ML20029C346
| ML20029C346 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 03/31/1991 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20029C344 | List: |
| References | |
| NUDOCS 9103270194 | |
| Download: ML20029C346 (378) | |
Text
{{#Wiki_filter:. __ _ _ __ _ . . _ . . _ _ _ _ .__ _ _ - _ __ _ _ _ _ _ ._ _ _ _ _ __ 1 l s
.v2.C soKW als CDees!6sion -
a r" U't U.S. soucil Am SIMULATION FACILITY CERTIFICATION '.%AtGVtA INSTRUCTIONS.
*ade affe, emitied submittai Th,5 fofo,e it to be ided for matial cert fication. recertification bl re'3uiredt. and for any changea to a timus tion f acihty performance tetting p tuco a olan 8,ovida the f ollowing enicfmat,on and Check the 800PC0f tett Don to ind,catt ,9340n for tu0m,ttal 8404UfY DOGalf %weta Grand Gulf Nuclear Station LiCENili S 416 DAf4 Enterov Onerations. Inc. 3[ir /91 ,
...i...........,,...e...~-,,.-.,,c,...,,,,e.,
e
. . . . . .= s.. .= s .i i. =.. . *ir. o.
.e a.e, o., .. s i.
. e. . .c .~ . . io c . . . . .,
Naut so, orne,,over.ocerec,u AND n,0CA fiON QF SiWut, A flON F ACiLITY Grand Gulf Nuclear Station, Port Gibson, MS
<<,-or-,e, e ,em eenoose - rae -.o r,. pea, -,a ,ae .,v e,,a,, c.,,,,,
W siou k A rioN . Acetirv et a s o a ANC: vist A est = A cts A rT Acul o e.,o , O t Sc a ir f ios os ei n p o pw ANc t f t $ f LNG COMPL E T t o 4 A rreen .ao,rione, onerisi .a nee.s *y. saa .oraviersae are* os.c >ar.oa ee,av rear,aueiri See Appendices VI, VII, Vill, and IX. J .54Wuta K sm, ene rne f 80N o.,nsno aciut ne,eyno 9t ,A.,
.ponwe.r ANct
. hee.,a ,f 45IING 6 SCHtovL6 a f T ACM80 s ee. a* cow # er sporo.,,*e,*'y .'3 % of o*r'or*a,re rara ne yee, nor rae ,oi,, , , o ,,oo Ot SCRie fiON OF Fla f onwaNct it$7 NG TO 8 8 CONoucit o #A rtwa .co,r,eae, pepe,es es a.co e,v. eae ,osar.fr rae ntem oe.rs.oren o.my conemu,o, See Attached.
l 91 a 90Ru ANCi iiST!NG 9LAN CH ANG E Ifor any moortwarcon to a oe,*o,mente Intons o,a, suom, sten are e ore eur teos,1,eet.ons O t 5C A'*t roN os e t a p opu ANC E ri stiNo st AN CH ANGE (d treca ears,reones seve,si as a.cnnery, o.,o eoenrety the reai o cr ps,or, es,av coari,iuses Not applicable at this time. l I .a l.c,,t,,e .rne
, noonse.nc,A t, ion ro,e, . . , ,u,,r.ae
, , . s.c
. . , , .w.n,,
, u.s,on, re.m.. orrun enu..,rs.o.o e,o,me,e.,r e.na,rmente rn,,nq ue ure,ance en to Con 4 ss sancte
.-..,,,,ee,,,
Not applicable at this time. 4..,,,.....o.,..,.....o~,,m..,.e,,.,w.,e.,-a,.....e...,u...o......e..,-.,,..e, me...,,...... . . . . . . . , , . . ,
, ,,,,,, , uno.. ...,,, .. , m .n . ,m., - . i .. m m.,
SIGNATURE . AUTHORIZE 0 REPR A TIVE fifkt
" # f0Af(
e/,'7/h/ W.' T. Cottle Vice President, Nuclear Operations in accoraance witn to CF R i SS 5. Commun cat,ons, inis form snan ce suom,etea to tne NRC as tonows. / BY MAIL ADORESSEO 70 Diceetor. office of Nucteet Rosetoe Regulation BY DELIVE AY IN PE ASON U S.Nucleer Regulatory Commm.on TO THE NRC C,F FICE AT- 792o Norfolk Avenue 9103270194 910331 ' PDR ADOCK 05000416 R PDR I
Entergy Operations, Inc. l Grand Gulf Nuc ear Station i Docket Number 50-416 l License Number NPF-29 ! i Simulator Cer:ification i i F Marc 1 1991 : l l i I
?
i s i 4 1 -r , , . - - . - -- , , , - - , ,,m... .. ,,4m ..,, s.m.. ,m.-..~, , . . . . . +,....w,....+.--. - - ~ - , . . - , . . . . , ,- . . - -. . ~ . .
' GRAND GULF NUC1 EAR STATION ;
SJMULATOILCERTIFICATION REPORT , TABLE;QE_QONTENTS Section Pace I. GGNS SIMULATOR 1 A. General Information .1 B. Instructor Station 10-C. Simulator Design Data. 14 G. Configuration Management 15 II. EXCEPTIONS TO ANSI /A14S 3.5 - 1985 16-A. General 16 B. Major Exceptions -17 III. SIMULATOR TESTS AND. TESTING SCHEDULE 20 A. Simulator Tests - General Descriptions 20 L B. Testing Schedule'. 24 1 IV. SIMULATOR UPGRADE PROJECT , A. Project Scope -24 B. Project Schedule 2 5 . :- - V.-APPENDICES Appendix I - Simulator 'nitial Conditions: Appendix II - Simulator Malfunctions Appendix III - _ Simulator RemotoEFunctions' - Appendix IV:- Simulator Instrument Overrides
-Appendix V - Simulator Trainee Proficiency Review Parameters--
Appendix-VI - Simulator Test-Abstract's Computer-RealiTime- l Tests
-Appendix VII - Simulator Test ~ Abstracts - Steady 1 State'and-Normal Operations Tests >
Appendix VIII - Simulator Test Abstracts > Transient., Tests Appendix IX - Simulator Test Abstracts:-LMalfunction; Tests-1
i 1 l Grand Gulf Nuclear Station Simu'.ator Certification Initial Report, March 1991 Page 1 of 25 l l I. GGNS Simulator A. General Information
- 1) General
- a. The owners of Grand Gulf Nuclear Station (GGNS) and the simulator are System. Energy Resources, Inc. and South Mississippi Electric Power Association. The operator'of '
GGNS and the simulator is Entergy Operations, Inc. The simulator vendor was ~S-3 Technologies' (previously Singer Link) of Columbia, Maryland,
- b. The reference plant is Grand Gulf Nuclear Station-Unit-1 located near Port Gibson, Mississippi. Grand Gulf Nuclear Station is . a General Electric - Boiling Water Reactor, rated at-3833. megawatts-thermal. The plant is a BWR-6 with a Mark III- Containment. The turbine ~
generator is a Kraftwerk Union design rated at 1306 MWo. (GGNS Unit 2 was not completed and is'unow being cancelled; the simul'v'r as discussed herein references - Unit 1).
- c. The simulator waslavailable for training in 1982. ,
- d. This is Entergy Operation's initial report for the GGNS simulator. ,
- 2) Control Room Front Panels;
- a. The simulated control room. replicates the reference plant control room to the greatest extent possible.EExceptions-and-differences are discussed below.
- b. The-following front panels are simulated:
H13-P680, OperatorJControl. Console H13-P870, Auxiliary Control Benchboard Panel-H13-P858, Isolation Valve Status Panel H13-P854, Plant Control VerticallBoard. H13-P827, SPDS Console (located at CRO work station)' H13-P864, Diesel Generator Benchboard' Panel
-H13-P601, Reactor Core Cooling Benchboard Panel i
e r , - - - > -
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Page 2 of_25 . r a H13-P807, Auxiliary Electrical Control Benchboard H13-P866, BOP Monitoring-Console (Located at-the Shift Superintendent Workstation)
- c. The front panels.are layed out in an arrangement and '
orientation similar to the reference plant control room. Differences in panel arrangements and dimensions do not detract from training. Indicators, meters, labels, and l controls are similar to the reference plant control room. They duplicate the size, shape, color, and configuration ;
-of the functionally simulated hardware of the reference r
- plant so as not to detract from-training.
- d. A panel review has been conducted of all simulator panels (front and back) . Photographs of the control . room-panels were compared to-thefsimulator panels. Differences in labeling, and arrangement, etc. were noted .cn1 the ;
photographs. Differences which have not been corrected ' i during the panel reviews will be documented on simulator ; discrepancy reports. These'are reviewed for training i , impact and dispositioned accordingly. All items are-l tracked until resolution 11s accomplished. i ( 3) Control Room'Back-Panelg
- a. In the simulated-backLpanel area,-there are currently two panels which contain_ a mix of the indicators, meters, and controls important to training licensedioperators,
- b. Panel H13-P642, Division .2 Leak Detection: Vertical Board !
panel-is_ fully simulated but:is not currently oriented- ; in the same relative location as'in the reference plant. During the cilaulator upgrade project,. this panel will be-reoriented to a position closer-to11ts-correct relative location (the remote shutdown panels currently.' occupy _ this exact _ location and; will remain- as' presently.
' located).
- c. The other simulator back panel'contains portions of'the -
. following back-panels from'the reference _ plant: _ j i) H13-P604 and H13-P600, Process . Radiation : Recorder Vertical Board Panels' >
- 11) H13-P822, Turbine i Supervisory Recorder Vertical-
-Board Panel: ,
r s -
, ,_, ~,-n.
Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 Page 3 of-25 lii) H13-P614, NSSS Temperature Recorder Vertical Board Panel iv) H13-P878, Division 4 ESF Logic Vertical Board Panel v) H13-P A9, Jet = Pump Instrumentation Panel vi) H13-P669, Neutron and ' Radiation Monitoring _ Panel Division I vil) H13-P844, Area Radiation Monitor Cabinet
- d. Details of tho above simulator back-' panels are as-follows: '
- 1) The following radiation monitors from panel H13-P604 are included:
D17-RITS-R601B, Offgas Post Treatment -Radiation Monitor-Channel B D17-RITS-R601A, Offgas Post Treatment Radiation Monitor Channel A D17-RITS-R607, CCW Radiation Monitor The following recorders and controls from H13-P600 are-included: D17-RR-R605,. Auxiliary Building Puel Handling Area Exhaust Radiation Recorder Monitor A/D Selector Switch-for D17-R605 ' Monitor B/C Selector-Switch for.D17-R605 D17-RR-R600, .Offgas/Radwaste, Building / Containment l Ventilation Radiation ~ Recorder Offgas and Radwaste Building Ventilation-l Purge On/Off' switch Containment Ventilation-Purge On/Off Switch D17-RR-R601, Offgas Post Treatment. Radiation Recorder Offgas Post Treatment: Purge On/Off switch
.. i 5 -. , , --
s -- -
Grand Gulf Nuclear 8tation Simulator Certification Initial Report, March 1991 Page 4 of 25 D17-RR-R604, Offgas Pre Treatment Radiation Recorder Offgas Pretreatment Purge On/Off switch D17-RR-R606, Auxiliary Building Fuel Handling Area Pool Exhaust Radiation Recorder Me' itor A/D Selector Switch for D17-R606 Mt tor B/ C Selector Switch for-D17-R606 D17-RR-R608, Control Room Ventilation Radiation Recorder Monitor A/D Selector Switch for D17-R608 Monitor-B/C Selector Switch for D17-R608 D17-RR-R602, SSW Effluent Radiation Recorder D17-RR-R607, Fuel Handing Area-Vent / Turbine 4 Building Vent Radiation Recorder Turbine Building Vent Purge On/Of f Switch Fuel Handling Area Vent Purge On/Off-Switch . , 11) All of the. turbine supervisory recorders from panel i H13-P822 are included as follows: N32-TJR-R615, Turbine Casing Metal Temperature Recorder N32-YJR-R638, Turbine Shaft Expansion and Vibration Recorder N32-TJR-R619, Turbine Bearing Metal Temperature Recorder N32-YJR-R637, Turbine. Relative Expansion Recorder N32-YJR-R639, Turbine Shaft Expansion and
- Vibration Recorder N32-TJR-R620, , Turbine Bearing Metal Temperature Recorder These recorders are arranged such that there is no negative impact to training, iii) All of the recorders from panel H13-P614 are
. . ~ . . . -
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Page 5 of 25 included as follows: l B33-TJR-R601, Recirculation Pump / Motor A and B l Temperature Recorder B21-TJR-R614, ADS Safety Relief Valve Temperature Recorder B33-TR-R604, Recirculation A and B~ Water Temperature Recorder B21-TR-R643, Reactor Vessel Temparature Monitoring. Recorder These recorders are arranged such that there is no negative impact-to training, iv) The following recorders from panel H13-P878 are included: l N21-UR-R618A, Reactor Feed ~ Pump Turbine A Eccentricity and Vibration Recorder N21-TR-R619A, Reactor Feed Pump Turbine A Lube. Oil and Bearing Temperature Recorder N21-UR-R618B, Reactor- Feed- Pump Turbine- B Eccentricity and Vibration Recorder N21-TR-R619B, Reactor Feed Pump Turbine'B-Lube Oil and Bearing Temperature Recorder These recorders are arranged such that_there is no negative impact to training. Seal Steam Generator-Controls, and Rosemount Trip Units 'from P878 are not simulated since they are not required for simulator training, v) All jet' pump instrumentation meters-from H13-P619 are included on the.end of the simulated backpanel. These indicators- will remain in their current location. vi) The following nuclear instrumentationnis included from panel H13-P669: 1 SRM Channel A-IRM Channel A
l. Grand Gulf Nuclear Station simulator Certification -! Initial Report, March 1991 i Page 5 of 25 l APRM Channel A , Main Steam Line Radiation Monitor A Power Supply Meter Offgas Radiation Monitor APRM A Status Lights- (includes associated LPRM status lights) i LPRM cards associated with APRM A are not simulated, l i Controls are simulated 'by instructor remote functions. All LPRM detectors and their status can l be determined from the RC&IS section of panel--H13-P680. All other indicators and controls from panel H13-P669 are not included since they are not required i for operator training. Panel H13-P669' is located in the upper control cabinet area in the reference plant. As part of,the simulatorLupgrade project, j the Nuclear Instrumentation c hannels will be converted to channels normally in the - back' panel area of the reference plant, 1.e. those associated with panel -H13-P670 or H13-P672, Neutron and -1 Radiation Monitoring Panel for Division 2 or 4.- This will enhance fidelity. vii) The following indicators.and recorders from-panel-H13-P844 are-included:- 1 All H13-P844 annunciators Area Radiation Monitoring (ARM) Modules: D21-RITS-K603, RCIC Room ARM-D21-RITS-K605,'TIP' Mechanism Area ARM D21-RITS-K609, CRD Hydraulic Units North ARM y D21-RITS-K629, Containment Fuel Area South ARM D21-RITS-K635, Reactor Feed Pump Area ARM D21-RITS-K636, Turbine Building Operating Floor ARM
i Grand Gult Nuclear Station simulator certification
- Initial Report, March 1991 Page 7 of 25 The' above ARM modules are sufficient to conduct required - training on the operation of the modules. i Both ARM chart recorders from the reference plant ,
are included in the simulator, with simulated input on all recorder points (24 points eac%): D21-R600A Area Rad Monitor Recorder (24 point).. D21-R600B Area Rad Monitor Recorder (24 point) This combination- of modules and recorders is sufficient.to conduct all required training in the operation of the' Area Radiation Monitoring system. A 'SPING/AXM terminal will be added into the simulator during the simulator upgrade project. This terminal . will be located in' an area more easily accessible. for . training than its - control room' counterpart.-
- 4) Remote Shutdown Panel Roomg
- a. The Remote Shutdown Panel rooms are incorporatedLinto
- .the rear area of the-simulator.'They are located in a different area in the reference plant-butlare. included in the simulator to enhance training.. The Coi Abort Switch is not included at the entrance to the-cimulated area. The environment present in. these rooms i.e.
lighting, access, book . and key locker,_ floors, walls, etc. are not simulated to the'same extent as-is present in the reference plant.- The rooms contain panels H13 ' P150 and H13-P151 to provide training _ capability _on i shutdown from the remote shutdown panels.
- b. The panels simulated provide-the controls and indicators
. used by licensed'. operators to perform a. safe -- shutdown outside of the. control room.
t c.- Alternate shutdown panels-outside of the remote shutdown' rooms in the reference plant are not required for the training of licensed operators and are not
-includod.
- 5) Other Features
- a. The BOP and NSSS computer is currently under a long term upgrade pro]ect in'the reference-plant. Displays, 4
terminals, software and hardware- changes will be incorporated into the simulator after the final reference-
. , , ,-,--n, - - , n- , -
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Page 8 of 25 plant upgrade is completed. This will be accomplished via normal simulator configuration control. The simulator presently has capability for performance of OD-3, 7 and 8 NSSS edits. There are two BOP computer stations available on panel P680 as in the reference plant and a third BOP computer station - at the Shift Superintendent work. station. These capabilities are sufficient until the computer systems which are being upgraded in the plant can be incorporated into the simulator.
- b. Emergency Procedure (EP)_ flowchart attachments which are located in the control room emergency. locker are located in a file cabinet in the simulator. This has proven to be adequate for training since implementation of Revision 4 of the BWROG Emergency Procedure Guidolines,
- c. The personal computer and associated printers used for protective tagging and access to the SIMS network are included,
- d. The Shif t Supervisor work station is not included in the simulator.since this area is' normally_ used to conduct administrative duties and is-not required for simulator training. Reference materials'normally located-at this work station are available. Communications equipment from this work station are -redundant to that. on other simulated work stations,
- e. Simulator Support Engineering work _ stations and five book lockers for s i m u l a t o r -- s u p p l i e s _. w h i c h . a r e n o t in - the reference plant control _ room are .. locoted in outlying areas of the simulator -so as ' to preclude impact on-training.
- f. A set of controlled drawings is included'in the simulator but are not located in fire proof horizontal drawing cabinets similar to the reference plant. Vertical-stick files located.in an outlying area of the simulator have been found to be acceptable _for training..
- g. Termination panels which are part of the reference plant control room and upper control cabinet area are not included in the simulator since they have no indicators or_ controls used by operators, h.-The upper control cabinet area present in the reference plant is-not incorporated. A review of these panels by l
l l
Grand Gulf Nuclear Station simulator:certificationi Initial Report, March 1991-Page-9 of 35 the simulator configuration review board found it unnecessary to-include them for training effectiveness,
- i. The two access control gates between panels H13-P807-and H13-P601-and-between panels H13-P855 and H13-P870-in the reference plant control room are not incorporated in,the simulator. Access = control into the reference - plant control room, 1.e. security : key _ card _ system is not-incorporated. The security' camera console located behind the STA work station in-the reference plant control room is not incorporated. The entrance'and exit doors:in the' simulator are not the same as in : the : reference- plant control room. These.are not required for: training,
- j. The GGNS Vistor center, Energy Central, located in.the-
-ESC building is visible through the: viewing gallery; windows.- -During training. sessionswhere this may _ be -
distracting, blinds can be lowered to avert interference. A similar viewing gallery is-part of the reference plant..
~
- k. The status board -behind the' Shift = Superintendent's workstation has.been relocated--to allow viewing:of_the simulated' control room-from the.. instructor' console. The-instructor console, TV cameras,= microphones ' and video equipment:used to record exams- and training 1 sessions, simulator.halon and fire protection system, . are not part of- the- reference- plant- but - are included for training 'and -
fire protection'1 purposes. The y simulatorT panels __ : ha_ve emergencyLpower off pushbuttons: located'on':their-access doors, which are not incorporated'on. control room panels-in the reference plant. These differencesidonnot impact
' training,
- l. The ' Control > Room Operator, -- Shif t Superintendent, and Shift Technical Advisor work stations are- included in the simulated control room. - The following communications equipment have been included;to = simulate reference plant.-
communications equipment:- Public Address System
- Radio SystemV(The radio console' is not included,- radio:
handsets are included.') Sound Powered-Phones and; extension reels-(Sound powered = l phones are used to' simulate communications with the. local . diesel ' room. - There are minor- differences between the circuits in the simulator'and -the plant which do not impact-training.) 7\ / /
Grand Gulf Nuclear 8tation simulator certification Initial Report, March 1991 Page 10 of 25 Fire and Evacuation Alarm NRC Emergency Notification System (Red Phone) Jackson and Pine Bluff Dispatcher Phones Operational Hot Line Phone (for notification to State and Local agencies) Talk-a-Phone Intercom Commercial Telephone Lines (only one line is incorporated)
- m. At least one controlled copy of the following reference materials are available in the simulator:
Technical Specifications and Operating License Plant Operations Manual (all volumes present in the reference plant control room) EP Flowcharts (laminated copies) and Attachments (controlled sets of jumper kits are not included in the simulator) Electrical Schematics, Logic Drawings and P& ids, and System Flow Drawings UFSAR and Current Cycle Safety Analysis System Descriptions
- n. The security and fire computer terminal at the CRO workstation is not functional but is included for appearance only. This difference has no impact on training.
- o. Noises associated with plant events including turbine trip and SRV lift are not audible in the control room.
Noise from a simulated control room HVAC system will not " be included since the system is not simulated. Noises from diesel generator operation and main turbine operation on the turbine deck are only audible through communications equipment and will not be added. These differences have no impact on training. B. Instructor Station
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991
'Page 11 of 25
- 1) An instructor station is provided in the simulator which is located behind the Shift Superintendent workstation.
The console is elevated above these areas to permit-viewing of the main simulator control area through tinted windows. Blinds are provided to-reduce instructor / student' eye contact during. training. or exam sessions, which reduces the potential for cueing or prompting. 2)- All types of communication equipment used in the
~
simulator are alse provided in the instructor station.
- 3) Existing video and audio recording equipment provide the capability for recording overviews of the activities at the panels during exam or, training sessions.-Tapes can be played back for critiquing training sessions.
- 4) A personal computer is incorporated for monitoring phone calls made by the operators. This PC alerts- the instructor to the number that . was called and the-corresponding plant location.
- 5) The instructor station will be-sound-proofed during the simulator upgrade project. ,
- 6) An= instructor console is provided for setting and-changing simulated initial-conditions, manipulation of remote functions, entering and deleting . malfunctions, and changing instrument overrides.
~
Provision is made'for the following features:
- a. Emergency power.off switch
- b. Acknowledging and resetting. computer faults.
- c. Resetting the instructoriconsole.
- d. Resetting BOP computer screens
- e. Resetting SPDS computer screens
- f. Resetting TDF computer screens
- g. Placing annunciator operation on a backup computer.
- h. PeformingLDaily Operational Readiness Test (Simulator maintenance checkc)
Grand Gulf Nuclear Station l Simulator Certification Initial Report, March 1991 Page 12 of 25
- 1. Reset
- j. Switch check Override,
- k. Freeze
- 1. Snapshot
- m. Backtrack with automatic or manual reverse or forward' stepping. 3
- n. Slow Time (X 1/8)
- o. Decay Heat,. Xenon, and Containment Hydrogen Fast Time (X 10)
- p. Malfunction Clear
- q. Malfunction Tableaux Inactive
- r. Annut.ciator Sound Inoperative
- s. . Annunciator Silence (Master)
- t. Annunciator Acknowledge (Master)
- u. _ Annunciator Reset (Master) -
- v. Instrument Noise Off l
- w. Recorder-Power Off i
- x. Malfunction Increase / Decrease (for variable
-severity malfunctions, up to 15 total) 1 l y. Operations Limits Exceeded-indicator l light r Non-Real-time indicator _ light
- 2. Alarm sound Inoperative (for silencing Operations Limits _ exceeded or Non-Real~
Time detected annunciator) l l
- 7) An .nstructor remote control Edevice is provided -which can remotely- freeze and reset simulation, and can activate pre-inserted malfunctions.
- 8) If simulator _ operation is detected to-not be in real 3 time, an indicator lignts:and a " school-bell" sounds to
.- . . . .. . _ ~, -,
Grand Gulf Nuolear Station Simulator certification Initial Report, March 1991 Page 13 of 25 alert the instructor of this limitation.
- 9) Provision is incorporated for monitoring simulator operating limits. Certain limits cause a "hard freeze".
The simulator must be reset or reinitialized before continuing after these occur. Certain other limits cause a " soft freeze". These cause the simulator to " freeze". Simulation may then continue by simply taking the simulator "out of freeze" 'as controlled-- by _ the instructor.
- a. Provision is made on the instructor console monitor to determine which simulator operating limit was exceeded. The following parameters are monitored:
Suppression Pool Temperature >212*F - Hard Freeze containment Temperature >185'F - Soft Freeze Drywell Temperature >330*F - Soft Freeze RPV Pressure >1563 psig - Hard Freeze Containment Pressure >15 psig - Soft Freeze. Drywell Pressure >30 psig - Soft Freeze
- b. Administrative means are provided to alert instructors- of other conditions which are- not automatically monitored. only one such' condition is currently controlled in this manner. After an injection of Standby Liquid: Ccntrol is terminated, the isolation of Reactor Nater Cleanup System can be reset and Reactor Waterocleanup can be restartedi This' scenario is allowed by electrical interlocks in both the simulator and in the reference plant.
The simulator code has been simplified such-that if this unlikely scenario were to occur, boron removkr by RWCU operation would_ not occur . cs it should. Plant procedures prohibi'. operation in' this manner. Instructors are informad of- this condition to preclude negative training.
- 10) Two CRT displays and a-keyboard are used to control the simulator. remote functions, initial entry yage, initial
- conditions, trainee proficiency- review, instrument l overrides, and malfunctions.
- a. The initial conditions (ICs) which are .tvailable _are
)
i Grand Gulf Nuclear Station simulator certification Initial Report,-March 1991 Page 14 of 25 listed in Appendix I. Provision is made for creating up to 10 additional temporary ICs. Up to fifteen malfunctions can be entered by the instructor at a-time. These can be activated and cleared on a timer. The severity of variable malfunctions can be controlled with increase / decrease pushbuttons on the instructor console.
- b. The malfunctions which are available for training are listed in Appendix II.
- c. Remote functions are provided for controlling functions not accessible from the simulated control panels. These are 1.isted in Appendix III.
- d. Instrument overrides are provided for . overriding certain instrument and handswitch functions. These are listed in Appendix IV.
- e. Certain parameters which can be monitored automatically by the simulator are listed in Appendix V.
C. Simulator Desian Data
- 1) Final Desian Specification
~
A final design specification docume/at.is maintained for each simulator software module which contains the following where applicable:
- a. System design data - lists -design drawings including P& ids, process diagrams, schematic-diagrams, design specification sheets, performance data, system data documents, instrumentation drawings, etc., the data source, and . identification control numbers which were used by the simulator vendor to l develop the software for that module.
l I
- b. Design asoumptions - specifies assumptions used, rationale, and-justification that were used to develop the software code by the simulator vendor,
- c. Design simplifications - Specifies assumptions -
and modeling simplifications used to develop the software code by the simulator vendor.
r
- s Grand Gulf Ruclear Station simulator certification Initial Report, March 1991 Page 15 of 25
- d. Parameter contro) Monitor (PCM) monitored parameters.
- e. Process computer monitored parameters,
- f. Simulation system description - brief description of the system function, flow paths, etc.
- g. Software interface diagrams.
- h. Math model block diagrams.
- 2) System Desian Data Base This is a data base which lists design data documents, data source, and document numbers.
D. G2Dfiguration Manacement
- 3) Sihiulator DiscreDancy ReDort and Software Procram
- a. Training Section Procedure 14-S-01-22 controls the handling and processing of simulator discrepancy reports and resulting modifications to software and hardware.
M
- b. Simulator Discrepancy Reports (DRs) are used to document all potential changes, modifications, or repairs to the simulator.
i i
- c. DRs are prioritized and assigned to'the simulator !
support group to resolve. Corrective actions are checked by a retest. Affected tests are rerun as i determined necussary during the DR closeout. Software changes are reviewed for approval, and software listings are annotated on their cover pages , for affected DRs and software revision numbers,
- d. Training loads with affected software changes are released for training after being- checked and approved.
- e. DRs are entered and tracked to completion in a-DR database.
- 2) Eipulator Desian Chances i
- a. Simulator modification and design is controlled by l
.i
i d J
! Grand Gulf Nuclear Ctation simulator certification Initial Report, March 1991 2
Page 14 of 35 Nuclear Plant Enginearing (NPE) Desk Top Procedure EDP-007.
- b. Modifications which have been incorporated into the reference plant are screened and reviewed for simulator applicability.
i c. If applicable, a Simulator Design Modification Package, (SDMP) is prepared to document changes to
- the simulator,
- d. Applicable hardware and software. changes are-controlled under Training Section Procedure 14-S-01-22 when a DR is generated to track the change to resolution. This controls applicable ratests and reviews of changes made to the simulator hardware and software.
II. Exceotions to ANS/ ANSI 3.5 - 1985
- 4 A. G.eneral
- 1) A simulator configuration review board _has reviewed the potential exceptions to ANS 3.5 - 1985 identified in this initial certification report. The committee is composed of Operations Department management representatives, l Training Section representatives, and representatives from the Simulator Support' staff. Potential exceptions to ANS 3.5 - 1985 were identified and documented during i simulator certification testing and fidelity reviews.
l This included the followings l A statement of the problem, , The impact on the training of licensed operators, l Justification for continuing training with the exception . as:is, and v Recommended. corrective action.
- 2) The review board reviewed each. exception. issue.
-The rajor. issues'id6ntified included:-
The: scops - of > simulated o panels ' versus panels 'in the-g reference pin.it contro1~ room,
. Major deficiencies noted during simulator certification.-
testing, and i
- - . . , , w : . . --, =. - . -.;. u-; . . ~ . - - . . - - . - . - . - . , - . _ . . . - - . . . . . . . - ~.
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Page 17 of 25 Major simulator and control room differences.
- 3) The committee reached a concensus for the recommended action for each issue based upon the impact on training and coat benefit. The recommendations of the review board formed the scope of the simulator upgrade project.
B. Maior EKtgotions
- 1) The list that follows contains the major exceptions to ANS 3.5 - 1985. A discussion concerning each issue is provided. For panels which are not included within the present scope of simulation or the scope of the simulator upgrade project, the committee recommended that training or examinations could be conducted using job performance measures in the reference plant control room as an appropriate alternative to training in the simulator.
- a. The actions taken as required by the following Emergency Procedure (EP) flow chart Attachments will not be simulated for the indicated reason:
i) EP-2 Attachment 25, Injection into the RPV with Condensate Transfer.
- 11) EP-2 Attachment 26, Injection into the RPV with the Fire Protection Water System.
iii) EP-2 Attachment 27, Injection into the RPV with the SLC Test Tank. Performance of these EP attachments would only be required in extreme cases. The plant conditions requiring use of these attachments would require the following: 1 Loss of all Emergency Core Cooling pumps which includes High Pressure Core Spray, Low Pressure Core Spray and Low Pressure Coolant Injection A, D, and C, and Loss of Reactor Core Isolation Cooling (RCIC), and Loss of both Control Rod Drive Hydraulic Pumps A and B, and Loss of Standby Liquid Control pumps A and B (Note: These pumps would be required for Attachment 27), and Loss of all condensate pumps, and
- _ - - . -. - . -.- - - - - - - -_--. ~ . - - - -
Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 Page 18 of 25 Power still available to supply condensate transfer (For Attachment 25 only). Loss of all these sources simultaneously and power being available to the remaining alternate souress would be extremely unlikely. The simulator configuration review board agreed that the need for these attachments would be expected to occur in excess of the time for an exam scenario duration and there was no practical training-bennfit for incorporation of these attacQments. Acceptable alternatives to training include job performance measures and plant walkthroughs.
- b. The following systems will not be simulated due to the=
Simulator Configuration Review Board determining a limited training benefit for licensed operators:
- 1) Control Room HVAC system.
- 11) Fire Protection Water, Halon, and CO systems, iii) Seismic Monitoring system.
iv) Security and Firo Protection computer systems. v) GETARS computer system.
- c. The following control room panels will not'be' included as part of the simulated control room due to the Simulator Configuration Review Board determining a limited training benefit for licensed operators:
i) H13-P855, Control Room Vent-Panel i
- 11) H13-P654 and P655, Livision.I and II MSIV Leakage Control Panels lii) RCIC Overspeed Test Device Panel iv) H13-P607, TIP Control and. Monitoring Instrument Panel v) _ H13-P625, HPCS Relay Panel vi) H13-P6E1, P652, P653 Rod Control and Information System Panel _,
vii) H13-P634, Recirculation Flow Control Panel
% - - e,, -u.-,n- -v. ,
e ,- -,--..e.ww . ,~ . . , - w- -
~ ,
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Page 19 of 25 viii) H13-P862, Fire Protection Panel 1x) All Halon Fire Protection System cabinets x) H13-P856, Seismic Instrument Cabinet xi) H13-P842, KVAC Control Panel xii) H13-P871 and P872, Division 1 and 2 ESF Logic Panels l xiii) H13-P618, Division 2 RHR B and C Relay Panel ' xiv) H13-P629, Division 1 LPCS and RHR A Relay Pel.el xv) H13-P877, Division 3 ESF Logic Panel xvi) H13-P612, Feedwater and Recirculation Panel xvii) H13-P613, NSSS Process Instrument Panel xviii) H13-P843, BOP Process Instrument Panel ' xix) H13-P849, BOP Logic Panel xx) H13-P630, Remote Annunciator Logic Cabinet xxi) H13-P851, Annunciator Logic Cabinet i xxii) H13-P852, Remote Annunciator Electronic. Cabinet xxiii) H13-P622, Inboard Valve Relay Panel t xxiv) H13-P623, outboard Valve Relay Panel xxv) H13-P628, ADS Channel A Relay Panel xxvi) H13-P801, Generator and Transformer Protection. Panel xxvii) H13-P808, ESF Transformer 11 Protection Panel xxviii) H13-P810, ESF Transformer 12 = and BOP Transformer Protection Panel xxix) H13-P811, ESF Transformer 21 Protection' Panel xxx) H13-P639, Division 1 Leak. Detection cabinet i
Grand Gulf Nuolear Station simulator Certification Initial Report, March 1991 Page 20 of 25 xxxi) H13-P670, P671, and P672, Neutron and Radiation Monitoring Cabinets for Division 2, 3, and 4 xxxii) H13-P691, P692, P693 and P694, h?S Division 1,2,3, and 4 Cabinets xxxiii) H13-P621, RCIC Relay Panel Alternative training methods may be used -to train- or examine operatora on any operations on these panels including job performance measures and plant walkthroughs. !
- d. The current simulator lighting fixtures are not the same as incorporated into the control room. The Simulator-Configuration Review Board determined that similar lighting fixtures would provide very little training benefit to licensed operators at unnecessary cost.
Simulator lighting levels comparable to the Control Room during station blackout conditions can and will be used for training when appropriate.
- e. The reactor mode switch from simulator panel H13-P680 has been modified to preclude loss, breakage and removal of the key, but is otherwise fully functional. The reactor mode switch key in the reference plant is removable.
This difference has posed no. impact on training. III. Simulator Teste and Testina Schedule A. Simulator Tests - General Descriotion
- 1) Comouter Real-Time Tests- 4
- a. This test is performed for the simulator master computer Central Processing Unit (CPU), the slave computer Central Processing Unit, and slave computer Internal Processing Unit (IPU).
- b. A program is - run _ which establishes non-real-time conditions in the execution of the sof tware modules.
Indications of non-real-time operation are verified. This includes a " School Bell" alarm and the non-real time status light on-the instructor console. A software variable- for master,. slave, and. IPU slippage is. verified to have incremented properly for the test. l
'l 4
I Grand Gulf Nuclear Station simulator certification
- Initial Report, March 1991 Page 21 of 25
- c. Abstracts of these tests are included in Appendix VI.
- 2) Sleadv State and Normal Ooerations Tggj;.g +
- a. Steady State tests were conducted at 100%, 60%, and 20%
power.
- b. Critical simulator parameters such as reactor power, total feed flow, reactor pressure, generator output, and recirculation drive flows are verified to be within 2%
of plant reference parameters.
- c. Non-critical simulator parameters are verified to be within 10% of reference plant parameters.
- d. Additional- parameters are monitored . and recorded for reference.
- e. Plots are made of critical, non-critical and additional-i parameters.
- f. Incremental Heat Rate Test Data gathered in the reference' plant was used for comparison where available.
- g. A one hour stability test was run at 100% power;using data gathered during the 100% steady state test.
- h. Abstracts of these tests are included in Appendix VII.
t
- 3) Transient Tests-(
-a. Tests were performed to verify correct- parameter responses and automatic . actions occurred for selected transients,
- b. Where applicable, acceptance criteriaL from plant startup test procedures were used.:
- c. Reference data was obtained from either plant startup data, RETRAN analysis, or UFSAR transient data.
- d. Plots'were made of all-critical parameters and reference data where available for evaluation. i e'. Abstracts of these tests are included in Appendix VIII..
f.-The following transient tests were_ performed: I-l' l
-. _, , s _... . - . . , _ . _ _ _ _ .. _ -. . . , . - _ - - , . -
Grand Gu\f Nuclear station simulator certification Initial Report, March 1991 Page at of 25 TT1A Reactor Scram from rated power TTlB Reactor Scram from 58% power TT2 Loss of normal FeedWater TT3 MSIV Isolation at power TT4 Double Recirculacion pump trip TTSA Single Recirculation pump trip TT5B Single Recirculation pump seizure TT6 Turbine trip with bypass valves operable TT7 Decrease then increase in the Recirculation flow TT8A Maximum Recirculation line break (100%) in the Drywell TT8B Small Recirculation line break (0.09 f t8 ) in the drywell TT9 Main Steam line break inside the drywell TT10 MSIV Isolation with a Ltuck open relief l valva l TT11 Stuck open SRV while at power TT12 Turbine trip without bypass at 863 MWt (22.5% power) ; TT13 Loss of vacuum TT14 Loss of offsite power.with operable Diesel Generators TT15 HPCS injection at high power . TT16 Normal low power Recirculation Pump startup ; TT17 Failure of reactor' pressure control
- 1 1
i
, - ~ . ~ , . , , , . , - - , , ,- ,, , , - - ,..-.-- .~ -- - . - , - - , - , . -
Grand Gulf Nuolear Station simulator Certification Initial Report, March 1991 Page 23 of 25 resulting in closure of turbine pressure control valves TT18 Failure of reactor pressure control resulting in opening of turbine pressure control valves TT19 Group I Isolation with a failure to SCRAM TT20 Feedwater line break outside the containment TT21 Feedwater line break inside the containment (drywell) TT22 Fast closure of both Recirculation flow control valves TT23 Failure of a single Recirculation flow control valve open TT24 Failure of both Recirculation flow control valves open TT25 Failure of osa Recirculation flow control valve closed TT26 Closure of both Recirculation flow control valves TT27 Feedwater control system failure to maximum demand TT28 Loss of Feedwater heating TT29 Load reject with no bypass TT30 Load reject with bypass
- 4. Malfunction Tests
- a. Written l tests were performed to verify the correct functioning of' alarms, parameters, and automatic actions for each malfunction available for training use. Criteria were ' established in each test to identify expected responses.
l
-e -
l 1 l Grand Gulf Nuclear. station i
-Simulator certification Initial Report, March 1991 Page 24 of 25-
- b. Test procedures were developed using.the. plant schematics, logics, P&ID drawings, and operations procedures.
- c. Where - applicable, various malfunction severities were tested for expected response.
- d. A list of all simulator ma) functions tested is included in Appendix II.
- e. Abstracts- of malfunction test procedures- are included in Appendix IX.
B. Testina Schedole
- 1. All transient tests, steady state ' tests and the l computer real time tests will be~ performed annually.
- 2. Approximately one-fourth of'the malfunction tests will be performed each year as follows:
- a. 1991-Malfunction Tests'l'through 45
- b. 1992-Malfunction Tests.46through-90 '
- c. 1993-Malfunction Tests 911through 135 ;
- d. 1994-Malfunction Tests 136 through 181' ~l 1
3.- -Additional-testing:will-be performed ~following simulator modifications 'and deficiency- resolution. as applicable. Ratests may include-portions.of the affected tests- or_ may include. several tests: depending on the problem resolution. <, 1 i
-IV. Simulator Uoarade Proiect -j A. Proiect Se m -
- 1) Entergy Operations L recognizes: theineedfto continually'
' improve'=and': update :our facilities, = including. the .,
i simulator. The upgrade project plan was developediby the
-Simulator Configuration Review Board'and is: planned?for
-implementation;during 1991-94 _MajorDf eatures described .
'i
-below are:.. subject > to- -change as: ~ planning; 'and; implementation. progresses.=
j
f Grand Gulf Nuclear station simulator certification Initial Report, Maroh 1991 Page 25 of 25
- a. The simulator upgrade is expected to replace several existing software models, instructor station features, and add several back panels.
The following system models currently are_ planned , to be replaced:
- 1) Core and recirculation system
- 2) Main steam
- 3) Turbino
- 4) Condensor
- 5) Condensate
- 6) Feedwater
- 7) Containment and drywell
- b. An offgas system model will be added.
- c. Radiation-transport will be incorporated in'all l
new models,
- d. The offgas Panel 1H13-P845, and the Division 2 ADS Panel 1H13-P631 will be added.
- e. The portion of the Control Rod Test Panel, 1H13-P610, including the RPS power supply indicator lamps and transfer switches will be added.
- 2) Proiect ScherJglg at The project is currently scheduled to begin in late 1991 af ter - the award of a contract for vendor support.
- b. The project is currently scheduled to be completed in 1994 and ready for training prior to simulator y recertification-in 1995.
Appendix I Simulator Initial Conditions ,2
' ~
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Appendix I - Initial Conditions Page 1 of 10 ADoendix I Initial Condition (IC) Inde.g I. Introduction Twenty six initial conditions are provided to supply a variety of plant operating conditions, fission product poison concentrations, and core exposure conditions. .The following table will help in understanding each Initial Condition (IC) description (All values are approximate): IL. Descriotion of-Table 100% Xenon = 100% power equilibrium value I = Increasing , D = decreasing EXAMPLE: 120% I means 20% above equilibrium value increasing. Decay Heat is given in % of 100% core power (3833 MWT) BOC = Beginning of Cycle (2/3 old fuel 1/3 new fuel) l MOC = Middle of Cycle EOC = End Of Cycle III. Notes A. IC-1, BOC is the only IC having no decay heat. This allows new students to perform start up operations without having_to deal with decay heat. B. Samarium is set at 66.6% ' for a new fuel load with no exposure. Samarium is ~ set at 100% for ' equilibrium conditions. C. The - step of the rod movement sequence is 'given for applicable -ICs. Reactor temperature (MOD TEMP) .and , pressure (RX PRESS) are given for each IC. D. Each of the initial conditions'provided were created by performing plant startep, shutdown, and power operations using the plant procedure governing evolutions required-to establish the conditions.^These ICs support training . for the normal plant evolutions ~specified in ANS 3.5 Section 3.1.1'
. ~ s , - - . , . , . , , . -- , , -, , , . , , . , , . . . ,
Grand Gulf Nuclear Station Mimulator certification Initial Report, March 1991 Appendix I - Initial Conditions Page 2 of 10 IV. Simulator Initial Condtions (ICs) IC-1 XENON = 0% SAMARIUM = 0% DECAY HEAT = 0% MOD. TEMP. = 85'F RX. PRESS. = 0 psig BOC POWER = 0% STEP: All rods in COMMENTS: All systems secured except electrical, instrument and service air, PSW, and TBCW systems. 19-2 XENON = 0% ' SAMARIUM = 66% DECAY HEAT = 0.15% MOD. TEMP. = 121*F RX. PRESS. = 0 psig POWER = 0% BOC STEP: All rods in COMMENTS: Preparations for reactor startup are almost complete. Ready to place the reactor mode switch in startup except for removing RHR shutdown cooling Loop A from operatien. IC-3 ' XENON = 158% D SAMARIUM = 100% DECAY HEAT = 0.7% MOD. TEMP. = 142*F RX. PRESS. = 0 psig POWER = 0% MOC STEP: All rods in COMMENTS: 10 hours after scram.-Preparations for restart of 1 the reactor are complete except forr placing the. ' reactor mode switch- in startup and removing RHR shutdown cooling loop A from operation. I
> t i
i Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Appendix I - Initial Conditions : Page 3 of 10 ! IC-4 I XENON = 0% ! SAMARIUM = 66% [ DECAY HEAT = 0.14% i MOD. TEMP. = 135'F RX. PRESS. = 0 psig l POWER = 0% 3 BOC ! STEP 34 COMMENTS: Plant startup is in progress. The reactor is $0.54 . suberitical. Control rods are in the process of ) being withdrawn for reactor criticality. .! f IC-5 ! i XENON = 154% D i SAMARIUM = 100% f DECAY HEAT = 0.7% i MOD. TEMP. = 302'F ! RX. PRESS. = 110 psig t i POWER = 0% ; MOC : STEP 102 COMMENTS:. Plant startup is in progress. The reactor is
$0.50 suberitical. Control rods are in the process ,
of being withdrawn for reactor criticality. - IC.-l XENON = 0% { SAMARIUM =.100% DECAY HEAT = .4% MOD. TEMP. = 188'F f l RX. PRESS. = 0 psig POWER = 0% EOC i STEP 74A _ COMMENTS: Plant startup is in progress. The reactor is $0.50 suberitical.- . Control. rods are in the process-'of' ; being withdrawn for reactor criticality. The reactor ! was~previously shutdown'for three weeks.- ! L h P
+
a ,--,-,.nm,w n m. 3
. , ,- . ,- - - , v.,an, -..~.r e.,, ,v- r,-e- ,;.,--,-n,n-,n,- , . . .
Grand Gulf Wuclear station simulator certification Initial Report, March 1991 Appendix I - Initial conditions Page 4 of 10 19 I XENON = 0.02% SAMARIUM = 66% DECAY HEAT = .15% MOD. TEMP. = 292'F RX. PRESS. = 29 psig POWER = 1% BOC STEP 40 COMMENTS: Plant startup is in progress. A 40'F/hr heat up is in progress. IC-8 i XENON = 0.1% SAMARIUM = 66% DECAY HEAT = 0.22% MOD. TEMP. = 447'F RX. PRESS. = 403 psig , POWER = 2% BOC STEP 67 COMMENTS: Power is on IRM range 7 or 8 with a heatup in progress. Bypass valves are approximately 10% open. Plant startup has progressed to the point of placing the first reactor feed pump in service. 1G-2. XENON = 0.2% . SAMARIUM = 65% DECAY HEAT = 0.24% MOD. TEMP. = 536'F RX. PRESS. = 951 psig Increasing POWER = 2% BOC ' ' .EP 82A COMMENTS: Plant startup has progressed to the point of raising reactor power to above >4%, so that the reactor mode switch can be placed to Run. Plant heatup is complete.. i b 1
Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 Appendix I - Initial Conditions Page 5 of 10 IC-10 XENON = 0.5% SAMARIUM = 64% DECAY HEAT = 0.8% MOD. TEMP. = 537'F RX. PRESS. = 957 psig ; POWER = 12% BOC STEP 100 COMMENTS Plant startup is in progress. The main turbine ; is ready to roll. TSE Influence is on. IC-11 XENON = 1% SAMARIUM = 64% DECAY HEAT = 1.5% MOD. TEMP. = 537'F RX. PRESS. = 974 psig POWER = 26% BOC STEP 107 COMMENTS Plant startup is in progress. The turbine generator is ready to be synchronized to the grid. IC-12 XENON = 10% I SAMARIUM = 60% i DECAY HEAT = 2.3% MOD. TEMP. = 540*F RX. PRESS. = 974 psig POWER = 37% BOC STEP 111 COMMENTS Plant startup is in progress. Plant conditions are at the point of shif ting recirculation pumps to fast speed and starting the second circulating water pump. The reactor rod line-is below 80%. -
l 4 Grand Gulf Nuclear station simulator Certification l Initial Report, March 1991 Appendix I - Initial Conditions Page 6 of 10 IC-13 XENON = 74% SAMARIUM = 51% DECAY HEAT = 4.5% I MOD. TEMP. = 544'F RX. PRESS. = 1003 psig POWER = 84% BOC STEP 138 COMMENTS Plant startup is in progress. .The plant is operating in the maximum extended operating domain. A reactor rod line of 108% has been established. Power ascension is limited by preconditioning constraints. IC-14 XENON = 73.5% I SAMARIUM = 50% DECAY HEAT = 5.3% MOD. TEMP. = 548'F RX. PRESS. = 1019 psig POWER = 99% BOC STEP 138 COMMENTS Plant startup is complete per 101-2. l IC-15 XENON = 100% SAMARIUM = 100% DECAY HEAT = 6.88% MOD. TEMP. = 54 8 ' F RX. PRESS. = 1020.psig POWER-= 100% BOC ST2P 140B COMMENTS: The- plant has been running continuously' for two months after a refueling outage. l .~ . . . - . - , . , , , . . , , , . . . - ~ , . -. , , . , --..-,.i.,, -
- . . . ~ , . -
4 1 Grand Gulf Nuclear Station , simulator certification Initial Report, March 1991 Appendix I - Initial Conditions Page 7 of 10 i ma XENON = 100% SAMARIUM = 100% l l DECAY HEAT = 6.8% l MOD. TEMP. = 548'F RX. PRESS. = 1020 psig POWER = 97% MOC STEP 139 (at position 24) COMMENTSt The plant is at mid cycle. The plant is operating in the maximum extended operating domain with high flow (105%) and less-than-rated rod line (97%). IC-17 XENON = 100% SAMARIUM = 100% DECAY HEAT = 6.8% MOD. TEMP. = 584'F RX. PRESS. = 1020 psig , POWER = 100% ' EOC STEP: All rods out COMMENTSt 99% core flow, 100% power. The plant is at end of cycle in a core coastdown. All rods - are fully withdrawn. IC-18 XENON = 107% Increasing SAMARIUM = 102% Increasing DECAY HEAT = 3.9%
- MOD. TEMP. = 540'F RX. PRESS s.973 psig.
POWER = 33% EOC STEP 138 -(reverse order)- COMMENTS: Plant shutdown is in progress.- Core flow is 47% and Reactor power is.35%. Rods are being inserted in sequence-per the rod-movement sequence. ! I i
-i
Grand Gulf Nuclear Station simulator certification i Initial Report, March 1991 Appendix I - Initial conditions Page 8 of 10 IC-19 XENON = 118% Increasing SAMARIUM = 106% Increasing DECAY HEAT = 2.7% MOD. TEMP. = 537'F RX. PRESS. = 964 psig POWER = 17% EOC STEP 131 (reverse order) COMMENTS: Plant shutdown is in progress. The generator is ready to be disconnected from the grid. LC. .2.Q XENON = 135% Increasing SAMARIUM = 113%-Increasing DECAY HEAT = 1.5% MOD. TEMP. = 538'F RX. PRESS. = 950 psig POWER = 1% EOC STEP 36 (reverse order) COMMENTS: A plant shutdown is in progress. Control rods are almost fully inserted. The last few rods need to . be-inserted beforo placing the reactor mode switch , to shutdowr.. IC-21 XENON = 145% Increasing SAMARIUM = 119% Increasing DECAY HEAT = 0.96% MOD. TEMP. = 358'F RX. PRESS. = 141 psig POWER = 0% EOC STEP: All rods in COMMENTS: . Plant shutdown to cold shutdown is in progress. RHR SDC loop "A" is flushed and ready'for warm up. ' I e t
-- , - . 4 . - - , , . . - ~ - . . . . - . . . , ,
4 I Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 Appendix I - Initial Conditions i Page 9 of 10 l' IC-22 XENON = 153% Increasing SAMARIUM = 133% DECAY HEAT = 1.57% MOD. TEMP. = 538'T RX. PRESS. = 951 psig POWER = 1% EOC STEP: 126 COMMENTS: Power has been reduced to Range 6 and 7 of the IRMs so that a Drywell entry can be made. IC-23 XENON =0% SAMARIUM a 66% DECAY HEAT = .3% MOD. TEMP. = 150'F RX. PRESS. = 0 psig , POWER = 0% MOC STEP: all rods in COMMENTS: Plant shutdown is complete. The Recirculation, CRD, and RWCU systems are secured. MSIVs'are shut.. IC-24 XENON = 107% Increasing SAMARIUM = 101% DECAY. HEAT = 2% MOD.. TEMP. = 530'F
- RX. PRESS. = 998 psig POWER = Ot-
.EOC
! STEP all rods in . . l COMMENTS: A reactoriseram on':high-main steam.line radiation occurred. The-results ofJcoolant samples indicate i fuel failure. The SCRAM ONEP was completed.
- a
, , - - ~ , , , , , , - m ,, - - --.- - - -~,..c
. . . - , . . . . . . . + , - , , - - . . - , . , - - - -- - .~- .--
i Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Appendix I - Initial Conditions Page 10 of 10 IC-25 XENON = 0% SAMARIUM = 100% DECAY HEAT = 1.5% MOD. TEMP. = 413*F RX. PRESS. = 437 psig POWER = 0% EOC STEP all rods in (except 4 stuck rods ) COMMENTS: This IC is set up to allow Hydrogen Control procedures to be used. Hydrogen Analyzers are turned on. Containment hydrogen concentration is 1.2%. Drywell hydrogen concentration is 0.5%. IC-26 XENON = 0% SAMARIUM = 66% DECAY HEAT = 1.5% MOD. TEMP. = 149'F i RX. PRESS. = 0 psig POWER = 0% IRS STEP all rods in COMMENTS: This IC is for use in _ displaying an initial startup after a refueling outage. RHR Shutdown cooling _was just removed from service. All conditions are met to place-the reactor mode switch in1startup. a
~
l
)
_g- _______________i-_i-._
N N n f 4 1 i Appendix II Sitnulator Malfunctions [ l l I t
-1
, _. -- - - ~ . .
- _ ---.__.-_ - - - . ~ . . . - . - - . _ . - - - - _ . . - - . . - - . - - . - - - - -
'l Grand Gulf Nuclear Station l
, Simulator certification
- Initial Report, March 1991 i Appendix II - simulator Malfunctions !
Page 1 of 14 ! Accendix II { Simulator Malfunctions f l The malfunctions listed in Appendix II are provided to simulate ; plant response and automatic-plant control functions for the rango 6 of malfunctions specified in section 3.1.2 of ANSI /ANS 3.5-1985. i The malfunct' ons are listed by malfunction page-number, malfunction ; selection number, and a short description. Malfunctions E- with ; variable severity are marked (VAR). The malfunctions are grouped j by system or category. Malfunctions are activated on the MST page ; of the instructor console by entering a line number, malfunction selection number, a time delay to activate or clear if ' appropriate, ! and a severity for certain malfunctions. Up to 15 malfunctions can j be activated at a time. ; i t Pace Selection Descriotign l M01 - NEUTRON MONITORING SYSTEM 1A SRM CHANNEL A FULLSCALE ! 1B SRM CHANNEL B FULLSCALE-1C SRM CHANNEL C FULLSCALE ID SRM CHANNEL D FULLSCALE l 1E SRM CHANNEL E FULLSCALE 1F SRM CHANNEL F FULLSCALE ' i
- 2A SRM CHANNEL A DOWNSCALE <
2B SRM CHANNEL B DOWNSCALE f 2C SRM CHANNEL C'DOWNSCALE I 2D -SRM CHANNEL D DOWNSCALE ! 2E SRM CHAT!NEL E DOWNSCALE j i 2F. SRM CHANNEL F DOWNSCALE 1 3A 'SIW CHANNEL A DETECTOR STUCK ; 3B SRM CRANNEL B DETECTOR STUCK 3C SRM CHANNEL C DETECTOR STUCK-3D SRM CHANNEL'D DETECTOR STUCK ~ ! 3E .SRM CHANNEL E DETECTOR STUCKu , 3F- SRM CHANNEL F= DETECTOR' STUCK l' t t i t
~i i
)
, - - . ..;- -A- .,_ ._._;,-..,a _ , _ . . . - - ..
Grand Gulf Nucisar Station simulator Certification
- Initial Report, March 1991 4
Appendix II _ simulator Malfunctions Page 2 of 14 i Pace Selection Descriotion
- M02 -
NEUTROH MONITORING SYSTEM 4A IRM CHANNEL A FULLSCALE- . 4B IRM CHANNEL B FULLSCALE , 4C IRM CHANNEL C FULLSCALE l 4D IRM CHANNEL D FULLSCALE 4E IRM CHANNEL E FULLSCALE . 4F IRM CHANNEL F FULLSCALE 4G IRM CHANNEL G- FULLSCALE 4H IRM CHANNEL -H FULLSCALE - SA IRM-CHANNEL.A DOWNSCALE , SB IRM CHANNEL B DOWNSCALE l SC IRM' CHANNEL C DOWNSCALE-SD IRM CHANNEL D DOWNSCALE I SE IRM CHANNEL E DOWNSCALE i
- 5F IRM. CHANNEL F DOWNSCALE SG- IRM- CHANNEL G -DOWNSCALE l SH IRM CHANNEL H DOWNSCALE 6A-H IRM CHANNEL.(A-H) DETECTOR' STUCK-M03 -
NEUTRON MONITORING SYSTEM-7(X) LPRM DETECTOR'(XX-YYZ) FULLSCALE-8(X) LPRM DETECTOR (XX-YYZ) - DOWNSCALE , 9A APRM CHANNEL A FULLSCALI 9B APRM CHANNEL: B FULLSCALE 9C APRM CHANNEL C FULLSCALE 90 APRM CHANNEL D FULLSCALE 9E APRM. CHANNEL E FULLSCALE 9F APRM CHANNEL F FULLSCALE ( 9G- APRM CHANNEL G FULLSCALE. 9H APRM CHANNEL- H -FULLSCALE MO4 - NEUTRON-MONITORING > SYSTEM-
' 10A
- APRM-CHANNEL A~DOWNSCALE:-
10B- APRM CHANNEL:B DOWNSCALE 10C APRM CHANNEL.C DOWNSCALE l - 10D . APRM~ CHANNEL D DOWNSCALE 10E APRM CHANNEL E DOWNSCALE
-10F - APRM' CHANNEL FJDOWNSCALE 10G APRM CHANNEL G DOWNSCALE--
10H APRM CHANNEL H .DOWNSCALE t 1 l y- y----- , , w
-yo_,<w . r y.+% y-. ,mm.,,4w w y.w_,.,._%.,, ,, ,--wg -
,-.,v,-y .,,,w, ,,4 m.g.,,,,.-rw.,.-wi#w-,-.-,-aw,,,9 . , - , , - , .-e,#-,U
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Appendix II - Dimulator Halfunctions Page 3 of 14 Pace Selection Qgaggiption M04 (Continued) RECIRCULATION SYSTEM 169A RECIRC PUMP A SHAFT SHEAR 169B RECIRC PUMP B SHAPT SHEAR 170A RECIRC PUMP A PUMP SEIZURE 170B RECIRC PUMP B PUMP SEIZURE 172A RECIRC LOOP A FCV RUNBACK 172B RECIRC LOOP B FCV RUNBACK 173A RECIRC LOOP A FCV FAILS OPEN 173B RECIRC LOOP B FCV FAILS CLOSED MOS - RECIRCULATION SYSTEM 11 JET PUMP FAILURE 12A REACTOR RECIRCULATION PUMP A TRIP 12B REACTOR RECIRCULATION' PUMP B TRIP 13A VALVE P42-F066 FAILURE (VAR) 13B VALVE P42-F114 FAILURE 14A RECIRC PUMP A HIGH VIBRATION 14B RECIRC PUMP B HIGH VIBRATION 15A1 RECIRC Al SEAL FAILURE 15A2 RECIRC A2 SEAL FAILURE 15B1 RECIRC B1 SEAL FAILURE 15B2 RECIRC B2 SEAL FAILURE 16A RECIRC LOOP A INCORRECT START 16B RECIRC LOOP B INCORRECT START 17A RECIRC LOOP A FLOW CNTRL VALVE FAILURE 17B RECIRC LOOP B FLOW CNTRL VALVE FAILURE 18A RECIRC LOOP A FLOW FEIDBACK SIGNAL FAIL l 18B RECIRC LOOP B FLOW FEEDBACK SIGNAL' FAIL. 19 FLUX FEEDBACK SIGNAL FAILURE 20 RECIRC MASTER CONTROL FAILS LOW M06 - CONTROL ROD DRIVE HYDRAULIC (CRLH) SYSTEM 161(X) CONTROL ROD (XX-YY) PRIFT GUT 21(X) CONTROL ROD (XX-YY) DRIFT IN 22(X) CONTROL ROD (XX-YY) STUCK 23(X) UNCOUPLED CONTROL ROD (XX-YY) 24(X) CONTROL ROD (XX-YY) ACCUM_ TROUBLE 25(X) SCRAM CONTROL 200 (XX-YY) 26(X) CONTROL ROD DRIVE (XX-YY) SEAL WORN 27 CRDH FCV A FAILURE (VAR) 28A CRDH HYDRAULIC PUMP A TRIP 28B CRDH HYDRAULIC PUMP B TRIP 1 6
Grand Gulf Nuclear Station
' Simulator certification Initial Report, March 1991 Appendix II - simulator Malfunctions Page 4 of 14 Ragg Selecti4D Descriotion M06 (Continued) REACTOR WATER CLEANUP SYSTEM (RWCU) 29A AEACTOR WATER CLEANUP PUMP A TRIP 29B REACTOM WATER CLEANUP PUMP B TRIP 30 RWCU DRAIN VALVE - FAI'LS SHUT 31 RWCU DEMIN HI DIFFERENT.IAL PRESSURE 32A RWCU DEMINERALIZED RESIN DEPLETION-A-32B RWCU DEMINERALIZER RESIN DEPLETION B 33 REACTOR WATER CLEANUP SYSTEM LEAK (VAR)
Mb7 - ROD CONTROL JWD INFORMATION SYSTEM (RC&IS) 34 RC&IS --CHANNEL DISAGREE 35 RC&IS - RPIS FAILURE (OPEN REED SWITCH) f 36 .RC&IS - RPIS FAILURE (CLOSED REED' SWITCH) 37 RC&IS - TIMER MALFUNCTION 38 RC&IS - SELF TEST FAILURE 39A RC&IS - RPC FAILS TO INITIATZ ROD BLK- N1
; 39B RCGIS - RPC FAILS TO INITIATE ROD BLE 02 39C RC&IS - RPC FAILS TO INITIATE ROD BLK N31 MOB -
NUCLEAR BOILER SYSTEM INSTRUMENTATION [ 40A DW PRESS XHTR B21-N094 A FAILS (VAR) 40B DW PRES 3 XMTR B21-N094 B FAIIS (VAR) 40E DW PRESS XMTR B21-N094 E FAILS (VAR) 40F DW PRESS XMTR B21-N094 : F- FAILS (VAR) 41A RPV LEVEL XMTR B:11-N091 A FAILS (VAR)' 41B RPV LEVEL XMTR B;1-N091 B TAILS (VAR) ' 41E RPV LEVEL XMTR L21-N091 E FAILS (VAR) 41F RPV LEVEL XMTR B21-N091. 7 FAILS (VAR) 42A RPV PRESS XMTR B21-N068 A FAILS (VAR) 42B RPV PRESS. XMTR B21-N068 B FAILS . (VAR) 42E RPV PRESS XMTR B21-N068 E FAILS (VAR) 42F RPV PRESS XMTR B21-N068 F FAILS (VAR) MO9 -
. REACTOR CORE ISOLATION COOLING .(PCIC) 43 RCIC AUTO START FAILURE-44- RCIC TURB SPEED CNTRL FAIL (VAR) 45 RCIC FLCW TRANSMITTER. FAILURE 46 RCIC-TURBINE OVERSPEED 47 RCIC TURBINE TRIP 48 RCIC SYSTEM ISOLATION 49: RCIC STEAM.LINE LIAK-(VAM) i
.. . - _ ._- _ ~ . . . _ , - - _
Grand dulf Nuclear Station siuulator Certification i Initial Repert, March 1991 Appendix II - Simulator Malfunctions Pa7e 5 of 14 2AE2 Eelection Jescriotion M10 - EMERGENCY CORE COOLING SYSTEMS 50A IRR PUMP A TRIP 50B FlR PUMP B TRIP 50C Ra P PUMP C TRi? 51 T1 'd f3 ESSURE CORE SPRAY PUMP TRIP 52 a3H'03 ESSURE CORE SPRAY PUMP TRIP 51 G*URIOUS HPCS INITIATION 54 TafR SYS A STM VALVE E12-F051A FAILL OPEN 54L RHR SYS B STM VALVE E12-F051B FAILS OG' 55A RHR SYS A STM VALVE E12-F051A FAILS JWT 55D RHR SYS B STM VALVE E12-F051B FAILS $90/ 50A RHR SYS A THROT VLV E12-F003A FAILS 56B THR SYS B THROT VLV E12-F003B FAILS 57A 16'il SYS A HX TUBE LEAK 57B RHk *1YS B HX TUBE LEAK 58A RHR L 'IS A HX LVL CNTRL FAILS DOWNSCALE 58D RHR \1':1 B HX LVL CNTRL FAILS DOWNSCALF, 177 DEFElt HPCS AUTO INITIATION M11 - LEAXP, l 59(X) SRV (X) FAILS OPEN (SEE M26) 60(X) SRV (X) LEAKS (SEE M26) 61(X) SRV (X) FAILS TO RESEAT (SEE M26) 62 INSTRUMENT LINE LEAK 63 RECIRCULATION LOOP RUPTURE (VAR) 64 1TFAM LEAK IN DRVWELL 65 LTP.AM LINE RUPWURE IN DRYWELL (VAR) 66 STElM udAK IF TUNNEL ! 67 STEAP LINE Pi 'TURE IN TUNNEL (VAR) l 68 STEAM ,'3 alRJ-IN TURPINE BUILDING (VAR) 69 INSTRUK%i, LINd LEAK OUTSIDE CONTAINMENT 70 FEEDWATES l 'ME RUPTURE OUTSIDE CNTNMENT 71 FUEL CLADDINT L2AK (VAR) 171 FEEDWATER LIN.' B BREA.KS IN DRYWELL (VAR) l
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Appendix II - simulator Malfunctions ' Page 6 of 14 Pace gelection Descriotion M12 - REACTOR PROTECTION SYSTEM (RPS) 72- MAIN STEAM LINE HIGH RADIATION 73AI MSIV VAILS . SHUT B21-F022 A - 73BI MSIV FAILS SHUT.B21-F022B 73CI MSIV FAILS SHUT B21-F022C 73DI 'MSIV FAILS SHUT B21-F022D 73AO MSIV FAILS SHUT B21-F028AL 73B0 MSIV FAILS SHUT B21-F028B 73CO MSIV FAILS SHUT B21-F028C 73DO MSIV FAILS SHUT B21-F028D 74 SPURIOUS REACTOR SCRAM- i 75 FAILURE TO SCRAM 76 FAILURE To SCRAM (MAN SCRAM OPERABLE) 77A RPS MG SET A FAILURE 77B RPS MG SET B FAILURE 162 RPS AUTO & MANUAL FAILURE'TO SCRAM 164A SDV HYDRAULIC BLOCK OF: CONTROL RODS 165 LaSALLE POWER OSCILLATION EVENT , M13 - PRESSURE REGULATOR 78 PRESS' REG FAILS LO/TRANS TO BACKUP 79 PRESSURE REGULATOR FAILS - LOW 80 PRESSURE REGULATOR FAILS HIGH 81- PRESSURE. REGULATOR OSCILLATION-82A TURBINE BYPASS CONTROL VLV A STUCK 82B TURBINE BYPASS CONTROL VLV'B-STUCK 820 ; TURBINE BYPASSE CONTROL VLV C STUCK-83A TURBINE BYPASS ' CONTROL 1VLV- A-- FAILS' OPEN 83B TURBINE BYPASS CONTROL VLV B FAILS OPEN-83C TURBINE BYPASS- CONTROL VLV C FAILS OPEN 84A TURRINE BYPASS- STOP VLV A FAILS = SHUT 84B TURUINE' BYPASS STOP VLV D FAILS SHUT 84C TURBINE BYPASS STOP VLV C FAILS SHUT t 4
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Appendix II - Simulator Malfunctions Page 7 of 14 Eage Selection Descrintion M14 - TURBINE / AUXILIARIES 85 TURBINE GOVERNOR FAILS HIGH 86 TURBINE GOVERNOR FAILS LOW 87 TURBINE ACCELERATION CONTROL FAILURE 88A TURBINE STOP VLV A FAILS SHUT 88B TURBINE STOP VLV B FAILS SHUT 88C TURBINE STOP VLV C FAILS SHUT 88D TURBINE STOP VLV D FAILS SHUT 89A TURBINE CV SERVO A FAILS (VAR) 89B TURBINE CV SERVO B FAILS (VAR) 89C TURBINE CV SERVO C FAILS (VAR) 89D TURBINE CV SERVO D FAILS (VAR) 90A MSR A FIRST STAGE REHEAT FAILURE 90B MSR b FIRST STAGE REHEAT FAILURE 91A MSR LOSS OF 2ND STAGE REHEAT TEMP CONTROLLER A 91B MSR LOSS OF 2ND STAGE REHEAT TEMP CONTROLLER B M15 - TURBINE / AUXILIARIES 92A MSR LOSS OF 2ND STAGE REHEAT STEAM A 92B MSR LOSS OF 2ND STAGF, REHEAT STEAM B 92C MSR LOSS OF 2ND STAGE REHEAT STEAM C 92D MSR LOSS OF 2ND STAGE REHEAT STEAM D 93 MAIN TURBINE TRIP 94 TURBINE GLAND SEAL REGULATOR FAILURE 95 LOSS OF TURBINE HP CONTROL OIL PRESSURE 96 TURBINE THRUST BEARING WEAR TRIP 97 TURBINE BEARING OIL LOW PRESSURE 90 TURBINE LUBE OIL TEMP CNTRL FAILURE 99A-L TURBINE BEARING X HIGH VIBRATION 100 TURBINE EXHAUST HOOD SPRAY FAILURE 101A TURBINE BELLOWS SEAL FILL VLV A FAILS SHUT 101B TURBINE BELIOWS SEAL FILL VLV B FAILS SHUT 101C TURBINE BELLOWS SEAL FILL VLV C FAILS SHUT 163 LOSS OF CONDENSER VACUUM (VAR)
1
~ l Grand Gulf Nuclear Station.
simulator-Certification Initini Report, March 1991 Appendix-II;- simulator-Malfunctions Page 8 of 14 PJlLqa Seleclip_D Descriotion> M16 -- GENERATOR / AUXILIARIES , 102 GENERATOR AUTO VOLTAGE REGULATOR -TRIP'~ 103 GENERATOR TRIPL 104 GEN ROTOR CLE WTR FLOW DECR-(VAR) 105 GEN PRI CLG TEMP CNTRL FAIL'(VAR) 106 LOSS OF GEN. PRIMARY-COOLING WATER 107 GEN H2 TEMP;CNTRL FAILURE _(VAR) 10SA ISOLATED PHASE BUS DUCT BLOWER A-TRIP : - 3' 108B ISOLATED PHASE BUS DUCT = BLOWER A TRIP M17 - CONDENSER /FEEDWATER' 110A SJAE A STEAM SuPPLYJFAILURE (TOTAL) 110B SJAE-B STEAM SUPPLY-FAILURE (TOTAL) 111 0-G HI ACT-AFTER CHAR ADSORB (VAR); 112 MAIN CONDENSER-HOTWELL. MAKEUP CNTL FAIL 113A MAINLCONDENGER_ TUBE RUPTURE AL(VAR)- 113B MAIN CONDENSER TUBE RUPTURE B (VAR) 114A MAIN CIRCULATING -- WATER PUMP: A TRIP - 114B MAIN . CIRCULATING WATER .' PUMP ( B TRIP
-115A CONDENSATE: PUMP.A TRIP 115B- CONDENSATE:PUMPfB TRIP j 115C CONDENSATE PUMPEC TRIP 116(A-H) 1 CONDENSATE DEMIN RESIN BED (A-H) FAILURE 117(A-H) <
COND DEMIN (X)'HI DELTA'P.(A-H)L M18- - CONDENSATE /FEEDWATER 118A CONDENSATE. BOOSTER PUMP A TRIP-118B CONDENSATE' BOOSTER PUMP B TRIP 118C- CONDENSATE BOOSTER PUMP C TRIP-119- HEATER DRAIN TANK LVL CNTRL FAILURE 120A LOW PRESSURE-HEATER 1A HI-HI-LEVELS 120B LOW PRESSURE' HEATER 1B HI-HI LEVEL 120C LOW PRESSURE HEATER 1CsHI-HILLEVEL ' 120D LOW PRESSURE HEATER 2A HI-HI ~ LEVEL 120E LOW PRESSURE HEATERc2B;HI-HI' LEVEL-
-120F LOW PRESSURE HEATER 2C HI-HI LEVEL l 120G LLOW PRESSURE HEATER.3A HI-HI! LEVEL 120H LOW PRESSUREiHEATER:3B HI-HI-LEVEL
! 120I_ LOW PRESSURE HEATER 3C HI-HILLEVEL-120J ' LOW PRESSURE HEATER 4A HI-HI LEVEL 120K- LOW PRESSURE HEATER 4B HI-HI LEVEL
-120L LOW PRESSURE HEATER-4C HI-TTI1 LEVEL L _ . . . _ . . _ . _ -- . _ . _ . . - - , . . _ , _ . , . . _ . _ . - ._
}
Grand Gulf Nuclear Station simulator certification
. Initial Report, March 1991 Appendix II - simulator Malfunctions Page 9 of 14 i Pace Selection Oescriotion M19 -
CONDENSATE /FEEDWATER 121A RFPT A SIGNAL FAILURE 121B RFPT B SIGNAL FAILURE 122A RFPT A HP STEAM SUPPLY-FAILURE 122B RFPT B HP STEAM SUPPLY FAILURE 123A RFPT A OVERSPEED TRIP-123B RFPT B OVERSPEED TRIP 124 FW '.,0 FLOW CV SIGNAL FAIL (VAR) 125 FAiaURE OF STM FLOW SIG TO FW CONTROL-126A FEEDWATER CNTRL VESS LVL A SENSOR FAILURE 126B FEEDWATER CNTRL VESS LVL B SENSOR FAILURE 127 FEEDWATER MASTER CONTROLLER FAILS OPEN 128 FEEDWATER MASTER CONTROLLER FAILS SHUT 129A5 HIGH PRESSURE HEATER A5 TUBE LEAK (VAR) 129A6 HIGH PRESSURE HEATER A6 TUBE LEAK-(VAR) 129B5 HIGH PRESSURE HEATER B5 TUBE LEAK (VAR) I 129B6 HIGH PRESSURE HEATER B6 TUBE LEAK--(VAR) M20 - ELECTRICAL 130 NETWORK LOAD INCREASE-131 NETWORK LOAD. DECREASE 132 NETWORK LOAD LOSS 133U1 SERVICE TRANSFORMER 11 LOCKOUT 13302- SERVICE TRANSFORMER-21 LOCKOUT 134(X) BOP & ESF TRANS-LOCKOUT BUS-(X) (11A, 11B, 12A,'12B, 13, 23, 11, 12) 135 SWITCHYARD FAULT (500 & 115KV) 136(X) -34.5 KV BUS'(X) OVERCURRENT-TRIP (21R, 11R, 12R, 13 R) - 180 ESF TRANSFORMER 21 LOCKOUT-
Grand Gulf Nuclear Station Simulator Certification-Initial Report, March 1991 Appendix-II - simulator Malfunctions Page 10 of 14-EA2g Selection Descriotion M21 - ELECTRICAL 137(X) 34.5 KVAC BUS (X) PHASE DIFFERENTIAL TRIP-(11R, 21R) 138(X) .6,.9 KVAC BUS (X) OVERCURRENT TRIP (11HD, 12HE) 139(X) 4160 VAC BUS-(X) OVERCURRENR TRIP (13 AD,14 AE,18AG, 28AG, .15AA,16AB, 17 AC) 140A DIV-I DG-11 FAIL TO START 140B DIV II DG 12 FAIL TO START. 140C DIV III DG 13 FAIL'TO START 141A DIV-I DG 11 TRIP 141B DIV II DG 12 TRIP 141C DIV III DG 13' TRIP 142(X) 480 VAC BUS (X) OVERCURRENT (11BD1-5, 7 & 12 bel, 2,4-6 & 13BD1, 2, 5) (14BE1,2 & 18BGl'& 15BAl-6) (16BB1-6 & 17B01) 178 480 VAC BUS 28BG1 OVERCURRENT TRIP M22 - ELECTRICAL
-143(X) 120 VAC BUS-(X) TRIP (1Y71, 1Y74, 1Y75, 1Y76,- 1Y78) 166(X)- 120V AC BUS (X)-TRIP (1Y87,'lY86, 1Y96, 1Y95)_
144 250 VDC BUS 111DF-TRIP 145 (X) 125 VDC BUS (X) TRIP (11DA, 11DB, 11DC, 11DD, 11DE) 146(X) 24/48 VDC: BUS (X) TRIP (11DH, 11DJ) 176(X) 125 VDC BUS (X) TRIP (11DK, 11DL) 179(X) 125 VDC BUS (X) _ TRIP (11DG, 21DG)
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Appendix II - Simulator Malfunctions Page 11'of 14 Eagg Selection Descriotion M23 - MISCELLANEOUS / AUXILIARIES 147 INSTRUMENT' AIR SYSTEM FAILURE 148A STANDBY SERVICE WATER-PUMP-A TRIP 148B -STANDBY SERVICE WATER PUMP-A TRIP 149 -HPCS STANDBY SERVICE WATER PUMP TRIP 150A PLANT SERVICE WATER PUMP A TRIP 150B PLANT SERVICE WATER PUMP B TRIP 150C PLANT SERVICE WATER PUMP C TRIP-150D P1 ANT SERVICE WATER PUMP D TRIP 150E PLANT SERVICE WATER. PUMP E TRIP 150F PLANT SERVICE WATER PUMP F TRIP 151A COMPONENT COOLING WATER PUMP A TRIP 151B COMPONENT COOLING WATER PUMP B TRIP 151C COMPONENT COOLING WATER PUMP C TRIP' M24 - MISCELLANEOUS / AUXILIARIES , 152A TURB BLDG COOLING WATER.PMP A TRIP 152B TURB^ BLDG COOLING WATER PMP B TRIP 152C TURB - BLDG COOLING WATER FMP C TRIP 153A LOSS OF-DW COOLER 1 FANS A &-B-153B LOSS OF DW COOLER 2 FANS A & B 153C LOSS OF DW COOLER 3 FANS A & B 153D LOSS OF-DW COOLER 4' FANS =A & B 153E- LOSS OF DW COOLER 5 FANS A:& B 153F LOSS OF DW-COOLER 6 FANS A & B 154A SGTS--TRAIN A - HIGH DIFFc PRESSURE 154B SGTS TRAIN B - HIGH DIFF PRESSURE 155A TRIP OF SGTS. EXHAUST FAN-A 155B TRIP OF SGTS EXHAUST FAN B ?
1 l Grand Gulf Nuclear Station Simulator certification-Initial Report, March 1991 i Appendix II - simulator Malfunctions Page 12 of 14 Eagg Selection Descriotion M25 - MISCELLANEOUS / AUXILIARIES 156 OFF-GAS POST TREAT LOG RM K601B DOWNSCALE 157A-S PROCESS RAD MONITOR (SEE M27) 158 AREA RAD (VAR) (SEE M28 -M30) 159A MOV FAILURE E22-F004 HPCS INJECTION 159B MOV FAILURE E12-F008 S/D COOLING SUCTION-159C MOV FAILURE S12-F028B ' RHR CONTAINMENT SPRAY 159D MOV-FAILURE P41-F005A SSW A RETURN 159E MOV FAILURE N19-F040B LP HEATER DISCHARGE 160(X) ANNUNCIATOR (X) 174A FAIL DIV 1 DRYWELL PRESS HI ISOLATION 174B FAIL DIV 2 DRYWELL PRESS HI ISOLATION 175A FAIL DIV 1 RPV WTR LVL LO-LO ISOLATION 175B FAIL DIV 2'RPV WTR.LVL LO-LO ISOLATION 181A INADVERTENT SPMU A INITIATION 181B INADVERTENT SPMU B INITIATION M26 - S R V - ASSIGNMENTS (Reference Malfunctions 59, 60, and 61 on page-Mll) A - F041D *(D) B - F041F *(B) C - F041K * (B) D - F047A *(A).- E - F047L ~ * (C) F - F051A *(A) G - F051B * (B) H - F051C *(C) I - F041A- _(A) J - F041B (B) K - F041C (C) L - F041E (A) M - F041G (C)- N - F047C (C) 0 - F047D (D) P'- F047G (C) Q - F047H (D) R - F051D (D) S - F051F (B) T F051K (B) 1
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 i Appendix II - simulator Malfunctions 1 Page 13 of 14 PAHA Selection Descriot_LSD M27 - MALF 157 - PROCESS RAD ASSIGNMENTS 157A MAIN STEAM LINE 1 157B OFF GAS PRE-TREAT 157C OFF GAS POST-TREAT A 157D OFF GAS POST-TREAT B 157E COMPONENT COOLING WATER 157F CONTAINMENT VENT 157G OG & RADW BLDO VENT 157H SERVICE WATER EFFLUENT A 157I SERVICE WATER EFFLUENT B 157J CNTMNT & DW VENT EXH B/C 157K CNTMNT & DW VENT EXH A/D 157L AUX BLDG FUEL AREA VENT EXH B/C 157M AUX BLDG FUEL AREA VENT EXH A/D 157N AUX BLDG FUEL POOL EXH B/C 1570 AUX BLDG FUEL POOL EXH A/D 157P TURB BLDG VENT 157Q FUEL HANDLING AREA VENT-157R CONTROL ROOM VENT B/C 157S CONTROL ROOM VENT A/D M28 - MALF 158 - AREA RAD ASSIGNMENTS A01 - RHR ROOM A A02 - RHR ROOM:B-A03 - RCIC ROOM A04 - COMP CLG WTR HX A05 --TIP MECHANISM AREA A06 - DRWL EQUIP: HATCH A07 - DRWL PERS AIRLOCK A08 - CTMT PERS AIRLOCK A09'- CRDH HYD UNITS NORTH A10 - CRDH HYD UNITS SOUT1: All - RHR HX A HATCH A12 - RHR HX B HATCH A13 - SGTS FILTER TRAIN A14 - CRDH REPAIR ROOM A15 -~OUTSIDE CRDH REPAIR ROOM A16 - AUX BLDG SAMPLE STATION I
.__.__m_____.m_-.____ _ -- -. -
7 i Grand Gulf-Nuclear Station Simulator Certification Initial Report, March 1991 Appendix II - simulator Malfunctions Page 14 of 14 Paq9 Selection Descriotion M29 - MALP 158 - AREA RAD ASSIGNMENTS A17 - CTMT VENT EQUIP ROOM A18 - H2 SAMPLE PANEL A A19 - H2 SAMPLE. PANEL B A20 - CTMT VENT FILTER TRAIN A21 - CTMT SAMPLE STATION A22 - FUEL HANDLING AREA NE A23 - FUEL HANDLING AREA E A24 - FUEL HANDLING AREA SW B01 - FUEL: HANDLING AREA W B02 - DRYER STORAGE AREA B03 - SEP STORAGE AREA B04 - CTMT FUEL AREA N B05 - CTMT FUEL AREA S B06 - CTMT PERS AIRLOCK B07 - TURB SLDG FILTER TRAIN B08 - TURB BLDG SAMPLE STATION M30 - MALF 158 - AREA RAD ASSIGNMENTS B09 - MECH VAC PUMP AREA B10 - TURB BLDG INST RACK Bil - RX FEED PUMP AREA-B12 - TURB BLDG OPER FLOOR NE B13 - TURB BLDG OPER FLOOR NW B14 - TRUB BLDG OPER' FLOOR SW B15 - TURB BLDG OPER FLOOR SE B16 - RMT SHUTDOWN AREA B17 - HOT MACHINE SHOP-B18 - RAD BLDG INST RACK B19 - RAD BLDG-SAMPLE STATION. B20 - RAD BLDG CONTROL STATION B21 - DISTLT SAMPLE TANK ROOM B22 - RAD BLDG'HVAC ROOM' B23 - SOLID RADWASTE AREA B24 - CONTROL ROOM i
Appendix III Simulator Remote Functions
. v , . ,- -
1 Grand Gulf Nuclear St> tion Simulator certification Initial Report, March 1991 Appendix III - Simulator Remote Functions Page 1 of 13 AcDendix III Simulator Remote Functions The remote functions listed in Appendix III-are provided for the instructor to act in the capacity of auxiliary or other operators remote from the control room. The remote functions are listed by remote function page number which are grouped by system or category. Eagg Number Description Condition R01 NUCLEAR INSTRUMENTATION' 263 SRM SHORTING LINK A INSTL/aEMV 264 SRM SHORTING LINK.B INSTL/REMV 26S SRM SHORTING - LINK C INSTL/REMV 266 SRM SHORTING LINK D INSTL/REMV 270 SRM CHANNEL B TEST-SW POSITION OPER/STDBY 271 SRM CHANNEL C TEST SW POSITION OPER/STDBY 272 SRM CHANNEL D TEST SW POSITION 'OPER/STDBY-273 SRM CHANNEL E TEST SW POSITION OPER/STDBY 274 SRM CHANNEL F TEST SW POSITION OPER/STDBY R02 NUCLEAR INSTRUHENTATION 275 IRM CHANNEL B TEST SW POSITION OPER/STDBY 276 IRM CHANNEL C TEST SW POSITION OPER/STDBY 277 IBM CHANNEL D TEST SW POSITION _OPER/STDBY , 278 IRM CHANNEL E TEST SW POSITION OPER/STDBY 279 IRM- CHANNEL F TEST SW POSITION OPER/STDBY 280 IRM CHANNEL G TEST'SW POSITION OPER/STDBY 281 IRM CHANNEL H TEST _SW POSITION OPER/STDBY l 256 APRM CHANNEL B TEST SW POSITION- .OPER/STDBY 257 APRM CHANNEL C TEST SW POSITION OPER/STDBY 258 APRM CHANNEL D TEST SW POSITION OPER/STDBY 259 APRM CHANNEL E TEST SW POSITION OPER/STDBY 260 APRM CHANNEL F TEST SW POSITION OPER/STDBY 261 APRM CHANNEL G TEST.SW POSITION OPER/STDBY 262 APRM CHANNEL H TEST SW POSITION OPER/STDBY
...,,a , - . . . . - , , y"-
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Appendix III - Simulator Remote Functions Page 2 of 13 Page Number Descriotion Condition R03 NUCLEAR INSTRUMENTATION 285 SRM CHANNEL A GAIN 0.1-10000 286 SRM CHANNEL B GAIN 0.1-10000 , 287 SRM CHANNEL C GAIN 0.1-10000 288 SRM CHANNEL D GAIN 0.1-10000 289 SRM CHANNEL E GAIN 0.1-10000 290 SRM CHANNEL F GAIN- 0.1-10000 291 APRM CHANNEL A GAIN 0.8-1.2
-292 APRM CHAUNEL B GAIN 0.8-1.2 293 APRM CHANNEL C GAIN 0.8-1.2 294 APRM CHANNEL-D GAIN 0.8-1.2 295 APRM CHANNEL E GAIN .0.8-1.2 296 APRM CHANNEL F GAIN 0.8-1.2 297 APRM CHANNEL G GAIN 0.8-1.2 298 APRM CHANNEL H GAIN 'O.8-1.2 R04 LPRM BYPASS TABLE 300 BYPASS LPRM POSITION XX-YY NORM / BYPASS-i 301 BYPASS LPRM POSITION XX-YY NORM / BYPASS 302 BYPASS LPRM POSITION XX-YY NORM / BYPASS.
303 BYPASS LPRM POSITION XX-YY NORN/ BYPASS 304 BYPASS LPRM POSITION XX-YY LNORM/BYPAS3 305 BYPASS LPRM POSITION XX-YY NORM / BYPASS 306 BYPASS LPRM POSITION XX-YY -NORM / BYPASS 307 BYPASS LPRM POSITION XX-YY. NORM / BYPASS 308 BYPASS LPRM POSITION XX-YY NORM / BYPASS 309 BYPASS LPRM POSITION XX-YY NORM / BYPASS
- Grand Gulf Nuclear Station-Simulator certification Initial Report, March'1991 Appendix III Simulator Remote-Functions
.Page 3 of-13 Pace Number Description Condition ROS EP ATTACHMENT TABLE-1 310 RCIC HIGH SP LEVEL SUCTION INTERLOCK. -_ OPER/ DEFEAT
, 311 HPCS HIGH SP LEVEL SUCTION INTERLOCK .OPER/ DEFEAT i 312~ E22-F004 LEVEL 18 CLOSURE-INTERLCCK_ ,OPER/ DEFEAT:
313- E12-F053A ISOLATION. OPER/ DEFEAT: 4 314 E12-F053B ISOLATION -OPER/ DEFEAT-315 RCIC DIV1 LOW RPV ISOL INTERLOCK OPER/ DEFEAT-316 RCIC DIV2-LOW RPV ISOLiINTERLOCK.
.OPER/ DEFEAT.
317- DW PURGE-CMP A - 15105-(EP2-ATT15)' -RESET / TRIP
-318- DW PURGE CMP B - 16204 . (EP2-ATT15)- RESET / TRIP J19 RCIC DIV1"ISOL-EP2 ATT3 OPER/ DEFEAT 4 l 320 RCIC DIV2 ISOL EP2 ATT3 -OPER/ DEFEAT = 1 321 ARI/RPT DC BRKR 72-11E64EP2 ATT 18 CLOSE/OPEN 322 ARI/RPT DC BRKR 72-11K26 EP2 ATT 18 CLOSE/OPEN
, 323 B21-K4A/K35A JUMPER EP2 ATT 9 RMVD/ INST , 324 B21-K4B/K35B JUMPER EP2 ATT~9- RMVD/ INST; 325~ B21-K4C/K35C JUMPER'EP2 ATT 9- 'RMVD/ INST-l 326 B21-K4 D/K35D JUMPER EP2 - ATT . 9 - RMVD/ INST. 327 C34A-K7A/F-RELAYS EP2 ATT~6 INST /RMVD-f .R06- EP ATTACHMENT TABLE l 328 BORON. ADDITION TO CST EP2 ATT 28- RMVD/ INST ! 329. GAG P11-F064 EP2 ATT 28 RMVD/ INST' i 330- GAG P11-F065 EP2'ATT 28 'RMVD/ INST-J 331 RELAY C11A-KA' JUMPER EP2 - ATT .- 2 0 - RMVD/ INST- , 332 RELAY C11A-KB JUMPER: EP2 ATT 20- RMVD/ INST 333 N2 BOTTLES TO ADS ~ AIR RECVRS EP2 ATTi7 RMVD/ INST 334- SCRAM AIR HEADER' VENT EP2,ATT 23 RMVD/ INST-
-l l
i 3 4 1 4 f v w -3 -w-+ _ ,.,-1 , ,-e,vw --w re .w e -e- :, e s #
- r-s- rs *-
-Grand Gulf-Nuclear Station Simulator Certification Initial Report, March 1991 Appendix III - simulator Remote Functions Page 4 of-13 Eagg Number Descriotion Condition R07 6.9 & 4.16 KVAC BRKR OC/BUV LOR TABLE 350 11HD OC LOR TRPD/ RESET-351 11HD BUV LOR TRPD/ RESET 352 12HE OC LOR TRPD/ RESET 353 12HE BUV LOR TRPD/ RESET 354 13AD OC LOR TRPD/ RESET-355 13AD BUV LOR TRPD/ RESET 356- 14AE OC LOR 'TRPD/RESETL 357 14AE BUV LOR TRPD/ RESET 358 18AG OC LOR TRPD/ RESET 359 18AG BUV LOR TRPD/ RESET 360 '28AG OC LOR TRPD/ RESET 361 28AG BUV LOR TRPD/ RESET R08 RADIAL WELL & SWITCHGEAR-370 MIRCROWAVE TRANSMIT SIGNAL TO-WELLS OFF/ON 371 INSTRUCTOR CONTROL OF SWITCHGEAR & PSW LCL/ REMOTE 372 BUS-18AG FDR FM BOP XFMR 13 - 152-1801 TRIP / RESET 373 BUS 28AG FDR FM BOP XFMR 23 --152-2801 TRIP / RESET 374 BUS 18AG FDR TO.LCC 18BG1 - 152-1802- TRIP / RESET 375 BUS 18AG-12AG XTIE;FDR - 152-1808- -TRIP / RESET 376 BUS 28AG FDR TO'LCC 28BG1 - 152-2802 TRIP / RESET-377 LCC 18BG1 FDR FM BUS 28AG 18101- TRIP / RESET 378 LCC 28BG1 FDR FM BUS 28AG. 28101 TRIP / RESET 379 LCC'18BG1/28BGl'XTIE FDR 18102 TRIP / RESET' 380 PLANT SERVICE' WATER' PUMP A STATUS OFF/ON 381 -PLANT SERVICE WATER PUMP B r !.TUS _OFF/ON:
382 PLANT SERVICE WATER PUMP C S1AlsS -OFF/ON 383 PLANT SERVICE WATER PUMP D STATUS OFF/ON 384 PLANT SERVICE' WATER PUMP E STATUS OFF/ON. 385- PLANT SERVICE WATER PUMP F' STATUS. -OFF/ON 386 PLANT SERVICE WATER PUMP J STATUS OFF/ON. 387 PLANT SERVICE WATER PUMP K STATUS OFF/ON
Grand Gulf Naclear 8tation simulater Certification , Initial Reps ,t, March 1991 Appendix III - Simulator Remote Functions {., Page-5.of 13 1 EaE2 Number DescriotiQD Condition R09 EP-2 ATTACHMENT 13, 16, & 17 390 M71-R1, M1 TO T1, EP-2, ATT 17 RMVD/ INST 391 M71-R1, M4 TO T4, EP-2,'ATT 17 & ATT-13 RMVD/ INST 392 M71-R14, M2 TO T2, EP-2, ATT 16 RMVD/ INST-393 M71-R16, M1 TO T1, EP-2, ATT'17 RMVD/ INST M71-R16, M4 TO T4, EP-2, ATT 17 & 13 394 RMVD/ INST 395 M71-R28, M4 TO T4, EP-2, ATT 16 RMVD/ INST 396 T48-R11, M1 TO T1, EP-2, ATT 16 RMVD/ INST 397 T48-R11, M2 TO T2, EP-2, ATT 16 &'ATT 13 .RMVD/ INST-398 T48-R22, M1 TO T1, EP-2, ATT 16 RMVD/ INST 399 T48-R22, M2 TO T2, EP-2, ATT 16 & ATT 13 RMVD/ INST-R12 REACTOR PROTECTION 200 RECIRC PUMP A RPT NORM / BYPASS 201 RECIRC PUMP B RPT NORM / BYPASS 202 ALTERNATE FEED TO RPS BUS-A OFF/ON' 203 ALTERNATE FEED TO RPS BUS-B OFF/ON 204 RESET RPS A MG SET OFF/ON-205 RESET RPS B MG SET OFF/ON 206 RESET RPS BUS A ALTERNATE ' EPA BRKR TRPD/ RESET 207 RESET RPS BUS B ALTERNATE EPA BRKR - TRPD/ RESET-208 ISOLATION VLV STATUS PANEL (P858)' OFF/ON 209 REM STDWN PNL ISOL SWITCH OFF/ON~ 214 SCRAM SOLENOID BRKRS CB2A &-CB8A OPEN/CLOSE 215 SCRAM SOLENOID BRKRS CB2B &.CB8B. OPEN/CLOSE 216 RPS SCRAM SETPOINTS IIDOP/21DOP LEAK DETECTION 210 RWCU ISOLATION A' BYPASS NORM / BYPASS' 211 RCIC ISOLATION.A BYPASS NORM / BYPASS 212 RHR ISOLATION A BYPASS NORM BYPASS 213 MSIV LEAKAGE CONTPL SYS ACT NORM / ACTIVE-f
- s. ..____._._m _ . . _ . . _ _ - - - - - - - - - - - - - - - - - - - - - --
1 Grand Gulf Nuclear station " simulator certification Initial Report, March 1991 > Appendix III - simulator Remot3 Functions-
-Page 6 of 13-
\
PAgg Number Descriotion Condition-- R13 CONTROL ROD DRIVE HYDRAULIC 030 CRDH PUMP A DISCHARGE VALVE C11-F217A :0-100%- 031 CRDH-PUMP B DISCHARGE VALVE C11-F217B 0-100% 032 DRIVE WATER FLOW CONTROL VLV SEL A/B 033 RECIRC PUMP A SEAL PURGE FLOW IN/0UT 034 RECIRC-PUMP B SEAL PURGE FLOW IN/ CUT-REACTOR WATER CLEANUP / INSTRUMENT AIR 035 RWCU FILTER /DEMIN~A: STATUS- _ IN/ CUP 036 RWCU FILTER /DEMIN B STATUS. 'IN/0Ur-- 037 CCW TO NRHX IN/0UT 038 SERVICE AIR COMPRESSOR A-LOCAL' CONTROL' OFF/CN 039 SERVICE AIR-COMPRESSOR B LOCAL CONTROL. OFF/CN 040 UNIT 1-' INST AIR COMPRESSOR STATUS OFF/STDBY: 041 UNIT 2 INST AIR COMPRESSOR STATUS -OFF/CH-042-SERVICE AIR COMPRESSOR'A-LOCAL / REMOTE III/REMCYPE - 043 SERVICE AIR COMPRESSOR B-LOCAL / REMOTE -LCL/ REMOTE R14 ROD CONTROL L INFORMATION SYSTEMi 060 POSITION' BYPASS-ROD XX-YY LNORM/ BYPASS 061 POSITION, BYPASS ROD-XX-YY _ NORM / BYPASS-062 POSITION. BYPASS ROD XX-YY NORM / BYPASS-063- POSITION -BYPASS ROD XX-YY: NORM / BYPASS. 064 POSITION BYPASS. ROD XX-YY -NORM / BYPASS 065 POSITION BYPASS ROD XX-YY. NORM / BYPASS-066 POSITION BYPASS ROD ~XX-YY NORM / BYPASS 067 _ POSITION-BYPASS ROD-.XX-YY NORM / BYPASS 068 --DISABLE-ROD XX-YY (DRIVE'BYP) NORM / BYPASS. 069 REFUEL BRIDGE POSITION- POOL / CORE
i l Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 Appendix III - simulator Remote Functions Page 7 of 13 Eagg Number Descriotion Q2ndition R15 RECIRCULATION SYSTEM-153 ISOL COOLDWN RATE TIME X (100) 1-100 154 RECIRC PUMP A LOCKOUT RELAY TRPD/ RESET-155 RECIRC PUMP B LOCKOUT RELAY TRPD/ RESET 156 RECIRC LOOP A, SUBLOOP 1 STATUS MAIN / READY 157 RECIRC LOOP B, SUBLOOP 2 STATUS MAIN / READY 158 RECIRC LOOP A, LEAD SUBLOOP LOOP 1/ LOOP 2 159 RECIRC LOOP B, SUBLOOP 1 STATUS MAIN / READY 160 RECIRC LOOP B, SUBLOOP 2 STATUS MAIN / READY 161 RECIRC LOOP B, LEAD SUBLOOP - LOOP 1/ LOOP 2 162- ARI/RPT TEST CHANNEL 1 OFF/ON 163 ARI/RPT TEST CHANNEL 2 OFF/ON 164 FLUX ESTIMATOR POWER SELECTOR SWITCH HIGH/ LOW 215 MASTER CONTROLLER SIG. ABN.-ANN. AT P634 RESET / TRIP 6 RECIRC FCV A MINIMUM POSITION JUMPER RMVD/ INST R16 RCIC, RHR, ECCS,-ADHRS ' 100 RCIC TURBINE OVERSPEED TRIP: - TRPD/ RESET 223 RCIC EXH DRN TRAP LEVEL NO 1EAY/IEAK EMERGENCY CORE COOLING SYSTEMS 103 FLUSHING WATER SUPPLY VLV E12-F020 OPEN/CLOSE , 104 WATER LEG-PUMP A -OFF/ON f 105 WATER LEG PUMP B - - OFF/ON 106 INJECTION' TEST PERM RHR A 0FF/ON 107 INJECTION TEST PERM RHR B OFF/ON~ 108 INJECTION-TEST PERM RHR C- OFF/ON 110 INJECTION TEST PERM LPCS' OFF/ON
-ALTERNATE DECAY HEAT REMOVAL SYSTEM (ADHRS) 198 ADHR BRKR STATUS CLSD/OPEN 199 ADHR A MODE TRIP ENABLE SWITCH . NORM /ADHRS 200 E12-F066B BRKR STATUS CLSD/OPEN 201 E12-F021-BRKR STATUS CLSD/OPEN
GrEnd Gulf Nuclear Station simulator certification Initial Report,-March 1991 Appendix III - simulator Remote Functions. Page 8 of-13 Page Number Descriotion Condition , R17 MAIN STEAM 020 SEAL STEAM CONDENSER EXHAUSTER OFF/ON~ 021 MAIN STEAM TO SSG N33-F003 OPEN/CLOSE 022 SSG SHELL-SIDE MOV, N33-F107 OPEN/CLOSE-023 MSR A/B 2ND STG EXCESS STEAM NORM / ALT 024 1ST STAGE MSR A WARMUP:VLV N11-F048A OPEN/CLOSE 025 2ND STAGE MSR A WARMUP VLV N11-F004A OPEN/CLOSE 026 IST' STAGE MSR B. WARMUP VLV N11-F048B OPEN/CLOSE 027 2ND- STAGE MSR B WARMUP VLV N11-F004B OPEN/CLOSE 028 B21-K1A/C BYPASS MSIV CLOSURE LVL 1 NORM / BYPASS 029 B21-K1B/D BYPASS MSIV CLOSURE LVL 1 -NORM / BYPASS R18. TURBINE / CONTROL 220 TURBINE ECCENTRICITY 0-15 MILE 221 START LOAD LIMIT DEVICE 0-100% 222 SPEED CHANGER 0-100% 223 BYPASS STARTING DEVICE '0-100% 224 OIL TANK LEVEL 0-100% 225 METAL TEMP LAG- NORM /NOLAG 226 DROOP UNIT OFF/ON 227 LOAD REFERENCE 110% BIAS OFF/ON 228 _OVERSPEED TRIP SEQUENCE . OFF/ON l 229 ELECTRIC BYPASS CONTROLLERS RESET NORM /COMM i 230 SEAL OIL PUMP A SWITCH OFF/ON: 231 SEAL OIL PUMP B SWITCH OFF/ON 232 SEAL OIL PUMP A,B,C IN STANDBY OFF/ON 233 SEAL OIL PUMP D SWITCH. OFF/ON. i
-0
. , . ~ _ . .
Grand Gulf Nuclear Station Simulator-Certification Initial Report, March 1991 i Appendix III - simulator Remote Functions-Page 9 of 13 Egg 2 Number Descriction Condition R19 GENERATOR 046 MAIN GENERATOR LOCKOUT RELAY RESET _ .TRPD/ RESET 047 MAIN GENERATOR H2 LINUP FILL / VENT 048 MAIN GENERATOR H2 FILL / VENT RATE -- 0/300 049 MAIN GEN H2. PURITY (VAR) 0-100% 050 ISOPHASE BUS DUCT BLOWER A STATUS STOP/RUN 051 ISOPHASE BUS DUCT BLOWER'B STATUS STOP/RUN 052 ISOPHASE BUS DUCT BLOWER A SWITCH OFF/ STANDBY 053 ISOPHASE BUS-DUCT BLOWER-B SWITCH -OFF/ STANDBY DIESELS 054 DG 11 LOR TRPD/ RESET 055 DG 12 LOR TRPD/ RESET 056 DG 13 LOR -TRPD/ RESET 070 DG DIV III EMERGENCY STOP RESET TRPD/ RESET 057 DG DIV I MAINTENANCE MODE OPER/MAINT 058 DG DIV II MAINTENANCE MODE OPER/MAINT 059 DG DIV III MAINTENANCE MODE OPER/MAINT 060 DG OIV 1 EMERGENCY START (P75-HS-MO21A)- DE-EN/ START 061 DG - DIV II EMERGENCY START (P75-HS-MO21B) - DE-EN/ START R20 ELECTRICAL DISTRIBUTION 085 SERVICE XFMRS LOCKOUT RELAYS TRPD/ RESET 086 BOP XFMRS LOR .TRPD/ RESET a 087 6.9 KV BOP-LOAD CTR XFMRS LOCKOUT _ TRPD/ RESET 088 4.16 KV BOP LOAD CTR XFMRS' LOCKOUT TRPD/ RESET 089 34.5 KV BUSES' LOR TRPD/ RESET-092 - ESF XFMRS LOR TRPD/ RESET 093 4.16 KV-ESF DIV I LOAD CTR XFMRS TRPD/ RESET 094 4.16 KV ESF DIV'II LOAD CTR XFMRS - TRPD/ RESET 095 4.16 KV ESF DIV III LOAD CTR XFMRS TRPD/ RESET 096 4.16 KV ESF DIV I-BUS LOCKOUT TRPD/ RESET 097 4.16 KV ESF DIV II BUS LOCKOUT TRPD/ RESET-098 4.16 KV ESF DIV III BUS' LOCKOUT TRPD/ RESET ' 099 480V MCC 15B42 FEEDER BREAKER TRPD/ RESET 100 480V MCC 16B42 FEEDER BREAKER TRPD/ RESET 217 ~ LOAD SHED SEQUENCING STATUS- OPER/INOP 218 DG III FEED 15AA/16AB INOP/OPER
'I l
l Grand Gulf'Nucl@#r station simulator certification-Initial Report, March 1991- . Appendix III - Simulator _ Remote Functions 4 Page 10 of 13 Eagg Number Descriotion ConditiQD R21 CONDENSER, AUX WATER 182 HOTWELL MAN MAKEUP VLV N19-F008 -OPEN/CLOSE 183 VACUUM-BKR WATER' SEAL N19-142A/CL OPEN/CLOSE-184 VAC.PMP COMMON SUCT VLV N62-F014 OPEN/CLOSE 185. CIRC WTR LUBE FM2 FR THRU P66-F942/F943 OPEN/CLOSE 186 CIRC WTR LUBE WTR FROM PSW P44-F867 OPEN/CLOSE 187 OFFGAS-SYSTEM STATUS NORM /ISOL 189 PLANT CHILLERS A,B, &-C RESET OFF/ON. . < 190 COOLING TOWER EAST HALF. NORM /ISOL 191 COOLING TOWER WEST HALF NORM /ISOL ! 192 TBCW VALVE P43-F289- .OPEN/CLOSE~ 219 PLANT CHILLER A LOCAL START /STOP STANDBY /ON 220 PLANT CHILLER B LOCAL START /STOP- STANDBY /ON i 221 PLANT CHILLER.C LOCAL START /STOP STANDBY /ON R22 FEEDWATER-149 HDT TO FP SUCTION-(N23-F054) CLOSE/OPEN' 150 RFPT A OVERSPEED TRIP RESET TRPD/ RESET 151 RFPT B OVERSPEED.TRIPERESET' TRPD/ RESET 152 RFPT A,B ECCENTRICITY 0-15 MILS 153- FW CLEAN-UP RECIRC -(N23-F078) CLOSE/OPEN
.154 CONDUCTIVITY HEATER; DRAIN TANK 0-10 UM/C 155 COND STORAGE TANK CONDUCTIVITY 0-10 UM/C 193 N21-F503A-VALVE CONTROL _ . AUTO / MAN-194 N21-F503A VALVE POSITION- 0-1 195 N21-F503B' VALVE' CONTROL AUTO / MAN l
196- N21-F503B. VALVE POSITION. . 0-1
- 222- FEEDWATER HEATERS 1-6 STARTUPLVENTS' CLOSE/OPEN 1
1 i
.m, - ,- , . _ , . . , , , . -
Grand Gulf Nuclear Station-Simulator certification-Initial Report, March 1991 Appendix III - Simulator Remote Functions Page 11 of 13 Page Number Description Condition R23 FEEDWATER 165 COND CLEANUP DEMIN A STATUS IN / OUT 166 COND CLEANUP DEMIN D STATUS IN / OUT 167 COND CLEANUP DEMIN C STATUS IN / OUT 168 COND CLEANUP DEMIN D STATUS IN-/ OUT 169 COND CLEANUP DEMIN E STATUS IN / OUT 170 COND CLEANUP DEMIN F STATUS IN / OUT 171 COND CLEANUP DEMIN G STATUS IN/ OUT 172 COND CLEANUP DEMIN M STATUS IN / OUT 176 PRECOAT FILTERS ISO VALVES OPEN/CICSE 177 PRECOAT A FILTER STATUS IN / OUT , 178 PRECOAT B FILTER STATUS IN /_OUT 179 PRECOAT C FILTER STATUS 197 IN / OUT CNDS BSTR PMP BYPASS N19-F043 CLOSE/OPEN R24 CONTAINMENT 120 CNMT VENT SUPPLY FANS 121 OFF/ON CGCS INLET-FROM CTMT CLG E61-F012 OPEN/CICSE 122 CGCS OUTLET FROM CTMT-CLG E61-F013 OPEN/CIDSE 123 CNMT/DW PURGE FANS 124 OFF/ON' CNMT COOLER FANS OFF/ON 125 RECIRC DAMPER M41-F005' OPEN/CLOSE 126 127 CTMT COOLING FILTER TRAIN INLET M41-F028 OPEN/CLOSE CNMT FILTER TRAIN FANS OFF/ON 128 CNMT COOLING SYSTEM EXHAUST M41-F003 OPEN/CLOSE 129 CNMT EXM FILTER TRAIN FANS 130 OFF/ON CTMT PURGE COMPRESSOR- OFF/ON-131 132 CTMT PURGE COMPRESSOR DISCH VLV E61-F015 OPEN/CICSE - CTMT PURGE VLV E61-F021 OPEN/CIOSE 133 PURGE-FAN DISCH TO CTMT M41-F014 OPEN/CIDSE 9 ______.______________m_ _ - . . . . _
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Appendix III - Simulator Renote Functions Page 12 of 13 Pace Number Descriotion Condition R25 CONTAINMENT 134 SUPPRESSION POOL CONDUCTIVITY .1-10 UM/C 135 FPCC PUMP /HX A STATUS OFF/CH 136 FPCC F/D A STATUS IN/ CUT 137 FPCC F/D B STATUS Dy0UP 138 HYDROGEN ANALYZER SAMPLE VLVS OPEN/CIOSE 139 DW H2 ANALYZER RANGE SEL A LO/HI 140 DW H2 ANALYZER RANGE SEL B LO/HI 141 CTMT H2 ANALYZER RANGE SEL A LO/HI-142 CTMT H2 ANALYZER RANGE SEL B LO/HI 145 M41-F013 & M41-F015 FUSE STATUS DVCUr 146 MAXIMUM DRYWELL TEMPERATURE 320-350F R26 SJAE VLVS 600 N62-F012A OPEN/CLOSE 601 N62-F001A -OPEN/CLOSE 602 N62-F002A OPEN/CLOSE 603 N62-F006A OPEN/CLOSE 604 N62-F011A OPEN/CLOSE-605 N62-F024A OPEN/CLOSE 606 N62-F003A P680/ INST 607 N62-F003 A VALVE POSITION 0-1 608 N62-PIC-R010A POSITION O - 150 ISIG 610 N62-F012B OPEN/CLOSE 611 N62-F001B OPEN/CLOSE 612 N62-F002B OPEN/CLOSE 613 N62-F0069 OPEN/CLOSE 614 N62-F01 -OPEN/CLOSE 615 N62-F024; OPEN/CLOSE-616 N62-F003B .P680/ INST 617 N62-F003B VALVE POSITION O- .1' 618 N62-PIC-RO10B-POSITION O - 150 PSIG t
)
l Grand Gulf-Nuclear Station simulator Certification Initial Report, March 1991 Appendix III - simulator Remote Functions Page 13 of 13 Eagg Number Dgggriotion Condition R27 MISCELLANEOUS VALVES 620 N23-F520A/B CONTROL AUTO / MAN-621 N23-F520A POSITION- 0-1 622 N23-F520B POSITION 1 SIMULATOR' OPERATING LIMITS 630 SUPPRESSION POOL TEMP >212'F LT/GT 631 CONTAINMENT TEMPERATURE >185'F LT/GT 632 DRYWELL TEMPERATURE ;>330*F LT/GT 633 CONTAINMENT PRESSURE >15 PSIG LT/GT 634 DRYWELL PRESSURE >30.PSIG LT/GT 635 REACTOR VESSEL PRESSURE >1563 PSIG LT/GT R28 BREAKER RACKOUTS 640 RHR PUMP A BREAKER IN/OUT . 641 RHR PUMP'B BREAKER IN/OUT 642 RHR PUMP C BREAKER IN/OUT 643 LPCS PUMP BREAKER IN/OUT-644 HPCS-PUMP BREAKER IN/OUT 645 CRD PUMP A BREAKER IN/OUT
- 646 CRD PUMP B BREAKER IN/OUT-647 SSW PUMP A BREAKER IN/OUT 648 SSW PUMP B BREAKER IN/OUT
Appendix IV Simulator Instrument overrides l 1 e d i 9
Grand Gulf Nuclear Station Simulator-Certification Initial Report, March 1991 I Appendix IV - Simulator Instrument Overrides Page 1 of 2 Accendix IV I i Simulator Instrument Overrides i The simulator instrument overrides listed in Appendix IV are providad for additional training capability. in addition to the malfunctions previously listed. These are listed by override -(OR) page numbers OR1 through OR3. Handswitches are designated with an S in their_ component number (e.g. E22A-S2 is-the HPCS initiation handswitch). Instruments,. recorders and- other. indicators are designated- with an R or N in their component number (e.g. -C41-R601 - is the Standby Liquid Control storage - tank level indicator). Overrides are activated on page IST of the instructor monitor by entering a line number and the instrument override number. - EAsp Number Comoonent I/O Override Parameters-OR1 1 E22A-S2 HPCS INITIATION 2 E51A-S37 RCIC INITIATION 3 E21A-S9 RHR A/LPCS INITIATION [ 4 E12A-S21 RHR B/RHR C INITIATION 5 C41-S1A SLC PUMP A: l 6 B21C-S30A ADS MANUAL INIT CHANNEL A 7 B21C-S31A ADS MANUALLINIT CHANNEL E= 8 B21C-S30B ADS MANUAL INIT' CHANNEL B 9 B21C-S31B ADS MANUALLINIT CHANNEL F 10 C71A-S1 REACTOR MODE SWITCH 11 C71A-S3A MANUAL-SCRAM DIV.1 12 C71A-S3C ' MANUAL SCRAM DIV 3 - 13 C71A-S3B MANUAL SCRAM DIV 2-14 C71A-S3D MANUAL SCRAM.DIV 4 - 15 B21H-S25A MANUAL ISOLATION DIV 1 16 B21H-S25C MANUAL ISOLATION DIV-3 17 B21H-S25B MANUAL. ISOLATION DIV 2 OR2 18 B21H-S25D MANUAL ISOLATION DIV 4 19 C41-R601 SDLC STORAGE TANK LEVEL' 20 E51-R602 RCIC TURBINE STEAM INLET PRESSURE. 21 B21-R605 SHUTDOWN RANGE LEVEL t 22 B21-R615B FUEL ZONE LEVEL 23 B21-R623A RECORDER WIDE RANGE LEVEL 24 B21-F623A RECORDER WIDE RANGE PRESSURE 25 B21-R623B RECORDER WIDE RANGE LEVEL 26 B21-R623B RECORDER, WIDE RANGE PRESSURE 27 C34-R605 WIDE RAhGE PRESSURE 28 B21-R604 WIDE RANGE' LEVEL a m-__. __.__m--.__.s.__._m___.m_--m__ -___
Grand Gulf NucleEr Station Simulator Certification Initial Report, March 19'Al Appendix IV - Simulator Instrument Overrides Page 2 of 2 Eggg Number Component I/O Override Parameters OR2 29 C34-R606A NARROW RANGE LEVEL
- 30 C34-R606B NARROW RANGE LEVEL 31 C34-R606C NARROW RANGE LEVEL 32 B33-R611A CALIBRATED JET FLOW A 33 B33-R611B CALIBRATED JET FLOW B OR3 34 C34-R609 RECORDER STEAM FLOW 35 C34-R609 RECORDER FEEDWATER FLOW 36 C34-R614 WIDE RANGE PRESSURE 37 C34-R614 WIDE RANGE LEVEL 4 38 C34-R615 NARROW RANGE PRESSURE-39 C34-R615 NARROW RANGE LEVEL 40 M71-R601B DRYWELL WIDE RANGE PRESSURE 41 M71-R601B CONTAINMENT NARROW RANGE PRESSURE #
42 B21-N669B-1 ATWS LEVEL B 43 B21-N658E-1 ATWS PRESSURE E 44 C61-R400A WIDE RANGE LEVEL 45 C61-R400D WIDE RANGE LEVEL , 46 C61-R401A WIDE RANGE PRESSURE 47 C61-R401B WIDE RANGE PRESSURE 48 C61-R402A SUPPRESSION POOL LEVEL 49 C61-R402B SUPPRESSION POOL LEVEL 50 C61-R403A SUPPRESSION POOL TEMPERATURE 4 h I i 4 1 1
.4
, - -4 . . r ,- .e - r 4 r - .
1D f (1 0 . 1 Appendix v Simulator Trainee Proficiency Review Parameters f b ( (
l Grand Gull' Uuoloar Station Simulator certification l Initial Report, March 1991 l Appendix V - l Trainne Proficiency Keview Par.tmeters rage 1 of 2 Apl2And.iK Y g , nee 1*oficiency Regview Paraneters ! Trainee proficiency review parametors listed in Appendix V can be l 7 elected on page TPR on the instructor monitor for tracking of I student or crew performance. Theso are listed by proficiency review p'.y' number (PR1 through PR3): Egye Number Parameter Unito PRI 1 SOURCE RAMGT MONITOR CHAN C CPM 2 SOURCE RANGE MONITOR CHAN E CFM 3 SOURCE RiU1GE MONITOR CHAN D CPM 4 BOURCE RANGE MONITOR CHAN F CPM S REACTOR PERIOD CHAN C SEC 6 REACTOR 1ERIOD CHAN D SEC 7 INTERMEDIATIf RANGE MONITOR CHAN C B INTERMEDIATE RANGE MONITOR CHAN E 9 INTERMEDI;.TE RAhGE MONITOE CHAN D i 10 INTERMEDIATE RANGE MONITOR CdAM F 11 AVERAGE POWER' RANGE MONITOR CHAN C l 12 A7ERAGE POWER RANGE MONITOR CHAN E l 13 A/FRAGE POWER FANGE MONITOR CHAN D 14 AVEilAGE POWER RANGE MONITOR CHAN F 15 MODE SWITCH POSITION t,6 SCRAM CHAN A (SCRAM / RESET) 17 SCBAM CHAN D (SCRAlf/REGET) PR2 Ifl NET REACTIVITY DOLLARS 19 REACTOR POWER - MWT MWt 20 RECIRC LOOP A ,FION LBM/HR 31 RECIRC LOOP B PLOW LBM/HR l 22 TOTAL CORE FLOW LBM/HR 23 ' ICTAL STEAM FLOW LBM/HR 24 TOTAL FEEOWATER FLOW LBM/HR 25 REACTOR WIDE RhNGE LIV'ZL INCHES 26 REACSOR FUIL ZONE LD EL INCHES 27 RSAC' CUR HIDH RAriGU ERESSURE PSIG 26 REACTOR REATUP/COOTAOWN RATE F/HR 29 FEEDWATER INEET TEMP TO RPV
- 7 30
- DRYWELL NEMPERATURE F 31 DRYWELL PRESSURE PSIG 32
- CONTAINMENT TEMPERATURE F 33 CONTAINMEl(T PRESSURE PSIG
'24 SUPPRESSION TEMPERATURE F
~. . Grand Gulf Wuolear station simulator certification Initial Report, March 1991 Appendix v - Trainee Proficiency Review Parameters Page 2 of 2 Page Number Parametar Units PR3 35 SUPPRESSION POOL LEVEL FEET 36 HP STOP VALVE CASING TEMP 'F j 37 HP TURBINE-CASING TEMP *F 38 HP TURBINE INLET PRESSURE PSIG 39 MAIN TURBINE SPEED RPM 40 MAIN GENERATOR MWe MWe 41 MAIN GENERATOR MVARs MVAR 42 HP CONDENSER VACUUM IN HG 43 CONDENSATE DEMINS DP PSID 44 REACTOR COOLANT CONDUCTIVITY UMHO 45 DIV I STBY D/G MWe MWe 46 DIV I STBY D/G MVARs MVAR 47 DIV II STBY D/G MWe MWe 48 DIV II STBY D/G MVARs MVAR 49 HPCS D/G MWe MWe 50 HPCS D/G MVARs MVAR l 4
\
. . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -- i
- . ~ . - . . _ . . . . - . . ..
Appendix VI Simulator Test Abstracts Computer Real-Time Tests I ( 1
l l Grand Gulf Nuclear station simulator certification Initial' Report, March 1991 simulator _ Test Abstracts , i Appendix VI -Computer Real Time Tests Page 1 of 2 i Computer Real-Time' Test 4 I. Introduction The computer real-time test verifies that simulation not running in real-time will alert the instructor to this condition. II. Purposer To test the simulntor master computer Central Processing Unit , (CPU), slave computer CPU, and slave computer Internal ; Processing Unit (IPU) operation to determine if non-real-time ( operation will be alarmed and indicated to the siinulator instructor at the instructor console. III. ANSI /ANS 3.5 Test recuirement ANSI 3.5 Appendix A'section A3.1 IV. Initial condit12DA: IC - 17 (See Appendix I) V. Epauence of events: A. This test is' performed on the simulator master computer CDU, slave computer CPU, and_ slave computer IPU. B. Fer the master computer CPU real1 time test, the 'following-l are-performed: The variable MSTSLIP -is set to zero. at the engineering i
- TDFfstation.
The ICP " Alarm Sound! INOP" light; is verified extinguished. 5 The program "MRT" is .run. The' ICP alarm bell is ; verified. to - sound.L The :ICP "Non' Real . Time" light ~ is varified to illuminate. The variabla "MASTSLIP" is verifled tof have -in' cremented by 4. 3
. g '
, w *~_---- ._w. w_.. -mmr, , ~_,,,%,, ,, ,-,eg..w. ,-,.-+c,--c-,.,,.i rmEyy '
--~..r-- ,
Grand Gulf Nuclear Station Simulator Certification Initici Report, March 1991 simulator Test Abstracts Appendix VI -Computer Real Time Tests Page 2 of 2 Af ter depressing the "Non Real Time" pushbutton, the "Non Real Time" light is verified to extinguish. C. For the slave computer CPU real time test, the following are performed: The variable SLAVSLIP is set to zero at the engineering TDF station. The ICP " Alarm Sound INOP" light is verified extinguishea. ' The program "SRT" is run. The ICP alarm bell is verifie6 to sound. The ICP "Non Real Time" light is verified to illuminate. The variable "SLAVSLIP" is verified to have incremented by 4. Af ter depressing the "Non Real Time" pushbutton, the "Non Real-Time" light is verified to extinguish. D. For the slave computer IPU real time test, the following are performed: The variable IPUSLIP is set to zero at the engineering TDF station. The ICP " Alarm Sound INOP" light is verified ; extinguished. l i The program "IRT" is run. The ICP ?larm bell is verified to sound. The ICP "Non Real Time" light'is verified to illuminate. 4 The variable "IPUSLIP" is 'terified to have incremented by 4.
)
, Af ter depressing the "Non Real Time" pushbutton, the "Non - Real Time" light is verified to extinguish. VI. Date of last test: February 1, 1991 VII. Concern (s)- found None XI. Resolution: None 1
I r I i i e
?
Appendix VII Simulator Test Abstracts i Steady State Tests t
?
k
-4
i Grand Gulf Nuotoar station ; simulator certification Initial Report, March 1991 l simulator. Test Abstracts Appendix VII - steady state Tests [ Page 1 of 10 , Appendix VII Simulator Steady State Test Abstracts ! l STEADY STATE TEST # 1 I. Introduction . The 100% steady state test verifies that simulator accuracies , are related to full power values over a sixty minute period. II. Purcoset To verify at 100% power conditions that simulator critical parameters aro stable and vary by not more _ than 2% from initial reference plant values and that the noncritical parameters are stable and va.y by not more than 10% from . reference plant values. ' III. ANSI /ANS 3.5 Test Recuirement e : ANSI 3.5 4.1, 5.4.2(2), Appendix A A3.2(1), , Appendix B Bl.1 IV. Initial conditionst l Steady state plant conditions for 100% power. - V. Sectuence of events:- A. The simulator is initialized to within baseline comparison data values, i.e. as close to critical and non - ! l critical plant parameters as possible. -- t i B. Data is collected for one hour. C. No operator actions are required. . D. plots are made of critical and non critical parameters and the results- are reviewed to verify critical and
- noncritical. parameters are within criteria except' as ,
noted. ' E. .The following critical parameters are verified to be ! I within 2% of their comparison values:- i t t i
. , . _ . . - _ _ . . _ _ . _ _ _ _ , ., _._I_ _
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VII - steady state Tests Page 2 of 10 Reactor Thermal Power Total Feed Flow Reactor Dome Pressure Generator Output MW Electric Recirculation Loop A Flow Recirculation Loop B Flow F. The following noncritical parameters are verified to be within 10% of their comparison valuest Reactor Neutron Flux Reactor Narrow Range Level Total Steam Flow Total Core Flow Total control Rod Drive Hydraulic Flow Reactor Water Cleanup system Inlet Flow Reactor Water Cleanup system Return Flow (Note RWCU return flow was not taken in plant data. Outlet flow was verified to be within 10% of inlet flow. Reactor Water Cleanup system Return Temp HP Condenser Pressure-IP Condenser Pressure LP Condenser Pressure i Main Generator Output Voltage (Note: Generator output voltage was not taken in plant data, but was verified to be within 10% of 22 KVAC. Main Generator Hydrogen Pressure (Notet Generator hydrogen pressure was not taken in plant data, but was verified to be within 10% of 60 psig.) Main .nerator Reactive Load MVAR Output (Note: Generator )W)@ was not taken in plant' data, but was verified to be within , 100 MVAR of 0 MVAR. ! G. The following parameters are monitored for data only' and no comparison Control Rod Drive Hydraulic Temperature
.______m_m.-_m_ _ - - - - - - _ - _ - - - - - - -
4 Grand Gulf Nuclear statiJn simulator Certification Initial Report, March 1991 Simulator Test Abstracts ' Appendia VII - steady state Tests :
- Page 3 of 10 i
Ambient Air Temperature
- Feed Line A Temperature to Reactor "
Feed Line B Temperature to Reactor Feed Temperature to Vessel Condenser Hotwell Temperature i LP Circ Water Inlet Temperature IP Circ Water Inlet Temperature ! HP Circ Water Inlet Temperature , HP Circ Wster Outlet Temperature SJAE A Steam Flow
- SJAE B Steam Flow 7 MSR lst Stage Steam Flow ,
MSR 2nd Stage Steam Flow ; Dewpoint Temperature VI. Comparison Data Grand Gulf " Incremental Heat Rate Testing" Procedure 17-S-03-11 run April - June of 1990. VII. Date of last test February 1, 1991 . , VIII. Concern (s) found: HP, IP, and LP condenser pressures could not be initialized within 10% of the reference plant values. These pressures , were slightly lower than the reference plant values and were > stable. - l MSR A and B first stage. steam flows were zero. These l l monitored parameters were' not used for comparison- with 4 reference plant data. IX. Rosolution: 4 l These deficiencies will be corrected during~model replacements in the simulator upgrade project. i t
, , , ~ - - -
e ., -,--er- --m, e e -s 9 , -,m,a ~w ,
- e,-
--.-- . - . - - . . . - . . . . - . . - - - - - . _ - . - - . . ~ . - - ~ - .-
I ! Grand Gulf Nuclear station l l simulator certification j Initial Report, March 1991 i simulator Test Abstracts ! Appendix VII - steady state Tests , I Page 4 of 10 { 6 STEADY STATE TEST # 2 i I. Introduction ! The 60% steady state test verifies that simulator accuracies , are related to full power values over a sixty minute period. , II. PURPOSEt i To verify at 60% power conditions that simulator- critical parameters are stable and vary by not more than 24 from initial reference plant values and that the noncritical ; parameters are utable and vary by not more than 10% from i reference plant values. , i III. ANSI /ANS 3.5 Test recuirements i ANSI 3.5 4.1, 544.2(2), Appendix A - A3.2 (1), ! Appendix B Bl.1 j IV. Initial Conditiqngt steady state plant conditions for 60% power. , i~ V. Secuence af eventst l l The simulator is initialized to within baseline comparison ! data values, i.e. as close to critical and non critical plant - parameters as possible, j Data is collected for one hour.
- No operator actions are required..
Plots are made of critical and non critical parameters and-the results are reviewed to verify - critical -and noncritical parameters are within criteria except as noted.. A. The following critical: parameters are Verified to.be
.within 2%.of their comparison valuest i Reactor _ Thermal Power
. Total Feed Flow
- Reactor Dome Pressure Generator. Output MW Electric Pecirculation Loop A Flow ,
4
- k i
. . . . - _ . _ . . , , -. . . . ,,.m.
4 ; l Grand Gulf Nuclear station I simulator certification ) Initial Report, March 1991 i simulator Test Abstracts 1 Appendix VII - steady state Tests Page 5 of 10 Recirculation Loop B Flow I D. The following noncritical parameters are verified to be within 10% of their comparison values: Reactor Neutron Flux Reactor Narrow Range Level Total Steam Flow Total Core Flow Tota? Control Rod Drive Hydraulic Flow Reactor Water Cleanup system Inlet Flow ReaStor Water Cleanup system Return Flow (Notes RWCU Return flow was not taken in plant data. Outlet flow was verified to be within 10% of inlet flow. Reactor Water Cleanup system Return Temp HP Condenser Pressure IP Condenser Pressure LP Condenser Pressure Main Generator output Voltage Main Generator Hydrogen Pressure l Main Generator Reactive Load MVAR-Output I (Note: Generator MVAR was.not taken in plant data is verified to be within 100 MVAR of 0 MVAR. C. The following parameters are monitored for data only and l no comparison: Control Rod Drive Hydraulic Temperature Ambient Air. Temperature Feed Line A Temperature to Reactor j Feed Line B Temperature to Reactor l Feed Temperature to Vessel l Condenser Hotwell Temperature LP Circ Water Inlet Temperature IP Circ Water-Inlet. Temperature ! HP Circ Water Inlet: Temperature i HP Circ Water Outlet Temperature l SJAE A Steam' Flow-l SJAE B Steam Flow MSR lst Stage Steam Flow MSR'2nd Stage Steam Flow , Dewpoint. Temperature !
-0
1 i l Grand Gulf Nuclear station
- simulator Certification i
~ Initial Report, March 1991 simulator Test Abstracts Appendix VII - steady state Tests Page 6 of 10 VI. Comparison Datat Grand Gulf " Incremental Heat Rate Testing" Procedure 17-S-03-11 run Apr11 - June of 1990. VII. Date of last test: February 1, 1991 VIII. Concern (s) found Generator output (MWe) could not be initialized Within 2% of the reference plant value. It was slightly higher than the reference value and was stable. HP, IP, and LP condenser pressures could not be initialized within 10% - of their reference r3 ant values. These were slightly lower than their referen o values and were stablo. IX. Resolution: These deficiencies will be corrected during model replacements in the simulator upgrade project. 4
-+
l
Grand Gulf Nuclear station simulator certification Initial Report, March 1991 ! simulator Test Abstracts Appendix VII - steady state Tests Page 7 of 10 STEADY STATE TEST # 3 I. INTRODUCTION The 20% steady state test verifies that simulator accuracies are related to full power values over a sixty minute period. II. PURPOSEt To verify at 20% power conditions that simulator critical parameters are stable and vary by not more than 2% from initial . - ref erence plant values and- that -the noncritical-parameters are stable and vary by not more than 10% from reference plant values. III. ANSI /ANS 3.5 Test requirement: ANSI 3.5 4.1, 5.4.2(2), Appendix A A3.2(1), Appendix B Bl.1 IV. Initial gonditions: Steady state plant conditions for 20%0 power. V. ? quange of events: The simulator is initialized to within' baseline comparison data values, i.e. as close to critical and non' critical plant , i parameters as possible. Data is collected-for one-hour. No operator actions are required. Plots are made of' critical and non critical parameters and the results tre reviewed. to verify critical and noncritical parameters vi'c within criteria except as noted.. .i
'A. The following critical parameters. are verified' to ' be within 2% of their comparison valuest:
Reacter Thermal-Power. Total Feed Flow Reactor Dome Pressure i i
l t l Grand Gulf Nuclear Station simulator Certifi stion Initial Report, March 1991 1 simulator Test Abstracts ; Appendia VII - steady state Tests l Page 8-of 10 i J. > , Generator output MW Electric- , Recirculation Loop A Flow 1 Recirculation Loop B Flow B. The following n.oncritical parameters'are verified to be within 10%.of their comparison values: Reactor Neutron Flux Reactor Narrow Range Level Total Steam Flow-Total Core Flow Total Control Rod Drive Hydraulic Flow: Reactor Water Cleanup system Inlet Flow = Reactor Water Cleanup system Return Flow (Note: RWCU return flow was not taken in
. plant data. Outlet flow was verified to be within-10% of inlet flou.
i Reactor Water Cleanup system Return TempL HP condenser Pressure IP Condenser Pressure LP Condenser Pressure-Main Generator Output Voltage (Note: Generator ~ output Voltage .was not-taken in plant data, but was verifiedL to be within 10%,of 22~KVAC. Main Generator Hydrogen Pressure-(Note: Generator hydrogen pressure was-not 4 taken in plant data, but was veriflad'
- to be within.10%fof-60 psig.)
Main Generator Reactive Load MVAR Output . , (Note: - Generator MVAR was not taken in plant -
- data, but was verified to be-within
- 100.MVAR of.O MVAR.
C.- The following; parameters are monitored' for data only and uno comparisons-Control Rod-Drive-Hy'draulic Temperature- ;
- Ambient' Air ~ Temperature Feed:LinelA-Temperature-to Reactor Feed-Line B-Temperaturefto Reactor
,,, r-ww >r-.-- , - - - pc-.- ,.,.m,,,,y- 9 - m- e+-. a-y+y- y oA-m y y g _a,yg4-.,.gs,.g..- 7 y ,,,3,eg, 9._g,an_c e s g o . c ye 9 y wW -9 m -- e af ca-yqq y wyr 9 egqq-y ,4 v-
orand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VII - Steady State Tests Page 9 of 10 Feed Temperature to Vessel Condenser Hotwell Temperature LP Circ Water Inlet Temperature IP Circ Water Inlet Temperature HP Circ Water Inlet Temperature HP Circ Water Outlet Temperature SJAE A Steam Flow SJAE B Steam Flow MSR lot Stage Steam Flow MSR 2nd Stage Steam Flow Dewpoint Temperature VI. Comparison Data: Grand Gulf " Incremental Heat Rate Testing" Procedure 17-S-03-11 run April - June of 1990. VII. Date of last test January 31, 1991 VIII. Concern (s) found/ status Generator output (MWe) could not be initialized within 2% of l the reference plant value. It ' was slight lower than the reference value and was stable. i l HP, IP, and LP condenser pressures could not be initialized within 10% of their reference plant val"ea. These were slightly lower than their reference values tore stable. Total steam flow could not be ini',talized , ..in 10% of its reference plant value. It was slightly lower than its reference value and was stable. IX. Resolution: These deficiencies will be corrected during model replacement in the simulator upgrade project. l G
1 l ! h Grand Gulf Nuclear Station ! simulator certification i Initial Report, March 1991 ! simulator Test Abstracts ! Appendix VII - steady state Tests ! Page 10 of 10 > ( STEADY STATE TEST # 4 i Introduction I. r The 100% stability tust verifies that simulator accuracies [ ] are related to full power values over a sixty minute perled. ! II. PURPOSEt To verify that the simulator computed values at 100% reference ! plant conditions will be stable and not vary more than 2% of i that initial value over a 60 minute time period. [ III. ANSI /ANS 3.5 Test recuirements f , i ANSI 3.5 4.1, 5.4.2(2), Appendix A A3. 2 (1) , , Appendix B Bl.1 i IV. Initial Conditionst : Steady state plant conditions for 100% power. V. Secuence of events: A. The simulator is initialized to within baseline comparison data values, i.e. as close to critical and non ; critical. plant parameters as possible. B. Data is collected for one hour. , C. No operator actions are required. f l D. Comparison is made between the initia1' simulator data at. l time 0 minutes.to all simulator data collected on that ' j parameter over a-time period of one_ hour. These values; i are verified to not dif fer by more than :2%. l l- VI. Comparison' Data [ Steady State Test number 1 simulator download data. t VII. Date'of last test: February 1, 1991 VIII._ Concern (s) foundt None IX. Resolution: None i t w m -m -, -
, , -ew-, -
+~ r , < , + , + , -
_ .m_ _ _ _ _. -. . _ . . I i 1 l l i 1 i f Appendix VIII Simulator Tests Abstracts Transient Tests
- i
( h 9
4 ) Grand Gulf Wuclear station ! simulator certification Initial Report, March 1991 ; simulator Test Abstracts Appendix VIII - Transient Tests : Page 1 of 70 l i Accendix VIII i Simulator Transient Test Abstracts l
-r Transient Test #1A ;
I.
Purpose:
To test the response of the simulator to a manual reactor scram at 100% power. t i i II. ANSI /ANS 3.5 Test requirements . ANSI 3.5 3.1.2 (19) ! ANSI 3.5 Appendix B, Bl.2 (1) ! l ? III. Initial Conditions: l Beginning of cycle ! Core thermal power 3993 MWt -[ Core flow 101% l i Feedwater temperature 425'F Reactor dome pressure 1045 psig [ i IV. Sequence of events:
]
The transient is introduced by manually Scramming'the reactor with manual scram pushbuttons'on; panel 1H13-P680.- -y Ten seconds after the manual ceram the reactor mode switch is I placed to Shutdown on panel 1H13-P680.- 3 After verifying reactor. level increasing toLapproximately 18E inches, one of the running reactor . feed pumps is manually i tripped. - i No;other operator actions are'taken. A. The following' critical parameters are recordedL for ' 3-minutes and later-plotted out for analysis: .
,v. , , . , _
i l i Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests Page 2 of 78 Reactor Power Total Steam Flow Total Feed Flow Reactor Wide Range Pressure Reactor Narrow Range Pressure Roactor Wide Range Level Reactor Narrow Range Level l Generator Electrical output MWe Turbine steam flow e Total Core flow Recirculation loop A flow rate Recirculation loop B flow rate ; Automatic actions are verified. V. Comparison data: No reference data is availableLfor a manual scram from rated-power. Since the test is similar to transient test 1B, the results of the plant data from this test were used for comparison-and expected trends. , VI. Date of last test 1/25/91 , VII.- . concern (s) foundt Level' did not shrink to below level- 3 as expected. . As
-a result,. Recirculation-pumps did not get a. signal-to transfer to slow speed and a setpoint setdown function did not occur in'feedwater control-.
The B RFP was manually tripped when level ' dropped - to about.18 inches.and was recovering. A recirculation FCV; runback 'occurredJsince -recirculation pumps were still- , i
, - - e - - -+
Grand Gulf Nuclear station i simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests
- Page 3 of 78 running in fast speed.
Total steam flow drop was not as fast as expected. Steam flow did not steady at about 1 MLBM/HR after the scram. VIII. Resolution: These deficiencies- will be corrected with tL9 model replacements in the simulator upgrade project. l l t r
~ _ , _. - ,- . - ,, , , , . , .,
Grand Gulf Nuclear station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests Page 4 of 78 Transient Test #1B I.
Purpose:
To test the response of the simulator to manual Scram from approximately 58% power. II. ANSI /ANS 3.5 Test requirement: ANSI 3.5 3.1.2 (19) III. Initial conditions: , Beginning of cycle Reactor power 57.4% Core fl/-- 36.64% Feedwater temperaturc 371*F Reactor dome pressure 969 psig Reactor narrow range level approximately 31 inches. IV. Sequence of events: The transient is introduced by manually Scramming - the reactor by placing the reactor mode switch is placed to Shutdown on panel 1H13-P680. After verifying reactor level increasing to approximately 18 inches, one of the running reactor feed pumps is . i manually tripped. No other operator actions are taken. A. The following critical parameters are recorded .for. 5- ; minutes and later plotted out for analysis: Reactor Power Total-Steam Flow
. Total Feed Flow.
g .-
Grand Gulf Wuolear Station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests Page 5 of 78 Reactor Wide Range pressure Reactor Narrow Range Pressure Reactor Wide Range-Level-Reactor Narrow Range Level l Generator Electrical output MWe q Turbine steam flow Total Core Flow Recirculation loop A flow rate Recirculation loop B flow rate Automatic actions are verified. V. ' Comparison data GETARS data from Scram number 58 for comparison of the following parameters: Reactor power Total steam flow i Total feed. flow , Reactor wide range pressure s Reactor narrow range pressure t Reactor wide range level Reactor narrow range level Total core' flow Generator electrical output KWe i Recirculation' loop:A flow rate
- Recirculation -loop--.B : flow rato- '
Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 simulator Test Abstracts I Appendix VIII - Transient Tests-Page 6 of 78 I VI. Date of last test 2/1/91 VII. Concern (s) found Scram 58 occurred in the middle of a transient on Feedwater temperature after a rapid downpower maneuver. The simulator initial conditions could not be established within the stated band for feedwater temperature. This did not appear to impact the test results. Total steam flow did not drop as fast as plant data indicates. Total steam-flow did not stabilize at about 1 MLBM/HR as expected after the scram. Level did not shrink below level 3 as expected. Plant data indicated a drop of about 40 inches ) should have occurred. As a result, the setpoint ' setdown function c2 feedwater control did not occur. Recirculation pumps did not automatically transfer to slow speed at the same relative time and was delayed until feed flow dropped to the cavitation interlock setpoint. One reactor feed pump was manually tripped when level dropped to +18 .nches and was recovering. A- recirculation FCV runback signal occurred since recirculation pumps were still running in fast speed. Recirculation FCVs were-already closed below 15% position. No valve movement resulted.- VIII. Resolution: These deficiencies will be corrected with the model ; replacements in the simulator' upgrade project. ! i I 4 a--_---.w.--,---- --ma-.-------- - _ _ - - - - _ - - - - - - - . _ . - - .
- =. . - . - . . - - _ _ . .- ..-.- _. _.- .- . . - - -
P Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests Page 7 of 78 Transient Test #2 I.
Purpose:
To test the response of the pimulator to a loss of normal feedwater. II. ANSI /ANS 3.5 Test requirement: ANSI 3.5 3.1.2 9 ANSI 3.5 Appendix B, Bl.2-(2) III. Initial Conditions: Beginning of Cycle conditions Reactor power (96.5%)
- 103% Core flow '
Feedwater level 34 inches, RPV dome pressure 1020 psig. IV. Sequence of events: The transient is-introduced by inserting the following malfunctions to occur siraltaneously: Malfunction 115 A, Condensate Pump'A Trip Malfunction 115 B, Condensate Pump B Trip Malfunction 115 C, CondensateLPump C Trip This'causes a loss of all condensate pumps, Vith-a trip of ; all- condensate booster pumps - and both reactor feed pumps on low suction pressure. l- RCIC is; manually initiated after-12 seconds. The main-turbine is tripped after-approximately 34 seconds. A. The following. critical' parameters are recorded for 5 minutes and later plotted out for-analysis. Reactor Power- ---
i Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991
.8imulator Test-Abstracts Appendix VIII - Transient Tests Page 8 of 78 Total Steam Flow Total Feed Flow ;
Reactor Wide Range pressure Reactor Harrow Range P.iessure I Reactor Wide Range level Reactor Narrow Range Level 1 Generator output in MWe Turbine steam flow Total core flow Recirculation loop A-flow rate Recirculation loop-B flow rate-Automatic actions are verified. V. Comparison data: GETARS data recorded from plant scram number 48 for the following-parameters: Reactor power. Total: steam. flow Total'feedwater flow. j Reactor wide range pressure , . Reactor narrow range pressure Reactor wide range level-Reactor narrow range level-Generator' electrical outputLMWe Turbine steam flow
'i
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendiz VIII - Transient Tests Page 9 of 78 Total core flow Recirculation loop A flow Recirculation loop B flow f HVI . Date of last test 2/1/91' VII. Concern (s) found Wide range level fell to approximately -70 inches. in good , agreement with plant data. Level did not reach level 2 as fast as it occurred -in the plant (*?ent. Recirculation pumps automatically transferred to slow speed for about-20 seconds before they received-a trip signal from ATWS-RPT in the simulator event.- This did not occur in the plant event. Plant data suggests that the Reactor Feed Pumps did not trip until about 10 seconds after level 3 was reached. Simulator data indicated that RFPs were already tripped. Feed flow continued in-the plant event as indicated.in the data for power, feed flow and' level. The plant event shows feed. pump response'from setpoint setdown before the reactor; feed pumps trip on loss of-suction.-This response. is - not seen in~ the simulator l event and does not impact training. The simulator event shows an MSIV isolation on' low steam line pressure .since the reactor mode switch was not required to be placed inishutdown. This did not impact-the result of the test-since it occurred almost at the end of the test.-Plant' data shows pressure-falling-also
-at that time. This is not' considered a deficiency..
VIII. Resolutiont None-I
^.
n u-
i Grand Gulf Nuolear Station ! simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests Page 10 of 78 1 Transient Test #3 I. Purpose To test the response of the simulator to a Group I Isolation while at power. II. ANSI /ANS 3.5 Test requirements ANSI 3.5 Appendix B, Bl.2(3) III. Initial Conditions: Beginning of cycle Reactor power 74.1% l Reactor dome pressure 991 psig
~
i l Core flow 112.5 M1bm/hr q Total steam flow 11.74 M1bm/hr IV. Sequence of events: i The-transient is initiated by_ arming and depressing the Division 1 and 2 NSSS Manual Isolation pushbuttons- on panel' 1H13-P680 and verifying a Group I' . Isolation is occurring before-releasing the pushbuttons. No other manual operator actions are taken. A. The following critical, parameters are recorded for 15 minutes and later plotted out for analysis. Reactor-Power
. Total Steam Flow Total ~ Feed Flow Reactor.Wido Range pressure Reactor Narrow Range Pressure i
~, .w~ .a., - - -- , , a
l I l l Grand qulf Nuclear. Station l simulator 1ertification i Initial 14reste, March 1991 ! l simulator Test Abstracts i Appendix VIII - Transient Tests ; Page la Of 78 t i Reactor Wide Pange Level ' l Reactor Narrow Range Level !
. Generator Electrical output-MWe Turbirie steam flow {
.i Total Core Flow- -
I Recirculation loop A flow rate i Recirculation loop B flow rate , Automatic actions are verified. > l V. Comparison. data: i GETARS data obtained durir.g the performance of Startup Test 1-B21-SU-25-fi which included the following parameters: Reactor power Total steam flow Total' feed flow l l Reactor wide range' pressure . l Reactor narrow range pressure l Reactor wide range level l Reactor narrow range level' l l Generator electrical output-(MWe)
. Total corefflow.
Recirculation Loop:A flow rate' . Recirculation loop B flow rate- ; i
- - - , , ,vp ,e ,-r-- r-n y n,
sw , , l' i Grand Gulf Nuclear station simulator certification . Initial Report, March 1991 y simulator Test Abstracts Appendix VIII - Transient Tests Page 12 of 78 VI. Date of last test! 2/5/,1 VII. Concern (s) found: The plant data shows that ATWS-RPT did not occur for the plant event. It did occur for the simulator event since the high pressure setpoint was lowered during Refueling outage number 2 from 1125 psig to 1095 psig. This is not considered to be a deficioney. VIII. Resolution: None i t
.I i
~
i 1 1
j l Grand Gulf Nuclear Station simulator certification Initial Report,-March 1991 Simulator Test Abstracts Appendix VIII - Transicat Tests Page 13 of 78 Transient Test 14 I.
Purpose:
To test the response of the simulator to'a simultaneous trip of both recirculation pumps II. ANSI /ANS 3.5 Test requirement: _ ANSI 3.5 3.1.2 4 [ ANSI 3.5 Appendix B,--B 1.2 (4) III. Initial Conditions: Beginning of Cycle 98.3% c'o're thermal power 108.4% core flow 1037. psia-RPV dome pressure 527*F Recirculation loop temperature RPV narrow-range. level 29.0 inches IV. Sequence of events: The tran.'ient is introduced by manually transferring both - recircul.1.icn pumps from fast to slow speed. This is the method used during Startup1 Test 1B33-SU-30-6., No other? operator actions are taken. A. ine following critical - parameters are recorded for-3 minutes and-later plotted out for analysis. m Reactor Power Total Steam Flow b
, Total Feed Flcw l
Grand Gulf lluolear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests . Page-14 of=78
]
;ctor Narrow Range Pressure Reactor Narrow Range Level Total Core Flow Recirculation loop A flow rate Recirculation loop B flow rate Individual calibrated jet pump flow rates (4 total)
V. Compeiltr.n data: GETARS test data collected during Startup test from 1B33-SU-30-6 for the following parameters: Reactor Power Total steam flow rate Total feedwater flow rate Total core flow h Recirculation loop A flow rate Recirculation loop B flow rate RPV Narrow range pressure
-RPV Harrow range level Calibrated jet pump' A; flow rate CL.ibrated jet pump B flow rate Calibrated" jet pump C flow rate Calibrated jet pump D flow rate VI. Date of lent test: 2/5/91 VII. Concern (et) .ound:
-{
1
Grand Gulf-Nuclear Station Simulator certification Initial Report,-March 1991 1 Simulator Test Abstracts Appendix VIII - Transient Tests Page 15 of 78 The initial power drop is'not as rapid as in the-plant event. Power is rising much faster 1 minute after the event initiation due to feedwater temperature response. Feedwater flow response and resultant narrow range level changes are different. There are two level peaks in the plant event -with only one in - the simulator event. VIII. Resciutions i Replacement of the core and feedwater models' are expected to resolve these deficienclea. These will be completed in conjunction with the simulator upgrade project. s
Grand Gulf Nuclear Station Simulator-Certification Initial Report, March 1991-Simulator Test Abstracts Appendix VIII - Transient Tests Page 16 of 78 Transient Test #5A I.
Purpose:
To test the response of the simulator to = a single recirculation pump trip. II. ANSI /ANS 3.5 Test requirement: ANSI 3.5 3.1.2 4 ANSI 3.5 Appendix B, B 1.2 (5) III. Initial Conditions: Beginning of cycle 3825 MWt 107.6% Core Flow 1025 psig Dome Pressure 529'F-Average Recirculation Loop Temperature 31 inches narrow range level IV. Sequence of events: The transient'is' introduced by manually tripping breEker CB-5B which results in a trip to off of the: B E recirculation pump. No other operator actions are taken. A. The following icritical parameters are . recorded -for 2 minutes-and later plotted out;for-analysis.
' Reactor Power Total Steam Flow Total Feed Flow-Reactor-Narrow Range Pressure Reactor Narrow Range 19 vel-Total Core Flow Recirculation loop A flow rate
Grand Gulf Nuclear' Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts-Appendix VIII - Transient Tests Page 17 of ' 7 8 - - Recirculation loop B flow rate Individual calibrated jet pump flow rates (4 total) V. Comparison data: - GETARS test data collected during Startup test fron 1B33-SU-30-6 for the following. parameters: Reactor Power Total-steam flow rate Total feedwater flow rate Total core flow Recirculation loop A flow rate Recirculation-loop B flow rate RPV Narrew range pressure RPV Narrow range level Calibrated _jetLpump-A flow rate Calibrated; jet pump B' flow rate Calibrated jet pump-C flow rate
- Calibrated jet pump D flow rate VI. Date of last test: =2/5/91 VII. Concern (s) found:
Power drop is . not as' ~ rapid initially. Power is increasing more-rapidly;one minute after the event due~toLfeedwater temperature response.- Feed' flow drops .too rapidly. The level rise is more rapid initially. Recirculation loop A flow does not increase as it should from a decreased resistance to two phase flow in the core.
b Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix-VIII - Transient-Tests Page 18 of 78 VIII. Resolution: -. These deficiencies will be corrected with the model replacements in;the simulator upgrade' project. i
Grand Gulf Nuclear Station simulator certification > Initial Report, March 1991 Simulator Test Abstracts hypendix.VIII1- Transient Tests Page 19 of-78 Transient Test #5B I.
Purpose:
To test the response of.the simulator.to a: seizure ofLa. single recirculation pump. II. ANSI /ANS 3.5. Test-requirement: ANSI 3.5 3.1.2- 4 III. Initial Conditions:- Beginning of Cycle Reactor power -'96 86% Core flow - 97'. 9 6% RPV narrow range level 37" IV. Sequence of events: The transient- is- introduced by inserting malfunction 170B, recirculation pump B' seizure. No operator actions are-performed. A. The following critical parameters .are' recorded for 3 minutes and-later plotted out for analysis. Reactor Power Total Steam' Flow Total Feed Flow Reactor Narrow Range Pressure Reactor NarrowlRange Level Total Core Flow Recirculation loop A flow rate .
Grand Gulf Nuclear Station. Simulator Certification; Initial Report,' March 1991 Simulator Test Abstracts l Appendix VIII : Transient: Tests Page 20_of 78 1 Recirculation loop B flow rate l- 4 l Individual calibrated jet pump flow rates (4 total) t
- V. Comparison data
l Data from a -RETRAN analysis: of the event .for the l following parameters:'. Reactor power Total- steam flow - Total feed ~ water flow-Narrow range RPV: level Narrow range RPV level Total core flow -- Calibrated jet pump flows VI. Date of last test: 2/5/91. l VII. Concern (s).found: The drop in recirculation drive flow in' the af fected loop resulted in - a sharp drop in APRM Thermal Flux . - RPS setpoint. - APRM -. thermal flux did; not -- drop : sharply -- as. expected and _ a reactor = scram on APRM-- thermal, flux resulted.1Recator level did not reach-level 8 on narrow range. RETRAN _ and USFAR ' analyses ' indicate that - a scram-should occur when-- level- reaches level R 8. Due~ to this basic. difference, the; overall response was different-
- between the RETRAN 'and: simulator ~ event. APRM ' flux experienced a short duration increase initially.1 The RETRAN data predicts pressurization and ' loss of feed
.-when ' level 8 is : reached. - These did not - occur in the
. simulator event.. -
' VIII. Resolution:- +
The model replacements in the. simulator upgrade project , are expected to resolve these deficiencies. L r---, --, , 3
Grand Gulf Nuclear Station Simulator Certification Initial Report,-March 1991 Simulator Test Abstracts Appendix:VIII - Transient Tests-Page 21 of 78. Transient Test #6 I.
Purpose:
To test the response of the simulator to a turbine trip within bypass valve capacity. II. ANSI /ANS 3.5 Test requirement: ANSI 3.5-3.1.2 (15) ANSI 3.5 Appendix B, B 1.2 (6) III. Initial Conditions: Beginning of cycle Core thermal power 1219 MWt Core flow 35.6% . Reactor dome pressure 960.8 psig-Narrow range level 34 inches IV. Sequence of events:
-The transient is introduced in a manner'similar to:its reference Startup test byLopening the1 generator output breakers. This causes' an initial response very similar to a turbine trip due-to the load re' ject which-is' sensed.
A. The following critical- parameters are recorded !for ' 3 minutes and later plotted out for analysis.. Reactor Power Total Steam Flow Total Feed Flow Reactor wide range pressure Reactor Narrow Range Pressure
Grand Gulf Nuclear Station Simulator Certification-Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 22 of 78 Reactor Wide Range . Level Reactor Narrow Range Level Generator' Electrical output MWe Turbine steam flow Recirculation loop A flow rate Recirculation loop B flow rate-Automatic actions are verified. V. Comparison data: None. No GETARS Data could be derived from Startup Test 1-000-SU-27-2. Section .4;2 of 1-000-SU-27-2 was used for comparison of the power peak following'the event. VI. Date-of last test: 2/5/91 VII. Concern (s) found: None VIII. Resolution: None 3
Grand Gulf Nuclear Station-Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 23 of 78 Transient Test #7-I.
Purpose:
To test the response _of the simulator to a decrease then an increase in recirculation system flow. II. ANSI /ANS 3.5 Test requirement: ANSI 3.5 Appendix B,--31.2 (7) III. Initial Conditions: End of cycle-Reactor power 98.86% Core flow 97.96% Narrow range level 37.4 inches-IV. Sequence of events: The transient is introducedLby placing both the-A and B Recirculation loop-controllers to-close using the fast speed detente to_close FCVs at about 1% per.second over a 38 second period and stopping'FCV movement when core power _ reaches approximately 75%.
'After a 12 second wait, both the A and B Recirculation loop- controllers are opened using-the fast speed detente to . open FCVs at about 11% _ per second over - a . 37 - second period and stopping FCV-movement when core. power reaches approximately 99%.
No other manual operator actions are taken. A. The following ' critical parameters are recorded for 2 minutes and later plotted out for analysis: Reactor-Power Total Steam Flow i Total Teed Flow
i Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 24 of 78 Reactor Narrow Range Pressure Reactor Wide Range Pressure Reactor Narrow Range Level Reactor Wide Range Level Total core flow Recirculation loop A flow rate Recirculation loop B flow rate Generator output MWe Turbine Steam Flow Automatic actions are verified. V. Comparison data: Data from a RETRAN analysis - of the same - event for the following parameters: Reactor power Total steam flow Total feed flow Reactor narrow range level Reactor narrow range' pressure Total core flow Recirculation-loop A flow rate Recirculation loop B flow rate VI. Date of last test: 2/6/91 VII. Concern (s) found: , Level did not swell or shrink as much during'the power
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 25 of 78 changes as compared to RETRAN data (approximately inches). VIII. Resolution: These will- be corrected with the model replacements in the simulator upgrade project. S 9
Grand Gulf Nuolear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 26 of 78 i Transient Test 8A I.
Purpose:
To test the response of the simulator- to a maximum recirculation line break (100%)- in the Drywell in conjunction with loss of- of fsite power and failure of the LPCI Diesel Generator (Division II) to start. II. ANSI /ANS 3.5-Test requirement being satisfied: ANSI 3.5 3.1.2 lb and ic and. Appendix B, B 1.2 (8) III. Initial Conditions: 3993 MWT 101% Core Flow 1045 psig Dome Pressure 425'F Feedwater Temperature End of cycle IV. Sequence of ovents: The transient is initiated by inserting. the following malfunctions to occur simultaneously: Malfunction 63 at 100%, Recirculation Line Break. The break size is 3.181 ft' which is based: on' flow rate out of-break. Malfunction 135, Loss of Offsite Power Malfunction 140B, Failure of Division 2 Diesel-Generator to Start No operator action is taken-since FSAR comparison data-assumes this does not occur for the first 10 minutes. A. The -following. critical parameters-are recorded e f or- - 10 minutes and later plotted out for analysis: Reactor Power Reactor Wide Range Pressure Reactor Wide Range Level
Grand Gulf Nuclear 8tation Simulator. Certification Initial Report, March 1993 Simulator Test Abstracts Appendix VIII;- Transient Tests Page 27 of 78-Reactor Fuel Zone Level Total Steam Flow-4 Total Feed Flow Containment Temperature Suppression Pool Temperature Containment Pressure Drywell Temperature Drywell_ Pressure Individual Low Pressure. Coolant--Injection flow rates from RHR A, B, and C Low Pressure Core Spray (LPCS) flow rate High Pressure Core Sprayf(HPCS) flow rate _ Reactor Core Isolation- Cooling - (RCIC)- flow rate Automatic actions ~are. verified. V. Comparison Data: RPV-Wide' Range Pressure from UFSAR Figure 6.3-15 RPV FueltZone Level from UFSAR-Figure-6.3-14 I Containment Pressure data from UFSAR Figure.6.2 ! - l Drywell Temperature data from.UFSAR Figure'6.2-3 Drywell. Pressure data from UFSAR! Figure 6.2-2 VI. Date of last test:- 2/25/91 VII. Concern (s) found: None VIII. Resolution: None
- - - . =- . . . . _. ..
Grand Gulf ~ Nuclear Station-simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII-- Transient Tests Page 28 of 78 Transient Test #83 I.
Purpose:
To test the response of the' simulator to a small-Recirculation line leak in the Drywell' in conjunction-with a loss.of offsite power and failure of the - HPCS Diesel Generator (Division _III).. II. ANSI /ANS 3.5' Test requirement: ANSI 3.5 3.1.2-1b and ic III. Initial Conditions: 3993 MWt 101% Core Flow 425'F Feedwater temperature 1045 psig Dome Press. End of-cycle IV. Sequence of events: g The transient-is initiated by inserting the _ .following . malfunctions to initiate-simultaneously:- Malfunction 63 at '8%, Recirculation Line Break. The break eize is .09 ft8 which is based on flow . rate out of' break. 1 Malfunction-135, Loss of.Offsite' Power Malfunction 140C, Failure of. Division 3 Diesel l Generator; to-Start.- No operator action is takenLsince_FSAR; comparison. data; assumes this does not occur for the first 10 minutes. l
- A. The 'following critical-Jparameters are ' recorded - for ,10 '
[ minutes and-later-plotted.out for analysis: Reactor Power-Reactor Wide Range-Pressure 4 Reactor Wide Range Level 1 i
- . . . . . - ~ ,, ,,c - - ,, _-
.. - - .. .~_ . . . -
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests Page 29 of 78 Reactor Fuel Zone Level Total Steam Flow , Total Feed Flow Containment-Temperature Suppression Pool Temperature Containment Pressure Drywell Temperature DryWell Pressure Individual Low Pressure Coolant Injection flow rate from RHR A, B, and C Low Pressure Core Spray (LPCS)Eflow rate High Pressure Core. Spray (HPCS)-flow-rate Reactor Core-Isolation-Cooling (RCIC) flow rate l Automatic actions are verified. V. Comparison Data: Suppression Pool temperature data from'UFSAR Figure 6.2-14 for first 180 seconds. Containment Pressure data from UFSAR Figure 6.2-13 Drywell Temperature data' from UFSAR Figure 6.2-14 Drywell Pressure data from UFSAR Figure-6.2-13 UFSAR sections 6.2 and 6.3 VI. Date of last-test: 1/25/91 VII. Concern (s) found: None VIII. Resolution. None
- - _ _ ~
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991' Simulator Test Abstracts Appendix VIII - Transient Tests Page130 off78 Transient Test #9 I.
Purpose:
To test the response of the simulator to a main:-steam line break inside the drywell concurrent.with a loss of offsite power and a failure of Division II. Diesel Generator. II. ANSI /ANS 3.5 Test requirement being satisfied: ANSI 3.5 3.1.2 (20) ANSI 3.5 Appendix B, Bl . 2 (9)- III. Initial Conditions: End of Cycle Core power 3993 MWt Core Flow 101% : Feedwater temperhture 425'F Dome pressure 1045 psig IV. Sequence of events: The transient is introduced by inserting the following malfunctions toLoccur simultaneously:. 65, Main Steam Line Break at 100%~ severity 135, Loss of Offsite Power 140B, Failure of' Division - II Diesel Gepcrator to start due-to loss of starting air. No. manual operator actions are taken. A. .The'following critical parameters are recorded for 5 minutes and later-plotted out for analysis: Reactor Power i _. - ,~ . . . . _ -
Grand Gulf Nuclear Stetion Simulator Certification Initial ~ Report,' March-1991 Simulator ~ Test Abstracts Appendix VIII - Transient Tosts Page 31 of 78 Reactor Wide Range Pressure Reactor Wide Range Level Reactor Fuel Zone Level Total Steam Flow Total Feed Flow Containment Temperature Suppression Pool Temperature Containment Pressure Drywell Temperature Drywell Pressure -; Individual Low Pressure Coolant Injection flow rates from RHR A, B, and C (3 Total) Low Pressure Core Spray (LPCS)' flow rate: High Pressure Core Spray-(HPCS) flow rate-Reactor Core Icolation Cooling (RCIC). flow rate Automatic actions are verified.
-V. Comparison-Data:i Drywell and Containment Pressure data from UFSARLFigure 6.2 Drywell and Suppression Pool Temperature .from UFSAR -
Figure 6.2-11' Reactor vessel pressure data from UFSAR Figure 6.2-100 VI. Date of last test: 1/31/91 ! 4 VII. Concern (s) found: There -is indicated. total steam flow af ter MSIV .
Grand Gulf Nuclear Station simulator certification
. Initial-Report, March 1991 simulator Test Abstracts Appendix VIII - Trarnient Tests Page-32 of 78 isolation apparently due to liquid flow out'of the break following vessel- reflood. This' is not in accordance with the description for malfunction 63.
The break location is between the RPV and the flow restrictor. Steam flow is sensed at the flow restrictor in the reference plant. VIII. Resolution: This deficiency will be corrected with the model replacement in the simulator upgrade project.
Grand Gulf Nuclear station simulator certification Initial' Report, March 1991 Simulator Test Abstracts -l Appendix VIII - Transient Tosts Page 33 of 78 Transient Test #10 I.
Purpose:
To test the response -of the simulator - to a Group I-Isolation in conjunction- with a stuck open relief valve. II. ANSI /ANS 3.5 Test requirement: ANSI 3.5 Appendix B, Bl.2 (10) III. Initial Conditions: End of core cycle Core power-100% Core flow 100% Narrow range level 37 inches IV. Sequence of events: The transient is introduced:by, pacing the.handswitch for all-MSIVs to close on panel:1H13-P601. When SRVs are open, one SRV handswitch is placed to open on panel'1H13-P601. No other manual operator. actions are=taken. , A. The ' following critical - parameters . are recorded for'30 minutes and later plotted out for analysis: Reactor Power Reactor Wide Range Pressure Reactor Wide Range Level Reactor Fuel Zone Level Total Steam Flow Total Feed Flow
) Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 34 of 78 Containment Temperature Suppression Pool Temperature Containment Pressure Drywell Temperature Drywell Pressure Individual Low Pressure Coolant Injection flow rate from RHR A, B, and C (3 Total) Low Pressure Core Spray (LPCS) flow rate High Pressure Core Spray (HPCS) flow rate Reactor Core Isolation Cooling (RCIC) flow rate Automatic actions are verified. 4 V. Comparison data: Data from a RETRAN analysis of the same event for the following parameters: (Note only about 25 seconds pf data is available) Reactor power Reactor Wide range pressure Total steam flow Total feed flow Reactor Wide range level VI. Date of last test: 2/1/91 VII. Concern (s) found: None VIII. Resolution: None
Grand Gulf Nuclear Station Simulator Certification Initial Report,-March 1991 Simulator Test Abstracts Appendix'VIII - Transient Tests Page 35 of 78 Transient Test'#11 I.
Purpose:
To test the response of the simulator to a stuck open Safety Reljef Valve while at power.- II. ANSI /ANS 3.5 Test requirement being satisfied: ANSI-3.5 3.1.2 id III. Initial Conditions: End of cycle 80*F Suppression Pool-Temperature 3789 MWt 97.96% Core Flow 1023 psig Dome Press. IV. Sequence of events: A handswitch for an SRV'is placed to open. The SRV flow rate should be about 775,000 lbm/hr. The SRV: remains :open for the transient. No operator action'is taken_since FSAR comparison data assumes this does not occur .for-the duration of-event-tested-(15 minutes). A. The following E critical parameters;are recorded for 15= minutes and later' plotted out for analysis:1-Reactor Power Reactor-Wide Range Pressure-Reactor Wide Range Level Reactor Fuel Zone Level Total' Steam Flow Total Feed Flow Containment. Temperature-
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 36 of 78 Suppression Pool-Temperature Containmant Pressure Drywell Temperature Drywell Pressure Individual Low Pressure Coolant Injection flow rate from RHR A, B, and C Low Pressure Core Spray (LPCS) flow rate High Pressure Core Spray (HPCS) flow " ate ( Reactor Core Isolation Cooling (RCIC) flow rate Automatic actions are verified. V. Comparison Data: Suppression Pool temperature data from UFSAR Figure 6.2-91 (approximately 820 seconds). RETRAN data for the following parameters:=
-(Note: Only about 45 seconds of data were available).
Reactor Power Total Steam Flow Reactor Wide Range Level = Total Feed-Flow VI. Date of last test: 2/5/91 VII. Concern (s) found: None-VIII. Resolution: None-b \ 1
+
Grand GJ1f Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 37 of 78 Transient Test #12 I.
Purpose:
To test the res'ponse of the simulator to a turbine trip without bypass at low power conditions. II. ANSI /ANS 3.5 Test requirement: ANSI 3.5 3.1.2 (15) III. Initial ~ Conditions: End of cycle Reactor power 25% IV. Sequence of events: To prevent turbine bypass, the Bypass starting device is lowered to zero output on panel 1H13-P680 which causes a closure of bypass stop valves. The transient is introduced- by- manually tripping the turbine using the turbine trip button on panel:1H13-P680. A. The following critical parameters. are Erecorded for 3 minutes and later plotted out for analysis:
-Reactor Power Total-Steam Flow Total Feed Flow Reactor Wide Range pressure Reactor Narrow Range ~ Pressure Reactor Wide Range-level Reactor Narrow Range Level Generator output-in MWe
l Grand Gulf Nuc3 ear Station simulator certification-Initial Report, March 1991-Simulator. Test Abstracts Appendix VIII - Trans$ent Tests ( Page 38 of'78 l i l Turbine steam flow-l Total core flow Recirculation loop A flow rate. Recirculation loop B flow rate
. Automatic actions are verified.
V. Comparison datat-Data from the current cycle safety analysisifor cycle 2 for a load reject event- with'no bypass :is used due to the similarity of events. This - includes-- the! following parameters: Reactor power i Total Steam flow Total Feed flow l RPV Wide range pressure-l Reactor Wide range level' VI. Date of last--test: 12/ 5/91-- VII. Concern (s)-found: None VIII., Resolution: None-1 I y , ,m.- ,+ e r w* ,
% m-- w
i Grand = Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator-Test Abstracts Appendix 'tII --Transient Tests
'Dage 39 of 78 0
Transient Test #13 I.
Purpose:
-l ' To test the response 9f tho . simalator . to a loss ' of - condenser vacuum. II. ANSI /ANS 3.5 Test requirement:- ANSI 3.5 3.1.2 5 - III. Initial conditions: Reactor power 3450 PMt Core flow 87% Feedwater temperature ~425'F Reactor Dome pressure 1025'psig- - Reactor-level 38.5 inches. IV. - Sec.yence of events:
.The-transient is initiated by--inserting malfunction;110A' and increasing s e v e r i t y .- t o - d e c r e a s e condenser -vacuum until the turbine-trips.'
)
While condenser vacuum is decreasing,- recirculation flow a is manually reduced to coincide with the-; power reduction performed during the re.ference plantievent. A. The following critical - parameters-~ are recorded for 3-minutes and later plotted out-for analysis. b Reactor Power 4 Total Steam Flow Total Feed Flow Reactor Wide Range pressure - Reactor Narrow Range Pressure
\ Grand Gulf Nuolear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests , . Page 40 of 79- i
' Reactor Wide Range level-Reactor Narrow Range Level-Generator output in MWe Turbino steam flew Total core flow Recirculation loop A flow rate Recirculation loop B flow rate r
Automatic actions are verified. L V. Comparison data: I GETARS data recorded from plant scram number 44 for. the following_ parameters: Reactor power Total steam flow Total feedwater flow Reactor wide = range pressure R;3ctor narrow range pressure Reactor. wide ~ range level-Reactor narrow range ~1evel <
, Generator electricalt output MWe-Total core flow Recirculation loop A flow Recirculation loop B flow VI. Date of last test: 2/5/91 l
u
4 Grand Gulf Wuolear station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 41 of 78 i VII. Concern (s) founds Level shrink during the pressurization and turbine trip was not as great as occurred in the plant. Simulator level dropped to 20 inches. IM.*al in the plant event dropped to 0 inches. Total steam flow exceeded bypass capacity of 35% following turbine trip. VIII. Resolution l Thema deficiencies will be corrected during model replacement in the simulator upgrade project. i i i i l I ' I L
- _ _ . . . _ _ . . - - . _ . ~ - . _ . . . . _ . _ _ _ _- __ _ -- . _ . . _
l Grand Gulf Nuclear Station Simulator Certification j Initial Report, March 1991 Simulator Test Abstracts i Appendix VIII - Transient Tests Pags 42 of 78 : Tra.isient Test #14 1 I.
Purpose:
To test the rimulator response to a loss of offsite power ' with all emergency diesel. generators operable II. ANSI /ANS 3.5 Test requirement being satisfied! ANSI 3.5 3.1.2 3 III. Initial Conditions: End of CFole 3993 MW7 101% Core./ low 425'F Feedwater temperature i IV. e *nce of events: The transient is introduced by inserting malfunction 135, loss of offsite power, which deenergizes all AC power to all BOP and ESF busses. ! No operator action is taken since FSAR comparison data assumes this does not occur for the durat:.on of event tested (15 minutes). A. The following critical parameters are recorded for 15 ! minutes and later plotted out for analysis. Reactor Power ' l Reactor Wide Range Pressure Reactor Wide Range Level I
};eactor Fuel Zone Level!
l Total Steam Flow Total Feed Flow- -j Containment Temperature ;
\
l Suppression Pool Temperature. p l}
i Grand Gulf Nuclear Station simulator Certification i Initial Report, March 1991 f simulator Test Abstracts Appandix VIII - Transient Tests Page 43 of 78 Containment Pressure ' j Drywell Temperature Drywell Pressure , Individual Low Pressure Coolant Injection flow rate from [ RHR A, B, and C ! Low Pressure Core Spray (LPCS) flow rate ; High Pressure Core Spray (HPCS) flow rate ; Reactor Core Isolation Cooling (RCIC) flow rate [ Automatic actions are verified, f V. Comparison data [ Reactor Power from UFSAR Figure 15.2-9 Reactor wide range pressure from UFSAR 15.2-9 Reactor wide range level from UFSAR Figure 15.2-9 Total steam flow from UFSAR Figure 15.2-9 Total feedwater flow from UFSAR Figure 15.2-9 i VI. Date of last test: 1/31/91 VII. Concern (s) found: Level did not drop to -level 2 until approximately 288 seconds after the event began.- Prior to that time, _ there , vers no vessel injection systems making up level. Level ; was swelling due to repeated SRV actuations. FSAR data i l indicates that the time to reach level-2 should be much-sooner. VIII.- Resolution: This deficiency will be corrected with the' model replacements in the fimulator upgrade project. ! L
~l
?
r-
Grand Gulf Nuclear station Simulator certification Initial Report,. March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 44 of 78 Transient test #15 I.
Purpose:
To test the response of the simulator to an injection of the HPCS system while at full power. II. ANSI /ANS 3.5 Test requirement ANSI /ANS 3.5 3.1.2 (23) III. Initial Conditions: End of cycle Core power 100% Core flow 100% IV. Sequence of events: The transient is introduced by inserting malfunction 53 which causes an inadvertent automatic initiation of the High Pressure Coro Spray System.- No other manual. operator actions-are taken. A. The following - critical parameters are recorded for.3 minutes and later plotted out for analysis.- Reactor Power Total Steam Flow Total Feed Flow $ -Reactor wide range pressure Reactor Narrow Range Pressure-Reactor. Wide Range Level
- Reactor Narrow Range. Level' Generator Electrical output-MWe
i Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 C/mulator Test Abstracts Appendix VIII - Transient Tests Page 45 of 78 Turbine steam flow Total Core flow-Recirculation loop A flow rate Recirculation loop B flow rate ; Automatic actions are verified. , V. Comparison data GGNS LER 99-019 reported a similar plant event involving-an inadvertent HPCS injection at power. There was not enough detail available from which to reconstruct the exact sequence of manual actions taken so that the event could be duplicated in the simulator. The inadvertent HPCS initiation event described in UFSAR t section 15.5.1 was instead used for a reference.- L t . VI. Date of last' test: 1/31/91 L VII. Concern (s) found:- HPCS Injection valve stroke time was excessive. VIII. Resolution: l F This deficiency will be corrected within:4 years of initial certif!::ation.
'S n
g r-- r- rwm-ye-w t-,,.- t e e T ,, >c -.e. -- - , +-me,-v y--=r --w.e & a,-
t i Grand Gulf Nuclear Station f simulator Certification : Initial Report, March 1991 , Simulator Test Abstracts , Appendix VIII - Transient Tests
- Page 46 of 78 l
Transient Test #16 [ I.
Purpose:
} To test the response of the simulator to a start of Recirculation pump B.
- II. ANSI /ANS 3.5 Test requirement being satisfied:
E ANSI /ANS 3.5 - 1985 3.1.2(4) > III. Initial Conditions: ' Beginning of cycle , Recirculation loop B secured Core power 1607 MWt Core flow 34.3% i Total Feedwater flow 6.94 Mibm/hr Total Steam' flow 6.94 Mlbm/hr- ; Reactor level 36-inches IV. Sequence of events: { i The transient- is introduced by manually starting the B j l Recirculation: pump. ' No other manual operator actions are taken. f A. The- following critical parameters .are . recorded .for 3 minutes and later plotted out for analysis. Reactor Power l Total Steam Flow l Total Feed Flow I Reactor Narrow Range Pressure-f f Reactor Narrow Range Level !
- _ _, _ ~- -., , -_ , _ _ _
Grand Gulf Wuolear station ! simulator certification Initial Report, March 1991 i simulator Test Abstracts >
.sppendix VIII - Transient Tests ;
Page 47 of 78 [ i Total core flow Recirculation loop A flow rate Recirculation loop B flow rate Individual calibrated jet pump flow rates (4 total) Automatic actions are verified. V. Comparison data GETARS Data obtained during the performance of Startup Test 1-B33-SU-30-6, Section 4.4 for the following-parameters: Reactor power Total steam flow Total feedwater flow Narrow range reactor pressure Narrow range reactor pressure Total core flow Recirculation loop A flow rate Recirculation loop B flow rate VI. Date of last test: 2/15/91' VII. Concern (s) found: _None VIII. Resolution: None l
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 48 of 78 Transient Test #17 I. Purpose ' To test the response of the simulator,to a failure of Reactor pressure control to minimum output resulting in , closure of all turbine and bypass control valves. 1 II. ANSI /ANS 3.5 Test requirement: P ANSI 3.5 3.5 3.1.2 (25) III. Initial conditions: End of cycle Core power 3993.MWt Core flow 101% Feedwater temperature 425'F Reactor dome pressure 1045 psig Reactor level 36 inches IV. Sequence of events: , The transient is introduced by inserting malfunction 79, pressure controller failure to minimum output. No manual operator actions are performed. A. The following critical parameters are recorded for 3 minutes and later plotted out'for analysis.- Reactor Power Total Steam Flow Total Feed Flow Reactor wide range pressure ! Reactor Narrow Range Pressure j.
t Grand Gulf Nuclear ','tation [ simulator certification ! Initial Report, March 1991 , simulator Test Abstracts [ Appendix VIII - Transient Tests , Page 49 of 78 j i Reactor Wide Range Level ! I Reactor Narrow Range Level 4 i Generator Electrical output MWe Turbine steam flow t I Total core flow , i i Recirculation loop A flow rate l Recirculation loop B flow rate Automatic actions are verified. ! V. comparison data Reactor power, total steam flow, narrow and wide range reactor I pressures, wide range - reactor level, narrow range reactor I level, and total core flow from UFSAR Figure 15.2-1. I , VI. Date of last test: 1/31/91 VII. Concerns found: None VIII. Resolution None P k o 4 n- n ec, ' - W-
. .._ _ ._ _ ~ .- _ .._ _ .. ... . _ _ _ . _ . _ _ _ . - . . _ .
Grand Gulf Wuolear station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests ! Page 50 of 78 Transient Test #18 I. Purpose To test the response of the simulator to a pressure control failure resulting in maximum output from the controller. II. ANSI /.\NS 3.5 Test requirement being satisfied: ANSI 3.5 3.1.2 (25) III. Initial conditions: End of cycle Core power 3993 MWt Core flow 1011 Feedwater temperature 425'F Reactor dome pressure 1045 psig P IV.- Sequence of events: The transient is introduce by inserting malfunction 80,
- pressure regulator failure to maximum demand.
No' manual operator' actions are taken. A. The following critical parameters are recorded for- 3 minutes and later plotted out for analysis. Reactor Power l Total Steam Flow l
- Total Feed Flow l Reactor Wide-Range pressure Reactor Narrow Range Pressure Reactor Wide Range Level m, -e. m y e w , 1
Grand Gulf Nuclear station simulator certification Initi_* Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests Page 51 of 78 React 3r Harrow Range Level Generatar Electrical output - Turbine steam flow Total Core flow J Recirculation loop A flow rate Recirculation loop B flow rate Automstic actions are verified.
't. Comparison data Reactor power, total steam flow, total feed flow, wide range reactor pressure, Larrow range reactor pressure, wide range reactor level, narrow range reactor level, turbine steam fitw, and total core flow from UFSAR Figure 15.1-4.
l VI. Date of last test -1/31/91 ; i' VII. Concern (s) found Power did not fall sharply at the beginning of-the event due to void formation. Power falls i as pressure falls. MSIV closure occurs on low-steam _ pressure as_ expected. Steam flow did_not _ drop sharply because power- and _ steam-production did not drop. Feed flow increased to match steam flow initially. Level- did not shrink as expected during the pressurization event _ caused by MSIV closure. Level instead' swells due to SRV actuation and increases. i Level did not ' drop _ to ~ level 2 as expected. , -Reactor feed pumps continue pumping 20 seconds ! after MSIV closure. Loss of RFP' supply _ steam should have occurred sooner. VIII. Resolution:. These problems will_ be corrected with the model ! replacements in the simulator upgrade, i i i
i i i ! Grand Gulf Nuclear Station i simulator certification Initial Report, March lb 1 : simulator Test Abstracts-Appendix VIII - Transient Tests j Page 52 of 78 ; Transient Test #19 t I. Purposes j i To test the response of the simulator to a closure of ! Main Steam Isolation valves with a failure to j automatically SCRAM. ; t II. ANSI /ANS 3.5 Test requirement being satisfied . j ANSI 3.5 3.1.2 (24) ! III. Initial Conditions i End of cycle i 100% core thermal power i t 100% core flow I r IV. Sequence of events: i An ATWS condition is created by inserting malfunction- 75, f which- causes a 'r $ 1ure of all control rods to insert on a i l Scram signal. , The transient is introduced by arming and depressing the Division 1 and 2 NSSSS isolation pushbuttons on panel 1H13-P680, which causes MSIVs to isolate. l No manual operator actions are-taken. ! A. The following critical. parameters are ~ recorded .for 10. . minutes-and later plotted out-for analysis: r Reactor Power , Reactor Wide Range Pressure i Reactor Wide Range Level Reactor Fuel 1 Zone Level i t
~
Total Steam Flow [ Total Feed Flow i
?
I
- , -,,,n.. ,, - . . , . . ,, ., --- - ,. , , - . , , .
~ . , , ,
Grand Gulf Nuclear Station ! simulator Certification ! Initial Report, March 1991 ! Simulator Test Abstracts - Appendix VIII - Transient Tests ! Page 53 of 78 ! l Containment Temperature : t Suppression Pool Temperature ' Containment Pressure i Drywell Temperature Drywell Pressure , Individual Low Pressure Coolant Injer' n flow rate from i RHR A, B, and C (3 Total) , t Low Pressure core Spray (LPCS) flow rate i L High Pressure Core Spray (HPCS) flow rate I Reactor Core Isolation Cooling (RCIC) flow rate Automatic actions are verified. f V. Comparison datat EPRI NP-5562 - Dec. 87, Analysis of ATWS in Severe BWR l Accidents (BWR-4). including the following parameters Reactor power i i Wide Range pressure f f Wide Range level i i Fuel Zone level l VI. Date of last testi. 2/1/91 < [' t VII. Concern (s) found None i VIII. . Resolution: None E l i
., I r- w - - --
m + ,, v s y 4 + < -- g
J Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts ' Appendix VIII - Transient. Tests Page 54 of 78 1 Transient Test #20 I. Purpose To test the response of the simulator to a feedwater line break in the turbine building. II. ANSI /ANS 3.5 Test requirement being satisfied ANSI 3.5 3.1.2 (20) III. Initial Conditions: Beginning of cycle Core power 3993 MWt Reactor dome pressure 1045 psig IV. Sequence of events: The transient . is introduced by inserting - malfunction -70, Feedwater line break in the turbine _ building. A. The following critical parameters are' recorded for 3 minutes and later plotted out for analysis: Reactor Power Reactor Wide' Range Pressure Reactor Wide Range Level Total Steam Flow Total Feed Flow Turbine Steam flow Total Core flow Recirculation loop'A flow Recirculation loop B flow Automatic actions 1are-verified.
.- . . , . -l
t Grand Gulf-Wuclear Station ! simulator certification i Initial Report, March 1991 : Simulator Test Abstracts i Appendix VIII - Transient Tests ! Page 55 of 78 ; h V. Comparison Data ' ( Since there are no other transient analyses available for
- comparison, this transient was evaluated by a panel of. >
experts consisting of at least two Senior Reactor [ Operators licensed on Grand Gulf. ;. Discussion data from UFSAR section 15.6.6 was used in evaluating this transient. l There are no UFSAR figures to use for containment ! response comparison for this transient. , t VI. Date of last test: 2/15/91 i VII. _ Concern (s) found None l VIII. Resolution: None i r
)
t s
- -~
, a a-.- ,-,.,n. . _ . - g,..- . . , . --,,,.y ,, , .,-n,.
.~ _
- ~. -- . _ _ _ _
l P i' f Grand Gulf Nuclear Station I simulator Certification Initial Report, March 1991 j simulator Test Abstracts ! Appendix VIII - Transient Tests i Page 56 of 78 ' 4 Transient Test #21 I.
Purpose:
To test the response of the simulator to a feedwater line i break in the Drywell. II. ANSI /ANS 3.5 Test requirement: , ANSI 3.5 3.1.2 (20) i III. Initial Conditions: Beginning of cycle Core power 3993 MWt , Core flow 101% Feedwater temperature 425'F . Reactor dome pressure 1045 psig IV. Sequence of events: The ' transient is introduced by inserting malfunction 171, feedwater line break at 100% severity. 1 l-No operator manual actions are taken. , A. The followin; critical . parameters are recorded for! S' minutes and:later plotted out for analysis: , Reactor Power ,
. Reactor Wide Range Pressure Reactor Wide Range Level' Reactor . Fuel Zone : Level-Total Steam. Flow-Total Feed Flow Containment Temperature
- 1
_ - - , - , , # .-c,*,-.,my- ,
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests i Page 57 of 78 Suppression Pool Temperature Containment Pressure Drywell Temperature i Drywell Pressure Individual Low Pressure Coolant Injection flow rate from ; RHR A, B, and C (3 Total) ' Low Pressure Core Spray (LPCS) flow rate High Pressure Core Spray (HPCS) flow rate Reactor Core Isolation Cooling (RCIC) flow rate Automatic actions are verified. V. Comparison Data The feedwater break location used in malfunction 171 is different from the break location described in the UFSAR. l The UFSAR event description could _ only be used for. reference and comparison in general. , VI. Date of last test: 1/25/91 VII. Concern (s) found: None VIII. Resolution: None I l I z
I Grand Gulf Nuclear Station sinulator Certification Initicl Report, March 1991 simulator Test Abstracts Appendix VIII - Transient Tests Page 56 of 78 Transient Test #2P I.
Purpose:
To test the response of the simulator to a fast closure of both Recirculation Flow Control valves. II. ANSI /ANS 3.5 Test requirement being satisfied: ANSI 3.5 3.1.2 (17) III. Initial Conditions: Reactor power 98.96% Core flow 97.96% Recirculation flow 96.54% Recirculation Flow Control valves at 65% (normal full' power position) Reactor Level narrow range 37.4 inches. l IV. Sequence of events: The transient is introduced by placing both Recirculation loop controllers in the close position using the fast speed-detente which results in a closure rate of _ . approximately 1% per second. No other operator actions are taken. The following critical parameters are recorded for 3 minutes ! and later plotted out for analysis. l Reactor Power Total Steam Flow Total Feed Flow Reactor Narrow Range Pressure I Reactor. Narrow: Range. Level -
-j
*: i
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 59 of 78 ! Total Core Flow Recirculation loop A flow rate Rocirculation loop B flow rate Individual calibrated jet pump flow rates (4 totai-Automatic actions are verified. V. Comparison datat Data from a RETRAN analysis is plot'ced'for comparison of the following parameters: Reactor power Total steam flow , Total feed flow , Reactor narrow range pressure Reactor narrow range level Total core flow VI. Date of last test: 2/6/91 i VII. . Problems found/ status: l Reactor level did not' swell as expected. Level roseDto 40 inches in the simulator as compared to 48 inches predicted by RETRAN analysis. o Core flow, power, steam flow and feed flow decreased lower i than predicted by RETRAN analysis. Power increases 'more rapidly following FCV ' closure than i expected by RETRAN analysis due to decreasing Feedwater ! temperature. Recirculation Pump' Temperature-interlock alarms " were received as a result. VIII. Resolution: These deficiencies will be corrected with-the model-replacements in the simulator upgrade project.
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator' Test Abstracts Appendix VIII - Transient Tests Page 60 of 78 Transient Test #23 I.
Purpose:
To test the simulator response to a single Recirculation Flow Control valve failing open. II. ANSI /ANS 3.5 Test requirement-being satisfied: ANSI 3.5 3.1.2 (17) III. Initial conditions: End of Cycle 100% Rod line Recirculation Flow Control valves at minimum position e Recirculation pumps in fast speed IV. Sequence of events: The transient is introduced by inserting _ malfunction 173B, This causes a failure in the control circuit resulting in opening of the "B" Flow Control valve. This is limited-hydraulically by.a-velocity limiting-orifice to approximately 25% per second.- No other operator actions are-taken. A. The following critical parameters are recorded for 3 minutes and later plotted out for-analysis. Reactor Power Total Steam Flow Total Feed Flow i Reactor Narrow Range Pressure' { Reactor NarrowLRange Level
-Total Core Flow-- - -
j
.j i
1
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 61 of 78 Recirculation loop A flow rate Recirculation loop B flow rate l Individual calibrated jet pump flow rates (4 total) Automatic actions are verified. V. Comparison data: Data from UFSAR Figure 15.4-4 is plotted for comparison of f the following parameters: Reactor power Total steam flow Total feed flow Reactor narrow range pressure Total core flow Qualitative results of the UFSAR- analyzed event in UFSAR section 15.4 were available for comparison. The UFSAR event assumes a FCV opening rate of 30% per second. The simulator assumed failure rate is 25% per second. This differences account for differences in response of the simulator and the UFSAR data. VI. Date of last test: 2/6/91 VII. Concern (s) found: None VIII. Resoulution: None I
! Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 ' 1 Simulator Test Abstracts-Appendix VIII - Transient Tests Page 62 of 78 Transient Test.#24 I. Purpose To test the response of the - simulator to a failure resulting in the opening of both Recirculation Flow control Valves II. ANSI /ANS 3.5 Test requirement being satisfied: ANSI 3.5 3.1.2 (17) III. Initial conditions: End of Cycle 100% rod line Recirculation pumps _in fast speed Recirculation Flow Control Valves at minimum position IV. Sequence of events: The transient is introduced by placing the A and B Recirculation Flow Control Valve loop controllers to open using their fast speed detente. No other operator actions are taken. The following critical parameters are recorded for 3 minutes and later plotted out for analysis. Reactor Power. Total Steam Flow Total Feed Flow Reactor Harrow Range Pressure
-Reactor Narrow Range _ Level Total Core-Flow-
- . _ ._ _ _ .. f. _ - _ . _ _ .
- - ..... - - .. - _ - . - .- .- . .. . . - - . . . . ~ - .- _ .
i Grand Gulf Nuclear atation ' simulator certification Initial Report, March 1991 simulator Test Abstracts ! Appendix VIII - Transient Tests t Page 63 of 78 T t Recirculation loop A flow rate i i Recirculation loop B flow rate 7 i Individual calibrated jet pump flow rates (4 total) l t V. Comparison data: l i Data from UFSAR Figure 15.4-5 was available for comparison i but not plotted due to significant differences in the assumed ! speed of the opening failure of 11% per second per the UFSAR and 1% per second in the simulator. [ i i VI. Date of last test:- 2/6/91 ! VII. Concern (s) found L Level did not shrink as expected following scram on High APRM flux. [ VIII. Resolution: I i This discrepancy will be corrected with. the~ model , replacements in the simulator upgrade project. - i z f l i i r I (. i
. u.~ . - _ .
. _ . _ . . . ~ . . . _
t , . , , , , , ,..,,,mm, , , _ , .,..,s ,, .,_
i Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests ! Page 64 of 78 l Transient Test #25 t I.
Purpose:
To test the response of the simulator to a closure of one Recirculation flow control valve. II. ANSI /ANS 3.5 Test requirement being satisfied: ANSI 3.5 3.1.2 (17) III. Initial Conditions: Reactor power 3993 MWt Core flow 101% Feedwater temperature 425'F Reactor dome pressure 1045 psig IV. Sequence of events: The transient is introduced by placing the A Recirculation Flow Control Valvc ' loop controller to close using its fast speed detente. No other operator actions are taken. The following critica1' parameters areLrecorded for 3 minutes and later plotted out for analysis. Reactor Power l Total Steam Flow i l Total Feed Flow Reactor Narrow-Range Pressure l Reactor Narrow' Range-Level-Total Core Flow.
, Recirculation loop A flow rate
i e Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 65 of 78 Recirculation loop B flow rate Individual calibrated jet pump flow rates (4 total) V. Comparison data Data from UFSAR Figure 15.3-3 which is for a closure of Recirculation Flow Control-Valve at 60% per second for the following parameters: Reactor Power Steam Flow Feed Flow Harrow Range Pressure Narrow Range Level Total Core Flow VI. Date of last test: 1/25/91 VII,-Concern (o) foundt l The final' power, core _ flow, steam and feed flow were not as expected. VIII. Resolution: These. deficiencies will be corrected with' model. replacements in the simulator upgrade project.
, . . , , . - . - J , ..
- . _ . _ ~ - . ~ - - .. - - . . . _ - . - . . . . - - - - .. . - . . . -.. . ._ - . - - .
, Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII - Transient Tests Page 66 of 74 Transient Test #26 I. Purposet To test the response of the simulator to a closure of both recirculation flow control valves. II. ANSI /ANS 3.5 Test requirement being satisfied ANSI 3.5 3.1.2 (17) III. Initial Conditions: Beginning or cycle Reactor powor 3778 MWt Reactor dome pressure 1032 psia Reactor level 35.4 inches Core flow 98% IV. Sequence of events: The transient is introduced by manually closing Recirculation Flow Controluvalve A and B using the slow.s;?eed detente.at intervals to match the operator' actions taken .n the=--reference-plant event.. No other manual actions are taken. The following critical parameters are recorded' for:20 minutes . and later plotted.out for-analysis.
! Reactor-Powarz Total' Steam Flow 1
Total Feed Flow Reactor Narrow Range Pressure , ReactorJNarrow-Range Level- - -' -
._i...., _ . _ _ . _
i i Grand Gulf Nuclear 8tation
' Simulator certification !
Initial Report, March 1991 ' Simulator Test Abstracts Appendix VIII - Transient Tests Page 67 of 78 , Total Core Flow Rec <.Julation icop A flow rate ; Recirculation loop B flow rate , Individual calibrated jet pump flow rates (4 total) ! V. Comparison data: ! GETARS test data collected during Startup test from 1B33-SU- l 29-6 for the following parameters: l f Reactor Power Total stean flow rate Total feedwater flow rate l' Total core flow t Recirculation loop A flow rate j Recirculation loop B flow rate i RPV Narrow range pressure ? l VI . . Date of last test: 2/6/91' VII. Concern (s) _ found: ! The final power, core flow, steam flow, reci :ulation loop flows were . lower than indicated by refer *1 plant L data. There was good agreement in overall tisno. l~ VIII. Resolution: ! .These deficiencies will- be ' corrected with the- model {
, replacements in the simulator upgrade project. ;
i
=
i f Grand Gulf Nuclear Station l Bimulator certification : Initial Report, harch 1991 ; simulator Test Abstracts
- Appendix VIII - Transient Tests :
Page '8 6 of 78 i Transient Test #27 l I. Purpose . To test the response of the simulator to a feedwater ; control system failure _to maximum demand. i' II. ANSI /ANS 3.5 Test requirement being satisfied: i ANSI 3.5 3.1.2 (17) i III. Initial Conditions: ' i End of cycle Core power 3993 MWt Core flow 101% . Feedwater temperature 425'F Reactor dome pressure 1045 psig ; i IV. . Sequence of events: The transient is introduced by entering malfunction 127, , ! Feedwater master controller fails to maximum output. ; I No other manual operator actions are performed.- < The following critical parameters are recorded for 3 minutes and later plotted out for analysis:~ , Reactor Power i Total Steam Flow ! Total: Feed Flow Reactor Wide Range pressure Reactor Narrow Range-Pressure ' l Reactor-Wide Range-level-L Reactor Narrow Range- Level , 1 J N v - , . . - vn e ,, r w.- 1 e-,+ - v,- -% r-.-.
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator-Test Abstracts Appendix VIII - Transient Tests Page 69 of 78 Generator. output in MWe Turbine steam flow Total core flow Recirculation loop A flow rate Recirculation loop El flow race Automatic actions are verified. V. Comparison data: Data from_b?SAR Figure.15.1-3 for the following parameters: Reactor power ' Total steam flow Total feed flow Rea:Ror wide range pressure Reeactor narrow range pressure Reactor wide range levell Reactol narrow range ',evel Total core flow The UFSAR assumed maximur. feedwater flow rate is 130% which; exceeds the simulator maximum flow -rate of about 123.6%. This l: accounts for the differences in repsonse times and- the- . subsequent events, l-L VI. Date of last-test: 1/31/91 l VII. Concern (s) found:- None VIII. Recolution: None
Grand Gulf Nuclear Station Simulator certification Initial Report, darch 1991 Simulator Test Abstracts Appendix VIII - Transient _ Tests-Page 70 of 78 Transient Tost #28 I.
Purpose:
To test the response of the simulator to a loss of feedwater heating. II. ANSI /ANS 3.5 Test requirement being satisfied: ANSI 3.5 3.1,I (17) III.- Initial Conditions: Beginning of cycle Reactor power 75.5% Core flow 96.4% Reactor dc3e pressure 993.6_psig Reactor level 36.1 inches IV. Sequence of events: The transient is introduced by isolating extraction steam to the number 6B High Pressure Feedwater heater. No other manual operator actions are; performed. A. The following critical ~ parameters are- recorded fore 3
-minutes and later plotted out for a:ialysis:
Reactor Power Total Steam Flow Total Feed Flow-Reactor Wide Range pressure e- . Reactor Narrow Range Pressure Reactor Wide Range level Reactor Narrow Range Level
l l l Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 Simulator Test Abstracts- ! Appendir-VIII -Transient Tests l Page 71 of 78 Generator output in Mhe L Turbine steam flow Total core flow Recirculation-loop A-flow rate Recirculation loop B flow rate Automatic actions are verified. V. Comparison data: GETARS data from Startup Test- 1-000-SU-23-6 for the -following parameters: l l Reactor power Total steam flow Total Feedwater flow Reactor wide range pressure Reactor narrow range pressure-Reactor wide range level
-Reactor narrow range 11evel Generator. electrical output MWe Total core flow.
l 1 Recirculation loop A flow Recirculation loop B flow-VI. Date-of last test:- 2/6/91 VII. Concern (s) found:- ~ Power rises more rapidly (180 seconds) -than- plant data i indicates. -Plant- data shows power steady at approximately 360
t 1 Granc Gulf. Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix VIII Transient Tests Page 72 of 78 seconds. Simulator core steam and feed flow were slightly more-responsive to the feedwater temperature reduction than plant data indicated. Wide range level was about 15-inches higher-than plant data indicated for comparable plant conditions. VIII, Resolution: These deficiencies will be resolved with the model replacements in the simulator. upgrade project. i i
8 Grand Gulf Nuclear Station ! simulator certification Initial Report, March 1991
- Simulator Test Abstracts Appendix VIII - Transient-Tests Page 73 of 78 l Transient Test #29
, I.
Purpose:
To test the response of the simulator to a generator load reject with no turbine bypris. II. ANSI /ANS 3.5 Test requirement b *" .atisfied: ANSI 3.5 3.1.2 (15) III. Initial Conditions: > End of cycle ! Reactor power 98.86%. Core flow 110.2 Mlbm/hr Reactor level 37.4 inches Reactor dome pressure 1022.7 psig-IV. Sequence of events: To fail turbine bypass, malfunctions 82A, 82B, and 82C are entered. This fails-- the A, B, and.C turbine stop-valves closed.- , The transient is -initiated by manually opening ~ both-generator output breakers, J5228 and-J5232, from panel lH13-P680. This results .in an immediate loss of load -and a resulting load' reject.
'No other manual operator actions are performed.
A. The following critical parameters are recorded for 3 minutes and-later plotted out.for analysis: Reactor Power Total Steam: Flow
-Total Feed Flow Reactor Wide Range pressure
- - ...,.,,..c.
-1.-.
Grand-Gulf Nuclear Station
-Simulator certification Initial Report, March-1991 simulator Test Abstracts Appendix VIII - Transient Tests-Page 74 of 78 Reactor Harrow Range Pressure-Reactor Wide-Range level Reactor Narrow Range Level Generator output in MWe Turbine steam flow Total core flow-Recirculation loop A flow rate Recirculation loop B flow rate Automatic actions are verified.- '
V. Comparison data: Data from an RETRAN analysis'- for- the- following-- parameters: Reactor power-Total steam flow Total Feedwater flow Reactor wide range pressure' Reactor: narrowLrange pressure-
- Reactor-narrow range level
-Total' core-flow Recirculation loop'A flow-Recirculation loop B flow.
VI. Date of-last test: 2/6/91) VII.-Concern (s) found:
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts. Appendix VIII - Transient Tests-Page 75 of 78 Level shrink from the pressurization was greater than indicated by RETPAN data - (approximately.10") VIII. Resolution: This deficiency will be corrected with the model replacements in the simulator upgrade project.
Grand sulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts-Appendix VIII - Transient Tests Page 76 of 78 Transient Test #30 I.
Purpose:
To test the response of the simulator to a load reject event of the main generator. II. ANSI /ANS 3.5 Test requirement being satisfied: ANSI 3.5 3.1.2 (16) III. Initial Conditions: Beginning of cycle core conditions Reactor power 3820 MWt Core flow 97.7% - Generator output 1226 Mwe Reactor dome pressure 1020.7 psig Peactor level 37 inches IV. Sequence of events:- To match the conditions of the startup test being used for reference data, the A bypass valve is failed:using-malfunction 82A. The transient is initiated - by manually opening. . both generator output breakers, J52281and J5232, from panel 1H13-P680. This results- in an immediate -loss of load and-a resulting load? reject.- A. The following critical ~ parameters are recorded for 3 minutes and later plotted out for analysis: Reactor Power Total Steam Flow Total Feed Flow Reactor. Wide Range pressure
i Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix-VIII . Transient Tests Page 77 of 78 i Reactor Narrow Range Pressure Reactor Wide Range level Reactor Narrow Range Level Generator output in MWe Turbine steam flow-Total core flow Recirculation loop A flow rate Recirculation loop B flow' rate Automatic actions are verified. V. Comparison data GETARS data gathered in the performance of startup test 1-000-SU-27-6 for the following parameters: Reactor power , Total steam flow l Total-feedwater flow' Reactor narrow range pressure' Reactor wide range pressure
- Reactor wide range-level-Reactor narrow range level Generator Electrical output MWe ,
Turbine steam flow Total core flow Recirculation' loop'A flow-
I Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts
-Appendix VIII - Transient' Tests Page 78 cf 78 Recirculation loop B flow VI. Date of last test: 2/6/91 VII. Concern found/ status:
Pressure exceeded startup test criteria of 71.G psi rise but did not exceed the plant maximum pressure. This is due to variance in initial conditions. The initial pressure was lower than the SRV actuation pressure setpoint. VIII. Resolution: None
Appendix IX Simulator Test Abstracts Malfunction Tests
~
Grand Gulf Nuclear Station
. . Simulator Certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests rage 1 of 210 Apoendix IX Simulator Test Abstracts Malfunction Tests Malfunction test abstracts are provided for- each malfunction tested. Each' test abstract contains ' the malfunction number and title, initial conditions, an abstract of the-sequence of events, and automatic actions and response verified, date of last test, concerns noted during each test, and planned resolution.
Malfunction Test #1, SRM Channel Failure Full Scale , Initial Conditions: Reactor approximately $0.50 subcritical Sequence of events: Malfunction #1 is -introduced for each of the- six SRM Channels. Automatic actions and alarm - response ' are -verified including the fol3 awing: Each SRM channel : fails- upscale which causes = SRM- Period response, Control Rod -Withdrawal . Block annunciation, LSRM Upscale annunciation. l Back panel indications are checked for SRMIA only. Attempts to withdraw a control rod fail. The Offected SRM channel 11s bypassed'. Affected annunciators and Control' Rod Withdrawal' Blocks clear.- Control Rod withdrawal:is subsequently permitted.. The malfunction i s-. c l e a r e d and - the - affected SRM channel returns to normal'. The affected SRM channel; bypass _ is - removed and -the test repeated 1for remaining' channels.
l Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Test Abstracts [ Appendix II - Malfunction Tests Page 2 of-210 l The simulator is reinitialized to'an IC with the reactor ( critical at 400 psig RPV pressure and all "*?s. on range 8 or l above. l r l Malfunction #1 is inserted and all previous responses. verified to occur except Control Rod Withdrawal Block. The test is repeated.for all SRM channels. Date of last test: 3/12/90 Concern (s) found/ Planned resolution: l Back panol indications could be verified on SRM channel A only. SRM channels B through F are not simulated..A simulator review board has recommended that SRM channel B b e -- i n c o r p o r a t e d into a revised simulator backpanel 1H13-P670 and that the other channels are not- needed for training. The revised panel is to be incorporated into an upgrade simulator project for expected completion within 4 years of initial certification. I
-)
* -,--<-4
Grand Gulf Nuclear Station
;8imulator certification Initial Report, March 1991 Test Abstracts i Appendix II - Malfunction Tests Page 3 of 210 Malfunction Test #2, SRM Channel Failure Downscale Initial Conditions:
Reactor approximately $0.50-subcritical Sequence of events: Malfunction 2 is introduced for each of the six SRM Channels. Automatic actions and alarm-response are verified including the following: Each SRM channel fails downscale causing SRM Period response,
~
Control Rod Withdrawal Block annunciation, SRM . Downscale - annunciation, and SRM Detector ERetract Not Perm!rsible Indication. Back panel indications are checked for SRM A .only. Attempts to withdraw a control rod and the affected SRM fail. i The affected SRM channel is-bypassed. Affected annunciators and Control Rod Withdrawal Blocks clear. SRM Detector Retract indicates withdrawal is permitted. Control Rod and SRM detector withdrawal are subsequently-permitted. The malfunction is cleared and the affectedL SRM channel returns to normal. The affected SRM channel bypass'is-removed. Control Rods are withdrawn until IRM channels are above Range ~ 2 but below Range 8. Malfunction number 2 is entered and all previous actions occur except the SRM Detector Retract Not Permitted annunciator does not actuate.
Grand Gulf Nuclear Station Simulator certification Initial-Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Pr.ge 4-of 210 Date of last test: 3/16/90 Concern (s) found/ Planned resolution: Back panel -indications could be verified on SRM channel A .- only. SRM channels B through F are not simulated.-A simulator review board has recommended that SRM channel B. be-incorporated into a revised simulator backpanel 1H13-P670 and-that the other channels are not needed for training. The revised panel is to be incorporated into an upgrade simulator project for expected completion within 4 years of initial certification. Panel P680 annunciator P680-7A-Cll, "SRM DET RTRACT NOT PERM" did not clear when the affected SRM ' was bypassed - with the malfunction present. Thia discrepancy will be corrected-on a schedule to be completed. within 4 . years =_of initial certification. I
..m ..,
Grand Gulf Nuclear Station Simulator Certificatior Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 5 of 210 Malfunction Test #3, SRM Channel Detector stuck Initial Conditions: Reactor approximately $0.50 suberitical Sequence of events: Malfunction #3 is introductd for each of the 6 SRM channels. Automatic actions and alarm response are verified: including the following: Control rods are withdrawn until SRM channels approach 1.0 - E+05 CPS. All SRM chant.ol detectors are selected for withdrawal. All SRM channels except the one with the stuck detector indicate they are not full in and experience decreasing count rate. Back panel-indications are checked for-SRM A only. ! Control rods are withdrawn to increase SRM,. count. rate -2.0 l E+05 CPS in the channel with the stuck detector, which causcs-SRM Upscale alarm and annunciation. Using an instructor remote function, the appropriate 'RPS Shorting link is removed which causes an RPS half scram condition. The RPf; shorting link is replaced with the remote function and thn half scram condition.is reset. ! The n.alfunction .is. cleared. And the SRM detector is l successfully - withdrawn -until . its indication Ereturns to the level of the other detectors and upscale ' annunciation and control rod withdrawal-blocks clear. A control _ rod is then withdrawn to verify the control - rod withdrawal block condition is clear. The malfunction is repeated for remaining channels.
,l* ..
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 6 of 210 Date of last test: 3/13/90 Concern (s) found/ Planned resolution: Back panel indications could be verified on SRM channel A only. SRM channels B through F are not simulated. A simulator review board has recommended that SRM channel B be incorporated into a revised simulator backpanel 1H13-P670 and that the other channels are not needed for training. The revised panel is to be incorporated into an upgrade simulator project for expected completion within 4 years of initial certification.
i Grand Gulf Nuclear-Station simulator Certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 7 of 210 Malfunction Test #4, IRM Channel Fails Upscale Initial Conditions: The reactor is critical with RPV Pressure at approximately.400 psig. Sequence of events: All IRM/APRM channel recorders are verified to be selected to record IRM response. Malfunction 4 is introduced into each channel. Automatic actions and alarm response are verified including the following: IRM flux level increases in the affected channel and results in IRM Upscale alarm for both Nuclear Instrumentation and RPS in the affected channel. Back panel indications are checked for.IRM A only. Ranging of the IRM' channel is verified to have no influence on the increasing flux level. A control rod block and RPS half scram condition occur. No control rod motion after attempting withdrawal. The affected IRM Channel is bypassed which causes control rod
-withdrawal block indication to-clear..
' Control rods are verified to be able to be withdrawn.
The RPS half Scram-condition is reset. The nalfunction test is repaated for remaining channels. The malfunction is introduced in combinations to verify half scram or full scram conditions occur for correct combinations. l
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 8'of 210 The simulator is reinitialized at an initial condition at about 10% power and the Reactor. Mode Switch in RUN. It is verified that the malfunction does not cause RPS actuation with these conditions for each' channel. Date of last test: 3/14/90 Concern (s) found/ Planned resolution: Back panel indications could be verified on -IRM channel A only. IRM channels B through H are not simulated. A. simulator review board has recommended that IRM channel B be incorporated-into a revised simulator backpanel 1H13-P670 and that the other channels are not needed for training. The revised panel is to be incorporated into an upgrade simulator project for expected completion within 4 years - of initial certification. i
& % ,-r - - - , , e- - . ,-
f Crand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 9 of 210 l Malfunction Test #5, IRM Channel Fails Downscale Initial Conditions: The reactor is critical with RPV Pressure at approximately 400 psig, sequence of events: All IRM/APRM channel recorders are verified to be selected to record IRM response. Malfunction 5 is introduced into each channel. l Automatic actions and alarm response are verified which include the following: IRM flux level decreases in the affected channel and results in IRM Downscale alarm in the affected channel. Back panel indications are checked for IRM A only. Ranging down to range 1 is verified to have no affect on the channel indication. Cont rol rod withdrawal block annunciation results in the inability to withdraw control rods. The affected IRM channel is bypassed which clears control rod withdrawal blocks. When the IRM bypass condition is cleared with the malfunction present the control rod withdrawal rod block condition returns. The malfunction is deleted and the affected IRM channel indications return to normal with all affected annunciation clearing. 1 The malfunction test is repeated for remaining channels. The simulator is reinitialized to 100% power.
Grand Gulf Nuclear Station simulator certification Initial Report,-March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 10 of 210-The malfunction is reintroduced into all IRM channels. The affected IRM channel responds ~as before but without the control rod withdrawal block indication. Date of last test: 3/15/90 Concern (s) found/ Planned resolution: Back panel indications could be verified. on IRM -channel A only. IRM channels B through H are not simulated. A simulator review board has recommended that IRM channel 'B- be incorporated into a revised simulator backpanel 1H13-P670 and that the other channels are not needed for training. The revised panel is to be incorporated into an upgrade simulator project for expected completion within -4 years of initial certification. l l
1 i l Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 11 of 210 Malfunction Test #6, IRM Channel (A - H) Detector Stuck Initial Conditions: Reactor startup in progress at approximately 10% power with the main turbine ready to brought on line. All IRM channel-detectors are withdrawn. Sequence of events: All APRM/IRM recorders are selected to record IRM response. All IRMs are ranged down to range 4. Malfunction 6 is introduced into each channel'. Automatic actions and alarm response _are verified including the following: The channel with the stuck detector is selected for insertion but is verified to not move by status light and= lack of IRM channel response. Back panel response is verified for_ channel A only. The malfunction is deleted and the detector is verified'to insert by status light-indication and-increasing IRM channel response. Back panel response. is verified for channel A only. The test is performed on remaining channels B through H. Date of last test: 3/16/90 Concern (s) found/ Planned resolution:- l Back panel indications could be verified on:IRM channel A only. IRM channels B-through H are-not, simulated. A' simulator review board-has recommended that IRM channel' B be incorporated into a revised simulator backpanel 1H13-P670 and that the other channels are not needed for training. The revised panel is to be incorporated.into an upgrade simulator project f o r -' e x p e c t e d c o m p l e t i o n within 4 years of initial certification.
l l I Grand Gulf Nuclear Station simulator .ertification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 12 of 210 Malfunction Test #7, LPRM Detector Fails Upscale Initial Conditions: Reactor startup in progress at approximately 10% power'with-the main turbine ready to brought on line. Sequence of events: Malfunction 7 is introduced for each of 144 LPRMs. RC&IS status lights and indicators are verified'to-indicate that both an LPRM string has an upscale LPRM detector, and the detector with the upscale trip has failed.by selecting each control rod adjacent to the detector and monitoring the RC&IS status lights and indicators. l The APRM with .the LPRM failed = detector input is verifed to l have increased. l LPRMs which i'iput to APRM A are verified to read upscale on the APRM back panel' instrument. NSSS' computer edit OD-8 is used to verify- that the - LDRM- is - upscale. i The failed LPRM detector is bypassed with instructor remote functions. RC&lS indicators and status lights are verified show the
-bypass condition by selecting'all' rods adjacent to-the LPRM detector.
l The-APRM with_the LPRM failed detector input is verifed to have returned near-to its former value. Malfunction 7 is' deleted and the LPRM level is determined to. > have returned to its former value. The= bypass condition is removed with instructor' remote function and status lights are verified to reflect the removal of the bypass.. The-test is performed on the remaining LPRMs, 144 total.-
l l Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 13 of 210 l l For each AT03 channel, A through H, the APRM INOP is verifed l to actuate when at least 9 LPRMs are bypassed which input into that APRM channel. Date of last test: 3/16/90 Concern (s) found/P!.anned resolution: Back panel indications could be verified on LPRMs associated with APRM channel A only. APRM channels B through H are not simulated. A simulator review board has recommended that APRM channel B be incorporated into a revised simulator backpanel 1H13-P670 and that the other channels are not needed for. training. The revised panel is to be incorporated into an upgrade simulator project for expected completion within 4 years of initial certification. The simulator does not.have the capability to perform core monitoring as is done in the control room using the various programs associated with the POWERPLEX core monitoring systep. The simulator review board has determined that the POWERPLEX core monitoring capability is not necessary for simulator training of licensed operators since core monitoring functions are the responsibility of the- Reactor Engineering section. Therefore, this capability will not be incorporated into the simulator. Furthermore, the simulator review board has i recommended that the only NSSS computer options to be retained I are OD 3, 7, and 8.- OD-3 provides core thermal power information from a calculated heat balance. 00-7 provides a scan of rod positions. OD-8 provides a scan of LPRM outputs. These are all that are necessary for training. l l 4
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page_14 of 210 Malfunction Test #8, LPRM Detector fails downscale Initial Conditions: 100% power Sequence of events: Malfunction 8 is introduced for each of 144 LPRM detectors. RC&IS status lights and indicators are verified to indicate. that both the LPRM string has a downscale LPRM detector, and the string with the downscale detector has failed by selecting - each control rod adjacent to the detector and monitoring the RC&IS-status lights and indicators. The APRM with the LPRM failed detector input is verifed to have decreased. LPRMs which input to APRM A are verified to read downscale on ' the APRM back panel instrument.. NSSS computer edit OD-8 is,used to verify _ that the,LPRM is downscale. The failed LPRM detector-is bypassed with instructor remote functions. The LPRM' bypass conditionTis verified on- RC&IS ind'icators and - status lights by selecting all rods adjacent to ' the LPRM detector. The APRM with the LPRM failed detector input is verifed to have-returned near to its former value. Malfunction 711s deleted and the'LPRM level is determined to have returned-to its.former value. The bypass condition: is removed with -instructor remote function and status lights are verified to reflect the removal of the bypass. The test is performed on the remaining LPRMs, 3*d . total.
Grand Gulf, Nuclear Station simulator certification Initial Report, March-1991 Test Abstracts Appendix IX - Malfunction Tests Page 15 of 210 Date of last test: 3/17/90 Concern (s) found/ Planned resolution: Back panel indications could be verified on-LPRMs associated-with APRM channel A only. APRM channels B through H are not simulated. A simulator' review board has recommended-that APRM ' channel B be incorporated into-a revised simulator *backpanel'
-1H13-P670 and- that the other - channels are not needed for training. The revised . panel is - to be incorporated into-an' upgrade simulator project for expected completion within 4 years-of-initial certification.
The simulator does not have the: capability to perform core monitoring _as is done in the control room using the various programs associated with the POWERPLEX core monitoring system. The simulator-review board has determined that-the POWERPLEX core monitoring 4 capability .is not _ nr.wsary for simulator training of licensed operators since ccre menitoring functions-4 are performed by_the Reactor Engineering-section.'Therefore,
' -this capability will not_be--incorporated into_the simulator.
Turthermore, the simulator review board has recommended that the only NSSS computer options to be retained are OD 3,. 7, and
- 8. OD-3 provides core thermal power information from a calculated heat balance. OD-7 provides a .- scan' of- rod positions. OD-8 provides.a scan =of LPRM--outputs._ These are all.that are necessary for training.
-\
I
1 Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Test Abstracts Appendix II - Halfunction Tests - Page 16 of 210 Malfunctior To-t w9, avaM Channel Fails Upscale
, Initial' 'Jonditions:
50% power: Sequence of everats: Malfunction 9 is introduced for each ApRM channel. The correct APRs chart recorder is verified to increase to 125%, and Panel P680 alarms and status lights are verifed to indicate the upscale condition.
- 1. hai scram in RPS and control rod withdrawal block in RC&IS are verified.
The affected channel is bypassed and the bypass condition is verified to occur. The Upscale trip condition and - the - control rod withdrawal block condition are verified to have. cleared.'The RPS half scra- is cleared. Control rod. withdrawal capability is verified. The bypass condition- is . cleared. and the upscale trip conditions are verified to return. The malfunction is cleared. The.- APRM channel indication is-verified to return to normal. The half scram is reset. The control rod withdrawal block is-verified.to have cleared. Correct combinations of APRM channel failures are verified to cause a full PPS Scram. The test is performed on the remaining APRM channels, 8 -tote.1, A th rough _ H.
Grand Gulf Muclear Station simulator certification Initial Rscort, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 31 of 210 Date of last test: 3/17/90 Concern (s) found/ Planned resolution: Back panel indications could be verified on AI9M channel ~A only. APRM channels B through H are not simulated. A simulator review board has recommended -that APRM channel B be incorporated into a revised simulator backpanel 1H13-P670 and that the other channels are not needed for training. The revised panel is to be incorporated into an upgrade simulator. project for expected completion ejthin 4 years of initial certification. I D. og - g g y , "ev r -'~%Y
___ __ __ __ _ _.___ __._ _ _ _ ~. _ _. _ _ _ _- ._ _ _ - _ . . . . Grand Gulf Nuclear Station l Simulator Certification Initial Report karch 1991 i Test Abstracts Appendix II - Malfunction Tests Page it of 210 Malfunction Test #10, APRM Channel Fails Downsc.d n Initial Conditions: 100% power Sequencs of eventet Malfunction 10 is introduced for each APRM channel. the APRM chart recorder is verified to have decreased tm zero. Panel P680 alarms and status lights are verited to indicate the downscale condition. Back panel indications for APRM A are checked, since only APRM A is simulated. A control rod withdrawal block is verified to have occurred. ' control rods are verified to not be able to be withdrawn. The affected channul is bypassed and the bypass condition is verified to occur. The cleared, control rod withdrawal block condition As verified to have control rod withdrawal capability is verified restored. The bypass condition is cleared. The control rod withdrawal l block is verified to return. The malfunction in deleted and. the affected APRM channel lindications are verified to return to ' normal'. Downscale indicators and alarms and the control rod withdrawal block are verified to have cleared. The test is performed on the remaining APRM channels, 8 total, A through H. , f
)
o s
. , . , ~,, .+... -
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 19 of 210 Date of last test: 3/17/90 concern (s) found/ Planned resolutl ~' Back panel indications could be verified on APRM channel A only. APRM channels B through H are not simulated. A simulator review board has recommended that APRM channel B be incorporated into a revised simulator backpanel 1H13-P670 and that the other channels are not needed for training. The revised panel is to be incorporated into an upgrade simulator project for expect 44 completion within 4 years of initial certification. { l
=
.f'
t Crand Gulf Muclear station ! simulator Certification l Initial Report, March 1991 ; Test Abstracts i Appendix II - Malfunction Tests ! Page 20 of 210 i Malfunction Test #11, Jet Pump Failure, . break in riser to jet pumps 5 and 6, in the vessel downcomer region. Initial conditions: ; 100% power
- Sequence of events: I L
Malfunction 11 is introduced. t The following indications are verified to respend to the i break in the riser: ' Calibrated jet pump A, .B, C, and D flow indicators l 1 P619 jet pump flow indicators i Recirculation pump A and-B drive flow recorders l Jet pump loop flow indicators l The total jet pump flow (core flow) recorder The reactor core differential pressure indicator Reactor powcr, level, and Reactor feed pump response . are verified to respond the transient, i After the malfunction is deleted, indicators are checked . bo ensure the effects of the break remain. Date of last test: 3/17/90 . Concern (s) found/ Planned resolutions- , No indication of reverse flow occurred on jet pumps 5 and 6. This deficiency has been corrected and restested. 1
, , - .m . s, .
I Grand Gulf Muclear station l simulator certification Initial Report, March 1991 Test Abstracts Appendix-II - Malfunction Tests Page 21 of 210 t Malfunction Test #12, Reactor Recirculation Pump Trip Initial Conditions: 100% power Sequence of events: Malfunction 12 is entered on each Recirculation puinp. Breaker CB-5 for the affected pump, pump motor amperage, bus feeder amperage, pump differential pressure, pump speed,. jet pump loop flow, and jet pump flow indicators are verified to-respond correctly to the pump trip. Reactor vessel level is verified to adell. Reactor power gnd generator load are vcrifed to drop. After the malfunction is deleted, the .affected loop FCV is closed to the minimum position, -and the Lockout on the pump motor breaker is reset with instructor remote function, the pump is verified to restart with a normal fast speed start sequence, j The simulator is reinitialized to 100% power before the second <
. pump is tested.
The simulator is-reinitialised-at.100% power. Malfunction-12 is introduced to' trip;both pumps simultaneously. Reactor water level is' verified to swell higherefor a dual l pump trip than for a single pump trip and is verified not to
- reach RPV Level 8 as was confirmed by Plant Startup Test' dhta. ,
l The simulator is reinitialized at ~100% power and- the recircula& ion pumps are manually . &,ransferred to ' slow speed : running on the LFMGs. ' Malfunction 12 is' entered for each Recirculation pump. C3-2 ' for the tripped ~ pump, pump speed, pump differentail ; pressure, -jet pump loop flow, and jet pump flow indicators nres verified to respond appropriately. L:
i Grand Pulf Nuclear. Station simulator certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 22 of 210 Reactor vessel-level is verified to swell. Reactor power and generator load are verifed to drop. After the malfunction is deleted, the affected Ivop FCV is closed to the minimum position. The Lockout on the pump motor breaker is reset with a instructor remote function. The pump is veri?ied to restart with a normal manual start sequence. Date of last test! 7/11/90 concern (s) found/ Planned resolutions i No indication of reverse tiow on jet pump flow in the loop affected by the recirculation pump trip. This deficiency has been corrected and retested. a yyc..- -4y ,yy - y v_ g44-'1yp9 = w' +t-
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 23 of 210 Malfunction Test #13, A - P42-F066 failure (variable severity) D - P42-Fil4 failure Initial Condit4.ons: 100% power Soquence of events: Malfunction 13A is introduced at 100% severity, P42-F066 full open. The severity is reduced to 55% and verified to have no effect except partial closure indication on the valve on the P870 panel. The severity is reduced to 50% and verified to cause an RWCU Filter Demineralizer High Temperature alarm at 130*F with no RWCU isolation. The severity is reduced to 35% and verified to cause an RWCU Filter Domineralizer High Temperature alarm at 140*F and isolation of the RWCU system, and a Recirculation Pump A/B Winding low cooling water flow alarm. The severity is reduced to 5% severity and verified to cause Recirculation Pump A/B Seal cooling water low flow and Recirculation Pump / Motor High temperature alarms,. and an increase in Recirculation pump and motor winding tempertures on recorders on back panel P614. The severity is reduced to 0% severity and verified to cause valve closure with no control over the valve it until the malfunction is removed. The valve is reopened on panel P870. Temperature and flow alarms are verified to clear, l l
Grand Gulf Muclear station simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 24 of 210 Malfunction 13B is introduced. This causes P42-F114 to stroke closed. Recirculation Pump A/B Winding low cooling water flow, Recirculation Pump A/B Seal- cooling water low- flow, Recirculation Pump / Motor High temperature are verified to alarm. Recirculation pump and motor winding temperatures are verified to increase on recorders on back panel P614. cfalfunction 13B is removed and temperature and flow alarms are verified to clear. Rocirculation pump / motor A and B ttmperatures are verified to return to normal. Dato of last test: 3/17/90 Concern (s) found/ Planned resolution: RWCU Non-regenerative Heat Exchanger outlet temperature or RWCU filter demineralizer inlet temperature read greater than
!50*F using point 5 of the RWCU temperature select switch with
! no alarm or isolation present. This deficiency will bo carrected on a schedule to be completed within 4 years of initial certification. l
- _ - _ - _-.. = . - - - . _ _. . .- . - = - - - - - - - . . . _
i i Grand Gulf Muclear Station [ Simulator certification ; i Initial Report, March 1991 l l Test Abstracts [ Appendix IX - Malfunction Tests l Page 25 of 210 Malfunction Test #14, Recirculation Pump high vibration l Initial Conditions [ t 100% power Sequence of events: Malfunction 14 is introduced for each Recirculation pump. [ Af ter 1 minute, the affected recirculation pump tigh vibration , alarm is verifed to alarm and not reset with vibration points i on the BOP computer are verified to increase ta approximately , 12 mils. , The affected recirculation pump high v l': ration alarm is verifed to clear when the FCV for the affected recirculation ' pump is closed and to realarm when the FCV is reopened. The affected recirculation pump high vibration alarm is verifed minute to clear approximately. one after the ; malfunction is cleared. > Date of last test: 3/21/90 Concern (s) found/ Planned resolution: No problems noted. l ' I l i i i
- . _ .. y
Grand Gulf Nuclear station simulator certification Initial Report, March 1991 Test Abstracts - Appendix II - Malfunction Tests Page 26 of 210 P ' unction Test #15, Recirculation pump seal failure Initial conditions . 100% power Sequence of events: , Malfunction 15 is introduced to fall the number 1 seal thcn the number 2 seal, then to fail both seals on both-recirculation pumps. Recirculation seal cavity pressures, seal staging high/ low flow and outer seal leak high alarms, and seal temperatures are verified to respond to each failure. When both seals on each pump are failed, recirculation se'al cavity pressures, ssal staging.high/ low flow and outer seal . leak high alarms, seal temperatures, drywell ficor drain flow alarms, drywell temperature and pressure, and drywell air cooler drain flow are verified to respond. Date of last test: 3/17/90 concern (s) found/ planned resolution: No problems noted. l l. l
. _ -_ _ _ ~ . . . _ . . _ _ _ _ ..
m Grand Gulf Muclear station simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 27 of 210 Malfunction Test #16, Recirculation Slow Speed Incorrect Start Initial Conditions: Plant startup in progress with the reactor subcritical and recirculation pumps in slow speed. i Sequence of events: For the affected pump to be tested, its FCV is reduced to minimum valve position and is tripped to off_by opening CB 1 and 2. Malfunction 16 is introduced and then a slow speed start sequence is initiated for the pump to be tested. , The slow speed start sequence is verified to commence up to the point of CB-2 closure. This includes the closure of CB-1 and 5 initially, increase in speed and motor amperage, automatic trip of CB-5 at 95% speed with speed an,' motor-amperage drop, but failure of CB-2 to close at approviaately 25% spoed. The Recirculation pump automatic transfer incomplete alarm is verified to alarm 40 seconds after the sequence was initiated and trip of CB-1. Attempts to restart the pump are verified to fail when the CB-5 start pushbutton is depressed. After the: malfunction is removed, a normal slow speed start is verified. The CB-5 stop pushbutton is used to reset the incomplete sequence. The supplyLbus, voltage isfraised to 7.2 KV.-The CB-5 start pushbutton is depressed. ' The other recirculation' pump is then similarly tested. The- simulator is reinitialized .to . 100%-- power with recirculation pumps in fast speed. Malfunction 16 is introduced for the pump- to be tested. 7L slow ~ speed Ltransfer , sequence is' initiated by depressing the_CB-5 1. and B TRANS TO LFMG pushbuttons. l
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests
.Page 28 of 210 C.-C A and B are verified to trip.=CB-1 A and B are verified !
to Jiase. Pump speed is verified to drop. Pump motor amperage is verified to drop when the transfer sequence is initiated. i CB-2 for the pump tested is verified to fail to close at approximately-25% speed resulting in an-' incomplete sequence for the affected loop. CB-2 for the other-pump is verited-to close to complete the= transfer sequence for the other loop. CB-1 for the. pump to be terted is verified to trip and the recirculation pump automatic transfer incomplete alarm is. verified to alarm. Attempts to restart the pump are verified to fail when the CB-5 start pushbutton is depressed. After the malfunction-is removed, a fast speed speed start is verified. (since the power. interlock will: be met) The CD-5 stop pushbutton is used to reset the incomplete sequence. The ; supply -bus voltage is raised to 7. 2 KV. ' The CB-5 start pushbutton is depressed. The-other-recirculation pump 16 then similarly tested after, reinitializing-to=100% power. Date of last test ~3/17/90 i i Concern (s) found/ Planned' resolution j No problems:noted.-.
~ . _ . . . . _-_l
i [ Grand Gulf Nuclear Station ' simulator Certification ; Initial Report, March 1991 ! Test Abstracts , Appendix II - Malfunction Tests i Page 29 of 210 Malfunction Tent #17, Recirculation Loop FCV Signal Failure } t Initial Conditionst i 100% power with Recirculation flow control on loop manual. requence of eventst Malfunction 17 is-introduced to the A Recirculation loop FCV l then the B Recirculation loop FCV. l For the affected FCV, servo error, FCV motion inhibit alarm i status light, hydraulic subloop IHOP and redundant subloop INOP alarms are verified to respond correctly. ! i The affected FCV is verified 1.ot to have moved.and not to be i able to move with the malfunction active. No change in ! recirculation drive flow is verified. The unaffected.FCV is i verified to function normally. . i When the malfunction is cleared, servo error is verified to f return to zero, and hydraulic subloops restarted with instructor remute functions and their-subloop .INOP alarms are verified to clear. FCV normal operation is verified.to occur when the FCV inhibit , reset pushbutton is depressed. The FCV inhibit alarm and status light-are' verified to clear. Date of last test 3/17/90 I r l Concern (s) found/ Planned resolutions. No problems noted. ! l
+
h
.h h
5 i m & - w 3 -+-my & W- (
I l Grand Gulf F.11 ear Station Simulator certification Initial Report, March 1991 Test Abstracts
!.gpendix II - Malfunction Tests l
Page 30 of 210 i l Malfunction Test #18, Recirculation Loop Flow Feedback Signal I failure Initial Conditions: 100% power with Recirculation flow control in Ivop manual. Sequence of events: Malfunction is is verified to have no effect with recirculation loop flow control in loop manual. Loop manual mode is the only allowed operating mode. Date of last test: 3/19/90 Concern (s) found/ Planned resolutiont No problems noted. l t
i i Grand Gulf Nuclear station sinulator certification ; Initial Report, March.1991 Test Abstracts l Appendix II - Malfunction Tests. Page 31 of 210 . Malfunction Test #19, Recirculation Flow Control Flux Feedback ,- Signal Failure l Initial conditions: '100% power with Recirculation flow control in loop manual. j sequence of events: Malfunction 19 is verified to have no effect with recirculation loop flow control in loop manual. Loop manual mo? ) is the only ; allowed operating mode. ; Date of last test: 3/19/90 concern (s) found/ Planned resolution: No problems noted. '
. t b
t l l I l l l l h I i l L - l-h t
l Grand Gulf Nuclear Station simulator certification InitiP1 Report, March 1991 ! Test Abstracts Appendix II - Malfunction Tests ) Page 32 of 210 Malfunction Test #20, Recirculation _ Flow Control Mas'er contro11or fails low Initial conditions: 100% power with Recirculation flow control in loop manual. Sequence of events: Malfunction 20 is verified to have no effect with recirculation loop flow control in loop manual. Loop manual mode is the only allowed operating mode. Date of last test: 3/19/90 Concern (s) found/ Planned resolution: No problems noted. I i i ( l i
.s,- , , . -
I Grand Gulf Muclear station simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 33 of 210 _; i Malfunction Test #21, Control Rod Drift In Initial Conditions: 100% power Sequence of events: Fifty control rods are selected at random to be tested. This provides assurance that the malfunction will work correctly , for any control rod. Malfunction 21 is inserted for each contrcl rod. The control rod drift alarm is verified to actuate. RC&IS display status lights and indicators are verified. APRM levels are verified to decrease as the rod inserts and that the rod drifts in at about 1/4 speed. i The drifting rod is verified to withdraw at the normal speed with the malfunction active. The rod drift annunciator and RC&IS rod drift' status lights are verified to clear with the withdrawal sequence active. The rod drift indications Jare verified to return when the - RC&IS withdraw pushbutton is released. The rod is verified to I stop inward rod movement when the malfunction is removed. 1 The test is repeated for all selected rods.
. Date of last test:. ' 4/30/90 Concern (s)'found/ Planned resolution: No problems noted.
1' l H 9 9 y y b h e' - r-9_q p. qw a v .-m 's
i
- 8 i '
Grand Gulf Nuclear Station simulator Certification -
- Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests l Page 34 of 110
- Malfunction Test-#22,' Control Rod Stuck Initial-Conditions
- 100% power Sequence of events:
Fifty control rods are selected at random.for testing. This , provides assurance that-the malfunction =will work correctly for any control rod. 1 Each rod is verified to be partially inserted or is positioned' as such, t Malfunction 22 is inserted. for each rod while it is . still-moving from a withdraw or insert command. The rod-is verified-to stop moving and to not be able to be inserted or withdrawn. , Stall flows for insert and withdrawal are-verified.- After increasing CRD drive water differential pressureLup-to. I its maximum of 350 psid, the rod is-. verified'to not be able
.to.be inserted or. withdrawn.-
l .
-After inserting a manual. Scram signal,.the affected rod ist -
verified to remain stuck. The test is-repeated for all-selected rods. i Date of last testi- 4/18/90
- Concern (s) found/ Planned resolutions- No problems noted.
.r
+
m-r - _ _ , , , _ _ , . . . - , . _ , _ . _
s -sn + - . . .- - , u.w.~n us -.a s . ..- Grand Gulf Nuclear station simulator! Certification-Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests ; Page 35 of 210 Malfunction Test #23, control Rod Uncoupled Initial cond' 'ons: Plant startup in progress with RPV pressure at about 400 psig. Sequence of events: Fifty control rods are selected at random. This provides assurance that the malfunction will work correctly for any-control rod. Each rod is withdrawn to position'"48". A coupling check is performed to verify the selected rod will not withdraw-past position "48". Stall flows-are verified. Malfunction 23 is inserted for the rod to be tested while it is at position "48". Another-coupling check is performed and loss of rod position is . verified' to occur with Raw Data selected on RC&IS. The Control. Rod Overtravel annunciator is. verified to alarm. The rod is then inserted full in to position "00" clearing the Control Rod Overtravel annunciator. Malfunction 22, Stuck Control Rod, is: inserted for the-rod to be tested, to stick the blade full in.- The _ drive is' fully withdrawn. APRM and LPRM flux levels are verified . to not-change with the blade' stuck. l j Malfunction 22 is removed to unstick the; blade and cause a. rod ~
- drop. Flux level is verified to: increase rapidly.
The'. test is repeated for all selected rods.- 5/1/90 Date of last test concern (s) _ found/ Planned resolution: No problems noted. i. i i l
l i Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 36-of 210 Malfunction Test #24, Control Rod-Accumulator Trouble Initial Conditions: 100% power with all -control rods fully withdrawn. Sequence of events:
- Fifty control rods are selected ' et random. This provides assurance that the malfunction will work correctly for any -
control rod. Malfunction 24 is entered for . each rod tested. The . Scram Accumulator Trouble alarm is verified to alarm. Depressing the "ACCUM FAULT" pushbutton is ' verified to identify the correct rod on-the RC&IS display.- The control rod is verified to insert from a manual Screm. The simulator is reinitialized: to plant' startup conditions with RPV pressure at 550 psig. Malfunction 24 is inserted for the rod to be- tested and CRD pumps are tripped. All rods except the rod being tested are verified to insert on manual Scram. Date of last test: 5/2/90 Concern (s)-found/ Planned resolution: No control rods moved when a manual Scram was-inserted with no CRD pumps running and RPV pressure less than 600 psig. -This simulator deficiency has been-corrected and retested. Control rods.will now -continue to insert on rods with' sufficient-accumulator pressure. Accumulator pressure has'been verified = - to remain at sufficient pressure to.cause Scramifor up to 10 minutes per Technical Specification requirements. No specific
. plant data is available for times beyond,this..
. .= -
4 . Grand Gulf Muulear station l Simulator certification Initial Report, March 1991 l Test Abstracts ; Appendix II - Malfunction Tests ' Page 37 of 210 l t i Malfunction Test #25, Individual Control Rod Scram i Initial Conditions: 100% power j i Sequence of events: ! i Fifty control rods are selected at random. This provides i assurance that the malfunction will work correctly for any -; control rod. , Malfunction 25 is entered for each rod tested. , control rod drift, Scram Valves, Rod full in and accumulator fault status indication are verified for the affected control . rod on the RC&IS display. Power and generator-load are verified to decrease. ! Withdrawal attempts - on the affected rod are verif h.d - to be !' ineffective. When the malfunction is deleted,. the Scram Valves status light - and the accumulator fault are verified to clear. Ability to ! withdraw the control rod is verified. ! Date of last test: 4/18/90-concern (s) found/ Planned resolution
~
- No problems noted, j f
i I i
?
a
-t t
e , m , ,-r-e. \v-
l I Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 38 of 210 Malfunction Test #26, control Rod Drive Seal Worn Initial Conditions: Plant shutdown with all rods inserted Sequence of events: Fifty control rods are selected at random. This provides assurance that the malfunction will work correctly for any control rod. Malfunction 26 is entered for each rod tested. The rod to be tested is withdrawn in Gang Mode along with the other rods in its gang. The tested rod is _ verified to lag behind the other rods in the gang and the gang misaligned status indicator is verified to function. The flow for the rod with the worn seal is verified to be greater.than the flows for other rods in the gang during withdrawal. The drive with the worn seal is verified to have ' excessive l stall flow during a coupling check. l The gang is inserted and ? the rod with the worn seal is verified to lag behind the ~ other rods in the gang and have - higher drive flow for insert than when it was withdrawn. Date of last' test: 3/19/90~ Concern (s) found/ Planned resolution: When 2 rods . in a 3 rod gang had worn seals, only 1 rod exhibited symptoms of the malfunction. The slow rod did not always have the highest stall flow. This deficiency'will be corrected on a - schedule l to be completed L within 4 years.of initial certification.
Grand Gulf-Nuclear Station Simulator Certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 39 of 210 Malfunction Test #27, CRD Flow Control Valve "A" Failure Initial Conditions: CRD Pump A and CRD FCV A in service at the P601 panel' Sequence of events: Malfunction 27 is entered.-The se* < .ity of the malfunction initializes to that corresponding to the current FCV position. Malfunction severity is reduced to zero. The CRD FCV is verified to fail closed on the P601 panel. CRD system flow and cooling water flow are verified to decrease to approximately zero (5 gpm minimum), and charging water pressure is verified to increase. CRD drive water and cooling water differential pressure are verified to decrease. The C;RD flow controller output is verified to increase to maximum output. The CRD Hydraulic Temperature High alarm is verified to alarm. Attempts to move the rod are verified to be unsuccessful. Control is shif ted to the B CRD FCV using instructor remote functions and flows 0nd pressures are reestablished to normal. Control rods are verified _to move normally. The simulator is reset to the initial conditions. Malfunction 27 is entered. The severity is increased to 100%.and the CRD FCV is verified to open fully on the P601 panel. CRD system flow and ' cooling water flow are ' verified to increase. Charging water differential pressure is verified to decrease slightly. CRD drive water and cooling water dif ferential . pressure are E verified to increase. CRD ' flow controller output is verified to decrease to minimum output. The CRD Drive water differential. pressure high alarm is 1 verified-to alarm. 1 Date of 3ast test: 3/19/90 concern (s) found/ Planned resolution: No problems were noted. l
I Grand G lf Muclear station sind stor certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 40 of 210 Malfunction Test #28, CRD Hydraulic Pump Trip Initial conditions: CRD Pump A in Service Sequence of events: Malfunction 28A is inserted to trip the running CRD pump. Applicable alarms on panel p601 and P680 are verified to alarm. The A CRD Auxiliary 011 Pump is verified to automatically start. CRD flows and pressures are verified to respond to the pump trip. . The CRD flow controller output is verified to have increased to maximum output. ! CRD accumulator fault status lights and alarms are verified to occur on panel P680 and acknowledged on the RC&IS display. The CRD FCV is closed by manually lowering controller output. Manual restart attempts of the A CRD pump are verified not to work. The B CRD pump -> manually started, the' CRD - FCV manually opened to restor: RD. flows and pressure to. normal. Affected annunciators are verified to clear. After deleting the malfunction, the A CRD pump is~ restarted and stopped. The test is re eated on the B CRD pump with malfunction 28B after establishing the B CRD pump running.~ Date of last'testt- 9/7/90 Concern (s)- found/ Planned- resolution:-- 'No problems.were noted.- _.__________________.____J
i I I Grand Gulf Nuclear station ; I simulator certification i Initial Report, March 1991 Test Abstracts i Appendix II - Malfunction Tests : Page 41 of 210 l Malfunction Test #29, RWCU Pump Trip' e Initial Conditions: 100% power with both RWCU pumps A and B in operation ' sequence of events: - Malfunction 29A is inserted and the pump tripped status light and appropriate alarms are verified on panel P680. i Filter demineralizer flows and bottom head drain flows are verified to decrease in response to'the RWCU pump trip._ i Using remote functions, Filter domineralizer_ A flow is reduced and Filter domineralizer B flow is verified u to increase. A-- ,
. Filter . demineralizer ef fluent . conductivity is verified to respond to the flow reduction. i 1
The A RWCU pump is verified to not be.able to be' restarted. RWCU temperatures are verified to have changed in response.to the pump trip. After deleting the malfunction, the A RWCU_ pump is restsrted j and the A filter domineralizer is restored to. operation. The test is repeated for the-'B_PWCU' pump;using' malfunction 29B. Both RWCU pumps are tripped using malfunction 29 A and B. ,
. Appropriate RWCU-alarms are verified on panel P680, the RWCU i< pump trip status lights are-verified to' light. RWCU flows and:
' bottom head drain line flows-~are-verified to decrease'. RWCU '
filter domineralizers; are verified Lto automatically _ goroff 1 lina-as-indicated by instructor remote function status change.
~
-Restart attempts for both pumps are verified to fail with the '
malfunction active.- Date of last test: 4/4/90 Concern (s)-founi/ Planned resolution: _ No_-problems noted.1 l J
i I f s t Grand Gulf Nuclear station simulator certification Initial Report, March 1991 ; Test Abstracts Appendix II - Malfunction Tests , Page 42 of 210 l i Malfunction Test #30, RWCU Drain Valve Fails Shut Initial Conditions: RWCU Blowdown in service for RPV lovel control. l I i Sequence of events: Malfunction 30 is inserted and RWCU blow'own d flow is verified , to decrease to zero on panel P680. i RWCU temperatures are verified to' respond to'the change in- ! system flow. RPV level is verified to increase since the amount of blowdown flow is decreased with no' change in RPV makeup. The RWCU blowdown controller output is increased to the
- maximum. RWCU blowdown flow is verified to stay at zero flow.
5 After the malfunction is deleted, RWCU blowdown flow is- ; l Verified to increase and RPV. level-to begin to decrease. ' RWCU Blowdown flow is adjusted to stabilize RPV level. C0te of last test: 4/5/90
~
concern (s) found/ Planned resolution: No-problems noted. ' t r
~
s
- , my ye, -e - ,~*v '
Grand Gulf Nuclear Station , Simulator certification Initial Report, Maren 1991 Test Abstracts ' i Appendix IX - Malfunction Tests Page 43 of 210 Malfunction Test #31, RWCU Demineralizer B High Differential Pressure Initial conditions: RWCU Filter Demineralizer B in service Sequence of events: Malfunction 31 is inserted. The "B" Filter demineralizer is verified to go off line by checking instructor remote function status. RWCU system flows, differential pressure and conductivities are verified to respond to the change. Appropriate alarms on P680 panel are verified to alarm. Af ter deleting the malfunction, B RWCU filer demineralizer is placed back in service using instructor remote functions. Flows and differential pressure are . verified to return to i normal and alarms are verified to clear. ! Date of last test: 3/21/90 Concern (s) found/ Planned resolution: None
- - ..- . .- . . - ._ . .. - .. . ~ - . - . - _ . . . - - .
Grand Gulf Nuclear station Simulator <!artification Initial Roport, March 1991 ! Test Abstracts Appendix IX - Malfunction Tests Page A4 of 210 Malfunction Test #32, RWCU Deminsralizer Resin Depletion Initial Conditions: 'RWCU systam in service' with both RWCU filter det '.nnralizers in service. Sequence of events: Malfunct3on 32A is inserted. RWCU conductivities and high conductivity alarms are verified to occur on the P580 panel. After deleting the malfunction, conductivities are verified to return to normal and conductivity alarms are verified to clear. i The test is repeated using malfunction 32B for the B filter dominuralizer. Date of last testi 4/6/90 Concern (s) found/ Planned _ resolution: RWCU Filter Demineraliser Effluent Conductit'ity on G33-CR-R603 did not respond to the malfunction on the A u.74 the_B filter i demineralizer. This deficiency will' hs correctsd on a schedule _to be' completed _w ithin 4 years of' initial-certification.
Grand Gulf Nuclear Station Simulator Corti $1 cation Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 45 of 210 Malfunction Test #33, RWCU System Leak (Variable) Initial Conditions: RWCU system in service Sequence of events: Malfunction 33 is inserted at 50% severity.' Readings on Riley temperature switches on back panel P642 are checked to verify that RWCU heat Exchanger room temperatures are increasing. The severity is increased to 100%. RWCU heat Exchanger room temperatures are verified to be increasing at a greater rats. RWCU temperature alarms are verified to alarm on the P680 panel. RWCU heat exchanger room Riley temperature switches are verified to trip on back panel P642. The RWCU system is verified to isolate. RWCU pumps are verified to trip and not be able to restart, and system flows verified to decrease. Filter domineralizers are verified to go off line by checking the status of instructor remote functions. Date of last test: 4/4/90 Concern (s) found/ Planned resolution: Riley temperature switchen on panel P632 in the upper control cabinet area are not simulated. The simulator review board has concluded that this panel is unneccessary for training and will not be included. There was no indication of differential flow indicated on back panel P642 while malfunction 33 was active. This deficiency will be corrected on a schedule to be completed within 4 years of initial certification. h
l , I i Grand Gulf Nuclear Ste. tion. .
. Simulator certificatlun Initial Report, March-1994 Test-Abstraftts
. Appendix II - Malfunction-Tssts.
Page 45,of 210 , Malfunction Test #34, RC&IS Channel Disagree , ( Initial conditions: 100% power l-Sequence of events: During the insertion sequence of-a control rod,, malfunction 34 is entered. The RC&IS:" Channel *isagree"1 status-1ight is verified .to -light. The "RC&IS INOP" alarm .is verified to alarm and the inserte'd control rod verified 'to settle at theilas' notch passed. . . An attempt to move the control rod-is vorified'to' fail;-RC&ISL channels are -verified to- . ofirm: a timer malfunction 'in channel 1 but not in channe : Af ter deleting malfL.J ion' 34, the -RC&IS "Cha nnel :. Di~sagree" ' l status-light and'"RCh1STINOP" a7 rm.are verified to-clear., L l Normal control rod movement is'ver!fie.l. Date of last test: 7/26/90 . Concern (s)-found/ Planned resolution: No problems noted. i l l L' _ - ..~.. . . . - . . . . . _ . .- .,c , , , _ . .-
Grand Gulf Nuclear St2 tion Simulator Certificntion Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction. Tests Page 47-of.210 Malfunction Test #35, RPIS Failure (Open Reed Switch) Initial Conditions: 100% power Channel-1 RC&IS data selectml ;
-Rav Oata and individual drive mode selected
~
Sequence t.f events: Molfunction 35 is inserted. A control rv -is inserted continuously then allowed to settle. "Chan: 31 L ;virrae", _ " Data-Fault", and "Pithdraw Inhibit" lights are verified on the RC&IS Display. The Control Rod Withdrawal. _ Block" alarm is verified to alarm on the P680 panel. After selecting channel 2 hta, rod-position'is verified to display a blank. After deselecting Raw Data,-channel 2 data is verified.to be displayed. After selecting.both channels to , alternately display data and . selecting Raw Data again, the-display siternately displays rod position and then_a blank. - AN NSSS edit OD-7 Option 2 is verified to show="-99" for the l position of the affected rod. , The: control rod 'is inserted one notch. Raw Data is deselected. l The "Chennel Disagree" status light is terified to light. With ! both channels- se19cted, the previous and new rod -position' are verified to alternc*.ely display. Channel :1' is _ verified. to - display the 'new positic3. Channel' 2.'is verified to display;the former position. Raw Data is selected.. e.- substitute position ent,'y 1.s verified to Jause:a-substitute position error status light. Control rod' withdrawal is verified to be blocked. When the malfunction is - deleted, the " Channel Disagree", Withdraw Block", " Withdraw-Inhibit" and " Data' Fault" RC&IS -
--status-lights are verifiea to clear.
r b
, s ..,s~. .r w - e +1 a < - , , -
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page-48 of 210-l The control rod gang is realfgned, and the tast is' repeated I to verify similar -ef fects in gang drP e mode. The " Gang Misaligned" status light on the RC&IS dj ' lay is verified-in-. this mode. The simulator is reinitialized to 20% power below the Low Power Setpoint, with both RC&IS data channels selected, Raw Data selected, and gang drive modo selected. Af ter ' entering malfunction 35', a rod' gang is selected and inserted one notch. The gang is verified to"e o misaligned. The
" Channel Disagree", " Withdraw Block" , " Insert Block", " Data Fault", " Withdraw Inhibit",.and " Insert Inhibit" RC&IS status lights are verified to light. The control rod withdrawal block annteci et .,r 16 verified to. alarm. Rod position is ~ verified to
-t alternately display new ' and blank rod! . position for the selected. rod. <
Date of'last test: 7/24/90 i Concern (s) tound/ Planned resolution: This malfunction failed-to cauce a Data Fault, and. Control Rod Blocks when required. . This deficiency will be corrected-on-a schedule to be- completed 'within L4 years of : initial-certification. 3 9 l-i I Ma a
Grand Gulf Nuclear Station-simulator Certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction; Tests Page 49 of 210 Malfunction Test #36, RPIS Failure (Closed Reed Switch) Initial Conditions: 100% power Channel 1 of RC&IS data selected , Raw Data selected
-Individual Drive Mode selected Sequence of events:
After entering malfunction'36, the selected control rod is inserted one notch. A." Channel Dicsgree". status light is verified -to light on RC&IS. After selecting Channel 2 data,_ a Data Fault ,is verified to exist on channel'2 'r the selected rod. An NSSS edit OD-7 option _2 is verit ~
'd=to display _a data fault for the selected: rod.-
After selecting both data channels,~ good data and data fault data are verified to alternately _ display.-- Attempts.to move the' rod are verified-to be unsuccessful.- After entering substitut4 data from the good' channel to_the bad channel and verifying substitutefdata, the1 selected. rod is moved-one notch. The Channel Disagree:and data fault are verified to return for the new rod position. . After deleting malfunction 36, the-channel disagree and data-fault are verified to clear-and rod position is verified'to i be restored. Date of last-test: 4/4/90 Concern (s) found/ Planned resolution: The control rod could be: moved with a data fault and channel disagree present. This deficiency will_ be corrected or a schedule to-- be completed- within '4 years- cf init?41-certification. - f i 1
Grand Gulf-Nuclear Station Simulator certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 50 of 210-Malfunction Test #37, RC&IS Timer malfunction Initial Conditions: Approximately 1% power Individual d*ive mode selected Both RC&IS data channels selected Sequence of events: The selected control rod is continuously withdrawn = 4. notcates. Malfunction 37 is inserted during the withdrawal sequence. A Channel Disagree condition and a. failure of the.RC&IS timer in-channel ; is verified. Attempts to select.another rod are verified
- to fail. Attempts to moveLthe selected. rod are veritted to' fail.
RC&IS indications are verified 'to return to normal af ter - the i malfunct. ion is deleted. - Normal . control. rod movement is : verified' restored. Date of last test: 3/22/90 concern (s) found/ Planned resolution: No problems'noted. l
l l l Grand' Gulf Nuclear' Station simulator Certification' Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page:51 of 210 , Malfunction Test #38, RC&IS Self Test Failure Initial Conditions: 100%.. power Sequence of events: Malfunction 38 is ' inserted during a control rod -insertion:
. sequence. An "RC&IS - INOP" alarm is verified to occur. A
" Channel Disagree" status light and'a flashing " Test Display" status light are verified on the RC&ls-section.of panel P680.-
Stabilizing valves.in the CRD hydraulic system are verified to close by drive flow differential pressure and cool!ng water flow indication on the P601 panel. While RC&IS is inoperative, malfunction 24, CRD Hydraulic Accumulator Trouble, and malfunction 25,' Individual Rod Scram, are inserted for two dif ferent rods. The scan function is verified to not update the status of these two events. !- - At. tempts at control rod motion are verii. ed to fail. l After malfunction 38 is deleted, the "RC&IS INOP" alarm is i reset. I Date of.last test: 7/26/90 Concern (s) found/ Planned resolution: Stabilizing _ valves did. not indicate closed- and the " Test-l Display" status light did not flash while the malfunction was active. The "RC&IS INOP".elarm cleared before being reset.- These deficiencies will be corrected - on a schedule to be-completed within 4 years of initial--certification. l l
, ... ~. . , , . -
. ~ --
l I Grand Gulf Nuclear Station simulator Certification-Initial Report, March 1991 l .
-Test Abstracts-Appendix II - Malfunction Tests Page 52 of 210 l Malfunction Test #39, Rod Pattern Controller-fails to initiate l rod block' Initial Conditiens: All Rods inserted with sequence A-selected Sequence of events:
With malfunction 39A,1 inserted, cont &ol rod withdrawal blocks are verified not to occur at "04" for a gang of group -1 rods. A control rod withdrawal. block is verified to occur for this ; gang at "08". The remaining gangs of rods in group 1 a're withdrawn to "06" and malfunction 39B,1 is entered. Control rod withdrawal blocks are verified not'to occur at "08" for a gang of group 1 rods but.is verifled to occur at ! "12". l l The remaining gangs-of rods in group.1 are withdrawn to "10". Malfunction 39C,1 is inserted.' Control rod withdrawal blocks .are verified ' 90t to ' occur at "12" for.a gang-of group 1 rods. The simulator ~is reinitialized and the test is performed on group 2 control rods c using malfunctions 39A,2, 39B,2, and 39C,2 similar to' the test for group 1 rods at "04", "08", and- .
"12".
The simulator is reinitialized to all ' rot;r, in with sequence "A" selected. Group 1:and'2;contro11 rods are-withdrawn.-The test is performed on group 3 control':'ds using malfunctions 39A,3, 39B,3, and 390, 3 similar to the test for-group 1 rods;- at "04", "08", and'"12". The simulator is reinitialized tc all-rods in with sequence "A" selected. Groups 1 and 2 control rsGs are withdrawn. The-test is performed on group-4 control rods using malfunctions , 39A,4, 39B,4, and 3?C,4 similar-to'the-test'for group:1 rods at "04", "08", and "12". l
=
_- -- ~-
l Grand Gulf Nuclear Station-
' simulator certification Initia11-Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page !3 of 210- ,
Grou). 3 and 4 control: rods are-withdrawn to es.tablish groups-1, 2, 3, and 4: 311 at."48".-The "GP1-4 Full'Out" status light, - is-verified. Using these -initial conditions groups 7,'8,9,--and: 10 are tested using malfunctions 39A,7, 39B,7,'39C,7, 3 9 A ,- 8 i . and -39B,8, 39C,8, 3 9A,9, . 3 9B,9, --39C,9, 3 9 A ,10, : ' 3 9 B ,10, andt 39C,10 similar to the test ~for group 1 rods at ' "04",~ "08"i and "12". ^ Group 3 and 4 control rods-are withdrawn to establish groups-1, 2, 3, and 4 all at "48". Malfunction'39 is-entered..
- Control rod withdrawal blocks are verified: not -to occur at-
~
- "12" for a_ gang of group 5:-control rods. The -gang. is ' inserted to "00".
l- Control' rod withdrawal- blocks; are verifiedinot Lto . occur _atL j '"12" for'a_ gang of group 6 control. rods.- .
.7 The-simulator-is initializedLto approximately'35%-power. This-l
-is above the Low Power:< Set' Point .and i below. the; High : Power Setpoint. Fifty control- rods- are selected -at random. '
Malfunction 39,11'isl entered.. For each control rod - tested Eat l position 48") thel rod . is inserted 5_ notches. Rod felect clear is, depressed.:The rod?is' ; rat. elected _ and'. then ' withdrawn ' 5 ; notches ? toi position 18." '. =
- Control rod 1 withdrawal blocks: are veritied7not L to1 occur at-
~ "46".-
For. each s control < rodJ at ? position ."38. or loss,f thQod. is - nithdrawn'to positior "48".. Control: rod withdrawaltblucks are ;i L verified.not..to'occubas-they should'at 4Jnotches. Th'e simulator 'is initialize to approximately< 100% power whichi l is~ above the High ; Power,' Setpoint. . . -50 control rods are.
- selectedv at random. Malfunction 139,12 iisL entered. >
For ~ each control. . rod tested " atL position ; '"40 . thei rod his
._ inserted.3; notches,: Red Select ClearLis-depressed, the rod is -
resolected t and- thenEkithdrawn - 3 notchesoto position "48". . - Controlf rod withdravalTblocks L;ake; verifieds not ' to Roccur. att:
"46".1 m
'h i
,p~ , ,s -,,r, ,
e.Ls , , .m. a u
,,..,,'i,, ,,-,wra . ., , , . Ln e , + . a
l l l l l Grand Gulf Nuclear Station j Simulator certification l Initial Report, March 1991 Test Abstracts Appendix II Malf metion Tests Pt 54 of 210 t ! For each control rod.at-position "42" or less, the rod is , l withdrawn to position "48". Control rod withdrawal blocks are verified not to occur as they-should at 2 notches. Date of last test: 4/9/90 Concern (s) found/Planr.ed resciution: With malfunction 39 active, control rod withdrawal was blocked at position "12" for group 5 and 6 control ' rods. These deficiencies will be cc_ acted on a schedule to be completed' within 4 years of initial certification. l e l l f L s A v c a a em - -t ,-w -# y-- . g w t- t- r - % -e=,=
Grand G:tif raclear Station simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 55 of 210 Malfunction Test #40, Drywell Pressure Transmitter, B21-N094 A, E, B, and F Failure (Variable Severity) Initial conditions: 100% power Sequence of events: 1 Malfunction 40A is entered with severity to fail as is. Malfunction 63, Recirculation Line Break is entered at 5% severity to establish a LOCA signal- from High Drywell pressure. The following systems are verified to have responded as ' indicated: Division 1 ADS High Drywell pressure logic E Division 2 ECCS (LPCI B and C) Div 2 ADS high Drywell pressure logic B and F Division 3 ECCS Recirculation system FCV Inhibit Reactor Scram Containment, Drywell, and Auxiliary Building Isolations SPMU Division 2 after a manual initiation The following systems are verified to have not initiated as indicated: Division 1 ECCS (LPCI A and LPCS) Division 2 ADS High Drywell prescure logic A Division 1 CGCS Division 1 SSW SPMU Division 1 after a manual initiation After manually initiating Division 1 ECCS logic, Division 1 ECCS (LPCI-A and LPCS) , SSW Division 1, CGCS Division 1, DG11, Division 1 LSS are verified to actuate. SPMU Division 1 is. verified to actuate after manually initiating it. The simulator is initialized to 100%. Malfunction 40A is a entered at 300% severity. Appropriate annunciators on panel P601 are verified as well as the A ADS High Drywell pressure logic status light. l
f L GrE36 G G f Muclear Station simulatcr Certification-Initial Report, March 1991-Test Abstracts Appendix II - Malfunction Tests Page 56 of 210 Division 1 Containment- Spray is . manually initiated. Appropriate annunciators _ on-panel P601 are verified to alarm. RHR A is verified'to auto start and align _to the containment spray mode. SSW A is verified to auto start and align to the RHR A heat Exchanger. Containment pressure is verified'= to respond to the-spray action. Division 2-Containment Spray is manually initiated. No other 9 effects except the alarm associated with arming the manual initiation pushbutton are-verified to occur. The. test is repeated similarly- for malfunction 40 . E .which affects division 1 ECCS logic. The t e s t -- i s _ r e p e a t e d for-malfunctions 39 B and F which' affect. Division 2~.ECCS logic. . The simulator is initialized'to 100%.- Combinations of logic failures are entered by _ inserting malfunction - 40 at 100% severity for these combinations as follows:- For malfunctions 40 A and E, Division.l'ECCS, DGil, Division-- 1 SSW, Division 1 CGCS:are' verified to initiate. For malfunctions 40 B and F, Division'2 ECCS, DG12, Division 2-SSW, Division 2 CGCS are' verified to initiate.
~
For malfunctions.40'.A and B,'no initiation's_are; verified-to _ ,- occur. For malfunctions _40 E and F, nofinitiations are verified to occur. For malfunctions 40 A and-F,.no initiations.are-verified to occur.
~
For malfunctions 40 B-and E, no initiations are verified to occur. The: simulator,is roset.to 100% power. Malfunction 40 A,-B, E, 2 and F are entered to fail as is. Malfunction 63,- Recirculation line-break is entered at 100% severity.
\
l
Grand Gulf Muclear Station simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 57 of 210 Division 1 and 2 ECCS actuations are. verified to occur on RPV-Level I. Division 1 and-2 High-drywell-pressure annunciation ott panel P601 for ECCS isl verified not to alarm. The-ADS high drywell pressure logic status lights are- verified not to-light. Division 3 ECCS is verified to have initiated on high-drywell pressure and low RPV Level. Date of last test: 3/23/90 concern (s) found/ Planned resolution: No problems were noted. i-t l-t . 1 i A
= - + t -
-t v as_.- ,, + -m -ww ,.vn < w e.,re-- r ,w A , w -e renoe < rw __w-
i 1 i Grand Gulf Muclear Station
-Simulator Certification Initial Report, March 1991 Test Abstracts Appendix _IX - Malfraction Tests Page 58 of 210 l
l l Ealfunction Test #41,- RPV Level. Transmitter, B21-N091 A, E, B, l and F Fails (Variable severity). Initial Conditions: 100% power Sequence of events:
' Malfunction 41A is entered to fail B21-N091A as is.
In order to establish a RPV Level T condition without a High-Drywell pressure condition, malfunct10n 28A,- A.CRD Pump Trip, malfunction 47, RCIC Turbine Trip, ma'. function 52, HPCS pump , trip, malfunction 70, Feedwater line_ rupture in the turbine building, and malfunction-68 at 50% aaverity,-main steam'line-rupture in the turbine building are entered to. actuate E l simultaneously. The reactor modo switch is placed to shutdown-as soon as Scram occurs,'which ensures.MSIVs-do not isolate
-on Low Main Steam Line pressure.
Whei Wide Range RPV flevel ' I' . is reached the following are. verirled to have'or have not occurred: Division 2 ECCS initiation-(LPCI Brand C) ! HPCS initiation with a HPCS pump' Trip. > RCIC initiation:with a RCIC_ turbine trip i Division-2 LSS actuation with DG12' auto start RPV ' level. stabilizing . af ter Group I. isolation at RPV Levul I Division 1 SSW initiation due to RCIC_initiaticn but not lining up t_o a LOCA;1ineup.-(SSW aligning to the RHR l- A Heat-Exchanger).
"RX LVL 1 (-150") .LO" alarm on the division l' .(RHR A/IUCS)- portion ofi the P601' palml but -failure of Division 1 ECCS pumps to start (7 PCS and LPCI : A)-
The simulator is initialized to 100%.The same' test isLrun exc'ept using malfunction 41E.1This-prevents Division 1 ECCS
-actuation.
l The slaulator -is initialif.ed to 100%. The same test tis run l using malfunction 41B. .This prevents Division 2 ECCS
- . actuation.
1
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Test Abstracts-Appendix IX - Malfunction Tests Page 59 of 210 The simulator is initialized to 100%. The same test is run using malfunction 41F, -This prevents Division 2 ECCS actuation. The simulator is initialized to 100%. The same. test is run using malfunction 41 A and E. Alarm "RX LVL 1:-.(-150") LO" is verified to not alarm on the division 1-(RHR A/LPCS) portion cf the is'1 panel. Division 1 ECCS pumps are verified not to initiate. _ Tht. ADS A High Drywel? Bypass Timer is confirmed to not actuate. The sinulator is initialized to 100%. The same test is run using malfunction 41 B and F. simultaneously. Alarm.'"RX LVL 1 (-150") LO" is verified to not alarm on the division 2 (RHR B/C) portion of the P601 panel. . Division 2 :ECCS pumps are verified not to initiate. _ The ADS B High Drywell Bypass Timer is confirmed to not actuate. The simulator is inttialized to 100%. The-same test is run; - using malfunction 41 A and B simultaneously. Division 1 and' 2 syatems are. verified not to initie.te. The simulator.is initialized to1100%. The same test'ia run using malfunction 41 E and F simultaneously. Division 1 and 2 systems are verified not to initiate. The simulator is initialized to 100%. In-order to establish a RPV Level I condition without a High - Drywell pressure condition, the following malfunctions'are-entered'to actuate simultaneously: Malfunction 28A, A CRD Pump Trip-Malfunction $2, HPCS pulp-trip
. Malfunction //0, Feedwater- line ' rupture in the turbine building.
Malfunction 68-at 50% severity, main steam line rupture.
- in the turbine. building
1
>m,, .r s s.
- tv --=4e'
- 1 Grand Gulf Nuclear Station
! simulator Certification Initial Report, March.1991 l l Test Abstracts Appendix II Malfunction Tests Page 60.of 210 Malfunction 41 A and F are entered simultaneously. Division 1 and 2 systems are verified not to initiate. RCIC is also verified not to initiate. This portion is repeated except using malfunction 41A and F and verifying the same outcome. The simulator is initialized to 100%. -Malfunction- 41A is entered ' and its severity reduced to 0%. Alarm'"RX-LVL 1 ( 150") Lo" alarm on the. division 1 (RHR A/LPCS) portion _of the - P601 panel and the ADS A High Drywell Byoass timer alarm is verified to alarm and the division 1 wide range level recorder-is verified to read -150". This portion is repeated except using malfunction ' 41E , and - verifying the same outcome. The simulator is initialized to 100%. Malfunction 41B .is l enterop and its severity reduced to 0%.-_ Alarm "RX;LVL 1I(-- ) 150") LO" alarm on-the division'2 (RHR' B/C) portion of-the P601 panel and the ADS B High-Drywell Bypass timer' alarm is-verified to alarm. The division 2 wide range: level recorder is verified.to read -150". This portion of the test- is repeated except using malfunction 41F and verifying the.same outcome. The cimulator is roset to 100% power. ' Ccmbinations of malfunction 41 are inserted. Severity is L varied. Automatic: actions and actuation of appropriatecalarms are verified at i each severity. Severities are varied from 53%, 0%:and then to 100%. A severit.* of 53% tested cortoct logic' combinations which cause acto j nitiation of RCIC at RPV level 2. L A severity of_100%-tested correct logic-combinations which-cause auto
-closure of E51-F045 at RPV lsvel 8. Malfonction'41 logic combinations tested included the.following':
A and B A and E B and.E B and.F E and f s
- -- , . i
Grand Gulf Nuclear Station simulator-Certification Initial Report,-March-1991 Test Abstracts-Appendix IX - Malfunction Tests. Page 01 of 210 t certain logic combinations of - RPVf Level 1_ and High Drywell' pressure will or will not initiate Division 1 or 2 ECCS. To test this malfunction-combinations of m.?lfunctions 40 and 41- ~ w.re inserted (with the severity-to cause RPV levelil 1 b !.gh - i orywell pressure) as follows: , 40A and 41A Division 1 ECCS' docs 3ct,1nitiate , 40A and 41E Division 1 ECCS initiat.es : 40E and 41A Division 1 ECCS initiat6H. 40E and 41E Division 1 ECCS does not: initiate ~ ' 40B and 41B Division'2 ECCS doesinot' initiate-40B-and 41F Division 2 ECCS initiates 40F and 41B-DivisionJ2 ECCS-initiates . - 40F and 41F Division 2 ECCS does-not initiate Date of last~ test: 3/30/90 Concern (s) found/ Planned resolution:- E51-F045 did not automatic close whenlmalfunctionso41A:and B'- were applied to fail with a wide range level-8 trip (condition. E"#1-F045 closure occurs when RPV Levele 8 onLNarrow-RangeLis reached. These deficiencies will be corrected-on<a schedele to be completed within;4 years'of'initialicertification. t E i i
.: f 5- t I
t -s I s v w br- r -'me-4% 'a-t'en<**n v"* t % m-
Grand Gulf Nuclear Station Simulator Certification Initial Repott, March 1991 Test Abstracts Appendix IX - Ma3 function Tests Page 62 of 210 Malfunction Test #42, RPV Pressure Transmitter BPl-N068 A, B, E, and F. Fails (variable severity) Initial conditions: 100% power. Sequence of events: Malfunction 42 is entered to fail each transmitter individually as 10. SRV logic A or B is verified to actuate or fail to actuate for each case.in conjunction-with a pressurization event caused' by malfunction 79, IPC Failura Low. The simulator is reset in between each event.- The simulator is reset to'100%. Malfunction 42 is then entered again to fail euch transmitter individually as is. Severityfis increase'd uc 97% then back to 0% and verifying proper alarm response. While _each transmitter is at 0% the low: pressure ECCS manual ~ initiation pushbuttons are depressed and automatic' injection ; valve-opening is verified.-This portion of the test is repeated for. each transmitter, resetting the simulator in-between events. l The combination cf . A = and E transmitter . failures .are then l tested ramping the severity to 0%, = then to 974, to 0%, to 99%, to 100%, back:to 99%, to 98%, to 93% t o 91%, then' to 85%. This verified proper actuation of SRV low-low set logic and- SRV - opening and closing in-SRV. logic-A. This test is then repeated forithe combination B.and.F-
' transmitters to test SRV logic B.
Date of11ast test; 3/30/90 l Concern (s) found/ Planned resolution: B21-F051D did not close when the combination -of, A and E transmitters had been ramped from 97% to 0%.-This detficiency will be corrected-on.a schedule to be. completed within 4 i :rs of initial certification. wwa , e -n s , -
< e
Grand Gulf Nuclear Station i simulator certification Initial Report, March 1991 ! Test Abstracts Appendix IX - Malfunction Tests Pr.ge 63 of 210 Malfunction Test #43, RCIC Auto Start Failure Initial conditions: 100% power. Sequence of events: Malfunction 43 is activated. A loss of feed event is simulated by manually tripping both Reactor Feed pumps. is verified not to automatically initiate at Level 2. TheRCIC Malfunction is deleted and proper system initiation is verified to occur. Date of last test: 3/20/90 Concern (s) found/ Planned resolution: None. P
Grand Gulf Nuclear Station Simulator Certification -r Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 64 of 210 Malfunction Test #44,-RCIC Turbine Speed Control Failure Initiel Conditions: 100% power. Sequence of events: RCIC is manually started with injection into the RPV. The RCIC flow controller is placed in auto with flow rate at 800 gpm. Malfunction 44 is inserted. The RCIC system is verified to respond by a_ drop in speed to.its low speed. limit, drop in-flow to O gym, drop in discharge pressure, and auto opening of the RCIC pump minimum flow valve. , The flow controller is placed in manual and lack of control is verified. The severity of the malfunction is increaseduto 25%. The RCIC turbine is verified to increase speed to 2000 RPM. Controller output is verified to increase to 100% output. The severity is increased to' 65% and RCIC turbine speed,. discharge-pressure and flow are verified to increase. When the malfunction is deleted, the flow controller is; verified to respond normally. L Date of Last Tes' c: 3/20/90 Concern (s) found/ Planned resolution: None. l {
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Test Abstracts Appendix IX - Malfunction Tests Page 65 of 210 Malfunction Test #45, RCIC Flow Transmitter Failure-Initial Conditions: 100% power. Sequence of events: The RCIC system in manually initiated withiinjection to the RPV. Malfunction 45 is activated and the controller is verified to ' increase to maximum output with no -indicated flow.
~
After switching to manual control, RCIC is verified to control properly. Malfunction 45 is deleted and RCIC parameters are. - confirmed to return-to normal. Date of last test 3/20/90 Concern (s) found/ Planned resolution: None. t i t r --- t- n yW w ~ e m
Grand Gulf Nuclear Station-Simulator Certification Initial-Report, March 1991-Test Abstracts Appendix IX - Malfunction Tests - Page 66.of 210 Malfunction Test #46, RCIC. Turbine Overspeed Initial Conditions: 100% power.- Sequence of events: RCIC is manually started and brought to rated flow in the CST Test Return flow-path. Malfunction 46 iF. entured and the RCIC. system is verified to respond byl increasing in flow. and speed until it trips. After deleting oalfunct;-on 46, the-overspeed trip device is reset with remote function 100. RCIC is restarted and verified to operate normally. i Date'last test: 3/21/90 Concern (s) found/ Planned resolution: None.
'l I'
. - . , . . - - - , . - .- .4 - . . .
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Test Abstracts Appendix II - Malfunction Tests Page 67 of 210 Malfunction Test #47, RCIC Turbine Trip Initial Conditions: 100% power. Sequence of-events: RCIC is manually started and brought to rated . flow and pressure. Malfunction 47 is activated. This simulates suction pressure transmitter (N053) failure. Indicated suction pressure is verified to remain unchanged while the "RCIC PMP Suct Press Lo" alarm activates. Af ter malfunction 47 is deleted, RCIC is reset and. restarted and verified to operate normally. Date of last test: 5/13/90 concern (s) found/ Planned rest -.ution: None.
Grand Gulf Muclear Station simulator certification Initial Report,- March 1991 Test Abstracts Appendix II - Malfunction Tests Page 68 of 210 Malfunction Test #48, RCIC System Isolation Initial Conditions: 100% power.- Sequence of events: RCIC is manually started with injection to the RPV-at rated flow. Malfunction 48 is entered. RCIC isolation and turbine trip indications are verified. The isolation is verified not to reset until malfunction . 48 is - reset. The RCIC- system isolation valves are reopened, and the RCIC system is verified to restart normally, Date tf last test: 3/22/90 concer: (s) found/ Planned resolution: None.
Grand Gulf Nuclear Station simulator Certification Initial Report,' March 1591 Test Abstracts Appendix IX - Malfunction Tests
~Page 69 of 210 Malfunction Test #49, RCIC Steam Line Break (Variable _ severity);
Initial Conditions: 100% power. Sequence of events: The RCIC system is manually started and brought to rated flow-in the CST return flow path.1 Malfunction-49-is inserted at 20%_ severity. Area temperatures and area radiation-monitors for the affected areas are verified to increase on backpanels H13-P642 and H13-P844. RCIC- isolation' and turbine . trip indications are verified. After resetting the simulator, RCIC is again manually. started and brought to -rated flow in the CST return flow - path. Malfunction 49 is. entered at 100%-severity, RCIC Division I , and II Steam line high differential, pressure alarms and tho' i effects of PCIC isolation and_ turbine trip are verified except for temperature and- radiation - alarms actuating. Date of last test: 3/22/90 Concern (s) found/ Planned resolution: None.
.1
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 1 Test Abstracts l Appendix II - Malfunction Tests Page 70 of 210 Malfunction Test #50, RHR Pump Trip Initial conditions: 100% power. SequenJa ,f '
. ants:
Malfunction 50A is entered and RHR A pump is manually started and verified to trip. Appropriate alarms and status lights are verified to actuate or light. Malfunction 50A is deleted. Appropriate alarms and status lights are verified to clear. After manually arming and depressing the LPCS/RHR Manual initiate pushbutton, proper actuation of the RHR A system is verified. The test is repeated with Malfunction 50 B and C, for pumps RHR B and C, and the RHR B/C Manual Initiate pushbutton. Date of last test: 3/21/90 Concern (s) found/ Planned resolution: None, i t
Grand Gulf Nuclear Station Simulator Certification Initial Report,-March-1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 71 of 210 Malfunction Test #51, LPCS Pump Trip Initial Conditions: 100% power. Sequence of events: Malfunction 51 is entered. When the LPCS pump is manually started, the pump is verified to trip. Appropriate alarms and status lights are verified to actuate or light. When the malfunction is deleted, the "LPCS PMP OVERLD" alarm is verified to clear. The L9CS/RHR A manual initiate pushbutton is armed and depressed. Proper LPCS system actuation is verified to occur, Date of last test: 3/22/90 Concern (s) found/ Planned resolution: None. ' i 4
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix 7X - Malfunction Tests Page 72 of 210 Malfunction Test #52, HPCS Pump Trip Initial Conditions: 100% power. Sequence of events: Malfunction 52 is entered. When the HPCS pump is manually started, the pump is verified to trip. Appropriate alarms and status lights are verified to actuate or light. After deleting the malfunction, the HPCS manual; initiate pushbutton is armed and depressed. Proper HPCS system actuation is verified to occur. Date of last test: 3/22/90 concern (s) found/ Planned- resolution: None. \ i l l 1 i 1
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page-73 of 210 Malfunction Test #53, Spurious HPCS Initiation Initial Conditions: 1 to 2% power at 950 psig pressure. Sequence of events: Malfunction 53 is entered, and the HPCS system is verified to automatically initiate, including Division 3 SSW, and the HPCS Diesel Generator. HPCS injection is verified followed by reactor level - increase, neutron - flux increase due to cold water addition, and reactor scram on high neutron flux. HPCS-injection is verified to terminate when level reacnes Level 8. Malfunction 53 is deleted and the initiation-signal is reset. The HPCS system is returned to standby. The simulator is reinitialized to 100% power and malfunction 53 is entered. The HPCS system, Division 3 SSW, and the HPCS Diesel Generator are verified to auto initiate. HPCS injection is verified to occur. Level is verified to increase'and steady out at near normal. Reactor neutron flux-is verified to increase slightly but below the APRM Plow Biased Scram point.- Malfunction 53 is deleted. The HPCS system is reset, shutdown and returned to standby. RPV level is verified to return to normal. Date of last test: 3/22/90 Concern (s) found/ Planned resolution: None. l
4 Grand Gulf Nuclear 8tation Simulator certification ! Initial Report, March-1991 Simulator Test Abstracts Appendix II - Malfunction Tests - Page 74 of 210 : Malfunction Test #54 (A,B), RHR Steam Reducing Valve Fails open. Initial Conditions: 100% power. Sequence of events: After alignment of the A RHR Heat Exchanger Valves, E12-F052A, _ ' F003A, and F047A, the steam pressure controller E12-R606A is-placed in auto with a 0%_setpoint and activated with its permissive switch. After entering malfunction 54A, the A RHR heat exchanger pressure is verified to increase regardless of controller status. After deleting the malfunction, the A RHR Heat _ Exchanger is depressorized with E12-F047A. The A steam pressure controller is verified to control the A RHR Heat Exchanger pressure normally. The test is repeated using malfunction 52B on the B RHR Heat Exchanger, using B RHR-Heat Exchanger valves and controller. Date of last test: 7/17/90 Concern (s) found/ Planned resolution: None.
-i
}
I l .
.. ,a +,,, . - , - ,r
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 75 of 210 Malfunction Test #55 (A,B), RHR System Steam Reducing Valve Fail Closed. Initial Conditions:-100% power. Sequence of events: After alignment of the A RHR Heat Exchanger valves E12-F052A, F003A, and F047A, steam pressure controller E12-R606A is placed in manual with 100% output. After inserting malfunction SSA and activating the controller with its. permissive switch,
- the A RHR Heat Exchanger pressure is verified to remain at
) zero. The controller is verified to function normally after deleting the malfunction._ The test- is repeated using malfunction SSB on the B RHR Heat Exchanger using B RHR Heat Exchanger valves and controller. l Date of last test: 7/17/90 Concern (s) found/ Planned resolution: None. ii f
,1 s.. - , - - - . .
4 Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction-Tests Page 76 of 210 , Malfunction Test #56 (A,B), RHR Hect Exchanger Throttle Valve fails
^
to throttle. (E12-F003 A/B) Initial Conditions: 100% power. Sequence of events: With malfunction 56A active, E12-F003A is verified to close when attempting to throttle the valve closed, but fail to open after attempting to throttle the valve back open. L inalfunction is deleted and reactivated while E12-F003A is closed. While attempting to throttle the valve open,:E12-F003A is verified to open fully then go to zero by position indication. E12-F048A is closed and positive flow indication is observed after starting RHR A in the Suppression Pool Cooling mode and confirming that E12-F048A is open. After deleting malfunction 56A, E12-F003A is verified full open. E12-F003A is then verified to throttle as desired. The test is repeated using malfunction 56B and E12-F003B. Date of last test: 3/22/90 Concern (s) found/ Planned resolution: None. 1
Grand Gulf Nuclear Station Simulator Certification . Initial Report, March 1991
-Simulator Test Abstracts Appendix II - Malfunction Tests Page 77 of 210 ;
Malfunction Test #57 A,B, RHR Heat Exchanger Tube Leak i Initial Conditions: RPV pressure at 100 psig. Sequence of events: RHR is placed in shutdown ~ cooling mode. SSW A is manually } initiated supplying SSW to the RHR A Heat Exchanger Malfunction 57A is inserted. RPV Level is verified to decrease. RHR and SSW Heat exchanger flow is verified to increase. SSW and RHR temperature.are verified to increase. SSW Loop A radiation monitors are verified to' alarm. After shutting the RHR Heat exchanger inlet and outlet valves, RPV level is verified to increase. RHR and-SSW heat exchanger ' flows are verified to decrease. RHR and SSW heat exchanger temperatures are verified to decrease. The SSW Loop A radiation alarm is verified to clear. After clearing malfunction 57A, the test is performed in:similar fashion on the RHR B heat exchanger using malfunction 57B. i Date of last test: 4/27/90 Concern (s) found/Plannned resolution: SSW fill tank Hi/ Low alarm came in' with P41-Fil3- closed. RPV level did not decrease with the RHR minimum flow valves open. The SSW ' Loop A/B High radiation alarms did not come in. -These l deficiencies will be corrected on a schedule to be' complete within 4 years of initial certification. L i I Y i
~ _ _ _ ~ . . __ _ _ _ _ _ . _ . - . _ . . . _ . ._. , . _ _ .
Grand Gulf Nuclear Station-Simulator Certification Initial Report, March 1991 simulator Test-Abstracts Appendix IX - Malfunction Tests - Page-78 of 210 i.
-t Malfunction Test #58, RHR Heat Exchanger Level Controller Fails Downscale.-
Initial Conditions: N/A Sequence of events: Since the steam condensing mode of RHR is inoperable and-its operation is-prnhibited-by licensing commitment, the-test of this malfunction is not applicable. Date of last test: -N/A. Concern (s) found/ Planned resolution: In the event licensing conditions are changed to allow RHR steam condensing mode operation,-this malfunction will be' tested. i
GrADd Gulf Nuclear Station-
%Leilator Certification Initial Report, March 1991 Simulator Test Abstrccts Appendix II - Malfunction Tests Page 79 of 210 Malfunction Test #59, Main Stt:am Relief Valve Fails Open Initial Conditions: 100% power.
Sequence of events: Malfunction 59 is activated for each of the 20 SRVs one at a time. Each SRV is verified to have failed open by attempting to close it. Total Steam flow, generator megawatt output, turbine control valve positions, and feedwater temperature are verified to decrease. Alarm response is verified. Suppression pool temperature and level are verified to increase. The suppression pool temperature indicator nearest the open SRV is verified to increase. An approximate 1 MLB/HR Steam flow / feed flow mismatch is verified. The Malfunction is deleted. SRV closed indication is verified and applicable alarms verified to clear. SRV tailpipe temperature is verified to decrease slowly. For the six Low Low set SRVs, Remote shutdown panel H13-P150 red indicators are verified to respond. With the malfunction active, the SRV it closed using the handswitch at the remote shutdown panel and SRV closure indicators verified. t Date of last test: 3/27/90 Concern (s) found/ Planned resolution: SRV Tailpipe temperature decreases too rapidly following closure of SRVs. This deficiency will be corrected on a schedule to be completed within 4 years of initial certification. Backpanel H13-P631 is not included in the simulator. This panel is to be incorporated into the upgrade simulator project. Suppression pool temperature nearest the SRV tailpipe did not increase the most. This deficiency has been corrected and retested.
i I Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 80 of 210 Malfunction Test #60, Main Steam Relief Valve Leaks . Initial Conditions: 100% power. Sequence of events: Malfunction 60 is entered for each SRV. Alarm response = is verified. SRV tailpipe and suppression pool temperature are verified to increase. After deleting the malfunction, alarms are verified to clear. SRV tail pipe temperatures are verified to decrease. Suppression pool temperature is verified to stabilize. Date of last test: 3/27/90 Concern (s) found/ Planned resolution: Suppression pool temperature nearest the SRV tailpipe did not increase as expected. This deficiency has been corrected and retested. SRV tailpipe temperature decreased too quickly following SRV closure. This deficiency will be corrected on a schedule to be completed within 4 years of initial certification. i J
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991-Simulator Test Abstracts Appendix II - Malfunction Tests Page 81 of 210-Malfunction Test #61, Main Steam Relief Valve Fails to Reseat Initial Conditions: 100%. power. Sequence of events: With malfunction 61 active, each SRV is manually opened and reclosed. Alarm response is verified. Indications that the SRV is not reseated-are verified. This includes reduced steam flow rate, generator megawatt output, SRV tail pipe temperature, turbine control valve position, suppression pool temperature and level, and SRV tail pipe pressure switch indication. Malfunction 61 is-deleted and parameters are verified to return to normal.-Alarms are verified to clear,. The test is repeated for all 20 SRVs. Date of last test: 3/29/90 Concen(s) found/ Planned resolution: None.
Grand Gulf Nuclear Station Simulator certification 1 Initial Report, March-1991 l simulator Test-Abstracts Appendix II - Malfunction Tests Page 82 of 210 Malfunction Test #62, Instrument Line Break Initial Conditions: 100% power. Sequence of events: Malfunction 62 is entered. CRD drive water, cooling water and core plate differential pressure instruments are verified to respond. Alarm response is verified. Drywell and containment parameters are verified to respond to the liquid line leak. The malfunction is allowed to run for 15 minutes without resetting any annunciators. A LOCA-checklist-is used to verify correct automatic actions and annunciation. Date of last test: 6/25/90 Concern (s) found/ Planned resolution: The following annunciators did not alarm as expected:
"SSW PMP DISCH PRESS LO" "PCW PMP A TRIP" "PCW PMP B TRIP" l "SSW PMP B TRIP"'
"SSW CLG TWR FAN C TRIP" "SSW CLG TWR FAN D TRIP" ,
These deficiencies-have been corrected and retested. L . l I l _ . _ . _ _
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts
- Appondix II - Malfunction Tests Page 83 of 210 Malfunction Test #63, Recirculation Loop A Rupture (Variable)
Initial Conditions: 100% power. Sequence of events: Malfunction 63 is entered at 8%, 20%, 50% and 100% severity for each test. In each case the transient is allowed to run for 15 minutes without resettirg any annunciators. A LOCA. checklist is used to verify correct automatic actions and annunciators for each severity level. This checklist specifies alarms, status lights, component status indications, and system and component conditions to aid in this evaluation. Date of last test: 6/26/90 Concern (s) found/Plsnned resolution: The following annunciators did not alarm as expected:
"SSW PMP A DISCH PRESS.LO" "PCW PMP A TRIP" "PCW PMP B TRIP" "SSW-PMP B TRIP"-
"SSW CLG TWR FAN C TRIP" "SSW CLG TWR' FAN D TRIP" These deficiencies have been corrected and retested.
t Unit 2 instrument air compressor did not -trip as expected. This item has been corrected and retested.
-Annunciators."CTMT CLG EXH RAD HI",-CTMT CLG'EXH DIV 1, 4 RAD HI-HI/INOP", "CTMT CLG EXH DIV 2, 3 RAD HI-HI/INOP" did not alarm as expected. These deficiencies will be corrected on a schedule to be completed within 4 years of initial certification.
I 4--.. _ ... ,4,---- ----y , v.
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix IX - Malfunction Tests e Page 84 of 210 , t i Malfunction Test #64, Steam Leak-in the Drywell Initial Conditions: 100% power with "A" Reactor Level instrument ' selected-for feedwater control. Sequence of events: Malfunction 64 is inserted. Feedwater control response is verified to result in decreased feed water flow. The ' instruments which share this reactor vessel instrument tap are verified to fail with level instruments failing upscale and pressure instruments failing downscale. Alarm response and status light responses are verified. RPV level is verified to - decrease on other unaf fected . instruments. The "B" Reactor level instrument is selected for feedwater control when level falls to 20 inches. RPV level is verified to stabilize. The malfunction is allowed to run for 15 minutes without resetting any annunciators. A LOCA checklist is used to verify correct automatic actions and annunciators. This checklist specifies alarms, status lights,-component status indications, and system and component conditions to-aid in this evaluation. ; Date of last test: 6/26/90 Concern (s) found/ Planned resolution: Unit 2 instrument air compressor did not trip as expected: '
"SSW PMP A DISCH PRESS ~LO" '
"PCW PMP A TRIP" "PCW PMP B TRIP"'
"SSW PMP B TRIP" '
"SSW CLG TWR FAN C.' TRIP" "SSW CLG TWR-FAN D TRIP" These deficiencies have been corrected and retested.
1 l
._ _ ,, .a., . -
.----~ - - - -
Grand Gulf Nuclear 8tation 81stOator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 85 of 210 Malfunction Test #65, Steam Line Rupture in the Drywell (Variable) Initial Conditions: 100% power. Sequence of events: Malfunction 65 is entered at 5%, 20%, 50% and 100% severity. for each test. For the 5% test,-main. steam-line B flow is verified to decrease below normal. In each case the transient-is allowed to run -15 minutes without- resetting any annunciators. A LOCA~ checklist is used to verify correct automatic actions and annunciators for each-severity level. This checklist specifies alarms, status lights, component' status indications, and system and component conditions to aid in this evaluation. Date of last test: 6/26/90 concern (s) found/ Planned resolution: The following annunciators did not alarm as expected.
"SSW PMP A DISCH PRESS LO" "PCW PMP A TRIP" "PCU PMP B TRIP" l "SSW PMP B TRIP" l "SSW CLG TWR FAN C TRIP" L "SSW CLG-TWR FAN D TRIP" l These deficiencies have been corrected.
1. Unit 2. instrument air compressor- did not trip as expected. This item has been corrected.
~
Annunciators "CTMT CLG EXH RAD'HI", CTMT CLG EXH DIV 1, 4 RAD HI-HI/INOP", CTMT CLG EXH DIV 2, 3. RAD-HI-HI/INOP" did not alarm as expected. These deficiencies will be corrected ~ on a schedule to be completed within 4 years of initial certification.
- . . . . _ . . ~ - . , _ ~_ . . _ _ _ _ . . - _ . - . . . . . , _.a
Grand Gulf Nuclear Station simulator Certification. Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests- ' Page 86 of 210 Malfunction Test #66, Steam Leak in Tunnel Initial Conditions: 100% power. Sequence of events: Malfunction 66 is entered. Steam tunnel temperatures-are verified to increase using backpanel~H13-P642 indications. Alarm response is verified._A group I isolation (MSIVs and drain valves) is verified to occur with correct automatic actions and alarms including SRV actuation, Low-Low set , actuation, and reactor scram. Date of last test: 3/26/90 Concern (s) found/ Planned resolution:. Leak detection switches and indication on panel H13-P632, Division I Leak Detection System are not-included in-the simulator. This panel will not be ir.:orporated as previously l discussed in the exceptions section of trLs-report. l t I
Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix IX - Malfunction Testa Page 87 of'210 Malfunction Test #67, Steam Line Rupture in Tunnel (Variable): Initial Conditions: 100% power. Sequence of events: Malfunction 67 is entered at 5%, 50% and-100% severity in ' separate sections of the test. Steam tunnel temperature are verified to increase'using backpanel H13-P642 indications . Alarm response is verified.LA group I-isolation is verified to occur with correct automatic actions and alarms-including SRV actuation, lou-low set actuations and-reactor scram.. Date of last test: 3/26/90 Concern (s) found/ Planned resolution: None. e .,-,~e, , e, e- " - ,- m w-r
[ Grand Gulf Nuclear Station
- Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II _ Malfunction Tests Page 88 of 210 i
i i Malfunction Test #68, Steam Rupture in the Turbine Building ! (Variable) , Initial Conditions: 100% power. Sequence of events: Malfunction 68 is entered at 5% severity. Turbine Building area and ventilation radiation monitor alarms are verified - along with increasing radiation levels in the turbine building. The severity is increased to 20% severity. Main steam flows are verified to increase and_the APRM Upscale alarm is verified to annunciate.- . The simulator is reset to 100% power and malfunction 68 is-entered at 50% severity. Alarm response'and automatic actions are verified to occur. A group I isolation on low main steam line pressure is verified to occur along with SRV actuation, , low-low set actuation, and reactor scram. The simulator is reset to 100% power and malfunction 68 is entered at'100% severity. A group I isolation on high main steam line flow is verified to occur along with other automatic actions and alarms, i Date of last test: 3/26/90 . Concern (s) found/ Planned resolution: None. i
?
k
Grand. Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 89 of 210 l I l Malfunction Test #69, Instrument Line Leak Outside Containment Initial conditions: 100% power. l Sequence of ovents: Malfunction 69 is entered. Feedwater-flow on the "A" indicator is verified to fail to zero with total feed flow decreasing by 50%. Turbine building floor drain sump alarms are verified to occur. Feedwater control is verified to respond with an increase to maximum output followed by an' increase jn level and power, a. reactor- scram on high neutron flux and finally a return of level to the normal, range. Alarm response is verified. Date of last test: 3/27/90 Concern (s) found/ Planned resolution: None.
Grand Gulf Nuclear Station-Simulator certification Initial Report, March 1991 simulator Test Abstracts , Appendix IX ~ Malfunction Tests Page 90 of 210 Malfunction Test #70, Feedwater Line Rupture Outside Containment. Initial Conditions: 100% power. Sequence of events: Malfunction 70 is entered. Feedwater flow on the B ii.dicator is verified to decrease significantly'followed by a decrease to zero on both the A and B indicators. Turbine Building sump alarms are verified to occur. Feedwater control is verified to respond to maximum output with no feed flow indication based upon. closed feedwater c'c.ack valves. Automatic actions and alaria are verified. An - automatic transfer of Recirculation pumps to slow speed and Reactor Scram are verified to occur at Level 3. HPCS and RCIC initiation are verified to occur at Level-2. The transient is allowed to run for 15 minutes without . resetting any annunciators. A LOCA checklist is used to verify correct automatic actions and annunciators. This checklist specifies alarms, status lights, component status indications, and system and component conditions to aid in this evaluation. I Date of last test: 6/26/90 concern (s) found/ Planned resolution: Unit 2 instrument air compressor did not trip as expected. This has been corrected and retested. The following annunciators did not alarm as expected:
"SSW PMP A DISCH PRESS LOW" "PCW PMP A TRIP 25 "PCW PMP B TRIP" "SSW PMP B TRIP" "SSW CLG TWR FAN C TRIP" "SSW CLG TWR FAN D TRIP" These deficiencies have been corrected and retested, i
m- g y**my c-- - ~ w ,o-~w *yw,,,-or--
Grand Gulf Nuclear Station Simulator Certification-Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests-. Page 91 of 210 f
. Malfunction Test #71, Fuel Cladding Leak (Variable)
Initial Conditions: 100% power. Sequence of events: Malfunction 71 is entered at 10%, 50%, and 100% severity. For each severity level, offgas pretreatment, offgas post t r e a t t.e n t , main steam line radiation, and area radiation monitor levels are verified to increase. For 50% severity, offgas is verified to isolate and loss of vacuum is verified to result in turbine trip, reactor scram,- EOC RPT trip actuation, and Group.I isolation.- For 100% severity, high main steam line radiation is verified to cause a trip with the resulting reactor scram and and group I isolation. Both loops of suppression pool cooling are placed in service 'and area radiation-levels in the containment, drywell, and auxiliary buildings are -verified to increase. Date of last test: 3/26/90 Concern (s) found/ Planned resolution: ARM response in the Drywell, containment, and auxiliary buildings did not increase as. expected. There was veryllittle or no time delay between Offgas pretreatment and post - treatment radiation monitor response. These deficiencies will be corrected by the simulator-upgrade project which will incorporate an offgas model and radiation transport during i fuel-leaks, on a schedule to be completed within 4 years of initial certification. l I !
Grand' Gulf Nuclear Station simulator Certification-Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction-Tects Page 92 of 210 l 1 I i Malfunction Test #72, Main Steam Line High Radiation. Initial Conditions: 100% power. Sequence of events: Malfunction 72 is entered. A new condensate /demineralizer is placed in service using instructor remote functions. Automatic actions and alarms are verified which include main steam line high radiation indications, reactor scram, SRV and low low set actuation, group I. isolation, and group 10 isolation valves (B33-F019 and F020). . The division 1 and 2 post accident sampling valve isolation is manually overridden-and B33-F019 and F020 are opened. When the manual override is reset, B33-F019 and F020 are verified to reisolate. Date of last test:- 7/12/90 Concern (s) found/ Planned resolution: I The division 1 and 2 post accident sampling valve isolution
. manual override handswitches are reversed in divisional function. Annunciators " PROC SMPL VLVS DIV I MAN OVERD" and l
" PROC SMPL VLVS-DIV'2 MAN OVERRD" did not alarm and reset as expected. These defici^7cies will be corrected on a schedule to be . completed wit' ( years of initial cartification.
l i l l i
- , - , .i,..
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 93 of 210 Malfunction Test #73, MSIV' Fails Shut Initial Conditions: 75% power. Sequence of events: Malfunction 73 is inserted to fail each MSIV individually. Valve closure time,. reactor pressure and neutron flux response, niain steam line flow and alarm response- are verified. The test is repeated for all 8 MSIVs. Date of last test: 3/26/90 Concern (s) found/ Planned resolution: None.
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 94 of 210 Malfunction Test #74, spurious Scram Initial Conditions: 100% power. Sequence of events: Malfunction 74 is entered. Automatic actions and alarm response are verified to occur. These include control rod insertion, scram solenoid status lights, power drop,, control rod hydraulic scram discharge volume vent and drain isolation, recirculation pump automatic transfer to slow speed, generator reverse power trip, bypass valve pressure control, and reactor level- response. Malfunction 74 is deleted and _ the reactor mode switch is placed to-shutdown, the.CRD Discharge Volume High level trip bypass switches are placed'to bypass. The scram is
~
reset. Control rod dri f t is reset. CRD accumulator - fault indication is acknowledged. APRM/IRM recorders'are selected to IRM mode. All IRM and SRM detectors are inserted. Automatic-actions and alarms are. verified to - occur as a result 'of ' these ' actions which are normally performed following any scram-condition. Date of last test: 3/26/90 Concern (s) found/ Planned resolution: None.- L
>M- S " na- M Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991-Simulator Test Abstracts Appendix IX - Malfunction Tests Page.95 of 210 Malfunction Test #75, Failure to Scram Initial conditions: 100% power.
Sequence of events: Malfunction 75 is entered. The main turbine is manually tripped. Automatic and alarm actions -are verified to occur including scram trip indication, turbine trip, neutron monitoring system trip, SRV actuation, ATWS-ARI/ RET trip actuation, and power decrease and stabilization. Scram valves are verified to indicate open on panel H13-P680.-When malfunction is deleted, full control rod insertion is verified. Date of last test: 7/12/';0 Concern (s) found/ Planned resolution: None, i {
, , + + e
. . . ..- _ = - - - . . . - - -
1 Grand Gulf Nuclear Station Simulator. Certification-Initial Report, March 1991 Simulator' Test Abstracts Appendix IX - Malfunction Tests
=Page 96 ofL210 Melfunctico Test #76, Failure to Scram (Manual Scram Function Operable)
Initial Conditions: 100% power Sequence of events: Malfunction 76 is entered. Malfunction 9A and 9B are entered to fail APRM channels A and B upscale. Neutron monitoring trip indication is verified without RPS. actuation. A half scram condition is inserted and verified by manually _ arming and depressing the-Division 1 and 3 manual scram pushbuttons. The half scram conditions are-reset. . A half scram condition is inserted and verified by manually arming and depressing ~the Division 2 and 4 manual scram push buttons. The half scram condition is reset. The reactor mode switch is placed to shutdown and a full scram condition is verified. The simulator is reset to 100% and malfunction 76 is entered.- A half scram condition is verified when CB2B and CB8B are opened using remote functions. A full scram is verified to occur when. additionally CB2B and CB8B are opened using remote functions. The simulator is reset to 100% with malfunction 76 active. With malfunctions 9A and 9B entered as before, a scram is verified to occur when malfunction 76 is deleted. The simulator is reset to 100% and malfunction 76 is entered. The main turbine is tripped with malfunction 93. ARI/RPT'is verified to cause control rod insertion. Date of last test: 3/27/90' Concern (s) found/ Planned resolution: None.
l Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 , Simulator Test Abstracts Appendix IX - Malfunction Tests Page 97 of 210 Malfunction-Test #77, A/$ RPS MG Set Failtwe Initial Conditions: 100% power. Sequence of events: ( Malfunction 77A is entered. Loss of RPS A indications are verified on . panel H13-P680. RPS A power is placed on ! alternate power using'remoteLfunctions. The half scram j condition is reset. Malfunction 77A'is deleted. RPS MG set A is restarted and RPS A power is placed back on normal power using remote functions.
~
The test is repeated for malfunction 77B and RPS MG set B', RPS B power,-and associated remote functions. Late of last test: 3/28/90 Concern (s) found/ Planned resolution: , RPS power indications and controls backpanel on.H13-P610 are 1 not currentl'/ included in the. simulator but are - simulated. using remote, functions.'The RPS power porcion of this panel will . be lucorporated with the' simulator upgrade = project, on a schev ale to . be completed within 4 years of initial certification. I
? Grand Gulf Nuclear station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix IX - Msifunction Tests Page 98 of 210 Malfunction Test #78, Pressure Regulator Fails Low / Transfers to Backup Initial conditions: 100% power. Sequence of events: Malfunction 78 is entered. Lack of effectu are verified on turbine steam flow, control valve position and generator output. The Turbine IPC Cabinet railure alarm and pressure controller tsult status light ura verified. When malfunction 78 is deletod., the above alarm and-status light are verified to clear. Date of last test: 3/28/90 concern (s) found/ Planned resolution: None. l l l
- .. .. -. .. . -- ._ - - . . . - . . _ . .- -. .- -. ._.~.- . -- . - .
Grand Gulf Nuclear station > simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 99 of 210 Malfunction Test #79, Pressure Regulator Fails Low Initial Conditions: 100% power. Sequence of events: Malfunction 79 is entered. Turbine Steam flow reduction, pressure control output reduction, turbine control valve closure, turbine and generator trip, pressure response, SRV actuation, ATWS-ARI/RPT, EOC-RPT actuation, reactor scram, and status light and alarm response are verified. Date of last test: 3/28/90 Concern (s) found/ Planned resolution: The following alarms did not actuate:
"RPT A Trip" "RPT B Trip" These deficiencies will be corrected within 4 years of initial certification.
l i
+m.. ... , . -..,e .= , , . . . . , - - , .,-gw-w. ~ve - v -w +-v -t* vt --
r**~-s--*-- T 'T* y 'r + * * ' = - vm- ' - +v f
i . Grand Gulf Wuolear station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix 2x - Malfunction Tests Page 100 of 210 4 Malfunction Test #80, -'ressure Regulator Fails High I Initial Conditions: 100% power. Soquence of events: Malfunction 80 is inserted. Pressure control output is verified to increase to maximum output. IPC alarwa and status lights are verified. Turbine conttv2 Valves are verified to open fully. Bypass valves are verified to open to approximately 40% to 50% in response to the increase in controller output. Generator load is verified to increase. Reactor pressure is verified to decrease .until a Group I isolation on low main steam line pressure occurs. Automatic and alarm response are verified including SRV ' actuation, reactor scram, and ARI/RPT initiation. Bypass - valves are verified to fully open following MSIV isolation. Turbine control valves are verified to close following load reference automatic switch off. Generator and Turbine trip are verified to occur. Date of last test: 3/28/90 concern (s) found/ Planned resolutions. None. i
-l %+ 7ye-. - - - -p-py- g.- ++ --omg.--w-w y y -wg~"wr-.- g y>M -rw- y e 't'g w --
1 l l e 1 4
' Grand Gulf Nuclear station simulator certification !
Initial Report, March 1991 simulator Test Abstracts . Appendix IX - Malfunction Tests i Page 101 of 220 ; ~ Malfunction Test #81, Pressure Regulator Occillation Initial conditions: 100% power. Sequence of events: Malfunction 81 is entered. Generator load, turbine steam pressure, control valve and bypass valves, reactor pressure, steam flow and reactor flux oscillations or perturbations are verified to occur. When malfunction 81 is deleted, -all parameters are verified to stabilize at normal values. Date of last test 3/28/90 Concern (s) found/ Planned resolution: None. I i
. - _ . - , . . . . . _ . . . . . . . . _ . . - . _ . . _-_...__,,.._.s,. . . . __ . , _ . _ _ _ _ , - - - - - . . . - , , - .
- 1 Grand Gulf Nuclear Station simulator certification Initial Report, March 1991
, simulator Test Abstracts Appendix II - Malfunction Tests . Page 102 of 210 ) l Malfunction Test #82, A,B,C Turbine Bypass-Control Valve Stuck Initial conditions: 100% power. Sequence of events: Malfunction 82A, B, and C, are entered to fail bypass control valves closed. Malfunction 93 is entered to trip the main turbine and create a bypass opening demand. Bypass valves are verified not to open. Bypass-control status lights a7 e verified to respond. Reactor pressure and SRV response are vorified. When all malfunctions are deleted, bypass valves are verified to control reactor pressure. Date of last test: 3/28/90 Concern (s) found/ Planned resolutioni None. [ E
, ,-,,,e- - , r.>..,,- , , , - - - -,-,,..,~w* *~w- s-v w ~ ,w x = ,s ~v~ =--**~',-v * ** '
. . . . . - . -. -.--. _ _ _.-- - . . . ~.-.. .. - ..-._ _ _- - . . _ . - . - _ - . -
l I Grand Gulf Nuclear Station i simulator certification l Initial Report, March 1991 l Simulator Test Abstracts Appendiz II - Malfunction Tests Page 103 of 210 i Malfunction Test # 83, A, B, C, Bypass Control Valve Fails Open Initial Conditions: 100% power. Sequence of events: Malfunctions 82 A, B, and C are entered and deleted for each bypass valve. While each is active, automatic - actions and alarm response are *.ri fied . This includes turbine control valve, byi ass valve, generator output, steam flow,, reactor power and byaass control status light response. After each malfunction .s cleared, the above parameters and status lights are verified to return to normal. The. test is repeated-for each bypass valve. Date of last test: 3/28/90 Concern (s) found/ Planned resolution: Bypass control valve status lights did not correctly update after malfunction 82A, B, or C_was deleted. This deficiency will be corrected within 4' years of initial certification. ll t b
-,----n, , , , . . . - , - . . , , . , . , , - , , . , , -..
~ . - - , , . , , , , , , . - , , ~ , , . - , - , - . . . . ~ - - - . , . , , - ~ . .
I Grand Gulf Nuolear Station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 104 of 210 Malfunction Test #84, Turbine Bypass Stop Valve Fails Shut Initial conditions: 100% power. Sequence of events: Malfunction 84 is entered for each turbine bypass stop valve in conjunction with a turbine trip and malfunction 82 for the other 2 bypass control valves. The Bypass stop valve is-verified to close for the affected valve. Bypass control valve response is verified for the remaining valves. Turbine pressure is verified to increase rapidly. When malfunction 84 is deleted while the other malfunctions are active, the applicable bypass stop and control valve is verified to open and the remaining bypass valve reduces pressure. The-test is rotated and repeated for all bypass stop valves. Date of last test: 3/29/90 Concern (s) found/ Planned resolution: None. 4
Grand Gulf Nuclekt station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests 'I Page 105 of 210 e i Malfunction-Test #85, Turbine Governor Fails High Initial Conditions: 10% power with the main turbine on the turning gear. Sequence of events: Af ter raising. turbine speed demand to 400 RPM, malfunction 85 is entered. Turbine control valve, status lights.and alarms are verified to respond which results in a turbine speed' increase until turbine MHC control limits speed. When-the mal- 7 function is deleted, turbine speed as verified to return.to ; the previous speed demand of 400 RPM.-The simulator is reset to approximately 10 to 20 MWe. Malfunction 85 is_ entered. Turbine control is verified-to respond correctly including a load increase until bypass valves are closed. When the malfunction is deleted, turbine load is verified to return to 10 to 20 MWe and annunciator and computer points reset. The simulator is reset to 100% and malfunction 85 is entered. ; Effects on plhnt parameters are verified not to occur. Alarm, computer points, and turbine control are verified to respond when load limit is reduced to-zero, load demand is reduced to ' ! zero, and speed demand is reduced to zero.-Generator. load is L verified not to change while the malfunction'is present. Date of last test: 3/29/90 f Concern (s) found/ Planned resolution: , The turbine accelerated to 1875 RPM instead of 1926-RPM when limited by MHC control. This deficiency will be corrected within 4 years of initial certification. o
-y i
9 P
?
.-.,.,__-.y .m,._y. , , , . . ,.,7,_m, y_.._,. , , , , - , , . , . , , _,,,,,,,.,.,,m...,m_..,,m.-m,-_,.,,-,
l Grand Gulf Nuclear Station Dinulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 106 of 210 l Malfunction Test #86, Turbine Governor Fails' Low Initial Conditions 35% power Malfunction 06 is entered. Automatic r6sponses are verified ircluding the following: Turbine control valve closuro Bypass control valve opening Generator load decrease Turbine steam flow decrease : Load reference auto switch off- ! Generator trip on reverse power , Alarm response l Absence of reactor scram The simulator is reinitialized to 35% power. Power is raised to 50% power. Malfunction 86 is entered. Automatic responses are ! verified including the following: Reactor pressure increaso ! APRM flux increase Scram Pressure control with bypass valves Alarm response Date of last test:- 3/28/90 Concern (s) found/ Planned resolution: None. l
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 107 of 210 Malfunction Test #87, Turbine Acceleration Control Failure Initall conditions: Same as for malfunction test 85 Sequence of events: The same sequence of events for malfunction 85 are repeated for malfunction 87, since the automatic actions are the same. Date of last test: 3/28/90 concern (s) found/ Planned resolution: None. 1
,- .- ,c - -
,----g,y,g n--9y .i w - 9 . , ,--y e-, y--y,m---
Grand Gulf Nuclear station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Me1 function Tests Page 108 of 210 Malfunction Test #88 A,B,C,D, Turbine Stop Valve Fails Shut Initial Conditions: 35% power, turbine generator _on the line. Sequence of events: Recirculation pumps are shifted to fast speed. Power-is increased to 50%. Malfunction 88 A,B,C, and D are entered singly and in combination to test automatic response and actions including: Reactor half scram and full scrsr. for correct Turbine Stop Valve closure combinations. EOC RPT actuations for correct Turbine Stop Valve closure combinations. Turbine pressure and control valve response. Generator output response. Valve Status light and alarm response. Automatic transfer of Recirculation pumps to slow speed on EOC RPT actuations. Date of last test: 3/28/90 Concern (s) found/ Planned resolution: EOC RPT actuations did not open RPT breakers -as expected during divisional testing. This deficiency will be corrected on a schedule to be completed within 4 years of initial certificat'on.
Grand Gulf Nuclear station simulator Certification , Initial Report, March 1991
' simulator Test Abstracts Appendix II - Malfunction Tests Page 109 of 210 Malfunction Test #89, Turbine Control Valve Servo Failure Initial Conditions 50% power Sequence of events:
Malfunction 89 is entered for ecch turbine control valve. Severity is varied from failure as is initially to 20%, and then to 100%. As severity is varied, automatic response and actions are verified including: Turbine pressure controller and control valve response. The simulator is reinitialized to 50%. Malfunction 89 is entered for all turbine control valves to fail as is. Power is increased to 754. Bypass control vulves are verified to open to control pressure. Date of last test 3/28/90 Concern (s) found/ Planned resolution: None. i
- a. - . - .- -.--- -.-. . - - . . . . - , - . . - . . . --
. _ - - - . . - . . - . - - . _ _ _ - . _. . .. - ~ . . . - . - . - . .-. -.. . - .
3 Grand Gulf Nuclear station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 110 of 210 Malfunction Test #90 A,B, Moisture Separation Reheater Losss of First Stage Reheat. Steam Initial Conditions: Sequence of events: Malfunction 90A is entered. Automatic actions and alarm response are verified including: Generator load reduction Extraction steam MOV and BTV closure MSR First Stage Reheater Drain Tank Hi Hi Level Alarm. Malfunction 90A is deleted and extraction steam valves are reopened. Generator load-is verified to increase. Alarms are verified-to clear. The test is repeated for the-B MSR using. Malfunction 90B. Date of last test: 3/29/90 Concern (s) found/ Planned resolution: None. l b
;e,
-. . , _ . , ..g ,....r... ....r -
._m. ., ,w....m., . , , , _ - . -7. . _ _ . . . . . . . . , . , , , _ _ , ., ,,,,m.-., ., ,,. ., ,,,c,, s.__.
- - - . _ _ . = . - . . _ _ _ _ _ . __ - - - _-_ . - - _ .- _ . _ _ . f Grand Gulf Nuclear Station ! simulator certification i Initial Report, March 1991 ' simulator Test Abstracts ; Appendix II - Malfunction Tests , Page ill of 210 i Malfunction Test #91 A,B, Moisture Separator Reheater Temperature , Control Failure Initial conditions: 20% p'ower. i Sequence of events: Malfunction 91A is entered. Automatic actions and alarm response are verified including:
"A" MSR Second Stage tube sheet pressure and temperature l response. l "A" MSR Heating Steam controller output failure to 100%
and cabinet failure alarm. ! Hi level in #6A HP FW Heater i opening of affected heating steam valves. The A MSR heating steam controller is placed in manual and , output reduced to stabilize tube sheet pressure and temperature. , Malfunction 91A is removed 'and the A MSR controller is , returned to normal automatic mode. The test repeated for the "B" MSR Heating Steam controller
- using malfunction 91B. Similar effects.on the "B" MSR are verified.
Malfunction 91A and B are entered simultaneously. - Extraction MOVs N11-F028A and B are verified open by computer point status. Date of last test: 3/29/90 Concern (s) found/ Planned resolution:- None._-
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests 1 Page 112 of 210 l I Malfunction Test #92 A,B,C and D, Moisture Separator Reheater, Loss of Second Stage Reheat Steam Initial conditions: 100% power. Sequence of events: Malfunction 92 is entered to close each second stage reheat MOV. Automatic actions are verified including: Second Stage Extraction Steam MOV closure. MSR second stage pressure drop for the affected MSR. Turbine pressure control and control valve response drop in generator output. The Malfunction is deleted and the affected MSR heating steam MOV is reopened. Turbine control valve and generator output are verified to respond toward their initial values. The test is repeated for each second stage reheat MOV. Date of last test: 3/29/90 Concern (s) found/ Planned resolution: None. s y -~,
,er-, .. -
-eng-v-- , -
4 , n
v
~
Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 113 of 210 Malfunction Test #93, Main Turbine Trip Initial Conditions: 100% power. Sequence of events: Malfunction 93 is entered. Automatic actions and alarm response are verified including: Turbine stop and control valve closure (high and low low pressure valves) Bypass control valve opening Reactor Scram and power drop EOC RPT actuation SRV actuation Low Low Set actuation Generator trip reverse power Extraction steam valve isolation and drain valve opening Turbine bearing oil pressure decrease on turbine coast down. Auxiliary lube oil pump auto start Shaft lift oil pump auto start Turning gear engagement on turbine coast down Auxiliary primary water pump auto start Turbine gear valve-FE01 speed interlocks are verified at 210 and 100 RPM i 1
Grand Gulf Nuclear station simulator Certification , Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 114 of 210 The simulator is reinitialized to 25% power, with the turbine at rated speed. The generator is synchronized to the grid and loaded until bypass valves are closed. Annunciator "Turb CV/SV Close Trip BYP" is verified to seal in. Malfunction 93 .ls entered. Turbine Stop and control valves are verified closed. Bypass valves are verified to open. The reactor is-verified to not scram. Date of last test 3/30/90 Concern (s) found/ Planned resolution: None. l l 1 i
Grand Gulf Nuclesr Station simulator certification Initial Raport, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 115 of 210 Malfunction Test #94, Turbine Gland Seal Regulator Failure Initial conditions: 100% power. Sequence of events: Malfunction 94 is entered. Automatic actions and alarm response are verified including: Zero output from the Seal Steam controller and decreased seal steam header pressure Seal Steam controller cabinet failure . alarm Turbino seal steam low pressure alarm Main Turbine Seal Steam Controller fault alarm closure of seal steam control valves Decrease in turbine vacuum / increase in condenser shell pressures. The seal steam control bypass valve-is reopened. Seal steam pressure and condenser vacuum are restored to normal , values. The seal steam control bypass valve is closed and seal steam 1 pressure is restored using manual control of the seal steam controller. Malfunction 94 is deleted. The seal steam controller is placed in auto and verified to f 4nction normally. Alarms are verified to clear. 1
Grand Gulf Nuclear station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 116 of 210 Date of last test! 7/12/90 Concern (s) found/ Planned resolution Condenser vacuum was not appreciably affected by the loss ~of seal steam. This deficiency will be corrected within 4 years of initial cortification. l l l l [ i t
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix IX - Malfunction Tests Page 117 of 210 Malfunction Test #95, Loss of Turbine HP Control Oil Pressure Initial Conditions: 100% power. Sequence of events: i Malfunction 95 is entered. Automatic actions and alarm response are verified including the following: Control fluid pressure decrease on the HP fluid side. Control fluid pressure remaining constant on the LP fluid side. Auto start of the standby control Fluid pump. HP control valves drifting cloned. Bypass stop and control valve closure, i Reactor pressure increase. APRM flux increase. HP Stop and control valve closure. LP Stop and control valve closure. Control Fluid Tank low level alarm. Reactor Scram; Generator trip. l ATWS- ARI/RPT Actuation. , EOC-RPT Actuation. 1 ( Alarms associated with above actions. Date of last test: 7/17/90 i Concern (s) found/ Planned resolution: _ l The low pressure output of the controlfluid pump did not remain constant and went to zero. This = deficiency will-- be corrected within 4 years of initial certification.
Grand Gulf Nuclear Station ' simulator certification Initial Report, March 1991 , Simulator Test Abstracts Appendix IX - Malfunction Tests i Page 118 of 210 Halfuriction Test #96, Turbine Thrust Bearing Wear Trip Initial Conditions: 100% power. ; Sequence of events: i Malfunction 96 is entered. The main turbine is verified to trip. Vibration alarms, turbine thrust bearing trip alarm, and , bearing metal high temperature alarm are verified to actuate. : Vibration levels above normal are verified. The turbine ; t supervisory recorder is checked for temperature indications of a wiped bearing. Date of last test: 3/30/90 Concern (s) found/ Planned resolution: " The thrust bearing temperatures on the turbine supervisory panel did not respond as expected indicative of a wiped bearing. This malfunction will be corrected within 4 years of initial certification. I t l i i b
- - . - . . - _ __ _. - .J
I Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 119 of 210 Malfunction Test #97, Turbino Bearing Oil Low Pressure l Initial Conditions: 100% power. Sequence of events: Malfunction 97 is entered. Automatic actions and alarm response are verified including: Turbine lube oil reservoir low level alarm. Bearing oil pressure decrease and low pressure alarm Auto start of auxiliary oil pressure pumps and stabilization of bearing oil pressure. Date of last test: 3/30/90 Concern (s) found/ Planned resolution: Nona.
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix IX - Malfunction Tests Page 120 of 210 Malfunction Test #98, Turbine Lube 011 Temperature Controller Failure Initial conditions: 100% power. Sequence of events: Malfunction 98 is entered. Automatic actions and alarm response are verified including: Increasing lube oil temperature Zero lube oil temperature controller output Turbine lube oil high temperature alarm Generator seal oil trouble alarm Increasing turbine bearing temperatures. Manual control of the turbine oil temperature controller is taken and lube oil temperatures are decreased. The controller is returned to auto. Malfunction 98 is deleted. l' Alarms are verified to clear. Turbine lube oli temperatures, bearing metal temperatures, and vibration levels are verified to return to normal values. l Date of last test: 3/30/90 Concern (s) found/ Planned resolution: None.. l 1 1
- , . . . _ . - - - . - - . .- , _ . . . _ . _ _ _ . - ~ . . .. _ , - . . - .
l Grand Gulf Muclear station ! simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests i Page 121 of 210 Malfunction Test #99 (A-L), Turbine Bearing High Vibration Initial conditions: 100% power. Sequence of events: Malfunction 99 is entered for each vibration recorder point of N34-YJR-R638. Vibrations above normal and the Turbine Bearing high vibration alarm are verified. Malfunction 99 is deleted. Vibration levels are verified to return to normal. Alarm are verified to clear. Date of last test: 3/30/90 t concern (s) found/ Planned resolution: None. l
- .-_- - .- - . _. . . = . . - - . - . .-- --
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 122 of 210 Malfunction Test #100, Turbine Exhaust Hood Spray Failure Initial Conditions: 10% p'ower, turbine ready to roll. Sequence of events: The turbine is rolled to 1800 RPM and no load conditions. Malfunction 100 is entered to activate the failure due to a blown fuse on the motor controller for MOV-N19-F141. Manual and automatic control of N19-F141'are verified to not allow opening of this valve. Alarm response on increasing hood temperature and hood spray failure are verified. Manual bypass valve N19-F140 is opened to reduce temperatures and clear the high temperature alarm. Malfunction 100 is cleared. N19-F141 is verified to auto open. The manual bypass valve is closed. Alarms are verified to clear. Malfunction 100 is entered and the generator is synchronized on the grid. Power is increased to 30%. Valve N19-F141 is verified to remain open after the load. increase. Date of last test 7/12/90 concern (s) found/ Planned resolution: None.. i l . . - .
l Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 123 of 210 Malfunction Test #101 (A-C), Turbine Bellows Seal Fill Valve rails Shut Initial Conditions: 100% power. Sequence of events: Malfunction 101 is entered for each LP turbine bellows. Condensor Turbine Expansion Joint low level alarms are verified. Condenser vacuum is verified to decrease and absolute condenser pressures are verified to increase after about 10 mintites. Generator load is verified to decrease. Malfunction 101 is deleted. Vacuum and load conditions are verified to return to normal-over a 30 minute-time period. Expansion joint now level alarms are verified to clear. The test is repeated for each bellows, A, B, and C. The Malfunction 101 is activated simultaneously for the A, B, and C bellows. The same parameter changes are verified to occur but over a shorter time reference. The main turbine is confirmed to trip on low vacuum. Date of last test: 3/30/90 Concern (s) found/ Planned resolution: None. i l-I
- is
- r-w Tr + * - - - N 7 -M-*FrT T 'T* - **' == =' * '-TT
Grand Gulf Muolear station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 124 of 210 Malfunction Test #102, Generator Auto Voltage Regulator Trip Initial conditionst 100% power. Sequence of events: With malfunction 102 active, the Exciter TVR cabinet failure alarm is verified with auto transfer to manual control. , i Manual backup control of voltage is verified. Malfunction 102 is deleted. Alarms are cleared. Auto TVR control is reestablished. Date of last test 4/1/90 Concern (s) f~ '4/ Planned resolution: None. I
l Grand Gulf Nuclear station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 125 of 210 Malfunction Test #103, Generator Trip Initial conditions: 100% power. Sequence of events: Malfunction 103 is entered. Automatic action and alarm response are verified including the followingt Generator trip, associated alarms and decreasing output. Generator output breaker opening. Reactor scram. ATWS-ARI/RPT actuation Turbine Control Valve fast closure alarm EOC RPT actuation Turbine coast down Generator lockout status on remote function status. Date of last test: 4/1/90 Concern (s) found/ Planned resolution: None. l
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts-Appendix IX - Malfunction Tests Page 126 of 210 Malfunction Test #104, Generator Rotor Cooling Water Flow decrease Initial conditions: 100% power. Sequence of events: Malfunction 104 is inserted to fail as is. The severity is raised to 50%. Automatic actions and alarm response are verified including the following: Generator rotor high temperature and primary water trouble alarms. Generator rotor low' primary water flow alarm. Turbine trip followed by reverse power generator trip. Reactor Scram. Turbine Control valve fast closure and alarm. i Gener?'.or lockout status doing remote function status indication. 1 EOC-RPT actuation. ATWS-ARI/RPT actuation. Generator output breaker opening. Generator output. decrease. Turbine coastdown. Date of last test: 4/1/90 l Concern (s) found/ Planned resolution: None.
Grand Gulf Nuclear Station Simulator certification Initial Report, March-1991 Simulator Test. Abstracts Appendix IX - Malfunction Tests: Page 127 of 210 Malfunction Test #105, Generator Primary Cooling Temperature Controller Failure Initial Conditions: 100% power. Malfunction 105 is inserted to fail as is. The severity is decreased to one half its initial value. Automatic actions and alarm response are verified including the following: Generator primary water, high stator temperature, and bushing temperature alarms. Increasing stator temperature indication. The generator stator cooling temperature controller-is placed in wanual and stator temperatures are_ stabilized and decreased-until all high temperature alarms are cleared. Af ter clearing malfunction 105, the generator stator cooling-temperature controller is returned to automatic mode.
- Malfunction 105 is reentered to fail as is. The: severity is reduced to 0%. Similar effects as noted above are verified to occur but with a faster rate. The Generator primary water-
-high temperature alarm and turbine / generator trip are verified _
to-occur. The simulator is reinitialized to 100% power. q Malfunction 105 is entered and its severity is increased.to 100%. Generator stator cooling temperatures are verified to-decrease. The generator primary- water / cold gas low differential temperature alarm is verified. Date of last test:- -4/12/90 1 L Concern (s) found/ Planned resolution:- None. l
. . _ - . . . _ . _ . . . __ _ _ .-, . . ~ _ -.
Grand Gulf Muclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II.- Mc1 function Tests Page 128 of 210 Malfunction Test #106, Loss of Generator Primary Cooling Water Initial Conditions: 100% power. Sequence of events: Malfunction 106 is entered. Automatic actions and alarm response are verified including: , All affected Generator Primary Water low flow alarms.. Generator Primary water trouble alarm while speed is above 1750 RPM. Auto start of the auxiliary primary water circulation pump, , Turbine trip and closure of all_ turbine steam valves. Turbine coastdown. Malfunction 106 is deleted. Primary water flow is verified
- restored during turbine coastdown.
j Date of last test: 5/12/90 Concern (s) found/ Planned resolution: Ncne. 1 o e B
. _ _ - . . - - ~ _ . . ... ..- . . . - . -. . . . _ . . , .
-l l
i Grand Gulf Ruclear Station j Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 129 of 210-Malfunction Test #107, Generator Hydrogen Temperature Controller Failure Initial Conditions: 100% power. Sequence of events: , Malfunction 107 is entered-to-fail the generator hydrogen temperature controller as is. The severity is reduced to zero. Automatic actions and alarm response are verified including the following: , Increased generator hydrogen temperature and pressure. Increased generator-seal oil pressure increase. Increased generator hydrogen temperature- controller- ' output. Increased generator stator cooling water temperature controller output. Various _ alarms- on generator hydrogen _and stator temperatures. Malfunction 107 is deleted. Alarms are verified to-clear. Generator hydrogen-temperatures and1 pressures are verified. to return to normal. Malfunction 107 is again entered to fail the-generator hydrogen temperature controller as is. Malfunction severity is increased to 100%. Generator-hydrogen temperature _is verified to- decrease. The hydrogen temperature controller and stator cooling temperature controller outputs arerverified to decrease. Generator load is verified to remain 1 constant. Date of last test: 5/21/90' L Concern (s) found/ Planned resolution:
- 4 Generator hydrogen'and seal oil pressure increased only slightly when controller output .was - f ailed to- zero. This- will be corrected with lthe model replacements in the simulator upgrade project on a schedule-to be-completed'within 4 years of initial certification.
l __ ~- _ - _
Grand Gulf Nuclear 8tation Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 130 of 210 Malfunction Test #108 (A,B), Isolated Phase Bus Duct Blower Trip Initial Conditions: 100% power. Sequence of events: Malfunction 108A is entered. Automatic actions and alarm response are verified including the following : Isolated phase bus cooling fan A trip and auto start of I the B fan Isophase Bus Trouble alarm. ' Attempts to restart the A fan with remote functions are verified to fail until the malfunction is deleted. , The test is repeated using malfunction 108B for the trip of the'B fan and auto start of the A fan.. Date-of last test: 4/1/90 Concern (s) found/ Planned: resolution: None. ? l ( , l t
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 131 of 210 Malfunction Test #110 (A,B), Steam Jet Ejector (SJAE) Steam Supply Failure Initial Conditione 100% power. Sequence of events: Malfunction 110 is entered for the operating SJAE. Automatic actions and alarm response are verified including the following: SJAE low pressure and low flow alarms Closure of the SJAE first stage suction valve Decreasing condenser vacuum and applicable alarms Turbine trip on low vacuum Reactor scram ATWS ARI-RPT actc.ation EOC RPT actuation SRV actuation The simulator is reinitialized to 100%,-the other SJAE is placed in service and the test is repeated. Date of last test: 4/2/90 < Concern (s) found/ Planned resolution: None. l l l l
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 132 of 210 Malfunction Test #111, Offgas High Activity Following Charcoal Absorbers Initial Conditions: 100% power. Sequence of events: Malfunction 71 is entered at 1% severity. Severity is increased to 100%. Automatic actions-and alarm response are verified including the following: The range of offgas post treatment high radiation alarms Increased offgas post treatment radiation levels. Auto closure of N64-F045. Date of last test: 4/2/90 , Concern (s) found/ Planned resolution: None.
Grand Gulf Nuclear Station Simulator certification Initial Report, March-1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 133 of 210 Malfunction Test #112, Main Condenser Hotwell Makeup Control f Failure Initial Conditions: 100% power. Sequence of events: 3 Malfunction 112 is entered to cause loss of air to normal and
- emergency hotwell makeup valves. An SRV is opened. Automatic actions. and alarm . response are verified including- the following:
Decreasing hotwell level No observed makeup flow - Hotwell level low alarms . Condensate pump trip and alarms Condensate booster pump trip and alarms Loss of feedwater and reactor feed pump trip and alarms i Reactor level decrease and associated' alarms Reactors Scram at level 3 , Auto transfer of recirculation pumps at level 3 and decrease in core flow. Date of last test: 4/2/90 Concern (s) found/ Planned resolution: None.
+
l l l i i i
. _ . = .
k Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 134 of 210 Malfunction Test #113 (A,B), Main Condenser Tube Rupture (Variable) . Initial Conditions: 100% power. Sequence of events: 4 Malfuncticn 113 is entered and tested individually on the A-and B circulating water loops. For each case, automatic actions and alarm response are verified including: Increased conductivity in the condensate, condensate demineralizer effluent, reactor feedwater,'RWCU and recirculation flow path with associated conductivity alarms. Increased Condenser hotwell level Increased feedwater turbidity Increased condensate demineralizer differential pressure Increased CRD, Recirculation and RWCU 61ssolved' oxygen CST High/ Low level alarm After deleting the malfunction, all plant- parameters are-
- verified to return to normal over an extended period of time.
Date of last test: 4/3/90 Concern (s) found/ Planned resolution: RWCU A and B conductivity did not increase due_to expected depletion of the RWCU filter resin. This -deficiency -will be corrected within 4 years of initial certification. l l
Grand Gulf Nuclear Station simulator certification Initiel Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 135-of 210 Malfunction Test #114 ( A, B) , Main Circulating Water Pump Trip Initial conditions: 100% power. Sequence of events: Malfunction 114 is entered to trip each circulating water pump. Automatic actions and alarm response are verified including the following: Circulating water pump stops Circulating water pump overload / trip alarm Auto closure of discharge valve Circulating water loop low flow alarm Decrease in condenser vacuum and generator-output Turbine trip After malfunction 114 is deleted, the appropriate BOP transformer tap settings-is raised. The circulating water pump is . restarted. Alarms are- verified to clear.. The t circulating water pump discharge valve is verified to automatically open. Date of last test: 9/5/90 Concern (s) found/ Planned resolution: None.- l l I r , - . - , , , . . - . - . . . - . ~ , - . - -, , , , .
~ .,,
l Grand Gulf Nuolear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunotion Tests Page 136 of 210 Malfunction Test #115 ( A, B, C) , Condensate Pump Trip Initial Conditions: 100% power. Sequence of events: Malfunction 115 is entered to trip each_ condensate pump. Automatic actionu and alarm response are verified including the following: Condensate pump trip and alarm. Decrease in condensate and condensate booster pump discharge pressure. Decrease in Reactor feed pump suction pressure. The af fected pump discharge valve is closed manually. The trip malfunction is deleted. Trip annunciation is cleared. The affected pump is restarted and- its discharge valve verified to automatically open.- The condensate' pump recirculation valve is verified to cycle open and closed. Condensate and feedwater parameters are verified to return to normal. Date of last test: 4/3/90 Concern (s) found/ Planned resolution: None.
$h/ /W%%
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991' Simulator Test Abstracts Appendix IX - Malfunction Tests Page 137 of 210 Malfunction Test 116 (A-H), Condensate Demineralizer Resin Bed (A-H) Failuts Initial Conditions: 100% power. i sequence of events: All 8 condensate demineralizers are placed in service using remote functions. Malfunction 113A is entered at 10% severity. This causes a condenser tube rupture and ancreased conductivity at the condensate pump discharge and condensate demineralizer inlets. Each demineralizer malfunction is tested by entering malfunction 116. This simulates resin bed failure due to depletion of the resin capability. Automatic actions and alarm response are verified including the following: Increasing conductivity in the condensate and feedwater by chart recorder indication on panel H13-P680. Condensate Demineralizer Effluent High conductivity alarm Condensate demineralizer trouble alarm. Malfunction 116 is deleted. This simulates-a return of the l affected demineralizer to normal. Conductivity 'is-verified to drop at the demineralizer effluent. Demineralizer are alarms are verified to clear. This test is repeated for all 8 demineralizers. Date of last test: 5/12/90 1 Concern (s) found/ Planned. resolution: None.
- . , , . . . . _ . - - .. - ~ . - - , _ _ . .
. ~ . - ._
Grand Gulf: Nuclear station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 138 of 210 Malfunction Test #117 (A-H), Condensate Demineralizer (A-H) High High differential pressure (Variable) Initial Conditions: 100% power. Sequence of events: All 8 condensate demineralizers are placed in service using remote functions. Each demineralizer malfunction is tested by inserting malfunction 117. This simulates high differential pressure in the affected demineralizer. ' The malfunction is-inserted on all demineralizers. Severity is set to fail differential pressure as is. One at a time, cach demineralizer differential pressure is increased by raising severity to 100%. Automatic actions and alarm response are verified, including the following: Increased demineralizer differential pressure and decreased flow in the affected demineralizer. Auto opening of the filter demineralizer bypass valve on excessive differential pressure. High differential pressure and condensate dimineralizer trouble alarms. The malfunction is deleted and the other domineralizers are tested. Date of last test: 4/4/90 Concern (s) found/ Planned resolution:. The condensate demineralizer high differential pressure alarm and condensate demineralizer trouble alarm did not occur as expected. These deficiencies will be corrected.within 4 years of initial certification.
)
l
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 l simulator Test Abstracts ' Appendix IX - Malfunction Tests Page 139 of 210 Malfunction Test #118 (A,B, and C), Condensate Booster Pump Trip Initial Conditions: 100% power. Sequence of events: Malfunction 118 is entered to trip each condensate booster pump individually on a motor phase differential trip. Automatic actions and alarm response are verified which include: Condensate booster pump trip and alarm indication. Decrease in condensate booster pump discharge pressure. Decrease in Reactor feed pump suction pressure. The affected pump's discharge valve is manually closed. The malfunction is deleted. The pump is restarted. The_ pump's discharge valve is verified to open. The condensate booster pump recirculation valve is verified to auto open and close. Condensate and feedwater parameters are verified to return to normal. The test is repeated for each pump. Date of last test: 4/4/90 Concern (s) found/ Planned resolution: None.
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991-simulator Test Abstracts Appendix II - Malfunction Tests Page 140 of'210 Malfunction Test #119, Heater Drain Tank Level Controller Failure Initial Conditions: 85% power. Sequence of events: Malfunction 119 is entered to-cause a' failure of the Heater Drain Tank level controller. Automatic actions and alarm response are verified including the following: Decrease in flow and opening of the recirculation valves for both the A and B heater drain pumps. Decrease in. Reactor Feed Pump suction pressure. ater drain tank high level alarm. Closure of both heater-drain tank level control valves with actual level-being maintained at approximately 30 inches. Malfunction 119 is deleted. Parameters are verified to return to normal. Date of last test: 4/20/90 Concern (s) found/ Planned resolution: None. I
c- - Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 141 of 210 Malfunction Test #120 ( A-L) , Low Pressure Fe<,dwater Heater High High Level Initial Conditions: 100% power. Sequence of events: Malfunction heaters 120 is entered to test each of the four feedwater in the three low pressure. feedwater strings. Malfunction 120 fails the Hi-Hi level switch for the respective heater. As this occurs, automatic actions and alarm response are verified including the following: Feedwater heater Hi-Hi level alarm. Feedwater heater string isolation (inlet and outlet MOV closure). Slight decrease in feedwater temperature. Reactor power and generator output are monitored for an increase in power. After the malfunction is deleted, the heater str ing is restored to normal. The Hi-Hi level alarm is verified clear. All condensate /feedwater parameters are veril: led to rcturn to normal. Control valves which isolate other feedwater heaters and the seal steam generator drain tank are verified to close when the malfunction is entered for affected heaters. s Date of last test: 4/4/90 Concern (s) found/ Planned resolution: N33-F019A/B/C did not isolate as expected on failure of the number 4 feedwater heater. These deficiencies will be corrected within 4 years of initial certification.
- _ _ _ _ _ - _ _ _ _ - _ - =
Grand Gulf Nuclear Station Simulator certification Taitial . Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 142 of 220 Malfunction Test #121 A,B, Feedwater Pump Signal Failure Initial Conditions: 50% power. Sequence of events: The A RFP is placed in service being controlled by the master controller in auto. Malfunction 121A is entered to simulate an electrical signal loss from the individual turbine speed control M/A transfer station. The RFP control signal failure alarm is verified. The amber status light abova the Control signal failure reset pushbutton is verified to net ~ reset. Power is increased by about 5%. Automatic actions are verified including the following: Increase in core flow and-APRM flux. Increase in feedwater master controller output.
-Increase in "B" RFP speed with no increase in "A" RFP l speed.
l l Increase in both "A" and "B" RFP speed controller outputs. Increase in "B" RFP suction flow with no increase in "A" RFP suction flow. The "A" RFP speed controller is placed in manual. Increasing-the output is verified to havo_no effect. The hydralic jack l on the "A" RFP is engaged and response using the manual speed changer is verified on the "A" RFP. The malfunction is deleted. The "A" RFPT control signal failure is reset and the "A" RFP is returned to automatic control.- The test is repeated for tr.a "B" RFP with malfunction.121B in similar fashion. Date of last test: 4/4/90 Concern (s) found/ Planned resolution: None.
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 simulator Test Abstracts-Appendix II - Malfunction Tests Page 143 of 210 Malfunction Test #122 A,B, Reactor Feed Pump Turbine Steam-Supply Failure Initial Conditions: 35% power. Sequence of events: Malfunction 122A is entered to simulate mechanical binding-of the "A" RFPT HP control valve. Power is reduced to reduce generator load to 150 MWe. A lack of response and decreased feed flow are verified for the affected RFP. The "B" RFP-is placed in service in Auto Control. The "A" RFP is verified to continue to decrease speed, and trip on low flow with respective low flow and RFP trip alarms. After level is stabilized, malfunction 122B is entered to cause failure of the "B" RFP. Level and feedwater flow are verified to decrease. The "B" RFPT is verified to also decrease speed and trip on low flow as with the "A" RFP. Date of last test: 8/24/90 Concern (s) found/ Planned resolution: None. I I
)
Grand Gulf Nuclear Station simulator certification ; Initial Report, March 1991 ' simulator Test Abstracts Appendix II - Malfunction' Tests Page 144 of 210 1 Malfunction Test #123 ( A, B) , . RFPT Overspeed Trip Initial Conditions: 100% power. Sequence of events: Malfunction 123 is entered for'each RFP individually. Automatic actions and alarm rasponse are verified including the following: RFP trip and alarm indication. RFP Discharge Valve Closure. RFP Recirculation Valve Opening. RFP Drain Valve Opening. Closure of RFPT control valves. Decrease in RFP Speed and discharge pressure. Increase in unaffected RFP response to restore level. Reactor Hi/ Low level alarm. Recirculation Flow control valve runback and alarm indication and-decrease in power and core flow. Malfunction 123 is' deleted, the affected RFP is reset and returned to service on the master level controll'er in auto. The-test is repeated for the "B" RFP. Date of last test: 4/5/90 Concerns found/ Planned resolution: None.
Grand Gulf Nuclear Station Simulator Certification : Initial Report, March 1991 Simulator Test Abstracts. Appendix IX . Malfunction Tests Page 145 of.210 Malfunction Test #124, Failure of Startup Level Control Valve Signal (Variable) Initial Conditions: 25% power. Sequence of events: Malfunction 124 is entered to simulate startup level control valve position failure. The severity is reduced to 10% to verify decreasing' level and alarm indication on low level. The severity is increased.to 25% to verify increasing level and alarm indication on high level. The severity is reduced to minimum to fail the startup level control valve fully closed and to verify decreasing level.and alarm indication on low level. The severity is increased to maximum to fail the startup level control valve open and verifying increasing level and alarm indication on high level. Level is cycled between 20 and 50 inches for these tests. Finally malfunction 124 is deleted and reactor level is verified to stabilize at 31 inches with the reactor high/ low level alarm clearing. Date of last test: 8/20/90 Concern (s) found/ Planned resolution: None. l
, --e, , ,4
, . - ~, e , ,
Grand Gulf Nuclear Station i Simulator Certification Initial Report, March 1991 Simulator Test Abstracts . Appendix II - Malfunction Tests ! Page 146 of 210 Malfunction Test #125, Failure of Steam Flow Signal to Feedwater- ' Control Initial Conditions: 100% power. Sequence of events: > Malfunction 125-is entered to fail the "C" main steam line - flow signal input to feedwater control. Automatic action and alarm response are verified which includes the following: ' Steam flow C indicator goes to zero flow, i Total steam flow drop to 75% of rated. Decrease in RFP A and B speed and drop in reactor level
- which stabilizes approximately 14 inches-below normal-Reactor High/ Low level alarm '
Feedwater control is switched to single element control. Level . is verified to return to normal. Feedwater control is returned-to 3 element control. Level is-again verified to drop below normal. Malfunction 125 is deleted. Steam and _- feedwater flow, and reactor level are - verified to return _ to. normal with the . reactor level alarm clear.
- Date of last test: 4/15/90 concern (s) found/ Planned resolution: None.
d l
.me- - - , . . c -. -
-n -
Grand Gulf Nuclear Station Simulator certification Initial Report,. March 1991 simulator-Test Abstracts Appendix IX - Malfunction Tests Page 147 of 210 Malfunction Test #126 (A,B), Feedwater Control Vessel Level Sensor Failure Initial Conditions: 100% power. Sequence of events: With the "A" level selected for control, the "B" level signal is failed to zero with malfunction 126B. Actuation of the Reactor water level signal failure alarm and status light are verified. Malfunction 126B is deleted to restore all narrow range level indicators to normal. The "B" level is selected for control and the "A" level signal failed to zero with malfunction 126A. The Reactor water level signal failure alarm is again verified to actuate. Malfunction 126A is deleted to restore all narrow range level indicators to normal, , , While each level is selected for control, malfunction 126 is entered to fail it to zero output. Automatic actions are verified which include the following: Increase in both RFP speeds. Increase in reactot level to level 8. Scram, Turbine, and RFP trip at level 8. Auto transfer of-recirculation pumps to slow speed. This portion is repeated to test both the "A" and "B" level
-sensor failure.
Date of last test: 4/9/90 Concern (s) found/ Planned resolution: None.
Grand' Gulf Nuclear station simulator certification Initial Report, March 1991-simulator Test Abstracts Appendix II -' Malfunction Tests Page 148 of 210 Malfunction Test #127, Feedwater' Master Controller Fails Open Initial Conditions: 100% power. Sequence of events: Malfunction 127 is entered to fail the feedvater master controllor to maximum output. Automatic actions and alarm response are verified which include the following: Increase in both RFP speed, discharge pressure for both RFPs. ' Flow decrease in RFP suction pressures. Increase in reactor level and ARPM flux. Reactor Scram on high APRM flux. Reactor Hi/ Low level alarm-Reactor Feed pump and Main turbine trip and alarm indications at level 8. Reactor and RPS level 8 alarms. Date of last test: 4/10/90 ; Concern (s) found/ Planned resolution: None. k l 1 l l l s
g + a a a _a - , ~ - nw w4, - - - - - - - , - -
, _ -w _ ~- - - - - - - - , - - - - - - - - - - , - - - , - - - - -
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 149 of 210 Malfunction Test #128, Feedwater Master Controller Fails Shut Initial conditions: 100% power. Sequence of events: Malfunction 128 is entered to fail the feedwater master controller to minimum output. Automatic actions are verified which include the following: Decrease in both RFP's speed and discharge-pressure. Reactor level decrease. Opening of both RFP recirculation valves. Date of last test: 4/10/90 Concern (e) 'Jound/ Planned resolution: None. i n 9
+ e e ~ w
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 simulator Test. Abstracts Appendix II - Malfunction Tests Page 150 of 210 Malfunction Test #129 (A,B), High Pressure Feedwater Heater Tube Leak (Variable) Initial Conditions: 100% power. Sequence of events: Malfunction 129 is entered at 100% severity for each #5 FW heater. Automatic actions and alarms are verified which include the following: FW Heater Hi and Hi-Hi level alarms for the affected heater. Extraction steam and Bleeder Trip Valve (BTV) closure for the affected heater. Decrease in FW temperature. Increase in APRM flux, steam -flow, feed flow and generator output. This test is repeated for both the SA and SB-heaters. The= simulator is initialized and malfunction 129 is-entered at 10% severity for each #6 FW heater. Heater Drain tank level-is verified to increase uncontrollably .from dump' valve
-opening. The severity is increased to 100% and.the above automatic actions and alarms are verified plus the-following additional items:
Restricting' orifice bypass drain valve opening for the affected heater. MSR second stage reheater drain tank high level alarm (A or B) for the affected heater. Bypass valve opening. Reactor Scram on high neutron flux. This test is repeated for both the 6A and 6B heaters. Date of last test: 4/10/90 Concern (s) found/ Planned resolution: None.
l 1
. Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 151 of 210 i Malfunction Test #130, Netverk Load Increase Initial Conditions: 100% power.
Sequence of events: Power is reduced to 80%. Malfunction 130 is entered. Automatic actions are verified including the following: Decrease in AC Incoming line voltage and generator terminal voltage. Decrease in generator frequency to 58.5 Hz and. turbine speed to less than 1800 RPM. Increase in_ generator field current, field voltage, and reactive load. Decrease in plant bus voltages. Malfunction 130 is deleted and plant parameters are verified-to return to normal. Date of last test: 4/10/90 Concern (s) found/ Planned resolution: None. 1
i l Grand Gulf Nuclear Station ' simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfun; tion Tests Page 152 of 210 Malfunction Test #131, Network Load Decrease Initial Conditions: 100% power. Sequence of events: , Malfunction 131 is entered. Automatic actions and alarm respor4ss are verified which include the following: Increase in AC incoming line voltage and generator terminal voltage. Increase-in generator frequency to 60.5 Hz. Increase in turbine speed. Decrease in generator field current and field voltage. Decrease in generator reactive load in negative direction. Partial closure of turbine control valves and decrease in generator load. i Opening of turbine bypass valves. Malfunction 131 is deleted. Plant parameters are verified to to return to normal. Date of last test: 4/10/90 Concern (s) found/ Planned resolution: None.
.. _ - - _ _ _ _ _ - _ a
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator tast Abstracts Appendix IX - Malfunction Tests Page 153 of 210 Malfunction Test #132, Network Load Loss Initial Conditions: 100% power. Sequence of events: Malfunction 132 .is entered. Automatic actions and alarm response are verified which include the following: 500 KV breakers J5228 and J5236 trip. Increase in bus 11R voltage. Increase in 500 KV frequency then decrease until the turbine / generator trips. Reactor Scram. Turbine CV fast closure trip alarm. ATWS-ARI/RPT actuation. Trip of generator output breaker J5232 when the generator trips on loss of Service Transformer 11. The 132 issimulator entered.is reinitialized to 35% power and malfunction to trip. No scram 500is KV breakers verified J5228 with to occur and J5236 are verified all steam flow going through bypass valves. The generator is verified to continue to carry in plant loads. Date of last test: 8/21/90 Concern (s) found/ Planned resolution: None. A
l Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 i simulator Test Abstracts Appendix IX - Malfunction Tests Page 154 of 210 Malfunction Test #133 A and B, Service Transformer Lockout Initial Conditions: 100% power. Sequence of events: All ESF and BOP buses are transferred to service transformer 11 (STil). Malfunction 133A is entered to trip STil. Automatic actions and alarm response are verified which include the following: Service Transformer trouble, sudden pressure, primary and secondary trip lockout alarms. Trip of breaker 552-1105 and associated breaker trip alarm. (buss 11R feeder breaker) Loss of bus 11R voltage and current' loss of bus voltage and undervoltage alarms on BOP busses 11HD, 12HE, 13AD, 14%E, 18AG, and 28AG. A*1 ESP and BOP busses are manually transferred to all other possible alternate power supplies, i l The test is repeated for Service Transformer 21 lockout using ! malfunction 133B in similar fashion for ST21 and bus 21R.
- Date of last test 4/11/90 Concern (s) found/ Planned rerolution
- None.
,-, wn r- .e e- ..-p ,. - - . - - - - n <- p ----- - e .y s v ,-- --e- n.-m p..
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Testa Page 155 of 210 1 Malfunction Test #134, BOP and ESF Transformer Lockout Initial Conditions: 100% power. I Sequence of eventst Malfunction 134 is entered to fail each transformer. This includes BOP transformers 11A, 11B, 12A, 12B, 13, and 23 and ESF transformers 11 and 12. For each DOP transformer lockout malfunction test, all applicable BOP buses are transfurred to the respective Transformer. Lockout, trouble, incoming feeder breaker trip alarms are verified. Bus doenergization and undervoltage alarms are verified. Affected buses are reenergized from alternate feeders. Undervoltage lockouts are reset using remote functions. Af fected Load Centers are reenergized. *u.e transformer lockout malfunction is deleted, lockouts reset using remote functions. Applicable buses are rennergized from the affected transformers. For loss of DOP busses 11HD, 12HE, 13AD, and 14AE, reactor scram, recirculatior, pump trip and loss of feedwater with level recovery by HPCS and RCIC are verified. For loss of BOP buses 18AG and 48AG, radial well indication and control is verified lori with alerm indication of low PSW header pressura For ESF transformer 11 and 12 lockout., malfunction tests, all ESF buses are aligned to the ESF transformer to be locked out. Applicable transformer lockout and trouble alarms and incoming feeder breaker trip and bus lockout trip alarms are verified. Buses are verified to doenergize and be reenergized by their respective diesel x
..erator. Proper Load shedding and t ' Sequencing (LSS) actuation on bus undervoltage and loss of power are verified along with correct status lights and alarm response. The respective bus power is transferred to all possible alternate AC sources.
Date of last test: 9/4/90 Concern (s) found/ Planned resolutiont Radial Well bus indication and control should be lost when LCC 18BG1 is lost. This item has been corrected and ratested.
~ ~ - - -
Grand Gulf Nuclear Station Simulator Certification
- Initial Report, March 1991 Simulator Test Abstractri Appendix II - Malfunction Testo Page 156 of 210 Malfunction Test #135, Switchyard Fault (500 KV and 115 KV)
Initial Conditions: 100% power. i Saguence of events: Malfunction 135 is entered. Automatic actions and alarm response are verified which include the following: Reactor Scram on TCV fast closure-ATWS-ARI/RPT actuation Recirculation pump trip to off Total loss of condensate and fecdwater MSIV closure on loss of RPS SRV actuation ESF bus deenergitation, diesel genarator start, and ESF l bus reenergization from their respective diesel ' generator. Auto start of SSW A, B, and.C. Deenergization of Service Transformer 11 and 21 LOP sequence from LSS 1 Deenergization of all BOP buses Reactor level decrease to level 2 1 Level recovery by HPCS and kCIC ! Group 1,2,3,6,7,8, and 10 isolations
- ESP transformer 12 115 KV source lost alarm i t
4
- - - - - - . - , , - .-- ,w.,e-w. ..y,...e . - .
7 um ,n.- -,a--. - - ,,g- ,, , , y,-- -
~ - . . _ - - . - . . _ . - . - . . . _ - . _ _ .. . - - ~ _ - . . . . . _ . . . . . __ _ _ _ _ _ - _ _ - _ -
Grand Gulf Nuclear Station 3 Simulator certification Initial Report, March 1991 simulator Test Abstracts ! I Appendix IX - Malfunction Tests Page 157 of 210 I t I Auto start of the DC Seal Oil pump Auto start of RFPT A and B DC cil pumps 1 Auto start :s! main turbine DC oil pump after turbine : coastdown. Dato of last test! 4/11/90 ; i Concern (s) found/ Planned resolution: None. ! i h t i t, b l \ t i i l f ! 6
?
}
.j i
i I i l. i
,. - - . ~ . - ,e - -- -.
,,--.n,,, ,,, ,. n c, , , , - . .ac., . . . , , .-.w. . . , . , - , , . ---..,.rrec.,,,a
1
- i Grand Gulf Nuclear Station )
simulator certification i Initial Report, March 1991 R simulator Test Abstracts Appendix II - Malfunction Tests l Page 158 of-210 i Malfunction Test #136, 34.5 KV Bus Overcurrent Trip Initial conditions: 100% power. Sequence of events: Malfunction 136 is tested for each 34.5 KV Bus which' includes: 21R,-11R, 12R and 13R For the overcurrent trip test of buses 11R and 21R, all ESF r 4 and BOP busses are aligned such that the loss of the 34.5 KV-bus will result in deenergitation of all BOP and ESF 4.16 and 6.5 KV buses. This included buses 11HD, 12HE, 13AD, 14AE, 15AA, 16AB, 17AC, 18AG, and 28AG. For the overcurrent test of busses 12R and 13R, BOP busses 11HD, 12HE, 13AD, and 14AE are aligned such that the loss of' the 34.5 KV bus will result in thel loss of BOP' busses 11HD, 12HE,13AD, and 14AE. Automatic actions and alarm response are verified for each bus trip which includes the following: Incoming feeder breaker trip alarms
- Loss of voltage to deenergized busses l
Bus undervoltage and lockout alarms Note: Bus loss effects are tested during malfunction-135 test. l The bus overcurrent trip malfunction-is-deleted. The bus . I overcurrent lockout is- reset with remote functions. The l deenergized busses ara reenergized from.the previously tripped 34.5 KV bus. Date of last test: 4/16/90 concern (s) found/ Planned resolution: None. yo..ww*--,1r-gr- +wi, y- w-4 w y- w - b- ,*w
. gn ,-eq+4e e e, wy.y f r- .,r---,. 4- --,w-,
. p.- r m- e4 . - - w w -s -m-- , . - . , - - .-e, e"-s,,= w .wn-.-e
i Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 159 of 210 Malfunction Test #137, 34.5 KV Bus Phase Differential Trip Initial conditions: 100% power. Sequence of events: Malfunction 137 is tested for 34.5 KV Buses 11R and 21R.- Prior to the test of each bus, ESF and BOP buses are aligned such that the loss of the 34.5 KV bus will result in deenergization of all BOP and ESF 4.16 and 6.9 KV buses. This includes buses 11HD, 12HE, 13AD, 14AE, 15AA, 16AB, 17AC, 18AG, and 28AG. Automatic action and alarm response are verified for each bus trip test which includes: Incoming feeder breaker trip alarms Loss of voltage to deenergized busses Undervoltage and lockout alarms Buss loss effects as tested with malfunction 135 test The bus phase differential trip malfunction is deleted. Bus lockouts are reset using remote functions. The 34.5 KV bus is reenergized, BOP and ESF busses are reenergized from the previously tripped 34.5 KV bus. Date of last test 4/16/90 Concern (s) found/ Planned resolution: None. l l l
Grand Gulf Nuclear station : simulator certification l Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests 1 Page 160 of 210 j r Malfunction Test #138, 6.9 KV Bus 11HD/12HE Overcurrent Trip Initial Conditions: 100% power. ! Sequence of events: [ Malfunction 138 is tested for'both 6.9 KV buses. For each test automatic actions and alarm response are verified. A i checklist of the proper load loss actions is used for each bus. These include the followings ; Incoming feeder trip. alarms.- Bus undervoltage and lockout alarms. Loss of voltage to deenergized buses and LCCs. Loss of current through bus feeder and LCC feeder breakers. : Specific loads lost applicable to each bus. Note: Verification of individual- LCC loads are tested under the Malfunction 142 test. i The alternate bus supply breaker is verified to be locked out with the malfunction active. The taalfunction is deleted. The bus lockout is recet using remote functions. The affected bus 1 is reenergized. Applicable loads are . verified to be locked.out ' until bus undervoltage lockouts are' reset using remote . functions.-LCCs are reenergized and bus loads are-restarted. ; Date of last test: 9/4/90 I l . Concern (s) found/ Planned resolution: None. ! l l
.~, , . , _ - , . . . , . , , _ . , , , . . _ . . , ,, ,.m., , <
l i Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts 1 Appendix II - Malfunction Tests Page 161 of 210 Malfunction Test #139, 4.16 KV Bus Overcurrent Trip Initial Conditions: 100% power. Sequence of events: Malfunction 139 is tested for 4.16 KV Buses 13AD,14 AE,18AG, 28AG, 15AA, 16AB, and 17AC. For each test automatic actions and alarm response are verified. A cnecklist of proper load loss actions is used for each bus. These include the following: Incoming feeder breaker trip alarms. Loss of voltage to applicable busses and LCCs. Loss of current to applicablo busses and LCC feeder breakers. Bus supply breaker trip. Specific loads lost applicable to each bus. Verification of individual LCC loads are tested under the Malfunction 142 test. The alternato bus supply breaker is verified to be locked out with the malfunction active. The malfunction is deleted. Bus lockouts are reset using remote functions. The applicable bus is reenergized. Applicable loads on lost BOP busses are verified to be locked out until bus undervoltage lockouts are reset using remote functions. LCCs are reenergized and bus loads are restarted. Date of last test: 9/4/90 Concern (s) found/ Planned resolution Radial:well indication and -control was not lost when LCC 18BG1 is deenergized. This has been corrected and retested.- ' l 1
.,..-..__.,___m . . ~ _ ~ , . - _ _ . , , . - . . . . . . , , _ , , , , _ , , , , . , , . _ , _ _ , , . , , _ , ,,_.,,.m , _ . . , ._._,e__. ,,
j Grand Gulf Nuclear Station simulator Certification
- Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 162 of 210 l
Malfunction Test #140 (A,B,C), Emergency Diesel Generator Fall to Start. Initial conditions: 100% power. Sequence of events: Mal. function 140 is tested for all emergency diesel generators including DG11, DG12, and DG13. Malfunction 140 is entered for the affected diesel and alarms and status lights are verified to actuate. The bus is deenergized by opening its supply breaker. Alarms and status lights are verified to actuate for the bus loss which include: Incoming feeder breaker trip alarms. Undervoltage alarms for the bus and LcCs.
. LSS failure alarms and status lights for loss of bus 15AA and 16AB.
An alternate offsite feedor breaker is closed to reenergize , the affected bus. For bus 15AA and 16AB, LSS sequencing is i verified to occur. The malfunction is deleted. The affected diesel is restarted and paralleled to its bus. l Date of last test 4/17/90 Concern (s) found/ Planned resolution: None.
t Grand Gulf Nuclear station . Simulator certification Initial Report, March 1991 Simulator Test Abstracts , Appendix IX - Malfunction Tests , Page 163 of 210 i Malfunction Test #141 (A,B,C), Emergency Diesel Generator Trip f Initial Conditions: 100t power. j Sequence of events: 1 Malfunction 141 is tested for all emergency diesel generators including DC '_1, 12, and 13. The diesel being tested is ' started, paralleled, and loaded to maximum allowed-load. ; Malfunction 141 is entered to trip the diesel. Automatic actions and alarm response are verified including the following: Incoming bus feeder breaker trip alarns. Various diesel trouble and trip alarms specific to : each diesel. Loss of diesel generator voltage, current, load and ' p frequency indication. The diesel is verified to be locked out from'a manual start. The malfunction is deleted. Lockouts are reset using remote , functions. The diesel is restarted, paralleled and loaded. Date of last test: 12/11/90 Concern (s) found/ Planned resolution: None. t I
}
l l i t k 9 i I
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 164 of 210 Malfunction Test 142, 480 VAC LCC Overcurrent Trip Initial conditions: 100% power. Sequence of events: Malfunction 142 is tested for each 480 Volt LCC. These include: 11BD1 12BE1 13BD1 14BE1 15BA1 16BB1 17B01 11BD2 12BE2 13BD2 14BE2 15BA2 16BB2 18BG1 11BD3 12BE4 13BD5 15BA3 16BB3 11BD4 12BE5 15BA4 16BB4 11BD5 12BE7 15 BAS 16BB5 11BD7 15BA6 16BB6 For each malfunction test, automatic actions and alarm response are verified. A checklist is used to verify proper load loss and ' actions, these include: Incoming feeder breaker trip indication and alarm LCC undervoltage alarms Loss of voltage on applicable LCC Loss of current for the applicable LCC incomiag feeder breaker. L Specific loads components, indicators, annunciators, valves, pumps, chart recorders, etc. applicable to the LCC-tripped. I The malfunction is deleted. The applicable LCC and equipment which l were lost are reenergized. All associated alarms are cleared. l l 1 Date of last test: 8/3/90 Concern (s) found/ Planned resolution: Various alarms, indicators, and other components did not respond as expected. These are - being resolved based on potential training impact and will be resolved or corrected within 4 years of initial certification.
4 t > I i Grand Gulf Nuclear $tation i simulator Certification l Initial Report, March 1991 i simulator Tect Abstracts ! Appendia II - Malfunction Tests ; Page 165 of 210 ! t i Malfunction Test # 143, 120 VAC UPS Trip - Initial Conditions: 100% power. ! I sequence of events: i Malfunction 143 is tested for the following 120 VAC UPS l distribution busses: l 1Y71 1Y92 i 1Y74 1Y76 l! 1Y75 1Y94 1Y76 i 1Y78 i I , For each malfunction, a checklist is used to verify proper j . ?,oad loss and results. These checklists include components,
- indicators, status lights, alarms, controls, controllers, 1 valves, recorders, etc.. .
Date of last test: 8/3/90 ! Concern (s) found/ Planned resolutions ; e various alarms, indicators and other components did.not : respond as expected. These-are being resolved based on i potential training impact. These-will be resolved or l corrected within 4 years of initial certification. ! i t l
}
}
4 i 9 i i 1 l l 1
, _ _ , . . . _ , _ _ , _ . . . _ _ . _ _ . . . . . . , , . . . , , , . . . _ _,..a...,,,._,..-
Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 166 of 210 Malfunction Test #144, 250 V DC Bus 11DF Trip Initial Conditions: 100% power. Sequence of events: Malfunction 144 is entered to trip bus 11DF. Automatic actions and alarm response are verified which include the following: 250 VDC Bus 11DF incoming feeder breaker trip alarms. Loss of pump indicator lamps for DC oil pumps for RFP A and B, Seal Oil, and the main turbine. Turbine DC Oil pump failure and generator seal oil trouble alarma. RFPT and DC oil pump protection / lockout alarms. Malfunction 144 is deleted. All indications are verified to return to normal. Date of last test: 4/18/90 Concern (s) found/ Planned resolution: Alarms "250 VDC BUS 11DF INCM FDR 72-11D05 TRIP " and "250 VDC BUS 11DF INCH FDR 72-11E05 TRIP" did not actuate as-expected. These deficiencies will be corrected within 4 years of initial certification. I
Grand Gulf Nuclear Station I Simulator certification Initial Report, March 1991 Simulator Test Abstracts 1 Appendix IX - Malfunction Tests 1 Page 167 of 210 Malfunction Test #145, 125 VDC Bus Trip Initial conditions: 100% power Sequence of events: Malfunction 145 is tested for each 125 VDC bus. These include 11DA, 11DB, 11DC, 11DD, and 11DE. For each bus loss a list of automatic actions, indicators, controls, status lights, alarms, computer points, valves, components, etc. is used to verify proper automatic actions and alarm response. Date of last test 8/3/90 Concern (s) found/ Planned resolution: Various alarms, indicators, and other components did not respond as expected. These are being resolved based on. potential training impact. These will be resolved or corrected within 4 years of initial certification. i
.- _ . _ . . _ . . ~ . _ , _ , .
Grand Gulf Nuclear Station j simulator Certification Initial Report, March 1991 ! Simulator Test Abstracts Appendix II - Malfunction Tests Page 168 of 210 Malfunction Test #146 A,B 24/48 VDC Bus Trip Initial Conditions: 100% power. Sequence of events: Malfunction 146 is tested for both 24/48 VDC busses 11DH and 11DJ. For each bus loss a list of indicators, controls, alarms, automatic actions, controllers, valves, etc. is used to verify proper automatic actions and alarm response. Date of last test: 7/28/90 Concern (s) found/ Planned resolution: All EHC turbine control indicators and the seal- steam controller did not fail upon a simultaneous loss of 11DH, 11DJ, and LCC 13BD1 as expected. These deficiencies will be corrected within 4 years of initial certification.
Grand-Gulf Nuclear station simulator Certification Initial Report, March 1991 simulator Test Abstracts-Appendix II - Malfunction Tests Page 169 of 210 Malfunction Test #147, Instrument Air System Failure Initial Conditions: 100% power. Sequence of events: The unit I instrument air compressor is placed in standby. Malfunction 147 is entered to rupture the Instrument air header. Automatic actions and alarm response are verified. A checklist of alarms, indicators, systems and components is used to verify actions. Overall effects which are verified include: Air Operated valve failure.(open, closed, fail-as-is). , Control rod drift Scram valves failure open Scram Loss of cooling caused by loss of PSW to components and systems cooled by PSW. Loss of vacuum Loss of Radial Well pumps Loss of drywell, containment and steam tunnel cooling Loss of RWCU and FPCCU Loss of plant chill water.and drywell chill veter. Date of last testi 5/24/90 concern (s) found/ Planned-resolution Unit I Instrument Air Compressor did not auto start on decreasing header pressure. The Instrument Air Receiver low pressure alarm did not actuate.
)
l 1 1 l l j
Grand Gulf Nuclear station simulator certification Initial Report, March 1991' simulator Test Abstracts Appendix II - Malfunction Tests Page 170 of 210 k control / rods did not begin to drift before scram occurreds N62-F060 did not fall open N19-F521 did not fail closed N22-F502 did not fail open N21-F513 did not fail open N62-F060 response has benn corrected. The remaining items will be corrected within .4 years of initial certification. AOV res;aonse is not realistic -for a gradual loss of Instrument A:.r. This deficiency will be corrected as systems are remodeled during the simulator upgrade project. l l l l- _ . _ - _ , . . _ _ . . . _ , - . . .,._,.m. _ _ . , _ , ,, .~ _ - _ . _ , _ . , , _-~.,_.,_..-_m,,. , -..m,-,,,.m,
Grand Gulf Nuclear station Simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 171 of 210 Malfunction Test #148 A,B Standby Service Water (SSW) pump Trip Initial conditions: 100% power. Sequence of events: For the divisional SSW pump being tested, the following are performed prior to entering malfunction 148: The respective SSW system is manually initiated using the pushbuttons on the pB70 panel. The respective emergency diesel generator is started, paralleled, and loaded. The respective drywell purge compressor is started in the test mode. Af ter malfunction 148 is entered, automatic ' actions and alarm response are verified including: SSW pump trip and associated alarms and status lights. Decreasing loop and pump flow including flow through the associated RHR heat-exchanger. SSW pump low discharge pressure alarm. Diesel generator trip and associated alarms and status lights. Drywell purge compressor trip and associated alarms. Malfunction 148 is deleted and the previously tripped' SSW pump is restored to operation. For the SSW pump B trip test, all instrument air and service; air compressors are started. Cooling for the instrument air compressors and the drywell chillers are aligned to SSW B. After malfunction 148 is entered, these components are verified to trip and associated alarms verified to actuate.
- ._ _ ~. -_ _...__.i__. . _ -
r t f Grand Gulf Muclear Station f simulator certification ! Initial Report, March 1991 1 Simulator Test Abstracts l Appendix IX - Malfunction Tests Page 172 of 210 l t [ I i
.i i
Date of last test 5/24/90 Concern (s) found/ Planned resolutions j i Drywell purge compressors did not trip while in test mode i without ssW cooling. Air compressors will run a lot longer l without SSW then without TBCW cooling. These deficiencies- ! will be corrected as system dynamics are remodeled during the : simulator upgrade project, on a schedule to.be completed within 4 years of initial certification. : i I
.i t
1 i
.?
t 1
?
s i k i e i
-, ,f y r -m y 'r-'-w y ' - - --> M + -+&-y u m se r +-7> -m - * - - - - rc -nm--- -----ve*-v *w %" ar -HU- .--w w+ % v r a
- e=r'- *- g--'y =-'
7
Grand Gulf Nuclear'8tation simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 173 of 210 Malfunction Test #149, HPCS Service Water Trip Initial Conditions: 100% power. Sequence of events: Division 3 Diesel Generator is started and loaded to 3300 KW. The HPCS Diesel Generator running alarm is verified. HPCS SSW pump start and discharge valve opening are verified. Malfunction 149 is entered. Automatic actione and alarm response are verified which include the following: Decrease in HPCS SSW flow and discharge pressure. , HPCS Service Water overload / power loss alarm. SSW Division 3 out of service alarm. HPCS Diesel Engine trouble, HPCS-Diesel Engine-trip, and HPCS Generator trip / lockout alarms. HPCS Diesel trip. l The HPCS SSW pump is verified to not manually start. The malfunction is-deleted. The HPCS SSW pump-is verified to successfully restart manually. Date of last test 5/23/90 Concern (s) found/ Planned resolution: None. 1 1
._%,, , ,,.e,.,..g.,#y . . - . . , _ , , _ - . , , . .e , .-., ., . , .
Grand Gulf Muolear station simulator certification ' Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 174 of 210 Malfunction Test #150, Plant Service Wator (PSW) Pump Trip _ ta-F) Initial conditions: 100% power Sequence of events: Radial Well pumps J and K are started using remote functions such that all radial well pumps are operating. (A through F, plus J and K). Malfunction 150 is tested for PSW pumps-A through F, one at a time. With the ralfunction active the following automatic actions and alarm responses are verified: Affected radial well pump trips. PSW Radial Well trouble alarm. Decrease in radial well discharge pressure. Decrease in PSW header pressure. After deleting the malfunction, the affected pump is restarted and parameters verified to return = to normal. After pump A-F have been individually tested, radial well' pumps J--and K are stopped using remote functions. Malfunction 150 is entered for
-pumps A through F simultaneously to simulate aitotal loss of PSW. Automatic actions and alarms response are- verified including the following:
Radial'well incoming feeder breaker trip alarms. Plant chiller trip. alarms. PSW header pressure low and low-low alarms. High temperature alaras- on components and systems affected. Decrease in condenser' vacuum indication and alarms.-- _ _ , _ _ . . _ _ , . ~ .
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 175 of 210 Turbine trip on. low vacuum and trip alarms. RWCU trip alarns. Date of last test 5/23/90 Concern (s) found/ Planned resolution: Instrument and Service Air compressors did not trip when PSW cooling was lost to TBCW. This deficiency has been corrected and retested. I l. I I 9' 9 t _ , _ . . . . . . . . . _ , ._. _ _ . . _ , _ . ._ ,, , _ , , - - . . _ . . . - . , . ~ . _ . - . _ , . , _ ._
Grand-Gulf Nuclear station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix IX - Malfunction Tests Page 176 of 210 , Malfunction Test #151, Component Cooling Water (CCW) Pump Trip ( A , B, C) Initial conditions:-100% power. Sequence of events: Malfunction 151 is tested for each CCW Pump. This verifies the auto start of the standby pump on decreasing CCW dischargo header pressure upon failure of one of the two running. pumps. A partial loss of CCW is tested whereby only one pump romains running. CCW and Recirculation pump motor high tenperature alarms and the various CCW low discharge pressure alarms are verified to actuate. A complete loss of CCW is tested whereby all pumps are tripped. Automatic actions and alarm response are verified-which includes the following: Recirculation motor and seal cooling low flow' alarms. RWCU filter demineralizer high temperature alarms. RWCU pump cooling water high temperature alarm. Fuel Pool Cooling Heat Exchanger low flow alarms.. 1 RWCU isolation. t CCW pump trip alarms. CCW discharge header low pressure alarms. Malfunctions are deleted and CCW pumps are restarted and returned to standby to test other auto start combinations. Date of last test: 8/21/90 Concern (s) found/ Planned resolution: None. t
+ - , p r,e-=.yw,,,---.-e,3,-w,-~,r +- . - - , ,,+ -+, ,e-, r w r-w--- w4. , . - . - , , - , ,-Sv.,----,-swvm~m, a -m -y- , , + - i,-.
Grand Gulf Nuclear Station ; simulator certification
- Initial Report, March 1991 ,
Simulator Test Abstracts i Appendix IX - Malfunction Tests ! Page 177 of 210 ' Malfunction Test #152, Turbine Building Cooling Water (TBCW) Pump Trip (A,B,C) Initial Conditions: 100% power. Sequence of eventes Malfunction 152 is tested for each TBCW pump. This verifles the auto start of the standby pump on decreasing TBCW discharge header pressure upon failure of one of the two running pumps. A partial loss of TBCW is tested whereby only one pump remains running. TBCW discharge pressure indications and low discharge pressure alarms are verified. A total loss of TBCW is tested for failure of all pumps. Automatic actions cad alarms response are verified which include the following: TBCW pump trip alarms. TBCW low discharge pressure alarms. High temperature alarms from components cooled by TBCW. Service air compressor trips.
- Instrument air compressor trips.
i l Failure to manually-start an air compressor. Reactor Scram resulting from a lous instrument air. Malfunction 152 is deleted. TBCW pumps are restarted and returned to standby to test other auto start combinations. Date of last test: 8/21/90 Concern (s) found/ Planned resolution: None.
Grand Gulf Nuclear station simulator Certification . Initial Report, March 1991 simulator Test Abstracts Appendix IX - Malfunction Tests : Page 174 of 210 l l Malfunction Test #153, Loss of Drywell Cooling Fans (A-F) _ Initial Conditions: 100% power. , i Sequence of events: ; Malfunction 153 is tested for each set of drywell cooling ! fans (A-F). With each malfunction active, automatic actions ! and alarm response are verified which includes: r Increasing Drywell temperature in the areas of the tripped fan and damper closure. Increasing drywell pressure and the Drywell High/ Low pressure alarm. j Malfunction 153 is deleted. The drywell cooler fan set is { restarted. Drywell parameters are verified to return to i normal. Alarms are verified to clear. , Date of last test: 4/16/90 Concern (s) found/ Planned resolution: None. l f L I l b
Grand Gulf Nuclear station simulator certification Initial Report, March 1991 simulator Test Abstracts l Appendix II Malfunction Tests Page-179 of 210 @ v Malfunction Test #154 (A,B) Standby Gas Treament System Train (A,B) High Differential Pressure 1 Initial Conditions: 100% power. Sequence of eventst Both trains of standby Gas Treatment are manually initiated. One train is placed in standby. Malfunction 154 is entered to ; create a simulated plugging of the second HEPA filter for the ' operating train. Automatic actions and alarm response are verified which include the following: ! Increased differential pressure on the filter differential pressure recorder on H13-P870. Standby Gas Treatment system train HEPA/ Charcoal filter high differential pressure alarm. Decreasing filter train flow. Increase in enclosure building pressure. Startup of the Standby filter train and associated alarm. Malfunction 154 is deleted. The affected filter train parameters are verified to return to normal. The test is repeated to test the malfunction on the. other i train and to. test the auto start function of the other train in standby. Date of last test: 4/17/90 Concern (s) found/ Planned resolution: None. l i
. - - . . - .- - - - - - . - - - . _ . - . = . - - - - . -
Grand Gulf Nuclear 8tation Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 180 of 210 Malfunction Test #155, Trip of SGTS Exhaust Fan (A,B) Initial Conditions: 100% power. Sequence of events: Both trains of Standby Gas Treatment are manually initiated. One train is placed in standby Malfunction 155.is entered to simulate tripping of the exhaust fan in the running filter train. Automatic action and alarm response are verified including the following: Stoppage of the exhaust fan in the running filter train and associated trip and out of service alarms and status light. Decrease in filter train flow and enclosure building negative pressure. Startup of the Standby filter train and associated alarm. Malfunction 155 is deleted. The affected filter train is verified to return to normal. The test is repeated to test the malfunction on the other train and to test the auto start function of the other train in standby. Date of last test 4/17/90 ; Concern (s) found/ Planned resolution: None. t l f i
.- . - - , , - ,r-, - . ~ . , - - =-4*.-- y e- - - --+m-- + - = = - -v--"e--r ' * ~ ' - - V > " + * * * * -
Grand Gulf Fuelear station simulator certification 4 Initial Report, March 1991 l simulator Test Abstracts H Appendix II - Malfunction Tests-Page 181 of 210 i Malfunction Test #156, off Gas Post Treatment Log Radiation Monitor K601B Fails Downscale Initial Conditions: 100% power. Sequence of events: Malfunction 156 is entered. Automatic actions and alarm esponse are verified including the following: Offgas peat treatment radiation monitor B recorder indicating downscale. Downscale indication on the monitor and recorder. Downscale and Hi Hi Hi/INOP alarms. Malfunction 156 is deleted. Indications are verified to return to normal. Annunciators are verified to clear. Date of last tests 4/17/90 Concern (s) found/ Planned resolution: 4one, i
.__m-__.___-_--,u---.--_---______--__w-m--------,,---a...-u-.-_-------.1. -
Grand Gulf Nuclear Station simulator certification Initial. Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 182 of 210 Malfunction Test #157, Process Radiation Monitor High Radiation (A-S) Initial Conditions: 1-2% power, 400 psig, with mechanical vacuum pumps running. Sequence of events: Malfunctjon 157 is tested for each of_the following process radiation monitors: if7A - Main Steam Line 157B - Offgas Pre Treat 157C - Offgas Post Treat A 157D - O*fgas Post Treat B 157E - Component Cooling Water 157F - Containment Vent 157G - Offgas and Radwaste Building Vent 157H - Garvice Water Ef fluent A 157I - Service Water Ef fluent B ' 157J - Containment and Drywell Ventilation exhaust Channels B and C 157K _ Containment and Drywell Ventilation _ exhaust-
""*7nels A and D 157L liary Building FRA vent exhaust channe? s and.C 157M - Auxiliary Building FHA vent exhaust channels A and D 157N - Auxiliary Building fuel pool exhaust channels l B and C l 1570 - Auxiliary Building fuel pool exhaust channels A and D 157P - Turbine Building Vent 157Q - Fuel. Handling Area Vent 157R - Control Room Vent channels B and C 157R - Control Room Vent channels A and D Malfunction test 158A is performed at 100 psig reactor pressure with 1 to 2% power. All other malfunction tests are performed at 100% power.
i For each malfunction,_ upscale alarms and recorder indication ! are verified to occur for each malfunction. When the malfenction is deleted, alarms are verified to clear and recorder indications are verified restored. 1 i _ _ . --m., -a
Grand Gulf Nuclear. Station simulator certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests Page 183 of 210 For malfunction 157A, a reactor scram, group I isolation, and a trip of the mechanical vacuum pumps are verified to also ocCMr. For malfunc tion 157J, isolation of Division 2 group 7 containment isolation valves are-verified. For malfunction 157K, isolation of Division 1 group 7. containment isolation valves are verified. For malfunction 157L, the following additional automatic actions are verified: Shift of the B Enclosure Building differential pressure recorder to fast speed. Isolation of Division 2 powered-Auxiliary Building ventilation' isolation valves. Standby Gas Treatment system B initiation. For malfunction 157M and 0, the following additional aute: itic isolations are verified: Shift of the A Enclosure Building _ differential pressure l recorder. Isolation of Division 1 powered Auxiliary Building Ventilation isolation valves. j Standby Gas Treatment system A initiation. l l Date of last test: 4/16/90 Concern (s, faund/ Planned resolution: Mal function 157J caused all group 7 isolation valves to close. Malfunction 157K caused all group 7 isolation valves to close. Malfunction 157 L,M,N, 'and O did not cause any-- valves to close. SBGT did not initiate.
l Grand Gulf Nuclear Station i simulator certification ! Initial Report, March 1991 i Simulator Test Abstracts i Appendix.II . Malfunction Tests Page 184 of 210 These deficiencies will be corrected within 4 years of initial certification. The A main steam line radiation monitor was not functional for the test of malfunction 157A. This monitor is currently being upgraded. . The monitor will be tested upon completion of this item which is expected by the end of 1991. The control room HVAC response can not be verified for malfunction 157 since the H13-P855 panel.and the system are not modeled. This has been identified as an exception to ANSI 3.5 standards.
's I
i l I f i
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 i Simulator Test Abstracts l Appendix IX - Malfunction Tests Page 185 of 210 Malfunction Test #158, Area Radiation Monitor Level Change (A01-A24, B01-B24) Initial conditions: 100% power. Sequence of events: Malfunction 158 is entered for each ARM simulated. Severity is varied to simulate an increase in radiation level in the respective area. Automatic actions and alarm response are verified which include the following: Area Radiation panel P844 Trouble alarm on the P680 panel. APM Recorder D21-R600A or B indicating higher than normal. High radiation alarm on backpanel H13-P844. Roset of the Area Radiation panel P844 Trouble alarm on panel H13-680 when the H13-P844 panel- alarm- is acknowledged.
- . Increase in the readings on the ARM monitor on panel H13-l P844 (only 6 are simulated).
Decrease in ARM recorder indication when the malfunction is deleted. l Date of last test: 4/16/90 l Concern (s) found/ Planned resolution: Recorder D21-R600A was inoperative during the test. The recorder is expected be repaired and retested in 1991. No ARM malfunction exists for -high - radiation - in . the -- Post Accident Sampling Station area or .the Technical Support center area. Malfunction 160 can be used to-cause the-appropriate high radiation alarm on panel H13-P844 to simulate a radiation problem in those areas. This is an acceptable alternative' for training.
l Grand Gulf Nuclear Station Simulator certification Initial-Report, March 1991 i l simulator Test Abstracts l Appendix II - Malfunction Tests Page 186 of 210 l Malfunction Test #159, MOV Failure A-E Initial Conditions: 100% power. Sequence of events: Malfunction 159 is tested for each MOV failure which includes' the following: 159A - HPCS Injection Valve E22-F004 159B - RHR Shutdown Cooling suction valve E12-F008 159C - RHR Containment Spray Valve E12-F028B 159D - SSW A Return valve P41-F005A 159E - Low' Pressure Feedwater Heater String. Discharge t valve N19-F040B Each valve is tested and verified to fail to close_or open with the_ respective malfunction active. Malfunction 159 A,B, C, and D are also tested in conjunction with the respective system MOV Test Switch in Test and Normal. l Malfunction 159A is tested to verify the HPCS injection valve failure during a HPCS initiation. Malfunction 159C is tested to ' verify that the Containment Spray valve E12-F028B fails to open during a containment spray. initiation. l For each test, overload or' power loss status-lights, MOV in test status light, system out of service alarms, and valve position indications'are verified to occur as-applicable to each valve. Date of last test: 4/17/90 Concern (s) found/ Planned resolution: None.
Grand Gulf Nuclear Station ! Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix _IX - Malfunction Tests Page 187 of 210 I Malfunction Test #160 Initial Conditions: 100% power. One annunciator from.each grid section of the following panels was selected for this test:. . H13-P601 H13-P680 H13-P807 I H13-P844 H13-P854 , H13-P864 H13-P870 Malfunction 160 is entered to activate each alarm. The alarm is verified to stay in and not clear after acknowledging.it. After malfunction 160 is deleted,- the alarm is verified to clear after reset. Date of last test: 7/26/90 Concern (s) found/ Planned resolution:- Some annunciators were.found to clear and come back in when - acknowledged. This deficiency will be corrected within 4 years of initial certification. l l
-..-w
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 188 of 210 Malfunction Test #161, Control Rod Drift Out Initial Conditions: 100% power. Sequence of events: Fifty control rods are selected at random.- For each rod tested, it is first inserted or withdrawn to position "12". Malfunction 161 is entered for the rod. The' Control rod' drift alarm and status light are verified. Increasing APRM flux level is verified. The Rod Crift pushbutton control on the RC&IS display is verified functional. The drifting rod is selected and inserted 1 notch.-Rod speed is verified to be less than normal with the drift malfunction in. The drift reset function is verified while an insert is attempted. The drift condition is verified'to return when the insert pushbutton is released. After malfunction 161 is deleted, the drif t condition ic verified to stop. The test is repeated for each selected rod. Date cf last test: 5/2/90 Concern (s) found/ Planned resolution: None. 4 l
I Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix IX - Malfunction Tests Page 189 of 210 , Malfunction Test #162, RPS Auto and Manual Failure to Scram Initial Conditions: 100% power. 1 Sequence of events: Malfunction 162 is entered to simulate an electric fault such that RPS automatic and manual Scram signals fail Ro-scram rods. Malfunction 9A and 9B are entered to cause an automatic-scram signal from a neutron monitoring system trip ao verified by alarm indication. The reactor is verified to not scram by- i absence of the Reactor Scram Trip alarm and scram solenoids ! lights on the H13-P680 panel remaining lit. The Division 1,2,3, and 4 manual scram pushbutton-are armed and depressed. The Reactor Manual Scram trip alarm and-absence of a scram are verified. The Re ctor mode switch is placed to shutdown. The absence of a scram is verified. The scram solenoids are deenergizered to cause reactor scram by-opening CB-2A,8A,2B, and 8B with remote functions.. The simulator is reinitialized to 100% power.. Malfunctions 162, 9A, and 9B are entered., The reactor is verified to scram when' malfunction 162-is deleted. The simulator is reinitialized to 100%. Malfunction 162 is entered. The-turbine is. tripped with Malfunction.93. The alternate rod. insertion function ir verified'to cause' rod insertion. Date of last test: 4/7/90
- Concern (s) found/ Planned resolution: None.
l 'I
)
i d I 9 y- ----',y-m-. -
-m -, :% a 7.- , 3 g y4m9 ,.f
L 1 N Grand Gulf Nucleur station simulator Certification Initial Report, March 1991 simulator Test Abstracts Appendix II - Malfunction Tests i Page 190 of 210 Malfunction Test #163, Loss of Condenser Vacuum (Variable) Initial Conditions: 100% power Sequence of events: Malfunction 163 is entered at 46% severity to simulate air inleakage to the condenser. Automatic actions and alarm response are verified which include the following: Increasing condenser;shell pressure and LP condenser shell high pressure alarm. Decreasing condenser vacuum. Offgas panel trouble alarm.. The severity is increased'to 50%. The IP condenser shell high pressure alarm is verified along with indications of additional vacuum loss. ' The severity is increased to 55%. The HP condenser shell high pressure alarm is verified along with indications of additional vacuum loss. The severity is increased to 70%. Indication of additional vacuum loss is verified. The following automatic actions are verified: Main Turbine trip RPFT trip Bypass stop valve closure Date of'last test: 12/11/90 Concern (s) found/ Planned resolution: None. ___._.__.._.____m.-- --.
i Grand Gulf Nuclear Station l Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 191 of 210 Malfunction Test 164A, CRD Hydraulic Block Initial Conditions: 100% power Sequence of events: Malfunctions 164A is entered at 0% severity. This simulates failure of Scram Discharge . volume level annunciation and piping such that there is insufficient volume for a scram. Malfunctions 9A and 9B are entered to cause neutron monitoring system trip and reactor - scram trip alarms. Partial control rod insertion-is verified by both RC&IS panel indication and NSSS process computer edit, OD-7 option 2. Malfunction 9A and 9B are deleted and the scram is reset. The reactor is manually scrammed several more times until all control rods are fully inserted. The test is repeated for 20%, 50% , and-100% severities. Date of last test: 4/17/90 Concern (s) found/ Planned resolution: , The amount of rod motion between successive -scram attempts is l independent of the time between scram attempts. .And-time to I vent and drain the scram discharge volume. This deficiency will be corrected within 4 years of initial certification. l r . r ~~
l 1 Grand Lif Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II --Malfunction Tests Page 192 of_210 Malfunction Test #165, La Salle Power Oscillation event , Initial Conditions: 100% power, beginning of cycle, less than 100% rod line. Sequence of events: Malfunction 165 is entered to simulate power oscillation due to thermal hydraulic conditions when plant parameters are in regions of potential instability. Recirculation flow control valves are closed to minimum position. Control rods are withdrawn one notch at a time. The following automatic actions are verified: Power beginning_to oscillate. oscillation amplitude increasing with each rod withdrawal. Reactor Scram on High neutron flux.- Date of last test: 6/5/90 Concern (s) found/ Planned resolution: None. I 1 l
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test-Abstracts Appendix IX - Malfunction Tests Page 193 of 210. Malfunction Test #166, 120 VAC UPS Trip (A,B,C, and D) Initial Conditions: 100% power. Sequence of events: Malfunction 166 is entered to fail each ESF Inverter distribution panel including 1Y87, 1Y88, 1Y96,'and 1Y95. Automatic actions and alarm response are. verified which include the following: Various alarms and status lights.
~
Half scram indications. Various recorder and indicator failures. Loss of SRM/IRM drive controls. i Loss of SRM/IRM/APRM Bypass status lights. For each inverter failure, a detailed list of alarms, status lights, indicators, component status changes are used to j verify correct response . After the response to each malfunction A, B, C, and D is verified at.100% power, the simulator.is initialized to less than 135 psig RPV pressure. Both loops .of shutdown cooling are placed in service. RHR Head spray is. established. Malfunction 166A is verified to cause a division one isolation of group 3 containment isolation valves. Malfunction'166B is verified to cause a division 2 isolation of group 3 containment isolation valves. l The simulator is reinitialized to a reactor startup with a 50 l second stable period. Malfunction 166 is verified to fail affected SRMs resulting in-an infinite period. i
i l Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 194 of 210 Date cf last test: 12/4/90 ; Concern (s) found/ Planned resolution: Recorder C51-R614 did not fail downscale on loss. IRM Range pushbuttons and range. indication did not fail cs expected. IRM and SRM drives did not fail to drive - as expected. These deficiencies will be corrected within 4 years of initial certification. D l l
1 I Grand Gulf Nuclear Station Simulator certification l Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page-195 of 210 Malfunction Test #169, Recirculation Pump A/B Shaft Shear Initial Conditions: 100% power. Sequence of events: Malfunction 169 is~ entered for each recirculation pump. Automatic. actions and alarm response are verified.which include the following: Decrease in. recirculation pump' differential pressure-drive flow, and amperage. Decrease in jet pump flows in affected loop. Reactor level swell and reactor Hi/ Low level alarm. Power drop to approximately 55%. Load drop. Decrease in 6.9 KV. Bus feeder amperage. Increase in 6.9 KV Bus voltage. The test is repeated to test both pumps. The simulator is reinitialized- to 100% power.- Malfunction.169 A and B are entered simultaneously. ScramLis' verified to occur-i on APRM flow biased thermal flux. A, reactor level swell higher
- than with the malfunction of a' single pump trip is-verified o
to occur. Reactor level is verified to not reach level-8. l Date of last test: 4/18/90 l Concern (s) found/ Planned resolution: None. 3
i Grand Gulf Nuclear Station l Simulator Certification l Initial Report, March 199. Simulator Test Abstracts Appendix IX - Malfunction Tests Page 196 of 210 Malfunction Test #170, Recirculation Pump A/B Seizure Initial Conditions: 100% power. Sequence of events: Malfunction 170 is entered for each recirculation pump. Automatic ; ions and alarm response are _ verified which include the following: Opening of affected motor breaker CB-5. Rapid decrease in affected motor amperage. Recirculation pump overload / trip alarm _for the affected pump. Decrease in pump speed, differential pressure, and loop I drive. flow to zero for the affected pump. Decrease in affected' log jet pump flows. ! Reactor level swell to level 8 and subsequent Scram. l l The test is repeated to test each recirculation loop. i Date of last test: 8/15/90 Concern (s) found/ Planned resolution: Reactor level did not swell to level.8. . No scram occurred. This deficiency will be corrected during the simulator upgrade l project which will involve replacement of the core and i I recirculation system models, on a schedule to be completed within 4 years of initial certification. l
+mir C ry
l Grand Gulf Nuclear Station-simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests Page 197 of 210 Malfunction Test #171, Feedwater Line B Rupture in the Drywell (Variable) Initial Conditions: 100% power Sequence of events: Malfunction.171 is entered at 100% severity. The tronsient is run for 15 minutes. The simulator is placed in freeze. Alarms are acknowledged but not reset. A detailed checklist-is used to verify correct automatic actions and alarm response as drywell pressure increases. Any unexpected responses are noted. Expected responses include the following: ,
- Trip of all condensate and reactor feed pumps.
l High drywell pressure. ATWS-ARI/RPT initiation. Recirculation FCV motion inhibit. Recirculation pump breakers CB1, 2, - 4, and.5 trip. Reactor Scram. Drywell Floor and Equipment drain sump high level. Loss of condenser = vacuum. Initiation of ECCS and RCIC. Auto start of DG11,12, --and 13. Automatic isolations. MSIV closure. Pressure control with SRVs.
Grand Gulf Nuclear Station Simulator Certification Initial Report, March 1991 Biaulator Test Abstracts Appendix II - Malfunction Tests Page 198 of 210 Loss of Instrument air, Plant Service - water, drywell chill water, plant chill water. Auto start of SSW, Standby Gas Treatment. LSS actuation. Level drop and recovery by HPCS. The simulater is reset to lobt power. Power is reduced to-00%. The "B" Feedwater line is isolated. Malfunction 171 is entered at 100% severity. A leak of approximately 200 GPM in the drywell is verified. Date of last test: 7/3/90 4 Concern (s) found/ Planned resolution: The following alarms did not occur as expected: SSW Pump A Low Dischargo Pressure PCW Pump A Trip PCM Pump B Trip SSW Pump B Trip SSW Cooling Tower Fan C Trip SSW Cooling Tower Fan D Trip The above problems have been corrected and retested. Containment Cooling Div 1,4. Rad Hi'lli/INOP Containment Cooling Div 2,3 Rad Hi Hi/INOP Containment Cooling Rad Hi The remaining problems will be corrected within 4 years of initial certification.
Grand Gulf Nuclear Station Simulator Cwrtification Initial Repr.ct, March 1991 Simulator Test Abstracts Appendix IX --Malfunction Tests Page 199.of 210 Malfunction Test #172, Recirculation Loop A/B FCV Runback-Initial Conditions: 100% power ' Sequence of events: Malfunction 172 is entered to cause a runback for each FCV. Automatic actions and alarm response are verified which include the following: Recirculation FCV Partial Close/RFP Trip alarm. FCV-closure to 15% indicated. Core flow decrease to approximately 52%. Power decrease to approximately_65%. Decrease in affected loop driving flow, jet pump flows and motor amperage. Reactor level swell and return to normal. Malfunction 172 is deleted. The - runback is reset and the affected FCV is reopened. The test is repeated for the other FCV. i Malfunction 172A and B are also tested simultaneously. A level swell higher than for a single FCV runback' is verified. Absence of a reactor scram is verified. Date of last test: 4/19/90 l Concern (s) found/ Planned resolution: None. < l _ __ - ~ , . , , . . . . , . . -.v.,y
1 Grand Gulf Nuclear Station simulator certification > Initial Report,. March 1991 Simulator Test Abstracts ! Appendix II - Malfunction Tssts Page-200 of 210 i Malfunction Test #173, Recirculation Loop A/B FCV Fails Increasing Flow Initial Conditions: 100% power Sequence of events: 1 Malfunction 173 is entered to cause a failure of each FCV . resulting in valve opening. Automatic actions and alarm response are verified which include the fo.11owing: FCV opening rapidly. Increase in core flow and reactor power. Increase in affected loop jet flows.and driving flows.- i l Reactor level decrease. Reactor scram on-high. neutron flux.- The test is repeated for the other FCV.- Date of last test: 5/23/90 Concern (s) found/ Planned resolution: None. t t I i
Grand Gulf Nuclear Station Simulator Certification Initial _ Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 201 of 210 Malfunction Test #174, Failure of High- Drywell' Pressure Isolations Initial Conditions: 100%-power. Sequence of events: Malfunction 174, A or B are entered to cause a failure' of high drywell pressure for division 1 or 2 isolations due to a relay failure. All Group 2, 6, 7 and auxiliary building isolation valves are
-opened on panels H13-P601 and H13-P870. RHR valves E12-F037A and F037B are opened. Malfunction 63 is entered at 5% to cause a high drywell pressure signal which__results in ECCS initiations. For malfunction 174A,- division 1 isolation valves are verified to remain open and division 2 valves closing.
For malfunction 174B, division 2 isolation valves are verified
- to remain open and-division l' valves closing.
For each division failure, . a Recirculation FCV inhibit and reactor scram are verified to occur onihigh drywell pressure. NSSS manual isolation is verified to cause the divisional isolation- that failed -to close. The test is repeated to test-the other division. Date of last test: 6/26/90-Concern (s) found/ Planned: resolution: .None. l l 1 1
Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appe'a$ix II - Malfunction Tests Page 202 of 210 I Malfunction Test #175,-Failure of Low Reactor Level Isolations Initial Conditions: 100% power. Sequence of events: Malfunction 175 A or B is entered to cause - a- failure-- of Reactor Level 2 isolation for division 1 or 2. All group 6, 7, 8, and.10 and auxiliary. building isolation-valves are opened on the H13-P601, P870, and P680-panels.-:A reactor level 2 condition is created by tripping both' reactor feed pumps until the HPCS and RCIC systems initiate. A reactor scram is verified to occur at Level 3. Isolation valves-in the failed division are verified to remain open. Isolation valv'es in the other division-are. verified to automatically close. NSSSS manual isolation pushbuttons are' verified to close'to-- isolation valves which failed to close. , The test is-repeated for the other' division.
~
Date of last test: 6/26/90 Concerns.found/ Planned resolution: None. l l l
l l Grand Gulf Nuclear Station simulator Certification Initial Report, March 1991 simulator Test Abstracts-Appendix II - Malfunction Tests Page 203 of 210 Malfunction-Test #176(A,B), 125VDC Bus 11DK, 11DL Trip. Initial conditions: 100%' power. Sequence of events: Malfunction 176A is entered to. trip 125VDC Bus 11DK. Automatic actions and alarm response are verified which-include the following: Loss of voltage on Bus 11DK. Static inverter trouble alarms for 1Y98 and 1Y80 which clear. Af ter ATWS-ARI/RPT is manually initiated on panel : H13-P680, ARI/RPT is verified to fail.'A LOCA signal is then inserted by. manually initiating R!!R A/LPCS. Automatic actions and alarm responsa are verified which include the following: ' Static inverter trouble alarms ~on 1Y98 and 1Y80. Loss of inverter 1Y98 loads. Loss of. inverter 1Y80 loads. Malfunction 176A is deleted. Bus voltage on 1-1DK-is verified to return. Alarms are verified to clear. 1Y98 and 1Y80 loads are verified to reenergize. l l l l
[ Grand Gulf Nuclear Station Simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Halfunction Tests Page 204 of_210 The simulator is reset to 100%. Malfunction 176B is entered to trip 125VDC Bus 11DL. Automatic action and alarm response are verified which include the following: Loss of voltage on Bus 11DL Static inverter trouble alarms for 1Y81 and 1Y97 which clear. A LOCA signal is- then inserted by manually initiating RHR B/C. Automatic actions and alarm response are verified which include the following: Static inverter trouble alarms on 1Y81 and 1Y91 which remain actuated. Loss of SPDS indication. Loss of isolation valve status panel. Malfunction 176B is deleted. Bus 11DL voltage is verified to return. Inverter trouble alarms are verified to clear. 1Y81 and 1Y91 loads are verified to reenergize. Date of last test: 5/25/90 Concern (s) found/ Planned resolution: Static inverter trouble alarms for 1Y98, 1Y80, 1Y98, 1Y80 did not. actuate as expected. The SPDS display was not lost on a loss of 11DL. These deficiencies will be ccrrected within-4 years of initial certification. i
Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Tests-Page 205 of 210 Malfunction Test #177, Failure of Automatic HPCS Initietion Initial conditions: 100% power Sequence of events: Malfunction 177 is entered to'aimulate a relay failure which prevents HPCS automatic initiation. Malfunction 63 is entered at 5% severity until high drywell pressure results in other ECCS initiations. HPCS initiation and auto start of DG13 and SSW C are verified not to occur. The severity of malfunction 63 is increased to 100%. HPCS initiation and auto start of DG13 and SSW C are verified not to occur until the HPCS manual initiato pushbutton is armed and depressed. Date of last test: 6/26/90 Concerns found/ Planned resolution: None. l l' l 1 l 1 e r=w e
l l Grand Gulf Nuclear Station simulator certification Initial Report, March 1991 Simulator Test Abstracts
- Appendix II - Malfunction Tests Page 206 of 210 ,
Malfunction Test #178, 480 VAC Bus 28BG1 Overcurrent Trip Initial Conditions: 100t power, Sequence of events:
- Malfunction 178_is entered with both the normal lineup or with Bus 28BG1 cross connected to Bus 18BG1.
Automatic actions and alarm response are verified- which - include the following: Loss of voltage to 2dBG1 and undervoltage alarm. Loss of current--through 28BG1 feeder breaker. Supply breaker openings (depending on lineup) and associated alarms, i Loss of radial well switchgear indication and control. When 18BCI:is also lost (when cross connected), inability to close incoming breckers with-malfunction active. l When the malfunction is deleted,-28BGl'is reenergized using remote functions. Alarms are verified to clear. Indications are verified-to return to normal. - Date of last test: 12/4/90 Concern (s) found/ Planned rerolution: =None. t
M A_ _. m Grand Gulf Nuclear 8tation Simulator Certification Initial Report, March 1991 Simulator Test Abstracts Appendix II - Malfunction Testr. Page 207 of 210 Malfunction Test #179, (A,B) 125 VDC Bus 11DG/23 DG T: lip Initial Conditions: 100% power. Sequence of events: Malfunction 179 is entered to trip either bus 11DG or 21DG. Automatic actions and alarms response are verified which include the following: Indication loss and loss of control to breakors dependent on bus lost. Loss of Radial well' computer indication for loss of 21DG. BOP transformer 13 or 23 DG control power loss alarm. Malfunction 179 is deleted. Breaker - control' is verified restored. a of last test: 12/14/90
<~ncern(s) found/ Planned resolution: None.
t i w w , e ~ W
l l Grand Gulf Nuclear Station Simulator Certificetion Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 208 of 210 Malfunction Test #180, ESF Transformer 21 Failure / Lockout Initial Conditions: 100% power. Sequence of events: Busses 15AA, 16AB, and 17AC.3re energized from ESF Transformer 21. Malfunction-180-is entered. Automatic. actions and alarm response are verified which i include the following: ESF Transformer 21 trouble and lockout alarms. Incoming feeder breaker trip alarms. Deenergization of Busses'15AA, 16AB, and 17AC. Reenergization of Busses 15AA, 16AB, and 17AC by DGs 11, 12, and 13. Bus undervoltage lockout trip alarme. DG 11/12 Ready to Acad status lights. , HPCS 111esel Runniag alarm. LSS panal status and, amber and red status lights. LSS panel white status lights going out. LSS BUV sequencing.. ' ESF Transformer 11 and 12 are paralleled to each diesel generator. i Malfunction 180 is deleted and the transformer lockout is- - reset with remote functions. -Bus: es 15AA, 16AB, and 17AC , l are manually transferred to all possible ESF power supplies. ! Date of last tent: 12/4/91. [ l f
- Concern (s) found/ Planned resolution
- None..
I l 1
, , < ,e - - - -
l Grand Gulf Nuclear Station Simulator Certification 1 Initial Report, March 1991 Simulator Test Abstracts : Appendix II - Malfunction Tests Page 209 of 210 Malfunction Test #181, Inadvertent SPMU Dump (A/B) Initial Conditions: 100% power Sequence of events: With suppression pool level in the normal band of 18.34 to 18.81 feet, an inadvertent dump test is performed on both A and B Suppression Pool Makeup (SPMU) lines by inserting malfunction 181. Automatic actions and alarm response are verified which include the followines SPMU valves open alarm SPMU valves opening and opening times-Suppression pool high level alarms HPUS and RCIC suction valve automatic alignment to the Suppression pool SPMU dump time of cpproximately 7.5 minutes Affected Fuel Pool Level alarms. After either of A or B SPMU line dump' tests, a dump recovery test is performed. Malfunction 181 is deleted, the SPMU dump-valves are manually closed after placing the divisional SPMU mooe switch to OFF. The RHR - A and. B pumps are started and valver: E12-F037A and B are opened to pump from the suppression pool to the upper containment pool. The time to recover level in the suppression ' pool to normal is verified to be approximttely 13 minutes. After resotting the simulator to 100% power, SPMU valve interlocks are tested in-conjunction with malfunction 181. SPMU valve movement:is verified to be precluded with malfunction 181 active and the SPMU mode switches in OFF. SPMU valves are verified to open automatically with malfunction 181 active and'the SPMU mode switches in Auto. SPMU valve strokes are verified to stop by placing the SPMU mode switches to OFF. 3- +-
\
Grand Gulf Muolear station simulator certification Initial Report, March 1991 Simulator Test Abstracts Appendix IX - Malfunction Tests Page 210 of 210 i The level increase in the suppression pool is verified to stop after placing the SPMU modo switches to off. The level in the suppression pool is verified to continue to increase. Manus 1 valve closure is verified to be prevented by interlocks in this condition. The SPMU mode switches are returned to Auto and SPMU valves are verified to open fully. Manual valve closure is verified to be prevented by interlocks in this condition. Manual valve closure is verified to be allowed only i l after malfunction 181 ir deleted and the SPMU mode switches placed in OFF. Alarms for SPMU out of service, SPMU division 1 or 2 abnormal, SPMU division 1 or 2 valves open and af fected SPMU status lights are verified for correct logic combinations. Date of last test February 13, 1991 . Concern (s) found/ Planned resolution: The time for an inadvertent SPMU dump was excessive at approximately 21 minutes. Time for dump recovery was excessive at approximately 15.5 minutes. These problems will be corrected within 4 years.of initial certification. 4 i i}}