ML20237D492
| ML20237D492 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1291, NUDOCS 8712230297 | |
| Download: ML20237D492 (261) | |
Text
Eve ~nt Descriptions !
l I
i i
Manuscript Completed: September 1987 Date Published: November 1987 ;
Division of Licensee Performance and Quality Evaluation Office of Nuclear Reactor Regulation
.C'4.S. Nuclear Regulatory Commission
(,heshington, DC 20555 pe "%g
~
l l
I I
l
'N 8712230297 871130 PDR i
NUREC i 1291 R PDR l
i l
L
! l NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.)
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013 7082
- 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.
The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.
Documents available from the National Technical information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.
Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.
Single copies of NRC draft reports are available free, to the extent of supply, upon written request
{
to the Division of Information Support Services, Distribution Section, U.S. Nuclear Regulatory 1 Commission, Washington, DC 20555.
I i
Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library,7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.
O i
I
,e f
! Ns')
ABSTRACT This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios.
Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence (s) of expected immediate and subsequent candidate actions, including communications, that'can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities.
i O i t N-i l
/ \
! I
'O ili
., /m
/ \.
k --
TABLE OF CONTENTS fage ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii PREFACE ............ . . . . ... . . . . . . . . . . . . . ix ACKNOWLEDGEMENTS . . . . . . . . .. . . .. . . . . . . . . . . . . . xi 1
ACRONYMS / ABBREVIATIONS . . . . .. . . .... . . . . . . . . . . . .
INSTRUCTIONS FOR USING EVENT DESCRIPTIONS . . . . . . . . . . . . . . 7 Organization of Event Descriptions . . . . . . . . . . . . . . . 7 The Lists of Abnormal-and Emergency Events . . . . . . . . . . . 9 Scenario Preparation . . . . . .. .. . . . . . . . . . . . . . 9 Example Scenario Development Using Event Descriptions . . . . . . 12 BWR Example . . . . . . . . .. ... .. . . . . . . . . . . . . 15 PWR Example . . . . . . . . .. . . .. . . . . . . . . . . . . . 31 BWR ABNORMAL EVENT DESCRIPTIONS .. .. . . . . . . . . . . . . . . . 43 A
/ Boiling-Water Reactor Abnormal Events . . . . . . . . . . . . . . 45'
\%
Master Feedwater Controller Failure . . . . . . . . . . . . 47 ]
Nuclear Instrument Channel Failure . . . . . . . . . . . . . 49 Rod Position Indicating System Failure . . . . . . . . . . . 51 One Reactor Recirculation Pump Trip . . . . . . . . . . . . 53 Trip of Both Recirculation Pumps . . . . . . . . . . . . . . 57 Recirculation Pump Seal Failure . . . . . . . . . . . . . . 61 Scoop Tube Lock . . . . . . . . . . . . . . . . . . . . . . 63 Increasing Suppression Pool Temperature . . . . . . . . . . 65 Drywell Cooler Failure . . . . . . . . . . . . . . . . . . . 67 Stuck Control Rod . . . . . . . . . . . . . . . . . . . . . 69 Uncoupled Control Rod . . . .. . . . . . . . . . . . . . . 71 Control Rod Drift . .. . . . ... . . . . . . . . . . . . 73 Control Rod Drive Hydraulic Pump Trip . . . . . . . . . . . 77 Loss of All CRD Hydraulic Pumps . . . . . . . . . . . . . . 79 CRD Flow Control Valve Failure ... . . . . . . . . . . . . 81 l Condensate or Condensate Booster Pump Trip . . . . . . . . . 83 Reactor Feedwater Pump Trip . . . . . . . . . . . . . . . . 85 Loss of Feedwater Heater Extraction Steam . . . . . . . . . 87 Stator Cooling Water Pump Trip . . . . . . . . . . . . . . . 89 ;
Steam Jet Air Ejector Malfunction . . . . . . . . . . . . . 91 J Loss of One Reactor Protection System Bus . . . . . . . . . 93 J Area Radiation Monitoring System Alarm . . . . . . . . . . . 95 High Main Steam Line Radiation . . . . . . . . . . . . . . . 97
/N High Ventilation Exhaust Radiation . . . . . . . . . . . . . 99
( Inadvertent HPCI or RCIC Initiation . . . . . . . . . . . . 101 Loss of One RBCCW Pump . . . . . . . . . . . . . . . . . . . 103 v I L. .
l l
l l
m 9
l BWR EMERGENCY EVENT DESCRIPTIONS . . . . . . . . . . . . . . . . . . 105 Boiling-Water Reactor Emergency Events . . . . . . . . . . . . . 107 Reactor Scram With MSIVs Open . . . . . . . . . . . . . . . 109 Reactor Scram With MSIVs Closed . . . . . . . . . . . . . . 113 Loss of Shutdown Cooling . . . . . . . . . . . . . . . . . , 117 Gross Fuel Failure ... . . . . . . . . . . . . . . . . . . 119 Excessive Reactor Cooldown Rate . . . . . . . . . . . . . . 123 Anticipated Transient Without Scram . . . . . . . . . . . . 125 Stuck Open Main Steam Safety / Relief Valve . . . . . . . . . 129 Small Break Loss of Coolant Accident . . . . . . . . . . . . 133 Reactor Coolant Leakage Outside Primary Containment . . . . 137 Jet Pump Failure . ... .. . . . . . . . . . . . . . . . . 14 1 High Suppression Pool Water Temperature . . . . . . . . . . 143 Main Turbine or Generator Trip . . . . . . . . . . . . . . . 147 Main Turbine or Generator Trip Without Bypass Valves . . . . 151 Loss of Condenser Circulating Water . . . . . . . . . . . . 155 Loss of Feedwater System . . . . . . . . . . . . . . . . . . 159 Loss of All High Pressure Feedwater . . . . - . . . . . . . 163 Loss of Plant Control / Instrument Air . . . . . . . . . . . . 167 EHC Pressure Regulator Failure (All Valves Open) . . . . . . 171 Loss of Nuclear Service Water . . . . . . . . . . . . . . . 175 Loss of Reuctor Building Closed Cooling Water System . . . . 179 Loss of Off-Site Power . . . . . . . . . . . . . . . . . . . 183 Loss of All AC Power (Station B1cckout) . . . . . . . . . . 187 PWR ABNORMAL EVENT DESCRIPTIONS ... .. . . . . . . . . . . . . . . 191 Pressurized-Water Reactor Abnormal Events . . . . . . . . . . . . 193 Loss of RCS Makeup . . . . . . . . . . . . . . . . . . . . . 195 Loss of Automatic Pressurizer Pressure Control . . . . . . . 197 Failure of Pressurizer Spray Valve . . . . . . . . . . . . . 199 Loss of Automatic Pressurizer Level Control . . . . . . . . 201 Progressive Failure of No. 1 Seal in RCP . . . . . . . . . . 203 Failure of Steam Dump to Open . . . . . . . . . . . . . . . 205 Steam Generator Safety Valve Fails Open and Fails to Rescat. 207 Steam Generator Level Control Failure High/ Low . . . . . . . 209 Dropped Control Rod .. . ...... . . . . . . . . . . . 211 Inoperable or Stuck Control Rod . . . . . . . . . . . . . . 213 Inadvertent Boration at Power . . . . . . . . . . . . . . . 215 Inadvertent Dilution at Power . . . . . . . . . . . . . . . 217 Failure of N-44 High . . . . . . . . . . . . . . . . . . . . 219 Loss of Instrument Air . . . . . . . . . . . . . . . . . . . 221 Failure of Turbine to Runback Automatically and Manually . . 223 Failure of Impulse Pressure Transmitter (Low) . . . . . . . 225 ;
Steam Generator Tube Leak Within Capacity of Charging Pump . 227 i Loss of Condenser Circulating Pump . . . . . . . . . . . . . 229 l
I i
vi
pg
)
Page Criticality Outside Expected Band . . . . . . . . . . . . . 231 Failure of Loop Temperature Instrumentation High/ Low . . . . 233 Loss of One Main Feedwater Pump at High Power . . . . . . . 235 Spontaneous Opening of the Main Generator Output Breakers . 237 Loss of RCP Without Reactor Trip . . . . . . . . . . . . . . 239 Main Steam Leak Inside Containment . . . . . . . . . . . . . 241 Rupture in Letdown Nonregenerative Heat Exchanger to CCW . . 243 Failure of Pressurizer Control Bank Heaters . . . . . . . . 245 PWR EMERGENCY EVENT DESCRIPTIONS . . . . . . . . . . . . . . .. . . . 247 Pressurized-Water Reactor Emergency Events . . . . . . . . . . . 249 Reactor Trip . . . . . . . . . . . . . . . . . . . . . . . . 251 Large Break LOCA -- Reactor Trip With Safety Injection . . . 253 PZR/PORV Failure to Open . . . . . . . . . . . . . . . . . . 257 Steam Generator Tube Rupture . . . . . . . . . . . . . . . . 259 Failure of Main Turbine to Trip . . . . . . . . . . . . . . 263 Small Break Loss of Coolant Accident . . . . .. . . . . . . 265 Anticipated Transient Without Scram . . . . . . . . . . . . 269 Loss of Auxiliary Feedwater -- Inadequate Core Cooling . . . 271 fg Loss of Off-Site Power . . . . . . . . . . . . . . . . . . . 275 i ) Station Blackout -- Loss of All AC Power . . . . . . . . . . 277
\s / Control Room Fire Requiring Evacuation . . . . . . . . . . . 281 Main Steam Break Inside Containment . . . . . . . . . . . . 283 RHR LOCA -- Complete Loss of All RHR . . . . . . . . . . . . 287 o
Dx I ) I
\s / 1 vii
i 1
l-
. fK l Y'
l-PREFACE This document presents PER and BWR Off-Normal Event Descriptions prodoced by l Oak Ridge National Laboratory and Physics Applications, Inc., under contract to the Nuclear Regulatory Commission. These event descriptions are intended as a reference document to assist operator licensing examiners in their development of simulator examination scenarios in accordance with the procedures described in the Examiner Standards (NUREG-1021).
Unlike much of the reference material that examiners must review in the course of preparing simulator scenarios, the descriptions include information specifically geared toward the selection and construction of NRC simulator !
scenarios. Therefore, they are intended to serve both as an aid to examiners in their compliance with the Examiner Standards,-and as a generic supplement to facility-specific reference material. l Descriptions are provided for a total of 87 emergency and abnormal events for BWR and PWR facilities. Each event description includes a cover shcot and a flow chart depiction of the progression of operator actions. The information provided in these descriptions can assist examiners in: (1) selecting events
,-, for examination scenarios; (2) constructing scenarios that provide ample opportunities to observe and evaluate candidates on each major operator or
(\
%,/ . senior operator competence; (3) identifying objective bases for use in candidate evaluation; and (4) developing expected candidate actions and behaviors, per the procederes in the Examiner Standards.
The event descriptions were developed by subject matter experts with the use of facility and vendor reference documents. ' Licensing examiners reviewed each draft description for accuracy, completeness, and appropriateness. -
However, these descriptions are intended as a' generic reference basis.
Further, they were not " field tested" prior to isruance. Therefore, examiners should review the descriptions for completeness and accuracy when developing plant-specific scenarios, o
ix
ACKNOWLEDGTEENTS This document is the product of work primarily performed by Oak Ridge National Laboratory, Physics Applications, Inc., and Hensley & Associates.
Primary contributors were C. R. Bovell, R. J. Carter, F. G. Gunnon and G. B.
Stewart.
O O
Xi
1 I
.J U S f% .
b ACRONYMS / ABBREVIATIONS AC - alternating current accum - accumulator add - additional adj - adjust admin -' administrative ,
ADS - automatic depressurization system AFW - auxiliary feedwater alt - alternate:
annun - annunciator -
AO - auxiliary operator A0P - abnormal operating procedure approx - approximate /approximately
'APRM - average power range monitor
' atmos - atmosphere / atmospheric auto - automatic aux - auxiliary avail - available B & W - Babcock'and Wilcox B/U - backup /burnup BIIT - boron injection initiation temperature f- s BIT - boron injection tank bkr/brkr - breaker
('- b1dg - building BOP:- balance of' plant brd - board
- i. brg - bearing BWR - boiling-water reactor C-E - Combustion Engineering CB - concentration of boron CCW condencer circulating water charg/chrg - charge / charging chem - chemistry CIA - containment isolation phase A CIB - containment isolation phase B cire - circulating / circulation CMFLPD - core maximum fraction limiting power density compli - compliance cone - concentration cond -. condenser / condensate
, cont - control contm't/contmt - containment CRD - control rod drive CRDM - control rod drive mechaninm CSBW - cold shutdown boron weight CSFST - critical safety function status tree
/'~' CSIP - charging safety injection pump
( '~
CST - condensate storage tank CVCS -- chemical' and volume control system 1
i DC - direct current DE - dose equivalent decreas - decreasing DG - diesel generator disch - discharge DNBR - departure from nucleate boiling ratio DRPI - digital rod position indication DW - drywell E-0/E0 - Westinghouse " Reactor Trip or Safety injection" procedure E-3.0 - Westinghouse " Steam Cencrator Tube Rupture" procedure ECA-0.0 - Westinghouse " Loss of All AC Power" procedure ECA-0.1 - Westinghouse " Loss of All AC Power Recovery Without SI Required" procedure ES-0.1 - Westinghouse " Reactor Trip Response" procedure ES-0.2 - Westinghouse " Natural Circulation Cooldown" procedure ES-1.2 - Westinghouse " Post LOCA Cooldown and Depressurization" procedure ES-1.3 - Westinghouse " Transfer to Cold Leg Recirculation" procedure ES-302 - Examiner standard " Scope of Operating Examinations Administered to Reactor Operators and Senior Reactor Operators - Power Reactors" ECC - emergency core cooling ECCS - emergency core cooling system ECP - estimated critical position EDG - emergency diesel generator e.g. - for example EHC - electrohydraulic control elec - electrical emerg/cmg - emergency energ - energy ENR - extraction non-return envir - environment E0I - emergency operating instruction E0P - emergency operating procedure EPG - emergency procedure guideline EPP - emergency preparedness plan equil - equilibrium ERF - emergency response facility estab - establish etc. - and others exch - exchange / exchanger F-0.2 - Westinghouse " Core Cooling" procedure FR-C.1 - Westinghouse " Response to Inadequate Core Cooling" procedure FR-H.1 - Westinghouse " Response to Loss of Secondary Heat Sink" procedure F - Fahrenheit fail - failure FCV - flow control valve fd - feed FSWL - flow stagnation water level FW - feedwater 2
,/ :
i .
. I.
5~ -
GE - General Electric gen - generator GP - general procedures gp/ group'- group gpm - gallons per minute H2 - hydrogen HCTL - heat capacity temperature limit HCU - hydraulic control unit-hd - head Hg - mercury hi - high hp - high pressure itPCI - high pressure coolant injection HPCS - high pressure core spray HSEW - hot shutdown boron weight htr - heater HVAC - heating, ventilation, and air conditioning 1 - iodine 1 & C - instrumentation and controls IC - initial condition i.e. - that is to say inc/incr/iners - increase / increasing
[' ind - indication / indicator
( '~'
inj - inject / injection inop - inoperable INP0 - Institute of Nuc1 car Power Operations inst /instrum - instrument / instrumentation IR - intermediate range IRM - intermediate range monitor IRP1 - individual rod position indication iso /isol - isolate / isolation lab - laboratory Kv - kilovolts LCO - limiting condition for operation lo - low lo-lo - double low LOCA - loss of coolant accident Lp - pressurizer level LPCI - low pressure coolant injection LPCS - low pressure core spray lvl -. level l
M-G - motor generator maint - maintenance malf - malfunction
{
man - manual / manually l mang/mgmt - management max - maximum agr - manager 3
- _ _ _ ____u
min - minimum en - main O
MS - main steam MSBWP - maximum cuberitical banked withdrawal position MSIV - main steam isolation valve MSL - main steam line MSO - maximum safe operating NWt - megawatts, thermal N - no N2 - nitrogen N/A - not applicable NlS - nuclear instrumentation system norm - normal NR - nuclear recorder NRC - Nuclear Regulatory Commission NRHX - nonregenerative heat exchanger NSSS - nuclear steam supply system nuc engr - nuclear engineer i
OD - on demand OP - overpressure / operating procedure OPC - overspeed protection control oper - operator 4 ops - operations OT - overtemperature P/ press - pressure PA - public address pcm - percent millirho pit - plant pmp - pump PORV - power operated relief valve pos - positive posit - position Pp - pressurizer pressure ppm - parts per r.1111on j pri - primary prot - protective / protection psid - pounds per square inch - differential psig - pounds per square inch - guage PWR - pressurized-water reactor pvr - power PZR - pressurizer OPRT - quadrant power tilt ratio rad - radiation RBCCW - reactor building closed cooling water RBli - rod block monitor RCIC - reactor core isolation cooling FCP - reactor coolant pump i RCS - reactor coolent system recirc - recirculation 4
! Y t
k, recircs - recirculation pumpn ref - refer / reference reg - regulation / regulator req'd - required req'ments/reqm'ts - requirements RFP - reactor feedwater pump TJIR - residual heat removal RlL - rod insertion limit RMS - radiation monitoring system rug - range RO - reactor operator RPlS - rod position indicating system
- rpm - revolutions per minute RPS - reactor protection system RPT - recirculation pump trip RPV - reactor pressure vessel RSCS - rod sequence control system RTD - resistance temperature detector RVLlS - reactor vessel level indicating system RWCU - reactor water cleanup RWM - rod worth minimizer RWST - refuel water storage tank Rx - reactor S/D - shutdown i S/U - startup
\- ') SBGT - standby gas treatment SDV - scram discharge volume see - secondary serv - service SG - steam generator SGT - steam generator tube SI - safety injection SJAE - steam jet air ejector SLC - standby liquid control SPLL - suppression pool load limit SR - source range SRM - source range monitor SRO - senior reactor operator SRV - safety relief valve l SS - shift supervisor stm - steam st pt - set point
, supp - suppression SW - service water sw - switch sys - system TAF - top of active fuel
,. Tauct - auctioneer temperature
[ Tave/Tavg - average temperature
\. TC - thermocouple temp - temperature ,
termin - terminate )
5 l i
I i
l 1
l therm - thermal T-hot - hot leg temperature O'
Tk -tank Tref - reference temperature TS/ tech spec - technical specification TSC - technical support center tur/turb - turbine uCi/gm - micro Curles/ gram V - volts vac - vacuum VCT - volume control tank vent - ventilation vs - versus w - with Wf - mass flow rate, feedwater Ws - mass flow rate, steam x-connect - cross-connect Xe - xenon x-fer - transfer Y - yes
@ - at
- - number
% - percent
& - and o - degrees
- feet
- inches
> - greater than
< - less than
/ - per Al - change in flux AP - change in pressure A T - change in temperature
- Westinghouse O
1 6
i i
l j
l
'1NSTRUCT10NS FOR USING EVENT DESCRIPTIONS l
Organization of Event Descriptions The boiling-water reactor and pressurized-water reactor event descrip-tions are organized into two major parts: a cover sheet and a progres-sion of operator actione flow chart. Together, these two major parts provide generic information to aid in the development of simulator examination scenarios.
Cover Sheet Each cover sheet contains the following information:
(1) Operating Sequence - The titic of the event.
(2) NSSS/ Type - The nuclear steam supply systen vendor (s) (General Electric, Westinghouse, Combustion Engineering, and/or Babcock
& Wilcox] and the type of reactor (BWR or PWR].
(3) Initial Plant State - The operating status of the plant at the time the event starts. The initial plant state may be obtain-ed either by use of the initial conditions input into the simulator computer, or by instructing the candidates to take l ( the plant into the desired plant state.
(4) Sequence Initiator - A brief description of the equipment failure that causes the event.
(5) Important Plant Parameters - Those plant parameters that should be monitored by candidates during the course of the event. The parameters listed are unique to the event; parameters that are important in virtually every off-normal condition, such as primary system pressure, water levels, and reactor power, are not repeated in each description. The important plant parameters are intended to provide objective bases for use in candidate evaluations, including the ability to diagnose plant conditions, comply with procedures, and observe technical specification limits.
(6) Progression of Operator Actiens - Discussed in detail below.
(7) Major Plant Systems - Those plant systems that are uniquely affected by the event. The plant systems listed either experience the failure or are used in mitigating the conse-quences of the failure.
(8) Tolerance Range - The tolerance range of operator actions represents the bounds within which the candidates must respond
( before technical limits are exceeded. Similar to "important plant parameters", tolcrance ranges are intended to provide objective bases for use in candidate evaluation. For exanple, failure to remain within the tolerance range may provide 7
L _- . _ _ - - _ _ _ _ _ - _ _ _ _ _ -
justification for an unsatisfactory evaluation on one or more of the competency categories.
O (9) Final Plant State - The possible plant conditions by which a judgment can be made to end the event and move on to the next part of the scenario / examination. The event may be ended before this point is reached, provided that enough information is gathered to adequately assess candidate performance.
However, if time permits, the event should be taken to the indicated final plant state.
(10) Competencies Tested - A list of competencies, by job posi-tion, demonstrated during the event by each candidate. Only competencies for which the event will provide ample opportu-nity to observe are listed.
Progression of Operator Actions Flow Charts The progression of operator actions depicts, in a flow chart manner, the representative sequence (s) of expected immediate and subsequent candi-date actions (including communications) that can be observed during tne event. These flow charts are intended to be as generic as possible l'or a given reactor / vendor type. The flow charts indicate that, in some cases, there is more than one path which the event can take. The path taken will depend on the likely perturbations of.the system, the deci-sions of the candidate, or choices made by the examiner. The objective of these multiple paths is to provide as much flexibility as possible, while retaining simplicity.
The progression of operator actions for the abnormal event descriptions were developed using available event-based plant procedures. The emergency event descriptions were developed using the symptom-based emergency procedure guidelines from the various owners groups. Because of the use of the EPGs, the emergency event descriptions are somewhat more generic than the abnormal event descriptions. The event descrip-tions are presr.nted in the four event description sections, "BWR Abnormal Event Descriptions", "BWR Emergency Event Descriptions", "PWR Abnormal Event Descript ions", and "PWR Emergency Event Descriptions".
The event descriptions were written to be as generic (i.e., apply to as many plants) as possibic. The expected operator actions are also intended to provide examiners with defined, observable behaviors for use during scenario development and candidate evaluation. There are events for which these two goals (generic applicability versus defined behav-fors) could not be achieved, particularly in the case of PWR events. ]
since three vendors are included in this reactor type. Therefore, some l pWR event descriptions are more applicable to Westinghouse facilities than to C-E or B & W plants. However, in many of these instances, l I
modification of the event description for another vendor plant may only )
require slight changes in terminology (e.g., " phase A" versus " phase B). j 8
__-___ -_~
l t
v The Lists of Abnormal and Emergency Events There are 87 abnormal and emergency events which are described.. They are shown in the Table of Contents and in the front of each event i description section. The breakdown of events in terms of reactor type and severity of event is as follows:
Reactor Type Severity BWR PWR Total Abnormal 26 26 52 Emergency 22 13 35 Total 48 39 87 Events were selected for description based on the following criteria:
(1) Together, the events should represent a broad base of abnormal and emergency conditions (note: major electrical system g malfunctions are underrepresented due to significant differ-ences in electrical systems across facilities, making the f%s_-
( w) development of generic descriptions infeasible);
(2) The events should be able to be replicated on the majority of simulators in use today; (3) The events should be able to be performed within the time limits of a typical simulator examination; (4) The events should allow for the demonstration of an ample number and variety of candidate competencies through the execution of activities required for the mitigation of the events.
Scenario Preparation Development of effective scenarios using the event descriptions is a five-step process:
(1) Select events for compliance with ES-302 and to test applicable competencies; (2) List the events in sequence and by scenario;
/'~'} (3) Complete the Scenario Events Form; (4) Complete the Operator Actions Form; (5) Check for adequacy of competency coverage.
9
1 I
1 Step One: Select the Events The event descriptions are intended to aid the examiner in selecting simulator events for compliance with the criteria described in ES-302.
These criteria include:
(1) Events requiring candidates to operate in normal evolutions, instrument failures, component failures, and major plant transients (Attachment 7. ES-302);
(2) Events requiring candidates to operate under a range of conditions within each category as listed in item #1 above, such as degraded heat removal, degraded electrical power, containment challenges, and degraded pressure control; (3) Events that impact important safety systems such as the systems identified in the PWR/BWR knowledge and ability catalogues (NURECs-1122/1123);
(4) Events that , together, will provide ample opportunity to evaluate each candidate on each relevant candidate competency (Attachment 8, ES-302);
(5) Events that will complement and/or supplement information gained on the candidates during the written and oral sections ,
of the examinations. For example, if the control rod drive system is the focus of a number of questions on the written and/or oral sections, it may be more effective to select l events involving another system responsibic for reactivity control, such as CVCS. On the other 1:and, if knowledge of CVCS components is tested for on the written and/or oral examination (s), it still might be important to include events in the simulator portion that will test the candidate's ability to operate that system's components, such as having him/her perform a boration or dilution reactivity change.
To aid in selecting events, a set of competency matrices has been included in this document (see the " Event s-By-Compe tencie s Matrices" section). The competency matrices summarize the information that appears at the bottom of each event description cover sheet. These matrices aid in selecting a sufficient number of events to ensure that each candidate demonstrates each of the applicable competencies over the course of the simulator examination, per item #4 above. At a minimum, select enough events such that each competency is demonstrated at least once, but preferably more than once.
j Step Two: List the Events i i
Each exam scenario should present the candidate with a logical and realistic set of problems to which he/she is to respond. For example, component and instrument failures can be used as precursors to major casualties. This will fulfill two or three examination requirements while achieving scenario realism. Make a rough list of the events that 10 l
l l
l
r-- - - - - - - - - - - - - - - - - - - - - - - - - - -
e c
are to be used in each scenario. If the sequence of events is not logical, the rough list should be revised until this objective is met.
Step Three: Complete the Scenario Events Form The Scenario Events Form (Attachment 3, ES-302) provides the simula- ,
tor operator with a set of instructions for entering initial conditions and malfunctions into the simulator computer. The information for this form is obtained from the event description cover sheets and simulator reference materials. The first item is to select the appropriate plant ,
condition from the I.C. menu. For ' example , if the event description specifies that the event should be initiated from high ,0wer, an initial condition for this power level may be selected from the menu, or a lower power level may be selected and the candidates directed to perform a power escalation. This will meet the requirements for a normal evolu-tion or reactivity change, and a major casualty. The malfunctions to be run during the scenario, along with the elapsed time that the malfunc-l tions should be initiated, should then be included on the Simulator Scenario Form.
Step Four: Complete the Operator Action Form The Operator Actions Form (Attachment 4, ES-302) should include the observable candidate behaviors for use in evaluating candidates. The progression of operator actions flow charts can be used as an aid in developing these expected actions / behaviors. This information should be compared to the plant specific procedures to ensure the appropriateness of the flow chart information for that facility.
Each action block on the flow charts indicates the candidate primarily responsible for the action. This is intended as a guide and may not be accurate for every situation. In general, the senior reactor operator is responsible for directing the actions of the reactor operator and the balance-of-plant operator, communication with the auxiliary operator and other support personnel, and all administrative duties. The RO is primarily responsible for the reactor and reactor auxiliaries within easy reach of the reactor panel. The BOP is responsible for all secon-dary plant systems, electrical distribution, emergency core cooling sys-tems, and process / area radiatien monitoring. However, when the workload on one operator becomes excessive, assistance may be given by another operator. When an action is entered on the Simulator Administration Form, the candidate responsible for the action is indicated in the
" position" column.
Step Five: Check for adequacy of competency coverage.
After the first scenario has been drafted, the expected actions /behav-iors listed on the Operator Actions Form should be reviewed against the competencies listed on page 4 of Form 157, to determine which competencies should be addressed for each candidate. If there are candidate competencies that are not addressed in the first scenario, the selection of events for the next scenarios F.hould be chosen, in part, to evaluate those competencies.
11
\
Exaraple Scenario Development Using Event Descriptions To demonstrate the use of event descriptions for scenario development, two examples, one consisting of BWR events and the other PWR events, are presented here. Each example was designed utilizing the five-basic steps discussed in " Scenario Preparation" above.
Step One: Select the Events The selection of events to be used during the simulator examination should be based on the procedures described in ES-302. In the exampics provided, the events along with the normal operations that are perform-ed meet several of the required operations as outlined in Attachment 7 to ES-302. In the BWR example the requirements for component malfunc-tions were fulfilled by using a recirculation pump seal f ailure and a loss of f eedwater heater extraction steam. For the major plant tran-sient a slowly developing fuel failure was' chosen. Starting at less than rated power 1cvel provides the opportunity to test candidates en a reactivity change of 10%. In the PWR example an instrument failure, N-44 high, and a component failure, condenser circulating water pump trip, were celected. For the maj or plant transient a steam generator tube leak that develops into a full steam generator tube rupture was chosen. Each of the events used in the two examples is described in the event descriptions sections of this document.
From the event description cover sheets, it was determined that the selected events are suitable for the plant conditions used during the scenario and that an acceptable number of competencies will be demon-strated by the candidates during the events. In neither of the ,
examples, however, will the candidates perform all of the minimum l simulator requirements nor all of the required competencies. Therefore events for the remaining scenarios in the examination should be selected to fill the gaps that renain after the first scenario.
Step Two: List the Events Once the events have been selected, a rough list should be made of the events placing them in the sequence in which they will be initiated during the scenario. In the BWR example the list consists of the recir-culation pump seal failure, then the loss of feedwater heater extraction steam, followed by a fuel failure of sufficient severity to produce a high main steam line radiation, and finally a gross fuel failure. The sequence of events in the PWR cxample i t, as follows: influre of N-44 high, steam generator tube Icak within capacity of charging pump, loss of condenser circulating pump, and steam generator tube rupture.
O 12 l l
1 a
l Step Three: Complete the Simulator Scenario Form j Simulator reference materials, particularly the initial conditions and the malfunctions lists, should be used to enter the appropriate information on the Scenario Events Form, . The event description cover sheets are useful for providing information on the initial plant state for a given event and the simulator malfunctions that should be used to initiate the event. In the BWR example malfunction 103 is a closure of an extraction non-return valve. The malfunction cause-and-effect descriptions would be beneficial for this step to ensure that the malfunctions will create the desired conditions. The timing of the malfunctions dependo on the complexity and number of operator actions in response to the malfunctions. The descriptions of the malfunctions entered on the Scenario Events Form should he as complete as possible so that the simulator operator can initiate the correct malfunctions in the manner in which the examiners wish. In the BWR exampic on malfunc-tion 43, fuel failure, it was estimated that a severity of 20% is required to produce a high steam line radiation alarm. In this case the malfunction cause-and-effect descriptions provided valuable information.
To ensure that the radiation levels increase slowly and that the feed-water heater problem is ended before the high steam line radiation, the example specifies that the severity should be increased at 1% per minute up to 20%. During the fifteen minutes from the time malfunction 43 is s initiated and the high steam line radiation alarm is received, the candidates would respond to the feedwater heater problem. When the
% alarm is received, the trace on the strip chart recorder would indicate l
a gradual increase instead of a drastic step increase. After allowing j time for the candidates to respond to the minor steam line high radia-tion, the severe fuel failure event would be started with a much quicker increase in severity, step changes of 10% per minute and only eight minutes needed to reach 100%.
Step Four: Complete the Operator Actions Form Perhaps the most important step in the preparation of a simulator exam scenario is entering information onto the Operator Actions Form. Using the event descriptions, plant procedures, and technical specifica-tions, the actions were entered onto this Form in the two examples for each of the events selected for the scenario, if one refers to the event description progression of operator actions for each of the events in the examples, he/she may note many similarities in the sequence of actions on the Form and in the flow charts. However, there may occa-sionally be differences. The actions should be listed on the Operator Events Form in the sequence in which they appear in the plant procedure should there be a difference between the procedure and the flow chart's l progression of actions. Also, the plant procedures provide important !
normal parameter values and set-points that should be noted on the Form. !
The two examples include additional information on procedure numbers l (normal, surveillance, and cmergency) and Tech. Spec. sectionc that I
] should be used during the events. This information is important in evaluating the use of procedures and technical specifications and as nn aid in observing whether the candidates have successfully diagnosed the conditions created by the malfunctions. >
13 i
__ ____o
Step Five: Check for adequacy of competency coverage.
The event description cover sheets and the Operator Action Forns can be
- used to check the adequacy of competency coverage. The cover sheets can serve as guides for identifying the competencies which are typically tested during the events. On a plant specific basis, however, they may require modification due primarily to control panel, system, and procedural differences. For exampic, in the PWR scenario the SRO is not expected to ' exercise diagnosis of events / conditions based on signals /
readings according to the event description cover sheets or the PWR competency matrices. However, it was determined that the steam l generator tube rupture provides adequate opportunity for event diagnosis on the part of the SRO to justify that eva7uation. In the BWR example the opposite is true. The event description for high main steam line radiation indicates that the RO would demonstrate compliance /use of technical specifications. However, it was determined that a singic scal failure would not exercise this competency adequately to justify an evaluation.
O O
14
l BWR EXAT'LE I
l I
I 15
-- - - - - - - - - _ _ _ - _ _ _ _ mr
SCENARIO EVENTS I Simulation Facility: k h ol k e Scenarto No.
O Examiners: k. AA4ds Candidates: b Iorvcc
- l. IaJA$4 T $rh r 1"$ !
D. Lham L . Anou>ni ,
Initial Conditions:
TIME- MALF. N0. DESCRIPTION i
O r. c . la est Pcwea , sua of we L,,, l
+2 mla 1% c: et We e tao Pu a W $1 cent FAlluno '
~1%l0 min thhL F. 103 ^ l-os< AP Esronertmd Strna Eo
}4.' P Nea nnt EB 1 t30 ma thatr u Fura Fa,'lu e - swr & E L '
n awl ineacnst & A %lmwors vmriI 30 % seuu, rm is nea&J O
%S mid Ihhtf 4 h ~
fup) FAik ~ I n H ACE h lo % /nimore 6o 100% seuesiru e
O 17 1
l
l l
l O
OPERATOR ACTIONS Scenario No. Event No. Page of /
Brief
Description:
k6 6 kC. . S6A f Phi /uet ;
PUmf Time Position Candidate Actions / Behavior 0 L en+nce Couleie usiive.
L s h s. 4 ' masin man %I. Wum Uper Goaabiltu Tecr uswc cr. 4. 'C. E' i 19
+2 w Rmine. fumo '
'A '
- I Seal Paibucs Ro Gera m wc Openupes & ninam < nAA r,enm cent cAvira.
bAl n a s Partem
~
Ro Re e oad inrem paro n Reacb c os!
O.13. h o,% R w r Da n Cu b .
I 6I R0 Ar G nio. inraunk recond im TE R R Arn/2- 12 eadinc IF Surn ,0 nuts S ode #A rrt, t y V
. 5(20 k roam Sh,cr sooenvoson, los evervr V
_ SRO .
NlT]ATV Oi h'l/U TdNA/UCA kfG udST~
't o h A v8- CeAl 126 D &lA f d Y b.L hie iT~ O V rfhs C b>noa r=.
18
o kl OPERATOR ACTIONS Scenario No. [ EventNo.Nk Page b 'of / 1
. Brief
Description:
' bhd#'9 b;//Ih r N NOI'.'E 5 't.5. E,d, b. l Time Position Candidate Actions / Behavior E { b,I 6 tDL/2-kh lAC.4!. Nikk tw buBD13 #CGla n heal 0 o ol}>v e f i -l aa'b i
VL Ate 'YPOT Unn hen ll9_d IN Yh AIJVAl & ' Aun bAck to d. ,
On&su OST beat f) AlUt.
r -
1 UfAsFH WAGT StyArw LINK I 4 0lA Siniv IJ A lot C oddA) \
v Smer & M Pc.r au v os I ovn o t I Gnen KPCLT tuohisus Srsnm.
'svDola vm vc :
ir o OIt L CoxirseBL LG12 h obc *A Ilt.
W Ao TTLC VALUf- OOCJT/QAl V
At He arrt r ODeA>s . l)2R I FH
$vabaule >jueREysts 'Sooed 0 v
0 oeiv HPer ouno' diRJdAde eJn/n.
O v '
s _ w a c a n aem v i 19
l l
O1 I
OPERATOR ACTIONS I Scenario No. -
Event No. h Page 3 or /A !
Brief
Description:
1 M bl / ,
I Time Position Candidate Actions / Behavior IMP Ash'usr toebia eacect bo chrdiv Kooo GPih 9 //60 Psin
_d_ c e.L a.
g n o L ,DkeKCuk 5 RFTfrL Cmwowss recored tuo ha w <O p p d ' Pto w lL A w .
x ekcRaola n%ccox '
L cinurio W I4Pn r su <run A de rv ens -h o wha ok1wcx s G
Clordowru 2 H /2 % < r m t1 J-mes s'a < rno% nendwe< <
% LPCI m ndo 0 Reivd tesr IVcults to fG HA r <hepsier m our nc I+ pc c on o 12 a h; I,'ru TEST Q
v O
20 l
< ---------------------j
, -- ~ .
'g, OPERATOR ACTIONS Scenario No. Event No. Page of 1 Brief
Description:
08 b OF bN/d" DA Time Position Candidate Actions / Behavior
+2a na Los s e>P Gwre ws ~ Cwnn l 6, M.P. A w n.r w $9 Ro unifa F Lu. ten o dereemiive o r u {
(Raf (% fc hun d, Il ones.
to det&muuc II.P '
~
idd &Arme s
\O) %am Gn hm low ,Gd9,os e eue
%./ ,
Ihn P Ukeeb TNQ U A}v t. 06SITl OAJ l 1ndi eigrw s ( m u n % S A is ~
nIo < ed) llo Certvtm nur if oxona vs. plou)
I% ou T_Ti o C CDo)lAT5ur YY1I& O V Q I SRO L ferz >x Aluc. Wik>ee. r O
$R6 Ob rn w Cwu vvt. 99, }v rou r oP l
& th PL P6 V Ro 0edut.e_ reeinc . PLoto but .20 % !
o RO .Tn ce #T* d odrie al bod < Yo
/,, e(na APRm + Ran Reol 6 loch i
21
l l
l OPERATOR ACTIONS Scenario No. Event No. Page of M Brief
Description:
M O U##
0$b (Ob C one T/A/ cf )
Time Position Candidate Actions / Behavior ih w b6SS 6 P &>-rDAt now $73D%k Ibd
- 14. p . Hrie q goo (2p a u rA T a u v . noMAnut to .
rLb} d in Cuoh /m 6o ED2 l 1V cT 1
ROP flu otsT <x u v. nn ef. Elose I n Ula o A> %. A ' HTR TR 9(t/l l 520 f/r nno f PAnuar to SAipr
<i,'n en vu e na los e veiv r I ! U
\
Q2D Lvo rta rc mmAnasAneC RtaocsY u
Fu o Th M F /03 .
1 I
22
i 1
',,\ 1
\_, l OPERATOR ACTIONS ;
Scenario No. Event No. Page of /2 1 Brief
Description:
him $nnent zws gad.
2NC46AS/Ng Tirne Position Candidate Actions / Behavior
+3a ow Kn Pa,ks - Slow /m ovaensed F20m Wo to 20 % sev2surm "
Mn-s T+L.w2 m /2FCrlu ed @ .20 h RCf
~
Chore rad n10No Ton $ k dNA/2T Rec.adeo w inMesswr 1 nod (mon irws & Rttn?d&L OSbR&k
(~T AhAh '
Q_) T' (2oP
~
(%Jr opp -os noe -wnrin avr
'2/el.
l rh ou e r@t
[]D Co<w m e +>Le nouse e nejucr,an? -
usiac r&clho . x pds l
0 OOP rn oiv ,rna rad. non irat< fon O 40. d1' C&tA T frS Cots)8d. is V
NP Au ted.
5Ro Aloro e + A)ve . Gries.c of
~
W3slde I:u e I FL oo r v
6(2 O QeneNT A$om Inb. %n/e Rx e on/Ad k nh -0 A< l'?pl <pe 4Nr c ooIAer has 9.W n 6/am dou ea ur J,1- 131.
!'D R1savs s hek ko9cc. henshu mini oo tum %MnE cRhed 'bv"PW s u HTR 'FMwk 23
O OPERATOR ACTIONS Scenario No. Event No. Page of Brief
Description:
1/vC2dAS/Mg O/N I IE29 / h LI N I /be d -
Time Position Candidate Actions / Behavior I
Vued Fen.v,ec - st w ine,e m g
((Otv7/rJue
&Q 0PLTEQ. MIN 2 7~ Y"WAA)fIk% T D t'.
I-I%l iG b4l0(.4) Y b,1 h L'lAh1 ~
Ark ,' r- nc T.S. 'U WNa Yh p Lf 4h W 10 Y, O
l 24
'N OPERATOR ACTIONS Scenario No. Event No. Page of Brief
Description:
g.of.5 ve I Fal/vae.
Time Position Candidate Actions / Behavior FueI PAituipe .Gsv u iu
/ht2f.AfeA Y/2Dth $lo D bb NOl}n '
V / 0 b o / m ; ,s u r r bop Uewv mst R a d . inie s e n s w c_
O ARo Demour chm. InA ci w n le.
f CC60s T h 6PS-4AG '
U YO bblNOVryet 60812 00 oG /D0 SS /blt* l h lC' L 6 A 2 A A Adi E'rboAf
~
O Q0 C bisrn eni + Kx nouMj2 Oedue T70s.S Usi nit AP C]}Q C_ G h I?$$
0 .
f40 N Aiu A L h Qcan DosdAlts op ron to o p uk w ' .2 XAI o
l th, 5'cemx d- feo nr,n I l enimaa KRO L vert 20T -$ 'R P V C oivis) "
4 Ro}Rol60P G eturf COI- A . A . "Le ut A C~rnol# l b Or- L, A "Perssuer G r neos^ '
,% For-6.c
- Powex Con nes /
- O Cru 6 UA (4 thsrLU 25
O' OPERATOR ACTIONS I Scenario No. i Event No. 1 Page of M Brief
Description:
[hf05.s Fue Fai/ vee Time Position Candidate Actions / Behavior 1
fucA Fan udE -/00'), C&eein j (conriivued)
%eol/80P lieeofa lh C1t)s closed U
AoP Pla e + 2di2 suraw in con nnis<,on coy /
lI ooo/isa U f
g I?o Place mod + su in swroami W KYZo I vo,f a Rwcu isolsim iP
- leurj O z + l 0 u l
20 lJe aufm all aoivneo / 12 els nocepra 9 bee eivol cosmw 04' o V 130P T h u m flu onem <Rt/< b /LeAx.c Rv Opp <6.Ad % 4 9 km DS)s .
. > VE, > A<sU,. A itr,/3,7 RO luoifa Qc cure nu nu, < l'unhnck ho twwG Lows LenKI thr A 4D =C1 "
l rop G wr nn vew nurn smarof V
R c z n ~ d- H 9 a 7 k_ 9 _ a_
26
T }
'J OPERATOR ACTIONS Scenario No. L Event No. Page [O of b Brief
Description:
40 G $, Uc( Fm/ age _
Time Position Candidate Actions / Behavior Yue ?Rs luR G ~ l00 bCueeo74 r
[c orvTiivupcA) 0 rop A Fiv72. 1_ r urL k E TVAUS $6 /V oe m A f.
nl Ate. RO TC tm th Ah!UA L b R c~n uc h"
't~un Arnif #G'Detd. AlAtr hlPor in/
[ S T~ ha d $7~' Flo to PATW bo (3 C vrmenT* OA nC . Oon;TR0L seo enrae s. o. s . "scanar Paocchwe '
$tt Ti orti 0. "$t AAm k) $A I '
Z4c)IAr i erv '
[bDf 05ACC lh $Itl SW n rr Hk t //v OLOSE
$O 00Th /ni D O - Y O AtN1Tou T~ 0~ UM I Pt/
Ry no m n d '3 th '
sop C oure el R c r e 'iiv ik o rtb ru to Dn+ veivT h& lerw1 turebiive
%ielp @ +& "
Ro Insur Serh n rern de tecreas i
Ko O ot>>x> e n n i 1 Taim s 1, n aninini,1>
\ A CAAr~tviG0 lDeTix>Niv' 2 0 & 90 (V u ;
27
O OPERATOR ACTIONS Scenario No. Event No. Page //of Brief
Description:
A.Oss Fve-( Fa.,/v,ee Time Position Candidate Actions / Behavior Fuel mitvas - /oo 9; (si/so im C
CCon rinven)
(BOP Tkg p ott varfy rnnin
,1,n e f w , tsul (b oF' Vu,fa e,rnnou loade X-fn wear s rnriA~ xenvier 60 o ne- s i r>=-
nowu? I/s/ k u fans raman <
\
P30P Ua, f a rno rmt suc rion nwn
%77Nh I '
l (HOP Cmpr
)if r avu c b) 900 RPA V V kD [~r. A~ 3 61'&ih~C WAT kk S C)Q A M
$80ul&l A)DT~ be- fe.ser IP P02L :
PA)L vre f I( SUSatc ted V
RolBof theor 6; ,a TE iso /Artw d-/?trvM clenrotio '
'su r rm ta s en uie e-o (40 (?esiner b
~ re miec. Oua as V /
6RO $]eeslAA A Alo 7~I Fi c A 77a s h 0E UNuG0Al iTVchpT. Jth nl#I?1e& T STMFFIN & 0 F $"R F NYSr n
,/~8
\m ,
OPERATOR ACTIONS Scenario No. Event No. Page[kof !A I Brief
Description:
4053 Fue) Fm/va Time Position Candidate Actions / Behavior Fup-l Farfue+ - 100% & PueAtN G
( con > rin>uod) :
rop Commwr RPV con /dovuv
@) 90
- F/h o u,e usin e w /c on MPer 'in 4e GT rnEde
, ~s
(
) Sk0 MT1 r:yl Slu'f r 200 6'a ut s o/2 no P,'ll o u r Acam n e oa s r i
WHO Yn A L F Lj %.
- JOO h, (1:\IngiW
(-
Xh!O ACEhliifI6
)
U _
29
I l
J l
i O
PWR EXAMPLE l
l O
1 1
l O
31
O SCENARIO EVENTS Simulation Facility: A'I t
> 0/N I Scenario No. b i 0
c a.ine,,: S. Aba Can.i.at... 6. kbl#- !
- m. maaw P. LM
> Q , Woe les (b. L oiv e. -
o o l Initial Cen.ition : O O ClNAA2 PO D f ! F o/ut' . s _
e ut of cakc (ya rr TIME MALF. N0. DE$rRIPTIDW thL 1h AL.P CA h/- 44 F4llvo# Mrd U
+I%w inm p, n t S .(',. T o h e L,.nw W 2 9. e f p
F>lI kntuos ( c C h h - in,taarv
(> C A n & dirta OP $2Ana//vc O(n50 ~~
t g QAi _
t TKto 'A ' C anden.gn.e_. O R c . Hitto.
+2,6 mw mALP.99 '
ft?'Pfn oVE in AL F. Civ euG) ..
+'iTino a MALF Of T nc.AFAt[ G A. Ti./he l.fAk (D ao Os /m mn> n:- t a 10 0 %
f(WIDE.if4 O
l
)
l 33 l
l
O DPERATOR ACTIONS l Scenario No. Event No. Page of Brief
Description:
bi~ 4 'l ?At UAE NllY y QPfROXIMAMf hp aprsrz. gods p/AceX w Aura. ,
.Ti fne Position Candidate Actions /Behavler N-4'l PhlLunc H v,H R.6 UETftrh/NE F/20m A t M rw 6 &-
/vIs &dwc s &r A>-## /r UDscnie. 0 '
1
\
ilo iJL Acc l2 ock c oivrieel m n1Anui/
~
- fo hd ius f flod < 4- Tuenin f co Treon'
~
ta %)(os r TAva. to eavne ygg 4 u Mml '
Ro Defear Wc llowave o : .
n 00e2 PnLOGYL R08[ Src)P R ()OLU D Q sh/S Yb Arni .C1.feviT -
cN Ov24EoT ConlPANATOL CtenutT N c Ittwoct c odpAAerort cincuar (DerhovE conmel powen Fusss
- __$/20 Alo TiFV _fsC. _
$flo 18171 Arc _ WapJC AtoueSY 0
920 -_T+forsor plaivr snapp4gmgor_
Sl26 Recott EVENT IN LoQ 1 22 ?olhoV f Courtnue- LoiiH DoDeiL esco/griGk '
E / > f> inhL i S0 v.
. - _ _ _- _ - - - _ - _ ~ ~
em.
OPERATOR ACTIONS Scenario No. Event No. -
Page of Brief
Description:
5.6. Tube leex wiruw C)p,v.47g 0P CHG. PutsPS Time Position Candidate Actions / Behavior 4lo n,.ip fgo/Ro/Boe feQfopm nowM~ E4c A LA'rlorV FoHowim 'G P- 10 0 soc -num vfi Plou t Lin Feedwarm numa.
liv sertvic.f @ KO, Dnwks : 's/Ar e
$2 0d 4 IN buT6
+I % ~ w p , ,: . % - S G. Tube Lean (a) c2 % -
At Aom s V
( (20,llboP ()sto Fy 5.&. 8LownoWN ISOL.
1 lerkww ko Zsol. ivoarn ni Ro t)vsvM mntsvP AVAitAALE To \lcr KO SThk iL o 2 E P2 R L-E\/EL
&(2O Rt &JE5Y Ci+EM- SAshDLE f+L L C G.S Fo/2 Acrivi T/
o (LO PCRFOQ1% LPAK (%A TE te.r7* De&_
S T. 4 . 11 . Al (Leak RArr 'Is w ell /m /o u
- E {oec. h4 e r-)
<20 /12 0
~
Derstmiir -ifo L EAK l< hv Wut
'B' s.G - ;
m hily S Tenth Lo qE Afh6 .
Cowd. Linc, kp/ Al/l CJECTott Afhb i Pktt L.EW1 O STetron
O OPERATOR ACTIONS Scenario No. Event No. Page of 0 0NktN$C & 0I Brief
Description:
(/)ggf og,g ggCompqo/3 by &$./mki%j0-
/9PTO Tirne Position Candidate Actions / Behavior A 20 C ALCUL A TC A AD. A&LG9 S T S2.0/R0 IntTIArt EPP PBL PRocEDr#?E thlO b1ht F. ') L
~
RoleoP De rumuus cino. Durun terhed ha U
ninams d- INdt CADbN $
(bop CJose ore vensfu nuro tuse or
'R ' ,Du n,o d th.} . llALUE ._
~
(bop /R O OpcusM fuepiin loAd A4 nnas Aar to lhnin tncou m aboV6 e;lN' M4 1
6op/R6 5two 22 12v/tunnwc powu (hO P (h 0 NIT 0fL TURAlivE RYHAt M MOO M. Yl6M D/O 6oP .riv, ria rr on varpv Auro lN I VI Artatv 0 P H 6 orb S C A A M L [N ,30 I F
CAO h)nTiPA Skti~ r SvDPA tf/Gn /2 0 V 612 0 .ZNr YlATV Yh AJ/V TD./A/Wf A&WUc5 f*
36
l O
OPERATOR ACTIONS Scenario No. Event No. Page of Brief
Description:
STEW $Duenn Ty6E l2up7vag awpra 4- 50% Powa Time Position Candidate Actions / Behavior sito Re ed roenT in he c (6OP id4w IMInsexi w Is Rmova t.Menvrr ovmD * '
_tna n m fU 20 14ftw flokao
'~
Oerstme~c Mom A A es
- IxdtcAir inu +6(Ar ca.11,e lexu le t an~,~ e -
A Pah Low /M err.
.6 P2/2 Low horss. && o 31 P?12_ leupl chviaru4 LN (kc ) W oivorm2c, e n Secd6 sWe G~
RO TRuP NY li.EDn) hDOAcArlia CcA4A1 IW f rop 1Rup OR wein Auro '
t urawr TRIP ,
rop %,si Dowes to M A 12ues ~
/ I 0 YPAlfJ h$ hCT&lATIO A)
/
KO Jeporv famwam2 iso /A riod O {LD Veas I i
PdhSE & iso % rtoAl 37
i O
OPERATOR ACTIONS Scenario No. Event No. Page of Brief
Description:
SN bMNN E CAPPox* EDVo Mu'$
Time Position Candidate Actions / Behavior it O l/ER)Fy AUY FemkArn2 p uwwuun Pol 60P
, vEttFV, hn SX D vnu,% RuwNuvi a
flo jfbop U EEIF)) 6.C .W /A) Sen uIcK 80NRo '
lleRIF4 Kenut c E wAma ovm n.s R o w itli n n SRo IRO Q
Vetn Fy C,0tantuurnOv7 eneDA Coo /Q R0 Uet2,fa Cou ntiiva ovr- t>eivT i c oI.
(QO \ lee u FrI towm1wn arr GORAv AloT (Jefi or R en
%RolKO '
UeenF4 ORaoM 5_T . Plou> Rnrrs V UAYve ' liv"wD I
(LO JednFV DFL@M Rfid Flow Ri9res
& VAldc '/s/m ,o lio vea,fa. fles t w decneanwi v v P.o vierfa -
Poevs a scam s char v u 0 ? f) GN7h% f-30 vt l
- _ _ - - _ _ _ - - _ -~
N ,
i OPERATOR ACTIONS l Scenario No.
/
- Event No.
hI
/
Page D of U 9 '
Brief
Description:
Sens 6mswa Mr @mM Time Position Candidate Actions / Behavior RO f+FTD2- UOLIFytan 5.I. Fla ta '.
Cro P % Rc)
R of (bop nemtm nuc S. G. 'R ' h n< 1TA e
' re v ')TvR e' ,
1 (bop lRo TsoI&rE S . 0 . '8 '
O A nurr\/ i G h (4rm os. /bR V f Alowdowit/
(2n180P Lucune. 6.G . 'A ' Jr uel > G% '
/
rhop llt0, ZSolnry All Pam To 56. % '
R0 tit'aify Pu P O R.V< 6 los e d 1
(2a V eR n Pvl 5 G . 'l3 ' ISotAn=0
(%o lleAt PV C .G . %' J 'c ' L leve25_?52'%
(2 o Ret >sr S . r. .
Ko (k <e r Phase. n/a Isoknaar
~
(Lo ()e Rn H, i tvs T. MR. To a onrmemar (J~
39 i
O OPERATOR ACTIONS
/
J-q Scenario No. Event No. D Page / of Brief
Description:
5tynm awzum TA Rvrvae
( A ff
)(2 0% i /%IEL} Y 0 Potuerib Time Position Candidate Actions / Behavior SD \)PA.L PJ hhh 00 bOSc6 Csven t2C'O eo n m 'ina lou > Ivad Sr enH he Gronoed iV khA0lW 0DYh m t* ttt E WG C.OblkCV)tt) DC /sv r.
cnmrw danuos o V
(to Urarq 'fb' S.G. leved % w 1
_ ni2 'tw chnswc &' l RolBoP ()en,rq Vcs subcooline civ c aa e K oN e o olnu t m onirk I
[k0 C e nR P<<ve r2 E Rf 6 . Pedud A f. '
P U vCA. IrAft RA W Ro gsrwausa pu wn > a K 9.
G00ll20 %A(nitun Tf" S 2. FLott)
(26 Granu <a Airamel ebnuin L u o Ro Pe u a,pm Date level > 9 5 %
0 '
Rb leoP l
hrnruisH nm2ml iMuud to Ik7~
f Ro fisrnauw ivearnni lerdow W 40
O OPERATOR ACTIONS Scenario No. Event No. Page of Brief
Description:
hM b@VNN2 T bC I2bC IYNC (ApfKodn1&rt%1' 50*), 90 wue'.)
Time Position Candidate Actions / Behavior PD P Ser upt $1ece I -
%er> ERAT 5Y2s u.
(40 Re-vedtP4 DRImann I <ecaudA A mo lenkn a'E ' unuia Wn l
cuhenirical/rm l
l R.o Tavax G (bop lAO Cwre ne c cos Irlowit) usiiv c OP - 91. A O CRo Nor1fa (VRO FofAivr nsvMcn,mr AF c9p fvr 4- Acr75/vs fMrg/tf G~N D %E A) A RI o l
41
O l
I l
f BWR ABNORMAL EVENT DESCRIPTIONS 4
O 1
f O
43
7x (v' )
LO1 LING-WATER REACTOR ABNORMAL EVDITS Master Feedwater Controller Failure Nuclear Instrument Channel Failure Rod Position Indicating System Failure One Reactor Recirculation Pump Trip Trip of Both Recirculation Pumps Recirculation Pump Seal Failure Scoop Tube Lock Increasing Suppression Pool Temperature Drywell Cooler Failure Stuck Control Rod
,,, Uncoupled Control Red
?
\
,/\ Control Rod Drift Control Rod Drive Hydraulic Pump Trip Loss of All CRD Hydraulic Pumps CRD Flow Control Valve Failure Condensate or Condensate Booster Pump Trip Reactor Feedwater Pump Trip Loss of Feedwater Feater Extraction Steam Stator Cooling Water Pump Trip Steam Jet Air Ejector Malfunction Loss of One Reactor Protection System Bus Area Radiation Monitoring System Alarm High Main Steam Line Radiation High Ventilation Exhaust Radiation O's Inadvertent ilPCI or RCIC Initiation Loss of One RBCCW Pump 45
Operating Sequence: Master Feedwater Controller Failure NSSS/ Type: GE/BWR Initial Plant State: Power Operations with Feedwater System in Automatic Level Control Sequence Initiator: Failure of Any Input Signal to the Feedwater Level Control Circuit (Total Steam Flow, Total Feedwater Flow, or RPV Water Level)
Important Plant Parameters: 1) RPV Water Level, 2) RFP Turbine Speeds, if applicable, 3) Feedwater Level Control Valve Position, if applicable Progression of Operator Actions: See Flow Chart Final Plant State: Quick operator action will be needed to prevent a reactor scram if the initial power level is near rated (high level turbine trip if failed open, reactor low water level scram if failed closed). Should operator actions prevent the reactor scram, the final plant state will be essentially unchanged from the initial plant state.
Major Plant Systems: Feedwater, Feedwater Level Control Tolerance Range: The operators must ensure that the RPV water level remains belou the main steam lines and above the top of active fuel at all times.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response R0 - Understanding / Interpretation of Annunciator Alarm / Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Control Board Operation Communications / Crew Interaction BOP - Understanding of Instrument / System Response Control Board Operation Communication / Crew Interaction 47
MASTER FCEDWATER CONTROLLER FAILURE Progression of Operator Actions:
Abnormal RPV level alarm (r on '
Observe RPV water level indications 90 Observe RFP turbit' -
speed or feedwater flow control valve position an/nnn U Place the master feedwater control-1er in MANUAL POP l'
[V)P Return RPV nater Did Did Did Return RPV water level to normal and control man-
.N 6
level reach urbine tri" the contro11. N all open?
RPV level reach low lev N level to normal and control man-ually .etpo . . cram set int ually Y
l' Y cpq 1r RO/SRO/ ROP pn/can/ 1PnqP can 1r Request I & C Did Perform scram Perform scram Request I & C investigate the the react' y actions action investigate the cause of the scram? cause of the coa-controller failure troller failure ti l' prp ir p<p 1r t' Reset and restart when level return: END tre reactor feed- to normal, reset mater pump, if and restart reac-tripped (turoine tor feedwater driven RFP) pump, if tripped rno o j 1
Return RPV nater j level to normal '
cPn '
Inform plant l nanagement and l the t#C of the event and actions j taren i
v EtO 48
i Operating Sequence: Nuclear Instrument Channel Failure
,~\
\ NSSS/ Type: GE/BWR l
\
Initial Plant State: Any STARTUP or RUN Mode of Operation Sequence Initiator: Failure of an SRM, IRM, APRM, or RBM in an Upscale, .
Downscale, or Inoperable Condition l
Important Plant Parameters: 1) Source Range Monitor Levels, 2) Inter- j mediate Range Monitor Levels, 3) Average Power Range Monitor Levels, j
- 4) Rod Block Monitor Levels 1
Progression of Operator Actions: See Flow Chart Final Plant State: No significant change from the initial plant state.
l Major Plant Systems: Source Range Monitor, Intermediate Range Monitor, Power Range Monitors (LPRM, APRM, RBM), Process Computer Tolerance Range: Operation with inoperable nuclear instrumentation is i permitted, provided that the Tech. Specs, for the minimum number of operable channels per trip system are satisfied. In the event that the minimum number of operable channels per trip system cannot be satisfied, the operation of the reactor may continue on the Tech. Spec. LCO that applies, provided that the trip system without the minimum number of operable channels is placed in the tripped condition until the require-
[N ment can be met.
( '-
Competencies Tested:
RO - Understanding / Interpretation of Annunciator / Alarm Signals l Compliance /Use of Technical Specifications
( Compliance /Use of Procedures Control Board Operation l
BOP - Compliance /Use of Procedures Control Board Operation i
l l
l l
l 1
7"'
%_,/
49
NUCLEAR INSTRUMENT CHANNEL FAILURE Progression of Operator Actions:
Channel Fall-ure and Rod )
n109 ilare Rn 1' Check control pan -
el indleating lights for failed channel nn o Verify the instr.
ument is upscale, downscale, or inoperable pq v Check the other instrument chan-nels to verify reactor power is in a safe con-dition Rn 1' Verify the func.
tion switch is in Is tht.
"OPLRATE" Y / failure inoperable an
-hannel'
/
U RO l' glfryp l '
Request instrteeni Perform functional technician check test of instrument high voltage, channel electronics modules, LPRM in-puts, etc.
l .I pn Bypass and log out the failed f4
(
the inst-instrument chan- 4 .went M' 6
"#I stored?
Y RO l' Re et t. cram sig-nai, if present, wid w r i f y the Corit ro] icd v<itn-1rewal block :s
<! rated FMI U Refer to Tech.
' "' for inst-
. S'it r d t t h.1 t init-
. 13tL c:1.e .c ,.
! - talDJ S I
~$~
50
l
. Operating Sequence: Rod Position Indicating System Failure y-
[ 1- NSSS/ Type: GE/BWR
'%l Initial P?snt State: Any Condition Sequence Initiator: Loss of Power to the RPIS Electronics Important Plant Parameters: 1) Full Core Rod Position Indications, if applicable, 2) Four Rod Position Indication, 3) Process Computer Rod Position Readout Progression of Operator Actions: See Flow Chart Final Plant State: Same as initial plant state.
Major Flant Systems: Electrical Distribution, RPIS, Nuclear Instrumen-tation Tolerance Range: No control rod movement is permitted without rod position indication.
Competencies Tested:
SRO - Compliance /Use of Procedures Compliance /Use of Technical Specifications (if applicable)
(} RO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures (m) l i
hs 4
51
RPIS FAILURE Loss of Rod Progression of Operator P ..on In-Actions: d on 90 l' 5 top all control roe movement in progress 1
i I
90 1F l
Monitor nuclear instrumentation l l
SPD t Reque3t auxilixar) operator ct.eck breaker alignment, fuses, circuit
}determine cards,etc.to problem P!) U When the problen is corrected, re-set alarins and re-initialJze the RWM (if below bypass setpoint) ,
9 END 52
Operating Sequence: One Reactor Recirculation Pump Trip 7"3 NSSS/ Type: GE/BWR (w/ )
Initial Plant State: Rated Conditions Sequence Initiator: One Reactor Recirculation Pump Inadvertently Trips Important Plant Parameters: 1) Reactor Power Level, 2) Core Flow, 3) Re-circulation Loop Flows, 4) Reactor Water Level, 5) Recirculation Pump Speeds, 6) Recirculation Loop Temperatures, 7) Reactor Steam Dome Pressure Converted to Saturation Temperature, 8) Bottom Head Drain Temperature Progression of Operator Actions: See Flow Chart Final Plant State: There are two possible plant states for this event.
The operators, if permitted, may return the plant to the operating state that existed prior to the trip by restarting the tripped recirculation pump. If not permitted to restart the recirculation pump, plant opera-tion will continue at a lower pc.wer level with one recirculation pump in operation. In this case, the continued operations will be restricted by the Tech. Spec. LCO on one loop operation (typically 12 or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i of continued operation).
Major Plant Systems: Reactor Recirculation, Neutron Monitoring, Control Rods, Reactor Vessel Instrumentation Feedwater (possibly) 4
(
Tolerance Range: Immediately following the recirculation pump trip, the
_, /
high APRM level alarms indicate that thermal hydraulic limits may be exceeded or approached. Therefore, the operators should take immediate action to insert control rods to reduce the reactor power to maintain the thermal parameters within their Tech. Spec. limits.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response ,
Compliance /Use of Technical Specifications {
Compliance /Use of Procedures j Supervisory Ability J Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response compliance /Use of Technical Specification Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Control Board Operation Communications / Crew Interaction bi i NOTE: This event description applies to plants with motor-generator set
\s_- powered recirculation pumps. Plants equipped with recirculation flow control will differ la the progression of operator actions.
53
OrG REACTOR RECIRCULATION PUW TRIP on U Progression of Operator Actions:
Page 1 of 2 r umps Check the Master Controller is on j MANUAL and Run '
Back to Minimum
.RO o on eenn < .
Vt:rify Recire. l Request Auxiliary Pump Tripped and l Operator Reset Running Pump Loop Lockout Relay (s)
Flow is Less Than or Equal to 100t of Rated an o RO o Place the Remain. Verity the Suc-ing Recirc. Pump tion Valve is SDeed Controller Open the Dis-on Individual charge Valve is Manual (if in Mas. Closed, and Reset ter Manu 1) the Scoco Tube Pq v Rn u Place the Control Reduce tne Speed Switch for the of the Running
, Tripped Pump in Pump to Less tne STOP Position Than 50%
and Run Back Con.
troller Rn w RO/ ROP 1r Monitor Reactor Verify the at on Water Level Re. Vessel, Loop to mains Within the Loop, and Loop to Normal Operating Vessel are Within Band Tech. Spec. i Limits j nn w on w r Open the Dis. Verify that Reae. Are charge Valve on tar Power All at N the Tripped Pump Decreases Limits Het Bl to Maintain Loop ?
Temperature Y
SRO u 40 u RO v Request an Insert Control Close the M-G Estinate of the Rods in Sequence Set Motor Dreaker Repair Time as Necessary to Verify the M-C Clear the APRM Set Starts, and A1 and RBM Rod Blocks the Field Breaker Closes j RPn ir RO/nno u 40 w Refer to Tech. Close the Dis- Open or verify Spec. LCO for charge Valve on Auto Opening of Time Limit on the Tripped Pump the Pump One Loop Discharge Valve Operation enn ir pn /enn gr RO w spo <r Notify Plant Request an Aux. Verify Lore flow Notify Plant Management of iliary Operator and Reactor Power Managementof Event and Actions Investigate the Increases as the Event and Actions Taken Cause of the Pump Pump Discharge Taken Trip Valve Opens
,, ' pn v v Can When the Dis-END .
the pump y Charge Valve is END I
be restarted Full Open, Return 7 to Normal Recirc.
l System Operation l
54
ONE REACTOR RECIRCULATION PUMP TRIP Progression of Operator Actions:
Page 2 of 2
>V
~
{
l B1 N
pn Refer to Tech.
Spec. LCO for
- i'\
is the Time Limit of N Vessel di Operation With Within Limi One Pump in CPrvice Y
SRO " RO 'r Notify Plant Open the Pump Management of Discharge Valve Event and Actions Taken ir pg ir END Allow Natural Circulation in the Loop to Raise the Temperature of the Idle Loop n, P Close the Pump (4 Discharge Valve v
A1 l
l l 55 l l l 1 !
i
)
7_s Operating Sequence: Trip of Both Recirculation Pumps i
(/ NSSS/ Type: GE/BWR Initial Plant State: Rated Conditions Sequence Initiator: Inadvertent Trip of Both Reactor Recirculation Pumps Important Plant Parameters: 1) Reactor Power Level, 2) Core Flow, 3) Re-circulation Loop Flows, 4) Reactor Water Level, 5) Recirculation Pump Speeds, 6) Recirculation Loop Temperatures, 7) Reactor Steam Dome Pres-sure Converted to Saturation Temperature, 8) Bottom Head Drain Temperature Progression of Operator Actions: See Flow Chart Final Plant State: For plants that have not performed a safety analysis of natural circulation operation, a reactor scram is required by Tech.
Specs, and procedures. The final plant state for these plants will be a hot shutdown condition. If the cause of the recirculation pump trips are corrected, the pumps may be restarted after the reactor has been scrammed. For plants that permit natural circulation operation, there are three possible final plant states. Operation may continue at a l
lower power level in natural circulation on a Tech. Spec. LCO time limit l
(typically 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />). If the operation of one recirculation pump can be restored, another Tech. Spec. LCO time limit must be used (typically 24
< '$ hours). If both recirculation pumps can be restored to operation, power
- / operations may continue indefinitely, i N.. / !
Major Plant Systems: Reactor Recirculation, Neutron Monitoring, Control Rods, Reactor Vessel Instrumentation, Feedwater Tolerance Range: If applicable, the reactor should be scrammed immedi-ately when it is determined that no recirculation pumps are operating.
If scram is not required, the operators must take immediate action to j insert control rods to ensure that the thermal hydraulic limits on the reactor fuel are not exceeded.
l Competencies Tested:
SRO - Understanding of Instrument / System Response Compliance /Use of Technical Specifications Compliance /Use of Procedures RO - Understanding of Instrument / System Response Compliance /Use of Technical Specifications l
Compliance /Use of Procedures
- Control Board Operation BOP - Control Board Operation
,__s NOTE: This event description applies to plants with motor-generated set
( powered recirculation pumps. Plants equipped with recirculation flow
}
(_,/ control will differ in the progression of operator actions.
57
TRIP OF DOTH RECIRC PUMPS Recirc. Pump Progression of Operator Actions: Trip Alarms Page 1 of 2 on Manually Scram the Reactor is (see Progression y Reactor of Operator Act- Scram Requir ions for Rx Scram) ?
l R0/80P P RO "
f Control Reactor Verify Reactor Water Level Above Water Level Hold-Normal to Increase ing in the Nor-Natural Circul- mal Operating ation Band y Rn <
Verify Reactor Can Power Level Dec-y One or reased to Natural 53 1 Both Pumps be Circulation Level Restart
?
N Al epn V ominno 8 Report Event and Actions Taken to Remove feedwater Pump (s) and Con-Plant Management densate System Pump (s) as Nec-essary enn o pn '
SRO "
Classify the Event and Notify Insert Control Refer to Tech.
Rods by the Rod Spec. LCO for Plant Personnel Pattern to Return Time Limit for to Within Operat- Operation in ing Limits Natural Circulaticq can i con i FRn "
Notify the NRC ad of the Event Direct the Aux-111arY Operator " ""98**
to Investigate Events and Action:
the Cause of the Taken Pump Trips y <
v END END One o N Both Pumps be Restar
?
Y IP B1 Oi, 58 i
- l
TRIP OF BOTH RECIRC. PUMPS Progression of Operator Actions:
Page 2 of 2 p
~
B1 i, 1 7
'w/ U '
nn RO Run the Master Open or Verify Controller Back Auto Opening of to Minimum Output the Pump Disch.
arge Valve pq l' '
RO Place the Indiv- Verify Core Flow idual Controllers and Reactor Power on MANUAL and Run 'ncrease as the Back to Minimum Jischarge Valve 81 Output ] pens u
PO "
, Y Llose the Pump ,
Dirtharge Valve Can and verify tne Are y the Secon I Suction valve is Both-Pumps Pump be Full Open unning? (tarted' at g RO l' cPO P SRO i r" a
Notify Plant Refer to Tech.
) sm Reset the Lockout Management of the Spec. LCO for l Vessel AT Event and Actions % Time Limit on Relay (s)
Within Limit Taken One Loop Opera-
? tion
/' N y l ) R0 Pn U
\ / Open the Pump Reset the Scoop V
Discharge Valve Tube END rm Pn l' PO/90P 1 '
Allow Natural Verify the AT on Circulation in the Vessel, Loop the Loop to Raise to Loop, and Loop the Loop Temper- to Vessel are ature Within Tech. Spec.
Limits en "
Close the Pump Discharge Valve Are N All Temper-ture Limite Met?
pq u Y Close the M-C Set Motor Breaker, B1 Verify the M-C Set Starts, and the Field Breaker Closes I
n v
59
f-~g Operating Sequence Recirculation Pump Seal Failure
( \
\h NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Failure of One or Both Shaft Seals on One of the Reactor Recirculation Pumps Important Plant Parameters: 1) Reactor Vessel Pressure, 2) Recirc. Pump Seal Cavity Pressures, 3) Drywell Equipment Drain Sump Flow Integrator Reading, 4) Drywell Floor Drain Sump Flow Integrator Reading Progression of Operator Actions: See Flow Chart Final Plant State: If only one seal on the affected pump fails, normal operation may continue. If both seals fail on the affected pump, the final plant state will likely be one loop operation at a reduced power level.
Major Plant Systems: Reactor Recirculation, Radiation Waste (Drywell Equipment Drain Sump), Drywell Atmospheric Monitoring Tolerance Range: Operation with total drywell leakage greater than the Tech. Spec. limit for identified and unidentified leakage is not per-mitted. Either the loop must be shutdown and isolated or the reactor shutdown if the leakage exceeds the Tech. Spec. limit.
(' Ng
\'j Competencies Tested:
SRO - Compliance /Use of Technical Specifications Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings BOP - Diagnosis of Events / Conditions Based on Signals / Readings O
l 61 l .___ . _____-_ ___________ -_______ .
RECIRCULATION PUMP SEAL FAILURE Progression of Operator Actions: ,
ute Alarm (s) !
RO F
}
i Creck seal cavity pressures to det-ermine wh1Ch seal (s) have failed RD Continue normal recirculation d pump operation 7,g Y
pc/non 1' nnntop u Monitor crywell mnitor drywell drain sump pumps drain sump pumps for excessive for excessive operation operation L
v Sno Y lDeterminethe flow rate from the affected pump seals to the dr) sell equipment en c_ ms pg o SRO u Isolate seal purgc shutdown and iso-fiefer to Tech.
Specs. for limits late the affected total drywell recirculation leakage (ident-pump ified + unident-s r s c. n gpg o o Refer to Tech. Is Specs. for the Y the Tech.
LCD on single Spec. limit pump operation
'aceede N
non v pq man 1P Check drynell Continue f reawnt atmosphere con- monitoring of the ditions to verify sump pumps for pump is isolated excessive opera-tion ran p nnn U Notjfy plant man- Monitor drywell agement of the temperature, p es-events and action sure, and humidity taken for increasing trends l
_ .x r
1 62 '
i l
_ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -_----J
Operating Sequence: Scoop Tube Lock NSSS/ Type: GE/BWR
(
U Initial Plant State: Any Condition with Recirculation Pumps Operating Sequence Initiator: Loss of M-G Set Speed Control Signal Important Plant Parameters: 1) Recirculation Pump Speed, 2) Recire. Pump Controller Output Signal Progression of Operator Actions: See Flow Chart Final Plant State: If the controller is restored to operation, there will be no difference from the initial plant state. Without operation of the controller restored, the final plant state will be the same as the initial plant state, except that one recirculation pump will be in local manual control.
Major Plant Systems: Reactor Recirculation Tolerance Range: None Competencies Tested:
SRO - Supervisory Ability i /N RO - Understanding / Interpretation of Annunciator / Alarm Signals
( p) Understanding of Instrument / System Response Control Board Operation i
r
~
63
SCOOP TUBE LOCK Scoop tube Progression of Operator Actions:
lock or signal finen nin- 5)
Rn Place the control..
ler for the aff-P"* 0" {S g tt 1 [na e manual?
mn V Request I & C investigate and m repair the failed pump co.itroller SRO P pq Request an aux.
' Adjust pump con.
111ary operator. Can remove power to N the M otrol- Y troller to match the scoop tube ler U. iepa output signal brake ', with the actual speed of the pump SRO pn i' Request auxiliary Reset the scoop I. operator operate the scoop tube tube lock signal
'-'I with the hand crank as needed j to contr?1 oumo ,
i s
t
}
i'
' RO u i; END Operate the pump k in manual to '
verify proper operation of the controller 6
t RD 1' l '
Return recire.
control to master manual, if app-licable END
' )
l
) 4 f
l O
w s 64
Operating Sequence: Increasing Suppression Pool Temperature NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Suppression Pool Temperature Increaces Above 95 degrees F Important Plant Parameters: 1) Suppression Pool Temperature, 2) RHR Flow Rate Progression of Operator Actions: See Flow Chart Final Plant State: No significant change in initial plant state. Sup-pression pool cooling will be in operation.
Major Plant Systems: Main Steam, RCIC, HPCI, Primary Containment, Resi-3 dual Heat Removal Tolerance Range: Appropriate action should be taken to limit the sup-pression pool temperature to less than 110 degrees F. Should the suppression pool temperature reach 110 degrees F, the operators should scram the reactor.
Competencies Tested:
O SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures BOP - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Technical Specifications Compliance /Use of Procedures Control Board Operation O
65
INCREASING SUPPRESSION POOL TEMPERATURE High supp. ,
Progression of Operator Actions: Mltemp. m nm "
verify suppressier i pool tempreature g j is ?-[95 f) (the ]
alarm point) j SPO u Refer to lech.
Specs, for temp-erature limits and actions for high suppression pool temperature nnn o vetermine if SW's. Is are open, HPCI C1 y RCIC is operating pocItgmp.
y [105 r]?
N BOP U ngo e Stop all testing Place one loop of of SRV's, HPCI, RHR in the supp-or RCIC
- ression pool cool-ing mode of oper-ation RSP 1' Monitor suppress-100 pool temper-ature for decrease r
END l
i l
4 O
1 1
66
a Operating Sequence: Drywell Cooler Failure
. A ..
k 'NSSS/Typei-GE/BWR Initial Plant State: Power Operations Sequence Initiator: Loss of Either the Drywell Cooling Unit Fan or Cool-ing Water Flow Through One or More Drywell Cooling Units Important Plant Parameters: 1) Drywell Temperature, 2) Drywell Pressure,
- 3) RBCCW System Flow Progression of Operator Actions: See F, low Chart Finai Plant State: There are three possible final plant states depending on the degree of the failure. Reactor' operation may continue at the initial plant state if the failure involves only one drywell cooler. If I more than one cooles is lost, the reactor may be operating at a' lower power level due to the requirement that one or both recirculation pumps be shutdown on high motor winding temperature. If several drywell coolers are lost, the reactcr mai scram on high drywell pressure.
Major Plant Systems: Reactor Building Closed Cooling Water System, Reactor Recirculation, High Pressure Coolant Injection or High Pressure Core Spray, Low Pressure Coolant Injection, Low Pressure Core Spray f-~g Tolerance Range: Operator actions should prevent drywell pressure and temperature from reaching the technical specification safety limits.
)
Competencies Tested:
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals
! Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction t
1 67
- - - - - - - - _ - - - - - - - - - - - - - - _ _ _ _ _ _ .I
DRYWELL COOLER FAILL'RE Progression of Operator Attlans:
11gh Drywell Temp. and/or we In n rmi, pf)D l' 'I Check operation 73 ys of' the drywell drywell N N cooler fans ressure Al2 ,
Q~[45
) e))0M F]? '
'\ END i
l psig)? \
s Y
( Y Hnp U SRn/00 / r 9no ano o Check cooling Perform scram Monitor IeCirc-water supply to actions ulation pump the cooling units motor minding temperatures nnp u nop u Start or verify verify ECCS act- Is auto start of Jated on high ,ither mot-standby cooling drywell pressure r winding temp N
[ END fan (s) 0 2[216 F
?
Y nnn U B'P D Rn U Verify valve line. Shutdown the High RedJce the speed up or align cool. Pressure Coolant of the affected ing water to the Injection (or High recirculation standby cooling Pressure Core pump (s) unit (s) Spray) system
/l
@Op/Rn ir 900 i' Rn if Monitor drywell Prevent automatic shutoown the pressure initiation of the affected recirc-Core Spray and ulation pump (s)
- .ow Pressure Cool-ant Injection systems goomo u y u Monitor drywell Refer to Tech.
temperature Specs for LCO on operation with inoperable recirc<
ulation pump (s) v END O
68
Operating Sequence: Stuck Control Rod NSSS/ Type: GE/BWR Initial Plant State: Any STARTUP or RUN Mode of Operation Sequence Initiator: Failure of the Selected Control Rod to Move in Response to an Insert or Withdraw Signel Important Plant Parameters: 1) Selected Control Rod Position, 2) CRD Drive Water Differential Pressure, 3) CRD Hydraulic System Total Flow Progression of Operator Actions: See Flow Chart Final Plant State: If there is no violation of the technical speci-fications for inoperable control rods, there will be no significant difference from the initial plant state. If there is a technical specification violation for the number of inoperable control rods, the reactor will be shutdown.
Major Plant Systems: Same as important plant parameters Tolerance Range: No control rod movement is permitted until the problem with the in-sequence rod is corrected or the rod is declared inoperable and the technical specifications for inoperable control rods is satis-fled.
Competencies Tested:
RO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction 69
STUCK CONTROL ROD Progression of Operator Actions:
'ontrol Hod Falls to Move
,,r emmand pq
no <r verify the select. Return trie crive ed control rod is water pressure to in secuence for normal i RSCS (less than l 50% reactor power) e 90 P 90 o verify CHD hydrau- Hequest the aux-lic pump operation "lliary operator pressures, flows, check valve line-and valve lineup up and operation are all normal of the solenoid valves at the HCU 90 1' '
40 Give the control Stop all control rod a continuous *S rod movement until insert signal for the HCU N repairs are made several seconds normal? to the HCU in an attempt to flush the drive Y
Pq
l '
id oeclare the cont-END
) Y the contru Tod move ?
rol rod inoperabit N
pn v on v "
no increase the drive Hequest tne aux-is Heavest adjustment
,*ater dif ferential 111ary operator continue Y to the Control aressure to [350 isolate the aff-rod movement rod pattern from asid) ected HCU (except the nuclear eng-for cooling water ermitte 7 ineer N
PO 1' en l no o j r Attempt to notch Request the aux- c.
the control rod lliary operator u n na t-reador su. END electrically djs- ,
arm the affected solenoid valves an o pn y ir leturn the crive \ Refer to Tecn, aater pressure to " '
Y e ntro N
'Tormal rod move? er ble control rods l
END 70
Operating Sequence: Uncoupled Control Rod
/~N I. ) NSSS/ Type: GE/BWR LJ
. Initial Plant State: Any STARTUP or RUN Mode of Operation Sequence Initiator: The Selected Control Rod Withdraws Past Position "48" When an Overtravel Check is Performed Important Plant Parameters: 1) Selected. Control Rod Position, 2) Neutron Flux on the Appropriate Nuclear Instrumentation Progression of Operator Actions: See Flow Chart Final Plant State: If there is no violation of the technical speci-fication for inoperable control rods, there will be no significant difference from the initial plant state. If there is a technical specification violation for the number of inoperable control rods, the reactor will be shutdown.
Major Plant Systems: Same as important plant parameters Tolerance Range: No control rod movement is permitted until the problem with the in-sequence rod is corrected or the rod is declared inoperable and the technical specifications for inoperable control rods is sat-isfied.
[#
\
h Competencies Tested:
/
RO - Compliance /Use of Technical Specifications Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction I
v 71
UfCOUPLED CONTROL ROD Progression of Operator Actions:
Overtravel Alarm _
Rn
on. 1' Check nuclear Insert the cont.
Instrumentation rol rod to pos-to ensure that ition "00" and neutron flux res- declare the rod ponded to the rod withdrawal l inoperable 40 0 P0 l' Insert the con. Request the aux-trol rod one iliary operator notch isolate the aff-ected HCU (except for cooling water RO l RD 1r Check for recoup- Request the aux-ling by attempt- iliary operator Ing to withdraw electrically dis-the rod past pos- arm the affected Atlon "48" solenoid valves Dd efer t Tech.
} Y the con. Specs. for inop-END trol rod re- erable control couple? rods N
no 1' Rn Repeat notch in- Request adjustment Is -
sert and coupling continue- y tc the control check for several rod movement rod pattern from attempts nermittedo the nuclear eng.
Ineer N
Did men e an END Y the contro o I actor END e shutdown j rod retouple
?
N l'
END 9
72
Operating Sequence: Control Rod Drift n NSSS/ Type: GE/BWR (N_-
Initial Plant State: Any STARTUP or RUN Mode of Operation Sequence Initiator: One Control Rod Begins to Drift from its Latched Position Important Plant Parameters: 1) Control Rod Position Digital Indication,
- 2) Rod Worth Minimizer Insert and Withdraw Errors, 3) Neutron Flux Level Indication on the Appropriate Neutron Monitoring Subsystem (s), 4) Con-trol Rod Drive Hydraulics Cooling Water Header Pressure Progression of Operator Actions: See Flow Chart Final Plant State: No significant difference in plant state from the initial plant condition. Power level may be slightly lower if the drifting rod is declared inoperable and disarmed in the fully inserted position.
Major Plant Systems: CRD Hydraulics, Rod Position Indicating System, Rod Worth Minimizer, Reactor Manual Control, Nuclear Instrumentation Tolerance Range: The operators should not move any other control rods or make any adjustments to reactor power level until the rod drift problem is corrected or the reactor engineer has provided a new rod pattern to f'~' compensate for the inoperable control rod.
(
\- Competencies Tested:
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Technical Specifications Control Board Operation I
f I
p.
N_- i 73
CONTROL ROD DRIFT Progression of Operation Actions: Control Rod PACE 1 0F 2 Drift Alarm Rn \'
Observe Full Core Display to Deter-mine Which Con-trol Rod is Drift-.
Ing on v Select the Drift-ing Control Rod nn y nn Attempt to With. Attempt to Insert draw the Control Is the Control Rod Rod to Its Orig- IN the Rod OtiT11ng In o" OUT to Its Original inal Position and Position Hold Out?
Rn "
eRn "
Check Control Report the Event Room Indication and Actions Taken y Does for Excessive to Plant Mant<ge-Cooling Water the Rod Con-ment and Record nue to Pressure in Daily Log rif' Y
nn 1' Adjust Cooling "
y Water Pressure and Restore the Y Cool ng END Affected Control Water Pressur j 03 Rod to Its Orig- High?
inal Posjtion N
RO/cA9
Direct Auxiliary Operator to Check (END for Leaking Scram Valves at the Affected HCU v
v N 81 Scram Valve : B2
' esking?
O 74
CONTROL ROD DRIFT j Progression of Operator Actions j m PAGE 2 OF 2
! l
% ,/
81 B2 ---* B3
' " U Rn/c.pn on/enn nn l Direct Auxiliary Direct the Auxil- Insert the Cont-Operator to Check lary Operator to rol Rod to the Control Air Press- Check for a Leak- "00" Position ure at the Affect. ing Insert Direct-ed HCU lonal Control Valve nn ienn e,Rn U Direct the Aux 11 Declare the Rod s IS Inoperable and Y lary Operator to Restore Control ' ^ ~
p on V iliary Operator Air Pressure to Leakin 7
the Affected HCU to Disarm the Di-7 rectional Valves N Y pn/cpn " nn 'I eRn 'r Direct the Aux 11- Cive the Control Direct the Aux 11 lary Operator to Rod an Insert lary Operator to Scram the Rod in Signal in an Isolate the HCU an Attempt to Re- Attempt to Reseat Except for Cool-seat the Leaking the Leaking Valve ing Water K Scram Valve v I
( )
son '
\ j ,,
Consult With the Reactor Engineer Does y to Determine the the Rod i g3 Effects of the Continue Inap Rod on Flux
- 1ft' and Pattern Adius .-
ments N 'P RO o ern Refer to Tech, Restore the Con- Specs. for Inop-trol Rod to Its etable Control Original Position Rods e,nn u enn v Report the Event Report the Event and Action Taken and Actions Taken to Plant Manage-and Record the ment and Record Event in the in Daily Log Daily Log v v END END gg
/ \
I ]
'%_,,,/
75 l
Operating Sequence:. Control Rod Drive Hydraulic Purp Trip NSSS/ Type: GE/BWR
.p Initial Plant State: Any STARTUP or RUN Mode of Operation L(
N Sequence Initiator: Trip of the Running CRD Pump on Low Suction Pressure or Motor Overload
.Important Plant Parameters: 1) CRD System Flow, 2) CRD Cooling Water
. Differential Pressure, 3) CRD Drive Water Differential Pressure,
- 4) Scram Accumulator Low Pressure Indicators Progression of Operator Actions: See Flow Chart Final Plant' State: No significant change from the initial plant state.
' Major Plant Systems: Control Rod Drive Hydraulics Tolerance Range: CRD hydraulic pressure and flow must be restored before the maximum number of low accumulator pressure alarms are received (typically five) or high control rod drive mechanism temperature alarms are received.
Competencies Tested:
RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of. Instrument / System Response Control Board Operation Communications / Crew Interaction
.t
\
i O
gT 1
77
- --- -- --------------- .J
i CONTROL ROD DRIVE HYDRAULIC PUMP TRIP Progression of Operator Actions:
Pump trip and O .
or low flow l l
A l m tm ' e. )
RO ir Rn 9 Request the aux- Restart the pump iliary operator or start the check the suction standby pump path valve lineup s
R0 P RO P Request the aux- Allow any dis-iliary operator charged accumul-check for high ators to fully suction strainer recharge differential pressure Rn 1' Rn 1r Request the aux- y Open the flow con-111ary operator Y diffe entia rol valve for b e place the standby required flow and press e 51 suction strainer place the flow in service controller in AUTO N
Rn ir RO l' Place the flow Adjust the cool-controller in ing water pressure c MANIJAL and reduce control valve for the controller the required diff-output to zero erential pressure Rn 1P PO l' Fully open the Adjust the drive cooling water water pressure pressure control control valve for valve the required diff-srential pressure Ro P "
p Fully open the drive Mater press- END ure control valve 1-s 78
Operating Sequence: Loss of All CRD Hydraulic Pumps (D
+ ; NSSS/ Type: GE/BWR
. Q./ -
Initial Plant State: Startup or Power Operations Sequence Initiator: Failure of the Running CRD Hydraulic Pump Important Plant Parameters: 1) CRD Scram Accumulator Pressures (Alarm),
- 2) Control Rod Drive Temperatures, 3) Reactor Vessel Pressure Progression of Operator Actions: See Flow Chart Final Plant State: If above 600 psig and no high drive mechanism alarm is received, reactor operation may continue without rod movement. If less than 600 psig or high drive temperature alarm is received, the reactor will be shutdown.
Major Plant Systems: Control Rod Drive Hydraulics Tolerance Range: With reactor pressure less than 600 psig the reactor must be scrammed immediately upon receiving the second low accumulator pressure alarm.
Competencies Tested:
SRO - Compliance /Use of Technical Specifications l Compliance /Use of Procedures RO - Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation BOP - Compliance /Use of Procedures Control Board Operation O(h. ,
, 79
-- ~
LOSS OF ALL CRO HYDRttULIC POPS Progression of Operator Actions:
Pump Trip Alarm (s) nno Attempt to start the spare CRD pump nnp Monitor control Is rod drive mecha- RX pressur.
nism temperatures N q ess than [600 PS193
?
Y
' NOTE: For scram followup actions, pt refer to the DWR energency N drive emp* N tha }
n ridon for reador alarm nw accum. pree S 8 #8
? m a
alar? /s Y
1P kX Exit to Scram N scram re- g y power N E: tit to Scram quired by pro - above R /
Procedure / rocedure cedure 7
? /
com/rm i, nno Kinually screm the reactor and enter RPV Control Procedure 590/R0 4 BOP 90/PD i, pnp SR0/90 4 900 Execute " Monitor Execute "Honitor Execute " Monitor and Control RPV and Control RPV and Control Reac-Level" subpro- Pressure" sub- tor Power" sub-Cedure procedure procedure enn AnP i PO ,r verify no isola- Verify turbine Place the reactor tion or ECCS ini- bypass valves mode switch in tlation signals are controlling SHUTDOWN exist reactor pressure pn ,, ,, pn ir Verify all con ~ ye y reador Exit to Scram recirculadon trol rods inser- Procedure ted beyond the pumps runback to MSBWp minimum speed and core flow
, o Exit to Scram Exit to scram Procedure (See note) s Pr~#Pocedu[e)
I N'2. 0 80
Operating Sequence: CRD Flow Control Valve Failure NSSS/ Type: GE/BWR Initial Plant State: Any Plant State Where Control Rods Are Being Manipulated Sequence Initiator: Closure of the CRD Flow Control Valve Due to Mechanical Failure Important Plant Parameters: 1) CRD Total System Flow, 2) Charging Header Pressure, 3) Drive Water Pressure, 4) Cooling Water Pressure, 5) FCV Position Indication Progression of Operator Actions: Gee l'10w Chart Final Ple7t State: No significant change from the initial plant state.
Major Plant Systems: Control Rod Drive Hydraulics Tolerance Range: At high reactor temperatures, the cooling water to the control rod drive mechanisms must be restored to avoid high temperature in the control rod drives. The alarm setpoint for the high temperatures is typically 350 degrees F.
Competencies Tested:
SRO - Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Control Board Operation 81
CRO FLOW CONTROL VALVE FAILURE Progression of Operator Actions: Charging Hea er Press. Al c:
ko v Verify charging l water header '
pressure is high Pn n Observe no cool-water flow, cool-water 6P, and no drive water P is indicated Rn "
Observe position indicating lights for the FCV in-dicate the valve is fully clased enn u Request auxiliary operator valve in air supply to the standby valvt '
i and open isolatico valves l
on u Place fLv contro] -
ler in MANUAL and zero the con-troller output signal epn u Request auxiliar:-
operator place the local selec-tor switch in position for the etv @ v va}ve Pn P Manually adjust controller to establish desired flows and 2P's Rn P Place the FCV controller in AUTO
't END r
82
l Cperating Sequence: Condensate or Condensate Booster Pump Trip
,~
i ) NSSS/ Type: GE/BWR
% .sI Initial Plant State: Power Operations or STARTUP with Reactor Water ,
Leve) Maintained by the Condensate System !
Sequence Initiator: Inadvertent Trip of One Condensate or Condensate .
Booster Pump Important Plant Parameters: 1) Reactor Water Level, 2) Total Feedwater System Flow, S) Condensate Pump Discharge Header Pressure, 4) Condensate Booster Pump Ditcharge Header Pressure, 5) Reactor Feedwater Pump Suc-tion Pressure Progression of Operator Actions: See Flow Chart Final Plant State: With quick operator action, the plant should continue operation at the initial power level, or slightly lower. If all operat-ing feedwater pumps trip, the reactor will scram on low water level. In this case, the reactor will be in hot shutdown with the possibility that the main steam isolation valves will be closed.
Major Plant Systems: Condensate and Feedwater, Reactor Core Isolation Cooling Reactor Recirculation, Control Rods Tolerance Range: Water level at all times should not be permitted to
/ ~s drop below the technical specification safety limit.
\-) Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings RO - Understanding of Instrument / System Response Control Board Operation BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures l
l i
f s ,
,s_- ,
j 83
1 CONDENSA~.E OR CONDENSATE BOOSTER PUbP TRIP Progression of Operator Actions: j l
l 1
on Booster j Pump or RTP i 9 ctior Pre" I
BOP '
Start or verify automatic start of standbv pump Ro U Monitor reactor water level and power level v
d
} N A FP's trip END j on low suctio
> ressure
?
Y nnP "
Reset RFP trips and restart feed-water purp(s) v Di END }1 g the reat-tot scram on
) on level
?
Y SRO/RO/ 1 ' BOP Perform Scram actions B0p 'f nnP Use the feedwater necessary, Did system to return N the main Y S art E and reactor water steam lines return reactor level to normal water level to so, late
, normal o 'r END END 84 I
J
l Operating Sequence: Reactor Feedwater Pump Trip fN NSSS/ Type: GE/BWR
{v) Initial Plant State: Power Operations Sequence Initiator: Inadvertent Trip of One Reactor Feedwater Pump Important Plant Parameters: 1) Reactor Water Level, 2) Reactor Power Level, 3) Recirculation Pump Speeds, 4) Total Feedwater Flow Progression of Operator Actions: See Flow Chart Final Plant State: With one feedwater pump tripped from full power, the final plant state will be reactor power leve), approximately 60%
(typical) because of the recirculation pump runback. Should the ex-aminer choose to restore operation of the tripped feedwater pump, the
. operators may increase power level to the initial condition.
Major Plant Systems: Feedwater, Reactor Recirculation, Control Rods, Feedwater Level Control Tolerance Range: In the case of a feedwater pump trip from full power, the best operator response is a hands-off approach. Should the opera-tors experience a reactor scram on low reactor water level, operator actions must ensure that the water level remains above the top of the active fuel at all times.
Competencies Tested:
SRO - Compliance /Use of Procedures Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operation B0P - Control Board Operation i
l 85 {
i,
REACTOR FEEDWATER P&P TRIP RFP Trip Progression of Operator Actions: Alarm (s) l Rn U Monitor reactor water level de-crease to the alarm setpoint Rn U j
Verify reactor I recirculation pumps run back to #2 speed limiter Rn
Verify reactor power decreased to within the capability of the remaining RFP(s) 1 Rn
Reduce reactor power with contro rods or recire, flow as necessary to remain within pump capabillity DOP U Place the control.-
1er for the trip-ped pump on MAN-UAL and zero the controller output 69 '
Start the idle RFP per the syster y IS operating proced, a standby ute urro avail-able?
" N BOP ERn U Adjust the control- 1 Request an aux-ler for the pump 111ary invest [-
placed in service , gate and repair and place in AUTO the cause of the level control pump trip 40 l' 1 As instructed by the SRO, begin E.ND power escalation
~
to the initial cower level O
86
~0perating Sequence: Loss of.Feedwater Heater Extraction Steam
- l
\ NSSS/ Type: GE/BWR
~
Initial Plant State: Power Operations
' Sequence-Initiator: Isolation of Feedwater Heater Extraction Steam Line Important Plant Parameters: 1) Feedwater Temperature, 2) Reactor' Thermal Power, 3) Feedwater Heater Shell Side Pressures, 4) Feedwater Heater-Levels, 5) Process Computer Thermal Hydraulic Parameters, 6)' Core Flow Progression of. Operator; Actions: See Flow Chart Final Plant State: The final plant state will depend on the contribution of the affected feedwater heater to the.overall feedwater. heating. Loss of the highest pressure heater will cause the greatest loss of effi-
'ciency, and, therefore, the greatest reactor power transient. Loss of the lowest pressure heater will have little_effect on the plant effi-ciency and reactor power.
Major Plant Systems: Feedwater, Feedwater Heater Extraction Steam and Drains, Nuclear Instrumentation, Recirculation, Control Rods, Process Computer Tolerance Range: Operation outside the power vs. flow operating map is.
not permitted. Ooeration outside the map indicates that one or more of I =the thermal hydraulic parameters is exceeding or approaching a technical specification limit. Should the process computer readouts verify that a thermal hydraulic parameter is being exceeded, operator actions must be initiated to return the affected parameter to within the Tech. Spec.
limit within the LCO time limit (typically two hours).
Competencies Tested:
SRO - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Technical Specifications ~
Compliance /Use of Procedures Supervisory Ability j Communications / Crew Interaction j i
RO - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation BOP - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response
' Control Board Operation Communications / Crew Interaction I
4 87
__ _ _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l l
I i
LOSS OF FEEDWATER HEATER Feedwater EXTRACTION STEAM Heater Troub Progression of Operator le ^'~'-(c.
Actions:
RO 9 Monitor feedwater temperature for decrease BOP U Check heater pres .
sures and levels to determine which heater has the arablem SRO U Monitor reactor thermal power and thermal hydraulic parameters sRO "
Inform the reacto engineer of the event R0 1r Determine if the reactor power vs.
core flow is within the oper-ating map R9 'r Adjust recirc.
pump speeds and N I8 control rods per power vs. flo procedure to re- within the map?
turn to the per-missible realon Y
SRO 9 Request auxiliary operator attempt to restore extr-action steam to the affected heater P9/ ROP If As directed by i procedure, close j heater drains, trip drain pump (s) ,
etc for the af f-erted heater i
Y I END l
l l
1 l
88 l
1 I
Operating Sequence: Stator Cooling Water Pump Trip
/ NSSS/ Type: GE/BWR Initial Plant State: Power Operations j l
Sequence Initiator: Inadvertent Trip of the Operating Generator Stator Cooling Water Pump '
Important Plant Parameters: 1) Generator Power, 2) Reactor Power Level,
- 3) Stator Cooling Water Conductivity, 4) Generator Load Set Setting i.
Progression of Operator Actions: See Flow Chart Final Plant State: If the initial plant state is one of relatively low generator output, operation may continue provided the stator cooling i water conductivity is less than the limits recommended by the generator )
l manufacturer. Higher initial power levels are difficult to cope with if I the stator cooling water cannot be restored. At high power levels the reactor may scram on turbine trip or high pressure following generator runback.
Major Plant Systems: Main Turbine-Generator, Stator Cooling Water, l Reactor Recirculation, Control Rods Tolerance Range: Operation of the main generator at high loads is not permitted without stator cooling water. Low load operation is not per-g mitted without stator cooling water if the cooling water conductivity l j is high. The generator manufacturer provides limits on generator opera-V tion for loss of stator cooling water.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding of Instrument / System Response Control Board Operation Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response l Compliance /Use of Procedures l- Control Board Operation Communications / Crew Interaction
[
t 89
STATOR CDDLING WATER PUMP TRIP Progression of Operator Actions:
-cc r > - r .,
89P
Verify stator cooling water pump tripped ROP D Start standby stator cooling water pump v
Did END
\, Y the standb f pump start 7
N ROP 1 Verify generator load set runback (if applicable) 90P
nn n Request the aux- Reduce reactor 111ary operator power level until check the stator all bypass valves cooling water close conductivity Is Did END Y the conduc. N the reacto
,j/ : tivity belo scram?
mits ?
N y nno 1' cAq/PN/U nnn Trip the main Perform scram turbine within actions the specified timc y limit v
END 90
l
,- s Operating Sequence: Steam Jet Air Ejector Malfunction e s
, i
'N ,/ NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Malfunction of the SJAE Steam Pressure Regulator or Pressure Controller Important Plant Parameters: 1) Main Condenser Vacuum, 2) SJAE Steam Pressure, 3) Seal Steam Pressure, 4) Reactor Power Level i
Progression of Operator Actions: See Flow Chart -
Final Plant State: The reactor will be operating at a lower power li vel j
I as the result of operator actions to slow the rate of pondenser vacuum decrease.
Major Plant Systems.: Off-Gas, Main Steau, Main Turbine-Generator, ,
Reactor Recirculation, Control Rods o
Tolerance Range: Operator actions should be quick enough on the slow condenser vacuum decrease to prevent a reactor scram on 104 condenser vacuum.
Competencies Tested: 4 RO - Control Board Operation (O)s_-
BOP - Understanding / Interpretation of Annunciator / Alarm Signals 1:
Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation ,
) >
ij I
b >
t 6
[
1 .
, ?
e d I f \
l __,
a,- ,
91
\
- - - - - - - - --_a
i I
l l
STEAM JET AIR EJECTOR MALFUNCTION Progression of Operator Actions:
Low Condenser Vacuum Alarm RO u 89P '
Reduce reactor Lineup the stand-
. power level to by steam jet air
, limit the Iate of ejectors condenser vacuum
,' decrease i! ' i
'4
s B9P P mno "
,, verify proper Place the stand '
I ,i operation of by steam jet air steam seals, circ. ejectors in ser-water rystem, vice
- vacuuc breakers, i,
etc.
/, JKE,.,,a " Ano P
-' /
3
/ Verify e i. aper Isolate the fault)
'/ valve, ; ira Jp of steam jet air I, the 0 fas systert ejectors t
/
>, / 5 '
.enP v se_[j_.x Ver t rf wr g , Veilfy condenser valve '(fxp o f vacuum is return.
f .
S 2 st/im supply Irvi co normal g adet,,at.tp /
, r rA,n pressu m n ' ,
t e regulatdt,, _j '
?an '_ ( ,,' '
o
, Check operat / n "
END I
of the ster /4 prefj. '
sure regulator '
.i and press e con-
) troller 4_ (,
/
4
.I'i s
/ ',
,n g
" I
-[ ,, i i i j
, / ,.
[ ,
/
1
!( /
k
(:
t .1 [ .
y ,
+
y 92
/?*
g, 1 ,
. _ - _ _ _ - _ _ _ _ _ _ _ . .t 1 <
l 5
I' 4s
.s . y
" Operating Segnhnte Loss of One Reactor Protection Systes Bus 3
.. i
' y t.
[^ NSSS/ Type:.GE/BWR -
3f 1
- t t
b Initial Pldnt State: Any Conditica Sequence Inltiator: Fa11'dre of the RPS Motor-Generator Set impor. tant Plant Parameters: 1). ,t:roup Solenoid Indicating Lights,
\')J Anf Plant Parameter Which Coubi Initiate a Reactor Scras l'rogression of Operator Actions: See flow Chart 9 ,.
- t. .
' final Plant State: Same as initial plant state.
[,( % l e
Electrical Distr!.bution, Reactor Protection System, 'f i
i)g t[ Major Plant Systems:Powee i6tnge Neutron Monitors, Process Radiation Monitors, C f
Ti 'l:sois t, ion System g
f
,e I ., . Toletance Range: None
<c.,q >
> ; Competencies Tested: ,
i Ti SRO - Communications / Crew Interaction
- , .r ,
y do' . Understanding / Interpretation of Annunciator / Alarm Signals
' Control Board Operation BOP - Control floard Operation
)
v
,u s %,
\
d I
+
A, I;
t h 1 U.O ,
, .5
' 93 i
.. 2w t u _ -
3
l LOSS OF ONE RPS BUS Progression of Operator Actions: Scram Alarms pq o verify no scram O
condition exist ROP /cRO1P Check or request aux. oper. Check breakers, fuses, M-G set, etc. to determine loss of power RDP/SRO 1r Align alternate power supply or Can request aux, oper N the M-G switch to alter- set provide nate power to the ower?
RPS bus ,
Y RO 'r When power is re-'
stored, reset the 7 tripped RPS bus BOP 4 Reset rad. monitoI alarms, APRM alarns, isolation signals, etc.
1r END O
94
Operating Sequence: Area Radiation Monitoring System Alarm
/'~1 k) v NSSS/ Type: GE/BWR Initial Plant State: Any Sequence Initiator: High Area Radiation in the Reactor Building Important Plant Parameters: 1) Reactor Building Area Radiation Levels,
- 2) Reactor Building Ventilation Exhaust Radiation Levels Progression of Operator Actions: See Flow Chart Final Plant State: Same as the initial. plant state.
Major Plant Systems: Area Radiation Monitoring, Ventilation Process Radiation Monitoring System, Standby Gas Treatment System Tolerance Range: All personnel should be evacuated from the reactor building when it becomes apparent that the limits of 10CFR20 will be exceeded (" Protection of Personnel from Raddatien"), The operators should ensure that the standby gas treatment system automatically starts (or manually starts) when radiation levels in the ventilation exhaust reach the setpoint (variable between plants).
Competencies Tested:
[ )
\s_,/
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Compliance /Use of Procedures BOP - Understanding / Interpretation of Annunciator / Alarm Signals Compliance /Use of Procedures
(
\s. )
95
_- ___- -- A
I AREA RADIATIDN MDNITCRING SYSTEM ALARM Progression of Operator Actions: )
High Rad.
O Monitor Alar nnp " BOP U Scan Area Rad. Start or Verify Monitor Panels Auto Stert of the to Determine Standby Cas Treat-which Monitor ment System if i
has the Alarm Vent. Rad. Levels P- _
nnF o can U Read Radiation Direct Health Level on the Physics to Deter-Alarming Monitor mine Radiation and Contamination Levels and to Becin Envir, Mon-itoring P '
9tn can I Evacuate the Refer to Admin.
Affected Area of Procedures for the Reactor Classification of
. Building the Event Rn9/Rn U can
Check Indications Notify Plant for Possible Management of the Causes of the Event and Actions High Radiation Taken Bop t can U Monitor ventil- Notify the NRC ation Exhaust of the Event and Process Radiatior Actions Taken Monitors for Increasing Rad.
Levels l
EPO O
96 l
l
l Operating S: quince: High Main Steaa Line Radiction NSSS/ Type: GE/BWR A
( ) Initial Plant State: Power Operations N,/ s i
Sequence Initiator: Gradual Fuel Failure Important Plant Parameters: 1) Main Steam Line Radiation, 2) Off-Gas Radiation, 3) Reactor Power Level l Progression of Operator Actions: See Flow Chart Final Plant State: If the' iodine concentration is less than the maximum limit permitted by technical specifications, operation may continue at a f substantially lower power level, provided that the main steam line radiation increase has ceased. If the technical specification limit on iodine is exceeded, the reactor will be in hot shutdown with the main steam isolation valves closed.
Kajor Plant Systems: Main Steam, Off-Gas, Reactor Recirculation, Control Rods, Balance of Plant Systems, Process Radiation Monitoring Tolerance Range: When it is determined that the main steam line radia-tion is increasing, the operators should immediately commence a reactor power reduction. Shod 1d main steam line radiation or off-gas radiation approach the isolation setpoints, the reactor should be shut down immediately. The technical specifications also impose restrictions based on the results of laboratory analysis of the reactor coolant. If the equilibrium value for dose equivalent I-131 is exceeded, reactor
[/}
g operation may continue for no more than the LCO time limit (typically 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) provided that the increased activity is due to a power transient.
Should the dose equivalent I-131 concentration exceed the maximum allowed by technical specifications, an orderly reactor shutdown should be initiated and the main steam isolation valves closed. Refer to the plant specific technical specifications for more details on these restrictions.
Competencies Tested:
SRO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Technical Specifications Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Understanding of Instrument / System Response 7-6 Compliance /Use of Procedures
\-. Control Board Operation Communications / Crew Interaction 97
HICH . MAIN STEAM LINE RADIATION Progression of Operator Actions:
High Main O
Steam Line Hi Rad, 01 arm RO/DOP 1' SRO Y Check radiation Request chemistry monitor trend re- laboratory calcu-corder to deter- late the off-gas mine the rate of release rates at MSL radiation in- frPquent intervals crease Bon u SR9 o Monitor off-gas Intorm nuclear radiation for engineer of the Increasing trend fuel failure RO e SRO P Recuce reactor Refer to Tech.
power until MSL Specs for limits radiation begins on coolant chem-to decrease istry 11mits on dose equir. lent I-131 BOP P O SRO/PD/RP j Monitor main stear :
33 Commence an order-33 and off-gas rad, oclant chem. ym or shu b monitors during N 1 131 with. N within 1.5. n max, allow- "O control rod inser" imits? able?
" " "# * "O tion to determine 40TE 1 procedures Ilocation er lenk_
Ing fuel assembly y Y SRO l' p S90 P P Heguest chemistry Refer to Tech, laboratory sample Et0 Soecs, for LCO END reactor coolant on operation ab-and off-gas ove the equillb-rium dose equival-ent I-131 value P
EPO NOTE 1: The Technical Specifications will give two vaIues for dose equivalent 1-131 that the operators should be concerned with during this event. The first value con-cerns the equilibrium (steady state) limit (f or example, 3.2 uC1/gm). The second, and higher, limit on dose equivaient I-131 1s the maximum allowed following power transients. Under no circumstances should reactor oper-ation continue if this higher value is exceeded.
O 98 1
. _ _ _ _ _ - _ _ _ _ - _ _ _ _ i
Operating Sequence: High Ventilation Exhaust Radiation NSSS/ Type: GE/BWR Initial Plant State: Any Normal, Abnormal, or Emergency Condition Sequence Initiator: Inadvertent Release of Gaseous and/or Volatile Radioactive Material into the Reactor Building or Refuel Floor Important Plant Parameters: 1) Ventilation Exhaust Radiation Levels,
- 2) Area Radiation Monitor Levels Progression of Operator Actions: See Flow Chart Final Plant State: No significant change from the initial plant state.
Major Plant Systems: Reactor Building and Refuel Floor Ventilation, Standby Gas Treatment System, Primary Containment Isolation System, Process Radiation Monitoring, Area Radiation Monitoring Tolerance Range: When it is evident that a release of radioactive material has occurred, the plant management and the appropriate off-site emergency response agencies must be notified immediately.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Compliance /Use of Technical Specifications
"* Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Diagnosis of Events / Conditions Based on Signals / Readings Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Compliance /Use of Procedures Communications / Crew Interaction 99
HICH VENTILATION EXHAUST RADIATION Progression of Operator Actions:
High Radiatio Alarm Roo P nop 1-Check the radia- Close or verify tion levels on automatic closure the ventilation f the normal radiation monitors ventilation damp-ers SRO y 80p y Start or verify Announce over the plant PA the pre- automatic start of sence of airborne the standby gas radioactivity in treatment system the reactor bldg.
nr vo r, iv ri nne son 1 nno 9 Request health Verify that all physics survey the other containment affected area (s) isolation that to determine the should occur have airborne activity occured levels SRO 1
' ROP /R0 P Classify the event Monitor plant par- j and notify the ameters and area ,
I appropriate on- radiation monitors site and off-site in an attempt to personnel per the locat3 the source emergency plan and location of the leakage nnp V SRO l' Trip or verify itefer to the env-automatic trip of iromental Tech. '
the normal ventil- Specs. for limits ation supply and an radioactive exhaust fans release from the ,
alant stack l P
END O
100
1 Operating Sequence: Inadvertent HPCI or RCIC Initiation NSSS/ Type: GE/BWR I
Initial Plant State: Power Operations l
Sequence Initiator: Failure of Initiation Logic on HPCI or RCIC Important Plant Parameters: 1) Reactor Vessel Water Level, 2) Reactor Vessel Pressure, 3) Reactor Power Level, 4) Drywell Pressure, 5) HPCI j Turbine Speed, 6) RCIC Turbine Speed I Progression of Operator Actions: See Flow Chart Final Plant State: No significant change from the initial plant state.
Major Plant Systems: High Pressure Coolant Injection, Reactor Core Isolation Cooling, Neutron Monitoring, Feedwater, Main Steam Tolerance Range: Suppression pool cooling should be placed in service if the suppression pool water temperature reaches 95 degrees F.
Competencies Tested:
SRO - Compliance /Use of Technical Specifications RO - Diagnosis of Events / Conditions Based on Signals / Readings e ' Understanding of Instrument / System Response
!.('
BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Control Board Operation O
101 1
I
INADVERTENT HPCI OR RCIC INITIATION Progression of Operator Actions:
Actuation Alarm (s) 9l '
nn 4 Verify reactor water level and drywell pressure (for HPCI) are normal BSP V Attempt to reset the initiation signal Anp gop '
Depress and Place the control-hold the turbine HPCI ler in MANUAL and trip pushbutton HPCI or RCIC reduce setpoint RCIC tape to zero 7
RnP pnp Rnp 1' At zero speed, Attempt to det- Depress the tur-place auxiliary ermine the cause bine trip push-oil pump in ___ , of the failure <____ button and verify PULL TO LOCK the turbine stop and release trip valve closes oushbutton SPO "
Request an inst-rument technician repair the cause of the failure nno u Assess the need for suppression pool cooling and place in service if required nn v Verify reactor water level, pre.
ssure, and power level return to normal I cpn U Refer to Tech.
Specs. for inop-erable ECCS END 102 l
~ Operating' Sequence: Loss of One RBCCW Purp l
p NSSS/ Type: GE/BWR
( ' Initial Plant State: Power Operations 'f Sequence Initiator: ' Inadvertent Trip of One Reactor Building Closed 1
. Cooling Water' System Pump
,Important Plant Parameters: 1) RBCCW System Pressure, 2) RBCCW Tempera-ture, 3) Reactor Water Cleanup Non-Regenerative Heat Exchanger Outlet Temperature, 4) Fuel Pool Temperature, 5) Primary ContainmentTempera-ture Progression of Operator Actions: See Flon Chart Final Plant State: No significant difference in the final plant state from the initial plant state.
Major Plant Systems: RBCCW, RWCU, Fuel Pool Cooling and Cleanup, Primary Containment Tolerance Range: The technical specification Limit on the temperature of the water _in the spent fuel storage fuel should not be exceeded during this event.
Competencies Tested:
._ ,/] BOP - Understanding / Interpretation of Annunciator / Alarm Signals t Control ^ Board Operation Q
103
LOSS OF DNE RBCCW PUMP i' Progression of Operator Actions:
Low RBCCW pressure alar
. BOP o verify loss of the running pump l
nno u Start the standby ,
Reset RWCU isol- 01 ation and return y RWCU iso the RWCU system ate?
to operation N
ROP 1P Monitor drywell j temperature and
- fuel pool temper-ature 1r END l
104
1 i
)
l BWR EMERGENCY EVENT DESCRIPTIONS l
l 105
7 t \
( / BOILING-WATER REACTOR EMERGENCY EVENTS
%J Reactor Scram With MSIVs Open Reactor Scram With MSIVs Closed Loss of Shutdown Cooling Gross Fuel Failure Excessive Reactor Cooldown Rate Anticipated Transient Without Scram Stuck Open Main Steam Safety / Relief Valve Small Break Loss of Coolant Accident Reactor Coolant Leakage Outside Primary Containment Jet Pump Failure High Suppression Pool Water Temperature
,~
Main Turbine or Generator Trip
( j)'
\
Main Turbine or Generator Trip Without Bypass Valves Loss of Condenser Circulating Water l
Loss of Feedwater System Loss of All High Pressure Feedwater Loss of Plant Control / Instrument Air EHC Pressure Regulator Failure (All Valves Open)
Loss of Nuclear Service Water Loss of Reactor Building Closed Cooling Water System I Loss of Off-Site Power Loss of All AC Power (Station Blackout) ls
- v 1
107 J
Operating Sequence: Reactor Scram With MSlvs Open NSSS/ Type: GE/BWR Initial Plant State: Any STARTUP or RUN Mode Operation Sequence Initiator: Any failure that will deenergize the Reactor Protection System trip channels or any condition that will require the operators to manually scram the rea.ctor as specified in Tech.
Specs, and/or procedures.
Important Plant Parameters: 1) Reactor Water Level, 2) Neutron Monitoring Subsystems (APRM, 1RM, SRM), s) Reactor Vessel Pressure,
- 4) Recirculation Pump Speed 5) Feedwater Flow to the Reactor,
- 6) Control Rod Positions, 7) Main Generator Output, 8) Main Turbine Speed, 9) Condenser Vacuum, 10) Reactor Vessel Pressure Converted to Saturation Temperature Progression of Operator Actions: See Flow Chart Final Plant State: Hot Shutdown Major Plant Systems: Nuclear Instrumentation, Control Rod Drive Hydrau-lics, Reactor Protection System, Condensate and Feedwater Systems, Reactor Recirculation System, Main Steam, Main Turbine / Generator, Elec-trical Distribution Tolerance Range: During any reactor trip, the operators should ensure that reactor water level remains above the active fuel, the reactor vessel pressure is maintained below the Tech. Spec. safety limit, and the reactor vessel cooldown rate does not exceed the Tech. Spec. limit.
Other tolerances apply as the situation may warrant.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction 109
~ - - - - - - _ _ ~ _ _ - - . _ _ _ _ _ _ - _ _ _ - -
REACTOR SCRAM WITH MSIV's CPEN l Progression of Operator Actions: l Page 1 of 2 Reactor Scram Alares 01 enn n 3
Enter RPV Control 1 Procedure )
i i
i i
l 4
SRO/Rf / k BOP enn/on/1r nno SPD/Rn/ h AnP Execute " Monitor Execute " Monitor Execute " Monitor and Control RPV and Control RPV and Control Reac-Water Level" Pressure" sub- tar Power" sub-subprocedure procedure p!oCedure spo U ROP P RO verify no isolat- Verify turbine Place the reactor ion or ECCS init- bypass valves m de switch in lation conditions are controlling SHUTDOWN exist reactor pressure RO RO Verify all control verify the reactor Exit to Scra recirculadon rods are inserted Procedure beyond the MSBWP pumps runback to minimum speed and core flow Exit to Scra Exit to Scra Procedure Procedure O
110
REACTOR SCRAM WITH MSIV's OPEN'
[D~ Progression of Operator Actions:
( Page 2 of 2 Enter from EPG's nn II nno 'I RO/ BOP 1' verify that react- Verify station- Return recirc.
or power decreases electrical loads system, RWCU, and to less than 3% transfer to the ventilation sys '
of rated off-site power tems to service supplies if tripped BOP' if BOP 1' SRO 1!
After level returns Start or verify Evaluate the 1 ne d- auto start of the need for reactor turbine lift pumps vessel cooldown water m motor suction pump and the turning i
- ^!1 ~ ~'
RO' 'r SRO 'f mn l' Insert SRM and IRM Determine the Log the events detectors and. cause of the reac. and actions taken switch recorders tor scram to monitor IRM's RO Ir RO/ BOP 1l SRO I' Downrange IRM's. Notify supervisor to maintain an Remove unnecessar) on-scale reading eQulpment from service BOP D BOP 1f Mn If Trip or verify Verify MSIV's Complete scram auto trip of the are open and con- report main turbine / gen- denser vacuum is erator normal ggp p ^RO ' '
verify generator areakers open When the scram END signals are clear, bypass the SDV high level scram and reset the tea-n n, e-r= m
'I 1
1 1
I
~
l 111
rx Operating Sequence: Reactor Scram With MSIVs Closed NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Closure of the MSIVs with the Reactor Mode Switch in RUN Important Plant Parameters: 1) Reactor Water Level, 2) Reactor Vessel Pressure, 3) Reactor Power Level, 4) RCIC System Flow, 5) HPCI System Flow, 6) SRV Positions, 7) Suppression Pool Temperature Progression of Operator Actions: See Flow Chart Final Plant State: Hot shutdown. If the cause of MSIY closure is not fuel failure or loss of coolant accident, the MSIVs may be reopened.
Major Plant Systems: Main Steam, Neutron Monitoring, Reactor Recircula-tion, Control Rods, HPCI, RCIC, Condensate and Feedwater, RHR (Suppres-sion Pool Cooling Mode)
Tolerance Range: Watea level must be maintained above TAF at all times.
Due to SRV, RCIC, and HPCI operation, suppression pool temperature will increase. Suppression pool cooling should be placed in service as soon as possible to maintain pool temperature below 95 degrees F.
,e w v
) Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures I Control Board Operation j Communications / Crew Interaction i BOP - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction i
G
\
l 113 l - - - - - - - - - - - . - - _ - - - - - - -
REACTOR SCRAM WITH MSIV's CLOSED Progression of Operator Actions:
Page 1 of 2 Scram Signal (s ,
MSIV's Close O
SPn o Enter RPV Control Procedure nnn
SPO/Rn/hRnP 9Pn/Rn/Anp %RO /Rn / nnD Start or verify Execute " Monitor Execute " Monitor Execute " Monitor automatic start and Control RPV and Control RPV and Control Reac-of RCIC Water Level" sub- Pressure" sub- tor Power" sub-procedure procedure procedure 90P if SDO V DDP U Rn U Start or verify Verify MSIV's Initiate isolat- Place reactor automatic start closed ion condenser or mode switch in of HPCI suppression pool SHUTDOWN cooling onP if ERn if RnP l' Pn if Restore reactor Verify all other Manually open Verify reactor water level to isolations occurr- SRV's in presc ib.. recirculation the normal oper- ed that should ed sequence to pumps runback to ating band using have occurred reduce reactor minimum speed HPCI and RCIC pressure to (935 and core flow psio}
ir pn o nnP v "
Verify all control Augment pressure Exit to Scram Exit to Scram control with Procedure rods are inserted procedure beyond the MSBWP HPCI and RCIC turbine operation v
Exit to Scram Procedure O
114
l l
REACTOR SCRAM WITH MSIV's CLOSED
.s Progression of Operator Actions
} Page 2 of 2 1
%)
Enter from EPG's AnP U Rn 31no Pn Place MSIV control Downrange 1RM's Reset the 1501- When all scram switches in the to maintain an ation logic signals cleared, CLOSE position on-scale reading bypass the SDV high level scram and reset the re-actor scram RO " BOP Rnp U
P Pn/AnP Verify all control Trip or verify open the int,oard Return recirc.
rods fully insert- auto trip of the MSIV's system, RWCU, and ed and reactor main turbine and ventilation sys-power decreased tc transfer of elec. tems to service less than 3% loads to off-site if tripped power RnD Rnp F Anp " U SRO Control injection Start or verify Pressurize the Evaluate the need from RCIC, HPCI, auto start of the main steam lines for reactor vessel l and feedwater (r4- tubine lift pumps and open the out- cooldown tot driven pumps) motor suction pump, board MSIV's to prevent high and turning gear level turbine tr1E oil numo 1
[\ j pn y c;pn H Rnp 9 cPn U V / Insert SRM and Determine the Place seal steam the v IRM detectors cause of the in service Lnklonsfakntand ac en scram pn P ' Rnp 4 j SRO U j Switch recorders Can
, Establish main Notify the super-j from APRM and RBM condenser vacuum to record IRM the MSIV visor j .colation be I with the mechan-l set? ical vacuum pump or SJAE's l sop U SRO D l
Return the feed- Complete scram water system to report service (turbine driven pumps) and secure RCIC and HPCI
! END 73 115
.,-~s Operating Sequence: Loss of Shutdown Cooling
/
k NSSS/ Type: GE/BWR Initial Plant State: Hot Shutdown with Reactor Vessel Pressure Less Than 100'psig Sequence. Initiator: Trip or failure to start of the RHR pumps aligned for shutdown-cooling, or closure or-failure to open of the shutdown cooling suction isolation valves Important Plant Parameters: 1) Reactor Coolant Temperature, 2) Reactor Vessel Pressure, 3) Reactor Water Cleanup System Total Flow, 4) Reactor Water Cleanup System Reject Flow, 5) Main Condenser Vacuum Progresolon of Operator Actions: See Flow Chart.
Final Plant State: The plant will be in cold shutdown with either the alternate RHR loop providing shutdown cooling or feed and bleed operation, Major Plant Systems: Residual Heat Removel, Reactor Water Cleanup, Condensate, Main Condenser Tolerance Range: Depending on the situation, the operators may violate technical specifications if the reactor is not placed in cold shutdown (reactor shutdown was performed to meet a technical specification LCO),
itf'{} ' Otherwise, the most limiting situation is one where the vessel head-is
j -
removed. In this condition, the coolant temperature must be maintained below a plant specific limit (typically 150 degrees F).
l Competencies Tested:
SRO - Compliance /Use of. Technical Specifications (possible)
Compliance /Use of Procedures Supervisory Ability RO - Compliance /Use of Procedures Control Board Operation l
BOP - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation I
l I 117
LOSS OF SHUTDOWN COOLING Progression of Operator Actions:
RHR Pump (s)
O Trip Rn/ Ann ir Verify reacter water level and drywell pressure are normal
.nnp 'r Realign RHR pump Is suction from the y the other suppression pool 4 F!HR loop to S/D cooling etable?
l nnp v nntnno o Establish RHA If directed by service water flou procedure, raise through the heat reactor water exchangers level to increase natural circula-tion nnp 9 BOP o i Start the RHR Place the main '
pump (s) and es- condenser in tablish flow to service the reactor SRn u nnp 'r Notify supervisor Establish reject flow through the 4 reactor water cleanup system to the main con-denser spn 'r nnp u Log events and Increase CRD flow actions takon to the reactor or use condensate system to maintain reactor water leve!
END O
118 1
q
\
I I
Operating Sequence: Gross Fuel Failure j 4
4 %
\d NSSS/ Type: GE/BWR Initial Plant State: Power Operations j Sequence Initiator: Gradual Increase in Main Steam Line Radiation to 3 Times the Normal Background Important Plant Parameters: 1) Main Steam Line Radiation, 2) Off-Gas Radiation, 3) Reactor Power Level, 4) Reactor Water Level, 5) Suppres-sion Pool Temperature, 6) HPCI or HPCS Flow Rate, 7) RCIC Flow Rate Progression of Operator Actions: See Flow Chart l i
Final Plant State: How shutdown with MSIVs closed, suppression pool j cooling in service, and reactor water cleanup system in service.
Major Plant Systems: Main Steam Process Radiation Monitoring, Primary Containment Isolation System, Residual Heat Removal (Suppression Pool Cooling Mode), High Pressure Coolant Injection or High Pressure Core l Spray, Reactor Core Isolation Cooling Tolerance Range: Following main steam isolation on high radiation, under i no circumstances should nuclear steam be discharged to the main conden-ser. The reactor water cleanup system should be returned to service, if e tripped, as soon as possible following fuel failure. Caution should be l(
I used in resetting the reactor scram, since highly radioactive material may be transported outside secondary containment when the scram is
- l reset.
Competencies Tested:
I SRO - Understanding / Interpretation of~ Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability ;
Communications / Crew Interaction RO - Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation )
Communications / Crew Interaction _
B0P - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation l
Communications / Crew Interaction i
4 i
l 119 !
JROSS FUEL FAILURE Progression of Operator Actions:
Page 1 of 2 i High Main Steam Line ua%
RnP r sRn SRn RO W Verity main stearr Request rad. chem, N line radiation lab sample reac-Notify plant'per-sonnel of pos-Reduce I actor poweruskngre- /\Has alarm and monitor w tor coolant and y sible high area circ. pumps and trend g Ieactor off-gas radiation control rods cram occurr
?
Y SRO d Enter RPV Control Procedure nno E SRO/Rn E AnP SRO/RO - Pnp SRO/RO d rop Start or verify Execute " Monitor Execute " Monitor Execute " Monitor eutomatic start and Control.RPV and Control RPV and Control reac-of RCIC Water Level" sub- Pressure" sub- tor Power" sub-procedure procedure procedure nnp ,, spn u nnp Pn ir Start or verify verify MSVI's Initiate iso- Place reactor automatic start closed or man- lation condenser mode switch in of HPCI ually close suppression pool SHUTDOWN cooling nnp ,, 99n er Anp u RO ir Restore reactor Verify all other Manually open Verify reacto:
water level to isolations occur. SRV's in pre- recirculation the normal ope- red that should scribed sequence pumps runback rating band usinq have occurred to reduce reac- to minimum or HPCI and RCIC tor pressure to tripped
[935 psig)
,, RO u nno w "
Verify all con- _ Augment pressure Exit to Scram trol rods inser- Exit to Scram Procedure tad beyond the control with Procedure HPCI and RCIC MSBWP turbine operatior u
Exit to Scram rucedure 9
120
i l
l l 1 GROSS FUEL FAILURE
'7N) Progression of Operator Actions:
127 ij Page 2 of 2 Enter from EPG's U '
AnP ROP f Place MSIV cont- Start or verify rol switches in auto start of the the CLOSE positior i turbine lift pumps ,
motor suction pump ,
and turning gear oil pump i Rn t nnp/pn 1runir Verify all contro.. Return the reactor rods fully insert -
water cleanup sys" ed and reactor tem to service NOTE: The reactor operator should power decreased g
if tripped to less than 3% -severe fuel failure preset until health physics and/or rad. chem, verify the contamination of the BOP if RO/ ROP U scram discharge volume water is below SCceptable radiation levels.
Control injection Commence reactor from RCIC, HPCI, cooldown using and feedwater SRV's and place (motor driven the shutdown cool-pumps) to prevent ing system in ser-high level tr1D vice when 41000sia r%
nn v n
( \ SRO
(' Insert SRM and Log events and IRM detectors actions taken RO " SRO U Switch recorders Notify supervisor from APRM and RBM to record IRM pn 1r enn e
' Complete the scram Downrange IRM's to maintain an report on-scale reading Rnp c,Pn 1I Trip or verify I Implement the auto trip of the radiological main turbine and emergency plan transfer of elec- as specified by trical loads to administrative off-site power procedures
,7 u 121
Operating Sequence: Excessive Reactor Cooldown Rate 7
(%j ) NSSS/ Type: GE/BWR Initial Plant State: Hot Shutdown Sequence Initiator: One turbine bypass valve opens to control reactor pressure following a reactor scram and falls to close as the reactor pressure decreases. Operators should discover the problem during performance of scram followup actions.
Important Plant Parameters: 1) Reactor Vessel Pressure, 2) Reactor Water Level Progression of Operator Actions: See Flow Chart Final Plant State: Cold Shutdown Major Plant Systems: Main Steam, RCIC, RHR (Shutdown Cooling Mode)
Tolerance Range: Operators should take corrective action as soon as it becomes evident that the cooldown rate LCO is being exceeded. Once corrective action is initiated, the cooldown rate should not exceed the LCO.
Competencies Tested:
/N SRO - Diagnosis of Events / Conditions Based on Signals / Readings
( ,)
Understanding of Instrument / System Response Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability RO - Understanding of Instrument / System Response l Compliance /Use of Technical Specifications Compliance /Use of Procedures BOP - Diagnosis of Svents/ Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Technical Specifications Compliance /Use of Procedures Control Board Operation l
l l
f~)
N__-)
123
- 7. .n
- i t*
EXCESSIVE 'NEACTOR LOOLDOWN RATE '/ f Progression of Operator Actions:
'q )
Cocidown Rate / e' A>,e LCO
\,i t,
-)
1t i:
.),
1 nP
- fi j , (<'
Close the Main ,
Steat Isrlation ip c
valves ,
. g. J
/ t' y' y, / .d
).
!?O .,
C Enter RPV Control i.
Procedure I
\
spn/pn f ann ~'~ e.pn/nn i nnn Execute " Monitor 1 Execute "Moriitor j and Control RPV and Control RDV Level"supro- 5 Pressure" sub- l cedure '
procedur7, l \ r 't ' ), l i -
ne . nm V - } 1, Place RCIC in l Place is,f ation 3- ;
service to con- conde iser or sup-trol reactor pression pool water level , ,
.g iing in , ( ~, l'
', t i<1ce
/ . t. t,
/ I ,
nm w ,I Return to Manyally open -
Scram Procedur W. .
I'lli ' PV
. n /s y ge drops
' . t [:1i'ssigj I \
y w Establish cool.
Uown rate at the LCO ,
i (L
( -l pn/enn .
l d Monitor crolcev3 5 rate ^
(,
I (.
/ . >
]:. 1
'JI / l
_pan . / 'i hhen shutdown cooling inter- /. ,j locks are clearec '
, / ,
i place shutdown l"
[
cooling in ser-vire ; 1 i ,
Heturn to j cram Procedu"
[.
124 1
t A w
_ _ _________ _ 7 I *
'f
\ v'I ',
f r<t ';
' ,l. ( l .
e
,/ /
vw s ;
,/
l' Operating SequeAme: f , Anticipated Tra,b.ienc M*thout Scram ~
I NSSS/ Type: GE/I p
" i,
[ Ill'tial Plant 4 ttr/. Poes' 00erations ! '
y' a t [' t, .
Sh pencej InJ/,E[or: Fail1tre of Nontrol Roch to F.'lly In3?rt ,on &ay Scr'am i SirJsl NcLust of Hydraulic Lock in the R: ram Div.harge Volume, , , .
1,ntetant Plant Paramet'ers: 1) Reactor Power /sevel, 2) Vteact)r Water ; i 1 N 1, 3) Su[pession Pool ' temperature, 4)I Pr ryy CotMith.ent Pressure,
.b ;,RI orq18 gt System Flow,,i) RCIC Sys't.?m Flow, 7) W R System Flow, ,
hJ FOdwu h System Flow, 9) Stn9dtn Liquid Coc.tro.). Tahk Level f g
, ,,, (; '
k A P6tcresdsw.dlgeratorActiony l[ Flor Chart St.e' f
t" ,
J. y '
, t ' . ,
/
t l
'} %naiAMt) /; State:\'lotSh.cdow (/
1 !
(
. Major P' ant (Sys r as: ; Control Rods, Staatby dquid Control, HPCI or hPCS, ,(
i RCIC, RHR (hppens1<k Pool Cooling) 4 Primary contrinsent Isolati'un , I
< Sutten, Neuttr$ Nun 1(aring, Reactor Recircuit.ti d, Main Steam s i
' ,' j,
~
- ;/
( ,,,
[hleraince Range! lear, tor vessel pressurn 1tust .br ariintlained belowyn
/.
[', sn()ty limit at all fluest Reactor water level n.ust be. siaintained agtve 5
' f' 1,). ..
/hq top of the actim? fuel,unlesstheprocedaredirectsthe.operatsffs
'to hwer water levej! fvlow the tcp of ehti.v.e fuel for, short periods of
~ .
/^ ' ;' time. Containment'prepsure and si;ppression pool load should ba
' nintained 'ralow the Wtilure If.aits.
, ,o e j Competerklas/kested: l' l f
', 6 ,.
Sko - Underttand."ng/Init/t trMation of Annuncia'.orWarm signals .
Di /dn.)s\d}N Events /Conditio,f DasV2 on Signal <e/Revlings s
Compliarre)/Jit of' Tee:mical Specif nations '
Complio$c'tOJjeofProcedures ,.
Superv2rary) Ability 7 Commuriz.:ations/ Crew Interaction r
i RO -Undersffnding/InterprMd.$onofAnnunt,lator/AlarmSignals t g} ,, . Diag /dels of Events /CJnditions Based on Signals / Readings 7
/' '
Unds/ stand 1ntr of In m ukant/ System Response
/ i Complianctobb of Procedures
/ ,
C)ntrol Pea'rd Operation ' ,
Camunications/ Crew Interaction e BOP 7 ndarstanding/
U Interpretation % Annunciator / Alarm Sign ds 1
Understanding of Instrument /Sy.etem Response
' , r; Compl bnce/Use of Procedures, Control Baard Operation / ,
Communications / Crew Interactico e
j I
'k f t
- ';?5 l ,
l' ,
- _ - _ - _ _ - _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ -__l_____.__ _ _
- Ij q
[
,c,-.
!$- g <
NOTE: Actions der:Dted cith "SRD" indicate that the SRD is responslD1e for directing the aux 111ery operator to perform this action.
Any Scram
'(Signal Cater JFV Control and Contairvnent Control pro-cedures
!t E
can L
' Verify all iso- nn s
-( lations occured Place reactor mode switch in 3 - i that should have SHUTDOWN occurred
)>
a ;j _
)
cAn n nno ,,
pn ir verify all ECCS Terminate all in- Trip the reactor
[ and EDG hitia- jection except tions occurred CRD And SLC recirculation pumps that should have occurred RnD tr ,f0P tr nn l' Initiate IC or Cpen ADS SRV's Before reacing suppression pool and depressurize B1IT, initiate cooling the reactor y SLC and prevent Approaching ADS Initiation BIIT?
N
_ nno o_ non ,, u Rn
'Open SRV's in Slnwly f v rease Verify the Reac-prescr1Md Jnjecfken to re. L tor Water Clean-
%gerce mtil ttam RPV water up system isolate:
RPV pressure is leiel to above less than [935 W osici
'\
nno v -
e,Rn+ u When BIIT is ex. Has Remove fuses that ceeded lower */ HS0W supply power to water level below injected : the scram sole-FSWL by termi- Ff 4
? noids nating all injec- / Restore reactor egn water level to the normal ope- l nnr> ir _ rating band eng, ,1 j
Inject es neces- '
Close the scram sw/ to hcM air header supply water level below B1 : I valve and open i
FSWL and above TAF the scram air hender vent X 1 cA }
maintaloof 1evel ts N h bove T4 / p ) 'l C1 I< ?r s
Y O'
} 126
ANTICIPATED TRANSIENT HITHm T SCRAM Progression of Operator ACU)ons:
Page 2 of 3
\
\ /
w/ B1 J
Has CSBW v injected
?
N nnP o 1P Maintain suppres- Depressurite the sion pool temp. RPV at the cool-or RPV pressure (---- down rate LCO below HCTL RSP 1r Maintain suppres-sion pool level or RPV pressure ,
l below SPLL gr pqp Anp er Terminate all in- Operate contain- 1
'~~'N Can.
ot maintai jection into RPV ment pressure l ) from sources ex- control and/or
('""'f e L below 11- SCT as needed to mit terminate contain.
7 ment except CRD maintain D.W. pr-and SLC ess, below [2 Osig]
N =
P nnp ir Can- Prevent injection Has Y from LPCS and CSBW N not mainta injected TL below 11 LPCI pumps not re-quired for ade- ?
mit
? quate core coolina Y N
ir ,,
nno .,
Open ADS SRV's Exit to Scram Ifdrywelltgmp. and depressurite exceeds (135 F) Procedure place all ava11- the reactor able drywell coolers in ser-vice u
I i
l ,
l4
~'s, i
)
s
~./
127 l l
ANTICIPATED TRANSIENT WITHOUT SCRAM Progression of Operator Actions:
Page 2 of 3 C1 C3 l
0 cnn* r Po ir nn --
When control rods Reset the reactor Start all CRD stop inward move. scram hydraulic pumps ment, replace fuses, close vent ,
and open the air tunnly on ., a ran* .
Reset the reactor i Close the HCU scram the scram scram accumulator
- C2 reset -
charging water i
? header Supply valve
, Y
, pn ir enne o verify the scram Can Defeat RSCS Any discharge volume interlocks scram be N C3 vent and drain ods nct in- y reset valves open etted beyond
? he MSB
?
Y Y MRO* ,e SRO* o RO ., cAn* ,,
l l Drain the SDV and for any control Insert contorl For rods not in-open the HCU rod not inserted rods not insertet setted beyond the charging water past the MSDWP past the MSBWP MSBWP, open with-header supply open the scrsm using teactor draw vent line to valve test switch manual control radmaste drain 90 gr Y ton
- gr r corie er Manually scram When the control Open the withdrsw the reactor Did Can control i iud "O IUD 9"I the scram be vent valve until ods move in- reset inward movement ward moves close theinward'am scr 7 stops
, test switch All C2 s I 05 E **
Setted beyond M$DWP 7
Y nn ,
Stop boron in.
Has jection by trip-CSBW ping the SLC injected F--
pump
?
, e Exit to Scram Exit to Scram Procedure Procedure 128
l
,- s Operating Sequence: Stuck Open Main Steam Safety / Relief Valve
! ) 1
\m _/ NSSS/ Type: GE/BWR 1
Initial Plant State: Power Operations )
Sequence Initiator: One Main Steam Safety / Relief Valve Fails Open and
~
Cannot Be Reclosed Important Plant Parameters: 1) Suppression Pool Temperature, 2) Reactor .
Vessel Pressure, 3) Main Steam Line Flow, 4) Feedwater Flow, 5) Reactor ]
Water Level, 6) Reactor Power Level, 7) Reactor Coolant Temperature Progression of Operator Actions: See Flow Chart Final Plant State: Cold Shutdown Major Plant Systems: Main Steam, Residual Heat Removal, Condensate and Feedwater Tolerance Range: The reactor must be shutdown prior to exceeding the boron injection initiation temperature. Suppression pool temperature or reactor vessel pressure must be maintained below the heat capacity temperature limit at all times.
Competencies Tested:
p_
( ) SRO - Compliance /Use of Technical Specifications
\- ' Compliance /Use of Procedures Supervisory Ability RO - Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction
(
L 129 I
J STUCK OPEN MAIN STEAM SAFETY / RELIEF VALVE Progression of Operator Actions:
Page 1 of 3 O
SRV Open 5 Alarm (s)
An U Verify reactor pressure is normal nnp p Determine which valve has failed open 7 .
nnp if Attempt to close the failed valve by cycling switet or pulling the fuse for the affected valve
' SP0/RO V Refore reaching
, re ce N Sup ession circulation
- r PoolTe8p* pumps to minimum 393 F speed and manually scram the reactor Y
%Rn l' Eag an Enter RPV Control Enter Contain- suppressi ment Control y procedure pool temp. be procedure aintained
<vJ N
s ERO/p0/ lf BOP RO/000 if M Execute " Monitor Maintain reactor and Control Supp- pressure less g) ression Pool than HCTL Temperature" sub-procedure nnp 9 F Place suppression Exit to i pool cooling in RPV Control service O
130
t i
x.,/
' STUCK OPEN MAIN STEAM SAFETY / RELIEF VALVE Progression of Operator Actions:
Page 2 of 3 B1 SRn/Rn/ RnP SRO/Rn / URnP SRn/Rn/'I Rnp Execute " Monitor Execute " Monitor Execute " Monitor and Control RPV and Control RPV and Control Reac-Water Level" Pressure" sub- tor Power" sub-subprocedure procedure procedure SRn BnP II Rn if Verify no isolat-Verify turbine Place the reactor ion or ECCS init- pa mode switch in lation conditions n ng SHUTDOWN ex st reactor pressure
[N Rn " "
Rn 1I
,,\_/ ) Exit to Scra Verify all cont- Verify the reac-rol rods are in- Procedure tor recirculation serted beyond pumps runback to MSBWP minimum speed and
- l. core flow l
l 1r lf Exit to Exit to Scra Scram Proced. Procedure-ute r\
tj 131 ;
J 1
STUCK OPEN MAIN STEAM SAFETY /RELIEr VALVE Progression of Operator Actions:
Page 3 of 3 Enter from EPC's l'
nnP U '
RO U Is Trip or verify When all scram N C001down auto trip of the signals are clear rate greater main turbine / gen- bypass the SOV an LC07 erator high level scram i and reset the rea-. l ctor scram l
g fBop P Bop iI BOP /RO U Reduct or stop verify generator Return recirc.
all other steam breakers open system, RWCU, and loads on the ventilation sys-reactor to limit tems to service the cooldown rate if tripped RD 1 BOP P BOP P Verify that reac. Verify station When reactor tor power decreas- electrical loads pressure decreas-
+ es to less than transfer to the es to less than 3% of rated off-site power 100 psig, place supplies shutdown cooling in service 80p 'r pnp 1I epq 1f After level re- Start or verify Log the events turns to normal, auto start of the and actions taken trip all but one turbine lift pumps ,
feedwater pump motor suction pump ,
and the turning aear nit numn RO P RO/ BOP U CR0 l' Insert SRM and Remove unnecessary Notify supervisor IRM detectors and equipment from switch recorders service to monitor IRM's RD Y BOP l' enn ir Downrange IRM's verify MSIV's are Complete the
. to maintain an open and condenser scram report on-scale reading vacuum is normal (if SJAE's have 1ot been secured) if END O
132
Operating Sequence: Small Break Loss of Coolant Accident A NSSS/ Type: GE/BWR
(
Initial Plant State: Power Operations Sequence Initiator: Recirculation Line or Main Steam Line Break Inside i Containment. Break Area No Larger Than 20% of the Design Basis Line Break.
Important Plant Parameters: 1) Drywell Pressure, 2) Reactor Water Level,
- 3) Suppression Pool Temperature, 4) Suppression Pool Level, 5) Drywell Temperature, 6) Suppression Chamber Pressure, 7) Mark III Containment Temperature, 8) Mark III Containment Pressure, 9) HPCI System Flow,
- 10) RCIC System Flow, 11) RHR System Flow, 12) Low Pressure Core Spray System Flow, 13) Reactor Vessel Pressure Progression of Operator Actions: See Flow Chart Final Plant State: Cold Shutdown l
l l Major Plant Systems: HPCI, RCIC, RHR (LPCI), RHR (Suppression Pool Cooling), Containment Cooling, Low Pressure Core Spray, Primary Con-tainment Isolation System, Standby. Gas Treatment, RHR (Shutdown Cooling)
Tolerance Range: At all times the reactor water level should be main-1
.tained above the top of active fuel. Containment temperature and g
pressure should be maintained below their respective safety limits at all' times. Suppression pool temperature and level should be maintained below their respective safety limits at all times.
Competencies Tested:
1 SRO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response compliance /Use of' Technical Specifications Compliance /Use of Procedures Supervisory Ability Communication / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures l
l Control Board Operation l
Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals l Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction 133
SMALL BREAX LOSS Or CCOLANT ACCIDENT Progression of Operator Actions:
Page 1 of 3 High D.W.
Press. or Low evg1 scram y u 80P ,
nnp 4 Entet RPV Con- verify EDC's Operate all aval-trol and Contain, started or, high lable drywell ment Control Pro. D.W. prersure coolers before cedures do wo ll tempera-tute greaches
[135 F]
RRn ir RO m nnp u Verify applic - Verify reactor As required, able group iso- recirculation operate contain-lations occurred pumps runback to ment pressure on high drywell minimum if not control systems, pressure or low tripped water level following app- I licable procedure fi bop /RO o pn /nnp . ,
Rt1P ir Manually isolate Augment HPCI and As required, systems that RCIC as needed operate SCT and failed to iso- using preferred drywell purge late when re- systems (CRD, (if space gtemp.
quired feedwater, etc.) below 212 F),fol.
to maintain levej lowing applicable proceoures Rn u BOP ,
Place reactor Place suppres-mode switch in sion pool coolinti Mark SHUTDOWN in service using III Y 02 RHR pumps not Containment required for ade- ?
auate core coollog N
nnprinr nno ,, NnTr non r v Start or verify Prevent injectior Prevent injection Maintain supp.
auto initiation LPCS and LPCI from LPCI and pool level or of HPCI pumps not requi- LPCS pumps not RPV pressure : Al red for core co- required for core below SPLL 011ng before rea- cooling l rhinn inj- n nte. l
-m enn y nno o anp ,,
Verify low pres- Maintain supp. Initiate IC, if '
sure ECCS have pool temp. or applicable received [2 RPV pressure gj psig) initiation below HCTL signal Bop l
, r nno ,
Start or verify Can Open ADS valves auto start of HCTL N and depressurize RCIC if level be maintained the reactor less than [-52") ? NOTE: With RPV level normal or near normal, the low pressure ECCS may Y
flood the reactor when RPV pressure decreases to [500 psig) with the
, [2 prig) drywell pressure signal present. The operators n.ust trip the pumps manually to prevent inject-lon in excess of that required for i adequate core cooling.
134
SMALL BREAK LOSS OF COOLANT ACCIDENT Progression of Operator Actions:
Page 2 of 3 I,, \
\ }
(/ B1 B2 nnP AnP Can Prevent injection When containment SPLL from LPCS and temperatgrerea-e maintained LPCI pumps not re- ches [90 F],
quired for ade- operate all aval-7 quate core coolinc lable containment y coolina nnp Initiate isola- Is tion condenser' L N contain-if applicable A1 - ment temperatur proaghi 90 Y
ROD r ROP Is Open all ADS val- Prevent injection 73 suppres- ves and depres- from LPCI and LPCI suppressic
" surize the reac- pumps not requirec N er pms, sion pool l, eve tor for adequate core ove [12 bove [2 ap cooling 33 Y Y can mo -, AnP o
.m
, Request analysis Initiate isolatior Initiate suppres.
\ of suppression condenser, if sion pool spray
[
1 v) pool water sariple to determine if discharge is applicable using only those RHH pumps not re-quired for ade-cermitted auste core coolina Anp ROP o RnP <r Reduce and main- Open all ADS val- When suppression tain pool level ves and depres- chamber pressure below high level surize the reac- drops below [2 LCO if discharge tor psig], terminate is permjtted suppression pool spray nno a lf containment hydrogen concen-tration exceeds -
0.5%, place re- Al combines in service Exit to scram Procedure
]
I,v) 135
SMALL BREAK LOSS OF COOLANT ACCIDENT Progression of Operator Actions:
Page } of 1 Enter from EPC's 1
pn u BOP o I Verify all con- Shutdown EDG's trol rods fully if off-site inserted and RX power is avail-power is less able than 3%
nno nnp ,,
IT MSVI's closed ,
Start or verify place control auto start of switches in the lift pumps, CLOSE position motor suction pump, and turning aear oil cump RO , RD Insert SRM and When all scram 1RM detectors signals cleared, bypass SDV high level scram and reset reactor scram Rn ir ph/nno ir Switch recorders Return recirc.
from APRM and system, RWCU, and RBM to monitor ventilation sys- '
1RM's tem to service RO ir BOP i, Downrange IRM's When RPV press. }
to maintain ar. is less than [100 on-scale reading psig), place shut-down cooling in service nno ,
enn ,,
Trip or verify Log events and auto trip of the actions taken hiain turbine nno , c
_og u Verify station Complete scram electrical loads report u transferred to off-site power END supplies -
I O
136
,_ Operating Sequence: Reactor Coolant Leakage Outside Primary Containment
\
( ,) _
NSSS/ Type: GE/BWR
{
Initial Plant State: Power Operations I f
Sequence Initiator: Line Break in a Reactor Auxiliary System (e.g. RWCU, HPCI, or RCIC Steam Line), Except Main Steam Line Inside Secondary Containment Important Plant Parameters: 1) Secondary Containment HVAC Exhaust i
Radiation, 2) Secondary Containment Area Radiation Levels, 3) Secondary Containment Area Temperatures, 4) Secondary Containment Floor and Equip- f ment Drain Sump Levels,.5) Reactor Power Level, 6) Reactor Pressure l
Progression of Operator Actions: See Flow Chart 4
Final Plant State: Cold Shutdown Major Plant Systems: Secondary Containment HVAC, SBGT, Primary Contain-ment Isolation System, Process Radiation Monitoring, Area Radiation l Monitoring l
Tolerance Range: Prompt operator action must be taken to secure the secondary containment HVAC system and place SBGT in service to prevent l off-site releases of radioactivity.
l ,m (N_ ) Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings j Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability Communication / Crew Interaction RO - Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation )
Communications / Crew Interaction i i
BOP - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction
/'~'T l
,v 137
i l
REACTOR COOLANT LEAKAGE OUTSIDE PRIMARY CONTAINMENT Progression of Operator Actions- I Page 1 of 3 4 Secondary Con '
tainment high temn_ ne r. .
SRO n Enter Secondary Containment Con-trol Procedure SRO/RO/ BOP cRn/Rn/ 9 RnP RRn/Pn# 3 Execute " Monitor Execute " Monitor and Control Sec- Execute " Monitor and Control Sec- and Control Sec-ondary Contain- ondary Containment ment Temperature" ondary Contain-Radiation Levels" ment Water Levels' subprocedure subprocedure subprocedure nRn '
y u yjenn 1' Request auxiliary Isolate or verify Operate or request operator operate auto isolation of radwaste operate all available HVAC and start sump pumps for secondary contain- SBGT system any sump with ment area coolers high level
' RO/py 1r y End Sec. Cont Isolate all sys- End Sec. Cont Temperature B1 tems dist7arging Water Level
_nntrn1 into the afiected ontrni area, except fire a and reactor safe-ty systems v
Are rad. leve . N End Radiatio
, bove MSO leve Control Pro-I more t. 94 '"a i
o ' a* a?
Y Rn / ROP 1' Rapidly depress-urize the reactor using the turbine bypass valves nnp "
When shutdown cooling interlock:
are cleaIed, place shutdown cooling in service
'End Radiatio Control Proc n4 ,. n 138
NEACTOR COOLANT LEAKACE OUTSIDE PRIMARY CONTAINMENT
, Progression of Operator Actions:
/ g Page 2 of 3 I
.\ ___/
B1 snn n Enter RPV Control l Procedure j Rn U If not initiated, initiate a reac-tor scram l
SRO/Rn/ ir AnP SRn/Rn/i' AnP SRn/Rn/ sr pnp Execute " Monitor Execute " Monitor Execute " Monitor and Control RPV and Control RPV and Control Reac-Level" subproced. Pressure" sub- tor Power" sub-ure procedure procedure P nnP/Rn " Rn U l j g SRO Place reactor i vj verify isolstions that should have Verify turbine bypass valves mode switch in occurred automat- are controlling SHUTDOWN ically initiated reactor pressure Rn v sap U Rn if Verify all control If SRV's are cycl- Verify reactor rods inserted be- ing, initiate isoj - recirculation l yond MS8WP ation Condenser pumps runback to l
and open SRV's un- minimum speed til reactor press-ure is below [935E sig) i i
nnp e o "
j If feedwater sys- Exit to Scra Exit to Scram I tem is unavailable ,
Procedure Procedure start or verify auto start of RCIC o
Exit to Scram Procedure i
)
r'N l' (v) l Q
139 !
1 i
, 1 l j
REACTOR COOLANT LEAKAGE O!JT51DE PRIMARY CONT AINMENT Progression of Operator Actions:
Page 3 of 3 l Enter from ,
EPC's !
nn w n ne, ., nnp e Verify that reac- Verify station Establish RPV tor poner electrical loads cooldown rate decreases to les5 transfer to the at LCO than 3% of rated off. site power supplies nnP u pn/nnp er enn F Atter level re- Remove unneces- Log the events turns to normal, sary equipment and actions trip all but one from service taken feedwater pump PG " M SRO g insert SRM and verify M5v1's Notify super-IRM detectors anc are open and con- visor switch recorders denser vacuum is to monitot IRM's normal pn y PD o can .,
Downrange IRM's When the scram Complete scram to maintain an signals are clear .
report on-scale reading bypass the SDv high level scram and reset the a reactor scram pop u pn/nnP er v Trip or verify auto trip of the Return recirc.
system, RWCU, and EO main turoine/ ventilation sys-generator tem te service if tripped 90p er verity generator breakers open o
O 140
/~~% l I Operating Sequence: Jet Pump Failure V )' .
NSSS/ Type: GE/BWR Initial Plant State: Power Operations I
Sequence Initiator: Structural Failure of a Jet Pump Diffuser Section or Ram's Head (Affects Two Jet Pumps) l J
Important Plant Parameters: 1) Core Flow, 2) Jet Pump Flows, 3) Reactor Recirculation Pump Speeds, 4) Reactor Recirculation Pump Flows,
- 5) Reactor Vessel Pressure, 6) Core Plate Differential Pressure, j
- 7) Reactor Water Level, 8) Reactor Power Level !
1 Progression of Operator Actions: See Flow Chart Final Plant State: The plant should be in cold shutdown with shutdown cooling in operation. The examiner may wish to end the exercise prior to complete shutdown and cooldown, if this scenario is not used to fulfill the normal evolution requirements.
Major Plant Systems: Reactor Recirculation, RHR (l'nutdown Cooling Mode) l Tolerance Range: Technical specifications specify that reactor operation l
, _s without'all jet pumps operable is not permitted. The reactor must be in
( cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a jet pump failure.
\s_
Competencies Tested:
SRO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Proceduces Supervisory Ability Communications / Crew Interaction RO - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Technical Specifications Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction l
BOP - Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction i 1
l r%
l s_ /
I 141 4 l
JET PUMP FAILURE Progression of Dperator Actions:
Unexpected Change in rnre rin-enn /nn/v nno RO/ ROP o Determine the Perform core flow cause if the rate calculation change in core surveillance flow test 1
SRO/RO/u B0p nn/nnp u Verify 1.hermal Perform diffuser parameters are to lower plenum in a safe region differential pressure survell-lance test RD u SRO e Verify power vs. Determine il flow in a safe Tech. Spec, cil-region teria for jet i pump operability I is met 90 r enn .
Insert control Declare the jet rods and adjust pump (s) inoper-core flow as nec -
able essary to main-i n operatino tajhinoperating l wl.2 j Rh/RnP u SRO/RO/c BOP Perform recire. Commence an 07-pump flow im- derly reactor balance measure- shutdown and ment survell- cooldown lance test nno _.
Plate shutdown cooling system in service in-jection into the unaf fected loop 1P END i
I 142
. _ - _ _ _ _ - _ - _ _ _ _ _ _ _ ~
Operating Sequence: High Suppression Pool Water Temperature 7k_,/ NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Leaking SRV Important Plant Parameters: 1) Suppression Pool Water Temperature,
- 2) Suppression Pool. Level, 3) Reactor Vessel Pressure, 4) Reactor Water Level, 5) Reactor Power Level Progression of Operator Actions: See Flow Chart Final Plant State: Hot Shutdown Major Plant Systems: RHR (Suppression Pool Cooling), Reactor, Main Steam Tolerance Range: The operators should scram the reactor before. reaching the boron injection initiation temperature (typically 100 degrees F).
Competencies Tested:
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures RO - Understanding of Instrument / System Response 7~ Compliance /Use of Procedures
(
i N Control Board Operation BOP - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Technical Specifications Compliance /Use of Procedures Control Board Operation NOTE: For additional information pertaining to this event, refer to the BWR abnormal event description for increasing suppression pool water temperature.
/
1
(
f 143 1 1 1.
HICH SUPPRESSION POOL WATE7 TEMPERATURE Progression of Operator Actions:
Page 1 of 2 High $uppres-sion Pool Tem-
.larm SRO P Entec Containment Control Procedure.
I fglz3n/hnon cnn/nn/ hnon Execute " Monitor Execute " Monitor and Control Supp- and Control Supp- i ression Pool Temp- ression Pool Wate:' !
erature",subpro- Level" subproced-cedure ure nne U V Operate available is suppression pool suppress-N cooling Jon pool leve
-bove (12' 2" '
RO if SRO Y Before reaching Request analysis BIIT, scram the of suppression reactor pool water sample to determine if discharge is permitted RRO h U BOP V Enter RPV Control Can Reduce and main.
Procedure pool temp. Y tain suppression he maintained - pool level below
' low FCL the high level y LCD N
U nnP P u Reduce and main- End Suppress B1 tain reactor ion Pool Level pressure to below N ' "'I FCLT v
End Suppress g y y1 Temc.
1 i
144 l
l
i l
l
/'~h HIGH SUPPRESSION POOL WATER TEMPERATURE i ) Progression of Operator Actions:
\, _,( Page 2 of ?
B1 SRO y Enter RPV Control Procedure RRO/Rn/ M9P SPO/RO/ U BOP tRn/Rn/ Ann Execute " Monitor Execute " Monitor Execute " Monitor l and Control RPV and Control RPV and Control Water Level" sub. Pressure" subpro- Reactor Power" procedure cedure subprocedure tRn U RSP V Rn U l Verify no isola- Verify turbine Place the reactor l
tion or ECCS init- bypass valves are mode switch in ation conditions controlling RPV SHUTDOWN exist pressute fo- x . --
! )
\_
'~~' ) Rn n Verify all control U Rn v Exit to Scram Verify the reacto!
rods are inserted Procedure rec 1IOulation beyond the MSBWP pumps run back to minimum speed and core flow v y Exit t." Scram Exit to Scra Procedure Procedure NGTE: For subsequent operator actions following the reactor scram, refer to the BWR emergency event description for reactor scram with MSIVs open, jG,
- 1 \
\v/ {J 145
j Operating Sequence: Main Turbine or Generator' Trip
/
.{ NSSS/ Type: .GE/BWR'
-Initial Plant State: Power'0perations Above 30% of Rated Power Sequence Initiator: Closure of the Turbine Stop or Control Valves Due ;
to Any Failure That Will Initiate a Turbine or Generator Trip (Loss of Lube Oil Pressure, Load Reject, etc.)
IImportant Plant Parameters: 1) Reactor Water Level, 2) Reactor Power Level, 3) Reactor Pressure, 4) Turbine Stop Valve Positions, 5) Turbine 1 Control Valve Positions ,
1 Progression of Operator Actions: See Flow Chart Final Plant State: Hot Shutdown
. Major Plant Systems: Main Steam, Main Turbine / Generator and Auxiliaries, Condensate and Feedwater Tolerance Range: Reactor water level must be maintained above the top of active fuel at all times. Reactor vessel pressure must be controlled below the safety limit at all times.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals
, (,^ Compliance /Use of Procedures Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures
. Control Board Operation
BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response ;
Compliance /Use of Procedures j Control Board Operation '
1 1 ,
\
i 147 1
1 MAIN TURBINE OR CENERATOR TRIP Progression of Operator Actions:
Page 1 of 2 Turbine Trip, RX Scram
)
epn v Enter RPV Control Procedure epn/An/hpnp cRn/Rn/ if AnP spn/Rn/ nnp Execute " Monitor Execute " Monitor Execute " Monitor and Control RPV and Control RPV and Control Reac-Water Level" 5.ub Pressure" sub- tor Power" sub-procedure procedure procedure c;Rn 9 BOP U 40 "
Verify no isola- Verify turbine Place the reactor tion or ECCS b'/ pass valves are mode switch in initiation condit- controlling RPV SHUTDOWN lons exist pressure RO P 'I Rn 'I Verify all contro] Exit to Scra Verify the reacto!
rods are inserted Procedure recirculation beyond the MSBWP pumps are tripped (if RPT breakers intalled) or run-back to minimum o P Exit to Scra, Exit to Scram Procedure Procedure 9
148
1 l
/ *.
j
= _ , MAIN TURBINE OR. GENERATOR TRIP j Progression of Operator Actions: l Page L o( L I
i Enter from EPC's Rn ir' nnp
pn l' j Verify that reac- Verify station When the scram i tar poner level- electrical loads signals are ci:,ar, j decreases to less transfer to the bypass the SDV i than 3% of rated of f-site power high level scram l supplies and reset the rea- .
ctor scram Roo V nnp U pnfqnP l' After reactor Start or verify Return recire, level returns to auto start of the system, RWCU, and normal, trip all turbine lift pump: , ventilation sytems but one feedwater motor suction pum; ., to service if ;
pump and turning gear tripped l oil pump 1
U l RO U SRn/nnP ' sRn insert SRM and IRF Determine the Log the events s
/ % detectors and cause of the tur- and actionu taken saltch recorders bine or generator
((-) to monitor IRM's trip and request repairs j Rn 9 Rn / Ann " SRO U Downrange IRM's Remove unnecessary Notify supervisor to maintain an equipment from on-scale reading service i
BOP y Anp " son U l I
Verify generator Verify MSIV's Complete the breakers open are open and con- scram report denser vacuum is normal
$f END g-g i
mj !
149
Operating Sequence: Main Turbine or Generator Trip Without Bypass Valves p_
/ i x ,,/ NSSS/ Type: GE/BWR Initial Plant State: Power Operations Above 30% of Rated Power Sequence Initiator: Closure of the Turbine Stop or Control Valves Due to Any Failure That Will Initiate a Turbine or Generator Trip (Loss of Lube !
}
Oil Pressure, Load Reject, etc.) and the Turbine Bypass Valves Fail to Open to Control Reactor Pressure i
Important Plant Parameters: 1) Reactor Water Level, 2) Reactor Power Level, 3) Reactor Pressure, 4) Turbine Stop Valve Position, 5) Turbine Control Valve Position, 6) SRV Acoustic Monitor 1
Progression of Operator Actions: See Flow Chart l
Final Plant State: Hot Shutdown Major Plant Systems: Main Steam, Reactor Recirculation, Main Turbine / ,
Generator and Auxiliaries, Condensate and Feedwater, HPCI, RCIC Tolerance Range: Operators must ensure that the reactor vessel pressure does not exceed the safety limit.
Competencies Tested:
,/ 3
. i SRO - Understanding / Interpretation of Annunciator / Alarm Signals
\_-}
~
Diagnosis of Eventc/ Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedurea Control Board Operation Communications / Crew Interaction l
-'y l l
'm_/
l l
151 l
MAIN TURBINE OR GEtJERATOR TRIP WITHOUT BYPASS VALVES Progression of Operator Actions Page 1 of 2 Turbine Trip, RX Scram sRn ir Enter RPV Control Procedure r
SRO/R3.e ROD y SRO/RO u BOP i
SP0/RO 4 BOP Execute "Moritor Execute " Monitor _ Execute " Monitor and Control RPV ard Control RPV and Control Reac-Water Levela sub- Pressure" sub- tor Power" sub-procedure procedure procedure SRO gr ROP gr Rn MP verify no iso. Initiate Isola- Place the reactor ,
lation or ECCS tion condenser or node switch in i initiation con- suppression pool SHUTDOWN dition exist cooling l l
l RO ir Ano er RO e j verify all con- Manually open verify recircu- i trol rods in- SRV's until RPV lation pumps are i serted beyond pressure drops tripped (if RPT the MSBWP to [935 psig) breakers lastal-led) or runback to minimum
,e ir ir Exit to Scram Enter to Scram Enter to Scra Procedure >rocedure rocedure ;
l l
l l
l 1
0, 152 1
1
MAIN TURBINE OR GENERATOR TRIP WITHOUT BYPASS VALVES Progression of Operator Actions:
\ Page 2 of 2
^( s._.--
)
Enter from EPC's j Rn 'r BOP ir RO/ BOP gr Verify that reaC- Start or verify Augment RPV tor power level auto start of the pressure control decreases to lesG turbine lift pumps ,
with HPCI and than 3% of rated motor suction RCIC turbine pump, and turning operation aear all oumo BOP i, can/nnP ,
BOP /R0 ,,
After reactor Determine the Commence FYV cool-level returns to cause of the tur- down at the cool-t normal, trip all bine or generator down rate LCO l trip and request but one feedwater pump repairs Dn v RO/ ROP gr 100 i, Insert SRM and Remove unneces- When shutdown IRM detectors and sary equipment cooling interlocks switch recorders from service are cleared, place to monitor IRM's shutdown cooling in service
[ v (D RO nnr> can
,r ir
) Downrange IRM's Verify MSIV's Log the events I V' to maintain an are open and con- and actions taken on-scale reading denser vacuum is normal BOP gr Rn 1, SRO ir verify generator When the scram Notify sypervisor breakers are signals are clear open bypass the SDV high level scram and reset the reactor scram BOP i- nninnn , SPD 1r verify station Return recire. Complete the scram electrical loads system, RWCU, and report transfer to the ventilation systen s off-site power to service, if supplies tr1pped ir END 1
l v) l 153 I
- < .J r t j 't .
Operating Sequence: Loss of Conde.nser Circulating Water ;
n.
/ NSSS/ Type: GE/BWR ,
j
( Initial Plant State: Greater Thanl50% Reactor Power I'
Sequence Init!ator: Simultaneous Trip of All Circulating Weter Pumps p, .
Important Plant Parameters: 1) Rede; tor Power Level, 2) R3 actor Water Level, 3) Reactor Pressure, 4) Main Condenser Vncuum. ?.) Main Turbine /
Generator Load
)( ,
Progression of Operator Actions: See Flow Ch'rt( a Final Plant State: The plant will be in hot shutdown with the MSIVs closed, reactor pressure control through tho arsin nteam relief valves, and reactor level control by RCIC (if the plast is equipped with turbine j driven feedwater pumps) ; j Major Plant Systems: Circulating Jaer, Reactor' Recirculation, Main _)
Turbine / Generator ,
g Tolerance Range: Under no circumstances should the operators permit discharge of nuclear steam to the main condenser.without adequate vacuum h.
and seal steam pressure. hithout vacuum and seal steam,' radioactivity ,
will become airborne.in the turbine building. '-l i-Competencies Tested:
- /
(% SRO - Understanding /tuterpretation of Annunciator / Alarm Signals Die.posis of Events / Conditions Based on SignalWMeadings ; ;
Understanding of Instrument / System Respon's,3 i Compliance /Use of Procedures /; .
M Supervisory Ability .
/
h j
' j' !
Communications / Crew Interaction ,
f ]
1, RO -- Understanding / Interpretation of Annunciator /Alaru Signals / '
Diagnosis of Events / Conditions Based on Signaln/ Readings Understanding of Instrument / System Response ,
Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction )
~
BOP - Understan(ing/ Interpretation of Annunciator / Alarm signtils i
<l Understanding,of instrument / System Response Compliance /Use of Procedures Control Board Operation >l Communications / Crew Interaction [ j 3
p
( )
\.
155 1
~
[
/ /'
/
f , , u.5 Ol' CONDENSER CIRCUL ATING WATER J Progression of Operator Actions; e , Page 1 of ?
, i,
'4 L'a rc Wateri)
Pumps Trip f ri
< i j nno P < yt jL_
t Place cce'rol Enter RPV (7ntrol
, switches ror the ' ocecfle m tripped pumps in the STOP position
(
, '. i ,
BOP P .' E]E nnP e.m/ Pn./ I ' non enn'nn' P nro Close or ve '.fy ;Execa e "Manitor E mcute " Monitor Execute " Monitor auto closure of the tripped pw'p 1.d control RPV W,.ter '_evel" sub-r 3 Coatcal RPV andControlReac-l r P '.*srure" subpro- tor Power" sub-discharge val es n' x ' cure cedure procedure I
enn t son i non U P0 P
' 'Ruqi,m *..an aux 11- Verify WIV'. Initiate IsrJatf or Place reactor laT' operator close x.er conden- .
condenser or bupp. mode switch in ire estigate the ser vac am reaches ression pool cool 5HUTDOWN cause of tr.; tne iso.ation set- ing
- pump trips thint '
,I i l 7,
/ noo e grm i nnP e nn e Attempt to restart 3rify all c.ther Manual]v open Verify reactor the tripped circ. Isc;1ticrS Occurrel SRV's li, che pre- recirculation
,,, water pumps that shottd have scribed sequence pumps are runback occurred to redo.e r reactor to minimum speed prd ac t to [9M and core flon EbY _ _ _ _
pn y e0p u PP o u
Reduce reactor '
Start or verify Augment prese,ure Exit to 5 cran.
powerlevelusingl ,n to start of control with Procedure recirc. flow and ' ACIC HPCI and RS C control rods ,.
turbine aneration
/
f
{ nno <' 4 xx\ ' Start or verify E r r t 9 s %m N sciamo} auto start of a n',- au;
- aporoat'aing HPCI _ , _ , _
scram 7 Y
Rn U nno ..y If nut automatic I ' Rest *re 2enctor all, scromed, ( watsi level to manually scram the normal oper-the reactor '
attig band using (HPCI and RCIC I J l
t
, Exit to Scre e Procedure s
r 156
l y ,,
,yr
'/
in LOSS OF CC M MSER CIRCULATING WATER Progression of Operator Actions:
9,, Page 2 of-?
Enter from T EPG's i}
1' rpq if c,pn 'I
's nno Place MSIV control Inform radwnste Evaluate the need
()
switer.us in the operators of the for reactor vessel y
CLOSE position circ. water syster cooldown failure and dir-ect them to stop any blowdown ops U cPn U
' AnP yg!_ Log the events (e;1Ty a.11 control When reactor watei
.ods fulh insert- level reaches not. and sctions taken
,, mal band, secure i
ed and mactor power decreased to HPCI if not used less than 3% for pressure con-trol nnp u cop s90 Trip or verify Notify the super I JControl injection visor fram RCIC, HPCI, auto trip of the and feedwater Oco- main turbinu and tor driven pums; transfer of elec.
to prevent high loads to off-site llevel turbine trip power y qqn U con p Start or verify Complete the stran
.in r t ditM and IR'4 uetectors auto start of the 7,. port turbine lift pump! ,
motor suction pum; ,
and turning gear oil pump y pn u o 90 Switch recorders When all scram END from APRM and ram signals are clear-to record 73M ed, bypass the SD\
high level scram and reset the re-actor scram RO/Onp 't nn a Downrange IRM's Return re, circ.
to maintain an tystem, RWCU, and on-scale reading ventilation sys-I tems to service, if tripped a
i 157
Operating Sequence: Loss of Feedwater System k NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Trip of All Feedwater Pumps During Power Operations.
The examiner may wish to trip condensate or condensate booster pumps individually to induce the feedwater pump trips on low suction pressure.
Important Plant Parameters: 1) Reactor Water Level, 2) Reactor Vessel Pressure 3) React 6r Power Level, 4) Feedwater Flow, 5) Reactor Feed-water Pump Suction Pressure, 6) Reactor Recirculation Pump Speed,
- 10) Main Condenser Vacuum, 11) Main Steam Line Pressure l
Progression of Operator Actions: See Flow Chart 1
Final Plant State: Hot Shutdown j Major Plant Systems: Feedwater and Condensate, Control Rods, Reactor 1 Recirculation, Main Steam HPCI, RCIC, Primary Containment Isolation System, RHR (Suppression Pool Cooling Mode)
Tolerance Range: Operators must maintain reactor water level above the top of active fuel at all times.
Competencies Tested:
I SRO - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedur es Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction 1
159
l LOSS OF FEEDWATER SYSTEM Progression of Operator Actions: i l
Page 1 of 2 Low RPV Wate i
Level Scram SRO 4 Enter RPV Control Procedure I
cRn/Rn/ h nnP SRO/RO/ " BOP spn/Rn/ h nnp Execute " Monitor Execute " Monitor Executs " Monitor and Control RPV and' Control RPV and Control Rea.
Water Level" sub- Pressure" subpro- Ctor Power" sub-procedure cedure procedure can l' B0o 4 Rn P If RPV water Initiate isolation Place reactor level reaches condenser or supp- mode switch in isolation set- ression pool cooJ- SHUTDOWN point, verify ing MSIV's close SRn
Rnp 1' Rn if Verify all other Manually open Verify reactor isolations occur- SRV's in prescrib. recirculation red that should ed sequence to pumps runback to have occurred reduce reactor minimum or trip-pressure to [935 ped pSiol AnP P ROP 1' 1r Start or verify Augment pressure Exit to Scram auto start of control with Procedure RCIC HPCI and RCIC turbine operation nap v v Start or verify Exit to Scram auto start of Procedure HPCI __
nnP "
Restore reactor water level to the normal oper-ating band using RCIC and FPCI P
Exit to Scra Procedure 1
160
l LOSS OF FEEDWATER SYSTEM Progression of Operator Actions:
Page 2 of 2 x
s 1
I
,\ /
' - ' ' Enter from EPG's 1
' RnP P BOP P BDP 1 Place the MSIV Start or verify Use turbine bypass control switches auto start of the valves to control in the CLOSE pos. turbine lift pumpt , reactor pressure ition motor suction pump ,
ard turning gear oil pump Rn U Rnp /Rn ir RO l Verify all control Reset the isolat- When all scram rods are fully in- lon logic signals cleared, serted and reactor bypass the SDV power has decreas. high level scram ed to less than and reset the red.
3% ctor scram BOP ' Ano P RO/RDP F When RPV water Open the outboard Return recire, level returns to MSIV's system, RWCU, and the normal band, ventilation sys-Secure HPCI if not tems to service used for pressure control pO_
U Rnp U SRO ir
,, _ s Insert SRM and IRM Pressurize the Log the event and
/ \ oetectors main steam lines actions taken and open the
(\m -) inboard MSIV's U Rnp H SRO 4 RO Switch recorders Place the seal Notify supervisor from APRM and RBM steam system in i
' to record IRM service en nnp u sRo y Downrange IRM's Establish main Complete the scram to maintain an condenser vacuum report on-scale reading using the mechan-ical vacuum pump and/or the SJAE's nnp U Bop P 1 Trip or verify After main conden-auto trip of the ser vacuum is es- END main turbine and tablished, attenot transfer of elec. to restart one loads to off-site reactor feedwater power pump t
j'"'s
/ \
! /
N_/-
161
l Operating Sequence: LossLof All High Pressure Feedwater p
k, NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Complete Loss of Reactor Feedwater System with HPCI l (or HPCS) and RCIC Systems Inoperable Important Plant Parameters: 1) Reactor Water Level, 2) Reactor Vessel Pressure, 3) CRD System Flow, 4) Suppression Pool Temperature Progression of Operator Actions: See Flow Chart Final Plant State: Cold Shutdown Major Plant Systems: Control Rod Drive Hydraulics, RHR (LPCI Mode), RHR (Suppression Pool Cooling Mode), RHR (Shutdown Cooling Mode), Main Steam Tolerance Range: Should reactor water level reach the top of active fuel, the operators should immediately perform an emergency reactor vessel depressurization.
Competencies Tested:
'SRO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures
[ Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of' Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction k'
163 l
1 t - - - - --------_ _ __---
LOSS OF ALL HIGH PRESSURE FEEDWATER Progression of Operator Actions:
Page J of 2 Low RPV Level Alarm SPD w Lnter RPv Control Procedure I
enn/an w nnn enn /nn 4 nnn enn/pn Jr nnP Execute " Monitor Can Execute " Monitor Execute " Monitor and Control RPV FV level LL and Control RPV Water Level" sub- and Control reac-Pressure
~
Y procedure maintained tor Fower" sub-above procedure Ar N
940 V POP V qr BOP R0 1r verify all 150-lations occurred Initiate suppres- Initiate isolation verify all sion pool cooling condenser or sup. control rods are that should have aression pool occurred fully inserted acaling Rn l
1r ,_
pop y qqp gy pg gg verify all cont- Open all ADS Manually open verify recir-rol rods inserted valves SRV's until RPV culation pumps beyond MS8WP pressure drops to are IJoback to (9.B psig) minimum speed and core flow ERn v 900 Rap ir .ir Verify ECCS and verify LPCS and Using SRv's EDC initiations LPCI inject when depressufite the Exit to Scram occur at [-152"] shutoff head is RX at the couldowr RPV level reached rate LCO (no emerg. oepres-surization)
EU 4F 000 w BOP 9r Start standby CRD when level return: when shitdown pump to normal, stop cooling inter-LPCS and LPCI locks are cleared, pumps not reQuiret place shutdown to maintain normal cooling in ser-level vice non I v w y Attempt to man-ually start HPCI Exit to Scram Exit to Scram procedure Procedure and RCIC O
164
LOSS OF ALL HIGH PRESSURE FEEDWATER Progression of Operator Actions:
Page 2 of 2 Rn i, AnP gr Downrange IRM's When RPV press.
to maintain an is reduced, align on-scale reading condensate system to feed the Teat-Enter from tor rpc't nnp e,Rn <r Pn w .,
verify all rods Trip or verify ILog the events fully inserted auto trip of the and actions taken and reactor power main turbine and less than 3% transfer of elec.
loads to off-site power 1 en u SRO ,,
nno ir ._
When all scram Notify supervisor Place M31V cor.-
trol switches in signals cleared, the CLOSE bypass SDV high postition level scram and reset rrrnm RO/ POP o epn ,,
nn v keturn recirc. Complete scram Insert SRM and system RWCU, and report IRM detectors ventilation systems to service RO ,, Ago ,, ,,
Switch recorders Establish cool-from APRM and down rate using EfC RBM to record SRy's IRM 165
1 Operating Sequence: Loss of Plant Control / Instrument' Air l'
\ NSSS/ Type: GE/BWR.
-Initial Plant State: Power Operations Sequence Initiator: Gradual' Decrease in Control / Instrument Air Pressure -l to O psig Important Plant Parameters: 1) Control / Instrument Air System Pressure,
Pressure, 5) Reactor Power Level Progression of Operator Actions: See Flow Chart-Final Plant State: Hot Shutdown Major Plant Systems: Control / Instrument Air, Plant Service Air, Control Rods, Reactor Recirculation, Condensate and Feedwater Tolerance Range: The operators must scram the reactor.when the air-pressure decreases to approximately 60 psig or control rods begin to drift inward.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals g
7- Understanding of Instrument / System Response Compliance /Use of Procedures Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation 167 i
__________________-__L
I l
LOSS OF PLANT CONTROL /INSTRUIENT AIR Progression of Operator Actions: 1 Page 1 of 2 Low Air Pres-sure Alarm (s)
AnP l' CRn 3
if Attempt to start Enter RPV Control standby air com- Procedure pressor spn 1' sRn/Rn/ RnP RPn/Rn/ U nnP RRn/An/ RnP Verify non-essent- Execute " Monitor Execute " Monitor Execute " Monitor ial air loads isol .
and Control RPV and Control RPV and Control Reac-ate on low air Water Level" sub. Pressure" subpro- tor Power" sub-pressure procedure cedure procedure sRn If sen u BOP U nn if Verify cross-tie Verify no isolat- Verify turbine Place reactor to service air or ion or ECCS init- bypass valves are mode switch in nitrogen system lation conditions controlling reac- SHUTDOWN opens (if applic- exist tcr pressure able) i Rn if pn U " Rn U When scram air Verify all contro: xit to Scram Verify the reactor header pressure rods are insert- Procedure recirculation decreases to (60 ed beyond the pu@s runback to psig) or rods MSBWP minimum speed and drift, manually core flow-scram the reactor o V Exit to Scram Exit to Scram Procedure Procedure O
168
LOSS OF PLANT CONTROL / INSTRUMENT AIR Progression of Dperator Actions:
Page 2 of 2 Enter from EPG's RO ' R7P i' nDP "
Verify reactor Verify station Verify main con-power decreases electrical loads denser vacuum is to less than 3% transfer to the normal of rated off-site power supplies ROP U BOP 'r nn U After level returns Start or verify When the scram tu normal, trip auto start of the signals are clear all but one feed- turbine lift pumps , bypass the SDV water pump motor suction pump , high level scram and the turning and reset the rea..
gear oil pump ctor scram Pn SRO " A0/BDP U Insert SRM and IRh Monitor MSIV pos- Return recirc.
detectors and itions for possib1 : system, RWCU, and switch recorder drifting to closed ventilation sys-to monitor IRM's position tems to service if tripped i
RO h RO/ ROP EPn U Downrange IRM's Monitor systems Log the events to maintain an with air operated and actions on-scale reading components for taken abnormal operation BOP 1' SRO " ERO P Trip or verify Request auxiliary Notify supervisor auto tria of the operator determine main turbine / gen- the cave,t of the erator loss of air and initiate repairs AGP 1' RO/ADP U ERO l' Verify generator Remove unnecessary Complete the scran breakers open equipment from report service u
END
/
/
169
Operating Sequence: EHC Pressure Regulator Failure (All Valves Open)
NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Failure of the In-Service Pressure Transmitter in the High Direction Causing the EHC Logic to Open All Turbine Control and Bypass Valves Fully l' Important Plant Parameters: 1) Main Steam Line Pressure, 2) Reactor l Water Level, 3) Reactor Vessel Pressure, 4) Reactor Power Level,
- 5) Suppression Pool Temperature Progression of Operator Actions: See Flow Chart Final Plant State: Hot Shutdown with MSIVs Closed Major Plant Systems: Main Turbine, Electrohydraulic Control, Main Steam.
HPCI, RCIC, RHR (Suppression Pool Cooling Mode), Co' n trol Rods, Reactor Recirculation Tolerance Range: To prevent an uncontrolled reactor depressurization of the reactor, the MSIVs should not be opened until the malfunction with the EHC system is corrected.
Competencies Tested:
Y SRO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction R0 - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation i Communications / Crew Interaction
/
171 1
EHC PRESSURE REGULATOR FAILURE (ALL VALVES OPEN)
Progression of Operator Actions:
Page 1 of 2 I
Heactor Scram, 1 MSIV's Close 1'
9D0 Enter RPV Control procedu*e I
Rno V cPn/Pn/dnnp son /an/ irnno snn/Pn/Ynnp Start or verify Execute " Monitor Execute " Monitor Execute " Monitor automatic start and Control RPV and Control RPV and Control Reac-of RCIC Water Level" sub- Pressure" subpro- tor Power" sub-procedure cedure procedure nno P spn U nne u on V Start or verify verify MSIV's are Initiate isolation Place reactor automatic start closed condenser or supp- mode switch in of HPCI ression pool cool- SHUTDOWN ing l nno 4 son p nnp u on y Restore reactor Verify all other Manually open Verify reactor water level to isolations occurr. SRV's it prescrib .
recirculation the normal oper- ed that should ed sequence to pumps runback to ating band using have occurred reduce reactor minimum speed and HPCI ard RCIC pressure to [935 core flow psigl u pq U nnp U U Exit to Scram Verify all control Augment pressure Exit to Scram Procedure Iods are inserted control with Procedure beyond the MSDWD HPCI and RCIC turhine operation 1
Exit to Scram Procedure O'
172
l EHC PRESSURE RECULATOR FAILURE (ALL VALVES OPEN)
, - -s Progression of Operator Actions:
}I Page 2 of ?
%J Enter from EPC's non o pn u nnn "
Place h61V contro. Downrange IRM's Log the event and switches in the to maintain an actions taken CLOSE position on-scale reading i nn " nnp y son U Verify all contro .
Trip or verify Notify supervisor rods fully insert.. auto trip of the ed and reactor main turbine and power decreased transfer of elec.
to less than 3% loads to off-site power supplies nno " nap P cnn U Control injection Start or verify Complete scram from RCIC, HPCI, auto start of the report and feed 4ater (mo-. turbine lift pumps ,
tor driven pumps) motor suction pump ,
to prevent high and turning gear level turbine trip oil pumo
['N pq y pq U y I } Insert SRM and Wher all scram
's__/
IRM detectors signals cleared, END bypass the SDV high level scram and reset the re-actor scram Pn l' Pn/RSP 1' Switch recorders Return recire, from APRM and system, RWCU, and RBM to recoro ventilation sys-IRM's tems to service if tripped i
l yy iI s l
173 1
1
i
.-g - Operating Sequence: Loss of Nuclear Service Water NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Failure of All Operating and Standby Service Water Pumps Important Plant Parameters: 1) Service Water System Pressure, 2) RBCCW Heat Exchanger Outlet Temperature, 3) Drywell Temperature, 4) Recircula-tion Pump Motor and Seal Cooler Temperatures Progression of Operator Actions: See Flow Chart Final Plant State: Hot Shutdown Major Plant Systems: Service Water System, RBCCW, Main Steam, Main Turbine and Generator, Fuel Pool Cooling and Cleanup System, Primary Containment, Reactor Recirculation Tolerance Range: Operator actions jn response to the service water system failure should be oriented toward the protection of plant equip-ment from operation without adequate cooling. The operators should also take action (or discuss' actions) concerning the Tech. Spec. limits.on fuel pool temperature.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response i Compliance /Use of Procedures Control Board Operation 175 L
[ . _ _ _ _ . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
LOSS OF NUCLEAR SERVICE WATER Progression of Operator Actions:
Page 1 of 2 Low S.W.
Press Alarm (s R9P i' Attempt to restort service water pumps to service Ro o Manually scram the reactor and close MSVI's SRO o Enter RPV Control Procedure RnP o SRO/RO/ 1 BOP RRn/Rn'dunp SPD/RD/ < r ADP Start or verify Execute " Monitor Execute " Monitor automatic start and Control RPV Execute " Monitor and Control Reac- and Control Reac-of RCIC Pressure" sub- tor Power" sub-procedure tor Power" sub-procedure procedure BOP u sRn ., nnp o Rn o Start or verify verify MSVI's Initiate isola- Place reactor automatic start closed tion condenser or mode suitch in of HPCI suppression pool SHUTDOWN cooling BOP o sRO ,, Rnn u pg y Restore reactor Verify all other Manually open Ve11fy reactor water leval to isolations occur- SRV's in pre- recirculation the normal oper- red that should scribed sequei.ce pumps runback to ating band using have occurred to reduce reactor minimum speed HPCI and RCIC pressure to [935 and core flow nsial
,, pg w nno e "
Verify all con- Augment pressure Exit to Scram trol rods are Exit to Scram rocedure control v>1th Procedure inserted beyond HPCI and RCIC the MSBWP turbine operation v
Exit to Scram Procedure 176
.s LOSS OF NUCLEAR SERVICE WATER i f, g Progression of Operator Actions:
t j Page 2 of 2-ss, -
Enter from EPC's R9 - RO ,r 80> a r NOTE verify all con. Insert SHM and Place RHR in fuel trol rods fully IRM detectors pool cooling inserted and resem assist tor power decrea-sed to less than 3%
400 er Pn gr BOP w NOTF Control injection Switch recorders Use fuel pool from RCIC from APRM and cooling as a and HPCI to pre- RBM to record temporary heat vent high level IRM sink for RBCCW turbine trip Ann er RO r Rn ,,
Trip or verify Downrange IRM's When all scram auto trip of main to maintain an signals cleared, turbine and trans- on-scale leading bypass SDV high fer of elec. level scram and loads to off-site reset reactor power scram
/ \ AnP u RnP gr %RO ar v
/ Trip RFP's con-densate and con-Start or verify auto start of the Log the event and actions taken densate booster lift pumps, motor pumps suction pump, and turning gear oil pump RO r BOP ., SRO r Trip reactor With chem. lab Notify surper-recirculation approval, init- vision pumps late drywell Nrge to maintain D.W. press. below i [2 psig]
nap v nnp r SRO Isolate or verify' Remove non-essen. Complete scram auto isolation tial RBCCW loads report of reactor water from service cleanup systen w
END NOTE: These are actions that may be specified in plant
, speelfic procedures as an alternate mehtod of operating
/ ,\
RDCCW without service water. Because of simulatnr limit-( ations, this may only be a discussion item.
I i
177 j
_ - _ . ._ _ -_-__ A
Operating Sequence: Loss of Reactor Building Closed Cooling Water Systeu 1 i NSSS/ Type: GE/BWR l V
Initial Plant State: Power. Operations j
Sequence Initiator: Loss of All Operating and Standby RBCCW Pumps Important Plant Parameters: 1) Drywell Temperature, 2) RWCU Non-Regen-erative Heat Exchanger Outlet Temperature, 3) Recirc. Pump Motor Temperatures, 4) Recirc. Pump Seal Cooler Temperatures Progression of Operator Actions: See Flow Chart ;
Final Plant State: Hot Shutdown Major Plant Systems: 1) Reactor Building Closed Cooling Water System.
- 2) Reactor Recirculation, 3) Reactor Water Cleanup System, 4) Drywell Cooling, 5) Control Rod Drive Hydraulics Tolerance Range: The operators should scram the reactor before drywell temperature reaches 135 degrees F or drywell pressure reaches 2 psig.
The reactor recirculation pumps should be tripped within ten minutes of the loss of RBCCW.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals
~( Compliance /Use of Procedures Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures l Control Board Operation l
BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation 179 i
LOSS OF REACTOR BUILDING CLOSED COOLING WATER SYSTEM Progression of Operater Actions: j Page .1 of.2 Low RBCCW Press Alarm BOP o Attempt to restore RBCCW to service RO i, Manually scrarh the reactor when is apparent that drywell cooling cannot be restored enn v Enter RPV Control Procedure i
enn /no E nnr> SRO/RO ar ROP cpn/pn 4 ano Execute " Monitor Execute " Monitor Execute " Monitor and Control RPV and Control RPV and Control Recc-Water Level" sub- Pressure" sub- tor Power" sub- i procedure procedure procedure can u BOP w on ,,
Verify no isola- Verify turbine Place the reactor tion or ECCS bypass valves mode switch in initiations con- are controlling SHUTDOWN ditions exist reactor pressure
"" " RO w Verify all con- Verify the reac-trol rods are Exit to Scram tor recirculation inserted beyond Procedure pumps runback to the MSBWP minimum speed and core flow i
y ., I Exit to Scram Exit to Scram Procedure Procedure O
180
LOSS OF REACTOR BUILDING CLC5ED COOLINO WATER SYSTEM Progression of Operatur Actions:
Page 2 of ?
Enter from EPC's nn u BOP qr BOP gr terify that reac- Trip or verify With chem. lab tor power auto trip of the approval, purge decreases to lest main turbine / gen- the drywell to than 3% erator maintain D.W.
l [2 press.
psig)less than nnp r nnp wr SRO ir After level re- Verify station Log event and turns to normal, electrical load actions taken trip all but one transfer to off-feedwater pump site power nn ur nnp v SRO ,r Shutdown the Start or verify Notify supervisor reactor recir- auto start of the culation pumps lift pumps, motor suction pump, and turning gear oil DumD nnn c noP gr enn ,r Isolate or verif) Verify main con- Complete scram auto isolation of denser vacuum is report reactor water normal cleanup system en ,, ,_ nno ,r ,,
After ensuring Establish cool-all control rods down rate at the END are fully inser- LCO and monitor ted, trip the the cooldown CRD pump (if required) nn ir SRO ,r Request rad.
Downrange IRM's to maintain an chem. lab obtain on-scale reading drywell sanole 181
Operating Sequence: Loss of Off-Site Power
'l 5Q NSSS/ Type: GE/BWR Initial Plant State: Power Operations Sequence Initiator: Load Reject and Loss of All Off-Site Power Supplies Important Plant Parameters: 1) Reactor Water Level, 2) Reactor Pressure,
- 3) Suppression Pool Water. Temperature, 4) RCIC Flow Rate 5) Diesel Generator Loads, 6) Diesel Generator Speeds, 7) Diesel Generator Voltages.
Progression of Operator Actions: See Flow Chart Final Plant State: Hot Shutdown Major Plant Systems: Control Rods, RCIC, RHR (Suppression Pool Cooling Mode), Electrical Distribution, Emergency Diesel Generators, DC Elec-trical System, Standby Gas Treatment l
Tolerance Range: Tec'hnical specification limits on reactor water level, reactor pressure, and suppression pool water temperature should not be exceeded at any time during the event.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Comr.unications/ Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction m
183
LOSS OF OrF. SITE POWER Progression of Operator Actions:
Page 1 of 2 Reactor Scra.
and Isolation 9Pn l' Enter RPV Control Procedure snn/Rn/hnns enn /nn t U nno sRn/Rn/ BOP Execute " Monitor Execute " Monitor Execute " Monitor and Control RPV and Control RPV and Control Reac-Water Level" Pressure" sub- tor Power" sub-subprocedure procedure procedure PO/ BOP U Bgp 'r pn U verify MSIV's Initiate isolatior Place reactor are closed condenser or supp. mode switch in ression pool cool. SHUTDOWN ing Rn/nnP " Rnp 1' F Verify all other Manually open Exit to Scra isolations occur. SRV's in prescrib- Procedure red that should ed sequence to have occurred reduce reactor pressure to (935 0s103 nnP U nno 't Verify emeIgency Augment pressure diesel generators control with started and con- HPCI and RCIC nected to their turbine operation assigned busses nnp " p l Start or verify Exit to Scram auto start of Procedure FPCI nno '
Start or verify auto start of RC1C nno P Restore reactor Exit to Scram water level to Procedure the normal oper-ating band using HPCI and RCIC 184
1 LOSS OF OFF-SITE POWER
-Progression of Operator Actions:
Page 2 of 2
'(O)\/
Enter from EPG's nnp If nnp U can H Place MSIV contro: Verify main turbin? Notify load dis-switches in the DC turning gear patcher of loss CLOSE position oil pump and DC of power and re-seal oil pump are quest power be running restored Rn II cre if can if Verify all contro .
Verify reactor Log events and {
rods fully insert -
feedwater pump DC actions taken ed and reactor lube oil pumps power decreased are running to less than 3%
sap - if Rn U cRn if j Control injection Verify recircula- Complete scram from RCIC and tion pump M-G set report FFCI to prevent DC lube oil pumps high level turbin e are running L. trip. Secure HPCI if not needed Rn i' Anp 1P Rn/Rnp If L Insert SRM and Verify standby When power is l -[
.t IRM detectors if possible gas treatment system running restored, return systems to norm-al operating status for the
's conditions 1
RO l' Anp if If Switch recorders Start or ver.fy from APRM and auto start of END RBM to record control / instrument IRM, if power air compressor is available Rn if nnp if Downrange IRM's As EDG load per-to maintain an mits, place fuel !
on-scale reading pool cooling in service, if needed 1
RO if AnP IF Start a CRD hyd- As EDG load per-raulle pump mits, place addit-ional suppression pool cocrling in ;
service i i
l 185
Op3rsting Stqurnce: Loza of All AC Power (Station B1Eckout) q NSSS/ Type: GE/BWR s
s~ / Initial Plant State: Power Operations Sequence Initiator: Loss of All Off-Site Power with Failure of the Emergency Diesel Generators to Either Start or Connect to Their Respective Emergency Busses Important Plant Parameters: 1) Reactor Water Level, 2) Reactor Vessel Pressure, 3) Suppression Pool Temperature, 4) HPCI System Flow, 5) RCIC System Flow Progression of Operator Actions: See Flow Chart Final Plant State: Hot Shutdown (Possibly Cold Shutdown), with Normal Power Restored, or Power Supplied by Recovered Diesel Generators Major Plant Systems: High Pressure Coolant Injection, Reactor Core Isolation Cooling, Main Steam. Electrical Distribution, RHR (Suppression Pool Cooling Mode), Emergency Diesel Generators Tolerance Range: Operators should maintain reactor water level above the' top of active fuel at all times. Without suppression' pool cooling, there is a possibility that the engineering limit heat capacity tempera-ture limit will be exceeded early in the event. When it is evident that this limit will be exceeded, the operators must depressurize the reactor
/' at a rate greater than the'LCO to preserve primary containment integ-
~( rity. In shutdown, the reactor must be depressurized to less than
'- (typically) 160 psig when the HCTL is exceeded on suppression pool temperature. Also, if the operators depressurize below the isolation setpoints for HPCI and RCIC, the only source of feedwater without AC power will be lost.
Competencies Tested:
SRO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Understanding of Instrument / System Response Compliance /Use of Procedures l Control Board Operation Communications / Crew Interaction BOP - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Technical Specifications fs Compliance /Use of Procedures ,
(s Control Board Operation Communications / Crew Interaction 187
LOSS Or ALL AC POWER (Station Blackout)
Progression of Operator Actions:
Page 1 of 2 Reactor Scram and Isolation l
sen ir Enter HPV Control Procedure BOP ., enn/nn Jr nnn RPn/Rn P AnP cnn/nn 1 Anp Start or verify Execute " Monitor Execute " Monitor Execute " Monitor auto start of and Control RPV and Control RPV and Control Reac-RCIC Water Level" sub. Pressure" sub- tor Power" sub-procedure procedure procedure RnP gr epn q, r pg ,,
Restore reactor Verify MSVI's is Place reactor water level to are closed supp, pool mode switch om normal operatin9 temp. above SHUTDOWN band using HPCI HCTL and RCIC 7 Y
RRn ir enn gr BOD gr BOP 1r gr If supp. pool Verify all other Manually open Manually open temp. high, isolations oc- SRV's to reduce SRV's in pre- Exit to Scram direct aux. oper. RPV press. to scribed seque7ce racedure Curred that to defeat suction should have oc. below HCTL to reduce reattor swap on HPCI and curred pressure to [935 RCIC psig) e nnP/nn ir nnP r Manually isolate Do not depres-Exit to Scram systems that surize to HPCI rocedure failed to auto. and RCIC isola-matically iso- tion late BOP 9r BOP w Determine that Augment pressure EDC's did not control with start or connect HPCI and RCIC i to emergency bus turbine operation BOP 1r y Attempt to start EDG's and/or Exit to Scram connect to emer- Procedure gency bus manually BOP ,r Start or verify auto start of HPCI 188
LOSS OF ALL AC POWER (Station Blackout)
Progression of Operator Actions:
/ % Page 2 of 2 (q)
Enter from EPC's n9p r nno ,r RO w Pluce MaIV con. Verify reactor Switch recorders trol switches in feedwater pump from APRM and the CLOSE DC lube o11 pumps RBM to monitor position are running IRM's f
RO w pn ir RO gr Verify all con- Verify reactor Downrange IRM's trol rods fully recire. pump M-G to maintain an inserted and set DC lube oil on-scale reading reactor power pumps are running decreased to less than 3%
nnp w ExAv1Nen, pn ,
Control injectior CLEAR EDG MAL- Start a CRD from RCIC and FUNCTION DN AT hydraulic pump, HPCI to prevent LEAST DNE DIESEL if possible high level tur-bine trip nop ,r rton v cFn v Maintain greater Place one loop of Notify super-
/"'g VISOf than [2200 RPM] RHR in suppres-l V} turbine speed on HPCI and RCIC sion pool cooling cpn mr DOP 'P*
Direct aux. If poSSiole. Log events and operator to start or verify actions taken locally start auto start of I
EDC's, i f pos- standby gas sible treatment system sp0 v Rap ,r cPn ir Direct load dis- As EDG load per ~ Complete scram patcher to re- mits, place fuel report store off-site pool cooling in power as soon as service possible E40P w on ir RO/ROPir Verify main tur- Insert SRM and When power is b1ne tE turn!ng IRM detectors if restored return gear oil pump power available systems to nor-and DC seal oil mal operating pump are running status for the cinditions I
v
/
/
END L
1 189 l 1
j
PWR ABNORMAL EVENT DESCRIPTIONS i
i i
1 I
I I
i i
1
)
\ l l
ll I
l 191 l u
A v
/ ' l?
M q
! PRESSUR1 ZED-WATER REACTOR ABNORMAL T.NENTS Loss of RCS Makeup Loss of Automatic Pressurizer Pressure Control Failure of Pressurizer Spray Valve F
Loss of Automatic Pressurizer Level Control -
(
D Progressive Failure of No. 1 Seal in RCP Failure of Steam Dump to Open ,, -
Steam Generator Safety Valve Faile Open and Fails to Reseat Steam Generator Level Control Failure High/ Low Dropped Control Rod Inoperable or Stuck Control Rod l
l Inadvertent Boration at Power
/~~'N Inadvertent Dilution at Power !
Failure of N-44 High l Loss of Instrument Air Failure of Turbine to Runback Automatically and Manually Failure of Impulse Pressure Transmitter (Low)
Steam Generator Tube Leak Within Capacity of Charging Pump Loss of Condenser Circulating Pump l
Criticality Outside Expected Band Failure of Loop Temperature Instrumentation High/ Low Loss of One Main Feedwater Fump at High Power Spontaneous Opening of the Ma1n Generator Output Breakers l', '
Loss of RCP Without Reactor Trip Main Steam Leak Inside Containment
(~~
Rupture in Letdown Nonregenerative Heatexchanger to CCW Failure of Pressurizer Control Bank Heaters ;
I 193 l
1
- _ _ _ _ _ _ - _ . .i
,L.
s 1
' Operating Sequence: Lot.s of RCS Makeup NSSS/ Type: hestinghoase/ PWP Initial Plant State: Normal Plant ,0perations 3e SequeacelInitiator: Loss of Charging Plow Control Valve (Fails Closed)
Important Plant Parameters: 1; Charging and Letdow'n Flow, 2) Pressurizer Level, 3) RCP Seal Flow Progression of Operator Actions: See Flow Chart Final Plant State: Charging through seals; etdown via alternate let-down; and pressurizer level in normal band .md stable.
Major Plant Systems: Pressurizer Level Control, Chemical Volume end Con- ,
tril, Excess Letdown, Seal Injection ,
j r
l
-X. Tolerance Range: Pressurizer level in normal range and no reactor trip.
.- S "
Ccapetencies Tested: '
\
K SRO - Compliance /Use cf Procedures Supersluory Ability RO - Under.st&nding/ Interpretation c% Annunciator / Alarm Signals l Diaginsis of Events / Conditions Jased on Signals /Readitzgs '
Control &;rird Operation ~/
et ,
t.
'\
i t. %
t
.k i i c
i
[
f J' j g 6
i k
i s
, 136 s I-
. {. ____ _ _ _ _ _ _ _ _ _ _ _ - _ _ .
i l
vt, l k,
1455 0F RCS MAKEUP Progression of Operator Actionst
- v. i ,
RO RO R0 Analyre Flow cwntrol Take manuel
- Atares velve f ailed
] >
altron ind control bu rd -F closed ->
control of ficw control
/ indications valve h
,\ s -
i RO RO RO RO RO Hatch Adjust flow en a Reduce flow charge / letdown control valve ,Ye.a valve No seal flow No to flow 4- 4---(' control? < 12 gpm < 12 gpm (per pump I +- #
Yes ,
RO y /# < R0 RO y RO Verify seal ik I Verify Initiate Secure flow / L.* 4own Yes ! letdown exceos normal normal -K f(rut ? temperature letdown -> letdown N Ol < st. pt. (per
\f procedure) g _ _ . _ _ . _
4 l valve failed?
I j
SRO RO R0 RO Init iat e Not if y Do es Trip repairs Shift spray No reactor
(- Supervisor valve k close?
q Yes g SRO 1r RO t. ko y RO Y Notify-Record Restore Verify all Trip RCP Management pressure to heaters associated End normal (- energized with failed spray valve A
I RO/SRO RO R0 LIE 0_
Refer to Did Refer to EOP for No plant Yes EOP for
> [ End -
~
te5ctor trip / 4 safety M ety recovery inj.? inj ec t ion
- Al a rms
- Pp Deviation
- Pp Low O
200
.A i Operating Sequence: Loss of Automatic Pressurizer Level Control NSSS/ Type: Westinghouse /PWR Initial Plant State: Any Power Level with Controls in Automatic and Plant Parameters Normal Sequence Initiator: Failure of Level Channel Selected for Automatic Control Important Plant Parameters: 1) Letdown Flow,.2) Pressurizer Level,
- 3) Charging Flow, 4) Pressurizer Heaters Progression of Operator Actions: See Flow Chart Final Plant State: Pressurizer level returned to program level. Letdown lineup and charging flow are normal. Repairs in progress or complete.
Major Plant Systems: Pressurizer Level Control System, Chemical Volume I and Control System )
Tolerance Range: Operator should diagnose the instrument failure and shift control to operable channel. He should reestablish letdown and charging and return level to normal.
( Competencies Tested:
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures RO - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Control Board Operation !
i b(
I l
l 1
)
l 201 l l 1 . _ _ _ _ _ _ _______-___ _ ____-
l LOSS OF ALTOMATIC PRESSURIZER LEVEL CONTROL Progression of Operator Actions:
- Alarms diagnose alternate
> event, observe -> channel level channels for control l
l RO RO RO B0 RO Verify EU Verify Press. Fatie Reenergiz e Verify heaters off pressure High Hi/Lo? Low heaters charging D normal 4 for auto. N decrease to control normal i
I:0 R0 R0 RO Y Verify chrg. Ensure seal Ensure Restore flow incrs. flow to level / press. letdown to restore N RCP's -> return to 4~ per level normal normal procedure l
Rn Y ERO ERD Notify Notif y 16C Verify Shift trip bistables Tech. Specs.
Supervisor N per -> compliance procedures l
FR0 YAC/RO/tRO Y SRO/I&C f
Not if y Return Init iat e End _
plant channel repairs management to service s
- Alarms
- Level Hi/Lo
- Level Trip (1 Channel)
- CVCS O
202
Operating Sequence: Progressive' Failure of No. 1 Seal in RCP
'( NSSS/ Type: Westinghouse /PWR Initial Plant State: RCPs in Service, Normal at Power Operation Sequence Initiator: #1 Seal Failure Such That Leakage Through #1 Seal Is No Longer Controllable Important Plant Parameters: 1) Seal Watsr Leakoff Flow Increasing,
- 2) Seal Water Leakoff Temperature Increasing, 3) High Seal Leakoff Flow Alarm, 4) Thermal Barrier P Low Alarm, b) Pressurizer Low Level Alarm,
- 6) Pressurizer Low Pressure Alarm, 7) VCT Level Decreasing, 8) RCP Bearing Temperature Increasing Progression of Operator Actions: See Flow Chart Final Plant State: Normal plant shutdown if RCP radial bearing tem-peratures remain less than 210 degrees F (secure affected RCP following plant shutdown). Plant is in hot-standby via RCP trip / reactor trip if RCP radial bearing temperatures exceed 210 degrees F.
Major Plant Systems: Reactor Coolant System, CVCS, CCW, Reactor Protec-tion System, Nuclear Instrumentation System Tolerance Range: The operator should be able to diagnose problem from A
! i1 control room indication. He should isolate seal water return. The operator should take appropriate actions listed in procedures - (a) l V monitor Tech. Specs. leakage limits, (b) monitor bearing temperature, (c) trip reactor if necessary, and (d) perform controlled shutdown.
Competencies Tested:
SRO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability l RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Comp,11ance/Use of Procedures Control Board Operation BOP - Compliance /Use of Procedures Control Board Operation NOTE: C-E has a different seal package, d
l 203
___ -- 1
PROGRESSIVE F<.ILL'RE OF NO. 1 SEAL IN (x) RCP
((x) - Picked by Examiner) !
Progression of Operator Actions! i 1
RO RO R0 RO Observe Observe Determine (1) Observe
- Alarms \ sealwater I leakoff >
sealwater which RCP thermal ;
y flow (incr.)
leako f f temperature
- 1 seal is failing
> barrier 6 P alarm (low) i (incr.)
> k , 4 - - - -
l i RO RO RO y Is Inform Shift Observe rate Yes Leakoff Supervisor of pressurizer Con t ru11a ble of problem 4 level
? decrease A J6 No SRO/ROY RO Y RO RO Evaluate Monitor RCP Verify / Monitor Close I . Tech. Spec s . '
radial bearing ,
CCW to sealwater (leakage) temperature r thermal > return Iso.
1rRO/ BOP R0 RO Perform con- Is Manually trolled shut- reactor No trip down per '
tripped reactor procedure (s) ?
Yes ;
i ,
ERO SRO/R0 Y BOP /R0 Notify RCS status Verify 1)Rx i)NRC Cont, status trip, 2)Turb.
- 2) Plant Mgmt. 4 9 Heat sink 4 trip, 3)Cen.
as req'd. by (EO) trip, 4)Stm.
procedure (s) dump ops.(EO) 1 End
- Al a rms
- 1 Seal Leakoff High RCP Standpipe level Alarm High Seal Inject ion Flow (1) 6P will decrease as magnitude increases.
O 204
i A Operating Sequence:-Failure of Steam Dump to Open
'i' NSSS/ Type: Westinghouse. C-E, and B & W/PWR (Capacity may vary).
Initial-Plant State: Reactor / Turbine!>-75% Power, Rod Control in
. Automatic Sequence Initiator: Turbine Runback of 254 - Means Determined by Examiner,-Steam Dump Will Not Open Important Plant Parameters: 1) RCS Parameters, 2) Reactor Power,
- 3) Secondary Power-(Turbine)
Progression of Operator Actions: See Flow Chart. ;
Final Plant State: Reactor power / turbine power matched at lower power level (approximately 25% lower).
Major Plant Systems: RCS, Rod Control, Pressurizer Pressure Control.
Steam Dump System, Turbine and Support Systems Tolerance Range: The operator should maintain plant in safe operating condition - plant should remain critical (the secondary steam generator shrink could cause a. reactor trip'if load rejection is excessive -
verify magnitude the individual plant can handle without reactor trip due to S/G' level)- . The plant parameters should be returned.to normal.
/ Competencies Tested:
4 ).
SRO - Compliance /Use of' Technical Specifications Compliance /Use of Procedures RO - Diagnosis of Events / Conditions Based on Signals / Readings-Compliance /Use of Procedures Control Board Operation i
h l
205
FAILURE OF STEAM DUMP TO OPEN Progression of Operator Actions:
Verify spray Operator Operator Perform E0P valve monitors observes
safety's rod motion 4' tripped?
approx. lifted?
2335# (RCS) l No RO l P R0 $ RO RO RO Manually Is Monitor Verify Ensure dump steam Yes TAVC RCS pressure safeties TAVC approx.
4 > TREF
?
I close A > equal to TREF No No r
SRO/R0 RO RO 90 R0 1P Init ia t e Inform Cond, Vac. Investigate Ensure sec./
repairs as ,
Shift , Circ. water cause of primary plant (cond) Control 4 required Supervisor = dump fallure , st abl e off (checked by RO) iP SRO/R0 SRO Log and Notify f record , plant mf End event 7 management Q
- Alarms Initiator Possible Alarms
- Turbine Runback
- SG Deviation
- Press. Alarm (RCS)
- Par, Level Dev, l
O 206
i
)
s- Operating Sequence: Steam Generator Safety Valve Fails Open and Fails to l e Resent
( NSSS/ Type: Westinghouse, C-E, and B & W/PWR j
Initial Plant State: Steady State Power (954) I Sequence Initiator: Safety Valve Lists Then Fails to Resent (Amount of Leakage as a Function of Severity Is Selected by Examiner)
Important Plant Parameters: 1) Reactor Power, 2) Steam Flow, 3) Feed Flow, 4) RCS Parameters Progression of Operator Actions: See Flow Chart Final Plant State: Plant cooldown in progress with affected steam generator isolated. The plant is at reduced power level with affected l safety gauged.
l
- l. Major Plant Systems: Main Steam, RCS, Steam Dump, RHR. Turbine' Control Tolerance Range: The operator should: identify the event; comply with Tech. Specs.; and follow procedures - (a) shutdown (if required), (b) cooldown (if required), and (c) isolation of faulted steam generator (if required).
!f^ Competencies Tested:
Ik SRO - Compliance /Use of Technical Specifications Co3pliance/Use of Procedures Suprvisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events /Cond.itions Based on Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Control Board Operation l BOP - Compliance /Use of Procedures l Control Board Operation
,\ i j
207
-- o
STEAM CENERATOR SAFETY VALVE FAILS OPEN AND FAILS TO RESEAI Progression of Operster Actions!
RO RO RO BOP /R0 Diagnose Inform Shift (1) 1s Redvce
- Alarms } >
alarms Supervisor RX Pwr
> 100%
Yes Turb/RX Pwr.
Stabilize ouh Attempt to Turb/RX Yes safety be gauge safety power 4 gauged (out side 1 operator)
Ne SRO y SRO/ y RO/ BOP RO/ BOP (2) Tech Specs Commenc e Isolate verif y controlled faulted compliance shutdown p S/C (3)
L SRO [ RO/ BOP RO/B0iv In it ia t e Place RHR Cooldown repairs in service to place 4 RHR in service SRO y log /Repon -
Plant Mang. End NRC when req'd Q
- Alarms Notes
- Steam Flow-Feed Flow Mismatch (1) Concerned about license MW g limit.
- Steam Generator Level Deviation (2) Inoperable safety
- reduce trip setpoints Possible
~ #' "#' "#"
- TAVG-TREF Deviat ion
( " "' "
- &<er Power Rod Stop
- Safety Valve Monitoring System em it ion (if available) Blowdown, O
208
I l
f3 Operating Sequence: Steam Generator Level Control Failure High/ Low N._ / NSSS/ Type: Westinghouse /PWR Initial Plant State: Reactor / Turbine Plant Steady State Steam Generator Level Control and All Control Systems in Automatic Sequence Initiator: Failure High/ Low of Steam Generator Level Trans-mitter for Main Control of Feed Regulation Vvalve Important Plant Parameters: 1) Steam Generator Level, 2) Steam Flow,
- 3) Feed Flow, 4) PZR Pressure Progression of Operator Actions: See Flow Chart Final Plant State: Steam generator level control has been returned to automatic - secondary control channel selected. The reactor / turbine plant is at steady state. The affected channel bistables have been tripped.
Major Plant Systems: Steam Generator Level Control, RCS, Turbine Tolerance Range: The operator should recover the steam generator levels.
He should return the controls to automatic. The reactor / turbine plant should be stable.
,/] Competencies Tested:
SRO - Compliance /Use of Technical Specifications Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Control Board Operation BOP - Control Board Operation NOTE: C-E and B & W control systems are different.
/"N
\s_- )
209 i
______________A
STEAM CENERATOR LEVEL CONTROL TAILURE HICll/ LOW Progression of Operator Actions l 1
- Alarms
- Alarms
- SG Level Deviation
- SC High/ Low 1.evel Trip
- Possible Steam Flow RO/ 1r BOP SRO Y Feed Flow Mismatch l
Ac knowl edge / Evaluate
(*} II""I S I 'CIIIC Y"1"' l diagnose Tech.
I alarms Spers.
RO/ ir BOP SRO / v 16C Take manual Trip affected l
control of channel main feed bistables reg. valve A l BOP / ,l RD SRO g Manually Initiate adjust flow repairs (affected per SC) procedure BOP / 1r RO spa Y Restore SC Not if y level to Plant normal Management (x)%
BOP / yl RO Y Hatch F.nd feed flow to ,'
steam flow h
level control channel on affected SC BOT / RO Place feed reg, valve in auto.
RO T Notify Shift Supervisor 0
210
1
'l l
f-- Operating Sequence: Dropped Control Rod s NSSS/ Type: Westinghouse /PWR Initial. Plant State: Plant in a Steady-State Condition at Power with Rod Control in Automatic l j
i Sequence Initiator: Dropped Rod - Loss of Stationary Gripper Coil Voltage or Mechanical Important Plant Parameters: 1) Reactor Power, 2) RCS Temperature (Tave),
- 3) Pressurizer Pressure, 4) Pressurizer Level, 5) Rod Position / Movement.
- 6) Alarms: (a) rod deviation, (b) rod bottom, (c) various NIS alarms possible, (d) pressurizer level, (e) pressurizer pressure 1
Progression of Operator Actions: See Flow Chart Final Plant State: If. reactor trip occurs, plant will be in hot-standby.
'If reactor remains critical and rod can't be recovered, plant opera-tion may continue p,rovided operation stays within the technical specifi-cations and plant procedures. If dropped rod can be recovered, plant should be returned to the initial power level or such a power level as permitted by technical specifications and plant procedures.
Major Plant Systems: Rod Control, Nuclear Instrumentation System, Reac-tor Coolant System, Chemical and Volume Control, Reactor Protection.
Turbine Control System
- \s 1
Tolerance Range: The operator should: recover dropped rod per procedure; maintain or restore axial flux and quadrant power tilt ratio; and iden-tify dropped rod condition. The shift supervisor should ensure Tech.
Spec. and procedure compliance. -
Competencies Tested:
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures Communications / Crew Interaction RO - Diagnosis of Events / Conditions Based on Signals / Readings -
Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation BOP - Compliance /Use of Procedures Control Board Operation l Procedure Compliance i
l
_('~
211
DROPPED CONTROL ROD Progression of Operator Actions:
RO kn Rn RO Reactor s Nc Id ent if y Inform Shift Verify rod
- Alarms trip? N dropped rod ,
Supervisor ,
motion out (IRPI) F r Returning plant to stable condit ion Yea g 3r RO R0 SRO/RO Reactor trip Rod control \ Ref. to End
\ Entry point E-0 to manual 4
No TAVO= TREF? Yes M
tecb. specs.
axial flux
[ Qptr control rods BOP y DOP/R0 SRO 1r Adjust Stabilize Determine
- Rod bottom / rod drop turbine load plant Qptr 6
- Possible urgent failure ,
g
~
- Possible NIS rate trip F requirements JDPl.Jta _ _.. RO SRO/RO/ BOP Maint/RO/SRO 1FSRO/R0 Reduce over Commence Cause of Determine reactor and No <(x)% rod recovery ,
dropped rod ,
cause of turbine ? 4 per plant ' corrected ' dropped rod power to (x) procedure y (x) = Power Level May Vary From Plant to Plant RO/BOPY RO/B0! SRO/R0 RO RO Stabilise Eusure Permission Select bank Record step plant
~
steady-state from SRO to mode to counter read- l' TAVG= TREF conditions ; commence 7 affected .
> ing for TAVC= TREF recovery bank (1) affected bank RO O RO T RO/SRO Zero pulse Is Reset open all .
I to analog Yes dropped rod affected 2 lif t coils converter in cont. step counter except dropped (2) bank? to zero (2) rod (2)
I &
No RO/ BOP RO y RO RO Change load Rod control Commence Ensure rod as required alarm reset rod is moving to " reset" ; withdrawal y (2)
No O RO RO SRO/RO/ BOP RO BOP Exercise Is Reconn ec t Return rod Adjust affected Yes power at lift co ils to bank posit. turbine load bank M desired -
(3) 4 previously 4 E ntain level? recorded TAVG= TREF SRO V Report and Notes log (1) CE - Individual mode
[ End k . (2) CE - No applicable (3) CE - To group mode O
212 I
I
I i
^, - ~
Operating Sequence: Inoperable or Stuck Control Rod
\
\
s_ / NSSS/ Type: Westinghouse and C-E/PWR Initial Plant State: Plant at Approximately 75% Power, Steady State, Rod Control in Automatic Sequence Initiator: Stuck Rod - Either Inoperable-Due to Mechanical Binding or Rod Control System Malfunction Important Plant Parameters: 1) Nuclear Power, 2) Power Distribution Limits, 3) Turbine Power Progression of Operator Actions: See Flow Chart Final Plant State: Steady state power rod (s) returned to proper align-ment. The plant has been placed in hot standby (turbine shutdown, reactor shutdown).
Major Plant Systems: Rod Control, NIS, Turbine Power / Control !
Tolerance Range: The operator should be able to realign the rod (option-al for examiner). The operator should be able to restore the plant to steady state condition - all parameters normal or plant placed in hot standby condition. During rod misalignment power distribution limits must be verified and appropriate corrective actions taken per technical p specifications, if required.
5 '
/
N- Competencies Tested:
LA0 - Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability RO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operation BOP - Compliance /Use of Procedures l
Control Board Operation G
213 j l
INOPERABLE OR STUCK CONTROL ROD Progression of Operator Actions BOP /R0 nn on SRO/R0 O
Minimiz e Determine Notify Shif t Verify power
- Alarm power change which rod (s) Supervisor dist r ibut ion turb/Rx -
is/are stuck b and Station ; limits satisfy Nuc. Engr. Tech. Spec.
(SNE) limits j, J 670TTr. cation Prior 4 g l to alignment I en v nn nn SRO/R0 Runi en Y Transfer rods Minimize Compare Crp. Rod Determine l to manual rod motion Mstep counter (s) No inop. or how long rod (s) l control >
- to 1RPI/DRPI untrippable have been out 7 of alignment Yes Run 1
0/SRO S"n'""
SNE determine >l rod Take action limits imposed; No misaligned by T'8 y 1 per Tech.
during rod (s) 12 steps Specs. - Inop.)
alignment ? \ untrippable
\
l mF y RO/ BOP ' l RO/SRO T RO/SRO
\
Refer to Per Procedure Will g Det ermine Plant Specifica Attempt to rod (s) gNo shutdown slign rod (s) align? g margin
\
s I*'
<Dn I 1 Yes >
1FRnn/RN'/R0 grBOP/RO/SRO Log / Record & Notify Plant Verify Pwr, Per Procedure report event ,
Management & , dist ribut ion Place plant in NRC J6 limit s hot standby (if req'd) satisfied < 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Tech. Specs.
V End
- Alarms Rod Deviation IRP!/DRPI Failure **
- Plant specific (may be digital or analog position ind.).
- (1) Quadrant Pwr Tilt possible (2) Rod urgent failure possible.
O 214
,_s Operating Sequence: Inadvertent Boration at Power
\
/
'(,,,.) NSSS/ Type: Westinghouse /PWR Initial Plant State: Reactor and Turbine at Power, Rod Control in Automatic Sequence Initiator: Emergency Boration Valve Opened at Power
- 3) Reactor Power Progression of Operator Actions: See Flow Chart l
Final Plant State: Rod Control in Automatic, Tave/ Tref Equal, Emergency Boration Valve Closed.
Major Plant Systems: RCS Makeup, Boron Concentration, RCS Rod Control Tolerance Range: The operator should maintair,and/or restore RCS and pressurizer boron concentration to within 50 ppm of each other. The shift supervisor / operator should ensure compliance with technical specifications - take actions as required. The plant should remain in operation - actions during scenario should leave plant safe and operating. The operator should establish new blend control for auto-matic makeup.
j Competencies Tested:
v SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures RO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operation BOP - Control Board Operation l
l
,em
{ l
\ '
\m.
215
INADVERTENT BORATION AT POWER' Progression of Operator Actions:
Close RWST RWST Verify Verify to charg. pmp. No o charg.pmp. blender primary &
suction euction 4 s et poin t
, , [
- Alarm
' secondary ,
valves closed power j '
matched l, Yes RO "^ Un I "^
Is Verify rods Verify Stop emg. borate N above RIL penalty dilution valve open W points I f accumulated Yes j, ir R0 RO RO RO RO Close emg. Request In Commence Restore borate RCS/ Par 6 flux No dflution Aflux to valve Y ; boron conc. - in target I target band sample band?
Yes ir RO SRO RO Inform Verify Moritor VCT Shift Tech. Spec. level TAVG.
Supervisor > compliance > rod positions SRO ERO SRO R0 1F Log, report Notify plant Investigate Adj ust End
\, record ,
management cause - ,
blender to f , ,
repair if , CB sample
/
required results 1
- Alarms i
Rod Hotion Out Possible TAVE-TREF Alarm O
216
1 l
i
< Operating. Sequence: Inadvertent Dilution at Power NSSS/ Type: Westinghouse /PWR Initial Plant State: Reactor and Turbine at Power, Rod Control in Automatic Sequence Initiator: ' Blender Bypass Valve Open (Manual Valve)
Important' Plant Parameters: 1) Reactor Power, 2) RCS Pressure, 3) RCS-i
. Boron Concentration, 4)'RCS Temperature Progression of-Operator Actions: See Flow Chart
' Final Plant State: Rod control in a'utomatic or manual; Tave-Tref ,
matched; and source of dilution isolated. l Major Plant Systems: RCS Temperature and Pressure, Nuclear Power, Rod Control, RCS Makeup i
Tolerance Range: The operator will be able to prevent reactor trip. !
He should be able to establish new boron concentration and setup of controls for automatic makeup. The operator should be able to take
. required actions listed in the' technical specifications (if rod inser-tion limits are exceeded). He should be able to take appropriate actions to maintain and/or restore boron concentration difference to f less than 50 ppe between RCS and pressurizer.
'- Competencies Tested:
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Control Board Operation I
i 217
_ . _ . . . _ _ ~ _ _ _ . _ _ . _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _
INADVERTENT DILUTION AT POWER Progression of Operator Actions:
0 Auto. rod con-trol inward
- Alarms '
rod motion j E ual rod cont. TAVC &
Pwr increase V R0 R0 RO pn Rod Lo-Lo Lo Stop makeup control Yes od insertion No rod insertio No stop makeup in auto
?
7 limit
?
alarm
?
7N pumps No Yes Yes imn ir un ir Dn ** 1ren/in Manually Emergency Borate per Dispatch insert rode borate per procedure outside to reduce procedure operator Rx pwr.
check valve lineup I . 4 Rn/RRn Dn kn en/ggn pg /gyn qr Log / record Adjust CB Egualize CB Sample Inform Shift report to restore heaters &
, , , RCS & Pzr Supervisor event ,
rods to ,
sprays boron conc. ' Verif y Tech.
maneuvering (pressurizer) Spec, compli.
band SRO y Inform plant management ,[
End
- Alarms ** Examiner Rodd in Auto, Rods Step in RIL Lo/Lo-Lo Results of investigation by operator should yield discovery of open valve Manual TAVG-TREF O
218
Operating Sequence: Failure of N-44 High NSSS/ Type: Westinghouse /PWR Initial Plant State: Automatic Rod Control At Power Sequence Initiator: Malfunction # , Power Range N-44 Fails High with Rod Control in Automatic l Important Plant Parameters: 1) Nuclear Power, 2) Axial Flux, 3) Quadrant l Tilt, 4) RCS Pressure, 5) RCS Temperature 1
Progression of Operator Actions: See Flow Chart Final Plant State: NIS (N-44) out of service; rod control returned to automatic; power operation continued; and bistables tripped (over-temperature delta T and overpower delta T).
Major Plant Systems: NIS, Rod Control, RCS Parameters (Pressure and Temperature)
Tolerance Range: The operator should be able to diagnose casualty. He should also be able to stabilize the plant (RCS parameters in normal range). Channel N-44 should be removed from service. The operator should restore delta flux to normal and refer to technical specifica-tions.
Competencies Tested:
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation BOP - Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction k
210
FAILURE OF N-44 IIICil Progression of Operator Actions:
Monitor ti!S ,,ingl e Did Manually T m power ranges channel No reactor , No trip reactor
- Alarms }
r determine failure trip >
f failed 7 7 channel Yes Yes 4 I BOP /R0 RO RO y RO/ y Sko g/SPD/R0 Use rods Does Place rod Inform Enter trip /
and turbine No TAVGaTREF cc nt rol in Shift Superv. ,
trip recovery control to 7 manual Plant Managmt. r procedure stabilize Ylt.
Yes y sRO/R0 SRO Defeat all Notify 16C functions for to trip End failed A OP AT/OT A1 channel (1) bistables -g l
SRO/R0 Are Yes ech. Spec req'ments mett (2)
No SRO/RO/ BOP SRO P SRO/R0 RO/SRO/ BOP SRO Cont inue Inform Can Take actions Infcrm operation Plant Manya ech. Spec.(s No required by Plant Mang.
4 Record / Log be L.L.O.(s) NRC atisfie as required A Yes A SRO l 1rRO/ BOP SRO/R0 R0/h0P Y Initiate Re t urn In Perform End repsirs y parameters to Plant shutdown per (work request) 4 3 normal (2) shutdown > appropriate (admin.regm'ts) require procedures No 4
- Al arms Notes
- NIS Channel Deviation TI N erpower Rod Stop
- NIS AT1ux Deviation Power Mismatch Circuit
- NIS Single Channel Trip Current Comparator Circuit
- NIS Positive Rate Single Channel Trip Channel Comparator Circuit
- Possible Rod Insertion Limit Alarms Control Power Tunes Removed (2) Rod Insertion I.imit s A Flux Quadrant Power 9
220
fs Operating Sequence: Loss of Instrument Air i
\ ,/ NSSS/ Type: Westinghouse /PWR Initial Plant State: Loss of Instrument Air, Control Systems in Automatic-Sequence Initiator: Instrument Air Break (Isolable if Possible) 1 J
Important Plant Parameters: 1) Instrument Air System, 2) Feedwater i Control. 3) Chemical Volume and Control. 4) RCS Pressure Control 5) RCS Pressurizer Level Control, 6) Main Steam Progression of Operator Actions: See Flow Chart ;
i Final Plant State: Either reactor / secondary steady state with leak isoleted or the reactor in a hot standby condition.
Major Plant Systems: RCS Pressurizer Pressure / Level Control, Feedwater System, Chemical Volume and Control, Instrument Air System Tolerance Range: The operator should save plant if leak is isolable.
If not isolable, the operator should maintain plant in safe and stable condition with reactor shutdown (hot standby).
Competencies Tested:
f
/
SRO - Compliance /Use of Technical Specifications
\~ ' Compliance /Use of Procedures RO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operation BOP - Control Board Operation NOTE: Each plant architectural engineer will have their own design, i /
\
%/
221
LOSS OF INSTRUMENT AIR e
Progression of Operator Actions:
R0/A0 RO/ BOP BOP /A0 RO/ BOP g Attempt to Verify inst. Check air isolate air / serv air dryer '** Is compressor .\.
- ~"- *
- Alarms leak & isol. X-connect n'rmal *~ running? ' ,
g shut -w,..m."
/
[
l_
No !
RO i RO DCP/R0 o RO/ BOP Verify N2 Can Are Start pressure to S/C leve Yes MSIVs standby Pzr. PORVs be main- closed? -. -
compressor (s) i t ained ? ,'
No C Ies Nn RO RO p R0 Use E0P Manually ormal for reactor trip the chrg & Yes trip e reactor letdown?
50 ',
DNo I
\
i U an P Maintain VCT Initiate i- Verify seal End Ivl be in normal manual flow maintained band *-- primary -> normal
? makeup it 4,
Yes SRO RO/SRO RO SRO/R0 RO Repott to Report and Restore Initiate Notify plant log event systems to maintenance Shift management % % normal % on air sys. Supervisor
(;
- Alarms
- Air Pressure Lov
- Compressor Trouble
- Compressor Auto. Start (S t and by)
- Service Air Isolailon O
222
--_--t__-. y i n,
\ \
i Operating Sequence: Faibtro ' '
of Turbine to Runback Automatically v.ud Manually - (
( ' v
\
O 3.\9
'NSSS/ Type: Westin(no9se.C-E,andB&W/PWR) i \ l , 563ctor and Turbith Conttvls in Automatic, Plant at
\,
t Initial Plant State: '
Steady State (Approrfaately 100%) p, 4 i
c '
L
( Sequence Initiator: Insert Condition that Anquires Turbine Runback e
(Overtemperature as Time Changes, Overpres'aare as Time Chaves, etc.)
Important Plant Patameteru: 1)'KCS Tdaperature and Pressure, 2) Reactor j and Turbine Power
\
l Progression of Opeintor Actions: is e s ? low Caart s i Final Plant State: Reactor in hot *uand5y. cutdition or reactor s innt stable below setpoint for runbac~r conti tion.
1 ,
Major Plant Systems: Turbine Control, lleactor'O.sntrol, Reactor hotec-
' t 3
[1fon '
I
! 'V,lerance Range: The operator should h cognizd'the event. %nuai tr tp .
- ,f the reactor, with runback condition and no automatic hid'manusl run-back. t 5
Competencies Tested:
{
t Understanding of Instrument /Systda Response Compliance /Use of Procedures I f
Control Board Operation p
1 j
BOP - Complitnca/W e of Procedures (
Cont, o? Rcot1 Operation
\
,L l
i E,'
s c,
y i
223 1 , (
\
s
L.
.s l
FATLURE'0F TURBINE TO RUNBACK AUTOMATICALLY AND MANUALLY s Progression of Operator Actionst t
RO RO/ BOP RO BOP Diagnose Turbine dfd Inform Shif t Atten.pt to alarma not runback l Supervisor manually
,.
- Alarms i
.\
j r r
l -@ > runback turbine 4
h(, .
RO V Investigate es Turbine does
- Alarms cause of No runback not runback
- Overt emp. g. or runbatN 4 condition '1
- Overpower AT faile e still
- Other Plant Specific 18t Alarms Which Denote Runback Conditions Yes
/ Initiate Manually
, 3 .refah trip the reactor j
Log / Report Enter Plant E-O Manager reactor trip v
W End s '
O 224 C _ _ _ _ _ _ _
Operating Sequence: Failure of Impulse Pressure Transmitter (Low) i
) NSSS/ Type: Westinghouse /PWR Initial Plant State: Steady State Power, Control Systems in Automatic )
Sequence Initiator: Impulse Pressure Transmitter Feeding Rod Control Fails Low Important Plant Parameters: I) RCS Temperature, 2) Tref, 3) Nuclear Power, 4) PZR Level, 5) PZR Pressure, 6) Delta Flux i
Progression of Operator Actions: See Flow Chart Final Plant State: Rod control in manual; Tave = Tref; plant in steady state; rods returned to greater than RIL lo or lo-Ll; and repairs under-way or complete.
Major Plant Systems:, Rod Control, Steam Dump, Nuclear Instrumentation, Turbine Control, RCS Tolerance Range: The operator should identify the failed channel, stop the automatic rod motion, maintain rods above RIL (final condition), and maintain / restore primary parameters at/to normal.
Competencies Tested:
['
\,_,)
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures
/
Supervisory Ability RO - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation BOP - Diagnosis of Events / Conditions Based on Signals / Readings Control Board Operation NOTE: C-E control systems differ significantly so that to make this description generJc was not possible.
,f l
v) l l l
- I 225 l l 1 _. ._ ._________-_a l
FAILURE OF IMPULSE PRESSURE TRANSMITTER (LOW)
Progression of Operator Actions:
- Alarms alarms
] >
control in
> mnual RO r an Are Emerg.
rods > No borate per RIL I' plant to-Lo? procedure I
Yes 4 RO P RO R0 Manually re Adjust rods adjust rode rods > to > RIL Restore D RIL b Rods in TAVC= TREF Lo? manual Yes I
, 4 BOP RO Adjust turb. Is load to No TAVG= TREF match ?
TAVC/ TREF
l Is Yes Inform Shift Verify Tech. Initiat e O I in Supervisor y Specs. p repairs l
Band **
4 satisfied (1)
?
l No A l
i BOP y R0 SRO y Adjust Pwr Report / Log level to restore AI -
End 1P Plant Management J
- Alarms (1) Rod Insertion Limits (RIL)
- TAVG-TREF Mismatch Delta Flux ( AT)
, - Low Imp. Pressure / Rod Withdrawal Block Reactor Protection Instrumentation
! - SC High/ Low Deviat io,-
l - SG 1-4 Loop High Stm. Iaw (Channel Alerts)
- Tech. Spec. Limit O
226
j Operating Sequence: Steam Generator Tube Leak Within capacity of
,,,\ Charging Pump l4/
NSSS/ Type: Westinghouse, C-E, and B & W/PWR-Initial Plant State: At Steady State Power Sequence Initiator: Primary to Secondary Leak (Rate Determined by Examiner) Within capacity of Charging Pump (s)
Important Plant Parameters: 1) Radiation Monitoring, 2) Pressurizer Level, 3) Charging Flow / Letdown, 4) RCS Makeup Progression of Operator Actions: See Flow Chart l' ' Final Plant State: The affected steam generator has been isolated.
l Plant cooldown is in progress. EPP may have been implemented l Major Plant Systems: RMS, RCS, CVCS, SG Blowdown, Auxiliary / Normal Feed Tolerance Range: The operator should be able to identify casualty and the faulted generator. He should minimize the release'to the atmosphere and meet technical specification requirements.
Competencies Tested:
- SRO - Compliance /Use of Technical Specifications g
Compliance /Use of Procedures N. Supervisory Ability Communications / Crew Interaction RO - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation BOP - Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation I
l l
j 227 l
1
_ _______o
STEAM CENERATOR TUBE LEAK WITHIN CAPACITY OF CHARCINC PUMP Progression of Operator Actions:
Verify Is Start add.
- Alarm
] >
steam gen, blowdown b ddit ional chrg. req'd Yes chtg. pump to restore M isolates ? (1) Pzr tv1
- Alarms 2 I SG. Blowdown Rms Main Steamline Ras ,
[ RO "
Cond. Vac. Pmp/ Air Ej ector Rms s isolate Pzr. Level Deviation letdown Yes normal isolat ion ' letdown i req'd? I 2)
No a !
RO q [R0 Stabilize Ensure Pzr level makeup 4 available to VCT I
SRO k Chem RO/B01 SRO/R0 SRO/R BOP /R0 Sample Perform leak Determine ri, to Continue all steam rate faulted sec. leak No operation gen. for su rve 111ance 4 steam
~
> Tech. I activity generator Spect Y'"
I & I RO/ BOP B0P BOP R0 q, BOP SRO y Close MSIV Adjust When turb. is Reactor /turb. Calculate on affected affected SC. off line plant shut- release SG. 2 PORV to value 2 per '
a down per _p
> no load procedure procedure press. (3)
RO y BOP BOP /R0 BOP SRO/RO/ BOP SRO RO Verify SC Lvl a Verify Commence A Is j Isolate Feed t eam t o No affected steam plant rool- emergency to affected Aux F.P. gen. blowdown p down and de- plan A4 p -
steam gen, turb from is isolated pressurization req'd?
ecced o.G. per procedure Yes BOP / 1r RO SR0 SRO SRO Y E0 1solate Stm. Init iat e/ No t ify Initlate supply from schedule NRC EPP per affected S.G. repairs 4 Plant Mang. ( procedure to Aux F.P.
Turb Notes RO y N (1) Some plants may have auto, start Img/ Record of second chrg. pmp on low level in Pzr.
as req'd by End )
procedure - -
(2) May be reg'd to initially restore Pzr l evel .
(3) Refer to plant procedure for specific number.
O 228
Operating Sequence: Loss of Condenser Circulating Pump p,
/ i
) NSSS/ Type: Westinghouse, C-E, and B & W/PWR Initial Plant State: Plant at Approximately 75% Power - Steady State, Control Systems in Automatic Sequence Initiator: Loss of Condenser Circulating Pump Important Plant Parameters: 1) Circulating Water Flow, 2) Condenser Vacuum, 3) Turbine System (Exhaust Hood Temperature /Possibly Load)
Progression of Operator Actions: See Flow Chart Final Plant State: Turbine load is slightly reduced; one circulating water pump is secured; and exhaust hood sprays are in service (if required).
Major Plant Systems: Condenser Circulating Water, Turbine Vacuum Tolerance Range: Reactor power / secondary power is stable and matched.
Competencies Tested:
SRO - Compliance /Use of Procedures RO - Diagnosis of Events / Conditions Based on Signals / Readings
,-y Compliance /Use of Procedures G 'j l
BOP - Compliance /Use of Procedures Control Board Operation i
iO s 1
1
'b 229
LOSS OF CONDENSER CIRCULATING PLHP Progression of Operator Actions
u BOP /R0 RO/ BOP RO/SRO/ BOP Is is Refer Cond. Yes reactor Yes E0P reactor vacuum turbine trip decreas? tripped No No BOP BOP y BOP /RO P BOP /R0 Initiate Monitor Decrease Stabilize exhaust hood turb. exhausc turbine load reactor /
sprays per e-- hood as required --> turbine procedure temperature power a i RO SRO SRO/RO/ BOP SRO Notify Initiate log / Inform Shift Super- 'repaits record plant End visor % management
- Ala rms Peeder Bkr. Trip Condenser Vacuum low O
230
Operating Sequence: Criticality Outside Expected Band NSSS/ Type.: Westinghouse /PWR Initial-Plant State: . Reactor Shutdown, Prepared to Take' Reactor Criti-cal, ECP/ECC Calculated Sequence Initiator: Criticality Not Achieved Within Expected Bank
+/- pcm of ECP/ECC Important Plant Parameters: 1) Rod Position, 2) Boron Concentration,
- 3) Reactor Power Progression of Operator Actions: See. Flow Chart Final Plant State: Reactor subcritical; rod inserted; and ECP/ECC recal-culation complete.
Major Plant Systems: Rod Control, Boron Concentration, RCS Tolerance Range: The operator must follow the approved procedure for taking the reactor critica). The reactor should not be maintained critical below zero power rod insertion limit. The operator should be able to' recognize criticality.- Appropriate actions (per procedure) must be taken when critical outside expected band.
Competencies Tested:
' SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures RO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operations 231
CRITICALITY OUTSIDE EXPECTED BAND Progression of Operator Actions:
RO/SRO RO Recalculate Take ECC/ECP reactor Alarm
-> critical per 4 procedure h
evnfrn go N.
14g, report is record Yes reactot er tical Rit' A No un/eq RO Not ify - Is Rx Engineer Yes reactor
& Shilt i N critical End Supervisor \500 pcv/
belot- ECC?
4 No RO l R0 y RO/SRO SRO Insert ,/ Ia With Rx ' Log, report control rods y,,,/ reactor 'N No c rit ical record c r it ical 'A--> continue ->
500 pcm/ S/U per procedure abo'vuQCP/1.CC?
O O
232
Operating Sequence: Failure of' Loop Temperature Instrumentation High/ Low
\ )
U NSSS/ Type: Westinghouse /PWR Initial Plant State: Reactor Controls in Automatic Power Level at About 75%, All Other Control Systems in Automatic j Sequence Initiator:' Loop (X) Hot Leg RTD (Narrow Range) Falls High/ Low Important Plant Parameters: 1) RCS Temperature / Pressure, 2) Reactor Power, 3) PZR Level, 4) Rod Position Progression of Operator Actions: See Flow Chart
' Final Plant State: The reactor / turbine plant is at steady state. The temperature defeat switches (delta T and Tave) in loop (x) are defeated.
The affected loop bistables for overtemperature/ overpower delta T have been placed in.the. tripped condition.
Major' Plant Systems: Rod Control, Reactor Proection and Control, RCS Tolerance Range: The reactor / turbine plant is stable. The operator must place the rods in manual to mitigate the casualty. The bistables should-be placed in the tripped position; the loop Tave and delta T inputs should be defeated.
j'] Competencies Tested:
SRO - Compliance /Use of Technical Specifications Supervisory Ability.
RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Control Board Operation BOP - Control Board Operation NOTE: Most C-E units have similar system response, but operator response and corrective actions are different.
l 233
FAILURE OF LOOP TEMPERATURE INSTRUMENTATION HICH/thW Progression of Operator Actions:
i R0 RO RO BOP /RO Acknowledge Place rod Restore Stabilize diagnose control to TAVE-TREF reactor /
- Alarms alarms
- manual *
Refer to TAVG & AT Inform Shift tech. specs, to defeat Supervisor Rx prot. - Defeated on -._-
channels Mn. Cont.
Brd.
- RCS TREF-TAUCT Hi/lo OT. AT control to
- RCS A T Deviation --- bfstables .----a- automatic
- TAVG-TREF Deviation
- OT AT Trip Alert
- OPd T Trip Alert (2) Failure Low (TH SRO SRO p 16C
- RCS AT Deviation
- RCS TREF-TAUCT Hi/La Not if y Initiate
- Possible Rod Insertion Plant repairs e
Limit Alarm End Managmt.
- Report /
per Record I procedure I
i 234
,e's Operating Sequence: Loss of One Main Feedwater Pump at High Power l \
'n NSSS/ Type: Westinghouse /PWR Initial Plant State: Normal Operation - Power >75%, Both Main Feed Pumps in Service Sequence Initiator: Main Feed Pump Trip Important Plant Parameters: 1) Steam Ger.erator Levels, 2) RCS Parameters Progression of Operator Actions: See Flow Chart Final Plant State: Reactor power is reduced; steam flow and feed flow are matched; steam generator levels are in normal band; and there may be a possible reactor trip (plant dependent).
Major Plant Systems: Feedwater, Steam Generator Level Control, Auxiliary Feed, Turbine Control (plant dependent)
Tolerance Range: The steam generator level is in program band. The j primary / secondary plants are stable. Reactor trip is not expected on units with automatic runback capability. Reactor trip is possible on other units.
Competencies Tested:
,e i,
) SP.0 - Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals i Diagnosis of Events / Conditions Based On Signals / Readings !
Compliance /Use of Procedures Control Board Operation BOP - Control Board Operation
,r \
k )
v 235
LOSS OF ONE MAIN TEEDWATER PUMP AT HICH POWER Progression of Operator Actions:
h T RO Observe and Verify feed )
trip \. Yes
~N Rx trip runback runback / 7 I ?,/
l Yes No No m VI nn'/RnP RO/ ROP nn/Rnp nnp T Reactor & Verify S/C Verify Ws Manually secondary level in approx. = runback End plant e program 4- Vf 41 t u rbine stable band I
l RO V RO/ BOP RO/ BOP RO SRO Secure Verify Verify Inform In it ia t e auxiliary We approx. = reactor /sec. Shift repairs feed pumps g Wf _p plants _ _p Supervisor ._.__g
- stable l
2 SRO l 9
Log and End report as required
- Al a rms
- Pump Trip
- Pump Trouble
- S/C Level 0
236
Operating Sequence: Spontaneous Opening of the Main Generator Output Breakers NSSS/ Type: Westinghouse, C-E, and B & W/PWR Initial Plant State: Dependent Upon Steam Dump Capacity for Plant:
85% Dumps - Power Level at Approximately 95%; 40% Dumps - Power Level at Approximately 50%.
Sequence Initiator: Both Generator Output Breakers Opened " Manually" Important Plant Parameters: 1) RCS Temperature and Pressure, 2) Turbine Load, 3) Nuclear Power, 4) Steam Generator Level Progression of Operator Actions: See Flow Chart Final Plant State: The reactor critical power level is approximately =
1% and Tave is at no-load value. When the plant stabilizes, the steam dump goes to " pressure mode", rod control to manual (approx. 15%), and feedwater to start up feedwater supply.
Major Plant Systems: Steam Dump, Rod Control, Steam Generator Level Control, Turbine Control Tolerance Range: The operator should diagnose the problem. He should ensure that the automatic systems, i.e., rod control, steam dump, turbine control, steam generator water level control, function properly.
The operator should also transfer the steam dump to " pressure mode" following transient and rod control to manual. He should align the startup feedwater supply to steam generators.
f Competencies Tested:
t SRO - Compliance /Use of Procedures Supervisory Ability RO - Diagnosis of Events / Conditions Based On Signals / Readings Understanding of Instrument / System Response Compliance /Use of Precedures Control Board Operation BOP - Compliance /Use of Procedures Control Board Operation NOTE: Generator output breakers must be manually opened to prevent generator / turbine trip which will result in reactor trip.
237
SPONTANEOUS OPEN7NC OF THE MAIN CENLRATOR OLTPUT BREAKERS Progression of Operator Actions:
R0 R0 Diagnose las Enter
- Alarms
] >
alarms / events l-reactor tripped Yes b
E-0 reactor trip
[End No BOP /R0lf RO/ BOP lave Aqtuate**
- Expset Rx trip if dump f ailed dumps No dumps to actuate automatically. actuated ' manually
?
\
Yes m l BOP 1r R0 B0P Verify Verify OPC (turbime) dump valves inward rod functions to open (1) > motion y shut control valves (2)
BOP /R0 SRO RO R0 RO 9P BOP Transfer Init iat e inform Verify RCS Verify SC Stm dump investigation Shift has level control to press. 4 Determine 4 Supervisor 4 stabilized 4 functions mode (3) cause properly R0 1P RO/ BOP SRO SRO/R0/ BOP Rod control Ensure Initiate Ing/ Report to manual feedwater repairs per _m Plant Mang. [ End
"(
r r procedures r (3) to S.C.s (4)
- Alarms (1) Number of valves req'd to open (plant specific).
- Gen. Trouble Alarm (s)
- Cen. Output Brkr(s) open (2) OPC - Overspeed Protection Control.
- Stm. Dump Actuation
- Stm. Gen. Level Deviation (3) Plant stable at about no-load value.
- Pzt Level Deviation
- Pzr Press. Deviation (4) Auxiliary feedwater or startup feedwater .
(plant specific startup supply of feedwater). l 1
0 238
l l
I i
f-~s Operating Sequence: Loss of RCP Without Reactor Trip
\s ') NSSS/ Type: Westinghouse and C-E/PWR Initial Plant State: Reactor / Turbine Plant at Approximately 20%, All Control Systems in Automatic Sequence Initiator: Malfunction Causes Loss of One RCP Important Plant Parameters: 1) RCS Flow, 2) RCS Temperature, 3) RCS Pressure, 4) Steam Generator Level Progression of Operator Actions: See Flow Chart Final Plant State: The plant is in hot standby; RCP is restarted; and the plant is ready for reactor startup. (Items 2 and 3 are at the discretion of the examiner.)
Major Plant Systems: RCS, Steam Generator, Turbine Tolerance Range: The operator should stabilize the reactor / turbine plant. He should perform shutdown of the reactor and turbine plant, and prepare to start the faulted RCP after the cause of the failure has been determined. At the discretion of the examiner, the operator should prepare for and conduct reactor startup.
() Competencies Tested:
' ~ '
SRO - Compliance /Use of Technical Specifications Supervisory Ability RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operation BOP - Compliance /Use of Procedures Control Board Operation
/-
L.)
239
LOSS OF RCP WITHOUT REACTOR TRIP Progression of Operator Actions:
O
- Alarms r
- Alarms
/ - RCS Loop low Flow
- TAVE - TAUCT Deviation
- RCS Loop 6T Deviation II*"' #" "
RO 1r - PZR Level Deviat ion Acknowledge alarms diagnose (1) Plant must be in hot
. standby (RX subcritical).
RO/ BOP v
Stabilite RX/ Turbine plant RO y Not if y Shift Supervisor BOP / 3r RD SRO l SRO y Commence Tech. spec. Obtain plant S/D for loops pe rmiss ion per operable for RX procedure startup 1 m
" l SRO/R0 JL RO/B04 init iat e Perform RX inves t igat ion startup into cause per procedure i
SRO/P1 nt Support V
Initiate repairs of End RCP h
RO 1r Start RCP per (1) procedure SRO y l Notify Plant management I m 240
i 1
l Operating Sequence: Main Steam Leak Inside Containment NSSS/ Type: Westinghouse, C-E, and B & W/PWR Initial Plant State: Steady-State Power. Level at Approximately 90%
Sequence Initiator: Small Steam Break Inside Containment - Leak Rate
-Will Have to be Slow Enough.to Raise Humidity First, (The Percentage of Break'is Dependent Upon the Examiner.)
j Important Plant Parameters: 1) Containment Humidity, Temperature and Pressure, 2) Reactor / Turbine Power, 3) Steam and Feed Flow Progression of Operator Actions: Se3 Flow Chart Final Plant State: 1) Operation continues within technical specification j guidelines, or 2) plant shutdown with cooldown in progress.
Major Plant Systems: Containment - temperature, pressure, humidity, and cooling, Main Steam Tolerance Range: The operator'should identify the steam leak'as non-
' radioactive. He should comply with technical-specifications. The operator should also comply with procedures - plant operation to con- ,
tinue or shutdown and cooldown. '
Competencies Tested:
O' SRO-- Diagnosis of Events / Conditions Based On Signals / Readings Compliance /Use of Technical Specifications Compliance /Use of Procedures Supervisory Ability RO - Understand / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based On' Signals / Readings Understanding of Instrument / System Response Control Board Operation BOP - Control Board Operation
[ t 241
MAIN STEAM LEAK INSIDE CONTAINMENT Progression of Operator Actions:
RO BOP /R0 RO RO Inform Shift Check con- Start Perform leak g Supervisor tainment temp. additional rate
- Alarms i
- y pressure & p containment y surveillance j humid it y cooling fans test RO/ BOP RO/ BOP BOP /R0 BOP /R0 RO/B0fi r Identify Chec k Compa r e Check for Check (1) affected S.C. Wa-Wf in reactor 6 recent sump rad iat ion (2) 2 each steam a turbine = pump operat. 4 monitoring generator power (containment) (containment) 1r RO/SRO SRO SRO/R0 SRO SRO/RO/ BOP Evaluate Refer to Evaluate f perati Shutdown /
need for Tech. Specs. continued / cooldown per
, to No containment r (containment) r operation continue? procedure entry Yes l
SRO SRO P VSRog Plan repairs Inform Log / Report manpa ,en t Plant Mang.
A SRO l iP SRO Y Log / Report Inspect ion of P' ant main st eam management lines (direct)
Maint.Y SFO V Init la t e )
End \ repairs F
- Alarma Notes Containment - Humidity (1) Notify Health Physics for airborna sample
- Temperature (containment atmosphere).
- Pressure (2) W,-Wf of affected S.G. is > W,-Wg of unaffected S.C.'s.
O l
242 I
1 L
l Operating Sequence: Rupture in Letdown Nonregenerative Heat Exchanger I
to CCW -
NSSS/ Type: Westinghouse /PWR Initial Plant State: Steady State Power Sequence Initiator: Primary to CCW Leak (Rate Determined by Examiner)
Important Plant Parameters: 1) Radiation Process Monitoring, 2) CCW Surge Tank Level, 3) PZH Level, 4) RCS Makeup, 5) Charging and Letdown-Flow.
Progression of Operator Actions: See Flow Chart Final Plant State: Normal letdown / charging is secured; excess letdown in p service; charging / makeup via seals; and CCW to letdown nonregenerative heat exchanger is secured.
Major Plant Systems: CVCS, Process Radiation Monitoring, RCS, CCW 1
Tolerance Range: The operator should identify the leak and the source.
l He should isolate the source and maintain PZR level at normal and the ._
plant at steady state.
I Competencies Tested:
I i SRO - Compliance /Use of Technical Specifications l
Compliance /Uae of Procedures Supervisory Ability RO - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Control Board Operation BOP - Control Board Operation i
O 243
RUPTURE IN LETCOWN NONRECENERATIVE HEATEXCHANGER TO CCW Progression of Operator Actions:
Alarms *
- Alarms
- CCW Radiation Monitor
- CCW Surge Tk. Level High
- Possible PZR Level Deviation RO qy Acknowledge alarms RO 3r Perform leak isolation procedure RO ,
Isolate normal charge and letdown RO y Place excess letdown in service RO tr Adj. Charge to restore par. level (N/A seals)
BOP / qr RO BOP 1r Perform Isolate leaktate CCW to surveillance NRHX RO q, Notify SRO Mgmt. < l SRO V Notify Rad Chem Sample CCW SRO 1r Schedule repairs p{ End 244
5 b
Operating Sequence: Failure of Pressurizer Control Bank Heaters
]
i
> sc :j l NSSS/ Type: Westinghouse /PWR
, Initial Plant State: Power Operation j Sequence Initiator: Failure of Control Bank Heaters Important Plant Parameters: Pressurizer Pressure Progression of Operator Actions: See Flow Chart j J
Final Plant State: Plant at steady state; pressurizer / plant pressure- J
. normal; and heaters (B/U) cycling to maintain pressure.
Major Plant Systems: RCS Pressure Tolerance Range: The operator should be able to maintain reactor criti-cal. The pressure should be restored to normal. The shift supervisor /
l reactor operator should be able to satisfy technical specifications ]
(press.~/DNBR limit). )
l' Competencies Evaluated:
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures i / RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings 1 Control Board Operation l
i i
)
4 1
i ,
_ i-I
\
i 245 .
FAILURE OF PRESSURIZER CONTROL BANK HEATERS Progression of Operator Ar.tions :
R0 R0 RO O
_ Verify Verify Verify B/U
- Alarms \ m Par. press. .
F control htra, m heaters low are fa11ed energized (not on)
RO R0 Did Manually B/U Htts. , No energize energize? heaters Yes 4
un SRO RO RO Init iat e Verify Inform Monitor **
repairs 2 press.) 2 Shift 2 pressure 16C/Elec. Tech. Spec. Supervisor increase to DNBR limit approx.
2225#
RO/SRQI Log, report record , [ End To management etc.
- Alarms O
Pressurizer Pressure Low Deviation
- Pressure Control on B/U Heaters may be at approximately 22238 (or set point / bandwidth) normal approximately 2235#.
O 246
-: (
x 1
l PWR ENERGENCY EVENT DESCRIPTIONS O
l O ,
247
,,-~~ ,
(
\'- PRESSURIZED-WATER REACTOR EMERGENCY EVENTS Reactor Trip Large Break LOCA - Reactor Trip With Safety injection PZR/PORV Failure to Open Steam Generator Tube Rupture Failure of Main Turbine to Trip Small Break Loss of Coolant Accident Anticipated Transient Without Scram Loss of Auxiliary Feedwater - Inadequate Core Cooling Loss of Off-Site Power Station Blackout - Loss of All AC Power
,-s Control Room Fire Requiring Evacuation
/
k, Main Steam Break Inside Containment RHR LOCA - Complete Loss of RHR s'%.
v
)
249
, . ~ .
[ \ Operating Sequence: Reactor Trip .
\ ]
NSSS/ Type: Westinghouse and C-E/PWR Initial Plant State: Power Level Above That Required to Have Control Systems in Automatic (About 20%) All Systems in Automatic Sequence Initiator: Reactor Trip - Cause Dependent on Examiner ;
Important Plant Parameters: 1) Nuclear Power, 2) Turbine Status, 3)
Emergency AC Power, 4) RCS Temperature, 5) PZR Pressure / Level, 6) Steam 4 Generator Level, 7) RCS Flow Progression of Operator Actions: See Flow Chart Final Plant State: The RCS temperature and pressurizer pressure / level are normal. The reactor power is in source range and the electrical shutdown lineup is normal.
Major Plant Systems: RCS, NIS, Rod Control, Auxiliary Feed, Turbine Control, Steam Dump, AC Power Tolerance Range: The operator should identify the reactor trip and the turbine trip. He should verify subcriticality and heat sink. The operator should also comply with procedures.
,-g Competencies Tested:
(
s
)
SRO - Compliance /Use of Procedures Supervisory Ability RO - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation x_-) ,
251
,S 1
- i N l,
s.,
- k. .
REACTOR TRIP C l Progression of Operator Actions:
1 5
I I Ii RO/ BOP R0/ BOP R0 BOY /R0 Verify trip / Is Manually Verify condition (s) # Reactor No ' trip rfactor trip
- Alarms tripped * \ '(1),
? (1)
Yes _
i RO/ BOP RO/ BOP BOP O BOP BOY Verify RCS Verify Is Itauu.111y heat removal oYmelfv emergency A.C. Yes turbine No trip ' tprp.:e (3) +-- busses *- tripped turbine *
" trip (2) energized ? (2) k' (
l RO o RO Is0 BOP /R0 RO/ BOP Verify RCS Verify Pzr Verify Pzr Verify S.C. Verify all temp. decrease level control press.
to No-Load * "
- levels (6) A.C.' losars s (4) control (5)
- energ %.d T-Avg. '
l
_. . J BOP /R0 BOP /R0/Outplant R0 R0 BOP y 3--
Maintain Shutdown Verify Verify one Transfer stable plant unnec essary riurce rug. RCP running atm. dump to conditions (8) % equip. -. r:nstgized (7) - press. mode SRO ir NR eac r Trip Plant Mang. - First Out Annunciator Notes ;
(1) Rods on Bottom
- DecreasinR neutron level ,k
- Rx trip 6' bypass brkrs open g (2) Turbine stop valves closed '
- Turbine speed decreasing '
'I bd (3) RCS temperature > PW Isol. Temp.
- Main IV control valves closed
- Total IV flow > (x)
- Rods on bottcm (4) Par level > 1etiow Iso. Inl.
- Chrg. & Ltds in serviu (5) Pzr press >SI StPt.
- Par press, trending to normal (x)
(6) In narrow range band
- Controlling to maintain level between (x) and 50%
(7) 1.R.< P-6 value, see plant specific value
- Verify $.R. energized
- NIS recot h r to S.R.
(8) Pzr level / press (x)
- A.C. level (x)
- RCS temp. (x) where (x) is a plant specific value.
O 252
, , e .
l l_ '
l
.t
[
<p p ,y
'y ' ,' o q
- OperatingSequence
- 'LargeBreakLOCA-E9actorgripWithSafety
(- 4
.Jnjection y .
,, 3 '
\
,5 I,Y N
NSSS/ Type:, Westinghouseiand C-E/PWR
't i, Initial Plant State: Pown:1 Level 75 - 100%. Ali Controls in Automatic i /g, ,
A Sequence Initiater: d ,Ieg Break - RCS (200%) \'<
C g'f,f c
Important Plant Paranctt/s: 1) PZR Level, 2) NR Pry.sure 3) Contain-ment Pressure, 4) Contajnmeat Radiation, 5) ContaintAint Surap' Level,
- 6) SG Levels 3 ..
E. i
'h Progression of Opei JNActions: See Flow Cnagt 1 p .s. I g+
Final Elant State: PIMt is in cohl leg recirhNintAhn and' prepared for hot leg recirculation.f Safety injection is runk Lg, Containment spray
<0, may be required,
} 'p g
,, [
Major Plant Systems: ItCSlhnergencyCoreCoolingSy(teas, Radiation Monitoring, Containment Qtteus, Steam Generators, Auxiliary I'eedvater i
t j
Tol%rance Range: The operator should: identify fadh event: follow apNooristes)rocedure(s)-ensuresafetyinjection.andauxiliaryfeed-wa6)idlow, Yerify CIA /CIB, cold and hot leg rec'3rQ dation, and core coLM utt and declare emergency event, ,T t s . /4/
j Compettucies Tested:
l 9 }l ?
DRO - Qvpliance/Use of Pro edures p
5 Supervisory Ability (
\
b i Communications / Crew Interaction \s
$ t,
.s L , .p RO - Understanding / Interpretation of Annunciator. Alarm Signals
! Diagnosis of.Fnents/ Conditions Based on Signals / Readings Compliance /Use,bf Procedures l Control Board Operation s c-Communications / Crew Interaction P rs I
grq - Compliance /Use of Procedures ,
t Control ~ Board Operation () ,
s J. (ammunications/ Crew Interaction u s
,r x hll s (r
- q,' y m ; r, Q ,
i J
$ '\ .
~ .
v[
( >
253
~ .
,. I Lt ,
ijL '
' i LARGE BREAK LOCA - REACTOR TR't WITH SAFETY INJECTICF h '! Pre r essio.t of Operator Act ions t la; s I %f 2 RO/ 5RO RO RO
,,,,, Diagnose ,, la Manually
^ alarms reactor '- No trip
- Alarm 8 fold out(s) tripped? reactor should l'e open ye,
/ i. ,
o BOP SRO/R0
_ BOP Verify Verify AC Verify S.I.
turbine busca actuated (5) trip
- et. erg tr ed ---*
I l RO/ BOP RO/ BOP RO#!JP RO RO g Verify CCW Verify 011 M .try aux. Verify - r if y i
in service S. I . p.tm f'.W. Ltnning Phase A feedwater
,,,e_, runnirs. % e isolation isolation 1
g BOP /R0 SRO/R0 SRO/R0 BOP /R0 RO Verify Verif y , Verify Verify Verify service conta inment containment MSIVs and containment
- w. ster 1r . emg. ,_ ) . ventila t ion % bypansre % spray running serv.*o cooling isolation isolate /
B07
F S. line up e
Verify proper Aux.
F.W. flows l l
pw -
Verify proper S.I.
+-
RO Verify Phase B isolation p
0l (1)'
I .
RO y 60 RO R0 SRO/R0 Verify S.I. l Verify RC5 Verify Pzt Reverify S.I. Verify valve I temp. PORVs and flows (1) Contmt.Pressa, line up % .mi,s. % sprays shut ---*' ---> Contmt. Rad.,
Contmt. Sump, are increasing
/'
1r SRO
- 4larms .-
g f - Pp low /S.I. - Trip I#" 1 reactor
~ P e O
S.I. Actuated coolant l - Containment Press Hi, Hi-H1 g,y
- Containment Radiation Hi
- Pha s e A & B I so la t ist.
! - ('ontainu nt Spre} Attuated h
Notes (1) Plant 3;etific Value(s)
(2) Power Avel'.able Prom Of f site f (3) Verifying Cold Leg Rectrc. Capability f (4) Evaluate Need to Vent Peactor Vessel Head (5) Emergency condition Should be Declared and l'pgraded as Required l
l 254 m .Aw _ _ _ _ _ ___._m m__.__ _ _ _ _ . _ _ __. J
I LA'CE EREAK LOCA - REACTOR TRIP VITH SAFETY INJECTION s Progression of Operator Actions:
l s.
e i Page 2 of 2 t )
l h. w_,/'/
BOP /R0 BOP /R0 BOP /R0 RO Reverify Restore and/ Verify Verify 1 S/Cs are or maintain secondary PORVs
- intact
- S/C level
- side radia- ---*" closed (1)% N.R. tion levels norm.
I RO/SRO RO Is Stop contain-contm't Yes ment spray pumps p<r1)?
ess.
No I
SRO/R0! BOP BOP /RO BOP /R0 (RO/SRO RO Evaluate Stop diesel Verify S/C is Stop low plant generators press. No RCS Yes head S.I.
status (3)
(2) stable or n press. pumps increasing > (1) ?
I p SRO/RO/ BOP RO Ro ho Large break Transfer to Verify Align S.I.
LOCA - Sys cold leg Reset 5.I. C.C.V. to for cold depressurized % recirc. % _ RHR heat leg rec tre .
ES-1,3 exch.
g~g
! \
\ ' - - ' ) SRO/R0 SRO/R0 SRO SRO/R0 SR0/R00 Transfer to Prepare Consult Align Start S.I.
hot leg for hot leg plant contm't pumps as recire.
- recirc. *-- engineering =*-- sprays as e necessary after (1) staff (4) necessary hrs.
SRO/R0 SRO/R0 Evaluate Report long term NRC End status
- INPO Plant Mang.
l l
l
( /s
- / \
- t. I 1 wJ 255
i Operating Sequence: PZR/PORV Failure to Open NSSS/ Type: Westinghouse and C-E/PWR Initial Plant State: PZR Pressure Control and Control Systems in-Automatic, System Pressure Normal:
Sequence Initiator: PORV Failed Closed and PZR' Pressure Channel Falls High Important Plant Parameters: PZR Pressure Progression of Operator Actions: See Flow Chart Final Plant State: An alternate pressurizer pressure control channel has L been selected. lThe pressure has been restored to normal and the PORV has been isolated and repairs are underway.
l Major' Plant Systems: Pressurizer Pressure Control Tolerance Range: The operator should recognize.and diagnose the event -
failed pressure channel and failure of PORV to open as required'. He should select an alternate channelEto mitigate the transient and stabilize the plant.
s Competencies Tested:
k,s SRO - Compliance /Use of Technical Specifications Supervisory Ability RO '- Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operation l
i b
257
PZR/PORV PAILURE TO OPEN Progression of Operator Actionst O:
- Alarm
indicators failed press " #"" I'" "' *
(1) channel OP erator should recognize this valve did M open.
RO i' SRO/R0 0 PORV did M Trip failed open, cha nnel sprays opened blatables Ro U sPo "
Select Repair alternate failed control channel chain PO i' SRO I' Restore Report Pzr. Press. Plant Mgr.
and >bintenance stabilize RO e Notify SRO F Shift Supervisor End SRO i' Ref to tech. specs.
for PORV(s)
SRO p Repair PORV j if possible l i l
258 i
________________--__----------J
j}
Operating Sequence: Steam Generator Tube Rupture i /
% NSSS/ Type: Westinghouse /PWR Initial Plant State: Control Systems in Automatic, Normal System Temper-ature and Pressure Power Level Greater Than 20%
Sequence Initiator: Steam Generator Tube Rupture Important Plant Parameters: 1) RCS Parameters, 2) RCS Inventory, 3)
Secondary Radiation Monitoring (Air Ejector Condenser, Steam Generator Blowdown), 4) Subcooling of RCS Progression of Operator Actions: See Flow Chart Final Plant Stato: The affected steam generator has been isolated and the primary / secondary delta P is at a minimum. RCS cooldown has com-menced and subcooling is being maintained.
l Major Plant Systems: RCS, Engineered Safety Features Actuation Systems, Radiation Monitoring, Steam Generator Tolerance Range: The operator should recognize the fault event and must identify the generator. He should also isolate the steam generator and reduce primary / secondary delta P within 30 min. The operator should declare an emergency condition.
,s y,, Competencies Tested:
SRO - Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based On Signals / Readings Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction f
x_ - ;
1 259
STEAM CENERATOR Tl'BE RUPTURE Progression of Operator Actions:
Page 1 of 2 RO/SRO RO R0 Diagnose Is With trip
- Alarms alarms reactor No imminen t p Fold out tripped 5 trip reactor 7
Yes l
BOP gr' BOP RO Verify Verify emg. Verify SI turbine trip AC buses actuated (7)
RO/ BOP RO/ BOP RO RO RO 1r Verify Verify all Verify aux. Verify Verify C.C.W. in S.I. pumps F.W. Phase A feedwater service 4 running 4 running 4 isolation 4 isolation BOP gr RO SRO/R0 RO R0 SRO/R0 Verify Verify Verify Containment Verify service water containment containment spray not, proper S.I.
p emg. cooling > vent ilat ion > required > flow (s) (1) isolat ion RO RO RO RO RO T Verify Pzr Verify RCS Verify S.I. Verify AFW Verify PORV's and temr. valve lineup valve lineup proper AFW spray's shut 4 decreasing ' ' '
flow (s) (1)
SRO gr Notes S/C tube (1) Plant Specific Values E (2) > (2) Operator received alar:r.s noted Should identif y S/C tube rupture (3) Rad. Monitors *
- Sample Line
- Blowdown
- Mn. Steam Line
- Alarms *S/C level increasing Rad. Monitor Alarms **
(4) , ggIy
- Condenser Evacuation System
- Steam Cenerator Blowdown - At mos. PORV
- Steam Line Rad. Monitors - Blowdown
' " # - Stm. Supply to Aux. Feed
, , p (If faulted S/C is the supply)
Pzr. Level Deviation (5) Plant specific requirements must be met (6) Method of cooldown dependent upon status of subcooling (7) Emergency condition should be declared and upgraded as required O
260
STEAM CENERATOR TUBE E!IPTURE Progression of Operator Actions:
Page 2 of 2 RO RO/ BOP BOP /R0 RO/ BOP Reverify Id en t if y isolate Ensure SI flow faulted faulted level in 01 > then stop RCP
> S/G (3) > S/G (4) e affected S/C
> (1)
RO RO PO RO BOP /RC Reset S.I. Verify Reverify Verify Pzr Isolate all (5) unaffected faulted S/G PORV's feed to G S/G 1evels 4 isolated 4 closed 4 affected S/C
> (1)
RO y RO RO RO SRO/RO/ BOP Reset Verify Verify all Determine Commence conta inment instrument AC buses if low head RCS cooldown Phase A/B ' air to ' energized y S.I. can be p (6)
(5) containment stopped (1)
I SRO/R0 RO RO RO/ BOP RO y Terminat e Establish Depressurize Verify RCS Verify S.I. flow Pzt level RCS - reduce subcooling faulted S/C 4 (1) 4 AP - reduce 4 (1) 4 level stable leak rate or inc.
I R0 y RO RO/ BOP RO BOP Establish Reverify Establish Establish Secure l normal Pzr level normal makeup normal diesel l charging > (1 ) > to VCT > letdown e generators l
SRO BOP /R0 RO 80 Y Notify Commence Ensure Reverify End NRC & cooldown by subtriticality primary /sec.
management 4 appropriate 4 4 leakage procedure minimized 261
Operating Sequence: Failure of Main Turbine to Trip
/ NSSS/ Type:_ Westinghouse and C-E/PWR
\
Initial-Plant State: Power Level at 75 to 100%, Control Systems in Automatic Sequence Initiator: Reactor Trip with Failure of the Main Turbine to Trip Automatically (and/or Manually)
Important Plant Parameters: 1) RCS Temperature and Pressure, 2) NIS,
- 3) Turbine Load, 4) Steam Generator Level Progression of Operator Actions: See Flow Chart Final Plant State: Emergency operation is complete with recovery from reactor trip and safety injection is in progress. The main turbine has tripped or~the steam is isolated from the turbine.
Major Plant Systems: Main Turbine / Control, RCS, Reactor Protection, Main Steam Tolerance Range: The operator should recognize and diagnose the event and declare the appropriate plant event. He should take actions to limit RCS cooldown'and reduce the possibility of safety injection on low pressure due to cooldown. The SRO should continue to check to ensure action (s) are being taken to trip and/or isolate the main turbine.
f-~s (Even though the turbine may not have tripped manually, the crew should not leave emergency operation but continue with required procedures.)
(d )
Competencies Tested:
SRO - Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Communications / Crew Interaction BOP - Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction 1
4 263
r l
l 1
FAILURE OF MAIN TURLINE TO TRIP Progre2sion of Operator Actiones s
- Alarms RO/SRO P O
Diagnose alarms Foldout (s) open (1)
}
3r Verify reactor trips I
l (Info)3r Turbine does NE l auto trip BCP Sr Attempt to manually trip turb. >
(2)
BOP BOP 1F SRO/A0 Manually (2) Dispatch (2) runback operator to load locally trip turbine (3)
SRO/ ROT BOP Cont inue E-0 Rx trip r or SI (4)
SRO 1P Repo rt NRC Plant Mgr.
V nd
- Alarms - Any Rx trip first out.
i (1) Declare event.
(2) The examiner can allow the manual trip to work or he may require additional steps to be taken to shutdown turbine.
(3) Manual trip f rom f ront standard or/and secure DiC system.
l (4) Reactor trip or SI and recovery 264
1 i
Operating Sequence: Small Break Loss of Coolant Accident j
7s j NSSS/ Type: Westinghouse and C-E/PWR
\ ,!
Initial Plant State: Any Power Level, Extended Power History (Decay Heat)
Sequence Initiator: Break Size (Unisolable) Greater Than the Capacity of One Charging Safety Injection Pump Important Plant Parameters: 1) NIS, 2) RCS Pressure, Temperature, and Level, 3) RCS Flow, 4) SI Flow, 5) Auxiliary Feed Flow, 6) Heat Sink,
- 7) Containment Parameters Progression of Operator Actions: See Flow Chart Final Plant State: The plant is stable and cooldown is in progress with subcooling established.
Major Plant Systems: RCS, Safety Injection, Auxiliary Feed, Containment, NIS, RMS, Electrical Distribution, Reactor Protection Tolerance Range: The . operator should recognize and diagnose event. He should also comply with procedures and declare the appropriate plant evet. t .
Competencies Tested:
'D I
) SRO - Compliance /Use of Procedures
'~ ' Supervisory Ability Communications / Crew Interaction R0 - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based On Signals / Readings Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction ym I \
LJ 265
l SMALL BRf4K LOSS OF COOLANT ACCIlsENT Progression of Operator Actiones Page 1 of 2 R0 R0 BOP Diagnose Rx. trip may Verify Verify q3, } j r alarms fold out(s) be manual or b automatic r reactor trip turbine
>* trip
/ open (1)
- Pp Low Press. Trip Verify F.W. Check if S.I. Verify PWR Pp Press. Low Alarm Containment Humidity Hi isolation actuated to A.C. emg.
4 Actuate S.I., 4 buses Containment Rad. H1 declare event Par Lvl Deviation High Chrg Flow (Plant Specific) 80 y RO/ BOP RO Notes Verify Phase Verify AFW Verify S.I.
"A"/"1" pumps running pumps running (1) Leak size will determine isolation > >
when operator will enter E-0 and when $1 is required.
Operator may trip reactor /
initiate SI if he feels l
press. Decrease is RO/ BOP RO/ BOP BOP /R0 kO/BOPT out of control l
Verify Verify Verify I Ve rif y CCW (2) Plant Specific Value containment ,
containment service ,
pumps running vent , emg. cooling m water M (3) Containment isolation pumps
- Radiation High running
- Sump Level Abnormal
- Press./ Temp Abnormal R0 RO R0/B0P y RO/ BOP (4) RCS Subcooling (2) Check if Verify Verify S.I. Verify aux.
Sec. Heat Sink Avail. (2) MSIV's should cont. spray flow (2) FD flow RCS Press. Stable (2) be isolated I not required i I (2)
Pzt. Level Stable (2) (2) at this (5) Break size should allow low HD to be placed in standby R0 R0 RO/ BOP RO/ BOP 4 (6) Additional SI pump (s) Check if Verify pzr. Verify tavg. Verify S.I.
may be started / stopped RCP's should ,
sprays and trending to and aux. FD as required to accom. be stopped , PORV's shut , no-load (2) 4 valve plish cooldown (2) alignment (7) Maintain adequate 2 Prr. level and RO/ BOP R0 SRO/R0 subcooling (2)
[
Verify Su's RCS not l E-1 Check if are not intact - ,lLossof y , RCP's should faulted or r (3) r" reactor r be stopped tube rupture coolant (2)
SRO/R0 RO DOP/R0 RO/ BOPy Check if S.I. Check pzr. Verify sec. Reverify flow can be sprays and radiation SG's intact reduced (4) 4 PORV's shut 4 normal 4 V
ES-1.1 S.I. _ p_ 1 termination 266 1
1 1
1 1
SMALL BREAK LOSS OF C00LANI ACCIDENT Progression of Op3rator Actions:
Page 2 of 2 j
4 k RO RO/ BOP RO/ BOP R0 Reset Reset Establish Stop all but S.I. containment instrument one CSIP 1 I k isolat ion (s,) b air b standby Phase A and/or B
R0/SRO RO/ BOP RO/ BOP RO 4 Pzr. level Isolate Establish RC decreases B.I.T. chrg. flow pressure
, , (2) stable /inc \ h
. r 2 Pm 1
?
RO/ BOP, , R0 BOP R0 Return B.I.T. ES-1.2 Reverify. S.I. Verify A.C. Place low to service post loca containment buses energ. Hd S.I. pumps I cooldown I iso. reset I from off- k in standby depressuri- instrum. air site (5) sation avail to con-entnmone RO RO R0/SRO RO/ BOP BOP g Par. htro. Reverify S.I. Verify RCS Initiate RCS Verify SG off pumps still subcooling cooldown per levels 4 in serv. 4 core exit 4 procedure 4 (2)I (6) TC's (2)*F RO 3r SRO/R0 RO RO RD
[ \
, Verify par. Start one Verify normal Establish
,f
) level is RCP per charging flow noreal charge Isolate B.I.T.
(2)% r procedure r path r flow r SRO/R0 SRO SRO/R0/ BOP RO RO qr Verify full Verify Reduce RCS Reverify RCP Control S.I. flow adequate pressure to is running charge to not required 4 shutdown 4 reduce break 4 4 maintain pzt.
margin flow (7) level (2)
SRO/R0p BOP RO RO RO/SRO Isolate Stop D.C..'s Verify RCP Establish Verify S.R.
accum. per cooling water RCP seal detectors procedure (2) > I normal k water return I energized flow SRO RO/ BOP SRO/RO/ BOP RO 3, Evaluate Place RHR in Shutdown Transfer plant service per unnecessary S.R.'s to status 4 procedure 4 equipment 4 recorders (2) (2) j l
SRO p
Report
/ NRC INPO - END
( plant agr v
267
Operating Sequence: Anticipated Transient Without Scram NSSS/ Type; Westinghouse and C-E/PWR Initial Plant State: Any Power Level Greater Than 20%, All Control Systems in Automatic Sequence Initiator: Manual or Automatic Reactor Trip Condition with Failure of Trip Breakers to Open Important Plant Parameters: 1) Nuclear Power, 2) Turbine Status,
- 3) Boration Flow Rate, 4) RCS Parameters, 5) Heat Sink (Auxiliary Feedwater)
Progression of Operator Actions: See Flow Chart Final Plant State: The reactor is subcritical and the turbine is tripped. There is normal shutdown electrical lineup and no safety injection. Subcooling is being maintained (RCS) and RCS is being borated to maintain subcriticality. The rods are on the bottom and auxiliary feedwater is running.
Major Plant Systems: Emergency Boration Via Makeup System, Reactor Coolant, NIS, Rod Control Including Trip Breakers, Main Turbine and Generator Tolerance Range: The operator actions should include placing the plant in safe condition by: making the reactor subcritical, tripping the turbine, performing emergency boration, and making the auxiliary feedwater flow.
Competencies Tested:
SRO - Diagnosis of Events / Conditions Based on Signals / Readings l
Supervisory Ability Communications / Crew Interaction RO - Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Control' Board Operation 269
ANTICIPATED TRANSIENT WITHOUT SCRAM Progression of Operator Actions:
Verify t
- Continue to
- Alarma
] >
condit1, exist s >
monitor CSFST (1)
V RO RO is Manually reactor No trip reactor tripped? k Yes SRO/R0 RO/ BOP R0 '
RO RO Dis pat ch Recheck Emergency Did Manually operators , reactor /tur- , borate per reactor No insert rods (2) ' '
bine trips procedure trip? b A & l RO "
y RO BOP /R0 Bol Verify Verify Verify is emergency aux feed heat sink Yes turbine boration flow > flow (aux, feed) 6 tripped?
running J6 No RO/ BOP 970/SRO BOP V Check for Verify RCS Manually Pos. reactivit) dilut ion trip turbine Addition 4 paths -
due to exces- isolated sive cooldown BOP I R0 9 RO/ BOP RO/SRO Check closed Verify Ma in ta in Repo rt MSIV's and reactor plant in event t bynasses y subcritical > stable cond it ion
> NRC & Plant management
>[ END
- Alarms Notes Reactor Trip First Out Annunciator (s) (1) Subcrit icality - ATWS ent ered from E-0.
CSFST entered when trip did not occur.
(2) Locally open trip breakers and trip turbine.
O 270
i i
Operating Sequence: Loss of Auxiliary Feedwater - Inadequate Core Cooling fi f NSSS/ Type: Westinghouse and C-E/PWR Initial Plant State: Any Power Level with Extended Power History Sequence Initiator: Total Loss of Feedwater Flow - Auxiliary Feedwater Problem Resulting in No Flow Important Plant Parameters: 1) Nuclear Power, 2) Feedwater Flow, 3) RCS Temperature / Pressure, 4) SG Levels, 5) PZR Level, 6) RCS Flow, 7) SI Flow, 8) AFW Flow, 9)' Thermocouple Temperatures, 10) RCS Subcooling Progression of Operator Actions: See Flow Chart Final Plant State: Adequate core cooling has been reestablished. Heat sink has recovered and safety injection termination is in progress.
L Major Plant Systems: RCS, PZR,.SGs, Reactor Protection, SI, Main Feed, Auxiliary Feed, RVLIS, Subcooling Monitoring, Incore Thermocouple, CVCS, Containment Support, Instrument Air, H2 Recombiners/ Monitoring Tolerance Range: The operator (s) should demonstrate his ability to diagnose the event and ability to successfully follow the appropriate procedure (s), E-0 ES-0.1, FR-C.1, FR-H.1, E-1.0, and ES-1,1, and CSFSTs,_F-0.2 and F-0.3. The operator should declare the appropriate
[h.
t plant emergency event.
Competencies Tested:
SRO - Compliance /Use of Procedures Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Understanding of Instrument / System Response Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction j B0P - Compliance /Use of Procedures Control Board Operation {'
Communications / Crew Interaction a
( I i
271
I LOSS OF AUXILit.CtY Progression FEEDUATER - INADEQUATE CORE COOLING of Operator Actions Page 1 of 2 R0 RO RO/ BOP O
RO/ BOP Diagnose Reactor eAlara ] alarms
> fold out(s) ,
tripped rods Turbine tripped A.C. power r inserted ,
r stop valves , is available should be power r open shut decrease
feed flow (none present)
RO/ BOP y7 BOP /R0 (2) Flant Specific Value SRO y,,ggy yy gg (3) At least 1 High Head status g
SI Running. Proper decreasing operator to Flow (no FW/AFW) > (2)1 I restore F.W.
(1) flow (4) Event should be declared by this point. Upgraded as required. e I SRO/RO l Continue to monitor CSFST's (4)
Verify S.I. BOP /R0 Inadequate ore valve Loss sec.
alignment core cooling y,, exit T/C's Attempt to i FO.2/FR-C.1 No heat sink restore AFW 1200*F b FO.3/FR-H.1 7
r to one or more SG's R0 Any Time Core TC's
,, R0 1200*F - Operator Goes RO RO Actuate to FR-C.! gr Check RCP S.I. Verify
, support Stop all r conditiona charging , RCP's available pump '
(2) available RO q, BOP BOP /RO Check S.I.
accumulators Attempt to Attempt to iso. valves establish , establish open nn . F. W. r condensate flow R0 RO/SR r RO RO RO Verify Core Monitor rvlis norm. Yes C's (exit No Actuate SC levels (2) containment S.I.
1200*F N Hy conc. I (2)%
?
(2) m
- I RO/SRO i
4 R0 ir Core exit T/C's No SG 1evel are Verify RCS
(
700*F (2)! feed path m
? flow (3) I Yes l
1 2 272 l
i
LOSS OF AUXILIARY FEEDWATER - INADEQUATE CORE COOLING Progression of Operator Actions Pags 2 of 2
- s u-RO Reset SI 1
2 4 R0 RO RO/ BOP RO/ BOP R0 1 ir Depressurize or Start RCPs Establish intact SCs exit TCs as necessary Reset CIA instrument
" ; air to (methods (2)) 4 < 1200F
?
containment RO RO/ BOP RO RO 37 3r y, Verify SI Feed / bleed Maintain Establish accumulators RCS until RCS feed RCS bleed .l can be TCs < 1200F and bleed 4 path via 'l g path PORVs {
isolated (2 open) l RO/ 3r SRO 4 RO/ yl BOP SRO/RO/ BOP Continue Restore Verify core l
SI flow heat sink TCs and RCS l
until liCS > hot leg (s) decreasing hot leg temp.
< 350F . _ _ .
l ,"\ R0 SRO/ RO -.
l j RO y 1r !
-'w '
/
Secure Shut PORVs Verify RCP(s) subcooling l
4 > (2)F RO RO R0 1r ir Check RVLIS Establish Maintain PZR and hot leg normal level > (2)%
temperature charge / y
< 350F letdown I
ES-1,1 E-1,0, Step 12 [ EO SI Evacuate plant Q termination status (4)
V END Notes
, -N (4) Decision to be made concerning recirculation mode j necessary continued depressurization.
\ /
~_s' 273 i 1
i
_ _ _ . _ {
-)
Operating Sequence: Loss of Off-Site Power l <~'s
'l I NSSS/ Type: Westinghouse and C-E/PWR l\j Initial Plant State: Any Power Level, Controls in Automatic Sequence Initiator: Loss of Feeder Breakers on Off-Site Power Supply Important. Plant Parameters: 1) Nuclear Power, 2) RCS Temperature and Pressure, 3) SG Levels, 4) Auxiliary Feedwater Flow, 5) Electrical Power, 6) RCS Flow (Natural Circulation)
Progression of Operation Actions: See Flow Chart Final-Plant State: Natural circulation cooldown and the plant is stable.
.. j Major Plant Systems: RCS, NIS, Auxiliary Feed, Electrical Distribution, MS PORVs, Reactor Protection, Turbine Generator Tolerance Range: The operator should recognize and diagnose the event.
He should follow the appropriate procedures - E-0, ES-0.1, and ES-0.2.
The operator should declare the appropriate plant ever,t. Natural circulation cooldown rate is such that inactive portions of RCS are cooled without drawing bubble in Reactor head.
Competencies Tested:
/N SRO - Compliance /Use of Procedures
( Supervisory Ability Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings l Compliance /Use of Procedures '
Control Board Operation Communications / Crew Interaction BOP - Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction !
t r
b 275
r LOSS OF OFF-SITE POWER Progression of Operator Actions:
RO RO/ BOP BOP BOP /R0 Diagnose Verify l Verify Verify PWR
- Alarms ) ,
r alarms fold out(s) m r
r lturbine ,
r to A.C.
trip emg. buses open RO/ BOP R0 RO Verify F.W. Verify RCS ES-0.1 Verify noo water source temp. trend Keactor safety to stm. I to (1) I trip 4 injection gen.'s response RO 3r R0 BOP /R0 BOP /R0 Verify all Verify per Verify SC's Verify A.C.
rods fully m level / press. m in narrow & power inserted F controls r range (1) r (off-site) functioning ng avail.
R0 RO/ BOP BOP R0/B0Py Verify S.R. RCP's not Transfer Verify NIS avail, verify / stm. dump diesels are energized 4 eotab. natura14 to press. 4 running and Transfer circulation mode loaded (1) recorder (2)
BOP /R0 3, SRO/ BOP /R0 SRO/R0 RO/ BOP Shutdown Maintain fNatural No RCP avail, unnecessary stable plant circ. Borate to plant > conditions >l required > cold shutdown equip. (1) (1) ES-0.2 per (3)
R0 SRO/R0 BOP /R0 RO/BOPy
- F~
Continue Continue to i Initiate verify CRDM cooldown verify j cooldown fans running per ES-0.2 4 natural circ. 4 stm. dump & 4 (3) flow aux, feed (3)
RO/SRoyl BOP /R0 RO/ BOP SRO I
Place RHR Continue to l Report in service (1) (3) >
cooldawn to 200*F El'offsite Attempt restore PWR iNRC plant angr.
I
- Alarms Notes I
, (1) Plant Specific Value END g,
Low Flow R4 Trip (2) Verify Natural Cire.
- Switchyard Trouble - RCS Subcooling (1)*F
- SG Press. Stable or Decreasing
- RCS Thot. Decreasing or Stable
- Core Exit TC Stable or Decreasing (3) Plant Specific Procedure 276
1 Op: rating SIquence: Station Blockout - Loss of All AC Porer l l
NSSS/ Type: Westinghouse and C-E/PWR
/ Initial Plant State: Any Power Level Controls in Automatic Sequence Initiator: Loss of Offsite Power With Failure of Diesel Genera-tors to Start Important Plant Parameters: 1) RCS Temperature and Pressure, 2) Reactor Power, 3) RCS Flow, 4) SG Level, 5) Natural Circulation, 6) AFW Flow,
- 7) Electrical Power Progression of Operator Actions: See Flow Chart Final Plant State: AC power has been restored and the plant is stable on forced circulation or natural circulation.
Major Plant Systems: RCS, NIS, Auxiliary Feedwater,. Electrical Distri-bution, Atmospheric Steam Dump, Reactor Protection l
Tolerance Range: The operator should recognize and diagnose event. He should follow the appropriate procedures - E-0, ECA-0,0, and ECA-0.1.
The operator should declare the appropriate plant event.
l Competencies Tested:
SRO - Compliance /Use of Procedures Supervisory Ability g
Communications / Crew Interaction
((-} RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Compliance /Use of Procedure Control Board Operation Communications / Crew Interaction
/ h 277 l I
STATION BLACKOUT - LOSS OF ALL A.C. POWER Progression of Operator Actions:
Diagnose is Manually
- ala rms } alarms
> Fold out(s) I reactor No trip
] open
tripped 7
reactor Yes ,
l BOP T RO/ BOP SRO/R0 Verify Verify AC Declare turbine trip buses are ECAD.O appropriate I g energized I emg. condition BOP /R0 RO BOP / ROV Are' Verify aux.
Yes C emg. buse Verify Reverify F.W. from tur. RCS reactor /
Enter E-0 -
energized 4 driven aux 4 I isolated (2) turbine
? F.W. pmp. (1) /
tripped No RO y SRO/R0 A0/ BOP /R0 A0/ BOP 2 V Pull to lock Dispatch 1solate RCP Verify SG are emg. core operator to END seals / therm. y not faulted cooling pumps I I barriers locally restore
> E are A.C. power locally isolated BOP BOP /R0 BOP /R0 R0 BOP /R0y Reverify Shed Verify and CST level Verify no SG Verify SG are nonessential maintain SG tube ruptures not faulted (1)% 4 D.C. loads ; levels (1)Z 4 4 and are isoisted BOP /R0 y SRO/R0 SR0/R0 RO RO Commence '
Verify Verify S.I. Initiate cooldown Verify reactor, not Phase "A" containment with atmos. I subcritical I required r dump isolation r isolation (ventilation)
BOP /R0 BOP /R0 RO y Is Attempt to Verify Verify y , A.C. power restore containment
' 4 , , containment restored A.C. power '
rad. levels ' pressure-
?
normal (1) psig Yes BOP y BOP /R0 BOP /R0 Stabilize Manually Verify ECA-0.1 S/G pressure load equip. service water Loss of all r as necessary I in service I A.C.-(No S.I.)
(3) l Notes (1) Plant Specific Values (2) PORV's Letdown and Excess Letdown (3) 480V buses, Battery Chargers Instrument Power, Emg. Lighting Communications, etc. (1).
278
STATION BLACKOUT - LOSS OF ALL A.C. POWER Progression of Operator Actionst j '
' \ Page 2 of 2
)
s_-
A0/ BOP /R0 RO R0 BOP /R0 Reverify Reset Establish Manually load RCP's seala Phase "A" inst. air to equipment on 1 y and thermal 7 isolation -
? containment > emg. buses barriers (4) isolated ,
1 R0 RO RO RO/SRO R0 1r Place ECCS Verify and Verify and Verify S.I. Establish equipment to maintain SG maintain not required charging standby (5) i levels I level and I (1) i flow (1)! pressure (1)
A0 RO RO RO/SRO SRO/R0 3r RO/ BOP Establish RCP Establish Maintain per Verify Verify SR.
seals and letdown per level and natural energized cooling water > proc 6 dure 5 pressure (1) I circulation 5 per procedures BOP SRO/R0 SRO/RO/ BOP RO/SRO1r Restore off- Verify Maintain Verify site power to ,
adequate ,
plant ,
shutdown A.C. buses , subcooling , conditions , margin stable (6)
,s i \4 (w_/ /
1r SRO is ES-0.2 Report natural natural NRC cire. cire. > plant mgmt.
required cooldown m
r 1r Refer to appropriate END plant procedure Notes (4) Instrument Air Compressor Comp. Cooling Water Pump Charging Pump Cont. Ventilation (1) Plant Specific List (5) High Head S.I. Pumps Low Head S.I. Pumps Containment Spray Pumps (6) RCS Pressure Par. Level RCS Temp.
Intact SC Levels I 1 1 /
w ./
279
Operating Sequence: Control Room Fire Requiring Evacuation NSSS/ Type: Westinghouse /PWR Initial Plant State: Power Level at 75 - 100%, All Controls in Automatic Sequence Initiator: Fire in Control Room Important Plant Parameters: 1) Nuclear Power, 2) RCS - Temperature and Pressure, 3) Auxiliary Feedwater, 4) Containment Pressure Temperature,
- 5) PZR Level, 6) Turbine Load, 8) SG Level Progression of Operator Actions: See Flow Chart Final Plant State: The plant is in shutdown and stable. Cooldown may be l
in progress (examiner's choice). Natural circulation cooldown is possible.
Major Plant Systems: RCS, NIS, Auxiliary Feedwater, Electrical Distri-bution, Charging and Letdown, Main Steam, Steam Generator Levels, RCS Makeup, Control Systems - PZR Level and Pressure, Steam Dump Control, Remote Shutdown Panel (s)
Tolerance Range: The operator should follow the appropriate procedure -
E-0, ES-0.1. He should also follow the abnormal procedure for safe shutdown with control room inaccessible. The operator (s) should use the appropriate logs / check lists.
Competencies Tested:
SRO - Compliance /Use of Technical Specifications Compliance /Use of Procedures Communications / Crew Interaction RO - Compliance /Use of Technical Specifications Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Compliance /Use of Procedures Control Board Operation Communications / Crew Interactions 281
CONTROL ROOM FIRE REQUIRING EVA1,UATION* l Progression of Operator Actions:
l RO SRO/R0 SRO/RO/ BOP Trip reactor Is ES-0.1 '
} from control immediate No Stab 111:e Fire > room (1) ** evacuation plant
] required I
Then evacuate Yes 4 i SRO/R0 ir BOP PO BOP /F0 Transfer Verify Verify emer-controls to reactor trip gency buses local > locally > energized (1) (2) (4) (3)
_ Verify RCS Verify SG Verif y Verify PZR temperature levels auxiliary level /
trending 4 (1) 4 feed flow (s) 4 pressure (1) down or stable (1)
No.loid BOP / g RO BOP /R0 RO RO Shut the Verify main Verify CCW Shut down MSIVs g enerat o r and ESW RCPs y output > operational )
breakers open SRO/R0 SRO/R0 SRO/RO/ BOP RO/ BOP Initiate Place RHR Commence Verify s
repairs in service cool down auxiliary 4 per 4 per (2) 4 equipment procedure needed for cool down available (1)
SRO y Report Plant Mang. END
- Simulators may or may not have auxiliary control panel (remot e shutdown panel) in which case the only items which can be accomplished will be plant trip and stabilization.
- Trip reactor, turbine and/or RCP may be included.
Notes (1) P'lan t specific list, value or requirement.
(2) Control room inaccessibility procedure.
(3) NIS local indication available locally in some plants.
(4) Ensure communications have been established.
O I
1
?82 L-__---------_----_---_-
- , Operating Sequence
- Main Steam Break Inside Containment NSSS/ Type: Westinghouse and C-E/PWR Initial Plant State:'Any Power Level, Control Systems in Automatic Sequence Initiator: Main Steam Break (Shear) Inside Containment Which Affects One Steam Generator Important Plant Parameters: 1) PZR Level, 2) PZR Pressure, 3) Contain-ment Pressure, 4) SG Levels Progression of Operator Actions: See Flow Chert Final Plant State: The plant is cooled down; the faulted steam generator is isolated; and safety injection and containment spray are stopped.
Major Plant Systems: RCS, Emergency Core Cooling Systems, Containment Systems, Steam Generators, Auxiliary Feedwater, Main Steam Tolerance Range: The operator (s) should identify the fault; follow appropriate procedures - E-0, E-2, E-1, and ES-1.1; and declare the emergency event.
Competencies Tested:
j e"'s SRO - Compliance /Use of Procedures I Supervisory Ability
(
'- / Communications / Crew Interaction RO - Understanding / Interpretation of Annunciator / Alarm Signals Diagnosis of Events / Conditions Based on Signals / Readings Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction BOP - Compliance /Use of Procedures Control Board Operation Communications / Crew Interaction I
\
283
MAIN STEAM BREAK INSIDE CONTAINMENT Progression of Operator Actionst Page 1 of 2 RO/SRO RO RO Diagnose Is Manually
- Alarms alarms reactor No trip fold out(s) b tripped? b reactor should be open Yes ; l i
BOP v RO/ BOP SRO/RO Verify Verify emg. Verify S.I.
tur bin e A.C. buses actuated (2) trip > energized >
BOP RO RO/ BOP RO RO/BOPV Verify Verify all Verify aux. Verify Verify C.C.W. in S. I . pumps F. W . P.hase A feedwater service 4 running i running 4 isolation 4 isolation I BOP /R0 9 S RO/ R0 SRO/R0 BOP /R0 RO Verify Verify Verify Verify MSIV's Verify I 6ervice conta inment containment and bypasses containment water in y emg. cooling y ventilation p isolated y spray running service isolation BOP /PO BOP /R0 En PO RO V Verify aux. Verify Verify Verify F.W. line up proper aux. proper S.I. Stop RCP's Phase B i F.W. flow 4 flows (1) i 4 isolation (1)
RO V ___
m Rn RO BOP /SRO/R0 Verify S.I. Verify RCS Verify Pzr Reverify Verify valve line up temp. PORV's and S.I. flows faulted S/C
> decreas. > sprays shut > (1) > depressurized BOP /R0 RO/SRO BOP /RO/SRO BOP /R0 V Isolate Id en t if y Verify '. er if y MSIV's Faulted S/G faulted S/C faulted S/C other S/C's & bypass on isolation (3) 4 4 not faulted i affected 4 (E-2)
S/C isolated (Info. Only)
BOP y BOP /R0 OP/RO RO RO i:nter Verify CST Verify sec. E-1.0 Ma int a in/ Verify Prr verify S.I.
> Tech. Spec. rad. levels restore S/C PORV's shut flows should value (1) y normal p level in ' ' be reduced s intact S/C's (4)
(1)%
Y ES 1.1 (S.I. Termination) t 1
284
MAIN STEAM BREAK INSIDE CONTAINMENT Progression of Operator Actions:
Fage 2 of 2 O RO RO RO I ermin.
Reset Phase Stop all Verify RCS y Reset S.I. A and B , high head press. stable S.I. pumpe but one
-) or incr-maning SRO/R0 RO RO i
RO RO y Verify S.I. Stop low Verify level Isolate Establish flow not head S.I. in Par. B.I.T. (5) normal required 4 4 4 4 charging flow path RO RO RO/ BOP R0 RO Reverify Establish Establish Transfer Establish Par. level / normal normal VCT charging to RUP seal press stable y letdown 7 mdkeup p VCT suctiou ; inj ec t ion BOP /R0 BOP /R0 R0 SRO/R0 BOP /R0 Verify RCP Verify Establish Verify RCS Transfer seal inject. intact SG's Pzr. HTRS. 6 T-Hot stable steam dump
& CCW norm. ( normal i sprays - 1 i to press.
(1) level (1) control P p mode
(
T BOP y R0 RO R0 BOP Verify all Start one Verify NIS Transfer NIS Step diesel A.C. buses RCP per source range recorder to g merators energized > procedure 7 energiz ed > source range >
SRO SRO/RO SRO/RO/ BOP BOP / yl RO r
Report Reverify Maintair Secure END \ NRC , S.I. not_ , plant in stable
, unnecessary j' INPO Plant Mgr.
' required '
cond it ions
' equipment
- Alarms Notes
~~- S'.I. Actuated ll)'Piant Specific Value(s)
," jg,7, (2) Emergency Condition Should be Declared and
- Steam Flow / reed Flow Mismatch IE'* * ** '9" #*
- S/C tow Pressure S.7. (3) Aux. F.W.
- Containment Hi/Hi-H1 Pressure Mn. F.W.
MSIV MSIV Bypass Sampling Blowdown (Plant Specific Lint, etc.)
(4) Subcooling - Via Incore Thermocouple Secondary Heatsink ,
RCS Press. Stable or increasing Prr. Level > (1) %
(5) B.I.T. May be N/A for Some Plants 285
rg ) s 3
Y b "
U p
}.,
(
\ - (k '
1
.i g ; . "ll (\ % ,
W Operating Sdquence: RHR CG'A t Complete Loss of all RHR , , /
l t.; ' '~
-pL b\ "y
-( NSSS/{yh? 'WestinghotqQt 'e C-6 L kR
- v. h c .4 a s -
A , {s K- 'Initia"1 Plant State: Reactcc4he.down,ExtendedPowerHidkory(,F,3Rin Ih
- n # q- Service **ti t
.n -
. Sequence Initiator: RuptureinRHdthepCausesTotalLossofRTIShut-L down. Cooling ,, ,
y :t; c ,
Important Plant Parameters: 1) RCS Temperature, patipure Flow, and Level,2)SGLevel,3).gdiationLevels,4)Auxiliar: Feed Flow, 5) '$ . ,
Containment Temperature 4, 3 ,' ( lh 3,
,1 .
Progressi6n of Operator Actions: See Flow Chart <
)
- Final Plant State: Core' cooling is bet h acchpflished by RCP. operation t.' t and steam generator feed and bleed. 'Ite audliary feed is supplying
~
the steam generaty . RHR is isolated'ans cd(:ainment integrity is established. -c ! (
s.^ ) ,
.)
/(/ t Major Plant Systems: Reactor Coolant, RHR, Aurillary Fet.derater Radiation kJ1otitoring,ContainmuutCoolingSteamGenerads p ,,
f j Tolerance Range 4 The plant.should be in a stable c'onditic,n at.about 212 (.
degrees F with>ths Tech. Specs. being satisfied and core c W41ng
' ?'aul RHR leak should be isolated and ret.uirs Jaltiated. The '
f f available.
site event klMild be declared. ql
/ Competencies Tested: ' / .,.
(l SRO - Compllance/Usj of Technh.a.i Specic,1fic ons s. I Compliance /Usi if Procedures
~
Supervisory Ability * '
.g.. Communications /C,rtMfInteraction,g
, , t R0 '- Understanding /$dterprefacin 'of Ary unciator/ Alar (s Signals Diagnosis of jfvents/Cond1Quns Based on Signals /Reasings CompliaJoe/ Uke of Procedur$c- ,
s <
s
- f. Control Board. Operation
\ Communications / Crew Interaction
- u3 v .,
' \ BOP - Compliance /Use of Procedures N Control Board Operation '
j l's- -( ' Cytmunications/ Crew Interaction 4
K. d<. 1 u ,
l> %d'< }' y
> ,a s-t ,
(' ?
(J l
,) .
., s (f+ .
f \ ,
f .'
r q, ,
s
( , av -
u j( f \ .i E s.
I p *. 4 e
s f $ $
287
.-.7
- . t> i
\ j RHU LOCA - COMP'iTE LOSS OF RHR JO Progression cf Operatoc Ar.t.kna:
3-e Q, R0
.R.O - - .., _RO _.. - - - - - .
?'
N Dia).ess e i Isolate tuR Isolate FAlk Leak stopped
- Alares. ~ p alar:asfind. , _ ,_.g.jvraia A/h __
m train B/A . with both Procedura O eak st ill. (leak still F trains E0P/AOP (1) l gresent) present) isolated (RHk
_j L,; inoperable)
,' SRO ,- 90 .SR0/A0 RO/A0 37 l
,- 1 Start CRDM Dianatch Establish (2)
. rarif y i Twh.,Spx. & Containment Operator to containment
?", ' f
- or p'laat e cooling fans find leak
- . . , 2 integrity (prior to 200 ' location and F) evaluate go k
Ma int a in RCS RO BOP Es t a blis'n Establish SC pressure using ) conditions N levels near opers,ble ... RCP (3) _g normal (3) chtg. pump r, .j)} rut.
p,
- p. _eu_ A> R0/B0f1r kua KCY Fte hug /
to equalize bloei .tr g RCS t emp. ' SCe ur.ing (4) aum, feed l
SRO feport SRO/F.0 b io M lr.
i , (T $
Nhc > Rat
} Plant Mgr 4 .h ordit ions _ (-
,8
[bistxteRF1
, , j , l l repairs
. ., _, u l
l Y
l ' END t
l'
- Ahrms
- Par Ptessure Decrease
- Area bf., Monitor
- Pzr Level "e: case
- Jump Levels ( M t.;!:
(1) Procedure plant spec ific .
(2) Si'.e emergency condition exist s - refer to site specific naag . plan.
(3) Flant specific value.
(4) Accomplished at about 212*F.
e t
288 1
NRC POAM 338 U S. NUCLEAR AlGULATORY COMMIEEIDN t MtPONT NvMata sampaea er reoc. eow ve, No , a anyi (2 841 Eo',"2E2 '
BIBLIOGRAPHIC DATA SHEET NUREG-1291
$tt INSimuCTIONS ON Yat mgvEnst 2 TITLE AND SU8f sTLt 3 LE AVE 8L ANK BWR and PWR Off-Normal Event Descriptions a DATE REPORT COMPLETED MONTH VEAR e Avi Daisi September 1987 6 D A T f N E POR T 185UE D MONTH vtAM November 1987 i Pen onMi o oacANirAnoN NAus AN6 MA L,No Aooans u,,v.,,,i. c , a causicT a As==oax uNif NvMesa Division of Licensee Performance and Quality Evaluation Office of Nuclear Reactor Regulation *"""'*"""
U.S. Nuclear Regulatory Commission Washington, DC 20555
- 10. BPONSORING QRGAmi2 ATaQN NAME AND MAILINo ADDht&S (#netwW /p tages ils TVPk DFPtPORT Technical o Pt asou covtnto tanctus,*e oewer 12 EUPPLEMtNT ARY NQf t$
13 ASSTMAC1 (200 wows er desst This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios.
l I Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence (s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities.
14 DOCUMtNT ANAsv$t8 - e at WWORD&!DESCRaPTOa5
't a vasLA8lLtf V reactor event descriptions "*""
operator licensing examiners BWR Unlimited PWR 15'cuaav c'^55'P'cA7'o*
ITng geget
. iDeNiiFieasmP N eNosD naMs Unclassified iF44 4portl Unclassified 17 NUM6t h Of P AGE 6 s
l 18 PHICE o p. 5. GD WC Phmt hf PA l h T ING Dr r l CC s 1987. 23 2-79 2 s G78 5 L______-_ - - - - - . - _ - - - - - - -