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| t APPENDIX 190 ABWR SHUTDOWN RISK EVALUATION f
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| b N
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| 1
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| -PDR 9207070276 920702 A ADOCK 05200001 PDR
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| . , . . - , , . . . . . - . . . . . . . ....-..:. . - . . =, - - - - .
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| l TABLE OF CONTENTS APPENDIX 19Q: ABWR SHUTDOWN RISK EVALUATION SECTI,0N PAGE 19Q.1 INTRODUCTION 3 19Q.2 EVALUATION SCOPE 3 19Q.3
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| ==SUMMARY==
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| OF RESULTS 5 _
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| 19Q.4 FEATURES TO MINIMIZE SHUTDOWN RISK 8 19Q.4.1 DECAY HEAT REMOVAL 9 19Q.4.2 INVENTORY CONTROL 18 19Q.4.3 CONTAINMENT INTEGRITY 24 19Q.4.4 ELECTRICAL POWER 25 19Q.4.5 REACTIVITY CONTROL 27 19Q.4.6
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| ==SUMMARY==
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| OF SHUTDOWN RISK CATEGORY ANALYSIS 30 19Q.5 INSTRUMENTATION 31 19Q.6 FLOODING AND FIRE PROTECTION 33 19Q.7 DECAY HEAT REMOVAL RELIABILITY STUDY 39 19Q.
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| ==7.1 INTRODUCTION==
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| 39 19Q.7.2 PURPOSE 40 19Q.7.3
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| ==SUMMARY==
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| 40 19Q.7.4 METHODOLOGY 42 19Q.7.5 CORE DIMAGE PROBABILITY GOAL AND RPV BOILING 43 19Q.7.6 SUCCEfS CRITERIA 45 19Q.7.7 ACCIDENT PROGRESSION AND EVENT TREES 46 19Q.7.8 SYSTEM FAULT TFIES 50 19Q.7.9 RESULTS AND CONCLUSIONS 69
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| l TABLE OF CONTENTS APPENDIX 19Q: ABWR SHUTDOWN RISK EVALUATION (Continued)
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| SECTION PAGE 19Q,0 USE OF FREEZE SEALS IN ABWR 76 19Q.9 SHUTDOWN VULNERABILITY RESULTING FROM NEW FEATURES 77 19Q.10 PROCEDURES 78 19Q 11 SUMMAR7 OF REVIEW OF SIGNIFICANT SHUTDOWN EVENTS:
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| ELECTRICAL POWER AND DECAY HEAT REMOVAL 82 19Q.12 RESULTS AND INTERFACE REQUIREMENTS 87 19Q.12.1 INSIGHTS GAINED FROM THE ANALYSIS 87 19Q.12.2 IMPORTANT DESIGN FEATURES 88 19Q.12.3 OPERATOR ACTIONS 89 19Q.12.4 RELIABILITY GOALS 90 19Q.
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| ==12.5 CONCLUSION==
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| S 91 ATTACHMENT 19QA DECAY FEAT RZMOVAL RELIABILITY STUDY FAULT TREES 92 ATTACHMENT 19QB DECAY HEAT REMOVAL RELIABILITY STUDY OFFSITE DOSE AND OPERATOR RECOVERY CALCULATIONS 93 ATTACHMENT 19QC REVIEW OF SIGNIFICANT SHUTDOWN EVENTS:
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| ELECTRICAL POWER AND DECAY HEAT REMOVAL 104
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| . . _. ,_ . . - . . - - - _ ~ . _ . . _ . , _ , . _ . _ . . . . - - . _ .- _ _ _ . . . _ _ . . __ _ _ . . _ _ _
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| APPENDIX 190t' ABWR SHUTDOWN RISK EVALUATION 19Q.1 INTRODUCTION Due to events at operating plants in the past several years such as the !
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| loss of off-site power at Vogtle on March 20, 1990 and the loss of decay heat removal (DER) at Diablo Canyon on April 10, 1987, the shutdown risk associated
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| .with nuclear power plants has become more of a concern to the industry. On
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| . January 17, 1992 the NRC issued Draft NUREG- 1449, "NRC Staff Evaluation of Shutdown and Low Power Operation". In NUREG-1449 the NRC staff Identified some
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| - safety. issues that may result in new regulatory requirements.
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| As part of the certification process for the Advanced Boiling Water Reactor (ABWR),-an evaluation of the shutdown risk associated with the ABWR was completed. This Appendix discusses the design and procedural features of the ABWR that contribute to the conclusion that the ABWR shutdown risks are negligible.~
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| 19Q.2 EVALUATION SCOPE The ASWR shutdown risk evaluation covers the important aspects of NUREG-1449 as well as specific items requested by the NRC.
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| The evaluation encompasses plant-operation in modes 3 (hot shutdown),
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| 4 (cold shutdown), and 5 (refueling). The ABWR full power PRA covered operation
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| - in Modes 1 (power operation) and 2 (startup/ hot standby). This evaluation ~
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| - addresses conditions for which there is. fuel in the reactor pressure vessel (RPV). It includes all aspects of the Nuclear Steam Supply System (NSSS), the containment, and all systems that support operation of the NSSS and containment.
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| D . It does not address events involving fuel handling outside the primary containment or fuel storage in the spent fuel pool.
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| _3
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| , - . . - . , . --,-,u._=- ,_ , ,, -.
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| The evaluation was broken down into several topics covering design- ,
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| pro:edures, and ABWR features that have the potential to prevent / mitigate past operating events that are' considered precursors to loss of decay heat removal capability and fuel damage. The design issues included decay heat removal,
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| ' inventory control, containment integrity, electrical , ver, reactivity control, and instrumentation. Guidelines for generation of ASWR procedures are covered in a separate section, as well as the risk implications of using freeze seals during ABWR maintenance.
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| In NUREG-1449 it was pointed out that due to the increased level of maintenance activity while shutdown, the potential for fires and flooding in operating nuclear plants is considered higher during shutdown. These topics are covered separately to highlight the ABWR features designed to minimize the shutdown risks from fires and flooding.
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| In order to evaluate the ABWR features that are capable of preventing or mitigating safety significant events that have occurred at operating plants in the past, a study was completed of specific past events that resulted either in
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| : a. loss of offsite power or a challenge to DHR. Loss of power events as described in NUREG-1410, " Loss of Vital AC Power.and the Residual Heat Removal 1
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| System During Mid-Loop Operation at Vogtle Unit 1 on March 20, 1990" were
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| ~
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| c evaluated and ABWR features which could have prevented / mitigated the event were ,
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| described. _A total of 74 loss of power events were evaluated. -In s like manner,-events described in NSAC-88, " Residual Heat Removal Experience Review and Safety Analysis - Boiling Water Reactors" were' reviewed along with certain
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| . loss of DER events from INPO Significant Evaluation Reports (SERs) and Significant Operating Experience Reports (SOERs) and NRC Information Notices.
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| ~Over 100 precursor events to loss of LHR were reviewed.
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| !? To ensure that new-features (i.e., different than current operating BWRs)
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| > of the ABWR do not' introduce any additional vulnerabilities to operation of the i
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| plant, a Failure Modes and Effects Analysis (FMEA) was completed on these new j-features. The FKEA focused on the potential safety impact of identified failure
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| ! modes and why these do not contribute to increased risk of ABWR shutdown operation.
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| ' . . - - . . - ~ .. - , . - . . . . - - . _ _ - . . _ . _ - . . . _ . .
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| v Lattly,'a detailed' reliability study was completed of the ABWR DHR_.
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| ; function. .Probabilistic risk assessment (PRA) models including Fault and Event Trees were completed for all DHR and makeup systems. Based on PRA results,.
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| 1 minimum sets of. equipment were identified that, if available, would result in acceptable shutdown risk.
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| Based on this shutdown risk evaluation, input has been provided to other parts of the SSAR. Systems and components important to safety were identified for inclusion in the reliability assurance program. COL action items such as a need for shutdown procedures-and important operator. actions were specified.
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| Plant' features important to risk reduction were 4dentified and made part of the ITAACS.
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| 19Q.3
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| ==SUMMARY==
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| OF RESULTS The ABWR design has been evaluated for risks associated with shutdown conditions (i.e., modes 3,- 4,.and 5). The. evaluation included the following
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| . shutdown risk categories discussed in NUREG-1449:
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| Decay heat removal
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| - . Inventory. control
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| - containment integrity
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| - Loss of electrical power Reactivity control The evaluation also included shutdown risk reduction features of the ABWR design
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| -due to instru. mentation, flooding and fire protection, use of freeze seals, and procedure guidelines. ABWR features that are not part of current domestic BWR designs were evaluated to determine if.any new shutdown risk vulnerabilities would'be introduced. -Finally, minimum sets of plant systems that if available would meet a goal of a conditional core melt probability-(i.e., given loss of
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| -one RER system) of 1.0 x 10~4 were identified.
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| - - __ - . .. - - . . . , - . - _ . . . . .~. .. . ~ - . ._ - . - .~ . - _ .. - ._ _ . .
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| 'I The results of this shutdown risk evaluation. demonstrate that the ABWR incorporates design features ihich make the plant risk during shutdown negligible. This conclusion is based on the following principal ABWR features which are capable of mitigating shutdown risket j
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| ' Shutdown Risk Concern Principal ABWE Feature j
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| ;=
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| Decay Heat Removal Three physically and electrically independent RHR and support systems i
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| -Inven',ory Control Multiplo makeup systems and sources ,
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| Loss of Electrical Power Two offsite and four onsite power sources
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| * Reactivity Control RPS and standby liquid control systems and interlocks to prevent accidental reactivity excursions r
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| The ABWR is' adequately protected from internal flooding by redundant floor drains, sump pumps, watertight doors, water level alarms, automatic isolation of flow sources, raised sills on doors, equipment mounted.on pedestals, and the Jability to fully contain potential flood sources (where appropriate).
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| Adequate pen
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| * action from fire is provided by means of fire barriers and physical separation of the three independent division. Use of fire detectors, alarms, s pri.. . systems, fire water system and a trained crew of fire fighters keep _the risk related to fire at a' negligible level.
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| To assure the flood and fire related risks are kept low during shutdown, the shutdown procedures that the-COL applicant is required to develop have been identified.
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| = - - . . - -
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| Based on a FMEA of the new features incorporated into t he ABWR that are different from operating domestic BWR glants, Jt is concluded that none of the new features will introduce additional shutdown vulnerabilitias.
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| Instrumentation was identified that is available during Thutdown to adequately monitor the st3tus of the plant and operation of systems which will result in low levels of shutdowe risk.
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| Guidance was precented on how freeze seals esuld be used during maintenance on unisoable valves to minimize the riax associated with loss of the freeze seal.
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| Recommendations on outage planning procedures were presented to ensure that activities scheduled during outages take into account plant status and potentially high risk periods or configurations during shutdown. It was pointed out that the single most important elenie af reducing shutdown risk is proper outage scheduling of maintenance on afstamc and support systems capable of removing decay heat or supplying inventory makeup.
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| An analysis of 70 loss of power and over 100 loss of DHR precursor events
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| .at operating BWRs confirmed that the ABWR design features would prevent or mitigate the most safety significant of these events.
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| The PRA model for analyzing the loss of DHR accident initiation identified about 12 systems that can be used to prevent core damage. The resultant core damage frequency was negligible (<<l.0E-7 per-year) but the focus of the study was to identify minimum combinations of systems that if available would result in a conditioned core melt probability which is less than the goal of 1.0E-4 per-year given a loss of RHR event. It was found that generally about four of the 12 systems are sufficient to met the above goal.
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| In all cases, the minimum type and number of systems required by technical specifications plus systems normally oporating during shutdown (e.g., CRD and fire water) are sufficient to maintain adequate shutdown safety margins.
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| , , ~ -. --. - - - _ ~ -.. .- - .- - - - . - . _ - . . - - . ~ ... ~ - - . - - ,.-
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| i Many such combinations are possible, but certain specific combinations of minimum sets of systems have been identified to provide guidance to tht COL i-applicant. - Additional-minimum sets of systems can be identified by the COL applicant,~if needed, by using the PRA model. These combinations of systems identified will allow COL owners much flexibility in pr3 paring outage plans to ensure that shutdown safety margins are adequate at all times. .
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| 190 4 = FEATURES TO MINIMIZE SHUTDOWN RISK As part of the process for certifying the ABWR design, the NRC requested that General Electric provide a specific discussion of ABWR features that -i minimize. shutdown risk, The list of ABWR shutdown risk features is presented in Table 19Q.4-1. The features are grouped by risk categories as discussed in NUREG-1449, " NRC Staff Evaluation of Shutdown and Low Power Operation". Fire protection was.not discussed in NUREO-1449 but was added to the list based on discussier.a with the NRC. The risk categories ares t
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| - Decay Heat Removal I
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| - Reactor Inventory
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| - Containment Integrity l
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| Electrical Power.
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| - - Flooding Control Reactivity Control Fire Protection l
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| NUREG-1449 also discussed reactor coolant system pressuritation but this l
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| was not included in the' list because-it is mainly a PWR issue. . BWR shutdown l
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| l . pressure control concerns are ultimately inventory (i.e., LOCA) concerns and are
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| - addressed under Reactor-Inventory.
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| 1 s- e --~e-<s- - n- ---nr- ,, w--- , ,wxr pw e,- m e- ,-.. <, y ,-m ------,r, - - -
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| 7
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| -- _ ~ . - - _ . - - - - . . - - - _ . ~ . . - . . - - - - _ _ - _ _ . ~ ~ _ .
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| i i
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| The ABWR has been: designed with the minimization of risk being a.high
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| . priority. PRA methods have been very influentlal in the design of the ABWR.
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| The ABWR features described in Table 19Q.4-1 along with appropriate Technical Specifications and' utility operating and maintenance procedures (which contain i insights. gained-from risk based evaluations) all result in the conclusion that l during shutdown conditions the ABWR is adequately protected against accidents and the estimated core damage frequency is negligible.
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| The following sections describe the shutdown risk concern, past experience at operating BWRs for each risk concern, and the ABWR features that contribute ,
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| towards minimizing shutdown risk for each concern.
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| 19Q.4.l_ Decay Heat Removal Shutdown Risk Loss of Decay Heat Removal (DHR) while shutdown can lead to fuel uncovery ,
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| and damage.- It can be initiated by loss of the operating RHR system or by loss of an intermediate or ultimate heat sink, If loss of DHR occurs shortly after shutdown, bulk boiling of reactor coolant and fuel uncovery can happen quickly '
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| (i.e., less than one half hour for bulk boiling and approximately five hours to core uncovery if no protective action is taken).
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| -East Experience There has never been a loss of DHR in a BWR which resulted in actual core
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| -uncovery but several precursors to such an event have occurred in the past.
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| ~
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| Section 19Q.ll discusses many of these precursor events and describes ABWR ifeatures that.could have prevented or mitigated each event.
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| 1
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| _ _ _ _ _ - ._ . _ -.._,_ . . __. ....m
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| , Table 19Q.4 1 ABWR FEATURES THAT MINIMIZE SHUTDOWN RISK CATEGORY FEATURE SHUTDOWN RISK CAPABILITY Decay Heat Residual Heat Three independent (100% capacity)
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| Removal (DHR) Removal (RHR) divisions of RHR and support systems System for normal DHR. Each RER division has several DHR modes (e.g., SDC, SPC).
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| Reactor Coolant During shutdown, reactor coolant tem-Temperature perature is determined by measuring Heasurement reactor water cleanup (CUW) inlet water temperature.
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| Shutdown Cooling The shutdown cooling mode of RER uses Nozzle suction piping that connects directly to a nozzle on the RPV instead of to an external piping system. This reduces the probability of losing RHR pump suction due to air entrapment or cavitation.
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| Safety Relief Can be used as alternate means of Valves decay heat. removal by venting steam to the suppression pool. They are also actuated to depressurize the RPV to allow use of low pressure FOIR or other low pressure systems.
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| Suppression Pool A potential heat sink and make-up source for decay heat removal. Pool temperature is monitored in the control room to indicate trends in pool temperature. This large heat sink allows sufficient time for appropriate ope ator actions.
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| Reactor Water Can be used under certain conditions Cleanup _ System to remove decay heat. See Section (CUW) 19Q.7 and Attachment 19QA for more
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| -details on this feature.
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| RPV Boiling When the RPV head is removed, boiling is an effective (although not preferred) heat transfer method as long as RPV water level can be main-tained by available make-up sources.
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| - - - - - - - . - . - .- - - ,---e , --
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| \
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| l l
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| l l
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| CATEGORY FEATURE SHUTDOWN RISK CAPABILITY Condenser The main condenser (if available) +
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| can be used for DHR.
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| Remote Shutdown Cold Shutdown can be achieved-and
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| -Panel (Two maintained from outside the control Divisions) room if the control room is uninhab.
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| itable due to fire, toxic gas, or other reasons. The remote shutdown panel is powered by Class lE power to ensure availability following a Loss Of Preferred Power (LOPP).
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| Controls are hard wired and thus not dependent on multiplexing systems.
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| A minimum set of monitored param-eters and controls are included to ensure the ability to achieve and maintain cold shutdown.
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| Instrumentation Adequate instrumcr.tation is avail-able to operators both inside and outside of the control room for monitoring shutdown conditions throughout the plant. Some of the safety significant' parameters moni-tored during shutdown include: RPV water level.. reactor coolant. tem-perature, neutron flux, drywell pressure, RHR flow, reactor pres-sure, and suppression pool tempera-ture and level. In addition to monitoring, signals are also avail-able to actuate ECCS functions on low RPV vater level, scram control rods ~ on high- flux, and close isola-tion valves on appropriate signals.
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| Four divisiont of instrumentation allow one division to be in mainte-nance without disabling the func-tion, thus assuring availability of instrumentation during shutdown.
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| Fuel Pool' The fuel pool cooling system (FPC)
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| L can be used for DHR'during mode 5 (refueling). The pool does not contain drains and includes anti-siphon devices to prevent inadvert-ent drainage. The RHRS can be in-terconnected to the FPC to aid cool-ing of fuel in the pool if required.
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| CATEGORY FEATURE SHUTDOWN RISK CAPABILITY Reactor High Strength low Low pressure piping connected to Inventory Pressure Piping high pressure piping has been redesigned to a higher pressure rating and is therefore expected to withstand full reactor pressure on a rupture criteria basis. This minimizes potential for loss of inventory while shutdown.
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| Interlocked RHR The RPV shutdown cooling suction _
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| Valves valve must be fully closed before the suppression pool return or suction valves can be opened.
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| Shutdown cooling suction valve cannot be opened until suppression pool suction and return valves are fully closed. This prevents inad-vertent draining of the RPV to the suppression pool.
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| RPV Isolation All large diameter (>2 inches) iso-Valves lation valves in the RHR and CUW systems that connect to the RPV (except injection lines) automati-cally close on a low RPV water level signal. This reduces potential for the core being uncovered due to an inadvertent RPV drain down event.
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| Make-up Control If RPV level decreases, High Pressure Core Flooder (HPCF),
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| Automatic Depressurization System (ADS), and Low Pressure Flooder (LPFL) systems initiate automati-cally If HPCF and LPFL systems are in the test mode and a RPV low level signal is received, the systems au-tomatically switch to the vessel injection mode.
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| Feedwater and Three electric driven pumps that can Condensate Pumps be used during shutdown for make-up.
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| Pumps will not trip on low NPSH thus increasing their potential availability.
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| - - --- . . - . , _ _ - . - . - ~ -- -- -__ . . , - . . . - . - .
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| a CATECORY FEATl'RE Sl{llJDOW tlISK CAPABILITY High Pressure / Low Controls position of RHR valves to l Pressure ensure that the RHR is not exposed Interlocks to pressures in excess of its design pressure.
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| Make-up Sources Multiple sources of RPV make-up are potentially available while the
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| ,' plant is shutdown (e.g.. main con-denser hotwell, condensate storage tank, suppression pool, control. rod drive system AC independent water addition system).
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| No Recirculation Elimination of Recirculation piping l Piping external to RPV reduces probability of LOCA both during normal operations and while shutdown.
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| RPV Level Permanently installed RPV water Indication level indication for all modes of shutdown. Sensors arranged in a 2-out of 4 logic to ensure high reliability.
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| Containment Containment Reinforced concrete structure sur-Integrity. rounds RPV to withsta I LOCA loads L
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| and contain radioactive product 1 from potential accidents during hot-shutdown. Secondary containment permits isolation and monitoring _all potential radioactive leakage from the primary containment.
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| _ Standby Cas Removes and treats contaminated-air
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| -Treatment System from the secondary containment t following potential accidents.
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| l Reactor Building Automatically closes isolation damp.
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| Isolation Control ers on detection of high radiation.
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| These dampers are potential leakage -
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| paths for radioactive materials to the environs following breach of nuclear system barriers or a fuel handling accident 11.
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| _ .. . -- . . .. - .. .-..- ~.. --._ -. .
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| CATEGORY _ ____ FEATURE SHUTDOWN RISK CAPABILITY Electrical Power 3 Diesel One diesel for eac., safety division, j Generators Independent, both electrically and physically, of each other to mini-mice common mode failure. Allows
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| .for diesel maintenance while still maintaining redundancy, Combustion Tur- Redundant and diverse means of sup-bine Generator plying power to safety and non-safety buses in event of loss of 4
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| offsito power and diesel generator failures.
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| 2 Sources of Reduces risk of LOPP due to equip-Off-Site Power ment failure or operator error.
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| Electrical Cable Vill prevent propagation of fire Penetratioas damage and water from postulated flooding sources, 4 Divisions of DC Elt-trically and physical'ly Power independent. Includes batteries and chargers, . Diverse ueans of electrical power for' control circuits and emergency lighting.
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| Flooding Control , Flood Monitoring Raactor, control, and turbine and Oontrol building floodin6 is monitored and alarmed in the control room. This alerts the operator to potential flooding during shutdown. Many .
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| flood sources (e.g., HVAC, EDG Fuel) are relatively small volume and are.
| |
| :self limiting. Operation of the-fire water system is alarmed in the ,
| |
| control room to help.the operator-differentiate between a break in the firewater system and the need to
| |
| -extinguish a' fire, Larger sources are mitigated by means of raised-sills on room doors, equipment mounted on pedescals. floor drains, watertight doors, pump trips, valves closing, anti-siphon valves, or operator actions.
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| I,
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| , . _ , . . , . . , . _ _ . . , , , , _ _ _ , - , _ , . _ , , . . . , , , , _ , . _ , . , _ . _ , , . -, m, , _
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| . . . ~ . - .. . - . - - . -
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| . . . . . ~ ,
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| )
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| i I
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| l l
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| C&TECORY FIATURE SjhfTDOWN RISK CAPABILITY j
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| ]
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| j= Room Separation The three divisions of ECCS are I Sysically separated and self con- ,
| |
| ained within flooding resistant -j
| |
| 'lls, floors, And doors. No ECCS +
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| all penetrations _are located belov l 5e highest potential flood level in
| |
| .he reae.or building first floor ,
| |
| corridor. No er.ternal potential !
| |
| flooding sources are routed through the ECCS rooms and potential flood-ing sources in other rooms vill not ;
| |
| overflow into the ECCS rooms and cause damage to ECCS electrical i equipment. If ECCS flood. barriers must be bread.- 1 during shutdovn, administrative nntrols ensure that at 1 cast one ECLS division is oper- ,
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| able and all barriers in that divi-
| |
| * sion are maintained intact, t
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| Reactivity Refueling A system of interlocks that Control _ Interlocks restricts movement of refueling ,
| |
| equipment and control rods during e refueling to prevent inadvertent criticality. When the modo switch is in the REFUEL position, only one i control rod or rod pair can bs with-drawn at a time.
| |
| Fuel Handling Fuel handling and storege facilities are designed to pre < ant inadvertent criticality sad to maintain adequate shielding and cooling for spent fuel.
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| CRD Supports and CRD supports limit the travri of a Brake control rod in the event a control rod housing is ruptured. The brake liraits the velocity at whi- a con-trol rod can fall out of c c core should a hydraulic line break or failure of flange bolts or a spool piece. _Roth of these limit reactiv-ity excursions and thus protect the fuel barrier.
| |
| F 15-4 s
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| ['i'-
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| w, , , - - - - - . - . . - , , . . . . . - -
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| .m-- - - - - . _ . ._ _ , . - , , .- - - - , ,
| |
| | |
| . . -. . - . . . . . . - . . . - - - . _ _ ~ -. - ~ . .- .. .
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| .l CATEGORY FEATURE SHUTDOWN RISK CAPABILITY ,
| |
| 'i instrumentation Reactor protection system (RUS) high flux (set down) and manual scram functions are operable during shut-down.
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| - Fire Protection Divisional The three ECCS divisions are physi, ,
| |
| Separation cally separated so that a fire 1 initiated in one division will not ye propagate to another division. Pro-cedures ensure that during shutdown if fire barriers between divisions must be breached due to maintenance, at least one diviu.aa vill be operable with barriers intact.
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| Detection Fire detection sensors that alarm in the control room are located thrrughout the plant and operate during shutdown. Actuation of the
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| -fire water system is alarmed in tha control room. Also, during shutdown more personnel are located throughout che plant to identify, extinguish, and report potential-fires, c Suppression Wster and chemical fire suppression
| |
| , systems are located at appropriate plant locations.
| |
| Water Supplies Multiple water supplies and both electric and diesel powered fire pumps can deliver water to any location in-the plant during' shut-down.
| |
| Multiplerod' Eliminates the need for a cable
| |
| - systems- spreading room which is a major fire ,
| |
| concern in most plants.
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| HVAC- Dual purpose HVAC/ SMOKE Control p -system, divistorally-separatad, to
| |
| : j. -concrol individual room pressure and l- essure clean air path'for fire L '
| |
| suppressi.n personnel. l L
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| l~
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| (
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| l.
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| i-I..'
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| , ... _ - - -- n- .. ,- - . - , , . . . ~ _ ....-m ...-__-..-._...._....-_...u_, . . _ _ . . . . . - . . . _ . . . . . . --
| |
| | |
| -Por BWRs, the most common precursor events-involved temporary loss of RHR due to various reasons-including inability to open Shutdown Cooling (SDC) valves inside containment and isolation of SDC due to low water level in the RPV or loss of power to the Reacter Protection System (RPS). In all of those cases, redundant loops of RHR or alternate DHR methods were available.
| |
| ABWR Features The ABWR contains many features to minimize the loss of DHR. The ABWR contains three divisions of RHR and aesociated support systems that are electrically and physically separated. This is the first line of defenns lu f maintaining DHR. Ono RHR loop could be in maintenance and if a single failure were to-occur to the operating loop, the third loop could be placed in service.
| |
| It-is also possible,-if conditions warrant, to run RHR loops in parallel. In this case, failure of one loop would not result in even a temporary lose of DHR.
| |
| In the unlikely event.that all RHR loops werb unavailabic, several alternate methods of DHR could be used. Steam from the RPV could be directed to
| |
| - the main condenser (if available). Hake-up to the RPV could be supplied by many sources as discussed in Section 19Q.4.2. Other heat sinks include the suppres-sien pool-_(via the safety relief valves), the reactor water cleanup system, or
| |
| .the spent fuel pool (if the reactor water level is raised to the refueling 1
| |
| ' level). As'a final method, 11 the RPV head was removed, bulk boiling of reactor
| |
| - coolant in-che RPV with adequate make-up would prevent _ fuel damage.
| |
| A= mentioned above, SDC flow has been temporarily interrupted at operating plants.in the past due to a-loss of RPS logic power. Loss of RPS power does not result.in isolation of the SDC system in the ABWR design. A loss of power to the multi-plexed ABWR safecy system logic would. result in SDC-isolation valves
| |
| , falling "as-is". It takes an actual need for completion of a' safety function, not'sidply a loss of power, in order for the ABWR safety system logic to cause
| |
| - Lctuation of safety systems (e.g., inolation of SDC).
| |
| _ =.
| |
| | |
| l From the above it can be seen that there are multiple methods to maintain DHR in the ABWR such that the shutdown risk associated with loss of DHR is negligible.
| |
| 19Q.4.2 Inventory Control Shutdown Riek Loss of inventory control can lead to uncovering the fuel and damage by overheating. Reduction of reactor coolant inventory is more likely when the plant is shutdown beca*1sa additional paths for diversion of coolant (e.g., RHR system) are operable. In addition, there are shutdown activities such as test and maintenance that require seldom used valve line-ups and plant configurations which increase the probability of operator errors associated with inventory control.
| |
| Past Experience As discussed in Section 19Q.11, events at operating plants have resulted in reduction of reactor coolant inventory. For BWRs this typically involved diversion of reactor coolant from the RPV to the suppression pool due to improper valve line-ups (e.g., opening suppression pool suction valve before SDC suction was fully closed) or valve leakage (e.g., RHR pump mini-recirc valve).
| |
| Other inventory losses were due to leaking RHR heat exchanger tubes, placing a partially drained RHR loop on-line following maintenance, and buckling of an RHR heat exchanger due to marine growth. In all cases, the loss of inventory was either recovered due to operator action or automatically stopped by isolation of SDC on low RPV level.
| |
| 1_ _-_ -___ _ -__- _ __ _ _ ___ __ ___ -
| |
| | |
| .-.m..___._ _m . _ ._ _ _ _ _ _ _ _ __ _ _ . . _ . _ , _ . . . _ . _ _ _ . , _ _
| |
| l l
| |
| hBWR Features The_ABWR contains several design features to minimize the potential for inventory loss. Indication of RPV level is displayed to the operator in the 7 i
| |
| control room during all shutdown configurations including refueling. To ensure adequate level is maintained in the RPV, multiple sources of make-up exist including Suppression pool, condensate storage tank, nmin condenser hotwell, T
| |
| and AC independent water addition system. ,
| |
| 4 To minimize the potential for pipe breaku, RHR system valves are inter-locked with reactor system pressure to ensure that low pressure RHR piping is not exposed to full system pressure. In the event that the interlocks fail or are bypassed, the RHR piping is capable of withstanding full reactor pressure without rupture. ;
| |
| During shutdown there are many maintenance taska and evolutions that could lead to potential draining.of the RPV. These include: CRD and Reactor Internal Pump (RIP) removal and replacement, and failures or operator errors associated with operation of the reactor water clean-up system and the RHR system. These '
| |
| i potential drainage paths are discussed below:
| |
| \
| |
| CRD Re51acement CRD repl acement for the ABWR will use the same procedure followed.for current operating BWRs. The CRD is withdrawn to the point where the CRD blade (back seats onto the CRD guide tube. This provides a metal to metal seal that prevents RPV drainage when the CRD is removed. The many years of BWR experience with CRD removal gives a high degree of assurance that the risk from this $
| |
| . operation will be negligible for the ABWR.
| |
| i RIP Motor and Impeller Replacement Nuclear plants with RIPS have been in operation for over 10 years. Over i
| |
| 100 RIPS and motors have Deen successfully removed and reinstalled in European 8
| |
| : i. _-. - _ _-- _ __ ._. _
| |
| | |
| I BWR plants. This has demonstrated that replacement activities can be carried out without dralning the vessel.
| |
| Replacement of RIP motor and impeller involves the following steps. The RIP lower bolta are loosened and the pump allowed to move downward approximately 1/4 inch to the point where the impeller becomes backsoated. An integral inflatable seal is then actuated as a backup sealing device to assure no RPV leakage. The RIP motor can then be removed. Following motor removal, a temporary cover plate is bolted to the bottom. The impeller is then removed from the top. The inflatable seal and bolted cover plate prevent leakage of coolant from the RPV. After the impeller is removed, a cap is installed on the RPV bottom head at the impeller shaft nozzle to provide additional protection against dralning the RPV.
| |
| For draining to occur, as a minimum, the Lmpeller backseat and the inflatable eeal have to fail when the motor is being replaced. Administrative procedures assure that impeller removal does not start until the RIP motor is removed and the temporary cover plate is bolted. In the most likely failure scenario, it is possible that the sealing between the impeller and its backseat and the asaling provided by the inflatable seal may not be perfect. However, auch failures are detectable, and result only in a small leakage (loss than one o
| |
| gallon per minute). Under these conditions, the operator can always bolt the temporary bottem plate if needed. For drainage to occur during impeller replacement, the impeller shaft nozzle cap must fail (or be dislodged), the inflatable seal must fall, and finally the bottom plate must also fail. Because of the multiple failures required, the probability of draining the vessel during RIP maintenance is negligibly low.
| |
| Contfc1 Rod Drive Hydraulic Svetem During operating modes 4 & 5, the control rod drive hydraulic system (CRDHS) continues operating with one pump running to provide purge water to the FMCRDs. With one pump in operation, the head of the pumping water can easily overcome the head of water in the RPV; hence, draining the RPV la I
| |
| . -____________________.___________-___.m_.__ . _ _ _ _ . - _ _ _ - . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _._ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ _
| |
| | |
| unlikely. In the event that neither pump is in operation, there are several potential paths for draining the RPV through the CRDdS.
| |
| With neither CRD pump operating, the scram valves will open due to low Hydraulic Control Unit (HCU) charging header pressure. The scram valves niay remain open due to operator error in net resetting the RPS logic or other system failures such as loss of instrument air to the scram valve. This combined with multiple mechanical f ailures to check valves and operator errors in CRD hydraulic system valve lineups could result in RPV drainage through the CRD --
| |
| hydraulic system. Multiple failurf .e required for RPV leakage to occur and even if a leak were to develop, 6niy two CRDs would be affected and the leak would be small since it would occur in a 1-1/4 inch line. Therefore, the probability of draining the RPV through the CRD hydraulic system is considered negligible.
| |
| 5 Reactor Water Clean-Up Svetem LCUW)
| |
| During shutdown, the CUW provides continuoup cleaning of the reactor coolant. Water is removed through a line attached to the RPV bottom head and after passing through a series or hoat exchangers and a tilter demineralizer is returned to the RPV either via an attachment to the upper head or through the -
| |
| feedwater lines and spargern.
| |
| Potential drainage paths exist due to several maintenance flush and drain valves and CUW discharge paths to the Low Conductivity Water (LCW) sump and the suppression pool. The latter two paths are used during reactor startup to control excees reactor water due to heat up and thermal expansion.
| |
| d for any of the potential flow paths described above to result in RPV drainage, multiple failures of equipment and operator errore must occur. In addition, if the RPV were to start draining all but one of the potential flow paths (LCW aump) would be automatically isolated on low RPV level. The flow path to the LCW sump is controlled by two valves in series one of which is locked closed and both are under administrative control. If drainage were to
| |
| | |
| . _ . _ _ _ .. _ _ .._~._ -._ _. _ m. _ _ . . _ . . _ - - _ _ _ _ . . . _ _ .
| |
| occur, LCW sump well level alarms would annunciate in the control room. Also, ;
| |
| I the line is'only.2" in diameter and so the flow rate would be slow enough to allow ample operator time to mitigate the leak.
| |
| Because of the multiple f ailures and operator errors that must occur to cause RPV drainage through the CUW and the automatic RPV isolation logic to stop most potential flow paths, the risk of RPV drainage though the CUW is considered negligible.
| |
| Residual Heat Removal System l
| |
| 4 The ABWR Residual Heat Removal (RHR)-System is a closed system consisting ,
| |
| of three independent pump loops (A, B, and C - where B and C are similar) which l inject water into the vessel and/or remove heat from the reactor core or
| |
| -containment. Loop A differs from B-and C in that its return line goes to the Reactor Pressure Veesel (RPV) through the feedwater line whereas loop B & C return lines go directly to the RPV. In addition, loop A does not have connections to the drywell or wetwell sprays or a return to the fuel pool cooling--system. .H74ever, for purposee nf thie anteisis,-the differences are minor and the three luep,cr: be cor s i '- 'ed identical. The RHR system has many
| |
| , modes of operation, each mode making use of common RHR system components. These components are actuated by the operator. Protective interlocks are provided to prevent the-most likely interactions of mode combinations.
| |
| The operator has-five mode relection switches available that will automatically perform the required valve alignment for the mode selected. This feature reduces the chance of operator error by'only requiring one action, the
| |
| '- ' mode selection switch, to realign-several valves. Only one mode at a-time can I '
| |
| be operational,.thus precluding potential-undesirable multiple mode
| |
| -interactions.- The five' modes are: 1) low pressure-flooding; 2) suppressio.t pool cooling;.3) shutdown cooling; 4) wetwell spray; and 5) drywell-spray.
| |
| =. - .- ... - _. . , .. .. - . _ . - . _ . . . . . - . . . - , _ _ ,
| |
| | |
| There are two basic ways that RPV water level can potentially be reduced through the RHR system in the ABWR during shutdown cooling. The first way is through operator error in opening manual isolation valves that are used for RER c system maintenance. These paths are to the High Conductivity Water sump and the Liquid Waste Flush system. These valve = are normally closed during the shutdown cooling mode of plant operation. These are 2 inch and 6 inch lines respectively and inadvertent opening of these valves would result in a relatively slow RPV level decrease which would be alarmed to the operator in the control room such that there would be adequate time to respond. If the operator failed to notice -
| |
| the decreased RPV level, an alarm would annunciate in the control room and the RPV isolation valves would automatically close on low RPV level. The fuel would remain covered with water and no fuel damage would occur.
| |
| The second way that RPV level could decrease would be for one of the Motor Operated Valves (MOVs) in the RHR system to open inadvertently or by operator error. Most of the MOVs in the Rl!R system are interlocked to prevent inadvertent diversion of RPV water (e.g., the Shutdown Cooling (SDC) suction
| |
| ~
| |
| line is interlocked so that the suppression pool suction and return valves and wetwell spray valve must be clooed before the SDC valve can be opened, the shutdown cooling suction valve must be fully closed before the suppression pool suction or return valve can be opened, the two series dry well spray valves
| |
| =
| |
| cannot be opened at the same time unless the drywell pressure is high). Thus loss of RPV level through these paths is not likely. Loss of RPV level through the wetwell spray valve requires a mechanical failure or an operator error to open the valve when not required. The only other potential path is via the RHR pump mini-flow valve. This valve is designed to open to al'.ow water flow back to the suppression pool if the RHR pump ic running at shutoff head. This is a pump protection fea'ure. The valve opens and closes automatically depending on measured RHR flow.
| |
| Whether the potential flow path is caused by mechanical failure or operator error, two features exist .o mitigate the loss of RPV level. On a low RPV level signal, both RPV isolation valves close to stop all flow out of
| |
| | |
| _ _ . . _, , _ _ _ . _ , _ . . _ . ~ _ _ . - _ _ . , _ _ . . _ _ _ _ . _ _ _ _ _ _ _ . _ _ - . , , ,_ . _ _ . _
| |
| C t'.se RPV. The.RPV low level setpoint is 2213 meters above the top of the fuel.
| |
| Even if the low RPV level isolation feature were to fail (after a previous valve j mechanical failure or operator error), flow out of the RPV would automatically stop when the RHR shutdown cooling nozzle is uncovered. At this peint, 1.7 meteru of water would still be above the top of the active tuel. Chereforc, the
| |
| - draining of the RPV via.the RHR system to the point of uncovering the fuel and causing fuel damage is not considered credible for the ABWR, Another potential for loss of inventory control is through-the use of-freeze seals on piping attached to the RPV. Section 19Q.8 discusses how freeze seals will be used on the ABWR and why the risks associated with freeze seals will be small. ,
| |
| i In summary, the ABWR contains many redundant and diverse features such that, along with the use of experience proven administrative controle,-loss of
| |
| . inventory control is not a significant safety concern.
| |
| 19Q.4.3 containment Inteority Shutdown Risk 4
| |
| .A breach of containment integrity is not by itself an inaue of high safety significance but, in conjunction with other initiating events, could increase the severity of the initiating event. A breach of containment integrity followed by breach of another radiological barrier or boiling of the reactor coolant could lead to a direct release to the atmosphere. Attachment 190B discusses potential of fsite releases following boiling in the RPV with the head ramoved and shows that releases would be a small fraction of normal operating In addition,-the'?RA results in section 19Q.7 indicate that the-risk of I limits.
| |
| RPV boiling is low.
| |
| During refueling of the BWR, the primary containment is open and cannot be readily closed since the drywell-head is removed. None-the-less, loss of containment integrity has-not been an issue for BWRs in the past.
| |
| i
| |
| | |
| ABWR Features During shutdown with the drywell head removed, the ABWR has the aecondary containment _which can be automatically isolated on high radiation from a radio-logical boundary breach or fuel handU na accident.
| |
| The standby gas treatment sy7 tem (SGTS) filters air from the sevondary containment to reduce potential contamination to the atmosphere.
| |
| The ABWR secondary containment and use of the SGTS results in a negligible
| |
| . risk concern for loss of containment integrity.
| |
| 19Q.4.4 Electrical Power Shutdown Risk
| |
| -A-loss of all off-site power challenges the on-site sources to power safety related equipment to maintain safe shu.down. Loss of individual buses (AC or DC) affects divisional train capability and results in loss of redundancy a to complete required safety functions.
| |
| Past Experience As discussed in Section 19Q.11,. loss of power events have occurred at many -
| |
| nuclear power plants. Thers have been several casea of a total loss of of f-site power which, in some instances, le: to loss of shutdown cooling and increases in coolant-temperatures of as much 4: 4 '14F.
| |
| The majority of total loss of off-site power events were due either to severe weather or operator errors. Several losses of on-site power events were due to objects f alling on transformers while performing maintenance activities in the switchyard. In other cases, - switching errors resulted in temporary loss of power to vital busses or off-site power.
| |
| 8 I
| |
| i ABWR Teatureo i
| |
| The ABWR electrical system has the 101199f .9 features to prevent or miti- ;
| |
| c' cate potential loss of power events:
| |
| - Three physically and electrically independent class IE emergency diesel generators
| |
| - Two independent sources of off-site power
| |
| - Three unit auxiliary transformers powering three Class 1E and- nos - 1E power buses Combustion Turbine Generator (CTG)-that can be used to power any of the class 1E or non-1E buses The ABWR electrical power eystem contains redundancy and diversity of electric power sources. This allows sources to be in mainter.anca during shut-down and still have adequate' power sources &o meet potential equipment failures. ,
| |
| Even in the case of loss of off-site power, the CTG has the ability to start a feedwater or .other pump for DHR or ir ie.cory makaup if required. This means j.
| |
| that the ABWP can use altet 96te sources of DHR with only on-site power sources, i
| |
| In the event.that one phaar of the main transformer were to fail, an installed spare is available to -) turn the preferred source of off-site rawer to service without.the need to procurn nnd deliver a new transformer.
| |
| As ,iscussed more fully in Section 19Q.11, the ABWR electrical power (P u ribution system has features that are_ capable of mitigating potential loss of power eveats thatchave occurred at oporating plants in the pa6t. The design
| |
| ~featurec describud above in conjunction with appropriate Technical
| |
| , Specifications and other administrative controle result in an electrical 1
| |
| distribution system that is able to maintain an adequate level of redundancy 1
| |
| | |
| and capacity even with equipment out for maintenance or testing. This ensures that saf6ty margins can be maintained at all times during shutdown and normal plant operation.
| |
| 19Q.4.5 Eeactivity control Shutdown Risk Reactivity control during_ shutdown may be a concern because local -
| |
| criticality can be achieved through movement of control rods or errors in fuel handling that may not be adequately detected by installed neutron detectors.
| |
| T.lso at-lower temperatures, the inherent negative feedback available at normal
| |
| - operating temperature and pressure is less able to mitigate potential power excursions.-
| |
| While overall core shutdown margins are adequate to protect the fuel as long as procedures are followed, inadvertent withdrawal of two adjacent CRDs or fuel handling errors can lead to fuel damage.
| |
| Past Exnerlence A few isolated cases of BWR shutdown reactivity control concerne have been identified in the past and were attributed to operator errors (e.g . withdrawing
| |
| -. the wrong control rod).
| |
| Reactivity excursion evente could occur'due to any one of the followingt
| |
| - Control Rod Drop
| |
| - Control Rod Ejection
| |
| - Refueling Error Rod Withdrawal Error
| |
| - -Fuel Loading Error
| |
| | |
| Control Rod Drop While shutdown, the vnly time a control rod drop could occur is during control rod testing. If one control rod is fully withdrawn, a rod block signal prevents withdrawal of a second control rod. If the rod block signal were to fail and the operator were to incorrectly select an adjacent control rod for withdrawal, a latch mechanism existe such that if the rod were to become stuck and decouple from its drive it could only drop a maximum of eight inches. In addition, a Class 1E eeparation detection system trould sense a separated control -
| |
| rod drive and initiate a rod block signal.
| |
| Due to the combination of evente required to cause a control rod drop including operator error coincide.a with multiple mechanical failures, the ABWR rod drop accident rish is considered negligible.
| |
| Control Rod Eiection d
| |
| For a control rcd ejection accident to occur while shutdown, RPV pressure would have to be increased (e.g., during a hydrostatic teet). The series of events that would have to occur are (1) During RPV hydrostatic testing a control rod is withdrawn for testing and, (2a) A break in the CRD Houoing of an adjacent rod occurs which also resulto in failure of the internal control rod anti-ejection supports (" shootout restrainte")
| |
| or (2b) A break in the CRD insert pipes coupled with failure of both its ball check valve and electro-mechanical brake.
| |
| | |
| .. - _- . ._ .- - _ - - .-_- , . - - - - _ ~ - . - ~ - - - . - . . - _ . -
| |
| i Due to the short amount of time that the RPV undergoes hydrostatic testing and the multiple failures required for a control rod ejection to occur, the risk i from this event is considered negligible.
| |
| Refuelino Error During refueling, inserting _a fuel bundle at the maximum fuel grapple speed into a fueled region of the core which has a withdrawn control rod blade could result in a reactivity accident.
| |
| The ABW2 features that prevent or mitigate refueling errors ares
| |
| ~
| |
| (1) An intarlock with the mode switch in the REFUEL position which prevents
| |
| -hoisting another fuel assembly over the vessel if a control blade has been removed.
| |
| (2) While in the REFLEL position, ,nly one rod can be withdrawn at a time.
| |
| Any attempt to withdraw a second control rod would result in a rod block l-signal being initiated by the refueling interlock.
| |
| (3) The_ operator would be alerted to a refueling error by the source range neutron monitoring system.
| |
| Due to the combination of operator errors, interlock failuren, and core configuration required for this event to occur, refueling accident risks are considered negligible.
| |
| l Bod Withdrawal Error
| |
| .If two adjacent control rods are withdrawn at the same time, the reactor may become critical. To prevent this the ABRR has a refueling interlock which prevents any more than one control rod being withdrawn at a time.- If the interlock fails and the rod is withdrawn, the rods would scram on a high flux signal..
| |
| | |
| 1 a
| |
| 1 1
| |
| The coincident failures of the refueling interlock and reactor protection ,
| |
| system in cGnjunction with operator error, which are required to cause a rod ,
| |
| +
| |
| withdrawal ector are considered improbable and the risk negligible.
| |
| Fuel Leadina Error ,
| |
| This event is similar to a refueling error, in this case the refueling procedure is not followed and a higher than design core reactivity configuration ,
| |
| is formed. If not identified by the core verification process, subsequent control rod testing may result in inadvertent criticality and power excursion.
| |
| A high flux scram would terminate the excursion.
| |
| The risk from a fuel loading error is considered negligible because of the combination of events req 11 red for the accident to occur.
| |
| Summary of Reactivity Concrol The ABWR refueling interlocks, control rod design, reactor protection system operability during shutdown, and strict administrative controle all combine to support the conclusion that shutdown Reactivity Control is a ,
| |
| negligible risk concern for the ABWR design.
| |
| 19Q.4.6 Summary of Shutdown Risk CateoJrv Analysis The ABWR design was evaluated against shutdown risk categories from ;
| |
| NUREG-14 4 9. -- The analysis took into account past experience at operating Bwas.
| |
| The conclusion _from this inalysis is that the ABWR design contains multiple !
| |
| features to minimize potential risk,during shutdown for the major shutdown risk
| |
| _ categories. ,
| |
| _. - . _ , ~ . . _ . , _ _ . _ - _ _ , , . . _ . , _ . _ . _ _ . . _ _ _ . _ , . . . _ _ . _ . _ _ _ _ _ . , . . . , . _ _-_
| |
| | |
| .; ., ..u...s-. ~ . . _ - = . - ~ . _ . - - - . . . - - - - - =. - - -= - .
| |
| I I
| |
| ~
| |
| l 1
| |
| s o
| |
| 19Q.5 IESTRUMENTATION The ABWR instrumentation system contains many features that help reduce shutdown risk. These features are contained in the basic design of the instru-ment systems and in the type and number of parameters monitored.
| |
| During shutdow.1, the main concern from a risk perspective is removal of decay heat from the fuel in the RPV. The large volume of water in the spent fuel pool and low probability of draining makes the risk associated with fuel pool
| |
| ' operation relatively low. The smaller reactor pressure vessel (RPV) volume and relatively high decay heat load of the fuel increases the cooling requirements and decreases the available time to recover from loss of decay heat removal (DHR). Thus, to minimize shutdown risk, the instrumentation system must monitor RPV level and water temperature, status of makeup sources and heat sinks, and >
| |
| -display these to the plant operators in a reliable and easy to understand manner.
| |
| j-Desian Features The ABWR utilizes tedundant chanaels of safety related instruments for initiating safety actiona and monitoring plant status. This is accomplished by a four division correlated and separated protection logic complex called the safety system logic and control (SSLC). The SSLC receives signale from the redundant channels of instrumentation, displays information to the operator, and makes decluions on safety actions.
| |
| The safety system setpoints are determined by analysis and experience factoring in instrument errors, drift, repeatability, safety margins, and the need to minimize spurious actuations. The system provides continuous automatic 4 on-line testing lof the logic and offline semi-automatic end-to-end (sensor input
| |
| .to trip actuator) testing. This combination meets all current regulatory requirements.
| |
| - . - - ., - .- , - - , , - . - ,. - , _ ._ _ _ - . - ,,. _ -. -+,
| |
| | |
| l Speelfic instrumentation features important to shutdown operations include:
| |
| - Automatic initiation of ECCS to ensure adequate RPV make-up.
| |
| - Four channels of instrumentation to allow for bypass during maintenance and testing while still retaining redundancy. (The two-out-of-four logic reverts to two-out-of-three during maintenance bypass).
| |
| - continuous monitoring for detection of fires or 71ooding in sa'ety related and other areas.
| |
| - Operability of the reactor protection system (RPS) during shutdown to mitigate potential reactivity excursions.
| |
| - Interlocked refueling bridge operation to prevent reactivity excursion.
| |
| - Automatic isolation of shutdown cooling (SDC) on low level in the reactor pressure vesoul (RPV) to ensc-e against fuel uncovery. .
| |
| - Interlocked residual heat removal (RHR) valves (SDC and suppression pcol) to reduce the potential for diversion of coolant from the RPV to the suppression pool.
| |
| - Ability to c*ntrol shutdown plant status from the remote shutdown panel in the event that the control room becomes uninhabitable.
| |
| Ability to monitor radiation levels throughout the plant to datect breaches in radiological barriers.
| |
| | |
| l l
| |
| Parameters Monitored The key shutdown parameters monitored by the ABWR instrumentation system include:
| |
| - RPV level, water temperature, and pressure
| |
| - Neutron flur l
| |
| 1
| |
| - Drywell and wetwell precoure and temperature
| |
| - suppression pool temperature and level i
| |
| - Reactor, turbine and control building flooding level
| |
| - RHR flow rate and temperature
| |
| ~ Fire detection in various buildings 4
| |
| - Electric power distribution system parameters (e.g., power, voltage, g current, frequency) !
| |
| - Operation of fire water system i
| |
| 190 6 FLOODING AND FIRE PROTECTION.
| |
| The ABWR has-been designed to minimize the risks associated with fires and flooding though the basic layout of the plant and the choice of systems to en-hance the plants tolerance-to fires and flooding.
| |
| - Plant Layout-The plant layout is such that' points of poasible common cause failure between safety related and non-safety related systems have been minimized. As 1 - . _ . - . . _ _ . , ,;._..._ _ . , . . . . . , _ . _ . - , _ _ . _ . . _ . - . . _ , _ _ . . . _ _ ._ . , _ _ , . _ . . . . _ . . _ . - , _ _ . . . . , _ _ ,
| |
| | |
| t i
| |
| t an example, the control room is situated between the reactor building and the j turbine building. Thus safety related equipment and controls that are used to shutdown and maintain long term cold shutdown of the plant cannot be impacted by failures of non-safety related systems in the turbine building. Likewise, I
| |
| non-safety related systems / equipment in the turbine building that could be used to reach and maintain cold shutdown (e.g., condensate, main condenser) are not ;
| |
| affected by failures of safety related equipment, therefore, interactions between reactor and turbine building systems are minimited.
| |
| I Normal and alternate preferred power is supplied through the turbine building to the reactor building for safety related loads. These non-safety re-lated power sources are backed-up by safety related diesel generators located in .
| |
| the reactor building. The diesel generators are thus_not affected by events in the turbine building.
| |
| The. buildings are laid out internally so that fire areas of the same divi- f r
| |
| sion are grouped together in block form as much as possible. This grouping is coordinated from building to building so that the divisional fire areas lineup adjacent to each other at the interface between the reactor and control build-ing. An arrangement of this fashion naturally groups piping, HVAC ducts, and ,
| |
| cable trays.together in divisional arrangemento and does not require routing of l services _of one division across space allotted to another division.
| |
| A major difference between the ABWR and current reactor designs is that due to'the multiplexing of plant systema, there is no need for a cable spreading room.- This removes a significant source of potential fires that could lead to core damage both during normal plant operation and shutdown conditions, systems
| |
| .The ABWR has_ three independent saf ety related divisione c any one of which is capable of maintaining the reactor in a safe cold shutdown condition. With l.
| |
| I_,__.___,__._.._.__._.___,._..__
| |
| | |
| this arrangement, a single division may le out for maintenance and a single random failure could occur which dienbled another division, but t he third division could be available to ensure continued DHR. In addition, there are non-safety relatud systems such as condensate that can be used to maintain cold shutdown.
| |
| In general, systems are grouped together by eaf ety division so that; with the exceptions of the primary containment, the control room, and the remote shutdown room (when operating f rom the remote shutdown panels); there is only -
| |
| one division of safe shutdown equipment in a fire area. Complete burnout of any fire area without recovery will not prevent continued DHR, therefore, complete ,
| |
| burnout of a fire area is acceptable from a public risk perspective.
| |
| The separation exception in the primary containment is made because it in not practical to divide the primary containment into three fire areas. The depign is deemed acceptable because:
| |
| : 1) Sprinkler coverage is provided by the containment spray system.
| |
| : 2) Only check valves and ADS /SRV valves (if the RPV head is on) are required to operate within containment to provide DHR. A fire could -
| |
| not prevent the operation of a check valve nor would it prevent a safety valve from being lifted on its spring by pressure. The high pressure pumps are capable of providing sufficient head to lift the SRV valves against their spring settings so that a fire could not prevent injection of water to and relief of steam from the reactor vessel.
| |
| : 3) In addition, maximum separation is maintained between the divisional equipment within primary containment.
| |
| All divisions are present in the control room and this cannot be avoided.
| |
| The remote shutdown panel provides redundant control of the DHR and ECCS functions from oateide of the control room. The controle on the remote shutdown panel ara hard wired to the field devicen and power supplies. The
| |
| | |
| . _ _ _ _ . _ _ _ _ - . _ _ _ _ _ . _ , _ _ . _ _ _ . . _ . _ _ _ _ _ _ _ . _ _ _ _ .___~
| |
| signals between the remote shutdown panel and the control room are multiplexed over fiber optic cables so that there are no power supply interactions between i the control room and the remote shutdown panel.
| |
| There are some areas where there is equipment from more than one safety l division in a fire area. Each of these cases is examined on an individual basis f to determine that the encroachment is required and that failure in the worst !
| |
| conceivable fashion is acceptable. These are documented in the SSAR Section [
| |
| 9A.S.5 under Special cases - Fire separation for Divisional Electrical Systems. l Divisions 1 and 2 125 VDC and 120 VAC power supplies, reactor building cooling water pumps and heat exchangersi emerg*ncy chillers and emergency HVAC systems are located in the control building. Cince these systems are required for DHR if the function of the control room is lost, they are separated from the control rocm complex-and its HVAc system by rated fire barriers. A fire result-ing in the loss of function of the control room will not affect the operation of the remote shutdown or remote shutdown support systems.
| |
| When the plant is uhutdown and if due to normal maintenance or other work fire barriers must be breached between two safety divisions, the third division ,
| |
| must be operable and its barriers checked to ensure they are intact.
| |
| Fire Containment l The fire containment system is a combination of structures and barriers that work together to confine the direct effects of a fire to the fire area in -
| |
| which the fire originates. The fire containment system is comprised of the fol-lowing elements:
| |
| : 1) Concrete fire barrier floors, ceilings and walls which must be at
| |
| -least-six inches thick if-made from carbonato and silicious i aggregates. Other aggregates and thicknesses are acceptable if the type of construction has been tested and bears a UL (or equal)- label for a'three hour rating.
| |
| ~ _ ... _ _ . _ ,., . _ . _ _ _, _ _ _ , , _ . . . _ _ . . _ _ , _ _ . _ . . _ . . _ _ _ . . _ _ _ _ _ ,
| |
| : 2) Fire doore, which are required to have a UL (or equal) label certifying that they have been tested for a three hour rating per ASTM E119, including a hope st ream test.
| |
| : 3) Electrical penetrations which are required to have been type tested to ASTM E119, including a hope strean test.
| |
| : 4) Piping penetrations which are required to have been type tested to ASTM E119, including a hoes stream test.
| |
| : 5) fire dampers for any HVAC duct penetrating a fire barrier and which must have a rating of three hours. The only fire dampers separating divisions are in the HVAC duct for secondary containment (six total). The plant arrangement minimizes fire dampera.
| |
| : 6) Fire rated columns and support beams, which are required to be of reinforced concrete construction or, if of steel construction, enclosed or coated to provide a three hour rating.
| |
| : 7) Backup of the fire barrier penetration seals by the HVAC systems operating in the smoke removal mode. This backup feature is accomplished in the reactor and control buildings by maintaining a positive static pressure for the redundant divisional fire areas with respect to the fire area with the fire. Leakage is into the fire impacted area under sufficient static pressure to confine smoke and huat to the fire area experiencing the fire, even if there is a major mechanical failure of the penetration seal.
| |
| Other aspects of the ABWR design that minimize the risk due to fires while shutdown are
| |
| - HVAC systems dedicated to the divisional areas which they serve.
| |
| - - A smoke control system to remove smoke and heat from the affected area, to control the pressure in a room due to a fire, assure that any fire barrier leakage is into the fire area experiencing the fire, and supply a clean air path for fire suppression personnel.
| |
| The HVAC system has been designed for the dual purposes of HVAC and smoke control.
| |
| -~ Fire alarm systems, i
| |
| - Fire suppression system to automatically initiate, where appropri-ate,_and extinguish fires.
| |
| -- Manual fire suppression equipment such as hand held CO2 or chemical fire extinguishers, and hoses.
| |
| - Administrative controls to ensure that-at least one safety division is avaliable with intact barriers at all times.
| |
| Floodina Many of the features that are designed to mitigate fires also serve to protect the plant from damage due to flooding. Physical separation of safety divinions not only prevents propagation of fires but also restricts or prevents flooding of safety related equipment. The fire barriers will also prevent potential water from entering a divisional area due to flooding from non-divisional sources or contain water in the fire area for diviolonal water-sources. !
| |
| Other aspects of the ABWR design that minimize the risk from flooding are
| |
| ~
| |
| Ethe practice of not routing unlimited sources of water (e.g., service water) through divisional-areas and ensuring that other large water sources (e.g.,
| |
| suppression pool) can be contained without damaging equipment in more than one safety division if-a flood were to occur.
| |
| An analysis has been completed of all ABWR internal flood sources and the results show that during shutdown conditions no more than one safety division would be affected by water damage for any postulated flood. Features, beside separation, that contribute to this low level of risk are: Adequat ely sized room floor drains, water level alarms and automatic isolation of flood cources for potentially affected rooms, mounting motors and other electrical equipment on pedestals above floor level, and water tight doors. As was discussed under fire protection, administrative controls will be implemented to assure that at least one safety division with intact barriers is available at all times during plant shutdown. Additional details on the ABWR flood mitigation capability is contained in Appendix 19R.
| |
| gurnary of Fire and Flood Features >
| |
| The ADWR has been designed to minimize the risk due to fires or floodin; during shutdown conditions by plant configuration and system design. Divisional separation, both physically and electrically, as well as fire / flooding mitiga-tion systems exist to reduce plant risks from these potential accidents. Along with these design features, administrative controle are implemented to ensure that at least on safety division is operable and its physical barriers are intact if the barriers between two safety divisions must be breached to complete necessary maintenance or plant modifications.
| |
| 19Q.7 DECAY HEAT REMOVAL RELTABILITY STUDY _
| |
| 19Q.7.1 Introduction As part of the ABWR shutdown risk evaluation, a reliability assessment of the decay heat removal (DHR) capability was completed. Decay heat removal reliability has received increasing attention due to events auch as those at Vogtle and Diablo Canyon where decay heat removal systems were made inoperable due to loss of electric power and other causes.
| |
| Attachment 19QC summarizes approximately 200 events at operating plants which were either loss of decay heat removal events or precursors to such events. The relatively large number of events underscores the potential for loss of decay heat removal events and the potential for associated core damage.
| |
| | |
| e t
| |
| I l
| |
| conservative assumption because it is unlikely that all systems allowed t o be in maintenance would all be in maintenance at the same time. In addition, some systems in maintenance might be returned to service in time. The fault trees used in this study are contained in Attachment 190A. A mission time of 24 hours was used for this study. The loss of RHR event is assumed to terminate successfully if the mitigating systemn start and run for a period of 24 hours.
| |
| It is assumed that provisions for long term maintenance of decay heat removal will be made within 24 hours. This assumption is consistent w.th other full power PRAs.
| |
| A number of deterministic analyses were performed and documented in Attachment 1908. These include the estimation of time available for operator action and human reliability analysis to estimate the probability of operator error under various conditions.
| |
| The event trees were quantified w.th an initiating event frequency of 1.0.
| |
| Thus the core damage probability that is obt ained by this evaluation yields the conditional probability of core damage given a loss of decay heat removal event.
| |
| The event trees were quantified assuming various complements of systems to be available. The various minimum complements of systems that met the 1.0E-4 goal were selected for inclasion in Table 19Q.7-1.
| |
| Maintenance of the suppression pool was not modeled in this study. If the suppression pool level must be lowered for any reason, several options exist, such as: off loading all fuel in the RPV to the spent fuel pool or setting up alternate sources of make up water and ensuring means of decay heat removal other than use of the suppression pool exist.
| |
| 19Q.7.5 Core Damace Probability Goal and RPV Boilina The conditional core damage probability goal of 1.CE-4 was selected for this study for the following reasons. The initiating event frequency for loss of an RHR system is not included in this probability goal, but can be conservatively assur d to be 0.1 per year. In the analysis, it is
| |
| | |
| These minimum sets were determined for an initiating event involving loss of an operating RHR system. The three primary causes of a loss of the RHR system were identified to be the following:
| |
| : 1) Mechanical or electrical failures in the operating RHR system.
| |
| : 2) Loss of service water associated with the operating RHR system.
| |
| : 3) Loss of offsite power.
| |
| Less of service water and offsite power were evaluated separately as the _
| |
| cause for loss of RHR because of their impact on othqr DHR systems. Each potential cause for loss of the RHR system was considered an initiating event.
| |
| Recommended minimum sets of systems were identified baned on the assumption that the loss of RHR was due to 1) above. Modifications to the sets were made for loss of service water as the initiating event. Separately, a recommended minimum set of electrical power sources were identified to protect against loss of offsite power.
| |
| Success criteria were determined for each initiating event, taking into account decay heat load and plant operating mode. Minimum complements of systems that will prevent core damage given the initiating event and the time -
| |
| dependent core decay heat generation rate were then identified.
| |
| Event trees were generated based on the assumed initiating event and applicable success criteria. System failure probabilities were determined with the help of fault tree analysis.
| |
| The results from the study as summarized in Tables 19Q.7-2, 190 7-3, and 19Q.7-4. The tables show that significant flexibility exists for completion of system maintenance during outages while still maintaining adequate safety margins. These minimum sets of systems can be used i y utilities for initial outage planning and for evaluating changes to n"tage schedules to ensure _
| |
| adequats satety margins are maintained at all times during the outage. The risk goal can, in general, be met by just those systems required to be operable (and therefore available) by the ABWR Technice.1 Specifications.
| |
| | |
| - . _ _ . - - . - - . . . , . _ - - _ - - _ , _ . - . . - - . . - _ _ . _ - _ . - ~ . _
| |
| 1 i
| |
| i I
| |
| 19Q.7.4 Esthodoloav
| |
| :l t
| |
| The methodology used in this study waa the same utilized in full power PRAs (i.e., event trees and fault trees). The plant is assumed to be shutdown ;
| |
| with decay heat being removed by the RHR system in the shutdown cooling (SDC) mode. Loss of the operating RHR system is then assumed. The loss could occur ,
| |
| due to mechanical or electrical component f ailures of the RHR system, loss of i
| |
| service (i.e., cooling) water in the same division as the operating RHR system, 3
| |
| i or loss of offsite electrical power. The three types of failures are assumed to be initiating events.
| |
| P For each initiating event, the success criteria were determined. The success criteria are the minimum camplement of systems that are capable of preventing core damage. As the decay heat load is dependent on the time following shutdown, the minLmum systems required to remove the decay heat will also be ti.me dependent. Therefore, the success criteria have been determined as a function of time. Section 190 7.6 discusses the succeso criteria in more detail. ,
| |
| With the help of the success criteria, event trees for each initiating ;
| |
| event.were developed for each period. -A total of 18 event trees were analyzed.
| |
| Section 190 7.7 discusses the event trees. '
| |
| The branch points on'the event trees model the probability of success.and failure for each system included in the success criteria. The failure probability for each system was evaluated by a fault-tree analysis. The fault trees model potential system failures due to mechanical failure of components,-
| |
| loss of electric power to pumps or valves, or operator errors aerociated with manual actions (e.g., valve line ups or remote control of pumps and valves), c Unavailability due to maintenance wae modeled as follows. For a systs included in a minimum set,'the maintenance unavailability was taken to be 0 (i.e., the system is assumed to not be in maintenance). For a system not an a minimum eet, J the maintenance unavailability was taken to be 1. In other words, the system was assumed'to be completely unavailable. This is a very
| |
| | |
| __._-_._..-_._.-_..__.____._.~.-...____.--._._._...m__m.-.
| |
| I conservative assumption because it is unlikoly that all systems allowed to be in maintenance would all be in maintenance at the same time. In addition, some I systems an maintena.*.ca might be returned to service in time. The fault trees !
| |
| used in this study are contained in Attachment 190A. A mission time of 24 hours was used for this study. The loss of RHR event is assuned to terminate successfully if the mitigating systems start and run for a period of 24 hours.
| |
| !It is assumed that provisions for long term maintenance of decay heat removal will be made within 24 hours. This annumption is consistent with other full power PRAs.
| |
| A number of aeterministic analyses were performed and documented in l Attachment 1908. These include the estimation of time availab*a for operator action and human reliability analysis to estimate the probability of operator error under various conditions. f The event trees were quantified with an initiating event frequency of 1.0.
| |
| - Thus the core damage probability that is obtained by this evaluation yields the conditional probability of core damage given a loss of decay heat removal event. ,
| |
| The event trees were quantified assuming various complements of systems to be available.- The various minimum complements of systems that met the 1.0E-4 goal ,
| |
| were selected for inclusion in Table 190 7-1.
| |
| Maintenance of the suppression pool was not modeled in this study. If the suppression pool level must be lowered for any reason, several options exist, such ass'off loading all fuel.in the RPV to the spent fuel poci or setting up 4 alternate sources of make up. water and ensuring means of decay heat removal other-than use of the suppression pool exist.
| |
| 190 7.5 core Damaae Probability Coal and RPV BoilinS ,
| |
| i l
| |
| ' The conditional core-damage probability goal of 1.0E-4 was eclected for this study for the following reasons. The initiating event frequency for loss of an RHR system is not included in this probability goal, but can be conservatively. assumed to be 0.1 per year. In the analysis, it is .
| |
| h I ,
| |
| L
| |
| , . . _ , . . _ . .m.._- . . , . . _ . - _ _ . _ __._.,-_.....___.--.__-....__.._...,.-.~.,,,.-:
| |
| | |
| conservativo1y assumed that all systems not explicitly required to be kept out of maintenance are totally unavailable (i.e., all in naintenance). Then, the conditional core damage probab'lity of 1.0E-4 f or t he remaining eystemn would result in a conservative core damage frequency of 1.0E-5 1.er reactor-year.
| |
| This compares to the NRC goal of an overall core damage frequency of 1.0E-4 and a large release goal of 1.0E-6 rer reactor-year. In reality, loss of P.HR events occur less than 0.1 timen per year ar.d more intertantly, not all systems allowed to be in maintenance will all be in maintenance at the er.me _
| |
| time. Typically, the results show that more than six to ten oystems r.ro allowed to be ander maintenaace and there is a very low probability that all the systems will be simultaneously under maintenance. A., analysis using more realistic maintenance unavailability assumptions rebults in core damage frequency estimates less than 1.0E-7 pec reactor-year. The simplifying assumption of 0 or 1 for maintenance unavailability allows for the c.ilculation of core damage probabilities without having to model maintenance unavaitability for each system. This avoids discussion of overlapping maintenance periods for systems during outages. These conservative assunptiens allow for a straightforward determination of minimum oystem availabilities that also meet NRC risk goals.
| |
| In Mode 5 with the RPV head removed, it is ausumed that successful DHR can be achieved by allowing water in the RPV to boil and making up lost water by various water sources. Boiling under these conditions is an effective means of DH4 but it is not desirable because the resultant pressure buildup in secondary conta8nment could cause loss of containment integrity. Calculations presented in Attr.chment 19QB show that the boiling release rates, assuming no coro damage, are well below allowable limits for normal plant operations.
| |
| In addition, no operator actions are required in the potential steam environment of the reactor building and equipment relied upon to function after boiling ara qualified for operation in a harsh environment (e.g., HPCF, RHR).
| |
| Other equipment relied upon for inventory makeup are not subjected to the steam environment (e.g., condensate, AC independent water addition
| |
| | |
| - - . - .- ~ _ _ _ - - ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
| |
| hystems. Thus bailing in Mode 5 does not violate any offeite done limits, l
| |
| equiiment relied upon will not be adversely af f ected by the steam, and all eyerator actions required af tcr bol:.ing are outside the reactor building.
| |
| 19 Q . 'i . 6 .S;ipeces Q31tedq
| |
| ~
| |
| In order to prevent core damage given an initiating event, sufficient i
| |
| systems must be available to ensure that the core decay heat is removed And the fuel remains covered by water. No fuel damage will occur as long as the f uel remains covered by water. There are three ways to achieve successi a) remove decay heat directly from the coolant in the RPV; b) remove decay heat indirectly by conueWJing the steam produced, and provide makeup water to the RPV; and c) allow the coolant to bull in the RPV and provide makeup water to the RPV to keep the core covered.
| |
| These three ways to achieve success are discussed in detail below
| |
| : 1) Direct Ducay Heat Removal from RPV - Recovery of the failed RHR system, use of one of the other two RHR systems (SDC) or the reactor water cleanup (CUW) system (under certain plant conditions) is sufficient for succeae. The CUW nyatem capacity in temperature .
| |
| dependent and requires both pumps and nonregenerative heat exchangers (the regenerative heat exchangere must be bypassed). In Mode 5, the fuel pool cooling and cleanup (FPC) system can be used after the reacter cavity is flooded. FPC alone after 4.5 days or in conjunction with CUW at earlier timee is sufficient to remove all the decay heat.
| |
| Both FPC pumps and heat exchangers and the supporting systems are requ ired .
| |
| : 2) Decay Heat Removal and RPV Water Makeup - Under certain plant conditions the main condenser if available can be used to remove decay heat by condonsing steam. The MSIVs must be opened and a condensate return path to the RPV is required.
| |
| | |
| l l
| |
| }
| |
| : 3) _ Coolant Makaup - High pressure makeup can be accomplis'.ed by the HPCF, CRD, or feedwater ar.d condensato systems. Low pressure makeup is available f rom the condensate, LPFL or AC independent water addition i systems. Low pressure makeup may require depressurisation of the RPV i by actuation of ADS or individual sRVs. l
| |
| ! In Mode 5 with the RPV head removed, boiling of water in the RPV with adequate nakeup from low or high pressure sources lo considered success for the :
| |
| purposes of this studye T!J tcasons for boiling being a viable success criteria are discussed in Section 190 10.5. j t
| |
| Table 190 7-1 summarizos the loss of RHR success criteria. t Mitigation of loss of of f site power requires recovery of of f site power or use of the emergency diesel generators or combustion turbine generator. The AC independent water addition system can be used for make up in the event of a loss of all AC power.
| |
| i 19Q.7.7 Accident Proaression and Event Trees Loss of RHR may initiate from a failure in the operating RHR System, loss !
| |
| of Service Water system, or loss of of f site power. The accident progression for each of the above initiators is discussed below.
| |
| 19Q.7.7.1- Loss of RHR Due to Failure in the Operatina RHR System i
| |
| Following reactor shutdown, the plant is' cooled down by rejecting steam to the main condenser and making up water loss in the RPV by t he feedwater system.
| |
| The RHR system in the SDC mode can be initiated at about 150 psia which corresponds to approximately 360bF. The RHR system is then used to cool down to either Mode 4 (less than 200\F) or Mode 5 (refueling). Loss of the operating RHR-loop is. assumed to occur sometime after it has been initiated.
| |
| r
| |
| | |
| Table 19Q.7 1 SUCCESS CRITERIA POR PREVENTION OF CORE DAMAGE System (s) Comment 1 FOUt (SDC) A'.1 times when available, or Main Condenser If available, open MSIVs and establish condansate return path to RPV. _
| |
| or JUV If temp >234'F or af ter 14 days (using 2 pumps and using 2 nonregenerative heat exchangers and regenerative heet exchanger bypassed). .
| |
| or CUV + TPC Mode 5 only, after 4.5 days FPC alone is success. Both pumps and heat exchangers in each system required.
| |
| or 1 Feedwater + High pressure injection.
| |
| 1 Condencate or ,
| |
| i HPCF High pressure injection.
| |
| or 1 CRD High pressure injection, or 1 Condensate Low pressure injection (may need ADS), ,
| |
| or 1 LPFL Low pressure injection (may need ADS),
| |
| or 1 AC Independent Water Low pressure injection (may need ADS).
| |
| Addition System 47-
| |
| | |
| . . . . _- _ _ -. m.-.. _ _ _ _ . .. -. ____ _ .. .m _ __ .._..- m. -
| |
| a l
| |
| l
| |
| -koss of RHH in_MA d e 3 or 4 ,
| |
| t l
| |
| Figure 19Q.7-1 and 190 1-2 are the event troop for loss of RHR in Mode 3 or 4, respectively. Following loss of the operating CHR loop (event tree node RHR), the operator has to recognize the event and start following the correct .
| |
| procedure (CP). The sequence of events following the successful outcome at this node is described first. The operator can identify the failed system and request the maintenance crew to restore it to operation. An analysis showed ,
| |
| that for the decay heat load at this time, water in the RPV would begin to boil in 1.3'houra. Using a typical mean time to repair for the RHR system, and 1.3 hours as the time for recovery, the system recovery probability was determined ,
| |
| (R8C).
| |
| If the fajled RHR system cannot be recovered, the operator could initiate or.e of the other two RHR systems, if available, in the shutdown cooling mode (R). If all RHR systems fall, the RPV would pressurite and the main condenser could be made available (V2) by opening the MSIVs, drawing a vacuum in the ,
| |
| condenser, and operating the f eedwater condensate putops for makeup.
| |
| t If the main condenser fails or is unavailable, the operator can use the -
| |
| (XAf system to remove the decay heat (W2) if the RPV temperature in above 234 F.
| |
| If all DHR means are unavailable, the only path to success la to keep the .
| |
| core covered by either high pressure or low pressure sources. The high pressure sources are feedwater and condensate (Q), HPCF (UH), or a CRD pump (C). The HPCF initiates automatically whereas the other two systems require operator ,
| |
| action. If all these fail, the operator must depressurite the RPV by actuation of individual SRVs or ADS will initiate automatically (X) on low water level in .
| |
| the RPV. Successful depressurization would make the LPFL (VI), condensate (CDS), or.AC independent water addition systems available (FW).
| |
| Failure to depressurize the reactor or failure of FW 1eads to core damage.
| |
| . ~ . , -.w .-. .a - - . ., . . - . - - . - .., _ .-. ,. - . --,..a.-,-,- .-,-_..a.,,..
| |
| | |
| m__., _.___._.._ _ -.. __ _ _ _ _ _ _ ... - _ ~ _ _. _ - - .. _ _ _ _ _ _ _ ,
| |
| 4 1
| |
| g 1
| |
| If at node OP, the operator fails to follow the correct procudure, t.
| |
| 4- reactor coolant temperature and pressure in the RPV will rive, the SRVs will open and discharge steam to the suppression pool and eventually the HPCF will .
| |
| J initiate (UH) on low RPV water level. If HPCF fails, ADS will actuate on low .
| |
| water level (X). Failure to depressurize will lead to core damage. Following ;
| |
| successful reactor depreusurization, LPFL s.ill injoet on low water level (VI). .
| |
| Failure to inject ~LPFL leads to core damage.
| |
| Loss of RHk in Mode 5 i f
| |
| Figure 190.7-3 shows the event tree for less of RHR in Mode 5 less than 3 days after shutdown. This sequance is the same as the previous one except that l Since the RPV head is removed, the main condenser and feedwater pumps are
| |
| . unavailable and ADS is not required as the RPV cannot become pressurized. Also, !
| |
| i at thia iow temperature, CUW is not capable of removing the decay heat.
| |
| i Figure 19Q.7-4 shows the event tree for loss of RHR in Mode 5 for 3-4.5 dayn;after shutdown. If the reactor cavity has been flooded ar.d the epent fuel {
| |
| pool gates are opened, the fuel pool cooling and cleanup (FPC) and CUW syotems 7 together (RF) have adequate capacity to remove the decay heat 3 days following uhutdown. Use of these systems requires operator action. The other success paths are_the same as.in the previous event tree (Figure 19Q.7-3). Figure 19Q.7-5 shows the event tree for loss of RHR in Mode [ for the period 4.5-14 days and Figure 19Q.7-6 shows the event tree for greater than 14 days.
| |
| The dif ferences in these event treen are that for the 1eriod 4.5-14 days FPC elone is success (FPC) and-beycnd 14 days CUW alone (W2) is success.
| |
| - 19Q.7.7.2. Loss of RHR Due to Loss of Service Water
| |
| * Figures 190 7-7 through 19Q.7-15 show the event truas for loso of service water.- The scenarios are basically the same as for a loss of RHR except that loss of service water may impact other DHR or makaup systems in addition to the operating RHR pump. For example, loss of Division A service water also results in loss of CUW and FPC. Likewise,. loss of Division C service water causes loss '
| |
| of HPCF(C) in addition to the Division C RHR pump.
| |
| | |
| I i
| |
| 19Q.7.7.3 Loss of RHR Dg.e to Loss of of f site PowtI Figcres 19Q.7-16, 19Q.7-17 and 19Q.7-18 show the event trees for loss of of f site power in Modes 3, 4 and 5 respectively. The success criteria are the
| |
| - same but longer time is available for recovery in.Hode S. Following a loss of offsite power, it is possible to recover power in time to prevent core damage.
| |
| If power is not recovered the available Do will start automatically, and if the DG fails, CTO can be manually initiated. Following loss of all AC power, the AC j independent water addition system can be used for make_up if the RPV can be depressurized by opening SRVs.
| |
| 19Q.7,8 System Fault Trees i
| |
| The unavailability of a system to perform its safety function on demand given a loss of RHR was evaluated by fault tree analysis. Eleven fault trees t were used in this analysis. The fault trees are contained in Attachment 19QA.
| |
| - Five of the fault trees. HPCF, RHR (SDC), RHR (LPFL), Reactor Building Service Water, and ADS were taken from the full power PRA with modifications
| |
| . (e.g., maintenance unavailability and operator actions) to reflect shutdown conditions. The other six fault trees were developed specifically for the shutdown PRA and includes Reactor Water Cleanup, Fuel Pool Cooling, Main Condenser, CRD, Condensate, and Feedsater, ,
| |
| The' fault trees model system unavailability due to mechanical failures, loss of power,.and operator errors. As previously mentioned, maintenance unavailability is either assumed to be 1 or 0 (i.e., system is in or out of f maintenance).
| |
| The unavailability of the AC independent water addition system was
| |
| - estimated based on the assumed' operator error in manually initiating the system. ;
| |
| g' gnn-w 9-JWrey y--- g g8ypyg -w vwy- w' $ PT'WWN''''* TTNyy TT"'M*"4"#'T*-T*W9 WhvN"$ M & 3 M' 'TV'*"9- t- T *?7--''=-*P?MvF T"'m -97Y-*MtW'M'PWY"*NT Tt'r'*'t*f' ^WP-N-YW'*PP'*FW''
| |
| | |
| 4 b N k NN A b b MY 'M kk MN ONE RHR FOLLCMrS OE THE RHR 9 OR RHRC CONDEN564 PUUPe HPCFC LOOP CONCXM"MTE 8". asp INS M f*OC) CONDENSnTE P dWP W
| |
| RNA OP AEC A '#2 W2 O UM C X Vf CDS FW 09
| |
| - Out
| |
| 'At mA CM WGr, eve %
| |
| , mics CW i,4 GJ
| |
| &~A wtw
| |
| . m
| |
| + a-se., ,
| |
| M
| |
| "" Ud* I R Rt2 *
| |
| * E e.w m a \
| |
| CM Cat Je%,
| |
| i desW
| |
| ~ * " "
| |
| g p 24 mAsem u
| |
| 'W 1
| |
| FIGURE 190.7-1 LOSS OF ONE RHR - MODE 3, STARTING AT 4 HRS t
| |
| | |
| ~~
| |
| "\ _
| |
| # ^
| |
| F5EWARR CLASS $EQ 9WLW WB OR 086 CAO ACS ANY M ONE ErntER 5 4R- IAAN CtfW FEEtwa7ER LOOP CONDENSATE PUWl5' LOSS OF OPETuTOR RECOVERY F'Jh8P
| |
| * HPCFC FOLLOWS OF T)E 5958 9 OR RHRC CONDENSER Pt W ONE RHR CONDEMSATE Pv57R (SOQ Pune x vs CDs N w2 O tM C RHR OP REC R V2 OK Out OW
| |
| +%n ,
| |
| O1 + n Fww e sk W !eV y
| |
| Ost
| |
| # m.
| |
| *iam g M
| |
| l we ' ,x ptr e we h dw e4 A e *1 r
| |
| s evn 4s x >3 W
| |
| l CW.
| |
| 1 i I g gg v tedTUri aAw M R?24
| |
| ~ <v n >>
| |
| 1 p*w sa er 1
| |
| FIGURE 190.7-2 LOSS OF ONE RHR - MODE 4,6 HRS TO 2 DAYS
| |
| | |
| I LOSS OF OPERA._ l RECOVERY E11HER RHR- HPCFB OR ONE CRD ANY RHR ONE FtREWATER CLASS SEQ NAME ONE RHR FOLLOWS OF THE RHR B GP RHR-C HPCFC LOOP CONDENSATE PUMP INSTRUCTION (SDC) PUUP S
| |
| RHR OP U. R LH C V1 CDS PN i
| |
| 4 OK i
| |
| OK REG 90 1
| |
| RSMTDBG OK H D BC RHR GROU OK i
| |
| WDG5a6 OK coss x gy,
| |
| , F:HbWAlH OK I
| |
| OPERATOR OK 4 MPCFOC X Rn2 WUG56G t
| |
| RHR2 TRE 6-N92 i FIGURE 190.7-3 LOSS OF ONE RHR - MODE 5,2 TO 3 DAYS t
| |
| i
| |
| | |
| 1 CONDENSATE FIRLWATER CLASS SEO NAME ,
| |
| CUW & FPC HPCFB OR ONE CRO AtN RHH OPERATOR RECOVERY l EITHER RHR- LOOP PUMP l LOSS OF HPCFC ONE RHR FOLLOWS OF THE RHR B OR RHR4 !
| |
| INSTR (SOCl V1 CDS FW R OF UH C l ,
| |
| RHA OP REC l OK OK OK REG 45 OK FtSHTOUC OK FtWCUF PC OK HPCFBG OK FiHR CRbu
| |
| ~
| |
| WDCSBC OK F iHL w AT H OK OK g
| |
| GPERATOR I X R34-2 HEO BC . . _ .
| |
| WDubem l
| |
| RHR3TRE 6 29-92 l
| |
| l FIGURE 190.7-4 LOSS OF ONE RHR - MODE 5,3 TO 4.5 DAYS
| |
| | |
| (
| |
| OrJE FIREWATER CLASS SEO NAME FPC HPCFD 05 Of4E CRD APN RHR OPERATOR RECOVERY EITHER RHR- LOOP cot 4DENSATE PUMP LOSS OF HPCFO ONE RHR FOt10WS OF THE RHR B OR RHR-C PUMP INSTR (SDC) ,
| |
| V1 CDs FW FPC U.1 C OP REC R RHR OK OK OK H Et.A 5 OK RSHTDBC OK FPG HPCFBG N
| |
| hMO C4DU OK WDGSOC y l # I X R414-1 p,g ;g OK l OK OPERATOR X R414-2 00 g RHA4 TRE 4 29-92 FIGURE 190.7-5 LOSS OF ONE RHR - MODE 5,4.5 TO 14 DAYS l
| |
| | |
| ONE FIREWATER CLASS SEO P4AME HPCFB OR ONE CRD AfiY Fe PLWP EffHER FWR- FUEL POOL CUW LOOP CONDENSATE LOSS OF OPERATOR BECOVERY FPCFC OF THE RHR 9 OR RN COOUNG PUMP ONE RHR FOt10WS NSTR (SOC)
| |
| V1 CDS FW l W2 LH C REC R FPC RHR OP OK OK OK REC 20 OK ftsHTD6C OK FFG l RWCtf OK HFif N RHR @
| |
| GRDU WDGir% g OK R14T-t gy X g
| |
| OK OK y #
| |
| CFERATCil I X R14D 2
| |
| #UfEU .
| |
| RHR5TRE &2492 f
| |
| FIGURE 190.7-6 LOSS OF ONE RHR - MODE 5,14 TO 33 DAYS
| |
| | |
| LOSS OF OPERATOR PECOVERY EITHER RHR- MAIN FE EDW ATER HPCFB OR ONE CRO ADS RHR B OR C ONE FIAEWATER CLASS SEC NAUT Div A FOLLOWS OF THE B OR RHRC CONDENSER PUMP + HPCFC CONDENSATE PUMP SERVICE INSTR SERVICE (SOC) CONDENSATE PUMP WATER WATER PUUP SWSA OP REC R V2 O UH C X V1 CDS Fw l
| |
| c.< -
| |
| OK OK REcwo RSHIDisG MGU OK E WMG O*(
| |
| ** b GF oG OK WuGtk% Og U"W Mk I X i.ni4Aip A12-1 g39-g3 x A12-7 OK i
| |
| OK OF EP A106 X AT2'O gpgg g v'wbGM6 X 412 4 H Abi ed53 SWSAIV3 TDE 6 N92 FIGURE 190.7-7 LOSS OF DIV A SWS - MODE 3, STARTING AT 4 HRS
| |
| | |
| LOSS OF OPERATOR . RECOVERY EITHER RHR- ' MAIN FEEDWATER HPCFB OR ONE ORD ADS RHR B OR C CNE Fi% VATER CLASS SEC NAME DIV A FOLLOWS OF TPE B OR RHRC CONDENSER PUMP + HPCFC CONDENSATE PUMP SERACE INSTR SERVtCE (SOC) CONDENSATE PUVP WATER WATER PUuP SWSA OP REC A V2 O G4 C X Vt CDS 1 fvt cK OK RECW3 og RCHTDBC OK VCU OK FWCOND N OK SWSA HPCFBC OK W3CcBC cnSJ mr i M I rtoEwsin g ,,,_q RXDODESS y 497 g OK OPEr1ATOR OK Hocrge IC9C X A12-3
| |
| ' TRESS y ggys SwSA1.TAE 6 M 92-FIGURE 190.7-8 LOSS OF DIV A SWS - MODE 4,6 HRS TO 2 DAYS
| |
| | |
| k LOSS OF OPERATOR RECOVERY EITHER RHR- HPCFB OR ONE CRD RHR B OR C ONE FIREWATER CLASS SEO NAME DIV A FOLLOWS OF THE B OR RHR-C HPCFC CONDENSATE PUMP SERVICE INSTR SERVICE (SDC) PUMP WATER WATER SWSA OP REC R UH C VI CDS FW OK OK OK RECW OK RSHTDBC OK HPCFBC CHOU OK SWSA WDCSBC OK CONF I FIREWAIR OK OFERATOR OK X A233 2 WDCSBC SWS A2 TRE 6-2492 FIGURE 190.7-9 LOSS OF DIV A SWS - MODE 5,2 TO 33 DAYS
| |
| | |
| ADS RNR A OR B ONE f rtEWA TER CLASS SEO NAW MA N CUW FEEDWATER HPCFB ONE CRO Pl%ED-LOSS Cf OPERATOR RECOVERY EITHER RHR. PUUp + CONDE NSA IE DfV C FOLLONS OF THE A OR RHR 9 CONE NSER PtNP SERvCE CONDE PtSATE SERVCE NSTM (SDC)
| |
| PUMP WATER WATER C X Vt CDs Fw REC A V2 W2 O UH SWSC CP OK C4 OK m.ve; OK R siiGha MOU OK H.'avuF OK 6 Av0f+.> OM sw si.; OK AASAs , OK UW ' y cer i
| |
| _NA_n e r it . ,
| |
| X CT2 2 Mi+ nc iG OK M
| |
| i u trMGn X C12-3 wha gjg - C,2 A
| |
| ....,a SWSC 5 W3. TAE 62992 FIGURE 190.7-10 LOSS OF DIV C SWS - MODE 3, STARTING AT 4 HRS
| |
| | |
| +
| |
| LOSS OF OPERATOR RECOVERY EITHER RHR- MAN CUW FEEDsATER HPCTa OPE r#9 ADS RHR A OR B ONE FITWATER lCLAS$5 SEQ NAME DIV C FOLLOWS OF THE A OR RHR4 CONDENSER PUMP + CONOENSATE PUwP i SERVCE NSTR SERVCE (SDC) CONDENSATE PUUP WAttA WATER PUW SWSC OP REC R Y2 W2 - 0 UH C 'M i VI CDs tw 04 OK M
| |
| mw F6HiUAa Wu OK tiAs OK
| |
| & 2.wtw 3AN f orte t$ OK euM5 j ,
| |
| 04 bevu w ., I
| |
| & ,sc eA T;,
| |
| M C121 M Ct2 2 iu %55 04 Cfivviair E
| |
| ~*
| |
| 4iMs A C'24 ws r,w SWSC1 TT-E 6 29-s2 FIGURE 190.7-11 LOSS OF DIV C SWS - MODE 4,6 HRS TO 2 DAYS
| |
| ~
| |
| | |
| t 4
| |
| LOSS OF OPERATOR RdCOVERY EITHER RHR- HPCFB ONE CRD RHR A OR B ONE ' FIREWATER CLASS SEO NAME' DIV C FOllOWS OF THE A OR RHR-B CONDENSATE PUMP..
| |
| SERV!CE INSTR SERVICE (SDC) PUMP WATER WATER ,
| |
| SWSC OP REC R ,
| |
| UH C V1 CDS _l FW
| |
| ^
| |
| OK OK REG 90 OK RSHIDAB OK HPCFB OK ;
| |
| SWSC CADU OK WDGSAS OK K C23-1 HHEWA!H OK OPERATOR OK HPCFB X C23-2 WDCSAB SWSC2.TRE 6-29-92 FIGURE 190.7-12 LOSS OF DIV C SWS - MODE 5,2 TO 3 DAYS i
| |
| e 4
| |
| 4
| |
| | |
| t LOSS OF OPERATOR RECOVERY ETTHER RHR- CUW & FPC HPCFB ONE CRO RHR A OR B ONE FIREWATER CLASS SEO NAME DfV C FOLLOWS OF THE A OR RHR-B CONDEtEATE PUMP SERVICE - INSTP SERVICE (SDC) PUMP WATER WATER SWSC OP REC R RF UH C VI CCS FW
| |
| 'l w
| |
| OK M ;
| |
| REC 45 OK R5HIDAB M
| |
| F1WC.UFPG OK HPCFB CRDO OK WLCSAB OK Core hHEWA1H M- .
| |
| l og-OPERATOH X C34-2 y,ggg SWSC3 THE 6- & 92 FIGURE 190.7-13 LOSS OF DIV C SWS - MODE 5,3 TO 4.5 DAYS
| |
| | |
| LOSS OF OPERATOR RECOVERY EITHER RHR- FPC HPCFB ONE CRD RHR A OR B OrJE FIREWATER CLASS SEO NAME DIV C FOLLOWS OF THE A OR RHR-B CONDEPGATE ' PUMP SERVICE INSTR- ' SERVICE (SOC) PUMP WATER WATER SWSC OP - REC R FPC UH C VI CDS FW w
| |
| m DEC45 W
| |
| RSHTDAB FFG HPGfB SWSG GADO @
| |
| wDesAs a p,gggggg X G414-1 OPERATOR OK l HNFO #~ 2 WENSAB SWSC4 TRE 6-29-92 FIGURE 190.7-14 LOSS OF DIV C SWS - MODE 5,4.5 TO 14 DAYS ;
| |
| | |
| ONE FIREWATER CIASS SEO NAME CUW HPCFB ONE CRO RHR A OR D RECOVERY EITHER FWR FPC CONDEt4 SATE PUMP LOSS OF OPERATOR PUup DIV C FOLLOWS OF THE A OR fRRO SERVKE INSTR SERVICE (SUC)
| |
| WATER WATER V1 CDS FW W2 LH C REC R FPC ,
| |
| SWSC OP OK OK OK REC 20 OK R5HIDAB OK FPG OK RWCOF hFCFB SWSC
| |
| * CFw (vDG5AB .
| |
| OK C# I g g4g gg7 g X mm OK g
| |
| OF ERATOFi 3 X C14D 2 6tFCFs g SW3C5TRE 6-2432 FIGURE 190.7-15 LOSS OF DIV C SWS - MODE 5,14 TO 33 DAYS
| |
| | |
| 1 LOSS OF RECOVERY ~ ALL DIESEL CTG ADS FIREWATER CLASS SEO NAME OFFSITE OF LOOP GENERATORS PUMP POWER NOT Fall LOSP1 REC DG CTG X FW OK LOSP OK i
| |
| REC 90 OK DGF OK CTGF X L12-1 RXDPRESS X L12-2 I
| |
| LOSP1M3 TFiE &n92 -
| |
| FIGURE 190.7-16 LOSS OF OFFSITE POWER - MODE 3, STARTING AT 4 HRS ,
| |
| | |
| i ADS FIREWATER CLASS SEQ NAME l RECOVERY ALL DIESEL CTG LOSS OF PUMP OFFSITE OF LOOP GENERATORS
| |
| , POWER NOT Fall CTG X FW LOSP1 REC DG OK l
| |
| OK LOSP OK REC 90 OK DGF X L12-1 CTGF RXDPRESS X L12-2 LOSP1TRE G N 92 FIGURE 190.7-17 LOSS OF OFFSITE POWER - MODE 4,6 HRS TO 2 DAYS l
| |
| | |
| IREWATER N
| |
| , LOSS OF RECOVERY OF ALL DIESEL CTG NOT Fall CLASS SEO NAME
| |
| .OFFSITE POWER LOOP GENERATORS NOT Fall i
| |
| LOSP1 REC. DG CTG FW OK LOSP OK REC 50 OK DGF OK CTGF 33-1 FIREWATR LOSP2_TRE 6 29-92 FIGURE 190.7-18 LOSS OF OFFSITE POWER - MODE 5,2 TO 33 DAYS
| |
| | |
| l 19Q.7.9 Fesults and Conclusions 19Q.7.9.1- Introduction The event trees described in_the previous section were evaluated and the
| |
| - core damage probability caleclated with certain systems assumed u_available due to maintenance. In general, the minimum set of equipment asst.ned to be available was initially taken as that required by the Technical Specifications for the given operating mode. Combinations of systems were made available until a set restited in a conditional probability of less than 1.0E-4. Each of these sequences that met the 1.0E-4 criterion is considered a minimum set for assuring acceptable shutdown risk.
| |
| Minimum sets were obtained for each of the three loss of RHR initiators Loss of Operating'RHR System, Loss of Operating SW System and Loss of Offsite Power.
| |
| Tables 19Q.7-2 through 19Q.7-4 list certain minimum sets of systems that meet.the 1.OE-4 criterion for loss of the operating RHR system initiator for the
| |
| - three major.configuratione during shutdown. The configurations are: Modes 3 or
| |
| ~ 4, Mode 5. prior to flooding the reactor cavity,.and mode 5 after the reactor 4 cavity has been flooded. The effect of changes in decay heat, as a function of timei will be discussed-for each of the three plant conditions.
| |
| ~
| |
| With about112 systems available, and about four needed to reeet the goal,
| |
| - many minimum sets can be identified. In order to simplify the selection of minimum set systems, the following maintenance philosophy was assumed: all of division C in maintenance, division B, ADS and combustion turbine generator are available. (other maintenance philosophies can be adopted and the model used to
| |
| - identify. appropriate minimum systems). Additional details of the plant configuration based on selected maintenance philosophy is as follows. The plant is being cooled through use of RHR "A" and its support systems (i.e., service water "A", RCW "A", electric power division "A"). Other
| |
| | |
| t Table 19Q.7-2 MINIMUM SETS OF SYSTEMS FOR MODES 3 AND 4 MAIN FIRE RHRB CONDENSER CW HPCFB CRD RHRB (CF) CONDENSATE VATER
| |
| : 1) * *
| |
| * 2)_ * * *
| |
| : 3) * * *
| |
| : 4) * *. *
| |
| : 5) * * * *
| |
| : 6) * * *
| |
| * 7)- * * * *
| |
| : 8) * *
| |
| * r 4
| |
| I I
| |
| ', w n r. , . - . , , - . , , - - . ---s.-,.. ~a...
| |
| | |
| ,, . . _ _ . . - _ , . . . ~. . . . . . . . . . . . , . . . . . _ . . _ . _ _ _ . . - . . . - , . . _ . . _ . _ .
| |
| y e, - s.
| |
| +
| |
| g F
| |
| Table-190 7-3.
| |
| MINIMUM SETS OF. SYSTEMS FOR MODE 5 (UNFLOODED) t
| |
| ~
| |
| RHRB' HPCFB- CRD_- RHRB (CF) CONDENSATE. FIREWATER 2)
| |
| : 2) *-
| |
| -a *
| |
| * 3)
| |
| .4) -
| |
| 4 6
| |
| e p
| |
| . . . . . . . , . . . . . .- . - . . . - . - . . - - . - ~ . , ~ . ~ . . . . . . . - . , . - . . . . - .
| |
| | |
| . _ _ ,. .. . _ __-.. . _ . . . .. . . . . . - .. . ~ . _ . ._
| |
| m j
| |
| J Table 190 7-4 HINIMUM SETS OF SYSTEMS FOR MODE.5 (FLOODED)
| |
| RHRB CUW & FPC - (, A J RHRB (CP) CONDENSATE. SPCFB FIRE WATER
| |
| *- a 1)
| |
| - 2) 3)-
| |
| : 4) -
| |
| F I
| |
| O i
| |
| , - v- - , - = -,-e, e,--- , , , , - * ,
| |
| | |
| _ division "A" systems, including EDG "A" may be in maintenance unless
| |
| ==
| |
| specifically included as a support system in one of the minimum sets. All division "B" systems are ascumed to be not in maintenance, although they may become unavailable due to random failures or operator errors. All division "C" systems are assumed to be in maintenance with the exctption of division "C" electric power to support opening of the suction valve for RHRB. Alternatively, this valve can be opened manually. This configuration was selected because it is one that meets minimum technical specification requirements (i.e., 2 ECCS and 2 RHR systems available). Other configurations could have been selected but _
| |
| this one is typical and the resulting minimum nets identified will demonetrate the low risk associated with loss of decay heat removal for the ABWR and the flexibility afforded utilities for outage maintenance scheduling while still ry. maintaining low risk levels.
| |
| ==
| |
| 19Q.7.9.2 Loss of RRR Initiator The minimum set for loss of RHR is discussed first. Table 19Q.7-2 lists ~
| |
| some minimum sets of systems that if available during mode 3 or 4 meet the core damage criterion. As can be seen, if the 2 ECCS systems are assumed to be RHR, then only a CRD pump or CUW plus AC independent water addition need be made available. This is not restrictive since one pump from CRD and firewater are usually available for other reasons (e.g., CRD to purge the FMCRDs and AC independent water addition for fire protection) and CUW is usually operable during this period. The table shows eight dif ferent minimum sets. This is indicative of the flexibility for performing ABWR shutdown maintenance while still maintaining risk margins.
| |
| --- Table 19Q.7-3 lists some minimum sets of systems for Mode 5 following 2-3 days after shutdown. In this configuration the RPV head bolts have been detensioned and the head is off but the reactor cavity has not been flooded.
| |
| For this Mode 5 configuration, fewer systems are available than during Mode 3 or 4 or after flooding the reactor cavity but enough systems are available to ensure adequate risk margins. Also, this is a relatively short duration of the outage. The main condenaar is not ;vailable since the RPV cannot be l
| |
| | |
| .l
| |
| . pressurized, fwel pool cooling cannot be used because the RPV and fuel pool have not'boen connected together and CUW capacity is not sufficient to remove'all the decay - heat due to the ' low RPV. temperature. The table shows four minimum sets of systems which~ meet the risk criteria. As was noted for Mode 3 and 4, the CRD pump which is-normally available in addition to minimum technical specification systems meet the core damage criterion. As time following shutdown increases beyond 14 days,fthe decay beat' load decreases to the point that the CUW system is capable of-removing all the decay heat. This allows even more flexibility in scheduling maintenance tasks.
| |
| Table 190 7-4 lists four minimum sets for mode 5 following 2-3 days af ter i
| |
| flooding of the reactor. cavity. CRD plus either RHRB or FPC. cod CUW meet the criterion. After-4-5 days, fuel pool cooling alone is sufficient and after 14 days CUW or fuel pool cooling are sufficient. As time following shutdown increases, more systems'bocome able to remove decay heat and greater time is available for' operator actions prior to boiling or core damage.
| |
| As-tables 190 7-2-through 19Q.7-4 illustrate, many combinations of systems can be made available to ensure adequate shutdown risk while still allowing for maintenance to be performed on systems. As previously mentioned, these minimum sets are only a few of the possible combinations-that will ere2re adequate shutdown risk margins, other minimum sets can be identified for different assumed plant conditions. P.n important point that-is illustrated by the minimum
| |
| ! sets:1dentified in this study is that under all shutdown plant conditions, minimum technical specification requirements plus systems that are normally
| |
| ' operating-during shutdown (i.e., CUW, FPC, CRD, and fire water) are enough to L ' ensure adequate shutdown risk margins.
| |
| l
| |
| .19Q.7.9.3 Loss of SW Initiator If loss of reactor building-service water (RSW) is aestmed to be the initiating event, Division "A" of CUW and FPC are also lost. AS both
| |
| | |
| , _-- . - - . = .-. . - . ~ . .-. . _-, .-_-. - _ _ ~ - . - . - ~ .. .~
| |
| M l
| |
| divisione of CUW and/or FPC pumps and heat exchangers are required to removeL decay heat, CUW and FPC must be assumed unavailable. Thus in table 19Q.7-2, the three minimum sets that assume CUW available would not apply for loss of RSW.
| |
| Also, the one set in Table 19Q.7-4 that assumes availability of-CUW plus RPC 1 1
| |
| I would not be applicable.
| |
| '19Q.7.9.4 Loss of offsite Power Initiator For a loss of offsite power initiator calculations have shown that having one EDG and the CTG available along with appropriate credit for recovery of offsite power results in a probability of less than 1.OE-4. For core damage to occur, loss ot AC independent water addition or ADS must occur. This scenario is less than 1.0E-4 and-thus meets the criterion. Since the above maintenance
| |
| -philosophy assumes only EDGB and the CTG are available, no additions to the minimum sets already identified are required for the loss of offeite power initiator.
| |
| 19QA.7.9.5 Adeouacy of Technical Specification From the above results, the following can be stated regarding adequacy of
| |
| 'the ABWR Technical Specifications. In Mode E the onset of boiling is most idependent on water level (or total inventory), and thus the most vulnerable condition is at low water level prior to flooding up of the reactor cavity. In this condition,-not only is the time to boiling relatively insensitive to decay.
| |
| -heat. level,.but RHR in shutdown cooling (SDC) is the only source of decay heat
| |
| -removal. This is the basis for the Technical Specifications requiring that two loops of RHA SDC be available in this condition; one normally operating and one in standby. Results of the analysis.show that given the loss of the operating RHR loop, with one RHR loop in standby there is a 5.OE-2 probability of the onset of boiling. Assuming a one in ten probability of the initiating event,
| |
| ' boiling would occur once in two hundred years. This is acceptable given the short time duration the plant is expected to be in this unique condition and the benign consequences that are calculated to result so long as core damage is avoided. Clearly, a utility could further reduce the likelihood of boiling by assuring that.the third division of RHR SDC provided in the ABWR design is
| |
| ~ .__ __ . _ ,
| |
| | |
| . - - . -- .. . ---- . . - . ~ . - . ~ - - - ~ .
| |
| l l
| |
| l l
| |
| l 1
| |
| available during these conditions. Thus, during the early stages of the transition from Hode 4 to Mode 5 (prior to flood-up), the availability of the third loop of-RHR SDC further reduces shutdown risk. However, given the other compensatory measures available to delay the onset of boiling and prevent core f
| |
| damage _(e.g., condensate, AC independent water addition, CRD),-the two loopt of RHR SDC required by ABWR Technical Specifications are more than adequate during +
| |
| these plant conditions.
| |
| The results of this study cen be used by utility personnel to optimize outage planning schedules while assuring acceptable shutdown risks at all times during a plant outage. These results should be useful in initial outage planning and rescheduling outages to meet unexpected system failures or other unplanned activities.
| |
| v 19Q.8 -USE OF FREEZE SEALS IN ABWR Freece seals are used for repairing and replacing such components as valves, pipe fittings, pipe stops, and pipe connections when it is impossible to isolate the area of repair any other way. Freeze seals have successfully been used in pipes as large as 28 inc. , in diameter.
| |
| The ABWR design has eliminated a significant amount of piping associated with the reactor _ coolant system.(e.g., no recirculation loops). This by_itself
| |
| .will reduce the necessity for freeze seals in ABWRs over other plant designs.
| |
| In addition to reduced RCS piping, the ABWR design has most piping con-nected to the reactor pressure vessel (RPV) enter at a level significantly higher (five-feet) than the top of active fuel. Inadvertent draining from these lines will automatically stop without exposing the fuel. 'The only piping con-nection below the top of active fuel (reactor water cleanup system) is small in size (< 2 inches). - If a freeze seal were required on this line and it were to fail, several sources of makeup are availiale to refill the RPV to prevent core r
| |
| uncovery.
| |
| | |
| Whenever freeze seals or other temporary boundaries are used in the ABWR, ,
| |
| administratxvc procedures will be necessary to ensure integrity of the temporary boundary. Also, mitigative meaaures will be identified in ad..nce and appropil-ate backup systems made available to ensure no loss of coolant inventory occurs.
| |
| ~An option that a utility could choose is to off-load all the fuel in toe RPV to the spent fuel pool when repair or maintenance of an unisoable valve must be completed.
| |
| The selected method for working on unisoable valves must take into account adequate safety margins, personnel experience with freeze seals, availability of backup systems, and the potential 14npact on other outage activities.
| |
| 19Q.9 $3 1QOWN-VULNERABILITY RESULTING FROM NEW FEATURES The ABWR has incorporated many new design features that do not exist in current operating domestic BWRs. These features have been added based on past operating experience, advances in technology since earlier designs were finalized, and the results of detailed probabilistic risk assessments (PRAs).
| |
| In order to evaluate the potential shutdown risk associated with these new ,
| |
| features, a Failure Modes and Effects Analysis (FKEA) was completed for each new feature. The feature is identified followed by potential failure mode (s). The possible method for detecting each f ailure mode is then presented followed by the potential impact on safe shutdown and any preventive or mitigating feature that may exist. Finally, the overall shutdown vulnerability evaluation is descrlhed.
| |
| The FMEA is contained in Table 19Q.9-1. As the results presented in Table 19Q.9-1 show, there are no identified vulnerabilities resulting from implementation of new design features in the ABWR that affect shutdown risk.
| |
| | |
| TABLE 19Q.9-1 SHUTDOVN VU1JERABILITY EVALUA' LION OF NtN ABVR FEARTRES Shutdown Potential Preventive /
| |
| ' Failure .
| |
| How Impact on Mitigative. Vulnerability Feature Mode Detected Safe Shutdown Feature' Evaluation Reactor' Inter- RPV leakage Visual Inventory loss, Multiple seals, None, past expe-nal Pumps' during mainte- indication of. fuel uncovery. admiristrative rience with (RIPS) nance. leakage. controls, maintenance on RIFs indicates o no concerns. E
| |
| -- 5,.
| |
| ~J m sJ . Combustion Tur- Fails to start No output. Logs of Two independent None, adequate 5 2( bine Generator (CTC) or pick up load. voltage on demand or test.
| |
| electrical power redundancy off-site power sources and off-site and on-site power
| |
| }'
| |
| three Emergency sources exist if '
| |
| Diesel Genera- CGT were to O tors (EDCs). fall. 4 h
| |
| w Improper syn- Less of bus Less of vital Two other divi- None,-redundant -
| |
| chronization to voltage when CGT power bus. sions capable of power' supplies existing power output breaker supplying vital and administra-sources. closes on demand power, auto syn- tive controls /
| |
| or test. chronization antisync circuit circuit, admin- prevent any 1strative impact on safe !
| |
| controls. shutdown.
| |
| Third EDC Fall to start or 'No voltage on Loss of power to CGT capable of None, increases pick up load. vital bus on one vital. bus. feeding any number of demand or test. vital bus, two on-site vital. ,
| |
| independent bus sources.
| |
| sources of
| |
| - off-site power.
| |
| | |
| T Shutdown Potential Preventive /.
| |
| Failure flow Impact on Mitigative Vulnerability Feature Mode Detected Safe Shutdown Feature, Evaluation Third ECCS Single failure Safety function Loss on one ECCS Two other -None, increases Division results in loss not completed division. divisions nember of ECCS +
| |
| of third- ECCS (e.g., no ECCS capable of com- divisions to division. flow given ini- pleting safety cooplete safety ,
| |
| tlation si 6nal) function. functions, on demand or allows for ECCS test. maintenance ,
| |
| without total n loss of redun- E
| |
| ~J dancy, separa- $
| |
| '# tion reduces r-common mode p.
| |
| failure suscep- p[
| |
| tibility. g ;
| |
| o 8
| |
| x Micro Processor Falls to ECCS function Loss of ECCS High reliability None.. increased "2 Based Safety initiate safety not completed on function. with self test reliability of h System Logic signal. derend or during feature. ECCS logic, test.
| |
| Fine Motion Falls to control CRD does not Reduced shutdown Only one CRD can None, adequate Control Rod CRD motion on move when margin. be withdrawn at preventive /
| |
| Drives demand. directed or spu- a time, RFS mitigative (FMCRDs), rious movement. active during features exist.
| |
| Alternate Rod shutdown (hi-Insertion (ARI) flux or manual trip). j r
| |
| i
| |
| | |
| l I
| |
| Shutdown P,tential Preventive /
| |
| Vulnerat ility l How Impact on Mitigative Failure Evaluaf.fon Detected Safe Shutdown Feature Feature Mode CGT and three None, increased Two Independent Loss of off-site No voltage on Loss of all safety division EDCs. number of on-Preferred Power power. bus.
| |
| site and off-power sources.
| |
| Sources site power sources.
| |
| O 1 None, increased j ECCS functions Loss of ECCS Self testing ca- h sj Multi-plex Con- Loss of control pability, high ECCS reliability g l trol System power to ECCS not completed on function.
| |
| and elimination r*
| |
| ss reliability.
| |
| demand or test.
| |
| [5 Sensor Inter- of cable p faces spreading room. M O
| |
| n o
| |
| Loss of safety Red undant heat None, closed g Closed Loop Re- Heat exchanger High temperature loop RCU M on RB equipment, equipment (e.g., exchanger can actor Building tuba failure.
| |
| EDCs, RHR heat supply necessary supplies cleaner h Cooling Water water accumula- water to sefety exchangers). cooling System (RCW) tion in RCW room equipment en-alarms in hancing cooling control room, capability (i.e., reduced fouling of heat transfer surfaces) as compared to direct cooling with service water.
| |
| I ,
| |
| l - - - -
| |
| g q ,j
| |
| | |
| Shutdown Potential Preventive /
| |
| Failure How Impact on Mitigative. Vulnerability Feature Mode Detected Safe Shetdown Feature Evaluation RCW Isolation High Temperature See Heat Three divisions See Heat Valve' failed on RB equipment. Exchanger of RCW Exchanger closed failure, failure.
| |
| RCW pump fails High temperature See Heat Redundant pump See Heat to supply water. on.RB equipment. Exchanger can supply Exchanger Failure necessary flow Failure n
| |
| E t'1 5
| |
| Other High None, nitrogen :#
| |
| High Pressure Cas leak. Loss of pressure Loss of ADS /SRV l pressure DHR Nitrogen Cas in accumulators. capability to supply instead 3 y
| |
| Supply to ADS reduce RPV means exist of air reduces
| |
| ] and SRVs pressure and (e . g. , HPCF, potential corro- g g allow use of low- condensate, sion of valves y pressure for RCIC). Can and loss of n Heat Removal reduce RPV system pressure g (DHR) Systems. pressure through due to compres-use of RCIC. sor failures.
| |
| More reliable h"t than air systems.
| |
| tle isolated Surve'.11ance See gas leak See gas leak See gas leak to valve test closure (operator error) i
| |
| | |
| Fctential Freventive/
| |
| Shutdown Mitigative vulner-bilir.y How Impact on Failure Feature Evaluation Detected Safe Shutdown Feature Mode Three SRV None, added fea-Safety equipment Loss of ability Enhanced Remote Trans fer controls exist, tures enhance fails to actuate to control shutdown safety.
| |
| Shutdown Panel switches fail to on demand or fourth SRV and local control of (e.g., 4 SRVs, actuate fourth HPCF iros the equipment is -
| |
| SRV and HPCF. during test.
| |
| HPCF) remote shutdown possible.
| |
| panel.
| |
| o E
| |
| es Loss of warning Lor.a1 alarms None, no safety p Fails to detect During test. function just r-Containment At- to operator on will actuate. m mosphere Moni- high radiation breach of monitoring.
| |
| torin5 System or t*wperature k in containment.
| |
| radiological ;j (CAMS) barrier. "
| |
| o as O x
| |
| Loss c f some 16 T/Cs remsin None, enhance- g Enhanced Sup- Fails to detect During operation to monitor tem- ment to supprea- q or test. eenudancy in sion pool moni-pression ?ool correct pool perature.
| |
| pool temperature toring function.
| |
| Temperature temperature. sonitoring.
| |
| Monitoring (48 T/Cs Instead of 16)
| |
| Local None, does not During operation Loss of pool Suppression Fails ta detect perform a safety correc. pool or test. water level Indication.
| |
| function.
| |
| Pool Level monitoring capa-Monitoring level.
| |
| bility.
| |
| i p
| |
| ~
| |
| ~
| |
| ^ m _-
| |
| | |
| i i
| |
| 19Q 10 PRocRREEgg The ADWR has been designed to minimize risk associated with plant opera- l tions both at normal power and shutdown conditions. As previously mentioned, PRA techniques have been employed to identify potential accident scenarios and, ;
| |
| where appropriate, design modifications have been included to rejuce estimated l risks. .In addition to the physical plant design and configuration, the ADWR will incorporate operating procedures that are based on rigorous engineering ;
| |
| evtluations including safety analyses. These procedures will be prepared consistent with NUMARC Guidelines presented in NUMARC 91-06, *cuidelines *.o-enhance safety during shutdown".
| |
| Each utility must generate plant specific operating procedures based on individual site characteristics and training program requirements. A procedures guideline _will-be completed for the ABWR to address shutdown conditions. The guideline will provide insight into two general areast
| |
| : 1) Effective outage planning and control.
| |
| : 2) Haintenance of key shutdown safety functions: Decay heat removal ca- ;
| |
| pability, inventory control, electrical power availability, reactiv-ity control, and-containment integrity (primary and secondary).
| |
| Outace Plannino and ConII21 Alti.ough design' features help,_ shutdown risk can best be minimited through appropriate outage planning and control procedures. Planning is important
| |
| {
| |
| l because'of the large number and diversity of tasks that must be completed during the' outage. Safety and support systems must be taken out of service for maintenance. This reduces redundancy of safety _ systems. If alternate means are r
| |
| not utilized to-back-up the-lost safety system, a reduction in safety margin may occur. The ABWR contains multiple normal and alternate systems to complete all required shutdown safety functions. Availability of normal and alternate systems must be made known to all personnel involved in planning and execution o
| |
| ~78-
| |
| | |
| personnel involved in planning and execution of the outage. This is an ,
| |
| ever-changing situation during outages and proper planning and tracking of activities is required to ensure safety .nargins are maintained.
| |
| The plant speelfic procedures for outage planning and control should ensure that the appropriate focus is maintained on the following activities:
| |
| : 1) Documentation of outage philosophy including organirations respon-sible for. outage scheduling. This should address not just the ini- -
| |
| tial outage plan but all safety significant changes to the schedule.
| |
| : 2) Ensuring that all activities, particularly higher risk evolutions, receive adequate resources. The plan should consider scope growth and unanticipated changes.
| |
| : 3) Ensuring that the " defense in depth" concept that is central to power a
| |
| operation be maintaired during shutdown to ensure that safety margins are not reduced. Safety systems must be taken out of service for maintenance but alternate or back-up systems can be made available if proper planning is completed.
| |
| '4) Ensuring that all personnel involved in outage planning and execution receive adequate training. This should include operator simulator training to the extent practicablo. Other plar.t personnel, including temporary personnel, should receive training commensurate with the ontage tasks they will be performing.
| |
| : 5) After completion of outage planning, but prior to final approval, a review of the scheduleTshould be completed by an independent safety
| |
| -review team. The main objective of this review is to assure that the defense in depth principal will not be violated at any time during the-outage.
| |
| | |
| l l
| |
| I i
| |
| l I
| |
| ghgtdown SafeAy 1seues J Procedures for outage planning and control address ger. oral aspects of risk )
| |
| reduction during shutdown. Specific shutdown procedures are-required to maintain key safety functions during shutdown. The following guidelines should
| |
| - e used for each key shutdown safety function.
| |
| : 1) Decay Heat Removal Capability - The normal method of Decay Heat Removal (DHR) is through use of the Residual Heat Removal syctem !
| |
| l (RHR) in the shutdown cooling mode. As discussed in section 190 7 and 19Q.11, there have been many events at operating plants that have resulted in partial or total loss of DHR. A recovery strategy should be established to address loss of normal RHR. This should include t
| |
| identification of alternate DHR systems as well as personnel =
| |
| responsible for execution of the recovery plan. In addition to recovery plans, outage planning yhould emphasize availability of DHP by postponing maintenance on RHR systems to later in the outage when decay heat loads have been reduced or to when ',he core has been off-loaded to the spent fuel pool. In the case of core off-load, procedures-should be prepared to ensure maintenance of spent fuel ,
| |
| . pool cooling.
| |
| : 2) Inventory Control - If DHR were to be lost, the time to reactor cool-ant boiling and core uncovery will be determined by the initial cool-ant inventory and make-up capability. Procedures should be prepared to ensure that adequate coolant _ inventory is maintained at all times i during shutdown. Also,' plant activities or configurations where a single failure can result in loss of inventory should be identified and compensatcry measures established. Specific activities for the
| |
| -ABWR that should be reviewed for the potential of inventory deduction are: Use of freeze seals (see Section 19Q.8 for a more complete
| |
| ' discussion); removal of control rods, control rod drives, and reactor internal pumps; RRR valve actuations or leakage leading to diversion
| |
| -of RPV coolant to the suppression pool (e.g.,
| |
| l - = - - ._. _-._.x-._-_---.-.-.---,-_.--- . - . . -
| |
| | |
| i RHR pump mini-flow valve failure / leakage, awitching shutdown cooling from one division to another); and inadvertent actuation of safety relief valves. ,
| |
| : 3) Electrical Power Availability - As discussed in Sections 19Q.4.4 and ,
| |
| 19Q.11, loss of electrical power during shutdown has resulted in loss of DHR in the past. The ABWR has two sources of off-site (preferred) and four sources of on-site electrical power. Procedures should be J utilized to ensure that defense in depti, for electrical power sources is maintained. Maintenance of power sources should reflect the current plant conditions. Availability of normal and alternate power sources should be ensured especially during periods of higher risk evolutions (e.g., RPV drain down following refueling). Many of the loss of power events discussed in Section 190 11 were caused by ,
| |
| operator errors (e.g., switching orrors, inadequate mainte-nance / testing procedures) and grounding of transformers in switchyards due to movement of equipment by cranes and trucks. All maintenance and switchyard activities should be reviewed to identify single- f ailures or procedural errors that could result in loss of power to-vital buses during shutdown. Procedures should be developed for implementation of alternate sources of power including applicable ,
| |
| breakers and bus locations, required tools, and sequence of steps to-be performed.
| |
| : 4) Reactivity Control - Shutdown reactivity control for the ABWR is maintained by core design analysis .snd interlocks that restrict fuel and control rod drive movements. Procedures are required to ensure that the core is loaded per design requirements and that unautherized fuel movement does not occur simultaneous with CRD mechanism maintenance. If the refueling sequence must be altered, new shutdown margin analyses-ehould be performed. All fuel movements should be verified by knowledgeable trained personnel.
| |
| : 5) Containment It'egrity - The ABWR primary containment wLil not be navailable during most of the refueling outage but procedures chould
| |
| | |
| 1 be developed to ensure its availability during Mode 3 and during Mode 4 (if appropriate). During all modes, procedures should be available to ensure that secondary containment can be maintained functional as required, sepecially during higher risk evolutions.
| |
| Procedure _ Reviews An important part of procedures implementation is a review of the adequacy of all operating procedures. All shutdown operating procedures should be reviewed periodically to ensure.that_the defense in depth concept is being maintained given the actual events occurring at each site. This review should include not only procedure adequacy but dissemination of the outage philosophy to all personnel involved in scheduling and executing the outage plan and
| |
| -training of personnel including temporary personnel. This review should be documented and retained as a permanent pl?.nt record.
| |
| 190 11
| |
| | |
| ==SUMMARY==
| |
| OF REVIEW OF SIGNIFICANT SHUTDOWN EVENTS:
| |
| ELFCTRICAL POWEP AND DECAY HEAT REMOVAL As part of thezcertification process for the ABWR design, the NRC has requested that General Electric complete a review of significant shutdown events in operating plants and discuss ABWR features which could prevent ar mitigate such events-To complete this evaluation, a review was made of operating events involving loss of off-site power (LOOP) and loss of Decay Heat Removal (DHR).
| |
| These two areas appear to have the greatest potential for causing core damage during shutdown based on past experience. The sources utilized for infctration on past-shutdown events were:-
| |
| ~
| |
| - " Residual Heat Removal Experience Review and Safety Analysis",
| |
| NSAC-88, March 1986 l
| |
| . ~ . _ - - . _ - _ _ _ - _ _ _ . . _ _
| |
| | |
| I
| |
| - " Loss of Vital AC Power and the Residual lleat Removal System during Mid-Loop Operations at Vogtle Unit 1 on March 20, 1990*,
| |
| NUREG - 1410, June 1990
| |
| - "NRC Staff Evaluation of Shutdown and Low Power Operation",
| |
| NUREG - 1449, March 1992
| |
| - Selected INPO SEO Reports and NRC Information Notices.
| |
| The results of this evaluation are contained in Attachment 1900, Tables 19QC-1 and 19QC-2 for LOOP and loss of Di(R respectively. The following is a discussion of the results for each event type.
| |
| 16292 NUREG - 1410 contains a discussion of 70 LOOP events at operating plants both PWR and BWR. Although the response to LOOP events will dif f er f or PWRs and BWRs, the initiating events are similar in that off-site and on-site power configurations are similar for both reactor types. The events evalusted in NUREG - 1410 occurred between 1965 and 1989. Two additional recent LOOP events were added to this list and are included in Table 190C-1. .
| |
| The LOOP events can be grouped into the following categories:
| |
| - Loss of all off-site power sources due to various reasons including weather, operator errors or grid upset Loss of one or more off-site sources with at least one off-site source remaining
| |
| - Isolation of off-site power due to on-site electrical faulte
| |
| - Degraded of f-site or on-site power sources resulting f rom errors in maintenance activities
| |
| | |
| As discussed in Table 190C-1, the ABWR electrical distribution system has several features which would prevent or mitigste every precursor event evaluated in this study. Prevent or mitigate in this case meer4s that at least l one class 1E power supply would be available to energire equit eent to maintain ;
| |
| plant cold shutdown.
| |
| L
| |
| 'A one line diagram of the ABWR electrical system is contained in Attachment 190C. The main features of the electrical system eres
| |
| - Two independent sources of off-site power
| |
| - Three physically and electrically independent Class IE emergency I diesel generators ;
| |
| - Three unit auxiliary transformers powering three Class IE and a three non-1E power buses
| |
| - Combustion Turbine Generator (CTG) that can be used to power any of the class IE or non-1E power buses The above features of the ABWR electrical distribution system, along with appropriate Technical Specifications and other administrative controle, assures that adequate power sources would be available to mitigate potential electrical events such as those described in Table 190C-1.
| |
| koos of DHR NSAC - 80 contains a discussion of 90 loss or degradation of DHR events during the seven year period 1977 through the end of 1983. The source for these events were_ Licensee Event Reports-(LERs). Other-events descrioed in INPO SEO-reports and NRC information notices were also reviewed and included in the !
| |
| atudy.
| |
| _. . - _ - - . . . . . _ _ . ~ . _ _ _ . _ . _ , _ _ . _ _ . . . _ . - . - . _ _ _ . - . - , . . . _ . . . _ .
| |
| | |
| _._.._.._.___.___m.._____.__.- ---.. _ _ __ - . _ _ _ _ _ _ _ _ _
| |
| i i
| |
| i 1
| |
| p,gmms rv_.o f ' DHAlv.tritt The results of this evaluation are contained in Attachment 1900 Table i 190C-2. Not all of the events discussed in NSAC - 88 are contained in Table 190C-2. Those events that were due to random failures of single components and did not result in loco of DHR or other significant plant effects were not evaluated further. If the single failure resulted in loss of coolant, over-pressurization, flooding, or loss of Shutdown Cooling (SDC) function, the event was included and the applicable ABWR feature to prevent or mitigate the event was discussed.
| |
| For the purposes of this study, prevention or mitigation means that, given the DHR challenge event, the ABWR design would either not be susceptible to the postulated failure or it has design f eatures that could be relied upon to ensure that the fuel in the RPV remained covered with water at all times.
| |
| of the events described in Table 190C-2, some were single failures of RHR system components that resulted'in either delayed achievement of Shutdown dooling (SDC), reduction in Reactor Pressure Vessel (RPV) water level, or a temporary loss of SDC. In all of these events, the fuel remained covered with ,
| |
| water and alternate means of DHR remained available (e.g., reactor water cleanup oystem, main condenser, and ECCS systems). In the cases of delayed or temporary loss of SDC, RPV water temperature increases ranged from 10 - 140 deg.F. In all cases, SDC was restored and alternate means of DHR were not used although available. Operator errors associated with improper valve lineups or incorrect maintenance were identified. In these cases, delays in implementing SDC or temporary loss of SDC occurred while the error was corrected. In a few caces, marine growth. caused failure of one or more RHR heat exchangers which resulted in temporary loss of SDC while other RHR loops or alternate cooling paths were
| |
| = implemented. . In-one-case,-a freece seal-failure in the RHRSW caused 15,000--
| |
| gallons of water to damage ECCS power supplies resulting in temporary isolation of-SDC.
| |
| | |
| None of the events described above and in tablo 1900-2 reaulted in fuel
| |
| , being uhcovered. The flexibility of the RHR system and the several alternate
| |
| ' means of DHR that were available served to mitigate the component failures or operator errors.
| |
| i Summary Significant shutdown events in operating plants have been reviewed to '
| |
| determine ABWR features which could prevont or mitigate the events. Loss of offsite power and loss or degradation events from published nuclear industry reports were the database for this review. The results of this review demonstrate that ADWR design includes many features that prevent or mitigate unacceptable consequences of typical past events.
| |
| The main features of the ABWR that will prevent or mitigate shutdown events ares
| |
| - Three divisions of ECCS and support systems tha*. are physically and electrically-independent '.
| |
| Two independent off-tite power sources
| |
| - Four on-site power sources (three emergency' diesel _ generators and one combustion turbine generator)
| |
| - Plant configuration and structural integrity to minimize common m mode failures due to fire and-floods l
| |
| - Appropriate Technical Specifications and_other administrative controls ~to ensure-availability of systems during periods of ,
| |
| potentially high risk operations j
| |
| { '
| |
| - Several alternate means of DHR if normal systema were to fail or be out of service for maintenance r
| |
| - . ,~._ _ _... _.._.;_ _.__. _ _ _ _ _ . _ _ . _ _ _ . _ _ . . . _ . _ _ _ _ _ . _ . _._ ..,_ _,_.,_,,___,_.-
| |
| | |
| - Instrumentation availability .iuring shutdewn to monitor plant safety status and initiate safety oystems when needed 19Q.12 BJMLLIS AND IFTNACE PJOUIRLMJJ[Tf!
| |
| 19Q.12.1 Janights Gained f rom the Analysis Completion of the ABWR shutdown risk analysis has resulted in the following insights:
| |
| : 1) The most important element in control of shutdown risk is adequate planning of maintenance on systems and support systems that can be used to remove decay heat or supply inventory makeup to the RPV.
| |
| : 2) The ADWR design has incorporated a significant number cf new design features relative to operating BWRs. Past events that have led to loss of decay heat removal capability or lose of offaite power can, in general, be mitigated by ABWR design features.
| |
| : 3) The ABWR design hac a very low risk associated with loss of decay heat removal. Adequate shutdown safety margine exist if only systems required by Technical specificationa and these that are already in operation (e.g., CRD, FPC, fire water) are relied upon. Minimum combinations of systems have been identified that, if availabio, will ensure adequate shutdown aatety margins.
| |
| Combinations other than those identified in this study may Sxist which also result in auequate shutdown risk margins. By taking advantage of these available decay heat removal and makeup systems, utilities can exercise much flexibility in outage maintenance scheduling while ensuring that adequate safety margins are maintained at all times during shutdown conditions.
| |
| | |
| . m m-_..___ _ . _ - , - - . . - - . - _. _ ---
| |
| : 4) The above safety margins were calculated using very conservative estimates for human error probabilities. For all ovents analyzed during shutdown, sufficient time is available to prevent core damage that no extraordinary operator actions are required. ABWR safety is designed into the plant.
| |
| 19Q.12.2 Important Desian Features finput to ITAAC)
| |
| The following ABWR features have been identified ad important contributors to the low level of risk associated with shutdown and have been includad as part of ITAAC:
| |
| - Lines attached to the RPV have isolation valves that close on a low RPV water level.
| |
| t Shutdown cooling piping connects to a nozzle in the RPV at an elevation that is above the top of the active fuel.
| |
| - The RHR system has mode switches to automatically realign valves in the RHR system.
| |
| - Two offsite and four onsite sources of electric power that are independent and physically separated.
| |
| - - Three divisions of ECCS and support systems that are independent and physically separated.
| |
| . Watertight doors on ECCS and RCW rooms.
| |
| - Floor drains in all floors above the first floors in control and .
| |
| reactor buildings.
| |
| t
| |
| - The RHR in the shutdown cooling mode does not isolate on loss of logic power.
| |
| AC independent water addition system.
| |
| l e~.-..- _ - . . - . . - . . - - . - - - . - . - .
| |
| | |
| __ _ -_ _. _ _.._. _ _ _ _ _ . - _ . _ ...__.. - .._ ,... - _ . _ .~. _ _ . _ _ _ _ ._._.
| |
| e i
| |
| - CRD purrp can be used for makeup under all shu*down conditions. l I
| |
| 19Q.12.3 Qperator Actione fi1p.u,t to C_Ql, hetion Items)
| |
| The following operator actions have been identified that are important :
| |
| to minimization of shutdown risk and have been included as COL acticn items: ,
| |
| - Ability to recognize failure of an operating RHR system.
| |
| - Rapid implementation of standby RHR systems following loss of ,
| |
| operating RHR system.
| |
| - Use of alternate means of decay heat removal using non safety grade equipment such at CUW, FPC, or main condenser. .
| |
| - .Use of alternate means of inventory makeup using non safety grado .
| |
| equipment such as AC independent water addition, CRD pump, feedwater, or condensate.
| |
| - How to utilize toiling for decay heat removal in Mode 5 with the RPV head removed including available makeup sources.
| |
| Implementation of fire / flood watches during periods of degraded safety division physical integrity.
| |
| - Fire fighting during shutdown.
| |
| - Use of remote-shutdown panei during shutdown.
| |
| (
| |
| ~ Instrumentation must be made.available during shutdown to support i the following functions:
| |
| --Isolation of RPV i .- ADS
| |
| -- HPCF
| |
| | |
| c , _ _ _ _ .- . . _ _ . _ . . . _ . . _ ... _._ _ . - - _ .._ . . _ _ . _ _ _ _ -. _ . _ . - . ..
| |
| 4 l
| |
| LPFL
| |
| - RPV water level, pressure, and temperature
| |
| - RHR system alarms
| |
| - EDG
| |
| - Refueling interlocks
| |
| - Flood detection and associated valve isolation and pump trips
| |
| - Procedures should be prepared to address the following tasks ,
| |
| during tasks during shutdowns
| |
| - Fire fighting with part of the-fire protection system in maintenance
| |
| - outage planning to minimize risk using guidance from NUMARC 91-016
| |
| - Use of freeze seals
| |
| - Replacement of RIPS and CRD blades
| |
| - Loss of offsite power
| |
| - Increasing CRD pump flow when using for inventory control
| |
| - Maintenance of suppression pool as it relates to maintaining safety margins for decay heat removal
| |
| - Ensure that one safety division is always available with intact f. ire / flood' barriers.
| |
| l.
| |
| ~ ^ + , a e -v w ., v- r, w ,m,,yv.gm- ,e-.g- .,e--n, n,y- , y w,. ,-v. .e w. rg- e, yy .,,---p, ,e,r-e.--w-ey , . , . - -,g s v--m r .,y e- v v~.n r--wr-rs -y
| |
| | |
| 1 i
| |
| 19.Q.12.4 -Be11 ability G_ofia (input to RAP)
| |
| The following assumed system unavailabilities were determined to be important in minimizing shutdown risk and are included 4.n the ABWR Reliability Assurance Programs System Unavailability RHR (SDC) 0.1 per reactor year RHR (LPFL) 1.0E-3 per reactor year HPCF 1.0E-3 per reactor year CRD 1.0E-3 per roactor year CTG 0.05/ demand EDG 0.07/ demand offsite Power 4.0E-3 per reactor year ADS 1.0E-4 per reactor year 19.Q.12.5 conclusions The ABWR has been evaluated for risks associated with shutdown conditions and for all postulated events the risk han been determined to be low.
| |
| Multiple means of removing decay heat and supplying inventory makeup have been identified that along with appropriate Technical Specifications and outage procedures result in acceptably low shutdown risk levels for the ABWR.
| |
| ._ . . , _ _ . . . . . _ . - - . _ .-.__,...._.u._. _ _ _ _ . . _ _ - . . . - . _ , _ - _ _ _ . - . . _ - _ . -_-
| |
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| L ATTACHHDTT QA DECAY HEAT RELIABILITY FAULT TREES 1
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| ll t
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| n+ . ,n- ,
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| l MAN CONDENSER UNAVALABLE s 1 l
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| I I I _ FEECWATER MO l
| |
| CIRCLAATNG WATER MAM COPOENGER OPEIMTOR FAILS TO LOSS OF MAN CONCEN%TE FALURE PUMPS FAL UNAVAILABLE DLE TO NITIATE MANUALLY COPOENSER YACtAJM MANTENANCE l*w u ca;ij
| |
| >#cvAc [ ucev*APs l l mm MN]
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| MCOP r~-9 7 f 100E41 2.ooE42 100E43 T ,
| |
| PUMP B FALO TO PUMP C FALS TO l PUMP A FALS START MC RUN START AND RLM l uceus*>e l luc +teeet t IuceouP41
| |
| %a ms I I PL%8P A FAILS TO RUN GROUP 4 POWER FALS ucAr4 [ P t tv M j 194E44 Paga 2 I
| |
| I 3 1 . s i 1 2 1 Page i I
| |
| .\SDTREEiMCU.CAF 6-30-92 MAIN CONDENSER SYSTEM
| |
| | |
| GROUP I PONER FALS
| |
| . l P sNcM I b
| |
| I I FAILURE OF GACUP A FAILURE OF GPOUP I 125 V DC BUS S 9 kV OUS
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| -T i
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| l EOc s tN I l EAcwecu l l
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| O 7.20E-06 [d_
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| 1 1 LOSS OF CFF-SITE FALURE OF Otv A POWER HECW TO PROVM ECCS CCOLNG l 1 EAct OSC I IM CW Afdl 1.50E43 i
| |
| orvisx:n a sem FALURE lEAc w >*al 5 7sE-os I i i 2 I MAIN CONDENSER SYSTEM .\SDTREE\MCU.CAF 6-30-92 Page 2
| |
| | |
| eUuP e Fu.s to srAar ma aus
| |
| ,, ou~m I I I
| |
| 1 PUntP LEG CNCX WJP W N*4 PUMP B FAR.S TO PUMP B FAE.S TO ALfN FALS VALVE FAR.S TO v%*
| |
| START
| |
| * l
| |
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| |
| raw %4 2 I 4 i i 2 1 1 i
| |
| .\SDTREELMCU.CAF l 6-30-92 Page 3 l MAIN CONDENSER SYSTEM
| |
| | |
| FAIL 1ME OF GROUP N 6D W BUS I 'xere I n !
| |
| LOSS OF OFF-StrE FALURE OF Dtv 8 POWER NECW TO PROVOE ECCS COCUNG I EActoso i Iw+ Eceul 6 .a.
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| __ . S_
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| FALURE IEAceauwNI 5 76E4)6 I $ 1 2 i MAIN CONDENSER SYSTEM .\SDTREE\MCU.CAF 6-30-92 Page 4
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| | |
| i l
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| PUMP C FALS TO START AND RUN lucPUMPC]
| |
| ~
| |
| I l I I e Pt.MP LEG CHECM GAOUP HI FCWER PUMP C FA.LS TO PUMP C FALS TO RUN VALVE FAILS TO OPEN FALS START T w,CD 1 P3m;.u l uCCS utkf4 227E M 194E M 1 doe m [
| |
| [-
| |
| i i
| |
| F ALURE OF GACL'P C FALUFE OF GW 125 Y DC BUS Mt 6.9 }V EUS I Eto w I I e ao+ v i 72cE a Pwp e e I 2 I . I i , I
| |
| .\SDTREE\MCU.CAF 6-30-92 Page 5 MAIN CONDENSER SYSTEM
| |
| | |
| c 1
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| I l
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| FALURE OF GROUP m 4 9 kV BUS WAC6%Mi 5
| |
| I I LOSS OF OFF-S!TE FALUAE OF Div C POWER HECW TO N_
| |
| ECCS COOUNG i
| |
| IEAClosEI IwHFCwct.1 O O ~~
| |
| Drvisscn a swGa FALURE I W.** m l O ~~
| |
| I 4 1 2 I I MAIN CONDENSER SYSTEM .\SDTREE\MCU.CAF 6-30-92 Page 6
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| | |
| e FEEDWATER AND CONDENSATE FALURE IFw wil h*4 I I FEEDWATER FALURE CONDENSATE FALL 9'E
| |
| *wo I cteon I
| |
| ~ ~~
| |
| l I I I j FEEDWATER FUMPS POWER FOR THE CCiNDENSATE PUUPS THE CONDCtSER FAL TO PROVOE FLOW PUECTION VALVE $ FAL TO PROVOE FLOW HOTWE!1 LEVEL FAIL INSUFFCENT
| |
| . pel isvawe a1 Ict+tel 4cmu1 A A A O ~~
| |
| PCWER FCH THE nmCToN ve_S Fall icvawtal I i i 2 I s I . I MAIN CONDENSER SYSTEM .\SDTREE\MCU.CAF 6-30-92 Page 7
| |
| | |
| Gate / Event Name Pace Zone Gate / Event Name Pace Zone Gate / Event Name Pace Zone Gate / Event Name Pama Zcne CE"LI 7 4 MCU 1 3 CCND1 7 4 MCVAC 1 2 COPUMPS 7 3 PlNCM 1 3 CVPOWER 7 4 PlNCH 2 2 EAC69Cmi 2 2 P2NCH 3 4 EAC69CM 2 2 P3NCM 5 4 EAC69Dmi 4 2 WHECWAN 2 3 EAC69DM 3 4 WHECWBN 4 2
| |
| ??C69DM 4 2 WHECWCN 6 2 ESC 59Emi 6 2 EAC692M S 4 i EAC69EM 6 2 EACLOSC 2 2 EACLOSD 4 1 EACLOSE 6 1 EDC11N 2 1
| |
| {
| |
| EDC12N 3 4 i
| |
| EDC13N 5 4
| |
| . FVPOWER 7 2 FWCOND1 1 5 FWCOND1 7 2 FWF1 7 2 FWPUMPS 7 1 i MCAR 1 2 MCBD 3 3 MCBR 3 2 i 1 MCBS 3
| |
| . MCCD 5 3 MCCR 5 2 MCCS 5 1
| |
| ~
| |
| NCMAIITT 1 4
| |
| ; MCOP 1 1 MCPUMPA 1 2 MCPUMPB 1 3 MCPLHPB 3 2
| |
| , MCPUMPC 1 4 i' MCPUMPC 5 2 i MCFUMPS 1 3 l
| |
| MAIN CONDENSER SYSTEM .\SDTREE\MCU.CAF 6-30-92 Page 8
| |
| | |
| i 1
| |
| FEEDWATER FALURE FWF 1 i e i i OPERATOR FALf1FE TO j FEEDWATER FEEDWATER PUMP 3 POWEM FOR THE PKTMTE MANUAU.Y UNAVAEJBLE DUE TO FAL TO PROYt0E FLOW PUECTON VALVES MANTENANCE FAL FWOP l FW%tAhl j l f'AM$FS l [GVG" M Hj 200E42 tooE41 I ' Pop 6 T a I
| |
| PUMP A FALS TO PUMP C FALS TO START AND RUN START AND MUN 15an w Al IF Wrte l P*P 2 Page 4 PUMP B FALS TO START AND RUN I 15 wm.w? I Pop 3 i
| |
| , ! , i 2 1 2 1 4 I FEEDWATER SYSTEM (SDC) .\SDTREE\FWF.CAF 6-29-92 Page 1
| |
| | |
| PUMP A FALS TO STAPT AND PLtd IF W #' A l
| |
| ' 1 3 i . .
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| FUMP A LEG CHECK GROUP f PCVEM FALS PUMP A FALS 'Jf- PUMP A FAL3 TO RUN ;
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| stanT Ar n a sTanT vAtvt Fats To ceEn I
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| . . Ao 1 94E-04 140E&
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| { F F.wo j
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| 3 1 4 I I 1 i 1
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| .\SDTREEiFWF.CAF 6-29-92 Page 2 FEEDWATER SYSTEM (SDC)
| |
| | |
| 6 PUM* B FALS TO START AND RUN Ir etus a1 I '
| |
| I i i i PUMP B FALS TO PUMP B FALS TO %W PWP G 1EG CHECK GROUP R PCW4R FALS '
| |
| START AFTER STIJTT VALVE FARS TO OPEN FWMJ FWBN l FWvivB { l PZNCO l 2 rem 194E& [ T 140EM \ !
| |
| U -J l i .
| |
| FAILURE OF GAOUP B t.OSS OF 69 KV 125 V DC BUS GROUP tt SWGA s&C CD POWTR T
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| l EC=;12*47 IEas.54 ol '
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| FEEDWATER SYSTEM (SDC) .\SDTREE\FWF.CAF 6-29-92 Page 3 l L
| |
| -, ,- --c- -
| |
| | |
| i PLMP C FAILS TO START AND RUN a
| |
| I FW5't*8PC 1 i
| |
| I I PUMP C FALS TO PL%F C LEG CHECK START VALVE FAILS TO OPEN i
| |
| Fwco I Fwvtvc 1 2rE c3 140E44 t I PLMP C FAILS TO P!UN Gur til POWER AFTER START FALS i-uwca i exo i ,.
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| l . e+ %s I
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| .: I ,
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| ) FEEDWATER SYSTEM (SDC) l .iSDTREE\FWF.CAF 6-29-92 Page 4
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| ll il l l 1 I1 5
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| |
| , i GROUP t POWER FALS GROLP 11 POWER T FALS T
| |
| rien b
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| I i 1 2 I FEEDWATER SYSTEM (SDC) .\SDTREE\FWF.CAF 6-29-92 .Page 6
| |
| | |
| Pace Long Gate / Event NaFC ES2e Zone Cate/ Event Name Pace Zone Gate / Event NamL Pace Zone Gate / Event Na:ne EAC69CD 2 4 i
| |
| EAC69DD 3 4 EAC69ED 5 2 EDC11N 2 4 j
| |
| EDC12N 3 4 EDCl3N 5 1 FVPOWER 1 4
| |
| _ ,1 FVPOJER 6 2 l
| |
| FWAD 2 1 FWAR 2 2 DIBD 3 1 FWBR 3 2 FWCD 4 1 FWCR 4 1
| |
| &lF 1 2 FWMAINT 1 2 FWOP 1 1 FWPUMPA 1 2 FWPUMPA 2 2 FWPUMPB 1 3 FWPUMPB 3 2 FWPUMPC 1 3 FWPUMPC 4 2 FWPUMPS 1 3 FWVLVA 2 3 FWVLVB 3 3 FWVLVC 4 2 PlN 6 1 PlNCD 2 4 P2N 6 2 P2NCD 3 4
| |
| ! P3NCD 4 2 P3NCD 5 2 6-29-92 Page 7 FEEDWATER SYSTEM (SDC) l .\SDTREE\FWF.CAF i ,
| |
| | |
| FEEDWATER AND CONDENSATE FALURE l FwCONol I I FEEDWATER FALURE CONDENeATE FALURE l CONF l b
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| I i I 2 1 FEEDWATER OR CONDENSATE PUMP (SDC) .\SDTREE\FWCOND.CAF 6-30-92 Page 1
| |
| | |
| Gate / Event Name Page Zone Gate / Event flame Pace Zone Gate / Event Name Page Znna' Gate / Event Name Eagg Zone CONF 1 2-FWCOND -1 2 FWF 1 1 I
| |
| 1 l
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| l FEEDWATER OR CONDENSATE PUMP (SDC) .\SDTREE\FWCOND.CAF 6-30-92 Page 2 1
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| | |
| l FPC FA! LURE l I I I FPC VALVES F All FPC PUMPS Fall TO FPC HEAT EXCHANGER POWER FALS COOLWO UNAVAILABLE PROVfDE FLOW I F a 24x I I Fu.vt vs l IHammi Page 6 Page 4 3
| |
| I I I STANDOY PUUP FAILS RUNNNG PUMP F AILS GROUP I POWER FAftS GROUP 11 POWER T FALS T
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| l 6 Wa.svg l l + e+mys l vin em Page 2 Page 3 s I 4 I s i i 2 I I i
| |
| .\SDTREE\FPC.CAF ' 6-29-92 Page 1 FUEL POOL COOLING (SDC) 4
| |
| | |
| i STAND 8Y PUMP FAR.S I Feseuue_J I I STRAINER D004B PLUGS PUMP C0019 H MANTENANCE l FM1W j I Fetxx248 l 3 40E45 1.00E41 i
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| i PUMP C001B FALS TO POMP C001B FAILS START DURHG MISSION I Fe00100.] l FP9016H I 227E-03 194E44 i
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| |
| 140E44 2 00E42 2
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| 1 2 i I i t
| |
| .\SDTREE\FPC.CAF 6-29-92 Page 2 FUEL POOL COOLING (SDC)
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| 1.? ?E-04 I i ! 2 i FUEL POOL COOLING-(SDC) .\SDTREE\FPC.CAF 6-29-92 Page 4
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| FAILLiE OF MX B00tB
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| j6H2tue[
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| Page 4 l' I HEAT EXCHAfJGER B N HEAT EXCHAF*3ER 8 MAINTENANCE FAILS DUFt2#3 CFERATfCN l FPo0
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| : 1. DOE 41 1.14E N 1 3 CFERATOR FAttfRE TO FALURE OF DIMf0N
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| FMP lWHCWBl
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| , 2 coEQ I i 1 2 1 y FUEL POOL COOLING (SDC) .\SDTREE\FPC.CAF 6-29-92 Page 5
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| ,y _' : . , .
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| I i FPG VALVES FAL
| |
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| 1 2 i l i
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| .\SDTREE\FPC.CAF 6-29-92 Page 6 FUEL POOL COOLING (SDC)
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| | |
| GatelRvent Narne Pace ICng Gate / Event Name Eagg long Gatss/ Event Name Eac2 Zone Gate / Event Name Pace Zone FNV021A- 6 1-FMV021B 6- 2 FPOO.i 4
| |
| 1 FP001AR 3 1 FP001B 5 2 FP001BD 2- 1 FP001BHM 5 1 FP001BM 2 2 FP001BR 2 2 FP003BD 2 1 FPC 1 4 FPCHX 1 5 FPCHX 4 2 FPCHX1A 4 2 FPCHX1B 4 3 FPCHX1B 5 2 FPCVLVS 1 6 FPCVLVS 6 2 FPD004A 3 2 FPD004B 2 1 FPOP 2 2 FPOP 5 1 FPPUMPC 1 4 FPRPUMP 1 4 FPRPUMP 3 2 FPSPUMP 1 3 2
| |
| FPSPUMP 2 !
| |
| P12 1 2 PIN 1 1 P2N 1- 2 WRCWA 4 2 WRCWB 5 2 FUEL POOL COOLING (SDC) l .\SDTREE\FPC.CAF 6-29-92 Page 7
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| t.'
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| CRC FLOW LPMVAILABLE i e
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| _ _ _ _ P6 LAP FAIL 1JRES CST WATER LEVEL -
| |
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| |
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| |
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| |
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| |
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| ; : ONE CRD. (SDC): .\SDTREETCRDU.CAF '6-29 Page-1 i t
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| STANDCY PL9AP C0018 FAIUS TO START AND RUN I cHPtA4Pt$1 I- t OPERATOR FAILdRE TO GROUPtiPOWER INITIATE MANUAUY FAILS T
| |
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| PUMP DtSCHARGE FAILURE OF OfVieM CHECK VALVE F0058 C RCW TO PROVOE FALS TO OPEN COOUNG FLOW T
| |
| l caoose 1 Inacwcl O ""* A I i I 2 i ONE CRD (SDC) .\SDTREE\CRDU.CAF 6-29-92 Page,2
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| Gate / Event'Name' Pace Zone Gate / Event Name Pace Zgng Gate / Event iame f Eagg Zone Gate / Eve _1t Name' Pace LQue CR001AR 1 1 CR001BD 2 1 CR001BM 2- 1 CR001BR 2 2 CR002AP 1 1 CR002BP 2 2 CR005B 2 1 CRD001F 1 2 CRDCST 1 3 CRDOP 2 1 CRDU 1 2 CRFLOP 1 2 CRPUMPA 1 1 CRPUMPB 1 2 CRPUMPB 2 2 P1NC 1 2 P2NC 2 2 WRCHA 1 1 WRCWA 3 1 WRCWAC 3 2 WRCWC 2 2 WRCWC 3 2 ONE CRD (SDC) .\SDTREE\CRDU.CAF 6-29-92 Page.4 i
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| l CONDENSATE FAfLURE CONI-l l f E THE CONDENSER CONDENSATE PUMPS PCWER FOR THE OPERATCT4 FAILURE TO CONDENSATE HOTWELL LEVEL FAIL TO PROVIDE FLOW NJECTION VALVES INfTIATE MANUALLY UNAVAILABLE DUE TO INSUFFICIENT MAINTENANCE FAL IcveowEal w wti lccwoel I cuAuf l Icoeuvesl 6SOE-06 2.ooE-02 1 o0E41 Page 2 I I GROUP 1 POWER FALS GROUP tl POWER T FALS T
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| .\SDTREE\ CONF.CAF 6-29-92 Page 1 ONE CONDENSATE PUMP
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| CONDENSATE PUMPS FAL TO PROVIDE FLOW Iconwtes i 9 i i i PUMP A FALS TO PUMP B FALS TO PUMP C FALS TO PUMP O FALS TO START AND RUN START AND RUN STAPT AND RUN START AND BUN lCPUMPAl- lCPUMP8l CPtw $ lCPUMPOl Page 3 Parje 4 Page 5 i --
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| i PUMP # FALS TO l PUMP A LEG CHECK SfART VALVE FAILS TO OPEN COAD l COVLVA l 227Em 1.40E-04 i i PUMP A FALS TO RUN GROUP 1 POWER FALS AFTER START T CoAR l P1NCO l 15,4E-04 I i 1 2 I a i 4 i ONE CONDENSATE PUMP -.\SDTREE\ CONF.CAF 6-29-92 Page 2
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| PUMP 9 FALS TO STAaT um NUN p ,
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| IcPvuee1 PUMP B FALS TO PUMP B LEG CHECK START VALVE FALS TO OPEN C000 l CovtvB l 2 P-o3 1.40E44
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| ! i i PUMP B FALS TO RUN GROUP R POWER l AFTER START FALS T
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| l l
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| l cosa I emo 1 1.9dE-04 I i I e I _,
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| l ONE CONDENSATE PUMP .\SDTREE\ CONF.CAF 6-29-l ' - ; Page 3 l
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| PUMP C FALS TO START AND RUN l CPLkAPC I L4 I I PUMP C FALS TO PUMP C LEG CHECK START VALVE FAILS TO OPEN COCD l COvtVC ]
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| 227E-03 1.40E W PLMP C FALS TO RUN GROUP Ifl POWER AFTER START ' FALS T
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| coca l Pmo I 1.94E-04 I i l 2 i ONE CONDENSATE PUMP .\SDTREE\ CONF.CAF 6-29-92 Page 4
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| PUMP D fad.S TO START AND RUN l
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| Iceoveol i i PUMD D FALS TO PUMP D LEO CHECK START VALVE FAR.S TO OPEN cdOO I COVLvDJ 2.27E 43 1.40E44
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| , i PUMP D FAILS TO RUN GROUP m POWER AFTER START FALS T
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| cdoH IP3NCOj 1.94E44 1 2 1 _
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| .\SDTREE\ CONF.CAF 6-29-92 'Page 5 ONE CONDENSATE PUMP
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| Gate / Event Name Pace Zone Gate / Event'Name Pace Zone Gate / Event Name Pace long Gate / Event Ih m Page. Zone CHWLI 1 5 CMADJT 'l 2 COAD 2. 1 COAR 2 1 COBD 3 1 COBR 3 1 COCD 4 1 COCR 4 1 CODD 5 1 CODR 5 1 CONF 1 3 CONOP 1 1 COPUMPS 1 3 COPUMPS 2 3 COVLVA 2 2 COVLVB 3 2 COVLVC 4 2 COVLVD 5 2 CPUMPA 2 2 CPUMPB 2 3 CPUMFB 3 2 CPUMPC 2 4 CPUMPC 4 2 CPUMPD 2 4 CPUMPD 5 2 CVPOWER 1 4 PIN 1 4 PlNCD 2 2 P2N 1 4 P2NCD 3 2 P3NCD 4 2 P3NCD 5 2 ONE CONDENSATE PUMP .\SDTREE\ CONF.CAF 6-2'J-92 Page 6
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| CLEANUP WATER UNAVALABLE I RWCUF l I I OPERATOR FAR.URE TO RWCU PUMPS FAL TO INTTtATE MANUALLY PROVIDE FLOW IRWCUOPl lHveMP5l 2.00E-02 Page 3 I i RWCU UNAVAILABLE RWCU NCN-REG. HEAT DUE TO MANTENANCE EXCW2ER COOLING LNAVAILABLE l RWMAe4T l lRwNaM al 100E41 Page 5 i i POWER TO PUMPS FAL RWCU VALVES FAL i l P12N l l 9WVLVS l Page 2 Page 6 I i i 2 i CLEAN-UP WATER SYSTEM (SDC) .\SDTREE\RWCUF.CAF 6-29-92 Page 1
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| POWER TO PUUPS FAL Ps2n e i GROUP I POWER FALS GROUP n POWER FALS 1N P2N 1 I I I FALURE OF GAOLP A LOSS OF GROUP A FALURE OF GROUP B LOS3 OF GROLP B 125 V DC BUS 480V MCC POWER 125 V DC BUS 480V MCC POWER T
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| IEocuNI I EAdE1N l 1 ELCt2N I IEALEF4l 0 '"" A O '"" A I i l 2 I s 1 4 l CLEAN-UP WATER SYSTEM (SDC) .\SDTREE\RWCUF.CAF 6-29-92 Page 2
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| RWCU PUMPS FAL TO PROVIDE FLOW IHWM) mpy} .
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| 1 I PUMP COotA FAO PUMP Co018 FALS l RWC001 A l l kWCOO!B l a
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| i I PUMP C001A FALS TO PUMP Dt3CHAFGE START OHECK VALVE F008A FAES TO OFe.N l Rwoo1AD I l RMEJ AD l 227Em 1.40E-04 I I P.!MP C00tA FALS STRANER 0001A plt)GG DUFtNG MGSION i awoot As l l Rwooni A i 1.94E A 3 40E-05 1 ,
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| i 2 I CLEAN-UP WATER SYSTEM (SDC) .\SDTREE\RWCUF.CAF 6-29-92 Page 3 z.-.-- . _ ._.___
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| .I-l l
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| PUMP C0018 FALS
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| I I I PUMP C0018 FALS TO PUMP OtSCHAAGE l START CHECK VALVE F0068 g FA13 TO C*EN I fien180j l nwxceto l O '"~ O """
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| l .\SDTRESRWCUF.CAF 6-29-92 Page 4 1
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| CLEAN-UP WATER SYSTEM (SDC)
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| RWCU NCPH4EG "r EAT ac>wnEn ccounG LNAVALAELE nanun n
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| . E F/ R17:V_ CF N0tMiGG FALURE OF NOf& REG HX 1 HX 2 I
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| I 1 1 2 I s 1 4 i CLEAN-UP WATER SYSTEM (SDC) ASDTREETRWCUF.CAF 6-29-92 Page 5
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| :___. _. ._. _ _. o
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| RWCU VALVES FAR.
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| I Rwv'. vs I l-- I RWCO ISOt.ATION MOV MOV Fo11 FALS F002 FALS CLOSED , CLOSED OJCFC}
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| [ sw row 1 3 ooE-os I e 1 2 l CLEAN-UP WATER SYSTEM (SDC) .\SDTREE\RWCUF.CAF 6-29-92 Page 6
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| Pace ZCne Gate /E.w t NE Ergt h Gate / Event Mare Pace Zone Gate / Event Nne Pace ZCne Gate / Event Nre 2 2 KWILVS 1 2 EACE1N 2 4 PF/LVS 6 2 EACE2N 2 1 h"1 CWA 5 2 EDC11N 2 3 WROWB 5 4 EDC12N P12N 1 1 P12N 2 2 PlN 2 2 l
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| l P2N 2 4 FX/002F 6 1 FF/003F 6 1 FK/011F 6 2 i Fr/015F 6 2 FM/016F 7 1 i FX/017F 7 2 RWOO1AD 3 1 KdOO1AR 3 1 RW001BD 4 2 RW001BR 4 1 RWOO21 5 1 RWOO22 5 3 RWOO8AD 3 2 RWOO8BD 4 1 TWO14D 6 2 RW1617F 7 2 RWC001A 3 2 RWC001B 3 3 RWC001B 4 2 RWCUF 1 2 RWCUOP 1 1 RWD001A 3 2 RWD001B 4 2 RhMAINT 1 1 RWNRHX 1 2 RWNRHX 5 2 RWNRHX1 5 2 Rht'RHX2 5 4 RWPUMPS 1 2 RWPUMPS 3 2
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| .\SDTREE\RWCUF.CAF 6-29-92 Page 8 CLEAN-UP WATER SYSTEM (SDC)
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| .\SDTREE\RWCUFPC.CAF 6-29-92 Page 2 CLEAN-UP WATER OR F'3EL POOL (SDC)
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| i FAR.1JRE OF F5e B
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| = m um i itw i I) lI I I FtHR LOOP B FALS RHR LOOP C FAR.S iwsoEtosei 1%xmtoxl Page 2 Page 16 1
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| RHR LOOP B FALS lusovvtussi Page 29 I I I I RHR PUMP 8 OfV1S80N 2 POWER FIHR LOOP B NSUFFCIENT DNA FAR.S TO FALS UNAVAR.ABLE DUE TO COCUPG FROM LOOP B FFACH RPV MANTENANCE I NSWBOfSC l P2 I wtwwaAWT l lOtmE3l Page 5 100E41 Pw;e 6 I I f RHR LOOP B FALS TO RPV OtSCHARGE NKCT NTO RPV FALURE lwt*WSI I M.x3 A B9 i Page 3 Pwps 4 I i I 2 I 3 I . I EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 2
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| M LOCP B FAILS TO PECT INTO FFV I wtem I I I MECHANICAL FALLHE MECHANCAL FALURE OF TESTABLE CHECK OF PMCTICtd VALVE YALVE Etif0068 E1150058 (NCFC)
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| I so%w j l Scsus I t soE44 17/CE43 1 2 1 I i
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| .\SD i REE\DECAYJ.CAF 6-29-92 Page 3 EITHER RHR-B OR RHR-C (SDC)
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| 9 RPY OtSCHA7G FAILUFIE I porsvtas1
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| -=
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| i I DW SPRAY VALVE E11- FUEL POOL VALVES WW SPRAY VALVE Ett-F0184 FALS OPEN FAL OFTN (LOOP B) F0148 FAILS OPEN (NCFO) (NCFO) l 5cteaFO I I L5'sve6O I l Fotw+O I
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| 1 I = 1 3 EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 4
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| Po 37 I I FALURE OF DNISaON FAtilRE OF 08VtS8CP4 {
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| I . I 2 I EFHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 5 i
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| ecL5nc:ENT COOUNG FBOM LOOP B I om*s l I I FALUFE OF LOOP B HX NSL*FICMEAT FOCW FLOW TO MXB iS FROM LOOP 8 PUMP BYPASSED d
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| , Y lewi I e wc I t rA43 I i i I 3 I . I _
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| ; EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92. Page 6
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| : w. _
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| PtA#P AlmuARIES FAL lWO3 AUX l-I 1 LOOP B Ft9AP C00 LNG RNR PUMP B FALS TO FALS RUP4 I wi+*+c I iwta vei 1.94E44 FALURE OF DMSON PUMP B � TOR B RCW TO PROYM BEARNG COOLER FALS COOUNG FLOW T
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| RHR t.OOP B SUCTION VALVES FAL
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| ,' I conisuv I I I MANUAL VALVE E11- VALVE E11f011B F0090 FALS TO FALS CLOSED (NCFC)
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| FOt2B FAILS CLOSED PacFC)
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| I Fotsc 1 l Foisc 1 170E43 170E43 I I OfVISION 3 POWER NTERLOCKS PREVENT FAILS OPENNG OF VALVE E11-FO129 l Ira CCASB l Page 12 Page 13 1 1 1 2 I EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 11
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| .\SDTREE\DECAYJ.CAF 6-29-92 Page 13
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| ~ iTHER RHR-B OR RHR-C (SDC)
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| OTHER VALVES FAL OPEN OR CLOSED
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| .\SDTREE\DECAYJ.CAF 6-29-92 Page 14 EITHER RHR-B OR RHR-C (SDC)
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| .\SDTREE\DECAYJ.CAF 6-29-92 Page 15 EITHER RHR-B OR RHR-C (SDC)
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| mRLOOP GFALS lw.w =x t I
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| E11+005C ('vCFC) (NCFO)
| |
| VALVE Ett-roceC i *."->v I * +o ! l I+=~*1 I 5" ~ we 1 1 '
| |
| 0 '"~ O '""' O '"~ O '"~ l l
| |
| FUE5 POOL VALVES FAL OPEN (LOOP C)
| |
| [e m.*vg l
| |
| Page 17 I
| |
| I 3 l 4 l 9 l l
| |
| * l 2 l I .\SDTREE\DECAYJ.CAF 6-29-92 Page 16 !
| |
| : j. j EITHER RHR-B OR RHR-C (SDC) 1 I
| |
| *- ' 4 . - 3 4*. ,y ,
| |
| s .-
| |
| ~
| |
| / -
| |
| | |
| l l
| |
| l FUEL POOL VAVE*
| |
| FAL OPEN (LOCe C) 1 '
| |
| Page 16 FUEL POOL VALVE E11- FUEL POCL VALVE E11-F014-C FALS OPEN F015C FALS OPEN (NCro) (NCrot i toioo 1 I stvc I 1.80E44 1 coE44 I i i = 1 i
| |
| .\SDTREE\DECAYJ.CAF 6-29-92 l Page 17 EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| NSumCrur COOLNG FROM LOOP C I oveces l FAILLTE OF LOOP C HX NSUFFICENT FLOW FLOW TO HNC IS FROu LOOP C PUMP BYPASSED 1 HxcFat i iuswtx>tol I wic se i I I I I EAT EXCHANGER HX TUBE SOE BYPASS UNiutgut FLOW BYPASS SERV 1CE WATER UNAVAE.ABLF TO HXC FAILS DURING VALVE E11f013C LINE OPEN OPERATON FALS OF94 (NCF(M I HEcqw,v,,g HNC [ F 0 ? )>O l lWCWAM i 1.14E44 t ecE44 Page 29 i i FAILURE CF HX IPA.E4 FAILURE OF DnCSON VALv ?217013C C ACW TO PROY0E (NCFC) COOUNG FLOW T
| |
| wo,2>c; i-i 1.70E43 I i I 2 I a i A I EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 18
| |
| | |
| FBOM LOCP C PUMP l wsM=>w n MIENT FLOW PUMP AUX 1]ARIE3 TO PUMP SUCTION FAL Icooiuts1 Ivoc^ux1 Pap to Page 23 g I M PUMP C FALS TO RHR LOOP C SUCTION START VALVES FAL I cooiURS I I CW''"Y I Pege 21 Page 24 PUup SUCTION VALVE OTER VALVES FAL E11MIC FALS OPEN OR CLOSED OPEN PCFO) l 5micw 1 I r ocvt vss i 1 ecE433 Perp 27 l
| |
| 1
| |
| * I 2 I
| |
| .\SDTREETDECAYJ.CAF 6-29-92 Page 19 EITHER RHR-B OR RHR-C (SDC)
| |
| A
| |
| | |
| i nstmcsEur rtow w r - sucrica
| |
| _ ,0 1 I FFV WATER 1.EVEL PEACTOR P5 ESSURE INADEQUATE DAOPS CAUS1PE3 SUCTION CAVfTATION l vstwevE l lNI 100E 6 [ 100EM i i 4
| |
| 1 I = I EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 20 __
| |
| | |
| r 4
| |
| i l fVr2 PJe#
| |
| C FAILS TO STAftT l1
| |
| _- . ,, i _ys i
| |
| ~
| |
| I I FNR F'JMP C "43 TC MAPFA FFAT:CM r STA J FALURE l
| |
| : I ic=cwi i we i O ~~ A'" :
| |
| a i
| |
| .i I-i i
| |
| i I . I 2 1 EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 21
| |
| | |
| ~
| |
| = ,<Twrm FALUFM i
| |
| 1 IKCI l
| |
| {--
| |
| FA!LL*T TO ntAppJALLY I S8GreAL TPNCSG FALS RKTMTE a .,sc<.,u i i~m x i t omai
| |
| - c curt er.tyd FAILS TO CLOSE I
| |
| i m.m i O '""
| |
| 4 i = I 1
| |
| .\SDTREE\DECAYJ.CAF 6-29-92 Page 22
| |
| (
| |
| EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| FUMP AWf; FAL
| |
| / s Iwt0*v=1 1 1 LCCP C PUMP COOUNG FtHR PLW C FAE.S TO FAILS RU*8 Mou t_wnw l 1,94E M V~.
| |
| 4 I I FAILLHE OF DMSON I PUUP C 6 '
| |
| C RCW TO PRCMOE BEAR!NG COCIER FAR,S le(wcl t*Lov 1 t.taE44
| |
| ? I RSAC PUMP ROOM PL9tP C MECHANICAL A.C. UNIT FALS SEALER COOLER FA15
| |
| , i i xomi i co, sci 4 76E43 1.14F44 1
| |
| 2 1 I i 1 ,
| |
| EITHER RHR-B OR RHR-C (SDC) .\SDTREBDECAYJ.CAF 6-29-92 Page 23
| |
| | |
| " tP LOOP C SLGY4 n
| |
| VALVES FA4.
| |
| I coo ttu v J Q ,
| |
| OfVSA 1 PCMER RPV J84 LPE FAILS tSCLATKt4 VALVE Eti-Fot2C FARE CLOSED t
| |
| h et i tot 2CN 5 f iccE N Page 25 [
| |
| f I t NTEr. LOCKS 6 VALVE E115011C OPENNG OF VALVE FAILS CLOSE9 4.'JCF'C)
| |
| E11-F012C 1 Mdd I Fo'm; I
| |
| ~ 2.
| |
| ( ,nm I h MANUAL V4WE 5.11 l VAL *K Ett-FoioC F009C FALS TO l FNLS (10 SED (NCFC) rEAN OEN l t
| |
| 1 Fotv>c j l_50 xFC I 0 '"~ O '""
| |
| l I I i i
| |
| .\SDTREE\DECAYJ.CAF 6-29-92 Page 24 EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| OfVtSON 1 W.A F'M e- ,a rwe ze
| |
| , i i
| |
| FalLLAE OF DVISION FA43JRE OF DrvtSION 1 125 V DC BUS 1480 V UCC T T s uc,, , p,ce , i
| |
| ^
| |
| L1 A w
| |
| I l
| |
| l t
| |
| i<
| |
| , I I 1
| |
| l .\SDTREE\DECAYJ.CAF 6-29-92 Page 25 EITHER RHR-B OR RHR-C (SDC)
| |
| I l .- u
| |
| | |
| I i
| |
| 3 7<rEmoCe enEvEm CeENHW3 OF VALVE E11f012C I 'NLCxASC I g
| |
| l , .
| |
| VALVE E11f001C VALVE E11f018C LDAfT SWTTCH FAILS (DATT SWITCH FAILS l I otsooic I I otsoiec 1 6 coE44 6 80E44 E
| |
| l VALVE E11-F00eC VALVE E11-F019C L9JfT SWITCH FAILS LIMfT SWTTCH FAILS l
| |
| l otdxwil I ou.oi9c I
| |
| (
| |
| 6 soE4s 6 00E44
| |
| }
| |
| l l
| |
| l I
| |
| I e i I i
| |
| .\SDTREE\DECAYJ.CTF l 6-29-92 Page 26 EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| OTHER VALVES FAL ceEN m etcSED i
| |
| p I wtacvtvss i
| |
| (- .
| |
| . i MANJAL VALVE E11- VALVE E11-F004C F003C FALS CLOSED FALS CLOSED (NOFC)
| |
| (NOFC) i Fooy>c l l Foo.cFc 1 1.30E-04 1.7E m i I PUMP DISCHARGE VALVE E11-FCoeC CHECK VALVE E11- FALS OPEN (NCFO)
| |
| F002C FAILS CLOSED I Foo2CFC l l Foom +C l 0 """ O """
| |
| l 1 1 2 1 EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 27
| |
| | |
| l MNIMUM PCW BYPASS LNE OPEN IWDd6PEJ I
| |
| I NO HIGH FLOW SIGNAL MNIMUM FLOW VALVE E11-F021C FA3.S OPEN (NOFO) 1 WLd:+ 1 I Fo2icro I Q O'"-
| |
| e i
| |
| MISCALIBRATION OF FLOW SKt*4AL FROM l FTooec FAILS FLOW TRANSMfTTERS E11-FT00eA.B.C l wt>C'mtF ] I CAUMMA I 2.90E-04 5 coE45 1
| |
| $ I 2 i l
| |
| .\SDTREE\DECAYJ.CAF 6-29-92 Page 28 EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| l HsHiDAs I r3 I I RHR LOOP A FAILS RHR LOOP B FAILS INSoWDGSAl [NS( OCS8l I I I I RHR PUMP A OfVtSION 1 POWER RHR LOOP A INSUFFICtENT DISCHARGE FAILS TO FAR.S UNAVAILABLE DUE TO COOLING FROM LOOP A REACH RPV MARRENANCE I NswAoisc I et I woA*AANT I I DiecAS l Page 25 1.00E41 Page 32 I I RHR LOOP A FAILS TO FEEDWATER LOOP A POECT PRO RPV VALVES Fall l WDAINS I lWokFW l Page 30 Page 31 1 : I 3 1 4 1 I 1 EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 29
| |
| | |
| 1 FDM LOOP A FALS TO NECT INTO RPV M ENSI
| |
| ~~
| |
| I I MECHANICAL FALURE MECHANICAL FALURE OF TESTABE CHECK OF MECTinN VALVE VALVE E11-F006A E11-F005A (NCFC) 1 FooeAuf_1 1 FocuMF i 1.coE44 1.70E43 1 1 1 2 i
| |
| .\SDTREE\DECAYJ.CAF 6-29-92 Page 30 EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| l FEEDWATER LOOP A VALVES FAL WOAFW Page 29 I f TESTABLE CHECK CHECK VALVE B21-VALVE B21-F003A F604A FALS TO CPEN FAILS TO OPEN I 8Foo'3AFc ] I eroo4 Arc l 1.soE-04 1.40E44 MANUAL VALVE B21-F005A FAAS CLOSED (NOFC)
| |
| I eroo w c I
| |
| , n-u I i I e i
| |
| .\SDTREE\DECAYJ.CAF l 6-29-92 Page 31 EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| NSiETCIENT CO^JLING FROM tDOP A p IOtPCAS l F.djuRE UF
| |
| * OOP A HX INSUFFCtENT R.OW FLOW TC MAA IS
| |
| ' FM LOOP A PUMP BYPASSED l HxhA3. l l NSWDAF LO l HXABP
| |
| ( Pr4}s 33 I I I s SERVICE WATER HEAT EXCHANGER HX TUBE SOE BYPASS MNIMUM FLOW DYPASS UNAVAA.ABLE TO HXA FA!LS DURING VALVE E11TO13A LINE OPEN OPERATICN FALS OPEN (NCFO) ksAsw HzA I Fos 3Afo I iwo**F]
| |
| 1.14 E-04 1_eOE44 Page 40 I
| |
| i i FAILURE OF 0:VtS50N F,'JLURE OF HX INLET A RCW TO PROVIDE VALVE P21fo13A cor1 SNG Flow pcFC)
| |
| Y twscwAt I wo[wc 1 g / 1.wo3 4._.1 I i i 2 1 3 1 4 i EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 32
| |
| | |
| nsurFcEnr FLOW FROM LOOP A PUMP I NswoAF LOI r
| |
| NSUFFCENT FLOW PUMP AUXLIARtFS TO PUMP SUCTION 'FAL l CootAFLS l l WDAAUX l Page 34 Page 36 I I flHFi PUMP A FALS TO RHR LOOP A SUCTON START VALVES FAL l cooshas l l COO 1 AMV l Page 35 Page 37 1 I PUMP SUCTON VALVE OTHER VALVES FAL Ett#001A FAILS OPEN OR CLOSED OPEN (NOFO) l F001 AFO l 1 WDA vtVSS I 1.e0E-03 Page 39 I i I 2 1 EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 33
| |
| | |
| e4StKFC:ENT FLC7f i
| |
| 'O P' Jkte SLCTION g L_C_c AFLS I L-RPV W/.TER LE%EL REACTOR PRESSUFE f BMADEOUATE DROPS CAUSN3 SUCTION CAVfTATON i
| |
| [NShWVF l-IWTMl iu. twE.
| |
| l I i I 2 I EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 34
| |
| | |
| l RHR PUMP A FALS TO START I cootAtus l j Page 33 I
| |
| 1 RHR PUMP A FAR.S TO MANUAL INITIATICt4 START FALunE I COO 1AA* 1 n45cs O'""'
| |
| I I
| |
| FALURE TO MAF#JALLY SGt4AL TRANSMISSON WNE FALS l
| |
| 1gualctmii iH esotR*1 l 2 00E42 l
| |
| ( I I
| |
| , 1
| |
| ( RHR CIRCUfT BREAVIR FALS TO C*M 1 0 EcootAce_I t ooE-06 l l I '
| |
| 6-29-92 Page 35 I
| |
| i \SDTREE\DECAYJ.CAF EITHER RHR-B OR RHR-C (SDC)
| |
| L ___
| |
| | |
| h I
| |
| PtAF AUXLIARIES FAL IwoAAux_]
| |
| I i
| |
| LOOP A PUMP COOLING RHR PLAP A FALS TO FALS RUN
| |
| \
| |
| ! lwoAnc] I wortus]
| |
| l Q O '~~
| |
| I i FALURE OF D! VISION PUMP A MOTOR A RCW TO PROV0E BE/Ji1NG "MR FALS COOUNG FLOW T
| |
| IWOAutscj IWRCWAI 1.14E-04 I
| |
| I MA PUMP ROOM PUMP A MECHANICAL A.C. UN'T FALS SEALER COOLER FALS IWDARACI lWDAMcM 4.76EM 1.14E N
| |
| \
| |
| l i
| |
| t 1 2 i i 1
| |
| .\SDTREE\DECAYJ.CAF 6-29-92 Page 36 EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| RHFt LOOP A SUCTION VALVES FAL l cooisuv I I I MANUAL VALVE E11- VALVE E11-F011A F009A FAILS TO FAILS CLOSED (NCFC)
| |
| REMAN OPEN I FooeAFC I I FotiAFc 1 1.30E-04 1.70E-03 l l RPV SUCTION LINE VALVE E11f010A I ISOLATION VALVE E11- FALS CLOSED (NCFC)
| |
| F012A FALS CLOSED (NCFC) l Foi5pf C I 1 FotoAFC l 1.70E4 1.70E43 I I DaAS10N 2 POWER NTERLOCKS PREVENT FAILS OPENING OF VALVE E11-FO12A lINLOCKSA]
| |
| Page 5 Page 38 I i 1 2 1 EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 37
| |
| | |
| mtnLoCxs enEvEur OPENING OF VALVE E11-F012A i moce i
| |
| ~
| |
| VALVE E11f001A VALVE E11f006A LNIT SWTTCH FALS UMIT SWITCH FAILS IOUNotAi l OLSooeA i 6 00E-04 6 90E44 I i I 2 I EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.C.AF l 6-29-92 Page 38
| |
| | |
| OTHER VALVES FAL OPEN OR CLOSED E>Avtyss I I MWAL VALVE E11 VALYE E11f004A F003A FALS CLOSED FALS CLOSED (NOFC) l Foo5AFC l l F004 AFC l 1.30E44 110E44 PUMP DISCHARGE VALVE EttfoceA CHECK VALVE E11- FALS OPEN (PCFO)
| |
| : F002A FALS CLOSED 1 Foo2AFC 1 l FOOSAFC l 1.40E-04 1 SOE44 I 1 1 2 I EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 39
| |
| | |
| MINIMUM FLOW BYPASS LDJE OPEN IWDA8PFI
| |
| ()
| |
| i NO HIGH FLOW SGNAL MINtMUM FLOW VALVE E11-F021A FAILS OPEN (NOFO) l WDAHF l l F021AFO l 1,80E43 I I FLOW SIGNAL FROM MISCAUBRATION OF FT00eA FAILS FLOW TRANSMffTERS E11f7008A.B.C l WDAf42LF 1 l CALNOO2A l 2.90E-04 5 00E45 l- 1 1 2 1 EITHER RHR-B OR RHR-C (SDC) .\SDTREE\DECAYJ.CAF 6-29-92 Page 40
| |
| | |
| Pace Zone Gate / Event Name EZLqe Zone Gate / Event Name jageZonel Gate / Event _Name Pace Zone Gate / Event Name 37 1 INLOCKSA 37 2 31 1 DLS001C 26 1 F009AFC BF003AFC 11 INLOCKSA 38 2 BF004AFC 31 2 DLS000A 38 2 F009BFC 1 2
| |
| 24 1 INLOCKSB 11 31 2 DLS008B 13 1 F009CFC 2 BF005AFC 37 2 INLOCKSB 13 35 2 DLS008C 26 1 F010AFC C001ACB 11 2 INLOCKSC 24 1 1 DLS018B 13 2 F010BFC C001AFLS 33 INLOCKSC 26 2 26 2 F010CFC 24 2 C001AFLS 34 2 DLS018C 2 NSDRPVF 8 1 13 2 F011AFC 37 C001AFRS 33 1 DLS019B 20 1 26 2 F011BFC 11 2 NSDRPVF C001AFRS 35 2 DLS019C 34 1 25 2 F011CFC 24 2 USDRPVF C001AMF 35 3 EACE1 29 3 5 2 F012AFC 37 1 MSDWDCSA C001AMV 33 2 EACE2 1 1 12 2 F012BFC 11 1 MSDWDCSB C001AMV 37 2 EACE3 2 NSDWDCSB 2 3 25 1 F012CFC 24 C001BCB 9 2 EDC11 29 4 )
| |
| 5 1 F013AFO 32 4 NSDWDCSB C001BFLS 7 1 EDC12 1 2 (
| |
| 12 1 F013BFO 6 4 NSDWDCSC C001BFLS 8 2 EDC13 18 4 NSDWDCSC 16 4 7 1 EMSCONN1 35 1 F013CFO I C001BFRS 4 2 NSWADISC 29 2 9 2 EMSCONN2 9 1 F014BFO C001BFRS 17 1 NSWEDISC 2 2 9 3 EMSCOtM3 22 1 F014CFO C001BMF 4 3 NSWCDISC 16 3 7 2 F001AFO 33 1 F015BFO C001BMV 17 2 USWDAFLO 32 3 2 F001BFO 7 1 F015CFO C001BMV 11 4 1 NSWDAFLO 33 2 22 2 F001CFO 19 1 F018BFO C001CCB NSWDBFLO 6 3
| |
| .39 1 F018CFG 16 3 C001CFLS 19 1 F002AFC 7 2 14 1 F019BFO 4 3 NSWDEFLO C001CFLS 20 2 F002BFC 4 NSWDCFLO 18 3 27 1 F019CFO 16 C001CFRS 19 1 F002CFC 19 2 F021AFO 40 3 MSWDCFLO 21 2 F003AFC 39 1 24 1 C001CFRS 15 3 P1 F003BFC 14 1 F021BFO C001CMF 21 1 28 3 P1 25 2 2 F003CFC 27 1 F021CFO C001CMV 19 32 2 P1 29 3 F004AFC 39 2 HXA
| |
| 'C001CMV 24 2 4 P2 2 3 14 2 HXABP 32 CALN002A 15 2 F004BFC P2 5 2 2 HXAFAIL 32 2 28 2 F004CFC 27 CALN002A 32 1 P2 37 1 30 2 !c ASW CALN002A 40 2 F005AMF .
| |
| 6 2 P3 11 1 F005BMF 3 2 HXB D1PCAS 29 4 6 4 P3 12 2 F005CMF 16 2 HXBBP
| |
| ' D1PCAS 32 3 6 2 F3 16 4 F006AMF 30 1 HXBFAIL D1PCBS 2 4 6 1 PDISVLBS 2 2 F006BMP 3 1 HXBSW
| |
| ! D1PCBS 6 3 18 2 PDISVLES 4 2 6 F006CMF 16 1 HXC DlPCCS 16 4 PDISVLCS 16 4 39 2 HXCDP IB D1PCCS 18 3 F008AFC 2 FAMIS 35 2 14 2 HXCFAIL 18 DLS001A 38 1 F008BFC 18 1 RBMIS 9 2 F008CFC 27 2 HXCSW DLS001B 13 1
| |
| .\SDTREE\DECAYJ.CAF 6-29-92 Page 41 EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| l Gate / Event Name Pace'Long Gate / Event Nara_ Egge Long- Gate / Event-Name Pace Long Gate / Event Name Pace Zone RCMIS 21 2 WDBLUB 10 3 22 2 WDBMAINT 2 4 RCMIS RHRSDERA 35 2 WDBMBC 10 2 9 2 WDBMSC 10 2 I RHRSDERB RHRSDERC 22 2 WDBN2LF 15 1 RSHTDA.B 29 4 WDBRAC 10 1
| |
| -1 2 WDBRC 10 2 RSHTDBC 4 2 WDBVLVSS 7 2 UPSVBFO 16 4 WDBVLVSS 14 2 UPSVCFO 17 2 WDCAUX 19 2 I UPSVCFO 32 2 WDCAUX 23 2 W013AFC 6 2 WDCBPF 18 5 WO13BFC 18 1 WDCBPF 28 2 W013CFC 33 2 WDCHF 28 2 WDAAUX WDAAUX 36 2 WDCINS 16 2 32 5 WDCLUB 23 3 WDABPF WDABPF 40 2 WDCMAINT 16 4 29 2 WDCMBC 23 2 WDAFW 31 2 WDCMSC 23 2 WDAFW 40 2 WDCN2LF 28 1 WDAHF 29 i WDCRAC 23 1 WDAINS 30 2 WDCRC 23 2 WDAINS 3 WDCVLVSS 19 2 WDALUB 36 WDAMAINT 29 4 WDCVLVSS 27 2 36 2 WDNPSC 8 2 WDAMBC WDAMSC 36 2 WDNPSC 20 2 40 1 WDNPSC 34 2 WDAN2LF WRCWA 32 1 WDARAC 36 1 WDARC 36 2 WRCWA 36 1 33 2 WRCWB 6 1 WDAVLVSS 39 2 WRCWB 10 1 WDAVLVSS 7 2 WRCWC 18 2 WDBAUX 10 2 WRCWC 23 1 WDBAUX WDBBPF 6 5 WDBBPF 15 2 WDBHF 15 2 WDBINS 2 1 <
| |
| WDBINS 3 2
| |
| .\SDTREE\DECAYJ.CAF 6-29-92 Page 42 EITHER RHR-B OR RHR-C (SDC)
| |
| | |
| 1 i
| |
| e g
| |
| N W) a C 4 I
| |
| SR P
| |
| ME T l
| |
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| |
| 2-6 .
| |
| 8 4
| |
| E 4 l D 4 E I 3 R
| |
| U T
| |
| A N
| |
| F n P U
| |
| R
| |
| +
| |
| E e s A R
| |
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| |
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| |
| 7 C.
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| |
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| M LS EI A L
| |
| 4 2 W EF A S
| |
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| , M5W0O i E
| |
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| |
| M *A l P T
| |
| D UR PE P s A
| |
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| WO W .
| |
| C 1 PO T L 5 4 O 4 NE OC 71 R TAF C) E 0
| |
| I1 0 NEO 2 SVW B O 0(
| |
| 0N 2 MOO l C DRLP F A
| |
| W fD EtE F
| |
| OOG E; ,
| |
| R PS 2 !
| |
| T IN 6 U O E L v TEL RWO I AVC RL UCO E A S.
| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
| [ T OC U CX T
| |
| ' A A1A R
| |
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| |
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| |
| * W AE E B HR c 'XL U NELRA 2
| |
| A
| |
| \ E i T CGA M 1 C
| |
| EA F
| |
| AN V AA W / I TH N I 1
| |
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| |
| A E 2 SCU RX V
| |
| I N E X
| |
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| |
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| 1 A 0 E E 0 i. l P a E CN RA Ni A
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| LOSS OF COCUNG TO LOSS OF COOUNG LOSS OF COCU*C TO THE FIRST ACTIVE FROM THE SWS TO THE THE SECOND ACTfVE HEAT E/ CHANGER STAPCBY HEAT HEAT EXCHANGER EXCHANGER I wwei I l wssmal I wama21 P g 19 I I I I FALURE OF CIVtSON SEPARATION VALVE FALURE OF OtVtS!ON SEPARATION VALVE B SWS TO PROVOE P41-F0058 FALS B SW$ TO PPOVOE F003 FOR THE COOLPG FLOW (NOFC) COOLPG FLOW STAPCBY N1 FALS CLOsEn ecFo Iwswse1 IwwvswHl I* sew 1 Iawvsml Page 17 3 00E4 Pup 17 5 40E44 I I SEPARATICH VALVE AtJ1TOMATC SEPARATION VALVE P41-F0038 FALS INmATION SGNAL F005 FOR THE (NOFC) AND OPERATCH FALUAE STANOBY HX FAL.S CLOSED eeCFQ I wuvswH l I wavi&e l I wwvswl 3 00EM Page 13 5 40E&
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| i I I ACTIVE HEAT SEPARATFJN VALVE l EXCHAF4R FALURE P21fC04F FAAS ctosto w ei I wue 81F H l l WSA VRaF H l 1.14E-04 3 coE46 l
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| tcSS & SWS COOten TO Div C RCW HEAT EXCHANGERS Pap 21 l 4 I LOSS OF COOUNG TO LOSS OF COOUNG LOSS OF COCUNG TO THE FRST ACTIVE FROM THE SWS TO THE THE SECONO ACTWE HEAT EXCHANGER STANDBY HEAT HEAT EXCHANGER EXCHANGER Iw mucil Iw wxcl I wan =c2 l I I I I FALURE OF OtVISION SEPARATION VALVE SEPARATION VALVE l.FAILURE C SWS TO OFPROVIDE OfvtS10N Foo3 FOR THE C SWS Te PFiOVIDE P414005C FALS COOUNG FLOW (NOFC) COOLING FLOW STANDBY HX FALS CLOSED #8CFQ twswscI I wvvs . CHI Iwswsci I wunw l Page 27 300Em Pa;pe 27 5 acEW I I SEPARATION VALVE AUTOMATIC SEPARATION VALVE P41To03C FALS NTIATION S3GNAL F005 FOR THE F40FC) AND OPERATOR FALURE STANOEW HX FALS CLOSED (NCFQ Iwvvsrst I wAurcn I l wisvwn 1 3 ocEm Par 23 5 40E-04 I i 1 2 1 3 I A I SERVICE WATER SYSTEM (SDC) .\SDTREE\SWS.CAF 6-29-92 Page 26
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| .\SDTREE\RHRCF.CAF 6-29-92 Page 18 RHR CORE FLOODING
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| I I
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| i 1 1 2 1 RHR CORE FLOODING .\SDTREE\RHRCF.CAF 6-29-92 Page 21
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| MPMAUM FLOW BYPASS LINE OPD4 A l W >B8141 M I I I NO HIGH FLOW SIGNAL M!N! MUM FLOW VALVE E11-F021B FA4.S OPEN (NOFO)
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| [vY @} I fo718ro l
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| \ 1.80E43 j
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| FLOW FTIOM M!SCAtt T10H OF FT00e8 FAILS FLCW TRANSMITTERS E11-FT008A.B.C
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| . . ~
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| l WD6N2LF l l CALNoo?A l 2.90E44 $ COE4 1 i I 2 i RHR CORE FLOODING .\SDTREE\RHRCF.CAF 6-29-92 Page 22
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| RHR LOOP C FALS p ,
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| Iwocsci I I I I RHR LOOP C NSUFTICENT RHR PUMP C OfV1S10N 4 POWER COOtJNG FBOM LOOP C F ' '- J UNAVALABLE DUE TO OtSCHARGE FAR. TO MANTENANCE REACH FtPV lWDCMAW11 I DiPCC l I WDCOiSC 1 P3 LME41 Page 26 Page 27 I I RHR LOOP C FAIL 3 TO SUPP POOL DISCHARGE PLECT NTO RPV FAILURE I .
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| I wocw I I Fiatsvtc 1 I I OPEN SIGNAL TO MECHAFTAL FALURE PLECTION VALVE E11- OF NJECTION VALVE Fo05A FAILS E11-F00*A (NCFC)
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| Iroowtoo 1 1 FoO*CMF 1 Page 24 1.70E43 MECHANICAL FAILURE OF TESTABLE CHECK VALVE E11-Foc4C l Fo06CMF j O '"~
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| 4 5 l 2 l 3 l l
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| l 1
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| .\SDTREEtRHRCF.CAF 6-29-92 Page 23 RHR CORE FLOODING
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| i CPEN SENAL TO EUECTION VALVE Eti-F005A FAILS Iroosctoo 1 I I VALVE AUTO CPENtNG MANUAL INmATION SIGNAL FAILS FAILURE lFoo'cAuT l l Rcus l I I GROUP 30tVtSION 3 I Ofv1SION 3 SYSTEM SYSTEMS RPV LEVEL RPV DOW PRESSUV SIGNAL FAILS SIGNA!L FAILS T T I evtic I I evec 1 b b I i 1 2 i RHR CORE FLOODING .\SDTREE\RHRCF.CAF 6-29-92 Page 24
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| MANUAL INITIATK)N FALL 9E f'tCMs Pege 24 I I FHR CIRCUIT BREMER SIGidAL TRANSMIS*M FAILS TO CLOSE FALS l Cecce l lEuscONN31 100E-06 FALURE TO MANUALLY INITtATE IN<4ENi t occA1 I i l 2 i RHR CORE FLOODING .\SDTREE\RHRCF.CAF 6-29-92 Page 25
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| i
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| *AJPP POOL JtSCHAPGE FAILURE p ,,
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| I Kasvt C l I I I
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| , FUEL POOL VALVES DW SPRAY VAtVES WW SPRAY VALVE E11-FAL OPEN (LOCP C) FAIL OPEN (LOOPS C) F014C FALS OPEN (NCFO) l r 9SVCFO 1 lDW5VCAO' l FotoCFO l 1NJE44 II I i I I FUEL POOL VALVE Eli- FUEL POCL VALVE E11- l DW SPRAY VALVE E11- DW SPRAY VALVE E11-F014-C FALS OPEN F015C FAILS OPEN t F017-C FALS OPEN F0184; FAILS OPEN (NCFO) (NCFO) (NCFO) (PCFO)
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| I Foi4cso l l Fosco 1 1 FoirCro l l Foiscro l 0 '"" O ~'~ O ~~ O '~~
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| l i I 2 1 3 1 4 i RHR CORE FLOODING .\SDTREE\RHRCF.CAF l 6-29-92 Page 26
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| Ii ll l
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| 's OlPCC '
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| ! I FA41)RE OF LOOP C HX NS*FICIENT FLOW FLOW TO HXC iS FFiOM LOOP C PUk? BYPASSED I
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| l HXCFAIL j l POCIF l ) HXCEP l Par 29 C- -
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| I I I I SERVICE WATER HEAV EXCHANGER HX TUBE SOE BYPASS MINNUM FLOW BYPAS9 UNAVAR.ABLE TO HXC FAN _S DURING VALVE E11-F013C LINE OPEN OPEfM1 SON FAILS OPEN (NCFO) lHA SWl HXC l FOl3CFO l
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| ~
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| l WDCBPf l 1.14E-04 1.80E44 Page :rJ I _
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| I l FALURE OF DIVISaCN FAILURE OF HX INtET VALVE P21fo13C C RCW TO PROV0E COCLING FLOW 94CFC)
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| T IwnCwc] l Woikac I 1.70E-03 I 1 1 2 I a I 4 i RHR CORE FLOODING .\SDTREE\RHRCF.CAF 6-29-92 Page 28
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| INSOFFICIENT FLOW l FROM LOOP C PUUP '
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| a l
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| ['
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| Page ?t rA f 1 I f INSUFFKDENT FLOW I RHR RiW2 C FALS TO Pt#p 5UCTiON VALVE PUMP AUXUAR$S OTHER VALVES Fall TO PUMP 3UCTION START AND FL*( E11-F%iO FAILS F A'L OC'EN 04 CLOSED CLOSED (NOFC)
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| ^
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| I notCRO I b7EMC Q $ MCICIC { l WD( A i#. l
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| ! WlCV! V% l Page 3c t.7CEos Page 32 1 i E SUPPRESSON POOL STPANER E11-D001C LOOP C PUMP COOLING F&R PUMP G FAILS TO
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| ' WATEP Ur*AVARABLE PWGGEO FAP_S RUN'
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| . DUE TO POOL RUPTURE
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| [Isps.w p l RM IMdyj WDuou 3.00EG 110E43 Pace 31 t04004 SUoPRESS80N POOL TEMP H!OH (LOSS Of M. IMP HEAO)
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| I rsvarw l 100E a l 1 l 1 l 3 l 4 I -5 l
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| RHR CORE FLOODING .\SDTREE\RHRCF.CAF 6-29-92 Page 29
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| Htfs P'WP C FALS TO SMW AND RUN
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| \
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| l cnotcra 1 I I PUMP START S!GNALS RHR PUMP C FA.U TO FAL START l coo.csio l l cooscuf I
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| .m.
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| , I umuAL un4 Ton I RHR LOOP C AUTO FALUBE PUMP START SGNAL FALS 1 Acus l l cooicatrr i b*" )
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| I I RHR CIRCUTT BPM l GROUP 30tVISION 3 FALS TO CLOSE SYSTEMS RPV LEVE1 SIGNAL FALS i T lcoofccal l evuc i e '-- A MAtAJAL OVERHIDE FA2 S IMTIATiON SIGMAL I C00 tcMOV l O'""
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| l 't l 2 l RHR CORE FLOODING .\SDTREE\RHRCF.CAF 6-29-92 Page 30 .
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| LOOP C Pt#AP COOUPAT FALS g3 b*_
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| ~
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| bAILURE DT/ISON PUMP C MOTOR C BCW TO P90t!OE EEARWG COOLER FA:LS C"XXJrKa FWV T
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| IVKwCI WDCUSJC 114E M RHRC P RCOM PUMP C CHANICAL l A.C. Uterr FALS SEALER COOLET4 FAJLS
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| [iFincnAc t [wccusci 476EC 1.14E M I
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| I 1 2 :
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| RHR CORE- FLOODING .sSDTREE\RHRCF.CAF 6-29-92 Page 31
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| OTHER VALVES Fall OPEN OR CLOSED i
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| p A I wocvt vE sI MAf0AL VALVE E11- VALVE E11-FOO4C Foo3C FAELS CLOSED FAILS CLOSED (NOFC)
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| (NOFC)
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| [ FNhCFC l l FOO4CFC l 1.30E44 1.70E-04 1 i PUMP DNE VALVE E11 FMcC CHECK VALVE E11- FAILS OPEN (PCFO)
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| Foo2C FA3LS CLOGED lFoo$cFC1 l FooaCFC l 0 '"~ O '"~
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| l i I = 1 RHR CORE FLOODING .\SDTREE\RHRCF.CAF 6-29-92 Page.32
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| t 4- ;
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| i- i e
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| f l
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| i- :
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| 1- !
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| l f
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| MNMUM FLOW BYPASS UNE OPEN '
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| IWD50PF l I I
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| ; NO HGH FLOW SGNAL MNMUM FLOW VALVE E11-F021C FALS OPEN (NOFO) i I FozicFo I i
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| ;, 1.80Em l l FLOW SG4AL FROM MISCAUBRATION OF .!
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| FT000C FALS FLOW TRANSMITTERS E11-FT006A.B.C ,
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| I WDdN2LF I l CALN002A l 2.00E-04 5.00E-OS 5
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| 2
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| , ,; ~
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| l- ... :1 2 'I
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| ^- -
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| - RHR CORE: FLOODING-
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| .\SDTREE\RHRCF.CAF f 6-29-92 'Page 33 4
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| q -
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| v _g y n - - - - - -
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| Gate / Event Nee Eagg Znne Gate /Fvent Name_ ?dce ZnIle Gate / Event._Uari a . Esse Z2ng Gate / Event Name EsLq e 13 Z OIle2 3 F008AFC 11 2 IPVPB BF003AFC 4 1 D1PCC 38 24 2 15 4 F008BFC 21 2 IPVPC BF004AFC 4 2 DNSVBFO P1 1 4 26 4 F006CFC 32 2 BF005AFC 4 ? DNSVCFC 12 3 4 F013AFO 5 4 P2 8 2 BACE1 1 C001AAUT 17 4 P2 16 2 3 1 EACE2 16 2 F013BFO C001ACB 28 4 P3 23 3 B 2 EACE3 27 2 F013CFO C0J1ACB 15 1 P3 27 2 7 2 EDC11 1 5 F014BFO C001AFLO 26 1 PCAIF 5 5 V
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| EDC12 14 3 F014CFO C001AFR 15 2 PCAIF 7 4 8 2 EDC13 27 1 F015BFO C001AFR 26 2 PCBIF 17 3 8 3 EMSCONN1 3 2 F315CFO CC01AMF 15 3 PCBIF 18 4 8 3 EMSCONN2 14 2 F017BFO C001AMOV 26 3 PCCIF 28 3 8 2 EMSCOUN3 25 2 F017CFO C001ASIG 15 4 PCCIF 29 4 19 2 7001AFC 7 4 F018BFO C001BAUT 26 4 PDISVLB 12 3 j 14 1 F001BFC 18 4 F018CFO '
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| C001BCB 15 5 PDISVLB 15 3 19 2 FOO1CFC 29 4 F019BFO C001BCB 26 5 PDISVLC' 23 3 18 2 F002AFC 11 1 F019CFO C001BFLO 6 3 PDISVLC 26 3 18 3 F0$2BFC 21 1 F021AFO C001BFR 22 3 RAMI 2 2 ;
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| 19 ? F002CFC 32 1 F021BFO C001BFR RAMI 3 2 j 11 1 F021CFO 33 3 C001BMF 19 3 F003AFC 8 1 HXA 5 3 RAMI 19 2 F003BFC 21 1 C001BMOV 5 4 REMI 13 3 ,
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| 2 F003CFC 32 1 HXABP C001BSIG 19 2 RBMI 14 2 l 11 2 HXAFAIL 5 C001CAUT 30 2 *004AFC 2 RBMI 19 1 21 2 HXASW 5 C001CCB 25 1 F004BFC 24 3 HXB 17 2 RCMI 30 2 FC04CFC 32 2 C001 COB 17 4 RCMI 25 2 l F005AAUT 2 1 HXBBP C001CFLO 29 2 RCMI 30 1 2 2 HXBFAIL 17 2 C001CFR 29 3 F005ALOG 3 2 2 4 HXBSW 17 1 RHRCFERA C001CFR 30 2 F005AMF 14 2 13 2 HXC 28 2 RHRCFERB C001CMF 30 3 F005BAUT 25 2 HXCBP 28 4 RHRCFERC 30 2 F005BLOG 12 1 C001CMOV 2 UPSVBFO 15 2 13 2 HXCFAIL 28 C001CSIG 30 2 F005BLOG 26 2 12 2 HXCSW 28 1 UPSVCFO CALN002A 6 2 F005BMF 2 1 WO13AFC 5 2 22 2 F005CAUT 24 2 IPVLlA CALN002A 8 2 WO13BFC 17 2 ;
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| 33 2 F005 CLOG 23 1 IPVLIA CALN002A 24 1 WO13CFC 28 2 1 5 F005 CLOG 24 2 IPVLIC DIPCA 30 3 WDAAUX 7 5 5 4 F005CMF 23 2 IPVLIC D1PCA 13 1 WDAAUX 9 2 12 5 F006AMF 2 3 IPVL2B D1PCB 19 3 WDABPF 5 5 17 3 F006bMF 11, 2 IPVL2B D1PCB 2 2 WDABPF 6 2 i 23 5 F006CMF 23 7 IPVPA ;
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| D1PCC .. '
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| 6-29-92 Page 34 RHR CORE FLOODING l .\SDTREE\RHRCF.CAF
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| I Gate /Evenc Man Pace Zgng Gate /Fyont Name Eagg ZQng Gate / Event Name Er322 Z2GS GdLO2.Enct Name Page Zone 1 2 WDCHF 33 2 WDADISC 1 3 WDCIN 23 2 WD?.TJ WDAFW 4 2 WDCLUB 29 5 WDAHF 6 2 UDCMAINT 23 4 WDAIN 1 2 WDCMBC 31 2 WDAIN 2 3 WDCMsc 31 2 2 WDCN2LF 33 1 WDALUB 9 l 1 1 WDCRAC 31 1 WDAMAINT WDAMBC 10 2 WDCRC 29 4 10 2 WDCRC 31 2 WDAMSC 1 WDCS 1 4 WDAU2LF 6 10 1 WDCSA 1 3 WDARAC 9 1 WDCSB 1 4 WDARC WDARC 10 2 WDCSB 12 4 7 2 WDCSC 1 5 KDASTPli 7 6 WDCSC 23 4 WDAVLVES NDA%LVES 11 2 WDCSTRN 29 2 WDa. AUX 13 5 WDCVLVES 29 6 17 5 WPCVLVES 32 2 WDBBPF 22 2 WD:tAnJT 7 2 WDBBPF 12 2 ' CWA 5 1 WCDDISC
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| ) 22 2 KECWA 10 1 WDSHF
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| [ ; WhCWB 17 1 WDBIN 1.2 1E S WRCWP 20 t hDELUL WDBHAINT 12 4 WRur 28 1 WDBMBC 20 W%WC 31 1
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| 'G 2 ZSPI:)tJDF 't 1 WDBMSC ,
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| 22 1 ZSP100LF 18 1 WDEN2Lv 20 1 OSPIUOOF 29 i WDBRAC 18 4 ZSP220LH 7 1 WDBRC 20 2 Z8P200% 3 2 WDBRC WDBSTUN 16 2 ZSP200Ja 29 2 WDBVI-/ES 18 6 WDB'JLVES 21 2 WPCAUX 29 5 WDCBPP 28 5 hDCBPF 33 2 23 2 fWDCPISC Page 35
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| .\SDTREE\RHRCF.CAFj G 29-92 l- .RHR CORE FLOODING .
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| l 1
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| ~~ '
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| -c-- _. .
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| q i .
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| Rr4% TFWHS A AND B FAL TO PftOVCE CORE FLOOOrG AksAsI, I '
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| RHR LOCp A FALS RHR LOOP B FAr S I *tm'AA I I*cC% I l
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| 1 i i 2 i I
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| .\SDTREL%RHRCFAB.CAF 6-30-92 Page !
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| RHR-A AND RHR-B CORE FLOODING (SDC)
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| - - - Wim
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| i 4
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| Cate/ Event ?hme Pace Zone Gate / Event Name Pace Zone Gate / Event name Pace Zone Gate / Event Name Pace Zone l WDCSA 1 1 I
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| WDCSAB 1 2 WDCSB 1 2 i i
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| , f i
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| '1 1
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| 1 i
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| i .
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| l t
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| i i
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| t RHR-A AND RHR-B CORE FLOODING (SDC) .\SDTREE\RHRCFAB.CAF 6-30-92 Page 2
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| llll R
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| H R -
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| B A
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| N D
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| R H
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| R-C C
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| O R
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| E F I L
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| O O R D
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| I H
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| R N I w
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| L O
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| (
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| w t ETW FON d L l
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| I l s s O RPS c O O B i D VA
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| .\ M10N S R G ED D H R
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| ! Gate /Evant Name Parte h Gate / Event Name Pace Zone Gate / Event Name Pace Zone Gate / Event Mame Paere Zone WDCSB 1 1 WDCSBC 1 2 j WDCSC 1 2 i
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| i i
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| .i i f L
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| 1 I
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| l 1
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| RHR-B AND RHR-C CORE FLOODING (SDC) .\SDTREE\RHRCFBC.CAF 6-30-92 Page 2
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| r FAtL1JRE OF HPCF
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| ! i i I 4 I i
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| 1 Mec t t i t
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| l I I FALURE OF HPCF-B FALURE OF HPCF-C y I
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| L q
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| ' HPCS E. -' lHW5Cl Page 21 j
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| ~ !
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| I '
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| i i "
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| DMSsC#4 2 POvdER FALURE OF etPCF4 HPCF4 UNAVAIULBLE FALS CUE TO M/INTENANCE ,
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| i t
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| ~ , ,ci a , , n ~~,un i s coE ^t FALURE OF DMSION FALURE OF DivtSK1N HPCF4 FALS TO HPCF4 PUMP COOUNG 2125 V DC SUS 2 480 V MCC START OR FIUN FALS 1 T T iacui in m i e, iami i b b b i 1
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| 1 2- 1 S I 4 I I i 1 HIGH PRESSURE CORE FLOODING SYSTEM .\SDTREE\HPCF.CAF 6-29-92 Page 1
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| I i
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| l l
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| 1 l
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| 4 e k
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| l HPCF-8 FALS TO START OR RUN I
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| Page 1 7
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| ; I I I i HPCF-S IPF'lATION PUMP OISCHAFKiE PUMP COO 18 FALS TO PUMP COO 18 SUCTION j SGNAL FALS FAILS TO REACH START OR RUN FALS
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| {
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| REACTOR 4 _. .
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| an n >+n i i
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| Pap 3 Page 4 Page 13 i i j! wCF Puur a FALS NeCF PuuP-s rAa.S I
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| : TO RUN TO START i I i, siemi i+ liao i ,
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| 1.94E-04 -
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| 6 90E-03 2
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| ij 4
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| I i l 2 1 3 1 4 I HIGH PRESSURE CORE FLOODING -SYSTEM .\SDTREE\HPCF.CAF 6-29-92 Page 2
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| I I
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| l l
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| l i
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| )
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| 4
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| W-8 HTIATION r SGNAL FALS I
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| ?
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| Page 2 Page 9 [
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| Page 11 [
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| lI I :
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| GROUP 2DivtSION 2 OPERATOR FALS TO ;
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| SYSTEMS RPV LEVEL ATTEMPT MANUAL !
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| SGNAL FALS INrnATION AFTER 30 }
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| T uat i I sit 2's I incx~N I 2.00E42 t
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| l i
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| ie t
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| 1 i i I 2 i HIGH PRESSURE CORE FLOODING SYSTEM .\SDTREE\HPCF.CAF 6-29-92 Page 3
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| PLMP CASCHARGE FALS TO REA'>*
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| HEACTOR Page 2 I I l 1
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| 1 ,
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| FtJLL FLOW TEST LhE S;.mysre Hureym1 CTFE.R WALVES NOT WNWUW FLOW SVPASS etJECTE.)N WALo'E E22- OPEN OPEN FALS FOD3B FALS TO OPEN I [ .,e u l [ *~, . . m l
| |
| -e ww m9 Page B 1 ME43 Page 5 I
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| f 1 I BYFASS LAE OPENS BYPASS LWWE F ALS LOF OF P ATER LEVEL VALVE E22FOQ3B AND FALS TO CLO6E TO OPEN 8 SENSORS F A3.S CLOSED MC-u m''m ERATED M VALwE ECCY ST'@
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| CN) 3 a6 (E*u 1 I-w. sea 1 dm i en'i ' l f 2 0N46 164E43 Page 7 NO HG4 FLOW SmAL h%e E22Fonos F ALS TO GOSE -
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| I -- 5 m 4 1. ws o en s [
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| g .. 9 --
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| 3 l 4 I s i 1 2 1 t ,
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| .\SDTREE\HPCF.CAF 6-29-92 Page 4 HIGH PRESSURE CORE FLOODING SYSTEM
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| I t
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| 1 4
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| i i =
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| OTER VALVES NOT OPEN
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| \
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| Page 4 i I I MANUAL GATE VALVE PUMP DISCHAAGE TESTABLE CHECK E22f005B NOT OPEN CHECK VALVE E22- VALVE E22-ForAB FC21B FAJLS TO CPEN FALS TO OPEN l
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| I l H8 s I ecv2iiw+ 1 l ncvm l 1 eaE-04 1soE &
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| v=vE E22ae. vuvE En=.
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| 'l M:SPOSIT80ED FALS TO REMAN OPEN N
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| Inesu e sl I =vw I thcEct t 30Ess I i i 2 I a l HIGH PRESSURE CORE FLOODING SYSTEM .\SDTREE\HPCF.CAF 6-29-92 -Page 5
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| | |
| l NO HGH FLOW SGNAL I I FLOW XRTTER E22- MtSCALIBRATON OF FT008&2 FALS LOW FLOW TRANSMITTERS
| |
| <cco i wt.we 1 i weaww i 2 90E-04 5 00Ec5 l
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| 3 1 2 I 1
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| .\SDTREE\HPCF.CAF 6-29-92 Page 6 HIGH PRESSURE CORE FLOODING SYSTEM
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| i 4
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| l i
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| )
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| 1 i
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| BYPASS LPE FALS 1
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| TO OPEN i
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| Pag. 4
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| ~
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| , i VALVE EWO10B NO HIGH PUMP FALS TO OPEN OtSCHARGE f% ESSURE S=
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| l iwv-i eno i f T 1.7oE.c3
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| ! U -
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| PPESSURE XMITTER ESCAUBRATION OF En8'Too78 FAAS LOW PRESSURE XuTTERS (CCF) a i -,r i i -o i s20E 4 5 00Ee i
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| i j
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| I i i I HIGH PRESSURE CORE FLOODING SYSTEM .\SDTREE\HPCF.CAF 6-29-92 Page 7
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| nu rum TtST uNE OPEN
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| -. +
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| t i TEST VALVE E22- TEST VALVE E22-F0088 OPEN F00GB OPEN j
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| HB as HB ,9 I
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| 1 4
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| . . i i
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| ; TEST VALVE E22- VALVE E22+0088 TEST VALVE E22- VALVE E22f0098 F00e9 NOT et mm FALS OPEN (NCFO) F00GB NOT CLOSED FAILS O*EN (NCFO) j
| |
| + ,. io ~i e i Hm- i 1.75EM i 75E-04 ll , ,
| |
| TEST VALVE E22- TEST VALVE E22- TEST VALVE E22- TEST VALVE E22-F0089 REMANS OPEN F0088 IS OPEN F0098 REMANS OPEN F000B IS CPEN 1
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| 1 ADS F AIL LAE DUE TO COMuGN CAUSE F AILuRES I i
| |
| [R' FS 1
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| ~
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| Page 1
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| ,~,
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| I I
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| Common CauSE CCF SvSTEM LOGIC F AILURE of SAv5 un!T (AILS I A Ai fj D ] O_.; f tJL) 2 %E - 0 7
| |
| ) 5 00E-CE, s
| |
| 1 I _
| |
| CCF Of inaNSulSSION CCF Of DEMa1E NE imON WJL i IPL E x ING UNI T S I ccr wo = 1 57,&rTG
| |
| , 3.9CE-07 m
| |
| } 2 %E -07
| |
| - 1 2 1 _-, . . . . .
| |
| 1 ~
| |
| FINAL ADS INPUT TO PRA E: 427 FINAL CAF 4 - 2 7 - 9El_J Pa g e ;> l
| |
| - x
| |
| | |
| l i
| |
| Gate / Event Name PE t Zone Gate / Event tJame P_;tg g Zo n g Gat e /E v er.t tJame Pan t ZQr3e Gateffy$ k th$n[ } Age ((gng ADSMAN 2 3 DIV3 MUX 18 1 EQUIPS 6 1 ADSMAt4 3 1 DIV3 MUX 19 1 EQUIP 9 5 6 :!
| |
| ADSMAta 4 1 DIV3 MUX 20 1 EQUIP 9 7 1
| |
| , ADSMAN 5 1 DIV.3 mux 20 .3 EQUIPA 1 1 i ADSMAN 5 3 EBY1CCF 2 1 EQUIPA 2 4 ADSMAF4 6 1 E8Y1CCF 2 5 .EOUIPB 1 1 j
| |
| ; ADSMAN 7 1 E8Y1CCF 5 6 EQUIP 8 5 4 j ADSMAN 8 1 ECA002H 5 7 EQUIPC 1 1 ADSMAra 8 3 ELDDP1 2 2 EQUIPC 8 4 .
| |
| ) ADSMAt4 to 1 ELDOP12 2 6 EQUIPD 1 2 '
| |
| ADSMAN 11 ! EQUIP 1 1 EQUIPD 12 4 J
| |
| ADSMAN 12 1 EQUIP 1 2 2 EQUIPE 1 2
| |
| ] EQUIPZ 16 4 a ADSMArd 12 3 EQUIP 10 5 6 j ADSMAt4 13 1 EQUIP 11 8 2 EQUIPF 1 2
| |
| ! ADSMAN 14 1 EQUIP 12 8 4 EQUIPF 20 2 i ADSMAN 15 1 EQUIP 13 8 5 ILCCCFH 21 2 1 ADSMAtt 16 1 EDUIP13 9 1 ILEPVCH 2 4 ANTMDivt 8 2 EQUIP 14 8 6 IRMU11 7 2 l IRMU11 10
| |
| { At4TMD1VL 13 1 EQUIP 14 to 1 1 i ANTMD1VL 14 1 EQUIP 15 8 6 IRMU11 12 4 j 11 IRMU11 1S 2 .;
| |
| At4T MO2VL 8 4 EQUIP 15 1 At4TMD2VL 15 1 EQUIP 16 12 2 IRMust 20 4
| |
| , at4TMD2VL 16 2 EQUIP 17 12 4 IRMU12 5 2 l 1
| |
| ARvCCFD 21 1 EQUIP 18 i 12 5 IRMU12 10 2 j CCFDTM 4 2 EQUIP 18 13 1 IRMU12 12 2 CCFMUX 21 1 EQUIP 19 12 6 IRM012 13 2 j ;
| |
| i CCFS 1 2 EQUIP 19 14 1 IRMU12 18 2
| |
| . CCFS 21 1 EQUIP 2 2 .4 IRMU13 16 4 1 CCFTLU 21 2 EQUIP 2O 12 6 IRMU21 5 4 DIv1 MUX 3 1 EQUIP 2O 15 1 IRMU21 11 1 !
| |
| ! DIV1 MUX 5 2 EQUIP 21 16 2 IRMU21 12 2 j' DIV1 MUX 6 1 EQUIP 22 16 4 IRMU21 16 2 'f i DIv1 MUX 8 4 EQUIP 23 16 5 IRMU21 20 2 4
| |
| DIV1 MUX 9 1 EQUIP 23 17 1 IRMU22 6 2 2 I DIV1 mux 16 3 EQUIP 24 16 6 IRMU22 11 l EQUIP 24 18 IRMU22 12 4
| |
| : DIV1 MUX 17 1 1 l j DIV 1Mu x- 18 1 EQUIP 25 16 6 IRMU22 14 2 l
| |
| ! DIV1 MUX 19 1 EQUIP 25 19 1 IRMU22 19 2 '
| |
| I
| |
| : DIV2Hux 3 2 EQUIP 26 20 1 IRMU23 17 2 l DIv2 MUX 5 4 EQUIP 27 20 3 AxDPRESS 1 2 .!
| |
| DIV2 MUX 7 1 . EQUIP 3 2 5 .'
| |
| 4 DIV2 MUX B 2 EQUIP 4 2 6 DIV2 MUX 9 1 EQUIP 4 3 1 2 7 i DIv2 Mux 16 4 EQUIPS l DIV2 MUX 17 1 EQUIPS 4 1 ;
| |
| ~
| |
| DIY2 MUX 20 1 EQUIP 6 5 2
| |
| : i. DIV2 MUX 20 3 EQUIP / 5 4 l DIV3 MUX 9 2 EQUIPS 5" 5 l - FINAL ADS-INPUT TO PRA E: 427 FINAL.CAF 4-27-92 Page 22 4 4 1
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| ,-. _ _ _ _ _ _ _ _ _ __ . ______________1_______.
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| l t
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| ALL DESEL GENERATORS FAL i i
| |
| +
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| i i FALURE 0' OlY B W2 TO WaE ECCS CDOLWG lw+ cang I I
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| i PuuP ca REFRGERATOR FAILURE Pags 5 I I i PUMP FAEfURE I l REFRIGERATOR FALURC I
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| |
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| ; FAtune OF orv c HECW TO PROV10E ECCS CDOLING In+;wt l HECW C FAILS TO ' LOSS OF COOLNG TO LOSS OF COOLNG TO l PROVfDE CHt. LED ESTENTIAL ELEC DIESEL GENCDATOR 1 WAMR EOLAP ROOM C ROOU C Indwic i IvF_t> w i I vt *.; I Fage 14 Page 15 I
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| FALURE OF DIV! SON b
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| t l va wm; g I vw u + 1 I vmma: I i v w m; 1 Page 11 Page 12 22CEM 2 eOEt:7 r
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| .\SDTREE\HVAC.CAF 6-29-92 Page 12 HVAC EMERGENCY COOLING WATER SYSTEM
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| HVAC EMERGENCY COOLING WATER SYSTEM .\SDTREE\HVAC.CAF 6-29-92 Page 15
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| Gate / Event Name .. Pace Zcne Gate / Event Name Pace Zone Gate / Event Name Pace Zone Gate / Event Name Pace Zcng EACU 1 1 2 VPUMPC 10 3 WPPHEBH 5 4 EAC6D1 5 2 VPUMPC 11 '2 WPPHECH 10 4 ERC6El 10 2 VPUMPE 5 WPRA 1 3 EMSCONN3 12 2 VPUMPE 7 2 WPF3 5 3 VBYPASSA 1 5 VPUPPF 10 4 WPRC 10 3 VBYPASSB 5 5 VPUMFr 12 2 WPv025A 2 2 VBYPASSC 10 5 VPVF012A 1 5 WPV025B 6 2 VDGA 1 5 VPVF012B 5 5 WPV025C 11 2 VDGA 4 2 VPVF012C 10 5 WPV025E 7 2 VDGB 5 5 VTE005A 1 6 WPV025F 13 1 VDGB 9 2 VTE005B 5 6 WRCWA 1 1 VDGC 10 5 VTE005C 10 6 WRCWB 5 1 VDGC 15 2 WCVH1F 12 2 WRCWC 10 1 VELECA 1 4 WHECWlA 1 3 WRFD1A 2 2 VELECA 3 2 WHECM1B 5 3 WRFD1B 6 2 VELECB 5 4 WHECW1C IO 3 WRFDIC 11 2 VELECB 8 2 WHECWA 1 4 WRFDlE 7 2 VELECC 10 4 WHECWB 5 4 WRFDlF 13 1 VELECC 14 2 WHECWC 10 4 WRFMNTF 13 ?
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| VF015A 3 1 WHSPA 2 1 WTE052 4 ?
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| VF016B 8 1 WHSPB 6 1 WTE056 9 2 VF016C 14 1 WHSPC 11 1 WTE060 15 2 VF022A 4 1 WHSPE 7 1 WTE113A 3 2 VF022B 9 1 WHSPF 12 2 idTE113B 3 2 VF022C 15 1 WHSRA 2 2 WTE113C 14 2 VOPERRF 12 1 WHSRB 6 2 VPR007A 1 5 WHSRC 11 2 VPR007B 5 5 WHSRE 7 2 VPR007C 10 5 WHSR" 12 3 VPRXXXA 2 3 WHSRF 13 2 VPRXXXB 6 3 WPMHC1A 2 1 VPRXXXC 11 3 WPMHC1B 6 1 VPRXXXE 7 3 WPMHC1C 11 1 VPRXXXF 13 2 WPMHC1E 7 1 VPUMPA 1 3 WPMHC1F 12 1 VPUMPA 2 2 WPMMUTF 12 2 VPUMPB 5 3 WPMSTRTF 12 1 VPUMPB 6 2 WPPHEAH 1 4 HVAC EMERGENCY COOLING WATER SYSTEM .\SDTREE\HVAC.CAF 6-29-92 'Page 16
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| DW PRESSURE StGNAL FOR Dtv 1 FAILS ItDWPAj
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| I _.wwA_) Page e I iDwmia ? Page 7 O
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| I'Dw d I I W0i" I 3 :cwet i I CNC iH I 1.77E45 42SE47 Page 6 Page 5 4 l 5 !
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| l 3 l l 2 l t
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| .\SDTREE\lNST.CAF 6-29-92 Page 1 l INSTRUMENTATION SYSTEM
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| 4
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| h MANu% TNITMTEDN
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| $GNAL FAE.S EusMwI Pogo 23 Pege 43 a a a u E i CCF OF REMOTE 88GNAL FROM FIRST SIGINAL FROM SECOND CCF SYSTEM LOGIC CCF OF TRANSamSSON FAILUAE OF DWTSPON l MULTIPLEXNG UNITS RMU
| |
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| 2 esE47 Page 4 2 95E47 5 00E4?
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| l INSTRUMENTATION SYSTEM .\SDTREE\lNST.CAF 6-29-92 Page 2
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| BYPASS FAILS DMSON1
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| , msrya
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| .d SLU 2 FALS DIV ! BYPASS UN'T Of/ISON 1 FALS l
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| l tstu21 l IBTP1 2.95E-04 1 e4E-03 2
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| CCF SLU BYPASS UNIT l CC[ eve i 183E-06 I i 1 2 1 INSTRUMENTATION SYSTEM .\SDTREE\lNST.CAF 6-29-92 Page 3
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| SIGNAL FBCA8 SECOND RMU FALS l tN31B1 ]
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| I I i _.
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| I 2ND ESF fWU OfV 1 OlV 1 TRANSMISSON SIGNAL FROhn TCONO , SLUIMS t. INK FOR FALS NETWORK FAILURE Stil FAILS DfV 1 SLU 2 FALS (EMS) 1 8 amu 21 I Q viuux IINSTD1 l l BUNU1 l 2.95E44 - # #
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| 6 I I SLU 2 FAL9 DIV 1 BYPAS FAILS DIVISION 1 l IStu21 ] l 'NSTF i l 2 95E-04 i i SLU 1 FAILS CAV 1 BYPASS UNIT OfVISION 1 FAILS I sn o;i l l evei l 2.9"E44 184E-03 CCF SLU BYPASS UNIT iCCFBYPl 1.83E-06 I i 1 2 1 3 1 4 I INSTRUMENTATION SYSTEM .\SDTREE\lNST.CAF 6-29-92 Page 4
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| DN 1 DW PRESStf4 SONM. FARURE I
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| LwP t Page 11 Page 17 -=
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| DN 1 REMOTE DW PRESSURE SENSOR MULTIPLEXNG ONfT FMLURE B21-PT025A FALS l uwuowPt l l iPktwis i 2.95E4+ 120E45 DN 1 TRANSMtSS,0N FAtuRE OF DN's NETWORK FALURE 1125 V DC BUS (EMS) T q
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| l Otytwux l l EDI.11_l 5.00E44 OtGITAL TRtP MODULE CCF OF OtGfTAL TRIP FAlt3 DNSION 1 UNITS ivist l ccd >TM 2 97E44 2 95E47 4
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| EMSOTW UNK FOR DW PRESSURE SENSORS DNISON 1 FAES CCF i ,t i. . , i i ,~ma i 4.25E47 2.40E4 DN i INSTRUMENT LWE BREAK
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| ] ttnavia 1 2.05E43 l -1 1 2 I INSTRUMENTATION SYSTEM .\SDTRE3 INST.CAF 6-29-92 Page 5 a_a. i. ..+m_.
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| DM 2 DW PRESS 11RE SGNAL FAIUJRE ilmP2 Page 11 Page f7 =
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| I I DM 2 REMOTE DW PRESSURE SENSOR MULTIPLEXING UNfT FAfLURE B?1-PT0258 FA19 I isuA >wr/ I I ptwm l 195E44 17]E-05
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| [
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| DfV 2 TRANSutSSON FAILUT OF OfV1 SON NETWORK FALURE 2125 V DC BUS (EMS) T I Otv2wux l IIoc o I 5.00E44 DGITAL TR8P MODULE CCF OF DIGITAL TRIP FAILS DNtSON 2 UNITS suiW J ct
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| * o t w l 4
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| 2.97E& 2.95E47
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| . I EMSOTM LINK FOR DW PRES 3URE SENSORS GIVtSON 2 FAILS CCF E 1 4 25E47 2 40F-06 DfV !! INSTRUMENT LINE BREAK l et><C7H I 2 05E43 I i I 2 I INSTRUMENTATION SYSTEM .\SDTREE\lNST.CAF 6-29-92 Page 6
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| l l 4 OlV 3 DW PRESSURE
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| . PUT SIGNAL FAILtJRE I KwvPon j
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| ~
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| I I Div 3 DN PRESSURE l OlV 3 TO 1 DTM TO SIGNAL FALURE SLU TRANSM!SSION I
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| F AILS l rowes 1 I nmm l Pa0s 8 1.77EM
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| ! 2 I i i
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| .\SDTREE\!NST.CAF 6-29-92 Page 7 INSTRUMENTATION SYSTEM
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| ' - - ;, a _
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| DV 3 DW PRESSURE SONAL FA:'URE s I CwP3 i 7
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| Page 15 Page 21 .
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| -m I I DM 3 REMOTE DW PRESSURE SENSOR UULTIPLEXNG UNIT FAILtPE B21-PT025C FALS QdwF3 [ WHLW 1H i 2.95E44 '. M 45 1
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| Div 3 TRANSMtSSION Frill URE OF DMSf0N NETWORK FALURE 3125 V DC BUS (E--S) T l DivAvux-l l flot3 I 5.00044 I I DIGITAL TRIP MODULE l CCF OF DiGITA+ TRIP FAILS DretS10N 3 UNITS 1 l 2.TE44 s
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| f2.sE47 EMSOTM UNK FOR DW FRESSURE SENSORS DMSION 3 FALS CCF l
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| I ti m4 3 I I si-Ht wcs '
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| 4.2SE47 2.40E 4!5 D:V 111 INSTFIUMENT LhE BREAK I st !wo 3H I 2,05E43 I i i 2 I INSTRUMENTATION SYSTEM .\SDTREE\!NST.CAF 6-29-92 Page 8 l
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| e l OtV 4 DW PfL*SURE
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| ! %T'UT SKANAL FAILURE
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| \
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| l PwPWA f
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| ^
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| I- __
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| I OfV 4 DW KtESSURE ' ON 4 TO 1 DTM TO SIGNAL FAILURE SLO THANSMtSSION FAILS l IDWP4 l FtM$41H I Page to 1.77E M l
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| ! 1 1 2 I lh STRUMENTATION SYSTEM l .\SDTREE\lNST.CAF 6-29-92 Page 9
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| I I I DV 1 RPV LEVEL DTM DN 2 RPV LEVEL UTM Dry 3 RPV LENEL DTM DV 4 R'V LEVEL DTV SONM. FALURE SIGNAL FALURE Sb3NAL FARURE SCPeAL FALUME l (vi13 l I r_vt23 I I (vi 33 I I *t vi 4 3 I.
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| i l 1 'l 2 l 3 'l 4 l 5 l INSTRUMENTATION SYSTEM .\SDTREE\lNST.CAF 6-29-92 Page 37 i
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| , _I i i 1 INSTRUMEivTATION SYSTEM .\SDTREE\!NST.CAF 6-29-92 Page 39
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| I , 1 2 I INSTRUMENTATION SYSTEM .\SDTREE\lNST.CAF 6-29-92 Page 42
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| y i s RPV PRESSURE SKDdA5 FAILURE PEPA I i i
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| INSTRUMENTATION ' SYSTEM .\SDTREE\lNST.CAF 6-29-92 Page 44 J n - - - - - - - - - - - . _ . .------ . - . - - -_. _-___ _ _ _ _ . _ _ _ _
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| 5 OfW 2 REMOTE RPV PRESSURE SENSOR MULTPLExte UNIT FAILURE 821-PT007B FALS.
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| 1-4 RPV PRESSURE SMAL FOR DIY 3 FAR.S inv'uc I E SGMAL FALURE MANUAL Wr."AT m l SO4AL FALS Itw w *eol I WNic I W '8 3
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| mv &- I.w/w 1 Page 44 117Em Page 45 LTTE u i t i 2 1 3 I 4 I s i INSTRUMENTATION SYSTEM .\SDTREE\lNST.CAF 6-29-92 Page 53
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| Gate / Event'Name': Eggg Zone .. Gate / Event Name Page' Zone Gate / Event Name Pace Zone Gate / Event Name Page Zone CCFBYP r '3. '2 .DIV2 MUX".
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| . - 40 1' IDNP4 ?.6 1 CCFDTM 10 2~ DIv3 MUX' -27 l' EMSCO!CI3 53 1 IDWP4 22- 1
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| , CCFDIM 24 2~ DIV3 MUX 34 1 iIBYP1' 3 2 IDWPA 1 '2 CCFUTM- 25 2' DIV3 MUX ~47 -1 IBYP1- 4 4 IDWPB 11' 2
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| , CCFDTM- 27' 2 DIV4 MUX 10 .1 IBYP2 13 2 IDNPC 17 2 l CCFDrM - 29 '2 1DIV4 MUX 29 1 IBYP2 14 4 ~IDWPI1A 1 2 CCFDTM 31' 2 DIV4 MUX 36 <1 IBYP3 19 2 IDNPI1B 11 2 CCFDTM: 32 2 DIV4 MUX -- 4 9 1 IBYP3 20 4 IDWPI1C 17 2 CCFDTM 34 2 EDC11- 2 7 IDTM1 5 1 IDNPI2A 1 4 CCFUrM 36 2 EDC11 5- 2- IDTM1 24 1 IDWPI2B 11 4 CCFUrM 44 2 EDC11 24 2 IDTM1 31 1 IDWPI2C 17 4 CCFDTM 45 2 EDC11 31- 2 IDTM1 44 1 IDWPI3A 1 5 CCFDTM 47 2- EDC11 44 2 IDTM2 6 1 IDWPI3A 7 2
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| , CCFDTM 49 2. EDC12. 6 2 IDTM2 25 1 IDWPI3B 11 5 CCFMUX 2 6 EDC12' 12 7 IDTM2 32 1 IDWPI3B 15 2 CCFMUX- 12 6 EDC12 25 2 IUrM2 45 1 IDWPI3C 17 5 CCFMUX 18 6 EDC12 32 2 IDTM3 8 1 IDhPI3C 21 2 CCFS3A 44 2 EDC12 45 -2 IDTM3 27 1 IDWPI4A 1 6 CCFS3A 45 2 EDC13 8 2 IDTM3 34 1 IDWPI4A 9 2
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| ) CCFS3A 47 '2 EDC13 18 7 IDTM3 47 1 IDWPI4B 11 6 CCFS3A 49 2 EDC13 27- 2 IDTM4 10 1 IDWPI4B 16 2 CCFT:- 3 2 4 EDC13 34 2 IDTM4 29 1 IDNPI4C '47 6 CCFTLU 12 4 EDC13 47 2 IDTM4 36 1 IDWPIfC 22 2 CCFTLU 18 4 EDC14 10 2 IDTM4 49 1 IDWPTA 1 4 DIV1 MUX 2 2 EDC14 29 2 IDWP1 1 1 IDWPTB 11 4 I DIV1MU2 4 2 EDC14 36 2 IDWP1 5 2 IDWPTC 17 4 DIV1MUF 5 1 EDC14 49 2 IDWP1 11 1 IIN012H 11 2 5-DIV1 MUX 24 1 EMSCONN1 1 1 IDNP1 17 1 IIN012H 30 2 DIV1 MUX 31 1 EMSCONN1 2 4 IDWP2 1 3. IIN012H 50 2 DIV1 MUX 44 1 EMSCONN1 23 1 IDWP2 6 2 IIN013H 17 2 i ,
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| INSTRUMENTATION SYSTEM .\SDTREE\lNST.CAF 6-29-92 .Page 56
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| h.
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| }l Pace Zone Gate / Event Name Pace Zone' Gate / Event Name Page Zone Gate / Event Name Pace Zone . Gate /EvenF Name 47 1 -ILvL41 28 2 L37- 2 ILE022H - 32 2. ILnE43 IIN013H '
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| :. 2 ' . ILnE44 10 1 ILVL42- 30 6 IIN013H 40 . 2- ILE023H '34 l -
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| 29 1 ILVL42 35 2 2 ILE024H 36' 2. ILDIK44 IIN013H: 53~
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| 36 1 ILVL43 37 ~6 4 ILEPVCH 24 2 ILINK44 IINO21H- L1 49 1 ILVL43 39 .2 IIN021H 23 4 ILEPVCH; 25 2 'ILUIK44' 5 2 ILVL44 40 6 4 ILEPVCH 27- -2 ILNOD1H IIN021H1 43 6 2 'ILVL44 ' 42' .2 17 4 ILEPVCH 29 2. ILNOD2H IINO23H 8 2 INSTA1 2 3
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| :37 4 ILEPVCH 31 -2. IuiOD3H IINO23H 10 2 niSTA2 12 3 IIN023H- 40 4 ILEPVCH. 32 '2 JILN0D4H 24 2 INSTA3 18 '3 53 4 ILEPVCH 34 ~2 ILNOV1H IINO23H 31 '2 INSTAL 2 4 IINO31H 7 ,. :2 ILEPVCH 36 2 Iui0v1H 44 7 INSTB1 4 2 26 2 ILDIK11 -- 2 4 ILNOVIH ,
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| IINO31H 25 2 INSTB2 12 4 46 2 ILUE1DI 1 2 "ILNOV2H IINO31H 32 2 INSTB2 14 2 15 2 ILDE11H 23 2 ILNOV2H IIN032H 45 2 INSTB3 18 4 33 2 . ILnallH 43 2 ILNOV2H IINO32H 27 2 UISTB3 20 2 51- 2 ILnE12 12 '4 ILtt0V3H IINO32H 34 2 INSTC1 2 3
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| :9 2 ILDE13 18 4 ILNOV3H IIN041H 47 2 INSTC2 12 3 28 2- ILINK21 4 4 ILNOV3H IIN041H 29 2 INSTC3 18 3 48 2 ILDIK22 14 4 ILNOV4H IIN041H 36 2 INSTD1 4 3 16 2 ILUE22H 11 4 ILNOV4H IIN042H 49 2 UISTD2 14 3 35 2 .ILINK22H 30 4 IuiOV4H IIN042H 23 2 INSTD3 20 3 52 2 ILINK22H 50 4 ILvL11 IIN047H 30 2 INSTE1 2 4 22 2 ILDIK23 20 4 ILVL12 IIN043H 37 2 INSTE1 3 2 39 2 ILUiK33H 21 2 ILVL13 IIN043H 40 2 niSTE2 12 4 42 2 ILINK33H '38 2 ILVL14 IIN043H 23 4 HISTE2 13 2 55 2- ILUJK33H 41 2 ILVL21 IIN043H 30 4 INSTE3 18 4 44 1 ILINK33H 54 2 ILVL22 ILC001H 37 4 INSTE3 19 2 ILC002H 45 1 ILUiK41 5 1 .ILVL23 4 3 24 1 ILVL24 40 4 INSTF1 ILC003H 47 1 ILDIK41 ,
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| 14 3 31 1 ILVL31 23 5 INSTF2 ILC004H 49 1 ILDE41 2 INSTF3 20 3 44 1 ILVL31 26 ILCCCFH 2 1 .ILlNK41 23 1 6 1 ILVL32 30 5 IPLvil ILCCCFH 12 1 ILUIK42- 24 2 25 1 ILVL32 33 2 IPLV11 ILC'TFH 18 1 ILnE42 IPLV11 37 1 32 1 ILVL33 37 5 ILE011H 24 2 ILDIK42 IPLV12 23 3 45 1 ILVL33 '38 2 ILE012H 25 2 ILDIK42 IPLV12 25 2 8 1 ILVL34 40 5 ILE013H 27 2 ILINK43 2 IPLV12 37 3 27 1 ILvL34 41 ILE014H 29 2- ILDiK43 6 IPLV13 26 1 34 1 ILVL41 23 ILE021H 31 2 ILHJK43
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| .\SDTREE\lNST.CAF 6-29-92 Page 57 INSTRUMENTATION SYSTEM
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| 38 -l - IPVP2 - 50 13 "IRMU22- 14" 1
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| , IPLV14 _ 2 8 .- l' IPVP2- 53 3 -.IRMU23 20 1 IPLV14 29 i2 - IPVP3~ '' 46 -1 IRMUWP1 5 1 IPLV14- . 1391 l' IPVP3' 47 2' IRMUWP2- 6 1 IPLV21~ 30 1 IPVP3 51 1 IRMUUNP3 8 1 IPLV21 :31 2 . IPVP3 - 54 1- IRMUDWP4" 10 1 IPLv21 40 -1 .IPVP4; ' 48 :14!IRMULV13 24. 1
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| 'IPLV221 -30 3, 'IPVP4 49- 2.,IRMULv12- 25 1 IPLV22 32 .2 :IPVP4 52 .17 ,IRMULV13 27 1 IPLV22 .40 '3' IPvP4 55 1 IRMULV14- 29 1
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| .IPLV23- 33 . l' IFVPA- 43 2.. IRMULv21 31 1 IPLV23' 34 L2 .IPVPB 50 2 IRMULV22 -32 1 IPLV23 41 1" IFVPC 53 IRMULV23 34 s '2 1 IPLV24' :35 1 IFVPI1A 43 2 IRMULV24 36 1 IPLV24: 36 2 IPVPI1B, 50- -2 ISLU11 2 3 IPLV24 -42 1 IFVPI1C' 53 2 ISLU11 4 3
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| -IPRDW1H. 5. 2 IFVPI2A 43 4 ISLU12 12 3 y' IPRDW2H. 6 2 IPVPI2B 50 4 ISLU12 14 3 l;
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| IPRDW3H 8 2 IFVPI2C- 53 4 ISLU13 18 3
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| -IPRDW4H 10 2 IPvPI3A: 43 5 ISLU13 20 3
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| . IPRDWCH 5 2 IPVPI3A 46 .2 ISLU21 3 1 j IFRDWCH 6 2 IPVPI3B 50 5L ISLU21 4 2 i IFRDWCH ,8 2 IPVPI3B' 51 2 ISLU22 13 1 I- IPRDNCH 10 2 IPVPI3C 53 5 ISLU22 14 2
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| ; IPVL1A 23 2 IPVPI3C 54 2 ISLU23 19 1 i IPVLIC 37 2 IPVPI4A 43 6: ISLU23- 20 2 IPVL2B 30 2 IFVPI4A 48 '2 PPP101 44 2 IPVL2C 40 2 IFVPI4B 50 6 PPP102 45 2 1 I IPVLTIA 23 4 IPVPI4B 52 .2 PPP103 47 2 IPVLT1C 37 4 IFVPI4C 53 6 PPP104 49 2 j IPVLT2B 30 4 IFVPI4C -' 55 2
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| ; IPVLT2C 40 4 IPVPTA 43 4 L IPVP1 43 1 IFVPT3 50 4 i IPVP1 44 2 IPVPTC 53 4 I IPVP1 50 1 IRMU11 2 1 IPVP1 53 1 IRMU12 12 1 j- IPVP2 -43 3 IRMU13 - 18 1 1
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| ; INSTRUMENTATION SYSTEM .\SDTREE\lNST.CAF 6-29-92 Page 58 n
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| LOSS OF DIV 1480V MCC POWER FOR HECN REFRIGERATORS) l EACEi l
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| S.76E 4 1 i I 1 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 4
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| LOSS OF OfV I 480V MCC POWER (FOR SERYtCE WATER)
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| % 26 Page 29 a i I FAILURE OF DIVISION BATTERY CABLE I DtSTRIBUTION PANEL FAILURE I Ece'ioiH I I EcAtsiH I 5.76E-Os 1.16E44 I I DIODE SO OPEN BATTERY FAILURE IEDE0A1Hl l E8v101H I 1.43Eas 2 53E44 BATTERY OFfT BKR BATTEAY CCF OPE IECB1BtHI I EBf1CCF I 3.2?E47 253E46 I i I = 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 25
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| D.G. PCMER NOT AVARABLE g
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| IEAcoGcl I I COMBINATION OF DG D.G. BKR 905 FAILS CCF YO CLOSE IEDtdCCFCi I Ecem i Page 27 1.00E46 I I OfV 1 DC CONmOL DG. C CONTROL I POWER LOST FAILURE I Eociix l R E LCLX4H l Page 25 475E45
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| , 1 D.G. C CABLE FALURE D.G. C FALS TO RUN lECAUCCHl E N+C 1.16E-04 4 80E42 2
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| I I D.G. FAILS TO START 125V DC OtSTRBt.TTON PANEL BKR D12 OPEN I ELx4 5CD I l Ectot2H I 2.50E-02 327E47 I i i 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30 Page'26
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| .1 FAwnE tw 1eOWER l EIFNuiH I IELOOP1l 324Ev5 1.0cE42 1 2 i =
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| 1 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 28
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| I I' DIV I DC CONTROL ST N TRANSF N POWER LOST , #2., FAILURE .,
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| 4 IEDU11X] I ETRSLW ] -
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| Pags 25 - 3 24E45 a n
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| - LOSS OF OFF-SITE - BKR CONTROL LOCHC '
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| * - LNE #2 FALURE .--
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| 'IELdOP2I. I ECL902H I
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| ' 1.00E-02 . 4.75E45 -
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| . I i stCOMING BKR 902 l OPERATOR FAILS TO FALS TO Ct.OSE - TRANSFER POWER
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| [ EC8902H I '. I EHueeC I 1.00E-06 ' 1.00E43 I I ;
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| CABLE 7.11. OR 12 . DfV.1 DC POWER BKR FAILURE ' 1 FALS OPEN- ,
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| I ECJb11H I IEC85DCHI ueEm sus 47 1 ,- l 2 l-
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| ' ELECTRIC ' POWER SYSTEM - .
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| .\SDTREE\ POWER.CAFl 6-30-92' Page 129 '.-
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| ' il POWER - . MCC POWER i
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| 12 W
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| .. 0 .
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| I EDCS12Y I I EDEoB1H I Page 31 1.43E os FAIUURE OF DMSION 11 DISTRSUTION PANEL i
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| - I EDP102H [
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| I STANOBY CHARGER OfV LOSS OF NORMAI ' LOSS OF BATTERY' l-H lKTT AVAR.ABLE CHARGER POWER .: ;
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| i'
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| ' I EBCS128Y I - l EBCN12Y I . M I B' JJrPtJT TRANSFER SKR STAND 8Y CHARGER H NORMAL CHARGER .FALURE FAILS TO CLOSE
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| . ELECTRIC: POWERJ SYSTEM .\SDTREE\ POWER.CAFL ' 6-30-92 ~ .Page 31 !
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| 4
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| LOSS OF NORMAL CHARGER I EBCNt2Y l I I 400V FEEDER BKR 425, 480V FEEDER CABLE OPEN 48 FAR.DRE I ECtE2SH l l ECAoseH ]
| |
| 3rE-07 1.16E44 5 I CHARGER FAR.URE CHARGER OUTPUT -i CABLE FAILURE l
| |
| _ i EscNt2H I I EcANi2H I 102E-04 1.16E44 I i I 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 32 ]
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| |
| * 490V MCC CABLE 41 FALURE SWGR PC 60 8 FEED DfV W 480V MCC LOSS OF 480V SWOR PC 631 POWER BLM FALURE BKR 354 OPEN INCOMING BKR 420 OPEN IEACAOil1 M IECA2aiHI IE(N9HI I ELEr* I E78646 1.1eE44 327E47 327E47 LOSS OF PC 6D4 SWW4PC S&1 POWER SOURCE FALURE IEAC5601Zl IER NO1H I s.76e46 I I LOSS OF 6.9 KV ON TRANSFORMER 1801 N SWGR MC 80 POWER fALURE IiM64)i I E meO1H I P as 324Eas PC 60 t INCOMING 6.9 KV BKR 931 OPEN BKR 341 OPEN l ECNais I I EC[d9i.41 S.27E47 127E47 CABLE 23 FALURE lECA'w e l 1.16E44 s I 4 I s I e I I i 1 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 34
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| i toss oF e.s xv Div fl SWGR BAC 60 POWER I EAc690 l LOSS OF POWER FAILURE OF OtV B SOURCES ~ HECW TO PROVIDE
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| - ECCS COOUNG T
| |
| I EAC64SD l [vo*CWBl b b DivtS80N 11 SWGR FAILURE IEAc69t*41 5.76E-06 I i i I
| |
| ' ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 35
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| LOSS OF POWER SOURCES
| |
| , i EAC6950 l i i t.1AS TURBINE - DA POWER NOT LOSS OF OFF-SITE GENERATOR POWER NOT AVAILABLE POWER AVARABLE l Gi([RBO l lEACOGDj i EActOSO I Page 38 Page 40 I ,
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| CTG MANUAL CTG FALS TO START DISCONNECT SWITCH AND RUN OPEN ICTGMANSw] l GilmNE l 3 00E-03 5.00E-02 CTG CABLE FARDRE MANUAL INITIATION IECAbiGMi lC1GMANl 1.16E44 2 00E-C2 DfV N DC CONTROL CTG DlV 2 BREAKER POWER LOST FAILS TO CLOSE I EDC12A l l ECaCIG2 l Page 37 tooE-06 I i 1 2 1 3 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30 Page 36
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| s DIV R DC CONTROL POWER LOST l EOC12X l .
| |
| Page 35 Page 41 p-d i I FAILURE OF DMSON BATTERY CAtYE R DISTRVUTION FAILURE PANEL I EEP102H ] IECA182Hl 576E-os t.18E44 I I DIODE SD OPEN BATTERY FA4.URE l EDE081H l lEevko2H_]
| |
| 143E-06 233E44 1 1 BATTERY OLT:PUT 8801 BATTERY CCF OPEN lECB182H l l EBv 1CCF l 327E-07 2.53E46 1 4 1 2 !
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| ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 37
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| t D.G. POWER NOT AVALABLE i
| |
| IEAcouoi 1
| |
| I I
| |
| COMB! NATION & DG DG. BKR 925 FALS CCF 70 CLOSE IEoocccFoi. I Ects9294_1 Page 32 1.coE46 DN N DC CONTROL O.G. O CONTROL POWER LOST FALURE I EDC12X l l Etrv30H_J Page 37 475 es O G. D CABLE FAfu)RE DG. D FALS TO RUN .
| |
| IECADGDHl ens 0 1.tsE-os ,
| |
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| I i
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| .\SDTREE\ POWER.CAFj d-30t2 ] Fage 38 ELECTRIC POWER ' SYSTEM
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| i 1 2 D G. S CCF 3 D.G. S CCF EDbc0 i ECw.ClE I 2 80E43 5.50E44 4
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| |
| ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 39 ,
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| c g. .. -
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| |
| ..IEACLOSOI' Page 87
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| ' LOSS OF OFF-SITE f COMMON MODE LOSS OF
| |
| -ER . --
| |
| SOURCES IEAdNAD1 I EtobP121
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| |
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| f '
| |
| STARTUP TRANSFORMER LOSS OF OFF-SITE
| |
| #1 FALURE . lJNE 1 POWER i
| |
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| |
| .I 1- l 2 i ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF- 6-30-92: - Page 41 ,
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| . Page 79 f''
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| STANDBY CHARGER DIV LOSS OF NOR4tAL LOSS OF BATTERY RHV NOT AVALABLE CHARGER POWER l EBC5'#AV I l EBCN13Y I [_EBY 13 i Page 44 Fine 45 CUTPUT TRANSFER BKR ' STANDBY CHARGER IN NORMAL CHARGER FALURE FALS TO CLOSE l ECONOM ] l Ef4CS34H l 1.00E46 1.02E44 4soV FEEDER BKR 434 CHARGER OUTPUT OPEN CABLE FAR.Ur1E l EBC4M4 l l ECASICH l 3.27E-07 1.16E44 I ,
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| OPERATOR FALS TO 480V INPLTT CABLE TRANSFER STANDBY FALURE CHARGER TO Div m l EHLa31CO l l ECAD"04 l 1.00E43 1.16E-04 1 1 I e 1 3 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-:'G-92 Page 43
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| |
| 5.7eE m 1.1eE44 3J7E47 327E 07 LOSS OF PC OE-1. SWOR PC eE-1 POWER SOURCE : <FALURE IEACU6E12t i EBS6EtH I
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| . SM#
| |
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| 327E47 3.27E47 1
| |
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| |
| I E C A'26M I
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| |
| m LOSS OF 6.9 KV DN
| |
| - W SWGR RC 6E POWER
| |
| , .L I EAd6eE I 3
| |
| c 't
| |
| ' LOSS OF POWER DMSONNSWGR SOURCES FAttJRE -
| |
| +
| |
| I ErmanF l . I EEeeEH I Page de '-
| |
| S.7eE m t d
| |
| L h
| |
| i '
| |
| t
| |
| ?
| |
| i t
| |
| I !
| |
| l' i 1- 2 1-ELECTRICL POWERL SYSTEM '.\SDTREE\ POWER.CAF- 30-92 Page'47 L
| |
| ~ ,. - ..e ,, p g. v v. or 6 1 w-ti=+1e-**''s r hr " ^# N 'I+- r'- -----------2--"
| |
| | |
| ) LOSS OF POWER SOURCES Pwp 47 GAS TURBedE DA POWER HOT LOSS OF OFTGTE GENERATOR POWER NOT AVAR.ABLE POWER AVAR.ABLE I GTuneE { lEActrk l lEudatl CTG MANUAL CTG FALS TO START D!SCONFECT SWITCH APD RUN OPEN Icisuwswl I endeeE {
| |
| sIXE43 5 00E42 I CTG CABLE FALURE MANUAL NTIATON lECAcTuwl I cT<m l 1.16E44 2.ccE-02
| |
| . 2--
| |
| ou u oC COnrnal Cn ce s m.o POWER LOST e 4.5 to CU::CE IEocusl l EcSs P.~ t .
| |
| Page es 1 coEas I i I _
| |
| 2 1 3 I CI ECTHF POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 48 ,
| |
| | |
| i av m tr: CONmOL POWER LCf I HC*n I Pap 50 Pop 53 -
| |
| I I i
| |
| FAPJJRE OF DMSION BATTERY CABLE l m cesTRIBUTION FALUFIE FAtJEL IEce' m J I Ec4 ww I S.MEos 1 tsE4e 09COE SO OPEN BATTERY FAILURE IEt h m i I E s' 'c." 1 143Eas 2514e I
| |
| I BATTERY OUTTUT SKR BATTERY CCF I
| |
| cPEN I Ece sw j i Ee v ict5 I 327E47 2 53E46 i 1 I 2 i
| |
| .\SDTREE\ POWER.CAF 6-30-92 Page 49 ELECTRIC POWER SYS1EM l- -
| |
| | |
| D.G. FOWER NOT AVA8.ABLE
| |
| \
| |
| l EAct=w I
| |
| ~
| |
| I I COMBfMATICN OF DG D.G. BKR MS FALS CCF TO CLOSE lEc=AccfE1 l Eceww I Page 51 1.00E4s
| |
| , I ,
| |
| DfV fil DC CONTROL D G, E N PCWER LOST FALURE s ;tra I I t' I E tct** H 1 Page 49 4 75E45 I I D.G. E CJeLE FALURE DG. E FALS TO RUN lECAL* Awl E cmA 1.16E-04 4 00E42 2
| |
| D G. FALS TO START 125V DC DtSTRSUTION PANEL BAR D32 OPEN 1 Etp# %D 1 1ECM7LN]
| |
| 2.soE42 s m 47 1 i I 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 50
| |
| | |
| COMBNATON CF DG CCr iu xwe i i i 2 D G. S CCF 3 D G. S CCF EDOCE l Eurti l 1sE43 5.50E44 2 D G S CCF I
| |
| ELwKt 4
| |
| ! 4 1 2 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 51
| |
| | |
| toss & OrF.stTE POWER I EActost 1 Py 88 I I LOSS OF OFF-STTE COMVCN MODE LOSS OF POWER BOTH OFFSTTE POWER
| |
| *OURCES I E AcNAE_] l ELoup +2 I l 1.00E43 II ._
| |
| i ,
| |
| LOSS OF PC5%AAL l ALTERNATE FOWER NOT PREFEPRED POrc.R AVAf_ASLE IEACm % l hPM i 1 I FCOMING BKR 941 NORMAL N 2 OPEN CABLE 2,4. OR 4 FA1.URE l Ects241H I [ ECAm2w]
| |
| amor 1.2eE.
| |
| START'JP TRAPGFORMER LOSS M (VF-STTE
| |
| #1 FAR.URE LNE 1 PLMER l ElRSU1H l l EL(z.w31 l 324E45 1.oCE M i i 1 2 1 3 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 52
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| | |
| I .
| |
| q
| |
| ]
| |
| ALTENGATE POWWER NOT
| |
| ! Avau ets s
| |
| 4 I EACASEE I l
| |
| I I (W W DC CONTROL STARTUP TRANSF0FAIER 3
| |
| i POWWER 1.OST ~ 82 FAttJC 5
| |
| 3 IEDC13XI I ETRSuaH I Page 48 32eE45 4 I I
| |
| ' LOSS OF OFF-SITE I BKR CONTROL LOGC UNE 82 FALURE i k I ELOOP2 I I Ect[pe2H I 1_00E45 475E45 f WGCOMING SKR 942 OPERA FALS TO FALS TO CLOSE TRANSFER POWEFI I EC8ee2M i - IEHusecI 1.00E46 103E4X3 4
| |
| I I
| |
| ., CABLE 7.11. OR 12 OfV 3 DC POWWER SKR FALUPE 3 FA4.S OF91 I Ec511H I I EC8h 1
| |
| : 1. tee 44 - 3rE47 r
| |
| 2 i.
| |
| l 1 1 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 53 4
| |
| i
| |
| .I 4:
| |
| | |
| i GROUP i POWER FAtLS I I FALURE OF GFIOUP A LOSS OF GROUP A 32sv oc eus d e uCC FOWrR T
| |
| ' i E LNi w I iLact",I I I I I CABLE 40 FAutAE SWGR PC 6Cet FEED Dry I asev eacC LOSS OF 480V SWGR ESF OfV I esov WCC PFC 6C 1 POWER BUS FAfLUPE BKR 312 OPEN NCOUme G BFR 410 Om i E Axthe 1 IEstmNI eA*d1.16E4e I t,x e m l 32rcAT I hi*i m i 3.zttr2
| |
| : 516EM I I LOSS OF PFC 6C-1 SFA PtC 6C-1 mER sounCE rAtene l L EvC1** I l tevC WN3 Page 56 SJ6EM 1 i l I 3 1 4 I s 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 54 w
| |
| | |
| FAR.UFIE & GROU* A 125Y DC BUS
| |
| _ S.
| |
| e.a. s<
| |
| Page 86
| |
| (
| |
| pe i
| |
| I I LOSS & NT.R SOURCE DeCOE $80 N ts ecsiiv i 1 E t(ca w I Page to 1.43E44 FALUFIE OF DfV1SrJN I OtSTRIBUDON PAhEL I ELS3ioiH l 57sE4s I 1 I = I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 55
| |
| | |
| L333 OF PC 6C-1 PC4YER SOURCE
| |
| ..)IEEcmciNi LOSS OF 6 9 KV TRANSFCFtMER T6CT GROUP 1 SWGR MC 6C FALURE POWER I EAccacN l l ETRoc1HN l Pues 57 324Ec5 i i PC 6C-1 P4COM24G 6 9 KV BKR 911 OPEN BKR 301 OPEN IEce30tm I I Eceeti++e I 3rE47 3rE47 _
| |
| CABLE 21 FALURE i
| |
| i eco,- i 1.16E44 I 1 I 1 4
| |
| ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 56
| |
| | |
| l.
| |
| 1-e i ..
| |
| toss oF ... xv a,wx ,iswaR =c.c i
| |
| =
| |
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| |
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| |
| i SOURCES HECW TO PROVOE 4 Eccs cocu,o I EAccesc*e I ppEEwAre e., se ime
| |
| )
| |
| 1 I DMSON I SWGR j FAtlEE i
| |
| i IEAccocHNI-57eE-06 l
| |
| 'l 1
| |
| :l i
| |
| I i I, a i ELECTRIC POWER SYSTEM . .\SDTREE\ POWER.CAF 6-30-92 Page 57 i
| |
| 4 w.. m w w ., - +3. e .e - 4 w, ,s w- s *w-- - *y - - ee . we., ,-ww.,- -e.- -.e,
| |
| | |
| LOSS OF POWER SOURCES Page s4 Page 57 I ")
| |
| * I TT I I
| |
| GAS TURBNE D G. POWER NOT s.OSS OF OFF-STTE GENERATOR POWER NOT AVAR.ABLE POWER AVARABE IorusecNI IEAcoccN1 I(ActoscN1 Page 65 Page 66 I 3 LOSS OF NORMAL AND COfAION MODE LOSS OF ALTEBNATE POWER BOTH OFFS 8TE PCMER REAcMacNI l Ett&tz l ll i .
| |
| LOSS OF NORRAAL ALTER 4 ATE POWER NOT PREFERRED POWER AVARABE I EpocN I I EAcAracN I Page 63 Page 70
| |
| $ I 2 1 3 1 I
| |
| ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 58
| |
| | |
| GROUP N PCWER FALS I
| |
| I I
| |
| LOSS OF GROUP O FALURE OF GROUP O 480V UCC POWER 125 V DC BUS T I La ta 1 I t Lc ta I i
| |
| i i
| |
| i SWGR RC 6G t FEED DV 11 em]v VCC ESF DV R 48N WCC CAE4.E 49 FAILURE PeCOpiNG Pp(R 420 LOSS Or 480V SWGR P>tn Me OPEM P/C 60-1 PCWER BUS FALURE OPEN 327E47 327E47 j 5.75E46 t.16E*M I
| |
| I LM OF PrG 60-1 SWGR N 631 R YE9 SOURCE FALURE l
| |
| 1 eat w nNI I E Wo?m I Pege 61 5 ME46 4 I s 1 2 1 3 1 1 1
| |
| .\SDTREE\ POWER.CAF 6-30-92 Page 59 ELECTRIC POWER SYS1EM
| |
| | |
| ~
| |
| s 5
| |
| : i. -
| |
| '- FALUME F GROUP 8 tas y oc sus .
| |
| i IEDb29e1 4
| |
| Page se
| |
| . Page 71
| |
| - Page s7 -
| |
| j s i
| |
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| |
| i LOSS & POWER SOURCE ' CNOCE SC CPEN I -
| |
| I EDCSizy I I E CEOS
| |
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| |
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| |
| i i
| |
| a j' . FAR.UfE & DIVISION N OtSTRSUTION PANEL 4
| |
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| |
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| |
| 4, i
| |
| 3 i
| |
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| |
| : I . I 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 60
| |
| ._,g _. _. ,, ._,.. , , , , _,
| |
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| |
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| |
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| |
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| |
| 1 i-PC 64194COhMNG SS KV BKR 931 CPEN f
| |
| : anR 3.i cPEn ,
| |
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| |
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| |
| { } 327E47 [- } 327E47
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| |
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| |
| i 1 . -l 2 I ,.
| |
| i ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 61 [
| |
| i 1
| |
| ,i s k- . < - . , . c., , s ., ~
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| toss c .. xv GROUP R SMRP4C
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| |
| i EAC*e i QHEcwN e., a , .==
| |
| FALUAE IEAcsw
| |
| * I 57sEm I i 1 e i ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 62
| |
| | |
| LOSS & POWER SOURCES I Em(AasLH I Page 62 I I E GAS TURemE Da NOT AVAR.ABLE LOSS & W-STTE GENERATOR POWER NOT NR AVAR.A8LE I Giusetw I I Em i I EMLM" I
| |
| - ~
| |
| 125 Y p SAFETY BMR FALS TO LOSS & NORMAL AND COMMON MCCE LOSS OF
| |
| - CtosE AutRnart POWEp eOTN OrrsnE PCatR OtSTRamON TEANEL SOURCES WsN IEcatr2 w I I Ec5M?*** I I EAML** I IELO *"21 3m4T ime % T. i=Ee i I NON-SAFETY BVR D & POWER NOT FALS TO CLOSE AVAR.ABLE I E G 22'54 i (E Accu s tal sum wn I i 1 2 I $ I
| |
| * I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30 Page 63
| |
| | |
| GROUP I POWET4 FALS P1HG
| |
| {-
| |
| . I FALURE OF GROUP A FA8LtEE OF GROUP I 125V DC BUS 6 9 kV BUS T
| |
| lELC11NI l e AtmoCG l b _ h-)
| |
| I I LOSS OF POWER FALUFE OF Dev A SOURCES HECW TC PROVO8E ECCS COOUNG IEAcsa%2+1 Im+ c
| |
| * e.1 Page se 1NE43
| |
| _t-FALURE I tac =>w i 5.78E46 I i 1 2 1 ELECTRIC POWER SYSTEM .\GDTREE\ POWER.CAF 6-30-92 Page S4
| |
| | |
| GAS TURBPE GENERATOR POWER NOT AVALABLE I v;Tte l Pa0e 58 I I CTG MANUAL CTG FALS TO START
| |
| * RIE4 A.O OtSCONFECT SWfTCH CPEN Icic a uswl I unem 1 3coE43 5 00E42 I I CTG CABLE FALURE MANUAL F#TLATION t
| |
| i Ecact rw 1 Icnsuasj 1.16E44 230E42 t I GROT.P I DC CONTROL CTG BREAKER FALS POWER LOST TO CLOSE i uc m,. i i umee, i Page 68 100E46 I 1 I 2 i ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 65
| |
| | |
| D G PCWER NOT AVARMLE I t. ;wNl
| |
| ~'
| |
| COMElm4ATON OF OG DA C CONTIOL CCF VAUPE l t:yc.Kh t EtJ * "? l Page 67 4.75Em 2
| |
| . I GROLF 3 OC CONTDCL Da C FAILS TO MUN POWEM LOST i E;s tir N I e t-_c; Page 68 49 4Zt 2
| |
| D G. C CABLE FAUSE 125v CC DSTRS.fiL8C PANEL eMR 012 CPEM i
| |
| Itc O.<_wl l ira >? M g 1 ,.E . g wEm D G. FAFLS TO START 4AANunt NTIAtom FAILtJRE 1 + tv .* WU n E /1MW '
| |
| g zu. , rocE .
| |
| O G BKR 906 FAE S To CtosE i t:N + t 1DGEM 1 T I I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 66
| |
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| |
| | |
| GROUP I DC CONTROL POWEM LOST I ELC11 nN I-PeGe 66 [ -
| |
| PeGe 70 h-
| |
| - FAELURE DMSION BATTERY CABLE i DISTRBUTION PAMEL FALUFIE I EDeto1HN l l ECA181W l S 76E-os i.26E44 i i 090DE SD OFEN SATTERY FALURE I EDEo'a1Hr4 l l E B v 101% l i cE o. zS3E .
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| I aAmRvcuTrurem exTTtav ccr ceEn l Ec8[81HN l l r.ev tcuN ]
| |
| 32T247 LS3E46
| |
| [
| |
| I i I 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 68
| |
| | |
| LOSS CF nom 4AL PREFERFED POWER g
| |
| l EAcw%N l
| |
| ~
| |
| wacownNG BM 901 NORMAL PREFERRED OPEN CABLE 2. 4 OR to FALURE I tcemtw l l ecece2+41 327E47 1.tsE-04 srARruP mmsroRuER toss or OrF.strE
| |
| #1 FALURE LlNE 1 POWER l E TH50TH l lELUU6'1l 324E-05 100E42 I $ I 2 i ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 69
| |
| | |
| ALTUNA1E POWER HOT AVARA8LE
| |
| , i EKA*>0CN I I I GROUP 1 DC CONTROL STARTUP TRANSFORMER POWER LCX3T 82 FAtlJRE IEDCt1AN I IETR932Hl Page 68 3M45 I I LOSS OF OFF-STTE SKR CONTROL LOGC LNE82 FALURE I ELECe2 I I EOtWM ]
| |
| 1.00EW22 4 75E45 I I NCOMNG BKR 902 0%8% TOR FALS TO FALS TO CLOSE TRANSFER POWER I EC&e i I EmeG l 1.00E-06 1 ocE43 I I CABLE 7.11. OR 12 i DfV 1 OC POWER BKR FALURE 1 FARS OPEN I EcAot tH l l Ecs ttAu f 116E-04 3 27E47 I i I 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 70
| |
| | |
| 1_
| |
| e GROU"ll POWER FALS T
| |
| P2NC e
| |
| I I FAILURE OF GROUP B FALURE OF GROLP 19 125 V DC BUS 6.9 kV BUS T
| |
| I EC012N l l E ACwoG l 1 i LOSS OF POWER FALURE OF DIV B SOURCES HECW TO PROVOE ECCS COCUNG I E*CERW I l*+cweNI
| |
| %= .=.
| |
| FAILLM IEAG54 m I 5 7sE-06 I i i I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 71
| |
| | |
| GAS TURBPE GDdERATO 4 POWEH .' h(
| |
| AVALABLE
| |
| , I Git
| |
| * son I Pop 63 1 I CTO MANUAL CTG FALS TO START DeSCONNECT SWNCH AND RUN OPD8 IcisuAnswl 1 GTum*+ l 3 00E-03 5MW CTG CABLE FALURE MANUAL FAT 1ATIC84 IEcacTGHI Ici++uaN1 1.16EW 2 0rK42 I I GROUP 18 DC CONTROL CTG DEV 2 BREAAER PCMNEM LOST FALS TO CLOSE Itoci2xNI EECRTww I
| |
| & "" O '""
| |
| I 1 1 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 72
| |
| | |
| D.G. POWER NOT AWAAAELE I t.mc.%=e !
| |
| 4 e CCMBfPGATON CF UG D.G. O CONTROL CCF FeAURE I
| |
| I t Law. t F L% l I t t ct= .'.= 1 Pepe 7A 415Em i d GFOU' W DC CONT 5tOL DA O FALS TO 54UN POWER LOST I E* $ 21% l *=o Page 75 * *EL" 2
| |
| s I I D.G. O CABLE FALURE 1N DC DSTINEU' TON PAa#L 1901 C12 OPEN Iitat iLaI L_Ei# %" 1 1.18E4e 3:1E 47 E I D.G. FAES TO STAPT taA m NTETON F A3LU8*E e
| |
| n . . ,i n, 2.50E Q # 0t #
| |
| 2 4
| |
| C C BKR 9C5 FAILS To CtesE I n w.w]
| |
| 1.oCE-08 I i 1 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 73
| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
| | |
| GROUP 81 M COWRCt.
| |
| pg4.w sq I EDC1204 l rage 73 Page 77 -
| |
| I I FAILURE OF OMScas BATTERf CABLE I a DtSTFnrW w FALURF PANES.
| |
| I topic 2*** 1 1 Ecaisql O """ O ""~
| |
| I .
| |
| i DODE SO OPEM BATTERY FALLSE 1
| |
| i e ncem i i e svie<< 1 43E m 233E44 i i BATTERY OUTFUT BAR BATTERY CCF cro.
| |
| l Ec8ie2>*e l l 69Y TCT F N l 32?E-c7 2_53EM i i 1 2 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF l 6-30-92 Page 75
| |
| ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --
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| |
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| |
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| |
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| |
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| |
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| I EACNMON I I EACA000N I
| |
| [
| |
| .' - q P4COMn4G BKR SE1 NORMAL PREFERRED i '-
| |
| * OPEN CABLE 2. 4. OF.10 FALURE 8
| |
| 1 - _
| |
| l EC8821H I I ECAco2H I l
| |
| 327E41 1.16E44 4
| |
| I I I STARTUP TRANSFORMER LOSS OF OFF-SITE
| |
| #1 FALUFE I.9E 1 POWER iETR9U1HI IELdP11
| |
| ' 324E-05 , 1M4
| |
| 'l 1 I .
| |
| 2- 1
| |
| ! ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 .Page 76-I
| |
| _ .. , . . . - ~ . . .. . , _ ..._-
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| I I GROUP W DC CONTHOL STARTUP TRANSFORMER POWER LOST - : ' 82 fat 1JRE 1
| |
| a IEDCI12XNI. . I ETRSU2H I Poco 75 324E45 ',
| |
| n I LOSS OF OFF-SITE - BKR CONTROL LOGIC -
| |
| .LNE82 FALURE ._
| |
| q
| |
| _ r IELOOP2I I Ecul22H I . ,
| |
| * .1.ccE42 4.75E45 I I WeCOMNG BNR 922 OPERATOR FALS TO FAILS TO CLOSE TRANSFER POWER ,
| |
| t s -
| |
| l EC8022H I I EHueoC 1 1M46 1 M 43 .
| |
| l CABLE 7.11. OR 12 ' OfV 2 DC POW.R BKR FALURE 2 FALS Of'EN IECA'tiHIo IEC85DCHI a.E. 3rE47 ,
| |
| 1- 1 1 2 -1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF ' 6-30-92. Page 77.
| |
| i i . , , , ,, ., , . . .,.
| |
| | |
| 4 x4 ,
| |
| , '_ /
| |
| WlOUP IN POWER -
| |
| s FAILS
| |
| - T; i- . .
| |
| FAILURE OF GROUP C . . FALURE OF OROUP 125 V DC BUS Ill SA kV BUS T. T I EDC13N I '' ' I EAcceEG l Pw. 7e -
| |
| I I LOSS CF POWER FALURE OF DIV C SOURCai1 HECW TO PROVIDE ECCS COOLING IEACdeGENI IWHEdwCf4 '
| |
| P.a. 2 s.rxa omson in swGR
| |
| $ FAILURE I EAccoEHN 1 5.7sE-os I i I 2 1
| |
| $ ELECTRIC . P3WER 4. SYSTEM .\SDTREE\ POWER.CAF 6-30-92' Page2-78 9 w y - ._.__n
| |
| | |
| FLLURE OF GROUP C
| |
| ' 125 V DC BUS l_,
| |
| Page 88 I I LOSS OF POWER SOURCE OtODE SID OPEN l EDCS13Y l l EDEoctH l Page 43 1.43E46 FA! LURE OF DIVIStON lit DISTRIBUTION PANEL EDP t(nH 3
| |
| i I i l 2 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 79
| |
| . s_ . _: _ _ _ _ _ _ _ _ _ _ .
| |
| | |
| GAS TURBNE GEhERATOR POWER NOT AVAILABLE laid @LNl i I MANUAL NTTIATK)N GROUP ft! DC CONTHCl POWER LOST lc # c5AANI l EOC13xN l 2.00Em Page 83 I I CTG MANUAL CTG FAILS TO START tySCONNECT SWTTCH AND RUN OPEN lCTGMANSWl l GIUhBINE I 3.0CE-03 5.00EM I I CTG CABLE F#tURE CTG DfV 3 BREAKER FARS TO CLOSE lECAbTGHl !ECOdiG3Nl 1.16E-04 1.coEm i 1 1 2 I ELECTRIC. POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 80
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| l'.p a , s.
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| ~
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| .4 4'
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| i:
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| DA POWER NCTT-
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| .. AVAs.Aat.E - e
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| ~.
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| ' s h .
| |
| Page 2 ''i s , ;
| |
| puhui -
| |
| : l. ' ~
| |
| , COMBINATION OF DG Da BKR 945 FAR.S .
| |
| CCF. , TO CLOSE
| |
| : l. '
| |
| IEDOCCCFENI' 'I ECEhM5H I Page 82 1 A0E46 '
| |
| E I GROUP HI DC . DA' E CONTROL CONTROL POWER LOST , FAILURE r
| |
| 4 iEDC132dI I ELCOGEH I Page. 83 - 4.75E45 '
| |
| D.G. E CABLE FAILURE D.G. E FAR.S TO RUN 4
| |
| :IECAbOEHI 1.10E-04 4soE42 2
| |
| I I D.G. FAR S TO START 125V DC DISTRSUTION PANEL exR Daz ceEN IEDCdSEDi. I Ectb32H i
| |
| ' 2.50E42 . 327E47 l- i l 2 I
| |
| . ELECTRIC ' POWER- SYSTEM .\SDTREE\ POWER.CAF 6-30 Page 81L
| |
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| , _ g 9
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| .g u ;;
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| T i
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| s
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| 2 ; Co 1xw oF DG -
| |
| .CCF-
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| -ieDocCcrtui I
| |
| E I
| |
| - 2 D.G. S CCF 3 D.G. S CCF r
| |
| iEDGCOEI.
| |
| . 2.80E-03 ' 530E44 '
| |
| 4 1
| |
| o , ,-
| |
| 2 D G. S CCF
| |
| "~
| |
| L b =
| |
| i-2 p
| |
| . -I'.. .- I : I j 4
| |
| ELECTRIC. POWER SYSTEM - .\SDTREE\ POWER.CAF 6-30 .Page.82. ;
| |
| A
| |
| ~ -
| |
| | |
| GROUP W DC CONTROL POWER LOST l EOC13xN l Page 81 Page 85 , a B I FAILURE OF DMSON BATTERY CABLE Ill DISTRSUTION FAILURE PANEL l EDP103HN l l ECA183HN l 5.76E-06 1.16EM DODE SO OPEN BATTERY FAILURE l EDEOC1HN ] l E B Y 103HN l 1.43Em 2.53EW I I BATTERY OUTPUT BKR BATTERY OCF OPEN l ECB183HN l l EBY ICCFN l 327E-07 2.53Em I i l 2 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 83
| |
| | |
| LOSS OF NORMAL PREFERRED POWER 1 EAU SOEN l INCOMING BKR 941 NORMAL PREFERRED OPEN CABLE 2,4. OR to FALURE l ECfN41H l l ECA002H l 327E47 1.16E-04 STARTUP TRANSFORMER LOSS OF OFF-SITE
| |
| #1 FAluRE LINE 1 POWER IETRSU1HI IELOOP11 324E-05 1.00E42 I 1 1 2 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 84
| |
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| r -~
| |
| ,.g a-r
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| #)
| |
| + , i
| |
| 'I''
| |
| . : AVAE.ABLE 1EACMMeeI 2
| |
| ~
| |
| m 6
| |
| I I GROUP NI DC STARTUP TPANSFORMEFI CONTROL POWER LOST 82 FALURE ' =
| |
| ' ^
| |
| I EDCt3XN I I ETRhiU2H ]
| |
| Paes 83 3.24E45 ' '
| |
| I i LOSS OF OFF-SITE -' BKS CONTHOL. LOGC - ;
| |
| j
| |
| ' UNE #2 . FAE.URE -
| |
| IELdOP21: I ECLM2H I
| |
| '1.00E-02 4.75Ee ,
| |
| 6 I I MCOMNG BKR 942 '. OPERATOR FALS TO FAE.S TO CLOSE TRANSFER POWER
| |
| * . I ECEh'#42H I I EHueeC 1 1.00E48 1.00E43 ~
| |
| CABLE 7,11. OR 12 DfV 3 DC POWER BKR FALURE ' 3 FALS OPEN -
| |
| I ECA011H I I ECBk)GH I #
| |
| ' .1.16E44 327E47 '
| |
| I i l- 2 1 4
| |
| ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30 :Page 85
| |
| | |
| GROUP 1 POWER FAILS P[NGO i i FAILURE OF GROUP A LOSS OF 6 9 KV 125V DC BUS OfVtSK)N I SWGR htO 6C POWER IEDC11NI lEAC69CO I Page 55 LOSS OF POWER DMSION I SWGR FA! LURE OF OfV A SOURCES FAILt>RE HECW TO PROVOE ECCS COOLING l
| |
| I EACebSCO l I EAG60e4 l lwHECWANl I3 5.76E 06 1.50E-03 II i ,
| |
| LOSS OF OFF-SITE GAS TURBENE POWER GENERATOR POWER NOT AVALABLE IEACLOSCl lGTURBCNl Page 26 Page 65 i
| |
| 2 3 I i i I 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF '6-30-92 Page 86
| |
| | |
| GROUP W POWER FAILS l
| |
| l P2NCD l I
| |
| I FALURE OF GROUP B LOSS OF 6.9 KV OfVISION 88 SWGR 125 v DC SUS SAC 6C POWER l EOC12N l l EAC 500 l I
| |
| 5 DIVISION N SWGR FALURE OF Div B LOSS OF POWER HF_CW TO PROVIDE SOURCES FAEURE ECCS COCt1NG lEAC699HNl lWHf CWBNI i EAC69sOO j t
| |
| 5.76E-Os 6 soEW f
| |
| GAS RJRBtNE LOSS OF OFF-SfTE GENERATOR PCM'ER NOT POWER AVArLABLE loTuEsong I Fxtoso I Parp 72 Page 40
| |
| {
| |
| I i 2 1 3 I I i 1
| |
| .\SDTREE\ POWER.CAF 6-30-92 Page 87 ELECTRIC POWER SYSTEM
| |
| | |
| GROUP m POWER FALS P3NCD F~
| |
| I I FALURE OF CROUP C LOSS 08 69 KV 125 V DC BUS GROUP W SWGR WC T 6E PCWER I E DC13N l ! EAC6 PED i Page 79 LOSS OF POWER DMSICH m SWGR FALURE OF OfV C SUURCES FAILURE HECW TO PROVIDE ECCS COCUNG 1 EAC69 SED _] l EAC69EMN l lWHECWChl I 5.76Em 6.73E-04 II I I GAS TURBPE LOSS OF OFF-SfTE GENERATOR POWER NOT POWER
| |
| , AVAILABLE IotudBENl l EACLOSE l
| |
| % 6o % s2 I i l 2 1 3 I ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 88
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| | |
| Gat e/ Event- Name Pace Zone Gate / Event Name Page Zone Gate / Event Name Pace Zone Gate / Event Name! Pace Zone CTGMAN 24 2 EAC69EHN 78 2 EACA69DN 76 3 EACLOSEN- 2 3 CTGMAN -36 2 EACC9Emi 88 3 EACA69DN 77 2 BACN69C 28 1 CTGMAN 48 2 EAC69EN 1 2 EACA69E 52 2 EACN69Cri 58 2.
| |
| CTGMAN 65 2 EAC69SC 23 1 EACA69E 53 2 EACN69CN 69 2 CTGMAN 72 2 EAC69sC 24 2 EACA69al 2 3 EACN69D 40 1 CTGMAN 80 1 EAC69SCD 86 2 EACA69ai 85 2 EACN69DN 76 2 CTGMANSW 24 1 EAC69SCN: 57 1 EACDGC 24 3 EACN69E 52 1 CTGMANSW 36 1 EAC69SCN 58 2 EACDGC 26 2 EACN69EN ~2 2 CTGMANSW 48 1 EAC69SCII 64 2 EACDGCN 58 2 EACN69EN 84- 2 CTGMANSW 65 .1 EAC69SD 35 1 EACDGCN 66 2 EACU6C1H 54 1 CTGMANSW' 72 1 EAC69SD 36 2 EACDGD 36 3 EACN6C1N 56 2 CTGMANSW 80 1 EAC69SDD B7 2 EACDGD 38 2 EACN6C1Z 22 2 EAC69C 22 1 EAC69sDN 62 1 EACDGDN 63 2 EACN6D1H 59 1 1" EAC69C 23 2 EAC69SDN 63 2 EACDGDN 73 2 EACN6 DIN 61 2 EAC69CD 86 2 EAC69SDN 71 2 EACDGE 48 3 EACN6D1Z 34. 2 EAC69CG 64 2 EAC69SE' 47 1 EACDGE 50 2 EACN6E1Z 46 2 EAC69CH 23 2 EAC69SE 48 2 EACDGEN 2 2 EACNAC 20 2 '
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| EAC69Cm! 57 2 EAC69 SED 88 2 EACDGEN 81 2 EACNACN 58 2 EAC69 Cal 64 2 EAC69SEN 1 1 EACDGN 63 2 EACNAD 40 2 EAC69CRI 86 3 EAC69SEN 2 2 EACE1 18 3 EACNADN 63 4 EAC69CN 56 1 EAC69SEN 78 2 EACE1 22 4 EACNADN 76 2 ;
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| EAC69CN 57 2 EAC6C1 3 2 EACE1N 54 4 EACNAE 52 2 EAC69D 34 1 EAC6C1Z 22 2 EACE2 30 3 EACNAEN 2 2 EAC69D 35 2 EAC6C1ZN 54 2 EACE2 34 4 EBC404H 19 1 EAC69DD 87 2 EAC6C2 5 2 EACE2N 59 4 EBC406H 15 1 EAC69DG 71 2 EAC6D1 7 2 EACE3 42 3 EBC426H 31 1 EAC69DH 35 2 EAC6D1Z 34 2 EACE3 46 4 EBC434H 43 1 EAC69DHN 62 2 EAC6D1ZN 59 2 EACLOSC 24 4 EBCN11H 20 1 EAC69DHN 71 2 EAC6D2 9 2 EACLOSC 28 2 EBCN11Y 19 3 EAC69DHN 87 3 EAC6El 11 2 EACLOSC 86 1 EBCN11Y 20 2 EAC69DN 61 1 EAC6E1Z 46 2 EACLOSCN 58 3 EBCN12H 32 1 EAC69DN 62 2 EAC6E2 13 2 EACLOSD 36 4 EBCN12Y 31 '3 EAC69E 46 1 EACA69C 28 2 EACLOSD 40 2 EBCN12Y 32 2 EAC69E 47 2 EACA69C 29 2 EACLOSD 87 2 EBCN13H 44 1 EAC69ED 88 2 EACA69CN 58 3 EACLOSDN 63 4 EBCN13Y 43 3 EAC69EG 78 2 EACA69CN ?0 2 EACLOSE 48 4 EBCN13Y 44 2 EAC69EH 47 2 EACA69D 40 2 EACLOSE 52 2 EBCN14H 16 1 EAC69EHN 1 2 EACA69D 41 2 EACLOSE 88 2 EBCN14Y 15 3 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 89
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| Gate / Event Name Pace Zone Gate / Event Name Pace Zone Gate / Event'Name Pace Zone Gate / Event flame . Eagg Zong EBCN14Y .16 '2 EBY103H 49 2L ECA023H 34 2 ECADGDH 73 1 EBCS12AY 19 2 EBY103mi 83 2 ECA023m! 61 2 ECADGEH SO 1 EBCS12BY 31- 2 EBY104H 17 2 ECA024H 10 2 ECADGEH 81- 1 EBCS12H 19 2 EBY11 19 4- ECA025H 12 2 ECAN11H 20 2 EBCS12H 31 2 EBY11 21 2 'ECA025H 46 2 ECAN12H 32 2 EBCS34AY 43 2 EBY12 31 4 ECA026H 14 2 ECAN13H 44 2 EBCS34BY 15 2 EBY12 33 2 ECA040H 22 4 ECAN14H 16 2 EBCS34H 15 2 EBY13 43 4 ECA040mi 54 4 ECAS1AH 19 2 EBCS34H- 43 2. EBY13 45 2 ECA041H- 34 4 ECAS1BH 31 2 EBS6C1H 4 2 EBY14 -15 4 ECA041mi 59 4 ECASICH 43 2 EBS6ClH 22 3 EBY14 17 2 ECA042H 46 4 ECAS1DH 15 2 54 2 EBY1CCF 17 2 ECA043H 15 2 ECAXX1H 5 1 EBS6C1ml EBS6C2H 6 2 EBYlCCF 21 2 ECA044H 16 2 ECAXX2H 9 1 EBS6D1H 8 2 EBY1CCF 25 2 ECA045H 19 2 ECAXX3H 13 1 EBS6D1H 34 3 EBY1CCF 33 2 ECA046H 20 2 ECAXX4H 3 1 EBS6D1ml 59 2 EBY1CCF 37 2 ECA047H 31 2 ECAXX5H 7 1 EBS6D2H 10 2 EBYlCCF 45 2 ECA04BH 32 2 ECAXX6H '11 1 EBS6E1H 12 2 EBY1CCF 49 2 ECA050H 43 2 ECB1B1H 21 1 EBS6E1H 46 3 EBY1CCFN 68 2 ECA051H 44 2 ECB1B1H 25 1 EBS6E2H 14 2 EBY1CCFN 75 2 ECA1B1H 21 1 ECB1B1mi 68 1 EBSE1H 22 3 EBY1CCFN 83 2 ECA1B1H 25 2 ECB1B2H 33 1 EBSE1HN 54 3 ECA002H 28 2 ECA1BlHN 68 2 ECB1B2H 37 1 EBSE2H 34 3 ECA002H 40 2 ECA1B2H 33 1 ECB1B2m1 75 1 EBSE2HN 59 3 ECA002H 52 2 ECA1B2H 37 2 ECBlB3H 45 1 EBSE3H 46 3 ECA002H 69 2 ECA1B2m1 75 2 ECB1B3H 49 1 EBSXX1H 5 -1 ECA002H 76 2 ECA1B3H 45 1 ECB1B3m1 83 1 EBSXX2H 9 1 ECA002H 84 2 ECA1B3H 49 '2 ECB1B4H 17 1 EBSXX3H 13 1 ECA011H 29 1 ECA1B3mi 83 2 ECB1DCH 29 2 EBSXX4H 3 1 ECA011H 41 1 ECA1B4H 17 1 ECB1DCH 70 2 EBSXX5H 7 1 ECA011H 53 1 ECACTGH 24 1 ECB2DCH 41 2 EBSXX6H 11 1 ECA011H 70 1 ECACTGH 36 1 ECB2DCH 77 2 EBY101H 21 2 ECA011H 77 1 ECACTGH 48 1 ECB301H 4 1 EBY101H 25 2 ECA011H 85 1 ECACTGH 65 1 ECB301H .22 1 EBY101HN 68 2 ECA021H 4 2 ECACTGH 72 1 ECB30lHN 56 1 EBY102H 33 2 ECA021H 22 2 ECACTGH 80 1 ECB312H 22 5 EBY102H 37 2 ECA021HN 56 2 ECADGCH 26 1 ECB312m1 54 5 EBY102m1 75 2 ECA022H 6 2 ECADGCH 66 1 ECB331H 6 1-EBY103H 45 2 ECA023H 8 2 ECADGDH 38 1 ECB341H 8 1 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 90
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| Gate / Event Name Page Zone Gate / Event Name Pace Zone Gate / Event Name Pace Zone Gate /Erent'Name Pace Zone ECB341H 34 1 ECB927H 63 1 ECBXX3H 13 2 EDC13N 79 2 ECB341HN 61 1 ECB931H 8 2 ECBXX4H 3 2 EDC13N 88 1 ECB354H 34 5. ECB931H 34 2 ECBXX5H 7 2 EDC13X 48 1 ECB354HN 59- 5 ECB931m1 61 2 ECBXX6H' 11 2 EDC13X 49 2 ECB361H 10 1 ECB941H 52 1 ECL902H 29 2 EDC13X 50 1 ECB371H 12 1 ECB941H 84 1 ECL902H 70 2 .EDC13X 53 1 ECB371H 46 1 ECB942H 53 1 ECL922H 41 2 EDC13XN 80 2' -
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| ECB380H 46 5 ECB942H 85 1 ECL922H 77 2- EDC13XN 81 1 ECB391H 14 1 ECB944H 14 2 ECL942H 53 2 EDC13XN 83 2 }
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| ECB3DCH 53 2 ECB945H 50 2 ECL942H 85 2 EDC13XN 85 1 ECB3DCH 85 2 ECB945H 81 2 EDC11 18 2 EPC14 15 4 ECB403H 20 1 ECB951H 12 2 EDC11N 54 1 EDCS11Y 18 1 ECB405H 16 1 ECB951H 46 2 EDC11N 55 2 EDCS11Y 19 2 ECB410H 22 6 ECBCTG1 24 2 EDC11N 64 1 EDCSllY 55 1 ECB410mi 54 6 ECBCTG1H 65 2 EDC11N 86 1 EDCS12Y 30 1 ECB420H 34 6 ECBCTG2 36 2 EDC11X 24 1 EDCS12Y 31 2 ECB420HN 59 6 ECBCTG2N 72 2 EDC11X 25 2 EDCS12Y 60 1 ECB425H 32 1 ECBCTG3 48 2 EDC11X 26 1 EDCS13Y 42 1 ECB430H 46 6 ECBCTG3N 80 2 EDC11X 29 1 EDCS13Y 43 2 ECB433H 44 1 ECBD12H 26 2 EDC11XN 65 1 EDCS13Y 79 1 ECB901H 28 1 ECBD12H 66 2 EDC11XN 66 1 EDCS14Y 15 3 ECB901H 69 1 ECBD22H 38 2 EDC11XN 68 2 EDE0AlH 18 2 ECB902H 29 1 ECBD22H 73 2 EDC11XN 70 1 EDE0A1H 25 1 ECB902H 70 1 ECBD23H 63 1 EDC12 30 2 EDE0A1H 55 2 ECB904H 6 2 ECBD32H 50 2 EDC12N 59 1 EDE0Almi 68 1 ECB905H 26 2 ECBD32H 81 2 EDC12N 60 2 EDEUBlH 30 2 ECB905H 66 2 ECBN01H 19 1 EDC12N 71 1 EDE0BlH 37 1 ECB911H 4 2 ECBN02H 31 1 EDC12N 87 1 EDE0B1H 60 2 ECB911H 22 2 ECBN03H 43 1 EDC12X 36 1 EDE081m1 75 1 ECB911mi 56 2 ECBN04H 15 1 EDC12X 37 2 EDEOC1H 42 2 ECB921H 40 1 ECBX11H 5 2 EDC12X 18 1 EDECC1H 49 1 ECB921H 76 1 ECBx12H 9 2 EDC12X 41 1 EDEOC1H 79' 2 ECB922H 41 1 ECBX13H 13 2 EDC12XN 72 1 EDEOClai 83 1 ECB922H 77 1 ECBX14H 3 2 EDC12XN 73 1 EDE0D1H 15- 4 ECB924H 10 2 ECBX15H 7 2 EDC12XN 75 2 EDGC 26- 2 ECB925H 38 2 ECBX16H 11 2 EDC12XN 77 1 EDGC 66 2 ECB925H 73 2 ECBXX1H 5 2 EDC13 42 2 EDGCCCFC 26 1 ECB926H 63 2 ECBXX2H 9 2 EDC13N 78 1 EDGCCCFC 27 2 ELECTRIC POWER SYSTEM .\SdTREE\ POWER.CAF 6-30-92 Page 91
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| Gate / Event Name Pace Zone Gate / Event Name Pagg Zone Gate / Event Name Pace Zone Gate / Event Name Pace Zone EDGCCCFD 38 1 .EDGMAN 66 2 ELCDGCH 66 '2 ETRSU1H 76 1 EDGCCCFD' 39 2' EDGMAN 73 2 ELCDGDH 38 .2 'ETRSU1H 84 1 EDGCCCFE '50 1 EDP101H 18 2 ELCDGDH 73 2 ETRSU2H 29 '2 EDGCCCFE 51 2 EDP101H 25 1 ELCDGEH 50 2 :ETRSU2n 41- 2 EDGCCCFEN 81 1 FDP101H 55' 2 ELCDGEH 81 2' ETRSU2H 53 2 EDGCCCFEN 82 2 EDP101mi 68 1 ELOOP1 28 2 ETRSU2H 70 '2 EDGCCFCN 66 1 EDP102H 30 2 ELOOP1 40 2 ETRSU2H 77 2 EDGCCFCH 67 '2. EDP102H 37 1 ELOOP1 52 2 ETRSU2H 85 2 EDGCCFDN 73 1 EDP102H 60 2 ELOOP1 69 2 GTURBC 24 2 EDGCCFDN 74 2 EDP102HN 75 1 ELOOP1 76 2 GTURECN 58 1 EDGCD 27 1 EDP103H 42 2 ELOOP1 84 2 GTURBCN 65 2-EDGCD 39 1 EDP103H 49 1 ELOOP12 2 3 GTURBCU 86 2 EDGCD 67 1 EDP103H 79 2 ELOOP12 28 3 GTURBD 36 2 EDGCD 74 1 EDP103mi 83 1 ELOOP12 40 3 GTUREDN 63 1 EDGCDE 27 2 EDP104H 15 4 ELOOP12 52 3 GTURBDN 72 2 EDGCDE 39 2 EGATE1 5 2 ELOOP12 58 3 GTURBDN 87 1 EDGCDE 51 2 EGATE1 6 2 ELOOP12 63 5 GTURBE 48 2 EDGCDE 67 2 EGATE2 L 2 ELOOP2 29 1 GTURBEN 2 1 EDGCDE 74 2 EGATE2 10 2 ELOOP2 41 1 GTURBEN 80 2 EDGCDE 82 2 EGATE3 13 2 ELOOP2 53 1 GTURBEN 88 1 EDGCE 27 2 EGAPE3 14 2 ELOOP2 70 1 GTURBItE 24 2 EDGCE 51 1 EGATE4 3 2 ELOOP2 77 1 GTURBINE 36 2 EDGCE 67 2 EGATE4 4 2 ELOOP2 85 1 GTURBINE 48 2 EDGCE 82 1 EGATES 7 2 ETR6C1H 4 1 GTURBIIE 6L 2 EDGD 38 2 EGATES 8 2 ETR6C1H 22 2 GTURBINE 72 2 EDGD 73 2 EGATE6 11 2 ETR6C1HN 56 2 GTURBINE 80 2 EDGDE 39 2 EGATE6 12 2 ETR6C2H 6 1 P1 18 2 EDGDE 51 2 EHU69C 29 2 ETR6D1H 8 1 P1N 54 2 EDGDE 74 2 EHU69C 41 2 ETR6D1H 34 2 PlNC 64 2 EDGDE 82 2 EHU69C 53 2 ETR6DlHN 61 2 PlNCD 86 2 EDGE 50 2 EHU69C 70 2 ETR6D2H 10 1 P2 30 2 EDGE 81 2 EHU69C 77 2 ETR6E1H 12 1 P2N 59 2 EDGFSCD 26 1 EHU69C 85 2 ETR6ElH 46 2 P2NC 71 2 EDGFSCD 66 1 EHUS1AD 19 1 ETR6E2H 14 1 P2NCD B7 2 EDGPSDD 38 1 EHUS10D 31 1 ETRSU1H 28 1 P3 42 2 EDGFSDD 73 1 EHUSICD 43 1 ETRSUlH 40 1 P3NC 78 2 EDGFSED 50 1 EHUS1DD 15 1 ETRSU1H 52 1 P3NCD 88 2 EDGFSED 81 1 ELCDGCH- 26 2 ETRSU1H 69 1 WHECWA 23 2 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 92
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| Gate / Event Name Paqa Zone Gate / Event Name Pace Zone Gate / Event Name Pace Zone Gate / Event Name Page Zone WHECWAN 57- 2 WHECWAN 64~ '3 WHECWAN 86 4 WHECWB 35 2 WEECWBN 62 2 WHECWPli 71 3 WHECWBN 87 4 WHECWCN 1 2 WHECWCN '78 3 WHECWCN 88 4 ELECTRIC POWER SYSTEM .\SDTREE\ POWER.CAF 6-30-92 Page 93
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| | |
| ATTACHMENT 19QB DHR RELIABILITY STUDY OFFSIT' . DOSE AND OPERATOR RECOVERY CALCULATIONS This attachment covers three different calculations that were comple*ed for various aspects of the ABWR Decay Heat Removal Reliability Study. The calculations are:
| |
| 1)- Offsite doses following RPV boiling in Mode 5.
| |
| : 2) Time to reach RPV boiling for specific plant conditions and decay haat loads.
| |
| : 3) Time for RPV water level to reach top of active fuel (TAF).
| |
| : 4) Human Reliability Analysis.
| |
| : 5) Decay heat removal capability of CUU and FPC
| |
| : 1. Offsite Doses For the ABWR Decay Heat Removal Reliability Study, the success criteria for Mode 5 allows boiling of water in the RPV or spent fuel pool. The following calculation of offsite doses assuming boiling in the RPV and spent fuel pool substantiates why boiling is a viable success criteria in Mode 5.
| |
| The equation for calculating offsite doses is:
| |
| Dose - RR
| |
| * DF
| |
| * BR
| |
| * DCF Dose - Offsite dose for 24 hour period (REM)
| |
| RR -
| |
| Release rate for 24 hours
| |
| | |
| l
| |
| - Decav Heat Load (Bru/hr1
| |
| * 0.00016 (uct/gm I 131
| |
| * 925 Btu /lb Water 0.015 (1 131 carryover)
| |
| * 454 Em
| |
| * 24 lb DF - Dispersion factor - 1,2x10' sec/m 3
| |
| Breathing rate - 3.47x10 BR -
| |
| m /see 6
| |
| DCF - Thyroid dose concentration factor - 1,08x10 REM /ci The values in the above equation such as I-131 carryover, I-131 concentration, and dispersion factor are conservative estimates based on ABWR S$AR analysis and regulatory guidance.
| |
| The decay heat load at 3 and 14 days following shutdown is 5.9x107 Btu /hr and 3.2x107 Btu /hr, respectively. Using the above equation, the doses for 24 hours at 3 and 14 days are 7.5x10-4 and 4.04x10-4 REM, respectively. This is significantly below the FEMA limit of 5 REM per 24 hours for normal plant operations. Thus boiling in Mode 5 will not exceed any offsite dose limits and is a viable success criteria.
| |
| : 2. Time to Reach Boillne The time for an operator to recover a failed RHR system in the ABWR Decay Heat Reliability Study is conservatively based on the time to boiling in the RPV or the spent fuel pool. The following discussion addresses the calculation of time to boiling for the RPV and RPV plus spent fuel pool at various times after shutdown.
| |
| It is assumed that the initial temperature of the RPV or spent fuel pool is 140*F. This is typical for normal Mode 4 or 5 operation.
| |
| l
| |
| | |
| Table 19QB 1 Decay lleat Mass of Vater Time to Reach
| |
| -d2d2 Days after Shutdown ( Btu /hr) (1bs) Boiline (hrs) 6 4 2 6.8x10 1.1x10 1.2 7 6 5 3 5.9x10 1.1x10 1,3 7 7 5 ' 5.9x10 1.2x10 15 7 7 5 14 3.2x10 1.2x10 27
| |
| | |
| The' equation for time to boiling is:
| |
| t - [AT/ heat up rate ('F/hr))
| |
| 212 - 140 t -
| |
| (Decay Heat Rate / Mass of Vaccr)
| |
| Table 19QB 1 shows the results for time to boiling for the RPV alone a't 2 and
| |
| : 3. days following shutdown and for the RPV plus spent fuel pool (i.e., reactor cavity flooded and fuel pool gates opened) at 3 and 14 days. As can be seen, the time for operator action varies from a little over an hour for the RPV alone to approximately one day for the RPV plus opent fuel pool 14 days after shutdown.
| |
| : 3. Time for RPV Water level to Reach Too of Active Fuel (TAF)
| |
| This section summarizes the calculations for the time to reach top of active fuel in modes 3, 4, and 5. The results show it will take 6.4 hours in mode 3, 13 hours in mode 4, 15 hours in the early part of mode 5 before flooding of the cavity, and more than a week after cavity flooding in mode 5. Assuming that it takes 925 Btu to vaporize 1 lb of water, the decay heat at a specific tim 6 is divided by 925 to find the rate of vaporization. Division of water mass by this vaporization rate results in the time for RPV water level to reach TAF. Table 19QB-2 shows the results.
| |
| i
| |
| | |
| Table 19QB.2 Decay Mass of After Heat Crater Time to Mode Shutdown -(Btu,hr) (1bs) Reach TAF 8 5 3 4 hrs 1.43*10 9.8*10 6.4 hrs -
| |
| 4 2 days 6.8*10 9.8*10' 13 hrs 5 3 days 5.9*10 9.8*10' 15 hrs 5 3 days 5.9*10 1.2*10 7.8 days 7
| |
| -5 14 days 3.2*10 1.2*10 7 14.5 days J
| |
| | |
| . - . --.. . ...... . ... .~.. - .
| |
| 4.0 Human. Reliability Analysis (HRA) 4.1 Purpose
| |
| ~The purpose of this HRA is to calculate the human error probabilities (HEPs).for the decay heat removal reliability study, l'
| |
| 4.2 Summary Tables 1 and 2 show the HEPs which were calculated for various time frames and plant modes for two cases.
| |
| Case a. Operator action required before water starts to boil Care b operator action required to prevent core damage (CD)
| |
| However, it was decided that more conservative values should be used in the PR\. These values are also shown in these tables.
| |
| l v ,
| |
| | |
| [ ,
| |
| Table 1 PROBABILITY OF FAILURE TO DIAGNOSE Case- Mode Time After Prob. (Fall to Diagnose)
| |
| Shutdown Calculated Used in PRA al 5 2-3 days 1.0E-04 1.0E-03
| |
| _a 5. >3 days <<1.0 E-05 1.0E-04 b; 3,4, and 5 Any Time (prior to flood- After Shutdown <<1.0E-Of 1.0E-04 i ing reacter cav '
| |
| ity) b 5 (after flood- Any Time <<1.0E-05 1.0E-04 ing. reactor cav-ity) i l
| |
| l
| |
| | |
| Table 2 PROBABILITY OF FAILURE T') START A SPECIFIE*v " MINIMUM-SET" SYSTEM Case Mode Time After Prob. (Fail to start)
| |
| Shutdown Calculated Used in PRA a All 4.0E-03 2.0E-02 b- -All 4.0E-03 2.0E-02 4.3 Hethodology The REP calculations were performed conservatively using the procedure for Nominal Human Reliability Analysis (HRA) in Table 8-1, Reference 1 with the following steps:
| |
| a) The displays and alarms availabit to the operator were identified.
| |
| b) _The times to boiling and core damage were identified.
| |
| c) The times for diagnosis and post-diagnosis actions were allocated.
| |
| d) .Tb6 HEPs for diagnosis and for post-diagnosis actions were calculated using- figure: 8-1 and table 8-3 and 8-5 of _ referenca 1.
| |
| e) Higher than calculated values were assigned conservatively for use in the PRA.
| |
| f) It is assumed that at least two operators are in the control room at all times during shutdown.
| |
| l l
| |
| l
| |
| | |
| 4.3.1 Control Room and Alarms Table 3 shows the relevant alarms thich are available in the CR
| |
| .(Reference 2). Operator is alerted to the failure of the operating RHR by means of one of the RHR specific alarms. If none of these alarms work, he will be alerted to the RPV parameters alarm 2 (though RPV pressure and water level may not be available prior to boiling). With these multiple alarms, it is reasonable to assume that all operators will be promptly alerted to the RHR failure.
| |
| TABLE 3
| |
| " CONTROL ROOM ALARMS AIDING DIAGNOSIS OF "ONE RHR LOST" RHR SPECIFIC RPV PARAMETERS
| |
| : 1. Pump discharge 'preusure high 1. Temperature
| |
| : 2. Pump motor ove: 2. Pressure
| |
| : 3. R9R loop powet failure 3. Water level 4, RHR loop loop logic failure
| |
| : 5. RHR pump motor trip
| |
| : 6. RCW outlet temperature high In mode 5 with the reactor cavity flooded, the operator would be made l
| |
| aware-of heating the fuel pool by many other indications. . Personnel on the refueling floor will all sence the increased temperature and L
| |
| will'see steam formation. If no personnel notice the fuel pool heat-up, the operator would receive an alarm of low fuel pool level E and initiation of fuel pool level make up.
| |
| r v 9~
| |
| | |
| 4.3.2 Allocation of Times To Diagnosis and Post-Diagnosis Actions The time atallable to the operators was allocated between time for diagnosis and time for post-diagnosis action. Table 4 shows the various times which were used to calculate the HEPs. Column three gives the calculated times before boiling (case a), and core damage (case b), or the total time available for allocation. Columns 4 and 5 show the results of the allocation. Enough tLme is allocated to poet-diagnostic act.ans, so that there is sufficient time for recovery of human errors, even if the required action must take place outside the control room.
| |
| TABLE 4 TIMES AVAILABLE (IN HOURS)
| |
| CASE MODE TOTAL TIME ALLOCATED DIAGNOSIS TIME FOR POST-DIAGNOSIS TO EVENT TIME (TD) ACTIONS (TA) a 5 (Days 2 to 3) 1.2 0.5 0.7 a 5 (after 3 days) 214 12 22 b All 26.4 22.4 24 4.4 Results and Conclusions The results of this HRA study is documented in table 1 & 2. It is concluded that the operator has adequate instrumentation and elarms to to diagnose the event. Adequate procedures and operator training will assure proper response to the loss of RHR event.
| |
| | |
| --. .. --- - _ . .~...-. . .
| |
| | |
| ==4.5 REFERENCES==
| |
| r
| |
| : 1) Swain, A. D., Accident Sequence Evaluation Program Human Reliabilit; Analysis Procedute, Sandia National Laboratories, NURE0/CR-4772, U.S.
| |
| Nuclear Regulatory Commission, Washington, D.C. February 1987
| |
| : 2) Interlock Block Diagram IBD 137C0326 Sh 10 Rev. 2 5.0 DECAY HEAT REMOVAL CAPABILITY OF CUW AND FPC The purpose of the following heat removal calculations is to determine heat removal capabilities of FPC and CUW after flooding the cavity up to 4.5 days as a function of time following shutdown. The results show that in mode 3, FPC and CUW, together, are capable of removing the decay heat. After 4.5 days, fuel pool cooling alone is sufficient and after 14 days CUW alone is sufficient. In Modes 3 and 4, FPC cannot be used but CUW is able to remove the decay heat because of increased capacity at higher temperatures. To perform these calculations for CUW, an outlet temperature of 212 F for the non-regenerative heat exchangers and inlet temperature of 95 F for cooling water 're assumed. Multiplying the temperature difference (212-95) .by mass flow rate through the pumps, 676 gpm, and the heat exchanger effectt<eness of 0.81 results in 9.39 mw which matches the decay heat 14 days after shutdown.
| |
| Similar calculations are performed for FPC, but this time the nians flow rate and heat exchanger are 2200 gpm and 0.39, respectively. Th4 result shows if FPC is used 4.5 days from shutdown, the water temperature will not reach 212 F.
| |
| The heat removal capability of FPC 4.5 days after shutdown in 14.72 mw which is equal to the decay heat at that time.
| |
| l.
| |
| | |
| ATTACllMENT 19QC REVIEW OF SICNITI': ANT SlWTDOWN EVENTS:
| |
| ELECTRICAL POWER AND DEC#Y llEAT REMCVAL i
| |
| i I
| |
| j-t
| |
| .- . _ .-.-... - , - ~ . - . _ , - . - . , m.c-.- _. - . , , _ , -. . . , . .
| |
| | |
| A review was made of operating events involving loss of off-site power (IDOP) and loss of Decay Heat Removal (DHR). These two areas appear to have the greatest potential for causing core damage during shutdown based on past experience. The sources utilized for information on past shutdown events were:
| |
| - " Residual Heat Removal Experience Review and Safety Analysis",
| |
| HSAC.88, March 1986
| |
| - " Loss of Vital AC Power and the Residual Heat Removal System during Mid Loop Operations at Vogtle Unit 1 on March 20, 1990",
| |
| NUREC 1410 June 1990
| |
| - "NRC Staff Evaluation of Shutdown and low Power Operation",
| |
| NUREC 1649, March 1)92
| |
| - Selected INPO Reports (e.g., SOE and SOCR summaries) and NRC Information Notices.
| |
| The'results of this evaluation are contained in Tables 19QC-1 and 19QC 2
| |
| . for.1DOP and loss of DHR respectively.
| |
| Siimmary of DHD Tre g Not all of the events discussed in NSAC - 88 are contained in Table 19QC 2. Those events that were due to random failures of single components and dii not result in loss 01 DHR or othsr significant plant effects were not evalu; tad furth:r. If. the sitigle failure resulted in loss of coolant, over pressurization, flooding, or loss of Shutdown Coolin5 (SDC) function, the event was included and the applicable ABVR feature to prevent or mitigate the event was discussed.
| |
| 19QC-1 4
| |
| _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___________.___.____.___________._____._.___._..____________-__.__.___._______________.________s
| |
| | |
| Summary l
| |
| The results of this review demonstrate that ABVR design includes many :
| |
| features that will prevent or mitigate unacceptable consequences of typical past events.
| |
| The main features of the ABWR that will prevent or. mitigate shutdovn events are:
| |
| - Three divisions of ECCS and support systems that are physically and electrically independent
| |
| - Two independent off site power sources
| |
| - Four on site power sources (three emergency diesel generators and one combustion turbine generator)
| |
| - Pitnt configuration to minimize common mode failures due to fire and floods
| |
| - Appropriate Technical Specifications and administrative cont-ols to ensure availability of systems during periods of potentiat.y ; rh
| |
| -risk operations-
| |
| - Several alternate means of DHR if normal systems were to fail or be out of service for maintenance
| |
| - Instrumentation availability durin5 shutdown to monitor plant safety status and initiate safety systems when needed 19QC-2
| |
| | |
| TABLE 19QC-1 IASS OF OFF-SITE POWER PRECURSORS EVEFT DESCRIPTION APPLICABLE ABUR FEATURES Indian Point 2 and Yankee " Great Northeast Blackout" ABUR has two independent offsite power sources. These are backed up by three Rowe (11/9/65) physically and electrically separate trains of Class IE AC power each contain-ing an emergency diesel gercrator. These are further backed up by a permanent onsite Combustion Turbine Generator (CTL) which is espable of powering any one of the three trains if all three diesels were to all fail. The CTG is alsa capable of supplying power to non-safet*r busses such that feedwater or condensate pumps can -
| |
| also be used to provide react or coolant jja make up.
| |
| & Point Beach 1 (2/5/71) Loss of all transmission lines, See discussion of Indian Poir.t 2 and failure of three transformer dif- Yankee Rowe (11/9/65).
| |
| ferential relays, causing trans-former lockout.
| |
| Ginna (3/4/71) Plant siding fell into 34.5kv line ABUR has two independent transformers pow-connecting sole startup ered by two independent of fsite power transformer. supplies which reduces the probability of losing offsite power. In the event of i losing offsite power, features described under Indian Point 2 and Yankee Rowe l (11/9/65) can mitigate the event.
| |
| Palisades (9/2/71) Transmission line fault, isolation See discussion of Ginna (3/4/71).
| |
| breaker failure. Backup relay isolated 345kv bus.
| |
| 4
| |
| | |
| TABLE 19QC-1 IDSS OF OFF-SITE POWER PRECURSORS
| |
| -Continued-j EVENT DESCRIPTION APPLICABLE ABWR FEATURES San Onofre 1 (6/7/73) 138kv auxiliary transformer out ABWR uses three auxiliary transformers.
| |
| for maintenance. Ground fault Each powers one of the three Class IE and operated differential relays, non-1E buses. In addition, a reserve deenergizing other auxiliary auxiliary transformer is available to transformers. power all three Class IE buses. CTG also
| |
| - available which can power IE and non-1E busses without using the auxiliary trans-formers.
| |
| Oconee 1 (1/4/74) 230kv switchyard isolated, 100kv The ASWR also has two sources of offsite off-site source remained energized power.
| |
| to supply power to the plant.
| |
| e-.
| |
| Fort Calhoun (3/13/75) Sole 161kv backup off-site trans- See Indian Point 2 and Yankee Rowe f mission line out for maintenance. (11/9/65) and ' a T3/4/71).
| |
| . 1 345kv output breaker tripped (faulty protective relays), open-ing remaining connection to i
| |
| off-site power. Off-site power could have been supplied from 345kv switchyard by opening gen-erator disconnects.
| |
| Turkey Point 4 (5/16/77) Ioss of Off-site Power (th0P) See Indian Psinc 2 and Yankee Rowe l
| |
| < (11/9/65).
| |
| Connecticut Yankee Protective relays operated shen See Indian Point 2 and Yankee Rowe .
| |
| 1 (6/26/76) lines were re-energized aftei ser- (11/9/65).
| |
| vice, causing IDOP.
| |
| i t
| |
| 4 __ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _ _ _ _ . _ _ _ _ . . _ _ _ _
| |
| | |
| TABLE 19QC-1 LOSS OF OFF-SITE POWER PRECURSORS
| |
| -Continued-APPLICABLE ABWR FFATITES UNH DESCRIPTION LOOP due to lightening strikes. See Indian Point 2 and Yankee Rowe Indian Point 2 (7/13/77) (11/9/65).
| |
| Emergency Diesel Generators (EDGs) operated.
| |
| Substation switching error. ABWR has two of fsite power sources so St. Lucie 1 (5/14/78) probability of one switching error result-ing in loss of all offsite power is low.
| |
| But if it were to occur, mitigation fea-tures exist as discussed in Indian Point 2 and Yankee Rowe (11/9/65).
| |
| Loss of all 7 transmission lines See Indian Point 2 and Yankee Rowe Turkey Point 3 (4/4/79) (11/9/65).
| |
| " due to weather.
| |
| $n ABVR has two sources of non-IE power. A Davis Besse (4/19/80) One EDG out for saintenance. One 0,
| |
| 13.8kv bus connected, other ener- ground fault on one would not result in loss of all non-lE power. In addition, if gized but not connected. Ground fault on 13.8kv bus caused loss of all non-lE power were to be lost, no valves connected to the RHR System would non-nuclear instruments. Air was pulled into DHR pump, and pump was automatically cycleAlso,and cause loss has the ABUR of NPSH three stopped by operator. Pump vented to any RHR pump.
| |
| and restarted after 2-1/2 hours. independent (100I) RHR Systems such that loss of one would not result in loss of the ability to remove decay heat.
| |
| Haintenance error caused LOOP. See St. Lucie 1 (5/14/78).
| |
| San Onofre 1 (4/22/80)
| |
| See Indian Point 2 and Yankee Fowe Prairie Island 1 (7/15/80) Weather related LOOP. (11/9/65).
| |
| l I
| |
| I
| |
| | |
| 1
| |
| [
| |
| j: ..
| |
| .TABtX 19QC-1 I
| |
| - ' IASS OF OFF-SITE FOUEK FRECURSORS j
| |
| -Continued-i EyllE ;ECSCRIPTIOl' APPLICABLE ABUR FEATURES' i f !
| |
| ( Maintenance error caused IDOP. See St. 1.ucie 1 (5/14/78).
| |
| San Onofre 1 (11/22/80) i I Diablo Canyon 1,(10/16/82) IDOP caused by b-2sh' fire.. See Indian Point 2 and Yankee Rowe L (11/9/65).
| |
| 1 .
| |
| 4-See St. Lucie 1-(5/14/78).
| |
| j .-Farley 2 (10/8/83) Switchyard breaker failure during [
| |
| i i- refueling.
| |
| 4 l
| |
| See Indian Point 2 and Yankee Rowe
| |
| . Palisades (1/8/84) . Deliberate deenergization of , l off-site power to isolate faulty (11/9/65) and cinna (3/4/71). ABER te_ch- '
| |
| . breaker. One EDG out for mainte- nical specifications require one offsite nance, other available but its and one onsite power source be available
| |
| }
| |
| [ .g service water pump was out for at all times l maintenance, and operators failed l
| |
| :g to. recognize this before authoriz- !
| |
| 4 ing work on breaker. Available l EDG overheated and was manually
| |
| [
| |
| j- tripped.. l
| |
| ' I i
| |
| 1 Sequoyah 1 (3/26/84) Cround short on 500kv switchyard A similar event at an ABWR could be more j l
| |
| breakar deenergized transformer. easily mitigated due to the existence of i I
| |
| F Startup transformer supplied the CTG and three EDCs.
| |
| power.
| |
| Yankee Rowe (5/3/84) One 115kv line out for mainte- See Sequoyah 1 (3/26/84).
| |
| nance, other energized. Normal ;
| |
| supply transformer energized. t l
| |
| Temporary fault detection relay ;
| |
| i-caused breakers from normal supply i transformer to open. f e ;
| |
| i I i .I i a' l t a- ,- - , - - - - . - . - , . - -_ .m. - . . . . . _ .._ _______ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ . _ . _ _ _ . . . .
| |
| e
| |
| | |
| i TABLE 19QC-1 IDSS OF OFF-SITE FOUER PRECURSORS
| |
| -Continued-EVENT DESCRIPTION APPLICABLE ABk'R FEATURES Oue of three safety buses was out Each of the three ABk'R safety trains have Salem 1 (6/5/84) of service for maintenance and one separate independent emergency power sup-of the batteries in the two plies and support systems so each diesel remaining safety trains was out of can supply power to its own cooling water service for replacement. pump. ABk'R technical specifications
| |
| - Automatic transfer relay which require one offsite and one onsite power should have energized this bus was supply be available at all times.
| |
| removed and placed in Unit 2 and ,
| |
| not replaced in Unit 1, loss of power to two buses resulted in two operable EDGs to start but loss of DC control to one of the trains i e prevoted closing of the EDG y output breaker. One EDG did energize one bus but EDG cooling 3 water pump was pavered by EDG which lost control power. EDCs ran for two hours without cooling water.
| |
| Connecticut Yankee One 115kv transmission line out See San Onofre 1 (6/7/73).
| |
| (8/24/84) for maintenance, one auxiliary transformer out for maintenance.
| |
| - Differential relay opened breakers to remaining auxiliary transformer.
| |
| Point Beach 2 (10/22/84) Breaker alignment errors during ABk'R design doe.s not allow cross-ties cross-tie between units caused between plants.
| |
| IDOP.
| |
| t
| |
| | |
| 1 TABLE 19QC LOSS OF OFF-SITE FOUER FRECURSORS 1
| |
| -Continued-i ,
| |
| _ 't
| |
| !. EVENT DESGIPTION APPLICABLE ABWR FEATURES f Indian Point 3'(11/16/84)' Object from roof fell onto startup See San Onofre 1~(6/7/73).and Ginna [
| |
| transformer. (3/4/71).
| |
| Turkey Point 3'(4/29/85) .Startup and C transformer were See Indian Point 2 and Yankee Rowe
| |
| ' both out of service. L Offsite (11/9/65).
| |
| . power supplied through main trans- t former. Relay failure resulted in f j.
| |
| l loss of main transformer and ;
| |
| ! LOOF. One EDG started and loaded >
| |
| l its safety bus.' ,
| |
| See Indian Point 2 and Yankee Rowe I
| |
| , Turkey Point 3 (5/17/85)- Brush fire disabled station.
| |
| t 'g- (11/9/65).
| |
| 7, .
| |
| : j. fg Vaterford 3'(12/12/85) Lightning caused loss of preferred ' See San Onofre 1 (6/7/73). ,
| |
| i offsite power source Two EDGs- i f
| |
| I started and loaded. Two sources of offsite power were available. !
| |
| i k I ,
| |
| { Fort Calhoun.(3/21/87) 1 EDG and alternate offsite power See: Indian Point 2 and Yankee Rowe !
| |
| I I
| |
| source were out for maintenance. (11/9/65) and Palisades (1/8/84).
| |
| ; controls for other EDG bypassed to-I prevent auto-start. Maintenance error tripped off-site power; EDG l
| |
| had to be manually loaded.
| |
| Yankee Rowe (6/1/87) Maintenance error caused loss of 2 See Sequoyah 1 (3/26/84).
| |
| f of 3 safety buses. l t
| |
| ! 'I l.i i.
| |
| I i
| |
| l i ,. . , - - _ . ._ __ -- , -. - _ __ -. . _ _ . _ _ _ . _ . - - _ _ _ - - _ _ _ . _ _ - .
| |
| 5
| |
| | |
| TABLE 19QG-1 IDSS OF OFF-SITE POWER TRECURSORS
| |
| -Gontinued-DESCRITTION APPLICABLE ABVR FEATURES EVENT 1 offsite power source and 1 EDG See Indian Point 2 and Yankee Rowe McGuire 1 (9/16/87) (11/9/65).
| |
| out for saintenance.- Test error caused loss of other offsite power source. Remaining EDG started and loaded. Offsite power restored
| |
| - after 25 minutes.
| |
| 1 EDG out for maintenance. Safety See Indian Foint 2 and Yankee Rowe Crystal River (10/14/87) (11/9/65) and San Onofre 1 (6/7/73).
| |
| buses cross-tied. Maintenance Also, ABWR design does not allow safety error caused loss of I safety j bus. Cross-connect breaker then buses to be cross-tied. The re f o re , this tripped and locked out. Dead bu- event cannot occur in AB'JR.
| |
| transfer was required to close one
| |
| .g cross-connect breaker. This j]
| |
| required shutting the running EDG 4 and resetting the under voltage lockout.
| |
| I safety bus and its EDG out for See Indian Point 2 and Yankee Rowe Crystal River (10/15/87) (11/9/65) and San Onofre 1 (6/7/73).
| |
| maintenance. Maintenance error grounded offsite power supply.
| |
| Remaining EDG started and loaded.
| |
| I safety bus and 1 EDG out for Anti-pump circuitry has been redesigned in Wolf Creek (10/16/87) maintenance. Error deenergized the ABVR to allow closure following other bus. EDG output breaker breaker trip when required.
| |
| opened and would not close due to anti-pump circuit preventing reclosure once it had been opened ,
| |
| after EDG started on under- j voltage. DHR lost for 17 minutes. l 1
| |
| i i
| |
| | |
| i TABLE 19QC-1 IDSS OF OFF-SITE POWE1 PRECURSORS
| |
| -Continued-EVENT DESCRIPTION APPLICABLE ABVR FEATUFES Oconee 3 (9/11/88) All off-site power going through 1 See Indian Point 2 and Yankee Rowe breaker. Maintenance error caused (11/9/65). ABVR RPV water level instru-this breaker to open, and it could ments are powered by batteries and at not be reclosed. No instruments least two divisions sre required to be to determine actual level and tem- operable during shutdown to support ECCS perature of water in reactor core automatic initiation functions.
| |
| region (incore thermocouples not
| |
| - yet reconnected, and no power to RPV level transmitters).
| |
| Surry 1 and 2 (4/6/89) Electrical fault and transformer See Indian Point 2 and Yankee Rowe lockout. This de-energized one (11/9/65) and Point Beach 2 (10/22/84).
| |
| w safety bus in each unit. Unit 2 y EDG started and loaded. Unit 1 EDG control in manual.
| |
| 4 o
| |
| Diablo Canyon 1 (3/7/91) Maintenance error caused power arc See Indian Point 2 and Yankee Rnwe and IDOP. EDGs started and (11/9/65).
| |
| Ioaded.
| |
| Nine Mile Point (11/17/73) I transmission line out for main- See Indian Point 2 and Yankee Rowe tenance. Maintenance error caused (11/9/65).
| |
| loss of other line.
| |
| Pilgrim (4/15/74) Lightning caused loss of all 345kv See Indian Point 2 and Yankee Rowe lines. 23kv line remain energized. (11/9/65).
| |
| Pilgrim (5/26/74) All 345kv lines deenergized (cause See Indian Point 2 and Yankee Rowe unknown). 23kv line remained en- (11/9/65).
| |
| ergized.
| |
| | |
| TABLE 19QC-1 IDSS OF OFF-SITE POWER PRECURSORS
| |
| -Continued-EVENT DESCRIPTION APPLICABLE ABkV FEATURES Brunswick 2 (3/26/75) I train of 230kv buses for each See Indian Point 2 and Yankee Rowe unit out for maintenance. Relay (11/9/65) error caused breakers on all five lines supplying remaining buses to open.
| |
| Quad Cities 2 (2/13/78) Reduced voltage on grid caused See Indian Point 2 and Yankee Rowe under-voltage relays to trip (11/9/65). ABkT has an alarm at 95% of breakers on both safety buses. rated voltage (degraded voltage). This Systen dispatcher incraased grid gives operator 5 minutes to restore full voltage. voltage before offsite breakers would open.
| |
| $ FitzPatrick (3/27/79) Maintenance error caused IDOP. See Indian Point 2 and Yankee Rowe
| |
| [ (11/9/65).
| |
| e Browns Ferry 1 and 2 Ice storm caused loss of both See Indian Point 2 and Yankee Rowe (3/1/80) off-site lines. Power supplied by (11/9/65).
| |
| Unit 3. ,
| |
| Monticello (4/27/81) 4.16kv breaker was racked out See San Onofre 1 (6/7/73). [
| |
| under load. Breaker then shorted, causing loss of both safety buses.
| |
| Quad Cities 1 (6/22/82) Not really an event: Unit 1 See Browns Ferry 1 and 2 (3/1/80) .
| |
| supplied Unit 2 when Unit 2 scrammed.
| |
| Pilgrim (10/12/82) Storms failed 345kv lines. 23kv See Indian Point 2 and Yankee Rowe remained energized. (11/9/65).
| |
| -~-
| |
| | |
| TABLE 19QC-1 LDSS OF OFF-SITE POWER PRECURSORS
| |
| -Continued-3 EVENT DESCRIPTION APPLICABLE ABWR FEATWES Brunswick 1 (4/26/83) 1 offsite power source out for See Indian Point 2 and Yankee Rowe test. Maintenance error caused (11/9/65)..
| |
| loss of second source resulting in i IDOP.
| |
| Fort St. Vrain (5/17/83) 1 EDG out for maintenance. 2nd For the ABWR, the CTG could be used to EDC in parallel with off-site power one of the safety buses if ofisite l power. Storm caused LOOP and 2nd power was not secure. In event of IDOP EDG tripped on overcurrent due to from any sources, features described under a faulty load sequencer and operat- Indian Point 2 and Yankee Rowe (11/9/65) ing non-essential loads. would mitigate the event.
| |
| 4 t
| |
| 5 Pilgrim (8/2/83) Lightning caused loss of all See Indian Point 2 and Yankee Rowe l
| |
| @ 345kv. (11/9/65).
| |
| l--
| |
| See Ginna (3/4/71) and Sequoyah 1
| |
| " Oyster Creek (11/14/83) Fire caused loss of power to 1 J
| |
| startup transformer. Switchyard (3/26/84).
| |
| 1 deenergized to permit cleanup.
| |
| Main generator disconnect links were removei. which allowed for use of unit transformer if neces-sary (wasn't used).
| |
| Monticello (6/4/84) I reserve transformer, I safety See Indian Point 2 and Lnkee Rowe l
| |
| j bus, 1 EDG out for maintenance. (11/9/65).
| |
| Prc::edure error caused loss of energized bus.
| |
| 4 t
| |
| i' . - - _ -
| |
| | |
| TABLE 19QC-1 LOSS OF OFF-SITE POWER PRECURSORS
| |
| -Continued-l EVENT DESCRIPTION 8PPLICABLE ABW FEATUPES Quad Cities 2 (5/7/85) Unit 2 dedicated EDG cut for main- See Indian Point 2 and Yankee Rowe tenance. Maintenance error caused (11/9/65) and Browns Ferry 1 and 2 LOOP to Unit 2. Unit 1 plus swing (3/1/80).
| |
| ! EDG powered Unit 2.
| |
| Millstone 1 (11/21/85) Reserve station transformer out See Indian Point 2 and Yankee Rowe for maintenance. EDG out for (11/9/65).
| |
| maintenance. Maintenance error caused loss of 345kv supply.
| |
| Peach Bottom 3 (4/13/86) Explosion and fire in transformer In the ABWR e sign, loss of the preferred caused loss of I startup trans- offsite power source would result in all y former. Alternate startup trans- three emergency diesels starting and pick-former supplied power. ing up respective IE buses. Power could Q be manually transferred to the alternate 4
| |
| " preferred power source (reserve trans-former) If desired depending on of fsite t
| |
| power reliability.
| |
| i Hope Creek (5/2/86' 2 of 4 EDCs out for maintenance. See Indian Point 2 and Yankee Rowe 1 of 3 off-site line out for main- (11/9/65).
| |
| tenance. Inadvertent relay actua-tion caused LOOP to safety buses.
| |
| i Pilgrim (11/19/86) Storm failed all 345kv. 23kv See Indian Point 2 and Yankee Rowe j remained energized. (11/9/65).
| |
| ]
| |
| Pilgrim (12/23/86) 1 345kv out for maintenance. See Indian Point 2 and Yankee Rowe Flashover caused loss of other (11/9/65).
| |
| 345h r. 23kv still available.
| |
| i.
| |
| | |
| TABLE ISQC-1 IDSS OF OFF-SITE POWER, PRECURSORS
| |
| -Continued-EVENT DESCRIPTION APPLICABLE ABUR FEATURES shoreham (3/18/87) 1 of 3 EDCs out for maintenance, See Peach Bottom 3 (4/13/86).
| |
| I safety bus out for maintenance.
| |
| Current transformers shorted as a safety measure. This unbalanced relays serving both service trans-
| |
| - formers, but without actuating differential current relays. 3 weeks later, condensate pump start caused differential relay trip, opening breakers from service transformer. Automatic fast transfer to reserve service trans-e former occurred, but unbalance c8 caused it to trip. 2 EDCs started 7 and loaded.
| |
| a Pilgrim (3/31/87) 134Skv ring bus breaker out for See Indian Point 2 and Yankee Rowe maintenance. 1 34 Sky line lost (11/9/65).
| |
| due to storm. Other line isolated due to resultant breaker open-ings. 23kv line still available.
| |
| I Peach Bottom 2 & 3 Lightning caused loss of 1 of 2 See Peach Bottom 3 (4/13/86).
| |
| (7/10/87) off-site. This caused loss of I startup transformer. Other trans-former remained in service.
| |
| 2 4
| |
| I
| |
| | |
| m TABLE 19QC-1 IASS OF OFF-SITF. POWER PRECURSORS
| |
| -Continued-EVEhT DESCRIPTIO$ APPLICABLE AB'w'R FEATURES Vermont Yankee (8/17/87) Both startup transformers and 1 of See Indian Point 2 and Yankee Rowe 2 345kv main generator output (11/9/65).
| |
| ' breakers out for maintenance.
| |
| Main generator disconnect links were removed. Unit auxiliary transformer energized by main
| |
| ~
| |
| transformer. Upset on grid caused other output breaker to open, causing IJDOP. EDGs started, and
| |
| . backup source was still available.
| |
| Pilgrim (11/12/87) 23kv line out of service. Snow See Indian Point 2 and Yankee Rowe y failed both 345kv lines. Startup (11/9/65).
| |
| .g transformer deenergized due to 4 arcing. EDCs started, and power w' was restored by removing main gen-erator disconnect links and backfeeding to auxiliary transformer.
| |
| FitzPatrick (10/31/88) 1 115kv line out for maintenance. See Indian Point 2 and Yankee Rowe ,
| |
| High winds interrupted other 115kv (11/9/65).
| |
| line. EDCs energized safety buses; efforts were directed at ,
| |
| other systems, so shutdown cooling was unavailable for 95 minutes (RCS temperature increased 10 deg. F).
| |
| | |
| TABLE 19QC-1 IDSS OF OFF-SITE POWER PRECURSORS
| |
| -Continued-EVENT DESCRIPTION APPLICABLE ABVR FEATt'RES Nine Mile Point 2 1 115kv line out for maintenance. See Indian Point 2 and Yankee Rowe (12/26/88) Current transformer failure caused (11/9/65).
| |
| loss of other line. Out of ser-vice line was returned to service and EDCs also started and loaded.
| |
| Pil oria (2/21/89) 345kv lost due to cable failure. See Indian Point 2 and Yankee Rowe 23kv line available, SB0 EDG (11/9/65) and Sequoyah 1 (3/26/84).
| |
| available. Disconnect links removed for backfeed.
| |
| Browns Ferry 2 (3/9/89) . Bus fault on secondary side of See Cinna (3/4/71) and San onofre 1 j g station transformer. EDCs (6/7/73).
| |
| gj started.
| |
| L Main generator disconnect links AEVR undervoltage load shed system will c'
| |
| Millstone 1 (4/29/89) removed. Loads had been trans- not inadvertently trip 6900 volt loads.
| |
| ferred to station service trans- ABVR undervoltage relays sense power on 1 former. Design error in relay of bus indepe dent of source.
| |
| : load shed system caused opening of 1 4.16kv breakers when reserve sta-j tion transformer was deenergizes.
| |
| Normal station transformer remained energized.
| |
| Browns Ferry 1 (5/5/89) Ground faults opened breakers from See Peach Bottaa 3 (4/13/86).
| |
| 1 500kv switchyard. Off-site power restored to safety buses from
| |
| ) 161kv iwitchyard through startup trans fo rmer.
| |
| i l
| |
| | |
| TABLE 19QC-1 1 DSS OF OFF-SITE POWER PRECURSORS
| |
| -Continued-EVENT DESCRIPTION APPLICABLE ABVR FEATt*FES 1 safety bus out for maintenance. In the ABWR, the two operable emergency VNP-2 (5/14/89) buses could have been energized from 2 EDGs out for maintenance.
| |
| Operator error caused LOOP to either the combustion turbine generator or other safety buses. EDG started the alternate preferred offsite reserve and loaded 1 safety bus. trans fo rme r.
| |
| River Bend (6/13/89) 1 of 4 preferred transformers See WNP-2 (5/14/89).
| |
| out. Maintenance error tripped 1 preferred transformer, causing loss of power to 1 safety bus.
| |
| EDG started and loaded. Mainte-nance stror tripped main generator 5 output breakers, causing LOOP to
| |
| $$ non-safety buses.
| |
| k ABUR has two offsite power sources, three Oyster Creek (3/9/91) One EDG and 1 bus out for mainte-nance. Routine check revealed diesel generators, and one combustion tur-other EDG had faulty head gasket bine generator.
| |
| which would have caused failure if required. This left plant with only 1 source of power, the startup transformer.
| |
| Vermont Yankee (4/23/91) LOOP due to improper maintenance ABVR procedures do not allow independent in switchyard. While installing a vital buses to be cross connected. The new battery on non-1E 125 VDC bus, multiple sources of on-site and of f-site two vital DC buses were cross con- power reduces the need to attempt cross nected through a battery charger connecting buses. The ABWR has four after defeating a mechanical physically separate and independent 125 interlock. When the battery VDC systems.
| |
| charger breaker was opened to install the new battery, a voltage
| |
| | |
| TABLE 19QC-1 IASS OF OFF-SITE POWER PRECLEORS
| |
| -Continued-EVENT DESCRIPTION APPLICABLE ABUR FEATtPES transient was sent through the entire DC control power system which caused both off-site power breakers to trip and lock open.
| |
| Diablo Canyon Unit 1 IDOP caused by boom of mobile ABUR has two independent preferred sources 3/7/91 crane shorting out 500kv trans- of off-site power.
| |
| i forme r. Standby startup trans.
| |
| former was out of service for maintenance. The three EDGs started and picked-up vital buses. Off-site power was i
| |
| >> restored in five hours.
| |
| e 3 Nine Mile Point LDOP while working on aux. boiler ABWR offsite power supplies are physically 4
| |
| a circuitry. Div. I diesel was out and electrically separated so loss of both :
| |
| 3/23/92 is not expected to occur due to common l for maintenance. Div. 11 diesel started and loaded. Div. Ill cause failure. Three independent electric
| |
| ?
| |
| (HPCS) started but tripped on over divisions (including instrument UPSs) '
| |
| temperature due to lack of cooling would reduce likelihood of simultaneous water.. All control room annuncia- failure of all three divisions.
| |
| tors lost due to loss of A and B t UPS.
| |
| t i
| |
| b i
| |
| i t
| |
| i
| |
| ~_
| |
| | |
| - ABWR Single Line Diagram Swtch Yard
| |
| : t. b Rev. 8l Mar 24,1992, JEM MPT 1500 MVA W
| |
| "4 g-
| |
| --> - - - --- -o o-- -
| |
| &- In N M'
| |
| OFFSITE UATA 37.5 MVA RAT o 37.5 WA OA65C m OA65C UATC mNNA MAIN L_ J
| |
| ^ ^d^ ^ 37.5 UATB MVA ^ ^1^ ^ 37.5 MVA ^iN^^
| |
| O -, OA65C i, OAS5C i 81 82 B3 C1 C2 C3
| |
| ')
| |
| A1 '
| |
| A3 (PG) j'')O) V (PG) ' ')0}'
| |
| A2
| |
| ')0) ' ') '
| |
| ') (PG) n (PIP) O
| |
| ')(PG)n (PIP) 9) ")(PG) n (PIP) 9 l JPG)'
| |
| RIP (2) CRD 8 RIP (2) CP CRD 0 RIP (3) l CP(2) i RIP (3)
| |
| TCW7} CWP : RFP HNCWo)
| |
| TCW 9 CWP CP TCWo) CWP RFP TSW ; '
| |
| CTMP TSW '
| |
| CTMP RFP TSW '; CTMP l ,
| |
| HNCW.(2) CTBP i .
| |
| HNCW (2) CTBP
| |
| ! -[. CTBP HDP j ' ,'
| |
| Op HDP l .
| |
| i MVP i
| |
| ; , i i .
| |
| ^ '
| |
| ^ ^
| |
| -------------S-------------'
| |
| ---o-c CTG . , . . :
| |
| ' i i i i 8
| |
| (9MW) .i i. i.
| |
| k I
| |
| ...---.-..-...--....i L.---.---...--- . . . . ,
| |
| i : .
| |
| ' i i i i .
| |
| - - - f- T - - -- - - - - - -- - - - - - -~ ~ i - ~ ~ r -
| |
| Racked Out Breders .~i E (D1)
| |
| ")r!i
| |
| ' r' o) ? F (D2) '")
| |
| ' *)
| |
| i o)
| |
| ? G (D3)o)? ' r' ") ''
| |
| ) ) )
| |
| (SMW) ( DG3 DG1 DG2 (D3)
| |
| (D1) (D2)
| |
| \ / \ / \ /
| |
| LOAD GROUP A LOAD GROUP B LOAD GROUP O cm e - .-
| |
| | |
| B' t
| |
| 1 I . EVENT CATEGORY: . LOSSES OR DEGRADATION OF RNRS DUE TO LOSS OF COOLANT FRGE REACTOR WESSEL A .
| |
| INITIAL PLANT COWITIceS EVENT DESCRIPTION REPORTED CAUSE APPLICASLE Asia FEATURE
| |
| : PLANT LER/DATE l .
| |
| I
| |
| ~ A slidt reactor unter levet drop uns - Feiture of the miniese ASIA component design and
| |
| ~
| |
| { Peech Sottom 3 flode 4, Cold Shut-79-002 -
| |
| doun. RuRS in ,p- : detected and determined to be coured tyr flow recirculation procurement ultt emphesiae
| |
| [ vetwe associated with ;
| |
| : i. l January 8, 1979- eretien on too;. 'A', teekage throuch the miniense flou recircu-- fabrication spelity and proper
| |
| ; letion vetve for the 'A' RNR pump the 'A' RNES pimp.- maintenance to minimize
| |
| )'
| |
| (No-164). Vesset tevet uns maintained tyr indivi& et component failures. -
| |
| use of the stay futt pressuriting system. Womever, if feiture occurs, Soc Attempts to eliminate the teskoge ty fur- . would be temporarity lost tiut '
| |
| I ther closing the miniese flou vetve two other Riet trains would be 1 reeutted in its felture to the ulde open eveitable to re-establish DM
| |
| ) position. This feiture caused a loss of before ery fuet demose j-' cootent to the suppression pool. ' The occurred. In addition other
| |
| * loss of veneet water tewet continued to heet removal systems (e.g., ,
| |
| l the point of footation of the shutdoun fuel pool cteerue and cooting l j cooling system on tou unter tewel, et (FPC), reactor meter cteerup) iditch time the unter tevet stabilized. ere eveitable for DM depending .
| |
| j The time respired to raise the reactor on plant conditions. Other r unter levet, wie the stay full system, makeip sources (e.g., ePCf. i cleer the Ruts isoletion eruf reestablish fee $seter, AC 1 4 M water l} shutdann cooling ulth the 'C' Rats pimp, addition, CROS) cari be used if J3 no cut system is eveitable and ettomed the cootent to rise to about l?
| |
| a --a - 2006F, ceueing a geoeeus releese vie dis- the reactor cootent begins to esseeaded RCIC steem isoletion vetwes. tioi t .
| |
| i8-
| |
| \
| |
| 3 notch 1 feede 5, Refueling. The 'B' toap RNRS use pieced in service tone reported. See Peech Sottom 3 (1/8/79). i Aurest 13, 1979 RMRS in operation. - . In the shutdann coeting mode and vesset l tevet une observed to be dropping. Velve i
| |
| l E11-F0064 uns determined to be teeking to the sigipreseion poet. A toce1 ieek rete test of the RNRS 'S' pump torus suction :
| |
| i i isolation votwe shomed the volve to be teeking in encess of specified criterie.
| |
| } ;
| |
| t Foltouing corrective action, the volve i was satisfactority retested.
| |
| t I
| |
| o I
| |
| j- t
| |
| , i i
| |
| 1 - r c-, -- - - - ,- - -, . - . __.
| |
| | |
| EVENT CATEGORY: LOSSES OR DEGRADATION OF RRRS DUE TO LOSS OF CDOLANT FROM REACTOR VESSEL INITIAL PLANT REPORTED CAUSE APPLICA8tE ABWR FEATURE COMolTIONS EVENT DESCRIPTION PLANT LER/DATE Circimferentist through See Peach setta= 3 (1/e/79).
| |
| pode 5, Refueling. This event consists actuelty of two Oyster Creek separate events irwetving shutdown esatt cracks in one ttbe 81-038 Rnts system in of the 'A' heet ex-operation on loop cooling heet exchanger ttbe teoks. On August 27, 1981 changer and one ttbe of August 28, 1961 'C'. Reactor had August 27, with reactor water teeperature the 'C' heet enchanger, been shutches for 13 et 197&F, the 'C' shutdown cooling best die to fetigue fsiture devs. exchanger developed a tube ieek resutting in reactor water teeking into the RSCCW caused ty flow indaced system as indicated by the RSCCW process vibration.
| |
| { radiation monitor. About 2 mirastes l (Mer, reactor water level began to i decrease. The decrease occurred over approsimetely 10 minutes, with an esti-I ented teek rete of 400 gpe. Reactor vesset water level was recovered by make tp surptied by the feedwater and corden-seta system. The 'C' toop was secured i and tesperature saintained below 212er ty j use of the 'A' shutdohet Cooling loop.
| |
| s I e On August 28, another BBCCW process N monitor alene was received and the RECCW 4 surge tank was reported to be overflow- J o irg. The 'A' shutdown (coting tocp was isolated. The 'B' heet enchanger was out of service but was made serviceable in a few hours. Temperature was eninteined by increasing flow to the RWCJ a-es e.e-tive best enchanger and increasing letdown to the metre condenser. Water was ptmped back to the reactor using a etn-densete pump. In addition to RWCU and seein condenser systems, the isolation condenser and ECCS systems were eti evellable.
| |
| Leeking flenge on spoot ASWR has three int >pmdmt BNR Mode 3, Not White placing RMF *A* toop in service in toops. Also, the main LaSalle 1 the shutdows cooting mode, leakege was piece at *A' RMR puno condmser ord RucJ ere capabte 82-039 shstdows. Plant Riet pig suction suction line, caused by June 9, 1982 cootdown in progress discovered at the 'A' thervuit growth on of removing decay heet in suode j RNR toop 'A' being line. RNR toop 'A' was teken out of 3.
| |
| 1 service for repairs. Alternate methods heettp and cooldown. I ptoced in service. of decoy heat removal were reactor recire (Prior to initial pumpe end irboerd mein steen Line drain crit 8cality.)
| |
| with RWCU.
| |
| l t
| |
| | |
| ! EVENT CATEGORTt LOSSES OR gEERAGATION OF anRS DuE To Lost OF C00UMif FROM REACTOR VE5SEL
| |
| - talTIAL PUMIT ;
| |
| i PLAuf LER/DATE cimetTIONS EVEuf gESCRIPTION REPORTED CAUSE APPLICAsLE Asut FEATURE 7 1
| |
| LaSette 1 stede 4, Cold The mit was in cold shutdoin fetteming Perseroiet did not ree- AguR processres mitt cleerly )
| |
| 82-042 Shutdann (Prior to perforemnce of reactor internets vibre- ognize the potentist describe proper operationet !
| |
| i **eps eius the technicet spect-June 11, 1982 initlet tien testing. *g* Rat system was operet- vesset drein path that t criticality). Ing in the shutdanni cooling mode with ett esisted w on returnity fications will be based on ff - J1ypnosing the 'e' RNR best enchanger the system to e nornet minimizing plant risks dsrirg ti - 'ntain reactor temperature between - linee from staney. normat futt pouer operation and 14C - and 2006F. The 'A' anR system uma operation. The test shutdonc conditions. The Agut I lined g for sten 6r shutdann cooling. procedsre feited to hos three i. 4 M ane '
| |
| The 'A' and 'g' RgR segpression pool suc -- recognize the current systems arus SDC is isolated on tion volves mere out of service electri- operating status of the' tou Rev tewet.
| |
| cetty for repelr arut the velves were Rat system in shutdoun
| |
| . manustty ctosed. se beettes seens of cooting. .The tevet
| |
| ! decer heet removel mes eveitable ese to instruments tap of f the i the reactor buildire ctooed cooling unter douncomer region idwre system belce out of service. (no actuet . shutdonc cootirg j decay heet entsted.) receives its suction. I q
| |
| The Tech Specs esere N Testing of the 'A' R51 drywett sprey interpreted such thet 3 4
| |
| y outinoord f ootetter- otve was approved and performed to accondarme mith ,- mM.
| |
| both shutdonc cootfrg Ioops more required j n After the test isos tengdeted, the systee operstde with one in
| |
| ; h operation, and that the
| |
| ! H uns returned to stan6y operation. The i resterotion processre directed the- idte pimp could be out opening of the Rua 'A' heet enchanger of service for crity 2 4 bypeos vetve. Wien this volve use hours. This wee a epened, water from the reector vesset conservettwo interpre-
| |
| {
| |
| fitted the previously drained RNR 'A' totion but it aggre- ,
| |
| ]-
| |
| i pipiris, dreining about 3,000 settens of voted the event by aseter from the asset. At 12.5 inches leposing an orbitrary [
| |
| i level, en aut a tic footetton of the time restreint on the I shutdoom cooting system occurred. The test. l
| |
| ' vesset Rat looptewet was was restored, verified andvented fitted and the 'e' , -
| |
| ; and shutdoem coolins system suction isolation vetwes reopened. Reector i vesset tewel agJin decreased to about 10 l t
| |
| inches arut a second isoletion occurred. - :
| |
| It uns determined that this second [
| |
| } ' lootetton reeutted from the startirg i 1
| |
| transient and resulting tewet drop in the i
| |
| i douncemer region. Vesset levet ases again i restored; and shutdoom cooting unisc- +
| |
| teted, vented, and restarted; and the 'A' [
| |
| i RnRs toop determined operable. j 1
| |
| I L
| |
| i !
| |
| i i
| |
| i
| |
| - . . . ,~ ,
| |
| | |
| EVENT CATEGnRY: ' LOSSES tR DEGRADAf t0N OF SuRS DLE TO LOSS OF CDOLANT Fatst REACTOR VESSEL' letilAL PLANT '
| |
| CtNSITIONS- EVEnf DESCRIPTI(31 REPORTED CAUSE AFPLICASLE ASWR TEATURE PLANT LER/DATE' The potentist for tteis operator Grand Gulf . Stade 4, Cold - Leap 'A' of the esas uns lined age in tfie Operator error misin .
| |
| N/A ' Shutdanet, efter LPCI mode, and toep 'O' isos Lined ey in tegretation of wetwe error has oeen eliminated in
| |
| _ April 3,'1983 .Inittet criticality. time shutdones cooling made for a surveit- positten indicotion. the Asut destyi isr providiN Rems Loop 't' in tence test. After canpletten of the F006 ' fully open* Indi- volve intertecks. neien 9pm j Shutdeun Cooting. test, the operator returned '9' toep to cator tight ises not - systems is in the shutdemos l the LPCI mode, telch roussired shuttirig burning, but neither cooline made (i.e., taking !
| |
| the toep 'B' W C suction volve (F006) and' . uos the "futty shut" suctiers fress the arv), the l opening the toep 't' sagspression poet Indicator. Vetwe esos discherpe wolves to the i suction volve (F006). Since o tight bulb probotzty in a partletty suppression poet are. }
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| ' uns burned out on the open indicator for operi position. Reason int ~aceked in the closed F906, tfie operator eseissed that F006 esos for F006 now breaker posMon to prevent inadvertent -
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| streedy shut, and opened F006. This ' trip not empteined. draining of the arv. To eponed a flou path from the reactor reettyi to the Lou Pressure vesset wie the 't' Est loop to the Flooder (LPFL) mode, the -
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| suppreselon poet. Approminertely 10,000 sigipression pool suction volve !
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| gotions of meter drained from the reactor cormat be opened until the SDC suction valve is fully closed. 'F wesset prior to outematic footetten of
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| - the RERS en tone esoter tewel. The opere-tor ettesyted to reehut F006 upon receiv- !
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| g ,
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| Ins e tons towet etene, but the wetwe's :
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| 9J su0t breaker tripped. ,
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| " i t
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| I i
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| k f
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| i t
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| L i
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| I i
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| b I
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| EVENT CATEGORY: LOSSES OR DEGRADATION OF RMS DUE TO LOSS OF COOLANT FROM REACT 0m VESSEL INITIAL PLANT PLANT LER/DATE CONDITIONS EVENT DESCRIPitDN REPORTED CAUSE APPLICASLE A8um FEATLArE Susquehenne 1 mode 3, not During a startw test to determine the Reactor Ceotent Systco See LaSatte 1 (6/11/82). Rat 83-056 Shutdom. capability of the shutdoun cooting mode shrinkoge caused by valve miset(gruents are April 7, 1963 of R M , the 'A' a m heat exchanger was rapid temperature minimized in the Agut design by valved in cousing a rapid temperature decrease. Watve line m mode switches for the five decrease. As a result of RPV water errer caused loss c* eperational RM modas.
| |
| voltsee shrinkege, the RM automaticatty triventory to stspres- Selection of a made (e.g., SCC isotated on tow reactor water level. CRD sien pool. causes automatic volve flow was used to restore levet; and reetlyments).
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| MStV's were opened to decrease the vesset dette-T. dutu was established to stop stratification. E n loop 'A' was restored, tut a velve linew error caused the ptsup miniflow valve to bypass RH flow to the stcpression pool, causim a second RM isolation on tow level. Level wes restored and R W reinitiated, but the inventory e&fitlen via condensate trans-M fer caused another temperature decrease
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| .y of 120e in 5 mirutes, so the RR system was isolated a third time to helt the n
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| O cooldown. The system tses restered egeln, Las and a fourth short isolation was received then starting the '1' RM pg.
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| LaSette 2 Cold shutdown. With the contret red drive system in ser- Using an tausuet volve $*e LaSatte 1 (6/11/82). The N/A Preoperationet vice ard the reactor water cteerup systese tirep erut bypesoing A8WR design has adeqmte safety August 15, 1983 testing prior to out of service, reactor water level was automatic safety features. However, trusuet fuel toed. being controtted try draining through the features. volve liveps and bypessirg of RMS '5' toop to the stapression poet. A safety features should be new drain path was being established wie perform =d urder strict the .' A' RM toop (F004 and F006). As administr6tive control.
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| soon se O.h m drein peth was timd w, the reactor vessel began draining rapidly. The event did not terminate automaticetty on tow RV water levet isolation of RNRS, because the low leve!
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| Isotation signet had been bypassed by transferring contret for the R=G shut h cooling isoletion volves to the remote shutchwn penet. this was done intention-etty to preve6t inadvertent isolations of the temporary drain path. The toss of cootent evant was terminated by operator oction, 32* ebove the top of the fuet region (fuel had not yet tw toeded).
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| j
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| 1 >
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| j '!
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| 1 I
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| 4
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| . EVENT CATEGORY:- LOSSES OR DEliRADATION OF RNRS IFK TO LOSS OF CDoutui FRon aEACTOR VES3EL [
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| 1 L
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| I tulTIAL Putui j PLANT LER/DATE. CDelitout EVENT DESCRIPTION REPORTED CAUSE APPLICA4LE ASWR FEATURE l
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| I
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| $ Lesetle 1. Cold shutdomm. enRS RuRS op d te, but shutdoun conting Trip fingers which bold See Peach settan 3 (1/3/79).
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| statue not stated. R4RS ptap 'A' minimum tSe actor operation in I 83-106 . operable.
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| September 1,1983 flow bypass velve (F0664) stueit open fot- hand &eet operation .
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| towirg a test. If shutdoun cooling wee were foemd brsken.
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| lined up to toop 'A' then a drain path to Velve actor dmenged. '
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| i the segression pool existed.
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| LaSatte 1 Cold sheldown. itM togic testing was in progress thich The LPC1 injection AguR component design ard
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| .83-105 respaired openine most toop a injection check votwe was stuck procurement will eesize September % ,. ,
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| and sprey vetves: drywett oprey velves open. Inspection of f abrication gastity ord proper [
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| 4 1963- (F016e end F01?B), soppressian pool aprey the vetwe reveeted meintenance to minis he !
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| ve!we and test return vetwas (F02?B and improper maintenance on individaat cessenent feitures. ,
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| i F0268), and 8 and C toop injection vetwes the votwe operetor. RMt topic testing does not .!
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| (F0628 and F062C). This lineege retted on The vetve had been require that RPv isotation rely i testable injection check volve F0618 to reessent>ted by tining on a single check vetwe during .
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| t prevent reactor vesset inventory loss wie to the wreng mark on RMt legic testing.
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| injection velve F0628 to the open sprey the spline shaft 'o the F
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| H and test return lines. When F0629 was - eir operator gears, opened, reactor wesset inventory ions rap- which held the check l
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| {* Idly tost to the drywell and swession waive 354 open. The i pecking gland was also !
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| }O A
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| pool because the testable check velve was stuck open. nost of the water lost from too tight to permit i
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| the reactor vestet went to the eigpres- futt closure. !
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| slon pool. The opeester terminated the- l j event af ter a 50= tewet drop to about 4 1e0= above the top of the active fuel. l l
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| Totet inventory toss was between 5,000 and 10,000 settons. It should be noted [
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| that no automatic footetton feeture asould.
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| have terminated this flow path; hewever, l
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| ' the LPC! injection line penetration is ,
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| l etre the top of the active fuel. ,
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| I Quad Cities 1 Cold shutdown. The RPV tevet decreesed 14' inches in two Operator error in A0WR proced;res witt highlight f l reteted events. The shutdoun cooling misaligning amt vetwes. Amt system velve stiyuernts I 1/26/91 daring assintenance. The keep I suctiert volve was stroked es e mainte-nonce check but some vent and drain fitt pump ord pressure etsra i velves in the loop were etso open, when assures a futt loop.
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| l the 30C suction velve esos open the arv t j drained 5 incbas.- The operator isolated 50C to stop the flow but when the loop [
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| I was returned to service an additionet 9 I. inches were drained frase the RPV into the i partletty empty enR toop. >
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| 4 i b i
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| 4
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| [
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| 1 L
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| I
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| . _ _ , _ _ .__ . .._ __..__________D
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| a EVENT CATEGatY: Losses OR DECRADATICM OF RMRS DUE TO Loss OF C00LAuf FROM ftEACTOR VESSEL INITIAL PLA 3 RtPW iEP CA835E APPLICABLE A!DA FEstVRE CD40!TIONS EvtNT DESCRIPTION PLANT LER/DATE operator error in not Asvt twR volves are interlocted Cold shutcknet. On Af ter isolating RCu the RPV tevet tegen to prevent SDC suction ard Quad cities 2 to increase. Operators attempted to following approved pre-8/17/87 shuttkan coetire in cedare for dreining the injection velves fram beirg one RMR toon, reec- redJte tevet by draining to the sogpres- opm et the same time es the RPY.
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| tor water cteen 4p sion pool using the RNR system test surpression poot return setres.
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| (RCW) syetem out for return vetve (14 inch velve). This meintenance. resulted in rapid decrease in RSV to los levet setpoint and en autenatic RPY isolation.
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| Operator error in piec- As',a Rua system keep fitt storm Hot Shutdown follow- $DC tcop was being put in servite but would eleet operator to e per-Fermi 2 normat loop heat-tp stignment could not ing SDC toop in service 3/17/87 ing toop test, ene using t<we.+d tietty drained toco condition RNR toop incoerebte. be used because one volve would not open procedJre.
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| (24 inch testable check vetves). A smetter (1 inch) valve was used to fitt the loop but the normal 4 inch drain line caused dreinsge f aster then the 1 inct:
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| tine could fitt the loop. This drained the tooo but operater could not telt.
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| 'y i
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| Wen proper SDC teop temperature was
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| .) reached operator opened SDC suttion vetwe
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| ?
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| M to the RPV and RPV tevet decressed to the tow levet setpoint eM RPV isoletion occurred. 1 i
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| Operator erroe in ret A8we twR stspression pool !
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| Cold $hutdown. SDC During the process of shif tirg SDC from suction and SDC suction velves Fermi 2 Division !! to Division 7, a RPV icw f ollowing proper pr9ce-8/2/87 on Division II. levet signet occurred tecause vetves were dure placing SDC in are intertocked to prevent service. inadvertent RPV dreinege.
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| misaligned resulting in en open flow path to the soppression pool from the RPV.
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| tverstor error in not sworessico poot suction velve Cold shutdoun in White returning fras SDC to s:errty tow cemet be opered mtit SDC suc- I WWP-2 knowing that stroke '
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| pressure injection ande c twe, the tion valve is fully closed.
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| 5/7/85 SDC.
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| operator opened the sagprsssion pool suc- time for each volve is 90 - 100 seconds.
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| tton vaive before the SDC suction vstwe was futty closed. This opened a drain path from the RPV to the stspression goot resulting in e low RPV and SDC isolation.
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| see wep-2 White returning one Rut toop to stoney, see wuP-2 Shoreham Cold shutdown both 5/T/35. 5/7/25.
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| RNR toops in SDC eperator opened surpression pool suction 7/26/85 vatwe whiie SDC suction vs1ve was per-mode, tietty open (see nap-2 5/7/85).
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| EVENT CATECORY: LOSSES 9 MGRADATIcel 0F RNRS DUE TO LOSS OF COOLANT FROM REACTCR VESSEL INITIAL PLANT EVENT DESCRIPTION REPORTED CAUSE APPifft9tE AEWt FEATURE PLANT LER/DATE CONDITIONS Loop "C" SDC suction valve remained open Operator error in mt A8WR RMR toops are independent Peach Bottaa 2 Cold shutdown. SDC en "A" RNR toop. after previous SDC operation. toop *A" knowing status of Rim. c.4 cross train flow camot ..
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| 9/24/85 system valves. occur.
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| regaired a futt flow test tkJe to ptsp problem investigation. SDC -A* isotsted and "A" pump atigned to stqvression guct for test. This opened path fror RPV to s@ pression pool through *C" SDC suction valve.
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| 'Jhi1e restoring SDC tcop to atantby, sp- See WNP-2 See NNP-2 Riverbend toId Sht "km. 5/7/85.
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| pression pool suction and SDC suction 5/7/S5.
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| 9/23/85 valves were open at the same time.
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| While placing "A" SDC Jn line a path was Operator error i @ roper ASWR proCedJres will Clearly Susq;ehanna 2 Cold Shutdoun.
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| 4/27/85 open fma the RPY t9 the main cmdenser. valve line-@. describe proper valve'linews.
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| RPV tevet dropped 35 Irdes resulting in -
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| RPV tow levet signet and isolation of g
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| SDC.
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| g O
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| Cold shutdown. SDC po p mini-flow valve falted open vetve failure. SDC would isolate on tow RPV e Susquehanna 1 attowing water to flow fras RPV to sm- level.
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| $ 5/10/85 5/20/85 pression pool.
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| Cold shutdown. White warning- w SDC toop, en isolation Operator error. 59 The keep fitt sterm would stert WNP-2 signet eccurred on high SDC ftow. Opera- loop isolation not the operator to ? oartletty 8/23/84 etermed in control drained RMR tono tor did not notice and loop drained to the rachssste system. Whm operator room.
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| placed loop in service water drained from RPV into empty SDC to@.
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| RHR toop in test mode with severa* valves Maintenance error. AswR RNR system tests would not LaSatle 1 Cotd t shutdown. regJire att watwes be open and 9/14/83 open. Loop check valve dapended upon to isolate RPV. Check valve falted cpen dJe rely on check valve to isolate to mis-es:;eebly and improper packing the RPV.
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| gland instattation.
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| Operater attempted to tower su pression Oprator error. RNR system drain to the Brunswick 2 Cold shutdown. radweste system contains two 9/24/84 poot levet to redwsste but loop was in SDC mode and resulted in unter diversion vatwes in series that from RPV to radweste. automaticatty close on tu RPV t ew t .
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| EVENT CATEGORT: LOSS OF RHR$ 00E TO INADVERTENT CLOSURE OF RMES SUCTION VALVES INITIAL PLANT EVENT DESCRIPTION REPORTED CAUSE APPLICA8tE A8WR FEATURE PLANT LER/DATE CONDITIONS Pilgrim Mode 5, Refueling White performing maintenance n a feeder Electrical contact in See Peach Bottom 3 (1/8/79) and 81-064 RNRS in operation. transfoemer, a live transfer of power was the paup trip logic LaSatte 1 (6/11/82).
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| Decenber 21, 1981 Coolant temperature attempted. Mat-operation of a power were cc-~=4ed to the at 704F. breaker deenergized a vital instrument extent that they seized panel, causing two shutdown cooling in the open position.
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| valves (MO-47 and MO-48) to close on 'C' Rhts pmp, there-receipt of a. reactor high pressure isota- fore diri not trip when tion signal. The 'C' RNRS pu g should the suction valves left have tripped lamediately when its suction their futt open posi-valves shut, but falted to do so. After tion. Inadegancies in about 5 hours, when the process ctmputer the implementation of was returned to service, abnormal heat adsinistrative caitrots exchanger temperatures alerted operator < for shift turnover, to a prot > tem. At this time, the PC' R A valve linetp checks, pmp was observed to be rtening with both ard board checks aggra-suction valves shut. The 'C' ptmo was voted the situation.
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| tripped, the valves opened, and the ' A' Extensive maintenance puup started to restore shutdown cooling. activities distracted e
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| ;g operators.
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| o Mode 4, Cold shut- The RMRS was operating in the shutdown The Reactor Protection Loss of power does not cause Susquehanna 1 4
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| w 83-030 down. RNRS in op- cooling mode. A Division I isolation System (R*S) was oper- isotation of SDC in the ABWR signal to the inboard Isolation valve to ating on alternate design. The multi-plexed Fet>ruary 16, 1983 erotion on loop 'A'.
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| safety system logic will only the 2NRS caused a loss of shutdown cool- power stypties Uite ing. The systems was reestablished by the RPS MG set was cause isolation if a valid resetting the signals. A second occur- undergoing maintenance. Isolaticn condition existed.
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| rence was experienced within an hour. Spurious trips of the RPd alternate power stopty breakers caused isotation signals.
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| Mode 4, Cold Shut- An RPS actuation caused RWR toop 'B' RPS actuation caused by See Susquehanna 1 (2/U'83).
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| Susquehanna 1 83-060 down. RMR in opera- cperating in the shutdown cooling mode to an inadvertent breaker April 11, 1983 tion on loop '8'. Isolate. RNR pusp 'O' tripped twice on trip (tusped by a con-attempts to restart. RMR cooling was struction worker). The a
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| established again on loop '8' using pump restart trips are
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| '8*. believed to be due to a f aulty shutdown cooling flow switch.
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| EVENT CATEGORY: LOSS OF RNRS DUE TO INADVERIENT CLOSURE OF RHRS SUCil0N VALVES INITIAL PLANT-PLANT LER/DATE CONDITIONS EVENT DESCRIPil0N RCPORTED CAUSE APPLICABLE ABWR FEATURE Grand Gulf Mode 4, Cold Following electric 61 maintenance during the power stopty fuses A8WR solid state logic 83-069 Shutdom. (During iAlch some shutdown cooling noter-oper- to the isolation logic minimizes use of fuses and May 23, 1983 initlet plant ated valves were blocked open, power was had not been replaced togic testing is easier such starttip gAsse). restored, and the valves were tsv> Locked. following conptetton of. that these types of operator the valves isolated as a result of a a design change. errors will be re &ced.
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| previously existing isolation si val from the valve isolation logic, causing a' toss of both shutdoun cooling loops.
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| Grand Gulf Mode 4, Cold Both RNR shutdown cooling loops isolated The solid state trip ABWR has three independent unit for the crmumon (botft physicat*y and 83-119 Shut h e1. RNRS toop on two occasions during attewpts to start August 18, 1983 'A' in operation. a controt room air-conditioning compres- 480V trip breaker had electricatty) RwR systens. No (During initiet sor. The systems interaction was due to fatted. commm power stwties tetween plant start w a comunon power source to the compressor RHR systems exist.
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| phase.) and to tenkage detection logic cir-cultry, which caused the isolaticri.
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| Grand Gulf Mode 4, Cold The RHR$ isolated after shifting the RPS The distribution trans- Sr 2 bsquehama 1 (2/16/83).
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| g e 83-137 Shutdown. RHRS toop power stpply to en alternate source. The former en the tairego-Septenber 1,1983 alternate stypty breaker tripped, causing lated RPS alternate
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| @ 'A' in operation.
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| (During initial en isolation of shutdom cooting. power source was
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| $ plant start y gAsse.)
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| stbject to transients.
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| Grand Gulf Mode 4, Cold During an instrtaient surveittance on the The cause of the isota- See Grand Gulf (8/18/83).
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| 83-193 Shutdown. Isolation logic for shutdown cooling, the tion was a tip breaking AsWR solid state togic Decenber 27, 1953 outboard suction valve 17008) closed, off a minitest clip eliminates need for test -
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| isolating both toops of the SDC system. used for juppering. jtspers. Surveillance is The system was returned to service in 49 automated to reduce chance of minutes. operstor error.
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| Suscpehanna 1 Mode 5, 0% Power. During the Unit 1 - Unit 2 tie-in outage, The cause of the trip AaWR see Susquehanna 1 83-172 one of the RPS 'B' breakers tripped, was a fatted breaker. (2/16/83).
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| December 30, 1983 closing SDC inboard and outboard isolation valves. Reactor coolant recir-culation was es*ablished through the fuel pool cooling system.
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| Hatch 2 Mode 4, Cold Received a low RPV water levet signal Permomet error iri not ABWR procedures will clearly Septenber 19, Shutdom. white volving out a RPV levet indicator. placing levet specify required maintenance 1986 This resulted in a scram signal and transmitter in bypass steps and precautions to isolation of SDC. SDC was restoced in before volving out preclude inadvertent SDC 10 minutes. detector. i sol at f or..
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| I EVENT CATEGORT': LOSS OF RNRS DUE TO INADVERTENT CLOSURE OF RMRS SUCTION VALVES INITIAL PLANT ' APPLICABLE AgWR FEATURE CONDIYlONS EVENT DESCRIPTION REPORTED'CAUSE PLANT LER/DATE Mode 4, Cold' Lost SDC for 1.5 hours due to inadvertent' Surveittance procedure AguR solid state logic'does not.
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| Hatch 2 require the use of jumpers to September 21, shutdown. RNR suction velve isolation during a required rrm vat.of-1986 surveittance test. ' instruners Links complete circuit logic checks.
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| instead cf Jtsspering them out, m er..tinks were opened, a RNE:
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| valve isolation signet was initiated.
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| Mode 4, Cold White transferring RPS power to en Inadequate processre See Susquehanna 1 (2/16/83).
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| Perry 1 tober 24, 1984 Shutdown. aiternete bus to comptete RPS MG set for trensferring power maintenance, a voltage transient occurred between buses.
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| A lch resulted in isolation of 10C.
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| River gerd 1 Mode 4, Cold SCC vetve was inadvertently closed when Personnet error. See Susquehema 1 (2/16/83).
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| October 28, 1986 Shutdown. technician accidently groimded a portion of the velve contret circuitry daring a surveillance test. The grotsd caused a e
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| o bloon control circuit fuse dich resulted in a vetve closure signet.
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| SDC isolated ckJe to toss of power to RPS Voltage fluctuation dse See Suscpjehanna 1 (2/16/83).
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| $ Perry 1 Cold Shutdown.
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| bus. RPS was being powered by alternate to starting one of the power single MG set was in maintenance. plant's circulation water pumps caused electrical protection
| |
| . devices (EPAs) to trip resulting in toes of power to the RPS.
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| Mode 4, Cold White performing a reactor coolant system The breeker controtter See notch 2 (9/19/86).
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| Clinton 1 for the high pressure January 22, 1987 Shutdoom. hydrostatic teek test. An isolation of SDC occurred ckse to high system pressure, interlock RNR vetve was racked out prior to the test to prevent valve closure. Following the test, the trip function t
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| was not reset prior to rocking in the breeker.
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| When the breaker was racked in the valve closed due to'the locked-in high pressure signal.
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| 4
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| -a
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| EVENT CATEGORY: ... LOSS Of RNRS DUE TO INADVERTENT CLOSURE OF RNRS SUC410N VALVES '
| |
| INITIAL PtANT' .
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| EVENT DESCRIPTION, REPORTED CAUSE, . APPLICAstE AgWR FEATURE PLANT LER/DATE COWITIONS
| |
| ' mode'5, Refueling.' Isolation of SDC occurred & ring Maintenance proce & re See Suse;eheme 1 (2/16/83).
| |
| Peach Botton 2 March 28, 1987 nelntenance on emergency bus reteys.
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| cetled for putting
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| ' fuses prior to reptocement of certain retey coils.- When one of the required fuses
| |
| ' was putted, the high pressure RMR interlock colt was de-energized.
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| This resulted in isolation of SDC.
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| Mode 5, Refueting. SDC isotsted when en isolation controt' The neutret wire for AeWR solid state is less-WP-2 severet relays, . . susceptible to'this type of retoy.for a non SDC function was
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| ~
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| April 21, 1987 Incitding the SDC feiture. Maintenance bypess ,
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| de-energized for maintenance.
| |
| relay, were ett does not require the lifting of conected together. teeds.
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| Lifting the neutret to
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| .e one retey caused a toss e
| |
| $ a of power to att relays with a consoon neutret.
| |
| ts O SDC toop was only See Match 2 (9/19/86).
| |
| Natch 1 Mode 3, not White placing a SDC toop in service, RPV tevet dropped from 62 to 3 inches.
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| portfatty futt prior to April 22, 1987 Shutdown.
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| placing in service.
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| SDC isolated when power was lost to the RPS MG set output See Susqueheme 1 (2/16/83).
| |
| Match 1 Mode 5, Refueling.
| |
| RPS bus. breaker inadvertently June 7, 1987 triped.
| |
| ' SOC isolated when power was removed from - Procedure did not See Susgaehenne 1 (2/16/83).
| |
| Perry 1 Mode 4, Cold the RPS bus for a surveittence ttst. recognize the igact on July 4, 1987 Shutdown.
| |
| SDC of removing power I from the RPS bus.
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| l SDC isolation occurred when the nonnet The cause of the loss See Susquehenne 1 (2/16/83).
| |
| j Peach Bottom 2,3 Mode 4, Cold r August 16, 1987 Shutdown. offsite power sigpty was lost and e of offsite power was transfer to en etternate source -not included in the temporarity de-energized electrical report.
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| tunes.
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| i i
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| I i :
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| ; a I
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| EVENT CATEGORY: LOSS OF RHRS DIE 70 INADVERTENT CLOSURE OF RHR$ SUCTION VALVES INITIAL PLANT PLANT LER/DATE CONDITIONS EVENT DESCRIPTIDW REPORTED CAUSE APPLICABLE ABWR FEATURE Peach Botten 2 Mode 4, Cold SDC isolated during maintenance on SDC isot tion colt see Suswehanna 1 (2/16/83) and August 28,'1987 Shutdown. edectric circuits, inadvertently WNP-2 (4/21/87).
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| de-energized daring maintenance.
| |
| Susquehanna 1 Mode 4, Cold While transferring SDC from the 'A' to A spurious high RNR ASWR solid state logic re w ires September 13, Shutdown. the 'C8 RHR ptsp, SDC isolated. flow signal caused the 2-out-of-4 signet to actuate a 1987 SDC isolation. safety firetion.
| |
| Peach Botton 2 Mode 4, Cold SDC isolated for 15 minutes. Loss of power to a MCC. See Suseema 1 (2/16/83).
| |
| September 16, shutdown.
| |
| - 1987 Perry 1 Mode 4, Cold SDC lactated during a pressure Personnet error in See Match 2 (9/19/86).
| |
| September 29, shutdoiss. transmitter response time test. attowing pressure 1987 signet from test instrunent to exceed SDC high pressure l isolation set point.
| |
| .)
| |
| ~>
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| ' 2 Pilgrim Mode 5, Refueting. SDC isotated on loss of power to 480V tus Cause for toss of power See Sus wehanne 1 (2/16/a3).
| |
| October 6, 1987 which st4pties power to the isolation not reported.
| |
| volve.
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| Pilgrim Mode 4, Cold SDC isolated dJring maintenance on An incorrect tend was See WieP-2 (4/21/87).
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| October 15, 1987 Shutdown. primary conteirment f ootation system. tifted which generated a false high reactor pressure signet.
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| Susquehanne Mode 5, Refueling. SDC isolated when RPS power stopty was Momentary loss of RPS See Susquehame 1 (2/16/83).
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| Novenber 1,1987 transferred between alternate sources. power.
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| Grand Gulf Mode 5, Refueling. SDC isolated during maintenance on power A temporary loss of See Susquehame 1 (2/16/83).
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| Novewber 30, 1987 buses. power occurred then tus was re-energized foltowing meintenance.
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| Peach Botters 2 Mode 4, Cold SDC isolated due to initiation of reactor Technicien caused a See Match 2 (9/19/86).
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| Decenber 6,1987 Shutdoest. scram signal. scram signal to be generated daring an ATWS togic pressure switch calibration.
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| a - - "
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| l EVENT CATECORT: LOSS OF RNES DUE TO IRADVERTENT CLOSURE OF RNRS SUCit0N VALVES INITIAL PLANT REPORTED CAUSE APPLICABLE ABWR FEATURE CONDITIONS EVENT DESCRIPT!0W PLANT ' G,*5 ATE Technician caused a See Match 2 (9/19/86).
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| Mode 4, Cold SDC isolated daring maintenance on RPV Nine Mile Ptint 2 pressure surge in the february 1, 1985 shutdones. tevet sensor. Instrument tine which c
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| ) resulted in a high Rht
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| ' system pressure signal to be generated.
| |
| Persomet error daring See Match 2 (9/19/86). I Mode 4, Cold SDC isolation signal generated during Pilgrim maintenance on emergency parameter maintenance.
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| February 2, 1988 shutdonat.
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| information conputer.
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| Maintenance persomet See crand cutf (5/23/83) and Mode 4, Cold SDC isolated durire refueling outage. Susenhama 1 (2/16/83)
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| LA.P-2 putted wrong set of May 30, 1988 Shutdomai. fuses.
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| Inadequate procedJre. See Match 2 (9/19/83).
| |
| Mode 4, Cold SDC isolated during maintenance on PCIS SDC isolation logic Peach Bottaa 2 topic circuitry.
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| July 29, 1988 Shutdown. should have been ;
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| blocked as part of j
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| [ maintenance task.
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| o 1 Q
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| w Technician See Susgehamn 1 (2/16/83).
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| Mode 4, Cold SDC isolated during modification work on inadvertently grounded i
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| N Mine Mlle Point 2 a RPS cabinet.
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| October 25, 1988 Shutdown. the RPS 24 Vdc power sigply.
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| toss of 295 power See Suagnhama 1 (2/16/83). )'
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| Mode 5, Refueling. CDC isolated fottowing a toss of two FitzPatrick offsite power lines and a 120 Vac UPS. caused SDC isolation.
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| October 31, 1988 momentary loss of power See Susgebama 1 (2/16/83).
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| Mode 5, Refueting. SDC pimp stofped when SDC isotation velve FitzPatrick Ieft its open position. to RPS caused SDC vatwe Noventer 9,19S8 to start closing.
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| Intertock of SDC isolation valve and pump caused control breaker to cpen.
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| Loss of power cause not See Susquahama 1 (2/16/83).
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| Mode 4, Cold SDC isolated when Div. 1 ESF power was Fermi 2 reported.
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| shutdown, lost.
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| January 10, 1989
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| EVENT CATEGORY: LOSS OF RHRS DtX TO INADVERTENT CLOSURE OF RHR$ SUCTION VALVES INITIAL PLANT CONDIT10HS EVENT DESCRIPTION REPORTED CAUSE APPLICABLE ABWR FEATURE PLANT LER/DATE Clinton Mode 5, Refueting. SDC isolated daring testing of RCIC White attempting to see Grand Gulf (5/23/83) and January 10, 1989 togic. . Jumper out the SDC Susquehama 1 (2/16/83).
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| Isolation signat, a technician inadvertently gromded the RPV tow level circuit. This caused a fuse to blow and SDC to isolate.
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| Nine Mite Point Mode 4, Cold SDC isolated daring a surveillance test Test procedre See Match 2 (9/19/86).
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| January 22, 1989 Shutdown. of the reactor building high tenperature specified the wrong isolation signet. Isolation signal be actuated.
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| Hope Creek Mode 4, Cold During perforinance of a surveillance ProcedJral error. See Match 2 (9/19/86) and March 1, 1989 Shutdown.' test, the SDC injection valve closed Leeds were tif ted to Hatch 2 (9/21/86).
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| resulting in a toss of SDC. atlow completion of RHR g
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| logic test without e
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| valve actuations. The
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| @e lead for the RNR injection valve teas
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| 'd inadvertently left off the list of leads to be lifted.
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| River Bend Mode 5, Refueling. SDC cooling isolated when 120 Vac Maintenance persomet See Susenhama 1 (2/16/83).
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| March 25, 1989 divisional logic was de-energized. de-energired logic power to conotete work on the reactor plant sampting system.
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| River send Mode 5. Refueling. SDC isolated dJe to toss of RPS power. A jusper fell off See Match 2 (9/21/86) and March 29, 1989 daring instattation Susg;ehema 1 (2/1/6/83).
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| causing a gromd of RPS power and a blown fuse in the RPS power stgply.
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| ' EVENT CATEGORY: Lost OF RNRS DUE TO INADVERTENT CLtme or tes SUCTION VALVES ;-
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| INITIAL PLAi.T
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| ' EVENT DESCRIPTION. . REPORTED CAUSE APPLICASLE AsuR FEATURE
| |
| - PLANT LER/DATE : l Co m ITIONS Grand Gulf -Mode 4hCold RW ptmp triped during surveillance test Technician lifted DC See Match 2 (9/21/86).
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| April 26; 1989 shutdown.- of RCtc trip throttle valve. power teed for RCic throttle valve but did -
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| not restire that the
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| - RNR pJup *no suction'' <
| |
| ' path" trip togic uns also on the circuit.
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| tRwn the teed uns
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| - tifted,'the R M ptmp
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| . tripped.
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| 4 River send Mode 5 Refueline. : soc isoleted dsrine a surveittance test :
| |
| Lead became - ~ see Netch'2 (9/21/86).
| |
| of manuel scrum function. disconnected during-April 27,1989 test and grounded out the R M high pressure
| |
| ' intertock circuit. ,
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| This caused the e isolation vetve to
| |
| * close.
| |
| : -6.m i
| |
| 4 i
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| ~
| |
| b 6
| |
| i 1
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| 4 6 - ,
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| | |
| EVENT CATEGORY: INA8ILITY TO ESTABLISM RHRS FLOW DUE TO INA8ILITY TO OPEN RNRS SUCTION VALVES INITIAL PLANT PLANT LER/DATE CONotTIONS EVENT DESCRIPTION REPORTED CAUSE APPLICABLE A8ut FfATURE Brunswick 1 Mode 3, Not A reactor cootckan was in progress RLptured flange gasket See Peach Bottom 3 (1/8/79).
| |
| 77-045 Shutdown, Plant following a scram. With reactor water on RNRSW toop 1A heat A8WR uses analog transmitters July 28, 1977 cooldown ist tesperature at 372eF preparations were exchanger cuttet vetve, instead of pressure switches progress. Tempera- commenced for placing RNRS toop 'A' in causing spray-induced for actuation circutts, so this ture at 372eF. shutdonc cooling. RHRS booster ptmps electrical damage. type of failure would not occur were started in conjunction with the 18 in the ABwn.
| |
| ruclear SW ptmp. A gasket rtetured on the RHR service water system as it was being placed in shutdown cooling. Water was observed spraying from the overhead cf the 20 f t. etevation in the reactor tuilding. The 18 toop of RnRS was ptoced in service at 3254F. When attempting to place the RHRS *18' loop in shutdown cooling, it was fotad that the inboard shutdown cooling suction valve would not open, due to a false signal from a g pressure switch.
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| e artnswick 2 Mode 3, Hot After a reactor shutdown, white estab- Electromechanical brake See Peach Bottom 3 (1/8/79).
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| @e 78-036 Shutdoen. Plant tishing shutdoen cooling, the shutdoun on valve operator The current level of the ASWR "gf, April 3, 1978 cooldown in cooling outboard suction valve (F008) failed, causing valve design does not generetty progress. would not open remotely. Valve was to bind and the motor acHress detalt component opened manuelty and reactor placed in operator to draw features. But it is espected cold shutdown. excessive current when that as is the case for energized. operating plants, MOvs will incitz$e hardAeets to mitigate events such as this.
| |
| Brunswick 2 Mode 3, Not During normal shutdown and cooldown, RHRS Cause for valve failure See Peach Bottuu 3 (1/8/79).
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| 78-052 Shutdown. Plant shutdown cooling valve tocated inside the not reported. Person-June 3, 1978 cooldown in containment (F009) would not open from net air lock inner door progress. the ccaitrol room. This valve sesst be would not open dJe to opened before the reactor can be placed sticky gaskets, caused in cold shutdown. Entry into the drywett by targe amount of via the personnet air tock was unsuccess- compressive force ful. Entry into the drywett was made esplied to gaskets by through the CRD hatch and the RHRS valve strongback instatted 2 was manuelty opened. days earlier for test.
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| Strongback removed on day of event.
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| | |
| EVENT CATECORY: INA8ILITY TO ESTABLISM RHRS FLOW DUE TO INABIL!iY TO OPEN RHR$ SUCTION VALVES titITIAL PLANT REPORTED CAUSE APPLICA8LE ABWR FEATURE CONDITIONS EVENT DESCRIPTION PLANT LER/DATE sticking microswitch See Srmswick 1 (7/28/77).
| |
| Reactor stema dcase high pressure switch Brunsw'ck 2 Mode 3. Not Caused instrument Shutdown. would not reset and would not atlow RNR$ fatture.
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| 78-074 vetve (F006) to open for shutdown cooling Novester 12, 1978 at a reactor pressure of 102 psig.
| |
| Thorough investigation See Peach Bottom 3 (f/8/79).
| |
| following a reactor shutdown, while Brunswick 2 Mode 3. Not revealed no catne for shutdom. Plant attempting to place RNRS shutdown cooling falted motor windings.
| |
| 81-019 .into service, the RNR stpply irboard
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| , February 14, 1981 cooldown in Isolation valve (F009) would not open progress.
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| electrically. Burned motor windings prevented the valve motor from opening the valve. Valve was annually opened and RNRS shutdown cooling placed in service.
| |
| Cold shutdown reached 8 hours af ter i
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| opening valve.
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| f See Peach 80ttcyn 3 (1/8/79).
| |
| White attuting to place Ruts shutdown Loose fastener on one Brmswick 2 Mode 3, Not of the overcurrent Shutdown. Plant coolirv ento service, RMRS shutdown p 81-070 coe'.ng stgply inboard isolation valve devices in the vetve c July 18,1981 cooldown in motor breaker, resutt-progress. (F009) would not open on a remote signet.
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| $e valve was manually opened, RNRS shutdown cooling placed in service and rotd ing in an overcurrent condition on two of the
| |
| $ motor phases, trigping shutdown achieved in 8 hours. the breaker.
| |
| Flow switch had teen See 3rmswick 1 (7/28/77).
| |
| Mode 3, Not Shutdown When tining tp for shutdom cooting isolated to perform LaSatte 1 operation, the RMR shutdcan rooting 82-034 at 2254F. (During calibration check; isolation valve (F009) would not opan dJe Jme 5,1982 initiat plant to en isolated RNR pusp suction flow maintenance tech failed starttp phase.) to unisolate instrument switch. after test.
| |
| Retaning torg;e switch See Peach Bottom 3 (1/8/79).
| |
| Mode 3, Not During starttp of shutdown cooling for a Monticetto refueling outage, the RMRS outboard problem. which caused 82-009 Shutdown. Plant shutdown cooling isolation valve contiruous close signet Septenber 2,1982 cooldown in to jam the valve gate progress. (MO-2030) motor failed. into the seat.
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| .}'
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| FfEhi CATEGORft r INAgILITY TO ESTAgtISM RNRS FLOW DUE TO INAgILITY TO OPEN RNRS SUCTION VALVES INITIAL PLANT-EVENT DESCRIPTION LREPORTED CAUSE APPLICAgLE AgWR FEATURE
| |
| ' PLAN! LER/DATE CONDITIONS Lase:te 1 Mode 3, Not The RNR Shutdoun cootIng suction fr60 erd During the test operat- .See LaSette 1 (6/11/82).
| |
| 83-142 Shutdown. . Isolation velve (F009) could not be -ing period, the vetve Mov e r 4, 1983 opened either by the motor operato. or' was normatty seated to . AgWR has 3 RNR systems. One of manuelty.' The satit was shutting doun for stop teekege. With the the two remaining SOC loops:
| |
| ptemed maintenance. plant at tower tempera- . would be avaltable to bring the -
| |
| ture, the velve would , - plant to cold shutdown.
| |
| ret open. ' Felture was attributed to high dif ferential tempere-tures resulting in therent contraction and pinching of the disc wedge into the vetve seat.
| |
| growns Ferry 1 Mode 3, hot While cooling doun to cold shutdown 'e' phase winding of See LaSatte 1 111/4/83) and 84-012 shutdown.' Plant following a manuet scram, the irboard RNR motor operator had Peach Bottom 3 (1/8/79).
| |
| February 14, 1984 cooldown to cold- shutdoun cooling isolation valve failed.- Apparently the gate had stuck in the
| |
| :1 shutdoun in (FCV-1-74-78) f alted to open, making it
| |
| ;$ progress. lagossible to achieve cold shutdown using valve seat and the >
| |
| nornet shutdoun cooling. An ALERT was motor could not s3a doctored, and the plant brought to cold generate enough torque '
| |
| shutdown through continued rermet to open the valve.
| |
| cooldoun to the moln condenser, and the Further investigation use of control rod drive ptops and RWCUS reveeted that the as alternete inventory addition and heat 'close' torque switch removat systems. Since the stuck shut setting was set higher suction vetve was inside contalrument, e ' than the marufacturer's containment entry was necessary to open recommended value the vetve manuelty. It took appromi- (2.5 vice 2.0). This mately five hours to de-Inert the drywett over-tightening prob-to permit entry, and another fours hours - ebly contributed to the to open the stuck vetve and establish stuck valve.
| |
| shutdown cooling, after d ich the ALEnf t was cancelled. Adfitional etternate means of heet removal were avellable.
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| i i
| |
| - - m _ _ _ _ _ _ _ _
| |
| | |
| m.
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| 4 y
| |
| -EVENT CATEGORT: LOSS OR DEgEADATION OF RNRS DUE 70 gTPASS FLOW, REVERSE FLOW, OR OTNEft VALVE PRogLENS L INITIAL PLANT-COWITIONS EVENT DESCRIPTION REPORTED CAUSE APPLICAstE Agut FEATURE .
| |
| PLANT'LER/DATE.
| |
| I Dresden 3- - Mode 4, cold ' An inadvertent heetty and pressurization- Vetve Line w error.- See teSette 1 (6/1T/82).' Agut shutdom. RNRS in . was caused try a volve linew error during Post maintenance' does not have enternet rectre Ney, W78 . operati m at 1606F.. . contelruent teek rete testing. About .18 testing of a rectre . pimps or vetves. aeector-
| |
| . internet pimps (RIPS) supply
| |
| ! hours efter.reeching test pressure, , pimp MG set regaired e -
| |
| J
| |
| ' reactor vesset flange temperature was . ; recirc pimp test rm. recire flow so this event could' discovered to be et approximately 3006F [ ;The motors were . not occur in the AguR.
| |
| and increasing. One loop of shutdom smcoupted from the cootIng was in service recording e recirc gamps for the temperature of appronlastely 1606F. The' test. The motors would i amt heet enchanger ehett . temperature and not start becomme
| |
| ;vesset flange temperature should have . piemp/ valve interlocks been egant. Investigation reveeted that - gave a trip signet.to the recirc pieps were off and recirc toop (the pimp motor since 4
| |
| . suction and discherse volves were open. the suction and This tinsic reeutted In the majority of- discharge vetves were RNRS flow circulating through the recirc closed. Conse m ently, 4 ..toop and not the core.' The vesset heetip maintenance persomet .
| |
| H and presourlastion caused a temperature opened the velves to j f and pressure increase in the drywett. perform the test. This o The canyuter program used to calculate permitted shutdoun the contelrument teak rete was tssing cooling flow to bypass
| |
| [ 6 i CD shutdeun cooling temperature to indicate the core via the recire-I condittens inside the vesset.. The loop,~cousing the computer misinterpreted vesset conditions ' inadvertent heetup eruf 4
| |
| and concluded there was e terse inteekage pressurization.
| |
| l condition.-
| |
| Mode 4, Cold With the reactor in the shutdown mode feutty euxillery Ag W hos three Ree loops, Hetch 1
| |
| , 80-057 Shutdown. RNR$ in - ' during testing, the shutdoun cooling contact block. The feiture of toop "g* with loop ,
| |
| suction vetve for the 'g' RNR$ pump norsetty ctooed retey =Aa in maintenance could be i May 25, 1980 w etlon. (F0063) feited to open. The'*g' pimp woe contact was foamd stuck . mitigated by using loop "C*..
| |
| j a ' doctored inoperable. Since the 'A' divi- in the open position. The WWCU system, FPC, and main !
| |
| elon of RNRS wee out for maintenance, condenser can etso be used for ' +
| |
| both guesps in the 'g8 division were ONR ander certain plant
| |
| , required to be operable. conditions.
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| I i
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| l' b
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| u i
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| u . -.}}
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