, NSCA Model Development and Component Selection, of Callaway Plant Calculation KC-26) identifies c omponents whose spurious operat ion alone (i.e., single spurious operation -
SO), or in combination with other components (i.e.,
multiple spurious operation - MSO), could be adverse to the NSCA functional requirements of one or more NSCA system / function / component, and includes these components (as NSCA components) in the respective system-to-co mponent logic success path and/or the re spective component-to-component logic success path. Consequently, "resolving" one success path for each NSPC performance goal in the deterministic analysis implicitl y includes addressing the spurious operation(s) that co uld be adverse to meeting each NSPC performance goal.* The deterministic NSCA for each resolved fire area in the NSCA database may rely on NSCA database equipment and/or cable re solutions.
* NSCA equipment resolutions identify and provide a traceable link for each component failure on a fire area basis that requires further engineering justification to be determined acceptable as-is (i.e., not having any adverse impact to the NSCA), or that requires further engineering review to identify and PROPOSE a plant change such as an OPERATOR MANUAL ACTION, or a physical plant modification.
Each equipment res olution includes descriptive text fields in the NSCA database to document the engineering review basis.* NSCA cable resolutions identify and provide a traceable link for protected cables in the fire area (i.e., raceway protected by ERFBS, raceway embedded in concrete with evaluation, raceway routed in buried ductbank through one or more manhole). Each cable resolution includes descriptive text fields in the NSCA database to document the engineering review basis.* Circuit analysis may be utilized to assess and disposition specific circuit failures modes (as documented in the NSCA equipment resolutions)
. Circuit analysis conforms to the criteria / assumptions identified in Section 3. 0 and Section 8.0, Circuit Identification and Analysis, of Callaway Plant Calculation KC-26. The circuit analysis may discount sp urious operation based on a fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage.The NSCA database generated hardcopy reports and other supplemental reports created to document the final "resolved" assessment for each fire area are included in Attachment 10-2 of Callaway Plant Calculation KC-26. These reports identify the credited success path for each NSPC performance goal.NSCA component position information is identified in the NFPA 805 NSPC Equipment List (Attachment 7-5 of Callaway Plant Calculation KC-26), and in the NSCA database.
Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-142 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Using the circuit analysis and evaluation criteria con tained in Section 3.5 of this document, determine the equipment that can impact safe shutdown and that can potentially be impacted by a fire in the fire area, and what those possible impacts are.NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.2.3Determine Safe Shutdown Equipment Impacts Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsCallaway Plant Calcu lation KC-26, Section 10.0, Dete rministic Fire Area Ass essment and Results, identifies the overall process utilized to perf orm deterministic fire area assessment and the fire area assessment results for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.From Section 10.2 of KC-26:* Where the initial NSCA analysis run identifies that there is no success path available to satisfy the NSPC in a given fire ar ea, then su bsequent iterations of the NSCA analysis is required, as necessary, to recover ("resolve") at least o ne success path to satisfy the NSPC in the given fire area. Separation requirements for the deterministic approach to demonstrate the NSPC are identified from NFPA 805, Section 4.2.3. These requirements must be satisfied for the fire area to be deterministically compliant.* The NSCA compliance strategy for these resolved fire areas in the NSCA database typically ends up crediting the least affected success path. This strategy for resolution generally involves the resolution of supporting functions first (i.e., electrical distribution, cooling water, etc.), followed by resolution of the supported front line systems / functions
/ components (i.e., Reactivity Control, Inventory Control, etc.).* The deterministic NSCA for each resolved fire area in the NSCA database may rely on NSCA database equipment and/or cable re solutions.
* NSCA equipment resolutions identify and provide a traceable link for each component failure on a fire area basis that requires further engineering justification to be determined acceptable as-is (i.e., not having any adverse impact to the NSCA), or that requires further engineering review to identify and PROPOSE a plant change such as an OPERATOR MANUAL ACTION, or a physical plant modification.
Each equipment res olution includes descriptive text fields in the NSCA database to document the engineering review basis.* Circuit analysis may be utilized to assess and disposition specific circuit failures modes (as documented in the NSCA equipment resolutions)
. Circuit analysis conforms to the criteria / assumptions identified in Section 3. 0 and Section 8.0, Circuit Identification and Analysis, of Callaway Plant Calculation KC-26. The circuit analysis may discount sp urious operation based on a fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage.
Page B-143 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Reference DocumentsCalculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-144 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment The available deterministic methods for mitigating the effects of circuit failures are summarized as follows (see Figure 1-2):- Provide a qualified 3-fire rated barrier.
- Provide a 1-hour fire ra ted barrier with automatic su ppression and detection.- Provide separation of 20 feet or greater with automatic suppression and detection and demonstrate that there are no intervening combustibles within the 20 foot separat ion distance.- Reroute or relocate the circuit/equipment, or per form other modifications to resolve vulnerability.
- Provide a procedural action in accordance with regulatory requirements.
- Perform a cold shutdown re pair in accordance with regulatory requirements.- Identify other equipment not affected by the fire capable of performing the same safe shutdown function.- Develop exemptions, deviations, Generic Letter 86-10 evaluation or fire protection design change evaluations with a licensing change process.
Additional options are a vailable for non-inerted containments as described in 10 CFR 50 Appendix R section III.G.2.d, e and f.NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.2.4Develop a Compliance Strategy or Disposition to Mitigate the Effects Due to Fire Damage to Each Required Component or CableApplicable NoneApplicability CommentsAlignment StatementAlignment BasisAligns with Intent Callaway Plant Calcu lation KC-26, Section 10.0, Dete rministic Fire Area Ass essment and Results, identifies the overall process utilized to perf orm deterministic fire area assessment and the fire area assessment results for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.From Section 10.2 of KC-26:* Where the initial NSCA analysis run identifies that there is no success path available to satisfy the NSPC in a given fire ar ea, then su bsequent iterations of the NSCA analysis is required, as necessary, to recover ("resolve") at least o ne success path to satisfy the NSPC in the given fire area. Separation requirements for the deterministic approach to demonstrate the NSPC are identified from NFPA 805, Section 4.2.3. These requirements must be satisfied for the fire area to be deterministically compliant.* The NSCA compliance strategy for these resolved fire areas in the NSCA database typically ends up crediting the least affected success path. This strategy for resolution generally involves the resolution of supporting functions first (i.e., electrical distribution, cooling water, etc.), followed by resolution of the supported Page B-145 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment front line systems / functions
/ components (i.e., Reactivity Control, Inventory Control, etc.).* The deterministic NSCA for each resolved fire area in the NSCA database may rely on NSCA database equipment and/or cable re solutions.
* NSCA equipment resolutions identify and provide a traceable link for each component failure on a fire area basis that requires further engineering justification to be determined acceptable as-is (i.e., not having any adverse impact to the NSCA), or that requires further engineering review to identify and PROPOSE a plant change such as an OPERATOR MANUAL ACTION, or a physical plant modification.
Each equipment res olution includes descriptive text fields in the NSCA database to document the engineering review basis.* NSCA cable resolutions identify and provide a traceable link for protected cables in the fire area (i.e., raceway protected by ERFBS, raceway embedded in concrete with evaluation, raceway routed in buried ductbank through one or more manhole). Each cable resolution includes descriptive text fields in the NSCA database to document the engineering review basis.* Circuit analysis may be utilized to assess and disposition specific circuit failures modes (as documented in the NSCA equipment resolutions)
. Circuit analysis conforms to the criteria / assumptions identified in Section 3. 0 and Section 8.0, Circuit Identification and Analysis, of Callaway Plant Calculation KC-26. The circuit analysis may discount sp urious operation based on a fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage.
* NSCA equipment resolutions that propose operator manual actions are identified as separation issues, and Variations from the Deterministic Requirements (i.e., VFDR) of NFPA 805, Section 4.2.3.The NSCA database generated hardcopy reports and other supplemental reports created to document the final "resolved" assessment for each fire area are included in Attachment 10-2 of Callaway Plant Calculation KC-26. These reports identify the credited success path for each NSPC performance goal.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-146 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.4 Fire Area Assessment Assign compliance strategy statements or codes to components or cables to identify the justification or mitigating actions proposed for achieving safe shutdown. The justification should address the cumulative effect of the actions relied upon by the licensee to mitigate a fire in the area. Provide each piece of safe shutdown equipment, equipment not in the path whose spurious operation or mal-operation could affect safe shutdown, and/or cable for the required safe shutdown path with a specific compliance strategy or disposition. Refer to Attachment 6 for an example of a Fire Area Assessment Report documenting e ach cable disposition.
NEI 00-01 Ref NEI 00-01 Sectio n 3 Guidance3.4.2.5Document the Compliance Strategy or Disposition Determined to Mitigate the Effects Due to Fire Damage to Each Required Component or Cable Applicable NoneApplicability CommentsAlignment StatementAlignment Basis AlignsSee the Alignment Basis for Item 3.4.2.2. The methodology for assigning compliance strategies (i.e., creating equipment and cable resolutions in the NSCA database) is consistent with the prescribed method.
Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0 Page B-147 August 2011 Ameren Missouri Callaway Plant N FPA 805 Transition ReportAttachment B -  Nuclear Safety Capability Assessment Methodology Review Page B-148 August 2011 to ULNRC-06060 
 
ATTACHMENT C: CHANGES TO THE TRANSITION REPORT ATTACHMENT C Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 198814.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptions Fire ZoneDescription1101General Floor Area No. 11102Chiller and Surge Tanks Area1103Letdown Chiller Heat Exchanger Room1104Letdown Reheat Heat Exchanger Room1105Valve Compartment1106Moderating Heat Exchanger Room1115Normal Charging Pump Room1120General Floor Area No. 21121Access Pit1122General Floor Area No. 3 & Auxiliary Building Tool Issue Area 1123Passage1124Valve Compartment1125Letdown Heat Exchanger Room1128Storeroom No. 31129Auxiliary Condenser Recovery and Storage Tank Room1130North Corridor1201Vestibule1202Access Area B & Chiller Surge Tank Area1203Pipe Space B1204Pipe Space A1205Access Area A1206Pipe Chase 1207Pipe Chase1329VestibuleAugust 2011 C-2 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881~"Process MonitoringRCS Pressure Channel IPressurizer Pressure Channel IPressurizer Level Channel I Ex-core Neutron Monitoring Channel I RCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel IISteam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel IAux. Feedwater Flow to Steam Gen. A Channel IRCS Loop D (4) T-hot Temperature Channels II and VI RCS Loop D (4) T-cold Temperature Channels I and VI Steam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVRCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel I Steam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and II RCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel III Steam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel ITurbine Driven Aux. Feedwater Pump Suction Pressure Channel IIRefueling Water Storage Tank Level Channel IDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A, or Steam Generators A, B, C, and D are supplied by the TDAFW Pump. Credited AFW Pump and Steam Generators depend on location of fire within Fire Area.August 2011 C-3LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881RCS Inventory ControlVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCP Thermal Barrier Cooling is credited, with cooling water provided by CCW Pumps A and C. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A, C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. A-01-001, A-01-002, A-01-003, A-01-004, A-01-005, A-01-006, A 007, A-01-008, A-01-009, A-01-010See VFDR No. A-01-012RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.See VFDR No. A-01-011Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-4 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A). 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&',%Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 2, dated 05/1984 based on the following:  1. Elevator and dumbwaiter doors are rated at 1-1/2 hours as required by ANSI A17.1. 2. The 1-1/2 hour doors are an industry standard and, as stated in ANSI A17.1, are acceptable for use in a 2-hour rated elevato r or dumbwaiter shaft. 3. For a fire to propagate from one floor elevation to another, it would have to penetrate two doors.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-5 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881%~&Deviation submitted per 6/29/1981 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, to justify the two sets of non-rated equipment hatchways in the northern and southern ends of the auxiliary building corridors, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Low fuel loading and configuration of equipment.2. Steel hatch covers are provided for each hatchway.3. Automatic sprinkler water curtains are provided for each hatchway at elevations 2000'-0", 2026'-0", and 2047'-0" to separate the corridor fire areas.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zones 1103, 1104, 1105, 11 06, 1123, 1124, 1125
, 1129, 1202, 1203, 1204, and 1329), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zone 1101 and no suppression in 1102, 1103, 1104, 1105, 1106, 1115, 1120, 1121, 1122, 1123, 1124, 1125, 1128, 1129, 1130, 1201, 1202, 1203, 1204, 1205, and 1329), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following: 
: 1. Minimal fire hazards.
: 2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-6LIC-25LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881....3.%654(5An excessive gap in the bottom of Door DSK11271 connecting Fire Areas A-1 and A-6 is acceptable based on the lack of intervenin g combustibles at/near the location of the door. Fire Area A-6 is a stairwell and, as such, no transient combustibles are expected near this doorway. Therefore DSK11271 is considered a non-rated feature commensurate with the fire hazards in the two areas and it provides an equivalent level of protection as a 3 hour rated fire door by prohibiting the propagation of fire between the two fire areas.&The removal of Thermo-Lag fire barriers from RHR and containment spray hatch covers in Fire Zones 1203 and 1204 is acceptable based on the fact that there are no fixed ignition sources or fixed combustibles in the vicinity of the hatches, and the fact that the areas are infrequently accessed due to the fact that they are "locked high radiation areas." Further, gaps between the individual plates between A-1 and A-8 are acceptable based on the fact that heat transferred would be rapidly dissipated in the large volume of A-8; there are no combustibles in the immediate vicinity of the hatch covers; combustible loading in the area is very low; there are no ignition sources in the area of the plates; ventilation is not affected based on gap size; an automatic pre-action suppression system is installed in the area of the hatch on the 2000' elevation; and the fire brigade is trained to aggressively control fires in this area.Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.August 2011 C-7LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881885((BBHV8141A - Cable damage (5BBK05AA) to BBHV8141A. Cable damage can spuriously close the Reactor Coolant Pump A Seal # 1 Water Outlet Isolation Valve, BBHV8141A (spurious closure is only credible assuming external hot shorts). This valve is required to remain open in order to maintain the effectiveness of thermal barrier cooling for the Reactor Coolant Pump Seal Package. This requirement is based on the latest Westinghouse Owners Group Analysis for the performance of the Reactor Coolant Pumps Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump Seal Injection may be interrupted (until recovered) in this area due to fire damage potentially affecting the Charging Pump Suction Supply Valves fro m the Volume Control Tank (VCT), BGLCV0112B and BGLCV0112C, and the Refueling Water Storage Tank (RWST), BNLCV0012D and BNLCV0112E (refer to the fire area A-01 VFDRS for equipment BGLCV0112B, BGLCV0112B-P,
 
BGLCV0112C-P, and BNLCV0112D). With the exception of spurious closure of BBHV8141A, component cooling water for the Reactor Coolant Pump A Thermal barrier is unaffected and available in this area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85(4BBHV8141B - Cable damage (5BBK05BA) to BBHV8141B. Cable damage can spuriously close the Reactor Coolant Pump B Seal # 1 Water Outlet Isolation Valve, BBHV8141B (spurious closure is only credible assuming external hot shorts). This valve is required to remain open in order to maintain the effectiveness of Thermal Barrier Cooling for the Reactor Coolant Pump Seal Package. This requirement is based on the latest Westinghouse Owners Group Analysis for the performance of the Reactor Coolant Pumps Seal Package following the loss of Seal Cooling. Note that Reactor Coolant Pump Seal Injection may be interrupted (until recovered) in this area due to fire damage potentially affecting the Charging Pump Suction Supply Valves fro m the Volume Control Tank (VCT), BGLCV0112B and BGLCV0112C, and the Refueling Water Storage Tank (RWST), BNLCV0112D and BNLCV0112E (refer to the fire area A-01 VFDRs for equipment BGLCV0112B, BGLCV0112B-P, and BGLCV0112C-P, and BNLCV0112D). With the exception of spurious closure of BBHV8141B, component cooling water for the Reactor Coolant Pump B Thermal barrier is unaffected and available in this area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-8 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 1988185(BGHV8149A - Cable damage (5BGK35AB and 5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85(:BGHV8149B - Cable damage (5BGK35BB and 5BGK35BD) to BGV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issueThe VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.BGHV8149C - Cable damage (5BGK35CB and 5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-9 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 1988185(7BGLCV0112B - Cable damage (1BGG12AD) to BGLCV0112B. Cable damage can spuriously close the Chemical Volume Control System Volume Control Tank Outlet Upstream Isolation Valve, BGLCV0112B (spurious closure is only credible assuming external hot shorts), and may also cause the valve to fail as-is (open). Similar failure mode(s) may also occur for BGLCV0112C
, the Chemical Volume Control System Volume Control Tank Outlet Downstream Isolation Valve, due to cable damage (4BGG12BD). Both of these valves are required open (i.e., to not spuriously close) to prevent failure of the Credited Charging Pump (when running) from a loss of suction. After the Charging Pump Suction flowpath has been aligned to the Refueling Water Storage Tank, either one of these valves is required to close on demand in order to isolate the Charging Pump Suction flowpath from the Volume Control tank to prevent gas binding of the Credited Charging Pump. Note that access to these valves requires the plant operator to transit through the fire affected area. The valve limit/torque switches for  BGLCV0112B can be bypassed by fire damage to cable 1BGG12AD. No loss of the offsite power or spurious undervoltage is expected to NB01 and NB02 in this fire area. The valve limit/torque switches for BGLCV0112C cannot be bypassed by fire damage to Cable 4BGG12BD. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85(;BGLCV0112B-P - Cable damage (1BGG12AD) to BGLCV0112B. Cable damage can spuriously close the Chemical Volume control System Volume Control Tank Outlet Upstream Isolation Valve, BGLCV0112B (spurious closure is only credible assuming external hot shorts), and may also cause the valve to fail as-is (open). Similar failure mode(s) may also occur for BGLCV0112C
-P, the Chemical Volume Control System Volume Control Tank Outlet Downstream Isolation Valve, due to cable damage (4BGG12BD). Both of these valves are required open (i.e. to not spuriously close) to prevent failure of the Credited Charging Pump (when running) from a loss of suction. After the Charging Pump Suction flowpath has been aligned to the Refueling Water Storage Tank, either one of these valves is required to close on demand in order to isolate the Charging Pump Suction flowpath from the Volume Control Tank to prevent gas binding of the Credited Charging Pump. Note that access to these valves requires the plant operator to transit through the fire affected area. The valve limit/torque switches for BGLCV0112B can be bypassed b y fire damage to Cable 1BGG12AD. No loss of offsite power or spurious undervoltage is expected to NB01 and NB02 in this fire area. The valve limit/torque switches for BGLCV0112C cannot be bypassed by fire damage to cable 4BGG12BD. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue. The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-10 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 1988185(6BGLCV0112C-P - Cable damage (4BGG12BD) to BGLCV0112C. Cable damage can spuriously close the Chemical Volume Control System Volume Control Tank Outlet Upstream Isolation Valve, BGLCV0112B (spurious closure is only credible assuming external hot shorts), may also cause the valve to fail as-is (open). Similar failure mode(s) may also occur for BGLCV0112B-P, the Chemical Volume Control System Volume Control tank Outlet Downstream Isolation Valve, due to cable damage (1BGG12AD). Both of these valves are required open (i.e., to not spuriously close) to prevent failure of the Credited Charging Pump (when running) from a loss of suction. After the Charging Pump Suction flowpath has been aligned to the Refueling Water Storage Tank, either one of these valves is required to close on demand in order to isolate the Charging Pump Suction flowpath from the Volume Control tank to prevent gas binding of the Credited Charging Pump. Note that access to these valves requires the plant operator to transit through the fire affected area. The valve limit/torque switches for BGLCV0112B can be bypassed by fire damage to cable 1BGG12AD. No loss of offsite power or spurious undervoltage is expected to NB01 and NB02 in this fire area. The valve limit/torque switches for BGLCV0112C cannot be bypassed by fire damage to cable 4BGG12BD. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85(+BNLCV0112D - Cable damage (1BNG01AA and 1BNG01AB) to BNLCV0112D. Cable damage cannot spuriously open or close Charging Pump A Suction from Refueling Water Storage Tank Isolation Valve, BNLCV0112D, but may cause the valve to fail as-is (closed). A similar failure mode may also occur for BNLCV0112E, the Charging Pump B Suction from Refueling Water Storage Tank Isolation Valve, due to cable damage (4BNG01BA and 4BNG01BB). Either one of these valves is required to open on demand to align the Refueling Water Storage Tank as the source of Reactor Coolant System Inventory makeup to the Credited Charging Pump, and to prevent failure of the Credited Charging Pump (when running) from a loss of suction. Note that access to these valves requires the plant operator to transit through the fire affected area. The valve limit/torque switches for BNLCV0112D can be bypassed by fire damage to cable 1BNG01AB. The valve limit/torque switches for BNLCV0112E can be bypassed by fire damage to cable 4BNG01BB. No loss of offsite power or spurious undervoltage is expected to NB01 and NB02 in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issu e.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-11 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 1988185(5EMHV8803A - Cable damage (1EMG02AA and 1EMG02AB) to EMHV8803A. Boron Injection Header Supply from Charging Pump A Isolation Valve, EMHV8803A, cannot spuriously open or close due to cable damage (1EMG02AA and 1EMG02AB). However, the valve may also fail as-is (closed). The desired position for this valve is throttled open to establish the train A Boron Injection Flowpath, which may become necessary to restore pressurize level (maintain positive control over RCS inventory and pressure) if RCS inventory makeup with alternate RCP Seal Injection is not sufficient. Note that access to this valve requires the plant operator to transit through the fire affected area. The valve limit/torque switches for EMHV8803A can be bypassed by fir e damage to cable 1EMG02AB. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85((PZR-HTR-BU-A - Backup Pressurizer Heater Groups A and B are subject to cable damage and/or loss of DC control power (cables 5BBG22AD and 5BBG22AG for the Group A Heaters, the breaker PG2101 close/trip control cables - cable 6PKG11BA for the Group B Heaters, the AC power cable to battery charger PK22, ultimately causing loss of DC control power for Breaker PG2201). Neither backup group of pressurizer heaters is available for safe shutdown in this fire area. Loss of pressurizer heater capability may adversely impact the ability of the plant to maintain safe and stable (potential adverse impact to RCS pressure control). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.  (Both backup groups of pressurizer heaters are recoverable with local manual operator actions).The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85(4SGK05B - No direct cable damage to SGK05B. Loss of battery charger - 125VDC No. 2, PK22, to Switchboard 125 VDC Bus, PK02, from cable damage (6PKG11BA) results in the eventual loss of 125 VDC power for the Fire Protection Interlock Circuit. Loss of 125 VDC power for the Fire Protection Interlock will result in a loss of ventilation from the Train B ESF Switchgear rooms air conditioning unit, SGK05B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue. The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-12 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881&$.>.?>..A&."&.&Ionization 100Detection NNNY1101NPre-action SKC43Suppression NNNYNERFBSNoneFeatureNNYYNIonization 100Detection NNNY1102NIonization 101Detection NNNYNPre-action SKC43Suppression NNNYNN/ANoneFeatureN/ANoneDetection 1103N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1104N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1105N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1106N/ANoneSuppression N/ANoneFeatureIonization 101Detection NNNN1115NN/ANoneSuppression N/ANoneFeatureAugust 2011 C-13 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881&$.>.?>..A&."&.&Ionization 101Detection NNNN1120NIonization 102Detection NNNNNPre-action SKC43Suppression NYNYNprovides water curtain for hatch N/ANoneFeatureIonization 101Detection NNNN1121NN/ANoneSuppression N/ANoneFeatureIonization 100Detection NYNY1122Nactivates hatch water curtainIonization 101Detection NNNNNPre-action SKC43Suppression NYNYNprovides water curtain for hatch N/ANoneFeatureN/ANoneDetection 1123N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1124N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1125N/ANoneSuppression N/ANoneFeatureIonization 117Detection NNNN1128NN/ANoneSuppression N/ANoneFeatureAugust 2011 C-14 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881&$.>.?>..A&."&.&
N/ANoneDetection 1129N/ANoneSuppression N/ANoneFeatureIonization 100Detection NYNY1130NPre-action SKC43Suppression NNNYNN/ANoneFeatureIonization 102Detection NNNN1201NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1202N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1203N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1204N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1205N/ANoneSuppression N/ANoneFeatureAugust 2011 C-15 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881&$.>.?>..A&."&.&Ionization 120Detection NNNY1206NWet PipeSKC48Suppression NNNYN20-ft Separation ZoneNoneFeatureNNNYNERFBSNoneFeatureNNYYNIonization 120Detection NNNY1207NWet PipeSKC48Suppression NNNYN20-ft Separation ZoneNoneFeatureNNNYNERFBSNoneFeatureNNYYNN/ANoneDetection 1329N/ANoneSuppression N/ANoneFeatureSLE RD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation
- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationAugust 2011 C-16 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-1Auxiliary Building - El. 1974, 19881To meet deterministic separation criteria  Fire Zones 1206 and 1207 have a 20-foot separation zone free of intervening combustibles with automatic detection and suppression. The 20-foot separation zone is clearly marked on the floor and designated as a "No Storage" location.&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-17 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-2Auxiliary Building Safety-Related Pump Area1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel IISteam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel I Steam Gen. A Atmos. Steam Dump Pressure Channel I Aux. Feedwater Flow to Steam Gen. A Channels I and IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VI Steam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription1111Residual Heat Removal Pump Room A1112Containment Spray Pump Room A1113Safety Injection Pump Room A1114Centrifugal Charging Pump Room AAugust 2011 C-18LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-2Auxiliary Building Safety-Related Pump Area1RCS Inventory ControlVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A).
: 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-19 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-2Auxiliary Building Safety-Related Pump Area1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-20 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-2Auxiliary Building Safety-Related Pump Area1&$.>.?>..A&."&.&Ionization 101Detection NNYN1111NN/ANoneSuppression N/ANoneFeatureIonization 101Detection NNYN1112NN/ANoneSuppression N/ANoneFeatureIonization 101Detection NNYN1113NN/ANoneSuppression N/ANoneFeatureIonization 101Detection NNYN1114NN/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-21 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-2Auxiliary Building Safety-Related Pump Area1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-22 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-3Boric Acid Tank Rooms1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel IAux. Feedwater Flow to Steam Gen. A Channels I and IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VI Steam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and II Condensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IV4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription1116Boric Acid Tank Room B1117Boric Acid Tank Room A1407Boric Acid Batching Tank AreaAugust 2011 C-23LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-3Boric Acid Tank Rooms1RCS Inventory ControlCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pump B and D, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.August 2011 C-24 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-3Boric Acid Tank Rooms1None8&$.>.?>..A&."&.&Flame101Detection NNNN1116NIonization 101Detection NNNNNN/ANoneSuppression N/ANoneFeatureFlame101Detection NNNN1117NIonization 101Detection NNNNNN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1407N/ANoneSuppression SLERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-3Boric Acid Tank Rooms1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-26 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-4Auxiliary Building Safety-Related Pump Area1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel IISteam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel II Steam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1107Centrifugal Charging Pump Room B1108Safety Injection Pump Room B1109Residual Heat Removal Pump Room 'B'1110Containment Spray Pump Room 'B'August 2011 C-27LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-4Auxiliary Building Safety-Related Pump Area1RCS Inventory ControlVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A).
: 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-4Auxiliary Building Safety-Related Pump Area1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-29 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-4Auxiliary Building Safety-Related Pump Area1&$.>.?>..A&."&.&Ionization 101Detection NNYN1107NN/ANoneSuppression N/ANoneFeatureIonization 101Detection NNYN1108NN/ANoneSuppression N/ANoneFeatureIonization 101Detection NNYN1109NN/ANoneSuppression N/ANoneFeatureIonization 101Detection NNYN1110NN/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-30 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-4Auxiliary Building Safety-Related Pump Area1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-31 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-5Auxiliary Building Stairway and Elevator (south)1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel I Ex-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel IISteam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel II Steam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and II RCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1119Stair A-11601Elevator No. 2 Machine RoomAugust 2011 C-32LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-5Auxiliary Building Stairway and Elevator (south)1RCS Inventory ControlMaintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-33 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-5Auxiliary Building Stairway and Elevator (south)1%~&',%Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 2, dated 05/1984 based on the following:  1. Elevator and dumbwaiter doors are rated at 1-1/2 hours as required by ANSI A17.1. 
: 2. The 1-1/2 hour doors are an industry standard and, as stated in ANSI A17.1, are acceptable for use in a 2-hour rated elevato r or dumbwaiter shaft. 3. For a fire to propagate from one floor elevation to another, it would have to penetrate two doors.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-34 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-5Auxiliary Building Stairway and Elevator (south)1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-35 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-5Auxiliary Building Stairway and Elevator (south)1&$.>.?>..A&."&.&Ionization 108Detection NNNN1119NN/ANoneSuppression N/ANoneFeatureIonizationNoneDetection NNNN1601NN/ANoneSuppression N/ANoneFeatureIonization 102Detection NNNNElevator Lobby Nthe elevator lobby is specified for location purposes only SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNone&There is no equipment susceptible to water damage in this area. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-36 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-6Auxiliary Building Stairway (North)1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel II Steam Gen. B Wide Range Level Channel II Steam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel IISteam Gen. C Wide Range Level Channel III Steam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. A-06-002, A-06-003, A-06-004, A-06-005, A-06-006, and A-06-007 Fire ZoneDescription1127Auxiliary Building Stairway (North)August 2011 C-37LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-6Auxiliary Building Stairway (North)1RCS Inventory ControlMaintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NoneAugust 2011 C-38 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-6Auxiliary Building Stairway (North)1....3.%654(5An excessive gap in the bottom of Door DSK11271 connecting Fire Areas A-1 and A-6 is acceptable based on the lack of intervenin g combustibles at/near the location of the door. Fire Area A-6 is a stairwell and, as such, no transient combustibles are expected near this doorway. Therefore DSK11271 is considered a non-rated feature commensurate with the fire hazards in the two areas and it provides an equivalent level of protection as a 3 hour rated fire door by prohibiting the propagation of fire between the two fire areas.&Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.August 2011 C-39 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-6Auxiliary Building Stairway (North)18857(Raceways 1J1L01 and 1U1K01 are provided with a Darmatt fire wrap in Fire Zone 1127. Per Callaway CAR 200607577, this fire wrap is improperly installed (i.e., inadequate due to framing issue). The fire rating of this ERFBS is degraded from the intended 3-hour rating of the design criteria to a 1-hour rating. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a degraded barrier issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8574ABPV0004-P - Cable damage (4ABI20HE and 4ABI20HH) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.857BMHV0001 - Cable damage (4BMK06AA and 1BMK06EA) to BMHV0001 (BMHY0001A and BMHY0001C). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-40 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-6Auxiliary Building Stairway (North)1857:BMHV0002 - Cable damage (4BMK06BA) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.857,BMHV0003 - Cable damage (4BMK06CA) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8577BMHV0004 - Cable damage (4BMK06DA and 1BMK06HA) to BMHV0004 (BMHY0004A and BMHY0004C). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805
, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-41 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-6Auxiliary Building Stairway (North)1857;FCHV0312-P - Cable damage (2FCK23AX) to FCHV0312-P. Cable damage cannot spuriously open Turbine Driven AFW Pump Trip and Throttle Valve FCHV0312-P. However, an Auxiliary Feedwater Actuation Signal (AFAS) could open the valve prior to cable failure and the valve could then fail open, which could result in the inability to remotely secure the non-credited Turbine Driven AFW Pump. If running, the non-credited Turbine Driven AFW Pump could become an uncontrolled source of inventory addition into Steam Generators 2(B) and 3(C), which could adversely impact the capability to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generators 2(B) and 3(C) are credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3
. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-42 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-6Auxiliary Building Stairway (North)1&$.>.?>..A&."&.&Ionization 109Detection NNNN1127NN/ANoneSuppression ERFBSNoneFeatureYNNNNSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere is no equipment susceptible to water damage in this area. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-43 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-7Boron Injection Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Alternate 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1126Boron Injection Tank and Pump RoomAugust 2011 C-44LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-7Boron Injection Room1RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A). 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain val id.August 2011 C-45 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-7Boron Injection Room1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-46 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-7Boron Injection Room1&$.>.?>..A&."&.&Ionization 101Detection NNNN1126NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. In the area, water would  drain through the open door to minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-47 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1~"Process MonitoringPressurizer Pressure Channel IIPressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel IISteam Gen. A Pressure Channel II4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. A-08-001, A-08-002, A-08-016, and A-08-017 Fire ZoneDescription1301Corridor No. 11302Filter Compartments1306Valve Compartments1307Corridor No. 21308Valve Compartments1311Auxiliary Building Sampling Room1312Boron Meter & RC Activity Monitor Room1313Volume Control Tank Room1314Corridor No. 31315Containment Spray Additive Tank Area 1316Valve Compartment 1317Seal Water Heat Exchanger Room1318Valve Compartment1319Demineralizer Compartments1320Corridor No. 41321Exit VestibuleAugust 2011 C-48LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1RCS Inventory ControlSteam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IVCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IIContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite Power to NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).See VFDR No. A-08-003, A-08-004, A-08-005, A-08-006, A-08-007, A-08-008, A-08-009, A-08-010, A-08-011, A-08-013, A 014, A-08-015, and A-08-018See VFDR No. A-08-012, A-08-019RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-49 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1%~&',%Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 2, dated 05/1984 based on the following:  1. Elevator and dumbwaiter doors are rated at 1-1/2 hours as required by ANSI A17.1. 
: 2. The 1-1/2 hour doors are an industry standard and, as stated in ANSI A17.1, are acceptable for use in a 2-hour rated elevato r or dumbwaiter shaft. 3. For a fire to propagate from one floor elevation to another, it would have to penetrate two doors.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-50 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1%~&Deviation submitted per 6/29/1981 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, to justify the two sets of non-rated equipment hatchways in the northern and southern ends of the auxiliary building corridors, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Low fuel loading and configuration of equipment.2. Steel hatch covers are provided for each hatchway.3. Automatic sprinkler water curtains are provided for each hatchway at elevations 2000'-0", 2026'-0", and 2047'-0" to separate the corridor fire areas.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zones 1302, 1306, 1307, 1308, 1313, 1318, and 1319), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following: 
: 1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zones 1301, 1312, 1316 and 1317 and no suppression in 1302, 1306, 1307, 1308, 1311, 1313, 1314, 1315, 1318, 1319, 1320 and 1321), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following: 
: 1. Minimal fire hazards.
: 2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-51LIC-25LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1....3.%+574(The non-rated configuration of Penetration 0P14151028 for the resin chute hatch cover in the floor/ceiling boundary of Fire Zones 1405/1319 is acceptable based on low combustible loading and the qualities of construction of the assembly and fire barriers that will limit heat and smoke transfer.&.%+5744The removal of Thermo-Lag fire barriers from RHR and containment spray hatch covers in Fire Zones 1203 and 1204 is acceptable based on the fact that there are no fixed ignition sources or fixed combustibles in the vicinity of the hatches, and the fact that the areas are infrequently accessed due to the fact that they are "locked high radiation areas." Further, gaps between the individual plates between A-1 and A-8 are acceptable based on the fact that heat transferred would be rapidly dissipated in the large volume of A-8; there is a no combustible zone at the hatches; combustible loading in the area is very low; there are no ignition sources in the area of the plates; ventilation is not affected based on gap size; an automatic pre-action suppression system is installed in the area of the hatch on the 2000' elevation; and the fire brigade is trained to aggressively control fires in this area.The removal of Thermo-Lag fire barriers from buttress hatch covers between areas A-8 and A-16 was determined to be acceptable based on the analysis in Calculation #01-0082-05-4087-01,  Addendum 1, which identifies that the steel will maintain its integrity without the need for fireproofing. Further, the minor gaps resulting from removal of the Thermo-Lag will be filled with Dow Corning 9-081 caulk, sealing minor gaps between the hatch covers and the supporting frames and automatic pre-action suppression is available above and below the hatch area.&August 2011 C-52LIC-25LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1.%555,54The detectors in beam pockets in Fire Zone 1314  that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable based on low combustible loading in the area and current NFPA 72-2010 guidance. NFPA 72-2010, Section 17.7.3.2.4.2 states that "for ceilings with beam depths of less than 10% of the ceiling height, smooth ceiling spacing shall be permitted." The corridor 18" deep beam is less than 10% of the ceiling height of 300". The detectors in beam pockets in Fire Zone 1320  that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable due to the presence of a "No Combustible Zone" and the fact that a n equipment hatch is present in the beam pocket which would allow smoke to rise to the floor above.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-53LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area18856(ABPV0001-P - Cable damage (1ABI20EA, 1ABI20EC, and 1ABI20ED) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a varian ce from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.ABPV0003-P - Cable damage (3ABI20GA, 3ABI20GC, 3ABI20GD, and 3ABI20GE) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.856BBPCV0455A-P - Cable damage (1BBK40AG and 1BBK40AK) to BBPCV0455A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-54 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1856:BGHV8105 - Cable damage (4BGG11AA and 4BGG11AB; and 1BGG11BC and 1BGG11BD) to BGHV8105 and BGHV8106 respectively. For Chemical Volume and Control System Charging Header to Regenerative Heat Exchanger Outer Containment Isolation Valve BGHV8105, cable damage cannot spuriously open valve; cable damage can spuriously close valve (spurious closure is only credible assuming external hot shorts). The valve may also close in response to a valid or spurious SIS. In the case of valve closure, cable damage could bypass the open / close limit / torque switches. However, the required position for this valve is closed. The valve may also fail as-is (open). For Chemical Volume and Control System Charging Header to Regenerative Heat Exchanger Outer Containment Isolation Valve, BGHV8106, cable damage can spuriously open or close the valve, but will not bypass the open / close limit / torque switches. The valve may close in response to a valid or spurious SIS. However, the required position for this valve is closed. BGHV8106 may also lose power from NG01B. The valve may also fail as-is (open). Either one of these valves is required closed to prevent potential adverse impact (i.e., flow diversion) to the NFPA 805 NSPC credited flowpaths of the chemical volume and control system (i.e., the boron injection and the alternate RCP seal injection flowpaths). Either one of these valves is required closed to also mitigate spurious operation of downstream valves BGHV8145 (Pressurizer Auxiliary Spray), BGHV8146 (Loop 1 Cold Leg Injection), and BGHV8147 (Loop 4 Cold Leg Injection), which are not included in the safe shutdown model, and which could adversely impact the ability to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.856,BGHV8149A - Cable damage (5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-55 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area18567BGHV8149B - Cable damage (5BGK35BD) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.856;BGHV8149C - Cable damage (5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-56 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area18566BGLCV0112B-P - Cable damage (1BGG12AA, 1BGG12AB, and 1BGG12AD); and loss of power (NG01A) to BGLCV0112B (Chemical Volume Control System Volume Control Tank [VCT] Outlet Upstream Isolation Valve). Cable damage (4BGG12BA and 4BGG12BB) to BGLCV0112C (Chemical Volume Control System Volume Control Tank Outlet Downstream Isolation Valve). Cable damage (1BNG01AA and 1BNG01AB); and loss of power (NG01A) to BNLCV0112D (Charging Pump A Suction from Refueling Water Storage Tank [RWST] Isolation Valve). No cable damage or loss of power to BNLCV0112E (Charging Pump B Suction from Refueling Water Storage Tank Isolation Valve). Cable damage (1BGI51CA) to BGLT0112 (Volume Control Tank Protection A Level Transmitter); loss of Channel I Vital AC instrument power (NN01) for Level Transmitter BGLT0112 (after battery depletion); the instrument sense line for Level Transmitter BGLT0112 is in the area. Cable damage (4BGI51DA) to BGLT0185 (Volume Control Tank Protection B Level Transmitter); the instrument sense line for Level Transmitter BGLT0185 is in the area. Cable damage to BGLCV0112B can spuriously close valve (spurious closure is only credible assuming external hot shorts); and may also cause the valve to fail as-is (open), or to close coincident with a spurious or valid SIS, Boron Dilution, or Low-Low VCT Level signal prior to valve BNLCV0112D reaching the fully open position. Cable damage to BGLCV0112C cannot spuriously close or open valve, but may cause the valve to fail as-is (open). Spurious closure of valve BGLCV0112B due to direct cable damage could isolate the suction flowpath from the VCT to the Credited Charging Pump, B. Cable damage and / or sense line heating to Volume Control Tank level transmitters can spuriously transfer charging pump suction from the VCT to the RWST.
This would not be adverse to safe shutdown. Cable damage and / or sense line heating to Volume Control Tank level transmitters could also result in a failure of transfer of charging pump suction from the VCT to the RWST on actual Low-Low VCT level. Charging Pump B is not normally running, but could start on an SIS signal, which it is conservatively assumed could occur for a fire in fire area A-08. These signals would also cause valve BNLCV0112E, unaffected by cable damage, to open. Direct cable damage to BNLCV0112D could cause the valve to fail as-is (closed). Both valves BGLCV0112B and BGLCV0112C are required open (i.e., to not spuriously close) to prevent failure of the credited charging pump (when running) from a loss of suction. After the charging pump suction flowpath has been aligned to the Refueling Water Storage Tank, either valve BGLCV0112B or BGLCV0112C is required to close on demand in order to isolate the charging pump suction flowpath from the Volume Control Tank to prevent gas binding of the credited charging pump. Either valve BNLCV0112D or BNLCV0112E is required to open on demand in order to align the charging pump suction flowpath to the Refueling Water Storage Tank for the credited charging pump. Note that BGLCV0112B and BGLCV0112C are both physically located in the fire affected area. The valve limit / torque switches for BGLCV0112B can be bypassed by fire damage to cables 1BGG12AB and 1BGG12AD. The valve limit / torque switches for BGLCV0112C can be bypassed by fire damage to cable 4BGG12BB. The valve limit / torque switches for BNLCV0112D can be bypassed by fire damage to cable 1BNG01AB. It does not appear that Train B 4kV Switchgear NB02 could sustain an immediate real or spurious loss of offsite power (LOOP) due to cable damage occurring in the fire area. As such, a LOOP start of the credited Train B charging pump has not been assumed in this fire area (i.e. LOOP start - auto start of the charging pump, but no auto opening of the pump suction valve from the RWST and no auto isolation of pump suction from the VCT). It should be noted that offsite power to 4kV Switchgear NB02 could be lost later in the event following loss of forced cooling for the start-up transformer, XMR01. However, by this time, it is assumed that the valve alignment for the normally idle Train B charging pump will have been performed locally to open the RWST suction valve and close the VCT suction valve prior to manual start of the credited charging pump. On loss of 480VAC motive and / or 120VAC valve control power VCT valve BGLCV0112B August 2011 C-57 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1will fail as-is (open). On loss of 480VAC motive and / or 120VAC valve control power RWST valve BNLCV0112D will fail as-is (closed). On loss of 120VAC vital instrument power VCT level transmitter BGLT0112 will initiate automatic opening of RWST valve BNLCV0112D. After BNLCV0112D reaches approximately 80% open, a valve limit switch permissive from BNLCV0112D will be generated, allowing VCT valve BGLCV0112B to close (coincident with the VCT low level signal generated from loss of power to BGLT0112). Loss of power only to BGLCV0112B, BNLCV0112D, and / or BGLT0112 cannot cause spurious isolation of the suction supply flowpath to the charging pumps, either BGLCV0112B will fail as-is (open) and BNLCV0112D will fail as-is (closed), or BNLCV0112D will open and BGLCV0112B will fail as-is (open) or will close. This addresses th e potential for the credited charging pump to start, and then immediately fail due to spurious isolation of the pump suction flowpath. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-58 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1856+BGLCV0112C - Cable damage (4BGG12BA and 4BGG12BB) to BGLCV0112C (Chemical Volume Control System Volume Control Tank [VCT] Outlet Downstream Isolation Valve). Cable damage (1BGG12AA, 1BGG12AB, and 1BGG12AD); and loss of power (NG01A) to BGLCV0112B (Chemical Volume Control System Volume Control Tank Outlet Upstream Isolation Valve). Cable damage (1BNG01AA and 1BNG01AB); and loss of power (NG01A) to BNLCV0112D (Charging Pump A Suction from Refueling Water Storage Tank [RWST] Isolation Valve). No cable damage or loss of power to BNLCV0112E (Charging Pump B Suction from Refueling Water Storage Tank Isolation Valve). Cable damage (1BGI51CA) to BGLT0112 (Volume Control Tank Protection A Level Transmitter); loss of Channel I Vital AC instrument power (NN01) for Level Transmitter BGLT0112 (after battery depletion); the instrument sense line for Level Transmitter BGLT0112 is in the area. Cable damage (4BGI51DA) to BGLT0185 (Volume Control Tank Protection B Level Transmitter); the instrument sense line for Level Transmitter BGLT0185 is in the area. Direct cable damage to BGLCV0112B can spuriously close valve (spurious closure is only credible assuming external hot shorts); and may also cause the valve to fail as-is (open), or to close coincident with a spurious or valid SIS, Boron Dilution, or Low-Low VCT Level signal prior to valve BNLCV0112D reaching the fully open position. Cable damage to BGLCV0112C cannot spuriously close or open valve, but may cause the valve to fail as-is (open). Spurious closure of valve BGLCV0112B due to direct cable damage could isolate the suction flowpath from the VCT to the Credited Charging Pump B. Cable damage and / or sense line heating to Volume Control Tank level transmitters can spuriously transfer charging pump suction from the VCT to the RWST.
This would not be adverse to safe shutdown. Cable damage and / or sense line heating to Volume Control Tank level transmitters could also result in a failure of transfer of charging pump suction from the VCT to the RWST on actual Low-Low VCT level. Charging Pump B is not normally running, but could start on an SIS signal, which it is conservatively assumed could occur for a fire in fire area A-08. These signals would also cause valve BNLCV0112E, unaffected by cable damage, to open. Direct cable damage to BNLCV0112D could cause the valve to fail as-is (closed). Both valves BGLCV0112B and BGLCV0112C are required open (i.e., to not spuriously close) to prevent failure of the credited charging pump (when running) from a loss of suction. After the charging pump suction flowpath has been aligned to the Refueling Water Storage Tank, either valve BGLCV0112B or BGLCV0112C is required to close on demand in order to isolate the charging pump suction flowpath from the Volume Control Tank to prevent gas binding of the credited charging pump. Either valve BNLCV0112D or BNLCV0112E is required to open on demand in order to align the charging pump suction flowpath to the Refueling Water Storage Tank for the credited charging pump. Note that BGLCV0112B and BGLCV0112C are both physically located in the fire affected area. The valve limit / torque switches for BGLCV0112B can be bypassed by fire damage to cables 1BGG12AB and 1BGG12AD. The valve limit / torque switches for BGLCV0112C can be bypassed by fire damage to cable 4BGG12BB. The valve limit / torque switches for BNLCV0112D can be bypassed by fire damage to cable 1BNG01AB. It does not appear that Train B 4kV Switchgear NB02 could sustain an immediate real or spurious loss of offsite power (LOOP) due to cable damage occurring in the fire area. As such, a LOOP start of the credited Train B charging pump has not been assumed in this fire area (i.e. LOOP start - auto start of the charging pump, but no auto opening of the pump suction valve from the RWST and no auto isolation of pump suction from the VCT). It should be noted that offsite power to 4kV Switchgear NB02 could be lost later in the event following loss of forced cooling for the start-up transformer, XMR01. However, by this time, it is assumed that the valve alignment for the normally idle Train B charging pump will have been performed locally to open the RWST suction valve and close the VCT suction valve prior to manual start of the credited charging pump. On loss of 480VAC motive and / or 120VAC valve control power VCT valve BGLCV0112B August 2011 C-59 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1will fail as-is (open). On loss of 480VAC motive and / or 120VAC valve control power RWST valve BNLCV0112D will fail as-is (closed). On loss of 120VAC vital instrument power VCT level transmitter BGLT0112 will initiate automatic opening of RWST valve BNLCV0112D. After BNLCV0112D reaches approximately 80% open, a valve limit switch permissive from BNLCV0112D will be generated, allowing VCT valve BGLCV0112B to close (coincident with the VCT low level signal generated from loss of power to BGLT0112). Loss of power only to BGLCV0112B, BNLCV0112D, and / or BGLT0112 cannot cause spurious isolation of the suction supply flowpath to the charging pumps, either BGLCV0112B will fail as-is (open) and BNLCV0112D will fail as-is (closed), or BNLCV0112D will open and BGLCV0112B will fail as-is (open) or will close. This addresses th e potential for the credited charging pump to start, and then immediately fail due to spurious isolation of the pump suction flowpath. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-60 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area18565BGLCV0112C-P - cable damage (4BGG12BA and 4BGG12BB) to BGLCV0112C (Chemical Volume Control System Volume Control Tank [VCT] Outlet Downstream Isolation Valve). Cable damage (1BGG12AA, 1BGG12AB, and 1BGG12AD); and loss of power (NG01A) to BGLCV0112B (Chemical Volume Control System Volume Control Tank Outlet Upstream Isolation Valve). Cable damage (1BNG01AA and 1BNG01AB); and loss of power (NG01A) to BNLCV0112D (Charging Pump A Suction from Refueling Water Storage Tank [RWST] Isolation Valve). No cable damage or loss of power to BNLCV0112E (Charging Pump B Suction from Refueling Water Storage Tank Isolation Valve). Cable damage (1BGI51CA) to BGLT0112 (Volume Control Tank Protection A Level Transmitter); loss of Channel I Vital AC instrument power (NN01) for Level Transmitter BGLT0112 (after battery depletion); the instrument sense line for Level Transmitter BGLT0112 is in the area. Cable damage (4BGI51DA) to BGLT0185 (Volume Control Tank Protection B Level Transmitter); the instrument sense line for Level Transmitter BGLT0185 is in the area. Direct cable damage to BGLCV0112B can spuriously close valve (spurious closure is only credible assuming external hot shorts); and may also cause the valve to fail as-is (open), or to close coincident with a spurious or valid SIS, Boron Dilution, or Low-Low VCT Level signal prior to valve BNLCV0112D reaching the fully open position. Cable damage to BGLCV0112C cannot spuriously close or open valve, but may cause the valve to fail as-is (open). Spurious closure of valve BGLCV0112B due to direct cable damage could isolate the suction flowpath from the VCT to the Credited Charging Pump B. Cable damage and / or sense line heating to Volume Control Tank level transmitters can spuriously transfer charging pump suction from the VCT to the RWST.
This would not be adverse to safe shutdown. Cable damage and / or sense line heating to Volume Control Tank level transmitters could also result in a failure of transfer of charging pump suction from the VCT to the RWST on actual Low-Low VCT level. Charging Pump B is not normally running, but could start on an SIS signal, which it is conservatively assumed could occur for a fire in fire area A-08. These signals would also cause valve BNLCV0112E, unaffected by cable damage, to open. Direct cable damage to BNLCV0112D could cause the valve to fail as-is (closed). Both valves BGLCV0112B and BGLCV0112C are required open (i.e., to not spuriously close) to prevent failure of the credited charging pump (when running) from a loss of suction. After the charging pump suction flowpath has been aligned to the Refueling Water Storage Tank, either valve BGLCV0112B or BGLCV0112C is required to close on demand in order to isolate the charging pump suction flowpath from the Volume Control Tank to prevent gas binding of the credited charging pump. Either valve BNLCV0112D or BNLCV0112E is required to open on demand in order to align the charging pump suction flowpath to the Refueling Water Storage Tank for the credited charging pump. Note that BGLCV0112B and BGLCV0112C are both physically located in the fire affected area. The valve limit / torque switches for BGLCV0112B can be bypassed by fire damage to cables 1BGG12AB and 1BGG12AD. The valve limit / torque switches for BGLCV0112C can be bypassed by fire damage to cable 4BGG12BB. The valve limit / torque switches for BNLCV0112D can be bypassed by fire damage to cable 1BNG01AB. It does not appear that Train B 4kV Switchgear NB02 could sustain an immediate real or spurious loss of offsite power (LOOP) due to cable damage occurring in the fire area. As such, a LOOP start of the credited Train B charging pump has not been assumed in this fire area (i.e. LOOP start - auto start of the charging pump, but no auto opening of the pump suction valve from the RWST and no auto isolation of pump suction from the VCT). It should be noted that offsite power to 4kV Switchgear NB02 could be lost later in the event following loss of forced cooling for the start-up transformer, XMR01. However, by this time, it is assumed that the valve alignment for the normally idle Train B charging pump will have been performed locally to open the RWST suction valve and close the VCT suction valve prior to manual start of the credited charging pump. On loss of 480VAC motive and / or 120VAC valve control power VCT valve BGLCV0112B August 2011 C-61 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1will fail as-is (open). On loss of 480VAC motive and / or 120VAC valve control power RWST valve BNLCV0112D will fail as-is (closed). On loss of 120VAC vital instrument power VCT level transmitter BGLT0112 will initiate automatic opening of RWST valve BNLCV0112D. After BNLCV0112D reaches approximately 80% open, a valve limit switch permissive from BNLCV0112D will be generated, allowing VCT valve BGLCV0112B to close (coincident with the VCT low level signal generated from loss of power to BGLT0112). Loss of power only to BGLCV0112B, BNLCV0112D, and / or BGLT0112 cannot cause spurious isolation of the suction supply flowpath to the charging pumps, either BGLCV0112B will fail as-is (open) and BNLCV0112D will fail as-is (closed), or BNLCV0112D will open and BGLCV0112B will fail as-is (open) or will close. This addresses th e potential for the credited charging pump to start, and then immediately fail due to spurious isolation of the pump suction flowpath. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-62 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1856(BNLCV0112E - No cable damage or loss of power to BNLCV0112E (Charging Pump B Suction From Refueling Water Storage Tank [RWST] Isolation Valve). Cable damage (4BGG12BA and 4BGG12BB) to BGLCV0112C (Chemical Volume Control System Volume Control Tank [VCT] Outlet Downstream Isolation Valve). Cable dama ge (1BGG12AA, 1BGG12A B, and 1BGG12AD); and loss of power (NG01A) to BGLCV0112B (Chemical Volume Control System Volume Control Tank Outlet Upstream Isolation Valve). Cable damage (1BNG01AA and 1BNG01AB); and loss of power (NG01A) to BNLCV0112D (Charging Pump A Suction From Refueling Water Storage Tank Isolation Valve). Cable damage (1BGI51CA) to BGLT0112 (Volume Control Tank Protection A Level Transmitter); loss of Channel I Vital AC instrument power (NN01) for Level Transmitter BGLT0112 (after battery depletion); the instrument sense line for Level Transmitter BGLT0112 is in the area. Cable damage (4BGI51DA) to BGLT0185 (Volume Control Tank Protection B Level Transmitter); the instrument sense line for Level Transmitter BGLT0185 is in the area.
Direct cable damage to BGLCV0112B can spuriously close valve (spurious closure is only credible assuming external hot shorts); and may also cause the valve to fail as-is (open), or to close coincident with a spurious or valid SIS, Boron Dilution, or Low-Low VCT Level signal prior to valve BNLCV0112D reaching the fully open position. Cable damage to BGLCV0112C cannot spuriously close or open valve, but may cause the valve to fail as-is (open). Spurious closure of valve BGLCV0112B due to direct cable damage could isolate the suction flowpath from the VCT to the Credited Charging Pump B. Cable damage and / or sense line heating to Volume Control Tank level transmitters can spuriously transfer charging pump suction from the VCT to the RWST.
This would not be adverse to safe shutdown. Cable damage and / or sense line heating to Volume Control Tank level transmitters could also result in a failure of transfer of charging pump suction from the VCT to the RWST on actual Low-Low VCT level. Charging Pump B is not normally running, but could start on an SIS signal, which it is conservatively assumed could occur for a fire in fire area A-08. These signals would also cause valve BNLCV0112E, unaffected by cable damage, to open. Direct cable damage to BNLCV0112D could cause the valve to fail as-is (closed). Both valves BGLCV0112B and BGLCV0112C are required open (i.e., to not spuriously close) to prevent failure of the credited charging pump (when running) from a loss of suction. After the charging pump suction flowpath has been aligned to the Refueling Water Storage Tank, either valve BGLCV0112B or BGLCV0112C is required to close on demand in order to isolate the charging pump suction flowpath from the Volume Control Tank to prevent gas binding of the credited charging pump. Either valve BNLCV0112D or BNLCV0112E is required to open on demand in order to align the charging pump suction flowpath to the Refueling Water Storage Tank for the credited charging pump. Note that BGLCV0112B and BGLCV0112C are both physically located in the fire affected area. The valve limit / torque switches for BGLCV0112B can be bypassed by fire damage to cables 1BGG12AB and 1BGG12AD. The valve limit / torque switches for BGLCV0112C can be bypassed by fire damage to cable 4BGG12BB. The valve limit / torque switches for BNLCV0112D can be bypassed by fire damage to cable 1BNG01AB. It does not appear that Train B 4kV Switchgear NB02 could sustain an immediate real or spurious loss of offsite power (LOOP) due to cable damage occurring in the fire area. As such, a LOOP start of the credited Train B charging pump has not been assumed in this fire area (i.e. LOOP start - auto start of the charging pump, but no auto opening of the pump suction valve from the RWST and no auto isolation of pump suction from the VCT). It should be noted that offsite power to 4kV Switchgear NB02 could be lost later in the event following loss of forced cooling for the start-up transformer, XMR01. However, by this time, it is assumed that the valve alignment for the normally idle Train B charging pump will have been performed locally to open the RWST suction valve and close the VCT suction valve prior to manual start of the credited charging pump. On loss of 480VAC motive and / or 120VAC valve control power VCT valve BGLCV0112B August 2011 C-63 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1will fail as-is (open). On loss of 480VAC motive and / or 120VAC valve control power RWST valve BNLCV0112D will fail as-is (closed). On loss of 120VAC vital instrument power VCT level transmitter BGLT0112 will initiate automatic opening of RWST valve BNLCV0112D. After BNLCV0112D reaches approximately 80% open, a valve limit switch permissive from BNLCV0112D will be generated, allowing VCT valve BGLCV0112B to close (coincident with the VCT low level signal generated from loss of power to BGLT0112). Loss of power only to BGLCV0112B, BNLCV0112D, and / or BGLT0112 cannot cause spurious isolation of the suction supply flowpath to the charging pumps, either BGLCV0112B will fail as-is (open) and BNLCV0112D will fail as-is (closed), or BNLCV0112D will open and BGLCV0112B will fail as-is (open) or will close. This addresses th e potential for the credited charging pump to start, and then immediately fail due to spurious isolation of the pump suction flowpath. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8564EGRV0009 - Cable damage (5EGK03AA and 5EGK03AB) to EGRV0009. Cable damage can spuriously open Component Cooling Water Tank A Surge Tank Vent Valve, EGRV0009 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-64 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1856EMHV8801B - Cable damage (4EMG02DA and 4EMG02DB) to EMHV8801B. Boron Injection Header Train B Outlet to Cold Legs Isolation Valve, EMHV8801B, cannot spuriously open or close due to cable damage (4EMG02DA and 4EMG02DB). However, the valve may also fail as-is (closed). The desired position for this valve is open to establish the Train B boron injection flowpath, which may become necessary to restore pressurizer level (maintain positive control over RCS Inventory and Pressure) i f RCS inventory makeup with alternate RCP seal injection is not sufficient. Note that the valve limit / torque switches for EMHV8801B can be bypassed by fire damage to cable 4EMG02DB. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.856:EMHV8801B-P - Cable damage (4EMG02DA and 4EMG02DB) to EMHV8801B. Boron Injection Header Supply from Boron Injection Header Train B Outlet to Cold Legs Isolation Valve, EMHV8801B, cannot spuriously open or close due to cable damage (4EMG02DA and 4EMG02DB). However, the valve may fail as-is (closed). Cable damage affects both Train B valves in the CVCS boron injection flowpath, EMHV8801B-P and EMHV8803B-P. Valve EMHV8803B-P cannot spuriously open or close due to cable damage (4EMG02BA and 4EMG02BB). However, the valve limit / torque switches may be bypassed due to cable damage. Valve EMHV8801B-P cannot spuriously open or close due to cable damage (4EMG02DA and 4EMG02DB). However, the valve limit / torque switches may be bypassed due to cable damage. Both of these CVCS boron injection flowpath valves could open in response to a spurious or valid SIS. This could be problematic for valves EMHV8801B-P and EMHV8803B-P as the valves could operate without limit / torque switch protection, which may prevent manual local closure of these valves should isolation of the CVCS boron injection flowpath be necessary. This flowpath may need to be isolated in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-65 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1856,EMHV8803B - Cable damage (4EMG02BA and 4EMG02BB) to EMHV8803B. Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B, cannot spuriously open or close due to cable damage (4EMG02BA and 4EMG02BB). However, the valve may fail as-is (closed). The desired position for this valve is throttled open to establish the Train B boron injection flowpath, which may become necessary to restore pressurizer level (maintain positive control over RCS Inventory and Pressure) if RCS inventory makeup with alternate RCP seal injection is not sufficient. Note that the valve limit / torque switches for EMHV8803B can be bypassed by fire damage to cable 4EMG02BB. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8567FCHV0312-P - Cable damage (2FCK23AX) to FCHV0312-P. Cable damage cannot spuriously open Turbine Driven AFW Pump Trip and Throttle Valve FCHV0312-P. However, AFAS could open valve prior to cable failure and valve could then fail open. The non-credited turbine driven AFW pump could become an uncontrolled source of inventory addition into Steam Generators A and D, which could adversely impact the capability to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generators A and D are credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.856;NB0105-P - Cable damage (1ALB01AD, 1ALB01AG, 1ALB01AR, and 1ALB01AS) to NB0105. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A), NB0105. Non-credited train Motor Driven Auxiliary Feedwater Pump A may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-66 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area18566PZR-HTR-BU-A - Backup Pressurizer Heater Groups A and B are subject to cable damage and / or loss of DC control power (cables 5BBG22AA, 5BBG2 2AB, 5BBG22AD, 5BBG22AE, and 5BBG22AF for the Group A Heaters, the breaker PG2101 close / trip control cables - cable 6PKG11BA for the Group B Heaters, the AC power cable to Battery Charger PK22, ultimately causing loss of DC control power for breaker PG2201). Neither backup group of pressurizer heaters is available for safe shutdown in this fire area. Loss of pressurizer heater capability may adversely impact the ability of the plant to maintain safe and stable (potential adverse impact to RCS Pressure Control). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.  (Both backup groups of pressurizer heaters are recoverable with local manual operator actions.)The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.856+SGK05B - No direct cable damage to SGK05B. Loss of battery charger - 125VDC No. 2, PK22, to Switchboard 125VDC Bus, PK02, from cable damage (6PKG11BA) results in the eventual loss of 125VDC power for the fire protection interlock circuit. Los s of 125VDC power for the fire protection interlock will result in a loss of ventilation from the Train B ESF Switchgear Rooms Ai r Conditioning Unit, SGK05B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-67 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1&$.>.?>..A&."&.&Ionization 103Detection NYNN1301NIonization 117Detection NNNNNPre-action SKC44Suppression NYNNNN/ANoneFeatureN/ANoneDetection 1302N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1306N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1307N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1308N/ANoneSuppression N/ANoneFeatureIonization 117Detection NNNN1311NN/ANoneSuppression N/ANoneFeatureIonization 103Detection NYNN1312NPre-action SKC44Suppression NYNNNN/ANoneFeatureAugust 2011 C-68 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1&$.>.?>..A&."&.&
N/ANoneDetection 1313N/ANoneSuppression N/ANoneFeatureIonization 103Detection NYNY1314Yactivates hatch water curtainIonization 117Detection NNNNNPre-action SKC44Suppression NYNYYprovides water curtain for hatch N/ANoneFeatureIonization 103Detection NYNN1315NIonization 117Detection NNNNNN/ANoneSuppression N/ANoneFeatureIonization 103Detection NYNN1316NPre-action SKC44Suppression NYNNNN/ANoneFeatureIonization 103Detection NYNN1317NPre-action SKC44Suppression NYNNNN/ANoneFeatureN/ANoneDetection 1318N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1319N/ANoneSuppression N/ANoneFeatureAugust 2011 C-69 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1&$.>.?>..A&."&.&Ionization 103Detection NYNY1320Nactivates hatch water curtainPre-action SKC44Suppression NYNYNprovides water curtain for hatch N/ANoneFeatureIonization 102Detection NNNN1321NIonization 103Detection NYNNNPre-action SKC44Suppression NYNNNN/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-70 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-8Auxiliary Building - El. 2000, General Area1None&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-71 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-9RHR Heat Exchanger Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1309Residual Heat Removal Heat Exchanger Room BAugust 2011 C-72LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-9RHR Heat Exchanger Room1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zone 1309), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.
None....3August 2011 C-73LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-9RHR Heat Exchanger Room1None8&$.>.?>..A&."&.&
N/ANoneDetection 1309N/ANoneSuppression N/ANoneFeatureSLERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationAugust 2011 C-74 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-9RHR Heat Exchanger Room1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-75 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-10RHR Heat Exchanger Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1310Residual Heat Removal Heat Exchanger Room AAugust 2011 C-76LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-10RHR Heat Exchanger Room1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B. Offsite Power to NB01 and NB02 credited. HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zone 1310), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.
None....3August 2011 C-77LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-10RHR Heat Exchanger Room1None8&$.>.?>..A&."&.&
N/ANoneDetection 1310N/ANoneSuppression N/ANoneFeatureSLERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationAugust 2011 C-78 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-10RHR Heat Exchanger Room1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-79 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-11Cable Chase, Auxiliary Building - El. 20001~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel I Steam Gen. A Narrow Range Level Channel IVSteam Gen. A Atmos. Steam Dump Pressure Channel IAux. Feedwater Flow to Steam Gen. A Channels I and IV RCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and II Condensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.See VFDR No. A-11-001, A-11-002, A-11-003, and A-11-0044.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription1335Cable ChaseAugust 2011 C-80LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-11Cable Chase, Auxiliary Building - El. 20001Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.August 2011 C-81 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-11Cable Chase, Auxiliary Building - El. 2000188(((BBPCV0455A-P - Cable damage (1BBK40AG) to BBPCV0455A; cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8((4BGHV8149A - Cable damage (5BGK35AB and 5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8((BGHV8149B - Cable damage (5BGK35BB and 5BGK35BD) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-82 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-11Cable Chase, Auxiliary Building - El. 200018((:BGHV8149C - Cable damage (5BGK35CB and 5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.&$.>.?>..A&."&.&Ionization 117Detection NNNN1335NWet PipeSKC35Suppression NNYNNN/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-83 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-11Cable Chase, Auxiliary Building - El. 20001None&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-84 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-12Auxiliary Building Cable Chase B, Auxiliary Building - El. 20001~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel I Steam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and II RCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel III Steam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and IICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1336Electrical ChaseAugust 2011 C-85LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-12Auxiliary Building Cable Chase B, Auxiliary Building - El. 20001flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.None8August 2011 C-86 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-12Auxiliary Building Cable Chase B, Auxiliary Building - El. 20001&$.>.?>..A&."&.&Ionization 117Detection NNNN1336NWet PipeSKC36Suppression NNNNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-87 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-13Auxiliary Feedwater Pump Room B1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel I Steam Gen. B Wide Range Level Channel II Aux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel ICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. A-13-001, A-13-002, A-13-003, and A-13-004 Fire ZoneDescription1325Auxiliary Feedwater Pump Room BAugust 2011 C-88LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-13Auxiliary Feedwater Pump Room B1Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A).
: 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain val id.August 2011 C-89 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-13Auxiliary Feedwater Pump Room B1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-90 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-13Auxiliary Feedwater Pump Room B188(ABPV0002-P - Cable damage (2ABI20FE and 2ABI20FH) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.BMHV0002 - Cable damage (4BMK06BA) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve B MHV0002, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separati on issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.BMHV0003 - Cable damage (4BMK06C A) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve B MHV0003, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Gene rator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separati on issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-91 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-13Auxiliary Feedwater Pump Room B18:FCHV0312-P - Cable damage (2FCK23AD, 2FCK23AR, 2FCK23AS, 2FCK23AT, 2FCK23AX, and 2FCK23AZ) to FCHV0312-P. Cable damage can spuriously open Turbine Driven AFW Pump Trip and Throttle Valve FCHV0312-P. If running, the non-credited turbine driven AFW pump could become an uncontrolled source of inventory addition into Steam Generators B and C, which could adversely impact the capability to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generators B and C are credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.&$.>.?>..A&."&.&Ionization 120Detection NNYN1325NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-92 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-13Auxiliary Feedwater Pump Room B1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-93 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-14Auxiliary Feedwater Pump Room A1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel IISteam Gen. A Wide Range Level Channel IAux. Feedwater Flow to Steam Gen. A Channel IV RCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel II Steam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels II and IVCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription1326Auxiliary Feedwater Pump Room AAugust 2011 C-94LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-14Auxiliary Feedwater Pump Room A1flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A).
: 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-95 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-14Auxiliary Feedwater Pump Room A1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-96 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-14Auxiliary Feedwater Pump Room A1&$.>.?>..A&."&.&Ionization 120Detection NNNN1326NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-97 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-15Turbine Driven Auxiliary Feedwater Pump Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel I Steam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel IAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IVCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron 4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. A-15-001, A-15-002, A-15-003, A-15-004, A-15-005, A-15-006, A 007, and A-15-008 Fire ZoneDescription1331Auxiliary Feedwater Pump Room CAugust 2011 C-98LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-15Turbine Driven Auxiliary Feedwater Pump Room1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-99 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-15Turbine Driven Auxiliary Feedwater Pump Room1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A). 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zone 1331), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-100LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-15Turbine Driven Auxiliary Feedwater Pump Room1....3.%+574;The non-rated blowout panel in the wall between Fire Areas A-15 and TB-1 is acceptable based on the fact that the combustible loading on both sides of the panel is not significant, there are no fixed ignition sources in the vicinity of the panel, and the panel is similar to a 3 hour UL listed configuration.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-101 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-15Turbine Driven Auxiliary Feedwater Pump Room188(,(ABHV0017 - Cable damage (4ABK28AB, 4ABK28AH, 1ABK29AC, 1ABK29AH, and 1ABK29AV) to ABHV0017 (ABHV0017V13A, ABHV0017V13B; ABHV0017V13C; ABHV0017V15A; ABHV0017V15B; and ABHV0017V15C) to Steam Generator B Main Steam Isolation Valve ABHV0017. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(,4ABHV0018 - Cable damage (1ABK23AC, 1ABK23AD, 4ABK23FC, and 4ABK23FD) to ABHV0018 (ABHY0018A and ABHY0018B) to Steam Generator B Main Steam Loop 2 ABHV0017 Bypass Isolation Valve ABHV0015. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts).
The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3
. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-102 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-15Turbine Driven Auxiliary Feedwater Pump Room18(,ABHV0020 - Cable damage (1ABK28BB, 1ABK28BH, 4ABK29BC, 4ABK29BH, and 4ABK29BV) to abhv0020 (ABHV0020V13A, ABHV0020V13B; ABHV0020V13C; ABHV0020V15A; ABHV0020V15B; and ABHV0020V15C) to Steam Generator C Main Steam Isolation Valve ABHV0020. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(,:ABHV0021 - Cable damage (1ABK23AC, 1ABK23AD, 4ABK23FC, and 4ABK23FD) to ABHV0021 (ABHY0021A and ABHY0021B) to Steam Generator C Main Steam Loop 3 ABHV0020 Bypass Isolation Valve ABHV0021. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts).
The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3
. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-103 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-15Turbine Driven Auxiliary Feedwater Pump Room18(,,ABPV0002-P - Cable damage (2ABI20FE) to pressure transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(,7BMHV0002 - Cable damage (1BMK06FA and 4BMK06BA) to BMHV0002 (BMHY0002 and BMHY0002C). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(,;BMHV0003 - Cable damage (1BMK06GA and 4BMK06CA) to BMHV0003 (BMHY0003A and BMHY0003C). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805
, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-104 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-15Turbine Driven Auxiliary Feedwater Pump Room18(,6FCHV0312-P - Cable damag e (2FCK23AD, 2FCK23AF, 2FCK23AG, 2FCK23AR, 2FCK23AS, 2FCK23AT, 2FCK23AV, 2FCK23AX, and 2FCK23AZ) to FCHV0312-P. Cable damage can spuriously open turbine driven AFW Pump Trip and Throttle Valve FCHV0312-P. The non-credited turbine driven AFW pump could become an uncontrolled source of inventory addition into Steam Generators 1(A), 2(B), 3(C), and 4(D), which could adversely impact the capability to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generators 1(A) and 4(D) are credited for Decay Heat Removal in this fire area. Note that Steam Generators 2(B) and 3(C) are not credited for Decay Heat Removal in this fire area.
This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issu e.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-105 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-15Turbine Driven Auxiliary Feedwater Pump Room1&$.>.?>..A&."&.&Thermal111Detection NNYN1331NIonization 120Detection NNNNNManual Spray SKC22Suppression NNNNNfor the TDAFP N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. A manually charged fixed water spray system is installed to protect the turbine and pump lubricating oil lines and bearings and manual system precludes any water damage to the turbine due to an inadvertent operation. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-106 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area1~"Process MonitoringNORTH:RCS Pressure Channel I Pressurizer Pressure Channels IPressurizer Level Channel IEx-core Neutron Monitoring Channel IV RCS Loop B (2) T-hot Temperature Channel I RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Narrow Range Level Channel ISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channel IAux. Feedwater Pump A Suction Pressure Channel ISee VFDR No. A-16-NORTH-0184.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBNORTH:Steam Generator C supplied by MDAFW Pump A.SOUTH:Steam Generators B and C are supplied by MDAFW Pump A.See VFDR No. A-16-NORTH-001, A-16-NORTH-002, A-16-NORTH-003, A-16-NORTH-004, A-16-NORTH-005, A-16-NORTH-006, A-16-NORTH-007, A-16-NORTH-014, and A-16-NORTH-015See VFDR No. A-16-SOUTH-002, A-16-SOUTH-003, A-16-SOUTH, 004, and A SOUTH-005 Fire ZoneDescription1401Component Cooling Water Pump & Heat Exchanger Area B1402Corridor No. 11406Component Cooling Water Pump & Heat Exchanger Area A1408Corridor No. 2August 2011 C-107LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area1RCS Inventory ControlRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel ICore Exit Thermocouples Train B (Channel IV and VI)SOUTH:RCS Pressure Channel I Pressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel ISteam Gen. B Pressure Channel ISteam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channel I Containment Pressure Channels I, III, and IVCore Exit Thermocouples Train A (Channel I and V)NORTH:Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. PORV (BBPCV0456A) is available for letdown of RCS inventory, if necessary. SOUTH:Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.See VFDR No. A-16-NORTH-011, A-16-NORTH-012, and A-16-NORTH-013See VFDR No. A-16-SOUTH-001, A-16-SOUTH-006, and A-16-SOUTH-009RCS Pressure ControlNORTH:Control pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.See VFDR No. A-16-NORTH-016August 2011 C-108 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area1Reactivity ControlNORTH:Trip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.SOUTH:Trip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesNORTH:Operate CCW Pumps B and D, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).SOUTH:Operate CCW Pumps A and C, and ESW Pump A.Onsite Power to NB01 credited.
HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. A-16-NORTH-008, A-16-NORTH-009, A-16-NORTH-010, and A-16-NORTH-017See VFDR No. A-16-SOUTH-007, and A-16-SOUTH-008SOUTH:Control pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-109 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area1%~&',%Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 2, dated 05/1984 based on the following:  1. Elevator and dumbwaiter doors are rated at 1-1/2 hours as required by ANSI A17.1. 
: 2. The 1-1/2 hour doors are an industry standard and, as stated in ANSI A17.1, are acceptable for use in a 2-hour rated elevato r or dumbwaiter shaft. 3. For a fire to propagate from one floor elevation to another, it would have to penetrate two doors.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-110 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area1%~&Deviation submitted per 6/29/1981 SNUPPS letter to the NRC, as supplemented by SNUPPS submittals to the NRC dated 2/1/1984 and 2/24/1984, justifying partial suppression with fire stops in intervening cable trays to provide adequate fire separation between redundant component cooling water system pumps, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Configuration of combustibles2. Fire stops installed in each cable tray communicating between rooms containing redundant shutdown components.&This deviation is active per Section A.16.2 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 6/29/1981 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, to justify the two sets of non-rated equipment hatchways in the northern and southern ends of the auxiliary building corridors, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Low fuel loading and configuration of equipment.
: 2. Steel hatch covers are provided for each hatchway.3. Automatic sprinkler water curtains are provided for each hatchway at elevations 2000'-0", 2026'-0", and 2047'-0" to separate the corridor fire areas.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zone 1401 and no suppression in 1402, 1406, and 1408), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.
: 2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-111LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area1....3.%+5744The removal of Thermo-Lag from cable tray fire stops in Fire Area 16 was determined to be acceptable based on the analysis in Calculation #01-0082-05-4087-01,  Addendum 1, which identifies that the existing cable trays and foam seals prevent potentially burning cable in one tray from igniting adjacent trays without the Thermo-Lag installed. Automatic sprinklers in the area provide additional assurance against fire propagation.The removal of Thermo-Lag fire barriers from buttress hatch covers between areas A-16 and A-19 is acceptable based on the analysis in Calculation #01-0082-05-4087-01, Addendum 1, which identifies that the steel will maintain its integrity without the need for fireproofing. Further, the minor gaps resulting from removal of the Thermo-Lag will be filled with Dow Corning 9-081 caulk, sealing minor gaps between the hatch covers and the supporting frames and automatic pre-action suppression is available on the 2026' elevation.The removal of Thermo-Lag fire barriers from buttress hatch covers between areas A-8 and A-16 was determined to be acceptable based on the analysis in Calculation #01-0082-05-4087-01,  Addendum 1, which identifies that the steel will maintain its integrity without the need for fireproofing. Further, the minor gaps resulting from removal of the Thermo-Lag will be filled with Dow Corning 9-081 caulk, sealing minor gaps between the hatch covers and the supporting frames and automatic pre-action suppression is available above and below the hatch area.&.%555,54The detectors in beam pockets in Fire Zone 1408 that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable based on the area being a "No Combustible Zone" and the fact that an equipment hatch is present in the beam pocket which would allow smoke to rise to the floor above. Detector locations in the other "non-equipment hatch" beam pockets are not considered deviations as they are provided for actuation of suppression systems.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-112LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area188(7ABPV0002-P - Cable damage (2ABI20FE, 2ABI20FG, 2ABI20FH, and 2ABI20FK) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.ABPV0003-P - Cable damage (3ABI20GA, 3ABI20GC, 3ABI20GD, and 3ABI20GE) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(7ABPV0004-P - Cable damage (4ABI 20HE, 4ABI20HG, 4AB I20HH, and 4ABI20HK) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-113 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area18(7BMHV0001 - Cable damage (4BMK06AA) to BMHV0001 (BMHY0001A). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator A is not credited for Decay Heat Removal in this fir e area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8(7BMHV0002 - Cable damage (4BMK06BA) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is not credited for Decay Heat Removal in this fir e area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8(7BMHV0003 - Cable damage (4BMK06CA) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-114 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area18(7BMHV0004 - Cable damage (4BMK06DA) to BMHV0004 (BMHY0004A). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator D is not credited for Decay Heat Removal in this fir e area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8(7EGHV0016 - Cable damage (4EGG05BA and 4EGG05BB) to EGHV0016. Cable damage cannot spuriously open or close Component Cooling Water Train B Supply/Return Isolation Valve, EGHV0016. The valve may fail as-is (open or closed). EGHV0016 (if failed as-is, closed) is required open to align Component Cooling Water Train B to the common service header (for cooling of the seal return heat exchanger). Note that EGHV0016 is located adjacent to the fire affected area. The valve limit/torque switches for EGHV0016 can be bypassed by fire damage to cable 4EGG05BB. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(7EGRV0009 - Cable damage (5EGK03AB) to EGRV0009. Cable damage can spuriously open Component Cooling Water Tank A Surge Tank Vent Valve, EGRV0009 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-115 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area18(7EGRV0010 - Cable damage (6EGK03BA) to EGRV0010. Cable damage can spuriously open Component Cooling Water Tank B Surge Tank Vent Valve, EGRV0010 (spurious opening is only credible assuming external hot shorts). This credited train valve is required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(7EMHV8801B - Cable damage (4EMG02DA and 4EMG02DB) to EMHV8801B. Boron Injection Header Train B Outlet to Cold Legs Isolation Valve, EMHV8801B, cannot spuriously open or close due to cable damage (4EMG02DA and 4EMG02DB). However, the valve may fail as-is (closed). The desired position for this valve is open to establish the Train B boron injection flowpath, which may become necessary to restore pressurizer level (maintain positive control over RCS Inventory and Pressure) i f RCS inventory makeup with alternate RCP seal injection is not sufficient. Note that the valve limit/torque switches for EMHV8801B can be bypassed by fire damage to cable 4EMG02DB. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-116 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area18(7EMHV8803B - Cable damage (4EMG02BA and 4EMG02BB) to EMHV8803B. Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B, cannot spuriously open or close due to cable damage (4EMG02BA and 4EMG02BB). However, the valve may fail as-is (closed). The desired position for this valve is throttled open to establish the Train B boron injection flowpath, which may become necessary to restore pressurizer level (maintain positive control over RCS Inventory and Pressure) if RCS inventory makeup with alternate RCP seal injection is not sufficient. Note that the valve limit/torque switches for EMHV8803B can be bypassed by fire damage to cable 4EMG02BB. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(7EMHV8803B-P - Cable damage (4EMG02BA and 4EMG02BB) to EMHV8803B. Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B, cannot spuriously open or close due to cable damage (4EMG02BA and 4EMG02BB). However, the valve may fail as-is (closed). Cable damage affects both Train B valves in the CVCS boron injection flowpath, EMHV8801B-P and EMHV8803B-P. Valve EMHV8803B-P cannot spuriously open or close due to cable damage (4EMG02DA and 4EMG02DB). However, the valve limit/torque switches may be bypassed due to cable damage. Valve EMHV8801B-P cannot spuriously open or close due to cable damag e (4EMG02DA and 4EMG02DB). However, the valve limit/torque switches may be bypassed due to cable damage. Both of these CVCS boron injection flowpath valves could open in response to a spurious or valid SIS. This could be problematic for valves EMHV8801B-P and EMHV8803B-P as the valves could operate without limit/torque switch protection, which may prevent manual local closure of these valves should isolation of the CVCS boron injection flowpath be necessary. This flowpath may need to be isolated in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-117 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area18(7FCHV0312-P - Cable damage (2FCK23A D, 2FCK23AR, 2FCK23AS, 2FCK23AT, 2FCK23AU, 2FCK23 AZ, 2RPK09AA, and 2RPK15CA) to FCHV0312-P. Cable damage can spuriously open Turbine Driven AFW Pump Trip and Throttle Valve FCHV0312-P. The non-credited turbine driven AFW pump could become an uncontrolled source of inventory addition into Steam Generators B and C, which could adversely impact the capability to maintain positive control over the rate of RCS cooldown, and to maintai n RCS sub-cooling. Note that Steam Generators B and C are credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(7NB0205-P - Cable damage (4ALB01B1, 4ALB01BD, 4ALB01BM, 4ALB01BN, 4ALB01BP, 4ALB01BR, and 4RPK15AA) to NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(7PZR-HTR-BU-A - Backup Pressurizer Heater Groups A and B are subject to cable damage and/or loss of 480V power (cable 5PGG05AA for the Group A Heaters, the Transformer XPG21 4kV power cable - cables 6BBG24AA, 6BBG24AD, 6BBG24AE, 6BBG24AG, 6BBG24AH, 6BBG24AJ, 6BBG24AM, 6BBG24AN, and 6RPK15AA for the Group B Heaters, the breaker PG2201 close/trip control cables). Neither backup group of pressurizer heaters is available for safe shutdown in this fire area (Backup Pressurizer Heater Group B can be recovered with a local operator manual action). Loss of pressurizer heater capability may adversely impact the ability of the plant to maintain safe and stable (potential adverse impact to RCS Pressure Control). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.  (One backup group of pressurizer heaters is recoverable with local manual operator actions.)The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-118 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area18(7SGK05B - Cable damage (4GKG13BE, 4GKG13BF, 6GKK31DA, and 6RPK09PA) to SGK05B. Cable damage may result in a loss of ventilation from the Train B ESF Switchgear Rooms Air Conditioning Unit, SGK05B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(7ABPV0003 - Cable damage (3ABI20GA, 3AB I20GC, 3ABI20GD, and 3ABI20GE) to pressure transmitter ABPT0003. Cable damage can spuriously open the Atmosphere Steam Dump Valve, ABPV0003, or cause the valve to fail closed. The valve is required open to establish a controlled cooldown of the RCS to maintain safe and stable plant conditions. Note that Steam Generator C(3) is credited for decay heat removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.BBPCV0456A-P - Cable damage (4BBK40BG and 4BBK40BK) to BBPCV0456A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0456A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-119 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area18(7&@BMHV0001 - Cable damage (4BMK06AC) to BMHV0001 (BMHY0001A). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator A is not credited for Decay Heat Removal in this fir e area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8(7&@BMHV0002 - Cable damage (4BMK06BC) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8(7&@BMHV0003 - Cable damage (4BMK06CC) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-120 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area18(7&@BMHV0004 - Cable damage (4BMK06DC) to BMHV0004 (BMHY0004A). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator D is not credited for Decay Heat Removal in this fir e area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8(7&@EGHV0053 - Cable damage (1EGG05CA and 1EGG05CB) to EGHV0053. Cable damage cannot spuriously open or close Component Cooling Water Train A Supply Isolation Valve, EGHV0053. The valve may fail as-is (open or closed). EGHV0053 (if failed as-is, closed) is required open to align Component Cooling Water Train A to the common service header (for cooling of the seal return heat exchanger). Note that EGHV0053 is physically located in the fire affected area. The valve limit/torque switches for EGHV0053 can be bypassed by fire damage to cable 1EGG05CB. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(7&@EGRV0009 - Cable damage (5EGK03AA) to EGRV0009. Cable damage can spuriously open Component Cooling Water Tank A Surge Tank Vent Valve, EGRV0009 (spurious opening is only credible assuming external hot shorts). This credited train valve is required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-121 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area18(7&@EGRV0010 - Cable damage (6EGK03BA and 6EGK03BB) to EGRV0010. Cable damage can spuriously open Component Cooling Water Tank B Surge Tank Vent Valve, EGRV0010 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(7&@EMHV8843 - Cable damage (4EMK04CA) to EMHV8843. Cable damage can spuriously open Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843 (spurious opening is only credible assuming external hot shorts). The valve is required closed in order to prevent potential diversion of RWST inventory to the SI test lines to ensure positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3.
This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-122 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area1&$.>.?>..A&."&.&Ionization 118Detection NNYN1401NN/ANoneSuppression N/ANoneFeatureIonization 104Detection NYYY1402NPre-action SKC45Suppression NYNYNN/ANoneFeatureIonization 118Detection NNYN1406NN/ANoneSuppression N/ANoneFeatureIonization 102Detection NNNN1408NIonization 104Detection NYYYYactivates hatch water curtainIonization 118Detection NNYNNPre-action SKC45Suppression NYYYYprovides water curtain for hatches20-ft Separation ZoneNoneFeatureNNNYNERFBSNoneFeatureNNYYNcable tray covers w/seals SLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-123 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-16Auxiliary Building El. 2026, General Area1To meet deterministic separation criteria Fire Area A-16,  Fire Zone 1408 is divided into two safe shutdown analysis areas A-16N and A-16S which are separated by a 20-foot separation zone. The 20-foot separation zone is clearly marked on the floor and designated as a "No Storage" location. Intervening cable trays within the 20-foot separation zone are provided with top and bottom tray covers to act as fire stops.&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. In addition, the 6-inch-high curb installed around Fire Zones 1401 and 1406 will protect the safe shutdown equipment against damage from water discharged by the sprinkler system in Fire Zones 1402 and 1408. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-124LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-17Electrical Penetration Room B1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IIISteam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and II RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel III Steam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump B Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and II Condensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IV Core Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.See VFDR No. A-17-001 and A-17-0024.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1409Electrical Penetration Room BAugust 2011 C-125LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-17Electrical Penetration Room B1Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 credited. Offsite and Onsite Power to NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated electrical penetrations in the reactor containment walls to Fire Areas A-17 and A-18 NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-126 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-17Electrical Penetration Room B1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&8BBPCV0456A-P - Cable damage (4BBK40BG) to BBPCV0456A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0456A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(;4EMHV8843 - Cable damage (4EMK04CA and 4EMK04CD) to EMHV8843. Cable damage can spuriously open Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843 (spurious opening is only credible assuming external hot shorts). The valve is required closed in order to prevent potential diversion of RWST inventory to the SI test lines to ensure positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-127 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-17Electrical Penetration Room B1&$.>.?>..A&."&.&Ionization 106Detection NNYN1409NIonization 113Detection NNYNNHalonSKC04Suppression NNYNNWet PipeSKC36Suppression NNNNNelectrical chase area only N/ANoneFeatureSLERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-128 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-17Electrical Penetration Room B1None&Halon system actuations are not expected to adversely affect electrical equipment. The effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-129 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-18Electrical Penetration Room A1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel I Steam Gen. A Narrow Range Level Channel IVSteam Gen. A Atmos. Steam Dump Pressure Channel IAux. Feedwater Flow to Steam Gen. A Channels I and IV RCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and IV Condensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.See VFDR No. A-18-003, A-18-004, A-18-005, and A-18-0064.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. A-18-001 and A-18-002 Fire ZoneDescription1410Electrical Penetration Room AAugust 2011 C-130LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-18Electrical Penetration Room A1Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated electrical penetrations in the reactor containment walls to Fire Areas A-17 and A-18 NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-131 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-18Electrical Penetration Room A1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-132 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-18Electrical Penetration Room A188(6(Raceways 1J1097, 1J3A1H, and 4J3C1C are provided with a Darmatt 1-hour rated fire wrap in Fire Zone 1410. Per Callaway CAR 200607577, the 1-hour fire wrap for conduit 1J1097 is degraded (i.e., notched). The fire rating of this ERFBS is degraded from the intended 1-hour rating of the design criteria. Conduit 1J1097 contains one safe shutdown cable, 1ABI20EE, which is the instrument signal cable for ABPT0001. Failure of this cable could cause spurious opening of Steam Generator A Atmospheric Steam Dump Valve ABPV0001. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a degraded barrier issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(64ABPV0002-P - Cable damage (2ABI20FE and 2ABI20FH) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.BBPCV0455A-P - Cable damage (1BBK40AG) to BBPCV0455A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-133 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-18Electrical Penetration Room A18(6:BGHV8149A - Cable damage (5BGK35AB) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(6,BGHV8149B - Cable damage (5BGK35BB) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.BGHV8149C - Cable damage (5BGK35CB) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-134 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-18Electrical Penetration Room A1&$.>.?>..A&."&.&Ionization 107Detection NNYY1410NIonization 114Detection NNYYNHalonSKC05Suppression NNYYNWet PipeSKC35Suppression NNNYNelectrical chase area only ERFBSNoneFeatureNNYYNSLERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-135 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-18Electrical Penetration Room A1None&Halon system actuations are not expected to adversely affect electrical equipment. The effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Should manual fire fighting be required, water damage to the electrical equipment in this area could result (with or without associated fire damage); however, the water damage would not adversely affect safe shutdown. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. Safety related electrical cable in tray is qualified for water exposure. The redundant equipment is located in another fire area. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-136 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-19Auxiliary Building El. 2047, General Area1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel IAux. Feedwater Flow to Steam Gen. A Channels I and IVSteam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVVolume Control Tank Level Channels I and IV Containment Pressure Channels I, II, and IIICore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron 4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. A-19-001 Fire ZoneDescription1504Containment Purge Exhaust and Mechanical Equipment Room B1506Containment Purge Supply Air Handling Unit Room No. A1513Control Bldg. Vent Supply A/C Unit RoomAugust 2011 C-137LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-19Auxiliary Building El. 2047, General Area1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B. Offsite Power to NB01 and NB02 credited. HVAC credited for Main Control Room (Train B credited), and Containment (Train A credited).See VFDR No. A-19-002, A-19-003, A-19-004, and A-19-005RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-138 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-19Auxiliary Building El. 2047, General Area1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.
: 2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 6/29/1981 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, to justify the two sets of non-rated equipment hatchways in the northern and southern ends of the auxiliary building corridors, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Low fuel loading and configuration of equipment.2. Steel hatch covers are provided for each hatchway.3. Automatic sprinkler water curtains are provided for each hatchway at elevations 2000'-0", 2026'-0", and 2047'-0" to separate the corridor fire areas.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-139 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-19Auxiliary Building El. 2047, General Area1%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated mechanical penetrations (process and sampling lines and containment purge penetration) in the reactor containment walls to Fire Areas A-19, A-20, A-23, A-24 and A-25, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.....3.%+5744The removal of Thermo-Lag fire barriers from buttress hatch covers between areas A-16 and A-19 is acceptable based on the analysis in Calculation #01-0082-05-4087-01, Addendum 1, which identifies that the steel will maintain its integrity without the need for fireproofing. Further, the minor gaps resulting from removal of the Thermo-Lag will be filled with Dow Corning 9-081 caulk, sealing minor gaps between the hatch covers and the supporting frames and automatic pre-action suppression is available on the 2026' elevation.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-140 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-19Auxiliary Building El. 2047, General Area188(+(ABPV0001-P - Cable damage (1ABI20EE) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.EFHV0052 - Cable damage (4EFG05BA and 4EFG05BB) to Essential Service Water Train B to Component Cooling Water Heat Exchanger Train B Valve, EFHV0052. Cable damage cannot spuriously open or close valve EFHV0052; the valve may fail as-is (open or closed), the valve auto opens on SIS and/or LOOP; the valve also could open (if closed) on a spurious or valid SIS or LOOP with the limit/torque switches bypassed due to cable damage; this would be acceptable as the valve is re quired to be open.
The valve is required open to maintain adequate Essential Service Water flow to the Train B Component Cooling Water Heat Exchanger. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-141 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-19Auxiliary Building El. 2047, General Area18(+EFHV0060 - Cable damage (4EFG04BA and 4EFG04BB) to Essential Service Water Train B from Component Cooling Water Heat Exchanger Train B Valve, EFHV0060. Cable damage cannot spuriously open or close valve EFHV0060; the valve may fail as-is (open or closed), the valve auto closes on SIS and/or LOOP; the valve also could close (if open) on a spurious or valid SIS or LOOP with the limit/torque switches bypassed due to cable damage; this would be acceptable as the valve is required to be closed. The valve is required closed to maintain adequate Essential Service Water flow from the Train B Component Cooling Water Heat Exchanger. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(+:EGHV0011 - Cable damage (1EGG04AC and 1EGG04BC) to EGHV0011 and EGHV0013 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train A Upstream Valve, EGHV0011, and Essential Service Water to Component Cooling Train A Downstream Valve, EGHV0013. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from loss of Component Cooling Water inventory into the Train A Essential Service Water header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-142 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-19Auxiliary Building El. 2047, General Area18(+,EGRV0009 - Cable damage (5EGK03AA) to EGRV0009. Cable damage can spuriously open Component Cooling Water Tank A Surge Tank Vent Valve, EGRV0009 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-143 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-19Auxiliary Building El. 2047, General Area1&$.>.?>..A&."&.&Ionization 108Detection NNNN1504NN/ANoneSuppression N/ANoneFeatureIonization 109Detection NNNN1506NN/ANoneSuppression N/ANoneFeatureIonization 109Detection NNNN1513NN/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-144 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-19Auxiliary Building El. 2047, General Area1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-145 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-20Personnel Hatch and CCW Surge Tank Area1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel IISteam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel II Steam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IV4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1502Component Cooling Water Surge Tank No. B1503Component Cooling Water Surge Tank No. A1505Corridor1507Personnel Hatch AreaAugust 2011 C-146LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-20Personnel Hatch and CCW Surge Tank Area1RCS Inventory ControlVolume Control Tank Level Channels I and IVContainment Pressure Channels I, II, and IIICore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-147 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-20Personnel Hatch and CCW Surge Tank Area1%~&',%Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 2, dated 05/1984 based on the following:  1. Elevator and dumbwaiter doors are rated at 1-1/2 hours as required by ANSI A17.1. 
: 2. The 1-1/2 hour doors are an industry standard and, as stated in ANSI A17.1, are acceptable for use in a 2-hour rated elevato r or dumbwaiter shaft. 3. For a fire to propagate from one floor elevation to another, it would have to penetrate two doors.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 6/29/1981 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, to justify the two sets of non-rated equipment hatchways in the northern and southern ends of the auxiliary building corridors, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Low fuel loading and configuration of equipment.2. Steel hatch covers are provided for each hatchway.3. Automatic sprinkler water curtains are provided for each hatchway at elevations 2000'-0", 2026'-0", and 2047'-0" to separate the corridor fire areas.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated mechanical penetrations (process and sampling lines and containment purge penetration) in the reactor containment walls to Fire Areas A-19, A-20, A-23, A-24 and A-25, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.
: 3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-148 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-20Personnel Hatch and CCW Surge Tank Area1%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated personnel hatch connecting reactor containment and Fire Area A-20, and the hatchways to YD-1, was approved by the NRC in NUREG-0830, Supplement 3, dated
 
05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid. Although the personnel hatch to Fire Area A-20 was approved, the containment emergency personnel and equipment hatchways to the yard, Fire Area YD-1, were not specifically called out in the SER. The emergency personnel hatchway is of identical construction to the personnel hatch to Fire Area A-20. The equipment hatch, while not identical, is equally robustly constructed, consisting of a welded steel assembly with a double gasketed, flanged, and bolted cover and provided with a moveable missile shield on the outside of the Reactor Building. Therefore, clarification regarding the approval of all containment hatchways is required. The clarification is being requested in the License Amendment Request Transition Report, Attachment T.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zones 1502, 1503, and 1601), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-149LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-20Personnel Hatch and CCW Surge Tank Area1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-150 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-20Personnel Hatch and CCW Surge Tank Area1&$.>.?>..A&."&.&Ionization 102Detection NNNN1502NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1503N/ANoneSuppression N/ANoneFeatureIonization 108Detection NNNN1505NN/ANoneSuppression N/ANoneFeatureIonization 108Detection NNNN1507NN/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-151 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-20Personnel Hatch and CCW Surge Tank Area1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-152 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-21Control Room AC and Filtration Unit B1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel I Pressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel III Aux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Refueling Water Storage Tank Level Channel IVolume Control Tank Level Channel IContainment Pressure Channels I, II, and IIICore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. A-21-001, A-21-002, A-21-003, A-21-004, A-21-005, A-21-006 and A-21-007Fire ZoneDescription1501Control Room A/C and Filtration Units Room BAugust 2011 C-153LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-21Control Room AC and Filtration Unit B1Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A. Onsite Power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. A-21-006(BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%555,54The detectors in beam pockets in Fire Zone 1501 that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable based on low combustible loading in the area. Large ductwork for the Control Room air conditioning system  is contained below the beam pocket and there is no space available for combustibles. In addition, high air flow in the area is anticipated to spread out smoke to adjacent beam pocket spaces containing detectors.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-154LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-21Control Room AC and Filtration Unit B1884((ABPV0004-P - Cable damage (4ABI 20HE, 4ABI20HG, 4AB I20HH, and 4ABI20HK) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.BMHV0001 - Cable damage (4BMK06AA) to BMHV0001 (BMHY0001A). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve B MHV0001, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator A is not credited for Decay Heat Removal in this fir e area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separati on issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.BMHV0002 - Cable damage (4BMK06BA) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve B MHV0002, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separati on issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-155 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-21Control Room AC and Filtration Unit B184(:BMHV0003 - Cable damage (4BMK06CA) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84(,BMHV0004 - Cable damage (4BMK06DA) to BMHV0004 (BMHY0004A). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator D is not credited for Decay Heat Removal in this fir e area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84(7EGHV0012 - Cable damage (4EGG04CC and 4EGG04DC) to EGHV0012 and EGHV0014 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train B Upstream Valve, EGHV0012, and Essential Service Water to Component Cooling Train B Downstream Valve, EGHV0014. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from loss of Component Cooling Water inventory into the Train B Essential Service Water header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-156 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-21Control Room AC and Filtration Unit B184(;NB0205-P - Cable damage (4ALB01BD, 4ALB01BM, 4ALB01BN, 4RPK15AA, 4ALB01B1, 4ALB01BP, and 4ALB01BR) to NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.&$.>.?>..A&."&.&Ionization 110Detection NNYN1501NN/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-157 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-21Control Room AC and Filtration Unit B1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-158 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-22Control Room AC and Filtration Unit A1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel IAux. Feedwater Flow to Steam Gen. A Channels I and IV RCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and IICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channels II Volume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent 4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. A-22-001 Fire ZoneDescription1512Control Room A/C and Filtration Units Room AAugust 2011 C-159LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-22Control Room AC and Filtration Unit A1flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).See VFDR No. A-22-002RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-160 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-22Control Room AC and Filtration Unit A18844(ABPV0001-P - Cable damage (1ABI20EE) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.EGHV0011 - Cable damage (1EGG04AC and 1EGG04BC) to EGHV0011 and EGHV0013 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train A Upstream Valve, EGHV0011, and Essential Service Water to Component Cooling Train A Downstream Valve, EGHV0013. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from loss of Component Cooling Water inventory into the Train A Essential Service Water header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-161LIC-24 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-22Control Room AC and Filtration Unit A1&$.>.?>..A&."&.&Ionization 110Detection NNNN1512NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-162 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel I Ex-core Neutron Monitoring Channel I RCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel IISteam Gen. B Wide Range Level Channel II Steam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and V RCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and II4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. A-23-001, A-23-002, A-23-003, A-23-004, A-23-005, A-23-006, A 007, A-23-008, A-23-009, A-23-010, A-23-011, A-23-012, A-23-013, A-23-014, A-23-015, and A-23-016 Fire ZoneDescription1411Main Feedwater Room No. 11412Main Feedwater Room No. 21508Main Steam Isolation Valve Room No. 11509Main Steam Isolation Valve Room No. 2August 2011 C-163LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment1RCS Inventory ControlCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-164 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment1%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated mechanical penetrations (process and sampling lines and containment purge penetration) in the reactor containment walls to Fire Areas A-19, A-20, A-23, A-24 and A-25, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.
: 3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.....3.%+574,The presence of non-rated missile resistant shields, non-rated access hatch, open steel drain piping, and air gaps caused by main steam line penetrations in the barrier between A-23 and TB-1 is acceptable based on the structural integrity of the barrier; the structural integrity of the unexposed steel; the dissipation of heat in the large volumes of Fire Areas A-23 and TB-1; limited quantities of combustibles in the vicinity of the non-rated features; low combustible loading in the areas; limited ignition sources; and automatic suppression is installed on several elevations of the Turbine Building .&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-165 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment188(ABHV0011 - Cable damage (4ABK28BB, 4ABK28BH, 1ABK29BC, 1ABK29BH, and 1ABK29BV) to ABHV0011 (ABHV0011V13A, ABHV0011V13B; ABHV0011V13C; ABHV0011V15A; ABHV0011V15B; and ABHV0011V15C) to Steam Generator D Main Steam Isolation Valve ABHV0011. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84ABHV0012 - Cable damage (1ABK23AA, 1ABK23AB, 4ABK23FA, and 4ABK23FB) to ABHV0012 (ABHY0012A and ABHY0012B) to Steam Generator D Main Steam Loop 4 ABHV0011 Bypass Isolation Valve ABHV0012. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts).
The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3
. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-166 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment18ABHV0014 - Cable damage (4ABK29AC, 4ABK29AH, 4ABK29AV, 1ABK28AB, and 1ABK28AH) to ABHV0014 (ABHV0014V13A, ABHV0014V13B; ABHV0014V13C; ABHV0014V15A; ABHV0014V15B; and ABHV0014V15C) to Steam Generator A Main Steam Isolation Valve ABHV0014. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8:ABHV0015 - Cable damage (1ABK23AA, 1ABK23AB, 4ABK23FA, and 4ABK23FB) to ABHV0015 (ABHY0015A and ABHY0015B) to Steam Generator A Main Steam Loop 1 ABHV0014 Bypass Isolation Valve ABHV0015. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts).
The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3
. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-167 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment18,ABHV0017 - Cable damage (4ABK28AB, 4ABK28AH, 1ABK29AC, 1ABK29AH, and 1ABK29AV) to ABHV0017 (ABHV0017V13A, ABHV0017V13B; ABHV0017V13C; ABHV0017V15A; ABHV0017V15B; and ABHV0017V15C) to Steam Generator B Main Steam Isolation Valve ABHV0017. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.87ABHV0018 - Cable damage (1ABK23AC, 1ABK23AD, 4ABK23FC, and 4ABK23FD) to ABHV0018 (ABHY0018A and ABHY0018B) to Steam Generator B Main Steam Loop 2 ABHV0017 Bypass Isolation Valve ABHV0015. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts).
The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-168 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment18;ABHV0020 - Cable damage (4ABK29BC, 4ABK29BH, 4ABK29BV, 1ABK28BB, and 1ABK28BH) to ABHV0020 (ABHV0020V13A, ABHV0020V13B; ABHV0020V13C; ABHV0020V15A; ABHV0020V15B; and ABHV0020V15C) to Steam Generator C Main Steam Isolation Valve ABHV0020. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.86ABHV0021 - Cable damage (1ABK23AC, 1ABK23AD, 4ABK23FC, and 4ABK23FD) to ABHV0021 (ABHY0021A and ABHY0021B) to Steam Generator C Main Steam Loop 3 ABHV0020 Bypass Isolation Valve ABHV0021. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts).
The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-169 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment18+ABPV0001-P - Cable damage (1ABI20EE) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85ABPV0002-P - Cable damage (2ABI20FE) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(ABPV0003-P - Cable damage (3ABI20GE) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-170 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment184ABPV0004-P - Cable damage (4ABI20HE) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8BMHV0001 - Cable damage (4B MK06AA, 1BMK06EA, and 5BMK06AA) to BMHV0001 (BMHY0001A and BMHY00 01C). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8:BMHV0002 - Cable damage (1BMK06FA, 4BMK06BA, and 5BMK06BA) to BMHV0002 (BMHY0002A and BMHY0002C). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is credite d for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805
, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-171 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment18,BMHV0003 - Cable damage (4BMK06CA, 1BMK06GA, and 5BMK06CA) to BMHV0003 (BMHY0003A and BMHY0003C). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.87BMHV0004 - Cable damage (4BMK06DA, 1BMK06HA, and 5BMK06DA) to BMHV0004 (BMHY0004A and BMHY0004C). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-172 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment1&$.>.?>..A&."&.&
N/ANoneDetection 1411N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1412N/ANoneSuppression N/ANoneFeatureFlame115Detection NNNN1508NN/ANoneSuppression N/ANoneFeatureFlame115Detection NNNN1509NN/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-173 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-23Main Steam and Feedwater Valve Compartment1None&There are no automatic fire suppression systems in the fire area. The safe shutdown instrumentation in this area has watertight enclosures. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-174 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-24Containment Mechanical Piping Penetration Room A1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel I Aux. Feedwater Flow to Steam Gen. A Channels I and IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VI Steam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IVCondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron See VFDR No. A-24-003 and A-24-0044.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. A-24-001 and A-24-002 Fire ZoneDescription1323Pipe Penetration Room AAugust 2011 C-175LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-24Containment Mechanical Piping Penetration Room A1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B. Offsite Power to NB01 and NB02 credited. HVAC credited for Main Control Room and Containment (Train B credited).RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated mechanical penetrations (process and sampling lines and containment purge penetration) in the reactor containment walls to Fire Areas A-19, A-20, A-23, A-24 and A-25, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-176 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-24Containment Mechanical Piping Penetration Room A1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-177 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-24Containment Mechanical Piping Penetration Room A1884:(ABPV0002-P - Cable damage (2ABI20FE and 2ABI20FH) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.ABPV0003-P - Cable damage (3ABI20GA and 3ABI20GE) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-178 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-24Containment Mechanical Piping Penetration Room A184:BGHV8105 - Cable damage (4BGG11AA and 4BGG11AB; and 1BGG11BA and 1BGG11BB) to BGHV8105 and BGHV8106 respectively. Cable damage cannot spuriously open or close Chemical Volume and Control System Charging Header to Regenerative Heat Exchanger Outer Containment Isolation Valves, BGHV8105 and BGHV8106. However, the valves may close in response to a valid or spurious SIS. In this case, cable damage could bypass the open/close limit/torque switches. However, the required position for these valves is closed. The valves may also fail as-is (open). Either one of these valves is require d closed to prevent potential adverse impact (i.e., flow diversion) to the NFPA 805 NSPC credited flowpaths of the chemical volum e and control system (i.e., the boron injection and the alternate RCP seal injection flowpaths). Either one of these valves is required closed to also mitigate spurious operation of downstream valves BGHV8145 (Pressurizer Auxiliary Spray), BG HV8146 (Loop 1 Cold Leg Injection), and BGHV8147 (Loop 4 Cold Leg Injection), which are not included in the safe shutdown model, and which could adversely impact the ability to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84::EMHV8801B - Cable damage (4EMG02DA and 4EMG02DB) to EMHV8801B. Boron Injection Header Train B Outlet to Cold Legs Isolation Valve, EMHV8801B, cannot spuriously open or close due to cable damage (4EMG02DA and 4EMG02DB). However, the valve may also fail as-is (closed). The desired position for this valve is open to establish the Train B boron injection flowpath, which may become necessary to restore pressurizer level (maintain positive control over RCS Inventory and Pressure) if RCS inventory makeup with alternate RCP seal injection is not sufficient. Note that EMHV8801B is physically located in the fire affected area. The valve limit/torque switches for EMHV8801B can be bypassed by fire damage to cable 4EMG02DB. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-179 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-24Containment Mechanical Piping Penetration Room A1&$.>.?>..A&."&.&Ionization 117Detection NNYY1323NN/ANoneSuppression ERFBSNoneFeatureNNYYNSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-180 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-25Pipe Penetration Room B1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels I, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1322Pipe Penetration Room BAugust 2011 C-181LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-25Pipe Penetration Room B1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated mechanical penetrations (process and sampling lines and containment purge penetration) in the reactor containment walls to Fire Areas A-19, A-20, A-23, A-24 and A-25, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-182 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-25Pipe Penetration Room B1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-183 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-25Pipe Penetration Room B1&$.>.?>..A&."&.&Ionization 117Detection NNNN1322NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-184 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-26Ops Storage/I&C Hot Shop1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1405Ops Storage/I&C Hot ShopAugust 2011 C-185LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-26Ops Storage/I&C Hot Shop1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NoneAugust 2011 C-186 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-26Ops Storage/I&C Hot Shop1....3.%+574(The non-rated configuration of Penetration 0P14151028 for the resin chute hatch cover in the floor/ceiling boundary of Fire Zones 1405/1319 is acceptable based on low combustible loading and the qualities of construction of the assembly and fire barriers that will limit heat and smoke transfer.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-187LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-26Ops Storage/I&C Hot Shop1&$.>.?>..A&."&.&Ionization 118Detection NNNN1405NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. There is no safe shutdown equipment susceptible to water damage in this area. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-188 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-27Reactor Trip Switchgear Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IVRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel I Steam Gen. C Narrow Range Level Channel I Steam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channel IAux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel ICore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. PORV (BBPCV0455A) is available for the letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite Power to NB01 credited.
HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. A-27-013See VFDR No. A-27-003, A-27-009, and A-27-010See VFDR No. A-27-011See VFDR No. A-27-008 and A-27-0124.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlUse PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generator C supplied by MDAFW Pump A.See VFDR No. A-27-001, A-27-002, A-27-004, A-27-005, A-27-006, and A-27-007 Fire ZoneDescription1403Load Center and MG Sets RoomAugust 2011 C-189LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-27Reactor Trip Switchgear Room1Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-190 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-27Reactor Trip Switchgear Room1....3.%555,54The detectors in beam pockets in Fire Zone 1403 that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable based on low combustible loading in the area and current NFPA 72-2010 guidance. NFPA 72-2010, Section 17.7.3.2.4.2 states that "for ceilings with beam depths of less than 10% of the ceiling height, smooth ceiling spacing shall be permitted." The corridor 18" deep beam is less than 10% of the ceiling height of 240".&Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.August 2011 C-191LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-27Reactor Trip Switchgear Room1884;(ABPV0002-P - Cable damage (2ABI20FE, 2ABI20FG, 2ABI20FH, and 2ABI20FK) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.ABPV0003-P - Cable damage (3ABI20GA, 3ABI20GC, 3ABI20GD, and 3ABI20GE) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84;BBPCV0456A-P - Cable damage (4BBK40BK) to BBPCV0456A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0456A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-192 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-27Reactor Trip Switchgear Room184;:BMHV0001 - Cable damage (4BMK06AC) to BMHV0001 (BMHY0001A). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84;,BMHV0002 - Cable damage (4BMK06BC) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84;7BMHV0003 - Cable damage (4BMK06CC) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-193 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-27Reactor Trip Switchgear Room184;;BMHV0004 - Cable damage (4BMK06DC) to BMHV0004 (BMHY0004A). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the determinis tic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84;6EGRV0010 - Cable damage (6EGK03BA and 6EGK03BB) to EGRV0010. Cable damage can spuriously open Component Cooling Water Tank B Surge Tank Vent Valve, EGRV0010 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84;+EMHV8843 - Cable damage (4EMK04CA) to EMHV8843. Cable damage can spuriously open Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843. The valve is required closed in order to prevent potential diversion of RWST inventory to the SI test lines to ensure positive control over RCS Inventory and Pressure. This condition represents a varianc e from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-194 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-27Reactor Trip Switchgear Room184;5PZR-HTR-BU-A - Backup Pressurizer Heaters Groups A and B will not be available for a fire in this area due to cable damage, loss of DC control power, and/or loss of 4kV power (for Heater Group A - cables 5PGG05AA [4kV power to XPG21], 5BBG22AG [breaker PG2101 close/trip control], 5PKK01AJ and 5PKK01AW [loss of DC control power for breaker PG2101]; for Heater Group B - cables 6PGG05AA [4kV power to XPG22], 6BBG24AD, 6BBG24AG, 6BBG24AH, 6BBG24AJ, 6BBG24AM, 6BBG24AN, 6RPK15AA, and 6RPK09TA [breaker PG2201 close/trip control], and loss of 4kV Switchgear NB02 for breaker PG2201) cable 6PGG05AA is likely embedded in Fire Area A-27, but this does not matter due to loss of 4kV Switchgear NB02. Loss of pressurizer heater capability may adversely impact the ability of the plant to maintain safe and stable (potential adverse impact to RCS Pressure Control). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84;(SENIR0061 - No cable damage to SENE0061, but loss of power (from MCC NG02A) to Neutron Flux Detector, SENE0061 (SENE0061 indicates on Neutron Flux Monitoring Recorder, SENIR0061). Cable damage (1SES07AC, 1SES07AD, 1SES07AE, 1SES07BA, 1SES07BB, and 1SES07CA) to Neutron Flux Detector, SENE0060. Cable damage and/or loss of power results in loss of all Main Control Room neutron flux monitoring indication. At least one channel of neutron flux monitoring instrumentation is desired for monitoring of reactivity from the main control room to satisfy the NFPA 805 Performance Goal of Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.SGK05A - Cable damage (5GKK31CA and 5RPK09NA) to SGK05A. Cable damage may result in a loss of ventilation from the Train A ESF Switchgear Rooms Air Conditioning Unit, SGK05A. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-195 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-27Reactor Trip Switchgear Room184;ABPV0003 - Cable damage (3ABI20GA, 3AB I20GC, 3ABI20GD, and 3ABI20GE) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003, or cause the valve to fail closed. The valve is required open to establish a controlled cooldown of the RCS to maintain safe and stable plant conditions. Note that Steam Generator C(3) is credited for decay heat removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84;:EJHV8811B - Cable damage (4EJG06BC, 4EJG06BG, and 4EJG06BT) and loss of power (NB02) to EJHV8811A; cable damage (4BNG03BE) and loss of power (NG02A) to B NHV8812B. Direct cable damage and loss of power to EJHV8811B and BNHV8812B may result in these valves failing as-is (EJHV8811B is normally closed; BNHV8812B is normally open), but cannot spuriously open or close either valve. However, cable damage to other plant equipment (PORV, ASD, etc.) and/or to pressurizer pressure instrumentation in the fire area (BBPT0456 - cable 2BBI16LB; BBPT0458 - cable 4BBI16NB) and containment pressure instrumentation in the fire area (GNPT0936 - cable 2GNI05BA; GNPT0935 - cable 3GNI05CA) may result in a valid or spurious safety injection actuation signal. Furthermore, cable damage to RWST level instrumentation in the fire area (BNLT0932 - cable 3BNI07EA; BNLT0933 - cable 4BNI07FA) may cause a spurious RWST LOW-LOW actuation signal. The combination of a valid or spurious safety injection actuation signal and a spurious RWST LOW-LOW level actuation signal could spuriously open EJHV8811B (assuming that EJHV8811B remains functional even with cables and loss of power in the fire area, and that BNHV8812B fails as-is due to cables and loss of power in the fire area). The spurious operation could create a gravity drain flowpath from the RWST to the containment sump resulting in the depletion of RWST inventory which is necessary to maintain safe and stable plant conditions for NFPA 805. This condition represents a variance from the deterministic requirements of NFP A 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-196LIC-26 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-27Reactor Trip Switchgear Room1&$.>.?>..A&."&.&Ionization 105Detection NNYY1403NIonization 112Detection NNYYNHalonSKC03Suppression NNYYNERFBSNoneFeatureNNYYNSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneHalon system actuations are not expected to adversely affect electrical equipment. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. Should manual firefighting be required, water damage could result to the electrical equipment in this area (with or without fire damage); however, the water damage would not prevent safe shutdown. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-197LIC-26 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-28Auxiliary Shutdown Panel Section A1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel I Steam Gen. A Wide Range Level Channel I Aux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron See VFDR No. A-28-003, A-28-004, and A-4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. A-28-001, A-28-002, and A-28-006Fire ZoneDescription1413AAuxiliary Shutdown Panel RoomAugust 2011 C-198LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-28Auxiliary Shutdown Panel Section A1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B, and D, and ESW Pumps A and B. Offsite Power to NB01 and NB02 credited. HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.28-005RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NoneNone....3August 2011 C-199 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-28Auxiliary Shutdown Panel Section A18846(ABPV0001-P - Cable damage (1ABI20EC and 1ABI20ED) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.ABPV0003-P - Cable damage (3ABI20GC and 3ABI20GD) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.846BGHV8149A - Cable damage (5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-200 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-28Auxiliary Shutdown Panel Section A1846:BGHV8149B - Cable damage (5BGK35BD) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.846,BGHV8149C - Cable damage (5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.NB0105-P - Cable damage (1ALB01AD and 1ALB01AS) to NB0105. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A), NB0105. Non-credited train Motor Driven Auxiliary Feedwater Pump A may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issu e.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-201 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-28Auxiliary Shutdown Panel Section A1&$.>.?>..A&."&.&Ionization 118Detection NNNN1413ANN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. Any fire can be extinguished manually, using the portable extinguisher. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-202 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-29Auxiliary Feedwater Valve Compartment, SG A&D1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IIAux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel III Aux. Feedwater Flow to Steam Gen. C Channel IAux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. A-29-001 Fire ZoneDescription1304Auxiliary Feedwater Pipe Chase1324Auxiliary Feedwater Pump Valve Compartment No. 11327Auxiliary Feedwater Pump Valve Compartment No. 2August 2011 C-203LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-29Auxiliary Feedwater Valve Compartment, SG A&D1RCS Inventory ControlMaintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-204 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-29Auxiliary Feedwater Valve Compartment, SG A&D1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A). 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zones 1324 and 1327), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-205LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-29Auxiliary Feedwater Valve Compartment, SG A&D1884+(ABPV0003-P - Cable damage (3ABI 20GA and 3ABI20GE) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in the fire area. The RA has been demonstrated t o be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-206 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-29Auxiliary Feedwater Valve Compartment, SG A&D1&$.>.?>..A&."&.&
N/ANoneDetection 1304N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1324N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1327N/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-207 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-29Auxiliary Feedwater Valve Compartment, SG A&D1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-208 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-30Auxiliary Feedwater Valve Compartment, SG B&C1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channels ISteam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel IAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VI Steam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IVCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. A-30-001 Fire ZoneDescription1305Auxiliary Feedwater Pipe Chase (East)1328Auxiliary Feedwater Pump Valve Compartment No. 31330Auxiliary Feedwater Pump Valve Compartment No. 4August 2011 C-209LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-30Auxiliary Feedwater Valve Compartment, SG B&C1RCS Inventory ControlMaintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-210 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-30Auxiliary Feedwater Valve Compartment, SG B&C1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A). 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zones 1328 and 1330), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-211LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-30Auxiliary Feedwater Valve Compartment, SG B&C188(ABPV0003-P - Cable damage (3ABI20GA and 3ABI20GE) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-212 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-30Auxiliary Feedwater Valve Compartment, SG B&C1&$.>.?>..A&."&.&Thermal120Detection NNYN1305NN/ANoneSuppression N/ANoneFeatureThermal120Detection NNYN1328NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 1330N/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-213 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-30Auxiliary Feedwater Valve Compartment, SG B&C1None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-214 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-33Auxiliary Shutdown Panel Section B1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel I Steam Gen. B Wide Range Level Channel II Aux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent 4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. A-33-001, A-33-002, A-33-003, and A-33-004 Fire ZoneDescription1413BAuxiliary Shutdown Panel RoomAugust 2011 C-215LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-33Auxiliary Shutdown Panel Section B1flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NoneNone....3August 2011 C-216 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-33Auxiliary Shutdown Panel Section B188(ABPV0002-P - Cable damage (2ABI20FG and 2ABI20FK) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.ABPV0004-P - Cable damage (4ABI20HG and 4ABI20HK) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8FCHV0312-P - Cable damage (2FCK23AD, 2FCK23AZ, and 2RPK15CA) to FCHV0312-P. Cable damage can spuriously open Turbine Driven AFW Pump Trip and Throttle Valve FCHV0312-P, which could result in the inability to remotely secure the non-credited turbine driven AFW pump. The non-credited turbine driven AFW pump could become an uncontrolled source of inventory addition into Steam Generators 2(B) and 3(C), which could adversely impact the capability to maintain positive contro l over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generators 2(B) and 3(C) are credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-217 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-33Auxiliary Shutdown Panel Section B18:NB0205-P - Cable damage (4ALB01B1, 4ALB01BD, 4ALB01BM, 4ALB01BN, and 4RPK15AA) to NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.&$.>.?>..A&."&.&Ionization 118Detection NNNN1413BNN/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-218 Ameren MissouriCallaway Plant NFPA 805 Transition ReportA-33Auxiliary Shutdown Panel Section B1None&There are no automatic fire suppression systems in the fire area. Any fire can be extinguished manually, using the portable extinguisher. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-219 Ameren MissouriCallaway Plant NFPA 805 Transition ReportAB-1Auxiliary Boiler Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription4315Auxiliary Boiler RoomAugust 2011 C-220LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportAB-1Auxiliary Boiler Room1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NoneNone....3None8August 2011 C-221 Ameren MissouriCallaway Plant NFPA 805 Transition ReportAB-1Auxiliary Boiler Room1&$.>.?>..A&."&.&Flame406Detection NNNN4315NWet PipeSKC09Suppression NNNNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThis area contains no cabling or electrically supervised equipment that is required for post fire safe shutdown: therefore, fire suppression activities will not adversely affect the plant's ability to achieve the nuclear safety performance criteria.August 2011 C-222 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-1Pipe Space and Tank Area, Control Building, El. 1974'1~"Process MonitoringNORTH:RCS Pressure Channel IIPressurizer Pressure Channel IIPressurizer Level Channel II Ex-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel IISteam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IV Aux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IVCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IV4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBNORTH:Steam Generators A and D are supplied by MDAFW Pump B.SOUTH:Steam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription3101Pipe Space and Tank Area3104Pipe Space and Tank AreaAugust 2011 C-223LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-1Pipe Space and Tank Area, Control Building, El. 1974'1RCS Inventory ControlContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train B (Channel IV and VI) SOUTH:RCS Pressure Channel I Pressurizer Pressure Channel I Pressurizer Level Channel I Ex-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel ISteam Gen. B Pressure Channel ISteam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel I Containment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V)NORTH:Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary. SOUTH:Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.RCS Pressure ControlNORTH:Control pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.August 2011 C-224 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-1Pipe Space and Tank Area, Control Building, El. 1974'1Reactivity ControlNORTH: Trip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.SOUTH:Trip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesNORTH: Operate CCW Pumps B and D, and ESW Pump B.
Onsite Power to NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling. SOUTH:Operate CCW Pumps A and C, and ESW Pump A. Onsite Power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.See VFDR No. C-01-NORTH-001See VFDR No. C-01-SOUTH-001SOUTH:Control pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-225 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-1Pipe Space and Tank Area, Control Building, El. 1974'1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. A fire test was performed in accordance with the Standard for Fire Tests of Door Assemblies, UL 10B. 2. The watertight doors without gaskets are classified by UL as Special-Purpose Type Fire Doors and Frame Assemblies, Rating 3 hour (A). 3. To achieve watertight integrity criteria, gasketing material was added to the door assembly in accordance with the manufacturer's recommendations.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zone 3104), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zone 3104), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-226LIC-25LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-1Pipe Space and Tank Area, Control Building, El. 1974'1....3.%555,54The detectors in beam pockets in Fire Zone 3101 that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable based on low combustible loading in the area, lack of ignition sources, and presence of a wet pipe sprinkler system.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-227LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-1Pipe Space and Tank Area, Control Building, El. 1974'1885(C-1-ESW-TRAIN-A-HDPE-PIPE - No cable damage; C-1-ESW-TRAIN-A-HDPE-PIPE (representing high density polyethylene [HDPE] piping material for the Train A ESW piping in Analysis Area C-1 NORTH) fails on location alone; a deterministic fire event in Analysis Area C-1 NORTH may cause loss of Train A ESW piping pressure boundary integrity due to fire damage affecting the HDPE piping material. This could potentially result in ESW and/or SW flooding of Analysis Area C-1 SOUTH and Analysis Area C-1 NORTH (causing the following ESW motor operated valves in Fire Area C-1 to fail as-is in the last positi on: EFHV0023, EFHV0024, EFHV0025, EFHV0026, EFHV0037, EFHV0038, EFHV0039, EFHV0040, EFHV0041, and EFHV0042). The flooding could adversely impact nuclear safety capability with the credited Train B ESW system. This condition represents a variance fro m the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85(&@C-1-ESW-TRAIN-B-HDPE-PIPE - No cable damage; C-1-ESW-TRAIN-B-HDPE-PIPE (representing high density polyethylene [HDPE] piping material for the Train B ESW piping in Analysis Area C-1 SOUTH) fails on location alone; a deterministic fire event in Analysis Area C-1 SOUTH may cause loss of Train B ESW piping pressure boundary integrity due to fire damage affecting the HDPE piping material. This could potentially result in ESW and/or SW flooding of Analysis Area C-1 NORTH and Analysis Area C-1 SOUTH (causing the following ESW motor operated valves in Fire Area C-1 to fail as-is in the last position: EFHV0023, EFHV0024, EFHV0025, EFHV0026, EFHV0037, EFHV0038, EFHV0039, EFHV0040, EFHV0041, and EFHV0042). The flooding could adversely impact nuclear safety capability with the credited Train A ESW system. This condition represents a variance fro m the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-228 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-1Pipe Space and Tank Area, Control Building, El. 1974'1&$.>.?>..A&."&.&Ionization 330Detection NNNY3101NWet PipeSKC46Suppression NNNYN20-ft Separation ZoneNoneFeatureNNNYNN/ANoneDetection 3104N/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-229 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-1Pipe Space and Tank Area, Control Building, El. 1974'1To meet deterministic separation criteria, Fire Area C-1 Fire Zone 3101 is divided into two safe shutdown analysis areas C-1N and C-1S which are separated by a 20-foot separation zone. The 20-foot separation zone is clearly marked on the floor and designated as a "No Storage" location.&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-230 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-2Control Building North Cable Chase, Control Building, El. 19741~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription3106North Vertical Cable ChaseAugust 2011 C-231LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-2Control Building North Cable Chase, Control Building, El. 19741Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&NoneAugust 2011 C-232 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-2Control Building North Cable Chase, Control Building, El. 19741&$.>.?>..A&."&.&Ionization 330Detection NNNN3106NWet PipeSKC37Suppression NNNNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThis area contains no cabling or electrically supervised equipment that is required for post fire safe shutdown: therefore, fire suppression activities will not adversely affect the plant's ability to achieve the nuclear safety performance criteria.August 2011 C-233 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-3Control Building Cable Chase B, Control Build ing, El. 19741~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription3105South Vertical Cable ChaseAugust 2011 C-234LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-3Control Building Cable Chase B, Control Build ing, El. 19741Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&NoneAugust 2011 C-235 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-3Control Building Cable Chase B, Control Build ing, El. 19741&$.>.?>..A&."&.&Ionization 330Detection NNNN3105NWet PipeSKC37Suppression NNNNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThis area contains no cabling or electrically supervised equipment that is required for post fire safe shutdown: therefore, fire suppression activities will not adversely affect the plant's ability to achieve the nuclear safety performance criteria.August 2011 C-236 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-5Control Building Access Control Area, Control Building, El. 19841~"Process MonitoringRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channels IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel II4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription3212Women's Locker Room3213Womens Restroom3214Hall3215Briefing Room3216Men's Locker Room3217Mens Restroom3218RWP Sign-In/Sign-Out Area3219First Aid Room3220Key Access Area3221Vestibule No. 1 3222Health Physicist's Office 3223Janitor's Closet3224Vestibule No. 23233Women's Locker Room3236Storage3237Communications ClosetAugust 2011 C-237LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-5Control Building Access Control Area, Control Building, El. 19841RCS Inventory ControlSteam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IVCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IV Containment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-238 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-5Control Building Access Control Area, Control Building, El. 19841%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zones 3213, 3214, 3217, 3221, 3224, and 3236), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following: 
: 1. Minimal fire hazards.
: 2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zones 3213, 3217, 3219, 3220, and 3224), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following: 
: 1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-239LIC-25LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-5Control Building Access Control Area, Control Building, El. 19841None8August 2011 C-240 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-5Control Building Access Control Area, Control Building, El. 19841&$.>.?>..A&."&.&Ionization 300Detection NNNN3212NWet PipeSKC34Suppression NNYNNN/ANoneFeatureN/ANoneDetection 3213Wet PipeSKC34Suppression NNYNNN/ANoneFeatureN/ANoneDetection 3214Wet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3215NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3216NWet PipeSKC34Suppression NNYNNN/ANoneFeatureN/ANoneDetection 3217N/ANoneSuppression N/ANoneFeatureIonization 300Detection NNNN3218NWet PipeSKC34Suppression NNYNNN/ANoneFeatureAugust 2011 C-241 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-5Control Building Access Control Area, Control Building, El. 19841&$.>.?>..A&."&.&Ionization 300Detection NNNN3219NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3220NWet PipeSKC34Suppression NNYNNN/ANoneFeatureN/ANoneDetection 3221Wet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3222NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3223NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3224NN/ANoneSuppression N/ANoneFeatureIonization 300Detection NNNN3233NWet PipeSKC34Suppression NNYNNN/ANoneFeatureN/ANoneDetection 3236Wet PipeSKC34Suppression NNYNNN/ANoneFeatureAugust 2011 C-242 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-5Control Building Access Control Area, Control Building, El. 19841&$.>.?>..A&."&.&Ionization 300Detection NNNN3237NN/ANoneSuppression N/ANoneFeatureIonization 301Detection NNNNAll¹N¹area above the drop ceiling is open to all zonesWet PipeSKC38Suppression NNYNN¹area above the drop ceiling is open to all zones SLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNone&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-243 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-6Control Building Access Control Area, Control Building, El. 19841~"Process MonitoringRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel ISteam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and V4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription3201Stair No. C-13202Controlled HP Tool and Instr. Storage Room3204Corridor No. 13205Respirator Maintenance/Hot Janitor's Closet3206Women's Hot Shower3207Women's Disrobe3208Respro. Issues/Storage and Laundry3209Hall3210Men's Hot Shower3211Frisk Area 3231Men's Disrobe 3232Decon Area3234Audio/Video Storage3235ALARA Brief RoomAugust 2011 C-244LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-6Control Building Access Control Area, Control Building, El. 19841RCS Inventory ControlSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channel I Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-245 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-6Control Building Access Control Area, Control Building, El. 19841%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zones 3201, 3206, and 3210), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.
: 2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zones 3201, 3202, 3205, 3206, 3210 and 3234), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following: 
: 1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-246LIC-25LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-6Control Building Access Control Area, Control Building, El. 19841....3.%654(5An excessive gap in the bottom of Door DSK32014 connecting Fire Areas C-6 and C-35 is acceptable based on the lack of intervening combustibles at/near the location of the door. The door leads to the stairwell in Fire Area C-6 and, since it is a stairwell, no transient combustibles are expected near this doorway. Therefore DSK32014 is considered a non-rated feature commensurate with the fire hazards in the two areas and it provides an equivalent level of protection as a 3 hour rated fire door by prohibiting the propagation of fire between the two fire areas&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-247 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-6Control Building Access Control Area, Control Building, El. 19841&$.>.?>..A&."&.&
N/ANoneDetection 3201N/ANoneSuppression N/ANoneFeatureIonization 300Detection NNNN3202NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3204NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3205NWet PipeSKC34Suppression NNYNNN/ANoneFeatureN/ANoneDetection 3206Wet PipeSKC34Suppression NNYNNN/ANoneFeatureN/ANoneDetection 3207Wet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3208NWet PipeSKC34Suppression NNYNNN/ANoneFeatureAugust 2011 C-248 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-6Control Building Access Control Area, Control Building, El. 19841&$.>.?>..A&."&.&
N/ANoneDetection 3209Wet PipeSKC34Suppression NNYNNN/ANoneFeatureN/ANoneDetection 3210Wet PipeSKC34Suppression NNYNNN/ANoneFeatureN/ANoneDetection 3211Wet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3231NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3232NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3234NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 300Detection NNNN3235NWet PipeSKC34Suppression NNYNNN/ANoneFeatureIonization 301Detection NNNNAll ¹N¹area above the drop ceiling is open to all zonesAugust 2011 C-249 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-6Control Building Access Control Area, Control Building, El. 19841SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-250 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-7Control Building North Cable Chase, Control Building, El. 19841~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron See VFDR No. C-07-001, C-07-002, and C-4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription3230Electrical Chase (North)August 2011 C-251LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-7Control Building North Cable Chase, Control Building, El. 19841Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).07-003RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.August 2011 C-252 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-7Control Building North Cable Chase, Control Building, El. 19841885;55(BGHV8149A - Cable damage (5BGK35AB and 5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministi c requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85;554BGHV8149B - Cable damage (5BGK35BB and 5BGK35BD) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministi c requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85;55BGHV8149C - Cable damage (5BGK35CB and 5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministi c requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-253 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-7Control Building North Cable Chase, Control Building, El. 19841&$.>.?>..A&."&.&Ionization 300Detection NNNN3230NWet PipeSKC37Suppression NNNNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-254 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-8Control Building Cable Chase B, Control Build ing. El. 19841~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription3229Electrical Chase (South)August 2011 C-255LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-8Control Building Cable Chase B, Control Build ing. El. 19841Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&NoneAugust 2011 C-256 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-8Control Building Cable Chase B, Control Build ing. El. 19841&$.>.?>..A&."&.&Ionization 300Detection NNNN3229NWet PipeSKC37Suppression NNNNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-257 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-9ESF Switchgear Room A1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VI Steam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.See VFDR No. C-09-001, C-09-002, C-09-003, and C-09-0044.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-09-005 Fire ZoneDescription3301ESF Switchgear Room No. 1August 2011 C-258LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-9ESF Switchgear Room A1Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.
: 2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-259 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-9ESF Switchgear Room A1....3.%555,54The detectors in beam pockets in Fire Zone 3301 that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable based on low combustible loading in the area and lack of ignition sources. In addition, high air flow in the area is anticipated to spread out smoke to adjacent beam pocket spaces containing detectors.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-260LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-9ESF Switchgear Room A1885+55(Valve BGLCV0112B may spuriously close due to valve control cable damage (1BGG12AC and 1BGG12AD). Train B 4kV Switchgear NB02 may sustain a real or spurious loss of offsite power due to the following cable failures: 1NFK01DA, 1NFY01FA, 3NFK01DA, and 3NFY01HA (affecting LSELS-GRP2 - Train B load shed/load sequencer, NB02 voltage monitoring inputs);
6NBB02AA, 6NBB02AB, and 6NBB02AC (power cables from XNB02 to NB01 and NB02 breakers NB0109 and NB0209); 6NBA11AE and 6NBB05AL (protective trip cables for XNB02 feeder breaker PA0201). A real or spurious loss of offsite power affecting Train B 4kV Switchgear NB02 could initiate a loss of offsite power (LOOP) actuation of the Train B load shed/load sequencer. This would start normally idle Charging Pump B (the credited charging pump for safe shutdown), which could then fai l due to loss of pump suction from spurious closure of BGLCV0112B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85+554EMHV8803A-P - No cable damage to EMHV8803A. The Boron Injection Header Supply from Charging Pump A Isolation Valve, EMHV8803A, fails due to loss of power (from MCC NG01B). This non-credited train valve cannot be re-closed if opened by SIS and then subject to loss of power. The valve may need to be closed, or non-credited train Charging Pump A may need to be secured in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-261 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-9ESF Switchgear Room A185+55NB0103-P - Cable damage (1EMB01AA, 1EMB01AB, 1NBK13AA, and 1NBK13AB) to NB0103. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump A (PEM01A), NB0103. Non-credited train Safety Injection Pump A may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85+55:NB0104-P - Cable damage (1BGB01AA, 1BGB01AB, 1NBK13AA, and 1NBK13AB) to NB0104. Cable damage may result in spurious closure of the feeder breaker to Charging Pump A (PBG05A), NB0104. Non-credited train Charging Pump A may need to be secured, or non-credited train valve EMHV8803A may need to be closed in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85+55,NB0105-P - Cable damage (1ALB01AA, 1ALB01AB, 1ALB01AW, 1NBK13AA, and 1NBK13AB) to NB0105. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A), NB0105. Non-credited train Motor Driven Auxiliary Feedwater Pump A may need to be secured in order to ensure positive control over the rate of RCS Cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-262 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-9ESF Switchgear Room A1&$.>.?>..A&."&.&Ionization 314Detection NNYN3301NIonization 315Detection NNYNNHalonSKC01Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneHalon system actuations are not expected to adversely affect electrical equipment. Any fire can be extinguished manually with the portable extinguishers and/or hose stations after high-voltage equipment is de-energized. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. The water associated with manual fire suppression will drain out doors and as such  standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-263 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-10ESF Switchgear Room B1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel ISteam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel III Aux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel I Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channel IContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.See VFDR No. C-10-003, C-10-004, C-10-006, and C-10-0074.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-10-001, C-10-002, C-10-005, and C-10-008 Fire ZoneDescription3302ESF Switchgear Room No. 2August 2011 C-264LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-10ESF Switchgear Room B1Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%555,54The detectors in beam pockets in Fire Zone 3302 that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable based on lack of ignition sources and the small width of beam pockets (4'). In addition, high air flow in the area is anticipated to spread out smoke to adjacent beam pocket spaces containing detectors.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-265LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-10ESF Switchgear Room B188(555(ABPV0002-P - Cable damage (2ABI20FE, 2ABI20FG, 2ABI20FH, 2ABI20FJ, 2 ABI20FK, 2ABI20FL, 2ABI20FM, 2ABI20FN, and 2RPY09BA) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8(5554ABPV0004-P - Cable damage (4 ABI20HE, 4ABI20HG, 4ABI20HH, 4ABI20HJ, 4ABI 20HK, 4ABI20HL, 4 ABI20HM, 4ABI20HN, and 4RPY09GA) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited f or Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-266 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-10ESF Switchgear Room B18(555Valve BGLCV0112C may spuriously close due to valve control cable damage (4BGG12BC and 4BGG12BD). Train A 4kV Switchgear NB01 may sustain a real or spurious loss of offsite power due to the following cable failures: 2NFK01CA, 2NFY01EA, 4NFK01CA, and 4NFY01EA (affecting LSELS-GRP1 - Train A load shed/load sequencer, NB01 voltage monitoring inputs);
5NBB06AA, 5NBB06AB, and 5NBB06AC (power cables from XNB01 to NB01 and NB02 breakers NB0112 and NB0212);
 
6NBA10AB and 6NBB03AB (protective trip cables for XNB01 switchyard feeder breaker MD523). A real or spurious loss of offsite power affecting train a 4kV Switchgear NB01 could initiate a loss of offsite power (LOOP) actuation of the Train A load shed/lo ad sequencer. This would start normally idle Charging Pump A (the credited charging pump for safe shutdown), which could then fai l due to loss of pump suction from spurious closure of BGLCV0112C. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(555:EMHV8803B-P - No cable damage to EMHV8803B. The Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B, fails due to loss of power (from MCC NG04C). This non-credited train valve cannot be re-closed if opened by SIS and then subject to loss of power. The valve may need to be closed, or non-credited train Charging Pump B may need to be secured in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-267 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-10ESF Switchgear Room B18(555,FCHV0312-P - Cable dam age (2FCK23AA, 2FCK23AP, 2FCK23AQ, 2FCK23A R, 2FCK23AS, 2FCK23AT, 2FCK23AU, 2FCK23AX, 2RPK09BA, and 2RPK15CA) to FCHV0312-P. Cable damage can spuriously open Turbine Driven AFW Pump Trip and Throttle Valve FCHV0312-P, which could result in the inability to remotely secure the non-credited Turbine Driven AFW Pump. If running, the non-credited Turbine Driven AFW Pump could become an uncontrolled source of inventory ad dition into Steam Generators B and C, which could adversely impact the capability to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generators B and C are credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(5557NB0201-P - Cable damage (4BGB01BA, 4BGB01BB, 4NBK15AA, and 4NBK15AB) to NB0201. Cable damage may result in spurious closure of the feeder breaker to Charging Pump B (PBG05B), NB0201. Non-credited train Charging Pump B may need to be secured, or non-credited train valve EMHV8803B may need to be closed in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8(555;NB0202-P - Cable damage (4EMB01BA, 4EMB01BB, 4NBK15AA, and 4NBK15AB) to NB0202. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump B (PEM01B), NB0202. Non-credited train Safety Injection Pump B may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operatio ns. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issu e.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-268 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-10ESF Switchgear Room B18(5556NB0205-P - Cable damage (4ALB01BA, 4ALB01BJ, 4ALB01BK, 4ALB01BL, 4ALB01BM, 4ALB01BN, 4RPK15AA, 4NBK15AA, 4ALB01BH, 4RPK09NA, 4ALB0 1B2, 4ALB01BG, and 4NBK15AB) to NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.&$.>.?>..A&."&.&Ionization 316Detection NNYN3302NIonization 317Detection NNYNNHalonSKC01Suppression NNYNNERFBSNoneFeatureNNYNNSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-269 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-10ESF Switchgear Room B1None&Halon system actuations are not expected to adversely affect electrical equipment. Any fire can be extinguished manually with the portable extinguishers and/or hose stations after high-voltage equipment is de-energized. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. The water associated with manual fire suppression will drain out doors and as such standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-270 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-11Control Building Cable Chase B, Control Build ing, El. 20001~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel I Containment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from See VFDR No. C-11-002, C-11-003, C-11-004, and C-11-0054.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-11-001 and C-11-006 Fire ZoneDescription3305Electrical Chase (South)August 2011 C-271LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-11Control Building Cable Chase B, Control Build ing, El. 20001the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-272 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-11Control Building Cable Chase B, Control Build ing, El. 2000188((55(ABPV0004-P - Cable damage (4 ABI20HE, 4ABI20HG, 4ABI20HH, 4ABI20HJ, 4ABI 20HK, 4ABI20HL, 4 ABI20HM, 4ABI20HN, and 4RPY09GA) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited f or Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8((554Valve BGLCV0112C may spuriously close due to valve control cable damage (4BGG12BC and 4BGG12BD). Train A 4kV Switchgear NB01 may sustain a real loss of offsite power due to the following cable failures: 6NBA10AB, 6NBB03AB, 6NBB03AD, and 6NBB03AE (protective trip cables for XNB01 switchyard feeder breaker MD523). A real loss of offsite power affecting Train A 4kV Switchgear NB01 could initiate a loss of offsite power (LOOP) actuation of the Train A load shedder/load sequencer. This would start normally idle Charging Pump A (the credited charging pump for safe shutdown), which could then fail due to loss of pump suction from spurious closure of BGLCV0112C. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8((55EMHV8803B-P - No cable damage to EMHV8803B. The Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B, fails due to loss of power (from MCC NG04C). This non-credited train valve cannot be re-closed if opened by SIS and then subject to loss of power. The valve may need to be closed, or non-credited train Charging Pump B may need to be secured in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-273 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-11Control Building Cable Chase B, Control Build ing, El. 200018((55:NB0201-P - Cable damage (4BG B01BB, 4NBK15AA, and 4NBK15AB) to NB0201. Cable damage may result in spurious closure of the feeder breaker to Charging Pump B (PBG05B), NB0201. Non-credited train Charging Pump B may need to be secured, or non-credited Train Valve EMHV8803B may need to be closed in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Sectio n 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8((55,NB0202-P - Cable damage (4EMB01BB, 4NBK15AA, AND 4NBK15AB) to NB0202. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump B (PEM01B), NB0202. Non-credited train Safety Injection Pump B may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8((557NB0205-P - Cable damage (4ALB01BM, 4ALB01BN, 4NBK15AA, 4NBK15AB, 4RPK15AA, 4ALB01BH, 4RPK09NA, 4ALB01B2, AND 4ALB01BG) to NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-274 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-11Control Building Cable Chase B, Control Build ing, El. 20001&$.>.?>..A&."&.&Ionization 301Detection NNNN3305NWet PipeSKC37Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-275 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-12Control Building Cable Chase A, Control Build ing, El. 20001~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VI Steam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.See VFDR No. C-12-002, C-12-003, C-12-004, C-12-005, C-12-006, and C-12-0074.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-12-001 and C-12-008 Fire ZoneDescription3306Electrical Chase (North)August 2011 C-276LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-12Control Building Cable Chase A, Control Build ing, El. 20001Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-277 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-12Control Building Cable Chase A, Control Build ing, El. 2000188(455(ABPV0001-P - Cable damage (1ABI20EA, 1ABI20EC, AND 1ABI20ED) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a varian ce from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.554BGHV8149A - Cable damage (5BGK35AB and 5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(455BGHV8149B - Cable damage (5BGK35BB and 5BGK35BD) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-278 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-12Control Building Cable Chase A, Control Build ing, El. 200018(455:BGHV8149C - Cable damage (5BGK35CB and 5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(455,EMHV8803A-P - No cable damage to EMHV8803A. The Boron Injection Header Supply from Charging Pump A Isolation Valve, EMHV8803A, fails due to loss of power (from MCC NG01B). This non-credited train valve cannot be re-closed if opened by SIS and then subject to loss of power. The valve may need to be closed, or non-credited train Charging Pump A may need to be secured in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(4557NB0103-P - Cable damage (1EMB01AB, 1NBK13AA, and 1NBK13AB) to NB0103. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump A (PEM01A), NB0103. Non-credited train Safety Injection Pump A may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-279 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-12Control Building Cable Chase A, Control Build ing, El. 200018(455;NB0104-P - Cable damage (1BG B01AB, 1NBK13AA, and 1NBK13AB) to NB0104. Cable damage may result in spurious closure of the feeder breaker to Charging Pump A (PBG05A), NB0104. Non-credited train Charging Pump A may need to be secured, or Non-credited Train Valve EMHV8803A may need to be closed in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Sectio n 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(4556NB0105-P - Cable damage (1ALB01AB, 1ALB01AW, 1NBK13AA, and 1NBK13AB) to NB0105. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A), NB0105. Non-credited train Motor Driven Auxiliary Feedwater Pump A may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-280 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-12Control Building Cable Chase A, Control Build ing, El. 20001&$.>.?>..A&."&.&Ionization 301Detection NNNN3306NWet PipeSKC37Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-281 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-13Class 1E Train B AC Equipment Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel I Pressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel I Containment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from 4.2.3.2 - Deterministic ApproachRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription3415Access Control and Electrical Equip. A/C Units Room No. 1August 2011 C-282LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-13Class 1E Train B AC Equipment Room1the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A. Onsite power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-283 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-13Class 1E Train B AC Equipment Room1&$.>.?>..A&."&.&Ionization 303Detection NNNN3415NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. A manual water spray system is provided in the charcoal adsorber unit. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. The water associated with manual fire suppression will drain out doors and as such standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-284 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-14Class 1E Train A AC Equipment Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VI Steam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription3416Access Control and Electrical Equip. A/C Units Room No. 2August 2011 C-285LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-14Class 1E Train A AC Equipment Room1Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-286 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-14Class 1E Train A AC Equipment Room1&$.>.?>..A&."&.&Ionization 303Detection NNNN3416NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. A manual water spray system is provided in the charcoal adsorber unit. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. The water associated with manual fire suppression will drain out doors and as such standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-287 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-15Battery and Switchboard Room B, Control Building, El. 20161~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel I RCS Loop B (2) T-hot Temperature Channel ISteam Gen. B Pressure Channel ISteam Gen. B Narrow Range Level Channel III Aux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel III Steam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channel I Core Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.See VFDR No. C-15-001, C-15-002, and C-15-0034.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-15-004 Fire ZoneDescription3403Non-Vital Switchgear and Transformer Room No. 13404Switchboard Room No. 4 (Rm. 3404)3405Battery Room No. 43410Switchboard Room No. 23411Battery Room No 2August 2011 C-288LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-15Battery and Switchboard Room B, Control Building, El. 20161Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.
HVAC credited for Main Control Room and Containment (Train A credited).RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.August 2011 C-289 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-15Battery and Switchboard Room B, Control Building, El. 2016188(,55(EMHV8803B-P - No cable damage to EMHV8803B. The Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B, fails due to loss of power (from MCC NG04C). This non-credited train valve cannot be re-closed if opened by SIS and then subject to loss of power. The valve may need to be closed, or non-credited train Charging Pump B may need to be secured in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.554NB0201-P - Cable damage (4NBK15AA and 4NBK15AB) to NB0201. Cable damage may result in loss of control power for the feeder breaker to Charging Pump B (PBG05B), NB0201. Non-credited train Charging Pump B may need to be secured, or non-credited train valve EMHV8803B may need to be closed in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(,55NB0202-P - Cable damage (4NBK15AA and 4NBK15AB) to NB0202. Cable damage may result in loss of control power for the feeder breaker to Safety Injection Pump B (PEM01B), NB0202. Non-credited train Safety Injection Pump B may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-290 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-15Battery and Switchboard Room B, Control Building, El. 201618(,55:NB0205-P - Cable damage (4RPK09NA, 4NBK15AA, and 4NBK15AB) to NB0205. Cable damage may result in loss of control power for the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. Thi s is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-291 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-15Battery and Switchboard Room B, Control Building, El. 20161&$.>.?>..A&."&.&Ionization 304Detection NNYN3403NIonization 305Detection NNYNNN/ANoneSuppression N/ANoneFeatureIonization 321Detection NNYN3404NIonization 322Detection NNYNNN/ANoneSuppression N/ANoneFeatureIonization 303Detection NNYN3405NN/ANoneSuppression N/ANoneFeatureIonization 324Detection NNYN3410NIonization 328Detection NNYNNN/ANoneSuppression N/ANoneFeatureIonization 303Detection NNYN3411NN/ANoneSuppression N/ANoneFeatureAugust 2011 C-292 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-15Battery and Switchboard Room B, Control Building, El. 20161SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. The water associated with manual fire suppression will drain out doors and as such standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-293 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-16Battery and Switchboard Room A, Control Building, El. 20161~"Process MonitoringRCS Pressure Channel IIPressurizer Pressure Channel IIPressurizer Level Channel IIEx-core Neutron Monitoring Channel IV RCS Loop A (1) T-cold Temperature Channel IISteam Gen. A Pressure Channel IISteam Gen. A Narrow Range Level Channel IV Aux. Feedwater Flow to Steam Gen. A Channel IV RCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels II and IV Condensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IIVolume Control Tank Level Channels IV Containment Pressure Channels I, II, III, and IV4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-16-004 Fire ZoneDescription3407Battery Room No. 13408Switchboard Room No. 13409Non-Vital Switchgear and Transformer Room  No. 23413Battery Room No. 33414Switchboard Room No. 3August 2011 C-294LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-16Battery and Switchboard Room A, Control Building, El. 20161RCS Inventory ControlCore Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.See VFDR No. C-16-001, C-16-002, and C-16-003RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.August 2011 C-295 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-16Battery and Switchboard Room A, Control Building, El. 2016188(755(EMHV8803A-P - No cable damage to EMHV8803A. The Boron Injection Header Supply from Charging Pump A Isolation Valve, EMHV8803A, fails due to loss of power (from MCC NG01B). This non-credited train valve cannot be re-closed if opened by SIS and then subject to loss of power. The valve may need to be closed, or non-credited train Charging Pump A may need to be secured in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.554NB0103-P - Cable damage (1NBK13AA and 1NBK13AB) to NB0103. Cable damage may result in loss of control power for the feeder breaker to Safety Injection Pump A (PEM01A), NB0103. Non-credited train Safety Injection Pump A may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.NB0104-P - Cable damage (1NBK13AA and 1NBK13AB) to NB0104. Cable damage may result in loss of control power for the feeder breaker to Charging Pump A (PBG05A), NB0104. Non-credited train Charging Pump A may need to be secured, or non-credited train valve EMHV8803A may need to be closed in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-296 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-16Battery and Switchboard Room A, Control Building, El. 201618(755:NB0105-P - Cable damage (1NBK13AA and 1NBK13AB) to NB0105. Cable damage may result in loss of control power for the feeder breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A), NB0105. Non-credited train Motor Driven Auxiliary Feedwater Pump A may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-297 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-16Battery and Switchboard Room A, Control Building, El. 20161&$.>.?>..A&."&.&Ionization 303Detection NNYN3407NN/ANoneSuppression N/ANoneFeatureIonization 325Detection NNYN3408NIonization 326Detection NNYNNN/ANoneSuppression N/ANoneFeatureIonization 323Detection NNYN3409NIonization 327Detection NNYNNN/ANoneSuppression N/ANoneFeatureIonization 303Detection NNYN3413NN/ANoneSuppression N/ANoneFeaturePhotoelectric 318Detection NNYN3414NIonization 320Detection NNYNNN/ANoneSuppression N/ANoneFeatureAugust 2011 C-298 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-16Battery and Switchboard Room A, Control Building, El. 20161SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. The water associated with manual fire suppression will drain out doors and as such standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-299 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-17Control Building Cable Chase B, Control Build ing, El. 20161~"Process MonitoringRCS Inventory ControlRCS Pressure Channels IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel I Containment Pressure Channels I, II, and III Core Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from See VFDR No. C-17-002, C-17-003, C-17-004, and C-17-0054.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-17-001 and C-17-006 Fire ZoneDescription3418Electrical Chase (South)August 2011 C-300LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-17Control Building Cable Chase B, Control Build ing, El. 20161the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-301 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-17Control Building Cable Chase B, Control Build ing, El. 2016188(;55(ABPV0004-P - Cable damage (4 ABI20HE, 4ABI20HG, 4ABI20HH, 4ABI20HJ, 4ABI 20HK, 4ABI20HL, 4 ABI20HM, 4ABI20HN, 4RPY09BA, 4RPY09CA, and 4RPY09GA) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the Main Steam Pressure Boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(;554Valve BGLCV0112C may spuriously close due to valve control cable damage (4BGG12BC). Train A 4kV Switchgear NB01 may sustain a real loss of offsite power due to the following cable failures:
6NBA10AB, 6NBB03AD, and 6NBB03AE (protective trip cables for XNB01 switchyard feeder breaker MD523). A real loss of offsite power affecting Train A 4kV Switchgear NB01 could initiate a loss of offsite power (LOOP) actuation of the Train A load shed/load sequencer. This would start normally idle Charging Pump A (the credited charging pump for safe shutdown), which could then fail due to loss of pump suction from spurious closure of BGLCV0112C. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(;55EMHV8803B-P - No cable damage to EMHV8803B. The Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B, fails due to loss of power (from MCC NG04C). This non-credited train valve cannot be re-closed if opened by SIS and then subject to loss of power. The valve may need to be closed, or non-credited train Charging Pump B may need to be secured in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-302 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-17Control Building Cable Chase B, Control Build ing, El. 201618(;55:NB0201-P - Cable damage (4BG B01BB, 4NBK15AA, and 4NBK15AB) to NB0201. Cable damage may result in spurious closure of the feeder breaker to Charging Pump B (PBG05B), NB0201. Non-credited train Charging Pump B may need to be secured, or Non-Credited Train Valve EMHV8803B may need to be closed in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 80 5, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(;55,NB0202-P - Cable damage (4EMB01BB, 4NBK15AA, and 4NBK15AB) to NB0202. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump B (PEM01B), NB0202. Non-credited train Safety Injection Pump B may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(;557NB0205-P - Cable damage (4ALB01BM, 4ALB01BN, 4NBK15AA, 4NBK15AB, 4RPK15AA, 4ALB01BH, 4RPK09NA, 4ALB01B2, and 4ALB01BG) to NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-303 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-17Control Building Cable Chase B, Control Build ing, El. 20161&$.>.?>..A&."&.&Ionization 303Detection NNNN3418NWet PipeSKC37Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-304 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-18Control Building Cable Chase A, Control Build ing, El. 20161~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VI Steam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.See VFDR No. C-18-002, C-18-003, C-18-004, C-18-005, C-18-007, C-18-008, C-18-009, and C-18-0104.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-18-001 and C-18-011 Fire ZoneDescription3419Electrical Chase (North)August 2011 C-305LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-18Control Building Cable Chase A, Control Build ing, El. 20161Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.See VFDR No. C-18-006RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-306 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-18Control Building Cable Chase A, Control Build ing, El. 2016188(655(ABPV0001-P - Cable damage (1 ABI20EA, 1ABI20EC, 1ABI20ED, 1RPY09CA, and 1RPY09DA) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the Main Steam Pressure Boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.554BBPCV0455A-P - Cable damage (1BBK40AK) to BBPCV0455A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(655BGHV8149A - Cable damage (5BGK35AB and 5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-307 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-18Control Building Cable Chase A, Control Build ing, El. 201618(655:BGHV8149B - Cable damage (5BGK35BB and 5BGK35BD) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, SECTION 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(655,BGHV8149C - Cable damage (5BGK35CB and 5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.557EGRV0009 - Cable damage (5EGK03AA and 5EGK03AB) to EGRV0009. Cable damage can spuriously open Component Cooling Water Tank A Surge Vent Valve, EGRV0009 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-308 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-18Control Building Cable Chase A, Control Build ing, El. 201618(655;EMHV8803A-P - Cable damage (1EMG02AC) to EMHV8803A. Cable damage can spuriously open or close the Boron Injection Header Supply from Charging Pump A Isolation Valve, EMHV8803A. This non-credited train valve may need to be closed, or Non-credited Train Charging Pump A may need to be secured in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(6556NB0102-P - Cable damage (1ENB01AB, 1ENB01AD, 1NBK13AA, and 1NBK13AB) to NB0102. Cable damage may result in spurious closure of the feeder breaker to Containment Spray Pump A (PEN01A), NB0102. Non-credited train Containment Spray Pump A may need to be secured in order to prevent potential diversion of RWST inventory to the containment spray headers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(655+NB0103-P - Cable damage (1EMB01AB, 1NBK13AA, and 1NBK13AB) to NB0103. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump A (PEM01A), NB0103. Non-credited train Safety Injection Pump A may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-309 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-18Control Building Cable Chase A, Control Build ing, El. 201618(65(5NB0104-P - Cable damage (1BG B01AB, 1NBK13AA, and 1NBK13AB) to NB0104. Cable damage may result in spurious closure of the feeder breaker to Charging Pump A (PBG05A), NB0104. Non-credited train Charging Pump A may need to be secured, or non-credited train valve EMHV8803A may need to be closed in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Sectio n 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8(65((NB0105-P - Cable damage (1ALB01AB, 1ALB01AD, 1ALB01AG, 1ALB01AR, 1ALB01AS, 1ALB01AW, 1NBK13AA, and 1NBK13AB) to NB0105. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A), NB0105. Non-credited train Motor Driven Auxiliary Feedwater Pump A may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-310 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-18Control Building Cable Chase A, Control Build ing, El. 20161&$.>.?>..A&."&.&Ionization 303Detection NNNN3419NWet PipeSKC37Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-311 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-19Control Building Cable Chase A at column C-3, Control Building, El. 20161~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel I Aux. Feedwater Flow to Steam Gen. A Channels I and IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VI Steam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels I, II, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescriptionC19Electrical Chase (North)August 2011 C-312LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-19Control Building Cable Chase A at column C-3, Control Building, El. 20161Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&NoneAugust 2011 C-313 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-19Control Building Cable Chase A at column C-3, Control Building, El. 20161&$.>.?>..A&."&.&Ionization 303Detection NNNNC19Nroom has no number so the fire area is usedWet PipeSKC37Suppression NNNNNroom has no number so the fire area is used N/ANoneFeatureroom has no number so the fire area is used SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-314 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-20Control Building Cable Chase B at column C-6, Control Building, El. 20161~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I Steam Gen. A Pressure Channel I Steam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel IAux. Feedwater Flow to Steam Gen. A Channels I and IV RCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and IV Condensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IV Containment Pressure Channels I, III, and IV Core Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-20-001 Fire ZoneDescriptionC20Electrical Chase (South)August 2011 C-315LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-20Control Building Cable Chase B at column C-6, Control Building, El. 20161Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-316 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-20Control Building Cable Chase B at column C-6, Control Building, El. 20161884555(ABPV0002-P - Cable damage (2ABI20FE, 2ABI20FG, 2ABI20FH, 2ABI20FJ, 2AB I20FK, 2ABI20FL, 2ABI20FM, 2ABI20FN, 2RPY09AA, and 2RPY09BA) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.&$.>.?>..A&."&.&Ionization 303Detection NNNNC20Nroom has no number so the fire area is usedWet PipeSKC37Suppression NNNNNroom has no number so the fire area is used N/ANoneFeatureroom has no number so the fire area is used SLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-317 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-20Control Building Cable Chase B at column C-6, Control Building, El. 20161None&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-318 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel IIPressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel IISteam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IV Aux. Feedwater Flow to Steam Gen. A Channel IV RCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels II and IVCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IVCore Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.See VFDR No. C-21-003, C-21-004, C-21-005, C-21-006, C-21-007, C-21-008, C 009, C-21-010, C-21-013, C-21-014, C 015, C-21-016, C-21-018, and C-21-0194.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-21-001, C-21-002, and C-21-017Fire ZoneDescription3501Lower Cable Spreading RoomAugust 2011 C-319LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room1Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).See VFDR No. C-21-011 and C-21-012RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-320 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.
: 2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 3
/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zone 3501), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-321LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room1....3.%555,54The detectors in beam pockets in Fire Zone 3501 that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable based on low combustible loading in the area, lack of ignition sources, lack of cable trays, and the very small width of beam pockets (1'-6").&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&.%~,+The cable trench in the floor of the Control Room (C-27) that does not conform to ASTM E 119 temperature rating requirements is acceptable based on calculations that determine that the maximum unexposed side cable trench temperatures will not exceed the thermal insult temperature of the cable, with an acceptable thermal margin such that the trenches will prevent the spread of fire from one fire area to another. The configuration of the cable trenches will not affect the plant's ability to achieve and maintain post-fire safe shutdown.&August 2011 C-322LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room1884(55(ABPV0001-P - Cable damage (1ABI20EA, 1ABI20EB, 1ABI20EC, 1ABI20ED, 1ABI20EE, 1RPY09CA, and 1RPY09DA) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS Cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84(554ABPV0003-P - Cable damage (3 ABI20GA, 3ABI20GB, 3 ABI20GC, 3ABI20GD, 3ABI20GE, and 3RPY09AA) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate o f RCS Cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84(55BBPCV0455A-P - Cable damage (1BBK40AE and 1BBK40AK) to BBPCV0455A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-323 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room184(55:BBPCV0455B - Cable damage (5BBI19AA) to BBPCV0455B. Cable damage can spuriously open the Reactor Coolant Pump A Pressurizer Spray Line Isolation Valve, BBPCV0455B (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84(55,BBPCV0455C - Cable damage (5BBI19BA) to BBPCV0455C. Cable damage can spuriously open the Reactor Coolant Pump B Pressurizer Spray Line Isolation Valve, BBPCV0455C (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.557BGHV8149A - Cable damage (5BGK35AA, 5BGK35AB, and 5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-324 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room184(55;BGHV8149B - Cable damage (5BGK35BA, 5B GK35BB, and 5BGK35BD) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84(556BGHV8149C - Cable damage (5BGK35CA, 5BGK35CB, and 5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-325 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room184(55+Valve BGLCV0112B may spuriously close due to valve control cable damage (1BGG12AC) and cable damage affecting VCT Level Transmitter BGLT0112 (1BGI51CA and 1SBS01AB). Train B 4kV Switchgear NB02 may sustain a real or spurious loss of offsite power due to the following cable failures: 1NFK01DA, 1NFY01BA, 1NFY01FA, 3NFK01DA, 3NFY01FA, and 3NFY01HA (affecting LSELS-GRP2 - Train B load shed/load sequencer, NB02 voltage monitoring inputs); 5PAA0 3AA, 5PAA10AB, and 5PAA12AB (spurious trip/close, or fail as-is of breakers PA0101 and PA0110); CA-00152 (spurious trip of switchyard breaker MDV41, feeder to start-up transformer XMR01); CA-00153 (spurious trip of switchyard breaker MDV43, feeder to start-up transformer XMR01); CA-00155 (spurious trip of switchyard breaker MDV45, feeder to start-up transformer XMR01); CA-00150 (failure to trip on demand of switchyard breaker MD523, safeguards switchgear 13kV circuit breaker 52-3); CA-00156 (spurious trip of switchyard breaker MDV51, 345kV breaker CAL-LSCR-2 to Bus A tie breaker); CA-00158 (failure to trip on demand of switchyard breaker MDV53, 345kV breaker main generator output CAL-LSCR-2); CA-00159 (failure to trip on demand trip of switchyard breaker MDV55, 345kV breaker main generator output Bus B tie breaker); CA-00160 (spurious trip of switchyard breaker MDV85, 345kV breaker MONT/CAL7 tie to Bus B); CA-00163 (spurious trip of switchyard breaker MDV75, 345kV breaker MTGY/CAL8 tie to Bus B); CA-00164 (spurious trip of switchyard breaker MDV81, 345kV breaker MONT/CAL7 tie to Bus A); and CA-00165 (spurious trip of switchyard breaker MDV71, 345kV breaker MTGY/CAL8 tie to Bus A). A real or spurious loss of offsite power affecting Train B 4kV Switchgear NB02 could initiate a loss of offsite power (LOOP) actuation of the Train B load shed/load sequencer. This would start normally idle Charging Pump B (the credited charging pump for safe shutdown), which could then fai l due to loss of pump suction from spurious closure of BGLCV0112B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-326 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room184(5(5BNHV8812A - Cable damage (1BNG03AC) to BNHV8812A. Cable damage may result in the Refueling Water Storage Tank to Residual Heat Removal Pump A Suction Isolation Valve, BNHV8812A, failing as-is (open). Cable damage (1BNI07CA, 1SBS01AC, and 1SBS02AC) to Refueling Water Storage Tank Level Transmitter BNLT0930 combined with cable damage (3BNI07EA, 3SBS01CA, and 3SBS02CA) to Refueling Water Storage Tank Level Transmitter BNLT0932 may generate a spurious RWST Low Level Permissive for the Containment Recirculation Sump A to Residual Heat Removal Pump A Suction Isolation Valve, EJHV8811A. Although the control circuit for valve EJHV8811A has been modified by plant modification MP 09-0025 such that the valve cannot spuriously open due to direct control cable damage in this fire area, valve EJHV8811A could sti ll spuriously open due to the spurious Refueling Water Storage Tank Low Level Permissive Signal coincident with a Safety Injection Actuation Signal, which could also occur in fire area C-21. Valve EJHV8811A could then fail as-is (open) due to direct cable damage and/or subsequent loss of power. This would cause a draindown of the Refueling Water Storage Tank to the containment sump through open valves BNHV8812A and EJHV8811A. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84(5((EGHV0011 - Cable damage (1EGG04AC and 1EGG04BC) to EGHV0011 and EGHV0013 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train A Upstream Valve, EGHV0011, and Essential Service Water to Component Cooling Train A Downstream Valve, EGHV0013. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from loss of component cooling water inventory into the Train A Essential Service Water Header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-327 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room184(5(4EGRV0009 - Cable damage (5EGK03AA and 5EGK03AB) to EGRV0009. Cable damage can spuriously open Component Cooling Water Tank A Surge Vent Valve, EGRV0009 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84(5(EMHV8803A-P - Cable damage (1EMG02AC and 1EMG02AD) to EMHV8803A. Cable damage can spuriously open or close the Boron Injection Header Supply from Charging Pump A Isolation Valve, EMHV8803A. This non-credited train valve may need to be closed, or non-credited train Charging Pump A may need to be secured in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NF PA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84(5(:NB0102-P - Cable damage (1 ENB01AB, 1ENB01AC, and 1ENB01AD) to NB0102. Cable damage may result in spurious closure of the feeder breaker to Containment Spray Pump A (PEN01A), NB0102. Non-credited train Containment Spray Pump A may need to be secured in order to prevent potential diversion of RWST inventory to the containment spray headers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-328 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room184(5(,NB0103-P - Cable damage (1EMB01AB and 1EMB01AC) to NB0103. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump A (PEM01A), NB0103. Non-credited train Safety Injection Pump A may need to be secured in order to prevent potential diversion of RWST. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fi ll lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84(5(7NB0104-P - Cable damage (1BGB01AB and 1BGB01AE) to NB0104. Cable damage may result in spurious closure of the feeder breaker to Charging Pump A (PBG05A), NB0104. Non-credited train Charging Pump A may need to be secured, or non-credited train valve EMHV8803A may need to be closed in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84(5(;NB0105-P - Cable damage (1ALB01AB, 1ALB01AC, 1ALB01AD, 1ALB01AG, 1ALB01AR, 1ALB01AS, 1ALB01AT, 1ALB01AV, and 1ALB01AW) to NB0105. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A), NB0105. Non-credited train Motor Driven Auxiliary Feedwater Pump A may need to be secured in order to ensure positive control over the rate of RCS Cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-329 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room184(5(6PA0107 - Cable damage (5BBA01AB, 5BBA01AJ, 5BBA01AK, 5BBA01AL, and 5BBA01AM) to PA0107. Cable damage may result in spurious closure or failure to trip on demand of the feeder breaker to the Reactor Coolant Pump A (PBB01A), PA0107. Non-credited Reactor Coolant Pump A may need to be secured in order to mitigate spurious pressurizer spray valve opening, establish natural circulation, and ensure positive control over RCS Decay Heat Removal capability. Reactor coolant pump seal cooling is unaffected in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84(5(+PA0108 - Cable damage (5BBA01BB, 5BBA01BJ, 5BBA01BK, 5BBA01BL, and 5BBA01BM) to PA0108. Cable damage may result in spurious closure or failure to trip on demand of the feeder breaker to the Reactor Coolant Pump B (PBB01B), PA0108. Non-credited Reactor Coolant Pump B may need to be secured in order to mitigate spurious pressurizer spray valve opening, establish natural circulation, and ensure positive control over RCS Decay Heat Removal capability. Reactor coolant pump seal cooling is unaffected in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-330 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-21Lower Cable Spreading Room1&$.>.?>..A&."&.&Ionization 306Detection NNYN3501YPre-action SKC39Suppression NNYNYN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-331 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel III Aux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Refueling Water Storage Tank Level Channel IVolume Control Tank Level Channel ICore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.See VFDR No. C-22-003, C-22-004, C-22-010, C-22-011, C-22-012, C-22-014, C-22-015, C-22-016, C-22-018, C-22-019, C-22-020, C-22-021 and C-22-0224.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-22-001, C-22-002, C-22-005, C-22-006, C-22-007, C-22-008, C 013, and C-22-017 Fire ZoneDescription3801Upper Cable Spreading RoomAugust 2011 C-332LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room1Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.
HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. C-22-009 and C-22-010(BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-333 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.
: 2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 3
/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zone 3801), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-334LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room1....3.%555,54The detectors in beam pockets in Fire Zone 3801 that are not installed in accordance with Section 4-3.7.3 of NFPA 72E-1978 Edition are acceptable based on low combustible loading in the area or limited presence of non-safety cable trays protected by 2 sprinklers
.&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-335LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room1884455(ABPV0002-P - Cable damage (2ABI20FB and 2RPY09AA) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS Cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.554ABPV0004-P - Cable damage (4ABI20HB, 4ABI20HJ, 4ABI20HL, 4ABI20HM, 4ABI20HN, 4RPY09BA, and 4RPY09CA) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over t he rate of RCS Cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.BBPCV0456A-P - Cable damage (4BBK40BE and 4BBK40BK) to BBPCV0455A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0456A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-336 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room184455:Valve BGLCV0112C may spuriously close due to valve control cable damage (4BGG12BC) and cable damage affecting VCT Level Transmitter BGLT0185 (4BGI51DA and 4SBS02DA). Train A 4kV Switchgear NB01 may sustain a spurious loss of offsite power due to the following cable failures: 2NFK01CA, 2NFY01CA, 2NFY01EA, 4NFK01CA, 4NFY01EA, and 4NFY01GA (affecting LSELS-GRP1 - Train A load shed/load sequencer, NB01 voltage monitoring inputs); CA-00128 (failure to trip on demand of switchyard breaker MD522, safeguards switchgear 13kV circuit breaker 52-2); CA-00130 (failure to trip on demand of switchyard breaker MDV41, 345kV breaker Bus A to start-up transformer); CA-00132 (failure to trip on demand of switchyard breaker MDV43, 345kV breaker start-up transformer CAL/BLAND1 tie breaker); CA-00134 (failure to trip on demand of switchyard breaker MDV45, 345kV breaker CAL/BLAND1 tie to Bus B); CA-00136 (failure to trip on demand of switchyard breaker MDV51, 345kV breaker CAL-LSCR-2 to Bus A tie breaker); CA-00137 (failure to trip on demand of switchyard breaker MDV53, 345kV breaker main generator output CAL-LSCR-2); CA-00138 (failure to trip on demand trip of switchyard breaker MDV55, 345kV breaker main generator output Bus B tie breaker); CA-00146 (spurious trip of switchyard breaker MDV85, 345kV breaker MONT/CAL7 tie to Bus B); CA-00144 (spurious trip of switchyard breaker MDV75, 345kV breaker MTGY/CAL8 tie to Bus B); CA-00145 (spurious trip of switchyard breaker MDV81, 345kV breaker MONT/CAL7 tie to Bus A); and CA-00141 (spurious trip of switchyard breaker MDV71, 345kV breaker MTGY/CAL8 tie to Bus A). A spurious loss of offsite power affecting Train A 4kV Switchgear NB01 could initiate a loss of offsite power (LOOP) actuation of the Train A load shed/load sequencer. This would start normally idle Charging Pump A (the credited charging pump for safe shutdown), which could then fail due to loss of pump suction from spurious closure of BGLCV0112C. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84455,BMHV0001 - Cable damage (4BMK06AA, 4BMK06AB, and 4BMK06AC) to BMHV0001 (BMHY0001A). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS Cooldown. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-337 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room1844557BMHV0002 - Cable damage (4BMK06BA, 4BMK06BB, and 4BMK06BC) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS Cooldown. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84455;BMHV0003 - Cable damage (4BMK06CA, 4BMK06CB, and 4BMK06CC) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS Cooldown. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.844556BMHV0004 - Cable damage (4BMK06DA, 4BMK06DB, and 4BMK06DC) to BMHV0004 (BMHY0004A). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS Cooldown. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-338 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room184455+EGHV0012 - Cable damage (4EGG04CC and 4EGG04DC) to EGHV0012 and EGHV0014 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train B Upstream Valve, EGHV0012, and Essential Service Water to Component Cooling Train B Downstream Valve, EGHV0014. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from loss of component cooling water inventory into the Train B Essential Service Water Header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8445(5EGRV0010 - Cable damage (6EGK03BA and 6EGK03BB) to EGRV0010. Cable damage can spuriously open Component Cooling Water Tank B Surge Tank Vent Valve, EGRV0010 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8445((EMHV8803B-P - Cable damage (4EMG02BC and 4EMG02BD) to EMHV8803B. Cable damage can spuriously open or close the Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B. This non-credited train valve may need to be closed, or non-credited train Charging Pump B may need to be secured in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NF PA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-339 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room18445(4EMHV8843 - Cable damage (4EMK04CA) to EMHV8843. Cable damage can spuriously open Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843. The valve is required closed in order to prevent potential diversion of RWST inventory to the SI test lines to ensure positive control over RCS Inventory and Pressure. This condition represents a varianc e from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8445(FCHV0312-P - Cable damage (2FCK23AA and 2FCK23AP) to FCHV0312-P. Cable damage can spuriously open Turbine Driven AFW Pump Trip and Throttle Valve FCHV0312-P. The non-credited Turbine Driven AFW Pump could become an uncontrolled source of inventory addition into Steam Generators B and C, which could adversely impact the capability to maintain positive control over the rate of RCS Cooldown, and to maintain RCS sub-cooling. Note that Steam Generators B and C are credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8445(:NB0201-P - Cable damage (4BGB01BB and 4BGB01BE) to NB0201. Cable damage may result in spurious closure of the feeder breaker to Charging Pump B (PBG05B), NB0201. Non-credited train Charging Pump B may need to be secured, or non-credited train valve EMHV8803B may need to be closed in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-340 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room18445(,NB0202-P - Cable damage (4EMB01BB and 4EMB01BC) to NB0202. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump B (PEM01B), NB0202. Non-credited train Safety Injection Pump B may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8445(7NB0203-P - Cable damage (4 ENB01BB, 4ENB01BC, and 4ENB01BD) to NB0203. Cable damage may result in spurious closure of the feeder breaker to Containment Spray Pump B (PEN01B), NB0203. Non-credited train Containment Spray Pump B may need to be secured in order to prevent potential diversion of RWST inventory to the containment spray headers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8445(;NB0205-P - Cable damage (4ALB01BC, 4ALB01BD, 4ALB01BH, 4ALB01B1, 4ALB01B2, 4ALB01BG, 4ALB01BP, 4ABL01BR, 4ALB01BY, and 4ALB01BZ) to NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS Cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-341 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room18445(6PA0204 - Cable damage (6BBA01DB, 6BBA01DJ, 6BBA01DL, and 6BBA01DM) to PA0204. Cable damage may result in spurious closure or failure to trip on demand of the feeder breaker to the Reactor Coolant Pump D (PBB01D), PA0204. Non-credited Reactor Coolant Pump D may need to be secured in order to establish natural circulation and ensure positive control over RCS Decay Heat Removal capability. Reactor coolant pump seal cooling is unaffected in this fire area. This condition represen ts a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8445(+PA0205 - Cable damage (6BBA01CB, 6BBA01CJ, 6BBA01CL, and 6BBA01CM) to PA0205. Cable damage may result in spurious closure or failure to trip on demand of the feeder breaker to the Reactor Coolant Pump C (PBB01C), PA0205. Non-credited Reactor Coolant Pump C may need to be secured in order to establish natural circulation and ensure positive control over RCS Decay Heat Removal capability. Reactor coolant pump seal cooling is unaffected in this fire area. This condition represen ts a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.844545PG2401 - Cable damage (6BBG20AB and 6BBG20AC) to PG2401. Cable damage may result in a spurious close signal to the non-credited train Pressurizer Heater Control Group C Breaker, PG2401, with the loss of remote trip control capability. Non-credited train Pressurizer Heater Control Group C may need to be secured in order to ensure positive control over RCS Pressure (to prevent pressurizer PORV challenge). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-342 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room184454(BBHV8141C - Cable damage (6BBK05CA) to BBHV8141C. Cable damage can spuriously close the Reactor Coolant Pump C Seal #1 Water Outlet Isolation Valve, BBHV8141C (spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to maintain the effectiveness of Thermal Barrier Cooling for the Reactor Coolant Pump Seal Package. This requirement is based on the latest Westinghouse Owners Group Analysis for the performance of the Reactor Coolant Pumps Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump Seal Injection may be lost in this area due to fire damage potentially spuriously closing the reactor coolant pump C Seal Water Supply Isolation Valve, BBHV8351C (Cable 4BBG04CC), and fire damage potentially affecting the Charging Pump Suction Supply Valve from the Volume Control Tank (VCT), and causing a loss of offsite power (refer to the Fire Area C-22 VFDR for equipment BGLCV0112C). With the exception of spurious closure of BBHV8141C, Component Cooling Water for Reactor Coolant Pump C Thermal Barrier can be recovered from the Main Control Room in this area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.844544BBHV8141D - Cable damage (6BBK05DA) to BBHV8141D. Cable damage can spuriously close the Reactor Coolant Pump D Seal #1 Water Outlet Isolation Valve, BBHV8141D (spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to maintain the effectiveness of Thermal Barrier Cooling for the Reactor Coolant Pump Seal Package. This requirement is based on the latest Westinghouse Owners Group Analysis for the performance of the Reactor Coolant Pump Seal Package following the loss of Seal Cooling. Note that Reactor Coolant Pump Seal Injection may be lost in this area due to fire damage potentially spuriously closing the reactor coolant pump D Seal Water Supply Isolation Valve, BBHV8351D (Cable 4BBG04DC), and fire damage potentially affecting the Charging Pump Suction Supply Valve from the Volume Control Tank (VCT), and causing a loss of offsite power (refer to the fire area C-22 VFDR for equipment BGLCV0112C). With the exception of spurious closure of BBHV8141D, Component Cooling Water for Reactor Coolant Pump C Thermal Barrier can be recovered from the Main Control Room in this area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-343 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-22Upper Cable Spreading Room1&$.>.?>..A&."&.&Ionization 307Detection NNYN3801NPre-action SKC40Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Floor penetrations in this area are provided with raised sleeves or curbs, and all penetrations have watertight seals to prevent water damage in the Control Room below during fire-fighting operations in this area. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-344 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-23Control Building Cable Chase B, Control Build ing, El. 20321~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel ISteam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel III Aux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel I Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channel IContainment Pressure Channels I, II, and IIICore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.See VFDR No. C-23-002, C-23-003, C-23-009, C-23-010, C-23-011, C-23-012, C-23-013, and C-23-0154.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-23-001, C-23-004, C-23-005, C-23-006, C-23-007, and C-23-014 Fire ZoneDescription3505Electrical ChaseAugust 2011 C-345LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-23Control Building Cable Chase B, Control Build ing, El. 20321Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. C-23-008Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-346 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-23Control Building Cable Chase B, Control Build ing, El. 2032188455(ABPV0004-P - Cable damage (4 ABI20HE, 4ABI20HG, 4ABI20HH, 4ABI20HJ, 4ABI 20HK, 4ABI20HL, 4 ABI20HM, 4ABI20HN, 4RPY09BA, and 4RPY09CA) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84554BBPCV0456A-P - Cable damage (4BBK40BK) to BBPCV0456A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0456A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8455Valve BGLCV0112C may spuriously close due to valve control cable damage (4BGG12BC). Train A 4kV Switchgear NB01 may sustain a real loss of offsite power due to the following cable failures:
6NBA10AB, 6NBB03AD, and 6NBB03AE (protective trip cables for XNB01 switchyard feeder breaker MD523). A real loss of offsite power affecting Train A 4kV Switchgear NB01 could initiate a loss of offsite power (LOOP) actuation of the Train A load shed/load sequencer. This would start normally idle Charging Pump A (the credited charging pump for safe shutdown), which could then fail due to loss of pump suction from spurious closure of BGLCV0112C. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-347 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-23Control Building Cable Chase B, Control Build ing, El. 203218455:BMHV0001 - Cable damage (4BMK06AC) to BMHV0001 (BMHY0001A). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8455,BMHV0002 - Cable damage (4BMK06BC) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requiremen ts of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84557BMHV0003 - Cable damage (4BMK06CC) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-348 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-23Control Building Cable Chase B, Control Build ing, El. 203218455;BMHV0004 - Cable damage (4BMK06DC) to BMHV0004 (BMHY0004A). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the determinis tic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84556EGRV0010 - Cable damage (6EGK03BA and 6EGK03BB) to EGRV0010. Cable damage can spuriously open Component Cooling Water Tank B Surge Tank Vent Valve, EGRV0010 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8455+EMHV8803B-P - No cable damage to EMHV8803B. The Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B, fails due to loss of power (from MCC NG04C). This non-credited train valve cannot be re-closed if opened by SIS and then subject to loss of power. The valve may need to be closed, or non-credited train Charging Pump B may need to be secured in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-349 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-23Control Building Cable Chase B, Control Build ing, El. 20321845(5EMHV8843 - Cable damage (4EMK04CA) to EMHV8843. Cable damage can spuriously open Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843. The valve is required closed in order to prevent potential diversion of RWST inventory to the SI test lines to ensure positive control over RCS inventory and pressure. This condition represents a varianc e from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.845((NB0201-P - Cable damage (4BGB01BB) to NB0201. Cable damage may result in spurious closure of the feeder breaker to Charging Pump B (PBG05B), NB0201. Non-credited train Charging Pump B may need to be secured, or non-credited train Valve EMHV8803B may need to be closed in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.845(4NB0202-P - Cable damage (4EMB01BB) to NB0202. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump B (PEM01B), NB0202. Non-credited train Safety Injection Pump B may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fi ll lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-350 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-23Control Building Cable Chase B, Control Build ing, El. 20321845(NB0203-P - Cable damage (4ENB01BB and 4ENB01BD) to NB0203. Cable damage may result in spurious closure of the feeder breaker to Containment Spray Pump B (PEN01B), NB0203. Non-credited train Containment Spray Pump B may need to be secured in order to prevent potential diversion of RWST inventory to the containment spray headers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.845(:NB0205-P - Cable damage (4ALB01BM, 4ALB01BN, 4RPK15AA, 4ALB01BH, 4ALB01B2, and 4ALB01BG) to NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.845(,PG2401 - Cable damage (6BBG20AB and 6BBG20AC) to PG2401. Loss of Breaker Control Power (PK42) (after battery depletion). Cable damage may result in a spurious close signal to the non-credited train Pressurizer Heater Control Group C Breaker, PG2401, with the loss of remote trip control capability. Non-credited train Pressurizer Heater Control Group C may need to be secured in order to ensure positive control over RCS pressure (to prevent pressurizer PORV challenge). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-351 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-23Control Building Cable Chase B, Control Build ing, El. 20321&$.>.?>..A&."&.&Ionization 303Detection NNNN3505NWet PipeSKC37Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-352 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-24Control Building Cable Chase A, Control Build ing, El. 20321~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VI Steam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.See VFDR No. C-24-002, C-24-003, C-24-004, C-24-005, C-24-008, C-24-009, C-24-010, and C-24-0114.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-24-001 and C-24-012 Fire ZoneDescription3504Electrical ChaseAugust 2011 C-353LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-24Control Building Cable Chase A, Control Build ing, El. 20321Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).See VFDR No. C-24-006 and C-24-007RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-354 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-24Control Building Cable Chase A, Control Build ing, El. 20321884:55(ABPV0001-P - Cable damage (1ABI20EA, 1ABI20EC, 1ABI20ED, 1ABI20EE, 1RPY09CA, and 1RPY09DA) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate o f RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84:554BBPCV0455A-P - Cable damage (1BBK40AK) to BBPCV0455A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84:55BGHV8149A - Cable damage (5BGK35AB and 5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-355 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-24Control Building Cable Chase A, Control Build ing, El. 2032184:55:BGHV8149B - Cable damage (5BGK35BB and 5BGK35BD) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84:55,BGHV8149C - Cable damage (5BGK35CB and 5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.557EGHV0011 - Cable damage (1EGG04AC and 1EGG04BC) to EGHV0011 and EGHV0013 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train A Upstream Valve, EGHV0011, and Essential Service Water To Component Cooling Train A Downstream Valve, EGHV0013. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from loss of component cooling water inventory into the Train A Essential Service Water Header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-356 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-24Control Building Cable Chase A, Control Build ing, El. 2032184:55;EGRV0009 - Cable damage (5EGK03AA and 5EGK03AB) to EGRV0009. Cable damage can spuriously open Component Cooling Water Tank A Surge Tank Vent Valve, EGRV0009 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84:556EMHV8803A-P - Cable damage (1EMG02AC) to EMHV8803A. Cable damage can spuriously open or close the Boron Injection Header Supply from Charging Pump A Isolation Valve, EMHV8803A. This non-credited train valve may need to be closed, or non-credited train Charging Pump A may need to be secured in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84:55+NB0102-P - Cable damage (1ENB01AB and 1ENB01AD) to NB0102. Cable damage may result in spurious closure of the feeder breaker to Containment Spray Pump A (PEN01A), NB0102. Non-credited train Containment Spray Pump A may need to be secured in order to prevent potential diversion of RWST inventory to the containment spray headers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-357 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-24Control Building Cable Chase A, Control Build ing, El. 2032184:5(5NB0103-P - Cable damage (1EMB01AB) to NB0103. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump A (PEM01A), NB0103. Non-credited train Safety Injection Pump A may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fi ll lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84:5((NB0104-P - Cable damage (1BGB01AB) to NB0104. Cable damage may result in spurious closure of the feeder breaker to Charging Pump A (PBG05A), NB0104. Non-credited train Charging Pump A may need to be secured, or Non-Credited Train Valve EMHV8803A may need to be closed in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.84:5(4NB0105-P - Cable damage (1ALB01AB, 1ALB01AD, 1ALB01AG, 1ALB01AR, 1ALB01AS, and 1ALB01AW) to NB0105. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A), NB0105. Non-credited train Motor Driven Auxiliary Feedwater Pump A may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-358 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-24Control Building Cable Chase A, Control Build ing, El. 20321&$.>.?>..A&."&.&Ionization 303Detection NNYN3504NWet PipeSKC37Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-359 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-25Control Building Cable Chase B at column C-6, Control Building, El. 20321~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I Steam Gen. A Pressure Channel I Steam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel IAux. Feedwater Flow to Steam Gen. A Channels I and IV RCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and IV Condensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IV Containment Pressure Channels I, III, and IV Core Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-25-001 Fire ZoneDescriptionC25Electrical Chase (South)August 2011 C-360LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-25Control Building Cable Chase B at column C-6, Control Building, El. 20321Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-361 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-25Control Building Cable Chase B at column C-6, Control Building, El. 20321884,55(ABPV0002-P - Cable damage (2ABI20FE, 2ABI20FG, 2ABI20FH, 2ABI20FJ, 2 ABI20FK, 2ABI20FL, 2ABI20FM, 2ABI20FN, and 2RPY09AA) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is not credited f or Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.&$.>.?>..A&."&.&Ionization 303Detection NNNNC25Nroom has no number so the fire area is usedWet PipeSKC37Suppression NNNNNroom has no number so the fire area is used N/ANoneFeatureroom has no number so the fire area is used SLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-362 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-25Control Building Cable Chase B at column C-6, Control Building, El. 20321None&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-363 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-26Control Building Cable Chase A  at column  C-3, Control Building, El. 20321~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel I Aux. Feedwater Flow to Steam Gen. A Channels I and IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VI Steam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels I, II, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron 4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-26-001 Fire ZoneDescriptionC26Electrical Chase (North)August 2011 C-364LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-26Control Building Cable Chase A  at column  C-3, Control Building, El. 20321Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B. Offsite power to NB01 and NB02 credited. HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-365 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-26Control Building Cable Chase A  at column  C-3, Control Building, El. 20321884755(ABPV0003-P - Cable damage (3ABI20GA, 3ABI20GC, 3ABI20GD, 3ABI20GE, and 3RPY09AA) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.&$.>.?>..A&."&.&Ionization 303Detection NNNNC26Nroom has no number so the fire area is usedWet PipeSKC37Suppression NNNNNroom has no number so the fire area is used N/ANoneFeatureroom has no number so the fire area is used SLE RD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation
- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationAugust 2011 C-366 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-26Control Building Cable Chase A  at column  C-3, Control Building, El. 20321None&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-367 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1~"Process MonitoringRCS Pressure Channel IVPressurizer Level Channels II Ex-core Neutron Monitoring Channel IVRCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVRCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Wide Range Level Channel II Steam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channel IIAux. Feedwater Pump B Suction Pressure Channel IVTurbine Driven Aux. Feedwater Pump Suction Pressure Channel IIRefueling Water Storage Tank Level Local Mechanical Instrument4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generator D is supplied by MDAFW Pump B, and Steam Generator B is supplied by the TDAFW Pump.See VFDR No. C-27-001, C-27-002, C-27-003, C-27-004, C-27-005, C-27-006, C-27-007, C-27-008, C-27-009, C-27-010, C-27-011, C-27-012, C-27-013, C-27-014, C 015, C-27-041, C-27-042, C-27-043, C 044, and C-27-092 Fire ZoneDescription3601Control Room3603Shift Managers Office3604Foyer3605Equipment Cabinet Area3606Emergency Equipment Storage Room3616Vestibule (South)August 2011 C-368LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1RCS Inventory ControlMaintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.
HVAC credited for Containment (Train B credited).See VFDR No. C-27-016, C-27-017, C-27-018, C-27-019, C-27-020, C-27-021, C-27-022, C-27-023, C-27-024, C-27-025, C 026, C-27-027, C-27-028, C-27-029, C 030, C-27-031, C-27-032, C-27-033, C 034, C-27-035, C-27-036, C-27-037, C-27-038, C-27-039, C-27-040, C-27-045, C-27-046, C-27-047, C-27-048, C-27-073, C 074, C-27-075, C-27-089, C-27-090, C-27-091, C-27-093, C-27-095, C-27-096, C-27-102, C-27-115, C-27-116, C-27-117, C 118, C-27-119, and C-27-120See VFDR No. C-27-049, C-27-050, C-27-051, C-27-052, C-27-053, C-27-054, C 055, C-27-056, C-27-057, C-27-058, C 059, C-27-060, C-27-061, C-27-062, C-27-063, C-27-064, C-27-065, C-27-066, C-27-067, C-27-068, C-27-069, C-27-070, C-27-071, C-27-072, C-27-076, C-27-077, C-27-078, C-27-079, C-27-080, C-27-081, C-27-082, C-27-083, C-27-084, C-27-085, C 086, C-27-087, C-27-088, C-27-094, C 095, C-27-096, C-27-097, C-27-098, C-27-099, C-27-100, C-27-101, C-27-103, C-27-104, C-27-105, C-27-106, C-27-107, C 108, C-27-109, C-27-110, C-27-111, C-27-112, C-27-113, C-27-114, C-27-121, C-27-122, C-27-123, C-27-124, and C-27-125 and C-27-126.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use Pressurizer Auxiliary Spray to depressurize, after performing necessary repairs.August 2011 C-369 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-370 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.
: 2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zone 3604), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~.( &Deviation submitted 2/2/1984 SNUPPS letter to the NRC, providing justification for insufficient separation between the  Load Shed Emergency Load Sequencer (LSELS) panels because the redundant panels are located in the same area of the Control Room and their output relays are mounted back-to-back in a common panel, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Procedures are available for manual starting of the diesel generators and for manual sequencing of the safe shutdown loads onto the Class 1E AC busses from outside the Control Room.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-371LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1%~.(Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, providing justification for lack of low level detectors in the Contr ol Room, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following: 
: 1. Union Electric (dba Ameren Missouri) committed to provide a duct detector in the Control Room HVAC exhaust duct.2. The HVAC exhaust inlets are near the floor level.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~.(Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, providing justification for lack of smoke detectors in all Control Room cabinets and consoles containing redundant equipment, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Union Electric (dba Ameren Missouri) committed to provide detection in the Control Room cabinets containing redundant safe-shutdown equipment.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&The design submitted 11/15/1982 and 8/23/1984 SNUPPS letters to the NRC, providing justification for the alternate shutdown capability. The design was approved by the NRC in NUREG-0830, Supplement 4, dated 1 0/1984 based on the following:  1. The phased procedural approach submitted August 23, 1984
: 2. Interim procedures identified for use until the installation of the five new isolation switches and the modifications to four of the existing switches.&This deviation is active; however, modifications to the Safe Shutdown strategy, procedures, and panel configuration have occurred since the original submittal. The current configuration will be clarified in Attachment T. Related implementation item 11-805-056 for plant modifications is summarized in Attachment S.August 2011 C-372 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&.%~,+The cable trench in the floor of the Control Room (C-27) that does not conform to ASTM E 119 temperature rating requirements is acceptable based on calculations that determine that the maximum unexposed side cable trench temperatures will not exceed the thermal insult temperature of the cable, with an acceptable thermal margin such that the trenches will prevent the spread of fire from one fire area to another. The configuration of the cable trenches will not affect the plant's ability to achieve and maintain post-fire safe shutdown.&.%+5;47Keeping Fire Door DSK36021 between the Control Room (Fire Zone 3601, Fire Area C-27) and the Control Room Pantry (Fire Zone 3602, Fire Area C-28) in the held-open position is acceptable based on the fact that the door is in the direct view of Control Room operators, the Control Room is continuously manned, and a heat detector is installed in Fire Zone 3602.&August 2011 C-373LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1884;55(ABHV0011 - Cable damage (1ABK26AB, 1ABK29BC, 1ABK29BH, 1ABK29BK, 1ABK29BN, 1ABK29BS, 1ABK29BV, 4ABK27AB, 4ABK28BB, 4ABK28BD, 4ABK28BH, 4ABK28BK, and 4ABK28BN) to ABHV0011 (ABHV0011V13A, ABHV0011V13B, ABHV0011V13C, ABHV0011V15A, ABHV0011V15B, and ABHV0011V15C) to Steam Generator D Main Steam Isolation Valve ABHV0011. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator D is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;554ABHV0012 - Cable damage (1ABK23AG and 4ABK23FG) to ABHV0012 (ABHY0012A and ABHY0012B) to Steam Generator D Main Steam Loop 4 ABHV0011 Bypass Isolation Valve ABHV0012. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-374 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;55ABHV0014 - Cable damage (1ABK26AB, 1ABK28AB, 1ABK28AD, 1ABK28AH, 1ABK28AK, 1ABK28AN, 4ABK27AB, 4ABK29AC, 4ABK29AH, 4ABK29AK, 4ABK29AN, 4ABK29AS, and 4ABK29AV) to ABHV0014 (ABHV0014V13A, ABHV0014V13B, ABHV0014V13C, ABHV0014V15A, ABHV0014V15B, and ABHV0014V15C) to Steam Generator A Main Steam Isolation Valve ABHV0014. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.55:ABHV0015 - Cable damage (1ABK23AG and 4ABK23FG) to ABHV0015 (ABHY0015A and ABHY0015B) to Steam Generator A Main Steam Loop 1 ABHV0014 Bypass Isolation Valve ABHV0015. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isola te the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This conditio n represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-375 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;55,ABHV0017 - Cable damage (1ABK26AB, 1ABK29AC, 1ABK29AH, 1ABK29AK, 1ABK29AN, 1ABK29AS, 1ABK29AV, 4ABK27AB, 4ABK28AB, 4ABK28AD, 4ABK28AH, 4ABK28AK, and 4ABK28AN) to ABHV0017 (ABHV0017V13A, ABHV0017V13B, ABHV0017V13C, ABHV0017V15A, ABHV0017V15B, and ABHV0017V15C) to Steam Generator B Main Steam Isolation Valve ABHV0017. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.557ABHV0018 - Cable damage (1ABK23AG and 4ABK23FG) to ABHV0018 (ABHY0018A and ABHY0018B) to Steam Generator B Main Steam Loop 2 ABHV0017 Bypass Isolation Valve ABHV0015. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isola te the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-376 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;55;ABHV0020 - Cable damage (1ABK26AB, 1ABK28BB, 1ABK28BD, 1ABK28BH, 1ABK28BK, 1ABK28BN, 4ABK27AB, 4ABK29BC, 4ABK29BH, 4ABK29BK, 4ABK29BN, 4ABK29BS, and 4ABK29BV) to ABHV0020 (ABHV0020V13A, ABHV0020V13B, ABHV0020V13C, ABHV0020V15A, ABHV0020V15B, and ABHV0020V15C) to Steam Generator C Main Steam Isolation Valve ABHV0020. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. The valve is also required closed to terminate steam flow to the Main Feedwater Pump turbine and thereby secure main feedwater flow (following Main Feedwater Pump coastdown) to all four steam generators. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.556ABHV0021 - Cable damage (1ABK23AG and 4ABK23FG) to ABHV0021 (ABHY0021A and ABHY0021B) to Steam Generator C Main Steam Loop 3 ABHV0020 Bypass Isolation Valve ABHV0021. Cable damage can spuriously open valve or prevent valve from closing on demand (spurious opening is only credible assuming external hot shorts). The valve is required closed to isola te the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This conditio n represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-377 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;55+ABPV0001-P - Cable damage (1ABI20EA, 1ABI20EB, 1ABI20EC, 1ABI20ED, 1ABI20EE, 1RPY09CA, and 1RPY09DA) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5(5ABPV0003-P - Cable damage (3 ABI20GA, 3ABI20GB, 3 ABI20GC, 3ABI20GD, 3ABI20GE, and 3RPY09AA) to Pressure Transmitter ABPT0003. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0003. The valve is required closed to isolate the main steam pressure boundary for Steam Generator C, to maintain positive control over the rate o f RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5((ALHV0006 - Cable damage (1ALI05AA, 1ALI05AB, 1ALI05AC, 1ALI05AD, 1ALI05AE, 1ALI05AF, 1RPY09CA, and 1RPY09DA) to ALHV0006. Cable damage can spuriously open Turbine Driven Auxiliary Feedwater Pump to Steam Generator D valve, ALHV0006. This valve is required closed to secure flow from the credited Turbine Driven Auxiliary Feedwater Pump to credited Steam Generator D. Credited Steam Generator D is monitored, but will be supplied by credited Motor Driven Auxiliary Feedwater Pump B. This action is performed to maintain positive control of RCS cooldown. Also, ALHV0006 could fail fully open, and may pose a pump runout concern for the credited Turbine Driven Auxiliary Feedwater Pump. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-378 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5(4ALHV0007 - Cable damage (4ALI03BA, 4ALI03BB, 4ALI03BC, 4ALI03BD, 4ALI03BE, 4ALI03BF, 4ALY09BD, 4RPY09BA, and 4RPY09CA) to ALHV0007. Cable damage can spuriously open Motor Driven Auxiliary Feedwater Pump B to Steam Generator A Valve, ALHV0007. This valve is required closed to secure flow from credited Motor Driven Auxiliary Feedwater Pump B to non-credited Steam Generator A. Non-credited Steam Generator A is not monitored. This action is performed to maintain positive control of RCS cooldown. Also, ALHV0007 could fail fully open, and may pose a pump runout concern for the credited Motor Driven Auxiliary Feedwater Pump B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5(ALHV0008 - cable damage (1ALI05BA, 1ALI05BB, 1ALI05BC, 1ALI05BD, 1ALI05BE, 1ALI05BF, 1RPY09CA, and 1RPY09DA) to ALHV0008. Cable damage can spuriously open Turbine Driven Auxiliary Feedwater Pump to Steam Generator A Valve, ALHV0008. This valve is required closed to secure flow from the credited Turbine Driven Auxiliary Feedwater Pump to non-credited Steam Generator A. Non-credited Steam Generator A is not monitored. This action is performed to maintain positive control of RCS cooldown. Also, ALHV0008 could fail fully open, and may pose a pump runout concern for the credited Turbine Driven Auxiliary Feedwater Pump. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-379 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5(:ALHV0012 - Cable damage (4ALI05BA, 4ALI05BB, 4ALI05BC, 4ALI05BD, 4ALI05BE, 4ALI05BF, 4RPY09BA, and 4RPY09CA) to ALHV0012. Cable damage can spuriously open Turbine Driven Auxiliary Feedwater Pump to Steam Generator C Valve, ALHV0012. This valve is required closed to secure flow from the credited Turbine Driven Auxiliary Feedwater Pump to non-credited Steam Generator C. The non-credited steam supply valve to the credited Turbine Driven Auxiliary Feedwater Pump from Steam Generator C (ABHV0006) may have spuriously opened. Non-credited Steam Generator C is not monitored. This action is performed to maintain positive control of RCS cooldown and also addresses the potential for overfill of Steam Generator C, whic h could flood the steam supply line to the credited Turbine Driven Auxiliary Feedwater Pump. Also, ALHV0012 could fail fully open, and may pose a pump runout concern for the credited Turbine Driven Auxiliary Feedwater Pump. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.5(,ALHV0036 - Cable damage (1ALG02CC and 1ALG02CE) to ALHV0036. Cable damage can spuriously open or close Condensate Storage Tank to Turbine Driven Auxiliary Feedwater Pump Valve, ALHV0036. This valve is required to open (to initially align the Condensate Storage Tank to the Turbine Driven Auxiliary Feedwater Pump suction) and close (to subsequently isolate the Condensate Storage Tank from the Turbine Driven Auxiliary Feedwater Pump suction). These actions are required to support the Performance Goal of Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-380 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5(7BBHV8001A- Cable damage (1BBK30AA and 1BBK30CA) to BBHV8001A and BBHV8002A respectively. Cable damage can spuriously open or prevent closure of the Reactor Coolant System Reactor Vessel Head Vent Protection A Upstream and Downstream Valves, BBHV8001A and BBHV8002A. At least one of these two valves is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5(;BBHV8001B - Cable damage (4BBK30BA and 4BBK30DA) to BBHV8001B and BBHV8002B respectively. Cable damage can spuriously open or prevent closure of the Reactor Coolant System Reactor Vessel Head Vent Protection B Upstream and Downstream Valves, BBHV8001B and BBHV8002B. At least one of these two valves is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-381 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5(6BBHV8141A - Cable damage (5BBK05AA) to BBHV8141A. Cable damage can spuriously close the Reactor Coolant Pump A Seal #1 Water Outlet Isolation Valve, BBHV8141A (spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to satisfy Westinghouse criteria for maintaining a fixed rate of 21 gpm from the Reactor Coola nt Pump Seal Package following the loss of all Reactor Coolant Pump Seal Cooling for greater than 13-minutes. This requirement is based on the latest Westinghouse Owners Group analysis for the performance of the Reactor Coolant Pump Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump A Seal Injection may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Chemical Volume and Control System. Reactor Coolant Pump A Seal Injection is isolated, and is not recovered for the Main Control Room fire evacuation event. Reactor Coolant Pump A Thermal Barrier Cooling may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Component Cooling Water System. Reactor Coolant Pump A Thermal Barrier Cooling is isolated, and is not recovered for the main control room fire evacuation event. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-382 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5(+BBHV8141B - Cable damage (5BBK05BA) to BBHV8141B. Cable damage can spuriously close the Reactor Coolant Pump B Seal #1 Water Outlet Isolation Valve, BBHV8141B (spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to satisfy Westinghouse criteria for maintaining a fixed rate of 21 gpm from the Reactor Coola nt Pump Seal Package following the loss of all Reactor Coolant Pump Seal Cooling for greater than 13-minutes. This requirement is based on the latest Westinghouse Owners Group analysis for the performance of the Reactor Coolant Pump Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump B Seal Injection may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Chemical Volume and Control System. Reactor Coolant Pump B Seal Injection is isolated, and is not recovered for the Main Control Room fire evacuation event. Reactor Coolant Pump B Thermal Barrier Cooling may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Component Cooling Water System. Reactor Coolant Pump B Thermal Barrier Cooling is isolated, and is not recovered for the Main Control Room fire evacuation event. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-383 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;545BBHV8141C - cable damage (6BBK05CA) to BBHV8141C. Cable damage can spuriously close the Reactor Coolant Pump C Seal #1 Water Outlet Isolation Valve, BBHV8141C (spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to satisfy Westinghouse criteria for maintaining a fixed rate of 21 gpm from the Reactor Coola nt Pump Seal Package following the loss of all Reactor Coolant Pump Seal Cooling for greater than 13-minutes. This requirement is based on the latest Westinghouse Owners Group analysis for the performance of the Reactor Coolant Pump Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump C Seal Injection may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Chemical Volume and Control System. Reactor Coolant Pump C Seal Injection is isolated, and is not recovered for the Main Control Room fire evacuation event. Reactor Coolant Pump C Thermal Barrier Cooling may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Component Cooling Water System. Reactor Coolant Pump C Thermal Barrier Cooling is isolated, and is not recovered for the Main Control Room fire evacuation event. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-384 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;54(BBHV8141D - cable damage (6BBK05DA) to BBHV8141D. Cable damage can spuriously close the Reactor Coolant Pump D Seal #1 Water Outlet Isolation Valve, BBHV8141D (spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to satisfy Westinghouse criteria for maintaining a fixed rate of 21 gpm from the Reactor Coola nt Pump Seal Package following the loss of all Reactor Coolant Pump Seal Cooling for greater than 13-minutes. This requirement is based on the latest Westinghouse Owners Group analysis for the performance of the Reactor Coolant Pump Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump D Seal Injection may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Chemical Volume and Control System. Reactor Coolant Pump D Seal Injection is isolated, and is not recovered for the Main Control Room fire evacuation event. Reactor Coolant Pump D Thermal Barrier Cooling may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Component Cooling Water System. Reactor Coolant Pump D Thermal Barrier Cooling is isolated, and is not recovered for the Main Control Room fire evacuation event. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;544BBHV8351A - Cable damage (4BBG04AC) to BBHV8351A. Cable damage can spuriously close the Reactor Coolant Pump A Seal Water Supply Isolation Valve, BBHV8351A. This valve is required to close in order to satisfy Westinghouse criteria for maintaining a fixed rate of 21 gpm from the Reactor Coolant Pump Seal Package following the loss of all Reactor Coolant Pump Seal Cooling for greater than 13-minutes. This requirement is based on the latest Westinghouse Owners Group analysis for the performance of the Reactor Coolant Pump Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump A Seal Injection may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Chemical Volume and Control System. Reactor Coolant Pump A Seal Injection is isolated, and is not recovered for the Main Control Room fire evacuation event. Reactor Coolant Pump A Thermal Barrier Cooling may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Component Cooling Water System. Reactor Coolant Pump A Thermal Barrier Cooling is isolated, and is not recovered for the main  fire evacuation event. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-385 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;54BBHV8351B - Cable damage (4BBG04BC) to BBHV8351B. Cable damage can spuriously close the Reactor Coolant Pump B Seal Water Supply Isolation Valve, BBHV8351B. This valve is required to close in order to satisfy Westinghouse criteria for maintaining a fixed rate of 21 gpm from the Reactor Coolant Pump Seal Package following the loss of all Reactor Coolant Pump Seal Cooling for greater than 13-minutes. This requirement is based on the latest Westinghouse Owners Group analysis for the performance of the Reactor Coolant Pump Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump B Seal Injection may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Chemical Volume and Control System. Reactor Coolant Pump B Seal Injection is isolated, and is not recovered for the Main Control Room fire evacuation event. Reactor Coolant Pump B Thermal Barrier Cooling may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Component Cooling Water System. Reactor Coolant Pump B Thermal Barrier Cooling is isolated, and is not recovered for the Main Control Room fire evacuation event. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;54:BBHV8351C - Cable damage (4BBG 04CC) to BBHV8351C. Cable damage can spuriously close the Reactor Coolant Pump C Seal Water Supply Isolation Valve, BBHV8351C. This valve is required to close in order to satisfy Westinghouse criteria for maintaining a fixed rate of 21 gpm from the Reactor Coolant Pump Seal Package following the loss of all Reactor Coolant Pump Seal Cooling for greater than 13-minutes. This requirement is based on the latest Westinghouse Owners Group analysis for the performance of the Reactor Coolant Pump Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump C Seal Injection may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Chemical Volume and Control System. Reactor Coolant Pump C Seal Injection is isolated, and is not recovered for the Main Control Room fire evacuation event. Reactor Coolant Pump C Thermal Barrier Cooling may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Component Cooling Water System. Reactor Coolant Pump C Thermal Barrier Cooling is isolated, and is not recovered for the Main Control Room fire evacuation event. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-386 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;54,BBHV8351D - Cable damage (4BBG 04DC) to BBHV8351D. Cable damage can spuriously close the Reactor Coolant Pump D Seal Water Supply Isolation Valve, BBHV8351D. This valve is required to close in order to satisfy Westinghouse criteria for maintaining a fixed rate of 21 gpm from the Reactor Coolant Pump Seal Package following the loss of all Reactor Coolant Pump Seal Cooling for greater than 13-minutes. This requirement is based on the latest Westinghouse Owners Group analysis for the performance of the Reactor Coolant Pump Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump D Seal Injection may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Chemical Volume and Control System. Reactor Coolant Pump D Seal Injection is isolated, and is not recovered for the Main Control Room fire evacuation event. Reactor Coolant Pump D Thermal Barrier Cooling may be interrupted for greater than 13-minutes or lost in this area due to fire damage potentially affecting the Component Cooling Water System. Reactor Coolant Pump D Thermal Barrier Cooling is isolated, and is not recovered for the Main Control Room fire evacuation event. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;547BBPCV0455A-P - Cable damage (1BBK40AE and 1BBK40AK) to BBPCV0455A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-387 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;54;BBPCV0455B - Cable damage (5BBI19AA) to BBPCV0455B. Cable damage can spuriously open the Reactor Coolant Pump A Pressurizer Spray Line Isolation Valve, BBPCV0455B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3.
This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;546BBPCV0455C - Cable damage (5BBI19BA) to BBPCV0455C. Cable damage can spuriously open the Reactor Coolant Pump B Pressurizer Spray Line Isolation Valve, BBPCV0455C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;54+BBPCV0456A-P - Cable damage (4BBK40BE and 4BBK40BK) to BBPCV0455A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0456A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-388 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;55BGHV8100 - Cable damage (4BGG24AC, 4BGG24AD, and 4BGG24AE) to BGHV8100. Cable damage can spuriously open or close the Seal Water Return Outer Containment Isolation Valve, BGHV8100. The valve is required closed to address the potential for loss of all RCP seal cooling for greater than 13-minutes in accordance with the Westinghouse mitigation criteria for loss of all RCP seal cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5(BGHV8105 - Cable damage (4BGG11AC and 4BGG11AD; and 1BGG11BC and 1BGG11BD) to BGHV8105 and BGHV8106 respectively. For Chemical Volume and Control System Charging Header to Regenerative Heat Exchanger Outer Containment Isolation Valve BGHV8105, cable damage can spuriously open or close the valve, but will not bypass the open/close limit/torque switches. The valve may close in response to a valid or spurious SIS. However, the required position for this valve is closed. The valve may also fail as-is (open). For Chemical Volume and Control System Charging Header to Regenerative Heat Exchanger Outer Containment Isolation Valve, BGHV8106, cable damage can spuriously open or close the valve, but will not bypass the open/close limit/torque switches. The valve may close in response to a valid or spurious SIS. However, the required position for this valve is closed. BGHV8106 may also lose power from NG01B. The valve may also fail as-is (open). Either one of these valves is required closed to prevent potential adverse impact (i.e., flow diversion) to the NFPA 805 NSPC credited flowpath of the Chemical Volume and Control System (i.e., the boron injection flowpath). Either one of these valves is required closed to also mitigate spurious operation of downstream valves BGHV8145 (Pressurizer Auxiliary Spray), BGHV 8146 (Loop 1 Cold Leg Injection), and BGHV8147 (Loop 4 Cold Leg Injection), which are not included in the safe shutdown model, and which could adversely impact the ability to maintain positive control over RCS Inventory and Pressure. This action will also isolate flow from the normal charging pump into the RCS. The RCP seal injection lines will be isolated with manual valves. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-389 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;54BGHV8111 - Cable damage (4BGG11DC, 4BGG11DD, 4BGG11DE, and 4BGG11DF) to BGHV8111. Cable damage can spuriously open or close the Charging Pump B Discharge Miniflow Isolation Valve, BGHV8111. The valve is required open to ensure a recirculation flowpath for Charging Pump B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5BGHV8149A - Cable damage (5BGK35AA, 5B GK35AB, and 5BGK35AD) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5:BGHV8149B - Cable damage (5BGK35BA, 5B GK35BB, and 5BGK35BD) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-390 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5,BGHV8149C - Cable damage (5BGK35CA, 5BGK35CB, and 5BGK35CD) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;57BGHV8153A-P - Cable damage (1BGK48CB and 1BGK48DB) to BGHV8153A and BGHV8154A respectively. Cable damage can spuriously open or prevent closure of the Reactor Coolant System to Chemical Volume Control System Excess Letdown Downstream and Upstream Isolation Protection A Valves, BGHV8153A an d BGHV8154A. At least one of these two valves is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5;BGHV8153B-P - Cable damage (4BGK48AB and 4BGK48BB) to BGHV8153B and BGHV8154B respectively. Cable damage can spuriously open or prevent closure of the Reactor Coolant System to Chemical Volume Control System Excess Letdown Downstream and Upstream Isolation Protection B Valves, BGHV8153B an d BGHV8154B. At least one of these two valves is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-391 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;56BGLCV0112B-P - Cable damage (1BGG12AC) to BGLCV0112B. Cable damage can spuriously open or close Chemical Volume Control System Volume Control Tank [VCT] Outlet Upstream Isolation Valve, BGLCV0112B. Spurious closure of valve BGLCV0112B due to direct cable damage could isolate the suction flowpath from the VCT to the credited Charging Pump, B, which is normally idle. Main Control Room evacuation procedure OTO-ZZ-00001 prescribes time critical operator actions upon evacuating the Main Control Room to locally secure normally idle Charging Pump B at the associated breaker on 4kV Switchgear NB02 as a precaution, until the charging pump suction flowpath from the Refueling Water Storage Tank is locally aligned (local opening of valve BNLCV0112E), and the charging pump suction flowpath from the Volume Control Tank is isolated (local closing of valve BGLCV0112C). The NFPA 805 NSPC compliance strategy for the Main Control Room (fire area C-27) is predicated upon the 10CFR50 Appendix R design criteria for alternate shutdown capability, which requires the design to ensure that safe shutdown can be achieved and maintained assuming one worst case spurious operation or signal at a time (until the Main Control Room evacuation is complete, and the plant operators have transitioned to the respective local control stations). There is no single spurious operation or signal originating from the Main Control Room that can automatically isolate all of the suction flowpaths for the charging pumps, and also start the safety-related charging pumps. A spurious Safety Injection signal originating from the Main Control Room (fire area C-27) will start Charging Pumps A and B and will isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C), but this signal will also align the Refueling Water Storage Tank (open BNLCV0112D and BNLCV0112E). Loss of offsite power and/or a spurious NB02 undervoltage signal originating from the Main Control Room (fire area C-27) will start Charging Pump B, but this signal will not isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C). A spurious Low-Low Volume Control Tank Level signal and a spurious Boric Dilution Accident Trip signal will isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C) and will also align the Refueling Water Storage Tank (open BNLCV0112D and BNLCV0112E), but this signal will not start the charging pumps. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-392 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5+BGLCV0112C - Cable damage (4BGG12BC and 4BGG12BE) to BGLCV0112C. Cable damage can spuriously open or close Chemical Volume Control System Volume Control Tank [VCT] Outlet Downstream Isolation Valv e, BGLCV0112C. Failure to close on demand (or spurious opening) of valve BGLCV0112C due to direct cable damage could for allow gas to enter the suction flowpath from the VCT to the Credited Charging Pump, B, which is normally idle. Main Control Room evacuation procedure OTO-ZZ-00001 prescribes time critical operator actions upon evacuating the Main Control Room to locally secure normally idle Charging Pump B at the associated breaker on 4kV Switchgear NB02 as a precaution, until the charging pump suction flowpath from the Refueling Water Storage Tank is locally aligned (local opening of valve BNLCV0112E), and the charging pump suction flowpath from the Volume Control Tank is isolated (local closing of valve BGLCV0112C). The NFPA 805 NSPC compliance strategy for the Main Control Room (fire area C-27) is predicated upon the 10CFR50 Appendix R design criteria for alternate shutdown capability, which requires the design to ensure that safe shutdown can be achieved and maintained assuming one worst case spurious operation or signal at a time (until the Main Control Room evacuation is complete, and the plant operators have transitioned to the respective local control stations). There is no single spurious operation or signal originating from the Main Control Room that can automatically isolate all of the suction flowpaths for the charging pumps, and also start the safety-related charging pumps. A spurious Safety Injection signal originating from the Main Control Room (fire area C-27) will start Charging Pumps A and B and will isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C), but this signal will also align the Refueling Water Storage Tank (open BNLCV0112D and BNLCV0112E). Loss of offsite power and/or a spurious NB02 undervoltage signal originating from the Main Control Room (fire area C-27) will start Charging Pump B, but this signal will no t isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C). A spurious Low-Low Volume Control Tank Level signal and a spurious Boric Dilution Accident Trip signal will isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C) and will also align the Refueling Water Storage Tank (op en BNLCV0112D and BNLCV0112E), but this signal will not start the charging pumps. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-393 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5:5BGLCV0112C-P - Cable damage (4BGG12BC) to BGLCV0112C. Cable damage can spuriously open or close Chemical Volume Control System Volume Control Tank [VCT] Outlet Downstream Isolation Valve, BGLCV0112C. Spurious closure of valve BGLCV0112C due to direct cable damage could isolate the suction flowpath from the VCT to the credited Charging Pump, B, which is normally idle. Main Control Room evacuation procedure OTO-ZZ-00001 prescribes time critical operator actions upon evacuating the Main Control Room to locally secure normally idle Charging Pump B at the associated breaker on 4kV Switchgear NB02 as a precaution, until the charging pump suction flowpath from the Refueling Water Storage Tank is locally aligned (local opening of valve BNLCV0112E), and the charging pump suction flowpath from the Volume Control Tank is isolated (local closing of valve BGLCV0112C). The NFPA 805 NSPC compliance strategy for the Main Control Room (fire area C-27) is predicated upon the 10CFR50 Appendix R design criteria for alternate shutdown capability, which requires the design to ensure that safe shutdown can be achieved and maintained assuming one worst case spurious operation or signal at a time (until the Main Control Room evacuation is complete, and the plant operators have transitioned to the respective local control stations). There is no single spurious operation or signal originating from the Main Control Room that can automatically isolate all of the suction flowpaths for the charging pumps, and also start the safety-related charging pumps. A spurious Safety Injection signal originating from the Main Control Room (fire area C-27) will start Charging Pumps A and B and will isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C), but this signal will also align the Refueling Water Storage Tank (open BNLCV0112D and BNLCV0112E). Loss of offsite power and/or a spurious NB02 undervoltage signal originating from the Main Control Room (fire area C-27) will start Charging Pump B, but this signal will not isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C). A spurious Low-Low Volume Control Tank Level signal and a spurious Boric Dilution Accident Trip signal will isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C) and will also align the Refueling Water Storage Tank (open BNLCV0112D and BNLCV0112E), but this signal will not start the charging pumps. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-394 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5:(BMHV0001 - Cable damage (4BMK06AA, 4BMK06AB, and 4BMK06AC) to BMHV0001 (BMHY0001A). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5:4BMHV0002 - Cable damage (4BMK06BA, 4BMK06BB, and 4BMK06BC) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5:BMHV0003 - Cable damage (4BMK06CA, 4BMK06CB, and 4BMK06CC) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator C is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-395 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5::BMHV0004 - Cable damage (4BMK06DA, 4BMK06DB, and 4BMK06DC) to BMHV0004 (BMHY0004A). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown. Note that Steam Generator D is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5:,BNHV8812A - Cable damage (1BNG03AC) to BNHV8812A. Cable damage can fail as-is (open) or spuriously open Refueling Water Storage Tank to RHR Pump A Suction Isolation Valve, BNHV8812A. This valve is required closed to mitigate draindown of the Refueling Water Storage Tank inventory into the containment sump, due to the spurious opening of valve EJHV8811A, resulting from cable damage (1EJG06AC and 1EJG06AG). Refueling Water Storage Tank inventory is required for Reactor Coolant System makeup, to maintain RCS Inventory and Pressure control. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5:7BNHV8812B - Cable damage (4BNG03BC) to BNHV8812B. Cable damage can fail as-is (open) or spuriously open Refueling Water Storage Tank to RHR Pump B Suction Isolation Valve, BNHV8812B. This valve is required closed to mitigate draindown of the Refueling Water Storage Tank inventory into the containment sump, due to the spurious opening of valve EJHV8811B, resulting from cable damage (4EJG06BC and 4EJG06BG). Refueling Water Storage Tank inventory is required for Reactor Coolant System makeup, to maintain RCS Inventory and Pressure control. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-396 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5:;BNLCV0112E - Cable damage (4BNG01BC and 4BNG01BD) to BNLCV0112E. Cable damage can spuriously open or close Charging Pump B suction from Refueling Water Storage Tank [RWST] Isolation Valve, BNLCV0112E. Failure to open on demand (or spurious closure) of valve BNLCV0112E due to direct cable damage could isolate the suction flowpath from the RWST to the credited Charging Pump, B, which is normally idle. Main Control Room evacuation procedure OTO-ZZ-00001 prescribes time critical operator actions upon evacuating the Main Control Room to locally secure normally idle Charging Pump B at the associated breaker on 4kV Switchgear NB02 as a precaution, until the charging pump suction flowpath from the Refueling Water Storage Tank is locally aligned (local opening of valve BNLCV0112E), and the charging pump suction flowpath from the Volume Control Tank is isolated (local closing of valve BGLCV0112C). The NFPA 805 NSPC compliance strategy for the Main Control Room (fire area C-27) is predicated upon the 10CFR50 Appendix R design criteria for alternate shutdown capability, whic h requires the design to ensure that safe shutdown can be achieved and maintained assuming one worst case spurious operation or signal at a time (until the Main Control Room evacuation is complete, and the plant operators have transitioned to the respecti ve local control stations). There is no single spurious operation or signal originating from the Main Control Room that can automatically isolate all of the suction flowpaths for the charging pumps, and also start the safety-related charging pumps. A spurious Safety Injection signal originating from the Main Control Room (fire area C-27) will start Charging Pumps A and B and will isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C), but this signal will also align the Refueling Water Storage Tank (open BNLCV0112D and BNLCV0112E). Loss of offsite power and/or a spurious NB02 undervoltage signal originating from the Main Control Room (fire area C-27) will start Charging Pump B, but this signal will not isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C). A spurious Low-Low Volume Control Tank Level signal and a spurious Boric Dilution Accident Trip signal will isolate the Volume Control Tank (close BGLCV0112B and BGLCV0112C) and will also align the Refueling Water Storage Tank (open BNLCV0112D and BNLCV0112E), but this signal will not start the charging pumps. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-397 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5:6BNLI0930 - Cable damage (1BNI07CA, 1SBS01AC, 1SBS02AC, 1SBS08AB, and 1SBY09CA) to Level Transmitter BNLT0930. Cable damage (2BNI07DA, 2SBS01BB, 2SBS02BB, 2SBY09DA, and 4SBS08AB) to Level Transmitter BNLT0931. Cable damage (3BNI07EA, 1SBS08CB, 3SBS01CA, 3SBS02CA, and 3SBY09EA) to Level Transmitter BNLT0932. Cable damage (4BNI07FA, 4SBS01DA, 4SBS02DA, 4SBS08DB, and 4SBY09FA) to Level Transmitter BNLT0933. Cable damage may result in the loss of all four channels of RWST level instrumentation. At least one channel of RWST level instrumentation is desired for monitoring of RWST level from the alternate shutdown panel to satisfy the NFPA 805 performance goal of process monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issu e.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5:+CEF01B - Cable damage (4EFY02ND, 4EFY02NF, and 4EFY02NG) to CEF01B. Cable damage (4EFI10NA, 4RPY09BA, and 4RPY09CA) to Temperature Element EFTE0068A. Cable damage may prevent proper operation of Essential Service Water Ultimate Heat Sink Cooling Tower Fan B, CEF01B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5,5CEF01D - Cable damage (4EFY02ND, 4EFY02NF, and 4EFY02NG) to CEF01B. Cable damage (4EFI10NA, 4RPY09BA, and 4RPY09CA) to Temperature Element EFTE0068A. Cable damage may prevent proper operation of Essential Service Water Ultimate Heat Sink Cooling Tower Fan D, CEF01D. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-398 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5,(CGD01B - Cable damage (4GDG01BF, 4GDY01BA, 4GDY01BB, 4RPY10BA, and 4GDY01BC) to CGD01B. Cable damage (4GDI05BA, 4GDI05BC, 4GDI05BE, 4RPY09BA, and 4RPY09CA) to Temperature Element GDTE0011. Cable damage may prevent proper operation of Essential Service Water Pump Room Supply Fan B, CGD01B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5,4CGD02B - Cable damage (4GDG02BF, 4GDY02BA, 4RPY10BA, and 4GDY 02BD) to CGD02B. Cable damage (4GDI05DA, 4GDI05DB, 4GDI05DC, 4RPY09BA, and 4RPY09CA) to Temperature Element GDTE0061. Cable damage may prevent proper operation of Cooling Tower Electrical Room Supply Fan B, CGD02B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5,CGM01B - Cable damage (4GMG01BD and 4GMG01BH) to CGM01B. Cable damage may prevent proper operation of Emergency Diesel Generator Ventilation Supply Fan B, CGM01B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-399 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5,:EFHV0026 - Cable damage (4EFG02DC, 4EFG02DD, 4EFG02DE, and 4EFG02DF) to EFHV0026. Cable damage to Service Water to Essential Service Water Train B Downstream Valve, EFHV0026, can spuriously open or close valve. The valve is required closed to support operation of the Essential Service Water System. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5,,EFHV0032 - Cable damage (4EFG07BC) to EFHV0032. Cable damage to Essential Service Water Train B to Containment Air Coolers Outer Containment Valve, EFHV0032, can spuriously open or close valve. The valve is required open to establish Essential Service Water flow for the Train B Containment Air Coolers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5,7EFHV0034 - Cable damage (4EFG09BC) to EFHV0034. Cable damage to Essential Service Water Train B to Containment Air Coolers Inner Containment Valve, EFHV0034, can spuriously open or close valve. The valve is required open to establish Essential Service Water flow for the Train B Containment Air Coolers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-400 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5,;EFHV0038 - Cable damage (4EFG06BC, 4EFG06BD, and 4EFG06BE) to EFHV0038. Cable damage to Essential Service Water Train B to Ultimate Heat Sink Valve, EFHV0038, can spuriously open or close valve. The valve is required open to support operation of the Essential Service Water System. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5,6EFHV0040 - Cable damage (4EFG03BC, 4EFG03BD, and 4EFG03BE) to EFHV0040. Cable damage to Essential Service Water Train B to Service Water Upstream Valve, EFHV0040, can spuriously open or close valve. The valve is required closed to support operation of the Essential Service Water System. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5,+EFHV0046 - Cable damage (4EFG09DC) to EFHV0046. Cable damage to Essential Service Water Train B from Containment Air Coolers Inner Containment Valve, EFHV0046, can spuriously open or close valve. The valve is required open to establish Essential Service Water flow for the Train B Containment Air Coolers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-401 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;575EFHV0050 - Cable damage (4EFG08BC and 4EFG08BD) to EFHV0050. Cable damage to Essential Service Water Train B from Containment Air Coolers Outer Containment Valve, EFHV0050, can spuriously open or close valve. The valve is required open to establish Essential Service Water flow for the Train B Containment Air Coolers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;57(EFHV0052 - Cable damage (4EFG05BC, 4EFG05BD, and 4EFG05BE) to EFHV0052. Cable damage to Essential Service Water Train B to Component Cooling Water Heat Exchanger B Valve, EFHV0052, can spuriously open or close valve. The valve is required open to establish Essential Service Water flow for the Train B Component Cooling Water Heat Exchanger. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;574EFHV0060 - Cable damage (4EFG04BC, 4EFG04BD, and 4EFG04BE) to EFHV0060. Cable damage to Essential Service Water Train B from Component Cooling Water Heat Exchanger B Valve, EFHV0060, can spuriously open or close valve. The valve is required closed to establish required Essential Service Water flow for the Train B Component Cooling Water Heat Exchanger. The valve limit/torque switches for EFHV0060 can be bypassed by fire damage to cables 4EFG04BC, 4EFG04BD, and 4EFG04BE. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-402 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;57EFHV0066 - Cable damage (4EFG05NC, 4EFY07NJ, 4EFY07NK, 4EFY07NL, and 4EFY07NM) to EFHV0066. Cable damage to Temperature Element EFTE0068A (4EFI10NA, 4RPY09BA, and 4RPY09CA). Cable damage to Essential Service Water Ultimate Heat Sink Cooling Tower Train B Bypass Valve, EFHV0066, cannot spuriously open or close valve. Cable damage to Essential Service Water Train B Ultimate Heat Sink Cooling Towers B and D Temperature Element, EFTE0068A, can spuriously open or close valve EFHV0066. The valve is required to open and close to support long term operation of the Essential Service Water System. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;57:EGHV0011 - Cable damage (1EGG04AC and 1EGG04BC) to EGHV0011 and EGHV0013 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train A Upstream Valve, EGHV0011, and Essential Service Water to Component Cooling Train A Downstream Valve, EGHV0013. Cable damage can spuriously open or close these valves with the limit/torque switches bypassed due to cable damage. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from loss of Component Cooling Water inventory into the Train A Essential Service Water header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-403 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;57,EGHV0012 - Cable damage (4EGG04CC and 4EGG04DC) to EGHV0012 and EGHV0014 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train B Upstream Valve, EGHV0012, and Essential Service Water to Component Cooling Train B Downstream Valve, EGHV0014. Cable damage can spuriously open or close these valves with the limit/torque switches bypassed due to cable damage. These credited train valves are required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from loss of Component Cooling Water inventory into the Train B Essential Service Water header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;577EGHV0016 - Cable damage (4EGG05BC) to EGHV0016. Cable damage can spuriously open or close the Component Cooling Water Train B Supply/Return Isolation Valve, EGHV0016. This valve is required open to establish Component Cooling Water flow from the Seal Water Heat Exchanger (for cooling of Charging Pump B recirculation flow). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-404 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;57;EGHV0054 - Cable damage (4EGG05DC) to EGHV0054. Cable damage can spuriously open or close the Component Cooling Water Train B Supply Isolation Valve, EGHV0054. This valve is required open to establish Component Cooling Water flow to the Seal Water Heat Exchanger (for cooling of Charging Pump B recirculation flow). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;576EGHV0070B - Cable damage (1EGK08AA, 1EGK08AC, 1EGK08AD, 1EGK08AE, 1EGK08BA, 1EGK08BC, 1EGK08BD, 1EGK08BE, 4EGK08CA, 4E GK08CC, 4EGK08CD, 4EGK08CE
, 4EGK08DA, 4EGK08DC, 4EGK08DD, and 4EGK08DE) to EGHV0069A, EGHV0069B, EGHV0070A, and EGHV0070B respectively. Cable damage can spuriously open the Component Cooling Water to/from Rad Waste Supply/Return Isolation Valves, EGHV0069A, EG HV0070A, EGHV0069B, and EGHV0070B respectively. One of these four valves is required closed to isolate Component Cooling Water flow to the associated non-critical Component Cooling Water loads to prevent potential adverse impact to the credited Train B Component Cooling Water System. The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-405 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;57+EGHV0071 - Cable damage (4EGG05DC) to EGHV0071. Cable damage can spuriously open or close the Component Cooling Water to Containment Outer Isolation Valve, EGHV0071. This valve is required closed to isolate Component Cooling Water flow to the RCP Thermal Barriers. The action to isolate Component Cooling Water flow to the RCP Thermal Barriers is a conservative measure to prevent further RCP seal damage following restoration of Component Cooling Water later in the event (i.e., water hammer, steam flashing, etc.) - isolation of CCW to RCP thermal barrier required to address loss of all RCP seal cooling for a time duration of greater than 13-minutes. All of the Component Cooling Water motor operated valves in the Component Cooling Water flowpath to/from the RCP Thermal Barriers are subject to spurious opening in the fire area C-27 fire event (EGHV0071, EGHV0058, EGHV0126, EGHV0127, EGHV0061, EGHV0062, EGHV0132, and EGHV0133). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.5;5EGHV0126-P - Cable damage (1EGG18AD) to EGHV0126. Cable damage can spuriously open or close the Component Cooling Water to Containment Bypass Isolation Valve, EGHV0126. The valve limit/torque switches for EGHV0126 can be bypassed by fire damage to cable 1EGG18AD. This valve is required closed to isolate Component Cooling Water flow to the RCP Thermal Barriers. The action to isolate Component Cooling Water flow to the RCP Thermal Barriers is a conservative measure to prevent further RCP seal damage following restoration of Component Cooling Water later in the event (i.e., water hammer, steam flashing
, etc.) - isolation of CCW to RCP Thermal Barrier required to address loss of all RCP seal cooling for a time duration of greater than 13-minutes. All of the Component Cooling Water motor operated valves in the Component Cooling Water flowpath to/from the RCP Thermal Barriers are subject to spurious opening in the fire area C-27 fire event (EGHV0071, EGHV0058, EGHV0126, EGHV0127, EGHV0061, EGHV0062, EGHV0132, and EGHV0133). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-406 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5;(EGRV0009 - Cable damage (5EGK03AA and 5EGK03AB) to EGRV0009. Cable damage can spuriously open Component Cooling Water Tank A Surge Tank Vent Valve, EGRV0009. This non-credited train valve is required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5;4EGRV0010 - Cable damage (6EGK03BA and 6EGK03BB) to EGRV0010. Cable damage can spuriously open Component Cooling Water Tank B Surge Tank Vent Valve, EGRV0010. This credited train valve is required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water system provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5;EMHV8801B - Cable damage (4EMG02DC and 4EMG02DD) to EMHV8801B. Boron Injection Header Train B Outlet to Cold Legs Isolation Valve, EMHV8801B, can spuriously open or close due to cable damage (4EMG02DC and 4EMG02DD). The valve may also fail as-is (closed). The desired position for this valve is open to establish the Train B Boron Injection flowpath, which is necessary to restore Pressurizer Level (maintain positive control over RCS Inventory and Pressure) as RCS inventory makeup with RCP seal injection is not available. Note that the valve limit/torque switches for EMHV8801B cannot be bypassed by fire damage to cables 4EMG02DC or 4EMG02DD. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-407 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5;:EMHV8803B - Cable damage (4EMG02BC and 4EMG02BD) to EMHV8803B. Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B, can spuriously open or close due to cable damage (4EMG02BC and 4EMG02BD). The valve may also fail as-is (closed). The desired position for this valve is throttled open to establish the Train B Boron Injection flowpath, which is necessary to restore Pressurizer Level (maintain positive control over RCS Inventory and Pressure) as RCS inventory makeup with RCP seal injection is not available. Note that the valve limit/torque switches for EMHV8803B ca nnot be bypassed by fire damage to cables 4EMG02BC or 4EMG02BD. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5;,EMHV8843 - Cable damage (4EMK04CA) to EMHV8843. Cable damage can spuriously open Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843. The valve is required closed in order to prevent potential diversion of RWST inventory to the SI test lines to ensure positive control over RCS Inventory and Pressure. This condition represents a varianc e from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5;7FEF02B - Cable damage (4EFG03NC) to FEF02B. Cable damage may prevent proper operation of Essential Service Water Self-Cleaning Strainer B, FEF02B. Emergency Service Water Self Cleaning Strainer B is required operable for long term operation of Essential Service Water Pump B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-408 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5;;GDTZ0011A - No cable damage to GDTZ0011A. Cable damage to Temperature Element GDTE0011 (4GDI05BA, 4GDI05BC, 4GDI05BE, 4RPY09BA, and 4RPY09CA). Cable damage can spuriously close ESW Pump Room Supply Fan B Outside Air Damper, GDTZ0011A. ESW Pump Room Supply Fan B Outside Air Damper is required open to support long term operation of Essential Service Water Pump B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5;6GDTZ0061A - No cable damage to GDTZ0061A. Cable damage to Temperature Element GDTE0061 (4GDI05DA, 4GDI05DB, 4GDI05DC, 4RPY09BA, and 4RPY09CA). Cable damage can spuriously close ESW Electrical Equipment Room Supply Fan B Inlet Damper, GDTZ0061A. ESW Electrical Equipment Room Supply Fan B Inlet Damper is required open to support long term operation of the Essential Service Water System. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5;+GDTZ0061B - No cable damage to GDTZ0061B. Cable damage to Temperature Element GDTE0061 (4GDI05DA, 4GDI05DB, 4GDI05DC, 4RPY09BA, and 4RPY09CA). Cable damage can spuriously open ESW Electrical Equipment Room Supply Fan B Recirculation Damper, GDTZ0061B. ESW Electrical Equipment Room Supply Fan B Recirculation Damper is required closed to support long term operation of the Essential Service Water System. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-409 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;565GMHZ0019 - Cable damage (4GMK0 4BB, 4GMK04BD, and 4GMK04BE) to GMHZ0019. Cable damage can spuriously close Diesel Generator Building B Exhaust Damper, GMHZ0019. Diesel Generator Building B Exhaust Damper is required open in support of the recovery of AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;56(GMTZ0011A - No cable damage to GMTZ0011A. Cable damage to Temperature Element GMTE0011 (4GMI02BA, 4GMI02BB, 4GMI02BC, 4RPY09BA, and 4RPY09CA). Cable damage can spuriously close Diesel Generator Building B Suction Outside Air Damper, GMTZ0011A. Diesel Generator Building B Suction Outside Air Damper is required open in support of the recovery of AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;564GMTZ0011B - No cable damage to GMTZ0011B. Cable damage to Temperature Element GMTE0011 (4GMI02BA, 4GMI02BB, 4GMI02BC, 4RPY09BA, and 4RPY09CA). Cable damage can spuriously open Diesel Generator Building B Suction Return Air Damper, GMTZ0011B. Diesel Generator Building B Suction Return Air Damper is required closed in support of the recovery of AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-410 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;56JELI032A - Cable damage (4JEI02PB, 4JEI02PC, 4RPY09BA, and 4RPY09CA) to Level Transmitter JELT0032. Cable damage may cause loss of indication from Emergency Fuel Oil Day Tank B Level Transmitter, JELT0032. Emergency Fuel Oil Day Tank B Level Indication is required operable in support of the recovery of AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;56:JELT0021 - Cable damage (4JEI04BA, 4RPY09BA, and 4RPY09CA) to Level Transmitter JELT0021. Cable damage may prevent proper automatic level control from Emergency Fuel Oil Day Tank B Level Transmitter, JELT0021. Emergency Fuel Oil Day Tank B Level Transmitter is required operable in support of the recovery of AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;56,KJPV0101A - Cable damage (4KJK03AH, 4KJK03AJ, and 4KJK03AK) to KJPV0101A. Cable damage to Diesel Generator B Starting Air Supply Pressure Control Valve A, KJPV0101A, may result in loss of electrical start capability for Emergency Diesel Generator NE02. Diesel Generator B Starting Air Supply Pressure Control Valve A is required operable to support the recovery o f AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-411 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;567KJPV0101B - Cable damage (4KJK03AH, 4KJK03AJ, and 4KJK03AK) to KJPV0101B. Cable damage to Diesel Generator B Starting Air Supply Pressure Control Valve B, KJPV0101B, may result in loss of electrical start capability for Emergency Diesel Generator NE02. Diesel Generator B Starting Air Supply Pressure Control Valve B is required operable to support the recovery o f AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;56;KJPV0108 - Cable damage (4KJK03AH) to KJPV0108. Cable damage may spuriously energize Diesel Generator B Fuel Rack Air Supply Pressure Control Valve, KJPV0108, due to control circuit damage. This could starve the diesel generator of fuel. Diesel Generator B Fuel Rack Air Supply Pressure Control Valve is required deenergized to support of the recovery of AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-412 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;566LSELS-GRP2 - Cable damage (1NFK01DA, 1NFY01BA, 1NFY01FA, 2NFK01DA, 2NFY01DA, 2NFY01FA, 3NFK01DA, 3NFY01FA, 3NFY01HA, 4NBB04AB, 4NBB05AC, 4NBB06AC, 4NEB02 AL, 4NFK01AA, 4NFK01DA
, 4NFK01HA, 4NFK01HB, 4NFK01HC, 4NFK01HD, 4NFK01HE, 4NFK01KA, 4NFY01FA, 4NFY01HA, 4SAK21BA, and 4SAY21BA) to LSELS-GRP2. Cable damage may prevent proper operation of Load Shed Emergency Load Sequencer Group 2, LSELS-GRP2, due to control circuit failures. Local operator actions will be taken to shed and re-energize loads in lieu of proper operation of Load Shed Emergenc y Load Sequencer Group 2 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;56+NB0102-P - Cable damage (1 ENB01AB, 1ENB01AC, and 1ENB01AD) to NB0102. Cable damage may result in spurious closure of the feeder breaker to Containment Spray Pump A (PEN01A), NB0102. Non-credited train Containment Spray Pump A may need to be secured in order to prevent potential diversion of RWST inventory to the containment spray headers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5+5NB0103-P - Cable damage (1EMB01AB and 1EMB01AC) to NB0103. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump A (PEM01A), NB0103. Non-credited train Safety Injection Pump A may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-413 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5+(NB0104-P - Cable damage (1BGB01AB and 1BGB01AE) to NB0104. Cable damage may result in spurious closure of the feeder breaker to Charging Pump A (PBG05A), NB0104. Non-credited train Charging Pump A may need to be secured, or non-credited train valve EMHV8803A may need to be closed in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill), and to minimize heatup of Charging Pump B recirculation flow. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5+4NB0105-P - Cable damage (1ALB01AB, 1ALB01AC, 1ALB01AD, 1ALB01AG, 1ALB01AR, 1ALB01AS, 1ALB01AT, 1ALB01AV, and 1ALB01AW) to NB0105. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A), NB0105. Non-credited train Motor Driven Auxiliary Feedwater Pump A may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5+NB0201 - Cable damage (
4BGB01BB and 4BGB01BE) to NB0201. Cable damage may result in spurious trip of the feeder breaker to Charging Pump B (PBG01B), NB0201. This breaker is required closed/tripped to provide makeup water for Reactor Coolant System Inventory and Pressure Control and long term Reactivity Control. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-414 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5+:NB0201-P - Cable damage (4BGB01BB and 4BGB01BE) to NB0201. Cable damage may result in spurious closure of the feeder breaker to Charging Pump B (PBG05B), NB0201. This breaker is required open for manual load shedding of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5+,NB0202-P - Cable damage (4EMB01BB and 4EMB01BC) to NB0202. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump B (PEM01B), NB0202. This breaker is required open for manual load shedding of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. Additionally, the action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fill lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5+7NB0203-P - Cable damage (4 ENB01BB, 4ENB01BC, and 4ENB01BD) to NB0203. Cable damage may result in spurious closure of the feeder breaker to Containment Spray Pump B (PEN01B), NB0203. This breaker is required open for manual load shedding of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-415 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;5+;NB0204-P - Cable damage (4EJB01BB and 4EJB01BC) to NB0204. Cable damage may result in spurious closure of the feeder breaker to Residual Heat Removal Pump B (PEJ01B), NB0204. This breaker is required open for manual load shedding of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;5+6NB0206 - Cable damage (4EGB01BB, 4EGB01BC, 4EGB01BD, 4EGB01BF, 4EGB01BG, and 4EGB01BK) to NB0206. Cable damage may result in spurious trip of the feeder breaker to Component Cooling Water Pump B (PEG01B), NB0206. This breaker is required closed to provide cooling water for Core Decay Heat Removal (i.e., cooling water for Component Cooling Water Heat Exchanger B and Residual Heat Removal Heat Exchanger B) and to provide oil/seal cooling for Charging Pump B and Residual Heat Removal Pump B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;5++NB0206-P - Cable damage (4EGB01BB, 4EGB01BC, 4EGB01BD, 4EGB01BF, 4EGB01BG, and 4EGB01BK) to NB0206. Cable damage may result in spurious closure of the feeder breaker to Component Cooling Water Pump B (PEG01B), NB0206. This breaker is required open for manual load shedding of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-416 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;(55NB0207 - Cable damage (EG B01DB, 4EGB01DC, 4EGB01DD, 4EGB01DF, and 4EGB01DG) to NB0207. Cable damage may result in spurious trip of the feeder breaker to Component Cooling Water Pump D (PEG01D), NB0207. This breaker is required closed to provide cooling water for Core Decay Heat Removal (i.e., cooling water for Component Cooling Water Heat Exchanger B and Residual Heat Removal Heat Exchanger B) and to provide oil/seal cooling for Charging Pump B and Residual Heat Removal Pump B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(5(NB0207-P - Cable damage (4 EGB01DB, 4EGB01DC, 4EGB01DD, 4EGB01DF, and 4EGB01DG) to NB0207. Cable damage may result in spurious closure of the feeder breaker to Component Cooling Water Pump D (PEG01D), NB0207. This breaker is required open for manual load shedding of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(54NB0208 - Cable damage (
4PGB12AA, 4PGB12 AC, and 4PGB12AD) to NB0208. Cable damage may result in spurious trip of the feeder breaker to Load Center PG22 (Pressurizer Heater Backup Group B), NB0208. This breaker is required closed in support of the recovery of AC electrical power to 480V Load Center PG22 for operation of Pressurizer Heater Backup Group B, to maintain Reactor Coolant System Inventory and Pressure Control. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-417 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;(5NB0208-P - Cable damage (4 PGB12AA, 4PGB12AC, and 4PGB12AD) to NB0208. Cable damage may result in spurious closure of the feeder breaker to Load Center PG22 (Pressurizer Heater Backup Group B), NB0208. This breaker is required open for manual load shedding of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(5:NB0209 - Cable damage (4NBB04AD, 4NBB05AC, 4NBB14AA, 4NBB14AB, 4NBB14AC, 4NBB14AE, 4NBB14AF, and 4NBB14AG) to NB0209. Cable damage may result in failure as-is (closed) or spurious closure of the normal feeder breaker to Essential 4kV Switchgear NB02 (from essential transformer XNB02), NB0209. This breaker is required open for manual alignment of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(5,NB0210 - Cable damage (4NGB10BB) to NB0210. Cable damage may result in spurious trip of the feeder breaker to Load Center NG04, NB0210. This breaker is required closed in support of the recovery of AC electrical power to essential 480V Load Center NG04. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-418 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;(57NB0211 - Cable damage (4NEB11AA, 4NEB11AD, and 4NEB11AL) to NB0211. Cable damage can spuriously trip or close the emergency feeder breaker to essential 4kV Switchgear NB02 (from Emergency Diesel Generator NE02), NB0211. This breaker is required open and then closed for manual alignment of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(5;NB0213 - Cable damage (4NGB10AB) to NB0213. Cable damage may result in spurious trip of the feeder breaker to Load Center NG02, NB0213. This breaker is required closed in support of the recovery of AC electrical power to essential 480V Load Center NG02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(56NB0215 - Cable damage (4EFB01NA, 4EFB01NB, 4EFB01ND, 4EFB01NG, and 4EFB01NH) to NB0215. Cable damage may result in spurious trip of the feeder breaker to Essential Service Water Pump B (PEF01B), NB0215. This breaker is required closed in support of the recovery of Emergency Diesel Generator NE02, and to provide cooling water for Core Decay Heat Removal (i.e., cooling water for Component Cooling Water Heat Exchanger B, cooling water for Containm ent Fan Cooler Units SGN01B and SGN01D, and Steam Generator Inventory Water for Auxiliary Feedwater Pump B and the Turbine Driven Auxiliary Fedwater Pump) and to provide cooling water for Room Coolers SGF02B, SGK05B, SG L10B, SGL11B, SGL1 2B, and SGL15B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issu e.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-419 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;(5+NB0216 - Cable damage (4NGB10SB) to NB0216. Cable damage may result in spurious trip of the feeder breaker to 480V Motor Control Center NG06E (Essential Service Water Pump House Motor Control Center), NB0216. This breaker is required closed in support of the recovery of AC electrical power to essential 480V Motor Control Center NG06E. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;((5NB0217 - Cable damage (4NGB11BB) to NB0217. Cable damage may result in spurious trip of the feeder breaker to Load Center NG08 (Ultimate Heat Sink Cooling Tower Building), NB0217. This breaker is required closed in support of the recovery of AC electrical power to essential 480V Load Center NG08. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(((NB0217-P - Cable damage (4NGB11BB) to NB0217. Cable damage may result in spurious closure of the feeder breaker to Load Center NG08 (Ultimate Heat Sink Cooling Tower), NB0217. This breaker is required open for manual load shedding of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-420 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;((4NE02 - Cable damage (4KJK07AE, 4NEK13AB, 4NEK13AD, 4NEK13AF, and 4NEK13AJ) to NE02. Cable damage may prevent proper operation of Emergency Diesel Generator, NE02, due to loss of 125 VDC power for field flashing, and potential spurious voltage and speed/frequency control signals. Emergency Diesel Generator NE02 is required operable in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;((NG0201 - Cable damage (4NGG11AA) to NG0201. Cable damage may result in spurious trip of the feeder breaker to Load Center NG02, NG0201. This breaker is required closed in support of the recovery of AC electrical power to essential 480V Load Center NG02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;((:NG0401 - Cable damage (4NGG11BA) to NG0401. Cable damage may result in spurious trip of the feeder breaker to Load Center NG04, NG0401. This breaker is required closed in support of the recovery of AC electrical power to essential 480V Load Center NG04. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-421 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;((,PA0107 - Cable damage (5BBA01AB, 5BBA01AJ, 5BBA01AK, 5BBA01AL, and 5BBA01AM) to PA0107. Cable damage may result in spurious closure or failure to trip on demand of the feeder breaker to the Reactor Coolant Pump A (PBB01A), PA0107. Non-credited Reactor Coolant Pump A may need to be secured in order to mitigate spurious pressurizer spray valve opening, establish natural circulation, and ensure positive control over RCS Decay Heat Removal capability. Reactor Coolant Pump seal cooling is affected in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;((7PA0108 - Cable damage (5BBA01BB, 5BBA01BJ, 5BBA01BK, 5BBA01BL, and 5BBA01BM) to PA0108. Cable damage may result in spurious closure or failure to trip on demand of the feeder breaker to the Reactor Coolant Pump B (PBB01B), PA0108. Non-credited Reactor Coolant Pump B may need to be secured in order to mitigate spurious pressurizer spray valve opening, establish natural circulation, and ensure positive control over RCS Decay Heat Removal capability. Reactor Coolant Pump seal cooling is affected in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;((;PA0204 - Cable damage (6BBA01DB, 6BBA01DJ, 6 BBA01DK, 6BBA01DL, an d 6BBA01DM) to PA0204. Cable damage may result in spurious closure or failure to trip on demand of the feeder breaker to the Reactor Coolant Pump D (PBB01D), PA0204. Non-credited Reactor Coolant Pump D may need to be secured in order to establish natural circulation, and ensure positive control over RCS Decay Heat Removal capability. Reactor Coolant Pump seal cooling is affected in this fire area. This conditi on represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-422 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;((6PA0205 - Cable damage (6BBA01CB, 6BBA01CJ, 6 BBA01CK, 6BBA01CL, an d 6BBA01CM) to PA0205. Cable damage may result in spurious closure or failure to trip on demand of the feeder breaker to the Reactor Coolant Pump C (PBB01C), PA0205. Non-credited Reactor Coolant Pump C may need to be secured in order to establish natural circulation, and ensure positive control over RCS Decay Heat Removal capability. Reactor Coolant Pump seal cooling is affected in this fire area. This conditi on represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.84;((+PB0301 - Cable damage (5BG B37AB, 5BGB37BA, and 6BGB37BA) to PB0301. Cable damage may result in spurious closure or failure to trip on demand of the feeder breaker to the Normal Charging Pump (PBG04), PB0301. Non-credited Normal Charging Pump may need to be secured (or flow from the normal charging pump isolated) in order to ensure positive control over RCS Inventory and Pressure (to prevent pressurizer overfill), and to minimize heatup of Charging Pump B recirculation flow. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(45PG2401 - Cable damage (6BBG20AB and 6BBG20AC) to PG2401. Cable damage may result in a spurious close signal or failure to trip on demand of the non-credited train Pressurizer Heater Control Group C Breaker, PG2401, with the loss of remote trip control capability. Non-credited train Pressurizer Heater Control Group C may need to be secured in order to ensure positive control over RCS Pressure (to prevent pressurizer PORV challenge). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-423 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;(4(PJE01B - Cable damage (4JEG01BB and 4JEG01BE) to PJE01B. Cable damage may prevent proper operation of Emergency Fuel Oil System Storage Tank B - Fuel Oil Transfer Pump B, PJE01B, due to control circuit failure. Fuel Oil Transfer Pump B is required operable in support of the recovery of AC electrical power to essential 4kV Switchgear NB02 from Emergency Diesel Generator NE02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(44SGK05B - Cable damage (4GKG13BC, 4GKG13BD, 4GKG13BE, 4GKG13BF, 4GKG13BL, 4SAZ20HA, and 4SAZ20PA) to SGK05B. Cable damage may result in a loss of ventilation from the Train B ESF Switchgear Rooms Air Conditioning Unit, SGK05B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(4SGL15B - Cable damage (4GLG12BC) to SGL15B. Cable damage may result in a loss of ventilation from the Auxiliary Building South Electrical Penetration Room Cooler, SGL15B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-424 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;(4:SGN01B - Cable damage (4GNG02BD, 4GNG02BE, 4GNG02BF, 4GNG02BH, 4GNG02BJ, and 4GNG02BK) to SGN01B. Cable damage may result in a loss of ventilation from Containment Cooler Unit B, SGN01B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.84;(4,SGN01D - Cable damage (4GNG02DD, 4GNG02DE, 4GNG02DF, 4GNG02DH, 4GNG02DJ, and 4GNG02DK) to SGN01D. Cable damage may result in a loss of ventilation from Containment Cooler Unit D, SGN01D. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.(47NB0214 - Cable damage (4NBB20AA AND 4NBB20AB) to NB0214. Cable damage may result in failure as-is (open) or spurious closure of the alternate emergency power system feeder breaker to essential 4kV Switchgear NB02, NB0214. This breaker is required open for manual alignment of Emergency Diesel Generator NE02 in support of the recovery of AC electrical power to essential 4kV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue. This VFDR has been identified based on the alternate shutdown action not being performed a t a "Primary Control Station" as defined in RB1.205, Draft Rev. 1 (Regulatory Position 2.4 (October 2009, ML092881133).A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-425 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area184;(4;NB0212 - Cable damage (4NBB15AA, 4NBB15AB, 4NBB15AC, 4NBB15AE, 4NBB15AF, and 4NBB15AG) to NB0212. Cable damage may result in spurious closure of the alternate feeder breaker to essential 4KV Switchgear NB02 (from essential Transformer XNB01), NB0212. This breaker is required open for manual alignment of Emergency Diesel Generator NB02 in support of the recovery of AC electrical power to Essential 4KV Switchgear NB02. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A risk evaluation with consideration of defense in depth and safety margin has determined that the delta risk associated with this VFDR is sufficiently low so as to require no further action. However, to enhance defense in depth, a recovery action (RA-DID) has been specified. The RA has been demonstrated to be feasible.August 2011 C-426 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1&$.>.?>..A&."&.&Ionization 308Detection NYYN3601NIonization 309Detection NYYNNIonization 319Detection NYYNNIonization 329Detection NYYNNHalonSKC07Suppression NNNNNfor the cable trenches N/ANoneFeatureIonization 308Detection NYYN3603NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 3604N/ANoneSuppression N/ANoneFeatureIonization 308Detection NYYN3605NN/ANoneSuppression N/ANoneFeatureIonization 308Detection NYYN3606NN/ANoneFeatureN/ANoneDetection 3616N/ANoneFeatureAugust 2011 C-427 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-428 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1The Alternate Shutdown Panel for Callaway Plant is RP118B. This is the Primary Control Station for implementation of the Alternate Shutdown Strategy in the event of a fire that requires the evacuation of the Main Control Room. NRC approval for the design of the Alternate Shutdown Panel, and for the overall Alternate Shutdown Strategy to meet the requirements of 10CFR50 Appendix R, Section III.G.3, was provided in NUREG-0830, SER Supplement No. 3, Docket No, STN 50-483, May 1984, and in NUREG-0830, SER Supplement No. 4, Docket No, STN 50-483, October 1984.Activation of the Alternate Shutdown Panel involves the transfer of control from the Main Control Room to RP118B through an operator action to manually position three isolation transfer switches and five control switches which are located on RP118B. Following activation of the Alternate Shutdown Panel, the plant operator is provided with the capability to control and monitor secondary side Decay Heat Removal capability utilizing the Auxiliary Feedwater System, the capability to control Reactor Coolant System pressure, and the capability to monitor critical Reactor Coolant System process parameters which are necessary to verify that natural circulation has been established in the RCS and that it is being successfully maintained thereafter. The NRC approved design for Alternate Shutdown Panel RP118B includes the following specific components and features:*Steam Generator B (2) pressure indication (ABPIC0002B)*Steam Generator B (2) wide range level indication (AELI0502A)*Steam Generator B (2) AFW flow indication (ALFI0003B)*Open control for Steam Supply valve from Steam Generator B (2) to TDAFP (ABHV0005)*Open and close control for Steam Generator B (2) Atmospheric Steam Dump Valve (ABPV0002)*Open and close control for Steam Generator B (2) AFW flow control valve from TDAFP (ALHV0010)*Open and close control for Essential Service Water to suction of MDAFW Pump B (ALHV0030)*Open and close control for Condensate Storage Tank to suction of MDAFW Pump B (ALHV0034)*MDAFW Pump B suction pressure indication (ALPI0024B)
*Trip and close control for MDAFW Pump B breaker (NB0205)*Steam Generator D (4) pressure indication (ABPIC0004B)*Steam Generator D (4) wide range level indication (AELI0504A)*Steam Generator D (4) AFW flow indication from MDAFW Pump B (ALFI0001B)&The only automatic fire suppression system is a Halon System for the nine cable trench/wall chase combinations in the floor and west wall of Fire Zone 3601. Halon system actuations are not expected to adversely affect electrical equipment. Any fire can be extinguished manually with the portable extinguishers and/or hose stations after high-voltage equipment is de-energized. The water associated with manual fire suppression will drain out doors and as such  standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-429LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-27Control Room Area1*Open and close control for Steam Generator D (4) Atmospheric Steam Dump Valve (ABPV0004)*Open and close control for Steam Generator D (4) AFW flow control valve from MDAFW Pump B (ALHV0005)*Open and close control for Essential Service Water to suction of TDAFP (ALHV0033)*TDAFP suction pressure indication (ALPI0026B)*Open and close control for TDAFP Governor Control valve (FCFV0313)*Open and close control for TDAFP Trip and Throttle valve (FCHV0312)
*Pressurizer level indication (BBLI0460B)
*Reactor Coolant System pressure indication (BBPI0406X)*Reactor Coolant System Loop 2 cold leg temperature indication (BBTI0423X)*Reactor Coolant System Loop 4 hot leg temperature indication (BBTI0443A)*Intermediate and source range neutron monitoring indication (SENI0061X and SENI0061Y)*Trip and close control for Pressurizer Backup Group B breaker (PG2201)VFDRs have been identified based on the alternate shutdown action not being performed at the Primary Control Station as defined in RG 1.205, Rev. 1 (Regulatory Position 2.4 (December 2009, ML092730314).Non-VFDRs have been identified based on the alternate shutdown action being performed at a Primary Control Station as defined in RG 1.205, Rev. 1 (Regulatory Position 2.4 (December 2009, ML092730314).August 2011 C-430 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-28Control Room Service Area1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and II Condensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IV4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A Fire ZoneDescription3602Pantry3607Restroom3608Janitors ClosetAugust 2011 C-431LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-28Control Room Service Area1RCS Inventory ControlCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage, (specifically no detection in Fire Zone 3607), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-432LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-28Control Room Service Area1....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&.%+5;47Keeping Fire Door DSK36021 between the Control Room (Fire Zone 3601, Fire Area C-27) and the Control Room Pantry (Fire Zone 3602, Fire Area C-28) in the held-open position is acceptable based on the fact that the door is in the direct view of Control Room operators, the Control Room is continuously manned, and a heat detector is installed in Fire Zone 3602.&None8August 2011 C-433LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-28Control Room Service Area1&$.>.?>..A&."&.&Thermal308Detection NNNN3602YN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 3607N/ANoneSuppression N/ANoneFeatureIonization 308Detection NNNN3608NN/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-434 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-28Control Room Service Area1None&This area contains no cabling or electrically supervised equipment that is required for post fire safe shutdown: therefore, fire suppression activities will not adversely affect the plant's ability to achieve the nuclear safety performance criteria.August 2011 C-435 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-29SAS Room, Control Building, El. 20471~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription3609SAS RoomAugust 2011 C-436LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-29SAS Room, Control Building, El. 20471Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&NoneAugust 2011 C-437 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-29SAS Room, Control Building, El. 20471&$.>.?>..A&."&.&Ionization 308Detection NNNN3609NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThis area contains no cabling or electrically supervised equipment that is required for post fire safe shutdown: therefore, fire suppression activities will not adversely affect the plant's ability to achieve the nuclear safety performance criteria.August 2011 C-438 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-30Control Building Cable Chase B, Control Build ing, El. 20471~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel ISteam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel III Aux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel I Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channel IContainment Pressure Channels I, II, and IIICore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.See VFDR No. C-30-002, C-30-009, C-30-010, C-30-011, C-30-012, C-30-013, and C-30-0154.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-30-001, C-30-003, C-30-004, C-30-005, C-30-006, and C-30-014 Fire ZoneDescription3617Electrical Chase (South)August 2011 C-439LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-30Control Building Cable Chase B, Control Build ing, El. 20471Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. C-30-007 and C-30-008Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-440 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-30Control Building Cable Chase B, Control Build ing, El. 2047188555(ABPV0004-P - Cable damage (4 ABI20HE, 4ABI20HG, 4ABI20HH, 4ABI20HJ, 4ABI 20HK, 4ABI20HL, 4 ABI20HM, 4ABI20HN, 4RPY09BA, and 4RPY09CA) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85554BBPCV0456A-P - Cable damage (4BBK40BK) to BBPCV0456A; cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0456A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8555BMHV0001 - Cable damage (4BMK06AA and 4BMK06AC) to BMHV0001 (BMHY0001A). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown.
Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance fro m the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-441 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-30Control Building Cable Chase B, Control Build ing, El. 204718555:BMHV0002 - Cable damage (4BMK06BA and 4BMK06BC) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown.
Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from th e deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8555,BMHV0003 - Cable damage (4BMK06CA and 4BMK06CC) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown.
Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from th e deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.85557BMHV0004 - Cable damage (4BMK06DA and 4BMK06DC) to BMHV0004 (BMHY0004A). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown.
Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance fro m the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-442 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-30Control Building Cable Chase B, Control Build ing, El. 204718555;EGHV0012 - Cable damage (4EGG04CC and 4EGG04DC) to EGHV0012 and EGHV0014 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train B Upstream Valve, EGHV0012, and Essential Service Water to Component Cooling Train B Downstream Valve, EGHV0014. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from loss of component cooling water inventory into the Train B Essential Service Water Header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85556EGRV0010 - Cable damage (6EGK03BA and 6EGK03BB) to EGRV0010. Cable damage can spuriously open Component Cooling Water Tank B Surge Tank Vent Valve, EGRV0010 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8555+EMHV8803B-P - Cable damage (4EMG02BC) to EMHV8803B. Cable damage can spuriously open or close the Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B. This non-credited train valve may need to be closed, or non-credited train Charging Pump B may need to be secured in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-443 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-30Control Building Cable Chase B, Control Build ing, El. 20471855(5EMHV8843 - Cable damage (4EMK04CA) to EMHV8843. Cable damage can spuriously open Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843. The valve is required closed in order to prevent potential diversion of RWST inventory to the SI test lines to ensure positive control over RCS inventory and pressure. This condition represents a varianc e from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.855((NB0201-P - Cable damage (4BGB01BB) to NB0201. Cable damage may result in spurious closure of the feeder breaker to Charging Pump B (PBG05B), NB0201. Non-credited train Charging Pump B may need to be secured, or non-credited train valve EMHV8803B may need to be closed in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.855(4NB0202-P - Cable damage (4EMB01BB) to NB0202. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump B (PEM01B), NB0202. Non-credited train Safety Injection Pump B may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fi ll lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-444 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-30Control Building Cable Chase B, Control Build ing, El. 20471855(NB0203-P - Cable damage (4ENB01BB and 4ENB01BD) to NB0203. Cable damage may result in spurious closure of the feeder breaker to Containment Spray Pump B (PEN01B), NB0203. Non-credited train Containment Spray Pump B may need to be secured in order to prevent potential diversion of RWST inventory to the containment spray headers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in the fire area. The RA has been demonstrated t o be feasible. Reliability is addressed within the FPRA using HRA methods.855(:NB0205-P - Cable damage (4ALB01BD, 4ALB01BM, 4ALB01BN, 4RPK15AA, 4ALB01BH, 4ALB01B1, 4ALB01B2, 4ALB01BG, 4ALB01BP, and 4ALB01BR) TO NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represe nts a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.855(,PG2401 - Cable damage (6BBG20AB and 6BBG20AC) TO PG2401. Cable damage may result in a spurious close signal to the non-credited train Pressurizer Heater Control Group C Breaker, PG2401, with the loss of remote trip control capability. Non-credited train Pressurizer Heater Control Group C may need to be secured in order to ensure positive control over RCS pressure (to prevent pressurizer PORV challenge). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-445 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-30Control Building Cable Chase B, Control Build ing, El. 20471855(7BBHV8141C - Cable damage (6BBK05CA) to BBHV8141C. Cable damage can spuriously close the Reactor Coolant Pump "C" Seal #1 Water Outlet Isolation Valve, BBHV8141C (spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to maintain the effectiveness of Thermal Barrier Cooling for the Reactor Coolant Pump Seal Package. This requirement is based on the latest Westinghouse Owners Group Analysis for the performance of the Reactor Coolant Pumps Seal Package following the loss of Seal Cooling. Note that Reactor Coolant Pump Seal Injection may be lost in this area due to fire damage potentially spuriously closing the Reactor Coolant Pump C Seal Water Supply Isolation Valve, BBHV8351C (Cable 4BBG04CA, 4BBG04CB, and 4BBG04CC). With the exception of spurious closure of BBHV8141C, Component Cooling Water for Reactor Coolant Pump D Thermal Barrier can be recovered from the Main Control Room in this area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.5(;BBHV8141D - Cable damage (6BBK05DA) to BBHV8141D. Cable damage can spuriously close the Reactor Coolant Pump "D" Seal # 1 Water Outlet Isolation Valve, BHV8141D (spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to maintain the effectiveness of Thermal barrier Cooling for the Reactor Coolant Pump Seal Package. This requirement is based on the latest Westinghouse Owners Group Analysis for the performance of the Reactor Coolant Pumps Seal package following the loss of seal cooling. Note that Reactor Coolant Pump Seal Injection may be lost in th e area due to fire damage potentially spuriously closing the Reactor Coolant Pump D Seal Water Supply Isolation Valve, BBHV8351D (Cable 4BBG04DA, 4BBG04DB, and 4BBG04DC). With the exception of spurious closure of BBHV8141D, Component Cooling Water for Reactor Coolant Pump D Thermal barrier can be recovered from the Main Control Room in this area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separati on issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-446 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-30Control Building Cable Chase B, Control Build ing, El. 20471&$.>.?>..A&."&.&Ionization 308Detection NNYN3617NWet PipeSKC37Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-447 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-31Control Building Cable Chase A, Control Build ing, El. 20471~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel IAux. Feedwater Flow to Steam Gen. A Channels I and IV RCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVAux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and IICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel II Volume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent 4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.See VFDR No. C-31-001 Fire ZoneDescription3618Electrical Chase (North)August 2011 C-448LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-31Control Building Cable Chase A, Control Build ing, El. 20471flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. C-31-002RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-449 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-31Control Building Cable Chase A, Control Build ing, El. 2047188(55(ABPV0001-P - Cable damage (1ABI20EE) to Pressure Transmitter ABPT0001. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0001. The valve is required closed to isolate the main steam pressure boundary for Steam Generator A, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator A is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.554EGHV0011 - Cable damage (1EGG04AC and 1EGG04BC) to EGHV0011 and EGHV0013 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train A Upstream Valve, EGHV0011, and Essential Service Water to Component Cooling Train A Downstream Valve, EGHV0013. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train B Component Cooling Water System (from loss of component cooling water inventory into the Train A Essential Service Water Header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-450 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-31Control Building Cable Chase A, Control Build ing, El. 20471&$.>.?>..A&."&.&Ionization 308Detection NNNN3618NWet PipeSKC37Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-451 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-32Control Building Cable Chase B at column C-6, Control Building, El. 20471~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels I, III, and IV Core Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-32-001 Fire ZoneDescriptionC32Electrical Chase (South)August 2011 C-452LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-32Control Building Cable Chase B at column C-6, Control Building, El. 20471Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-453 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-32Control Building Cable Chase B at column C-6, Control Building, El. 2047188455(ABPV0002-P - Cable damage (2ABI20FJ, 2ABI20FL, 2ABI20FM, 2ABI20FN, 2RPY09AA, and 2RPY09BA) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate o f RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.&$.>.?>..A&."&.&Ionization 308Detection NNNNC32Nroom has no number so the fire area is usedWet PipeSKC37Suppression NNNNNroom has no number so the fire area is used N/ANoneFeatureroom has no number so the fire area is used SLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-454 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-32Control Building Cable Chase B at column C-6, Control Building, El. 20471None&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-455 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-33Control Building Cable Chase B, Control Building, El. 2073-61~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel IPressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel ISteam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel III Aux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel I Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channel IContainment Pressure Channels I, II, and IIICore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.See VFDR No. C-33-002, C-33-009, C-33-010, C-33-011, C-33-012, C-33-013, and C-33-0154.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-33-001, C-33-003, C-33-004, C-33-005, C-33-006, and C-33-014 Fire ZoneDescription3804Electrical Chase (South)August 2011 C-456LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-33Control Building Cable Chase B, Control Building, El. 2073-61Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. C-33-007 and C-33-008Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&August 2011 C-457 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-33Control Building Cable Chase B, Control Building, El. 2073-618855(ABPV0004-P - Cable damage (4ABI20HJ, 4ABI20HL, 4 ABI20HM, 4ABI20HN, 4RPY09BA, and 4RPY09CA) to Pressure Transmitter ABPT0004. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0004. The valve is required closed to isolate the main steam pressure boundary for Steam Generator D, to maintain positive control over the rate o f RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8554BBPCV0456A-P - Cable damage (4BBK40BK) to BBPCV0456A; cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0456A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.855BMHV0001 - Cable damage (4BMK06AA and 4BMK06AC) to BMHV0001 (BMHY0001A). Cable damage can spuriously open Steam Generator A Blowdown Isolation Valve BMHV0001, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator A inventory to maintain positive control over the rate of RCS cooldown.
Note that Steam Generator A is not credited for Decay Heat Removal in this fire area. This condition represents a variance fro m the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-458 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-33Control Building Cable Chase B, Control Building, El. 2073-61855:BMHV0002 - Cable damage (4BMK06BA and 4BMK06BC) to BMHV0002 (BMHY0002A). Cable damage can spuriously open Steam Generator B Blowdown Isolation Valve BMHV0002, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator B inventory to maintain positive control over the rate of RCS cooldown.
Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from th e deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.855,BMHV0003 - Cable damage (4BMK06CA and 4BMK06CC) to BMHV0003 (BMHY0003A). Cable damage can spuriously open Steam Generator C Blowdown Isolation Valve BMHV0003, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator C inventory to maintain positive control over the rate of RCS cooldown.
Note that Steam Generator C is credited for Decay Heat Removal in this fire area. This condition represents a variance from th e deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.8557BMHV0004 - Cable damage (4BMK06DA and 4BMK06DC) to BMHV0004 (BMHY0004A). Cable damage can spuriously open Steam Generator D Blowdown Isolation Valve BMHV0004, or can prevent the valve from closing on demand. The valve is required closed to prevent diversion of Steam Generator D inventory to maintain positive control over the rate of RCS cooldown.
Note that Steam Generator D is not credited for Decay Heat Removal in this fire area. This condition represents a variance fro m the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-459 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-33Control Building Cable Chase B, Control Building, El. 2073-61855;EGHV0012 - Cable damage (4EGG04CC and 4EGG04DC) to EGHV0012 and EGHV0014 respectively. Cable damage can spuriously open Essential Service Water to Component Cooling Train B Upstream Valve, EGHV0012, and Essential Service Water to Component Cooling Train B Downstream Valve, EGHV0014. These non-credited train valves are required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from loss of component cooling water inventory into the Train B Essential Service Water Header). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8556EGRV0010 - Cable damage (6EGK03BA and 6EGK03BB) to EGRV0010. Cable damage can spuriously open Component Cooling Water Tank B Surge Tank Vent Valve, EGRV0010 (spurious opening is only credible assuming external hot shorts). This non-credited train valve is required closed to prevent potential adverse impact to the credited Train A Component Cooling Water System (from fluctuation of Component Cooling Water Surge Tank level and pressure). The Component Cooling Water System provides a support function for the other NFPA 805 Nuclear Safety Performance Goals of RCS Inventory and Pressure Control, and Core Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.855+EMHV8803B-P - Cable damage (4EMG02BC) to EMHV8803B. Cable damage can spuriously open or close the Boron Injection Header Supply from Charging Pump B Isolation Valve, EMHV8803B. This non-credited train valve may need to be closed, or non-credited train Charging Pump B may need to be secured in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-460 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-33Control Building Cable Chase B, Control Building, El. 2073-6185(5EMHV8843 - Cable damage (4EMK04CA) to EMHV8843. Cable damage can spuriously open Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843. The valve is required closed in order to prevent potential diversion of RWST inventory to the SI test lines to ensure positive control over RCS inventory and pressure. This condition represents a varianc e from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85((NB0201-P - Cable damage (4BGB01BB) to NB0201. Cable damage may result in spurious closure of the feeder breaker to Charging Pump B (PBG05B), NB0201. Non-credited train Charging Pump B may need to be secured, or non-credited train valve EMHV8803B may need to be closed in order to ensure positive control over RCS inventory and pressure (to prevent pressurizer overfill). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85(4NB0202-P - Cable damage (4EMB01BB) to NB0202. Cable damage may result in spurious closure of the feeder breaker to Safety Injection Pump B (PEM01B), NB0202. Non-credited train Safety Injection Pump B may need to be secured in order to prevent potential diversion of RWST inventory. The action to secure flow from the Safety Injection Pump is a conservative measure taken to mitigate the potential for pumped RWST flow diversion through the SIS test lines and/or the SIS accumulator fi ll lines, which have not been fully analyzed for all of the possible spurious valve operations. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-461 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-33Control Building Cable Chase B, Control Building, El. 2073-6185(NB0203-P - Cable damage (4ENB01BB and 4ENB01BD) to NB0203. Cable damage may result in spurious closure of the feeder breaker to Containment Spray Pump B (PEN01B), NB0203. Non-credited train Containment Spray Pump B may need to be secured in order to prevent potential diversion of RWST inventory to the containment spray headers. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.85(:NB0205-P - Cable damage (4ALB01BD, 4ALB01BH, 4ALB01B1, 4ALB01B2, 4ALB01BG, 4ALB01BP, and 4ALB01BR) to NB0205. Cable damage may result in spurious closure of the feeder breaker to Motor Driven Auxiliary Feedwater Pump B (PAL01B), NB0205. Non-credited train Motor Driven Auxiliary Feedwater Pump B may need to be secured in order to ensure positive control over the rate of RCS cooldown, and to maintain sub-cooling. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85(,PG2401 - Cable damage (6BBG20AB and 6BBG20AC) to PG2401. Cable damage may result in a spurious close signal to the non-credited train Pressurizer Heater Control Group C Breaker, PG2401, with the loss of remote trip control capability. Non-credited train Pressurizer Heater Control Group C may need to be secured in order to ensure positive control over RCS pressure (to prevent pressurizer PORV challenge). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-462 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-33Control Building Cable Chase B, Control Building, El. 2073-6185(7BBHV8141C - Cable damage (6BBK05CA) to BBHV8141C. Cable damage can spuriously close the Reactor Coolant Pump "C" Seal #1 Water Outlet Isolation Valve, BBHV8141C (Spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to maintain the effectiveness of the Thermal Barrier Cooling for the Reactor Coolant Pump Seal Package. This requirement is based on the latest Westinghouse Owners Group Analysis for the performance of the Reactor Coolant Pumps Seal Package following the loss of Seal Cooling. Note that Reactor Coolant Pumps Seal Package following the loss of Seal Cooling. Note that Reactor Coolant Pump Seal Injection may be lost in this area due to fire damage potentially spuriously closing the Reactor Coolant Pump C Seal Water Supply Isolation Valve, BBHV8351C (Cable 4BBG04CC). With the exception of spurious closure of BBHV8141C, Component Cooling Water for Reactor Coolant Pump D Thermal Barrier can be recovered from the Main Control Room in this area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.5(;BBHV8141D- Cable damage (6BBK05DA) to BBHV8141D. Cable damage can spuriously close the Reactor Coolant Pump D Seal #1 Water Outlet Isolation Valve, BBHV8141D (spurious closure is only credible assuming external hot shorts). This valve i s required to remain open in order to maintain the effectiveness of Thermal Barrier Cooling for the Reactor Coolant Pump Seal Package. This requirement is based on the latest Westinghouse Owners Group Analysis for the performance of the Reactor Coolant Pumps Seal Package following the loss of seal cooling. Note that Reactor Coolant Pump Seal Injection may be lost in this area due to fire damage potentially spuriously closing the Reactor Coolant Pump D Seal Water Supply Isolation Valve, BBHV8351D (Cable 4BBG04DC). With the exception of spurious closure of BBHV814D, Component Cooling Water for Reactor Coolant Pump D Thermal Barrier can be recovered from the main Control Room in this area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-463 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-33Control Building Cable Chase B, Control Building, El. 2073-61&$.>.?>..A&."&.&Ionization 308Detection NNYN3804NWet PipeSKC37Suppression NNYNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-464 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-34Control Building Cable Chase B at column C-6, Control Building, El. 2073-61~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels I, III, and IV Core Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescriptionC34Electrical Chase (South)August 2011 C-465LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-34Control Building Cable Chase B at column C-6, Control Building, El. 2073-61Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0None....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-466 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-34Control Building Cable Chase B at column C-6, Control Building, El. 2073-61&$.>.?>..A&."&.&Ionization 308Detection NNNNC34Nroom has no number so the fire area is usedWet PipeSKC37Suppression NNNNNroom has no number so the fire area is used N/ANoneFeatureroom has no number so the fire area is used SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-467 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-35Control Building Corridor, Control Building, El. 20161~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel I RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel I Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription3401Corridor No. 13406Corridor No. 23412Emergency Shower and Eyewash AreaAugust 2011 C-468LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-35Control Building Corridor, Control Building, El. 20161Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.
HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.See VFDR No. C-35-001(BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-469 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-35Control Building Corridor, Control Building, El. 20161%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.
: 2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zones 3401 and 3412), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-470LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-35Control Building Corridor, Control Building, El. 20161....3.%654(5An excessive gap in the bottom of Door DSK32014 connecting Fire Areas C-6 and C-35 is acceptable based on the lack of intervening combustibles at/near the location of the door. The door leads to the stairwell in Fire Area C-6 and, since it is a stairwell, no transient combustibles are expected near this doorway. Therefore DSK32014 is considered a non-rated feature commensurate with the fire hazards in the two areas and it provides an equivalent level of protection as a 3 hour rated fire door by prohibiting the propagation of fire between the two fire areas&.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&88,55(SGK05A - Cable damage (1GKG13AA, 1GKG13AB, 1GKG13AH, 1GKG13AJ, and 1SAZ19KA) to SGK05A. Mechanical thermal link fire dampers GKD0057, GKD0065, GKD0161, and GKD0162 may trip closed in response to a fire in area C-35. These cable failures and/or fire damper actuation(s) may result in a loss of ventilation from the Train A ESF Switchgear Rooms Air Conditioning Unit, SGK05A. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-471 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-35Control Building Corridor, Control Building, El. 20161&$.>.?>..A&."&.&
N/ANoneDetection 3401N/ANoneSuppression ERFBSNoneFeatureN/ANoneDetection 3406N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 3412N/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-472 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-35Control Building Corridor, Control Building, El. 20161None&There are no automatic fire suppression systems in the fire area. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-473 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-36Control Building Cable Chase B at column C-6, Control Building, El. 20001~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-hot Temperature Channels II and V RCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel III Steam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and II Condensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.See VFDR No. C-36-001 Fire ZoneDescriptionC36Cable Chase at column line C-6August 2011 C-474LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-36Control Building Cable Chase B at column C-6, Control Building, El. 20001Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-475 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-36Control Building Cable Chase B at column C-6, Control Building, El. 20001....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&855(ABPV0002-P - Cable damage (2ABI20FE, 2ABI20FG, 2ABI20FH, 2ABI20FJ, 2ABI20FK, 2ABI20FL, 2ABI20FM, and 2ABI20FN) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over t he rate of RCS cooldown, and to maintain RCS sub-cooling. Note that Steam Generator B is credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-476 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-36Control Building Cable Chase B at column C-6, Control Building, El. 20001&$.>.?>..A&."&.&IonizationNoneDetection NNNNC36Nroom has no number so the fire area is usedWet PipeSKC37Suppression NNNNNroom has no number so the fire area is used N/ANoneFeatureroom has no number so the fire area is used SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-477 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-37Control Building Cable Chase A at column  C-3,  Control Building, El. 20001~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel I Aux. Feedwater Flow to Steam Gen. A Channels I and IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VI Steam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channels I and IV Aux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescriptionC37Electrical Chase (North)August 2011 C-478LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-37Control Building Cable Chase A at column  C-3,  Control Building, El. 20001Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-479 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-37Control Building Cable Chase A at column  C-3,  Control Building, El. 20001....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-480 Ameren MissouriCallaway Plant NFPA 805 Transition ReportC-37Control Building Cable Chase A at column  C-3,  Control Building, El. 20001&$.>.?>..A&."&.&
N/ANoneDetection C37room has no number so the fire area is usedWet PipeSKC37Suppression NNNNNroom has no number so the fire area is used N/ANoneFeatureroom has no number so the fire area is used SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Safety related electrical cable in tray is qualified for water exposure. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-481 Ameren MissouriCallaway Plant NFPA 805 Transition ReportD-1Diesel Generator A, Diesel Generator Building, El. 20001~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VI Steam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription5203Diesel Generator Room AAugust 2011 C-482LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportD-1Diesel Generator A, Diesel Generator Building, El. 20001Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite power to NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, justifying the diesel fuel oil day tank containment dikes  was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Fuel tank and all piping are seismic Category I.
: 2. The fuel oil system is a gravity-feed-type system, therefore, no pressurized sprays will occur as a result of a leak.3. Floor adjacent to the dike has floor drains.4. The day tank is provided with level indication that alarms in the Control Room if there are more than 3-gallons of leakage.&This deviation is active per Sections D.1.2 and D.2.2 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid, except for the dike capacity, which is less than 100%. The clarification is being requested in the License Amendment Request Transition Report, Attachment T.August 2011 C-483 Ameren MissouriCallaway Plant NFPA 805 Transition ReportD-1Diesel Generator A, Diesel Generator Building, El. 20001....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-484 Ameren MissouriCallaway Plant NFPA 805 Transition ReportD-1Diesel Generator A, Diesel Generator Building, El. 20001&$.>.?>..A&."&.&Flame500Detection NNYN5203NThermal503Detection NNNNNPre-action SKC26Suppression NNNNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-485 Ameren MissouriCallaway Plant NFPA 805 Transition ReportD-2Diesel Generator B, Diesel Generator Building, El. 20001~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel I Containment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from 4.2.3.2 - Deterministic ApproachRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription5201Diesel Generator Room BAugust 2011 C-486LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportD-2Diesel Generator B, Diesel Generator Building, El. 20001the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, justifying the diesel fuel oil day tank containment dikes  was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Fuel tank and all piping are seismic Category I.
: 2. The fuel oil system is a gravity-feed-type system, therefore, no pressurized sprays will occur as a result of a leak.3. Floor adjacent to the dike has floor drains.4. The day tank is provided with level indication that alarms in the Control Room if there are more than 3-gallons of leakage.&This deviation is active per Sections D.1.2 and D.2.2 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid, except for the dike capacity, which is less than 100%. The clarification is being requested in the License Amendment Request Transition Report, Attachment T.August 2011 C-487 Ameren MissouriCallaway Plant NFPA 805 Transition ReportD-2Diesel Generator B, Diesel Generator Building, El. 20001....3.%55+5(Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.&None8August 2011 C-488 Ameren MissouriCallaway Plant NFPA 805 Transition ReportD-2Diesel Generator B, Diesel Generator Building, El. 20001&$.>.?>..A&."&.&Flame501Detection NNYN5201NThermal502Detection NNNNNPre-action SKC27Suppression NNNNNN/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThe effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-489 Ameren MissouriCallaway Plant NFPA 805 Transition ReportFB-1Fuel Handling Building1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel IISee VFDR No. FB-1-0014.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription6101Stair F-16102Laydown Area6103Cask Loading Pool6104Fuel Pool Cooling Heat Exchanger Room6105Fuel Pool Cooling Heat Exchanger Room6106Spent Fuel Pool and Storage Racks6201Passage6202Electrical Equipment Room6203Air Handling Equipment Room6204Cask Washdown Pit 6205Fuel Transfer Canal 6210New Fuel Storage Area6301General Floor Area6302Laydown Area6303Exhaust Filter Adsorber Room 'B'6304Exhaust Filter Adsorber Room 'A'August 2011 C-490LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportFB-1Fuel Handling Building1RCS Inventory ControlSteam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and II Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Local Mechanical InstrumentVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-491 Ameren MissouriCallaway Plant NFPA 805 Transition ReportFB-1Fuel Handling Building1%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated fuel transfer tube connecting reactor containment and the fuel building, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following: 
: 1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.
: 3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area suppression coverage (specifically partial suppression in Fire Zone 6101 and no suppression in Fire Zones 6102, 6103, 6106, 6201, 6204, 6205, 6210, 6301, and 6302), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zones 6101, 6201, and 6210), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-492LIC-25LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportFB-1Fuel Handling Building1%~&',CDeviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for plant areas without full-area suppression and detection coverage, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following: 
: 1. Low fuel loading.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&',Deviation submitted per 6/29/1981 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, providing justification for a non-rated cover on the trench connecting the fuel building and radwaste tunnel, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Combustibles in this area are separated by more than 50 feet.
: 2. Low combustible loading.3. The trench opening in this room is closed by a heavy steel cover plate approximately 4-feet x 8-feet.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.None....3August 2011 C-493 Ameren MissouriCallaway Plant NFPA 805 Transition ReportFB-1Fuel Handling Building1885(BNLI0930 - Cable damage (1BNI07CA) to Level Transmitter BNLT0930. Cable damage (2BNI07DA) to Level Transmitter BNLT0931. Cable damage (3BNI07EA) to Level Transmitter BNLT0932. Cable damage (4BNI07FA) to Level Transmitter BNLT0933. Cable damage may result in the loss of all four channels of RWST level instrumentation. At least one channel of RWST level instrumentation is desired for monitoring of RWST level from the Main Control Room to satisfy the NFPA 805 Performance Goal of Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-494 Ameren MissouriCallaway Plant NFPA 805 Transition ReportFB-1Fuel Handling Building1&$.>.?>..A&."&.&
N/ANoneDetection 6101N/ANoneSuppression N/ANoneFeatureThermal600Detection NNNN6102NPre-action SKC25Suppression NNNNNRailroad bay/lay down area N/ANoneFeatureN/ANoneDetection 6103N/ANoneSuppression N/ANoneFeatureIonization 601Detection NNNN6104NN/ANoneSuppression N/ANoneFeatureIonization 601Detection NNNN6105NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 6106N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 6201N/ANoneSuppression N/ANoneFeatureAugust 2011 C-495 Ameren MissouriCallaway Plant NFPA 805 Transition ReportFB-1Fuel Handling Building1&$.>.?>..A&."&.&Ionization 601Detection NNNN6202NN/ANoneSuppression N/ANoneFeatureIonization 601Detection NNNN6203NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 6204N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 6205N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 6210N/ANoneSuppression N/ANoneFeatureFlame602Detection NNNN6301NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 6302N/ANoneSuppression N/ANoneFeatureIonization 601Detection NNNN6303NN/ANoneSuppression N/ANoneFeatureAugust 2011 C-496 Ameren MissouriCallaway Plant NFPA 805 Transition ReportFB-1Fuel Handling Building1&$.>.?>..A&."&.&Ionization 601Detection NNNN6304NN/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNone&The effects of moderate energy line break and flooding which i nclude rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-497 Ameren MissouriCallaway Plant NFPA 805 Transition ReportLDF-1Laundry Decontamination Facility1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel IISteam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel II Steam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription1332Decontamination Room and Trash Sorting Area1333Laundry Room A1334Equipment Room1337Clean Laundry Sorting AreaAugust 2011 C-498LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportLDF-1Laundry Decontamination Facility1RCS Inventory ControlVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-499 Ameren MissouriCallaway Plant NFPA 805 Transition ReportLDF-1Laundry Decontamination Facility1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.
: 2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.None....3NoneAugust 2011 C-500 Ameren MissouriCallaway Plant NFPA 805 Transition ReportLDF-1Laundry Decontamination Facility1&$.>.?>..A&."&.&Ionization 116Detection NNNN1332NWet PipeLDFSuppression NNNNNN/ANoneFeatureIonization 116Detection NNNN1333NWet PipeLDFSuppression NNNNNN/ANoneFeatureIonization 116Detection NNNN1334NWet PipeLDFSuppression NNNNNN/ANoneFeatureIonization 116Detection NNNN1337NWet PipeLDFSuppression NNNNNN/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-501 Ameren MissouriCallaway Plant NFPA 805 Transition ReportLDF-1Laundry Decontamination Facility1None&This area contains no cabling or electrically supervised equipment that is required for post fire safe shutdown: therefore, fire suppression activities will not adversely affect the plant's ability to achieve the nuclear safety performance criteria.August 2011 C-502 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1~"4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBRB1Steam Generators B and C are supplied by MDAFW Pump A and/or Steam Generators A and D are supplied by MDAFW Pump B. Credited AFW Pump and Steam Generators depend on location of fire within Fire Area (Steam Generator A or B or C or D, or Steam Generators A and D, or Steam Generators B and C).RB2:Steam Generator B is supplied by MDAFW Pump A and/or Steam Generators A and D are supplied by MDAFW Pump B, or Steam Generator B is supplied by MDAFW Pump A and Steam Generator D is supplied by MDAFW Pump B. Credited AFW Pump and Steam Generators depend on location of fire within Fire Area (Steam Generator A, or Steam Generators B and D, or Steam Generators A and D).RB3:Steam Generators B and C are supplied by MDAFW Pump A.RB4:Steam Generators B and C are supplied by MDAFW Pump A.RB5:Steam Generators B and C are supplied by MDAFW Pump A and/or Steam Generators A and D are supplied by MDAFW Pump B. Credited AFW Pump and Fire ZoneDescriptionRB1Reactor Building - El. 2000', Rx Coolant PumpsRB2Reactor Building - El. 2000', Outer Annulus; Reactor Building - El. 2026', Above Accumulators A & D and Eastern Semi AnnulusRB3Reactor Building - El. 2026', North Electrical Penetration AreaRB4Reactor Building - El. 2026', South Electrical Penetration AreaRB5Reactor Building - El. 1974', Tendon Access Gallery; El. 2047', Main Floor and Reactor Vessel Area; El. 2051', Cable Tray Area; El. 2068', Reactor BuildiAugust 2011 C-503LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1Process MonitoringRB1:RCS Pressure Channel IPressurizer Pressure Channel II Pressurizer Level Channel I Ex-core Neutron Monitoring Channel I or IIRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel ISteam Gen. A Atmos. Steam Dump Pressure Channel I Aux. Feedwater Flow to Steam Gen. A Channels I and IV RCS Loop D (4) T-hot Temperature Channels II and VI RCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IVRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and II RCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VISteam Generators depend on location of fire within Fire Area (Steam Generator B or C, or Steam Generator B and C, or Steam Generators A and D).August 2011 C-504 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1Refueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)RB2:Pressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channels I and IVRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel IISteam Gen. A Pressure Channel ISteam Gen. A Wide Range Level Channel I Steam Gen. A Atmos. Steam Dump Pressure Channel I Aux. Feedwater Flow to Steam Gen. A Channels I and IVRCS Loop D (4) T-hot Temperature Channels II and VIRCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV RCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel IISteam Gen. B Pressure Channel I Steam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and II Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVAugust 2011 C-505 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1RB3:RCS Pressure Channel IIPressurizer Pressure Channel IIPressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel I Steam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VSteam Gen. C Pressure Channel ISteam Gen. C Narrow Range Level Channel IV Steam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IV Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train B (Channel IV and VI) RB4:RCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel I Ex-core Neutron Monitoring Channel I RCS Loop B (2) T-hot Temperature Channel ISteam Gen. B Pressure Channel ISteam Gen. B Narrow Range Level Channel III Steam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel IAugust 2011 C-506 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and II Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) RB5:RCS Pressure Channels I and II Pressurizer Pressure Channel I or IIPressurizer Level Channel I or IIEx-core Neutron Monitoring Channel IRCS Loop A (1) T-hot Temperature Channel I RCS Loop A (1) T-cold Temperature Channel IISteam Gen. A Pressure Channel I Steam Gen. A Wide Range Level Channel I Steam Gen. A Atmos. Steam Dump Pressure Channel IAux. Feedwater Flow to Steam Gen. A Channels I and IVRCS Loop D (4) T-hot Temperature Channels II and VI RCS Loop D (4) T-cold Temperature Channels I and VISteam Gen. D Pressure Channel ISteam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV RCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel IISteam Gen. B Pressure Channel I Steam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel IIAux. Feedwater Flow to Steam Gen. B Channels I and II RCS Loop C (3) T-hot Temperature Channels II and VAugust 2011 C-507 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1RCS Inventory ControlRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVRB1:Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary. RB2:Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary. RB3:Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary. RB4:Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary. RB5:Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath.See VFDR No. RB-01-001, RB-01-002, RB-01-003, and RB-01-004See VFDR No. RB-02-001, RB-02-002, RB-02-003, RB-02-004, RB-02-005, RB 006, and RB-02-007See VFDR No. RB-03-001, RB-03-002, RB-03-003, and RB-03-004See VFDR No. RB-04-001, RB-04-002, RB-04-003, and RB-04-004See VFDR No. RB-05-001August 2011 C-508 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1Reactivity ControlRB1:Trip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.RB2:Trip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.RB3:Trip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.RB4:Trip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.RB5:RCS Pressure ControlRB1:Both Pressurizer Heater Backup Groups unavailable. RCS pressure control can be satisfied with AFW and CVCS.RB2:Control pressure using Pressurizer Heater Backup Group B. RCS pressure control can be satisfied with AFW and CVCS.RB3:Control pressure using Pressurizer Heater Backup Group B. RCS pressure control can be satisfied with AFW and CVCS.RB4:Both Pressurizer Heater Backup Groups unavailable. RCS pressure control can be satisfied with AFW and CVCS.RB5:Control pressure using Pressurizer Heater Backup Group A. RCS pressure control can be satisfied with AFW and CVCS.August 2011 C-509 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1Trip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesRB1:Operate CCW Pumps A and C, and ESW Pumps A and B.
Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).RB2:Operate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).RB3:Operate CCW Pumps A, B, C, and D, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).RB4:Operate CCW Pumps A and C, and ESW Pumps A and B.
Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).RB5:Operate CCW Pumps A, B, C, and D, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A or Train B credited, depending on location of fire within Fire Area).Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-510 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1"There is no deviation associated with the requirements of Appendix A of BTP ASB 9.5-1; however the manual suppression system is a modified automatic system, which is in fact a deviation from NFPA 13 requirements. The fire protection for Containment, including the manually charged sprinkler system and dry standpipe system, was found to meet the guidelines of Appendix A to BTP ASB 9.5-1 by the NRC in NUREG-0830, dated 10/1981 based on the following: 
: 1. The reactor building is separated from adjacent buildings by 3-hr fire barriers. 2. There are no physical boundaries enclosing localized fire hazards within the reactor building. 3. Automatic line-type detection is installed above each reactor coolant pump.4. Line type thermal detectors are also installed in all areas where cable trays are concentrated.5. Ionization type detectors are installed in the containment cooler ducts.6. Union Electric (dba Ameren Missouri) committed to additional hose stations so that every hose station will be spaced no more than 100 ft. from an adjacent hose station.
: 7. A fixed, manually charged closed head sprinkler system is provided over the two cable tray pe netration areas.&There is no deviation associated with the requirements of Appendix A of BTP ASB 9.5-1; however the manual suppression system is a modified automatic system, which is in fact a deviation from NFPA 13 requirements. The system was specifically approved by the NRC. Some of the bases identified in the SER, and accepted by the NRC, have been modified since the configuration was approved. Therefore, clarification regarding the acceptability of the manual system is required. The clarification is being requested in the License Amendment Request Transition Report, Attachment T.August 2011 C-511 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1%~&Deviation submitted per 6/29/1981 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying the reactor coolant pump oil collection system, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The system has been seismically analyzed and qualified to remain functional during and after the safe shutdown earthquake.2. Collection tanks are provided with level indication and high level alarm in the Control Room.3. Should leakage exceed the collection tank capacity before corrective actions are completed, the tank would overflow into the containment sumps.4. Oil would not come into contact with hot surfaces and would not pose a significant fire hazard.5. The tanks are constructed to the requirements of ASME Code Section VIII.
: 6. The tanks have flame arrestors on the vents.
: 7. The drain piping meets American National Standards Institute (ANSI) Standard B-31.1.&This deviation is active per Section RB.4 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated electrical penetrations in the reactor containment walls to Fire Areas A-17 and A-18 NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-512 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated mechanical penetrations (process and sampling lines and containment purge penetration) in the reactor containment walls to Fire Areas A-19, A-20, A-23, A-24 and A-25, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated fuel transfer tube connecting reactor containment and the fuel building, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-513 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1%~&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated personnel hatch connecting reactor containment and Fire Area A-20, and the hatchways to YD-1, was approved by the NRC in NUREG-0830, Supplement 3, dated
 
05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid. Although the personnel hatch to Fire Area A-20 was approved, the containment emergency personnel and equipment hatchways to the yard, Fire Area YD-1, were not specifically called out in the SER. The emergency personnel hatchway is of identical construction to the personnel hatch to Fire Area A-20. The equipment hatch, while not identical, is equally robustly constructed, consisting of a welded steel assembly with a double gasketed, flanged, and bolted cover and provided with a moveable missile shield on the outside of the Reactor Building. Therefore, clarification regarding the approval of all containment hatchways is required. The clarification is being requested in the License Amendment Request Transition Report, Attachment T.None....3August 2011 C-514 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1885(D54D5:D5VFDR for intervening combustibles between Reactor Building deterministic analysis areas. The Reactor Building, Fire Area RB-1, is separated into five deterministic analysis areas (RB1, RB2, RB3, RB4 and RB5) based on physical barriers and separation. These analysis areas were defined to approximate the Reactor Building compliance strategies from the previous Appendix R licensing basis, to simplify the deterministic analysis for the NFPA 805 NSCA, and to provide a meaningful assessment of the Reactor Building Safe Shutdown capability for the NFPA 805 NSCA. The separation between these analysis areas contains intervening combustibles in the form of open and/or partially enclosed cable trays containing IEEE-383 qualified cables. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3.4(a). This is a separation issu e.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85(55(BGHV8149A - Cable damage (5BGK35AE, 5BGK35AF, 5BGK35AG, 5BGK35AH, and 5BGK35AI) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.554BGHV8149B - Cable damage (5BGK35BE, 5BGK35BF, 5BGK35BG, 5BGK35BH, and 5BGK35BI) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-515LIC-26 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building185(55BGHV8149C - Cable damage (5BGK35CE, 5BGK35CF, 5BGK35CG, 5BGK35CH, and 5BGK35CI) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.85(55:PZR-HTR-BU-A - Cable damage (5BBG23AB, 5BBG23BB, 5BBG23CB, 5BBG23DB, 5BBG23EB, 5BBG23FB, 5BBG23GB, 5BBG23HB, 5BBG23JB, and 5BBG23KB) to PZR-HTR-BU-A; cable damage (6BBG25AB, 6BBG 25BB, 6BBG25CB, 6BBG25DB, 6BBG25EB, 6BBG25FB, 6BBG25GB, 6BBG25HB, 6BBG25JB, and 6BBG25KB) to PZR-HTR-BU-B; neither backup group of pressurizer heaters is available for safe shutdown in this fire area. Loss of pressurizer heater capability may adversely impact the ability of the plant to maintain safe and stable (potential adverse impact to RCS Pressure Control). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.  (Neither backup group of pressurizer heaters is recoverable with local manual operator actions.)The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85455(BBPV8702A-P - Cable damage (4BBG12AH) to BBPV8702A-P. Power cable damage can spuriously open valve (even though the valve breaker is administratively locked open); this assumes that external hot shorts are possible for high/low pressure interface valves; power cable damage to EJHV8701A-P (1EJG05AD) can spuriously open valve (even though the valve breaker is administratively locked open); this assumes that external hot shorts are possible for high/low pressure interface valves; the power cables for valves BBPV8702A-P and EJHV8701A-P are routed in cable trays (with other potentially energized 480V three phase power cables). The cable trays are located within 20 feet of each other (Ref. E-2R2908A, E-2R2312C, and E-2R2312D). As such, the potential exists for both valves to spuriously open for a fire in fire area RB-02. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-516 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1854554BGHV8149A - Cable damage (5BGK35AC) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.85455BGHV8149B - Cable damage (5BGK35BC) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.55:BGHV8149C - Cable damage (5BGK35CC) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-517 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building185455,BGHV8153A-P - Cable damage (1BGK48CA) to BGHV8153A. Cable damage can spuriously open or prevent closure of the Reactor Coolant System to Chemical Volume Control System Excess Letdown Downstream Isolation Protection A Valve, BGHV8153A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.854557EJHV8701A-P - Cable damage (1EJG05AD) to EJHV8701A. Power cable damage can spuriously open valve (even though the valve breaker is administratively locked open); this assumes that external hot shorts are possible for high/low pressure interf ace valves; power cable damage to BBPV8702A-P (4BBG12AH) can spuriously open valve (even though the valve breaker is administratively locked open); this assumes that external hot shorts are possible for high/low pressure interface valves; the power cables for valves BBPV8702A-P and EJHV8701A-P are routed in cable trays (with other potentially energized 480V three phase power cables). The cable trays are located within 20 feet of each other (Ref. E-2R2908A, E-2R2312C, and E-2R2312D). As such, the potential exists for both valves to spuriously open for a fire in fire area RB-02. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85455;EMHV8843 - Cable damage (4EMK04CC) to EMHV8843. Cable damage can spuriously open or prevent closure of the Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843. The valve is required closed to maintain positive contro l over RCS inventory and pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-518 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building18555(BBPCV0455A-P - Cable damage (1BBK40AH) to BBPCV0455A; cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.85554BGHV8149A - Cable damage (5BGK35AC and 5BGK35AE) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.BGHV8149B - Cable damage (5BGK35BC and 5BGK35BE) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.August 2011 C-519 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building18555:BGHV8149C - Cable damage (5BGK35CC and 5BGK35CE) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.A Recovery Action (RA) is credited for this VFDR to reduce risk due to a fire in this fire area. The RA has been demonstrated to be feasible. Reliability is addressed within the FPRA using HRA methods.85:55(BBHV8001B - Cable damage (4BBK30BB) to BBHV8001B. Cable damage can spuriously open or prevent closure of the Reactor Coolant System Reactor Vessel Head Vent Protection B Upstream Valve, BBHV8001B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.554BGHV8153B-P - Cable damage (4BGK48AA) to BGHV8153B. Cable damage can spuriously open or prevent closure of the Reactor Coolant System to Chemical Volume Control System Excess Letdown Downstream Isolation Protection B Valve, BGHV8153B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-520 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building185:55EMHV8843 - Cable damage (4EMK04CC) to EMHV8843. Cable damage can spuriously open Boron Injection Header Outlet Upstream Test Line Isolation Valve, EMHV8843. The valve is required closed in order to prevent potential diversion of RWST inventory to the SI test lines to ensure positive control over RCS Inventory and Pressure. This condition represents a varianc e from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85:55:PZR-HTR-BU-A - Cable damage (5BBG23AB, 5BBG23BB, 5BBG23CB, 5BBG23DB, 5BBG23EB, 5BBG23FB, 5BBG23GB, 5BBG23HB, 5BBG23JB, and 5BBG23KB) to PZR-HTR-BU-A; cable damage (6BBG25AB, 6BBG 25BB, 6BBG25CB, 6BBG25DB, 6BBG25EB, 6BBG25FB, 6BBG25GB, 6BBG25HB, 6BBG25JB, and 6BBG25KB) to PZR-HTR-BU-B; neither backup group of pressurizer heaters is available for safe shutdown in this fire area. Loss of pressurizer heater capability may adversely impact the ability of the plant to maintain safe and stable (potential adverse impact to RCS Pressure Control). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.  (Neither backup group of pressurizer heaters is recoverable with local manual operator actions.)The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.85,55(BBPCV0455A-P - Cable damage (1BBK40AH) to BBPCV0455A; cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the determinist ic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-521 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1&$.>.?>..A&."&.&Line-Type 203Detection NNYNRB1NLine-Type 206Detection NNYNNLine-Type 215Detection NNYNNLine-Type 216Detection NNYNNLine-Type 218Detection NNYNNN/ANoneSuppression 20-ft Separation ZoneNoneFeatureNNNYNN/ANoneFeatureLine-Type 201Detection NNYNRB2NLine-Type 202Detection NNYNNLine-Type 203Detection NNYNNLine-Type 204Detection NNYNNLine-Type 205Detection NNYNNLine-Type 215Detection NNYNNN/ANoneSuppression 20-ft Separation ZoneNoneFeatureNNNYNRadiant Energy ShieldNoneFeatureNNNYNReactor Vessel as Radiant Energy ShieldLine-Type 215Detection NNYYRB3NPre-action SKC41Suppression NNNYNover cable trays; manually actuated from MCR N/ANoneFeatureAugust 2011 C-522 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1&$.>.?>..A&."&.&Line-Type 203Detection NNNYRB4NLine-Type 204Detection NNNYNLine-Type 205Detection NNNYNLine-Type 216Detection NNNYNPre-action SKC42Suppression NNNYNover cable trays; manually actuated from MCRRadiant Energy ShieldNoneFeatureNNNYNwrapped conduit as Radiant Energy ShieldLine-Type 217Detection NNYNRB5NLine-Type 218Detection NNYNNIonization 219Detection NNYNNLine-Type 220Detection NNYNNN/ANoneSuppression 20-ft Separation ZoneNoneFeatureNNNYNRadiant Energy ShieldNoneFeatureNNNYNwrapped conduit as Radiant Energy Shield SLERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationAugust 2011 C-523 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRB-1Reactor Building1None&RB1:There are no automatic fire suppression systems in the fire area. Adequate drainage capability exists in the reactor building to prevent the accumulation of fire-fighting water. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.RB2:There are no automatic fire suppression systems in the fire area. Adequate drainage capability exists in the reactor building to prevent the accumulation of fire-fighting water. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.RB3:A fixed, manually charged, closed head sprinkler system is provided over the cable trays in zone RB3. Adequate drainage capability exists in the reactor building to prevent the accumulation of fire-fighting water. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.RB4:A fixed, manually charged, closed head sprinkler system is provided over the cable trays in zone RB4. Adequate drainage capability exists in the reactor building to prevent the accumulation of fire-fighting water. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.RB5:There are no automatic fire suppression systems in the fire area. Adequate drainage capability exists in the reactor building to prevent the accumulation of fire-fighting water. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-524 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRSB-1RAM Storage Building1~"Process MonitoringRCS Inventory ControlRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel II Steam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and IICondensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription8501RAM Storage BuildingAugust 2011 C-525LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRSB-1RAM Storage Building1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0&Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain val id.August 2011 C-526 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRSB-1RAM Storage Building1None....3None8&$.>.?>..A&."&.&Ionization 801Detection NNNN8501NWet PipeNoneSuppression NNNNNN/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-527 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRSB-1RAM Storage Building1None&This area contains no cabling or electrically supervised equipment that is required for post fire safe shutdown: therefore, fire suppression activities will not adversely affect the plant's ability to achieve the nuclear safety performance criteria.August 2011 C-528 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1Fire ZoneDescription7101Waste Gas Compressor Room No. 17102Hydrogen Recombiner Room No. 17103Valve Room No. 17104Recycle Evaporator Feed Pump Room7105Recycle Hold-up Tank Room No. 17106Waste Gas Compressor Room No. 27107Hydrogen Recombiner Room No. 27108Valve Room No. 27109Corridor No. 17110Recycle Hold-up Tank Room No. 2 7111Waste Gas Decay Tank Room No. 1 7112Valve Room No. 37113Load Center and General Area7114Stair RW-17115Waste Gas Decay Tank Room No. 27116Valve Room No. 47117Corridor No. 2 (East, West)7118Steam Generator Blowdown Surge Tank and Pump Room7119Radioactive Pipe Chase7120Chemical Drain Tank and Pump Room7121Waste Evaporator Feed Pump Room7122Waste Hold-Up Tank Room 7123Waste Evaporator Bottoms Tank Room (Primary)7124Waste Evaporator Bottoms Tank Pump Room7125Floor Drain Tank Pump Room No. 17126Floor Drain Tank Room No. 17127Waste Monitor Tank and Pump Room7128Floor Drain Tank Pump Room No. 2August 2011 C-529 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1Fire ZoneDescription7129Floor Drain Tank Room No. 2 7130Waste Evaporator Condensate Tank and Pump Room 7131Vestibule7132Stair RW-27133Electrical Chase, Non-Radioactive Pipe Tunnel & Personnel Access7134Radioactive Pipe Tunnel7135Gaseous Radwaste Drain Collection Tank and Gas Decay Tank Drain Pump Room7201Recycle Evaporator Room7202Recycle Evaporator Valve Gallery7203Corridor No. 17204Waste Evaporator Room7205Waste Evaporator Valve Gallery7206SLWS Evaporator Reagent Tank Room 7207SLWS Valve Gallery7208MCC Equipment Load Center and General Area7209Control Room7210Nuclear Sample Panel Room7211Sample Laboratory7212Spent Resin Storage Tank Room (Primary)7213Corridor No. 37214Spent Resin and Evaporator Bottom Tank and Pump Room (Secondary)7215Decant Tank7216Corridor No. 27217SLWS Monitor Tank Room 7218Drum Processing Enclosure 7219Solidification Control Panel Room7220Valve Room7221Emergency Shower and Eyewash AreaAugust 2011 C-530 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1Fire ZoneDescription7222Misc. Storage Area 7223Vestibule 7224High Level Drum Storage Area7225Low Level Drum Storage Area7226Empty Drum Storage Area7227Filter Drop Station7228Drywaste Compactor Area7229Concentrates Pump Room7230Instrument Rack Area7231Subcoolers and Condenser Room7232Electrical Chase7233Area Over Valve Room7301Radioactive Pipe Chase Area 7302HVAC Equipment Area7303Vestibule7304MCC Equipment Area7305Electrical Chase7401Filter Compartment7402Valve Compartments7403Corridor7404Valve Compartments7405Demineralizer Compartment7406Valve Compartment7407Fuel Pool Cleanup Demin Compartment 7408Laundry and Hot Shower Tank Area 7409R.O. Unit Recycle Tank Area7410General Floor Area7411Waste Monitor Tank and Pump AreaAugust 2011 C-531 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I RCS Loop B (2) T-cold Temperature Channel IISteam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel II Steam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and VRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I Steam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel I Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)
Channels I and II4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription7412Caustic Tank Area 7413HVAC Platform 7501General Floor Area (2040 elev.)7502General Floor Area (2041 elev.)7503General Floor Area (2047 elev.)7504Platform (2051 elev.)August 2011 C-532LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1RCS Inventory ControlCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-533 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1%~&',Deviation submitted per 6/29/1981 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, providing justification for a non-rated cover on the trench connecting the fuel building and radwaste tunnel, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Combustibles in this area are separated by more than 50 feet.
: 2. Low combustible loading.
: 3. The trench opening in this room is closed by a heavy steel cover plate approximately 4-feet x 8-feet.&This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.None....3None8August 2011 C-534 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1&$.>.?>..A&."&.&
N/ANoneDetection 7101N/ANoneFeatureN/ANoneDetection 7102N/ANoneFeatureIonization 705Detection NNNN7103NN/ANoneFeatureN/ANoneDetection 7104N/ANoneFeatureN/ANoneDetection 7105N/ANoneFeatureN/ANoneDetection 7106N/ANoneFeatureN/ANoneDetection 7107N/ANoneFeatureIonization 705Detection NNNN7108NN/ANoneFeatureIonization 705Detection NNNN7109NN/ANoneFeatureN/ANoneDetection 7110N/ANoneFeatureN/ANoneDetection 7111N/ANoneFeatureAugust 2011 C-535 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1&$.>.?>..A&."&.&
N/ANoneDetection 7112N/ANoneFeatureIonization 705Detection NNNN7113NN/ANoneFeatureN/ANoneDetection 7114N/ANoneFeatureN/ANoneDetection 7115N/ANoneFeatureN/ANoneDetection 7116N/ANoneFeatureIonization 705Detection NNNN7117NN/ANoneFeatureN/ANoneDetection 7118N/ANoneFeatureN/ANoneDetection 7119N/ANoneFeatureN/ANoneDetection 7120N/ANoneFeatureIonization 705Detection NNNN7121NN/ANoneFeatureN/ANoneDetection 7122N/ANoneFeatureN/ANoneDetection 7123N/ANoneFeatureAugust 2011 C-536 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1&$.>.?>..A&."&.&Ionization 705Detection NNNN7124NN/ANoneFeatureIonization 705Detection NNNN7125NN/ANoneFeatureN/ANoneDetection 7126N/ANoneFeatureIonization 705Detection NNNN7127NN/ANoneFeatureIonization 705Detection NNNN7128NWet PipeSKC33Suppression NNNNNN/ANoneFeatureN/ANoneDetection 7129N/ANoneFeatureIonization 705Detection NNNN7130NN/ANoneFeatureIonization 705Detection NNNN7131NN/ANoneFeatureN/ANoneDetection 7132N/ANoneFeatureIonization 705Detection NNNN7133NN/ANoneFeatureN/ANoneDetection 7134N/ANoneFeatureAugust 2011 C-537 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1&$.>.?>..A&."&.&
N/ANoneDetection 7135N/ANoneFeatureN/ANoneDetection 7201N/ANoneFeatureN/ANoneDetection 7202N/ANoneFeatureIonization 700Detection NNNN7203NN/ANoneFeatureN/ANoneDetection 7204N/ANoneFeatureN/ANoneDetection 7205N/ANoneFeatureN/ANoneDetection 7206N/ANoneFeatureN/ANoneDetection 7207N/ANoneFeatureIonization 700Detection NNNN7208NN/ANoneFeatureIonization 700Detection NNNN7209NN/ANoneFeatureN/ANoneDetection 7210N/ANoneFeatureN/ANoneDetection 7211N/ANoneFeatureAugust 2011 C-538 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1&$.>.?>..A&."&.&
N/ANoneDetection 7212N/ANoneFeatureN/ANoneDetection 7213N/ANoneFeatureN/ANoneDetection 7214N/ANoneFeatureN/ANoneDetection 7215N/ANoneFeatureIonization 700Detection NNNN7216NN/ANoneFeatureN/ANoneDetection 7217N/ANoneFeatureN/ANoneDetection 7218N/ANoneFeatureN/ANoneDetection 7219N/ANoneFeatureN/ANoneDetection 7220N/ANoneFeatureN/ANoneDetection 7221N/ANoneFeatureIonization 700Detection NNNN7222NN/ANoneFeatureIonization 700Detection NNNN7223NN/ANoneFeatureAugust 2011 C-539 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1&$.>.?>..A&."&.&
N/ANoneDetection 7224N/ANoneFeatureN/ANoneDetection 7225N/ANoneFeatureIonization 700Detection NNNN7226NN/ANoneFeatureN/ANoneDetection 7227N/ANoneFeatureIonization 700Detection NNNN7228NWet PipeSKC33Suppression NNNNNN/ANoneFeatureN/ANoneDetection 7229N/ANoneFeatureN/ANoneDetection 7230N/ANoneFeatureN/ANoneDetection 7231N/ANoneFeatureIonization 700Detection NNNN7232NN/ANoneFeatureN/ANoneDetection 7233N/ANoneFeatureN/ANoneDetection 7301N/ANoneFeatureAugust 2011 C-540 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1&$.>.?>..A&."&.&
N/ANoneDetection 7302N/ANoneFeatureN/ANoneDetection 7303N/ANoneFeatureN/ANoneDetection 7304N/ANoneFeatureIonization 700Detection NNNN7305NN/ANoneFeatureN/ANoneDetection 7401N/ANoneDetection N/ANoneFeatureIonization 701Detection NNNN7402NN/ANoneFeatureIonization701, 702Detection NNNN7403NN/ANoneFeatureN/ANoneDetection 7404N/ANoneFeatureN/ANoneDetection 7405N/ANoneFeatureN/ANoneDetection 7406N/ANoneFeatureN/ANoneDetection 7407N/ANoneFeatureAugust 2011 C-541 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1&$.>.?>..A&."&.&
N/ANoneDetection 7408N/ANoneFeatureN/ANoneDetection 7409N/ANoneFeatureIonization 702Detection NNNN7410NN/ANoneFeatureN/ANoneDetection 7411N/ANoneFeatureN/ANoneDetection 7412N/ANoneFeatureN/ANoneDetection 7413N/ANoneFeatureN/ANoneDetection 7501N/ANoneFeatureN/ANoneDetection 7502N/ANoneFeatureN/ANoneDetection 7503N/ANoneFeatureN/ANoneDetection 7504N/ANoneFeatureAugust 2011 C-542 Ameren MissouriCallaway Plant NFPA 805 Transition ReportRW-1Radwaste Building1SL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThis area contains no cabling or electrically supervised equipment that is required for post fire safe shutdown: therefore, fire suppression activities will not adversely affect the plant's ability to achieve the nuclear safety performance criteria.August 2011 C-543 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1Fire ZoneDescription3102Pipe Space, Tank, and Storage Area3103Stair No. CC-13225Corridor No. 23226Counting Room3227Vestibule No.33228Hot Laboratory3303Corridor3304General Floor Area3307Combustible Liquids Storage Room3402Corridor No. 3 3502Lobby3503General Floor Area3611Corridor No. 23612Conference Room3613Computer Room3613AWork Control Area3613BEquipment Operators Room3614Corridor No. 33619General Floor Area3620Women's Toilet3621Men's Toilet3701General Floor Area 3702Battery Room3703Radio Equipment Room3704General Floor Area3705Battery Room3706EO's Office3802Elevator No. 1 Machine RoomAugust 2011 C-544 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1Fire ZoneDescription3803Corridor4101Stair T-1 4201Condenser Pit - General Floor Area4203SGFP Turbine Lube Oil Conditioners - Cond Pump Area4204Secondary Liquid Waste Collection Tank Pumps4205High TDS and Low TDS Tank and Pump Area4301General Floor Area SW and SE4302Condenser Vacuum Pump Area4303Air Compressor Area4304Mens Toilet4305Womens Toilet4306Janitors Closet4308Lube Oil Storage Tanks 4309Stair T-24310Stair T-34312Stair T-54313Stair T-44314Stair T-64316Condensate Polishing Area4317Process Sampling Lab4318Closed Cooling Water Heat Exchanger Area4319Condensate Chemical Add Units4321Railroad Bay and Laydown Area4322Truck Bay and Laydown Area 4323Cold Chemistry Lab 4351Floor4401General Floor Area4402Battery Room (SE)August 2011 C-545 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1~"Process MonitoringRCS Pressure Channel IIPressurizer Pressure Channel IIPressurizer Level Channel IIEx-core Neutron Monitoring Channel IV RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel IISteam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IV RCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IVAux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) 4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription4403Lube Oil Reservoir Room 4404Battery Room (NW) 4501General Floor Area (North and South of Column Line T-6)4502Womens Toilet4503Mens Toilet4504EHC Control Cabinet Room4505General Floor Area4506Room 45064601Elevator Machine RoomAugust 2011 C-546LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1RCS Inventory ControlChannels II and IVRefueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite Power to NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.See VFDR No. TB-01-003RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.See VFDR No. TB-01-001 and TB-01-002Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-547 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.
: 2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, providing justification for lack of full-area detection coverage (specifically no detection in Fire Zone 3227), was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. Minimal fire hazards.2. Availability of manual firefighting equipment.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.August 2011 C-548LIC-25 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1....3.%+574,The presence of non-rated missile resistant shields, non-rated access hatch, open steel drain piping, and air gaps caused by main steam line penetrations in the barrier between A-23 and TB-1 is acceptable based on the structural integrity of the barrier; the structural integrity of the unexposed steel; the dissipation of heat in the large volumes of Fire Areas A-23 and TB-1; limited quantities of combustibles in the vicinity of the non-rated features; low combustible loading in the areas; limited ignition sources; and automatic suppression is installed on several elevations of the Turbine Building .&.%+574;The non-rated blowout panel in the wall between Fire Areas A-15 and TB-1 is acceptable based on the fact that the combustible loading on both sides of the panel is not significant, there are no fixed ignition sources in the vicinity of the panel, and the panel is similar to a 3 hour UL listed configuration.&August 2011 C-549 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building188(PG2401 - Cable damage (6BBG20AE) to PG2401. Cable damage may result in a spurious close signal or failure to trip on demand of the non-credited train Pressurizer Heater Control Group C Breaker, PG2401, with the loss of remote trip control capability. Non-credited train Pressurizer Heater Control Group C may need to be secured in order to ensure positive control over RCS pressure (to prevent pressurizer PORV challenge). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.PZR-HTR-BU-B - Backup Pressurizer Heater Groups A and B are subject to cable damage, loss of motive power (for the Group A Heaters only), and loss of DC control power (cables 5PKK01AJ and 5PKK01AW and DC Switchboard Bus PK01 for the Group A Heaters, the breaker PG2101 DC control power cables - cables 6PKK02AG and 6PKK02AV and DC Switchboard Bus PK02 for the Group B Heaters, the breaker PG2201 DC control power cables). Neither backup group of pressurizer heaters is available for safe shutdown in this fire area. Loss of pressurizer heater capability may adversely impact the ability of the plant to maintain safe and stable (potential adverse impact to RCS Pressure Control). This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.  (Backup group B of pressurizer heaters is recoverable wi th local manual operator action.)The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.8SGK05B - Cable damage (6GKK31DA) to SGK05B. Cable damage may result in a loss of ventilation from the Train B ESF Switchgear Rooms Air Cond itioning Unit, SGK05B. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-550 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&Photoelectric 311Detection NNNN3102NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 3103N/ANoneSuppression N/ANoneFeaturePhotoelectric 311Detection NNNN3225NN/ANoneSuppression N/ANoneFeatureIonization 310Detection NNNN3226NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 3227N/ANoneSuppression N/ANoneFeatureIonization 310Detection NNNN3228NN/ANoneSuppression N/ANoneFeatureIonization 310Detection NNNN3303NPhotoelectric 311Detection NNNNNN/ANoneSuppression N/ANoneFeatureAugust 2011 C-551 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&Ionization 310Detection NNNN3304NN/ANoneSuppression N/ANoneFeatureIonization 310Detection NNNN3307NWet PipeN/ASuppression NNNY*N*required for Chapter 3 compliance N/ANoneFeatureIonization 302Detection NNNN3402NPhotoelectric 311Detection NNNNNN/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3502NPhotoelectric 311Detection NNNNNN/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3503NN/ANoneSuppression N/ANoneFeaturePhotoelectric 311Detection NNNN3611NN/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3612NN/ANoneSuppression N/ANoneFeatureAugust 2011 C-552 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&Ionization 312Detection NNNN3613NPhotoelectric 313Detection NNNNNPre-action SKC47Suppression NNNNNN/ANoneFeatureIonization 313Detection NNNN3613ANPre-action SKC47Suppression NNNNNN/ANoneFeatureIonization 312Detection NNNN3613BNIonization 312Detection NNNNNThermal313Detection NNNNNPre-action SKC47Suppression NNNNNN/ANoneFeatureIonization 302Detection NNNN3614NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 3619N/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3620NN/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3621NN/ANoneSuppression N/ANoneFeatureAugust 2011 C-553 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&Ionization 302Detection NNNN3701NN/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3702NN/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3703NN/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3704NN/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3705NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 3706N/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3802NN/ANoneSuppression N/ANoneFeatureIonization 302Detection NNNN3803NN/ANoneSuppression N/ANoneFeatureAugust 2011 C-554 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&
N/ANoneDetection 4101N/ANoneSuppression N/ANoneFeatureIonization 400Detection NNNN4201NPhotoelectric 401Detection NNNNNWet PipeSKC32Suppression NNYNYN/ANoneFeatureThermal409Detection NNNN4203NThermal410Detection NNNNNWet PipeSKC32Suppression NNYNYN/ANoneFeatureIonization 411Detection NNNN4204NWet PipeSKC32Suppression NNYNYN/ANoneFeatureN/ANoneDetection 4205Wet PipeSKC32Suppression NNYNYN/ANoneFeatureAugust 2011 C-555 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&Photoelectric 401Detection NNNN4301NThermal403Detection NNYNYThermal405Detection NNYNYThermal409Detection NNNNNThermal410Detection NNNNNWater Spray SKC23Suppression NNNNNWater Spray SKC24Suppression NNNNNPre-action SKC29Suppression NNYNYPre-action SKC31Suppression NNYNYN/ANoneFeatureThermal405Detection NNYN4302YPre-action SKC29Suppression NNYNYN/ANoneFeatureThermal405Detection NNYN4303YPre-action SKC29Suppression NNYNYN/ANoneFeatureN/ANoneDetection 4304N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 4305N/ANoneSuppression N/ANoneFeatureAugust 2011 C-556 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&
N/ANoneDetection 4306N/ANoneSuppression N/ANoneFeatureFlame400Detection NNNN4308NWet PipeSKC08Suppression NNNNNN/ANoneFeatureN/ANoneDetection 4309N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 4310N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 4312N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 4313N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 4314N/ANoneSuppression N/ANoneFeatureThermal403Detection NNYN4316YPre-action SKC31Suppression NNYNYN/ANoneFeatureAugust 2011 C-557 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&Ionization 414Detection NNNN4317NN/ANoneSuppression N/ANoneFeatureThermal403Detection NNYN4318YPre-action SKC31Suppression NNYNYN/ANoneFeatureThermal403Detection NNYN4319YPre-action SKC31Suppression NNYNYN/ANoneFeatureThermal403Detection NNYN4321YPre-action SKC31Suppression NNYNYN/ANoneFeatureThermal405Detection NNYN4322YPre-action SKC29Suppression NNYNYN/ANoneFeatureIonization 414Detection NNNN4323NN/ANoneSuppression N/ANoneFeatureThermal405Detection NNYN4351YPre-action SKC29Suppression NNYNYN/ANoneFeatureAugust 2011 C-558 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&Photoelectric 401Detection NNNN4401NThermal402Detection NNYNYThermal404Detection NNYNYThermal407Detection NNNNNWater Spray SKC11Suppression NNNNNfor the Hydrogen Seal Oil skidPre-action SKC28Suppression NNYNYPre-action SKC30Suppression NNYNYN/ANoneFeatureIonization 400Detection NNNN4402NN/ANoneSuppression N/ANoneFeatureFlame400Detection NNNN4403NWet PipeSKC10Suppression NNNNNN/ANoneFeaturePhotoelectric 412Detection NNNN4404NN/ANoneSuppression N/ANoneFeaturePhotoelectric 401Detection NNNN4501NThermal408Detection NNYNYPre-action SKC21Suppression NNYNYfor the turbine bearings N/ANoneFeatureN/ANoneDetection 4502N/ANoneSuppression N/ANoneFeatureAugust 2011 C-559 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1&$.>.?>..A&."&.&
N/ANoneDetection 4503N/ANoneSuppression N/ANoneFeaturePhotoelectric 413Detection NNNN4504NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection 4505N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 4506N/ANoneSuppression N/ANoneFeatureN/ANoneDetection 4601N/ANoneSuppression N/ANoneFeatureSLERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-560 Ameren MissouriCallaway Plant NFPA 805 Transition ReportTB-1Turbine Building1None&The safe shutdown components (Turbine Stop Valves) in the Turbine Building would not be impacted by the automatic suppressions or manual fire fighting. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-561 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUNCTUHS North Cooling Tower1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VISteam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IVCondensate Storage Tank Level Channel VIRefueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IV Containment Pressure Channels II, III, and IVCore Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron 4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription U-301North Side Electrical Room U-302North Side Electrical Room U-306UHS North Cooling TowerAugust 2011 C-562LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUNCTUHS North Cooling Tower1Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite Power to NB02 credited.HVAC credited for Main Control Room and Containment (Train B credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NoneNone....3None8August 2011 C-563 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUNCTUHS North Cooling Tower1&$.>.?>..A&."&.&Ionization 002Detection NNNNU-301NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection U-302N/ANoneSuppression N/ANoneFeatureN/ANoneDetection U-306N/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-564 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUNCTUHS North Cooling Tower1None&There are no automatic fire suppression systems in the fire area. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. The water associated with manual fire suppression will drain out doors and as such standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-565 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUNPHEssential Service Water Pump Room A1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IIPressurizer Pressure Channel II Pressurizer Level Channel IIEx-core Neutron Monitoring Channel IVRCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel II Steam Gen. A Narrow Range Level Channel IVAux. Feedwater Flow to Steam Gen. A Channel IVRCS Loop D (4) T-hot Temperature Channels II and VI Steam Gen. D Pressure Channel IISteam Gen. D Wide Range Level Channel IVSteam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IVAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels II and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel IIVolume Control Tank Level Channel IVContainment Pressure Channels II, III, and IV Core Exit Thermocouples Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump B via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train B is available for letdown of RCS inventory, if necessary.4.2.3.2 - Deterministic ApproachDecay Heat Removal - HSBSteam Generators A and D are supplied by MDAFW Pump B.
Fire ZoneDescription U-104Pump Room AAugust 2011 C-566LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUNPHEssential Service Water Pump Room A1Reactivity ControlTrip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps B and D, and ESW Pump B.Onsite Power to NB02 credited.
HVAC credited for Main Control Room and Containment (Train B credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NoneNone....3None8August 2011 C-567 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUNPHEssential Service Water Pump Room A1&$.>.?>..A&."&.&Ionization 002Detection NNNNU-104NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. The water associated with manual fire suppression will drain back into the Ultimate Heat Sink through the floor grating and as such standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-568 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUSCTUHS South Cooling Tower1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-cold Temperature Channels I and VSteam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel IContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.4.2.3.2 - Deterministic ApproachRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription U-304South Side Electrical Room U-305South Side Electrical Room U-307UHS South Cooling TowerAugust 2011 C-569LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUSCTUHS South Cooling Tower1Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite Power to NB01 credited.
HVAC credited for Main Control Room and Containment (Train A credited).Thermal Barrier Cooling remains available for RCP Seal Cooling.(BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NoneNone....3None8August 2011 C-570 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUSCTUHS South Cooling Tower1&$.>.?>..A&."&.&Ionization 001Detection NNNNU-304NN/ANoneSuppression N/ANoneFeatureN/ANoneDetection U-305N/ANoneSuppression N/ANoneFeatureN/ANoneDetection U-307N/ANoneSuppression N/ANoneFeatureSLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBAugust 2011 C-571 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUSCTUHS South Cooling Tower1None&There are no automatic fire suppression systems in the fire area. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. The water associated with manual fire suppression will drain out doors and as such standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-572 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUSPHEssential Service Water Pump Room B1~"Process MonitoringRCS Inventory ControlRCS Pressure Channel IPressurizer Pressure Channel I Pressurizer Level Channel IEx-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel I Steam Gen. B Pressure Channel I Steam Gen. B Narrow Range Level Channel IIIAux. Feedwater Flow to Steam Gen. B Channel IRCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel ISteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel III Aux. Feedwater Flow to Steam Gen. C Channels I and III Aux. Feedwater Pump A Suction Pressure Channel IRefueling Water Storage Tank Level Channel IVolume Control Tank Level Channel I Containment Pressure Channels II, III, and IV Core Exit Thermocouples Train A (Channel I and V)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from 4.2.3.2 - Deterministic ApproachRCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Decay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescription U-105Pump Room BAugust 2011 C-573LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUSPHEssential Service Water Pump Room B1the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pump A.Onsite Power to NB01 credited.HVAC credited for Main Control Room and Containment (Train A credited).
Thermal Barrier Cooling remains available for RCP Seal Cooling.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0NoneNone....3None8August 2011 C-574 Ameren MissouriCallaway Plant NFPA 805 Transition ReportUSPHEssential Service Water Pump Room B1&$.>.?>..A&."&.&Ionization 001Detection NNNNU-105NN/ANoneSuppression N/ANoneFeatureSL ERD- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action
- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNoneThere are no automatic fire suppression systems in the fire area. Safety related electrical motors are on pedestals and are designed and sealed to be water resistant. The water associated with manual fire suppression will drain back into the Ultimate Heat Sink through the floor grating and as such standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.August 2011 C-575 Ameren MissouriCallaway Plant NFPA 805 Transition ReportYD-1Yard Area~"Process MonitoringRCS Pressure Channels I and IIPressurizer Pressure Channel I Pressurizer Level Channel I Ex-core Neutron Monitoring Channel IRCS Loop B (2) T-hot Temperature Channel IRCS Loop B (2) T-cold Temperature Channel IISteam Gen. B Pressure Channel ISteam Gen. B Wide Range Level Channel IISteam Gen. B Atmos. Steam Dump Pressure Channel II Aux. Feedwater Flow to Steam Gen. B Channels I and IIRCS Loop C (3) T-hot Temperature Channels II and V RCS Loop C (3) T-cold Temperature Channels I and V Steam Gen. C Pressure Channel I4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptionsDecay Heat Removal - HSBSteam Generators B and C are supplied by MDAFW Pump A.
Fire ZoneDescriptionCSTCondensate Storage Tank AreaCWPHCirculating Water Pump HouseEX1Essential Transformer/Capacitor Bank Train AEX2Essential Transformer/Capacitor Bank Train BFPHFire Water Pump House MXFRMain Transformer AreaRWSTRefueling Water Storage Tank AreaSWYDMain SwitchyardSXFRStation Service Transformers XPB03 and XPB04 Area UHSUHS Retention PondYardYardAugust 2011 C-576LIC-28 Ameren MissouriCallaway Plant NFPA 805 Transition ReportYD-1Yard AreaRCS Inventory ControlSteam Gen. C Wide Range Level Channel IIISteam Gen. C Atmos. Steam Dump Pressure Channel IIIAux. Feedwater Flow to Steam Gen. C Channels I and IIIAux. Feedwater Pump A Suction Pressure Channel IAux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW) Channels I and II Refueling Water Storage Tank Level Local Mechanical Instrument Volume Control Tank Level Channels I and IVContainment Pressure Channels II, III, and IVCore Exit Thermocouples Train A (Channel I and V) and Train B (Channel IV and VI)Maintain inventory and RCP seal integrity using Charging Pump A via the Boron Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head Vent flowpath Train A is available for letdown of RCS inventory, if necessary.Reactivity ControlTrip reactor from Control Room. Use Charging Pump A to inject borated water from the RWST.Vital AuxiliariesOperate CCW Pumps A and C, and ESW Pumps A and B.Offsite or Onsite Power to NB01 and NB02 credited.
HVAC credited for Main Control Room and Containment (Train A credited).See VFDR No. YD-001RCS Pressure ControlControl pressure using Pressurizer Heater Backup Group A. Use PORV (BBPCV0455A) to depressurize.Calculation KC-26, Nuclear Safety Capability Assessment, Rev. 0August 2011 C-577 Ameren MissouriCallaway Plant NFPA 805 Transition ReportYD-1Yard Area%~&',Deviation submitted per 2/1/1984 SNUPPS letter to the NRC, as supplemented by SNUPPS submittal to the NRC dated 3/14/1984, justifying non-rated doors to maintain the 3-hr fire rating of barriers in which they are installed, was approved by the NRC in NUREG-0830, Supplem ent 3, dated 05/1984 based on the following:  1. The door is extremely rigid due to the 1-1/2" thick door plate and the reinforcing beam box assembly, which will resist the tendency for the door to bow towards the fire.
: 2. The maximum possible force exerted on the door due to thermal growth will not result in buckling of the door.3. The calculated deflection due to postulated heat exposure does not exceed the maximum allowed in ASTM E-152.4. The doors have been tested to a degree which ensures they would meet the acceptance criteria established in the ASTM E-152 3-hour fire test.&This deviation is active. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.&Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated personnel hatch connecting reactor containment and Fire Area A-20, and the hatchways to YD-1, was approved by the NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:  1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner. 2. Construction is capable of withstanding a 60-psig overpressure without failure.3. Penetrations serve special nuclear safety-related purpose.This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid. Although the personnel hatch to Fire Area A-20 was approved, the containment emergency personnel and equipment hatchways to the yard, Fire Area YD-1, were not specifically called out in the SER. The emergency personnel hatchway is of identical construction to the personnel hatch to Fire Area A-20. The equipment hatch, while not identical, is equally robustly constructed, consisting of a welded steel assembly with a double gasketed, flanged, and bolt ed cover and provided with a moveable missile shield on the outside of the Reactor Building. Therefore, clarification regarding t he approval of all containment hatchways is required. The clarification is being requested in the License Amendment Request Transition Report, Attachment T.August 2011 C-578 Ameren MissouriCallaway Plant NFPA 805 Transition ReportYD-1Yard AreaNone....3885(BNLI0930 - Cable damage (1BNI07CA) to Level Transmitter BNLT0930. Cable damage (2BNI07DA) to Level Transmitter BNLT0931. Cable damage (3BNI07EA) to Level Transmitter BNLT0932. Cable damage (4BNI07FA) to Level Transmitter BNLT0933. Cable damage may result in the loss of all four channels of RWST level instrumentation. At least one channel of RWST level instrumentation is desired for monitoring of RWST level from the Main Control Room to satisfy the NFPA 805 Performance Goal of Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.August 2011 C-579 Ameren MissouriCallaway Plant NFPA 805 Transition ReportYD-1Yard Area&$.>.?>..A&."&.&Wet PipeSKC1040Suppression NNNY*FPHN*required for Chapter 3 compliance N/ANoneFeatureThermal003/004Detection NNNNXMA01ANtransformer ID used in lieu of a fire zone for location purposeWater Spray SKC12Suppression NNNNNtransformer ID used in lieu of a fire zone for location purpose N/ANoneFeaturetransformer ID used in lieu of a fire zone for location purposeThermal005/006Detection NNNNXMA01BNtransformer ID used in lieu of a fire zone for location purposeWater Spray SKC13Suppression NNNNNtransformer ID used in lieu of a fire zone for location purpose N/ANoneFeaturetransformer ID used in lieu of a fire zone for location purposeThermal007/008Detection NNNNXMA01CNtransformer ID used in lieu of a fire zone for location purposeWater Spray SKC14Suppression NNNNNtransformer ID used in lieu of a fire zone for location purpose N/ANoneFeaturetransformer ID used in lieu of a fire zone for location purposeThermal011Detection NNNNXMA02Ntransformer ID used in lieu of a fire zone for location purposeWater Spray SKC15Suppression NNNNNtransformer ID used in lieu of a fire zone for location purpose N/ANoneFeaturetransformer ID used in lieu of a fire zone for location purposeThermal012/013Detection NNNNXMR01Ntransformer ID used in lieu of a fire zone for location purposeWater Spray SKC16Suppression NNNNNtransformer ID used in lieu of a fire zone for location purpose N/ANoneFeaturetransformer ID used in lieu of a fire zone for location purposeThermal016Detection NNNNXNB01Ntransformer ID used in lieu of a fire zone for location purposeWater Spray SKC19Suppression NNNNNtransformer ID used in lieu of a fire zone for location purpose N/ANoneFeaturetransformer ID used in lieu of a fire zone for location purposeThermal017Detection NNNNXNB02Ntransformer ID used in lieu of a fire zone for location purposeWater Spray SKC20Suppression NNNNNtransformer ID used in lieu of a fire zone for location purpose N/ANoneFeaturetransformer ID used in lieu of a fire zone for location purposeAugust 2011 C-580 Ameren MissouriCallaway Plant NFPA 805 Transition ReportYD-1Yard Area&$.>.?>..A&."&.&Thermal014Detection NNNNXPB03Ntransformer ID used in lieu of a fire zone for location purposeWater Spray SKC17Suppression NNNNNtransformer ID used in lieu of a fire zone for location purpose N/ANoneFeaturetransformer ID used in lieu of a fire zone for location purposeThermal015Detection NNNNXPB04Ntransformer ID used in lieu of a fire zone for location purposeWater Spray SKC18Suppression NNNNNtransformer ID used in lieu of a fire zone for location purpose N/ANoneFeaturetransformer ID used in lieu of a fire zone for location purpose SLER D- Required for Chapter 4 Separation Criteria- Required for NRC-Approved Licensing Action- Required for Existing Engineering Equivalency Evaluation- Required for Risk Significance
- Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk EvaluationBNone&This area contains cabling or electrically supervised equipment that is required for post fire safe shutdown; however those cables and equipment are located in sealed underground cable duct banks or in an enclosed structure with no automatic suppression (i.e., the RWST valve house), and they are unaffected by manual or automatic fire suppression activities at grade level. Therefore, fire suppression activities will not adversely affect the plant's ability to achieve the nuclear safety performance criteria.August 2011 C-581 to ULNRC-06060 
 
ATTACHMENT G: CHANGES TO THE TRANSITION REPORT ATTACHMENT G Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-22 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 ALHV0008 Turbine Driven Auxiliary Feedwater Pump to Steam Generator A Valve Action to isolate potentially spuriously open TDAFW Pump to SG A Valve, ALHV0008. Close manual valve ALV0056 to secure flow from credited TDAFW Pump to SG A. Action taken to isolate flow from credited TDAFW Pump to SG A. TDAFW Pump is credited to supply only SG B (steam supply for TDAFW Pump is credited from SG B only). Motor Driven AFW Pump B is credited to supply only SG D. Action taken to maintain positive control over the rate of RCS cooldown. C-27-013 RA C-27 ALHV0012 Turbine Driven Auxiliary Feedwater Pump to Steam Generator C Valve Action to isolate potentially spuriously open TDAFW Pump to SG C Valve, ALHV0012. Close manual valve ALV0071 to secure flow from credited TDAFW Pump to SG C. Close manual valve ABV0087 to secure steam flow from SG C to credited TDAFW Pump. Action taken to isolate flow from credited TDAFW Pump to SG C, and steam flow from SG C to credited TDAFW Pump. TDAFW Pump is credited to supply only SG B (steam supply for TDAFW Pump is credited from SG B only). Motor Driven AFW Pump B is credited to supply only SG D. Action taken to maintain positive control over the rate of RCS cooldown, and to address the potential for development of excessive differential pressure across the tube sheet of SG C. C-27-014 RA C-27 ALHV0036 Condensate Storage Tank to Turbine Driven Auxiliary Feedwater Pump Valve Action to open potentially spuriously closed Condensate Storage Tank to TDAFW Pump Valve, ALHV0036. Open ALHV0036 feeder breaker NG03CEF4 at Motor Control Center NG03C, and then manually operate (open/close) ALHV0036 with the hand wheel. C-27-015 RA C-27 BBHV8001A Reactor Coolant System Reactor Vessel Head Vent Protection A Upstream and Downstream Valves Action to isolate Reactor Coolant System Reactor Vessel Head Vent Protection A Upstream and Downstream Valves, BBHV8001A and BBHV8002A, due to potential spurious opening of BBHV8001A and BBHV8002A. Fail BBHV8001A and BBHV8002A closed by opening 125VDC breaker NK5109. C-27-016 RA LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-33 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 EGHV0011 / EGHV0013 Essential Service Water to Component Cooling Water A Upstream and Downstream Isolation Valves Action to isolate Essential Service Water to Component Cooling Water A Upstream and Downstream Isolation Valves, EGHV0011 and EGHV0013, due to potential spurious opening of EGHV0011 and EGHV0013. Close manual valve EGV0182. C-27-064 RA C-27 EGHV0012 / EGHV0014 Essential Service Water to Component Cooling Water B Upstream and Downstream Isolation Valves Action to isolate Essential Service Water to Component Cooling Water B Upstream and Downstream Isolation Valves, EGHV0012 and EGHV0014, due to potential spurious opening of EGHV0012 and EGHV0014. Close manual valve EGV0185. C-27-065 RA C-27 EGHV0016 Component Cooling Water Train B Supply/Return Isolation Valve Action to align potentially spuriously closed/failed as-is closed Component Cooling Water Train B Supply/Return Isolation Valve, EGHV0016. Manually locally open EGHV0016 feeder breaker NG04CJF3 at Motor Control Center NG04C, and then locally open EGHV0016 with hand wheel. Action taken to align Component Cooling Water Train B to/from CCW Common Services Header (for Seal Water Heat Exchanger, to support cooling of credited Charging Pump B recirculation flow). C-27-066 RA C-27 EGHV0054 Component Cooling Water Train B Supply Isolation Valve Action to align potentially spuriously closed/failed as-is closed Component Cooling Water Train B Supply Isolation Valve, EGHV0054. Manually locally open EGHV0054 feeder breaker NG04CKF1 at Motor Control Center NG04C, and then locally open EGHV0054 with hand wheel. Action taken to align Component Cooling Water Train B to CCW Common Services Header (for Seal Water Heat Exchanger, to support cooling of credited Charging Pump B recirculation flow). C-27-067 RA LIC-23LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-40 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0102-P Feeder Breaker to Containment Spray Pump A (PEN01A) Action to secure potentially spuriously running non-credited Containment Spray Pump A, to prevent pumped diversion of RWST inventory into Containment. Pull close control fuses for breaker NB0102 at Switchgear NB01 and then manually trip the breaker. C-27-089 RA C-27 NB0103-P Feeder Breaker to Safety Injection Pump A (PEM01A) Action to secure potentially spuriously running non-credited Safety Injection Pump A, to prevent pumped diversion of RWST inventory through SI test lines and/or SIS accumulator tank relief valves. Pull close control fuses for breaker NB0103 at Switchgear NB01 and then manually trip the breaker. C-27-090 RA C-27 NB0104-P Feeder Breaker to Charging Pump A (PBG05A) Action to secure potentially spuriously running non-credited Charging Pump A, to prevent PORV challenge due to Pressurizer overfill. Pull the close circuit fuses for breaker NB0104 at Switchgear NB01 and then trip the breaker. C-27-091 RA C-27 NB0105-P Feeder Breaker to Motor Driven Auxiliary Feedwater Pump A (PAL01A) Action to secure potentially spuriously running non-credited Motor Driven AFW Pump A. Pull the close circuit fuses for breaker NB0105 at Switchgear NB01 and then trip the breaker. Action taken to maintain positive control over the rate of RCS cooldown. C-27-092 RA C-27 NB0201 Feeder Breaker to Charging Pump B (PBG05B) Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear Feeder Breaker to Charging Pump B (PBG01B), NB0201, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0201 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0201 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0201 with the local close/trip control hand switch C-27-093 RA LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-41 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. Action taken to start credited Charging Pump B. C-27 NB0201-P Feeder Breaker to Charging Pump B (PBG05B) Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear Feeder Breaker to Charging Pump B (PBG01B), NB0201, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0201 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0201 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0201 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. Action taken to start credited Charging Pump B. C-27-094 RA C-27 NB0202-P Feeder Breaker to Safety Injection Pump B (PEM01B) Action to secure potentially spuriously running non-credited Safety Injection Pump B (PEM01B), to prevent pumped diversion of RWST inventory through SI test lines and/or SIS accumulator tank relief valves, and to manually load shed Train B Switchgear NB02 prior to re-energization of NE02 with the Train B Emergency Diesel Generator, NE02. Pull close control fuses for breaker NB0202 at Switchgear NB02 and then manually trip the breaker. C-27-095 RA LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-42 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0203-P Feeder Breaker to Containment Spray Pump B (PEN01B) Action to secure potentially spuriously running non-credited Containment Spray Pump B (PEN01B), to prevent pumped diversion of RWST inventory into Containment, and to manually load shed Train B Switchgear NB02 prior to re-energization of NE02 with the Train B Emergency Diesel Generator, NE02. Pull close control fuses for breaker NB0203 at Switchgear NB02 and then manually trip the breaker. C-27-096 RA C-27 NB0204-P Feeder Breaker to Residual Heat Removal Pump B (PEJ01B) Action to trip potentially spuriously closed 4kV Switchgear Feeder Breaker to Residual Heat Removal Pump B (PEJ01B), NB0204, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0204 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0204 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-097 RA C-27 NB0206 Feeder Breaker to Component Cooling Water Pump B (PEG01B) Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear Feeder Breaker to Component Cooling Water Pump B (PEG01B), NB0206, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0206 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0206 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0206 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-098 RA LIC-23LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-43 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0206-P Feeder Breaker to Component Cooling Water Pump B (PEG01B) Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear Feeder Breaker to Component Cooling Water Pump B (PEG01B), NB0206, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0206 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0206 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0206 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-099 RA C-27 NB0207 Feeder Breaker to Component Cooling Water Pump D (PEG01D) Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear Feeder Breaker to Component Cooling Water Pump D (PEG01D), NB0207, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0207 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0207 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0207 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-100 RA LIC-23LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-44 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0207-P Feeder Breaker to Component Cooling Water Pump D (PEG01D) Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear Feeder Breaker to Component Cooling Water Pump D (PEG01D), NB0207, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0207 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0207 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0207 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-101 RA C-27 NB0208 4kV Switchgear NB02 Feeder Breaker to Load Center PG22 Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear NB02 Feeder Breaker to Load Center PG22, NB0208, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0208 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0208 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0208 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-102 RA LIC-23LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-45 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0208-P 4kV Switchgear NB02 Feeder Breaker to Load Center PG22 Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear NB02 Feeder Breaker to Load Center PG22, NB0208, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0208 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0208 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0208 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-103 RA C-27 NB0209 Normal Offsite Power Feeder Breaker to Essential 4kV Switchgear NB02 (from Essential Transformer XNB02) Action to trip potentially spuriously closed/failed as-is closed Normal Offsite Power Feeder Breaker to Essential 4kV Switchgear NB02 (from Essential Transformer XNB02), NB0209, to manually load shed Train B Switchgear NB02 prior to re-energization of NE02 with the Train B Emergency Diesel Generator, NE02. Pull close control fuses for breaker NB0209 at Switchgear NB02 and then manually trip the breaker. C-27-104 RA LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-46 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0210 4kV Switchgear NB02 Feeder Breaker to Load Center NG04 Action to close potentially spuriously tripped 4kV Switchgear NB02 Feeder Breaker to Load Center NG04, NB0210, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0210 fro m the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally close breaker NB0210 with the local close/trip control hand switch to load NB02 in preparation for, or following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-105 RA C-27 NB0211 Emergency Feeder Breaker to Essential 4kV Switchgear NB02 (from Emergency Diesel Generator NE02) Action to trip and then close potentially spuriously closed/tripped Emergency Feeder Breaker to Essential 4kV Switchgear NB02 (from Emergency Diesel Generator NE02), NB0211, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0211 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0211 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02 (if EDG NE02 not already supplying power to NB02). Locally close breaker NB0211 with the local close/trip control hand switch to load NB02 following start of EDG NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-106 RA LIC-23LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-47 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0212 Alternate Offsite Power Feeder Breaker to Essential 4kV Switchgear NB02 (from Essential Transformer XNB01) Action to trip potentially spuriously closed Alternate Offsite Power Feeder Breaker to Essential 4kV Switchgear NB02 (from Essential Transformer XNB01), NB0212, to manually load shed Train B Switchgear NB02 prior to re-energization of NE02 with the Train B Emergency Diesel Generator, NE02. Pull close control fuses for breaker NB0212 at Switchgear NB02 and then manually trip the breaker. C-27-127 RA C-27 NB0213 4kV Switchgear NB02 Feeder Breaker to Load Center NG02 Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear NB02 Feeder Breaker to Load Center NG02, NB0213, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0213 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0213 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0213 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-107 RA LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-48 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0214 Alternate Emergency Power System Feeder Breaker to Essential 4kV Switchgear NB02 (from 4kV Switchgear PB05) Action to trip potentially spuriously closed Alternate Emergency Power System Feeder Breaker to Essential 4kV Switchgear NB02, NB0214, to manually load shed Train B Switchgear NB02 prior to re-energization of NE02 with the Train B Emergency Diesel Generator, NE02. Pull close control fuses for breaker NB0214 at Switchgear NB02 and then manually trip the breaker. C-27-126 RA C-27 NB0215 Feeder Breaker to Essential Service Water Pump B (PEF01B) Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear NB02 Feeder Breaker to Essential Service Water Pump B (PEF01B), NB0215, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0215 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0215 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02 (if EDG NE02 not already supplying power to NB02). Locally close breaker NB0215 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02.
Action taken to start credited Essential Service Water Pump B. C-27-108 RA LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-49 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0216 4kV Switchgear NB02 Feeder Breaker to Motor Control Center NG06E Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear NB02 Feeder Breaker to Motor Control Center NG06E, NB0216, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0216 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0216 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0216 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-109 RA C-27 NB0217 4kV Switchgear NB02 Feeder Breaker to Load Center NG08 Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear NB02 Feeder Breaker to Load Center NG08, NB0217, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0217 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0217 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0217 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-110 RA LIC-23LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-50 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 NB0217-P 4kV Switchgear NB02 Feeder Breaker to Load Center NG08 Action to trip and then close potentially spuriously closed/tripped 4kV Switchgear NB02 Feeder Breaker to Load Center NG08, NB0217, for re-energization of NB02 from the Train B Emergency Diesel Generator, NE02. Locally isolate control circuit for breaker NB0217 from the Main Control Room using hand switch NBHS0014 on Switchgear NB02, and then locally trip breaker NB0217 with the local close/trip control hand switch to load shed NB02 in preparation for re-energization from NE02. Locally close breaker NB0217 with the local close/trip control hand switch to load NB02 following re-energization from NE02. Action taken to align the Train B Class 1E Electrical Distribution System from the credited AC power source, Train B Emergency Diesel Generator B, NE02. C-27-111 RA C-27 NE02 Emergency Diesel Generator NE02 Action to recover local control of Emergency Diesel Generator NE02 and associated auxiliaries (Starting Air Supply Pressure Control Valves A and B, and Starting Fuel Rack Air Supply Pressure Control Valve). If Switchgear NB02 already energized from generator NE02, locally isolate control circuit(s) for generator NE02 from the Main Control Room using switches KJHS0203 and NEHS0032 located on Panels KJ122 and NE106 respectively. Transfer generator NE02 to loc/man using switch KJHS0109 located on Panel KJ122. If Switchgear NB02 not already energized from generator NE02, ensure breaker NB0211 is open, ensure lockout relays 186-1/DG and 186-2/DG are reset on Panel NE106, locally isolate control circuit(s) for generator NE02 from the Main Control Room using switches KJHS0203 and NEHS0032 located on Panels KJ122 and NE106 respectively. Transfer generator NE02 to loc/man using switch KJHS0109 located on Panel KJ122, and start generator NE02 using switch KJHS0101D or KJHS0101C located on Panel KJ122. C-27-112 RA LIC-23 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page G-53 Table G Recovery Actions and Activities Occurring at the Primary Control Station(s)
FireArea Component Component Description Actions VFDR RA/PCS C-27 PJE01B Emergency Diesel Generator Fuel Oil System Storage Tank B - Fuel Oil Transfer Pump B Action to recover Emergency Diesel Generator Fuel Oil System Storage Tank B - Fuel Oil Transfer Pump B, PJE01B. Locally isolate control circuit for pump PJE01B from the Main Control Room using switch JEHS0021A at Motor Control Center NG04D, and start pump using switch JEHS0021A at Motor Control Center NG04D. C-27-121 RA C-27 SGK05B Train B ESF Switchgear Rooms Air Conditioning Unit Action to recover Train B ESF Switchgear Rooms Air Conditioning Unit, SGK05B. Locally isolate control circuit for SGK05B from the Main Control Room using switch GKHS0103 located on the east side of the HVAC room (isolation position also starts unit). C-27-122 RA C-27 SGL15B Auxiliary Building South Electrical Penetration Room Cooler Action to recover Auxiliary Building South Electrical Penetration Room Cooler, SGL15B. Locally isolate control circuit for SGL15B from the Main Control Room using switch GLHS0035 at Motor Control Center NG02B, and then start cooler when directed using pushbutton switch GLHS0195 at Motor Control Center NG02B. C-27-123 RA C-27 SGN01B Containment Cooler B Action to recover Containment Cooler B, SGN01B. Locally isolate control circuit for SGN01B from the Main Control Room using switch GNHIS0009A at Motor Control Center NG02T, and then start unit when directed using switch GNHIS0009A at Motor Control Center NG02T. C-27-124 RA C-27 SGN01D Containment Cooler D Action to recover Containment Cooler D, SGN01D. Locally isolate control circuit for SGN01D from the Main Control Room using switch GNHIS0017A at Motor Control Center NG04T, and then start unit when directed using switch GNHIS0017A at Motor Control Center NG04T. C-27-125 RA C-30 BMHV0001 Steam Generator A Blowdown Isolation Valve Action to isolate potentially spuriously open Steam Generator A Blowdown Isolation Valve, BMHV0001. Fail BMHV0001 closed by opening 125VDC breaker NK4111. Action taken to maintain positive control over SG level. C-30-003 RA LIC-23LIC-23 to ULNRC-06060 
 
ATTACHMENT S: CHANGES TO THE TRANSITION REPORT ATTACHMENT S ItemUnitDescriptionLAR Section / SourceTable S-3 Implementation ItemsAmeren Missouri Callaway Plant N FPA 805 Transition Report 13-805-008The requirement to enter Technical Specification 3.0.3 will be removed from FSAR Table 9.5.1-2, Note 2. The NFPA 805 3.2.3(2) requirement to establish compensatory actions is included in procedure APA-ZZ-00703 Fire Protection Operability Criteria And Surveillance Requirements. The requirement to enter Technical Specification 3.0.3 will also be removed from this procedure.
Attachment X 1August 2011 Page S-21 FPE RAI 20 to ULNRC-06060 
 
ATTACHMENT W: CHANGES TO THE TRANSITION REPORT ATTACHMENT W Fire AreaArea DescriptionNFPA 805 BasisFire Area CDF/LERF VFDR(Yes/No)RAs(Yes/No)Fire Risk EvalFAttachment W - Table W-2  Fire Area Risk SummaryAmeren MissouriCallaway Plant NFPA 805 Transition ReportN/AC-34Control Building Cable Chase B at column C-6, Control Building, El. 2073-6N/A/NoNo1.10E-111.85E-15/4.2.3.2C-35Control Building Corridor, Control Building, El. 2016/NoYes/4.2.4.25.97E-10C-36Control Building Cable Chase B at column C-6, Control Building, El. 20005.27E-13/NoYes5.97E-105.27E-13/4.2.4.2N/AC-37Control Building Cable Chase A at column  C-3,  Control Building, El. 2000N/A/NoNo1.15E-111.65E-14/4.2.3.2N/AD-1Diesel Generator A, Diesel Generator Building, El. 2000N/A/NoNo1.08E-086.71E-11/4.2.3.2N/AD-2Diesel Generator B, Diesel Generator Building, El. 2000N/A/NoNo1.10E-087.23E-11/4.2.3.24.56E-08FB-1Fuel Handling Building7.98E-12/NoYes5.39E-081.99E-11/4.2.4.2N/ALDF-1Laundry Decontamination FacilityN/A/NoNo3.78E-095.42E-12/4.2.3.22.36E-07RB-1Reactor Building1.93E-09/YesYes2.36E-071.93E-09/4.2.4.2N/ARSB-1RAM Storage BuildingN/A/NoNo/4.2.3.2August 2011Page W-20 LIC-26 Fire AreaArea DescriptionNFPA 805 BasisFire Area CDF/LERF VFDR(Yes/No)RAs(Yes/No)Fire Risk EvalFAttachment W - Table W-2  Fire Area Risk SummaryAmeren MissouriCallaway Plant NFPA 805 Transition ReportN/ARW-1Radwaste BuildingN/A/NoNo3.76E-085.40E-11/4.2.3.20.00E+00TB-1Turbine Building0.00E+00/NoYes6.54E-061.36E-07/4.2.4.2N/AUNCTUHS North Cooling TowerN/A/NoNo2.73E-094.61E-13/4.2.3.2N/AUNPHEssential Service Water Pump Room AN/A/NoNo2.69E-092.62E-11/4.2.3.2N/AUSCTUHS South Cooling TowerN/A/NoNo2.73E-094.61E-13/4.2.3.2N/AUSPHEssential Service Water Pump Room BN/A/NoNo2.72E-092.69E-11/4.2.3.21.68E-08YD-1Yard Area2.94E-12/NoYes1.03E-062.18E-08/4.2.4.22.03E-053.99E-072.18E-064.24E-08TOTALS://August 2011Page W-21 LIC-26 to ULNRC-06060 
 
ATTACHMENT X: CHANGES TO THE TRANSITION REPORT ATTACHMENT X Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page X-1 X. Other Requests for Approval 7 Pages Attached LIC-26 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page X-2 Approval Request 1 On February 19, 1987 Callaway Plant submitted a license amendment request via ULNRC-01447 to delete fire protection Technical Specifications and relocate those requirements to the FSAR under licensee control in accordance with Generic Letter 86-10, Implementation of Fire Protection Requirements. On October 30, 1987 Callaway Plant responded to NRC questions related to this license amendment request via ULNRC-01667. Specifically, the following question and response is documented in ULNRC-01667.
NRC Question The shutdown requirement of Specification 3.7.10.1 ACTION b should be retained in an appropriate commitment document.
Response As part of implementing the proposed revisions to the Technical Specifications, the requirements of Specification 3.7.10.1 ACTION b will be retained and will not be modified without prior approval from the Nuclear Regulatory Commission (NRC). The requirements of Specification 3.7.10.1 ACTION b will be added to FSAR (USAR for Wolf Creek) Table 9.5.1-2 with a statement that no modifications to these requirements will be made without prior approval of the NRC. On January 13, 1988 the NRC issued Amendment No. 30 to Facility Operating License No. NPF-30. In the accompanying safety evaluation the staff noted the following:  The licensee had originally proposed to delete the shutdown requirement of Specification 3.7.10.1 Action b. The staff's position is that the loss of the normal fire protection water supply and the inability to establish a back-up fire suppression water system within 24 hours warrant plant shutdown. The licensee responded that the requirements of Specification 3.7.10.1 Action b. will be added to the FSAR with commitment that no modifications to these requirements will be made without prior approval from NRC. The staff considers this response to be acceptable. In response to the above, Callaway Plant has maintained the following statement in FSAR Table 9.5.1-2 for the Fire Suppression Water System, requirements a, b and c. With the Fire Suppression Water System in this condition, establish a backup Fire Suppression Water System within 24 hours. If this required action cannot be met, the requirements of Technical Specification 3.0.3 shall be initiated. Modifications to these requirements shall not be made without prior approval of the NRC. As part of the transition to NFPA 805, it is being requested that the NRC Staff review and approve the elimination of the requirement currently listed in FSAR Table 9.5.1-2 to enter Technical Specification LCO 3.0.3 for an inoperable Fire Suppression Water System coupled with the inability to provide a backup fire suppression water system within 24 hours.
Basis for Request: The current application of Technical Specification LCO 3.0.3 for plant configurations that do not meet the specific criteria of 10 CFR 50.36(c)(2) for inclusion into the plant Technical Specifications is inappropriate. Technical Specification LCO 3.0.3 is intended to be applied when a Technical Specification LCO is not met and the associated Technical Specification required actions are not met, an associated Technical Specification required action is not provided, or if directed by the Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page X-3 associated Technical Specification required actions. Technical Specification LCO 3.0.3 was not meant to be applied to non-technical specification plant configurations. The requirements for FSAR Table 9.5.1-2, Fire Protection System Requirements, were previously approved for relocation from the Technical Specifications to the FSAR in an NRC SER dated January 13, 1988. The commitment made by Union Electric in ULNRC-01667 was that the requirements of Specification 3.7.10.1 ACTION b would be added to the FSAR with a statement that no modifications to these requirements will be made without prior approval of the NRC. The current FSAR Table 9.5.1-2 text is as follows: With the Fire Suppression Water System in this condition, establish a backup Fire Suppression Water System within 24 hours. If this required action cannot be met, the requirements of Technical Specification 3.0.3 shall be initiated. Modifications to these requirements shall not be made without prior approval of the NRC. The existing requirement to enter Technical Specification LCO 3.0.3 would not be consistent with NFPA 805 which indicates that compensatory actions should be appropriate with the level of risk created by the unavailable equipment. NFPA 805, Section 3.2.3 states: "Procedures shall be established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established: (2) Compensatory actions implemented when fire protection systems and other systems credited by the fire protection program and this standard cannot perform their intended function and limits on impairment duration." Section 3.2.3 is supplemented by the following guidance from NFPA 805 Appendix A: "A.3.2.3(2) Compensatory actions might be necessary to mitigate the consequences of fire protection or equipment credited for safe shutdown that is not available to perform its function. Compensatory actions should be appropriate with the level of risk created by the unavailable equipment. The use of compensatory actions needs to be incorporated into a procedure to ensure consistent application. In addition, plant procedures should ensure that compensatory actions are not a substitute for prompt restoration of the impaired system." As stated in Attachment A, NEI 04-02 Table B-1, Callaway Plant procedure APA-ZZ-00703, Fire Protection Operability Criteria and Surveillance Requirements, will be used to establish the required compensatory actions and impairment durations following the transition to NFPA 805. FSAR Table 9.2.1-2 will be eliminated.
Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page X-4 Nuclear Safety and Radiological Release Performance Criteria: An inoperable Fire Suppression Water System with the inability to establish a backup system within 24 hours is an off-normal occurrence. NFPA 805 Section 3.2.3(2) requires compensatory actions to be implemented when fire protection systems and other systems credited by the fire protection program cannot perform their intended function. NFPA 805 Section 3.2.3(2) also requires that limits be established on the impairment duration. Callaway Plant procedure APA-
 
ZZ-00703, Fire Protection Operability Criteria and Surveillance Requirements, will be used to establish the required compensatory actions and impairment durations following the transition to NFPA 805. There is no impact on the nuclear safety performance criteria. An inoperable Fire Suppression Water System with the inability to establish a backup system within 24 hours has no impact on the radiological release performance criteria. The radiological release performance criteria are satisfied based on the determination of limiting radioactive release (Attachment E), which is not affected by impacts on the fire protection system due to this condition. Safety Margin and Defense-in-Depth: An inoperable Fire Suppression Water System with the inability to establish a backup system within 24 hours is an off-normal occurrence. NFPA 805 Section 3.2.3(2) requires compensatory actions to be implemented when fire protection systems and other systems credited by the fire protection program cannot perform their intended function. NFPA 805 Section 3.2.3(2) also requires that limits be established on the impairment duration. Callaway Plant procedure APA-ZZ-00703, Fire Protection Operability Criteria and Surveillance Requirements, will be used to establish the required compensatory actions and impairment durations following the transition to NFPA 805. Because actions resulting from this condition will not deviate from the approved fire protection program, there is no impact on safety margin and defense-in-depth.
 
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Conclusion:==
NRC approval is requested to eliminate the current requirement to enter Technical Specification 3.0.3 for an inoperable Fire Suppression Water System with the inability to establish a backup system within 24 hours. NFPA 805 Section 3.2.3(2) requires compensatory actions to be implemented when fire protection systems and other systems credited by the fire protection program cannot perform their intended function. NFPA 805 Section 3.2.3(2) also requires that limits be established on the impairment duration. As identified in the Section 3.2.3(2) compliance basis in Attachment A, Callaway Plant procedure APA-ZZ-00703, Fire Protection Operability Criteria and Surveillance Requirements, will be used to establish the required compensatory actions and impairment durations following the transition to NFPA 805. FSAR 9.2.1-2 will be eliminated following NRC approval of this request and during the transition to NFPA 805.
Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page X-5 Approval Request 2 Approval is requested for a deviation from common enclosure analysis requirements of NFPA 805 Section 2.4.2 for specific Current Transformer (CT) configurations where a fire-induced open-circuit failure could result in a secondary fire. A fire in plant fire area C-21, Lower Cable Spreading Room, Control Building El. 2032' or in plant fire area C-27, Main Control Room, Control Building El. 2047', results in an open circuit failure for circuits associated with the Main Generator Current Transformers (CTs) TVMA10A, B,
 
and C. Due to the design of the CTs (turns ratio, relaying accuracy class, isolation circuit design) a secondary fire due to overheating can occur in plant fire area TB-1 at the CTs. Requirements Section 2.4.2 of NFPA 805 requires consideration of fire-induced open-circuit failure modes and specifies that circuits which share a common enclosure with circuits required to achieve nuclear safety performance criteria, be evaluated to ensure that such electrical faults will not cause the fire to extend beyond the immediate (initial) fire area. As discussed in NFPA 805 B.3.4.2 the evaluation of common enclosure issues should include consideration of Current Transformers that are constructed such that an open secondary circuit could cause ignition of the transformer.
Specific details of the requirements are addressed below. NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis Section 2.4.2.2.2 of NFPA 805 states: "Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries." NFPA 805 Section B.3.4.2 of appendix B to NFPA 805 states in part: A special type of common enclosure issue involves current transformers.  -An opening in the secondary circuit causes excessively high voltages in the current transformer secondary circuit in an attempt to maintain this ratio, which can result in an ignition of the transformer materials.
Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page X-6 Regulatory Guide 1.205 Rev 1 Section 3.3, "Circuit Analysis," states: Chapter 3 of industry guidance document NEI 00-01, "Guidance for Post-Fire Safe Shutdown Circuit Analysis," Revision 2, issued May 2009 (Ref. 12), when used in conjunction with NFPA 805 and this regulatory guide, provides one acceptable approach to circuit analysis for a plant implementing an FPP under 10 CFR 50.48(c). NEI 00-01 Section 3.5.2.1, "Circuit Failure Due to An Open Circuit," states: This section provides guidance for addressing the effects of an open circuit for safe shutdown equipment. An open circuit is a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit will typically prevent the ability to control or power the affected equipment. An open circuit can also result in a change of state for normally energized equipment. For example, a loss of power to the main steam isolation valve (MSIV) solenoid valves [for BWRs] due to an open circuit will result in the closure of the MSIV. NOTE: The EPRI circuit failure testing indicated that open circuits are not likely to be the initial fire-induced circuit failure mode. Consideration of this may be helpful within the safe shutdown analysis. Consider the following consequences in the safe shutdown circuit analysis when determining the effects of open circuits:  Loss of electrical continuity may occur within a conductor resulting in deenergizing the circuit and causing a loss of power to, or control of, the required safe shutdown equipment. In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or other device. This loss of power may change the state of the equipment. Evaluate this to determine if equipment fails safe. Open circuit on a high voltage (e.g., 4.16 kV) ammeter current transformer (CT) circuit may result in secondary damage.
Basis for Request: The following assumes a plant fire results in an open circuit and resultant overheating and secondary fire at one or more of the Main Generator CTs TVMA10A, B, and C. Additionally, when the secondary fire occurs and the site fire brigade is assumed to be unavailable to respond due to the initial fire in either C-21 or C-27, the overheating and secondary fire at the CTs will result in a Main Generator trip and subsequent plant trip and also result in de-energizing the CTs. Therefore, the basis below addresses the secondary fire and its effects. The TVMA10A, B, and C CTs are doughnut CTs placed around the generator bushings located under the Main Generator. The generator sits over a rectangular opening in the concrete turbine deck. The generator bushings are in the opening that comes out of the bottom of the generator. Figure X-1 shows the typical arrangement for the CTs. There are no redundant systems, cables or components required to meet the Nuclear Safety Capability Assessment that will be affected by the secondary fire. The fire will be limited to the CTs due to the lack of insitu combustibles surrounding the CTs. The area around the CTs is either concrete or metal. Other than the CTs there are no close exposed combustibles. The field leads and wiring to the CTs are in enclosed metal conduit and will not propagate fire. The generator bushings are composed of ceramic material and are not affected by the CT overheating or fire.
Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page X-7  The fire will not be challenging or severe due to the lack of insitu combustibles design of the CTs. Due to the lack of insitu combustibles the heat release rate resulting from the fire will be not be significant. Due to the size and open nature of the Turbine Building a hot gas layer is not possible from this fire. The fire will not result in ignition of any adjacent transient combustibles that could result in fire growth beyond the CTs. Access to the CTs is by ladder from the 2033 elevation to an elevated walkway with grated flooring. As indicated in Figure X-1, there is significant free space between the CTs and the elevated walkway. As defense-in-depth, the area of the TB-1 elevation 2033 is protected by a full area wide automatic pre-action suppression system.
Figure X Typical Configuration of Main Generator CTs Acceptance Criteria Evaluation:
Nuclear Safety and Radiological Release Performance Criteria: The secondary fire will be specific to the CT transformers and will not affect any redundant systems, cables or components required to meet the Nuclear Safety Capability Assessment for fires in C-21 or C-27 the initiating event fire areas. The secondary fire would have no impact on the radiological release performance criteria. The CTs and the area surrounding the CTs is outside the permanent radiological controlled area Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page X-8 (RCA) and it is not a storage location for radioactive materials. The access to the CTs is by ladder to a raised platform with grated flooring. Safety Margin and Defense-in-Depth: There are conservatisms in the circuit failure analysis. The safety margin in the analysis for the fire event in C-21 and C-27 has been preserved. The postulated secondary fire will not affect any assumptions or analysis utilized for the evaluation of fire affects in fire areas C-21 and C-27. As defense-in-depth, the area of the TB-1 elevation 2033 is protected by a full area wide automatic pre-action suppression system. The defense-in-depth analysis for the fire events in C-21 or C-27 has been preserved and is not impacted by the secondary TB-1 CT fire.
 
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Conclusion:==
Approval is requested for a deviation from common enclosure analysis requirements of NFPA 805 Section 2.4.2 for specific CT configurations where a fire-induced open-circuit failure could result in a secondary fire. A fire in plant fire area C-21, Lower Cable Spreading Room, Control Building El. 2032' or in plant fire area C-27, Main Control Room, Control Building El. 2047', could result in a secondary fire at the Main Generator CTs TVMA10A, B, and C located in fire area TB-1. However the secondary fire will have no adverse impact on meeting the Nuclear Safety Capability criteria for fires in C-21 and C-27.}}

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