Text

Simulation Facility Certification
	ML20077G626
Person / Time
Site:	Perry
Issue date:	06/28/1991
From:	; CENTERIOR ENERGY
To:	;
Shared Package
ML20077G624	List: ML20077G626; PY-CEI-NRR-1358, Forwards Perry Power Plant Simulation Facility Certification;
References
	NUDOCS 9107020106
	Download: ML20077G626 (200)
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'~a*c^ ~ *ao'~'*"**'o'n' ' ~' oa SIMULATION FACILITY CERTIF CATION M 41 aON AND RtCOHDS MANAGE MINT BR A NC H (MN85 7 7141, US NOCLEAR HI C,UL a t DR Y COMMis$ tun, h ? r.St NGtON OC Mf4 AND 10 1 e* E cAPt HWOHst 54f Dut t tOhe PROJE CT litt.6 01 'at t OF i l( f OF MANAGtWENT AND putwit, h at mHUtON. Dt M03 NSTRUCTiONS. TNs form is to tw med 9 met.e4 coq 4anort receq Aetion Of reawren ami for say tr ante to a umu'enon tent ty performarse tssung pian maste sher mmat i I evemstial of sah a pies P onde the fonowiN m'ormation, oms che<A the apprope ate t.oa to nevnte eee.pn for sutwruttal. MP,(I NUMbk b p g ggy

440 Perry Nuclear Powgr Plant. Unii No. 1 DA1E LICE NSE E The Cle/ eland El?ctric liluminating Co. and Centerior Service Co.e et al.

6/26/91 TNs is to cert,fy that: t, The abo.e name1 tentay beensee is unr>s e umularen taal tv conmoag soiety of a pant veierenced s muistor that er eetiis rimweementi et 10 CF R t,8> 45. s 2. Documentation is evenatde for 16 C rwow m ettorcaw Mb 10 CF R M 4Mid Pledse refer to Tab B This simutet.on fealu, meets the gucese conta ned m ANS)i AN5 3 6, ly)8*a, as estornwt by N'tC Recitatory Guide 1.149 3. g gq q, c, u If there are e iy ewvpteons to the cert Asoon of this atom. ched bees [ v e<ui stweit e Nily nn edd tional pages as noteuer y ny mgj _m N AME brother Woorhet.oS/ ANO t.OCAf foN OF SIMULAllON F ACILIT Y PeTry Nuclear Powtr Plant Unit i Simulator (Please refer to Tab A "CeTtification 10 Ce. iter Road Overview" for additional simulator Perry, OH 44081 information.) }$1MULATlON F ActLITY PE RFOPM ANCE T t ST ABST R ACT S ATT ACHE D (For performence rests s 9ndacted e the perrod endmg wi otscsurnON Of PE vou AN;E TEs t iNG eow Ett to gnech edd,t.nw eemss on nern.ry.,st eenr+ the nem dm re te.rvenuwt> Please refer to Tab C " Simulator Tests" LIMULATION F ACILIT Y PE RFvRMANCE TESTING SCHlOULE ATT ACHED, (f or t8e cond#t of 80 pro 4r%Ife/r 25% of ferf nMSTe festsper Fee' for the four year l Q X oered wwwnencmq with the date of thi,, certsfaceton.) ~ DESCRIPTtON OF PE RFORMANCE T ESTING TO BE CONDUCTED. Mrtsch axAfttone/segersies nerecary, vid utentW the nem descnoten beme conteved) Pleasr refer to Tab D "Four Year Test Schedule" PE RTORMAkCE TEST ING FLAN CH ANCE. lfor sny mo#fration to s penormance terrmg plan semitted on s presmus certifocaten) DESCRIPTION OF PtnFORMANCE TESTING PLAN CHANG 5 ttach additoonalpagetsins necessery, endident% the trw orscnotaan t'ee'a conteved) N/A RECE RTIF M.AT SON (Describe cor'ect ve e: tens saa en. etrach results af comoreredper%rmance testme m accordance with to CFR j 55 bibHSHvf. Attach add tionalpagets) na necessary, and identify the itern descenpren being contmued i N/A 91070:0106 9106x PDR ADOCP c5000440 F PDR Ary faise statemeu t" cess.or in this document,includmg atta(hments, may be subject to civil and enama'saactens. I cen% urvier pena'ty of perjury that the ma rmation m o tNs document ut e.rMnntsis true and correct, AfuRE - Ataf HCM12EV EPRE,$ENTATIV TITLE DATE / f \\ ) / Vire President - Nuclear (p p h4 /YW p y g in accercance wah to CF R i 55.6, CommumcatU. this $rm shau os submrtted to the NRC as fonows: j 8Y MAIL ADDRESSED TO: Director, Off N ' or'ser R= ecto' Regutetaon BY DEllVERY IN PE RSON One vyhete Fhnt Nortti UA hesear M #T Commene6ria TO THE NRC OFFICE AT: 115% Rockotte Pike yteshmeten. E ~ 2M Rochelle MD NRC FORM a7e i1401 J

p CERTIl 'ATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 ( NRC Fot. 474 Supporting Documentation Table of Contents TAB A CERTIFICATION OVERVIEW Part 1. Simulator Information Part 2. Simulator Design Data Part 3. Simulator Discrepancy Resolution and Upgrading TAB B EXCEPTIONS TO ANS 3.5 Part 1. Permanent Exceptions Part 2. Temporary Exceptions (Discrepancies) Part 3. Schedule for Correcting Discrepancies TAB C SIMULATOR TESTS Part 1. Certification Test to ANSI /ANS 3.5 Section Cross Reference Fart 2. Computer Real Time Test Abstract Part 3. Steady State and Normal Operations Test Abstracts A. Core Performance Tests B. Continuous Plant Operation Tests C. Steady State Performance Tests Part 4. Transient Performance Test Abstracts Part 5. Malfunction Test Abstracts A. System Level Failures B. Malfunction Scenarios TAB D FOUR YEAR TEST SCHEDULE On the Cover Plant and Completion of simulator hardware milestones and installation at history S3 Technologies in Columbia, MD Factory testing Training and at S3 Education Center Technologies at Perry. PERRY POWER PLANT O SIMULATION FACILITY CERTIFICATION l JUNE 1991

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Certification Overview TAB A Table of Contents TAB A CERTIFICATION OVERVIEW Part 1. Simulator Information A. General Information B. Control Room Physical Fidelity C. Instructor Interface D. Operating Procedures Part 2. Simulator Design Data Part 3. Simulator Discrepancy Resolution and Upgrading O

1 CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO, 50-440 O Certification Overview - Simulator Information TAB A - Part 1 PAGE 1 of 4 A. GENERAL INFORMATION 1. ggner/Onerator/Hanufacturet 1 Owners: Cleveland Electric Illuminating Company 10 Center Road Perry, OH 44081 l Centerior Service Company 6200 Oaktree Blvd. Independence, OH 44131 Operator: Cleveland Electric Illuminating Company Perry Nuclear Power Plant Nuclear Training Section Manufacturer: S3 Techr. ologies Co. 8930 Stanford Blvd. Columbia, MD 21045-4752 Notes Former company names which may appear in this document include Singer, Link-Miles Simulation, and Singer Link-Hiles. 2. Reference Plant /Tvoe/Ratino -O \\,,j/ Reference Plant Perry Nuclear Power Plant, Unit No. 1 Type Boiling Water Reactor, GE Rating: 3579 MWth 3. Date Ready For Training After completion of Factory Acceptance Testing, delivery, installation, and site acceptance testing which is scheduled for completion before August 31, 1991. 4. Iyce of Reoort Initial Certification 5. Acronyms Uged Throuchout the Rooort ATP - Acceptance Test Procedure CMS - Configuration Management System (computer system) CUN - Configuration Update Notico (plant drawing change) DCP - Design Change Package (plant change document package) DCN - Drawing Change Notice (plant drawing change) IC - Initial conditions PE - Permanent Exception to ANSI /ANS 3.5 PEI - Plant Emergency Instructions SCR - Setpoint Change Request 'SDP - Simulator Discrepancy Package (CEI document) SDR - Simulator Discrepancy Report (S3 Technologies document) TE - Temporary Exception to ANSI /ANS 3.5 O TIE - Training Impact Evaluation TGIS - Third Generation Instructor's Station

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 g Certification Overview - SimJlator Information TAB A - Part 1 %g, PAGE 2 of 4 5. Acronyms (continued) THA - Training Manual Administrative Instruction TSR - Training Significance Review B. CONTROL ROOM PHYSICAL FIDELITY 1. The physical scope of the plant-referenced simulator was established in accordance with section 3.2 of ANSI /ANS 3.5-1985 with the active participation of two SRO's, one from plant operations and one from operator training. Tab A, Part 1, Attachments 1 and 2 illustrate the physical arrangement of both the Simulator Room and Control Room panels, respectively. Those Control Room panels included in the simulator design were the only panels determined to be required for conducting normal plant evolutions and for responding to the malfunctions listed in ANSI /ANS 3.5 section 3.1.2. Panels not included in the Simulator Room were evaluated as having no or minimal training impact. These as well as other simulator design limitations which were established as part of the Simulator scoping process are listed in Tab B, Part 1 " Permanent Exceptions to ANS 3.5." 2. Physical fidelity of the controls on the simulated panels, Ref. ANSI /ANS 3.5 section 3.2.2, has been verified against design- {h freeze photographs (dated July 1, 1989) of Perry Unit 1 panels in y,/ the scope of the plant-referenced simulator. 3. Design changes which physically affect Perry Unit I have been evaluated since the July 1, 1989 design-freeze date through the Training Significance Review (TSR) process (Ref. Tab A, Part 3, Section B). The plant changes which affect panels that are simulated have been handled as follows: a. Change has been or will be incorporated in the plant-referenced simulator prior to the time it is declared ready for training. Photographs of Perry Unit 1 are used in verifying the physical fidelity of the plant-referenced simulator with respect to the plant change, b. Change remains as outstanding work to be completed before June 28, 1992 and is documented under Tab B, Part 2, Section B " Temporary Exceptions to ANS 3.5" (post design freeze discrepancies). l 4. The ongoing program to ensure physical fidelity of the plant-referenced simulator consists of the TSR process (Tab A, Part 3, Section B) and PNPP Simulator Element Desk Guide 015 " Environ-mental Comparison of Simulator to Unit 1" which provides for periodic direct examination of physical fidelity of the simulator. C. INSTRUCTOR INTERFACE 1. Rescription i /s I The Perry Unit 1 Simulator Instructor Interface consists of an S3 Technologies Third Generation Instructor's Station (TGIS) version l

CERTIFICATIOt* OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Certification Overview - Simulator Information TAB A - Part 1 PAGE 3 of 4 I C. INSTRUCTOR INTERFACE (continued) 4.1, which incorporates the Advanced Han-Machine Interface (AMMI) software design. The TGIS consists of 3 fixed Workstations and 1 roll-around instructors station, one color copier, one 8 channel strip chart recorder, one eight-pen color plotter, a Laser Printer, two types of remote transmitters (full feature and limited feature), and various peripherals. The Instructors Station computer system is independent of the main simulation computer complex. An interface is provided to permit direct memory access botween the two computer systems. 2. Simulator Trainino Canabilitiqa 2.1 Initial Conditions The Simulator possesses the capability to store 230 Initial Conditions. 200 of the IC's are reserved for training use and 30 are designated as " discrepancy" snapshots which an instructor can use to store up to 15 minutes of operating data when a problem is observed. 50 IC's are designated as double password protected permanent IC's. 22 IC's are delivered from the factory for immediate use by the Perry O Training Staff. These 22 IC's include a variety of 5 operating conditiens, fission product poison concentrations, and various times in core life. 2.2 Malfunctions Malfunctions can be conveniently inserted by the instructor from either the fixed workstations or the hand-held remote devices. A series of malfunctions can be inserted (up to 300 discrete failures) simultaneously or sequentially using time delays and triggers. The capability for adding additional malfunctions is a basic design feature of the simulator. Attachment 3 to TAB A, Part i lists the generic malfunctions included in the 4 year performance test plan. TAB D is the schedule of testing for the next 4 years. The individual test abstracts (TAB C) list the malfunctions included in each test. 2.3 Other Control Features The Perry Siraulator design includes capabilities to freeze simulation, run simulation in variable slow time, and backtrack. Fast time is a selectable feature for various paran.eters and evolutions, such as xenon concentration, turbine coastdown, and drawing a vacuum. The TGIS includes many other useful features not required by ANSI /ANS 3.5. A series of tests were run at the factory to ensure all features work as intended. Documentation of testing and all design specifications are available for review. O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 [ Certification Overview - Simulator Information TAB A - Part 1 \\ PAGE 4 of 4 c. INSTRUCTOR INTERFACE (continued) 2.4 simulation of Items Outside the contr21 Room The simulation complex was d3 signed to provido as mueb realism as possible in the way the trainee interfaces to the systems and componente located outside the Control-Room. The main instructor control station includes a communication system which allows the instructor to communicate with the trainees exactly as they would communicate with local operators, including phones, radios and the GAI-Tronics internal communication system. The remote functions available allow for complete simulation of local operator actions required to perform the evolutions listed in Section 3.1 of ANSI /ANS 3.5. Remote functions can be controlled from the fixed work stations or keyed from the hand-held transmitter devices. D. OPERATING PROCEDURES Reference plant controlled procedures are used in the simulator. There are no differences. O (::)

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.m .._.s- ,s ______.__.m ~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.. _ i CERTIFICATION OF PERRY BIMULATION FACILITY DOCKET NO. 50-440 Certification Overview - Simulator Information TAB A - PART 1 ATTACHMENT 3 PAGE 1 of 3 i l - Ge.npric Mal [ unction List Each malfunction identified below is verified in one or more of the tests listed in TAB C Parts 4, 5A, and 5B. An asterisk (*) in the " Test #" column indicates that the malfunction is included in more tests than the one shown. For more complete information, ref er to the " Half unctions Tested" section of i each test abstract included in TAB C to determine the specific malfunctions utilized in any particular certification Test. Malf. ID 1211_f Malfunction Descriotion AD01 T.4.5.3.22* CYCLING SAFETY RELIEF VALVE AN01 7.4.5.1.02 ANNUNCIATOR INPUT OPTICAL ISOLATOR FAILURE AV02 T.4.5.3.27* AIR OPERATED VALVE FAILS CLOSED AV03 T.i.5.3.25 AIR OPEhATED VALVE FAILS AS IS BS01 T.4.5.3.28* BISTABLE FAILS TO TRIP BS02 T.4.5.3.27* BISTABLE SPURIOUS TRIP CB01 T.4.5.3.08* SPURIOUS BREAF.ER TRIP CB03 T.4.5.3.23 BRFAKER AUTO TRIP LOGIC FAILURE CD04 T.4.5.3.34 PREAKER AUTO CLOSE LOGIC FAILURE CB05 T.4.5.3.27* BREAKER FAILS IN CURRENT POSITION (MECHANICAL SEIEURE) O CB06 T.4.5.3.22* BREAKER FAILS IN CURRENT POSITION (LOSS OF CONTROL POWER) CB07 T.4.5.3.24 BREAKER FAILS TO CLOSE CN01 T.4.5.3.03* CONTROLLER AUTO /KANUAL FAILURE CN02 T.4.5.3.30 CONTROLLER AUTO FAILURE CP01 T 4.5.3.16 PUMP SHAFT BREAKS CP02 T.4.5.3.14* PUMP SHAFT SEIZES CP03 T.4.5.3.25* PUMP HEAD LOSS (FLOW DEGRADATION) CUO3 T.4.5.1.04 RWCU SYSTEM PIPE BREAK OUTSIDE CONTAINMENT (STEAM TUNNrl) r DG03 T.4 5.3.26* DIESEL GEN SPEED GOVERNOR FAILS DG06 T.4.5.3.26* FUEL OIL DAY TANK LEAK ED05 T.4.5.3.26 LOSS OF 4.16 KV BUS ED06 T.4.5.1.06 LOSS OF 480V BUS ED07 T.4.$.3.22 LOSS OF 120V BUS ED09 T.4.5.1.06 LOSS OF 125V DC BUS ED16 T.4.5.3.34 LOSS OF 480V MOTOR CONTROL CENTER ED17 T.4.5.1.06 LOSS OF 125V DC DISTRIBUTION PANEL EG01 T.4.5.1.07 MAIN GENERATOR LOCKDUT RELAY TRIP FWO2 T.4.5.1.09 FEEDWATER SYSTEM PIPE BREAK INSIDE DRYWELL FWO3 T.4.5.1.09 FEEDWATER SYS1EM PIPE BREAK OUTSIDE CONTAINMENT FWO4 T.4.5.3.29 FEED PUMP LOGIC FAILURE FWO8 T.4.5.3.28 FEEDWATER PUMP LOSS OF LUBRICATING OIL HXO2 T.4.5.3.25 HEAT EXCHANGER TUBE LEAK

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 ) Certification Ovorview - Simulator Information TAB A - PART * \\j ATTACllMENT 3 PAGE 2 of 3 M i n tment 3 - Generit_tiallunttien Litt Malf. ID Irat_t Halfunction Decerintion IA01 T.4.5.1.12 AIR RECEIVER LEAK IA02 T.4.5.1.12 INSTRUMENT 1.IR LINE LEAK MC01 T.4.5.1.15* CONDFNSER AIR INLEAKAGE MCO2 T.4.5.3.29 CONDENSER TUBE LEAK MS03 T.4.5.3.04 KAIN STEAM ISOLATION VALVE CLOSURE TIME VARIANCE MS11 T.4.5.1.16 STEAM SEAL HEADER PRESSURE REGULATOR FAILURE MV01 T.4.5.3.31 MOTOR OPERATED VALVE FAIL AS IS (LOSS OF CONTROL POWER) MV02 T.4.5.3.30 MOTOR OPERATED VALVE SPURIOUS VALVE OPENING MV03 T.4.5.3.01* MOTOR OPERATED VALVE SPURIOUS VALVE CLOSURT HV04 T.4.5.3.37 MOTOR OPERATED VALVE FAILURE OF AUTO OPEN 04RCUIT MV06 T.4.5.3.12* MOTOR OPERATED VALVE FAIL AS IS (MECHANICA BINDING) NM01 T.4.5.1.17 SOURCE RANCE MONITOR DETECTOR (PRE-AMP) FAILURE NM02 T.4.5.1.17 INTERMEDI ATE RANGd MONITOR DETECTOR (PRE-AMP) FAILURE NM03 T.4.5.1.17 LOCAL POWER RANGE MONITOR DETECTOR FAILURE NH04 T.4.5.3.22' AVERAGE POWER RANGE MONITOR OUTPUT FAILURE HM10 T.4.5.1.17 NEUTRON MONITORING DETECTOR DRIVE STUCK 7-~ OG03 T.4.5.1.18 OFF GAS SYSTEM LEAK UPSTREAM OF ADSORDERS OG04 T.4.5.1.18 OFF GAS SYSTEM LEAK DOWNSTREAM OF ADSORDERS PC01 T 4.5.3.25 INCREASED DRYWELL/CONTAINHENT BYPASS LEAKAGE PCOS T.4.5.3.40 SUPPRESSION POOL LEAK PTO1 T.4.5.3.04* PROCESS TRANSMITTER VARIABLE FAILURE PT03 T.4.5.3.38 PROCESS TRANSMITTER VARIABLE OUTPUT CLAMP RC04 T.4.5.3.26 REACTOR CORE ISOLATION COOLING GOVERNOR VALVE FallRE RD01 T.4.5.3.18* STUCK CONTROL ROD RD02 T.4.5.3.18* UNCOUPLED CONTROL ROD RD03 T.4.5.1.21 CONTROL ROD DRIFT - IN RD04 T.4.5.1.21 CONTROL ROD DRIFT - OUT RDOS T.4.5.3.23* CONTROL ROD ACCUMULATOR FAULT RD12 T.4.5.3.36 SCRAM OUTLET VALVE LEAK RD15 T.4.5.3.20* ANTICIPATED TRANSIENT WITHOUT SCRAM RD17 T.4.5.3.22* LOSS OF CONTROL ROD DRIVE PUMP LUBE OIL RD18 T.4.5.3.36* SCRAM DIScilARGE VOLUME DRAIN BLOCTAGE RH02 T.4.5.1.22* RESIDUAL HEAT REMOVAL SYSTEM PIPE BREAK RP01 T.4.5.3.23* ELECTRICAL PROTECTION ASSEMBLY TRIP RP02 T.4.5.1.23 INADVERTENT INITIATION OF ALTERNATE ROD INSERTION RP03 T.4.5.3.23* FAILURE OF ALTERNATE POD INJECTION TO INITIATE RPO4 T.4.5.3.23 INADVERTENT REDUNDANT REACTIVITY CONTROL SYSTEM FEEDWATER RUNBACK, REACTOR RECIRCULATION DOWNSHIFT, LOW FREQUENCY MOTOR GENERATOR TRIP ) RV02 T.4.5.3.33 RELIEF VALVE STUCK \\'~'j RV03 T.4.5.3.23' RELIEF VALVE FAILS OPEN

CERTIFICATION OF PERRY SIMUI.ATION FACILITY DOCKET NO. 50-440 Certification Overview - Simulator Information TAB A - PART 1 ATTACHMENT 3 PAGE 3 of 3 htachment 3 - Ggngrie Halfunction lig Malf. ID Test # Eg11 unction Description RYO1 T.4.5.3.24 RELAY FAILS DE-ENERGIEED RYO2 T.4.5.3.22d RELAY FAILS AS IS SLOS T.4.5.3.26* STANDBY LIQUID CONTROL INJECTION PIPING LEAK SWO1 T.4.5.1.25 NUCLEAR CLOSED COOLING SYSTEM PROCESS PIPING LEAKAGE SWO2 T.4.5.1.25 SERVICE WATER SYSTEM PROCESS PIPING LEAKAGE SWO3 T.4.5.3.31 TURBINE BUILDING CLOSED COOLING SYSTEM PROCESS PIPING LEAKAGE SWO7 T.4.5.3.29 LOSS OF COMPONENT COOLING - TURBINE BUILDING CLOSED COOLING TC04 T.4.5.3.09* BYPASS VALVE FAILURE TF01 T.4.5.3.11* LOSS OF TRANSFORMER TH01 T.4.5.1.27* RECIRC LOOP RUPTURE (DESIGN BASIS ACCIDENT LOCA) TH02 T.4.5.3.25 RECIRC LOOP PIPING BREAK TH14 T.4.5.3.33* RECIRC TLOW CONTROL VALVE HYDRAULIC POWER UNIT OIL O HIGH TEMP THIS T.4.5.3.26 GROSS FUEL FAILURE TH19 T.4.5.1.27 REACTOR PRESSURE VESSEL LEVEL INST REFERENCE LEO BREAK TH2O T.4.5.1.27 REACTOR PRESSURE VESSEL LEVEL INST VARIABLE LEG BREAK TH21 T.4.5.1.27 POWER / FLOW INSTABILITIES (IEB 88-07 SUPPLEMENT 1) TH26 T.4.4.2.10* MAIN STEAM LINE RUPTURE INSIDE DRYWELL TH27 T.4.5.1.27 MAIN STEAM LINE RUPTURE IN STEAM TUNNEL TH28 T.4.5.3.25 MAIN STEAM LINE BREAK INSIDE GUARD PIPE TUO1 T.4.5.1.28 MAIN SHAFT OIL PUMP DEGRADATION t O

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Certification Overview - Simulator Design Data TAB A - Part 2 PAGE 1 of 1 l SINULATOR DESIGN DATA i Data used to design the Perry Plant simulator is available through review of baseline information in the Configuration Management System (CMS). Types of data include (1) drawings and their change documents, i.e., Configuration Update Notices (CUNs) and Drawing Change Notices (DCNs), (2) Design Change Packages (DCPs), (3) Setpoint List information, (4) vendor and technical manuals, (5) miscellaneous plant data requested through formal " Data Requests," (6) Photos taken of Control Room panels for comparison purposes, and (7) Perry Plant System Operating Instructions. This information is available either in whole by logging e,. to the CMS computer system and viewing it or by reference to specific hardcopy documentation which is available in binders or on microfilm. The design of the simulator reflects the configuration of the referenced plant as of July 1, 1989 (the date of design freeza) with the exception that some later (updated) design information was incorporated when determined necessary and practical. The Configuration Management System provides the actual date of baselining for each design input item. i t _.... =. _ _ _ __, ,_..- - _ _.. ~ _ -,. -. _ _,

CERTIFICATION OF PERRY SIMULATION TACILITY DOCKET NO. 50-440 Certification Overview a Simulator Discrepancy TAD A - Part 3 t Resolution and Upgrading PAGE 1 of 1 SIMULATOR DISCREPANCY RESOLUTION AND UPGRADING A. Identification,_porrectina, and Testina of Discrecaneing The Perry Training section, through a certification Administrator, maintains the Perry Plant Simulator by interfacing with a matrix organisation which consists of personnel from the software engineering, j hardware engineering, instrumentation and controls, and operator training groups. Simulator discrepancies a.u identified in several ways by the certification Administrator or Software Engineer through direct review of I plant change documents shortly af ter completion of the changes in the plant; by the Software Engineer and/or Hardware Engineer during the plant change cost erstimate process when the simulator group provides simulator change cost estimates related to plant design changes; by operator trainees or instructors who make an observat ion and feed it back to the Certification Administrator using a Studen' 'notructor Teodback Report; or by anyone in the matrix organisation obours a problem while m0 king modifications or conducting daily, annual, eriodic testing. i For each discrepancy a Simulator Discrepanc,,

kage (SDP) is initiated in accordance with instruction OM14: TMA-4206 -

is processed through analysis, implementation, and testing phases by the certification Admini-i strator who also maintains appropriate tracking information in the Update Data Base portion of the computerised Configuration Hanagement System. O Instructions are available on site which describe this process in more detail. B. Trackina of Plant Desian Chances Shortly after completing a changc in the plant and declaring the modified item or syst em operable, the Perry Training Section is sent a copy of the change documents for review to determine any impact on training. The certification Administrator completes a Training Significance Review (TSR) in accordance with instruction OH14: TMA-4204 and, for those changes affecting the Simulator, initiates a Simulator Discrepancy Package and i enters the SDP number into the UN ate Data Rage for tracklng purposes. i The default due date for completing the change in the simulator is one i l year from the date the TSR was completed, however, the Operator Training Unit may assign an earlier date for higher priority changes. On an annual basis prior to the Simulator certification anniversary date, l a comparison of the current revision of design drawings, CUNs, and DCNs will be made with the revision listed in the simulator Configuration Management System. For any discrepancy identified, the certification Administrator will determine whether the item is already being tracked on an SDP initiated through the TSR process or whether a new SDP is needed. i This annual review process ensures that plant changes are implemented within the simulator preferably within one year of the associated TSR l review date but definitely no later than the next annual review date or r two years from the date of tbo change in the plant, tvhichever is earlier. l o l l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 Exceptions to ANS 3.5 TAH B Table of Cont ents TAB B EXCEPTIONS TO ANS 3.5 Part 1. Permanent Except ions Part 2. Temporary Exceptions (Discrep6nc eJ) A. Exceptions Re.sult i ng from Filing Form 474 Prior to Simulator Delivery and Installation D. Exceptions Related to Work Identified after Design Frontes were Establist>d Part 3. Schedule for Correcting Discrepancien A. Work to be Completed Prior to Declaring the Simulator " Ready for Training" D. Work to be Completed Defore June 28, 1992 O O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 (, } Exceptions to ANS 3.5 - Permanent Exceptions TAB B - Part 1 f PAGE 1 of 3 i i PERMANENT EXCEPTIONS TO ANSI /ANs 3.5 Permanent Exceptions to ANSI /ANS 3.5 are listed by *PE" number and identify the applicable ANSI /ANS 3.5 section or paragraph to which the exception is being i taken. Identified dif ferences between the plant and simulator were in the aren of control room environmental differences only. These differences, augmented with compensatory training given to operator trainees where necessary, have been determined by the Simulator Review Board to have minimal impact on training and are, therefore, acceptable exceptions. The Training Impact Evaluatione (TIES) referenced for each exception provide specific analysis and review documentation and are available at the Ferry site. PE 001 ANS 3.5 Section: 3.2.1 Degree of Panel Simulation Component Affected: Control Room Panels 1H13-P613, P618, P621, P622, P623, P625, P628, P629, P631, P652, P654, P655, P669, P670, P671, P691, P692, P693, P694, and P873 (Ref. Control Room Layout, Tab A, Part 1, ).

== Description:== Panels are physically there but are not functionally included in the scope of the new simulator. (Ref. Simulator Room 123 Floor Plan, ( Tab A, Part 1, Attachment 1). Related TIES: 91-016, 91-017, 91-018, 91-019, 91-020, 91-021, 91-022, 91-023, 91-025, 91-028, 91-029, 91-030, 91-031, 91-032, 91-033, 91-034, 91-035, 91-036, 91-037, 91-049 Testing / Training Impact: With the exception of off-normal and emergency controls (addressed by PE 002), the controls and indications on these panels that are manipulated or monitored by operators have minor impact and need not be simulated for one or more of the following reasons: a) panels for redundant divisions are simulated, b) operator actions are routine (performed daily), c) classroom training or OJT adequately covers three controls and 1 indications, d) panel has no impact on any training scenario. PE 002 ANS 3.5 Section: 3.2.2 controls on Panels component Affected: Panels 1H13-P691, 692, 693, 694, 625, 618, 629,

631, 640,

866, 871,

872, and 628 with the switches, roldys, fuses, and terminal boards that are operated during Plant Emergency Instructions (PEI's) and Off Normal Int'. uct'ons (ONI's).

== Description:== These 13 panels have been incorporated into one panel in the simulator (Unit 21,

1H13PEI, O

insta' led in the approximate location of control room panel 1H13P640).

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Exceptions to ANS 3.5 - Permanent Exceptions TAB B - Part 1 Pact 2 of 3 PERMANENT EXCEPTIONS TO ANSI /ANS 3.5 (PE 002 continued) Related TIES 91-005 Testing / Training Impact: These component s are manipulated nore in the simulator (during training) than in the plant, therefore, the operator'a knowledge of their actual plant location is important. Job Performance Measures (JPM's) have been or will be written to provide training on operation and actual location of the devices operated on the above panels. These JPM's are a required part of the training program to ensure that operators know the actual location of each device. $E003 ANS 3.5 Section: 3.2.1 Degree of Panel Simulation Component Af f ected: Unit 1 Control Room Panels 1H13-P610, Po12, P630, P637, P640, P821, P822, P840, P864, P865, P866, P867, P868, P869, P871, P872, P913 (Ref. Control Room Layout, Tab A, Part 1, Attachment 2)

== Description:== Panels are not included in the scope of tre new O simulator. (Ref. Simulator Room 123 Floor Plan, Tab A, Part 1, Attachment 1) Related TIES: 91-014, 91-015, 91-024, 91-026, ci-027, 91-038, 91-039, 91-040, 91-041, 91-042, 91-043, 91-044, 91-045, 91-046, 91-047, 91-048, 91-052 Testing / Training Impact: The absence of thera panels in the simulator has minor impact for one or more of the following reasons: a) operator actions are routine (perf ormed daily) and are addressed by OJT, b) off-normal and emergency controls are simulated on the PEI panel (see PE 002), c) classroom training and/or Job Performance Measures (JPHs) adequately address tre location and function of the panel, d) no exterr al controle or indications exist on the panel. PE 004 ANS 3.5 Section 3.2.3 control Room Eavironment component Affected: Ceiling differences in Unit 1 Control Room vs. Simulator Room.

== Description:== The Control Room upper and lower ceilings are 17' 6" and 10' 6"

higt, respectively, with l'x1' hidden spline acoustic tile in the lower ceiling and a motorized light dimming control for the upper ceiling.

The Simulator Room ceilings are 10' and 9' high, with 2'x4' lay-in acoustic tile O in the lower ceiling and air dif fuser slots in both. The Simulator lighting does not have a i dinning control.

i CERTIFICATION OF PERRY SIHULATION FACILITY DOCKET NO. 50-410 O Exceptions to ANS 3.5 - Permanent Exceptions TAB B - Part 1 PAGE 3 of 3 PERMANENT EECEPTIONS TO ANSI /ANS 3.5 (PE 004 continued) Rols'9d TIES 91-000, 91-009, 91-010, and 91-013 Testing / Training Impact: There is no impact to training due to the differences in ceiling configuration. Area lighting levels can be controlled at the Simulator Room Unit Superv isor 's workstation consintent with available lighting levels in the control Room but by using different means. Emergtney lighting levels at the Simulator Rof.1 c ont ro.'. panels will be maintained within NUREG 0700 gt idelines. Other visual dif forences in the ceilings, although noticeable upon inspection, have.egligible impact to training since no operr. tor interface with ceilings is required. PE 005 ANS 3.5 Section: 3.2.3 control Room Environment component Affected: Structure and site of Unit 1 Control Room (EL 654'-6") vs. Simulator Rooru (TEC123).

== Description:== The simulator Room is smaller in size and O different in structure f rom the Control Rocm in that 1) outside aisles to the north and south of simulator Room back panolo are missing, 2) exit and office doors are by photo mockup only, 3) some miscellaneous furniture cannot be included, and 4) an additional 15"x15" column exists in the simulator Room. Reasted TIEst 9' ~04, 91-006, 91-011, and 91-012 Testing / Training Impact: There is no impact or minor impact to training from the difference in environment. Where evaluat ions determined minor impact (i.e., with respect to missing aisles and the extra column), compensatory training has been or will be incorporated into t he. training program. The column causes negligible visibility impact since panel components were not designed to be distiguishable from the column area. The column and missing aisles cause minor impact in that a trainee may become accustomed to a slightly longer path to panels than required in the control room. O

CERTIFICATION OF PERRY SIMULATION FACIL1?Y DOCKET No. 50-440 (/) Exceptions to ANS 3.5 - Temporary Exceptions TAB D - Part 2 (,j (Discrepancies) Page 1 of 5 TEMPORARY EXCEPTIONS 70 ANSI /ANS 3.5 (DISCREPANCIES) The following lists of Temporary Exceptions (designated by "TE" numbers) represent all exceptions to ANSI /ANS 3.5 for which a schedule has been established to correct the discrepancies that created the exception. These exceptions f all into one of two categories: A) those renulting f rom filing Form 414 prior to shipment of the simulator from the vendor *s facility; or B) those resulting f rom discrepancies identified af ter f reeting design input to the new simulator on July 1, 1989. Followup to ensure closure of these exceptions is through the use of a simulator Discrepancy Report (SDR) issued by the simulator vender, S3 Technologies, or through une of a Simulator Discrepancy package (SDp) issued by CEI at Perry. Once perfonnance testing has been completed (ref erences exception TE 001), the makeup of these lists will likely have changed. Therefore, a supplement will be made to this Form 474 submittal within one month after declaring the Simulator " Ready for Training" in order to update this portion of the certification report. V (Remainder of page is intentionally blank) bd

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Exceptions to ANS 3.5 - Temporary Exceptir 1 TAB B - Part 2 (Diecrepancies) Page 2 of 5 i TEMPORARY EXCEPTIONS 'N) ANSI /ANS 3.5 (DISCREPANCIES) i geetion A The exceptions in Section A (TE 001 through TE 004) represent the work to be done l prior to declaring the Simulator " Ready for Training". They are the result of I filing Form 474 prior to shipment of the simulator f rom the vendor's f acility and installation /in-place testing of the simulator at Perry. Theco discrepancies will be handled on a high priority basis as soon as feasible but before declaring the simulator " Heady for Training." TE 001 ANS 3.5 Section

3. General Requirements; 5.4 Simulator Testing Nature of Exception:

Simulator Performance Testing has not been completed and/or major Simulator Discrepancy i Reports (SDRs) remain open rendering the Simulator "not ready for training." The " Tests included in this Section" listing in each Part to Tab c indir:ates which testo are incomplete by displaying an asterisk and either "(H/C)" for those that have not been successfully run or a number for the number of open SDRs associated with completed tests which constitute exceptions to ANSI /ANS 3.5. O Anticipated Resolutions 07/20/93 Related SDPs/SDRs SDRs See " Tests included in this Section" at the beginning of each Part in Tab C. TE 002 ANS 3.5 Section:

3. General Requirements; 5.4 Simulator Testing Nature of Exception: The Simulator is not yet installed or tested in its permanent location and, therefore, cannot function acceptably as a training device /exami-nation tool.

Remaining work to be conducted after installation includes On-Site Reverifi-cation Testing (including re-run of ANSI /ANS 3.5 App. B transients; ERIS Interface Test; and adjustment and operational verification of peripheral interfaces such as communication and lighting systems. Anticipated Resolution: 08/31/91 Related SDPs/SDRs SDPs 91-1060 and 91-1062

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Exceptions to ANS 3.5 - Terrporary Exceptions TAU B - Part 2 (Discrepancies) Page 3 of 5 TEMPORARY EXCEPTTONS TO ANSI /ANS 3.5 (DISCREPANCIES) TE 003 ANS 3.5 Section: 3.2.2 controle on Panels Nature of Exception: Various hardware items currently in use on the old simulator must be reinstalled on the new simulator after simulator installation at Perry. These items include pushbutton switches and annunciator windows. Anticipated Resolution: 08/31/91 Related SDPs/SDRs SDFs 91-1010, 91-1024, and 91-1071 t TE 004 ANS 3.5 Section: 3.3 Systems to be simulated and Degree of Completeness Nature of Exception: Hardware and Software not within the vendor's scope of work but required for proper simulation must be.:. stalled after delivery of the now simulator. Anticipated Resolution: 08/31/91 I Melated SDPs/SDRe SDPs 90-1012, 90-1031, 91-1014, 91-1015, 91-1016, 91-1023, 91-1039, 93 1051, 91-1057, 91-1058, and 91-1059 I t O f

l I i l CERTIFICATION OF PERRY SIMULATICA FACILITY DOCKET NO. 50-440 Exceptions to ANS 3.5 - Temporary Exceptions T4D D - Part 2 (Discrepancies) Page 4 of 5 TEMPORARY EXCEPTIONS TO ANSI /ANS 3.5 (DISCREPAhCIES) Sectica_D The exceptions in Section a (TK 005 through TE 008) resulted from discrepancies identified after design freezes were established. These discrepancies will be resolved prior to June 28, 1992 using the schedule shown in Part 3 to Tab B. T2 005 ANS 3.5 Section 3.1 Simulator Capabilities t Nature of Exception Setpoint changes completed in the plant between 11/07/89 and 04/10/91 munt be inplemented in the simulator. Anticipated Repolution See Tab D, Part 3 Related SDPs/SDRat SDPs 90-1045 through 90-1079, 90-1086, 90-1087, 90-1090, 90-1091, 91-1011, 91-1017 through 91-1022, 91-1027 through 91-1034, 91-1047 through 91-1050, 91-1056, and 91-1063 through 91-1060 TE 006 ANS 3.5 Section 3.1 Simulator capabilities Nature of Exception: Various changes to software must be made to implement plant changen made between 12/08/89 and 05/15/91 and to modify computer systemo used by Control Room operators. Anticipated Resolution: See Tab B, Part 3 Related SDPs/SDRs SDPs 90-1028, 90-1042, 90-104 3, 90-1044, 90-1081, 90-1084, 90-1088, 90-1092, 90-1094, 90-1098, 91-1001, 91-1002, 91-1003, 91-1004, 91-1005, 91-2006, 91-1007, 91-1008, 91-1012, 91-1013, 91-1025, 91-1040, 91-1041, 91-1069, and 91-1070 [ TE 007 ANS 3.5 Section: 3.2 Simulator Environment; 3.3 Systems to be Simtalated and Degree of Completeness Nature of Exception: Various hardware changes with related software changes need to be made to incorporato changes made to the plant after 11/15/90. Anticipated Resolution: See Tab D, Part 3 Related SDPs/S9Ros SDPs 90-1022, 90-1085, 91-1044, 91-1045, and l 91-1046 l O l

1 CERTIFICATION OF PERRY l'IMUI.hTION FACILITY DOCKET NO. 50-440 [ Exceptions to ANS 3.5 - Temporary Exceptions TAB D - Part 2 (Discrepancios) Page 5 of 5 TEMPORARY EXCEPTIONS TO ANs!/ANs 3.5 (DISCREPANCIES) i TE 008 ANS 3.5 Sectiont 3.2 Simulator Environn.ent Nature of Exception: Miscellaneous changes or additions to hardware neod to be made to make labels, nameplates, tage, operator aides, panel hardware, and furniture consistent with those items in the control Room. Anticipatsd Resolutions see Tab B, Part 3 Related SDPs/SDRs: SDPs 90-1018, 90-1080, 90-1082, 90-1083, 90-1089, t 90-1093, 90-1097, 91-1009, 91-1035, 91-1036, 91-1037, 91-1038, 91-1042, 91-1043, 91-1052, 91-1053, and 91-1061 } O I r i l l p-c. ,-m n e

CERTIFICATION OF PERRY SIHULATION FACILITY DOCKET NO. 50-440 Exceptions to ANS 3.5 - Schedule for Correcting TAB B - part 3 \\ Discrepancies PAGE 1 of 15 i SCHEDULE FOR CORRECTING DISCREPANCIES The Schedule for correcting Discrepancies provides dates for corepleting work related to Temporary Exceptions TE 002 through TE 008 (i.e., modifications not associated with performance testing at the Simulator vendor's facility). A supplement will be made to this Form 474 submittal within one month af ter declaring the Simulator " Ready for Training (RFT)" in order to update this schedule for any work related to resolving test discrepancies (i.e. Simulator Discrepancy Reports). Section A of this schedule describes work which will be completed prior to declaring the simulator " Ready for Training." Each " scheduled completion" date is listed as 08/31/91 which is the anticipated RFT date. Section B of ti.is schedule deveribes work which will be completed af ter declaring the simulator " Ready for Training" but before June 28, 1992. This work is necessary to bring simulator data and hardware into conformance with plant data and hardware. e i O (Remainder of page is intentionally blank) O l IE--

CERTIFICATION OF PERRY SINULATION FACILITY DOCKET NO. 50-440 gO Exceptions to AUS 3.5 - Schedule for correcting TAB B - Part 3 Discrepancies PAGE 2 of 15 schedule for correcting Discrepancies Tracked by SDP Number section A - To coeplete Prior to RrT fcP 90 1012 initiating Doewent N/A Plant Date N/A sche & ted Conpletion: 08/31/91 Descriptions Attach "$1mtator (mergency Power Of f' labels within 3 inches of emergency power off switches. Also instatt protective collars on each switch. SDP 90 1031 Initiating Docwent DCP $6 0075/000/00 Plant Date: 07/21/89 Scheduled Conpletion: 08/31/91

== Description:== G33

Change MPL G33f 0041 f rom f elt OPtw to f ait CLost.

SDP 91 1010 Initiating Documents h/A Plant Date W/A Scheduled Conpletion: 08/31/91 Descriptions Replacceent of Cutler Naniner, type E20 eras / depress pushtutton sultches, located on 1M13P0601 and 1N13P0680. Nordware from the old simulator is used. CDP 91 1014 Initiating Doewents N/A Plant Dates N/A Schedsted Conpletion: 08/31/91 Description LAll Display 033, ' Hydrogen Concentration' regstres en coerotional test. The previous sinulation did not provide the verlebtes necessary. Port of (Rll ATP. SDP 91 101$ initiating Doewents k/A Ptant Dates h/A scheduled Conpletion: 08/31/91 Description $ef tware sodifications to support interf ace of the 32/771Rll computer to the Encore 2040's in the new sinulator. $DP 91 1016 Initiating Doewent N/A Plant Date N/A Scheduled Conpletion: 08/31/91

== Description:== Back panet amunciators sust be re engraved to match haan factors annmelator engraving standards. Current engravings use leproper (cnts. $DP 91 1023 initiating Doceent: N/A Plant Date N/A $cheduled Conpletion: 08/31/91

== Description:== Instattation of Superphones and Plant Pubtle Address on new sipulator panels 1H13P0870, 1H13P0601, and CM13P0969. $DP911024{Initleting Doewent N/A Plant Dates N/A $cheduled Conpletion: 08/31/91

== Description:== Replacenent of simulator annunciator windows on 1N13P0680, IN13PO470, IM13P0577, and 1h13P0601, sImutator units 1 through 5. Wtrdows from the etd ainutator ere used. $0P 91 1039 Initiating Docwent: N/A Plant Date: N/A scheduled Conpletion: 08/31/91 Description Instatt two versatec aultiplexers in 1H13P0874 (Unit 14) erd associated cabling between basic control units in IC22P0001 & iC22P0002 (Unit 6) & V 60 Printer /Ptotter. l .m ..-__.,_y ._,-._.,_.~,,m.

CERTIFICATION OF PERRY SIMULATION TACILITY DOCKET NO. 50-440 /mT Exceptions to ANS 3.5 - Schedule for Correcting TAB B - Part 3 () Discrepancies PAGE 3 of 15 ) Schedule for correcting Discrepancies Tracked by SDP Number Section A - To Complete Prior to RPT SDP 91 1051 Initiating Docment: N/A Plant Date: W/A sche < bled Ccmpletion: 08/31/91

== Description:== Provide a neens for simulator comrmication egalgenent to operate in either "E Plan Made" (tied into plant conen, systen) or " Training Mode" (isolate /stnulate plant ccerunication systee). SDP 91 1957 Initiating Doctrent: N/A Plant Date N/A scheduled Conptetton: 08/31/91

== Description:== Revise the stis valve position calculations for AC powered valves that use the Contr<,L Room volve status lights. SDP 91 1058 Initiating Docments k/A Plant Date: N/A scheduled cceptetton: 08/31/91

== Description:== Update scales f or neters th27R0411A/s,1043R0062A/B and iC34t0608 per Mmen Factors initiated W s 90 4328, 90 4325, 90 4326 and 90 4329. SDP 91 1D59 initleting Docwent: DCP 68 0052/001/00 Plant Date: 07/15/89 Scheduled cceptetton: 05/31s91

== Description:== Reverse Sailey to trot ter Mcter output so "Close" is on lef t of scale ard a pen a u p) is on right of L. ale. (v SDP 91 1060 Initiating Docment N/A Plant Date: N/A schedAed Cocpletion: 08/31/91

== Description:== Af ter Installation of the new slaulator, adjust ceiling sixrited energency light cans to illwilnate appropriate areas of the control room. SDP 91 1062 Initiati M went: N/A Plant Datet k/A scheduled Conpletion: 03/31/91 ~ Descripttr.+. 53 Technotegies will te responstele for renovel of the old sisulator ard installation of the new sleutator. At ter power *@, post instattation testing must te cordac t ed. sDP 91-1071 initiating Docunent: N/A Plant Date: W/A scheckled Conpletion: 08/31/91

== Description:== Renove two black tels screen berets f rcvs entsting sipulator panel iH13P0680 ard Instatt on reptocement sinulator tritt 2 (1H13P06BO). ym v

i CERTIFICATION OF PERRY SIHULATION FACILITY DOCKET lio. 50-440 j I Exceptions to ANS 3.5 - Schedule for Correcting TAB B - Part 3 Discrepancies PAGE 4 of 15 i Schedule for Correcting DiscrepancAes Tracked by SDP Number section B - To Complete Before June 28, 1992 i SDP 50 1018 initleting Doement: N/A Plant Date W'A scheduled Cc q letion: 09/15/91

== Description:== Attach MPL rumplates on proper ewigment ef ter delivery of new slauMor Sir the latest copy of D 128 051, Cceputer Roca and instructor's Room EcN rm.t La yout. i $DP 90 1022 Initiating Doceent: DCP 90 s e42/001/00 Plant Date: 12/22/90 sche & ted Conptetton: 09/15/91 Description Replace current R61 Sequence of Events Recorder CPU & LPU (Rll) and printer (terminet) with a Beta Prodxts $tt and DEC LA 210 printer, i P SDP 90 1028 Initiating Doement: DCP 89 0259/000/00 Plant Date: 12/08/89 scheeled Copptetton: 12/08/91

== Description:== E31 Relocate RWcu pmp roco tecperature elements. $DP 90 1042 initteting Docment: CSCO 30 Plant Dates N/A sche &ted Cmpletion: 06/28/92

== Description:== Modify program AUX to run on the new sinutator. OV $DP 90 1043 Initiating Docment: DCP 88 0124/002/00 Plant Date: 05/18/90 scheduled Cc.pletion: 09/15/91

== Description:== TBCC Auto tecp regulator volve to Mf P heat exchanger. t $0P 90 1044 Initiating Doement DCP t8 0174/001/00 Plant Date: 03/13/90 scheduled Conpletion: 03/13/92

== Description:== 1N27 Ad:f terperature control heating element to MFP ttbe oil reservoir. SDP 90 1045 Initiating Document SCR 1 89 1508 Plant Da'en 02/05/90 sche &ted Copptetton: 10/04/91

== Description:== 1P45 Reise strainer dif ferentlet pressure for backwash to 3.75 PSID iner. MPL 1P45N0235 SDP 90-1046 Initiating Docment: SCR 1 89 1037 Plant Date: 11/13/89 sche &ted Conpletion: 10/04/91

== Description:== 1P45

Revise MPL IP45N0251B setpoint.

$DP 90 1047 Initiating Document: SCR 1 89 1045 Plant Date: 01/08/90 Scheduled Coppteticm 10/04/91

== Description:== 1P51 Lower service air corpressor autostart setpoint to 107 PSIG dec. MPL 1P51N0185 SDP 90 1048 Initleting Docment: SCR 1891046 Ptont Date: 01/05/90 $cheduled Conpletion: 10/04/91

== Description:== 1P51 Reise service air receiver tank (cw pressure alarm to 112 PslG dec. MPL x iP51N0070

t CERTIFICATION OF PERRY SIMULATION TACILITY DOCKET No. 50-440 ( Exceptions to ANS 3.5 - Schedule for correcting TAB B - Part 3 Discrepancies PAGE 5 of 15 schedule for correcting Discrepancies Tracked by SDP Number Section B - To Cooplete Before June 28, 1992 SDP 90+1049 Initleting Docwent: SCR 1891047 Plant Date: 01/29/90 sche &ted Cceptetton 10/04/91

== Description:== 1P51

Revise MPL iP51h0195 setpoint to reflect reset value.

$DP 90 1050 Initiating Docwent: SCR 0 891046 Plant Date: 12/05/89 sche &ted Conpletion: 10/04/91

== Description:== OP43 Revise MPL OP43h0351 A setpoint ard reset value. $DP 90 1051 Initiating Docwent: SCR 0 89 1047 P1 ant Datei 12/13/89 sche &ted Cenytetton: 10/04/91

== Description:== OP43

Revise MPL 0F43NO3518 setpoint and reset value.

SDP 90 1052 Initiating Doewert: SCR 0 891048 Plant Date: 07/07/90 oche&ted Conpletion: 10/04/91

== Description:== OP43 Revise MPL OP43h0351C setpoint and re, i value. PJ SDP 90 1053 Initiating Docunent: SCR ' 89 0070 Plant Date: 02/06/90 sche & Led Coppletion: 10/04/91 Descript ion: 1017 Revise MPL 1017t070 setpoints $0P 90 1054 Initiating Docwent SCR 2 89 1019 Plant Date: 01/29/9', scheduled Copptetton: 10/04/91

== Description:== 2P51 Lower service air conpressor autostart setpoint to 107 PSIG dec. MPL 2P51N0185 SDP 90 1055 Initiating Docwent: SCR 2 89 1021 Plant Date: 01/29/90 scheduled Conpletion: 10/04/91

== Description:== 2P51 Revise MPL 2P5th0195 setpoint to incorporate reset Information. SDP 90 1056 Initiating Docunent: SCR 2 89 1022 Plant Date: 01/29/90 schedated Corpletion: 10/04/91

== Description:== 2P52 Revise MPL 2P52h0195 setpoint to incorporate reset inf ormation. SDP9010h initiating Docwent SCR 1 90 1014 Plant Date: 02/06/90 ScNdsted Copptetton: 10/04/91

== Description:== 1E31 Reise RCic Equipnent Room isolation dif ferential tenperature setpoint to 69.4 DEG F. MPL 1E31h0603A SDP 90 1058 Initiating Docunent: SCR 1 90 1015 Plant Date: 02/06/90 scheduled Copptetton: 10/04/91

== Description:== 1E31 Raise RCic Iquirrent Room isolation dif ferentist teoperature setpoint to 69.4 DEG F. MPL 1E31h0603B

CEPTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 ,m Exceptions to ANS 3.5 - Schedule for Correcting TAB B - Part 3 (G} Discrepancies PAGE 6 of 15 Scheduln for Correcting Discrepancies Tracked by 8DP Number Section B - To Complete Before June 26, 1992 $DP 90 1059 Initiating Doewent: SCR 1 P9 1030 Plant Date: 01/16/90 scheckled Cceptetton: 10/04/91

== Description:== 1M41 Raise MPL iM41h0280 setpoint to 40 Of G F. SDP101060lInitiating Docwent SCR 1 89 1031 Plant Date: 01/16/90 scheckled Cceptetion: 10/04/91

== Description:== 1M41 Reise MPL iM41N0281 setpoint to 40 DtG F. $DP 90 1061 Initiating Doewent: SCR 1 89 1032 Plant Date: 01/16/90 Sche W ied Ccmpletion: 10/04/91 Descriptl.a 1M41 Reise MPL iM41h0282 setpoint to 40 DIG f. $DP 90 1062 Initiating Doewwnt SCR 1 89 1033 P. ant Date: 01'16/90 schecbled Cceptetton: 10/04/91

== Description:== 1M41 Reise MPL iM4th0283 setpoint to 40 DEG F. /3 ? ) J SDP 90 1063 Initiating Docwent: SCR 1 89 1458 Plant Date: 11/07/89 scheduled Cceptetton 10/04/91

== Description:== IP42 - AM setpoint f or MPL 1P42h0044A. ^ SDP 90 1064 Initiating Dxwent: SCR 1 09 1459 Plant Date: 11/22/89 Scheduled Coppletion: 10/04/91

== Description:== 1P42 AM setpoint for MPL 1P42h00448. $0P 90 1065 Initiating Doewent: SCR 1 89 1506 Plant Date: 01/22/90 scheckled Cceptetton: 10/04/91

== Description:== 1P45 Raise strainer dif f erential pressure f or backwash to 3.75 Pl!D inc. MPL 1P45h02204 hP901066 Initleting Doewent SCR 1 89 1507 Plant Date: 02/23/90 Schecbled Crepletion: 10/04/91

== Description:== 1P45 Reise strainer dif f erential pressure f or backwash to 3.75 PstD inc. MPL 1P45h02208. SDP 90 1067 Initiating Doewent: SCR 0 89 0058 Plant Date: 11/28/89 scheduled Ccepletion: 10/04/91

== Description:== MPL OG4th0368A - Reise low levet annunciator setpoint to 608'45" which corresponds to a tank levet of approximately 90 inebes. CDP 90 1068 Initiating Docwent $CR 0 89 0059 Plant Date: 11/28/89 Scheduled Ccopletion: 10/04/91 O (

== Description:== MPL 0041h03688 Raise low levet annmelator setpoint to 608'45" which corresonnds to a tank levet of arpecalmately 90 inches. N l l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 Exceptions to ANS 3.5 - Schedulo for Correcting TAB B - Part 3 Discrepancies PAGE 7 of 15 Schedule for Correcting Discrepancies Tracked by SDP Number Section B - To complete Before June 28, 1992 EDP 90 1069 initleting Docunent: SCR 1 88 1505 Plant Date: 12/05/89 Sche &ted cceptetton: 10/04/91

== Description:== MPL 1R2300614R Change the setpoints of tus urdervoltese eterm relay WGV15A(271) located in EF1D01 es f otlows: dropout: 96.0V, +4.0V/ 1.0V; pick w 109.0V men. $DP 90 1070 Initiating Doewent: SCR 1 88 1566 Plant Date: 12/05/89 Sche &ted Cocpletion: 10/04/91

== Description:== MPL 1E2200001 - Change the setpoints of retey lit 27H(27G) located in It?2P0001 es follows: pic k up 95.5v, +0.5v/ 5.CV; dropout 86.0V mininun. $DP 90 1071 Initleting Dxunents sta 2 89 0004 Plant Date: 12/05/89 sche &ted Conotetton: 10/04/91

== Description:== MPL 2R1107001C Correct setpoint of tine delay drop out relay 271 (CE type CR28208120AA4, 480v) to 4.0 +/ 0.2 sec. EDP901072l Initiating Docwent: SCR 2 89 0002 Plant Date 11/07/89 Sche &ted Ccepletion: 10/04/91

== Description:== MPL 2R11Q7001A - Correct setpoint of tine deley drop out relay 271 (GE type CR2820B120AA4, 480V) to 4.0 +/ 0.2 sec. $DP 90 1073 Initiating Doement: SCR 2 89 0003 Plant Date: 10/16/90 Sche &ted Cceptetion: 10/04/91

== Description:== MPL 2R1107001B Correct setpoint of tine deley drop out relay 271 (CE type CR2820B120AA4, 480V) to 4.0 +/ 0.2 sec. SDP 90 1074 Initiating Doewent: SCR 2 89 0005 Plant Date: 11/07/89 Scheduled Cceptetton: 10/04/91

== Description:== MPL 2R1107002A Correct setpoint of tine deley drop out relay 27 2 (CE type CR28208120A44, 480V) to 4.0 +/ 0.2 sec. $DP 90 1075 Initiating Docunent: SCR 2 89 0007 Ptent Date: 11/07/89 Sche &ted Coppletion: 10/04/91

== Description:== MPL 2R1107002C Correct setpoint of time deter drop out relay 27 2 (CE type CR28208120AA4, 480V) to 4.0 +/ 0.2 sec. $DP 90 1076 Initiating Doewent: SCR 1 88 1565 Plant Date: 11/13/89 sche & ted Cceptetton: 10/04/91

== Description:== MPL 1R2?00804 - Change the setpoints of relay hCV22(59D) located in penet (H1301 es follows: pickup: 95.5v, *0.5V/ 5.0V; dropout 86.0V minleun. SDP 90 1077 Initiating Doewent: SCR 1881579 Plant Date: 02/17/90 Schecksted Coppletion: 10/04/91

== Description:== MPL 1R4?o0601R Cherege the setpoints of retey 12NCv18A2A(27DC) located in ED 1 A es follows: dropove: 113.0v, +/ 0.5v; pickw: 130.0V maalaun. $DP 90 1078 Initiating Docwent. SCR t ?^ W S Plant Date: 11/07/89 sche &ted Conotetton 10/04/91

== Description:== MPL 1R45N01909 Change the low, low low and storm setpoints arv3 reset values for the fuel ott storage innk levet transmitter per LCR 1 89 1438.

I CERTIFICATION OF PERRY SIHULATION PACILITY DOCKET NO. 50-440 Exceptions to ANS 3.5 - Schedule for Cc'rrecting TAB D - Part 3 i Discrepancies PAGE 8 of 15 Schedule for correcting Discrepancies Tracked by SDP Number Section B - To Complete Before June 28, 1992 $DP 90 1079 Initiating Docunent SCR 2 89 0006 Plant Date: 11/05/89 sche &ted Conpletion: 10/04/91

== Description:== MPL 2R11070028 - Correct setmlnt of time detty drop out relay 27 2 (ct tyre CR28208120AA4, 480v) to 4.0 */ 0.2 sec. SDP 90 1080 Initiating Docment: $01 B13/000/01 Plant Date: 09/28/90 sche &ted Copptetton: 09/15/91

== Description:== Fcbricate and instat t on 1H13P0&B3 Operator Aid naaeplates for suppression Pool f erperature Recorders 1D23R0000A & B per TC 1 of $01813 Rev. 0 ( Attachment 2). SDP 90 1081 Initiating Docwent: Dtr 89 0003/001/00 Plant Date: 12/12/90 scheduled Cceptetton: 09/15/91

== Description:== Revise B33 Reactor Recirc Ptry breakers 3A and 4A control power stopty to tus 0 i* 8 (was powered by D 1 A). SDP 90 1082 Initiating Doment: DCP 90 0209/000/00 Plant Date: 11/01/90 Scheduled Cceptetton: 09/15/91

== Description:== Acks to, correct, or relocate various naneptates on Control Room Penets. SDP 90 1083 Initiating Docwent: Dtu 3317 Plant Date: 10/30/90 sche &ted Copptetton: 09/15/91

== Description:== Revise ts.Mts on various control room penets, includes (kacwentation to prevlcus namepl a t e/ annunc i a t or / mi mi c c or r ec t i ons. SDP 901034 Initiating Docunent t DCP 90 0086/000/00 Plant Date: 11/15/90 Scheduled Cwptetion: 12/06/9'

== Description:== P42/P45 ESW rotation of 8 spectacle flanges on ECCW sysiera. Af fects P42//45 System nodels. SDP 90 1085 Initiating Docunent: DCP 88 0163/000/00 Ptant Date: 11/15/90 Sche &ted Conpletion: 09/15/91

== Description:== E12 RMR A:M sliding link terminal blocks to eliminate lif ting leads during svi. Affects 1H13 Pfl. SDP 90 1086 Initiating Document: SCR 1 90 1221 Plant Date: 09/25/90 sche & led Conpletion: 12/06/91

== Description:== Decrease MPL tt34K0636 setpoint (which annunciates 1M13P0650 7 D1) to 1040 psig increasing; LAll +/ 4.5 psig; reset >1029.5 psig decreasing. SDP 90 1087 Initiating Doctrent: SCR 1 90 1135 Plant Date: 09/12/90 Sche &ted Conpletion: 12/06/91 Ducription: MPL 1$1100023 - Add setpoint for Relay SPR (63) located in penet 110 PV 8. Relay setting: Stardard (Similar to 1R1100004,1R1100014,1R1100024,1511000154. B,C) SDP 90 1088 Initiating toewent: DCP 89 0189/001/00 Plant Date: 11/26/90 scheduled Cceptetton: 12/06/91

== Description:== M42 Turbine Power Corplex vent system DCP Rev i accepts flow of register RPC515 at 290 cfm as balanced.

l i CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 ( Exceptions to ANS 3.5 - Schedule for Correcting TAB D - Part 3 Discrepancies PAGE 9 of 15 schedule for correcting Discrepancies Tracked by BDP Number Section B - To Cooplete Before June 28, 1992 SDP 90 1089 Initteting Docunent: DCN 03119 Plant Date: 05/24/90 scheouted Cceptotion: 09/15/91 Description Correct latet on seltch plate RT1 $1 (Control Room Lighting, Master Controt $ witch) on 1N13P0895 to agree with Drawing B 208 219 01 ord field conditione. SDP 90 1090 Initiating Doement: SCR 1 88 1524 Plant Date: 08/27/90 Sche & ted Completion: 12/19/91

== Description:== Modif y dropout /plchp of MPL 1R2200815 relay. $0P 90 1091 Initiating Docunent: ECR 1 89 0007 1 Plant Date: 07/05/90 sche & ted Completion: 12/19/91

== Description:== Change NCC Purp to RWCU B to flow storm to 17 gra (was 20 gin). Ref. MPL iP43N01938 $DP901092l Initiating Docuwnt DCP 87 0002/001/00 Plant Date: 12/06/90 scheduled Coupletion: 12/19/91

== Description:== Holes cut in DW floor dratn surp welr box change fitt rates with respect to levet inst rutent at ion. MPL G61/1!31 SDP 90 1093 Initiating Docunent: DCP 90 0072/000/00.Lant Date: 12/05/90 scheeled Ccrptetton: 09/15/91 Descriptices Correct deficiencies in 1N13 penets. (tabel changes / corrections) $0P 90 1094 Initiating Docuiient: DCN 02840 Plant Date: 02/15/90 scheduled Conpletion: 12/18/91

== Description:== Logic to low cordenser pressure rurbacks. Contact used to close when BPV81 tef t fuit closed position; now contact closes when BPvf1 reaches full open. SDP 90 1097 Initiating Docunent DCN 3015 Plant Date: 03/29/90 Scheduled Conpletion: 09/15/91

== Description:== Correct MPL ard recorder pen designations to 017 radiation monitors on 1N13P0600. $DP 90-1098 Initiating Docunent: DCN 3134 Plant Date: 07/11/90 Scheduled Cceptetton 12/17/91

== Description:== Correct operating points for F43. SDP 91 1001 Initiating Docwent DCP 89 0157/000/0A Plant Date: 12/23/90 scheduled Conpletion: 01/15/92 Descriptions feedwater seet injection Pwps 1W27C0005A & B changed frca 100 HP to 125 HP motors; lepetter increased f rca 121/2" to 13". $0P911002lIr ;;bting Document: DCN 03343 Rev. O Ptont Date: 11/09/90 scheduled Cceptetton: 12/21/91

== Description:== Optical isolator on B 205 222 sh. 103 la given the identity "It12A A12."

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 f} Exceptions to ANS 3.5 - Schedule for Correcting TAD D - Part 3 V Discrepancios PAGE 10 of 15 schedule for correcting Discrepaccles Tracked by sDP Number section B - To complete Before June 28, 1992 l$0P911003 Initiating Document DCN 03228 Rev. O Plant Date: 09/10/90 Pche&ted Compl et ion: 12/21/91

== Description:== HPL swap on B 208 015 Sh. 16 & 15. Per 1$R, change in FDS & Data Listings only. Affects $3 verlebles, setpoints, 1/0 kap and Instructor station (t/0 override). SDP 91+1004 Initiating Documentt DCN 03220 Rey, O Plant Date: 09/11/90 Scheduled Conpletion: 12/21/91

== Description:== DG toad / time secunce changes incorporated into plant refe ence drawings. SDP 91 1005 Initiating Docunent DCP 85 0295/001/01 Plant Date: 11/27/90 Sche &ted Coppletion: 12/21/91 Descriptions itectrical and Instruretation for Intermediate Building tub Exhaust incitding enrunciators and Process Conwter points. $DP V1 1006 initiating Docunent DCP 89 0066/000/00 Plant Date: 01/04/91 scheduled Conpletion: 01/15/92

== Description:== Revision of 1N25 tevet Atorm Devices n V SDP 91 07 Initiating Docunent: DCP 83 0377/000/00 Plant Date: 01/21/91 Scheduled conviction: 01/31/92

== Description:== IC85 Bourdon tube type pressure transmitters replaced with Rosemount type 1151 electronic pressure transmitters. Af fected MPLt iC55N0001A/S. Response of loop changes SDP 91 1008 initiating Docunent: DCP 90 0225/000/00 Plant Date: 12/11/90 Scheduled Conpletion: 12/21/91

== Description:== Increase ptmp Inpetter slie for 1P45C00018. EDP 91 1009 Initiating Docment: N/A Plant Dates N/A Scheduled Congletion: 09/15/91

== Description:== Addition of non functional relays and wiring herraceses in Panel N13 Pfl. Added electrical components in panel will present a more realistic appearance to operators. $0P 91 1011 Initiating Document: N/A Plant Dates N/A $chestel Cc T Mion: 06/28/92

== Description:== Restore the cRIS setpoints for p.rp run status to the plant values. f; 1012 Initiating Docunent: DCN 03,13 Plant Date: 01/10/91 ScSeduled Conotetton: 01/31/92

== Description:== Revision to MPL of optical isolator on B 208 222 00414, mey af fect MF AN01. Previously the drawing bhowed isolator 1[51 Ai$ as an inrut to annuncletor P601 21A C3. L

I CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Exceptions to ANS 3.5 - Schedule for Correcting TAB B - Part 3 d Discrepancies PAGS 11 of la Schedule for Correcting Discrepancies Tracked by SDP Nuimber section B - To Complete Before June 28, 1991 SDP 91 1013 initiating Docunent: DCN 03192 Plant Date: 01/02/91 scheduled Conpletion: 02/04/92

== Description:== Solenoid valves OM25F0250A/8 taken of f of $809 354; J930 heeder. These are swl1ed through H51P0152 & H51P0153 (J902 & J904 respectiveLy). SDP 91 1017 Initiating Document: SCR 1 90 1637 Plant Date: 01/09/91 scheduled Corpletion: 02/05/92

== Description:== Change of setpoints for 1C91N0652A/8/C/D. SCR's 1 90 1637 thru 1 90 1640 are covered by this ist and SDP. SDP 91 1018 Initiating Document: SCR 1 90 1154 Plant Date: 01/09/91 scheduled Conpletion: 02/05/92

== Description:== SCR changes setpoints for IE31N0680A/B/E/F. At the time of review, only 1E31N0680F required revision per SCR 1 90 1154. 5DP 91 1019 Initiating Docunent: SCR 1 90 1011 Plant Date: 07/05/90 schedated ccepletion: 02/05/92

== Description:== Change to 1019K0300. This SDP also covers ist 10781, SCR 1901012, for changes to 1019K0400. i SDP 91 1020 Initiating Docu,ent: SCR 1 89 1270 Plant Date: 05/17/90 scheduled Conpletion: 02/05/92

== Description:== Revise Diviolon 1 Dieset Generator HI/LO Lube Olt Tenperature setpoints. SDP 91 1021 Ir.itleting Document: SCR 1 89 1008 Plant Date: 06/07/90 scheduled Conptetton: 02/05/92

== Description:== Added setpoint for HB ventilation heaters LO Tenp swltch MPL iM41N0284 Switch added by DCP 88-0293. ( Also check setpoints for 1M41N0280, WO281, N0282, & N02&3.) SDP 91 1022 Initiating Dor atent: SCR 1 90 1008 Plant Date: 05/14/90 schedated Conpletion: 02/12/92 Descriptlero incorporate setpoint 1 90-1008 for 1P12N0010A ' 7el switch. This $DP will also incorporate TSR 10890, SCR 1-901009 for iPt N los levet switch. SDP 91 1025 Initiating Docunent: Revision to PEls Plant Dates N/A schedsted Coupletion: 06/28/92

== Description:== Revision of PEls odds additional operator actions to perform system logic bypass operations (overrides). SDP 9P 1027 Initiating Document: SCR 1 88 1482 Plant Date: 12/04/90 sche /.iled Conpletion: 03/07/92

== Description:== New setpoint of 1E31NO351A 135 Deg.F incr. Same for li31NO3518 (TSR 10976,$CR l 88 1483); 1E31NO351C (isR 10979,$CA 1 88 1484); 1F31NO3510 (TSR 109M9,$CR 1 88 1485) SDP 91 1028 Initiating Docunent: SCR 1-90 1013 Plant Date: 05/09/90 scheduled Conpletion: 03/08/92 Descriptions setpoints for sununer and winter flow alarms for 1019K0500. l l

1 CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Exceptions to ANS 3.5 - Schedule for correcting TAB B - Part 3 \\ Discrepancies PAGE 12 of 15 Schedule for correcting Discrepancies Tracked by SDP Number Section 5 - To Complete Before June 28, 1992 SDP 91 1029 Initiating Docunent: $CR 1 90 1252 Plant Date: 01/11/91 scheduled Cospletion: 03/07/92

== Description:== Change setpoint of 1N23N0041 from 35 to 39 paid. Also include SCR 1 90 1253 (1st 10973) to change setpoint of 1Ni3N0042 f rom 30 to 34 paid. SDP 91 1030 Initiating Docunent: SCR 1 90 1632 Plant Date: 02/11/9' scheduled Completion: 03/07/92

== Description:== Revise setpoint of 1P45N0251B to 7000 spa. SDP 91 1031 Initiating Docunent: SCR 1 90 1708 Plant Date: 12/18/90 sc. duled Conpletion: 03/07/92

== Description:== Revise MI/LO alarm setpoint of 1R45N0080. SDP 91 1032 Init ating Docunent: $CR 1 90 1808 Plant Date: 12/21/90 Scheduled Completion: 03/07/92

== Description:== Revise setpoint of 1G61N0025. d CDP 91 1033 initiating Docunent $CR 1 90 1054 Plant Date: 12/18/90 Scheduled Conpletion: 03/10/92

== Description:== Modify setpoint for 1G41K0026. This $DP also captures SCR 1891494 for 1G41N0111 (1sR 0011030). SDP 91 1034 Initiating Document: SCR 1 90 1709 Plant Date: 12/18/90 sche & Led Conpletion: 03/10/92

== Description:== Revise setpoints for 1R45N0090 Secondary Fuel of t Pu@. Also; this SDP captures SCR 1 90 1710 for 1R45N0100 (TSR 11022) setpointe for Primary Transfer Pupp. $DP 91 1035 initiating Docunent: N/A Plant Date: N/A Scheduleu Conpletion: 09/15/91

== Description:== Operations regaest to correct / add three labels for h41 relays on 1H13P0847. SDP 91 1036 Initiating Docunent: N/A Plant Date N/A Scheduled Coppletion: 09/15/91

== Description:== Operations request to correct change to nameplate for recorder IN31R0005 on 1M13P0823. SDP 91-1037 initiating Docutent: N/A Plant Dates N/A Scheduled Conpletion: 09/15/91

== Description:== Operations Request added two supplementary operators aldes (Labels) on 1H13P0632 at recorders 1E31R0608 and 1E'i1R0611. SDP 91-1038 Initiating Docunent: N/A Plant Date: N/A Scheduled Conpletion: 09/15/fr1 Descriptio7: Three red f amiculd tags with white lettering installed on 1H13P0680 per Operations request. l

CERTIFIOATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 [s V} Exceptions to ANS 3.5 - Schedule for Correcting TAB B - Part 3 Discrepancies PAGE 13 of 15 Schedule for Correcting Discrepancies Tracked by SDP Humber Section B - To Complete Before June 28, 1992 SDP 91 1040 Initiating Docunent DCN 03424 Plant Date: 01/31/91 Scheduled Conpletion: 09/11/91

== Description:== Instrunent air supply for SCv's OM25F0022A & B are corrected to J Headers J944 & J943, respectively. These valves supply danpera OM25F0020A & B, respectively. SDP 91 1041 initiating Document: DCN 03291 Plant Date: 02/15/91 Scheduled cceptetton: 03/18/92 Description Lifted lead #39 00277, docuented on DtNs 3208 & 3291, disables clearwell pep trip, OP20C0001A, B, C, on to lo basin levet as sensed by level switch OP20N0044. SDP 91 1042 Initiating Docutent: N/A Plant Date N/A Scheited Conpletion: 09/15/91

== Description:== Fabricate and attach "dynomarker" tape to identify the location of IRM's and SRM's on the full core display on Unit 2 (1H13P0680). SDP 91 1043 Initiating Docunent: N/A Plant Date N/A Scheduled Cceptetton: 09/15/91

== Description:== Fabricate and instat t fuse clips on the 76 fuses f or the SRV's and the 8 fuses for the RPS soters:ds Located in the rear of Unit 21 (1H13 PEl). SDP 91 1044 Initiating Document: N/A Plant Date: N/A Scheduled Ccopletion: 06/28/92

== Description:== Add LE0s for ' first hita indication f or Reactor Recirculation Flow Control analog lockups (trips) inside rear of panet 1H13P0614 SDP 91 1045 Initiating Document: N/A Plant Date* N/A Sche &ted Completion: 06/28/92

== Description:== Add 2 red LED's and 2 push buttons to sinutate 1933K0657A&B inside H13P0634 The devices are mounted on cards marked R/C/L in Racia 1 and 2 of Nest 5. SDP 91 10 Initiating Docunent: N/A Plant Date: N/A Scheduled Cceptetion: 06/28/92

== Description:== Add 4 ' tux Estimator Status LEDs (MAINT/ FAIL /APRM/EST) in the back of panel 1H13P0t34 and make sof tware changes as necessary to drive the lights (4 t/0 points). SDP 91-1047 Initiating Docuwnt SCR 191-0007 Plant Date: 01/31/91 Scheduled Copptetton: 09/15/91

== Description:== Turbine 1st stage pressure switches set to 212 psig as input for RPS (1C71N0652C). This SOP also covers TSR 11344 for sieller change to 1C71N06520. SDP 91 1048 19ttleting vocunent: SCR 1 90 1842 Plant Date: 01/15/91 Scheduled Conpletion: 09/15/91

== Description:== Pressure setpoint decreased from 126 psig to 114 psig for pressure switch 1h43N0110A - Software only. SDP 91 1049 Initiating Docunent: SCR 0 91-1010 Flant Date: 02/19/91 Schasuled Conpletion: 04/08/92 (

== Description:== Revise ALERT setpoir.t of Oc21K0322 to 14 MR/Hr f rom 10 MR/Hr. Similar change to 0021K0332 i= included in this SDP under ISR 11288; SCR 0 91 1011. i l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 \\ Exceptions to ANS 3.5 - Schedule for Correcting TAB B - Part 3 (d Discrepancies PAGE 14 of 15 Schedule for Correcting Discrepancies Tracked by SDP Nussber Section B - To Complete Before June 28, 1992 SDP 91 1050 Initlating Document: SCR 1 90 1235 Plant Date: 02/26/91 Scheduled Carmletion: 09/15/91

== Description:== Revise setpoint for 1R42000015 tion delay to 5 sec. At the time of TSR, $3 Setpoint list gave this value as 2 sec. SDP 91 1052 initiating Docunent: N/A Plant Date: N/A Scheduled Conpletion: 09/15/91

== Description:== Dbtain switch and knob for unit 02A10 (knob to the left of PPC crt located on 1H13P0680). Install switch and knob on new aisulator untt 02A10. SDP 91 1053 Initiating Document N/A Plant Date: N/A f.Cie&ted Completions !/15/91

== Description:== Install lead seats and saf ety wire on simulated and photo sockup relay glass covers on 1H13POS07,1H13P0808,1H13P0809, arwf 1H13P0810, 42 affected relays. SDP 91 1056 Initiating Docunent: SCR 0 90 1035 Plant Date: 02/19/91 Scheduled Cormletion: 04/08/92

== Description:== Revise setpoint f or OP45N0255B to 565' 71/8" (565.59') Elev. at tire of TSR; $3 setpoint list gave this value to te 565.34' Elev. !q 1 J SDP 91 1061 initiating Docunent: N/A Plant Date: N/A Scheduled Ccanptetton: 06/28/92

== Description:== Instalt oiscellaneous furniture and equipment in the $1 mutator Room 123 after installation of the new Simulator. Coordinate with Control Roers enhancement activities. SDP 91-1063 Initiating Docunent: SCR 1 89 0010 Plcnt Date: 03/07/91 Scheduled Cosmtetton: 09/15/91

== Description:== Revise (Add) Setpoint for 1C85N0702, emergency high level on EHC fluid reservoir. Switch is an input to annunciator P680-7A B1

steam Bypass HPU trouble".

SDP 91 1064 Initiating Documentt SCR 1-90-1236 Plant Date: 03/11/91 Scheduled Cornpletion: 09/15/91

== Description:== Change setpoint of 1R4200101S to 5 seconds +/ 1 sec. D1B undervoltage time delay relay 27 DCx. SDP 91 1065 Initiating Document: SCR 1 90 1234 Plant Date: 03/15/91 Scheduled Completion: 09/15/91

== Description:== Change setpoint of 1R42008015 to $ seconds +/- 1 sec. ED1B undervoltage time delay relay 27 DCX. SDP 91 1066 Initiating Docunent: SCR 1 88 1563 Plant Date: 04/10/91 Scheduled Cormletion: 06/28/92

== Description:== Pickup and dropout vottages updated for 1R22c0605 SDP 91-1067 Initiating Document: SCR 0 91 1020 Plant Date: 04/04/91 Scheduled Completion: 06/28/92 m (

== Description:== Revise alert setpoint of 0021K0302 to 15 MR/HR (Mi S.P. is 20 MR/He) setpoints U for OD21K0332 on TSR 0011702: SCR 0 91-1021 are also included.

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 [T Exceptions to ANS 3.5 - Schedule for correcting TAD B - Part 3 i 'yl Discrepancies PACE 15 of 15 Schedule for Correcting Discrepancies Tracked by SDP Number section B - To Complete Before June 28, 1992 SDP 91 1068 Initiating Document SCR 1 90 1237 Ptant Date: 05/10/91 Schedated Conpletion: 06/73/92

== Description:== setpoint f or 1P4200601s chaved 'o 5 sec. +/ 1 sec. Bus EDIA undervottage reley SDP 91 1069 Initiating Document DCP 90 0127 Plant Dates N/A Scheduled Conpletter.: 06/28/92

== Description:== Permanent deactivation of ASB27/59A relays froan 4.16k'/ M11 and H12 normal and alternate sources. SDP 91 1070 Initiating Docunent DCP 90 0011/000/00 Plant Date: 12/24/90 Scheduled Cceptetton: 06/18/92 Descriptim Revise off Cas Loop Seat Levet instrsnentation by replacing levet switches 1N64W0016,1W64N0035, and 1N64WD040 with MPLs 1W64NOT72 through 1N64N0789. O n(>

r-CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests TAB C i Table of Contents i TAB C SIMULATOR TESTS Part 1. Certification Test to ANSI /ANS 3.5 Section Cross Reference Part 2. Computer Real Time Test Abstract - T.2.7.1 Part 3. Steady State and Normal Operations Test Abstracts A. Core Performance Tests - T.4.2 B. Continuous Plant Operation Tests - T.4.3 C. Steady State Performance Tests - T.4.4.1 Part 4. Transient Performance Test Abstracts - T.4.4.2 Part 5. Malfunction Test Abstracts A. System Level Failures - T.4.5.1 B. Malfunction Scenarios - T.4.5.3 0

i ) CERTIFICATION OF PERKY SIMULATION FACILITY DOCKET NO. 50-440 O ' Simulator Tests - Certification Test to IRSI/ANS 3.5 TAB C - Part 1 Section Cross Reference PAGE 1 of 13 ANSI /ANS 3.5 SECTION: 3.1.1 Normal Plant Evolutions Test T.4.3, CONTINUOUS PLANT OPERATION ANSI /ANS 3.5 SECTION: 3.1.1 (01) Normal Plant Evolutions; Plant Startup - Cold to Hot Standby Test T.4.3.1, COLD SNUTDOWN TO REACTOR CRITICAL ANSI /ANS 3.5 SECTION: 3.1.1 (02) Normal Plant Evolutions; Nuclear Startup from Not Standby to Rated Power Test T.4.3.2, REACTOR CRITICAL TO TURBINE SYNCHRONIZED Test T.4.3.3, POWER INCREASE TO 100% POWER ANSI /ANS 3.5 SECTION: 3.1.1 (03) Normal Plant Evolutions; Turbine Oenerator Startup and Synchronization Test T.4.3.2, REACTOR CRITICAL TO TURBINE SYNCHRONIZED ANSI /ANS 3.5 SECTION: 3.1.1 (04) Normal Plant Evolutions; Reactor Trip Test T.4.4.2.01, MANUAL SCRAM ANSI /ANS 3.5 SECTION: 3.1.1 (05) Normal Plant Evolutions; Operations at Hot Standby Test T.4.3.2, REACTOR CRITICAL TO TURBINE SYNCHRONIZED ANSI /ANS 3.5 SECTION: 3.1.1 (06) Normal Plant Evolutions; Load Changes Test T.4.3.3, POWER INCREASE TO 100% POWER ANSI /ANS 3.5 SECTION: 3.1.1 (07) Normal Plant Evolutions; Operations with Less than Full Coolant Flow Test T. 4. 5.1. 27.TH21, POWER /F' " INSTABILITIES (IEB 88-07 SUPPLEMENT 1) Test T.4.3.3, POWER INCREASE 'c ' 100% POWER ANSI /ANS 3.5 SECTION: 3.1.1 (08) Normal Plant Evolutions; Shutdown and i Cooldown to cold Shutdown Conditions l-l- Test T.4.3.4, POWER DECREASE TO TURBINE / GENERATOR UNLOADED l Test T.4.3.5, PLANT COOLDOWN TO COLD SHUTDOWN V

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 Section Cross Reference PAGE 2 of 13 ANSI /ANS 3.5 SECTION: 3.1.1 (09) Normal Plant Evolutions; Core Performance Test T.4.2.1, REACTOR CORE XENON TRANSIENT TEST Test T.4.2.2, CORE FLUX DISTRIBUTION TEST Test T.4.2.3, CORE THERMAL POWER VS. RECIRC FLOW TEST Test T.4.2.4, CORE FLUX RESPONSE TO ROD MOVEMEh.' Test T.4.2.5, CORE THERMAL PERFORMANCE TEST Test T.4.2.6, CORE SUBCRITICAL MULTIPLICATION TEST Test T.4.2.7, REACTOR CORE LIFE TEST Test T.4.2.8, SHUTDOWN MARGIN DEMONSTRATION Test T.4.4.1, STEADY STATE PERFORMANCE ANSI /ANS 3.5 SECTION: 3.1.1 (10) Normal Plant Evolutions; Operator Conducted Surveillances on SR Equipment Test T.4.3.3, POWER INCREASE TO 100% POWER ANSI /ANS 3.5 SECTIOM: 3.1.2 Plant Malfunctions Test T.4.5.1, SYSTEM LEVEL FAILURES Test T.4.5.3, MALFUNCTION SCENARIOS ANSI /ANS 3.5 SECTION: 3.1.2 (01)(b) Plant Malfunctions; LOCA inside ( containment Test T.4.5.1.27.TH01, RECIRC LOOP RUPTURE (DBA LOCA) ANSI /ANS 3.5 SECTION: 3.1.2 (01)(b) Plant Malfunctions; LOCA outside containment Test T.4.5.1.04.CUO3, REACTOR WATER CLEANUP SYSTEM PIPE BREAK OUTSIDE CONTAINMENT (STEAM TUNNEL) ANSI /ANS 3.5 SECTION: 3.1.2 (01)(c) Plant Malfunctions; LOCA Large Break Test T.4.4.2.08, MAXIMUM SIZE LOCA WITH LOSS OF OFFSITE POWER Test T.4.5.1.27.TH01, RECIRC LOOP RUPTURE (DBA LOCA) ANSI /ANS 3.5 SECTION: 3.1.2 (01)(c) Plant Malfunctions; LOCA Small Break Test T.4.5.3.25, PEI MALFUNCTION SCENARIO #4 ANSI /ANS 3.5 SECTION: 3.1.2 (01)(d) Plant Malfunctions; Failure of Safety and Relief valves Test T.4.5.3.05, INADVERTENT SAFETY / RELIEF VALVE OPENING Test T.4.5.3.22, PEI MALFUNCTION SCENARIO #1 a t

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Certification Test to ANSI /ANS 3.5 TAD C - Part 1 >( Section Cross Reference PAGE 3 of 13 ANSI /ANS 3.5 SECTION: 3.1.2 (02) Plant Na1 functions; Loss of Instrument Air Test T.4.5.1.12 IA01, AIR RECEIVER LEAK ANSI /ANS 3.5 SECTION: 3.1.2 (03)(a) Plant Malfunctions; Loss of Off-Site Power Test T.4.5.3.21, LOSS OF OFF-SITE PO:iER ANSI /ANS 3.5 SECTION: 3.1.2 (03)(b) Plant Ms,/gnetions; Loss of Emergency Power Test T.4.5.3.26, PEI MALFUNCTION SCENARIO #5 (PAhT 1) ANSI /ANS 3.5 SECTION: 3.1.2 (03)(c) Plant Halfunctions; Loss of Diesel Generators Test T.4.5.3.26, PEI MALFUNCTION SCENARIO #5 (PART 1) Test T.4.5.3.29, EVALUATION MALFUNCTION SCENARIO #3 ANSI /ANS 3.5 SECTION: 3.1.2 (03)(d) Plant Malfunctions; Loss of Distribution Bus Test T.4.5.1.06.ED06, LOSS OF 480V BUS Test T.4.5.1.06.ED17, LOSS OF 125V DC DISTRIBUTION PANEL ANSI /ANS 3.5 SECTION: 3.1.2 (03)(e) Plant Malfunctions; Loss of AC Instrument Bus Test T.4.5.3.22, PEI MALFUNCTION SCENARIO #1 ANSI /ANS 3.5 SECTION: 3.1.2 (03)(e) Plant Malfunctions; Loss of DC Instrument Bus Test T.4.5.1.06.ED09, LOSS OF 125V DC BUS ANSI /ANS 3.5 SECTION: 3.1.2 (04)(a) Plant Malfunctions; Loss of Coolant Plow, dual pump Test T.4.4.2.04, SIMULTANEOUS TRIP OF ALL RECIRC PUMPS ANSI /ANS 3.5 SECTION: 3.1.2 (04)(a) Plant Malfunctions; Loss of Coolant Flow, single pump l Test T.4.5.3.16, RECIRCULATION PUMP SHAFT SHEAR t O 1

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 1O Simulator Tests - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 Section Cross Reference PAGE 4 of 13 i ANSI /ANS 3.5 SECTION: 3.1.2 (05)(a) Plant llalfunctions; Loss of Condenser Vacuum Test T.4.5.1.15.MC01, CONDENSER AIR IhLEAKAGE ' Test T.4.5.3.22, PEI MALFUNCTION SCENARIO #1 Test T.4.5.3.27, EVALUATION MALFUNCTION SCENARIO #1 Test T.4.5.3.3), EVALUATION MALFUNCTION SCENARIO #5 Test T.4.5.3.36, EVALUATION MALFUNCTION SCENARIO #10 ANSI /ANS 3.5.SECTION: 3.1.2 (06) Plant Malfunctions; Loss of Service Water Test T.4.5.1.25.SWO2, SERVICE WATER SYSTEM PROCESS PIPING LEAKAGE ANSI /ANS 3.S SECTION: 3.1.2 (07) Plant Malfunctions; Loss of Shutdown Cooling Test T.4.5.1.22.RH02, RESIDUAL HEAT REMOVAL SYSTEM PIPE BREAK Test T.4.5.3.12, FAILURE OF RESIDUAL HEAT REMOVAL SHUTDOWN COOLING Test T.4.5.3.40, EVALUATION MALFUNCTION SCENARIO #14 ANSI /ANS 3.5 SECTION: 3.1.2 (08) Plant Malfunctions; Loss of Component Cooling Test T.4.5.1.25.SWO1, NUCLEAR CLOSED COOLING SYSTEM PROCESS PIPING O: LEAKAGE ANSI /ANS 3.5 SECTION: 3.1.2 (09) Plant Malfunctions; Normal Feedwater System Failure Test T.4.5.3.01, LOSS OF FEEDWATER HEATING (REACTOR'RECIRC FLOW CONTROL IN AUTO) Test T.4.5.3.02, LOSS OF FEEDWATER HEATING (REACTOR RECIRC FLOW CONTROL-IN MANUAL) Test T.4.5.3.03, FEEDWATER CC'ITROLLER /AILURE-MAXIMUM DEMAND Test T.4.5.3.28,' EVALUATION MALFUNCTION SCENARIO #2 ANSI /ANS 3.5 SECTION: 3.1.2 (10) Plant Malfunctions; Loss of All Feedwater Test T.4.4.2.02, SIMULTANEOUS TRIP OF ALL FEEDWATER PUMPS . Test T.4.5.3.22, PEI MALFUNCTION SCENARIO #1. Test T.4.5.3.23, PEI MALFUNCTION SCENARIO #2 Test T.4.5.3.28, EVALUATION MALFUNCTION SCENARIO #2 ANSI /ANS 3.5 SECTION: 3.1.2 (11) Plant Malfunctions; Loss of Protective System Channel Test T.4.5.3.23, PEI MALFUNCTION SCENARIO #2 Test T.4.5.3.26, PEI MALFUNCTION SCENARIO #5 (PART 1) ANSI /ANS 3.5 SECTION: 3.1.2 (12)(a) Plant Malfunctions; Stuck Control Rod O-Test T.4.5.3.24, PEI HALFUNCTION SCENARIO #3 Test T.4.5.3.27, EVALUATION MALFUNCTION SCENARIO #1 4 ,-er--,.,- .m-. ,.,,v,..-.,-cr--m-w,,,-.. ,,,,,,_<..ws--.., g- .y----_.- _,ww--

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 /'^T Simulator Teste - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 (,,/ Section Cross Referenco PAGE 5 ef 13 ANSI /ANS 3.5 SECTION: 3.1.2 (12)(b) Plant Malfunctions; Uncoupled Control Rod Test T.4.5.3.27, EVALUATION MALFUNCTION SCENARIO #1 ANSI /ANS 3.5 SECTION: 3.1.2 (12)(c) Plant Malfunctions; Drifting control Rod Test T.4.5.1.21.RD03, CONTROL ROD DRIFT - IN Test T.4.5.1.21.RD04, CONTROL ROD DRIFT - OUT ANSI /ANS 3.5 SECTION: 3.1.2 (12)(d) Plant Malfunctions; control Rod Drop Test T.4.5.3.18, CONTROL ROD DROP ACCIDENT ANSI /ANS 3.5 SECTION: 3.1.2 (13) Plant Malfunctions; Inability to Move control Rods Test T.4.5.3.22, PEI MALFUNCTION SCENARIO #1 Test T 4.5.3.23, PEI MALFUNCTION SCENARIO #2 Test T.4.5.3.34, EVALUATION MALFUNCTION SCENARIO #8 Test T.4.5.3.37, EVALUATION MALFUNCTION SCENARIO #11 ANSI /ANS 3.!' SECTION: 3.1.2 (14) Plant Malfunctions; Fuel Cladding Failure Test T.4.5.3.26, PEI MALFUNCTION SCENARIO #5 (PART 1) ANSI /ANS 3.5 SECTION: 3.1.2 (15) Plant Halfunctions; Turbine Trip Test T.4. 5.1.28.TUOl, MAIN SHAFT OIL PUMP DEGRADATION Test T.4.5.3.10, TURBINE TRIP ANSI /ANS 3.5 SECTION: 3.1.2 (16) Plant Malfunctions; Oenerator Trip Test T.4.5.1.07.EG01, MAIN GENERATOR LOCKOUT RELAY TRIP Test T.4.5.3.08, GENERATOR LOAD REJECT WITH BYPASS VALVES Test T.4.5.3.09, GENERATOR LOAD REJECT WITHOUT BYPASS VALVES Test T.4.5.3.25, PEI MALFUNCTION SCENARIO #4 ANSI /ANS 3.5 SECTION: 3.1.2 (17) Plant Halfunctions; Failure in Reactivity Control System Test T.4.5.1.23.RP02, INADVERTENT INITIATION OF ALTERNATE ROD INSERTION Test T.4.5.3.13, RECIRC FLOW CONTROL FAILURE-DECREASING (BOTH FCV'S) Test T.4.5.3.17, RECIRC FLOW CONTROL FAILURE-INCREASING (BOTH FCV'S) Test T.4.5.3.23, PEI MALFUNCTION SCENARIO #2 ANSI /ANS 3.5 SECTION: 3.1.2 (19) Plant Halfunctions; Reactor Trip Test T.4.4.2.01, MANUAL SCRAM

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 (N Simulator Tests - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 's,) Section Cross Reference PAGE 6 of 13 ANSI /ANS 3.5 SECTION: 3.1.2 (20)(a) Plant Na1 functions; Nain Steam Line Break, inside containment Test T.4.5.3.38, EVALUATION MALFUNCTION SCENARIO #12 ANSI /ANS 3.5 SECTION: 3.1.2 (20)(b) Plant Na1 functions; Nain Steam Line Break, outside containment l Test T.4.5.1.27.TH27, MAIN STEAM LINE RUPTURE IN STEAM TUNNEL ANSI /ANS 3.5 SECTIONS.3'.1.2-(20)(c) Plant Na1 functions; Feedwater Line Break, inside containment Test T.4.5.1.09.FWO2, FEEDWATER SYSTEM PIPE BREAK INSIDE DRYWELL ANSI /ANS 3.5 SECTION: 3.1.2 (20)(c) Plant Na1 functions; Feedwater Line Break,. outside containment Test T.4.5.1.09.FWO3, FEEDWATER SYSTEM PIPE BREAK OUTSIDE CONTAINMENT ANSI /ANS 3.5 SECTION '3.1.2 (21) Plant Na1 functions; Nuclear Instrumentation Failures Test T.4.5.1.17.NM01, SRM DETECTOR (PRE-AMP) FAILURE i Test T.4.5.1.17.NM02, IRM DETECTOR (PRE-AMP) FAILURE l Test T.4.5.1.17.NM03, LPRM DETECTOR FAILURE Test T.4.5.1.17.NM10, NEUTRON MONITORING DETECTOR DRIVE STUCK l Test T.4.5.3.22, PEI MALFUNCTION SCEhARIO #1 Test T.4.5.3.28, EVALUATION MALFUNCTION SCENARIO #2 ' test T.4.5.3.39, EVALUATION MALFUNCTION SCENARIO #6 ANSI /ANS 3.5 SECTION: 3.1.2 (22)(a) Plant Na1 functions; Process Instrument System Failures Test T.4.5.3.24, PEI MALFUNCTION SCENARIO #3 Test T.4.5.3.27, EVALUATION MALFUNCTION SCENARIO #1 Test T.4.5.3.28, EVALUATION MALFUNCTION SCENARIO #2 Test T.4.5.3.29, EVALUATION MALFUNCTION SCENARIO #3 Test T.4.5.3.30, EVALUATION MALFUNCTION SCENARIO #4 j Test T.4.5.3.33, EVALUATION MALFUNCTION SCENARIO #7 Test T.4.5.3.34, EVALUATION MALFUNCTION SCENARIO #8 Test T.4.5.3.35, EVI.LUATION MALFUNCTION SCENARIO #9 Test T.4.5.3.37, EVALUATION MALFUNCTION SCENARIO #11 Test T.4.5.3.38, EVALUATION MALFUNCTION SCENARIO #12 Test T.4.5.3.39, EVALUATION MALFUNCTION SCENARIO #13 Test T.4.5.3.40, EVALUATION RALFUNCTION SCENARIO #14 Test T.4.5.3.41, EVALUATION MALFUNCTION SCENARIO #15 ) Test T.4.5.3.42, PEI MALFUNCTION SCENARIO #5 (PART 2) l 0

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Teste - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 N- / Section Cross Reference PAGE 7 of 13 Al'SI/ANS 3.5 SECTION: 3.1.2 (22)(b) Plant Malfunctions; Process Alarm System Failure Test T.4.5.1.02.AN01, ANNUNCIATOR INPUT OPTICAL ISOLATOR FAILURE ANSI /ANS 3.5 SECTION: 3.1.2 (22)(c) Plant Malfunctions; Process control System Failures Test T.4.5.3.22, PEI MALFUNCTION SCENARIO #1 Test T.4.5.3.26, P3I MALFUNCTION SCENARIO #5 (PART 1) Test T.4.5.3.30, ENALUATION MALFUNCTION SCENARIO #4 Test T.4.5.3.39, EVJ.LUATION MALFUNCTION SCENARIO #13 ANSI /ANS 3.5 SECTION: 3.1.2 (23) Plant Malfunctions; Passive and Active Component Failure in Plant Systems Test T.4.5.3.08, GENERATOR LOAD REJECT WITH BYPASS VALVES l Test T.4.5.3.22, PEI MALFUNCTION SCENARIO #1 Test T.4.5.3.23, PEI MALFUNCTION SCENARIO #2 Test T.4.5.3.24, PEI MALFUNCTION SCENARIO #3 Test T.4.5.3.25, PEI MALFUNCTION SCENARIO #4 Test T.4.5.3.26, PEI MALFUNCTION SCENARIO #5 (PART 1) Test T.4.5.3.27, EVALUATION MALFUNCTION SCENARIO #1 Oi Test T.4.5.3.28, EVALUATION MALFUNCTION SCENARIO #2 Test T.4.5.3.29, EVALUATION MALFUNCTION SCENARIO #3 Test T.4.5.3.30, EVALUATION MALFUNCTION SCENARIO #4 Test T.4.5.3.31, EVALUATION MALFUNCTION SCENARIO #5 T9st T.4.5.3.32, EVALUATION MALFUNCTION SCENARIO #6 Test T.4.5.3.33, EVALUATION MALFUNCTION SCENARIO #7 Test T.4.5.3.34, EVALUATION MALFUNCTION SCENARIO #8 Test T.4.5.3.35, EVALUATION MALFUNCTION SCENARIO #9 Tcat T.4.5.3.37, EVALUATION MALFUNCTION SCENARIO #11 Test T.4.5.3.38, EVALUATION MALFUNCTION SCENARIO #12 Test T.4.5.3.39, EVALUATION MALFUNCTION SCENARIO #13 Test T.4.5.3.40, EVALUATION MALFUNCTION SCENARIO #14 Test T.4.5.3.41, EVALUATION MALFUNCTION SCENARIO #15 Test T.4.5.3.42, PEI MALFUNCTION SCENARIO #5 (PART 2) ANSI /ANS 3.5 SECTION: 3.1.2 (24) Plant Malfunctions; Failure of the Automatic Reactor Trip System Test T.4.5.3.20, ANTICIPATED TRANSIENT WITHOUT SCRAM (ATWS) Test T.4.5.3.23, PEI MALFUNCTION SCENARIO #2 Test T.4.5.3.24, PEI MALFUNCTION SCENARIO #3 Test T.4.5.3.26, PEI MALFUNCTION SCENARIO #5 (PART 1) Test T.4.5.3.36, EVALUATION MALFUNCTION SCENARIO #10 Test T.4.5.3.37, EVALUATION MALFUNCTION SCENARIO #11 Test T.4.5.3.39, EVALUATION MALFUNCTION SCENARIO #13 Test T.4.5.3.41, EVALUATION MALFUNCTION SCENARIO #15 Test T.4.5.3.42, PEI MALFUNCTION SCENARIO #5 (PART 2) v

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 / Simulator Teots - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 Section Cross Reference PAGE B of 13 ANSI / ANS 3.5 SECTION: 3.1.2 (25) Plant Malfunctions; Reactor Pressure Control System Failure (BWR) Test T.4.5.3.04, PRESSURE REGULATOP. FAILURE-OPEN Test T.4.5.3.07, PRESSURE REGULATOR FAILURE-CLOSED ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.1.1 Test T.4.5.3.01, LOSS OF FEEDWATER HEATING (REACTOR RECIRC FLOW CONTROL IN AUTO) Test T.4.5.3.02, LOSS OF FEEDWATER HEATING (REACTOR RECIRC FLOW CONTROL IN MANUAL) ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.1.2 Teot T.4.5.3.03, FEEDWATER CONTROLLER FAILURE-MAXIMUM DEKAND ANSI /ANS 3.5 SECTION: 3.1.2 Plant Halfunctions, USAR Accident 15.1.3 Test T.4.5.3.04, PRESSURE REGULATOR FAILURE-OPEN ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.1.4 ) Test T.4.5.3.05, INADVERTENT SAFETY / RELIEF VALVE OPENING ANSI / ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.1.6 Test T.4.5.3.06, INADVERTENT RESc UAL HEAT REMOVAL SHUTDOWN COOLING OPERATION ANSI /ANS 3.5 SECTION: 3.1.2 Plant Halfunctions, USAR Accident 15.2.1 Test T.4.5.3.07, PRESSURE REGULATOR FAILURE-CLOSED ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.2.10 Test T.4.5.1.12.IA02, 1HSTRUMENT AIR LINE LEAK ANSI /ANS 3.5 SECTION: 3.1.2 Flant Malfunctisns, USAR Accident 15.2.2 Tr ot T.4. 5. 3.08, GENERATOR LOAD REJ2CT WITH BYPASS VALVES Test T.4.5.3.09, GENERATOR LOAD REJECT WITHOUT 'YPASS VALVES ANSI /ANS 3.5 SECi' ION: 3.1.2 Plant Malfunctions, USAR Accident 15.9.3 Test T.4.5.3.10, TURBINE TRIP O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 Section Cross Reference PAGE 9 of 13 ANSI /ANS 3.5 SFCTION: 3.1.2 Plant Malfunctions, USAR Accident 15.2.4 Test T.4.4.2.03, SIMULTANEOUS CLOSURE CF ALL MAIN STEAM ISOLATION VALVES Test T.4.4.2.10, SIMULTANEOUS CLOSURE OF MAIN STEAM ISOLATION VALVES W/ SINGLE STUCK OPEN SAFETY / RELIEF VALVE ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.2.5 Test T.4.5.1.15.MC01, CONDENSER AIR INLEAKAGE ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.2.6 Test T.4.5.3.08, OENERATOR LOAD REJECT WITH BYPASS VALVES Test T.4.5.3.11, LOSS OF AC POWER (LOSS OF AUX TRANSFORMER) Test T.4.5.3.21, LOSS OF OFF-SITE POWER ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.2.7 Test T.4.4.2.02, SIMI/LTANEOUS TRIP OF ALL FEEDWATER PUMPS ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.2.8 ('"N Test T.4.4.2.09, MAXIMUM SIZE UNISOLABLE MAIN STEAM LINE RUPTURE i ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.2.9 Test T.4.5.1.22.RH02, RESIDUAL HEAT REMOVAL SYSTEM PIPE BREAK Test T.4.5.3.12, FAILURE OF RESIDUAL HEAT REMOVAL SHUTDOWN COOLING ANSI /ANS 3.5 SECTION: 3.1.2 Plant Halfunctions, USAR Accident 15.3.1 Test T.4.4.2.04, SIMULTANEOUS TRIP OF ALL RECIRC PUMPS Test T.4.4.2.05, SINGLE RECIRC PUMP TRIP ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.3.2 Test T.4.5.3.13, RECIRC FLOW CONTROL FAILURE-DECREASING (BOTH FCV'S) ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.3.3 Test T.4.5.3.14, RECIRCULATION PUMP SEIZURE ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.3.4 Test T.4.5.3.16, RECIRCULATION PUMP SHAFT SHEAR ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.4.4 i Test T.4.5.3.15, ABNORMAL STARTUP OF IDLE RECIRCULATION PUMP

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Certification Teet to ANSI /ANS 3.5 TAB C - Part 1 g \\_/ Section Cross Roferenco PAGE 10 of 13 ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.4.5 Test T.4.5.3.17, RECIRC FLOW CONTROL FAILURE-INCREASING (DOTH FCV'S) ANSI /ANS 3.5 SECTION: 3.1.2 Plant '.4alfunctions, USAR Accident 15.4.9 Test T.4.5.3.18, CONTROL ROD DROP ACCIDENT ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.5.1 Test T.4.5.3.19, INADVERTENT HPCS STARTUP ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.5.3 Test T.4.5.3.03, FEEDWATER CONTROLLER FAILURE-MAXIMUM DEMAND ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.6.1 Test T.4.5.3.05, INADVERTENT SAFETY / RELIEF VALVE OPENING ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.6.2 Test T.4.5.1.27.TH19, RPV LEVEL INST REFERENCE LEO BREAK Test T.4.5.1.27.TH20, RPV LEVEL INST VARIABLE LEG BREAK ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.6.4 Tout T.4.5.1.27.TH27, MAIN STEAM LINE RUPTURE IN STEAM TUNNEL ANSI /ANS 3.5 SECTION: 3.1.2 Plant Melfunctions, USAR Accident 15.6.5 Test T.4.4.2.08, MAXIMUM SIZE LOCA W/ LOSS OFFSITE POWER ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.6.6 Test T.4.5.1.09.FWO2, FEEDWATER SYSTEM PIPE BREAK INSIDE DRYWELL Test T.4.5.1.09.FWO3, FEEDWATER SYSTEM PIPE RREAK OUTSIDE CONTAINMENT ANSI /ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.7.1 Test T.4.5.1.16.MS11, STEAM SEAL HEADER PRESSURE REGULATOR FAILURE Test T.4.5.1.18.OG03, OFF GAS SYSTEM LEAK UPSTREAM ADSORSERS Test T.4.5.1.18.OG04, OFF GAS SYSTEM LEAK DOWNSTREAM ADSORBERS ENSI/ANS 3.5 SECTION: 3.1.2 Plant Malfunctions, USAR Accident 15.8 Test T.4.5.3.20, ANTICIPATED TPANSIENT WITHOUT SCRAM (ATWS) O

_.__._.___._.______._..__mm - - ~ _ _ _ _ _ _. - _ _ _. _. -. I CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O. Simulator Teste - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 Section Cross Reference PAGE 11 of 13 1 r ANSI /ANS 3.5 SECIION: 3.1.2 Plant Halfunctions; Other Malfunctions Required to l Support Operator Training Test T 4.5.1.27.TH21, POWER / FLOW INSTABILITIES (IEB 88-07 SUPPLEMENT 1) . Test T.4.5.3.22, PEI MALFUNCTION SCENARIO #1 Test T.4.5.3.23, PEI MALFUNCTION SCENARIO #2 Test T.4.5.3.24, PEI MALFUNCTION SCENARIO #3 Test T.4.5.3.25, PEI MALFUNCTION SCENARIO #4 Test T.4.5.3.26, PEI MALFUNCTION SCENARIO #5 (PART 1) Test T.4.5.3.27, EVALUATION MALFUNCTION SCENARIO #1 Test T.4.5.3.28, EVALUATION MALFUNCTION SCENARIO #2 Test T.4.5.3.29, EVALUATION MALFUNCTION SCENARIO #3 Test T.4.5.3.30, EVALUATION MALFUNCTION SCENARIO #4 Test T.4.5.3.31, EVALUATION MALFUNCTION SCENARIO #5 Test T.4.5.3.32, EVALUATION MALFUNCTION SCENARIO #6 Test T.4.5.3.33, EVALUATION MALFUNCTION SCENARIO #7 Test T.4.5.3.34, EVALUATION MALFUNCTION SCENARIO #8 Test T.4.5.3.35, EVALUATION MALFUNCTION SCENARIO #9 Test T.4.5.3.36, EVALUATION MALFUNCTION SCENARIO #10 ~ Test T.4.5.3.37, EVALUATION MALFUNCTION SCENARIO #11 Test T.4.5.3.38, EVALUATION MALFUNCTION SCENARIO #12 Test T.4,5.3.39, EVALUATION MALFUNCTION SCENARIO #13 Test T.4.5.3.40, EVALUATION MALFUNCTION SCENARIO #14 Test T.4.5.3.41, EVALUATION MALFUNCTION SCENARIO #15 Test T.4.5.3.42, PEI MALFUNCTION SCENARIO #5 (PART 2) O ANSI /ANS 3.5 SECTION: Appendix A 3.1 Computer Real Time Test ' Test T.2.7.1, SPARE TIME VERIFICATION ANSI /ANS 3.5 SECTION: Appendix A 3.2 Normal Operations Test T.4.3, CONTINUOUS PLANT OPERATION ANSI /ANS 3.5 SECTION: Appendix A 3.2 Steady.. State Performance Test T.4.4.1, STEADY STATE PERFORMANCE Test T.4.4'.1.1, 25% POWER HEAT BALANCE Test T.4.4.1.2, 50% POWER HEAT BALANCE l Test T.4.4.1.3, 75% POWER HEAT BALANCE Test T.4.4.1.4, 100% POWER HEAT BALANCE Test T.4.4.1.5, 100% POWER STABILITY TEST - ANSI /ANS 3.5 SECTION: Appendix A 3.3 Transient Tests Test T.4.4.2, TRANSIENT PERFORMANCE Test T.4.4.2.01, MANUAL SCRAM Test T.4.4.2.02, SIMULTANEOUS TRIP OF ALL FEEDWATER PUMPS Test T.4.4.2.03, SIMULTANEOUS CLOSURE OF ALL MAIN STEAM ISOLATION VALVES Test T.4.4.2.04, SIMULTANEOUS TRIP OF ALL RECIRC PUMPS Test T.4.4.2.05, SINGLE RECIRC PUMP TRIP O Test T 4.4.2.06, MAIN TURBINE TRIP W/O REACTOR SCRAM Test T.4.4.2.07, MAXIMUM RATE PCWER RAMP (100% - 75% - 100%) Test T.4.4.2.08, MAXIMUM SIZE LOCA W/ LOSS OF OFFSITE POWER

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 [ Simulator Tests - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 ( Section Cross Reference PAGE 12 vf 13 Test T.4.4.2.09, MAXIMUM SIZE UNISOLABLE MAIN 4 TEAM LINE RUPTURE Test T.4.4.2.10, SIMULTANEOUS CLOSURE OF MAIN STEAM ISOLATION VALVES W/ SINGLE STUCK OPEN SAFETY / RELIEF VALVE ANSI /ANS 3.5 SECTION: Appendix A 3.4 Halfunction Tests Test T.4.5, MALFUNCTION AND COMPONENT FAILURE TESTS ANSI /ANS 3.5 SECTION: Appendix B Simulator Operability Tests Test T.4.4, BWR SIMULATOR OPERABILITY TESTS ANSI /ANS 3.5 SECTION: Appendix B 1.1 Steady State Performance Test T.4.4.1, STEADY STATE PERFORMANCE Test T.4.4.1.1, 25% POWER HEAT BALANCE Test T.4.4.1.2, 50% POWER HEAT BALANCE Test T.4.4.1.3, 75% POWER HEAT BALANCE Test T.4.4.1.4, 100% POWER HEAT BALANCE Test T.4.4.1.5, 100% POWER STABII,ITY TEST (N ANSI /ANS 3.5 SECTION: Appendix B 1.2 Transient Performance Test T.4.4.2, TRANSIENT PERFORMANCE ANSI /ANS 3.5 SECTION: Appendix B 1.2 (01) Transient Performance Test T.4.4.2.01, MANUAL SCRAM ANSI /ANS 3.5 SECTION: Appendix B 1.2 (02) Transient Performance Test T.4.4.2.02, SIMULTANEOUS TRIP OF ALL FEEDWATER PUMPS ANSI /ANS 3.5 SECTION: Appendix B 1.2 (03) Transient Performance Test T.4.4.2.03, SIMULTANEOUS CLOSURE OF ALL MAIN STEAM ISOLATION VALVES ANSI /ANS 3.5 SECTION: Appendix B 1.2 (04) Transient Performance Test T.4.4.2.04, SIMULTANEOUS TRIP OF ALL RECIRC PUMPS ANSI /ANS 3.5 SECTION: Appendix B 1.2 (05) Transient Performance Test T.4.4.2.05, SINGLE RECIRC PUMP TRIP ANSI /ANS 3.5 SECTION: Appendix B 1.2 (06) Transient Performance \\- Test T.4.4.2.06, MAIN TURBINE TRIP W/O REACTOR SCRAM

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Certification Test to ANSI /ANS 3.5 TAB C - Part 1 I Section Crose Reference PAGE 13 of 13 ANSI /ANS 3.5 SECTION: Appeadix B 1.2 (07) Transient Performance Test T.4.4.2.07, MAXIMUM RATE POWER RAMP (100% - 754 - 2004) ANSI /ANS 3.5 SECTION: Appendix B 1.2 (08) Transient Performance Test T.4.4.2.08, MAXIMUM SIZE LOCA W/ LOSS OF OFFSITE POWER ANSI /ANS 3.5 SECTION: Appendix B 1.2 (09) Transient Performance Test T.4.4.2.09, MAXIMUM SIZE UNISOLABLE MAIN STEAM LINE RUPTURE ANSI /ANS 3.5 SECTION: Appendix B 1.2 (10) Transient Perf ormance Tent T.4.4.2.10, SIMULTANEOUS CLOSURE OF MAIN STEAM ISOLATION VALVES W/ SINGLE STUCK OPEN SAFETY / RELIEF VALVE O O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulr. tor Tests - Computer Real Time Test Abstract TAB C - Part 2 V PAGE 1 of 1 ATP TEST SECTION T.2.7.1 COMPUTER REAL TIME TEST ABSTRACT Test T.2.7.1, SPARE TIME VERIFICATION Revision Number Later

ANSI /ANS 3.5 Section: Appendix A 3.1 Computer Real Time Test Date Tested:

Not-RUN

Run Times Later hours
Test

Description:

This test measures the available spare time on each of the main simulation computer nodes while the simulator is running a scenario containing a large number of malfunctions and involves the interaction of as many BOP and Safety systems as possible. Two measurements are taken in each node (CPU): total spare time, and spare titae execution per frame. The criteria for each CPU are a minimum of 40% spare tirce total, and a minimum of 20% spare time per frame. The test has no effect on the simulation and runs as a background task. Baseline Data used For Reference OQ Reference Type Other Simuistor Specification Halfunctions Tested: None Discrepancies: Unknown

Evaluators Later
This test has not-been completed (N/C).

O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Steady State and Normal Operations TAD C - Part 3A Tests Abstracts PAGE 1 of 9 ATP TEST SECTION T.4.2 CORE PERFORMANCE TESTS Overview of testina nerformed in this section This section of testing verifies the nuclear characteristics of the simulated core model are in compliance with established laws of nuclear physics and also matches the parameters of the Perry Cycle 1 core. Separate tests are run to determine the relative reactivities associated with fission product poisons and control rods. Neutron Flux profiles, thermal performance and flux response to core flow (power to flow map) are also checked at power. The shutdown margin verification and sub-critical multiplication tests are performed on the shutdown Core. No. of Tests included in this SectLqu: Pacle: Open SDR's: T.4.2.1, REACTOR CORE XENON TRAkSIENT TEST 2 None T.4.2.2, CORE FLUX DISTRIBUTION TEST 3 1* \\ T.4.2.3, CORE THERMAL POWER VS. RECIRC FLOW TEST 4 None T.4.2.4, CORE FLUX RESPONSE 10 RCD MOVEMENT None T.4.2.5, CORE THERMAL PERFORMANCE TEST 6 1* T.4.2.6, CORE SUBCRITICAL MULTIPLICATION TEST 7 None T.4.2.7, REACTOR CORE Life TEST 8 mone T.4.2.8, SHUTOOWN MARGIN DEMONSTRATION 9 None ANSI /ANS 3.5

Reference:

3.1.1 (09) Normal Plant Evolutions; Coro Periormance This test has not been completed (N/C) or has been performed but has one or more unresolved major Simulator Discrepancy Eeports associated with it. Please see the individual test abstract for these discrepancies which constitute exceptions to ANSI /ANS 3.5 section 3.1.1. O

_ _ _ _ _ _ ~. _.. _.... -.. - ~ _.. CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Steady State and Normal Operations TAB C - Part 3A Tests Abstracts PAGE 2 of 9 Test T.4.2.1, REACTOR CORE XENON TRANSIENT TEST Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.1 (09). Normal Plant Evolutions; Core Performance Date Tested: 06/04/91 Run Time 8.00 hours Test

Description:

This test verifies the nuclear characteristics of the simulated core model with respect to the reactivity effects of the major fission product poison, Xenon. Baseline Data used For Reference Reference Type Plant Data Cycle 1 Benchmark Data GENERATED BY PERRY REACTOR ENGINEERING SECTION Malfunctions Tested: None Discrepancies: None Evaluators: B.

Panfil, B.

Stetson r O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Steady State and Normal Operations TAB C - Part 3A Tests Abstracts PAGE 3 of 9 Test T.4.2.2, CORE FLUX DISTRIBUTION TEST Revision Humber: 02 ANSI /ANS 3.5 Section: 3.1.1 (09) Normal Plant Evolutions; Core Performance Date Tested: 04/23/91 i Run Times 1.50 hours Test

Description:

This test verifies that the simulated core axial flux shape is correct for the BOC and EOC co.es. Baseline Data used For Reference Reference Type Plant Data Cycle 1 Benchmark Data GENERATED BY PERRY REACTOR ENGINEERING SECTION Malfunctions Tested: None Major Discrepancies: F-Ol69 CANNOT PERFORM (PASS) T.4.2.2 AND T.4.2.5 DUE TO MFLCPR GREATER THAN 1. SHOULD DE ABLE TO OPERATE O'l PCTLLP UP TO 110% WITHOUT LIMIT VIOLATIONS. Evaluators: P. Bordley, C. Persson, B. Panfil, R.

Kearney, G.

Minshall, J. Pierson O

f CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Steady State and Normal Operations TAB C - Part 3A Tests Abstracts PAGE 4 of 9 Test T.4.2.3, CORE THERMAL POWER VS. RECIRC FLOW TEST Revision Number: 02 ANSI /ANS 3.5 Section 3.1.1 (09) Normal Plant Evolutions; Core Performance Date Tested: 06/10/91 Run Times 3.00 hours Test

Description:

This test verifies that the power to flow characteristics of the core model are valid in both forced circulation and natural circulation, and verified using the Percy Unit One Power to Flow Map. Baseline Data used For Reference Referenco Type: Plant Dat Cycle 1 Benchmark Data GENERATED BY PERRY REACTOR ENGINEERING SECTION PDB-A006 POWER TO FLOW MAP PDB-A010 LOOP FLOW VERSUS FCV POSITION PDB-A012 TOTAL CORE FLOW VERSUS RECIRCULATION DRIVE FLOW Malfunctions Testedt None Major Discrepancies: None Evaluators: B.

Panfil, C.

Persson, H. DeBoer O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Teste - Steady State and Normal Operations TAB C - Part 3A Tests Abstracts PAGE 5 of 9 Test T.4.2.4, CORE FLUX RESPONSE TO ROD MOVEMENT Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.1 (09) Normal Plant Evolutions; Core Performance Date Tested: 06/05/91 Run Time: 3.00 hours Test

Description:

Tais test verifies that the movement of shallow, intermediate, and deep control rode, produce reactor power changes of the correct direction and magnitude. Baseline Data used For Reference Reference Type Plant Data Cycle 1 Ber:hmark Data GEiERATED BY PERRY REACTOR ENGINEERING SECTION Malfunctions Tasted: None Discrepancies: None Evaluators: B.

Panfil, D.

Stetson O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Steady State and Normal Operations TAB C - Part 3A Tests Abstracts PAGE 6 of 9 Test T.4.2.5, CORE THERMAL PERFORMANCE TEST Revision Number C2 ANSI /ANd 3.5 Section: 3.1.1 (09) Normal Plant Evolutions; Core Performance Date Tested: 04/24/91 Run Time: 5.00 hou.s Test

Description:

This test verifies that core thermal limits are within Perry Unit 1 Technical Specification limits at various combinations of reactor power and core flow. Baseline Data used For Reference Reference Type: Plant Data Cycle 1 Benchmark Data GENERATED BY PERRY REACTOR ENGINEERING SECTION Technical Specifications PERRY UNIT 1 TECHNICAL SPECIFICATIONS Malfunctions Tested: None Hajor Discrepancies: P-1335 DURING CORE TEST TO VERIFY THAT THERMAL LIMITS WERF HUT ?.XCET.OED DURING OPERATION ALLOWED BY THE POWER / FLOW MAP, THERMAL LIMIT

WERE EXCEEDED AT SOME CORE FLOW /RX POWER CONDITIONS.

THERMAL LIMITS SHOULD O NOT BE EXCEEDED. Evaluators: C.

Persson, B.

Panfil, R.

Kearney, G.

Minshall, J.

Pierson O

1 ] l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Steady State and Normal Operations TAB C - Part 3A Tests Abstracts PAGE 7 of 9 Test T.4.2.6, CORE StBCRITICAL MULTIPLICATION TEST Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.1 (09) Normal Plant Evolutions; Core Performance Date Tested: 05/23/91 Run Times 0.50 hours Test

Description:

This test verifies the proper response of the core model neutron population during a reactor startup with power in the source range. Baseline Data used For Reference Reference Type Plant Data Cycle 1 Benchmark Data GENERATED BY PERRY REACTOR ENGINEERING S ECTIO*: Malfur.ctions Tested: None Discrepancies: None Evaluators: C.

Persson, B.

Panfil, J.

Steward, B.

Stetson, J.

McHugh . - - - _ _ _ - _ _ _ _ _ _ - ~ _ _ _ _. _ _ _ _ _

CERTIFICATION OF PERRY CATION FACILITY DOCKET NO. 50-440 Le and Normal Operations TAD C - Part 3A Simulator Tests - Steady Tests Abstracts PAGE 8 of 9 Test T.4.2.7, REhCTOR CORE LIFE TEST Revision Number: 02 ANSI /AN3 3.5 Section: 3.1.1 (09) Norr:sl Plant Evolutions; Core Performance Dato 'fested 05/23/91 Run Times 2.50 hours Tout

Description:

This tost verifies that the amount of control rod withdrawal needed to reach criticality, as well as Nuclear Instrumentation response, change appropriately as core age incroaces. Basoline Data used For Reference Reference Typo Plaitt Data cycle 1 Benchmark Data GENEPATED BY PERRY REACTOR ENGINEERING SECTION Malfunctions Tested: None Discrepancies: None Evaluators: C.

Persson, B.

Panfil, J.

Steward, H.

Stetson, J.

McHugh iv s l v

a l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET 110. 50-440 Simulator Tests - Steady State and Normal Operationu TAB C - Part 3A Tests Abstracts PAGE 9 of 9 =N Test 7.4.2.8, SHU7DOWN MARGI!4 DEMONSTRATION Revision !Jumber: 02 ANSI /'.lS 3.5 Section 3.1.1 (09} tiotmal Plant Evolutions; Core Performance Date Tested: 05/23/91 Run Times 0.50 hours Test

Description:

This test verifles that the shutdown margin for the simulated BOL and EOL cores comply with Perry Unit 1 Technical Specification limits. Baseline Data used For Reference heforence Type: Plant Data Cycle 1 Benchmark Dt.ta GENERATED BY PERRY REACTOR ENGItJEERIfJG SECTION Technicel Sr(cifications PERRY UNIT 1 TECH!i! CAL SPECIFICATIONS Malfunctions issted lione Discrepaneleon Hone Evaluators: C.

Persson, B.

Panfil, J.

Steward, B.

Stetson, J. McIlagh O

CERTIFICATION oF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Steady State and Normal operations TAD c - Part 3B Tests Abstracto PAGE 1 of 13 ATP TEST SECTION T.4.3 CONTINUOUS PLANT OPERATION Overview of testina performed in this section The purpose of this series of tests is to demonstrato the simulator's ability to perform a completo nuclear plant startup and shutdown starting from cold shutdown conditions (with a minimum of support systens operating) up to 100% rated power, and back to the cold shutdown condition. The tests were written directly from the PNPP Unit 1 Operations Manual, and only operations directed by procedures in use in the Perry Unit 1 Control Room are performed. The tests are written such that the entire startup and shutdown evolution is continuous, with each test proceeding from the end of the prior tost. All indicators, computer roadouts, logs, alarms, and other forms of man-machine interface available in the Perry Unit 1 Control Room were verified to the critoria of ANSI /ANS 3.5. All operator conducted survoillancos and other periodic tests were also tested within this section. These tests woro performed per section 3.2(2) of Appendix A of ANSI /ANS 3.5 as required by section 5.4.1 for initial construction performance testing. No. of Tests included in this Section IMgg t Onon SDR'st T.4.3.1, C0!O $HUTDOWN 10 RE ACfDR CRiflCAL 2 9' T.4.3.2, REACTOR CRl11 CAL 70 TUR$lkt $YhthRONIZED 5 14' T.4.3.3, POWER thCRE Att to 100% POWER 7 9' f.4.3.4, POWER DiCREA$t to TURBlWE/GtWERATOR UWLOADED 10 Wone T.4.3.$, PLAWT COOLDOWW TO COLD $NU100WN 12 8* ANSI /ANS 3.5 ReferenCat Appendix A 3.2 Normal Oporations This test has not been completed (N/C) or has been performed but has one or' more unrosolved ma-)or Sir.ulator Discrepancy Reports associated with it. Please see the individual test abstract for these discropancies which constituto exceptions to ANSI /ANS 3.5 section 3.1. !O _ _ ~,. _ _ _ _

1 i l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 i O Simulator Teots - Steady State and Normal Operations TAB C - Part 3B Tests Abstracts PAGE 2 of 13 Test T.4.3.1, COLD SHUTDOWN TO REACTOR CRITICAL [ Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.1 (01) Normal Plant Evolutions; Plant Startup - Cold to Hot Standby Date Tested: 05/22/91 Run Times 14.20 hours i Test

Description:

The purpose of this test is to demonstrate the f simulator's ability to conduct a normal plant startup from cold iron i i }, conditions using only operator action normal to Perry Unit 1. At the l beginning of the test, the reactor is shutdown, as are most supporting { systems (2001). Supporting BOP systems are started per the cold startup lineup (attachment I to 101-2). Safety systems are placed in the Standby Readiness mode. A reactor startup to reactor critical is t then performed per Perry Operations Manual Integrated Operating Inntruction 201-1. All instructions used as references for this test i are the same ones used by Operators in the Perry Unit 1 Control Room, l At the completion of the test, the reactor is critical and rcady for Nuclear Heatup. Baseline Data used For Reference O Reference Types Normal Operating Procedures I FTI-B02 CONTROL ROD MOVEMENTS 101-1 COLD STARTUP e SOI-B21 NUCLEAR STM SUPPLY SHUTOFF, AUTO DEPRESSURIEATION, & NSSS [ SOI-B33 REACTOR RECIRCULATION SYSTEM SOI-C11(CRDH) CONTROL ROD DRIVE HYDRAULIC SYSTEM (UNIT 1) SOI-C11(RCIS) CONTROL ROD & INFORMATION SYSTEM (UNIT 1) SOI-C85 STEAM BYPASS & PRESSURE REGULATOR SYSTEM (UNIT 1) SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-E21 LOW PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E22A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E22B DIVISION 3 DIESEL GENERATOR (UNIT 1) SOI-E31 LEAK DETECTION SYSTEM (UdIT 1) SOI-033 REACTOR WATER CLEANUP SYSTEM (UNIT 1) SOI-041(FPCC) FUEL POOL COOLING & CLEANUP SYSTEM (UNIT 1) SOI-G41(FPFD) FUEL E30L FILTER DEMINERALIEER SYSTEM (UNIT 1) i SOI-M11 CONTAINMENT VESSEL COOLING SYSTEM (UNIT 1) SOI-M13 DRYWELL COOLING SYSTEM (UNIT 1) SOI-M14 CONTAINHENT VESSEL & DRYWELL PURGE SYSTEM (UNIT 1) l SOI-M15 ANNULUS EXHAUST GAS TREATHENT SYSTEM (UNIT 1) LOI-M21 CONTROLLED ACCESS & MISC EQUIPMENT AREA (CA& MEA) HVAC SYSTEMS SOI-M23/24 MCC, SWGR, & MISC ELECTRICAL EQUIPMENT AREA HVAC SYSTEMS SOI-M25/26 CONTROL ROOH HVAC & EMERGENCY RECIRCULATION SYSTEM l SOI-M28 EMERGENCY CLOSED COOLING PUMP AREA COOLING SYSTEM SOI-M32 EMERCENCY SERVICE WATER PUMP HOUSE VENTILATION SYSTEM (UNIT 1) l SOI-M33 INTERMEDIATE BUILDING VENTILATION SYSTEM [ 1

i r CERTIFICATION OP PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests i Steady State and Normal Operations TAB C - Part 3B Tests Abstracts PAGE 3 of 13 Test T.4.3.1, COLD SHUTDOWN 10 REACTOR CRITICAL SOI-H35 TURBINE BUILDING VENTILATION SYSTEM (UNIT 1) SOI-Pk6 OFF-GAS BUILDING EKHAUST SYSTEM (UNIT 1) l SOI-H38 AUXILIARY BUILDING VENTILATION SYSTEM (UNIT 1) SOI-H39 ECCS PUMP ROOMS COOLING SYSTEM (UNIT 1) i SOI-H40 FUEL HANDLING BUILDING VENTILATION SYNTEM SOI-H41 HEATER BAY VENTILATION SYSTEM (UNIT 1) SOI-H42 TURBINE POWER COMPT,EX VENTILATION SYSTEM SOI-H43 DIESEL GENERATOR BUILDING VENTILATION SYSTEM (UNIT 1) SOI-H47 STEAM TUNNEL COOLING SYSTEM (UNIT 1) SOI-N11 MAIN & REHEAT STEAM SYSTEM (UNIT 1) SOI-N21 CONDENSATE SYSTEM (UNIT 1) SOI-N23 CONDENSATE FILTRATION SYSTEM (UNIT 1) 801-N24 CONDENSATE DEMINERALIZER SYSTEM (UNIT 1) SOI-N27 FEEDWATER SYSTEM (UNIT 1) kvI-N32/39 MAIN TURBINE & TURNING GEAR SYSTEM (UNIT 1) SOI-N33 STEAM SEAL SYSTEM (UNIT 1) SOI-N34 MAIN LUBE OIL SYSTEM (UNIT 3) i SOI-N42 HYDROGEN SEAL OIL SYSTEM (UNIT 1) SOI-N43 STATOR WATER COOLING SYSTEM (UNIT 1) SOI-N64(OGVRS) OFF-GAS VAULT REFRIGERATION SYSTEM (UNIT 1) SOI-N64/62 OFF-GAS / CONDENSER AIk REMOVAL SYSTEM SOI-N71 CIRCULATING WATER / CONDENSER HECHANICAL CLEANING SYSTEM (UNIT 1) SOI-Pil CONDENSATE TRANSFER & STORAGE SYSTEM (UNIT 1) O SOI-P12 CONDENSATE SEAL SYSTEM (UNIT 1) SOI-P40/41 SERVICE WATER & SERVICE WATER SCREEN WASH SOI-P42 EMERGENCY CLOSED COOLING SYSTEM (UNIT 1) SOI-P43 NUCLEAR CLOSED COOLING SYSTEM SOI-P44 TURBINE BUILDING CLOSED COOLING SYSTEM (UNIT 1) SOI-P45 EMERGENCY SERVICE WATER SYSTEM (UNIT 1) SOI-P46 TURBINE BUILDING CHILLED WATER SYSTEM (UNIT 1) SOI-P4' CONTROL COMPLEX CHILLED WATER SYSTEM SOI-P50 CONTAINHENT VESSEL CHILLED WATER SYSTEM (UNIT 1) SOI-P54 FIRE PROTECTION SYSTEM SOI-R13 ISOLATION PHASE BUS DUCT COOLING SYSTEM (UNIT 1) f SOI-R43 DIVISION 1 & 2 DIESEL GENERATOR SYSTEM (UNIT 1) SOI-R44 DIVISION 1 & 2 DIESEL GENERATOR STARTING AIR SYSTEM (UNIT 1) SOI-R44/E22B DIVISION 3 DIESEL GENERATOR STARTING AIR SYSTEM (UNIT 1) SOI-R45 DIVISION 1 & 2 DIESEL GENERATOR FUEL OIL SYSTEM (UNIT 1) SOI-R45/E22B DIVISION 3 DIESEL GENERATOR FUEL OIL SYSTEM (UNIT 1) Reference Types Surveillance Procedures SVI-B33-T1160 JET PUMP OPERADILITY SVI-C11-T1019 ROD PATTERN CONTROLLER SYS TEST BELOW LOW POWER SETPOINT SVI-M14-T2003 CONTAINHENT INBOARD & DRYWELL PURGE SUPPLY / EXHAUST ISOLATION DAMPERS OPERADILITY TEST I l i ' O I 1 . ~....-

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Steady State and Normal Operations TAB C - Part 3B Tests Abstracts PAGE 4 of 13 Test T.4.3.1, COLD SHUTDOWN TO REACTOR CRITICAL Halfunctions Testedt None Major Discrepancies: I F-0003 M14 SYSTEM NEVER SHIFTED TO THE DP CONTROLLER. F-0024 UNABLE TO KEEP CRDH PUNP RUNNING ON INITIAL START, TRIPS ON LOW SUCTION PRESSURE. F-0032 WITH NO FLOW THROUGH THE OFF GAS SYSTEH, COOLER CONDENSER i OUTLET TEMP 'B" WENT TO 40 DEG AS SOON AS GLYCOL WAS COOLED l DOWN AND "A" SHOWED NO CHANGE. F-0034 WHEN OFF GAS RECOMBINERS HEATERS WERE STARTED, RECOMD A TEMP SHOWED NO INCREASE OVER 3 HOURS AND D WENT TO NORMAL TEMP PROFILE IN 5 HIN. F-0035 THE VENTING OF RR PUMP SEALS HAD AN INAPPROPRIATE EFFECT ON CRD AND ON RR SYSTEM (OUTER SEAL LEAKAGE). i F-0039 CRD FLOW / PRESSURE OSCILLATIONS AND INTERMITTENT SUCTION FILTER DP ALARM. HI DP ALARM NOT CLEARING UNTIL FLOW IS <50 GPH. i F-0048 WHEN ESTABLISHING FLOW TO HOT SURGE TANK, OBSERVED FLOW, PRESS INCREASES WERE SEEN ON CRDH SYS W/ NO CRDH PUMP IN OPERATION. F-0049 WHEN THE HECHANICAL VACUUM PUMPS WERE STARTED, OBSERVED NO CHANGE IN OG VENT FLOW. F-0050 ROD BLOCKS ARE RECEIVED WHEN GROUP 1 AND 2 ARE WITHDRAWU TO 48. Evaluators: C. Persson, B. Panfil, J. Steward, B. Stetson, J. McHugh 4 i I i 4 O L ~.....

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Steady State and Normal Operations TAB C - Part 3D Tests Abstracts PAGE 5 of 13 Test T.4.3.2, REACTOR CRITICAL TO TURBINE SYNCHRONIEED Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.1 (02) Normal Plant Evolutions; Nuclear Startup from Hot Standby to Rated Power Date Tested: 05/25/91 Run Times 16.65 hours Test

Description:

The purpose of this test is demonstrate the simulators ability to conduct a normal plant startup and heatup from reactor criticality to generator on-line, using only operator action normal to Perry Unit 1. At the beginaisng of this test, the simulated l plant has just completed a reactor startup to criticality. A nuclear i plant heatup is performed and BOP systems are started at various Reactor Pressures in accordance with the System Operating Instructions and IOI-1. Special periodic tests are also performed on NSSS Systems such as RCIC, RCGIS, and SRV's during the startup. The Reactor Feed Pumps, Off-Gas system, Steam Seal System, and various other auxiliaries are started in sequence. At rated system pressure, the Main Turbine is warmed up and a Turbine Roll is conducted. At the conclusion of the test, the Reactor is at 12% Power, and the Main Turbine Generator is connected to the Grid supplying approximately 100 HWo. Baseline Data used For Reference Reference Type Normal Operating Procedures 201-1 COLD STARTUP SOI-B21 NUCLEAR STM SUPPLY SHUTOFF, AUTO DEPRESSURIEATION, & NSSS SOI-C34 FEEDWATER CONTROL SYSTEM (UNIT 1) SOI-C51(IRM) INTERMEDIATE RANGE MONITORING SYSTEM (UNIT 1) SOI-C51(SRM) SOURCE RANGE MONITORING SYSTEM (UNIT 1) SOI-E51 RCACTOR OORE ISOLATION COOLING SYSTEM (UNIT 1) SOI-033 REACTOR WATER CLEANUP SYSTEM (UNIT 1) SO1-N11 MAIN & REHEAT STEAM SYSTEM (UNIT 1) SOI-N23 CONDENSATE FILTRATION SYSTEM (UNIT 1) SOI-N24 CONDENSATE DEMINERALIEER SYSTEM (UNIT 1) SOI-N27 FEEDWATER SYSTEM (UNIT 1) SOI-N32/39 MAIN TURBINE & TURNING GEAR SYSTEM (UNIT 1) SOI-N33 STEAM SEAL SYSTEM (UNIT 1) SOI-N64/62 OFF-GAS / CONDENSER AIR REMOVAL SYSTEM SOI-S11 POWER TRANSFORMERS Reference Type Surveillance & Periodic Test Procedures PTI-N32-P0001 TURB OVERSPEED PROT DEVICES TRIP.& EHC/TURB LUBE OIL PUMP STARTS / STATOR WATER PUMP START, & ROTATIONS WEEKLY TEST SVI-D21-T2001 MSIV FULL STROKE OPERABILITY TEST SVI-B21-T2005 SRV EXERCISE TEST l SVI-E51-T1272 PROC / INSTR PERIODIC REVIEW - RCIC SYSTEM LOW PRESS TEST SVI-E51-T2001 RCIC PUMP & VALVE OPERABILITY TEST SVI-N31-T1151 MAIN TURBINE VALVE EXERCISE TEST l O i l .. ~ - -... - -.

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Steady State and Normal Operations TAB C - Part 3B Tests Abstracts PAGE 6 of 13 Test T.4.3.2, REACTOR CRITICAL TO TURDINE SYNCHRONIEED Malfunctions Tested None Major Discrepancies: 1 P-1514 IF RWCU TRIPS AT APPROXIKATELY 250d, PRESSURE AND FLOWS ARE UNSTABLE P-1515 NUMEROUS 00 PARAMETERS OUT OF BAf 'T WITHIN LAWS OF PilYSICS. F-0028 IN MANUAL AND AUTO CONTROL WAS AD1 3 CllANGE RFPT SPEED, STEAM FI 'M WITH NO CHANG % IN TURBINE 00: , VALVE POSITION. F-0059 RL LEVEL RESPONSE TO RPV HEATUP IE APPROPRIATELY LARGE. F-0060 RWCU FLOW OSCILLATES WITil NO CHANGE IN INPUT PARAMETERS. F-0062 RCIC SYSTEM CANNOT ACHIEVE RATED FLOWS AND PRESSURE AT 150-200 PSI RPV PRESSURE. F-0074 Ti!E RHR HEAT EXCHANGERS DID NOT COOL THE SUPP POOL AT A RAPID ENOUGH RATE. F-0080 PTS 1-5, 7-13, 15, 16, 19 ON RECORDER N31-R001 AREN'T PRINTING PROPERLY. F-0083 THE MSL LOW PRESSURE ALARM WAS STILL IN AT 920#. F-0086 DURING HEATUP OF MSL A/B/C/D, TIIE C MSL WENT TO 465 DEGREES IN SECONDS, THE OTHER 3 DO NOT INCREASE AT ALL. F-0091 CONTROL ROD GROUP 6 FROM 00 TO 08 !!AS A DRASTIC EFFECT ON REACTOR LEVEL. F-0091 PLACING SJAE A IN SERVICE NOTED SEVERAL PROBLEMS. t O F-0094 THE TURDINE PRESSURE DOES NOT DROP OFF TO 0# WHEN THE NUMBER 2 STOP VALVE PILOT IS CLOSED. F-0115 SEE ATTACHED LIST OF CONTROLLERS (AUTO /KANUAL). WilEN CONTROLLER IS IN AUTO, THE OPEN AND CLOSE PUS!!DUTTONS MOVE Ti!E OUTPUT. Evaluators B. Panfil, C. Persson, D. Stetson, J. Steward, J. McHugh i O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Dimulator Tests - Steady State and Normal Operations TAD C - Part 38 Tests Abstracts PAGE 7 of 13 Test T.4.3.3, POWER INCREASE TO 1004 POWER Revision Nunber: 02 ANDI/ANS 3,5 Section: 3.1.1 (02) Normal Plant Evolutions; Nuclear Startup from Hot Standby to Rated Power Date Tested: 05/31/91 Run Times 8.70 hours Test Description The purpose of this test is to demonstrate the simulator's ability to conduct a plant power increase to 100% rated power and operator conducted surveillances, using only operator action normal to Perry Unit 1. At the beginning of this test, the main Turbine-Generator is synchronized to the grid and Reactor Power is 12%. The Unit is maneuvered in accordance with Perry Procedures 101-1 and IOI-3 to the full power condition. The Power increase is achieved by Control Rod withdrawal and core flow increases. Supporting equipment such as additional feed pumps and MSR's are started as directed by the procedures as power is increased. Once the 100% operating condition is reached, additional sub-tests are performed to validate simulator performance. These sub-testa consist of operator conducted surveillances on safety related equipment and system operating instruction sections not performed as part of a normal startup and shutdown, such as shifting pumps, fans, controllers, etc. At the conclusion of this test, the Reactor is operating at 1004 power and the Main Generator is supplying Unit rated electrical power. s Baseline Data used For Reference Reference Type: Normal Operating Procedures FTI-B02 CONTROL R00 MOVEMENTS 101-1 COLD STARTUP 101-3 POWER CMANGES SOI-833 REACTOR RECIRCULATION SYSTEM SOI-C11(CRDH) CONTROL ROD DRIVE HYDRAULIC SYSTEM (UNIT 1) SOI-C11(RCIS) CONTROL ROD & INFORMATION SYSTEM (UNIT 1) SOI-C22 REDUNDANT REACTIVITY CONTROL SYSTEM (UNIT 1) SOI-C34 FEEDWATER CONTROL SYSTEM (UNIT 1) SOI-C51(APRM) AVERAGE POWER RANGE MONITORING SYSTEM (UNIT 1) SOI-CBS STEAM BYPASS & PRESSURE REGULATOR SYSTEM (UNIT 1) SOI-D21 AREA RADIATION MONITORING SYSTEM (D21) SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-E21 LOW PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E22A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E22B DIVISION 3 DIESEL GENERATOR (UNIT 1) SOI-E51 REACTOR CORE ISOLATION COOLING SYSTEM (UNJT 1) SOI-033 REACTOR WATER CLEANUP SYSTEM (UNIT 1) SOI-036 RWCU FILTER / DEMINERALISER SYSTEM (UNIT 1) SOI-G41(FPCC) FUEL POOL COOLING & CLEANUP SYSTEM (UNIT 1) l SOI-G42 SUPPRESSION POOL CLEANUP SYSTEM (UNIT 1) l SOI-M11 CONTAINMENT VESSEL COOLING SYSTEM (UNIT 1) Soi-M13 DRYWELL COOLING SYSTEM (UNIT 1) i SOI-M14 CONTAINMENT VESSEL & DRYWELL PURGE SYSTEM (UNIT 1) ,__,m....__ ..- -.s

CERTIFICATION OF PERRY SIMUT.ATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Steady State and Normal Operations TAB C - Part 3B Testa Abstracts PAGE 8 of 13 Test T.4.3.3, POWER INCREASE TO 200% POWER SOI-M15 ANNULUS EXHAUST GAS TREATMENT SYS7EM (UNIT 1) SOI-M21 CONTROLLED ACCESS & HISC EQUIPMENT AREA (CA& MEA) HVAC SYSTEMS SOI-M23/24 MCC, SWGR, & HISC ELECTRICAL EQUIPMENT AREA HVAC SYSTEMS SOI-M25/26 OONTROL ROOH HVAC & EMERGENCY RECIRCULATION SYSTEM SOI-M28 EMERGENCY CLOSED COOLING PUMP AREA COOLING SYSTEM SOI-M32 EMERGENCY SERVICE WATER PUMP HOUSE VENTILATION SYSTEM (UNIT 1) SOI-M36 OFF-GAS BUILDING EXRAUST SYSTEM (UNIT 1) i SOI-M39 ECCS PUMP ROOMS COOLING SYSTEM (UNIT 1) SOI-M40 FUEL HANDLING BUILDING VENTILATION SYSTEM j SOI-M43 DIESEL GENERATOR BUILDING VENTILATION SYSTEM (UNIT 1) 1 SOI-M51/56 COMBUSTIBLE GAS CONTROL SYSTEM & HYDROGEN IGNITERS (UNIT 1) SOI-N11 KAIN & REHEAT STEAM SYSTEM (UNIT 1) SOI-N21 CONDENSATE SYSTEM (UNIT 1) SOI-N23 CONDENSATE FILTRATION SYSTEM (UNIT 1) t SOI-N24 CONDENSATE DEMINERALIZER SYSTEM (UNIT 1) SOI-N27 FEEDWATER SYSTEM (UNIT 1) SOI-N32/39 MAIN TURBINE & TURNINO GEAR SYSTEM (UNIT 1) SOI-N33 STEAM SEAL SYSTEM (UNIT 1) SOI-N41/51 KAIN GENERATOR & EXCITATION SYSTEM (UNIT 1) SOI-N42 HYDROGEN SEAL OIL SYSTEM (UNIT 1) SOI-N43 STATOR WATER COOLING SYSTEM (UNIT 1) SOI-N64/62 OFF-GAS / CONDENSER AIR REMOVAL SYSTEM SOI-N71 CIRCULATING WATER / CONDENSER MECHANICAL CLEANING SYSTEM (UNIT 1) O SOI-Pil CONDENSATE TRANSFER & STORAGE SYSTEM (UNIT 1) SOI-P42 EMERGENCY CLOSED COOLING SYSTEM (UNIT 1) SOI-P45 EMERGENCY SERVICE WATER SYSTEM (UNIT 1) SOI-P47 OONTROL COMPLEX CHILLED WATER SYSTEM SOI-P51/52 SERVICE & INSTRUMENT AIR SYSTEM SOI-R10 PLANT ELECTRICAL SYSTEM SOI-R13 ISOLATION PHASE BUS DUCT COOLING SYSTEM (UNIT 1) SOI-R14 120V AC VITAL INVERTERS (UNIT 1) SOI-R42(DIV 1) DIV 1 DC DISTR BUSES ED-1-A & ED-2-At BATTERIES, CHARGERS, & SWITCHGEAR SOI-R42(DIV.2) DIV 2 DC DISTR BUSES ED-1-B & ED-2-B BATTERIES, CHARGERS, & SWITCHGEAR SOI-R42(DIV 3) DIV 3 DC DISTR BUSES ED-1-C & ED-2-C BATTERIES, CHARGERS, & SWITCHGEAR SOI-R42(SYS A) NON DIV DC SYS A DISTR BUSES D-1-A & D-2-At BATTERIES, CHARGERS, & SWITCHGEAR SOI-A42(SYS B) NON DIV DC SYS B DISTR BUSES D-1-B & D-2-B BATTERIES, CHARGERS, & SWITCHGEAR SOI-R43 DIVISION 1 & 2 DIESEL GENERATOR SYSTEM (UNIT 1) Reference Types Survalliance & Periodic Test Procedures PTI-N27-P0001 REACTOR FEED PUMP TURBINE STOP VALVE TEST PTI-N27-P0002 REACTOR TEED PUMP TURBINES THRUST BEARING WEAR TEST PTI-N27-P0005 RTP TURBINES LOCKOUT SUPPRESSED OVERSPEED TRIP TEST PTI-N27-P0006 EMERGENCY LUBE OIL PUMP TEST FOR MOTOR DRIVEN FEED PUMP PTI-N32-P0002 TURBINE - GENERATOR OIL SYSTEMS MONTHLY TESTING PTI-N36-P0001 WEEKLY NON-RETURN CHECK VALVE TEST SVI-B33-71160- JET PUMP OPERABILITY SVI-C11-T0009 CONTROL ROD SCRAM ACCUMULATOR PRESS / LEAK DETECTION TESTS O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simuistor Tests - Steady State and Normal Operations TAB C - Part 3B Teste Abstracts PAGE 9 of 13 Test T.4.3.3, POWER INCREASE TO 100% POWER SVI-C11-71003 CONTROL ROD EXERCISE SVI-C11-T1022 ROD PATTERN CONTROL SYSTEM - ROD WITHDRAWAL LIMITER SVI-C41-T2001 STANDBY LIQUID CONTROL PUMP & VALVE OPERABILITY TFST SVI-C51-T0024 APRM GAIN & CHANNEL CALIBRATION SVI-C85-T1314 TURBINE BYPASS VALVE OPERABILITY TEST SVI-E22-T1319 DIESEL GENERATOR START & LOAD DIVISION SVI-M26-T1259 CONTROL ROOM EMERGENCY RECIRCULATION OPERABILITY T2ST SVI-M51-72003 COMBUSTIBLE CAS MIXING SYSTEM OPERABILITY TEST SVI-R10-T5217 ELECTRICAL DISTRIBUTION SYSTEM ENERGIEATION CHECK SVI-R43-T1317 DIESEL GENERATOR START & LOAD DIVISION 1 Malfunctions Tested None Major Discrepancies: F-0119 CAN'T GET JET PUMP OPERABILITY SVI-B33-T1160 TO PASS. F-0126 COULD NOT TRANSFER B33-K603B TO MANUAL, AFTER IT WAS IN AUTO. F-0127 COULD NOT GET THE A LOOP FLOW CONTROL FOR REACTOR RECIRC ON THE AUTO CONTROLLER WITHOUT A SK LOOP FLOW MISMATCH. F-0128 201-1 REQUIRES FEEDWATER TEMPERATURE TO BE GREATER THAN PDB-A011. THIS IS AT 325 DEG F, SIMULATOR SHOWED 292 DEG F. F-0130 WHILE SHIFTING FROM RX FEED PUMP A ON THE S/U LEVEL CONTROLLER TO THE MASTER LEVEL CONTROLLER, NOTED A SEVERE BUMP OCCURRED. O F-0133 THE OUTER SEALS ON BOTH REACTOR RECIRC PUMPS ARE FAILED. F-0139 OSCILLATIONS IN TURBINE IST AND 2ND STAGE REHEAT FLOW STM FLOW OCCUR WHENEVER TURBINE LOAD IS CHANGED. F-0141 PTV 2P41F003 IS NOT RESPONDING TO MAINTAIN TBCC TEMrERATURE. F-0144 COULD NOT BRING ON ONE REACTOR WATER CLEANUP FILTET. AT A TIME. BOTH CAME UP TOGETHER WHEN B WAS PLACED ON LINE. Evaluators: B. Panfil, C. Persson, J. Steward, J. McHugh t O

CERTIFICATION OF PERRY DIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Steady State and Normal Operations TAB C - Part 3B Tests Abstracts PAGE 10 of 13 Test T.4.3.4, POWER DECREASE TO TURBINE / GENERATOR UNLOADED Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.1 (08) Normal Plant Evolutions; Plant shutdown from Rated Power to Hot Standby Date Tested: 06/10/91 Run Times 8.00 hours Test

Description:

The purpose of this test is to demonstrate the simulator's ability to conduct a normal plant shutdown from 2004 power to generator off-line, using only operator action normal to Perry Unit 1. At the beginning of this test, the simulated plant is operating at rated 100% Power capacity. The test consists of performing a normal reactor plant power decrease per Operating Procedures IOI-3, 101-4 and supporting SOI's contained in the PNPP Unit 1 Operations Manual. Reactor Powet-is decreased by inserting control rods and lowering reactor racirculation core flow. Supporting plant systems are similarly maneuvered per the IOI. The turbine is completely unloaded and shutdown. At the completion of the test, Reactor Power is approximately 84, steam is bypassed to the Main Condenser, and the turbine is coasting down from 1800 RPM. Baseline Data used For Reference Reference Type Normal Operating Procedures FTI-B02 CONTROL ROD MOVEMENTS IOI-3 POWER CHANGES 10I-4 SHUTDOWN SOI-B33 REACTOR RECIRCULATION SYSTEM SOI-C11(RCIS) CONTROL ROD & INFORMATION SYSTEM (UNIT 1) SOI-C34 FEEDWATER CONTROL SYSTEM (LNIT 1) SOI-C51(IRM) INTERMEDIATE RANGE MONITORING SYSTEM (UNIT 1) SOI-033 REACTOR WATER CLEANUP SYSTEM (UNIT 1) SOI-N11 MAIN & REHEAT STEAM SYSTEM (UNIT 1) SOI-N21 CONDENSATE SYSTEM (UNIT 1) SOI-N23 CONDENSATE FILTRATION SYSTEM (UNIT 1) SOI-N24 CONDENSATE DEMINERALIZER SYSTEM (Ulf1T 1) SOI-N27 FEEDWATER SYSTEM (UNIT 1) SOI-N32/39 MAIN TURBINE & TURNING GEAR SYSTEM (UNIT 1) SOI-RIO PLANT ELECTRICAL SYSTEM SOI-R13 ISOLATION PHASE BUS DUCT COOLING SYSTEM (UNIT 1) SOI-Sil POWER TRANSFORMERS Reference Type Surveillance Procedures SVI-C11-T1019 ROD PATTERN CONTROLLER SYS TEST BELOW LOW POWER SETPOINT SVI-C11-T1022 ROD PATTERN CONTROL SYSTEM - ROD WITHDRAWAL LIMITER O

___________._.___._~m__- _. - ~ CERTIFICATION OF PERRY SIHULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Steady State and Normal Operations TAB C - Part 3B Tests Abstracts PAGE 11 of 13 Test T.4.3.4, POWER DECREASE TO TURBINE / GENERATOR UNLOADED Malfunctions Tested: None Major Discrepancies: None Evaluators: B. Panfil, C. Persson, c. Minshall, H. DeBoer ( r I I t l f a { f l l f

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Steady State and Normal Operations TAD C - Part 3B Tests Abstracto PAGE 12 of 13 Tent T.4.3.5, PLANT COOLDOWN TO COLD SHUTDOWN Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.1 (08) Normal Plant Evolutions; Shutdown and cooldown to Cold Shutdown Conditions l Date Tested: 06/10/91 Run Time 5.50 hours Test

Description:

The purpose of this test is to demonstrate the simulator's ability to conduct a ncimal reactor plant shutdown and cooldown from generator off-line to the cold shutdown condition (<200 deg F), using only operator action normal to Perry Unit 1. At the beginning of this test, the reactor is at about 8s power. Control rode are inserted to complete the reactor shutdown. Auxiliary DOP systems are shutdown as instructed by the Operating Procedure IOI-4. Af ter i he reactor is shutdown, the plant is lined up to support a plant cooldown. Steam is dumped to the main condenser and the i cooldown is controlled <100 deg F/hr as it is in the reference plant. The RHR systems are flushed and prepared to support the shutdown cooling isode of operation. At the prescribed temperature, cooldown is shifted to the RHR system, and the steam plant is shutdown. Cooldown t is continued with RHR until the reactor temperature is in the normal cold shutdown band of 120-140 deg F. At the test conclusion, the steam systems are shutdown, vacuum in the main condenser has been broken, RRC pumps are secured, and temperature is being controlled by RHR in SDC mode. Baseline Data used For Reference Reference Type Normal Operating Procedures FTI-B02 COUTROL ROD MOVEMENTS IOI-12 MAIhTAINING COLD SHUTDOWN IOI-4 SHUTDOWN SOI-B21 NUCLEAR STM SUPPLY SHUTOFF, AUTO DEPRESSURIZATION, & NSSS SOI-B33 REACTOR RECIRCULATION SYSTEM SOI-C11(CRDH) CONTROL ROD DRIVE HYDRAULIC SYSTEM (UNIT 1) SOI-C34 FEEDWATER CONTROL SYSTEM (UNIT 1) SOI-C51(SRM) SOURCE RANGE MONITORING SYSTEM (UNIT 1) SOI-C71 RPS POWER SUPPLY DISTRIBUTION (UNIT 1) SOI-C85 STtAM BYPASS & PRESSURE REGULATOR SYSTEM (UNIT 1) SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-G33 REACTOR WATER CLEANUP SYSTEM (UNIT 1) SOI-N21 CONDENSATE SYSTEM (UNIT 1) SOI-N24 CONDENSATE DEMINERALIZER SYSTEM (UNIT 1) SOI-H27 FEEDWATER SYSTEM (UNIT 1) i SOI-N33 STEAM SEAL SYSTEM (UNIT 1) SOI-H64(OGVRS) OFF-GAS VAULT REFRIGERATION SYSTEM (UNIT 1) SOI-N64/62 OFF-GAS / CONDENSER AIR REMOVAL SYSTEM SOI-P12 CONDENSATE SEAL SYSTEM (UNIT 1) O l l

j CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Steady State and Not tal Operations TAB C - Part 3B Tests Abstracts PAGE 13 of 13 Test T.4.3.5, PLANT COOLDOWN TO COLD SHUTDOWN SOI-P42 EMERGr.NCY CLOSED COOLING SYSTEM (UNIT 1) SOI-P45 EMERGENCY SERVIs,E WATER SYSTEM (UNIT 1) SOI-P54 FIhE PROTECTION SYSTEM Reference Types Surveillance Procedures SVI-B21-T1176 OH7A SVI RCS HEATUP h COOLDOWN SURVEILLANCE Malfunctions Tested: None Major Discrepancies: P-0125 STEPS 28-35 GF T.4.3.5 HUST BE REPEATED. Ri!R INDICATIONS FOR FLUSH PRIOR TO INITI ATIt'G SDC MODE IMPPOPER. F-0190 WHILE PERFORMING HEAD SPRAY SHELL AND t!EAD FLANGE TEMPS ARE 100 DEGREES I.ESS THAh BOTTOM DRAIN TEMPS. F-0191 WHILE FILLING RX VESSEL TO 490-500", COULDN'T FILL ABOVE 400" PITHOUT CAUSING RPV PRESS TO INCR1iSE. F-0193 WP.LE DECREASING PRESSURE WITH BYFASS VALVES A 10# PRESS Di,REASE CAUSES 25" TO 30" LEVEL INCREASE (SWELL). F-0198 WnILE PERFORMING RHR S/D COOLING FLUSil WAS UNABLE TO MAINTAIN RHR HEADER PRESS WITH MINIMAL FLOW TilROTTLE POSITION (F040). F-0199 UNABLE TO OBTAIN PROPER FLOW RATE (6900-7100) WHILE IN SUPPRESSION P00L COOLING WITH ALL SYSTEM FLOW TilROUGH RHR A O HEAT EXCllANGERS. F-0201 REMOTE FUNCTION Ril24 (IE12-F315) NEEDS TO !! AVE EITHER THROTTLING OR MANUAL OPERATION CA? ABILITY TO ALLOW FOR VESSEL TEED AT AN ADJUSTABLE RATE TO SUPPORT COLD SHUTDOWN OPERATION. F-0208 WHEN ATTEMPTING TO FEED THE VESSEL WITH HOTWELL PUMPS TilROUGH THE CHEMICAL CLEANING LINE, CONDENSATE FLOW READS 10,006 GPM, BUT NO LEVEL INCREASE WAS OBSERVED. Evaluators B. Panfil, C. Persson, G. Minshall, H. DeBoer O

CERTIFICATIo!4 oF PERRY SIMULATIO!4 FACILITY DOCKET 110. 50-440 O Simulator Tests - Steady Stato and flormal operations TAB c - Part 3C Testo Abstracts PAGE 1 of 6 ATP TEST SECTIOli T. 4. 4.1 STEADY STATE PERFORMA11CE overview of testina nerformed in this section This series of tests verifles the critical paramotors for the Perry Unit 1 Simulator to be in compliance with the accuracy requirements of Section 4 of the Standard for the steady state heat balancos. This section also includes the tests for steady stato stability verification. I' No. of Tests included in this Section: Engs: open Son's: 1.4.4.1.1, 25% POWER HEAT BALAPCE 2 (W/C)* T.4.4.1.2, 50% K5):R HfAT BALANCE 3 (N/C)* T.4.4.1.3, 751 POWER MEAT BALANCE 4 (N/C)* i T 4.4.1.4, 100% POWER NEAT BALANct 5 (m/C)* f.4.4.1.5, 100% POWER STABillTY 1t$1 6 (N/C)* Alls 1/AttS 3.5

Reference:

Appendix B 1.1 Stoady Stato Performanco O This test has not boon completed (li/C) or has been performed but has one or more unresolved major Simulator Discrepancy Reports associated with it. Please soo the individual test abstract for these discrepancias which constituto exceptions to ANSI /AliS 3.5 section 3.1. ? P l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Steady State and Normal Operations TAB C - Part 3C i Tests Abstracts PAGE 2 of 6 Test T.4.4.1.1, 25% POWER HEAT BALANCE Revision Number: Later ANSI /ANS 3.5 Section: Appendix A 3.2 Steady State Performance Date Tested: Not RUN Run Times Later hours Test Description This test verifies the critical parameters for the Perry Unit 1 Simulator to be in compliance with the accuracy requirements of Section 4 of the Standard at a power level of 25%. The simulator to initialized to 25% power, slow speed pumps. Various primary and secondary plant parameters are recorded and compared to design and actual plant values, where available. Baseline Data used For Reference Reference Type Normal Operating Procedures FTI-805 CORE HEAT BALANCE Reference Type Other GE Thermal Kit BALANCE OF PLANT THERHAL MIT (4 HS/R VESSELS) Reference Type: Plant Data DR-176-S GE REACTOR SYSTEM HEAT BALANCE - RATED O Heat Balance SPECIAL LOG SERVICES HODIFY SPECIAL LOG 9 FT1 B-05 HEAT Malfunctions Tested: None i Discrepancies: Unknown Evaluators: Later I O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Steady State and Normal Operations TAB C - Part 30 Tests Abstracts PAGE 3 of 6 Test T.4.4.1.2, 50% POWER HEAT DALANCE Revision Huinbert Later A!!SI/ANS 3.5 Section Appendix A 3.2 Steady State Performance Date Tested: Not RUN Run Tine Later hours Test Description This test verifies the critical parameters for the Perry Unit 1 Simulator to be in compliance with the accuracy regulrements of Section 4 of the Standard at a power level of 50%. The simulator is initialized to 50t power, f ast speed pumps. Various i primary and secondary plant parameters are recorded and compared to design and actual plant values, where available. Baseline Data used For Reference !.eference Type Normal Operating Procedures FTI-BOS CORE HEAT BALANCE I Re.forence Type Other CK Thermal Kit BALANCE OF PLANT THERMAL KIT (4 MF/R VESSELS) Reference Type Plant Data r~g DR-176-S GE REACTOR SYSTEM HEAT BALANCE - RATED Heat Balance SPECIAL LOG SERVICES MODIFY SPECIAL LOG 9 FT1 D-05 HEAT j Malfunctions Tested None Discrepancies: Unknown Evaluators: Later t O

l CERTIFICATION OF ??RRY SIHULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Steady State and Normal Operations TAB C - Part 3C Tests Abstracts PAGE 4 of 6 Test T.4.4.1.3, 75% POWER HEAT BALANCE 2 Revision Number Later ANSI /ANS 3.5 Section: Appendix A 3.2 Steady State Performance i Date Tested Not RUN Run Times Later hours Test Description This test verifies the critical parameters for the Perry Unit 1 Simulator to be in compliance with the accuracy requirements of Section 4 of the Standard at a power level of 75%. The simulator is initialized to 75% power, fast speed pumps. Various primary and secondary plant parameters are recorded and compared to design and actual plant values, where available. I Baseline Data used For Reference j Reference Type Normal Operating Proceduren FTI-DOS CORE HEAT DALANCE i Reference Type Other GE Thermal Kit BALANCE OF PLANT THERMAL KIT (4 MS/R VESSELS) Reference Type Plant Data DR-176-S GE REACTOR SYSTEM HEAT BALANCE - RATED Heat Balance SPECIAL LOG SERVICES MODIFY SPECIAL LOG 9 FT1 B-05 HEAT I Halfunctions Tested: None Discrepancies: Unknown Evaluators: Later i I t O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Teets - Steady State and Normal Operations TAB C - Part 3C Tests Abstracts PAGE 5 of 6 l Test T.4.4.1.4, 200% POWER HEAT BALANCE i Revision Number: Later ANSI /ANS 3.5 Section: Appendix A 3.2 Steady State Performance Date Testeds Not RUN [ Run Times Later hours Test Description This test verifies the critical parameters for the Perry Unit 1 Simulator to be in compliance with the te uracy I requirements of Section 4 of the Standard at a power level of 100%. The simulator is initialized to 100% power, fast speed pumps. Various primary and secondary plant parameters are recorded and compared to design and actual plant values, where available. Baseline Data used For Reference Reference Type Normal Operating Procedures FTI-B05 CORE HEAT BALANCE Reference Type Other GE Thermal Kit BALANCE OF PLANT THERMAL KIT (4 MS/R VESSELS) Reference Type Plant Data O DR-176-S GE REACTOR SYSTEM HEAT DALANCE - RATED i Heat Balance SPECIAL LOG SERVICES MODIFY SPECIAL LOG 9 PT1 B-05 HEAT Malfunctions Tested: None Discrepancies: Unknown Evaluators Later s 4 9

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 Simulator Tetts - Steady State and Normal Operatior.s TAB C - Part 3C Tests Abstracts PAGE 6 of 6 Test T.4.4.1.5, 100% POWER STABILITY TEST Revision Number: Later ANSI /ANS 3.5 Section: Appendix A 3.2 Steady State Performance Date Tested: Not RUN Run Times Later hours Test

Description:

Thig test verifles the etability of the simulation at 1004 power level. The simulator le initialized to 1004 stoady state power Ic. The simulator is run for a minimum of one hour at a constant power level with no operator or instructor actions per ormed. The J value of critical parameters and various other parameture are recorded throughout the evolution and vorified to meet the critoria of the Standard for etability. Daeoline Data used For Reference Reference Type Other Simulator Specification Malfunctions Tested: None Discrepancies: Unknown Evaluators Later

i CERTIFICATION oF PERRY SIMULATION FACILITY DOCKET Ho. 50-440 O Simulator Tests - Transient Performance Test Abstracts TAB c - Part 4 PAGE 1 of 11 ATP TEST SECTION T.4.4.2 TRANSIENT PERFORMANCE l Qyprview of testinct. Derformed in this section I This series of testa verifies the Perry Unit i Simulator to bo in compliance with the transient critoria of Section 4 of the Standard for the 10 benchmark transients described in Appendix B of the Standard. No. of Tests included in this Section: Ea92: Qpen SDR's: l 1.4.4.2.01, MANUAL SCRAM 2 None f.4.4.2.02, ElWJL1ANEOUS 1 RIP OF ALL FEEDWATER PUMPS 3 1* 1.4.4.2.03, $1MJLTAhE00$ CLOSURE OF ALL MslV's 4 None f.4.4.2.04, SINJLT AhE0JS TRIP OF ALL RECIRC PUMPS 5 kone T.4.4.2.05, $1NGLE RECIRC PUMP TRIP 6 None f.4.4.2.06, MAIN TURBlhE TRIP W/0 RX SCRAM 7 None f.4.4.2.07, MAXlRJM R ATE POWER R AMP (100% + ?$1 100%) 8 1' T.4.4.2.08, MAXIMUM $1ZE LOCA W/ LOSS OF F$ lit ?OWER 9 (4/C)* 1.4.4.2.09, MAXlMJM Si!E UNI $0LABLE MSL RUPTURE 10 (4/C)* O. f.4.4.2.10, SIMULTANEOUS CLOSURE OF MSIV'$ V/ SINGLE STUCK OrEN 11 2 SAFEff/ RELIEF VALVE ANSI /ANS 3.5 Rotorence: Appondix B 1.2 Transient Porformanco This test has not boon completed (N/C) or has been performed but has one or more unresolved major Simulator Discrepancy Reports associated with it. Please see the individual test abstract for these discrepancios which constituto exceptions to ANSI /ANS 3.5 section 3.1. O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Transient Performance Test Abstracts TAB C - Part 4 PAGE 2 of Il Test T.4.4.2.01, MANUAL SCRAM Revision Number: 00 ANSI /ANS 3.5 Section: Appendix A 3.3 Transient Tests Date Tested: 06/15/91 Run Times 1.5 houre Test

Description:

The purpose of this test is to verify the proper response of the simulator to a manual Reactor Scram from Full Power. The simulator is initialized to 100% Power IC. A manual scram is performed. Critical parameters are recorded and analyzed for comparison to plant data. At the end of the test, the reactor is stabilized in a post-scram condition. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures ONI-C71-1 REACTOR SCRAM (UNIT 1)* Reference Type Other Best Estimate Reference Type Plant Data O SER-1-88-5 SCRJ.H E'.'ALUATION REPORT SER-1-88-7 SCP.AH EVALUATION REPORT Malfunctions Tested: None Major Discrepancies: None Evaluators: B. Panfil, C. Persson, G. Hinshall, N. DeBoer [ t 1 I O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Transient Performance Test Abstracts TAB C - Part 4 PAGE 3 of 11 1 Test T.4.4.2.02, SIMULTANEOUS TRIP OF ALL FEEDWATER PUMPS Revision Number: 00 ANSI /ANS 3.5 Section: Appendix A 3.3 Transient Tests Date Tested: 06/15/91 Run Times 1 hour Test

Description:

The purpose of this test is to verify the proper response of the simulator to a simultaneous trip of all normal feedwater pumps. The simulator is initialized to 100% steady state power IC. The Motor Feed Pump is placed in 'OFF' and both running RFPT's are tripped. Critical parameter are recorded per the ANS list, and compared to plant data. At the end of the test, the plant is stabilized in the post-scram condition with RPV level being maintained by HPCS and RCIC systems. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures ONI-N27 FEEDWATER PUMP TRIP (UNIT 1) i I Reference Type: Other i Best Estimate Reference Type Plant Data SER-1-88-2 SCRAM EVALUATION REPORT 5 SER-1-88-5 SCRAM EVALUATION REPORT i Reference Type Plant Data - Analyses USAR 15.2.7 PNPP UPDATED SAFETY ANALYSIS REPORT Malfunctions Tested: None i Major Discrepancies: F-0239 DURING LOSS OF FW ACCIDENT, FOLLOWING HPCS/RCIC L8 SHUTOFF, APPROX. 100" OF INDICATED LEVEL IS " LOST" IN "1 MIN. LEVEL SHOULD INCREASE Evaluators: D.

Panfil, C. Persson, G.

Minshall, H. DeDoor i

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Transient Performance Test Abstracts TAB C - Part 4 PAGE 4 of 11 Test T. 4. 4. 2. 03, SIMULTM'EOUS CLOSURE OF ALL MSIV' S Revision Number: 00 ANSI /ANS 3.5 Section Appendix A 3.3 Transient Testa Date Tested: 06/15/91 Run Times I hour Test

Description:

The purpose of this test is to verif y the proper response of the simulator to a simultaneous closure of all MSIV's. The simulator is initialized to 1004 power IC. The MSIV's are simultaneously closed. Critical parameters are recorded and compared to expected results f or validation. Baseline Data used For Reference Reference Typo Plant Data - Analyses USAR 15.2.4 PNPP UPDATED SAFETY ANALYSIS REPORT Reference Type: Plant Data - Startup Test Results STI-D21-025D VESSEL ISOLATION Malfunctions Tested: None O Major Discrepancies: Ncne Evaluators: B.

Panfil, C.

Persson, G.

Minshall, H.

DeBoer O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Transient Performance Test Abstracts TAB C - Part 4 PAGE 5 of 11 Test T.4.4.2.04, SIMULTANEOUS TRIP OF ALL RECIRC PUMPS Revision Nunber: 00 ANSI /ANS 3.5 Section: Appendix A 3.3 Transient Teste Date Tested 06/16/91 [ Run Times 1 hour t Test Descriptions The purpose of this test is to verify simulator response to a simultaneous trip of both Reactor Recirculation Pumps. The simulator is initialized to 100% steady state power IC with recirc pumps in fast speed. Both pumps are tripped simultaneously while critical parameters are recorded. The plant stabilizes in a post-Scram condition with natural recirculation core flow. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures ONI-833-2 LOSS OF ONE OR BOTH RECIRCULATION PUMPS (UNIT 1) Reference Type: Plant Data - Analyses USAR 1$.3.1 PNPP UPDATED SAFETY ANALYSIS REPORT Halfunctions Tested: None ) Major Diacrepancies: None Evaluators: B. Panfil, C. Persson, G. Hinshall, H. DeBoer i t

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Transient Performance Test Abstracts TAB C - Part 4 PAGE 6 of 11 Test T.4.4.2.05, SINGLE RECIRC PUMP TRIP Revision Number: 00 ANSI /ANS 3.5 Section Appendix A 3.3 Transient Testa Date Tested: 06/*.6/91 Run Times 1.5 hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to a single Reactor Recirculation Pump trip. The simulator is initialized to 100% Steady Stato power IC. The selected pump is tripped while critical parameters are recorded. Parameter response is compared to plant and USAR data for validation of response. Baseline Data used For Reference Reference type Off-Norma' operating Procedures ONI-B33-1 REACTOR RECIRv-. TION FLOW CONTROL HALFUNCTION (UNIT 1) Reference Type Plant Data - Analyses USAR 15.3.1 PHPP UPDATED SAFETY ANALYSIS REPORT Reference Type Plant Data - Startup Test Results STI-B33-030A/2 DATA REQUEST RESPONSE " ONE RECIRC PUMP TRIP & RESTART I Malfunctions Tested: None Major Discrepancies: None Evaluators: B. Panfil, C. Persban, G. Minshall, H. DeBoer E l 6 r l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Transient Performance Test Abstracts TAB C - Part 4 PAGE 7 of 11 l Test T.4.4.2.06, MAIN TURBINE TRIP W/O RX SCRAM Revision Number: 00 ANSI /ANS 3.5 Section: Appen6ix A 3.3 Transient Tests Date Tested 06/16/91 Run Times 0.5 hour Test Description The purpose of this test is to verify the proper response of the Simulator to a main turbine trip initiated from a power level which does not result in a Reactor Scram. Initial conditions: The simulator is initialized to 30% steady state power. 1 The main turbine is tripped. Tucbine bypass valves are verified to open to maintain reactor pressure constant. Critical parameters are recorded and compared to analysis and best estimate data. At the completion of the test, the reacter is operating at about 304 power with bypass valves open. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures ONI-N32 TURBINE AND/OR GENERATOR TRIP (UNIT 1) Reference type Plant Data - Analyses USAR 15.2.3 PNPP UPDATED SAFETY ANALYSIS REPORT Reference T/pe Plant Data - Startup Test Results STI-B21-027/2 DATA REQUEST RESPONSE "URD TRIP & GEN LD REJ Malfunctions Tested: None Major Discrepancies: Nona Evaluators: B. Panfil, C.

Persson, G. Hinshall, H. DeBoer

( l . -.. ~

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simula' or Tests - Transient Performance Test Abstracts TAB C - Part 4 PAGE 8 of Il Test T.4.4.2.07, MAXIMUM RATE POWER RAMP (100% - 75% - 100%) Revision Nc 5er: 00 ANSI /ANS 3.5 Section: Appendix A 3.3 Transient Tests Date Tor.te.1: 06/15/91 Run Times 0.7 hour rest

Description:

The purpose of this test is to verify simulator response to power ramps initiated by changes in Reactor Recirc flow. The simulator is initialized to 100% Steady State power IC. Using the normal mode of recire flow control, reactor power is reduced at the maximum rate to 75% power, then increased at the maximum rate to 100% power. Critical carameters are recorded and analyzed to validate the response. Plant patameters at the end of the test are the samo as initial, a xenon transient is expected. 'dasollne Data used For Reierence Reference Type: Other Best Estimate Reference Type Plant Data - Startup Test Results STI-B33-029B/3 DATA REQUEST RESPO;iSE - LECIRCULATION FLOW CONTROL f( STI-B33-030C/4 DATA REQUEST RESPONSE - RECIRCULATION SYSTEM Malfunctions Tested: None Major Discrepancies: F-0236 IF PWR IS REDUCED TO 75% A MOMLNTARY STM SEAL EVAP DNTK LVL HI SIGNAL INAPPROPRIATELY IGOLATES EXTRACTION STEAM TO SSE CAUSING A DECREASE IN SUPPLY TO STEAM SEAL HEADERS AND HDR PRESS IS LOST. Evaluators: B. Panfii >' 7ersson, G.

Minshall, H.

DeBoer l l l I I l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 Simulator Testa - Trannient Performance Test Abstracts TAB C - Part 4 PAGE 9 of 11 Test T.4.4.2.08, HAXIMUM SIZE LOCA W/ LOSS OF OFFSITE POWER Revision Number Later ANSI /ANS 3.5 Section: Appendix A 3.3 Transient Tests Date Tested: Not completed Run Times Later hours Test

Description:

The purpose of this test is to verify proper response of the Simulator to a DBA LOCA concurrent with a loss of Off-Site Power. Initial Conditiens: The simulator in initialized to 100% power steady state IC. A simultaneous Recirc Loop piping rupture and Loss of oft-site power is initiated while recording critical parameters. The transfent is allowed to continue until a somewhat stabilized condition is ati,ained. At the end of the test, the reactor is shutdown with cor. trol rods inserted, the reactor vessel is depressurized, Low Pressure ECCS is injecting to provide adequate core cooling. Containment parameters are out of their normal ranges, but are verified to not have exceeded any design limitations. U n; Data used For Reference Def rence Type: Off-Normal Operating Procedures 'I-Sil LOSS OF OFF-SITE POWER i Rcf _ence Type: Other dent Estimate Reference Type Plant Data - Analyses USAR 15.6.5 PNPP UPDATED SAFETf ANALYSIS REPORT Malfunctions Tested: TH01 RECIRC LOOP RUPTURE (DBA LOCA) Discrepancies: Unknown Evaluators: Later t O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 I\\ Simulator Tests - Transient Performance Test Abstracts TAB C - Part 4 \\s / PAGE 10 of Il Test T,4.4.2.09, MAXIMUM SIZE UNISOLABLE MSL RUPTURE Revision Number: Later ANSI /ANS 3.5 Section: Appendix A 3.3 Transient Tests Date Tested: Not completed Run Times Later hours Test

Description:

The purpose of thin test is to verify proper simulator response to a design basis Steam Line Rupture Accident in the Drywell. Initial Conditions: The simulator la initialized to 100% Steady State Power IC. Malfunction TH26 is activated. Critical parameters are recorded and compared to analysis and best estimate data. At the end of the test, the reactor is shutdown and depressurized, low pressure ECCS has flooded the RPV to the level of the break to provide adequate core cooling; Containment parameters are out of normal ranges, but are verified to not have exceeded design limits. Baseline Data used For Reference Reference Type: Other Best Estimate O-Reference Type: Plant Data - Analyses USAR 15.6.5 PNPP UPDATED SAFETY ANALYSIS REPORT Malfunctions Tested: TH26 MAIN STEAM LINE RUPTURE INSIDE DRYWELL Discrepancies: Unknown Evaluators Later l l i O

1 CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Transient Performance Test Abstracts TAB C - Part 4 PAGE 11 of 11 Test T.4.4.2.10, SIMULTANEOUS CLOSURE OF MSIV'S W/ SINGLE STUCK OPEN SAF Revision Nuraber: 00 t ANSI /ANS 3.5 Section: Appendix A 3.3 Transient Tests Date Tested: 06/16/91 Run Times 1.5 hours Test

Description:

The purpose of this test is to verify proper Simulator response to a simultaneous closure of the MSIV's with a stuck open SRV and no high pressure injection available. Initial conditions: the simulator is initialized to 100t power IC. Prior to activating the transient, the HPCS and RCIC systems are disabled, and the Motor Teod Pump is placed in secured status. One SRV is overridden open. The MSIV's are closed and critical parameters are recorded until a somewhat stabilized condition is obtained. At the completion of the test, the reactor is shutdown and depressurized due to ADS initiation. Low pressure ECCS pumps are injecting to the RPV to maintain adequate core cooling. Containment parameters are out of normal ranges, but are verified to not exceed design limits. Baseline Data used For Reference Reference Type: Other Best Estimate Reference Type: Plant Data - Startup Test Results STI-B21-025B VESSEL ISOLATION STI-B21-026/1 DATA REQUEST RESPONSE - SAFETY RELTEF VALVES Malfunctions Tested: None Major Discrepancies: F-0229 WITH LOCA SIGNAL PRESENT SUPP POOL MAKEUP VLVS AUTO OPEN UPON SENSING LO SP LVL ALARM SETPT(18-18.l'). SHOULD OPEN AT LEVEL <16.75' W/LOCA. F-0230 UPON AUTO NITIATION OF LPCS DUE TO LOW RPV LEVEL, THE LPCS PUMP MOTOR STARas. AND THEN IMMEDIATELY TRIPPED OFF DUE TO OVERCURRENT. Evaluators: B. Panfil, C.

Persson, G.

Minshall, H. DeBoer O

CERTIFICATION oF PERRY SIMULATION FACILITY DOCKET No. 50-440 Simulator Tests - Malfunction Test Abstractn TAB C - Part 5A PAGE 1 of 34 ATP TEST SECTION T.4.5.1 SYSTEM LEVEL FAILURES Overview of testdnu nerformed in this section This series of Tests was performed to verify the proper response of the Simulator to the set of System Level Failures provided in the design of the Simulator. The test methodology was to first verify proper "first principles" modc. ling of the simulated failure on the system level, then to verify the proper response of any interfacing systems in the integrated environment. Finally, each failure was to be verified to provide the proper overall plant response as predicted for the particular type of failure. Test procedures were written to check each of these items. These tests were written by Utility (CEI) employees holding a current SRO license on the Perry Unit 1 Plant and have operational and training experience. (These personnel were on loan to the Simulator Vendor from July 1989, and personally observed or performed all Operational Testing of the Simulator.) The test abstracts C) include a description of which generic cases were tested for 'v each generic malfunction, and the severities which were tested for those having adjustable rates. Please refer to the Failure Cause and Effects Manual-(available at Perry) for a complete description of the failure modes, the generic cases available, and the range of severities applicable to each System Level Failure. No. of Tests included in this Section: Page: Onen SDR's: T.4.5.1.02.AN01, ANNUNCIATOR INPUT OPTICAL ISOLATOR FAILURE 3 2* T.4.5.1.04.CUO3, RWCU SYSTEM PIPE BREAK OUTSIDE CONTAINMENT (STEAM TUNNEL) 4 2* T.4.5.1.06.ED06, LOSS OF 480V BUS 5 None T.4.5.1.06.E009, LOSS OF 125 VDC BUS 6 6* 7.4.5.1.06.ED17, LOSS OF 125V DC DISTRIBUTION PANEL 8 (N/C)* T.4,5.1.07.EG01, MAIN CENERATOR LOCKOUT RELAY TRIP 9 None T.4.5.1.09.FW02, FEEDWATER SYSTEM PIPE BREAK INSIDE DRYWELL 10 4* T.4.5.1.09.FWO3, FEEDWATER SYSTEM PIPE BREAK OUTSIDE CONTAlliMENT 11 4* T.4.5.1.12.lA01, AIR RECEIVER LEAK 12 9' T.4.5.1.12.lA02, INSTRUMENT AIR LINE LEAK 14 6* T.4.5.1.15.MC01, CONDENSER AIR INLEAKAGE 15 (N/C)* T.4.5.1.16.MS11, STEAM SEAL HEADER PRES $URE REGULATOR FAILURE 16 None T.4.5.1.17.NM01, SRM DETECTOR (PRE-AMP) FAILURE 17 1* 7.4.5.i.17.kM02, IRP8 DETECTOR (PRE AMP) FAILURE 18 None T.4.5.i.17.NM03, LPRM DETECTOR FAILURE 19 None T.4.5.1.17 NM10, NEUTRON MONITORING DETECTOR DRIVE STUCK 20 None T.4.5.1.18.0G03, OFF GAS SYSTEM LEAK UPSTREAM OF ADSOR8ER8 21 3* T.4.5,1,18.0G04, OFF GAS SYSTEM LEAK DOWNSTREAM OF ADSORRERS 22 None i v l l

i i I CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 0 Simulator Teste - Halfunction Test Abstracte TAB C - Part SA PAGE 2 of 34 i No. of Tests included in this Section: Paqo: 9 pen SDR's: 1 T.4.5.1.21.RD03, CONTROL ROD DRIFT IN 23 (N/C)* T.4.5.1.21.RD04, CONTROL ROD ORIFT 007 24 (N/C)* 1.4.5.1.22.RH02, RISIDUAL HEAT REMOVAL SYSTEM PIPE BREAK 25 (N/C)* f.4.5.1.23.RP02, INADVERTENT INITI AT10W OF ALTERNATE ROD INJECil0N 26 None T.4.5.1.25.SW01, NUCLEAR CLOSED COOLING SYSTEM Pt0 CESS PIPING LEAKAGE 27 1* T.4.5.1.25.SV02, SERvlCE WATER SYSTEM PROCESS P!VING LEAKAGE 28 2* T.4.5.1.27.TH01, RECIRC LOOP RUPTURE (DBA LOCA) 29 (N/C)* 1.4.5.1.27.f M19, RPV LEVEL INST REFERENCE LEG BRE AK 30 1* 1.4.5.1.27.TH20, RPV LEVEL INST VARI ABLE LEG BRE AK 31 (N/C)* T.4.5.1.27.TH21, POWER / FLOW INSTABILillES (IEB 88 07 SUPPLEMENT 1) 32 1* 1.4.5.1.27.TH27, MAIN STEAM LINE RUPTURE IN STEAM TUNNEL 33 2* 1.4.5.1.28.TUO1, hAIN SHAFT Oil PUMP DEGRADAfl0N 34 None ANSI /hNS 3.5

Reference:

3.1.2 Plant Malfunctions i i This test has not been completed (N/C) or has been performed but has one or more unresolved major Simulator Discrepancy Reports associated with it. Please see the individual test abstract for these discrepancies which contitute exceptions to ANSI /ANS 3.5 t y section 3.1.2. l O 1 i

_ _ ~ _. _ _ CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 I~' Simulator Tests - Malfunction Test Abstracts TAB C - Part SA \\ PAGE 3 of 34 Test T 4.5.1.02.AN01, ANNUNCIATOR INPUT OPTICAL ISOLATOR FAILURE Resision Number: 02 ANSI /ANS 3.5 Section: 3.1.2 (22)(b) Plant Malfunctions; Process Alarm System Failure Date Tested: 05/06/91 Run Times 1.00 hour Test

Description:

The purpose of this test is to verify the propor response of the simulator to annunciator failures. This failure simulates a loss of instrument power to annunciator system optical isolator units on the input (system) side. The failure has 39 generic cases representing a variety of isolators in numerous safety related systems (isolators are not required in BOP systems). The test is initiated from 100% power operating condition. During the test, the failure is randomly selected by the test personnel. Proper response is checked by driving the affected system parameters / conditions to the alarm setpoint/ condition, and observing that the affected annunciators are not received. Annunciation of alarms identifying the power loss are verified to annunciate. The failure is deletud, simulating repairs to the. isolator have been completed, and the receipt of appropriate alarms is verified. At the completion of the test the plant remains at 100% power condition. Baseline Lata used For Reference 1 Reference Type Off-Normal Operating Procedures OM6: ARI PERRY; UNIT ALARM RESPONSE INSTRUCTIONS Reference Type: Other Best Estimate Malfunctions Tested: AN01 Annunciator Input Optical Isolator Failure l-Major Discrepancies: P-1350 MALFUNCTION AN01 CASE Z HAD NO EFFECT ON THE ANNUNCIATORS DRIVEN BY l THE OPTICAL ISOLATOR. (1E22AT4). l l P-1351 CERTAIN ALARMS WERE NOT RECEIVED DURING THE RUNNING OF AN01 MALFUNCTION, CASES T, V, W, AB, AC, AE, AG. Evaluators: R. Libra, C.

Persson, B.

Stetson l 1 l O i .~.

} CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part SA kyY PAGE 4 of 34 Test T.4.5.1.04 CUO3, RWCU SYSTEM PIPE BREAK OUTSIDE CONTAINMENT (STM TUNNEL) Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.2 (01)(b) Plant Malfunctions; LOCA outside containment i Date Tested: 06/19/91 Run Time: 1.00 hour Test

Description:

The purpose of this test is to verify proper simulator response to a reactor water cleanup (RWCU) pipe break outside the containment (steam tunnel). The test is initiated from 100% power operating condition. Alarms, isolation valves and positions, recorders, timers, and temperature recorder and alarms are verified to operate properly. At the end of the test, the plant is shutdown with the MSIV's and Main Steam Line Drain Valves closed. Baseline Data used For Reference Reference Type: Off-Normal Operating Procedures ONI-Nil HIGH ENERGY PIPE BREAK OUTSIDE CONTAINMENT (UNIT 1) Reference Type Other Best Estimate Malfunctions Tested: CUO3 RWCU SYSTEM PIPE BREAK OUTSIDE CONTAINMENT (STEAM TUNNEL) Major Discrepancies: F-0255 WITH CUO3 ACTIVE AND DELTA FLOW AT GREATER THAN 100 GPM, THE ISOLATION TIMER TIMES OUT. WHEN SYSTEM SHOULD ISOLATE, DELTA FLOW INDICATION SPIKES DOWN TO O GPM AND THEN RETURNS TO 100 GPM F-0257 " STEAM TUNNEL LD AMB TEMP P632" ANNUNCIATOR WAS NOT RECEIVED UNTIL AMB TEMP WAS 155 DEGF AS' READ ON RECORDER 1E31-R608 THIS ALARM SHOULD BE RECEIVED AT AN AMBIENT TEMPERATURE OF 135 DEGF IN STEAM TUNNEL. Evaluators: C.

Persson, B.

Panfil, H. DeBoer, G. Minnhall O I

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SA PAGE 5 of 34 Test T.4.5.1.06.ED06, LOSS OF 480V BUS Revision Number: 03 ANSI /ANS 3.5 Section 3.1.2 (03)(d) Plant Malfunctions; Loss of Distribution Bus Date Tested: 06/19/91 Run Times 2.00 hours Test

Description:

The purpose of this test is to verify proper response of almulator to the following 480v bus failures EF1A, EFIB, EFIC, EFID, FIA, FIB, FIC, FID, F1E, FIF, FIG, XFIA. Initial conditions are as follows: 100% power, steady state. Failure is inserted to give respective bus power loss and alarms, recorders, indicating lights, i'eolations are verified to occur as associated plant procedure ONI-R23 1 os 2 states. Failure mode is such that the bus cannot be recovered by operator actions until trainer allows.. Failures are verified to be inserted and deleted properly. At completion of tests, the plant will be at condition determined by severity of loads lost by respective bus as stated in ONI-R23 1 or 2. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures ONI-R23-1 LOSS OF AN ESSENTIAL 480V BUS (UNIT 1) /y ONI-R23-2 LOSS OF A NON-ESSENTIAL 480 VOLT BUS (UNIT 1) Malfunctions Tested: ED06 LOSS OF 480V BUS Discropancies: None Evaluators: C.

Persson, B.

Panfil, H. DeBoer, O.

Minshall

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SA PAGE 6 of 34 Test T. 4. 5.1.06.ED09, LOSS OF 12 5V DC BUS Revision Number: 02 ANSI /ANF 3.5 Section: 3.1.2 (03)(e) Plant Halfunctions; Loss of DC Instrument Bus Date Tested: 06/19/91 Run Time 1.00 hour Test

Description:

The purpose of this test is to verify proper response of simulator to the following 125V DC bus failures EDIA, ED1B, ED1C, DIA, and O1B. Initial conditions are 100% power, steady state. Failure is inserted to give respective bus power loss. For EDIA, ED1B, and EDIC, D1B alarms, recorders, indicating lights and for EDIB - RX SCRAM on Main Turbine trip from Reactor Core Isolation Cooling initiation signal are verified to occur as per ONI-R42 1,2, 3 or 5 respectively. For D1A annunciation loss except for " ANN PWR SUPPLY FAIL", recorders, indicating lights, Lil and L12 bus transfers to L10, Rx Recirc pumps tripped to of f, are verified to occur as per ONI-R42-4. Failured are verified to be inserted and deleted properly. At completion of tests, the plant will be at 100% power with power restored to the buses, except for ED1B which will be recovering from a scram with power restored to ED1B Lus; and for DIA, Reactor power will be lower due to loss of RX Recirc pumps, with power restored to D1A bus. O Baseline Data used For Reference Reference Type: Of f-Normal Operating Procedures ONI-R42-1 LOSS OF DC BUS ED-1-A (UNIT 1) ONI-R42-2 LOSS OF DC BUS ED-1-B (UNIT 1) ONI-R42-3 LOSS OF DC BUS ED-1-C (UNIT 1) ONI-R42-4 LOSS OF DC BUS D-1-A (UNIT 1) ONI-R42-5 LOSS OF DC BUS D-1-B (UNIT 1) Halfunctions Tested: ED09 LOSS OF 12SV DC BUS Major Discrepancies: F-0241' ON LOSS OF ED1B, DID NOT LOSE INDICATION OR CONTROL OF 1P45-C001B (EH1205). WE LATER FAILED EDIA AND EH1205 INDICATION WAS LOST - l BELIEVE WRONG POWER SOURCE IS SIMULATED. F-0245 ON LOSS OF ED1B,.DID NOT LOSE CONTROL OR INDICATION OF (EF1D04). WE LATER FAILED EDIA AND EFID04 INDICATION WAS LOST - BELIEVE WRONG POWER SOURCE IS SIMULATED. F-0246 ON LOSS OF EDlB, DID NOT LOSE INDICATION OR CONTROL OF 1E12-C002B (EH1208). WE LATER FAILED EDIA06 AND EH1208 INDICATION WAS LOST - BELIEVE WRONG POWER SOURCE IS SIMULATED. F-0247 ON LOSS OF EH11 - LOST INDICATION OF P45-C001B SINCE P45-C001B SHOULD BE POWERED FROM EH12, LOSS OF EH11 SHOULD MAVE NO EFFECT ON P45-C001B F-0248 RCIC SYSTEM LOSES DIV 2 LOGIC POWER ON A LOSS OF EDIA (INDICATED ON OOS MATRIX AMBER LIGHTS) RCIC DIV 2 LOGIC / POWER LOSS SHOULD BE () DRIVEN FROM ED1B, NOT EDIA.

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 'O Simulator Tests - Malfunction Test Abstracts TAB C - Part SA PAGE 7 of 34 Test T.4.5.1.06.ED09, LOSS OF 125V DC BUS Major Discrepancies (continued): F-0265 DID NOT RECEIVE POWER LOSS OOS INDICATORS FOR ESW B/ECC D WHEN LOSS OF ED1B WAS RUN. ED1B PROVIDES CONTROL POWER FOR BREAKERS, SHOULD RECEIVE OUT OF SERVICE LIGHTS AND ALARM. Evaluators: C.

Persson, B. Panfil, H. DeBoer, G. Minshall

' O h IL.

\\ \\ CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O-Simulator Tests - Malfunction Test Abstracts TAB C - Part SA PAGE 8 of 34 %s Test T.4.5.1.06.ED17, LOSS OF 125V DC DISTRIBUTION PANEL Revision Number: later ANSI /ANS 3.5 Section: 3.1.2 (03)(d) Plant Malfunctions; Loss of Distribution Bus Date Tested: Not Complete Run Times Later hours . Test

Description:

The purpose of this test is to verify the proper response of the Simulator to a loss of 125V DC distribution panels. ED17 has seven cases (panels) in order to fail both safety-related and non-safety 125V DC distribution. The simulator is initialized to 100% power IC and individual cases of ED17 are inserted. Plant load lists are used to evaluate simulator performance and proper electrical distribution modeling. This includes checking for proper receipt of alarms, automatic actions, and loss of power to control circuits, components, and instrumentation. Each malfunction case is deleted to ensure proper restoration is achieved. Baseline Data used For Reference Reference Type: Of f-Normal Operating Procedures ONI-R42-1 LOSS OF DC BUS ED-1-A (UNIT 1) ONI-R42-2 LOSS OF DC BUS ED-1-B (UNIT 1) i , 'O ONI-R42-4 LOSS OF DC BUS D-1-A (UNIT 1) ONI-R42-5 LOSS OF DC BUS D-1-B (UNIT 1) Malfunctions Tested: ED17 LOSS OF 125V DC DISTRIBUTION PANEL Major Discrepancies: Unknown Evaluators Later l O l l i .m., v

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Tout Abstracts TAB C - Part 5A PAGE 9 of 34 Test T.4.5.1.07.E001, MAIN GENERATOR LOCKOUT RELAY TRIP Revision Number: 01 ANSI /ANS 3.5 Section: 3.1.2 (16) Plant Malfunctions; Generator Trip Date Tested: 06/20/91 Run Times 0.50 hour Test

Description:

The purpose of this test is to verify the proper response of the simulator to a spurious trip of the Main Generator Lockout Relay, 86-1 and 86-2. The simulator is initialized to an IC at which it in operating at 100% steady stato power. Malfunction EG01A is activated and operator action in accordance with ONI-N32 (Turbine and/or Generator Trip), and ONI-C71-1 (Reactor Scram) will be taken. All automatic actions associated with a generator trip from power are checked against those listed in the aforementioned ONI's. At the conclusion of the test, the plant will be stabilized in a post scram condition. EG01A is then deleted and it is verified that the 86-1 relay can be reset. EG01B is tested to the extent necessary to verify it can be inserted and deleted from the instructor station. Baseline Data used For Reference Reference Types -Off-Normal Operating Procedures ONI-N32 TURBINE AND/OR GENERATOR TRIP (UNIT 1) Reference Type: Other Best Estimate Malfunctions Tested: EG01 KAIN GENERATOR LOCKOUT RELAY TRIP Major Discrepancies: None Evaluators: C. Persson, B. Panfil, H. DeBoer, G. Minshall O - -,,, - - ~. -,, -. -, ~ -, - - - - - - -.. - - -.. - _, - -. -..,... > ~.. ,,r

.~ ~. -- CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-140 [} Simulator Tests - Halfunction Test Abstracts Tt.B C - Part SA ( ,f PAGE 10 of 34 Test T.4.5.1.09.FWO2, FEEDWATER SYSTEM PIPE BREAK INSIDE DRYWELL Revision Number: 01 ANSI /AHS 3.5 Section: 3.1.2 (20)(c) Plant Malfunctions; Feedwater Line Break, inside containment Date Tested: 06/19/91 Run Times 0.50 hour Test

Description:

The purpose of this test is to verify the response of the simulator to a large feedwater line break within the drywell structure. This malfunction is a discrete variable with 100% severity representing a 20 inch double end pipe failure. The simulator is initialized to a 100% IC and the malfunction is inserted at 100% severity. Containment and Reactor Pressure Vessel response will be evaluated against Perry USAR Chapter 15.6.5. Automatic actions initiated in response to a large break LOCA are also verified to perform correctly. At the end of this test, the reactor is shutdown with High Pressure Core Spray and/or Reactor Core Isolation Cooling maintaining RPV level between Level 2 and Level 8. Reactor pressure will be slowly decreasing as a result of spray system injection. Drywell and containment parameters degrading due to the loss of cooling systems. Lastly FWO2 is verified as a non-recoverable type by attempting to delete it. Baseline Data used For Reference Reference Type Normal Operating Procedures SOI-E22A HIGH PRESSURE CORE S" RAY SYSTEM (UNIT 1) l 'JOI-E51 REACTOR CORE ISOLATION COOLING SYSTEM (UNIT 1) Malfunctions Tested: FWO2 FEEDWATER SYSTEM PIPE BREAK INSIDE DRYWELL Major Discrepanciee F-0252 WITM MALF FWO2 INSERTED AND N27-F200 OPEN & ALL FW PUMPS EXCEPT ONE RFBP SHUTDOWN, NOTED FEED FLOW PEGGED HIGH (*20 MLB/HR). F-0253 AFTER FEEDWATER BREAK INTO THE DRYWELL WAS ISOLATED AND NCC RESTORED TO THE DRYWELL AHU'S, THE DRYWELL TEMPERATURES DID NOT DECREASE. F-0270 INDICATED FW FLOW ON IC34R607 PEGGED HIGH (>20 MLBM/HR) REGARDLESS L OF THE AMOUNT OF "REAL" FW FLOW WHEN A FW BREAK IN DW IS ACTIVE. l' F-0285 HPCS PUMP TRIPPED ON OVERCURRENT WHEN HPCS HI DRYWELL PRESSURE AUTO INITIATION SIGNAL WAS RECEIVED. Evaluators: C.

Persson, B. Panfil, H. DeBoer, G. Minshall O

l t

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part 5A PAGE 11 of 34 Test T.4.5.1.09.FWO3, FEE 7 WATER SYSTEM PIPE BREAK OUTSIDE CONTAINMENT Revision Number: 01 ANSI /ANS 3.5 Section: 3.1.2 (20)(c) Plant Malfunctions; Feedwater Line Break, outside containment Date Tested: 06/20/91 Run Times 1.75 hours Test

Description:

The purpose of this test is to verify the response of the simulator to a large feedwater line break outside the contairu_'t Structure within the Turbine Building Steam Tunnel. This malfunction is aiscrete, variable, with 100% severity representing a 20 inch double end pipe failure. The simulator is initialized to a 1004 Power IC and the malfunction is inserted at 100% severity. The sequence of events and key parameters are checked against those in the Perry USAR Chapter 15.5.6. System response and isolations expected to occur on a break in the Stean Tunnel are also verified. The plant stabilizes in a post scram condition with the MSIV's closed. RPV level is maintained between level 2 and level 8 by the operation of High Pressure Core Spray and/or Reactor Core Isolation Cooling. RPV pressure is maintained initially by safety / relief valve operation and slowly decreases threagh the operation at spray systems. Lastly, FWO3 is v3rified as a non-recoverable type by attempting to delete it. Baseline Data used For Peterence Reference Types Normal Operating Procedures SOI-E22A HIGH PRESSURE CORE LPRAY SYSTEM (UNIT 1) SOI-E51 REACTOR CORE ISOLATION COOLING SYSTEM (UNIT 1) SOI-033 REACTOR WATER CLEANUP SYSTEM (UNIT 1) Malfunctions Tested: FWO3 FEEDWATER SYSTEM PIPE BREAK OUTSIDE CONTAINMENT Major Discrapancies: F-0114 WITH THE MSIV'S ISOLATED DUE TO MF FWO3 ACTIVE AT 10% (FW BREAK IN TUNNEL) MS LINE TEMPS OSCILLATED, THEN PEGGED HIGH WITHIN 1 MINUTE. (CAUSE SIM FAILURE ONE TIME, NOT THE NEXT). F-0274 WITH A FW LINE BREAK IN THE STEAM TUNNEL, FW HDR FLOW AS SENSED BY 1C34N002B DID NOT INCREASE. F-0275 WITH A FW PIPE BREAK IN THE STFAM TUNNEL, MALFUNCTION FLOW DROPPED TO O AFTER THE RFPT'S TRIPPED. RFBP'S, MFP WAS IN OPERATION WITH A FLOW PATH STILL AVAILAELE TO THE PIPE BREAK LOCATION. ENERGY INPUT OUT BREAK SHOULD CONTINUE UNTIL THE LEAK FLOW PATH IS ISOLATED. F-0289 VERY LITTLE INCREASE IN STEAM TUNNEL AMBIENT TEMP WAS SEEN AND NO l INCREASE WAS SEEN IN TB AMBIENT TEMP DURING A FW LINE BREAK IN i STEAM TUNNEL. 400 DEG F FW FLASHING TO STEAM OUT THE GREAK SHOULD CAUSE A RAPID AND LARGE INCREASE IN STEAM TUNNEL /TB TEMPS. Evaluators: C.

Persuon, B.

Panfil, H. DeBoer.

G. Minshall . O 1 \\ l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAD C - Part SA \\ PAGE 12 of 34 Test T.4.5.1.12.IA01, AIR RECEIVER LEAK Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.2 (02) Plant Malfunctions; Loss of Instrument Air Dcte Tested: 06/20/91 Run Time 2.00 hours Test

Description:

The purpose of this test is to verify the simelator's response separately to a loss of Service and Instrument Air System receivers. This malfunction containe 4 cases of which two, IA01A and IA01C will be evaluated for plant response. The other two, IA01B and D will be tested sufficiently to ensure they can be inserted and deleted only. The plant is initialized to a 100% IC and the malfunctions inserted at 200s severity which simulates a receiver end weld failure. Automatic actions and component failures are evaluated against those listed in ONI-P51/PS2, loss of service and/or Instrument Air. For IA01A the plant stabilizes still at power following the automatic closure of service air to instrument air system cross-connects. In the case of IA018, the plant stacilizes in a post scram condition with the MSIV's closed and the instrument air system depressurized. Baseline Data used For Reference O Teference Type Off-Normal Operating Procedures ONI-P52 LOSS OF SERVICE AND/OR INSTRUMENT AIR Malfunctions Tested: IA01 AIR RECEIVER LEAK Major Discrepancies: F-0277 ON A LOSS OF AIR, N23 SYSTEM DP DID NOT DECREASE TO APPROXIMATELY 0 PSID. N23 BYPASS VALVES IN23F020 FAILS OPEN ON LOSS OF AIR, CAUSING SYSTEM DP TO DECREASE TOWARD 0 PSID. F-0278 WITH A LOSS OF AIR EJECTORS DUE TO A LOSS OF INSTRUMENT AIR, MAIN CONDENSER VACUUM REMAINED CONSTANT. EXPECT TO SEE SOME AMOUNT OF VACJUM DEGRADATION WITH NO AIR EJECTORS IN OPERATION AND STEAM BEING DUMPED TO THE CONDENSER. F-0279 DW VACUUM BKR MOV'S M1fF010A AND F010B WERE CYCLING IN THE PRESENCE OF AN NS4 ISOLATION SIGNAL. F-0280 WITH 80# AT FRONT STANLARD, THE MAIN TURBINE DID NOT TRIP. F-0283 1C11F002A DID NOT FAIL CLOSED ON LOSS OF AIR. F-0291 WHEN U1 SERVICE AIR RECEIVER RUPTURED, THE IA COMPRESSORS STARTED BUT UNLOADED AT 95#, COMPRESSOR START IN AUTO /ON-OFF MODE AND l SHOULD NOT UNLOAD TILL ABOUT 101.5#. l F-0292 ON A LOSS OF INSTRUMENT AIR, THE AUX BOILER WAS STILu AVAILABLE. l O

l 9A !f ' ?AVI;ON OF PERRY SIMULATION l'ACILITY DOCKET NO. 50-440 O Miraolator \\osts - Malfunction Test Abstracts TAB C - Part SA PAGE 13 of 34 Test T.4.5.1.12.IA01, AIR RECEIVER LEAK Major Discrepancies (continued): F-0294 ON LOSS OF AIP N26 VALVES DID NOT FAIL AS REQUIRED. i P-0252 AFTER MALF IA01 WAS INSERTED, THE FOLLOWING PROBLEMS WERE SEEN: Evaluators: C.

Persoon, B.

Panfil, H. DeBoer, G. Minshall O O

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SA PAGE 14 of 34 Test T.4.5.1.12.IA02, INSTRUMENT AIR LINE LEAK Revision Nuuber: 02 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions, USAR Accident 15.2.10 Date Tested: 06/20/91 Run Times 1.25 hours Test

Description:

The purpose of this response is to verify the proper response of the Simulator to an instrument air system line leak. IA01 has five cases (A-E), to simulate line breaks in 5 separate areas of the plant. The malfunctions are variable (0-1001) where 100% equsle a complete line severance (2 inch pipe). During the test, Cases A-C are fully tested for plant response. Cases D&F are verified to insert and delete from the instructor station. For iach case tested, the simulator is initialized to a 100% power IC. IA02A(B,C is inserted at 100% severity. The proper response of the plant to

e line break is checked using Perry off-normal instructions. Major e.tet is checked include air operated components f ailing to the fail position air ompressors starting at proper setpoints, and consequential ef fect s of loss of air to various systems.

Breaks occurring within the containc.ent are verified to cause a pressure increase in the affected area..Due loss of air to the SCRAM air header, the reactor SCRAMS. At the end of.the test, the plant is shutdown and stabilized in a post SCRAM condition. () Baseline Data used For Refarence Reference Type: Off-Normal Operating Procedures ONI-PS2 LOSS OF SERVICE AND/OR INSTRUMENT AIR Malfunctions Tested: IA02 INSTRUMENT AIR LINE LEAK Major Discrepancies: F-0271 WHEN AIR DISTRIB HDR PRESS IS LOW, AIR OPERATED VLVS GO TO FAIL POSITION BUT ONLY IN PROPORTION TO AIR HDR PRESS. THIS IS OK TO A POINT BUT RESPONSE IS UNREALISTIC IN RANGE OF.001 PSIG TO 1 PSIG. F-0276 WITH A COMPLETE LOSS OF INSTRUMENT AIR TO LOADS OUTSIDE CONTAINMENT, THE CLEARWELL " UMPS DID NOT TRIP. F-0281 ON A COMPLETE LOSS OF INSTRUKENT AIR TO LOADS OUTSIDE THE CONTAINMENT, IN64F051C AND IN64F051D DID NOT FAIL OPEN. F-0286 WITH AN INSTRUMENT AIR LINE BREAK IN THE DRYWELL, DRYWELL PRESSURE DID NOT INCREASE. F-0287 WITH AN AIR BREAK IN CONTAINMENT, CONTAINMENT PRESSURE DID NOT INCREASE. F-0288 WITH A LOSS OF AIR TO THE DRYWELL, HEAT EXCHANGER COILS DID NOT AUTO SWAP TO THE B COILS FOR THE DRYWELL COOLERS. Evaluators: C.

Persson, B. Panfil, H. DeBoer, G.

Minshall I /O 'O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part 5A g PAGE 15 of 34 Test T.4.5.1.15.MC01, CONDENSER AIR INLEAKAGE Revision Numbers later ANSI /ANS 3.5 Section: 3.1.2 (05)(a) Plant Malfunctions; Lose of Condenser Vacuum Date Tested: Not Complete Run Times Later hours Test

Description:

The purpose of this test is to verify the simulator's response to a loss of condenser vacuum. Malfunction MC01 has 3 cases: HC01A - Main Condenser HP Shell, MC018 - Auxiliary condenser A, and MC01B - Auxiliary Condenser B. All three are variable in nature 0-100s with 100s equivalent to 300 scfm at 3 inHgA. MC01A is tested, B and C are verified to be inserted and deleted from the Instructor Station. The simulator is initialized to a 100s steady state IC with two circulating water pumps in service, and MC01A is inserted at los severity. Main and Auxiliary condenser vacuums are then verified to slowly degrade. Changes in plant Offgas system are also verified. The automatic actions outlined in ONI-N62 are verified to occur. The plant response is evaluated against that listed in the Perry USAR Chapter 15.2.5. The plant stabilizes in a post scram condition with MSIV's closed. RPV pressure is being maintained by use of safety / relief valves and RPV level control is on the Motor Feed Pump augmented as necessary by High Pressure Core Spray and/or Reactor Core Isolation Cooling. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures ONI-N62 LOSS OF MAIN CONDENSER VACUUM (UNIT 1) Reference Type Other Best Estimate Malfunctions Tested: MC01 CONDENSER AIR INLEAKAGE Major Discrepancies: Unknown Evaluators Later O

~. _. ~. - CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SA PAGE 16 of 34 Test T.4.5.1.16.MS11, STEAM SEAL HEADER PRESSURE REGULATOR FAILURE Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions, USAR Accident 15.7.1 Date Tested: 06/20/91 Run Time: 0.60 hour Test

Description:

The purpose of this test is to verify the proper response of the simulator to a failure of the Steam Seal Evaporator Header Pressure control valve, IN33-F070. The malfunction is variable 0-100% and both extremes are tested to view the affects. The simulator is initialized to a 100% IC with the SSE in service. The regulating valve is first failed full open and then full close. Indicating lights, alarms and metered instrumentation is then evaluated. Malfunction MS11 is then deleted and proper operation of the N33-F070 valve is observed. The plant remains stable at 100% power at the completion of this test. Baseline Data used For Reference Reference Type Normal Operating Procedures SOI-N33 STEAM SEAL SYSTEM (UNIT 1) Reference Type: Of f-Normal Operating Procedures N ONI-N62 LOSS OF MAIN CONDENSER VACUUM (UNIT 1) (O Malfunctions Tested: MS11 STEAM SEAL HEADcx PRESSURE REGULATOR FAILURE Discrepancies: None Evaluators: C.

Persson, B.

Panfil, H. DeBoer, G. Minshall O l l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part 5A PAGE 17 of 34 Test T.4.5.1.17.NM01, SRM DETECTOR (PRE-AMP) FAILURE Revision Number: 01 ANSI /ANS 3.5 Section 3.1.2 (21) Plant Halfunctions; Nuclear Instrumentation Failures Date Tested: 06/20/91 Run Time: 0.50 hour Test

Description:

The purpose of this test is to verify the proper response of the simulator to a failure of Source Range Monitor Pre-amplifier. This malfunction has four cases: HM01A through D, are specific for each SRM. Each malfunction is variable C 100% output of the pre-amp. NM01A verifies upscale failures while NM01B, C, and D verify downscale failures. The simulator is initialized in a startup IC with power still in the scarce range. NM01A is inserted to 100% severity and the response is verified. SRM A counts increase to offscale, Peactor Period goes towards zero and remains positive until saturation of the pre-amp. Remote Function RP01 is used to remove the RPS shorting links to verify the scram function occurs. The plant stabilizes in a post scram condition and NM01A is deleted. Baseline Data used For Reference Reference Type Normal Operating Procedures SOI-C51(SRM) SOURCE RANGE MONITORING SYSTEM (UNIT 1) Reference Typan Other Best Estimate Malfunctions Tested: NM01 SRM DETECTOR (PRE-AMP) FAILURE Major Discrepancies: F-0272 NM01 (FAIL TO 0%) WANTED TO ACTIVATE & IMMEDIATELY RAMP TO 100%; FAIL ACTIVATION (%) & SEVERITY / RAMP RATE DON'T MAKE ANY SENSE Evaluators: B. Panfil, H. DoDoer, G. Minshall O l. __..__msia_m.

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part SA PAGE I8 of 34 Test T.4.5.1.17.NM02, IRM DETECTOR (PRE-AMP) FAILURE Revision Number: 01 ANSI /ANS 3.5 Section: 3.1.2 (21) Plant Malfunctions; Nuclear Instrumentation Failures Date Tested: 06/21/91 Run Times 1.25 hours Test

Description:

The purpose of this test is to verify proper response of the simulator to individual IRM failure upscale /oownscale. Initial conditions are as follows: approximately 2% power, plant heatup in process. Failure is inserted to give IRM upscale (200% severity). Alarms, recorders, and indicating lights are verified to operate properly with 1/2 scram i received. Failure is inserted tn give IRM downscale (ramp to 0%).

Alarms, recorders and indicating lights are verified to operate properly with rod block received.

Failures are verified to be inserted and deleted properly. At completion of the tests, the plant is at 2%. Baseline Data used For Reference Reference Type Normal Operating Procedures SOI-C51(IFGI) INTERMEDIATE RANGE MONITORING SYSTEM (UNIT 1, I Reference Type: Other ( Best Estimate Malfunctions Tested: NM02 IRM DETECTOR (PRE-AMP) FAILURE Discrepancies: None Evaluators:. B. Panfil, H. DeBoer, G. Minshall l i!O --,,m,- ,m- ... -, - ~..

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 {} Simulator Tests - Malfunction Test Abstracts TAB C - Part SA s,j PAGE 19 of 34 Test T.4.5.1.17.NM03, LPRM DETECTOR FAILURE Revision Number: 01 ANSI /ANS 3.5 Section: 3.1.2 (21) Plant Malfunctions; Nuclear Instrumentation Failures Date Tested: 06/21/91 Run Times 1.00 hour Test Descriptjoat The purpose of this test is to verify proper response of the simulator to individual LPRM failure upscale. Initial conditions are as follows: 75% power, all APRM's within 2% of each other with all LPRM signals to APRM's valid. Failure is inserted on one of the LPRM's to APRM "D" (severity level at 100%). Alarms, recorders and indicating lights are verified to operate properly with APRM "D" reaching approximately 77.5%. The failure is verified to be inserted and deleted properly. At completion of the test, the plant is at 75% power. Baseline Data used For Reference Reference Type Normal Operating Procedures SOI-C51(APRM) AVERAGE POWER RANGE MONITORING SYSTEM (UNIT 1) Reference Type: Other Dest Estimate O" Malfunctions Tested: NM03 LPRM DETECTOR FAILURE 1 Discrepancies: None Evaluators: B. Panfil, H. DeBoer, O. Minshall i 1 O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 3 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SA PAGE 20 of 34 Test T.4.5.1.17.NM10, NEUTRON MONITORING DETECTOR DRIVE STUCK Revision Number: 01 ANSI /ANS 3.5 Section: 3.1.2 (21) Plant Malfunctions; Nuclear Instrumentation Failures Date Tosted: 06/21/91 Run Time: 1.5 hours i Test

Description:

The purpose of this test is to verify the proper response of the simulator to a stuck reutron detector (SRM/IRM). There are 12 cases of NM10 (A-L), one for each moveable detector. The failure is variable to allow for sticking the detector at any position between full in (0%) and full out (100%). The simulator is initialized to a shutdown IC in which the permissives for detector movement are satisfied. The malfunction is activated for various detectors at different severities and the detector is moved and verified to stick at the position corresponding to the selected severity (full in, full out, intermediate) by observing reactor period, detector counts, status lights, etc. Each case of NM10 is verified to activate and delete properly from the instructors console. Baseline Data used For Reference Reference Type Normal Operating Procedures SOI-C51(SRM) SOURCE RANGE MONITORING SYSTEM (UNIT 1) Reference Type: Other Best Estimate Malfunctions Tested: NM10 NEUTRON MONITORING DETECTOR DRIVE STUCK Discrepancies: None Evaluators: B.

Panfil, H. DeBoer, G. Minshall t

r(

r CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 [N Simulator Tests - Malfunction Test Abstracts TAB C - Part SA \\_,) PAGE 21 of 34 Test T.4.5.1.18.OG03, 00 SYSTEM LEAK UPSTREAM OF ADSORBERS Revision Number: 01 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions, USAR Accident 15.7.1 Date Tested: 06/23/91 Run Times 0.25 hour Test

Description:

The purpose of this test is to verify the proper response of the simulator to a break in the Offgas process piping upstream of the charcoal adsorbers. Halfunction OG03 is variable, 0-100s, and the break location is within the holdup pipe room. The sinulation is initialized to a 100% steady state IC and the malfunction is inserted with 100% severity. Offgas parameters such as flow, component differential pressures and component temperatures are verified to decrease. Transport of radioactive material is verified by increases in holdup pipe room radiation levels and increases in the offgas vent pipe radiation levels. The plant remains at 100% steady state with offgas being processed to the offgas vent pipe via Offgas Building Ventilation System (M36). OG03 is then deleted to verify it can be cleared from the instructor station and a sitw return to normal pre-test conditions 10 checked. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures /"')T ONI-D17 HIGH RADIATION LEVELS WITHIN Pt. ANT (UNIT 1) ( Reference Type Other Best Estimate Malfunctions Tested: OG03 OG SYSTEM LEAK UPSTREAM OF ADSORBERS Major Discrepancies: F-0323 WHEP A D17 MONITOR EXCEEDS THE ALERT SETPOINT, THE HIGH ALT.RM LIGHT BLINKS, AND VICE VERSA. THIS WAS NOTED ON 2 OG RAD MONITORS. COULD BE A GENERIC PROBLEM. F-0330 WITH MF OG03 ACTIVE, PARTICULATE AND IODINE RAD LEVELS DECREASED PROPORTIONAL TC GASEOUS ACTIVITY, PARTICULATE AND IODINE ACTIVITY CONTINUE TO INCREASE SINCE THEIR VALVES ARE CUMULATIVE (INTEGRAL) DUE TO FILTER IN SAMPLE SKID. F-0332 OBSERVED NO INCREASE IN OG HOLD UP PIPE AREA RADIATION WHEN MALFUNCTION OG03 WAS ACTIVE AT 100% (LEAK IN HOLD UP PIPE). Evaluators B.

Panfil, H.

DeBoer, G.

Minshall O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Teste - Malfunction Test Abstracts TAD c - Part 5A PAGE 22 of 34 Test T.4.5.1.18.OG04, OG SYSTEM LEAK DOWNSTREAM OF ADSORDERS Reviolon Number: 01 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions, USAR Accident 15.7.1 Date Tested: 06/23/91 Run Times 0.25 hour Test

Description:

The purpose of this test is to verify the proper response of the simulator to an offgas process piping failure downstream o' the adsorbers. Malfunction OG04 is variable, 0-100%, and th( break location is at the inlet to the after filters upstream of N64-F061. The simulator is initialized to a 100% steady state IC and the malfunction is inserted with 100s severity. Offgas pcrameters such as individual train flow, indicated total system flow and component differential pressures are verified to change in the correct direction for the given break location. Offgas vent pipe radiation levels are verified not to change as this process has been treated by the adsorbers. The plant remains stable at 100s throughout this test. OG04 is deleted and the parameters return to their protest values. Baseline Data 'ased For Reference Reference Type: Off-Normal Operating Procedures ONI-D17 HIGH RADIATION LEVELS WITHIN PLANT (UNIT 1) O Reference Type Other Best Estiinate Malfunctions Tested: OG04 OG SYSTEM LEAK DOWNSTREAM OF ADSORBERS Discrepancies: None Evaluators B. Panfil, H. DoDoor, G. Minshall l l t l O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Teste - Malfunction Test Abstracts TAB C - Part SA PAGE 23 of 34 Test T.4.5.1.21.RD03, CONTROL ROD DRIFT - IN Revision Numbers later ANSI /ANS 3.5 Sec'; ion: 3.1.2 (12)(c) Plant Malfunctions; Drifting Control Rod Date Tested: Not Complete Run Times Later hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to a control rod drift caused by Directional Control valve (DCV) EP-123 failure during rod motion. RD03 may be selected for any one or more of the 177 control rods. The plant is initialized to a 100% steady state IC and a partially withdrawn control rod is selected at the instructor station. With no rod motion, RD03 is verified to be passive. When rod motion is commanded by the operator, the selected rode's EP-123 fails to reclose following the completion of the insert (or withdrawal) sequence and an inward rod drift occurs. Changes in rod position and appropriate RC&IS alarms are verified to occur. Rod motion continues until RD03 is deleted from the instructor station. The plant remains steady at approximately 100% power after the test is cotapleted. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures OM6 ARI PERRY UNIT ALARM RESPONSE INSTRUCTIONS (_ Reference Type Other Best Estimate Malfunctions Tested: RD03 CONTROL ROD DRIFT - IN Major Discrepancies: Unknown Evaluators Later l l l l l l I l l O i l 1 l l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part 5A PAGE 24 of 34 Test T.4.5.1.21.RD04, CONTROL ROD DRIFT - OUT Revision Numbers later ANSI /ANS 3.5 Section: 3.1.2 (12)(c) Plant Malfunctions; Drifting Control Rod Date Tested: Not Complete Run Times Later hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to a control rod drift outward caused by a stuck collet piston. RD04 may be selected for any one or more of the 177 control rods and the failure occurs following a command signal for rod motion. The simulator is initialized at a 100% steady state IC and a control rod that is not fully withdrawn is selected for failure at the instructor station. With no rod motion, the malfunction is verified to be passive. Following a withdraw sequence, the rod continues outward at a rate of about 1 inch per secc7d. Changes in rod position and other RC&IS indications as well sa a Rod Drift alarm are verified to occur. The rod drift can be stopped by the command of an insert signal or single rod scram using a remote function. The plant remains steady at approximately 100% power after the test is cocpleted. Baseline Data used For Reference i Reference Type: Off-Normal Operating Procedures OM6 ARI PERRY UNIT ALARM RESPONSE INSTRUCTIONS Reference Type: Other Best Estimate Malfunctions Tested: RD04 CONTROL ROD DRIFT - OUT Major Discrepancies: Unknown Evaluators Later O 1

1 CERTIFICATION OF PERRY SIMULATION TACILITY DOCKET No. 50-440 O Simulator Tects - Halfunction Test Abstracts TAB C - Port $A PAGE 25 of 34 i Test T.4.5.1.22.RN02, RHR SYSTEM PIPE DREAK Revision Number later ANSI /ANS 3.5 & action: 3.1.2 (07) Plant Halfunctions; Loss of Shutdown Cooling DG Tested: Not Complete Run Times Later hours Test Deecription: The purpose of this test is to verify the proper response of the simulutor to a creak in a RHR Pump discharge pipe. RH02 has three cases: RH02A, B and C which simulate a. break in each of the three RHR loops. The malfunction is variable, 0-100% with 100% equivalent to an 18 inch break. Since RHR has the ability to have up to three different suction sources, this test looks at all three separately. The simulator is initialized in a cold Shutdown state to test the fuel pool suction pa-and a Hot Shutdown IC to test suppression pool cooling ruction path and a sin to test the shutdown cooling flowpath. In all three rosets, the malfunc* ton i RH02 is teeted to 100% severity. Flowratos, pressures, water inventory transfers and automatic system activations and isolations are verified to occur dependent on suction source and fluid temperatures. At the completion of each test, RH02 is deleted to ensure it can be cleared from the instructor station. O Baseline Data used For Reference Reference Type Normal Operating Procedures SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) Halfunctions Tested RHQ2 RHR SYSTEM PIPE BREAK Hajor Discrepancies: Unknown Evaluators: Later j O 1 ~e-r- -+ s-w v-~ r--, w, w,w w w-----,-n g w-e-s--wor nz,e .v-em-m --wn---,--m_~~---m\\m. w.m-----_,,-- --,,--,--.e,--_-.e-n--w--. - - - - n-ns. ~... ~ - ---+-

CERTIFICATION OF PERRY S!HULATION FACILITY DOCKET NO. 50-440 Simulator Teste - Malfunction Test Abstracts TAB C - Part $A l ( PAGE 26 of 34 i Test T.4.5.1.23.RP02, INADVERTENT INITIATION OF ARI Revision Humber: 02 l ANSI /ANS 3.5 Section 3.1.2 (17) Plant Malfunctions; Failure in Reactivity i . Control System i Date Tested: 06/22/91 Run Times 0.50 hour Test

Description:

The purpose of this test is to verity the propets response of the simulator to un inadvertent initiation of Alternate Rod Insertion (ARI). RP02 has four cases A, B, C, and D, one for vach channel of RRCS. t All four cases will be inserted during this test. The simulator is 8 initialized to a 100% steady state Ic and each case is inserted individually. It is then verifled that alarms, indications and automatic actions occur per system design resulting in a full core scram. After the ARI is completed, the malfunctions are deleted and following the appropriate time delay, it is verified that the ARI can be reset. The plant stabilizes in a post scram condition at the completion of the test. Baseline Data used For Reference Reference Type Off-Normal Operating Proceduras ONI-C71-1 REACTOR SCRAM (UNIT 1) O Reference Type Other .Best Estimate Malfunctions Tested RP02 INADVERTENT INITIATION OF ARI f Discrepancies: None Evaluators D. Panfil, H. DeBoer, G. Hinshall N b i O

CEMTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Teuts - Halfunction Test Abstracts TAB C - Part SA PAGE 27 of 34 Test T.4.5.1.25.SWO1, NCC SYSTEH PROCESS PIPING LEAKAGE Revision Hurrbor: 02 ANSI /ANS 3.5 Section: 3.1.2 (08) Plant Half unctions; Loss of Component Cooling Date Tested: 06/22/91 Hun Times 1.00 hour Test

Description:

The pur pose of this test is to verify the proper response of the simulator to a loss of duelear Closed Cooling caused by a discharge header break. This malfunction is variable, 0-100% with 100% equivalent to a 30 inch pipe break. The simulator is initialized to a 100% steady state IC and SWO1 is inserted with 100% severity. The response of NCC is verified to include a lose of system inventory, system flow and pressure. Components cooled by NCC are verified to heatup and the automatic actions listed in ONI-P43, Loss of Huclear Closed Cooling are also verified. The plant stabilizes in a poet scram condition with NCC cooled components continuing t to degrade. Haltunction SWO1 is then deleted and an increase in NCC surge tank level is verified. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures ONI-P43 LOSS OF NUCLEAR CLOSED COOLING (UNIT 1) Reference Type Other Best Estimate i Halfunctions Tested: SWO1 NCC SYSTEM PHOCESS PIPING LEAKAGE Hajor Discrepancies F-0303 WHEN P47 A CHILLER TRIPPED (P904) TH2 CHILLER TRIP ANNUNCIATOR WAS NOT RECEIVE 0. j Evaluators: B. Panfil, H. DeBear, G. Hinshell l f { l 1 0

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part SA PAGE 28 of 34 Test T.4.5.1.25.SWO2, SERVICE WATER SYSTEM PROCESS PIPING LEAKAGE Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.2 (06) Plant Malfunctional Loss of Service Water Date Tested: 06/22/91 Run Times 3.00 hour Test

Description:

The purpose of this test is to verif y the proper response of the simulator to a loss of Plant Service Water caused by an underground pipe break. This malfunction is variable, 0-100% with 100% equivalent to a 54 inch pipe break. The simulator is initialized to a 100% steady state IC and SWO2 is inserted with 100% severity. The response of Service Water is verified to include a loss of system flow and discharge pressure. Components cooled by Service Water are verified to heatup and the automatic actions listed in ONI-P41, Loss of Service Water, are also verified. The plant stabilizes in a post scram condition and the loads serviced by Service Water continue to degrade. Halfun: tion SWO1 in then deleted. Baselino Data used P ! Reference Reference Type Off-Normal Operating Procedures ONI-P41 LOSS OF SERVICE WATER (UNIT 1) g Reference Type Other Best Estimate Malfunctions Tested: SWO2 SERVICE WATER SYSTEM PROCESS PIPING LEAKAGE Major Discrepancies: F-0304 BEFORE FAILURE TBCC AND MAIN LUBE OIL TEMP CONTROLLERS ARE ALREADY FULLY OPEN WITH LAKE TEMP AT 55 DEG. THEY ARE NORKALLY AT SOME THROTTLED POSITION, ESP. AT 55 DEO LAKE TEMP. NEED ABILITY TO CONTROL MLO AND TBCC TEMPS "IN BAND" UP TO LAKE TEMP OF 00 DEG F. F-0306 WITH SWO2 INSERTED COOLING TWR MAKE UP VLV HAD TO CLOSE DOWN BECAUSE IT STILL MAL FLOW. FLOW TO CT BASIN WENT UP WHEN WEIR LEVEL WENT TO 0. Evaluators: D. Panfil, H. DeBoer, G. Minshall l 1 1 .m.

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET HO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part 5A PAGE 29 of 34 Test T.4.5.1.27.TH01, RECIRC LOOP RUPTURE (DBA LOCA) Revision Humber: later ANSI /ANS 3.5 Section: 3.1.2 (01)(b) Plant Halfunctions; LOCA inside containment Date Tested: Not Complete i Run Times Later hours j Test

Description:

The purpose of this test is to verify the proper response of the simulator to a DBA Recire Loop Rupture (LOCA). This malfunction has two cases, one for each recirc loop. Each simulates a catastrophic failure of the recire piping between the RPV and the pump suction valve. The simulator is initialized to a 100% steady state IC and TH01A is inserted. The response of the RPV, Containment and Drywell are evaluated against the Perry USAR Chapter 15.6.5. Following core reflood to above TAF, all ECCS except LPCS is terminated to verify a f1vodable volume exists such that RPV water level does not decrease to less than 2/3 core height. Peak cladding temps are verified not to exceed 2200 deg F during this transient. The containment Isolation function and ECC3 initiation functions are verified to occur. The plant stabilizes in a post scram condition. TH01A is then verified to not delete until the simulator is reset. THolB is incerted after reinitialization and verified to also be non-recoverable. Baseline Data used For Reference Reference Type Other-Best Estimate Halfunctions Tested: TH01 RECIRC LOOP RUPTURE (DBA LOCA) Major Discrepancies: Unknown Evaluators: Later 1 I 3 O -,.v-. m.___.3 ,,,--.__-.y-__m--,,-_,,,,-.,,._,_,--,-,y ,_-,.,~,--,,_-,-,,.--3mv,,m,,,,y,,,-_%,,_,mm,,...,,,,,,. __,,,,,_,,,,,,-_,.,_,,w,,,,,, ,,m,-,-

i CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. Su-440 Simulator Tests - Halfunction Test Abstracts TAB C - Part 5A PAGE 30 of 34 Test T.4.5.1.27.TH19, RPV LEVEL INST REFERENCE LEG BREAK Revision Number: 01 ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.6.2 Date Tested: 06/23/91 Run Times 1.00 hour Test

Description:

The purpose of this test is to verify the proper response of the simulator to a rupture of a reactor vessel level inotrument sensing line (reference leg). This malfunction has several cases which allow the failure of different lines, either inside or outside the drywell. Tne simulator is initialized at 100% Power IC. Individually, each generic case is activated and the response of the simulator is checked. For each caso, proper indication for a small break LOCA are observed. The overall plant response is judged depending on the location of the break (inside/outside the Drywell). The response of attached level and pressure sensing instruments is also verified to be proper. The malfunction is non-recoverable, so the simulator must be reset to test each case. At the end of each tost run, the reactor is in a Post-Scram condition, with ECCS systems injecting to the RPV to maintain water level. Baseline Data used For Reference Reference Type: Other Best Estimate Halfunctions Tested: TH19 RPV LEVEL INST REFERENCE LEO BREAK Hajor Discrepancies: F-0298 TRANSHITTERS FOR CERTAIN VESSEL PRESSURE / LEVEL INSTRUMENTS ARE NOT ASSIGNED TO ANY INDICATOR " LEG" OR ARE ASSIGNED TO WRONG RET / VARIABLE LEG. Evaluators: B. Panfil, H. DeBoer, O. Hinshall O l . _ ~ - ~. - -, - -. - -,,,. -. -, -. -.,

i i CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part 5A PAGE 31 of 34 l l Test T.4.5.1.27.TH20, RPV LEVEL INST VARIABLE LEG BREAK l Revision Number later ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.6.2 1 Date Tested: Not Complete Run Times Later hours Test

Description:

The purpose of this test is to verify the proper responso of the simulator to a rupture of a reactor vessel level instrument sensing line (reference leg). This malfunction has several cases which allow the i failure of different lines, either inside or outside the drywell. The simulator is initialized at 2004 Power IC. Individually, each generic case is activated and the response of the simulator is checked. For each case, i proper indication for a small break LOCA are observed. The overall plant [ response is judged depending on the location of the break (inside/outside the Drywell). The response of attached level and pressure sensing instruments in also verified to be proper. The malfunction is non-recoverable, so the simulator must be reset to test each case. At the end of each test run, the reactor is in a Post-Scram condition, with ECCS 1 systems injecting to the RPV to maintain water level. Baseline Data used For Reference l O. Reference Type Other Best Estimate Halfunctions Tested: TH2O APV LEVEL INST VARIABLE LEG BREAK Major Discrepancies: Unknown Evaluators Later t i O t ...-.+o ~.-..-,-.-----.---.-e...----.-. . _ - ~,. - -. ~ - -.. - ~. - - - -.. - - -.. - - - ---m. e-m.,,,- - - - - - - -. - -,.. - -. _ -.,

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Testa - Malfunction Test Abstracts TAB C - Part 5A PAGE 32 of 34 Test T.4.5.1.27.TH21, POWER / FLOW INSTABILITIES (IEB 88-07 Supplement 1) Revision Number: 02 ANSI /ANS 3.5 Sections 3.1.1 (07) Normal Plant Evolutions; Operatiens with Less than Full coolant Flow Date Tested: 06/23/91 Run Times 0.40 hour Test

Description:

The purpose of this test is to verify the proper response of the simulator to neutron flux instabilities that may result from operation in the region of the BWR-6 power to flow map known as the " red zone" or high power / low flow region. While the exact magnitude of the instabilities is currently unknown for the Perry Plant, and intentional operation in this region is prohibited (which accounts for lack of comparison cata), some data is available from the LaSalle Plant event, and predictionc of possible response have been made by General Electric. This malfunction is available to allow training of Perry Operators on the possible effects of this phenomena. The effects can be varied from minor oscillations to severe oscillations which result in a Scram. The test is initiated from the 100% Power operating condition. The malfunction is inserted at 100% severity. No offects are seen since initial operation is occurring outside the prohibited region of the power to flow map. Flow is then intentionally lowered to enter the red zone. Out of phase neutron flux 0 oscillations are verified to occur on individual LPRH's. Oscillations are also verified on the APRM's, resulting in a Reactor Scram due to high flux. At the end of the test, the plant is stabilized in a post-Scram condition. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures ONI-B33-1 REACTOR RECIRCULATION FLOW CONTROL MALFUNCTION (UNIT 1) ONI-CSI UNEXPLAINED CHANGE IN REACTOR POWER OR REACTIVITY (UNIT 1) Reference Type Other Best Estimate Halfunctions Tested: TH21 POWER / FLOW INSTABILITIES (IEB 88-07 Supplement 1) Major Discrepancies: F-032' WHEN MALFUNCTION TH21 WAS INSERTED WHILE OPERATING IN THE REGION OF INSTABILITY OF THE POWER / FLOW MAP, DID NOT RECEIVE LPRM/APRM UPSCALE /DOWNSCALE ALARMS IN AN OSCILLATING FASHION. WHEN OSCILLATIONS OCCUR THEY SHOULD BE RANDOM UPSCALE AND DOWNSCALE IN P#IFFERENT RELIONS OF THE CORE AS FLOW OSCILLATIONS OCCUR IN OIFFERENT REGIONS OF THE CORE. P-0966 WHEN AT 100% POWER ACTIVATION OF THIS MALF CAUSES A SCRAM IKMEDIATELY. Evaluators: B. Panfil, H. DeBoer, G. Hinshall O

CERTIFICATION OF PERRY SIHJLATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part SA PAGE 33 of 34 Test T.4.5.1.27.TH27, MAIN STEAM LINE RUPTURE IN STEAM TUNNEL Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.2 (20)(b) Plant Halfunctions; Main Steam Line Break, outside containment Date Tested: 06/23/91 Run Times 0.50 hour Test

Description:

The purpose of this test is to verify the Simulator response to a Main Steam Line (HSL) Rupture in the Steam Tunnel. The break location is downstream the HSIV's and isolable. The malfunction is single case, nonrecoverable. The test is initiated f rom 100% Power operating condition. The failure is activated and plant response is verified against Perry USAR chapter 15.6.4 analysis. Proper response of alarms, automatic actions, and key parameter trends are checked. At the end of the test, the plant is stabilized in a Post-Scram condition with the containment and Reactor Vessel isolated, pressure controlled by safety / relief valves, and RPV level being maintained by the feedwater system, augmented by t he High Pressure Core Spray and Reactor Core Isolation Cooling systems. Failure TH27 is verified to be non-recoverable by attempting to delete it from the Instructors Station. Baseline Data used For Reference Reference Type: Other Best Estimate Halfunctions Tested: TH27 KAIN STEAM LINE RUPTURE IN STEAM TUNNEL Hajor Discrepancies: F-0327 NO RADI A'. ?FECTS OBSERVED DURING STEAM LEAK IN STEAM TUNNEL. EXPECT Ih. ED RAD LEVELS FOR TB/HB VENT EXHUAST, AUX BLDG EXHAUST, TU. DEA RAD MONITORS. F-0328 ALL WESTRONICS MULTIPOINT TEMP RECORDERS ARE DISPLAYING TEMPS THAT ARE MORE THAN 10% IN ERROR ON THE HIGH END. SCALE NORMALIZATION IS LINEAR, HETER SCALES ARE NOT. Evaluators: B.

Panfil, H. DeBoer, G.

Hinshall O

CERTIFICATION OF PERRY SIMULt. TION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part $h PAGE 34 of 34 l Test T.4.5.1.28.7001, MAIN SHAFT OIL PUMP DEGRADATION i Revision Number: 02 ANSI /ANS 3.5 Section: 3.1.2 (15) Plant Malfunctions; Turbine Trip Data Tested: 06/23/91 Run Times 0.50 hour Test

Description:

The purpose of this test is to verify the proper response of the Situulator to a Turbine Trip caused by a f ailure of the Main Shaf t Oil Pump (MSOP). The malfunction is single case, variable severity where 100% severity represente complete pump failure (O PSI discharge head). The test is initiated from the 100% Power operating condition. The malfunction is activated on a 2 minute ramp from 0 to 100% severity. During the pump degradation, oil pressures in the Main Turbine Lube Oil system (HTLO) are monitored. Automatic actions are verified as pressure lowers such no standby and emergency pumps otarting. A Main Turbine trip on low MSOP pressure is observed to occur, resulting in a plant shutdown. At the end of the test, the plant is stabilized in the Post-Scram condition with the motor suction pump, turning gear oil pump, and all bearing lif t pumps running. Baseline Data used For-Reference i Reference Type Off-Normal Operating Procedures ONI-N32 TURBINE AND/OR GENERATOR TRIP (UNIT 1) Reference Type Other Best Estimate Malfunctions Tested: TU0I KAIN SHAFT OIL PUMP DEGRADATION Discrepancies: Nonu Evaluators: B. Panfil, H. DeBoer, G. Minshall O

CERTIFICATION oF PERRY SIMULATIoM FACILITY DOCKET lio. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part SB Pact 1 of 62 ATP TEST SECTIO!1 T.4.5.3 14ALFUllCTIO!1 SCEllARIOS Overview of testinct performed in this seclipa This series of Tests woro performed to verify the proper response of the Simulator to the set of Malfunction Scenarios for which the simulator is being certified. These malfunction scenarios are comprised of one or more System Level and/or Component Level Failures, combined with 1/o overrides, Event Triggers, and/or Remote Functions. Each 14alfunction Scenario can ba used singly or in combination during the course of an Operator Training or Evaluation session. Prior tests in the Acceptance Test Procedure (ATP) tested the individual Failures which are contained in a scenario, so the primary purpose of the section 4.5.3 tests were to verify the proper overall plant response as predicted for each malfunction scenario. The test abstracts include a description of which specific combination of failures (including rates and severities), overrides, triggers, and remote functions were tested. The O tests are divided into two basic categories: those which validate malfunctions required for simulation of types of Perry USAR accidents (those which result in observable indications on control room instrumentation and for which the simulator has boon determined to be applicable), and those which support the current performance based operator training curricula and were not tested in previous ATP sections T.4.5.1 or T.4.4.2. 11 0. of Tests included in this Section: 1190: Open SDR's: 1.4.5.3.01, Loss of fitDWAf te HEAtthG (PRf C IN AUTC) 3 (N/C)* 1.4.5.3.02, Loss of fitSWAt[R HE AllhG (RRf C IN MANUAL) 4 (W/C)* T.4.5.3.03. fitDWAlf R CONTROLLER F AILUPt MAXIMUM DIMAND (N/C)* 1.4.5.3.04, PRESSURE REGULATOR FAILURt OPEN 6 (N/C)* 1.4.5.3.05, thADvtRTENT SAf t1Y/RttlEf VALVE OPEhlhG 7 (W/C)* f.4.5.3.06, thADvtRitNT RNR SHUTDOWN C00LlhG OPER A110N 8 (N/C)* T.4.5.3.07, PRESSURE REGUL ATOR F AILUtt ClotID 9 (N/C)* 1.4.5.3.08, CENERATOR LOAD PfJECT WitN BYPASS VALVts 10 (N/C)* 1.4.5.3.09, CEWERATOR LOAD RIJECT WITHOUT BYPA11 VALVE $ 11 (N/C)* 1.4.5.3.10, TUR8 t ht 1R IP 12 (k/C)* T.4.5.3.11, LOSS of AC POWtk (LOS$ of AUX IPANSf 0RMER) 13 (N/C)* T.4.5.3.12 F AILU6E Of R $1 DUAL HE AT REMOVAL $HUTDOWN COOLING 14 (N/C)* 1.4.5.3.13, RfCfRC FLOW CONTROL FAILURE DECREA$!WG (BOTH FCV'$) 15 (W/C)* 1.4.5.3.14, RECIRCULATION PUMP SEllVRE 16 (N/C)* 1.4.5.3.15, ABNORMAL $1ARfUP Of IDLE RECIRCULA110N PUMP 17 (N/C)* 1.4.5.3.16 REClRCULAi!ON PUMP SHAf f SME AR 18 (N/C)* f.4.5.3.17, RECIRC f t0W Cohf ROL F AILUst.lkCRI A5thG (Bof H FCV's) 19 (N/C)* T.4.5.3.18, CONTROL ROO DROP AtCIDENT 20 (N/C)*

CERTIFICATIott or PERRY SIMULATIoft FACILITY DOCKET llo. 50-440 O Simulator Tests - Malfunction Test Abstracts Tall C - Part LB PAGE 2 of 62 lio. of Tests included in this Sectipl): Bige: R2en SDILLu: 1.4.5.3.19, th ADVERit=1 HIGH FRt55uti CD&E 5F R AT $1 AR10P 21 (k/C)* 1.4.5.3.20, An1lClrattD 1RAksithi vitHauf stRAM (Afv5) 22 (w/C)* 1.4.5.3.21, LOS$ Of Off $ lit PostR 23 (W/C)* 1.4.5.3.22, til MAlpuhC110N SCthARIO 81 24 (W/C)' 1.4.5.3.23, til MAttuhtflow SCtkAtlo 82 26 (w/C)* 1.4.5.3.24, Pfl MAltuhC110N ScthAkl0 #3 28 (W/C)* 1.4.5.3.25, Pfl MAttuwC110N ScthARio 84 30 (N/C)' 1.4.5.3.26, l'El WAlfuhC180N $CtkARIO #5 (PART 1) 32 (h/C)* 1.4.5.3.27, tvAtuallON MAttuwtilDN scthAtl0 #1 31 (w/C)* i.4.5.3.28, tvAtuall0N MAlfuhCIICW LCthARIO #2 35 (N/C)* i.4.5.3.29, tvAluAllON MAlpuhtil0N StthARIO r3 37 (m/C)' 1.4.5.3.30, EVALUATION MAlfuhCTION SCthARio 84 39 (W/C)* T.4.5.3.31, tvALUAll0N MAttuhcit0N SCth ARio #5 40 (W/C)' 1.4.5.3.32, tvAtuAll0N MAttuhtilON SCthatl0 86 42 (N/C)* 1.4.5.3.33, tvAtuA110N MAttuhC110N SCthAtto 87 44 (W/C)* 1.4.5.3.34, EVALUAll0N MAlfuNCilON SCthARIO #8 46 (W/C)* 1.4.5.3.35, (VALUAllCW MAlfuhCil0N SCth AR10 #9 48 (W/C)* 1.4.5.3.36, tvALUA110N MAlfuhC110N SCthAtto #10 50 (N/C)* 1.4.5.3.37, tvALUA110N MAlluhC110N ScthARio #11 52 (Wic)* T.4.5.3.38, tvatuA110W MAltuwC110N SCthARIO #12 54 (w/C)* 1.4.5.3.39, Ev&LUAllow MalfuhCil0N sith4Rio 813 56 tu/C)* f.4.5.3.40, tvALuA110N MAlfuhCilON SCthAtl0 #14 58 (N/C)* t.4.5.3.41, tvAtuA110N MAlf uhC110N SCth ARio s15 60 (N/C)* 1.4.5.3.42, l't! MAlfukCilON scthARIO 85 (FART 2) 62 (N/C)* AliSI/ Alls 3. 5 Itoforence: 3.1.2 Plant Malfunctions This test has not been completed (11/ C) or has boon performed but has one or more unronolved major Simulator Discrepancy Itoports associated with it. Please see the individual test abstract for those discrepancies which constituto exceptions to AliSI/AliS 3.5 section 3.1.2. O l 1

t i CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 l O Simulator Tests - Halfunction Test Abstracts TAB C - Part $B i PAGE 3 of 62 i Test T.4.5.3.01, LOSS OF FEEDWATER HEATING (REACTOR REC 1RC FLOW CONTROL IN AUTO) Revistora Number Later r ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.1.1 i Date Tested: Not RUN s Run Times Later hours i I Test

Description:

The purpose of this test is to verify the proper i response of the simulator to a partial loos of Feedwater Heating accident with Reactor Recirc Flow Control in AUTO as described in the Perry USAR, section 15.1.1. The test is initialized at 100% Power ope rat ion. Extraction Steam to the number SA FW heater is isolated by inserting a component failure on the supply MOV. The sequence of events and key parameter response is checked as per the USAR 2 description. At the end of the test, conditions are as follows: Lore flow is about 80% of initial, Reactor Power is 106% of initial, i feodwater injection temperature has lowered, Turbine load and steam flow are nearly the same as initial. Baseline Data used For Reference Reference Types Normal Operating Procedures FTI-B10 PREPARATION FOR FINAL FEEDWATER TEMPERATURE REDUCTION OPERATION Reference Type Off-Normal Operating Procedures ONI-C51 UNEXPLAINED CHANGE IN REACTOR POWER OR REACTIVITY (UNIT 1) ONI-N36 LOSS OF FEEDWATER HEATING (UNIT 1) Reference Type Plant Data - Analyses USAR 15.1.1 PNPP UPDATED SAFETY ANALYSIS REPORT USAR FIGURE 15.1-1 PNPP UPDATED SAFETY ANALYSIS REPORT 'USAR TABLE 15.1-1 PNPP UPDATED SAFETY ANALYSIS REPORT j Malfunctions Tested: MV03 MOV Spurious valve closure Discrepancies: Unknown Evaluators: Later l e r 4 __._._.._.~._-..._._..-,,.$

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 i O Simulator Tests - Halfunction test Abstracts TAB C - Part 5B PAGE 4 of 62 l i Test T.4.5.3.02, LOSS OF FEEDWATER HEATING (REACTOR RECIRC FLOW CONTROL IN HANUAL) Revision Number: Later ANSI /ANS 3.5 Section 3.1.2 Plant Haltunctions, USAR Accident 15.1.1 ? Da':e Teeted: Not RUN I Run Times Later hours Test

Description:

The purpose of this test is to verify the proper i response of the simulator to a partial loss of reedwater Heating accident with Reactor Recirc Flow Control in Manual as described in the i Perry USAR, section 15.1.1. The test is initialized at 89% Power operation. Extraction Steam to the number 6A FW heater is isolated by inserting a component failure on the supply HOV. The eequence of events and key parameter response is checked as per the USAR description and the results of Startup Test STI-N27-023D. At the end of the test, conditions are as follows: Core flow is the same as initial, Reactor Power has increased, but a SCRAM has not occurred, feedwater injection temperature delta is less than or equal to 100 dog F. The acceptance critoria for this test is identical to the STI acceptance criteria for the plant startup test. Baseline Data used For Reference O Reference Type Normal Operating Procedures FTI-B10 PREPARATION FOR FINAL TEEDWATER TEMPERATURE REDUCTION OPERATION Reference Type Off-Normal Operating Procedures ONI-C51 UNEXPLAINED CHANGE IN REACTOR POWER OR REACTIVITY (UNIT 1) ONI-N36 LOSS OF FEEDWATER HEATING (UNIT 1) Reference Type Plant Data - Analyses i USAR 15.1.1 PNPP UPDATED SAFETY ANALYSIS REPORT i Halfunctions Tested: MV03 h0V Spurious valve closure Discrepancies: Unknown Evaluators Later P i O l

CERTIFICATION OF PERRY S!HULATIOn FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts ThD C - Part SD PAGE 5 of 62 Test T.4.5.3.03, FEEDWATER CONTROLLER FAILURE-MAXIMUM DEMAND Revision Number: Later ANSI /ANS 3.5 Section 3.1.2 Plant Malfunctions, USAR Accident 15.1.2 Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verif y the proper response of the simulator to a reedwater Contcoller Failure accident (maximum demand) as described in the Perry USAR, section 15.1.2. The test is initialized at 100% power. The malfunction la 1.sitiated by inserting component level failure CH01 for the feedwater Master Level Controller IC34R0600 at 2004 severity. This causes both operating Rep' tor Feed Pumps to ramp to the high speed stop (5450 RPH), injecting the maximum amount of feedwater. The reactor scrams and the feed pumps automatically trip on high level (219.5 inches above TAF), mitigating the transient. The sequence of events and trend of key parameters is checked per the USAR description. At the end of tho tcst, the simulated plant is stabilized in the Post-Scram condition with reactor level being controlled by the HPCS and RCIC systems. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures O ONI-C51 UNEXPLAINED CHANGE IN REACTOR POWER OR REACT 1VITt (UNIT 1) t ONI-N36 LOSS OF FEEDWATER HEATING (UNIT 1) Reference Type Plard Data - Analyses USAR 15.1.2 PNPP UPDATED SAFETY ANALYSIS REPOhT USAR FIGURE 15.1-3 PNPP UPDATED SAFETY ANALYSIS REPORT t USAR TABLE 15.1-3 PNPP UPDATED SAFETY ANALYSIS REPORT Malfunctions Tested: CN01 Controller Auto / manual failure i Discrepancies: Unknown l Evaluators: Later k O 1 l

i l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 i O l Simulator Tests - Halfunction Test Abstracts TAB C - Part 5B PAGE 6 of 62 l I f Test T.4.5.3.04, PRESSURE REGULATOR FAILURE-OPEN Revision Number Later J ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.1.3 l Date Tested: Not RUN k Run Times Later hours Test

Description:

The purpose of this test is to verif y the proper response of the simulator to a Reactor Vessel Pressure Regulator Failure (open) accident as described in the Perry USAR, section 15.1.3. The test is initialized at 1004 Power, EOC life. To simulate l the initial conditions of the plant as described in the USAR analys11, the HSIV stroke time is set to maximum allowable (5 seconds) with System Level Failure HS03. The Pressure regulator is placed in TEST l mode (simulating that one channel is out of service) with remote r function TC08. This prevents an automatic transfer to the non-f ailed channel. The malfunction is initiated by failing the main steam line pressure transmitter input to the inservice regulator with component level failure PT01 at 100% severity or failed upscale. The pressure regulator responds by rapidly opening all the valves under its controls the Turbine Control Valves, and the Steam Bypass Valves. The rapid rise in steam flow causes a RPV depressurization and vousel level i swell, initiating a Reactor SCRAM due to high level. The sequence of O events and response of key parameters are verified per the USAR description. At the end of the test, the reactor is shutdown, the Main feedwater pumps and the Main Turbine are tripped, the HSIV's are isolated, reactor pressure is being controlled by the Safety and Relief Valves, and reactor level is being controlled by the HPCS and RCIC systems. Baseline Data used For Reference Reference Type: Off-Normal Operating Proceduros ONI-C85-2 PRESSURE REGULATOR FAILURE - OPEN (UNIT 1) Ret 6tence Types Plant Data - Analyses USAR 15.1.3 PNPP UPDhTED SAFETY ANALYSIS REPORT 2 USAR FIGURE 15.1-4 PHPP UPDATED SAFETY ANALYSIS REPORT L USAR TABLE 15.1-4 PHPP UPDATED SAFETY ANALYSIS REPORT Halfunctions Tested: MS03 MSIV CLOSURE TIME VARIANCE PT01 Process Transmitter Variable failure Discrepancies: Unknown Evaluators ' Later

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - lla1f unction Test Abstracts TAB C - Part 50 PAGE 7 of 62 Test T.4.5.3.05, INADVERTENT SkV OPENING Revision Number Later ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.1.4 Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to an inadvertent SRV (Safety and Relief Valve) opening as described in the Perry USAR, section 15.1.4. The test is initialized at the 100s Power Operating condition. One SRV control switch is failed in the open position (simulating a short in the switch) by use of I/O override. Plant parameters are verified to respond as described in the USAR analysis, and as backed up by Perry Startup Test Data f rom the perf ormance of SVI-D21-T2005. At the completion of the test, Reactor Power remains at 100s, and RPV pressure is slightly lowered from initial. Baseline Data used For Reference Reference Type Of f-Normal Operating Procedures ONI-D21-1 SRV INADVERTENT OPENING / STUCK OPEN (UNIT 1) O Reference Type: Plant Data - Analyses USAR 15.1.4 PHPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.1-5 PNPP UPDATED SAFETY ANALYSIS REPORT Halfunctions Tested: None Discrepaneles: Unknown Evaluators: Later

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part SD PAGE 8 of 62 Test T.4.5.3.06, INADVERTENT RHR SHUTDOWN COOLING OPERATION Revision Nurrbert Later ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.1.6 Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to a improper initiation of Shutdown Cooling flow as described in the Perry USAR, section 15.1.6. The tout is initialized with the plant in a cold startup, IC08. At the point of reactor criticality (low in the source range), actions are taken to initiate one RHR Loop in the shutdown cooling mode. At the reference plant, this would require operator error, as there are no plant analfunctions which could result an this condition. Plant parameters are verified to respond as per the USAR description. At tho end of the test, th) reactor is verifled to have scrammed on high nr.itron flux. Baseline Data used For Reference Reference Type Normal Operating Procedures 101-1 COLD STARTUP SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) Reference Type Plant Data - Analyses USAR 15.1.6 PHPP UPDATED SAFETt ANALYSIS REPORT Malfunctions Tested: None Discrepancies: Unknown Evaluators: Later O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SD PAGE 9 of 62 Test T.4.5.3.07, PRESSURE REGULATOR FAILURE-CLOSED Revision Nunbert Later ANSI /ANS 3.5 Section 3.1.2 Plant Halfunctions, USAR Accident 15.2.1 Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to a Reactor vessel Pressure Regulator Failure (closed) accident as described in the Perry USAR, section 15.2.1. The test to initialized at 1004 Power, Eoc life. The C85 Pressure regulator is placed in TEST mode (simulating that one channel is out of service) with remote function TCOB. This prevents an automatic transfer to the non-failed channel. The malfunction in initiated by failing the main steam line pressure transmitter input to the inservice regulator with component level failure PT01 (variable severity transmitter output failure) at Os severity or failed downscale. The pressure regulator responds by rapidly closing all the valves under its controls the Turbine Control Valves, and the Steam Dypass Valves. The rapid reduction in steam flow causes a RPV pressuritation and neutron flux excursion. A Reactor SCRAM occurs due to high flux. The sequence of events and response of key parameters ate verified por the USAR description. At the end of the test, the reactor is shutdown, the Main feodwater pumps and the Main Turbine are O tripped, and reactor pressure is being controlled by the Safety and Relief Valves. Baseline Data used For Reference Reference Type Of f-Normal Operating Procedures ONI-C85-1 PRESSURE REGULATOR FAILURE - CLOSED (UNIT 1) Reference Type Plant Data - Analyses USAR 15.2.1 PNPP UPDATED SAFETY ANALYSIS REPORT USAR FIGURE 15.2-1 PNPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.2-1 PNPP UPDATED SAFETY ANALYSIS REPORT Malfunctions Tested: PT01 Process Transmitter Variable failure Discrepancies: Unknown Evaluators: Later O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAD C - Part 50 PAGE 10 of 62 r Test T.4.5.3.08, GENERATOR LOAD REJECT WITH BYPASS VALVES Revision Hunbers Later ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.2.2 { Date Tested: Not RUN 3 Run Timel Later hours Test Descriptions The purpose of this test is to verify the prorer response of the simulator to a Generator Load Reject (Bypass available) as described in the Perry USAR, section 15.2.2. The test is initialized at the 100% Power operating condition, IC19. The load reject is initiated by opening the Generator Output Dreakors. The main turbine is verified to trip due to Power / Load unbalance. The plant response is verified ter the USAR description in that a SCRAM and f Rectre Pump Trip is initiated due to the turbine trip. At the end of r the test, the reactor plant is shutdown, and pressure is being l maintained by the bypass valves. I Daseline Data used For Reference Reference Type Of f-Normal Operating Procedures ON!-N32 TURBINE AND/OR CENERATOR TRIP (UNIT 1) r ONI-S11 LOSS OF OFF-SITE POWER Reference Type: Plant Data - Analyses I. USAR 15.2.2 PNPP UPDATED SAFETY ANALYSIS REPORT i USAR FIGURE 15.2-2 PNPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.2-2 PNPP UPDATED SAFETY ANALYSIS REPORT j Reference Type Plant Data - Startup Test Results STI-821-027/2 DATA REQUEST RESPONSE - TURD TRIP & GEN LD REJ Malfunctions Tested: CB01 Spurious breaker trip 7 Discrepancies: Unknown i Evaluators Later t i l ((() w-e,e- +e w%=,m-.%- .rw,,w~.,-e,-,-,r- -,.r e .w--+---,----~,,,,.w.,+--- ,--wec.,e-e~-we<=--w----.-,, ,-..---.,+,---,.,.----.--,-.4-%..-- w,y.--,ee=,--w---, .--r

~ CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAD C - Part SD PAGE 11 of 62 Test T.4.5.3.09, GE.lERATOR LOAD REJECT WITHOUT DYPASS VALVES Revision Number Later ANSI /ANS 3.5 Sectirn: 3.1.2 Plant Halfunctions, USAR Accident 15.2.2 e Date Tested: Not RUN Run Times Later hours Test Descriptions The purpose of this test is to verify the proper response of the simulator to a Generator Load Reject (Dypass not available) as described in the Perry USAR, section 15.2.2. The test is initialized at the 300% Power operating condition, IC19. System level failure TC04 is activated at 0% severity for all Turbine Bypass Valves to prevent them f rom opening. The load reject is then initiated by opening the Generator Output Dreakers. Tne main turbine is verified to trip due to Power / Load unbalance. The plant response is verified per a the USAR description in that a SCRAM and Recirc Pump Trip is initiated due to the turbine trip. The peak pressure observed is verified to not t exceed the USAR predicted maximum. At the end of the test, the reactor plant is shutdown, and pressure is being maintained by the SRV's. Baseline Data used For Reference Reference Type: Of f-Normal Operating Procedures O ONI-N32 TURB?NE AND/OR GENERATOR TRIP (UNIT 1) ONI-S11 LOSS OF OFF-SITE POWER r Reference Type: Plant Data - Analyses l USAR 15.2.2 PHPP UPDATED SAFETY ANALYSIS REPORT I USAR FIGURE 15.2-3 PNPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.2-3 PHPP. UPDATED SAFETY ANALYSIS REPORT t Halfunctions Tested: CB01 Spurious breaker trip TC04 BYPASS VALVE FAILURE . Discrepancies: Unknown l Evaluators Later ) F O

i i CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Halfunction Test Abstracts TAB C - Part 5B \\ PAGE 12 of 62 i f Test T.4.5.3.10, TURBINE TRIP i Revision Humbert Later l ANSI /ANS 3.5 Section 3.1.2 Plant Halfunctions, USAR Accident 15.2.3 l t Date Tested: Not RUN i Run Times Later hours i Test

Description:

The purpose of this test is to verify the proper response of the simulator to a Turbine Trip as described in the Perry USAR, section 15.2.3. Three turbine trip scenarios are performed by this tost. The first and second are initiated from the 1004 Power Operating condition, EOC. The third is initiated from 30% Power. The first turbine trip is initiated with bypass valves available. The second and third trips are initiated after first failing the Turbine i Bypass valves closed with failure TC04 at 0% severity. For each turbine trip, the sequence of events is verified per the USAR description. The response of key parametern is verified by comparison to the USAR figures and tables. At the end of each trip, the Simulator is stable in the Post-SCRAM condition. j Baseline Data used For Reference { Reference Type Off-Normal Operating Procedures O ONI-N32 TURBINE AND/OR GENERATOR TRIP (UNIT I) Reference Type Plant Data - Analyses { USAR 15.2.3 PNPP UPDATED SAFETY ANALYSIS REPORT USAR FIGURE 15.2-4 PNPP UPDATED SAFETY ANALYSIS REFORT i USAR FIGURE 15.2-5 PNPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.2-4 PHPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.2-5 PNPP UPDATED SAFETY ANALYSIS REPORT Halfunctions Tested: None i Discrepancies: Unknown Evaluators Later i 9 F l i .. ~.

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 ' O l Simulator Tests - Halfunction Test Abstracts TAD C - Part $8 PAGE 13 of 62 Test T.4.5.3.11, LOSS OF AC POWER (LOSS OF AUX TRANSFORMER) Revision Nunbert Later ANSI /ANS 3.5 Section 3.1.2 Plant Halfunctions, USAR tecident 15.2.6 Date Tested Not RUN Run Times Later hours i Test Description The purpose of this test is to verify the proper i response of the simulator to a Loss of Normal AC Power as described in the Perry USAR, section 15.2.6. The test is initialized at the 100% Power operating condition, IC19. The transient is initiated by inserting Failure TF01 on Aux Transformer 110-PY-D, resulting in a loss of electrical power to 13.8KV busses L12 and L12, and all non-1E (BOP) 4160V and 480V AC bucees. The sequence of events and response of key parameters is verified to occur as per the USAR description, with allowance for assumed worst case conditions in the USAR analysis. At the end of the test the Reactor is shutdown and isolated, normal i feedwater has tripped off, reactor pressure is being controlled by i SRV's, and RPV level is being maintained by the HPCS and RCIC systems, j Baseline Data used For Reference Reference Type Of f-Normal Operating Procedures ONI-R22-1 LOSS OF AN ESSENTIAL AND/OR A STUB 4.16KV DUS (UNIT 1) O Reference Type Plant Data - Analyses USAR 15.2.6 PNPP UPDATED SAFETY ANALYSIS REPORT USAR FIGURE 15.2-8 PHPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.2-12 PNPP UPDATED SAFITY ANALYSIS REPORT Halfunctions Tested: TF01 Loss of Transformer Discrepancies: Unknown Evaluators Later l l l l l l l l l O

l l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 l O simulator Tests - Halfunction Test Abstracts TAB C - Part 58 PAGE 14 of 62 f Test T.4.5.3.12, FAILURE OF RHR SHUTDOWN COOLING Revision Number Later [ ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.2.9 I Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of ti,is test is to verify the proper i response of the simulator to a Loss of Shutdown Cooling (RHR) following a loss of off-site power as described in the Forry USAR, section 15.2.9. The test is initialized at the 100% power condition, l 1C19. Initial conditions are established to represent the USAR l analysis conditions: Suppression Pool level of 18.0 ft, Suppression Pool Temperature of 90 deg F, and lake temperature of 80 dog F. One i emergency diesel generator is taken out of service. Valve IE12-F008 is f ailed with HOV component f ailure MV06 (single passive f ailure). The transient is initiated by tripping all off-site grid tie breakers, resulting in a loss of of f-site power. The immediate effects are verified per the USAR description and ONI-S11, Loss of Of f-Site Power. The plant stabilizes in a post-scram condition with the RPV isolated, i pressure being maintained by SRV's and level being maintained with the high pressure injection system HPCS and RCIC. After about 30 minutes, actions are taken to commence a forced fast cooldown using SRV's. O After 12 minutes, the Reactor Pressure has been loworod to 100 peig. Due to failure of normal shutdown cooling, an alternate (emergency) method is established per the USAR description, following directions l given in ONI-E12-2, Loss of Shutdown Cooling using LPCI injection and flowing out through open SRV's. The ability to maintain long-term fuel cooling and reduce reactor pressure below 100 peig in this lineup is verified. At the end of the test, plant conditions are as described above. Baseline Data used For Reference Reference Type Normal Operating Procedures SOI-E12 RESIDUAL HEAT REHOVAI. SYSTEM (UNIT 1) Reference Type Off-Normal Operating Procedures ONI-E12-2 LOSS OF SHUTDOWN COOLING (UNIT 1) Reference Type Plant Data - Analyses USAR 15.2.9 PNPP UPDATED SAFETY ANALYSIS REPORT Malfunctions Tested: MV06 HOV Fail as is (mechanical binding) Discrepancies: Unknown. Evaluators: Later i O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part 5B .( PAGE 15 of 62 Test T.4.5.3.13, RECIRC FLOW CONTROL FAILURE-DECREASING (BOTH FCV'S) Revision Number Later ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions, USAR Accident 15.3.2 Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to a Reactor Recirculation Flow Control Valve Closure as described in the Perry USAR, section 15.3.2. The test is initialized at 100% Power operating condition, IC19. The RRC flow control valves controllers are failed to 0% output with controller Failure CN01. The flow control valves are verified to stroke closed at their maximum rate. When the flow control valves are at their minimum position, total core flow has been reduced to about 25% of rated. The rapid decrease in power and flow results in a level excursion causing a SCRAM on high reactor level. The simulator response is verified per the USAR description. At the end of the test the plant is stabilized in the Post-SCRAM condition. Baseline Data used For Reference Reference Type Off-Normal Operating Procedures O ONI-B33-1 REACTOR RECIRCULATION FLOW CONTROL MALFUNCTION (UNIT 1) v Reference Type Plant Data - Analyses USAR 15.3.2 PNPP UPDATED SAFETY ANALYSIS REPORT USAR FIGURE 15.2-4 PNPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.3-4 PNPP UPDATED SAFETY ANALYSIS REPORT Malfunctiono Tested: CN01 Controller Auto / manual failure Discrepancies: Unknown Evaluators: Later O% U

CERTIFICATION OF PERRY SIHULATION FACILITY DOCKET NO. 50-440 ['_,) Simulator Tests - Halfunction Test Abstracts TAR C - Part SD i (,,/ PAGE 16 of 62 i Test T.4.5.3.14, RECIRCULATION PUMP SEIEURE Revision Number Later ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.3.3 Date Tested Not RUN Run Times Later hours Test Description The purpose of this test is to verify the proper response of the simulator to a Recirculation Pump Shaft Suiture as described in the Perry USAR, section 15.3.3. The test is initialized at 2004 Power, IC19. The transient is initiated by incerting Pump component f ailure CP02 on Reactor Recirc Pump A. The sudden loss of flow in the A Loop results in a vessel level excursion, causing a Reactor SCRAM due to high level. The response of key parameters is verified per the USAR description. At the end of the test, the plant in stabilized in the post-SCRAM conditions. Baseline Data used For Reference Reference Type Plant Data - Analyses USAR 15.3.3 PNPP UPDATED SAFETY ANALYSIS REPORT USAR FIGURE 15.3-5 PNPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.3-5 PNPP UPDATED SAFETY ANALYSIS REPORT ( Halfunctions Tesced: CP02 Pump Shaft seizes Discrepancies: Unknown Evaluators Later 9 l . -.~ -..

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAD C - Part SD PAGE 17 of 62 Test T.4.5.3.15, ABHORMAL STARTUP OF IDLE RECIRCULATION PUMP Revision Number Later r ANSI /ANS 3.5 Section 3.1.2 Plant Halfunctions, USAR Accident 15.4.4 Date Tested: Not RUN Run Times Later hours \\ Test Description The purpose of thin test is to verify the proper response of the simulator to an improper Recirculation Loop restoration as described in the Perry USAR, section 15.4.4. The test is initialized at 100% Power, IC19. To prepare for the transient, Reactor Recire Pump A is tripped, the A Recire Loop is isolated and 'i allowed to cool down to 100 deg F. The temperature transmitter which normally provides an interlock to prevent restoring a cold, idle loop to service is failed to read normal (hot) temperature with transmitter failure PT01. The idle loop is then unisolated, and Recire Pump A is started normally per the system operating instructions (with exception i of verifying proper delta temperatures), resulting in a reactivity excursion. F.ey parameter response is verified per the USAR description (the reactor does not SCRAM). Final plant conditions are similar to initial. Baseline Data used For Reference O Reference Type Normal Operating Procedures 501-B33 REACTOR RECIRCULATION SYSTEM Reference Type Plant Data - Analyses USAR 15.4.4 PNPP UPDATED SAFETY ANALYSIS REPORT USAR FIGURE 15.4-1 PHPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15.4-1 PNPP UPDATED SAFETY ANALYSIS REPORT Halfunctions Tested: PT91 Process Transmitter Variable failure Discrepancies: Unknown Evaluators Later s l O .,e-4.e--w~++-+-e,+-++er-,e .-.m. .+----..--,++w,m,.: +-,.m r.+.v=v--- .,,,v. -v~---c---.---me--r-me-e-,-r-s=-e,,w e.v v +f

j 4 i 1 CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part SB PAGE lo of 62 Test T.4.5.3.16, RECIRCULATION PUMP SHAFT SHEAR Revision Number Later ANSI /ANS 3.5 Section 3.1.2 Plant Malfunctions, USAR Accident 15.3.4 Date Tested: Not RUN Run Times Later houre Test Description The purpose of this test is to verify the proper response of the simulator to a R4 circulation Pump Shaft Shear as doacribed in the Perry USAR, section 15.3.4. The test is initialized at 100% Power, IC19. The transient is initiated by inserting Pump i component f ailure CP01 on Reactor Recire Pump A. The suddon loss of t flow in the A Loop results in a vessel level excursion, causing a Reactor SCRAM due to high level. The response of key parameters is verified M r the USAR description. At the end of the test, the plant is stabilized in the post-SCRAM condition. Baseline Data used For Reference Reference Type Plant Data - APAly6es USAR 15.3.4 PNPP UPDATED SAFETY ANALYSIS REPORT Malfunctions Tested: CP01 Pump Shaft breaks () Discrepancies: Unknown Evaluators Later l . O l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCrET NO. 50-440 O Dimulator Tests - Halfunction Test Abstracts TAB C - Part 5D PAGE 19 of 62 l 1 Test 'O. 4. 5. 3.17, RECiF.C FLOW CONTROL FAILURE-INCREASING (DOTH FCV'S) i Revision Number: Later ANSI /AN*, 3.5 Section: 3.1.2 Plant Malfunctions, USAR Accident 15.4.5 Date Tested het RUN Run Times Later hours Test

Description:

The purpose of this test is to verify the proper recponse of the simulator to a Reactor Recirculation Flow Control Valve Opening as described in the Perry USAR, section 15.4.5. The test is initialized at 100% Power operat.'.ng condition, IC19. To approximate the initial canditions of the Analysis, reactor power is reduced to about 54% by reducing Reactor Core flow to 33% of rated. The RRC flow control valveu Master Controller is failed to 100% output with controller Failure CN01. The flow control valves are verified to stroke open at their maximum rate. Prior to reaching their full open position, a reactor SCRAM occurs due to the power transient. The simulator responee le verified per the USAR description. At the end of the test the plant is stabilized in the Post-SCRAM condition. Baseline Data used For Roference Reference Type Off-Normal Operating Procedures ONI-B33-2 LOSS OF ONE OR BOTH RECIRCULATION PUMPS (UNIT 1) Reference Type: Plant Data - Analyse 6 USAR 15.4-5 PNPP UPDATED SAFETY ANALYSIS REPORT USAR FIGURE 15.4-5 PNPP UPDATED SAFETY ANALYSIS P ORT USAR TABLE 35.4-5 PNPP UPDAfED SAFETY ANALYSIS RLv0RT Malfunctions Tested: CN01 Controller Auto / manual failure Discrepancies: _ Unknown Evaluators Later ~% ---my-r-

CERTIFICATION OF PERRY SIMULATION FACILITY DochtT W0. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part SD PAGE 20 of 62 i Test T.4.5.3.18, CONTROL ROD DROP ACCIDENT Revision Number: Later ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions, USAR Accident 15.4.9 i Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to a Control Rod Drop accident described in the Perry USAR, section 15.4.9. The to't is initialized with a reactor startup in progress, and the reactct critical. Control Rods are adjusted to achieve worst came conditions per the USAR analysis (50% rod density). Two failures are activated on a high worth rod; RD01 to stick the red fall in, and RD02 to uncouple the rod from its drive mechanism. The Rod Pattern Controller is byposeed and the selected CRDM is fully withdrawn (rod stays full in). Failure RPOl is deleted, resulting in the rod travelling at a rapid speed from full in to full out. The resulting neutron flux transient is verified to initiate a reactor SCRAM. The final conditions approximate the initial conditions with the exception of all Control Rods being inserted due to SCRAM. Baseline Data used For Reference Reference Type Plant Data - Analyses O USAR 15.4.9 PNPP UPDATED SATETY ANALYSIS REPORT USA 3 TABLE 15.4-9 PNPF UPDATED SAFETY ANALYSIS REPORT Malfunctions Tested: RD01 STUCK CONTROL )<OD RD02 UNCOUPLED CONTROL ROD Discrepancies: Unknown Evaluators: Later O

CERTIFICATION OF PERRY SIMULATION PACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part SB l PAGE 21 of 62 Test T.4.5.3.19, INADVERTENT HIGH PRESSURE CORE SPRAY STARTUP Revision Number Later ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions, USAR Accident 15.5.1 Date Tested: Not RUN Run Times Later hours Test Descriptions. The purpose of this test is to verify the proper response of the simulator to an tradvertent High Pressure Core Spray system initiation as described in the Perry USAR, section 15.5.1. The test is initialized at 100% Power, IC19. The transiene is initiated by failing the HPCS manual initiation switch with I/O override (simuls ;es short in switch). The response of key plant parameters is verified tc respond as per the USAR description. The final plant conditions are similar to the initial with reactor power slightly higher and reac or pressure slightly lower. Baseline Data used For Referenc's Reference Type: Off-Normal Operating Procedures ONI-E12-1 INADVERTENT INITIATION OF ECCS/RCIC (UNIT 1) Reference Type Plant Data - Analyses O USAR 15.5-1 PNPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 35.5-1 PNPP UPDATED SAFETY ANALYSIS REPORT Malfunctiona Tested: None Discrepancies: Unknown Evaluators Later i i l-O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Teste - Malfunction Test Abstracts TAB C - Part SB PAGE 22 of 62 Test T.4.5.3.20, ANTICIPATED TRANSI2NT WITHOUT SCRAM (ATWS) Revision Number: Later ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions, USAR Accident 15.8 Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to a Failure of the Reactor Protection System to initiate and complete a SCRAM during a major plant transient (vessel isolation) as described in the Perry USAR, section 15.8. Initial conditions: The simulator is initialized at 100% Power, EOC. The ATWS malfunction is activated at 100% severity (passive failure); 100% severity will result in no rod motion. A transient is then induced by inserting a spurious MSIV closure (vessel isolation) as described in the USAR. Critical parameters are recorded and compared to USAR analytical data. Automatic actions and alarms are verified. Two minutes after the MSIV closure, the SLC (boron injection) pumps are started. Final Conditio;.si The reactor is shutting down due to boron injection, the control rods are full out, the RPV is isolated. SRV's are controlling Rx Pressure, augmented by RCIC operation. As the core shuts down, reactor vessel level is restored by HPCS and RCIC injection. () Baseline DRia used For Reference Reference Type Plant Data - Analyses USAR 15.8 PNPP UPDATED SAFETY ANALYSIS REPORT USER APPENDIX 15C PNPP UPDATED SAFETY ANALYSIS REPORT USAR FIGURE 15C-1 PNPP UPDATED SAFETY ANALYSIS REPORT USAR TABLE 15C-3 PNPP UPDATED SAFETY ANALYSIS REPORT Malfunctions Tested: RDIS ATWS Discrepancies: Unknown Evaluators: Later O

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SD PAGE 23 of 62 Test T.4.5.3.21, LOSS OF OFF-SITE POWER Revision Number Later ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions, USAR Accident 15.2.6 Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify the proper response of the simulator to a Loss of Off-Site Power as described in the Perry USAR, section 15.2.6. Initial Conditions: The simulator is initialized to 18% steady state power IC with the main turbine on-line. To facilitate comparison to plant Startup Test data, reactor power is increased to 23%. The transient is initiated by simultaneously tripping the main turbine and opening the off-site Supply breakers L1003 and L2003. Pertinent parameters are recorded and analyzed for comparison to Perry Unit 1 startup test data. The acceptance criteria for this test is the same criter's used for the startup test STI-R43-031. Final Conditions: the reactor is shutdown and isolated, RCIC and HPCS are maintaining RPV water level greater RPV Level 2, RCIC operation is controlling RPV pressure <1033 psig (SRV operation is allowed by the test criteria). Parameters are suf ficiently stable that operators may take action to restore power to the Plant. Baseline Data used For Reference Reference Type: Off-Nr.' mal Operating Procedures ONI-S11 LOSS OF OFF-SITE POWER Reference Type: Plant Data - Analyses USAR 15.2.6 PNPP UPDATED SAFETY ANALYSIS REPORT Reference Type Plant Data - Startup Test Results STI-R43-031/2 DATA REQUEST RESPONSE - LOSS OF TURBINE GEN AND OFFSITE POWER Malfunctions Tested: None Discrepancies: Unknown Evaluatore: Later O

CERTIFICATION OF PERRY SIMULATION FACILITl DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part SD PAGE 24 of 62 Test T.4.5.3.22, PEI MALFUNCTION SCENARIO #1 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; other Halfunctions Required to Support Operator Training Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify proper simulator response while conducting PEI (Plant Emergency Instructions) Malfunction Scenario #1. This scenario is used during the training of Perry Plant Licensed Operator candidates. The scenario guido consists of 3 parts, each designed to be run in 50 minutes. Each part is tested separately. Each part is begun with the almulator initialized at the 100% (EOC) power operating condition. Part 1 starts with RPS division 1 tripped to meet a Technical Specification LCO. APRM B is failed upscale, but a scram fails to occur. Test aporators take action por PEI-B13 to initiate and complete a reactor shutdown using ARI (alternate rod insertion). At the end of Part 1, the plant is shutdown. Part 2 begins with a loss of main condenser vacuum, forcing a rapid plant shutdown. Vacuum continues to degrade and initiates a Reactor Vessel isolation, forcing entry into PEI-B13. A cycling SRV complicates both RPV pressure and level control, and is mitigated by removing the SRV control power fuses. The ability to exercise procedures PEI-E12 and PEI-G42 is validated when they are used O to control Suppression Pool Level and Temperature. Part 3 begins by failing all Feedwater flow to the RPV. The loss is caused by a failed level switch tripping all Reactor Foodwater Booster pumps, which in turn trips all running normal Feedwater Pump Turbines. The Reactor scrams on low level and forces entry into PEI-B13. Reactor Level restoration is attempted by the use of HPCS and RCIC systems. These systems are failed, resulting in a complete loss of normal and emergency feed. A rapid reactor depressurization is performed to allow injection by low pressure ECCS. At the end of each part, the Reactor is shutdown with all control Rode inserted with plant conditions being controlled as directed by the Plant Emergency Instructions. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL Reference Type Normal Operating Procedures SOI-E22A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) Reference Type Of f-Normal Operating Procedures ONI-B21-1 SRV INADVERTUNT OPENING / STUCK OPEN (UNIT 1) Reference Type Other PEI Malfunction Scenario OT-3034-01A O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 (~ Simulator Tests - Malfunction Test Abstracts TAB C - Part 50 PAGE 25 of 62 Tent T.4.5.3.22, PEI HALFUNCT.ON SCENARIO #1 Malfunctions Tested: AD01 L-ING SRV CD01 Spurious breaker trip CB06 Breaker fails in current position (lone of control power) CN01 Controller Auto / manual failure CP02 Pump Shaft seizes ED07 LOSS OF 120V BUS MC01 CONLENSER AIR INLEAKAGE NM04 APRM OUTPUT FAILURE RD17 LOSS OF CRD PUMP LUBE OIL RYO2 Relay Fails as in Discrepancies: Unknown Evaluators Later i t [ 1 l l

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part 5B PAGE 26 of 62 Test T.4.5.3.23, PEI MALFUNCTION SCENARIO #2 Revision Number 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; other Malfunctions Required to Support Operator Training Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify proper simulator response while conducting operator training in accordance with PEI Malfunction Scenario #2. This scenario consists of three 50 minute exercises (parts). Each part is tested separately. At the beginning of each exercise, the simulator is initialized at the 100% Power operating condition, EOC. Part 1 starts with a spurious RRCS feedwater runback which results in a Reactor Scram due to low water level. While controlling reactor parameters per PEI-B13, a spurious MSIV isolation forces operators to change the mode of pressure control. Part 2 starts by failing 2 SRV's open, forcing entry into PEI-G42 and PEI-E12 to control Suppression Pool Temperature and Level. A fast reactor shutdown is performed when pool temperature limits are exceeded. A passive failure of the Reactor Protection System to initiate a SCRAM occurs, and alternate rod insertion (ARI) is used to insert control rods and complete the reactor shutdown. A failure of all high pressure injection systems occurs, resulting in the operators rapidly lowering reactor pressure to O allow low pressure ECCS injection. Part 3 begins by falling the Control Rod Drive (CRD) pump and causing CRDM HCU accumulators to discharge, forcing operators to perform a reactor shutdown. Failures in the normal RPS system and the Alternate Rod Inserticn system result in an ability to shutdown the reactor with control rods. Actions are performed in accordance with PEI-B13 to reduce Reactor core flow and RPV 3evel to lower power. A CRD pu.np is recovered and control rods are inserted by the normal drive method. In each of the above exercises, the test ends with the reactor shutdown and plant parameters being controlled as directed by the plant normal and emergency instructions. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL Reference Typer Normal Operating Procedures 10I-8 SHUTCOWN BY MANUAL REACTOR SCRAM SOI-C11 (CRT H ) CONTROL ROD DRIVE HYDRAULIC SYSTEM (UNIT 1) SOI-C71 RSS POWER SUPPLY DISTRIBUTION (UNIT 1) SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-E2T.A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E51 REACTOR CORE ISOLATION COOLING SYSTEM (UNIT 1) O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part SB PAGE 27 of 62 Test T.4.5.3.23, PEI MALFUNCTION SCENARIO #2 Reference Type Off-Normal Operating Procedures ONI-B21-1 SRV INADVERTENT OPENING / STUCK OPEN (UNIT 1) Reference Type Other PEI Malfunction Scenario OT-3034-02A Malfunctions Tested: CB01 Spurious breaker trip CB03 Breaker Auto trip logic failure CB06 Breaker fails in current position (loss of control power) HV03 MOV Spurious valve closure RDOS CONTROL ROD ACCUMULATOR FAULT RD15 ATWS RD17 LOSS OF CRD PUMP LUBE OIL RP01 EPA TRIP RP03 FAILURE OF ARI TO INITIATE RPO4 INADVERTENT RRCS FW RUNBACK, RX RECIRC DOWNSHIFT, LOW FREQ MG TRIP RV03 Relief Valve Fails open RYO2 Relay Falls as is Discrepancies: Unknown Evaluators Later O

CERTIFICATION CF PERRY SIMULATION FACILITY DOCKET NO. 50-140 O Simulator Tests - Malfunction Test Abstracts TAD C - Part SB PAGE 28 of 62 Test T.4.5.3.24, PEI MALFUNCTION SCENARIO #3 Revision Number: 00 AN!I/ INS 3.5 Section: 3.1.2 Plant Malfunctions; Other Halfunctions Required to Support Operator Training Os Not RUN Run times Later hours Test Liscription: The purpose of this test is to verify proper simulator response while conducting Licensed Operator Training in accordance with PEI.31 function Scenario #3. The scenario is divided into two parts. In each, 'he simulator is initialized at power, 50% power for the first, and 9%% power near end of life for the second. The sequence of events for tne first part of the scenario is: while operating at 50% power, two SRV's inadvertently open on their safety settings and remain stuck open. Operator actions are performed per PEI-E12 and PEI-G42 to control Suppression Pool temperature. Rising pool temperature necessitates a f ast reactor shutdown. Control rod insertion fails by RPS and ARI. The Rx is shutdown by injecting boron with SLC (Standby Liquid) and normal rod insertion. Actions are also taken to reduce power by limiting injection and lowering Reactor Level. The sequence for the second part is: While operating at 95% power and inadvertent MSIV isolation occurs and the reaccor is shutdown when ARI is initiated on high RPV pressure. Seven (g)) of all high prescure injection systems result in RPV level decreasing to concrol rods fail to insert but are later inserted using CRDH. The loss top of active fuel. At this point, the reactor is depressurized. Low prqssure ECCS systems inject to restore reactor water level. The plant is then aligned for normal shutdown cooling. At the end of each part of the test, the reactor is shutdnwn using boron in case one and rode in case two. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR NtESSURE VESSEL CONTROL Reference Type Normal Operating Procedures IOI-8 SHUTDOWN BY MANUAL REACTOR SCRAM SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-E21 LOW PRCSSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E22A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) Reference Type Off-Normal Operating Procedures ONI-B21-1 SRV INADVERTENT OPENING / STUCK OPEN (UNIT 1) Reference Type: Other PEI Malfunction Scenario OT-3034-03A O V

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SB PAGE 29 of 62 Test T.4.5.3.24, PEI MALFUNCTION SCENARIO #3 Malfunctions Tested: C801 Spurious breaker trip CB07 Breaker faile to close MV06 MOV Fall as is (mechanical binding) PT01 Process Transmitter Variable failure RD01 STUCK CONTROL ROD RD15 ATWS RP03 FAILURE OF ARI TO INITIATE RV03 Relief Valve Falls open RYOl Relay Falls de-energized RYO2 Relay Fails as is Discrepancies: Unknown Evaluators Later O O

- c. l i l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SB PAGE 30 of 62 Test T.4.5.3.25, PEI MALFUNCTION ENARIO #4 Revision Humber: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; other Malfunctions Required to Support Operator Training Date T1sted: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify proper simulator response while conducting Licensed operator Training in accordance with PEI Malfunction Scenario #4. The scenario is divided into three parts. In each, the simulator is init ialized at 100% power near end of core life. The sequence of events for :he first part of the scenario: A small break LOCA (malfunction TH02A at 14 severity) results in increasing Drywell Pressure. ECCS initiation occurs at 1.68 peig sensed drywell pressure. The leak severity is increased to cause worsening containment conditions. Actions are performed in accordance with PEI-B13 to mitigate the high containment pressures. The sequence of events for the second part of the scenario: A small break in MSL between MSIV's occurs while operating at power. A failure of the MSL guard pipe coupled with a failure of the inboard MSIV to close results in a high energy release into the containment. When drywell pressure exceed 1.68 peig, PEI-B13 and PEI-D23 are entered. The Reactor is shutdown, and systems restored to remove heat from the containment structure. The reactor is O depressurized and containment sprays are initiated to control containment pressure. The sequence of events for the third part of the scenario A loss of drywell cooling is caused by a series of equipment failures associated with the drywell cooling system. The reactor is shutdown while drywell parameters are controlled using PEI-D23-2 and PEI-D23-3. Final plant conditions: In all three above scenarios, the reactor is shutdown and conditions are improving due to actions taken in response to PEI direction. Baseline Data used For Reference Reference Type: Emergency Cperating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL PEI-D23-1 CONTAINMENT TEMPERATURE CONTROL PEI-D23-2 DRYWELL & CONTAINMENT PRESSURE CONTROL Reference Type: Normal Operating Procedures SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-P51/52 SERVICE & INSTRUMENT AIR SYSTEM Reference Type: Other PEI Malfunction Scenario OT-3034-04A O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 [ Simulator Testa - Malfunction Test Abstracts TAB C - Part SB ( PAGE 31 of 62 Test T.4.5.3.25, PEI HALFUNCTION SCENARIO #4 Malfunctions Tested: AV03 Air Ormrated valve Falle as is CB06 Breaker fails in current position (loss of control power) CP03 Pump Head loss (flow degradation) HXO2 Heat Exchanger Tube leak PC01 INCREASED DW/CNTMT BYPASS LEAKAGE TC04 BYPASS VALVE FAILURE TH02 RECIRC LOOP PIPING BREAK TH28 HAIN STEAM LINE BREAK INSIDE GUARD PIPE Discrepancies: Unknown Evaluatore: Later O O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part 58 ( PAGE 32 of 62 Test T.4.5.3.26, PEI MALFUNCTION SCENARIO #5 (PART 1) Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; Other Malfunctions Required to support Operator Training Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify proper simulator response while conducting Licensed operator Training in accordance with PEI Malfunction Scenario #5 (part 1). Initial Plant conditions the simulator is initialized at 100% power near end of core life. Sequence of events: During severe weather a loss of feedwater and station blackout occur which results in a reactor shutdown. PEI-B13 actions are taken and steam cooling is conducted. PEI-M51/56 is entered in order to mitigate the effects of hydrogen generation. The Division 1 Diesel Generator is restored following fuel oil system repairs and RPV water level is restored promptly. The containment in vented until actual hydrogen concentrations can be determined. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL PEI-M51/56 HYDROGEN CONTROL Reference Type: Normal Operating Procedures SOI-C71 RPS POWER SUPPLY DISTRIBUTION (UNIT 1) Reference Type Other PEI Malfunction Scenario OT-3034-05A Malfunctions Tested: CB01 Spurious breaker trip CB06 Breaker falls in current position (loss of control power) CN01 Controller Auto / manual failure DG03 DIESEL GEN SPEED GOVERNOR FAILS DG06 FUEL OIL DAY TANK LEAK ED05 LOSS OF 4.16 KV BUS RC04 RCIC GOVERNOR VALVE FAILURE RDIS ATWS RP01 EPA TRIP RP03 FAILURE OF ARI TO INITIATE RYO2 Relay Fails as is SLOS SLC INJECTION PIPING LEAK TF01 Loss of Transformer THIS GROSS FUEL FAILURE l l l Discrepancies: Unknown l Evaluators: Later l { l /* l l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 f) Simulator Tests - Malfunction Test Abstracts TAB C - Part 5B PACE 33 of 62 (,) Test T.4.5.3.27, EVALUATION MALFUNCTION SCENARIO #1 Revision Number: 00 ANSI /ANS 1.5 Section 3.1.2 Plant Malfunctions; Other Malfunctions Required to support Operator Training Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to validate the Simulator response during an operator training / evaluation exercise scenario. The initial test conditions are Unit operating at full power at EOL conditions (coastdown), with the B Steam Air Ejector, and the B Control Rod drive pump unavailable. A surveillance instruction (SVI-Cll-T1003) is performed which exercises control rods. A single rod is uncoupled (system level failure RD02) and then stuck (system level failure RD01). The rod is inserted and disarmed. An inadvertent HPCS Initiation signal is generated (active failure of bistable component IB21N0667C/G). The HPCS Diesel Generator is failed when the start signal is received (passive failure DG03, case C). ' failure of the A Steam Air Ejector is simulated (active failure of AOV components IN62F0140A and IN62F0170A) resulting in a loss of Condenser Vacuum. The Main Turbine trips on low vacuum. Normal operator actions are performed following a Reactor SCRAM per Perry Operating Procedures. Condenser vacuum continues to degrade, r" resulting in a full vessel isolation. The RCIC system is started to restore RPV level and control RPV pressure following the Main Steam \\- Isolation. At the end of the test, the reactor is shutdown and isolated, the turbine has been shutdown, the RCIC system is running to maintain RPV level. This test allows for verifying the response of the simulator to a multiple failure scenario which may be used during operator examination evaluations. The ability to mitigate the effects of the various failures using Perry Plant Operating Procedures is also validated. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL Reference Type Normal Operating Procedures SOI-B33 REACTOR RECIRCULATION SYSTEM Reference Type: Off-Normal Operating Procedures ONI-E12-1 INADVERTENT INITIATION OF ECCS/RCIC (UNIT 1) Reference Type: Other Evaluation Malfunction Scenario OT-3058-ES-01A l

i CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part SB PAGE 34 of 62 Test T.4.5.3.27, EVALUATION MALFUNCTION SCENARIO #1 Malfunctions Tested: AV02 Air Operated Valve rails closed BSO2 Bistable Spurious trip CD05 Breaker fails in curretit position (mechanical seizure) DG03 DIESEL GEN SPEED GOVERNOR FAILS MC01 CONDENSER AIR INLEAKAGE RD01 STUCK CONTROL ROD RD02 UNCOUPLED CONTROL ROD Discrepancies: Unknown Evaluators Later l llO l l l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 9 Simulator Tests - Malfunction Test Abstracts TAB C - Part 5B PAGE 35 of 62 Test T.4.5.J.28, EVALUATION MALFUNCTION SCENARIO #2 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; Other Malfunctions Requi ed to Support Operator Training Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify proper simulator response while conducting Fvaluation scenarios used during Initial License Operator training and Licenso Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator actions are required to properly verify simulator response. The simulator is initialized to 96% reactor power with a division 3 outage in progress and minor BOP equipment out of service. RHR loop B is in the Suppression Pool Cooling mode of operation. APRM channel A falls downscale and a half scram is inserted to meet the requirements of Technical Specifications. A sequential loss of condensate and feedwater pumps occurs causing a Reactor Scram on low reactor water level. The Motor Feed pump is initially used for level control, but trips. The Reactor Core Isolation Cooling System (RCIC) fails to automatically start, but a manual start is successful. RPV pressure is maintained by the Pressure control system and RPV level is restored prior to reaching RPV Level 1 and level is brought back to the normal operating band of 9 185 to 215" using the Plant Emergency Instructions. At the end of the scenario, the Reactor scram procedure is being used to restore plant conditions. Baseline Data used For Reference Reference Type Normal Operating Procedures 101-3 POWER CHANGES SOI-B33 REACTOR RECIRCULATION SYSTEM SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-E51 REACTOR CORE ISOLATION COOLING SYSTEM (UNIT 1) SOI-N21 CONDENSATE SYSTEM (UNIT 1) Reference Type Off-Normal Operating Procedures ONI-N27 FEEDWATER PUMP TRIP (UNIT 1) Reference Type: Other Evaluation Malfunction Scenario OT-3058-ES-02A O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SB PAGE 36 of 62 Test T.4.5.3.28, EVALUATION MALFUNOTION SCENARIO #2 Malfunctions Tested: BS01 Bistable Fails to trip CD01 Spurious breaker trip CB06 Breaker fails in current position (loss of control power) FWOO FEEDWATER PUMP LOSS OF LUBRICATING OIL FWO8 FEEDWATLR PUMP LOSS OF LUBRICATING OIL NM04 APRM OUTPUT FAILURE Discrepancies: Unknown Evaluators: Later O O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB O - Part 5B O PAGE 37 of 62 Test T.4.5.3.29, EVALUATION HALFUNCTION SCENARIO #3 Revision Number: 00 ANSI /ANS 3.5 Section 3.1.2 Plant Malfunctions; Other Malfunctions Required to Support Operator Training Date Tested: Not RUN Run Times Later hours " * ~ Test

Description:

The purpose of this test is to verify proper simulator response while conducting Evaluation Scenarios used during Initial License Operator training and Licensed Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator actions are required to properly evaluate simulator response. The simulator is initialized to 80s reactor power with one of the RFPT's out of commission. Other minor BOP equipment is out of service as well as RHR loop B. The "A" pressure regulator fails low, resulting in a transfer to the backup regulator. The Motor Feed Pump experiences a signal failure requiring entry into Off-Normal Instruction ONI-C34. A leak develops in division 1 diesel generator day tank requiring that the diesel be dcclared inoperable and placed in secured status. A Main Cire Water tube rupture occurs and a Fast Reactor Shutdown is performed. The Reactor Protection System falls to shutdown the reactor, however Alternate Rod Insertion is successful in inserting all control rods. The Reactor Core Cooling Isolation System is started for RPV level control O and Plant Emergency Instruction PEI-B13 is exited to the Reactor Scram procedure. Actions are taken to isolate the feedwater and condensate systems in accordance with ONI-N61. Bareline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL Reference Type: Normal Operating Procedures IOI-3 POWER CHANGES SOI-C34 FEEDWATER CONTROL SYSTEM (UNIT 1) SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-E51 REACTOR CORE ISOLATION COOLING SYSTEM (UNIT 1) SOI-N27 FEEDWATER SYSTEM (""IT 1) SOI-R43 DIVISION 1 & 2 DIEoEL GENERATOR SYSTEM (UNIT 1) SOI-R45 DIVISION 1 & 2 DIESEL GENERATOR FUEL OIL SYSTEM (UNIT 1) Reference Type Off-Normal Operating Procedures ONI-N61 CONDENSER TUBE LEAK / ORGANIC INTRUSION (UNIT 1) Reference Type Surveillance Procedures SVI-R10-T5217 ELECTRICAL DISTRIBUTION SYSTEM ENERGIZATION CHECK Ref?rence Type Other Evaluation Malfunction Scenario OT-3058-ES-03A G

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tosts - Malfunction Test Abstracts TAB C - Part SB PAGL 38 of 62 Test T.4.5.3.29, EVALUATION KALFUNCTION SCENARIO #3 Malfunctions Tested CB06 Breaker fails in current position (loss of control power) DG06 FUEL OIL DAY TANK LEAK FWO4 FEED PUMP LOGIC FAILURE MCO2 CONDENSER TUBE LEAK PT01 Process Transmitter Variable failure RYO2 Relay Fails as is SWO7 LOSS OF COMPONENT COOLING - TBCC Discrepancies: Unknown Evaluators Later O i L { l i l O l l l l

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Halfunction Test Abstracts TAB C - Part SB PAGE 39 of 62 Test T.4.5.3.30, EVALUATION KALFUNCTION SCENARIO #4 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; Other Halfunctions Required to Support Operator Training Date Tested: Not RUN Run Time Later hours Test

Description:

The purpose of this test is to verify proper simulator response while conducting Evaluation Scenarios used during Initial License Operator training and License Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator action is required to properly evaluate simulator response. The simulator is initialized to 100% power with minor BOP equipment out of service. A power reduction is performed to conduct a Turbine Valve Exercise Test. Control Rod accumulator problems are encountered, followed by a trip of the operating CRD Pump. The pump is successfully restarted. The Hot Surge Tank level control valve fails causing a high level and isolation of heater 4, as well as Heater 6A with subsequent entry into ONI-N36. A sequential loss of Stator Water Cooling occurs causing a turbine load set runback and a Reactor Scram on high reactor pressura. PEI-B13 is used to control RPV level and pressure. During the pressure increase SRV's open as required, but one SRV fails to reclose. Actions are taken to close the valve by removing the appropriate fuses. l The plant is stabilized and actions addressed by ONI-C71-1 (Reactor Scram) are being used at the end of the scenario. Baseline Data used For Reference Reference Type Normal Operating Proceduren SOI-Cll(CRDH) CONTROL ROD DRIVE HYDRAULIC SYSTEM (UNIT 1) Reference Type: Off-Normal Operating Procedures ONI-C71-1 REACTOR SCRAM (UNIT 1) Reference Type: Ourveillance Procedures SVI-N31-T1151 MAIN TURBINE VALVE EXERCISE TEST Reference Type: Other Evaluation Malfunction Scenario OT-3058-ES.. A Malfunctions Tested: BSO2 Bistable Spurious trip CB01 Spurious breaker trip CB06 Breaker fails in current position (loss of control power) CN02 Controller Auto failure MV02 MOV Spurious valve opening MV03 MOV Spurious valve closure RDOS CONTROL ROD ACCUMPLATOR FAULT Discrepancies: Unknown Evaluators: Later \\

. _ ~ CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part SB PAGE 40 of 62 Test T.4.5.3.31, EVALUATION MALFUNCTION SCENARIO #5 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; Other Malfunctions Required to Support OperatcV Training Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify proper simulator response while conducting Evaluation Scenarios used during Initial License training and License Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator action is required to properly evaluate simulator response. For initial conditions, the simulator is initialized to 90% power, EOC conditions. The Motor Feed Pump is out of service for repair. The suppression pool is at an elevated temperature due to leaking SRVs. ESW and ECC loops "A" are in operation to support the anticipated startup of RHR loop "A" in the supprossion' pool cooling mode of operation. The sequence of events for this test are as follows; RHR loop "A" is started up in the suppression pool cooling mode. A shaft seizure occurs on Circ Water Pump A. Condenser vacuum degrades to the point of a Turbine Load Set Hunback and a Recirc Flow Control Valve Runback. After these runbacks are reset, reactor power is increased to the maximum allowed by the existing O condenser vacuum. A spurious Recire Pump trip occurs and actions are taken to recover the tripped Recirc pump. The TBCC suction line pipe ruptures, causing a complete loss of TPCC. The temperatures of cooled components rapidly increase, and a Fast reactor Shutdown is performed. Vacuum further degrades and the MSIVs automatically close. PEI B13 is entered to stabilize the plant. RPV level is maintained by operation of RCIC and HPCS, however the HPCS Pump trips on overcurrent shortly after starting. RPV pressure is maintained by operation of RCIC and intermittent SRV operation. Final plant conditions are as follows; The reactor is shutdown and isolated with RPV level and pressure under Control. Baseline Data used For Reference Reference Type Normal Operating Procedures IOI-3 POWER CHANGES SOI-833 REACTOR RECIRCULATION SYSTEM SOI-E51 REACTOR CORE ISOLATION COOLING SYSTEM (UNIT 1) SOI-N21 CONDENSATE SYSTEM (UNIT 1) SOI-N27 FEEDWATER SYSTEM (UNIT 1) SOI-N32/39 MAIN TURBINE & TURNING GEAR SYSTEM (UNI 7 1) SOI-N64/62 OFF-GAS / CONDENSER AIR REMOVAL SYSTEM SOI-P42 EMERGENCY CLOSED COOLING SYSTEM (UNIT 1) 1 SOI-P45 EMERGENCY SERVICE WATER SYSTEM (UNIT 1) Reference Type: "ff-Normal Operating Procedures ONI-B33-2 LOSS OF ONE OR BOTH RECIRCULATION PUMPS (UNIT 1) ONI-P44 LOSS OF TURBINE BUILDING CLOSED COOLING (UNIT 1) Reference Type Surveillance Procedures O SVI-B33-Tll68 IDLE RECIRCULATION LOOP TEMPERATURE & FLOW

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 (} ditnciator Tests - Malfunction Test Abstracts TAB C - Part SB (,f PAGE 41 of 62 Test T.4.5.3.31, EVALUATION MALFUNCTION SCENARIO #5 Reference Type Other Evaluation Halfunction Scenario OT-3058-ES-05A Halfunctions Tested: CB01 Spurious breaker trip CB06 Breaker fails in current position (loss of control power) CP02 Pump Shaft seizes MC01 CONDENSER AIR INLEAKAGE MV01 MOV Fail as is (loss of control power) SWO3 TBCC SYSTEM PROCESS PIPING LEAKAGE Discrepapr ese Unknown i Evale Later ks O

i CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Malfunction Test Abstracts TAB C - Part 5B ( PAGE 42 of 62 Test T.4.5.3.32, EVALUATION MALFUNCTION SCENARIO #6 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions; Other Halfunctions Required to Support Operator Training Date Tested: Not RUN Run Time: Later hours Test

Description:

The purpose of this test is to verify proper simulato-response while conducting Evaluation Scenarios used during Initial License training and License Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator action is required to properly evaluate simulator response. For initial conditions, the simulator is initialized to 95% power, EOC conditions. One Automatic Depressurization System (ADS) valve is inoperable with its fuses removed and the Division 2 diesel generator is operating unloaded. During the test an MSIV stroke test is performed and the diessi is loaded to bus EH12. A fire in an ECCS room cooling panel occurs and the room cooling fans controlled from that panel become inoperable. A plant shutdown is started. An inadvertent division 2 initiation signtil occurs, causing, among other things, nuclear closed cooling (NCC) to be isolated to the containment and drywell. Temperatures rise to all the c:ooled components in these areas, eventua'2y forcing the equipment te be .Os manually or automatically secure' iome ECCS equipment fails to start on the initiation signal. Plant F cy Instructions are entered to restore NCC to the containment ywell. Final plant conditions are the plant shutdown with leve) e seure being controlled in the normal band in accordance with PEI-t 4 ontainment and drywell parameters improving due to actions taker. ..cordance with PEI-D23-2 and PEI-D23-3. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-D23-2 DRYWELL & CONTAINMENT PRESSURE CONTROL PEI-D23-3 DRYWELL TEMPERATURE CONTROL Reference Type Normal Operating Procedures SOI-Cll(CRDH) CONTROL ROD DRIVE HYDRAULIC SYSTEM (UNIT 1) SOI-G41(FPCC) FUEL POOL COOLING & CLEANUP SYSTEM (UNIT 1) l SOI-M15 ANNULUS EXHAUST GAS TREATMENT SYSTEM (UNIT 1) l SOI-H25/26 CONTROL ROOM HVAC & EMERGENCY RECIRCULATION SYSTEM SOI-P43 NUCLEAR CLOSED COOLING SYSTEM ( SOI-R43 DIVISION 1 & 2 DIESEL GENERATOR SYSTEM (UNIT 1) Reference Type: Off-Normal Operating Procedures ONI-B21-1 SRV INADVERTENT OPENING / STUCK OPEN (UNIT 1) 1 ONI-P43 LOSS OF NUCLEAR CLOSED COOLING (UNIT 1) l l l l

_. ~.. - _ -. CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 -O Simulator Teste - Malfunction Test Abstracts TAB C - Part 5B PAGE 43 of 62 Test T.4.5.3.32, EVALUATION KALFUNCTION SCENARIO #6 Reference Type Surveillance Procedures SVI-B21-T2001 MSIV FULL STROKE OPERABILITY TEST Reference Type Other Evaluation Malfunction Scenario OT-3058-ES-06A Malfunctions Tested: CB05 Breaker fails in current position (mechanical seizure) RYO2 Relay Falls as in Discrepancios: Unknown Evaluators Later O , U t I i l 1

i CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part 5B PAGE 44 of 62 Test T.4.5.3.33 EVALUATION MALFUNCTION SCENARIO #7 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; Other Malfunctions Required to support operator Training Date Tested: Not RUN Run Times Later hours Test Descriptions The purpose of this test is to verify proper simulator response while conducting Evaluation Scenarios used during Initial License training and License Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator action in required to properly evaluate simulator response. For initial conditions, the simulator is initialized to 96% power, EOC conditions, with minor BOP equipment out of service for repair. The sequence of events for the test are as follows; A periodic test instruction is performed of the reactor feed pump turbines. A failure of the reactor recirculation flow control valve position device causes a FCV lockup. RFPT B trips due to a failure of its lube oil system, and power is automatically run back by the operable RR FCV. Recovery of the failed RR FCV loop is accomplished and power is' increased to level allowed by the feedwater pumps available (80%). An SRV opens inadvertently and actions taken in accordance with ONI-B21-1 close the valve but it fails to fully l. reseat. PEI G42 and PEI E12 are used to cool the suppression pool and g,,' restore suppression pool level. Suppression pool temperature approaches 110 deg F and a Fast Reactor Shutdown is performed. The reactor scram procedure is entered and while the plant is being stabilized, a fire breaks out in the control room. The control room is evacuated and ONI C61 actions are completed. Final plant conditions are the reactor shutdown with level and pressure being controlled in accordance with ONI C71-1, one SRV leaking and the suppression pool slowly heating up. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-E12 SUPPRESSION POOL TEMPERATURE CONTROL Reference Type Normal Operating Procedures SOI-B33 REACTOR RECIRCULATION SYSTEM SOI-Cll(CRDil) CONTROL ROD DRIVE HYDRAULIC SYSTEM (UNIT 1) i SOI-E12 RESIDUAL llEAT REMOVAL SYSTEM (UNIT 1) SOI-N21 CONDENSATE SYSTEM (UNIT 1) SOI-P43 NUCLEAR CLOSED COOLING SYSTEM Reference Type: Of f-Normal Operating Procedures ONI-B21-1 SRV INADVERTENT OPENING / STUCK OPEN (UNIT 1) ONI-B33-1 REACTOR RECIRCULATION FLOW CONTROL MALFUNCTION (UNIT 1) ONI-C71-1 REACTOR SCRAM (UNIT 1) I' f O l [

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O_ Simulator Testa - Malfunction Test Abstracts TAB C - Part SB PAGE 45 of 62 Test T.4.5.3.33, EVALUATION MALFUNCTION SCENARIO #7 Reference Type Periodic Test Procedures PTI-N27-P0003 REACTOR FEED PUMP TURBINE STANDBY OIL PUMP OPERATION Reference Type Other Evaluation Malfunction Scenario OT-3058-ES-07A Malfunctions Tested: CB01 Spurious breaker trip CB06 Breaker fails in current position (lose of control power) PTO1 Process Transmitter Variable failure RV02 Relief Valve Stuck RVO3 Relief Valve Fails open TH14 RECIRC FCV HYDRAULIC POWER UNIT OIL HI TEMP Discrepancies: Unknown Evaluators Later O l l 1

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 [ Simulator Tests - Malfunction Test Abstracts TAB C - Part SB PAGE 46 of 62 Test T.4.5.3.34, EVALUATION MALFUNCTION SCENARIO #8 Revision Humber: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions; Other Malfunctions Required to Support Operater Training Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify proper simulator response while conducting Evaluation Scenarios used during Initial License training and License Operator Requalifiiation examinations. The scenario is designed to be run in 50 minutes and operator action is required to properly evaluate simulator response. For initial conditions, the simulator is initialized to 96% power, EOC conditions, with minor DOP equipment out of service for repair, as well as the High Pressure Core Spray System (HPCS) out of service. The sequence of events for this test are as follows: A periodic test instruction is performed on the Main Lube Oil system. A failure of the RCIC exhaust diaphragm pressure transmitters causes an inadvertent RCIC isolation to occur. A loss of an electrical bus causes RFPT A to trip, and shortly thereafter, a loss of all feedwater occurs due to a loss of all RFBPs. Plant Emergency Instruction PEI B13 is used in an attempt to restore water level, and when the last HP injection system is lost (CRDH A), the reactor is depressurized, first using the turbine bypass valves, then when level decreases to less than top of active fuel, an emergency depressurization is performed. The RPV Level 1 LOCA signal causes the containment and drywell parameters to degrade and actions are taken to restore these parameters using the containment, drywell, and suppression pool PEIs. Final plant conditions are the reactor shutdown and depressurized with RPV level restored to 185" to 215" using low pressure injection systems. Conditions in the drywell, containment, and suppression pool are improving. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL PEI-E12 SUPPRESSION POOL TEMPERATURE CONTROL Reference Type Normal Operating Procedures SOI-B33 REACTOR RECIRCULATION SYSTEM SOI-C51(IRM) INTERMEDIATE RANGE MONITORING SYSTEM (UNIT 1) SOI-C51(SRM) SOURCE RANGE MONITORING SYSTEM (UNIT 1) SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-M51/56 COMBUSTIBLE GAS CONTROL SYSTEM & HYDROGEN IGNITERS (UNIT 1) Reference Type Off-Normal Operating Procedures ONI-R23-2 LOSS OF A NON-ESSENTIAL 480 VOLT BUS (UNIT 1) Reference Type Other Evaluation Malfunction Scenario OT-3058-ES-01E O

CERTIFICAT'0H OF PERRY SIMULATION FACILITY DOCKE7 NO. 50-440 himulator Tests - H: tunction Test Abstractc TAB C - Part 58 i PAGE (? Sf 67 i t 4-Test T.4.5.3.34, EVALUATION MALFUNCTION SCENARIO #8 Halfunctions Tested Ds02 Bistable spurious trip CB01 spurious breaktr trip CD04 9toaker Auto close logic fullure ED16 LOSS OF 480V v40 TOR CONTROL CENTER (MCC) RD!7 LOSS OF CRD PflhP LUDE OIL Disc $epancies: Unknoivn Evaluators: Later i e O O v.n. ,,..,uw.- n. ,.,e.r-r-

I t CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 i O. Simulator Teste - Halfunction Test Abstracts TAD C - Part 58 l PAGE 40 of 62 l i ~-. Test T.4.5.3.35, EVALUATION MALFUNCTION SCENARIO #9 Revision Numbar 00 ANSI /ANS 3.5 Section 3.1.2 Plant Halfunctions; other Halfunctions Required to Support Operator Training Date Testeds Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verif y proper simulator i response while conducting Evaluation Scenarloo used during Initial License training and License Operator Requalification examinations. The scenario 1-designed to be ran in 50 minutes and operator actions are required to properly evaluate simulator response. The simulator is initialized to 96% power, EOC conditions. The Motor Feed Pump, as well as other minor BOP components are out of commission for repair. An l operating Hotwell Pomp strainer clogs, requiring a Jhift of HW pumps. An SRV inadvertently opens, causing entry into ONI-D21-1. Actions taken in I this Of f Normal Instruction are successful in closing the SRV (fuses to the SRV solenoid are removed). A failure of the manual initiation f6ature associated with the RCIC system cause the Reactor Core Isolation Cooling System to initiate. This initiacion causes a trip of the Main Turbine and the Reactor Feed Pump Turbines. The reactor scrams and level and low reactor level 3 is quickly reached, causing entry into PEI-B13. As the plant conditions are being stabilized, a complete lors of level O indication occurs. RPV flood is performed to ensure adeguate core cooling. Final plant conditions are the plant shutdown with all control rods inserted, with adequate core cooling assured by maintaining MRF s preesure with available low pressure and high pressure injection systems with the reactor depreocurized. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEl-D13 REACTOR PRESSURE VESSEL CONTROL Reference Type Normal Operating Procedures SOI-B33 KEACTOR RECIRCULATION SYSTEM SOI-C51(IRM) INTERMEDIATE RANGE MONITORING SYSTEM (UNIT 1) 1 SOI-C51(SRH) SOURCE RANGE HONITORING SYSTEM (UNIT 1) SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-E22A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E51 REACTOR CORE ISOLATION COOLING SYSTF.p (UNIT 1) l SOI-H51/56 COMBUSTIBLE GAS CONTROL SYSTEM & HYDROGEN IGNITERS (UNIT 1) l SOI-N21 CONDENSATE SYSTEM (UNIT 1) SOI-N27 FEEDWATER SYSTEM (UNIT 1) ~SOI-P42 EMERGENCY CLOSED COOLING SYSTEM (UNIT 1) SOI-P43 HUCLEAR CLOSED COOLING SYSTEM SOI-P45 EMERGENCY SERVICE WATER SYSTEM (UNIT 1) 1 I ( l' O r

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CERTIFICATION OF PERRY S!HULATION FACILITY DOCKET NO. 50-440 i (O Simulator Tests - Halfunction Test Abstracts TAB C - Part 5B ) PAGE 49 of 62 f Test T.4.5.3.35, EVALUATION MALFUNCTION SCENARIO #9 Reference Type Off-Normal Operating Procedures l ONI-B21-1 SRV INADVERTENT OPENING / STUCK OPEN (UNIT 1) ONI-C71-1 REACTOR SCRAM (UNIT 1) Reference Type Other Evaluation Halfunction Scenario OT-3058-ES-02E Halfunctions Tested: BS01 Bistable Tails to trip I DS02 Bistable Spurious trip CP03 Pump Head loss (flow degradation) PT01 Process Transmitter Variable failure Discrepancies: Unknown Evaluators Later 1 O s O

i i l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O i Simulator Tests - Malfunction Test Abstracts TAB C - Part 58 l PAGE 50 of 62 l Test T.4.5.3.36, EVALUATION KALTUNCTION SCENARIO #10 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; Other Halfunctions Required to Support Operator Training Date Tested: Not RUN Run Times Later hours 1 Test

Description:

The purpose of this test is to verify proper simulator t response while conducting Evaluation scenarios used during Initial License training and Licensed Operator Requalification examinations. The scenario is designed ta be run in 50 minutes and operator actions are required to properly evaluate simulator response. The simulator is initialized to 38% reactor power, ready to shift Reactor Recirculation 3 Pumps to f ast speed in order to continue power ascension. A division 3 outage is in progress and minor DOP equipment is out of service for r repair. APRM E fails upscale causing a half scram. The APRM channel is bypassed and the half scram is reset. Problems are encountered with main condenser vacuum and a power reduction is required. As vacuum continues to degrade, a full scram signal is generated but the reactor protection system fails to shut down the reactor. Actions are taken to initiate Alternate Rod Insertion but the reactor stays at power. Further actions are taken in accordance with PEI-B13, including SLC injection and manual O rod insertion. A stuck open SRV further degrades plant conditions. As power is reduced to less than 4% by the actions previously mentioned, ARI J is finally successful in inserting all control rods. Final plant conditions include actions in progresu to cool the suppression pool. l Baseline Data used For Reference Reference Typra Emergency Operating Procedures PEI-813 REACTOR PRESSURE VESSEL CONTROL Reference Type Normal Operating Procedures 201-3 POWER CHANGES IOI-8 S!!UTDOWN BY MANUAL REACTOR SCRAM SOI-E22A HIGli PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E22B DIVISION 3 DIESEL GENERATOR (UNIT 1) 1 SOI-P45 EMERGENCY SERVICE WATER SYSTEM (UNIT 1) Reference Type Off-Normal Operating Procedures ONI-B21-1 SRV INADVERTENT OPENING / STUCK OPEN (UNIT 1) ONI-N32 TURBINE AND/OR GENERATOR TRIP (UNIT 1) Reference Type Off-Normal Procedures ARI-H13-P680-5 REACTOR CONTROL (LEFT) (UNIT 1) Reference Type Other Evaluation Malfunction Scenario OT-3058-ES-03c O

l CERTIFICATION OF PERRY S!HULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Halfunction Test Abstracts TAB C - Part SE PAGE $1 of 62 l Test T.4.5.3.36, EVALUATION MALFUNCTION SCENARIO #10 Halfunctions Tested: HC01 CONDENSER AIR INLEAKAGE RD12 SCRAM OUTLET VALVE LEAK RD15 ATWS RD18 SDV DRAIN BLOCKAGE RV03 Relief Valve Fails open RYO2 Relay Fails as is Discrepancies: Unknown Eva'uators !,at e r O P O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part $B t PAGE 52 of 62 Test T.4.5.3.37, EVALUATION MALFUNCTION SCENARIO #11 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; Other Malfunctions Required to support Operator Training Date Tested: Not RUN Run Times Later hours Test Description The purpobe of this test is to verify proper simulator 6 response while conducting Evaluation Scenarios used during Initial License training and License Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator action is required to properly evaluate simulator response. For initial conditions, the simulator is initialized to 96% power, EOC conditions, with minor DOP equipment out of service for repair, as well as the HPCS Pump motor. One control rod HCU has low accumulator pressure. The sequence of events for this test are as follows: A stroke of the Main Steam Line drain valves is performed. The operating CRDH Pump trips and the alternate pump is not available. Other withdrawn control rods experience accumulator f aults, and reactor power is decreased to perform a manual scram from 50% power. The reactor protection system and alternate rod insertion fail to insert control rods. Actions are taken in accordance with PEI B13 to lower reactor power. Both trains of O Standby Liquid Control are started, however the SLC B suction valve fails to open preventing SLC Pump B f rom starting. Water level is maintained with normal feedwater systems and the main turbine and bypass valves are available for pressure control. Efforts to restore SLC B suction valve are successful by stroking the valve at the MCC. Final plant conditions are reactor power in the source range and decreasing, all rods out, both SLC trains tripped on low storage tank level, with RPV level and pre,esure under control. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL Reference Type: Normal Operating Procedures 201-8 SHUTDOWN BY MANUAL REACTOR SCRAM SOI-B33 REACTOR RECIRCULATION SYSTEM SOI-E22A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-N11 MAIN & REHEAT STEAM SYSTEM (UNIT 1) SOI-N27 FEEDWATER SYSTEM (UNIT 1) SOI-N33 STEAM SEAL SYSTEM (UNIT 1) SOI-N34 MAIN LUBE OIL SYSTEM (UNIT 1) SOI-R10 PLANT ELECTRICAL SYSTEM Reference Type: Surveillance Procedures SVI-B21-T2006 MAIN STEAM DRAIN ISOLATION VALVE EXERCISE & STROKE TIME TEST Reference Type Other Evaluation Malfunction Scenario OT-3058-ES-04E O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Halfunction Test Abstracts TAB C - Part 5D PAGE $3 of 62 Test T.4.5.3.37, EVALUATION MALFUNCTION SCENARIO #11 Malfunctions Tested: MV04 HOV Failure of auto open circuit PT01 Process Transmitter Variable failuro RDOS CONTROL ROD ACCUMULATOR FAULT RD15 ATWS RD27 LOSS OF CRD PUMP LUBE OIL RP03 FAILURE OF ARI TO INITIATE RYO2 Relay Fails as is Discrepancies: Unknown Evaluators Later O i r O ,__..,...._.,.m..__,,_

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CERTIFICATION OF PERRY S!HULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Halfunction Test Abstracts TAB C - Part 5B PAGE 54 of 62 Test T.4.5.3.38, EVALUATION MALFUNCTION SCENARIO #12 Revision Humber: 00 ANSI /ANS 3.5 Sectlon 3.1.2 Plant Haltunctions; Other Halfunctions Required to Support Operator Training Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test is to verify proper simulator response while conducting Evaluation Scenarios used during Initial Licenee training and License Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator action is required to properly evaluate simulator response. For initial conditions, the simulator is initialized to 96% power, EOC conditions, with minor BOP equipment out of service for repair, as well as the HPCS Pump motor. The "A" pressure regulator is out of service. The sequence of events for this tset are as follows: A surveillance is performed on the Turbine control System. A failure of the temperature control valve associated with the lead subloop for A RR flow control valve occurs, causing an automatic subloop transfer. The standby subloop starts but fails to restore temperature resulting in a FCV lockup. When a spurious Reactor feedwater pump turbine trip occurs, An automatic Flow control valve runback occurs on B RR loop, A RR loop valve position does not O change. The "B" pressure regulator fails downscale, causing all TCVs and BPVs to close. This results in a reactor scram. RPV level is brought under control by operation of the Motor Feed Pump; RPV pressure is controlled by operation of safety relief valves and/or bypass valves using the bypass valve jack. When conditions have stabilized, a non-isolable main steam line rupture occurs in the drywell, causing a rapid decrease in RPV pressure and a rapid degradation of containment, drywell, and suppression pool parameters. Plant Emergency Instruction actione are taken in accordance with PEI's D23-1,2,3, G42, E12, and B13. The Low Pressure Core Spray Pump trips after receiving an automatic start signal. Final plant conditions are as followe; adequate core cooling is assured by either submergence cooling or maintaining reactor pressure 120# greater than containment pressure. Conditions in the containment and drywell are improving due to actions addressed by the PEIs. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL PEI-D23-1 CONTAINMENT TEMPERATURE CONTROL PEI-D23-2 DRYWELL & CONTAINMENT PRESSURE CONTROL PEI-D23-3 DRYWELL TEMPERATURE CONTROL PEI-E12 SUPPRESSION POOL TEMPERATURE CONTROL PEI-042 SUPPRESSION POOL LEVEL CONTROL O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part SB PAGE 55 of 62 Test T.4.5.3.38, EVALUATION MALFUNCTION SCENARIO #12 Reference Types Normal Operating Procedures SOI-833 REACTOR RECIRCULATION SYSTEM SOI-C11(CRDH) CONTROL ROD DRIVE HYDRAULIC SYSTEM (UNIT 1) SOI-E22A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-P43 NUCLEAR CLOSED COOLING SYSTEM Reference Type Periodic Test Procedures PTI-N32-P0001 TURD OVERSPEED PROT DEVICES TRIP & EHC/ TURD LUDE O!L PUMP STARTS / STATOR WATER PUMP START & ROTATIONS WEEKLY TEST Reference Type Other Evaluation Halfunction Scenario OT-3058-ES-05E Halfunctions Tested: CB01 Spurious breaker trip PT03 Process Transmitter Variable output clamp TH14 RECIRC FCV HYDRAULIC POWER UNIT OIL HI TEMP TH26 MAIN STEAM LINE RUPTURE INSIDE DRYWELL Discrepancies: Unknown Evaluators Later O (::)

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Halfunction Test Abstracts TAD C - Part 5B PAGE 56 of 62 Test T.4.5.3.39, EVALUATION HALTUNCTION SCENARIO #13 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions; Other Halfunctions Required to support Operator Training Date Tested: Not RUN Run Times Later hours Test Descriptions The purpose of this test is to verif y proper simulator response while conducting Evaluation Scenarios _used during Initial License training and License Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator action is required to properly evaluate simulator response. For initial conditions, the simulator is initialized to 40% reactor power, with a plant startup in progress. Two APRM channels are inoperable and a half scram signal has been inserted. The sequence of events for this test are as follows: A CROH Pump shift is perforened. After the shift, a failure of two HPCS bistables causes an inadvertent HPCS pump start. The division 3 diesel generator fails to start. The bistables are replaced and the HPCS and Division 3 D/G are returned to standby readiness. When the plant is in a stable condition, Turbine Building Closed Cooling pump A seizes, necessitating starting the standby pump. The common signal to the two RFPT controllers from the master level controller fails, causing the speed of both turbines to increase to their high speed stop. Reactor O level rapidly increases to the high level trip setpoint, but the reactor fails to scram. Alternate Rod Insertion (ARI) is successful in partial rod insertion. The Reactor Recirc Pumps are tripped, further lowering power to about 14%. Reactor level is stabilized at 185" by the use of the Motor Fe#d Pump on the Startup Level Controller. After bypassing the Low Power setpoint (LPSI'), manual rod insertion is succcesful in decreasing power to less than 12%. At that point, the scram and ARI can be reset and the next several manual ARI signals are successful in inserting all control rods. Final plant conditior.s are the reactor shutdown with RPV level and preocure under control and PEI B13 ready to be exited to ONI C71-1. Baseline Data used For Reference Reference Type: Emergency Operating Procedures P"I-D13 REACTOR PRESSURE VESSEL CONTROL Reference Type Plant Data - Analyses USAR TABLE 15.5-1 PNPP UPDATED SAFETY ANALYSIS REPORT Reference Type Other Evaluation Halfunction Scenario OT-3058-ES-06E O

l CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Tent Abstracts TAB C - Part 5B i PAGE 57 of 62 Test T.4.5.3.39, EVALUATION MALFUNCTION SCENARIO #13 Halfunctions Tested: BS02 Distable Spurious trip CB01 Spurious breaker trip CN01 Controller Auto / manual failure CP02 Pump Shaft seises DG03 DIESEL GEN SPEED GOVERNOR FAILS NH04 APRM OUTPUT FAILURE RD15 ATWS RD10 SDV DRAIN BLOCKAGE RYO2 Relay Falls as is Discrepancies: Unknown Evaluators Later O 4 l O .,..,,-,----_...---,,_,-,,.,....r_--..,,

-__.m.__.___-___-__.._______ CERTIFICATION OP PERRY SIMULATION FACILITY DOCKET NO. 50-440 ) O Simulator Teuts - Halfunction Test Abstracts TAB C - Part SB i PAGE 58 of 62 i Test T.4.5.3.40, EVALUATION MALFUNCTION SCENARIO #14 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Halfunctions; Other Halfunctions Required to Support Operator Training Date Tested: Not RUN Run Times Later hours 'est

Description:

The purpose of this test is to verify proper simulator responoe while conducting Evaluation Scenarios used during Initial License training and Licenae Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator action is required to properly evaluate simulator response. For initial conditions, the simulator is initialized to full power, EOC conditions. Hinor DOP equipment is out of service for repair as well as the HPCS Pump Hotor. The "B" narrow range level detector is out of service and equalized. The sequence of events for this test are as follows: A bypass valve suryt:111ance is performed. During performance, an SRV cycles open and closed several times until ito control power fuses are removed. RHR loep A is placed in the suppression pool cooling mode of operation in ordt:r to decrease suppression pool temperature. An RHR pipe break ocenre just downstream of the SP suction valve, and the valve cannot be closo4. As suppression peol level continues to decrease out the pipe break and into the RHR toom and Aux Building (room watertight ( door jammed ope s), efforts are attempted to raise SP level with the normal fill mode, the RCIC system on minimum flow, and the suppression pool makeup system. At a level of 12.2 feet in the pool, the reactor la manually scrammed and an Emergency Depressurization is performed in accordance with Plant Emergency Instructions PEI B13 and PEI G42. Finsi plant' conditions are as follows: the reactor is shutdown with RPV level maintain 3d by high pressure and/or low presaure systems, 8 SRVs are open with the reactor depressurized, and suppression pool temperature is about 145 degrees with level decreasing. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL PEI-G42 SUPPRESSION POOL LEVEL CONTROL Reference Type Normal Operating Proceda we SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-E22A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E51 REACTOR CORE ISOLATION COOLING SYSTEM (UNIT 1) SOI-G43 SUPPRESSION POOL MAKEUP SYSTEM (UNIT 1) SOI-P43 NUCLEAR CLOSED COOLING SYSTEM Raference Type: Off-Normal Operating Procedures ONI-B21-1 SRV INADVERTENT OPENING / STUCK OPEN (UNIT 1) O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 Simulator Tests - Halfunction Test Abstracts TAB C - Part SB PAGE 59 of 62 Test T.4.5.3.40, EVALUATION MALFUNCTION SCENARIO #14 Reference Types Surveillance Procedures SVI-C05-71314 TURBINE BYPASS VALVE OPERABILITY TEST Reference Type: Other Evaluation Malfunction Scenario OT-3058-ES-07E Halfunctions Tested: AD01 CYCLING SRV MV06 MOV Fail as is (mechanical binding) PC04 SUPPRESSION POOL LEAK PT01 Process Transmitter Variable failure R1102 RHR SYSTEM PIPE BREAK Discrepancies: Unknown Evaluators Later O 4 1 0

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET No. 50-440 O Simulator Tests - Malfunction Test Abstracts TAB C - Part 5B PAGE 60 of 62 Test T.4.5.3.41, EVALUATION MALFUNCTION SCENARIO #15 Revision Number: 00 ANSI /ANS 3.5 Section: 3.1.2 Plant Malfunctions; Other Halfunctions Required to support Operator Training 4 Date Tested: Not RUN Run Times Later hours Test

Description:

The purpose of this test in to verify proper simulator response while conducting Evaluation Scenarios used during Initial License training and License Operator Requalification examinations. The scenario is designed to be run in 50 minutes and operator action is required to properly evaluate simulator response. For initial conditions, tho simulator is initialized to End of Core CJastdown conditions at the 80% rod line. Hinor BOP equipment is out of service for repair as well as the HPCS Pump motor. The suppression pool is at an elevated temperature due to leaking SRVs. ASW and ECC loops "D" are in operation to support the anticipated startup of RHR loop "B" in the suppression pool cooling mode of operation. The sequence of events for this test are as follows: RHR loop "B" is started up in the suppression pool cooling mode. The local division 2 remote shutdown with is bumped, causing a trip of the RHR pump and drainage of the high elevation portions of the piping. After a fill and vent is performed, RHR loop B is returned to stsndby readiness. A failute of the level transmitter for O HSR 1A drain tank occurs. Reactor power is rapidly reduced in accordance l with IOI-14 in an attempt to reduce power to less taan the scram setpoint ~ for a TCV/TSV closure. The attempt is unsuccessful and a scram signal is generateo when the Main Turblie trips due to high MSR level. The reactor falle to cerkm and PEI B13 is entered in order to shutdown the reactor and stabilize the plant. ARI is unsuccessful in inserting control rods. RPV levei is maintained by operation of normal feedwater and preesure is maintained by operation of the bypass valves and SRVs. The Standby Liquid control system is initiated. Manual control rod insertion with i CRDH is successful after bypassing the low power setpoint. Final plant conditions are as follows: Reactor power is less than 4% and decreasing due to operation of SLC and manual rod insertion. All SFNs are closed i and supprossion pool temperaturo is decreasing. Normal reattor water level is being maintained. Baseline Data used For Reference Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL Reference Type Norma) Operating Procedures 101-14 FAST UNLOAD & TRIP OF MAIN TURBINE SOI-833 REACTOR RECIRCULATION SYSTEM SOI-C41 STANDBY LIQUID CONTROL SYSTEH (UNIT 1) SOI-E12 RESIDUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-P42 EMERGENCY CLOSED COOLING SYSTEM (UNIT 1) SOI-P45 EMERGENCY SERVICE WATER SYSTEM (UNIT 1) O

I CERTIFICATION OF PERRY S!HULATION FACILITY DOCKET NO. 50-440 [ Simulator Tests - Halfunction Test Abstracts TAB C - Part SB \\ Pact 61 of 62 Test T.4.5.3.41, EVALUATION HALFUNCTION SCENARIO #15 Reference Type Other tvaluation Halfunction Scenario OT-3050-ES-00E Halfunctions Tested: CD01 Spurious breaker trip PT01 Process Transmitter Variable failure RD15 ATWS Discrepancies: Unknown Evaluators: Later O U an.-

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Simulator Tests - Halfunction Test Abstracts TAB C - Part SD PAGE 62 of 62 Test T.4.5.3.42, PEI KALFUNCTION SCENARIO #5 (PART 2) Revision Number: 00 ANSI /ANS 3.5 Section 3.1.2 Plant Halfunctions; Other Halfunctions Required to Support Operator Training Date Tested: Not RUN Run Times Later hours Test Dancription: The purpose of this test is to verify proper simulator reoponse while conducting Licensed Operator Training in accordance with PEI Halfunction Scenario #5 (part 2). Initial Plant Conditions: the simulator is initialized at 100% power near end of core life. Sequence of events: A complete loss of RPV level instrumentation and failure to scram causes an entry into PEI-B13. With no indication of level, the reactor is depressurized and flooded. SLC (boron injection) is initiated, but a failure in that system results in the reactor remaining at power. Conditions at end of test: The reactor remains at power, actions are being taken to keep reactor power as low as practical. RPV level is unknown but adequate core cooling is being maintained. Plant recovery is not performed during this scenario. Bacoline Data used For Reference ( ( Reference Type: Emergency Operating Procedures PEI-B13 REACTOR PRESSURE VESSEL CONTROL i Reference Type Normal Operating Procedures SOI-C51(IRH) INTERHEDIATE RANGE MONITORING SYSTEM (UNIT 1) SOI-C71 RPS POWER SUPPLY DISTRIBUTION (UNIT 1) SOI-E12 RESIOUAL HEAT REMOVAL SYSTEM (UNIT 1) SOI-E21 LOW PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-E22A HIGH PRESSURE CORE SPRAY SYSTEM (UNIT 1) SOI-H51/56 COMBUSTIBLE GAS CONTROL SYSTEM & HYDROGEN IGr4ITERS (UNIT 1) Reference Type Other PEI Halfunction Scenario OT-3034-05A Halfunctions Teated: AV02 Air Operated Valve Fails closed BS01 Bistable Fails to trip BS02 Biotable Spurious trip PT01 Process Transmitter Variable failure RD15 AT*9 RYO2 Relay Falls as is SLOS SLC INJECTION PIPING LEAK Diseropancies: Unknown Evaluators: Later O 1 1

L i CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 O Four Year Test Schedule TAB D PAGE 1 of 5 Four Year Test schedule Below is a list of the performance tests which will be conducted annually. Pages 2 through 5 list (by year) the tests that will be conducted in years 1 i through 4 after filing Form 474. These schedules represent performance of approximately 25% of total tests required for certification per year. Any deletions or changes to the following schedule will require refiling of Form 474 as wl11 changes to the scope or ANSI /ANS 3.5 acceptance criteria related to these tests. 1.0 hFNUAL TESTING A. Computer Real Time Test 1. T.2.7.1 - Spare Time verification B. Steady State Performance Tests 1. T.4.4.1.1 - 25% Power Heat Balance 2. T.4.4.1.2 - 50% Power Heat Balance 3. T.4.4.1.3 - 75% Power Heat Balance 4. T.4.4.1.4 - 100% Power Heat Balance 5. T.4.4.1.5 - 100% Power Stability Test C. Denchmark Transient Testa 1. T.4.4.2.01 - Manual Scram 2. T.4.4.2.02 - Simultane.ius Trip of All Feedwater Pumps 3. T.4.4.2.03 - Simultaneous closare of All MSIV's 4. T.4.4.2.04 - Simultaneous Trip of All Recirc Pumpo 5. T.4.4.2.05 - Single Recirc Pump Trip 6. T.4.4.2.06 - Main Turbine Trip without Reactor Scram 7. T.4.4.2.07 - Maximum Rate Power Ramp (200% - 75% - 1004) 8. T.4.4.2.08 - Maximum Size LOCA with Loss of Offsite Power 9. T.4.4.2.09 - Maximum Size Unisolable Main Steam Line Rupture 10. T.4.4.2.10 - Simultaneous closure of MSIV's with Single Stuck open Safety / Relief Valve l

CERTIFICATION OF PERRY SIMULATION PACILITY DOCKET NO. 50-440 (s) Four Year Test Schedule tau D x PAGE 2 of 5 Four Year Test Schedule (continued) 2.0 PERIODIC TESTINQ A. Year 1 (June 29, 1991 through June 28, 1992) 1. Core Performance Tests

a. T.4.2.1 - Reactor Core Xenon Transient Test 2.

Normal Plant Evolutions i

a. T.4.3.3 - Cold shutdown To Reactor Critical

b. T.4.3.2 - Reactor critical To Turbino Synchronized 3.

Transient Tests

a. T.4.5.1.02.AN01 - Annunciator Input Optical Isolator Failure

b. T.4.5.1.06.ED17 - Loss of 125V DC Distribution Panel

c. T.4 5.1.12.IA01 - Air Receiver Leak

d. T.4.5.1.17.NH01 - SRM Detector (pre-amp) Failure

e. T.4.5.1.18.OG03 - Off Gas System Leak Upstream Adsorbers

f. T.4.5.1.22.RH02 - Residual Heat Removal System Pipe Break

g. T.4.5.1.27.TH01 - Recirc Loop Rupture (DBA LOCA)

/

h. T.4,5.1.27.TH27 - Main Steam Line Rupture In Steam Tunnel b) 3 4.

Halfunction Scenarios

a. T.4.5.3.01 - Loss of Feedwater Heating (RRFC In Auto)

b. T.4.5.3.02 - Loss of Feedwater Heating (RRFC In Hanual)

c. T.4.5.3.03 - Feodwater Controller Failure-Haximum Domand

d. T.4.5.3.04 - Pressure Regulator Failure-open

e. T.4.5.3.22 - PEI Halfunction Scenario #1

f. T.4.5.3.23 - PEI Halfunction Scenario #2

g. T.4.5.3.24 - PEI Halfunction Scenario #3

h. T.4.5.3.25 - PEI Halfunction Scenario #4 t

O

CERTIFICATION OF PERRY SIMULATION FACILITY DOCKET NO. 50-440 { Four Year Test Schedule TAB D PAGE 3 of 5 Four Year Test schedule (continued) D. Year 2 (June 29, 1992 through June 28, 1993) 1. Core Performance Tests

a. T.4.2.2 - Core Flux Distribution Test

b. T.4.2.5 - Core Thermal Performance Test 2.

Normal Plant Evolutions

a. T.4.3.3 - Power Increase To 100% Power 3.

Transient Tests

a. T.4.5.1.04.CUO3 - Reactor Water Cleanup System Pipe Break outside containment (Steam Tunnel)

b. T.4.5.1.07.EG01 - Main Generator Lockout Relay Trip

c. T.4.5.1.12.IA02 - Instrument Air Line Leak

d. T.4. 5.1.17.NH02 - IRH Detector (pre-amp) Failure

e. T.4.5.1.18.OG04 - Off Gas System Leak Downstream Adsorbers

f. T.4.5.1.23.RP02 - Inadvertent Initiation of Alternate Rod Insertion

g. T.4.5.1.27.TH19 - RPV Level Inst Reference Leg Break

(

h. T.4.5.1.28.TUO1 - Hain Shaft Oil Pump Degradation 4.

Halfunction Scenarios

a. T.4.5.3.05 - Inadvertent Safety Relief valve Opening

b. T.4.5.3.05 - Inadvertent RHR Shutdown Cooling Operation

c. T.4.5.3.07 - Pressure Regulator Failure-closed

d. T.4.5.3.08 - Generator Load Reject With Bypass Valves

o. T.4.5.3.09 - Generator Load Reject Without Bypass Valyce

f. T.4.5.3.10 - Turbino Trip

g. T.4.5.3.11 - Loss of AC Power (Loss of Aux Transformer)

h. T.4.5.3.26 - PEI Halfunction Scenario #5 (part 1)

1. T.4.5.3.27 - Evaluation Halfunction Scenario #1

j. T.4.5.3.28 - Evaluation Halfunction Scenario #2

k. T.4.5.3.29 - Evaluation Halfunction Scenario #3

1. T.4.5.3.30 - Evaluation Halfunction Scenario #4

m. T.4.5.3.31 - Evaluation Halfunction Scenario #5

(~) s_ , ~ _

CERTIFICATION OF PEPRY SIHULATION FACILITY DOCKET Ho. 50-440 (-- Four Year Test Schedu.e TAB D PAGE 4 of 5 Four Year Test Schedule (continued) C. Year 3 (June 29, 1993 through June 28, 1994) 1. Core Performance Tests

a. T.4.2.3 - Core Thermal Power vs. Recirc Flow Test

b. T.4.2.4 - Core Flux Response to Rod Hovement 2.

Normal Plant Evolutions

a. T.4.3.4 - Power Decrease to Turbine / Generator Unloaded 3.

Transient Tests

a. T.4.5.1.06.ED06 - Loss of 480V Bue

b. T.4.5.1.09.FWO2 - Feodwater System Pipe Break Inside Drywell

c. T.4.5.1.15.MC01 - Condenser Air Inleakage

d. T.4.5.1.17.NH03 - LPRH Detector Failure

o. T.4.5.1.21.RD03 - Control Rod Drift - In

f. T.4.5.1.25.SWO1 - Nuclear Closed cooling System Process Piping Leakage

g. T.4.5.1.27.TH2O - RPV Level Inst Variable Leg Break 4.

Halfunction Scenarios

a. T.4.5.3.12 - Failure of RHR Ehutdown Cooling

b. 1.4.5.3.13 - Recire Flow control Failure-decreasing (both Flow Control Valves)

c. T,4.5.3.14 - Recirculation Pump Seizure

d. T.4.5.3.25 - Abnormal Startup of Idle Recirculation Pump

o. T.4.5.3.16 - Recirculation Pump Shaft Shear

f. T.4.5.3.32 - Evaluation Halfunction Scenario #6 q, T.4.5.3.33 - Evaluation Malfunction Scenario #7

h. T.4.5.3.34 - Evaluation Malfunction Scenario #8

1. T.4.5.3.35 - Evaluation Malfunction Scenario #9

j. T.4.5.3.36 - Evaluation Halfunction Scenario #10

~~ > l l i r

-~s CERTIFICATION OF PERRY SIHULATION FACILITY DOCKET NO. 50-440 ( ) Four Year Test Schedule TAB D x' PAGE 5 of 5 Four Year Test Schedule (continued) D. Xgar_1 (June 29, 1994 through June 28, 1995) 1. Core Performance Tests

a. T.4.2.6 - Core Suberitical Hultiplication Test

b. T.4.2.7 - Reactor Core Life Test

c. T.4.2.8 - Shutdown Margin Demonstration 2,

Normal Plant Evolutions

a. T.4.3.5 - Plant Cooldown to Cold Shutdown 3.

Transient Tests

a. T.4.5.1.06.ED09 - Loss of 125V DC Bus

b. T.4.5.1.09.FWO3 - Feedwater System Pipe Break Outside Containment

c. T.4.5.1.16.HS11 - Steam Seal Header Pressure Regulator Failure

d. T.4.5.1.17.NH10 - Neutron Monitoring Detector Drive Stuck

e. T.4.5.1.21.RD04 - Control Rod Drift - Out

f. T.4.5.1.25.SWO2 - Service Water System Process Piping Leakage

('~}

g. T.4.5.1.27.TH21 - Power / Flow Instabilities (IEB 88-07

(_,/ Supplement 1) 4. Halfunction Scenarios

a. T.4.5.3.17 - Recirc Flow Control Failure-increasing (both Flow Control Valves)

b. T.4.5.3.18 - Control Rod Drop Accident

c. T.4.5.3.19 - Inadvertent High Pressure Cc re Spray Startup

d. T.4.5.3.20 - Anticipated Transient Without scram (ATWS) e.

T.4.5.3.21 - Lors of Off-Site Power

f. T.4.5.3.31 - Evaluatior Halfunction Scenario #11

g. T.4.5.3.38 - Evaluation Halfunction Scenario #12

h. T.4.5.3.39 - Evaluation Halfunction Scenario #13

1. T.4.5.3.40 - Evaluation Halfunction Scenario #14

j. T.4.5.3.41 - Evaluation Halfunction Scenario #15

k. T.4.5.3.42 - PEI Halfunction Scenario #5 (part 2) 4}}

ML20077G626

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