ML101590074
| ML101590074 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/03/2010 |
| From: | NRC/RGN-II |
| To: | Progress Energy Carolinas |
| References | |
| 50-324/10-301, 50-325/10-301 | |
| Download: ML101590074 (301) | |
Text
1.1. The TheCRDCRDsystemsystemisisbeing beingrestarted restartedlAW lAW20P-08, 20P-08,Control ControlRod RodDrive DriveHydraulic HydraulicSystem System Operat ing Proced ure, following a trip of Operating Procedure, following a trip of the runninggCRDthe runnin CRDpump pumpwith withthe thefollowing followingplant plant conditions:
conditions:
2A2ACRD CRDpumppump Running Running C12-FOO2B C12-F002B, , Flow FlowControl ControlValve Valve2B 2B Auto Auto CRD CRD system system flow flow 45 gpm 45gpm C12-PCV-F003, Drive C12-PCV-F003, Drive Pressure Pressure Vlv Vlv Full open Full open The operator The operatorisis directed directed toto throttle throttle the the C12-PCV-F003 C12-PCV-F003 to to establish establish drive drive water water DPDP betwee n 260 and between 260 and 275 psid. 275 psid.
Which one Which one ofof the the following following choices choices correctly correctly completes completes the the statement statement below below asas the the operato r throttle s the C12-P operator throttles the C12-PCV-F003? CV-F0 03?
The C12-F002B The C12-FOO2B will will throttle throttle (1)
(1) and drive and drive water water DP DP will will (2)
(2)
A. (1) open A. (1) open lower (2) lower (2)
B. (1)
B. (1) closed closed (2) lower (2) lower C
C~ (1)
(1) open open (2)
(2) rise rise D. (1)
D. (1) closed closed (2) rise (2) rise
Feedback K/A: 201001 KIA: 201001 A1.03 A1.03 Ability to predic predictt and/or and!or monito r change changess inin parame parameters ters associ associa ted with operat ated operatiing ng the CONTR CONTROL OL ROD DRIVE ROD DRIVE HYDRA HYDRAULICULIC SYSTEM contro controls includi including:
ng:
CRD CRO system flow (CFR: 41.5 145.5)
/45.5)
ROISRO Rating:
RO/SR Rating: 2.9/2.8 2.9/2.8 Objective:
Objecti CLS-LP-O8O ve: CLS-LP bj 6c
-080bj Given plant and CROHS CRDHS conditio conditions, ns, predict the values for the followin following CRDH system parameters:
g CROH CRDHS Total System Flow Rate
- c. CROHS
Reference:
Referen SD-08 ce: SO-08 Cog Level: low Explanation: With the given conditio Explan conditions ns (F003 full open) the drive water pressure will be low. The closing the F003 would reduce the size of the hole in the flowpat of flowpathh thereby raising pressur pressure.e. With the F002 in auto, it would have to open to maintai maintain n the desired flowrate flowrate.. All of the plausibilities deal with the relation the flow control control valve to the pressure control relationship ship of control valve making making any of them possiblpossiblee depending on where the student thinks the valves are.
Distractor Analysi Oistrac Analysis:
s:
Choice A: Plausible because the F002 will open and if the F003 is in a differen differentt portion of the flowpat flowpathh the pressure would drop.
Choice B: Plausible because if the F002 and F003 were in a differen differentt alignme alignmentnt in the flowpa flowpatthh this would possible.
be possibl e.
Choice C: Correct answer, see explana explanation tion Choice 0:D: Plausible because if the F002 was in a differen differentt alignme alignment nt this would be possibl possible, e, i.e. on the drive water header.
Notes Notes Closing the Closing the F003 F003 would would reduce reduce the the size size of ofthe hole inin the the hole the flowpath flowpath thereby thereby raising raising pressure pressure so sothe the F002 F002 would have would haveto open to to open maintain the to maintain the desired desired flowrate.
flowrate.
FIGURE 013-'
FIGURE 8- 1 CD HYDRAULIC CRD HYDA..L{D SYSTEM SYTA
- lifltr;;t"f,;;,ti1i(1fi
,Iflttfl 4it. -
r;_tr cn.v:
.rf'tl'lt,tIiJ'lY l:n~(r~VIJ;;"1 So.*OB Page Ok of 6B Categories KJA:
KIA: 201001 Al.03 Tier / Group:
Group: T2G2 RO Rating:
RORating: 2.9 SRO SRORating:Rating: 2.8 2.8 LP LP Obj:
Obj:
2.9 08-6C 08-6C W ""
Source:
Source: BANK BANK Cog Level:
Cog Level: HIGH \ Category Category 8: 8: Y Y
t1
.~~"
flU
~
FIGURE 08-1 FIGURE 08- 1 CRD HYDRAULIC SYSTEM CRD HYDRAULIC SYSTEM CRD DRIVE CRD PISTON ~
DRIVE PISTON INLETSCRAMVALVE SCRAM INLET ~~!...
~RPS 1><1'"
Y'J F101-t!> ~ ~~DJI --i .~02
~~V'VALVE i
~g~~
I i, CV126 101 - - - ;,
I "I', '"." ~. ,,err
'I 1 D'<I--c-~'. *"' ~
'" /
'T ~20 I I f~PS Ic~F010
~ ~-~ m'i 'r"'J,~-Cy~,," ~
121 I CV127 V139 I ; '" 1$1 I """"" J I 7,
,,l. +
~112~!i~~RPS!
1136 I
Z-~
(4)
TO OTHER HCU'S .... ~--~-l AL F034 I... -+I X'103 COOLING J
- 138 '.
Ii I
I /
I I~MC kLs\--
(1) I IV I !
RPS L _L-_ _ ~~T DRIVE r 104 \I, I~RAM I
~LS Q ~ , ~~
~
~ Ir:l@~-l.'6 ,
~,,~
CHARGING DISCHARGE COR-V1 i;:;;l TO INST ( I DRAIN tx1 1+- L"I '" " -'7 L'~~
E
~
IFJl rfT\ \ ( I~~
~
F004_ ****~*--l. . . . ~ OTHER F 0 0 6 ,i TO/FROM .9 f!J, .
l-W- FI I FOO2A -~---- HCU'S V140 CVF011
! D003A FT -, ,
RO r:~
L~~,!/
i I '"", ._\ ~:
I L I, '--'----~ ~
1_ PRIMARY CONTAINMENT L~ 6 ABOVE CORE PLATE N--
F0028 I _._:: ~ ____----.REACTOR PRESSURE ,
D0038 ~-ert=~~j~~~;_y~A DETAIL A RECIRC PUMP SEALS i I
' b:Xg07 L EQUALIZING VALVES EQUAliZING
-{><~><}~ ><
PCV-4105 I .
~
-= ,',--.-JF007.
- LB~__ 1 F070 F070 I F150A F15OA SET@75psid
- IF071 vr--&-,I F0678 I-~ SET @75 psid
~ !
REJECT ~-HEADE~SATE Q i CONDE L ~
F150B F1508 STABILIZING VAlVES
- . SET@85psid SET @85 psid ISD-08 Rev. 10 p~;61 Page 61 Of6!j of 68
- 2. Which
- 2. Which one one of of the the following following identifies identifies how how the the Reactor Reactor Manual Manual Control Control System System will will be be affected by a total loss affected by a total loss of the of the Uninteruptible Uninteruptible Power Power Supply Supply (UPS)?
(UPS)?
Control rods Control rods (1)
(1)
Control rod position Control rod position (2)
(2)
A. (1)
A. (1) can can bebe inserted inserted using using thethe Emergency Emergency In In switch switch (2) cannot be determined from (2) cannot be determined from any any location location B
B~ (1) cannot (1) cannot be inserted by be inserted by any any method method other other than than scram scram (2) cannot be (2) cannot be determined determined from from any any location location (1) can C. (1)
C. can bebe inserted inserted using using the Emergency Emergency In In switch (2) can (2) can be be determined determined only from from ERFIS ERFIS or or the process process computer computer D. (1) cannot be inserted by any method other than scram (2) can be determined only from ERFIS ERFIS or the process process computer computer Feedback Feedback K/A: 201002 K6.01 KIA: K6.0l Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR MANUAL CONTROL SYSTEM:
Select matrix power (CFR: 41.7/45.7) 41.7 /45.7)
ROISRO Rating: 2.5/2.6 RO/SRO Objective: CLS-LP-07 Obj 12 List the power supplies to the Reactor Manual Control and Rod Position Indication Systems.
Reference:
SD-52 1/ SD-07 Cog Level: low Explanation: This meets the ka by a loss of UPS which is the power supply to the select matrix and then asking how rods can be moved.
UPS provides power to the select matrix. With no rod being able to be selected, the operator can only UPS insert insert rods via aa scram. UPS also provides power to the Full core and four rod displays, which would be lost.
lost. RPIS isis also also lost.
lost. With the rod posiltion posiition indication gone then ERFIS indication gone ERFIS /1process computer will display display show unkown unkown (??)for
(??) for each rod.
Distractor Analysis:
Analysis:
Choice Choice A: A: Plausible Plausible because because thethe Emergency In In switch switch bypasses thethe RMCS RMCS logic, logic, but but there there is is no no power power to to select select aa rod to move.
rod to move.
Choice Choice B: B: Correct Correct answer, answer, see see explanation explanation Choice Choice C: C: Plausible Plausible because because the the Emergency Emergency In In switch switch bypasses bypasses the the RMCS RMCS logic, logic, but but there there isis no no power power to to select select aa rod rod to move. ERFIS/proce to move. ERFIS/process ss computer computer has has power power but but the the RPIS RPIS input input is is lost.
lost.
Choice Choice D: D: Plausible Plausible because because ERFIS/proce ERFIS/process ss computer computer hashas power power butbut the the RPIS RPIS input input is is lost.
lost.
Notes Notes 4.2.1. Vital 4.2.1. UPS Failure Vital UPS Failure complete loss AA complete lass of of Vital Vital UPS UPS povY'er power willwill have have the the following following effects effects (Additional information available (Additional information available in in 001-50.5):
001-50.5):
Reactor Manual Reactor Manual Control Control - Control
- Control rods cannot be rods cannot be moved moved by by normal normal means (scram means (scram function function isis unaffected).
unaffected). Power Power isis lost lost to to the the rod rod position position display panel.
display panel. Full Full core core display display isis lost.
lost. Since Since the the rodrod position position information system information system is is lost the Nls lost the Nis must must be be closely closely monitored monitored to to ensure ensure the reactor the reactor is shutdown and is shutdown and remains remains shutdown shutdown during during any any subsequent subsequent coolciown.
cooldown.
1 SD-52 SO-52 Rev.
Rev. 33 Page 21 of Page 2'1 of 431 43 3.1.1 Rod Selection As was briefly mentioned earlier, a control rod is selected by the operator at the RMCS select pushbutton and relay module. This module, which is located on on P603, t"las has 137 magnetically latched double-pole, double throw pushbutton switches arranged in an overhead view of the core. Power is provided to the switches by VDC power supplies that, in turn, receive power from the UPS. A 28 VOC rod select power switch at PS03 P603 controls the power to the select panel.
SD-07 1 SO-07 Rev.S Rev. 6 Page 10 of 57157 Categories K/A:
KIA: 201002 K6.01 Tier / Group: T2G2 RORating:
RO Rating: 2.5 2.5 SRO SRO Rating: 2.6 2.6 LP Obj: 07-12 07-12 Source: BANK BANK Cog Level: LOW Category Category 8:8: Y Y
- 3. Which one
- 3. Which of the one of following defines the following defines the purpose of the purpose the Rod of the Worth Minimizer Rod Worth Minimizer (RWM) (RWM) lAW Technical Specification lAW Technical Specifications? s?
A. Ensures that A. Ensures that fuel fuel enthalpy enthalpy does does notnot exceed exceed 280 280 cal/gm cal/gm during during aa control control rod rod drop drop accident when power accident when power is ~ 19.1 %.is> 19.1%.
Bb B~ Ensures that Ensures that fuel fuel enthalpy enthalpy does does notnot exceed exceed 280 280 cal/gm cal/gm during during aa control control rod rod drop drop accident when reactor accident when reactor power power isis ~ 8.75%.
8.75%.
C. Ensures that C. Ensures that the the MCPR MCPR remains remains ~1.11
>1 .11,, while while withdrawing withdrawing control control rods, rods, whenwhen power is poweris~ 19.1%. 19.1%.
D. Ensures D. Ensures that the the MCPR MCPR remains remains ~1.11 1 .11,, while withdrawing withdrawing control control rods, rods, when reactor power reactor power is 8.75%.
is ~ 8.75%.
Feedback K/A: 201006 KIA: 201006 K5.01 K5.0l Knowledge of Knowledge of the the operational operational implications implications of the following of the following concepts concepts as as they they apply apply to to ROD ROD MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC)
WORTH MINIMIZER SPECIFIC)::
Minimize clad damage if a control rod drop accident (CRDA) occurs P-Spec (Not-BWR6) (CFR: 41.5 145.3) /45.3)
RO/SRO Rating: 3.3/3.7 Objective: LOI-CLS-LP-07.1 LOl-CLS-LP-07.1 Obj. 11 State the purpose of the RWM
Reference:
TS Bases Cog Level: low Explanation:
OPERABILITY of the RWM, is required in MODES 1 1 and 2 when THERMAL POWER is :5. < 8.75% RTP.
When THERMAL POWER is >> 8.75% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA.
Since the failure consequence consequences s for U02 have shown that sudden fuel pin rupture requires a fuel energy deposition of approximately 425 cal/gm, the fuel damage damage limit of 280 cal/gm provides a margin of safety from significant core damage which would result in release of radioactivity. Generic evaluations evaluations of aa design basis CRDA (i.e., aa CRDA resulting in in aa peak fuel energy deposition of 280280 cal/gm) have have shown that ifif the peak fuel enthalpy enthalpy remains below below 280 280 cal/gm Distractor Analysis:
Choice Choice A: A: Plausible because the RWM because the RWM enforces enforces control control rod movement movement from all rods full-in from all full-in to to the the Low Low Power Power Setpoint Setpoint (LPSP).
(LPSP). (19.1%)
(19.1%)
Choice Choice B: B: Correct Correct answer, see explanation answer, see explanation Choice Choice C: C: Plausible Plausible because because the the RWM RWM enforces enforces control rod movement control rod movement fromfrom all all rods rods full-in full-in to to the the Low Power Setpoint Low Power Setpoint (LPSP).
(LPSP). (19.1%)
(19.1 %) and and the the RBM ensures MCPR RBM ensures MCPR limits.
limits.
Choice Choice D: D: Plausible Plausible because because the the RBM RBM is what ensures is what ensures the the MCPR MCPR limits.
limits.
Notes Notes Control Rod Control Rod Block Block Instrumentation I nstrumentaton B3.32.1 B 3.3.2.1 BASES BASES APPLICABLE APPLICABLE 2. Rod
- 2. Rod WOlth Worth Minimizer Minimizer SAFETY ANAL SAFETY ANALYSES, YSES, LCO, and LCO, and The RWM The RWM enforces enforces the the banked banked position position withdrav withdrawal ...al sequence sequence (BPWS)
(BPWS) toto APPLICABILITY APPLICABILITY ensure that ensure t:he initial that the initial conditions conditions of of the the CRDA CRDA analysis analysis are are not not violated.
(continued)
(continued} analytical methods The analytical methods and and assumptions assumptions used used inin evaluating the the eRDA CRDA are summarized are summarized in in References'11 References 11 and 12. 12. The The BPWS BPWS requires requires that control rods be moved moved in in groups, groups, with all all control rods rods assigned to a specific group group required to be 'Jiithin within specified specified banked banked positions.
positions.
Requirements that the Requirements the control control rod rod sequence sequence is is in in compliance compliance with the BPWS are specified in LCO [CO 3.'1.6, 3.1.6, "RodRod Pattern Control."
Control.
The RWM Function satisfies Criterion 3 of '10 10 CFR 50.36(c)(2)(ii}
50.36(c)(2Xii) (Ref. 3).
microprocessor-based system with the principle The RWM is a microprocessor-based principle task to reinforce procedural control to limit the reactivity worth of control rods under lower power conditions. Only one channel of the RVVM RWM is available and required to be OPERABLE. Special circumstances circumstances provided for in the Required Action of LCO 3. 3.1.3, Control Rod OPERABILITY,"
-1.3, "Control OPERABILITY, and LCQ 3:1.6 LCO 3 1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control Ol}eration pattern not in compliance with the BPWS. As required by these rod pattem conditions, one or more control rods may be bypassed in the RWM or the RWM may be bypassed. However, HO'....ever, the RWM must be considered inoperable and the Required Actions of this LCO followed since the RvVll.~ RWM can no longer enforce compliance with the SPWS. BPWS, Compliance with the BPWS, BP\tVS, and therefore OPERABILITY of the RWM, is required in MODES 1 1 and 2 when THERMAL POWER is ::;; 8.75% RTP.
When THERMAL THERtvlAL POWER is> is > 815%
8.75% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm callgm fuel damage limit during aa CRDA (Refs. 55 and 6). In In
- r. .10DES 33 and 4, all control rods are required to be inserted into MODES into the core:
core; Categories K/A:
KIA: 201006 201006 K5.01 K5.0 1 Tier!
Tier / Group:
Group: T2G2 RO Rating:
RORating: 3.3 3.3 SRO Rating:
SRORating: 3.7 3.7 LP Obj:
LP Obj: 7.1-1 7.1-1 Source:
Source: BANK BANK Cog Level:
Cog Level: LOW LOW Category 8:
Category 8: Y Y
- 4. Unit
- 4. Unit One One isis at at rated rated power power when when thethe RO RO receives receives thethe following following alarms:
alarms:
RHR LOOP RHR LOOP BB SYS PRESS LOW SYS PRESS LOW SUPPRESSION CHAMBER SUPPRESSION CHAMBER LLVL HI/LO VL HIILO Which one Which one ofof the the following following identifies identifies thethe causecause of of these these alarms?
alarms?
Suppression chamber Suppression chamber water water level level isis _---.1.(....:..1 (1))L--_ due due to to improper improper seating seating of of (2)
(2) only.
only.
A. (1)10w A. (1) low (2) RHR (2) RHR HX HX 1I BB Drain Drain To To Suppression Suppression Pool Pool Valve, Valve, EEli -FOl I BB 11-F011 B. (1)
B. (1)10w low (2) Loop (2) Loop BB Minimum Minimum Flow Flow Bypass Bypass To To Suppression Suppression Pool Pool Valve, Valve, E11-F007B Eii-FOO7B c.
C. (1) high (2) RHR (2) RHR HX HX 1I BB Drain Drain To Suppression Suppression Pool Pool Valve, E11-F011 El I -FOl 18B D
D~ (1) high (2) Loop B Minimum Flow Bypass To Suppression Pool Valve, E11-F007B Eli -FOO7B
Feedback Feedback K/A: 203000 KIA: 203000 K1.02 K1.02 Knowledge of Knowledge the physical ofthe physical connections connections and/or andlor causeeffect causeeffect relationships relationships between between RHRlLPCI:
RHRILPCI:
INJECTION MODE INJECTION MODE (PLANT (PLANT SPECIFIC)
SPECIFIC) and and the the following:
following:
Suppression Pool Suppression Pool (CFR: 41.2 to (CFR: 41.2 41.9 //45.7 to 41.9 45.7 toto 45.8) 45.8)
ROISRO Rating:
RO/SRO Rating: 3.9/3.9 3.9/3.9 Objective: CLS-LP-32.1 Objective: CLS-LP-32.1 Obj Obj 66 Identify the Identify the function function of of the Keepfill system, the Keepfill system, the the systems/components systems/components supplied, supplied, and and the the effects effects of of itit becoming inoperable.
becoming inoperable.
Reference:
Reference:
1APP-A-03 Cog Level:
Cog Level: high high Explanation:this meets Explanation:this meets the KA KA because because the RHR RHR system system is is in in aa standby standby lineup lineup itit needs needs to be be kept kept filled to prevent water hammer prevent hammer as itit isis started. this question is is asking asking about about the connection to the torus ifif the keepfill pressure keepfill pressure thenthen is is reduced reduced and is leaking and itit is leaking to the torus.
to the torus.
Keep Keep fill is supplied is supplied to the RHRRHR loop, loop, ifif the F007 F007 (single (single valve in in the flowpath) flowpath) is is leaking leaking by by the the low low pressure alarm would occur while filling the suppression pool so level will be high.
physical connection between the standby RHR loop for injection and the Suppression This identifies the physical pool.
Distractor Analysis:
Choice A: Plausible because the F011 FOl 1 is a drain flowpath, but it has a double isolation valve. This drain path is to Radwaste also.
Choice B: Plausible because the examinee may think that due to head pressure the torus may backfill into depressurized.or that this line may be external to the torus as the drain lines may the RHR line since it is depressurized.or go to radwaste.
Choice C: Plausible because the F011 FOl 1 is a drain flowpath, but it has a double isolation valve. This drain path is to Radwaste also.
Choice D: Correct answer, see explanation
Notes Notes RHR IflOP BB SYS RHR U80P 575 PRESS PRESS LOW LOW AUTO _~.CTI
!>.UTO ACTIONS ONS NONE NONE CAUSE CAUSE 1.
- 1. Keepf ill Station Keepfill Station Pressure Pressure Control Control Valve, Valve, El1-PC'l-F100, Ell-PCV-FlOO, failure failure or or valve lineup valve lineup incorrect.
incorrect.
22.. Discharge header Discharge header not not charged charged or or leaking.
leaking.
3.
- 3. If in If in &.utdown shutdown cooling ccling andand the RPV is the RPV is NOT NOT pressurized, pressurized, discharge discharge header pressure header pressure isis below below alarm alarm setpoint setpoint duedue to to E11-F003B Ell-FOO3B and/or and/or Ell-F048B yalve EII-F048B valve position.
position.
4.
- 4. If operating If operating in full flow in full flow test test mode, mode, discharge discharge header header pressure pressure isis below be alarm setpoint 10.." alarm setpoint when when flO\\l" flow is is increased increased to to near near maximum.
maximum.
S.
S. Circuit malfunction.
Circuit malfunction, OBSERVATIONS OBSERV.~TI ONS 1.
- 1. Keepf ill Station Keepfill Station Pressure Pressure Cont.rol Control Vahre, Valve, E11-PC'!-F100, E1l-PCV-F100, outlet outlet pressure as as read locally locally on on E11-PI-2676 is is less less than 41 psig.
ACTIONS
_~.CTIONS 1.
- 1. Verify that the Loop B l*linimum Verify Minimum Flow Bypass To Suppression Pool Valve, B11-F007B, Valve, E1l-F0078, is closed.
2.
- 2. If the If the Lc..op Loop B t>linimum Minimum Flow Bypass To Suppression Pool Ptol 'lal ve, Valve, Ell-FOO7B, is not properly seated, cycle the EII-F007B, the valve.
valve, Categories KJA:
KIA: 203000 K1.02 Tier!
Tier / Group: T2G1 RO Rating:
RORating: 3.9 SRO Rating: 3.9 SRORating:
LP LP Obj: 31.2-6 Source: NEW Cog Cog Level: HIGH Category 8: Y
FIGURE 17-5 Shutdown Cooling Warm-Up Typical for Both Loops of RHR 300 PSIG V32 422° F 422°F t t F017 F060 B32-F031B RCR LOOP B
, F049 F040 RN t
125 PSIG DRYWELL F04B 325 of
°F - - - - - - - - - - - -r---~El_-_l RX BUILDING TO SAMPLE r--r-II---r..--; STATION PASS SYSTEM DIW 4 FOB3 FOB4 F006A F006B F007
-27"
-27
'<---+/---+/---i A
-31"
-31 F004 F020 TORUS 1 SD-17 SO-17 14 Rev. 14 Page 103 103 of 1271 127
FIGURE 17-2 Suppression Pool Cooling Mode TO FUEL POOL V39 FO16B F016B FOI6A F016A V32 V33 F049 F040 F015A F060A JJO15j, F060B F015B o FOI7A F017A I--t><)-I'-V'I--'-...........,r.-v F017B FO17B RADWASTE FO5OA F050A FO5OB F050B "A" "B" RECIRC RECIRC PUMP PUMP F028A .... F024A F027A F024B F028B SPRAY HEADER i);,14 i);,14 t A WATER LEVEL B t
SUPPRESSION POOL FOD7A FD2DA F048A F048B FOO3A F047A TYPICAL FOD6 *4 F047B FOO3B F004B RHRA RHRB A B HX HX F004C F004D F014A F014B
- FOO2A RHRC RHRD FC02B
- RHR SERVICE RHR WATER F068A F068B SERVICE WATER SD-17 Rev. 14 Page97of127 Page 97 of 127
17-2A FIGURE 17-2A LPCI Mode G4l*F!l36 TO FUEL POOL FD1SB FOl6A V33 F04(I FD4{)
Rll~ F060A FOeoS FOl5B TO F017A F017B RAOWASTE t
Rl47B FOO3B F014B RHR
~~ _ _ _ _L -_ _~~~
FOO2A RHRD FOO2B RHR RHR SERVICE SERVICE ---9+----J WAT FOOBA L.-F068B~--- ~~CE WATER WATER ISD-17 SO-17 Rev. 14 14 Page98of127 Page 98 of 127
- 5. Unit Two
- 5. Unit Two hadhad just just been placed in been placed in Cold Cold Shutdown Shutdown whenwhen off-site off-site power power isis lost.
lost.
Operators are having difficulty re-establishing Shutdown Operators are having difficulty re-establishing Shutdown Cooling.
Cooling.
Which one Which one ofof the the following following parameters parameters must must be monitored for be monitored for determination determination of of aa mode mode change to Hot Shutdown?
change to Hot Shutdown?
A A~ Reactor saturation temperature.
Reactor B. Reactor bottom head B. Reactor head temperature.
C. Reactor recirculation loop temperature.
D. RHR heat exchanger inlet temperature.
D.
Feedback K/A: 205000 G2.04.21 KIA:
Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
Shutdown Cooling System (RHR Shutdown Cooling Mode) 41.7 /43.5 /45.12)
(CFR: 41.7/43.5/45.12)
RO/SRO Rating: 4.0/4.6 Objective: CLS-LP-307-B Obj. 19 ig given plant conditions, monitor cooldown rate per PT PT-01 .7
-01.7
Reference:
AOP-1 AOP-155 Cog Level: High (assessing plant conditions and determining the appropriate indication that is available)
Explanation:
Natural circulation cannot be depended on to provide adequate flow through the bottom head region or the recirculation loops. The recirculation loop suction temperatures and bottom head temperatures therefore cannot be utilized for vessel coolant temperature monitoring for indication of boiling. Under natural circulation conditions, reactor vessel pressure must be monitored for coolant temperature determination. (AOP-15.0).
(AOP-15.0). The logic is that in a saturated system pressure is equivelantto equivelant to the saturation temperature.
Distractor Analysis:
Choice A: correct answer, see explanation Choice B:B: Plausible Plausible because because under other circumstances (RWCU under other (RWCU isis in in service) this is is aa viable option.
Choice C:C: Plausible because because under under other other circumstances (recirc is is in in service) service) this is is aa viable viable option.
Choice Choice D:D: Plausible because under under other circumstances (SDC is in in service) this is is aa viable option.
Notes Notes CAUTION CAUTION Natural circulation Natural circulation can can NOT NOT bebe depended depended on on to to provide provide adequate adequate flow flow throuQh throuqh the the bottom head bottom head reQion reqon or or the the recirculation recirculation loops.
loops The The recirculation recirculation loop loop suction suction temperatures and temperatures and bottom head temperatures therefore can bottom head NOT be can NOT be utilized utilized for vessel coolant temperature monitorinQ coolant rnonitorinq for indication indication of boilinQ.
boilincj LftJIt1tI&i conditions, re conditions, '~"~j
~BfMo. If cGaflt temperature was initially less ess than pressure must tie an 212 F, ptessure'l1lust be closely monitored clo for indications of
- !gr~~~~qr:)(ldi(;Cltions of aa trend trend ofof increasing increasing pressure.
pressure. IfII this this trend is is established, est it must be assumed that 212°F has been exceeded, boiling is occurring,
............... ,'ifl11ust occurring, and a mode change has '~ken andia".Il1Ode;chahgetlas,. taken. plac~
place...
QAOP-15.O IOAOP-15.0 Rev.
Rev.. 23 Page Page 4 of21 of 21 I 3.3.11 3.3. 0 Reactor Recirc Pump running AND loop is NOT isolated from the Reactor, THEN use Recirculation Suction Temperatures read on B32-TR-R650.
832-TR-R650.
3.3.2 RHR Pump is running in Shutdown Cooling mode l,vith with the Heat Exchanger aligned as follows:
- 1. RHR HX in service: Use RHR HX 2A(8) 2A(B) Inlet Temperature as read on E41-TR-R605 Point 01(2), 1(2), on Panel H12-P6140 H12-P614.
3.4 Bottom Head temperature during heatup and cooldown may be determined in a a nLlmber number of \'v'ays ways depending on tile the status of the RWCU System and the Reactor Recirc Pumps 2PT-0i .7
!2PT-01.7 Revo Rev. 7 Page 44 of 11 !
Categories K/A:
KIA: 205000 G2.04.21 Tier / Group: T2G1 RO Rating:
RORating: 4.0 SRO Rating:
SRORating: 4.6 LP Obj:
LPObj: 307B-1G 307B-IG Source: BANK Cog Level: HIGH Category Category 8: Y
- 6. Unit One
- 6. Unit One isis at at rated rated power power when when aa steam steam lineline break break causes causes thethe temperatures temperatures inin the the ECCS pipe EGGS pipe tunnel tunnel to exceed 190°F to exceed two minutes 190°F two minutes ago.
ago.
Which one Which one ofof the the following following identifies identifies the the current current status status of of the the Group Group 44 and and Group Group 55 isolation valves?
isolation valves?
A-:' Group 44 valves A Group valves closedclosed only.
only.
B. Group 55 valves B. Group valves closedclosed only.
only.
G. Both Group C. Both Group 44 and and Group Group 55 valves valves closed.
closed.
D. Neither Group D. Neither Group 4 nor nor Group Group 55 valves valves closed.
closed.
Feedback Feedback K/A: 206000 K4.02 KIA: K4.02 Knowledge of HIGH Knowledge HIGH PRESSURE PRESSURE COOLANT COOLANT INJECTION INJECTION SYSTEM SYSTEM design feature(s) feature(s) and/or andlor interlocks which interlocks which provide provide for for the following:
the following:
System isolation: BWR-2,3,4 (CFR: 41.7)
RO/SRO Rating: 3.9/4.0 Objective: LOI-CLS-LP-012-A LOI-CLS-LP-O1 2-A Obj 6 Given plant conditions, determine if a Group Isolation should occur.
Reference:
SO-12 SD-12 Cog Level: low Explanation:
RCIC Group 5 has 27 min mm time delay to allow HPCI isolation to occur and possibly isolate the leak leaving RCIC available.
Distractor Oistractor Analysis:
Choice A: Correct answer, see explanation Choice Choice B: Plausible because it may be thought that the time delay is on the Grp 4 insteadinstead of the Grp Grp 5.
Choice C: Plausible ifif they do not apply any do not any time delays.
delays.
Choice Choice D:0: Plausible ifif they they do do apply the the time time delay delay to to both both groups.
groups.
Notes Notes ISOLATION ISOLATION ISOLATION ISOLATION TRI TRIPP SETPOINT SETPOINT NOTES NOTES GROUP GROUP SIGNAL SIGNAL Tech Spec.
Tectl Spec. Act~
Actual Allowable Allowable 4Note11}
~fi)t*
Value Group 4 High Steam Flow High Flow 5275%
275% 220% Note Note 55 Steam Pressure Low Steam Pressure ~ 104 psig 104 psig '1'15 115 psig psig High Turb Tuib Exh Pressure 9 59 psig psig 7 psig Steam Line Area Hi Temp 5200')F 200°F '165°F 165°F Steam line Steam Line Tunnel High 200°F 5200°F -165°FtI90°F 1 65°F/I 90°F Amb Temp AmbTemp Line Tunnel dT High Steam line 50°F 550°F 4?OF 47°F Equip Area High Temp 175°F 5175°F -165°F 165°F Group 5 High Steam Flow 5275%
275% 220% Note 5 Low Steam Pressure LO\v ~ 53psig S3psig 70 psig Exh Pressure High Turb Exll 566 psig 5 psig Steam Line Area Hi Temp s 175 175°F Q F 165 165°F Q
F Note 4 Steam Line Tunnel High S 200:'F 200°F 165':'Ft190°F Note 4 165°F/190°F Amb Temp AmbTemp Steam Line Tunnel dT High S 50 50°FCo F 47°F Note 4 Equip Area High Temp s 175 175°F Q F 165°F EquipAreadTl-ligh Equip Area dT High 50°F
< 50°F 4?OF 47°F 1:
Note 'j: Actual values from TRM All "Actual" Note 2: high level is calculated in accordance with the Otfsite Stack radiation higllievel Ofisite Dose Calculation Manual.
Note 3.:
3: After a 28.5 minute time delay Note 4: After 27 minute time delay delay Note 5: After a 5S second time delay nkNote 6:
"Note Specific Actual "Actual" values from EOP User'sUsers Guide, Attachment Attacllment -11 Categories KJA:
KIA: 206000 K4.02 Tier / Group: T2G1 RO Rating:
RORating: 3.9 3.9 SRO Rating:
SRORating: 4.0 LP Obj:
LPObj: 12-6 12-6 Source:
Source: BANK Cog Cog Level:
Level: LOW Category Category 8: 8: Y
- 7. AA Dual
- 7. Dual Unit Loss of Unit Loss of Offsite Offsite Power Power occurs occurs with DG1 under with DG1 under clearance clearance and and the the following following electrical plant lineup:
electrical plant lineup:
44 kV E-Busses kV E-Busses Energized from Energized their respective from their respective available available DGs DGs 480 VV E-Busses 480 E-Busses E5 and E5 and E8 E8 only only are are de-energized de-energized Then aa LOCA Then LOCA signal signal isis received received on on Unit Unit Two.
Two.
Which one Which one of of the the following following correctly correctly completes completes the the statement statement below below concerning concerning thethe ability of Unit ability of Two Core Unit Two Core Spray Spray to restore reactor to restore reactor water water level?
level?
The Core Spray The Core Pump in Spray Pump in (1) running and (1) running and injection injection is is available available through through the the (2) Core (2) Core Spray Spray Inboard Inboard Injection Injection Valve.
Valve.
A. (1)
A. Loop A (1) Loop A only only isis (2) 2E21-F005A (2) 2E21-FOO5A B. (1)
B. (1)LoopBonl Loop B onlyyis is 2E21-FOO5B (2) 2E21-F005B C
C~ (1) both Loops are (2) 2E21-F005A D. (1) both Loops are 2E21-FOO5B (2) 2E21-F005B
Feedback Feedback K/A: 209001 KIA: K3.01 209001 K3.01 Knowledge of Knowledge of the the effect effect that that aa loss loss or or malfunction malfunction of of the the LOW LOW PRESSURE PRESSURE CORE CORE SPRAY SPRAY SYSTEM SYSTEM will have will have on on following:
following:
Reactor water Reactor level water level (CFR: 41.7/45.4)
(CFR: 41.7/45.4)
RO/SRO Rating:
RO/SRO Rating: 3.8/3.9 3.8/3.9 Objective: CLS-LP-18 Obj.
Objective: CLS-LP-18 Obj. 13b 13b List the List the power power supplies supplies for for each each of the following of the following Core Core Spray Spray System System components:
components:
- b. MOV's
- b. MOVs
Reference:
SD-18
Reference:
SD-18 Cog Level:
Cog Level: High High Explanation: This meets the KA because the power loss is is causing a loss of CS and then determining what loops are available to inject (raise reactor water level) based on this loss.
A Core Spray Initiation Initiation Signal is present present and power is available to both Core Spray pumps, E3 E3 and E4 are energized. With the power loss to E8, MCC 2XD will not have power and the B B loop Core Spray valves will be de-energized. So both both pumps pumps would be running and injection would only be be available from A Loop Loop of Core Spray.
Distractor Analysis:
Choice A: Plausible because loop A pump is running and the injection path is available through A loop.
Wrong because the B loop pump is also running.
Choice B: Plausible because the loop pump has power but the injection valve does not. May think that valve power comes from the opposite unit same division similar to the RHR arrangement. The configuration of RHR pumps has power from the opposite unit for the pumps so with a loss of E1 El makes a loss of CS pump A plausilbe.
Choice C: Correct answer, see explanation Choice D: Plausible because both pumps do have power but the B injection valve does not. May think that valve power comes from the opposite unit same division like RHR does.
Notes 3.2.1 Automatic Initiation Initiation The following auto initiation signals will cause the Core Spray System to operate as necessary to perrormpeliorm its intended tunction:
function:
Reactor Vessel Low Tect"'!. Spec.
Tech. 13"
> 13 Water Level 3, LL3 Actual 45" 45 OR High Drywell High Spec.
Tech. Spec <1.8 psig 1.8 psig Pressure Pressure Actual I1.7
.7 psig AND Reactor LowLow Spec.
Tech. Spec. 402 psig 2: 402 pSlg Pressure Pressure &&
< 425 psiQ
<_425_psIg Actual Actual 410 psig 410 psig
ISO-18 SD-18 Rev. 44 Rev. Page 2'1 Page 21 of 531 of 53 The inboard The inboard and and outboard outboard isolation isolation valves valves are are pOvvered powered from from the the motor control following motor control centers:
centers:
Valve Number Number Mcc fvlCC Compartment Compartment Number Number E21 -FQO4A E2'!-FOO4A 1XC(2XC) 1XC(2XC) OTO DTO E21 -F0046 E21-FOO4B I XD(2XD) 1XO(2XO) OW5 DW5 E21 -FOO5A E21-FOO5A 1 XC(2XC) 1XC(2XC) OT1 DT1 E21 -FGO5B E21-FOO5B 1XD(2XD) 1XO{2XO} OW6 DW6 FROM SWGR E-4 V
J%A)
SLJSTAT1ON E-8 SU8STATION J I I MeC 2XD MCC 2XM
~
~
t......A:..A..J MCC MCC MCC MCC MCC
~ 2XB DGD 1XK 2PB 2XH
~ STBY E11 STBY E1i2
Categories Categories K/A:
KIA: 209001 K3.01 209001 K3.01 Tier Tier/Group:
/ Group: T2G1 T2G1 RO Rating:
RORating: 3.8 3.8 SRO Rating:
Source: NEW NEW Cog Level:
Cog Level: HIGH HIGH Category 8:
Category 8:
Emergency Bus E1/DG1 EIIDGI conditions are:
DGI DG1 No Load Light lit DGI DG1 Available Light lit Bus E1 El UUndervoltage ndervoltage Alarm sealed in Which one of the following identifies the cause of the above indications?
A. DG1 failed to reach rated speed.
B. DG1 failed to reach rated terminal voltage.
C. Substation E5 feeder breaker failed to trip.
D~
D Blown control power fuses for CS Pump 1 1A.
A.
Feedback K/A: 209001 K3.03 KIA:
Knowledge of the effect that a loss or malfunction of the LOW PRESSURE CORE SPRAY SYSTEM will have on following:
Emergency generators (CFR: 41.7 41.7/45.4) 145.4)
RO/SRO Rating: 2.9/3.0 Objective: CLS-LP-39 Obj. 12 Given plant conditions, determine if permissives are satisfied for the DG output breaker to close.
Reference:
SD-39 Cog Level: high Explanation:this meets the KA because the malfunction (loss of control power to the CS pump) causes the DG not to be able to automatically close onto the E-bus.
CS pump 1A is powered from bus E1 El and must load strip prior to EDG #1 OIP0/P breaker closure.
Distractor Analysis:
Choice A: Plausible because the output breaker will not close if this condition is not met, but rated speed is 514 rpm.
Choice B: Plausible because the output breaker will not close if this condition is not met, but the No Load light is lit which is from speed and voltage.
Choice C: Plausible because E1 El is required to load strip before the output breaker will close, but E5 is not one of the breakers that will load strip.
Choice D: Correct answer, see explanation
Notes Notes Automatic closure Automatic closure ofofthe the Diesel Diesel output output breaker breaker onto onto its its respective respective emergency bus emergency bus will will occur occur if:it
NORMAL.
- The Diesel The Generator isis operating Diesel Generator operating at at proper proper voltage voltage and and frequency.
frequency.
- No electrical No electncal faults exist on faults exist on the the Diesel Diesel Generator.
Generator
- Undervoltage on Undervoltage on tile the E E bus exists.
bus exists.
- All E All E bus loads have bus loaels have been been stripped s1pped (with (with the the exception exception of of the the 480 VAC 480 VAC Substation)
Substation) the the Slave Slave breaker breaker isis open open and and nono cross cross tie tie breakers are breakers are closed.
closed.
SD-39 1 8D -39 Rev. 10 Rev. 10 Page39ofi25 Page 39 of "1251 Unit Available Unit Available Run.ning Running Indicates the diesel is running at rated speed and voltage 1A1th Mth the olltput output breaker open.
breaker Categories Categories K/A:
KIA: 209001 K3.03 Tier / Group: T2G1 RO Rating:
RORating: 2.9 SRO Rating:
SRORating: 3.0 LP Obj:
LP 39-12 Source: BANK Cog Level: HIGH Category 8: Y
- 9. Which one of the following identifies the relationship between the SLC system and Core Spray Line Break Detection differential pressure instrument?
The (1) leg of this DP instrument senses _-->(=2) (2)___ core plate pressure via the SLC/Core Differential Pressure penetration.
A. (1) variable (2) below B. (1) variable (2) above C. (1) reference (2) below D
D~ (1) reference (2) above Feedback K/A: 211000 K1.01 KIA: Kl.0l Knowledge of the physical connections and/or andlor cause effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following:
Core spray line break detection: Plant-Specific 41.9// 45.7 to 45.8)
(CFR: 41.2 to 41.9 RO/SRO Rating: 3.0/3.3 Objective: CLS-LP-18 Obj. 10 Explain the prinCiple principle of operation of the CS Line Break Detection Instrumentation
Reference:
SD-18 Cog Level; low Explanation:
This system is comprised of a differential pressure detector which provides Control Room annunciation on detected high DP. The high pressure reference leg of this instrument is exposed to above core plate pressure via the SLC/Core Differential Pressure penetration. The low pressure of this instrument is normally exposed to above core pressure via the Core Spray injection line. This results in the instrument normally measuring core DP (not including core plate DP).
Distractor Analysis:
Choice A: Plausible because the examinee may confuse the reference and variable legs and SLC does discharge below the core plate Choice B: Plausible because the examinee may confuse the reference and variable legs Choice C: Plausible because it is the reference leg and SLC does discharge below the core plate.
Choice D: Correct answer, see explanation
Notes Notes This system is This system is comprised comprised of of aa differential differential pressure pressure detector detector which which provides Control provides Control Room Room annunciation annunciation on on detected detected high high LlP.
P. TheThe high high pressure reference pressure reference legleg of this instrument of this instrument is is exposed exposed to to above above core core plate plate pressure via pressure the SlC{Core via the SLCiCore Differential Differential Pressure Pressure penetration.
penetration. TheThe 1m",
low pressure of pressure instrument is of this instrument is normally normally exposed exposed to above above core core pressure pressure via tile via the Core Spray Spray injection injection line.
line. This This results results in in the the instrument instrument normally normally measuring core measuring core APP (not including core (not including core plate plate LlP).
P).
A break A in the break in Core Spray the Core injection line Spray injection line bev.....een the between the reactor reactor vessel vessel penetration and the core shroud would expose the low pressure penetration pressure side of the instrument to the lower pressure of the region outside the shroud.
would be sensed as an increased differential pressure This would pressure and Control Room Room annunciator would alert the Operator.
Operator Although other indications would be be available, tl1is this alarm would also indicate indicate a break in in the line E21-F0068(A) check valve and the reactor vessel between the E21-FOOB8{A) penetration..
penetration.
The Core Tile Core Spray pipe break detection instruments are located on op the Reactor Building 20' 20 elevation.
1SO-18 SD-18 Categories Rev.
Rev.44 I Page 29 of 531 Page29of53 K/A:
KIA: K1.01 211000 Kl.Ol Tier / Group: T2Gl T2G1 RO Rating:
RORating: 3.0 SRO Rating:
SRORating: 3.3 LP Obj: 18-10 Source: BANK Cog Level: LOW Category 8: Y
- 10. Which one of the following identifies which EPA breakers that will trip on a loss of 10.
480 VAC Substation E7?
RPS MG set (1) EPA breakers EPA (2)
(1)A A. (1) A (2)11 &
(2) &2only 2 only B. (1)
(1)BB (2)3&4only (2) 3 & 4 only C(1)A C~ (1) A (2) 11 & 2 and alternate source EPA breakers 5 & 6 D. (1)
(1)BB (2) 3 & 4 and alternate source EPA breakers 5 & 6 Feedback K/A: 212000 K2.01 KIA:
Knowledge of electrical power supplies to the following:
RPS motor-generator sets (CFR: 41.7)
RO/SRO Rating: 3.2/3.3 Objective: CLS-LP-03 Obj 18a1 8a State the power supplies for the following:
Reference:
SD-03 Cog Level: Low Explanation:
Power for the Motor Generator Sets is tapped off two phases of the normal 480 VAC MC 11 CAl1 CA/l CB GB (2CA/2CB) power supply for the motor through a stepdown transformer (480V to 120V) from E5/E6 (2CAl2CB) E51E6 (E7/E8). Selectable reserve power to the Bus is provided from 120 VAC 1 I E5(2E7) or 11 E6(2E8), and is normally selected to Division I. In the event that either RPS M-G Set fails to operate, the alternate power source must be manually selected.
Two EPAs in series are installed downstream of the generator output breaker for each Motor Generator Set and the alternate power supply for the RPS buses. Bus A is protected by EPA-i EPA-1 and -2; Bus B by EPA-3 and -4. Alternate power is protected by EPA-5 and -6 and-6 Distractor Analysis:
Choice A: Plausible because A MG set is lost along with EPA breakers 1 1 & 2, but these are not the only EPA breakers to trip.
Choice B: Plausible if the examinee picks the wrong power supply and EPA breakers 33 & 4 are powered from RPS MG Set B.
Choice C: Correct answer, see explanation Choice Choice D: Plausible if the examinee picks the wrong power supply and EPA Choice EPA breakers 33 & & 4 are powered from RPS MG Set B.
z Notes I
I RPS PowrSuppy
- 0
"'I!
I c
Categories KIA: 212000 K2.01 Tier / Group: T2Gl I
cIc,Dc.) (Th RORating: 3.2 SRORating: 3.3 LPObj: 03-18A Source: BANK Cog Level:
C LOW Category 8: Y cc
- 11. A
- 11. A control control rod rod is is notched notched out from position position 12.
- 12. The The operator operator observes observes the 12 12 indication indication on the four rod on rod display go go out, out, come come back back on, andand then go out out again.
again.
The operator then observes the 13 13 indication indication come come on then go out. out. No No additional rodrod position is position is displayed on the four rod rod display.
display.
Which one of the following identifies the rod position that will be displayed on the RWM? (assume no additional operator action)
A. Position 12 in inverse video.
B. FF with no inferred position.
c.
C. Position 14 in inverse video.
D D~ FE with an inferred position of 14.
FF Feedback K/A: 214000 A3.04 KIA:
Ability to monitor automatic operations of the ROD POSITION INFORMATION SYSTEM including:
RCIS: Plant-Specific (CFR: 41.7/45.7) 41.7 / 45.7)
RO/SRO Rating: 3.5/3.8 Objective: CLS-LP-07.1 Obj. 8 Explain how control rod position is inferred and substituted in the RWM SD-07. 1
Reference:
SO-07.1 Cog Level: low Explanation: For Brunswick the rod contol system is RWM which supplies rod blocks and such and indications of the selected rod and position of that rod.
On a rod withdrawal if the even notch position (in this case 14) is failed, As long as RWM detects the previous odd reed switch (13) RWM will provide an inferred position of 14 since RWM also receives data from RMCS that the operator initiated a withdraw motion. If there is an inferred position available, it will not be automatically substituted into RWM. A substitute rod position will be displayed on RWM in inverse video.
Distractor Oistractor Analysis:
Choice A is incorrect since 12 12 will not be displayed but is plausible since position 1212 was the last good even position sensed by RWM Choice BB is incorrect since position 13 13 was detected but is plausible since no inferred position would be available if 1313 failed or if no rod motion command was sensed by RWM and the last sensed reed switch was odd Choice C C is incorrect since is incorrect since RWM RWM will not substitute inferred position not automatically substitute position but but plausible since since the the is actually rod is at 14, actually at 14, and and this this would be be the the display display once once the the operator accepts the inferred operator accepts inferred position position Choice Choice D 0 is is correct, see see explanation explanation
Notes Notes 2.
- 2. Operator-Driven Rod "Operator-Driven Rod Motion" Motion Case Case Operator-driven rod Operator-driven rod motion motion can can result result in in an an unknol;vn unknown rod rod position position (Fr on
("FF" on the the RWM RWM screen) screen) from from either either anan electrical/reed electrical/reed switch switch failure failure or an or an interruption interruption ofof the the timer.
timer. Only Only the the first rst ofof these these need need bebe addressed as addressed as the the second second condition condition results results in in only only anan intermittent intermittent unknown position unknown position indication.
indication. InIn the the presence presence of of rod rod motion motion the the RWfvl RWM will offer will offer an an inferred inferred position position based based on on the the follo\lving following rules:
rules:
RWM I;vill RWM will recall recall the the last last known known position, position, andand
- IfIf odd, on an insert insert motion, motion, offer the nextnext more more inserted inserted even notch.
notch.
- IfIf odd, on a \vithdraw withdraw motion, offer the next more withdrawn even notch.
notCh.
- 11 T rdi&s of direction of motion, offer no inferred position.
1 SD-07.1 SO-07.-1 Rev. 7 Page 63 of 1251 125 Categories K/A:
KIA: 214000 A3.04 Tier!
Tier / Group: T2G2 RU RO Rating: 3.5 SRU Rating:
SRORating: 3.8 LP Ubj:
Obj: 7.1-8 Source: BANK Cog Level: LOW Category 8: Y
- 12. A plant
- 12. plant startup is is in in progress progress when aa low low voltage on on the the high high voltage power power supply supply for for IRM G IRM G occurs.
occurs. AlllRMs All IRMs are are on range range 1. 1.
Which one of the following identifies identifies the impact impact of of this condition condition and and the action required to clear required clear the alarm(s)?
alarm(s)?
The expected plant response is a (I)
(1) and the action required to clear the cause of the alarm(s) is placing the _-.\(=2)L-_ (2) ----
A. (1) rod block only (2) joystick on P603 for IRM G to Bypass B. (1) rod block only (2) operate switch on the IRM G drawer to STANDBY Cv C~ (1) rod block with a half scram (2) joystick on P603 for IRM G to Bypass D. (1) rod block with a half scram (2) operate switch on the IRM G drawer to STANDBY Feedback K/A: 215003 A2.02 KIA:
Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM)
SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
mop condition IRM inop (CFR: 41.5 /45.6)
/ 45.6)
RO/SRO Rating: 3.5/3.7 Objective: CLS-LP-09.1 Obj. 13c Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event:
mop alarm
- c. IRM Inop
Reference:
IAPP-A-05, 1APP-A-05, SD-9.1SO-9.1 Cog Level; High Explanation: A loss of power to the high voltage supply is an mop Inop trip of the IRM. In order to clear the cause of this alarm per the annunciator procedure would be to place the IRM in bypass using the joystick joystick on the P603 panel. The question can not state lAW lAW the procedure because that would give the answer to the first part of the question.
Distractor Oistractor Analysis:
Choice A: Plausible because because the the actions actions are are correct and and rod block block will occur, occur, but but also also aa half half scram will occur.
occur.
Choice Choice B: B: Plausible becasue becasue aa rodrod block block will occur occur and and some some components placing itit in in standby standby will remove thethe signal from the the trip circuit (i.e.
(Le. standby gasgas in in standby standby removes removes the the train train from from the the logic).
logic).
Choice Choice C: C: Correct Correct answer, answer, seesee explanation explanation Choice Choice D: 0: Plausible Plausible becasue becasue these these will will occur occur and and some some components components placing placing itit in in standby standby will will remove thethe signal signal from from the the trip trip circuit circuit (i.e.
(Le. standby standby gasgas in in standby standby removes removes thethe train train from from the the logic).
logic).
Notes Notes
IRt-1 UPScALE/
IRN A UPSCALE / INOP AUTO ACTIONS 1.
l Rod ~ithdra~al withdrawal block (bypasaed (bypassed when reactor mode switch is in RUN)
RUN)..
- 2. Reactor half half-Scram (bypassed ...
-Scram (b:i'passed when
- hen reacter reactor mode switch is i inth RUloO.
RUN:I.
CAUSE
- 1. IRM Q'lannel IRl-t Channel (a)(s) A, C,C E, o.r or G indicat indicatinging greater than er or equal to 117 Ol'lon the 0-125 O-12S scale.
- 2. IRM Q'lannel IRl-t Channel(s) (a) A, C, E, or G inoperative signals:
a
- a. IP~
lEN drawer selector s~itch switch not in operate.
- b. lEN drawer module unplugged.
- c. IP~
lEN detector high voltage power supply lew low voltage.
3
- 3. IRM .;,
IRl-1 E, or G detector failure.
A, C, E,er failure, 4
- 4. Improper ranging of IR?'1 Impl.*oper lEN A, C, E, or Grange range awit chea during switches durinq reactor startup o.r shutdown.
or ahutdown.
S.
- 5. Circuit malfm'lction inalfunct ion..
- CTIONS (Cc.ntinued)
.ACTIONS (Ccnt inued)
S.
- 5. If the alarm still exists and Olle one channel is affected, perfol."ID perfoi-m the follO"willg~
following
- a. Refer to Tech Specs and TRM TEN for IRM lEN ChCUlll.el channel cperability operability requirements requiremellts.
- b. Unit sro.
Notify the Ullit SCO.
- c. Bypass the affected channel using the IRM IRN bypass switch.
d
- d. Reset the reactor half-Scram si91'laI. signal.
6
- 6. If IIRNRl'l detector failure o.r or circuit malfunction malfullctiatl. is suspected, ensure that aa 'ii'/R W/R is prepared prepared.
NEUTRON t1ON SYS TRIP NEUTROl-J MON AUTO ACTIONS ACTIONS l
- 1. If alarm is initiated by IRl-ls IRMa and both RPS trip systems are affected, a a reactor Scram will .:>ccur. occur
- 2. If alarm is init iated by IF.Ms initiated IF-Ms and only one RPS trip system is affected, a a reactor half-Scram will occur, OCC1.lr.
3
CAUSE l
- 1. Any lENIRM channel upscale!
upscale/inop mop trip (bypassed when ~hen the reactor mode switch is in RUN) *
- 2. A ccnOination combill.ation of any two two. unbypassed APPJ4s APF.Ms with an upscale upacale or mop inop trip.
- 3. A conibination cembination of any two t ...*o unbypassed OPRN OPRM channels ti-ipped, tripped.
- 4. Any voter po~er supply vo.ter power supplJ:" failure, failure.
TABLE AGLE t)~i.l-O1 llCont'd)1 ConCd INSTRUMENT AND INSTRUMENT CONTROL SETPOINTS AND CONTROL SETPOINTS STARTUP RANGE STARTUP RANGE NEUTRON NEUTRON MONITORING MONITORING SYSTEM SYSTEM INTJUE4T 051l:GN IN~mi:UMENT CE ONAT ON ",,",0
...n.;)N AD TRIFTRl T aElN JrCTlO4, ADCI DT1ON M4C :.4MEN FUNcrl~ON ON LRM Inop
'.:;'M mop TI'IP Tr .. sa:: iC c 'hide Inlaa IIa rod Inlll,1i;M blook <~nd hlk1 rad 1I1()ok I11 i<orllm aoram Illhe IChw r; t~n<llllO!",:<
Il lIla Icl'<W.*:r,q Ira m~l:
cr6llICc: lire mal:
C51-IRI.l-JC>:1:l14~+il~~
1 Q &~F'" ..* 3AII:t:1 -ctj1 ,'r(r tR~ctcr MODE <ll.'Ii'TOH; Is !3:2! h'l ;:;'UN RactcrSCC!aTCal .JN C2Fl (.';.-H!I
- Cn-Kl':; o;:~"".IE OETE *I\!i=<<I~led
.IalaJ IRM 1PM Is !32!.
cl !)lli1ne~
)ça.a knrclar *;;;.M
";llnuntl~;); MAB) UCLPP
....IS) IJFllCld.<E,lNQP*
..* Ire1,.! i"7,cc!,,;je nM
'~"'~ilE ::-4 '& 4"~':
4*
unplupgea Lrpluoj IRM DIVNn'~DlIlo laM DcNrooaID S1 2 l.a Inl:~~~
Inltaa IIa rod rod blook bicok ,~'T~1~lo""ing "\:tnd~cn~ 1cra >lOPe ara rc;et:
r,a ca1mpM-Ka:i .tA+H
';;S1-IRM-JC>:1:l1 ;+lTh mil .tReactcr nactcr h~CC;= CCE a'.f~J0TCH Is ti"t:at r; m ,::f;UN JN nIrcl *~'RM
.ilt.nnund31jr H AiS:l CI'f;')'.'JNSJ.:..,I..Li:- rcr,
~sccl:!ll~rI IRtA PH I~ 1'701 ::"l=a'4s.~::f a..
A-! -4
'~".-05!-4J .,..,tC1Y~ ~
- stcja Range ianae -
LM Ullcoale
'RM Alarm U.z,a AIIIII:TI 7011'25a :'2.: Inl:=
lnl aa .. 11a red rod blook,~
blook 'T.~ a ~;)IDY.ing C;)ll:mtn~ : lcno >lOPe ara l7:e'::
ra CS1-IRM-K~':;1
- aipaci ,i ....+ H.J mil tfi<~cb:!r
.1eactcr MeCE Tc1 I~ :!hlljhr: .RUN
%CC <I'.'II:'TOH. JN
."'.rimmcll'br 4~;'~- Ai9) C4L A:B: Ul=aC.4LE-
""'.li:ccl~led
11".-052-4) nM UII~Re'Trlll IlRM U aTrlp I Inll~'le~
Intna /Ia red blcok ~nd rod bloOk i h.lf haiaoram uram Illhe r lIla rar:QN:r~
rcEcRr ~CMlIl:m ccraller: ar~ .:n~:
aranat:
,.-.+:
CzlIRM&jA+ 1:~ &,1t>I CSHRM-JC>:01 *Faactcr ~!OCE tRe~.;:b:!r CC TC1lIIl
- 3'.'1"Off I~ !3:2! r *.RUN LlN i::7,~~K1J!i V"."H) i!:
"""s::c~I:!4letl
- cclalai IRM 1PM I~Ia rot rpaaa;:)'!,~I1 .. =-~:t MrunclairoJ,"=tM
.It.nnundl1or .s: UF~3C~"'LEJ1NCj:P~
M ""-IS;, C1IOP A-OS ~-4
.*.~.-~5 -4 .1:. 4-£'.
4-OooLor NO<l DlrioDtor pull In 14c* Full NM Inn'.!i;M lnlta, II a red rod blook it 'T,e r 1olaiIa.ln ....1ng I:o,,:mcn~
1cIro aPe ara r:-~~<
rat:
CSIIRM-Ki CS1-IRM-K6')'1 "'-til m *1aadcr <MOC:;:
tR.e.!Ctcr 2tTCl 1$a !3:2! I".:;'UN OC :::I'.'nTOff .JN
.lRMeta:tcrMIn OIAll,l .:.ete:tcr.::at lUll In Nate: :r:"~r!;:~t. dll~~
Ncte: :rarck ,l ;:>r-ev!!nl aaa r."):)1 jOnl doIar-dl!'l~t1;),i mcamoil mt:v~,n:"t~lil ro:trnlaI CpaII :.:o, ía alad TIr1caI ft aIa-ra Marwa raIloc TT NOTE: With ','J'nh the shrtingllnks lhe shorting links rnived. an~' aingle removed. any single SRM SM UpsaIe Upscale Tr, Trio. Cf cr IRM RM Upscale UpsaIe cr or INO? signal will cause a lull INCP aignaIwilIcae full scran.
ISD-Ec.1 SO-09.1 I Reuj 6 Page age4a 40 of 81 o161 I Categories KJA:
KIA: 215003 A2.02 Tier/Group:
Tier / Group: T2GI T2G1 RO Rating:
RORating: 3.5
3.5 SRORating
SRO Rating: 3.7 3.7 LP Obj: 9.1-13C Source: NEW NEW Cog Cog Level:
Level: HIGH HIGH Category Category 8: 8: Y
- 13. AA plant
- 13. plant startup startup isis in in progress.
progress. AA control control rod rod block block has has occurred.
occurred. The The following following nuclear nuclear instrument indications instrument indications are are noted:
noted:
SRM SRM Counts Counts Position Position IRM IRM Counts Counts Range Rance A A 5 3x10 3x10 5 Full In Full In AA 25/1 25 25/125 33 BB 190 190 Mid Position Mid Position BB 65/125 65/125 2 CC 44 6x10 6x10 Full In Full In CC 35/125 35/125 33 0D 125 125 Mid Position Mid Position 0D 15/1 25 15/125 3 EE 12/125 12/125 22 FF 55/125 55/125 22 G 30/125 3 H 25/125 25/125 3 Which one of the following is the minimum action that is required to clear the ROD OUT BLOCK?
A Withdrawing SRM A only.
A':I B. Ranging IRM E to range 3.
C. Withdrawing SRM A and C.
D. Ranging IRM Band B and F to range 3.
Feedback Feedback K/A: 215004 KIA: 215004 K5.03 K5.03 Knowledge of Knowledge of the the operational operational implications implications ofof the following concepts the following concepts as as they they apply apply to to SOURCE SOURCE RANGE MONITOR RANGE MONITOR (SRM) (SRM) SYSTEM:
SYSTEM:
Changing detector Changing detector position position (CFR: 41.5 (CFR: 41.5 /45.3)
/ 45.3)
RO/SRO Rating:
RO/SRO 2.8/2.8 Rating: 2.8/2.8 Objective: CLS-LP-09.1 Objective: CLS-LP-09.1 Obj. Obj. 9a 9a Describe the Describe insertion/withdrawal of the insertion/withdrawal the SRM of the SRM detectors, detectors, including including the the following:
following:
- b. Reason for maintaing
- b. Reason for maintaing counts between 125 between 125 and .
5 2x10 5 2x10 .
Reference:
SD-09.1
Reference:
SD-09. 1 Cog Level:
Cog Level: high high Explanation:
o To clear the rod block SRM must be below 2x1 05 or IRMs must be be> > range 7. The retract permit is bypassed with IRMs IRM5 :::
> range 3. Withdrawing SRM A will cause the rod block to clear when less than 5
2x10 2x10 5.
DistractorAnalysis:
Distractor Analysis:
Choice A: Correct answer, see explanation Choice B: Plausible because IRM E is the only Div II IRM below range 3. If all Div I IRMs are above range 3 then the rod block from SRM Retract Permissive in would be bypassed, not the signal from SRM upscale. Also ranging IRM E to range 3 will cause a IRM downscale which is a rod block.
Choice C: Plausible because SRM A does need to be withdrawn and C is above the old setpoint for the upscale alarm. (recent change, old setpoint was 5x104). ).
4 5x10 Choice D: Plausible because IRM B & D are the only Div II IRMs below range 3 and these do meet the requirements for ranging them to 3. If all Div IIIRMs II lRMs are above range 3 then a rod block from SRM Retract Permissive would be bypassed, not the signal from SRM upscale.
Notes Notes TABLE TABLE 09.1-a. 1- 11 INSTRUMENTAND INSTRUMENT AND CONTROL CONTROL SETPOINTS SETPOINTS STARTUP RANGE STARTUP RANGE NEUTRON NEUTRON MONITORING MONITORING SYSTEM SYSTEM iF4ThLMEiT D::SIGN".TlCN INSTRI,;I\.1EKT DES1GNAT1CN TRIP SE7POINT TRIP SE7POiT ...ND ND FUNC7ICN FJNC7iC,....ADiX4L JDlTlO:v.L CCNJITICKS CCNlTlCN ...A4D :-<0 AND TRIP FU'<CTlOO ASO HllP FUNCflCN UNCTON
- UI'>CnON CClll ...1
- :NTS CClY1ET3 SRM lrcp Trip SRMIOOr.rr~ HlJFSS - 1CI%",
- 1C 1%
1%+ nrs a rod
.lntla:ESa roe blOCk lr :he r01C"'"r.;,t c~~01l1ons;;re rmot C51-SR~
CS1--S2C A-D) T'I;O "K5' [) (A-D) Sw4 net Swt.d1 nct In CPEATE Cf Ir CPERATE r .Reacicf MCOE
.Reaclcf MCOE SWITCH sirrci- ISa ro: roi ~1 RUN RU Mnu1a
....1nuno::l;;Jn:- SRM Mie lI,ptJgg,;.j SM MXlcie pu iMabnai
-ANY tlMs~)n2lIRM iM .., ~= nga 6S 3"13 a N::>T b)pa~ b3pasel..
'SRM RLtCLEmOP U;>SC,AL5'IKep- (."-OS 2-3) fA-OS 2-3) Nte:pasaae Note: Eljpass:tllf al'~I,;~lor2lIRIl5 lk1alrl i/s ~rE a:e aoo'\\!!
l>e R..*~nge aie 77 SRM Owr3ca1Tr SRM Downscale Trip T 5 1.Scpa 5",1.5Cps nrasa rod Intla:es.3 roe blOCk oc Ir :ne 1'o1C'.~l'I9 oonlllllor,s;;re
- n1cng oorclbraa-rnat O1et cS1-Rq-S2D (A-u',
C51-SRM-K50[) (A-Di T'" .Re3cftr 'lCOE
-R2aclor COE S\'flTCHS flCI- 15a r:o: ro: 1Il RUI'>
RU Mrurcla[cr "SF,M "mur:dalcf SfM OCWI'\SC.IIlE" DONSCAIr .4NYiManai IRM
-ANY,:lMslOn"J iRM '" Rir.;,t2
!nga :53 a'l::t N:!L!1)pa~.
a1i2b3p3aaeJ.
(A3E I-l't
,A,;): Note: 8pasaei it all
-4c: El)'Pass:tl1f l !lMsl:lnallRlIs 1Waiai iM5 ;;IE abe R..1'lge are ilb:)\\!! Ral;e 22 SRM Retrad!
SRM PGrrnlaaNe Reiraet PermISSIYE!
12: cJ:6 (lill1O , IEOI ;nra:es a roriblOCk lntla:esa roe eioc Irr: me: rOIC~~:r.;}
tclcwrJ conCillcrs;;re rcllura ate O1Et rlel CS1-FM-<S2CA-Di C51-SR\!-K511C' (A-D) TJ" - ./eecurncLLLlN
-3RM Ilie::alcr net flU IN A-im.n.zfa-
.oI,nncnd;;Jn:- .NiMsbal
.ANY 1JMslOn;;J IRM iM ..,RDm: :53 an:!
N;l[j)}pa~.
a11N2Lt1\p3eei M RETRACT
"$RM ETACT NCT C Note: 5pasaeo it all 4c: El}'Pass:tl1f IkslYul lMs ;;re ai tll,lslcn2llRlls ace Ra'ge are ;;oo,\!! Fa1e 22 EM;TTED ,:"'-3:
?::RMmED' -E ol.-:5]1-.i SRM Upscale Upecale Al3Im larm -
2Xl~~
2X :nea aa roCl.blOCk lntla:es rod ic Itr:neotcW- :he: I'oIC\'j~ oom::lllO~.s rdflure ;;re a-c 01:1:
rieL c51SR<ecqAy TO" C51-3RWKSOIHA-Di 1:i.3X
- .3x 1D-3i 3.2 ..CxX 1Q~f 1O: .Rcactcr MCOE
-Reaclcf SVicrCI- a no:
MCOE S\'IITCH!S ro:i ", RUN RU Annundaro.~
.kLraE '31'.\1 ' .AJthaiiM
-ANY !lMSlOnitIIRM .., Rir.;,t2 eS ~l
- au NOT NDT b)pa...,.:J.
b)paaaJ.
JFScALE-iN3 (A-Q5 UFSC."LE!lNO.'" AJE 2-3) 2-31 Note: EljpaSS:tl1fit al cte:p3Sae1 h-aIDrai iME ;le at !lMSlcr,21IRlls abo F.'l'lge are aoo\\!! .aue 7 SRM Upscale SRM Upacate trip Trip 5X 1(~!:f~
5X 1E ccc FlJII Full Scram Ir ~'J=rr.g sncfll'lg ScranuHreJeing eflc1-; In\s huc rEIil~vea rcrn:ee C51-SR\f"K6a[)(A-~) T,'" .3X (3.3 X tlYW--1.5 7.5 X X 1J')
12)
SRM PeriodPerIod so sccres-1O+16sec 5O&;;CO~ClS -10, +16 So?O Ar4aSFM
."nnuno::l;;Jn:-' SRM FERIOD' FEROD [A-C3-3) .....1:.53-3) csi-R<s:c C51-3RWK5!l[)("'-~)
j ... ClticaJ TfliI . -.
F3uIrmrr R_QuL Tn ..
rAnuaI tiM t!
... m_ntMiI!nll3:lliRMlr_~~c ', iSRM 3IFA In Imir .
...rum,enla~-Qn Il,-
.. Talec-,rfC"! ,lOOpe rc-r ,.DI.
,-..In:aw.. on ~};,1:teill"'C'II:~-er 2p:m,:n ." .. :ec ~)_In
.1;);o-lII: ... .
r .tt_ TM'JJI
, ~ !II TF5iY P2O I~
.. HV?:3::.3 :ua !il;t:! 'r'DM;);Jie i ~!i:e atiaa J)D'Aereer :UI=1=4)' :eLr3 (~~'J~61!r.
- urer :eteng t a-etc '"h1C mng;eJ an Ihe ra,eej 3n:i p rtaCe are ~:~i1i Y~'lle_
In pero:er,tage~ ace NO'!e: rap: AA ~(:cp{etel::.:.-:
- f If b cr Cf
,':loAer fr YiH wl oJcdlJ.ce apçarflt rI, Df CeCf an appi1rent-:rt,:Jo all trp t al1 .rIf '(I.e.
lrç I.;r,!f!; (I.n. FI.:H :crer Ifr ~hcrtjng FJl :cr~ flDrng 11->;:.
lr *;1ie re4 Cl,:,<!
ae r~:n:hed' C..f i)j a~~M up:aa rr:p)
.M 'iJp~:itle Tllp ISD-E.1 SD-W.1 Rev.
Rev 6 age ~
Page S of61 of I Categories K/A:
KIA: 215004 K5.03 Tier / Group: T2G1 RO Rating:
RORating: 2.8 SRO Rating: 2.8 LP Obj: 9.1-9A Source: NEW Cog Level: HIGH Category 8: Y
- 14. Which one of the following identifies the impact a loss of RPS MG Set B will have on Monitoring system and identifies the action the Unit One Power Range Neutron Monitoring required to energize RPS B from its alternate power supply?
A half scram will occur with (1) 2 and 4 losing power and by placing the RPS Power Source Select Switch on Panel P610 in the (2) position will re-energize RPS B lAW 10P-03, I OP-03, Reactor Protection System Operating Procedure.
A. (1) APRMs (2) ALT A (1)APRM5 B. (1) APRMs (2) ALT B C. (1) Voters (2) ALT A D
D~ (I) Voters (1)
(2) ALT B
Feedback Feedback K/A: 215005 KIA: 21 5005 A2.04 A2.04 Ability to Ability to (a)
(a) predict predict the the impacts impacts ofof the the following following on the AVERAGE on the AVERAGE POWERPOWER RANGERANGE MONITOR!
MONITOR!
LOCAL POWER LOCAL POWER RANGE RANGE MONITOR MONITOR SYSTEM; SYSTEM; and and (b)
(b) based based on on those those predictions, predictions, use use procedures procedures to correct, to correct, control, control, oror mitigate mitigate the consequences of the consequences of those those abnormal abnormal conditions conditions oror operations:
operations:
SCRAM trip SCRAM signals trip signals (CFR: 41.5 (CFR: 41.5 /45.6)
/ 45.6)
RO/SRO Rating:
RO/SRO Rating: 3.8/3.9 3.8/3.9 Objective: LOI-CLS-LP-09.6 Objective: LOI-CLS-LP-09.6 Obj.Obj. 12b 12b Given plant Given plant conditions, conditions, predict predict the response of the response of the the PRNMS PRNMS to to aa malfunction/failure malfunction/failure of of the the following following systems/components:
120 VAC Distribution
- b. 120 b.
Reference:
SD-09.6
Reference:
SD-09.6 Cog Level: High Explanation:
Each APRM instrument receives power from two power supplies, LVPS 11 and LVPS 4. LVPS 11 is fed from RPS RPS BusBus A while LVPS LVPS 4 is fed from RPSRPS Bus B. Therefore, a loss of an RPS Bus 8. RPS Bus Bus will not not affect operation of the APRM NUMACS. NUMACS. Each Each of the four VOTERS corresponds to a channel of the A 1, A2, B1, Al, Bi, and B2 RPS logic. The VOTER outputs to the RPS logic are: A1 Al (VOTER 1), A2 (VOTER 3), B1 Bi (VOTER 2), and B2 (VOTER 4). Voters 2 and 4 are powered from RPS B. OP-03 contains the steps to re-energize the RPS MG Set in which transferring to alternate power supply can be performed. If this is done then the switch will be placed in Alt B position. Some confusion usually happens as this procedure is performed because the light above the Alt B position is unlit. Students usually think then that Alt B has no power available to energize the RPS Bus and want to take the switch to Alt A which is the energized bus.
Distractor Analysis:
Choice A: Plausible because the APRM lose one power source but have a redundent power supply. The procedure action is plausible because the AL ALT T A is a position switch that is used for transferring the A RPS to alternate. The student may confuse this with transferring to the A RPS power supply because the light will be extinguished above the Alt B position and be on above the Alt A position.
Choice B: Plausible becasue the APRM lose one power source but have a redundent power supply. Alt B is the correct switch powition for the transfer switch.
Choice C: Plausible because the RPS system is the 120 VAC emergency power. The procedure action is plausible because the ALT ALT A is a position switch that is used for transferring the A RPS to alternate. The student may confuse this with transferring to the A RPS power supply because the light will be extinguished above the Alt BB position and be on above the Alt A position.
Choice D: Correct answer, see explanation.
Notes 2.8.8 PRNMS Power Supplies The Power Range Neutron monitoring Systern System uses one Quadruple Voltage Power Supply (QLVPS) chassis and four Dual Low Voltage Power Supplies (OLVPS),
(DLVPS), one for each bay of tile the PRNMS panel, to provide redundant power to the NUMAC instruments. These L VPS LVPS convert 120 VAC to lovi ~C. See Figure 09.6-'15.
low voltage DC. 09.6-15.
Each APRM instrument receives power from t\'y'o two pow'er power supplies.
supplies, L VPS 11 and L LVPS VPS 4. L LVPS VPS 'I1 is fed from RPS Bus A while LVPS 4 LVPS is fed from RPS Bus B. 8. Therefore, a loss of an RPS Blis Bus will not affect operation of the APRM NUMACS. Each RBM instrument also SD-09.6 1 SO-09.6 Rev. 5 Page 3'1 31 of 93 1 1.3.8 1.3.6 Two-Out-of-Four Logic System (VOTERS)
The VOTERS serve as the interface betw'eenbetween tile the APRM/OPRM APRMIOPRM channels, which generate safety safe trips, and the RPS. Each of the fOllr four VOTERS corresponds to a channel of tile the Ai, Al. A2, B'I.
Bi, and B2 RPS logic. The VOTER outputs to the RPS logic are: Ai Al (VOTER 1), A2 (VOTER 3), B1 BI (VOTER 2), and B2 82 (VOTER 4). VOTERS cannot be bypassed.
The VOTER logic does not latch a trip condition, condition. This means no reset is required and no input trip Signa!
signal occurs if one APRM instrument generates a a trip input and then clears before a second APRM generates a a trip input.
input A trip output occurs only if two or more inputs indicate a trip concurrently.
SD-09.6 1 SO-09.6 Rev. 5 Page 1100 of 931 93 From 10P-03:
CAUTION Transferring RPS Bus B to alternate power following a loss of power on RPS Bus B B shall always be accomplished by placing the RPS POWER SOURCE SELECT AL T B. A Scram will result if the switch is placed in AL T SWITCH in ALTB. TA.
A.
- 5. PLACE the RPS POWER SOURCE SELECT SWITHon SWITCH on Panel H12-P610 in ALTB.
Hl2-P610 inALTB.
Categories K/A:
KIA: 215005 A2.04 T2G1 Tier / Group: T2Gl RO Rating:
RORating: 3.8 SRO Rating: 3.9 SRORating:
LP Obj: 9.6-12B Source: NEW Cog Level: HIGH Category 8: Y
FIGURE 09.6-13 VOTERIRPS Interface Diagram (All VOTERS)
OPRM COA/INOP CDA/INOP TRIP OUTPUTS TO RPS ARE DEFEATED. REACTOR SCRAM ON 2-0F-4 2-OF-4 OPRM COAIINOP CDA/INOP TRIPS WLL NOT OCCUR r**-;~~******T****;;~*******!****"*;;~"*****l 2/4 2/4 2/4 2/4 2/4 r****;~4 ******r***;;~ ***!*******;;~******l 2/4 2/4 !OPRM 2/4
.................. \
2/4 2/4!
r****;~~******l*****;;~***
2/4 2/4
. *T. ***;;~2/4
. ******1
~~~~ I~~~ ! O~~~ I I~~~P~ I~~~ i O~~~ I APRM OPRM OPRM APRM OPRM APRM OPRM OPRM i TRIPS i DIDA DIOA i COA CDA i i TRIPS i DIOA DIDA i CDA COA i TRIPS DIDA CDA TRIPS DIDA CDA i i TRIPS i TRIPS i i i TRIPS i TRIPS i j TRIPS ' TRIPS ! i: i. TRIPS i TRIPS i:
"................... i.. ...................l ........I :.................. 1 ......... l. ... l.................j L***T**************~****l*********** ...: :'***r************** ..
- r*******************~********* .......:
j ~
~ ~ 1 1 1 1
IIWQ~!IK4Y II K43Y' IIW {IK4Y II K43Y QL 111~IIK4Y II K43Y'
!... IIf'=1IIK4X II K431, JIIf'=1IIK4X I K43x1 i:1:::: iii::: :i ::::1 /11f'=111 K4X I K43X
.........Yo.r..E.R ..l ...........Yo.r..E.R.. 2........ L. . . . . YOTER..3 ...................: !. . . . . .YOTER..4... !
RPS CH. Al RPS CH. BI RPS CH. 1\2 RPS CH. B2 I~:=---tl [11
!KI2D 1!
1...*****.. *************;*****1. . . *:. ..*** ..*..........] '-------;::-II,
! 1 KI2F!
i
'1 KI2H j
,...................................1 ,................................... .
KI5D K 1 4F lIeu !leu SOLENOIDS SOLENOIDS GROUP2A GROUP3A
\.......--~y------) \.......--~y~----) \.......----y~----) \.......--~y,----)
TRIP LOGIC A AlI TRIP LOGIC B BII TRIP LOGIC A2 TRIP LOGIC B2 SD-09.6 180-09.6 Rev. 6 Page 78 of 931 93
FIGURE 09.6-14 VOTERIRPS nterface Interface Diagram (VOTERS 1 I and 3) r******;;~********r*-*;;~** *****'****-**;*~~********1 j APRM j OPRM TRIPS -?~fp~
OPRM T~~:S i
i 214
- TRIPS
!~i;'o!j;-, 2-of-4 concurrent APRM trips .Q!
or 1 1 I
- TRIPS TRIPS i 2-of-4 concurrent OPRM DIDA trips or l,l,~,~L,;,~~Y II K~yl
the TRIP X output and TRIP Y output for that function (APRM trip or OPRM DIDA TRIP .Q!
or OPRM CDA trip) for each VOTER K4Y [i4x HK to change state and trip the RPS logic III~I~I K43X III~I~I KlX X f9 4
IK K43X related to the VOTER.
CDAlINOP TRIP OUTPUTS NOTE: OPRM CDAIINOP TO RPS ARE DEFEATED. REACTOR
........VO.TER ...1....... L. ........... VOTER3 ... . SCRAM ON 2-OF-4 2-0F-4 OPRM CDA/INOP CDAlINOP TRIPS WILL NOT OCCUR.
RPS CH. A1 RPS CH. A2 VOTER 1 I outputs to RPS Al A1
['". . . . . . . . . .1. . 1 VOTER 3 outputs to RPS A2 VOTER 22 outputs to RPS B1 Bi I K12A I I VOTER 4 outputs to RPS B2 L---ri: J .
iI K12E i Kl 5C Kl 4E HCU HCU SOLENOIDS SOLENOIDS GROUP1A GROUP4A HCU HCU SOLENOIDS SOLENOIDS GROUP2A GROUP3A
~~----~'<~------/ ~~----~'<~----~/
TRIP LOGIC A1 Al TRIP LOGIC A2 SD-09.6 180-09.6 Rev. 6 Page 79 of 931 93
0 N FIGURE 09.6-15 PRNMS Power Supplies
()
C. a)
U)
RPS RPS QLVPS BUSA BUS B DLVPS LVPS1 I I LVPS 1 RBMA I II LVPS4 I I LVPS4 APRM 1 2/4 LOGIC DLVPS MODULE (RPS A1)
LVPS2 1 I LVPS 1 RBMB I III LVPS3 I I LVPS4 APRM2 RBMB 2/4 LOGIC INTERFACE MODULE MODULE DLVPS (RPS B1)
I LVPS 1 APRM3 RBMA I LVPS4 2/4 LOGIC INTERFACE MODULE MODULE DLVPS (RPSA2)
"I LVPS 1 APRM4 I LVPS4 2/4 LOGIC
>-J MODULE 120VAC Cl) 0<
(I) ID 9
18D-09.6 C) Rev. 6 ci) (0 a. co C
Page 80 of 931 ci) C C) C)
FIGURE 09.6-16 Panel P603 Layout Operator Operator Display Display C51-R603A CS1-R603A C51-R603C CS1-R603C CS1-R603B C51-R603D CS1-R603D Assembly Assembly IRM A IRMA APRM1I APRM APRM2 IRMB RBM A RBMA RBMB RBM IRM C IRMC APRM APRM33 APRM4 IRMD IRM ER RBM A RBMA RBMB IRMF IRM 0 IRMG IRMH IRM H (Flux Level Only) (Flux Level Only) (Flux Level Only) (Flux Level Only)
Operator Operator Display Display Assembly Assembly APRM 113 APRM 214 Reactor
, Mode Switch Manual Manual Scram RBM IRM Scram SN Sw Scram SN Sw Reset Sw\
ResetSN"""\ APRM IRM BypSw BypSN BypSw BypSN BypSw BypSN BypSN BypSw
@ @ Trip 0 00 0 @ @ @
m m :1 m m
/"" Upscl Alarm
/
Dnscl Alarm
~ ~
Dnscl Alarm m
-j-W m I-W IRM/APRM IRMIAPRM IRMIAPRM IRM j-- By w
lRMROM IRMIRBM ass Bypass Indicator (typ)
InWcator m
IRMIRBM IRMRBM IRM m
IRM/APRM IRMIAPRM m
IRM/APRM IRMIAPRM A/l Ai1 C13 CO E B 0/A GIA BiB D D F12 Fi2 H/4 HI4 Z3A Z3C Z3E Z4A Z4B Z3B Z36 Z3D Z3F 80-09.6 E-09.6 Rev. 6 Page 81 of 93
FIGURE 9.6-22a Stab Wty Screen - DSS-CD OPRM Stability APRM 1 1 IOPRM I TRIP IALARMI RUN MODE OPERATE tRAPH OPRM DSS-CD GRAPH 0D 0C 5TP:
STP: 55.0% >6 r /
FLOW: 40.0% C C 6 E
ES 5 L 4
CELLS ------- L 3 OPERABLE: 24 S 2 RESPONSIVE: 19 1 I I I L 1
1 5 10 15 20 25 COUNTS SD-09.6 180-09.6 Rev. 6 93 Page 88 of 931
- 15. Unit One is at 94% power when one recirc flow input to APRM 2 fails downscale (zero).
15.
Which one of the following identifies:
(1) the OPRM response to the recirc flow failure and (2) the required action lAW the annunciator procedures?
A (1) OPRM 2 only is enabled.
A:'
(2) Bypass APRM 2.
B. (1) OPRM 2 only is enabled.
(2) Verify RBM B auto transfers to APRM 1.
c.
C. (1) All OPRMs are enabled.
(2) Bypass APRM 2.
D. (1) All OPRMs are enabled.
(2) Verify RBM B auto transfers to APRM 1.
Feedback Feedback K/A: 215005 KIA: 21 5005 A2.05 (a) predict Ability to (a) predict the impacts impacts of the following on the AVERAGE POWER POWER RANGE RANGE MONITOR!
MONITOR!
LOCAL POWER LOCAL POWER RANGE RANGE MONITOR MONITOR SYSTEM; SYSTEM ; and (b)(b) based based on those predictions, predictions, use procedures mitigate the consequences of those abnormal conditions or operations:
to correct, control, or mitigate Loss of recirculation Loss recirculation flow signal (CFR: 41.5 / 45.6)
ROISRO Rating: 3.5/3.6 RO/SRO LOl-CLS-LP-09.6 Obj 12g Objective: LOI-CLS-LP-09.6 Given plant conditions, predict the response of the PRNMS to a malfunction/failure of the following systems/components:
- e. Recirc Flow Module SD-9.6, 1APP-A-06
Reference:
SO-9.6, Cog Level: High Explanation:
Each Numac processes the signals from one sensor in Loop A and one in Loop B and averages the signals to obtain total recirc flow rate. If one of the two recirc flow signals to an APRM failed to a zero signal with reactor power at 100%, its OPRM becomes enabled because the calculated flow is reduced to one half of its initial value. The other APRM/OPRMs will be unaffected. The RBM has a primary reference from APRM 2 with the primary alternate from APRM 4 and a secondary alternate from APRM 3. APRM 1 I is the primary reference for RBM A.
Distractor Analysis:
Oistractor Choice A: Correct answer, see explanation Choice B: Plausible because the RBM B will transfer from APRM 2 to its alternate reference which is APRM 4 or its secondary alternate of APRM 3. Prior to the Numacs the transfer of the flow units was a manual transfer. APRM 1 1 is not used for RBM B but is the primary for RBM A.
Choice C: Plausible because the alarm for OPRM enabled will be in alarm, and the Voters will see the OPRM enabled on all 4 voters for only OPRM 2 though.
Choice 0:D: Plausible because the alarm for OPRM enabled will be in alarm, and the Voters will see the OPRM enabled on all 4 voters for only OPRM 2 though. The RBM B will transfer from APRM 2 to its alternate reference which is APRM 4 or its secondary alternate of APRM 3. Prior to the Numacs the transfer of the flow units was a manual transfer. APRM 1 1 is not used for RBM B but is the primary for RBM A.
RBMA.
Notes
4.23 4.2.3 Recirculation Flow Transmitters Failed Recirc Flow transmitters could result in control rod blocks or trip signals to be generated by the associated APRM, depending on the direction of failure and the initial reactor power level. For Far two recirc flow signals to an APRM failed to a example, if one of the t\'\I'o zero signal with reador reactor power at "100%,
100%, its OPRM becomes enabled because the calculated flow is reduced to one half of its initial value, and its STP rod block and trip set paintpoint will be exceeded because the flow used to calculate the STP rod block and trip set points is also reduced to one half of its initial value. Tile The other APRM channels are not affected since they use separate recirc flow signals.Signals.
Since the APRM 4 4 instrument provides Loop A and Loop B B Flow signals for meters, if either of its flow inputs fail to zero the associated meter will indicate zero. Since the APRM '1 instrument provides Loop A and Loop B flow signals for the flow recorder, if either of its flow inputs fail to zero the associated recorder pen will indicate zero.
Recirc Flow Module failures can be bypassed using the APRM bypass switch. This method, however, will bypass all functions associated with the affected APRM channel.
Flow signals and flow upscale alarm signals are bypassed when the corresponding APRM channel is bypassed. The RBM disregards flow signals from a a bypassed APRM wilenwhen processing flow compare logic.
1 SD-09.6 SD-09.6 Rev. 5 Page 45 of 931 93 Each RBM channel designates a hierarchy of normal and alternate APRM channels to use as their reference APRM channel. The alternate channels are used in hierarchical order when the preferred channels are not available. The primary reference APRM for RBM A is APRM 'I1 with the first alternate as APRM 3 and the second alternate as APRM 4. The primary reference APRM for RBM B B is APRM 2 with the first alternate as APRM 4 and the second alternate as APRM 3. The RBM circuitry will automatically transfer to an alternate APRM on failure of the primary reference APRM (Critical Self Test Fault). No operator action is required for this transfer.
RBM Channel A RBM Channel B B Primary Reference APRM 11 APRM2 APRM 2 First Alternate APRM3 APRM 3 APRM4 APRM 4 Second Alternate APRM4 APRM 4 APRM3 APRM 3
IONES ACTTIN AC H BA3ARG .GP PH R.;:._PR dis play5 di~pay at:; APRt1 5 at t ns of FLG cns FLON N (~) en en APR ?PRM derss do not noc If cbs erva~i ebserv atic ica tio tio n~
ns c.n on RSM P.BM OO~_ header 3DA and FLO FLOW N C3 .EE aia CO~lP;'_9. re ind alar rr:;
00As, OD 5, and ver forfor m ~he the foll fofl ow owi ing ng::
caisse e of the che aia alarrn.m, then the n t.er TRIPP PJ STAusTOS identify fy cau ETC so*f seftt key ke::l to to obt an IcLI obt: ;ain Cn OJ:". ~he the APF APfl d'l
}1 OD:?
OD- .s, press
-_5, ss ZTC a_
50f~
sof t key.
s sTR IP IP S7A TT J sof TOS sc.ftt ke key..
b_
ALA -_9.M AH c_
- c. Check FLD FLOil uPS UPS play.
di~t..a dis T1UJ;> STATUS ain, TPP c.:otaizo TUS sofsoftt-OD? s, ttespre ss E7C sof s EUC 50I t- !'.ey to obt t key d_
- d. Cr.
On Br: RBM ODA ke!l _
ke -
kev..
Pres sssTR TPI IP P STA TtJS sof S7A TOS sc.ftt key P STh TiJS S7..,_T]S e_
- e. N ElOti N FLm iC01PA -1PAP RE sta ~'ilS fron
~-t:;at.is from UPL TRIIP I _ Cbe C1-e ck 9.EC CIA IRG CU ULA LA TIO TID di~ppay dis la3{. . form. the per form the n per ;:;he car. be identifie ll can fied, d, then 2.
2 iilhe n the failed channe when fc)1 1ctvin'q :
a_
- a. ify the Uni Hot- if::,'
Not uni t 3CC seo_.
ected
-affect ed AP APrM uf b_
b.. Bypass the aff .-AL ann lclORt-L",-L uncciat anJ: ",un iatoorr cle ars..
Confirm f.rr. the FLO FLOWI'iRE REFE' OFF NDP ers,r
',io~ers c._
~ Con dit
"-dit iot il::>n ,,-as was initiat ini ~iated ed to t:;c. the Vot AP~ M Usc1e l iJps cale trip cC(I con
- 3. If an AP e.3.c eac hh Vot Vot .er_
er.
ss TRI l0RYI RES NE?IOP.
P l-IEt- AESED ET at t:;he then n pre~s TRIP Rev. 46 Page 66 of 78 781 APP "11AP P-A-A-0 6
-06 ies egorries Catego Grouup:
Tier I/ Gro p: T2G1I T2G K/A:
KIA 215005 A2.05 SRO Rating: 3.6 g:
atinng:
RO Rati ROR 3.5 35 Sourrce:
Sou ce: NEW bj:
LP Obj:
LPO 2G 9.6-112G 9.6-Category 8: Y Leveel:
Cog Lev l: HIGH
One RPS A analog
- 16. Which one of the following identifies the power supply to the Unit 16.
trip cabinets?
A. 120 VAC Panel 31AB B. 120 VAC Panel 32AB C~ 125 VDC 11A C hA D. 125 VDC 12A Feedback 216000 K2.01 K/A: 216000 KIA:
Knowledge Knowle dge of electrical power supplie s to the followi ng:
following:
Analog trip system : Plant-S Plant-S pecific pecific (CFR: 41.7)
Rating: 2.8/2.8 ROISRO Rating:
RO/SR CLS-LP-03 ve: CLS-LP Objective:
Objecti 18h
-03 Obj 18h following:
State the power supplies for the followin g:
- h. Analog Trip System Logic Cabinets
Reference:
Referen SD-03 ce: SO-03 Cog Level: Low Explanation: XU-68)).. Cabinets channell (XU-6 housing a separate channe (XU-65 through XU-68 5 through cabinetss for the RPS, each housing There are four cabinet 12A(B) for unit 2. An OC panels 12A(B) 1 IA(B) for Unit 1I and DC NLI /Topaz receive power from OC DC panels 11A(B) cabine t.
. In order to meet complete the comple te (backup (backu p)) inverter and a Lambda power supply are located in each cabinet supply failure in designedd to be shared in the event of a power criteria, the power supplies are designe redundancy criteria, PCIS on a loss of power. 31AB(3 31AB( 2AB) 32AB) feed trip logic for cabinetss cause a trip cabinet.. These four cabinet one cabinet (cause) a group 6 isolation on a loss of power.
group 6 isolation and also trip (cause)
Analysiis:
Distractor Analys Oistrac s:
supplies to trip systems in which this feeds Choice A: Plausible because there are 120 VAC (UPS) power PCIS Group 6 isolation tip logic.
Choice B: Plausible because there are 120 VAC (UPS) power supplies to trip systems in which this feeds PCIS Group 6 isolation tip logic.
explanation Choice C: Correct answer, see explana tion cabinetts.
cabine s.
D: Plausible because This is the feed to Unit two analog trip system Choice 0:
Notes
120V Distribution Panel 'J-3'IAB Load: '120V i-31AB (He7)
(HC7)
Location: Control Building Suiding 49'49 NW Drawing
Reference:
LL-93041-25 Upstream Power Source: 120V Emergency Distribution Panel 1 I E5 ES (Normal) 120V Emergency Distribution Panel 1 I E6 (Alternate)
CKT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER POWER 1 CAC System Inboard LOC.A. LOCA Signal Signa Trip Logic Group 6 Div II AC-powered CAe CAC valves will (XU-53, H12-P606, P622) (CAC-V172, V5, close (CAC-\l172, \/5, V6, V7, 17, V162, 1162, \1'163, V163, (TS 3.3.6.1,3.3.6.2,3.3.3:1,3.4.4,3.4.5, 3.3.6.1, 3.3.6.2, 3.3.3.1, 3.4.4, 3.4.5, V9), all Group 6 CAM, RIP and RXS valves
\(9),
00CM 7.3.2}
3.6.1.3, TRM 3.4, ODCM 7.3.2) will close 2 CAC System Division '1 (AC) Isolation Trip Will receive a a full Group 6 isolation {all (all CAC-CS-41 78 and Stack Override Switch, CAC-CS-4178 Group 6 6 valves will close, reactor building vu willI iso ation Override Switch Rad Monitor isolation S8GT tr&ns isolate, SBGT trains will start due to stack rad rod CAC-CS-5519 (TS 3.3.6.1, 3,3.6.1, 3.3.6.2, 3.3.3.1, monitor trip signal), override capability will 3.4.4, 3.4.5, 3.6.1.3, TRM 3.4, ODCM00CM 7.3.2) be disabled Load: '120V 120V Distribution Panel 2-32AB (HXO)
Location: Control Building 8uilding 49' 49 SE LL-0934i-25 Drawing
Reference:
LL-09341-25 Upstream Power Source: 120V Emergency Distribution Panel 2E7 (Nomlal) (Normal) 120V (Alternate 120V Emergency Distribution Panel 2E8 (Alternate)
CKT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER 1
1 CAC System Inboard LOCA Signal Signa Trip Logic Group 6 Oil,'
Div lAC-powered I AC-powered CAC valves will (XU-53, H12-P606, P622) close (CAC-\l172, (CAC-V172, V5, V6, V7, V162, V163,
/7, \l162, (TS 3.3.3.1,3.3.6.1,3.3.6.2,3.4.4, 3.3.3.1, 3.3.6.1. 3.3.6.2. 34,4, 3.4.5, V9), all Group 6 CAM, RIP and RXS valves 3.6.1.3. TRM 3.4, ODCM 3.6.1.3, 00CM 7.3.2}
7.3.2) will close 2 CAC System Division 1 I (AC) Isolation Trip Will receive a full Group 6 isolation on Unit Override Switch, CAC-CS-4178 and Stack I and Unit 2 {all 1 (all Group 6 valves will close, clcse.
Rod Monitor Isolation Rad Iso ation Override Switch reactor building wilwill isolate, SBGT trains wi wi!!I (TS 3.3.3.1, 3.3.6:1, 3.3.6.1. 3.3.6.2, 34.4, 3.4.5, 3.3.6.2. 3.4.4, monitoririp start due to stack rad monitor trip signal),
3.6.1.3. TRM 3.4, ODeM 3.6.1.3, 00CM 7.3.2}
7.3.2) override capability will be disabled on Unit 112-UA-03-5-4 and '1t2-UA-05-6-10 2 only. 1/2-U.A.-03-S-4 112-UA-05-6-iO will alarm.
Categories KIA:
K/A: 216000 K2.01 Tier / Group: T2G2 RO Rating:
RORating: 2.8 SRORating:
SRO Rating: 2.8 LPObj:
LP Obj: 03-18H Source: BANK Cog Level: LOW Category 8: Y
- 17. Following
- 17. Following aa loss loss of of feedwater, RCIC RCIC initiated initiated on on low low reactor reactor water level, level, then shut down subsequently shut down on on high high reactor reactor water level.
level.
Current plant Current plant status is:
is:
Reactor water level Reactor level isis 170 170 inches inches RCIC steam supply valve (E51-F045) is is closed RCIC injection valve (E51-F013) is closed RCIC flow controller in Auto set at 200 gpm The RO opens the E51-F045 and then depresses the PF push button on the RCIC flow controller. No other actions are performed.
Which one of the following identifies the indicated flow on the RCIC flow controller that would be observed for these conditions?
A 0 gpm A't B. 200 gpm C. 400 gpm D. 500 gpm
Notes 3.4.3 Pump Discharge Valve, E51-F012, RCIC Injection Valve, E51-F013 (Figures 16-21, 22)
RCIC Pump injection to the vessel is controlled by the nomlally open Pump Discharge Valve, E5'1-F012, E51-F012, and the normally closed RCIC Injection Valve, E51-FO*13. E51-F0i3. Tile The RCIC .Injection injection Valve will automatically open upon receiving a low reactor water level signal (LL2) provided tllat that neither the Turbine Trip and Throttle Valve, E5i-V8, nor the Turbine steam E51-V8, Steam Supply Valve, E5*t-F045, E51-F045, is full closed. The RCIC Injection Valve will automatically close once eitller either the Turbine Trip and Throttle Valve or the Turbine Steam Supply Valve is fully closed. If the Pump Discharge Valve is closed upon receiving the 1m\' low reactor water level signal, it, too, will IA~II receive an automatic open signal. The Pump Discllarge Discharge Valve may be controlled manually, using the RTGB P601 keylocked three-position CLOSEJAUTO/OPEN control switch, key removable in AUTO). The CLOSE/AUTO/OPEN RCIC Injection Valve may also be controlled manually, using the RC1C RTGB P601 three-position CLOSE/AUTO/OPEN CLOSEJAUTO/OPEN control switch.
tI I I
~ K40 - energized on '.18 k4
- V8 full s.~ ::u Jpg closed n
cosed E n
= I - energiized on F045 JQI} K20 - energized
- J full closed tii" n*
KJ K3 - Ll2
- LL2 signal sigri& """
CJ*
_:i
- J."
<G:l
!i.l-c I
115 /i:::o 1U
_rn II _Jl____ 5!'91 t.:Il'-.l
~I'-.l I
n 0
X' 21t CJ*
- J
.,Moo o
(2F:\,1 s2F I
i(:2R 2RI I r
' - . - / WI!II JcLrAE
~l;Lf;!jE c
_______________~~_oom_*_~_,roF 1Iir1t __________________~J(O~ ]ieI4T1R UJ n
JIHIClIIiI~IIIIIJIIWIIf_ * . . , . * 'WI Uf-U:IIIN..iXIII'IWI1l11mH '....... * .....1III1UL~
_~illfllll'miH """""""'WHlMI'IIMINLdtJIII!!D Categories KJA:
KIA: 217000 A1.01 Tier/Group:
Tier / Group: T2G T2G11 RO Rating: 3.7 SRO Rating: 3.7 LP Obj: 16-12A Source: BANK Cog Level: HIGH Category 8: Y
- 18. Given the following plant conditions with RCIC in pressure control mode:
RCIC controller output 70%
E51 -F022 Bypass to CST Vlv, E51-F022 Throttled RCIC Flow 300 gpm RPV pressure 990 psig, slowly rising RCIC controller Automatic set @ @ 300 gpm Which one of the following identifies how RPV pressure can be stabilized?
(1)L-)_ _ direction, or by _->(=2),--_
The RO can throttle the E51-F022 in the _->(..:..1 (2) the RCIC Flow Controller auto setpoint.
A. (1) open (2) lowering B. (1) open (2) raising C. (1) closed (2) lowering D
D~ (1) closed (2) raising
Feedback Feedback K/A: 217000 KIA: 217000 A1.04 Al .04 Ability to Ability to predict predict and/or andlor monitor monitor changes changes in in parameters parameters associated associated with with operating operating the the REACTOR REACTOR CORE ISOLATION CORE ISOLATION COOLING COOLING SYSTEM SYSTEM (RCIC)
(RCIC) controls controls including:
including:
Reactor pressure Reactor pressure (CFR: 41.5 (CFR: 41.5 1/ 45.5) 45.5)
RO/SRO Rating:
Rating: 3.6/3.6 3.6/3.6 Objective: LOI-CLS-LP-016-A Obj. 17b 17b Describe how the following evolutions are performed during operation of the RCIC system:
- b. Adjusting RCIC flow in the reactor pressure control mode.
Reference:
RCIC Hard Card Cog Level: High
..* ~
Explanation:
There are two ways to reduce the RPV pressure with the conditions given. One way is to open the 22 valve, there by increasing the size of the hole and mantaining the same flowrate, this will work until the controller is at 100% output. The second is to raise the controller setpoint thereby causing the turbine to work more (increase flow through the same size hole).
Distractor Analysis:
Choice A: Plausible becasue these are the opposite of the actual answers and if the operator was trying to raise RPV pressure this would be correct.
Choice B: Plausible because raising is correct and the operator could have a misconception about the 22 valve.
Choice C: Plausible because closing the 22 is correct and the operator could have a misconception about the flow controller.
Choice D: Correct answer, see explanation.
Notes RCIC PRESSURE CONTROL (OP-16 SECTION (OP-16 SECTION 8.2) 8.2) 1.
- 1. ENSURE THE ENSURE THE FOlLOVVING VALVES ARE OPEN:
FOLLOWING "ALVES OPEN: ES1-V8 E51-V6 f'/AL\lE (VALVE POSITION},
POSlTION E51-V8 ES1-V8 (ACTUATOR POSITION),
(ACTU.A.TOR POSITION). AND E51-V9.
2.
- 2. EE1-F046 OPEN E51-F046 OPEN
- 3. START VACUUM PUMP AND LEAVE SWITCH IN START.
LEAVE S'NITCH
- 4. E51-F013 IS CLOSED ENSURE ES'1-F013 5.
S. ENSURE E41-F011 IS OPEN
- 6. THROTTLE OPEN E51-FD22 DUAL INCATION IS OBTAINED E51-F022 UNTIL DU.A.LINCATION
- 7. OPEN E51-F045 6.
- 8. THROTTLE OPEN E51-FD22E51-F022 OR ADJUST RCIC FLO'N FLOW CONTROL, E51-FIC-R600, E51-FIC-R600. TO OBTAIN OBT DESIRED SYSEM PARAMETERS All'! DESIRED PARAMETERS .A.ND AND REACTOR PRESSURE.
- 9. ENSURE E51-F01 9 IS ENSURE ES1-FO'19 IS CLOSED WITH FLO'N FLOW ABOVE 80 GPM.
10.
to. ENSURE THE FOLLO'JVING ENSURE FOLLOWING VALVES ARE CLOSED: E5t-F025, ES1-F026, E51-F004, ESi-F026, E51-FD04,
.A.ND E51-F005.
AND E51-FD05.
- 11. START S8GT SBGT (OP-10}
- 12. ENSURE BAROMETRIC CNDSR CONDENSATE PUMP OPERATES FOR SHUTDOWN REFER TO OP-16 FOR TRANSFER BETWEEN PRESSURE AND LE'VEL LEVEL CONTROL REFER TO OP-16 1/1086 1 OP-i 6 l'lOP-16 Rev_
Rev 7'1 71 891 Page 86 of 89 Categories K/A:
KIA: A1.04 217000 Al.04 Tier/Group:
Tier / Group: T2Gl T2G1 RO Rating: 3.6 SRO Rating:
SRORating: 3.6 LP Obj: 16-17B Source: NEW Cog Level: HIGH Category 8: Y Y
- 19. Unit One is operating at power with CS pump 1 I B under clearance.
A small break LOCA occurs simultaneously with a Loss of Off-site Power to both units.
Only DG2 and DG3 start and tie onto their respective E E bus.
The following plant conditions exist on Unit One:
AUTO DEPRESS TIMERS INITIA INITIATED TED In alarm REACTOR LOW WTR LEVEL INITIA INITIATION TlON In alarm RPV pressure 600 psig Drywell pressure 13 psig Based on the conditions above, which one of the following identifies the status of ADS to depressurize the RPV for low pressure injection?
A. Not auto initiate due to loss of power to the Fluid Flow Detection cabinet.
B. Will auto initiate when RPV pressure lowers to 410 psig.
C~ Not auto initiate due to the logic not made up.
C D. Will auto initiate in 83 seconds.
Feedback Feedback K/A: 218000 KJA: 218000 K3.01 K3.01 Knowledge of Knowledge the effect of the effect that loss or that aa loss or malfunction malfunction of the AUTOMATIC of the AUTOMATIC DEPRESSURIZATION DEPRESSURIZATION SYSTEM will SYSTEM will have have onon following:
following:
Restoration of Restoration of reactor reactor water water level level after after aa break break that that does does not not depressurize depressurize the the reactor reactor when when required required (CFR: 41.7 (CFR: 41.7/45.4) 1 45.4)
RO/SRO Rating:
RO/SRO Rating: 4.4/4.4 4.4/4.4 Objective: CLS-LP-20 Objective: CLS-LP-20 Obj. 16b Obj. 16b Given plant Given conditions, predict plant conditions, predict how how the the following following will will be affected by be affected by aa 1055 loss or or malfunction malfunction of of ADS/SRVs:
ADS/SRVs:
- b. Reactor
- b. Reactor water water level level
Reference:
SD-20 Cog Level:
Cog Level: High High Explanation:
loss of offsite power and 1A CS pump under clearance this would leave only one pump With the 1055 pump available in each RHR loop. Therefore ADS logic is lost. Level will continue to lower until the ADS valves are manually opened (emergency depressurization) at which time the running low pressure pumps will be able to add water.
Distractor Analysis:
Choice A: Plausible because the FFD cabinet is powered from E8 and it has lost power. This only affects the acoustic monitoring system 50 so alternate means of determining if the SRVs are open would have to be utilized.
Choice B: Plausible because 410 psig and hi drywell pressure is a LOCA signal for starting the pumps.
The alarm though is the LL3 actuation 50 so the pumps would already be running. The logic will not be made up with only one RHR pump in each loop.
Choice C: Correct answer, see explanation.
Choice D: Plausible because if the logic was made up this would be the correct answer.
Notes Notes 4.t2 4.1.2 Automatic Operation Operation The ADS logic The logic automatically automatically opens opens thethe ADS ADS valves valves ill in tile the event event the the HPCI System HPCI System fails fails to to maintain maintain reactor reactor level level during during aa LOCA LOCA. The The seven ADS valves open automatically when all the following either of two logic cllannels conditions are met on eitller channels (A or 8) B) associated with ADS:
- (LL3 from B2'1-L Reactor low water level (Ll3 B21-LTS-NQ31A TS-N03-1 A and C or B and D).
- Reactor confirmatory low water level (lU (LLl from 821-LTS-N042A B2l-LTS-N042A or 8). B)
Operation of botll both pumps of an RHR loop or one Core Spray pump as indicated by a pump discharge pressure of 1'15 115 psig (either El 1-PS-ND1GA AND Cor E'11-PS-NO'16A C or 8B AND D or E11-PS-N020A El 1-PS-NO2OA AND Cor C or 8B for RHR or either E21-PS-NOOBA AND D for E21-PS-NQO8A AND E11-PS-N009A El 1-PS-NDO9A or E21-PS-NOO8B AND E21-PS-N0098 for CS).
E21-PS-NOOBB Cs).
A time delay of 83 seconds has elapsed (timer B21-TDPU-K5A or B).
AUTO/INHIBIT switches in AUTO for either or both logic channels A and B.
ISD-20 SD-20 Rev. 2 PAGE 26 of61of 61 Categories K/A:
KIA: 218000 K3.01 Tier/Group:
Tier / Group: T2GT2G11 RO Rating: 4.4 SRO Rating: 4.4 LP Obj: 20-16B Source: NEW Cog Level: HIGH Category 8:
- 20. Reactor
- 20. Reactor Recirculation Recirculation pumps pumps have have tripped duedue to aa low low level level condition.
condition.
G31-F001, G31-F001, RWCU RWCU Inboard Inboard IsolIsol Vlv, is Closed.
is Closed.
G31-F004, RWCU G31-F004, RWCU Outboard Outboard Isol Isol Vlv, isis Open.
Open.
Which one of the following identifies identifies what the Group Group 33 Isolation Isolation Status Status Box Box on ERFIS ERFIS will display in in five minutes?
minutes?
A A green GROUP ISOL A":I B. A red NO GROUP ISOL C. A yellow GROUP ISOL CMND D. A green NO GROUP ISOL CMND Feedback K/A: 223002 A3.03 KIA:
Ability to monitor automatic operations of the PRIMARY CONTAINMENT ISOLATION SYSTEM1NUCLEAR STEAM SUPPLY SHUT-OFF including:
SYSTEM/NUCLEAR SPDS/ERIS/CRIDS/GDS: Plant-Specific 41.7 /45.7)
(CFR: 41.7/45.7)
RO/SRO Rating: 2.5/2.8 Objective: LOI-CLS-LP-060-A Obj 4d Describe the methods used to do the following on the ERFIS/SPDS Computer:
- d. Obtain Group Isolation status including valve position
Reference:
SD-60 Cog Level: High Explanation:
ERFIS relies on the isolation signal to determine if an isolation is required. Since RWCU did receive a signal, ERFIS will recognize a valid isolation signal with at least one valve closed in the penetration path and remain Green and display GROUP ISOL.
Distractor Analysis:
Choice A: Correct answer, see explanation.
Choice B: Plausible because this is what would be expected with an isolation signal and no valves closed.
Choice C: Plausible because the isolation signal and valve closure time has not expired and can be confused with an incomplete isolation of the penetration flow path (both valves not closed).
Choice D: Plausible because the candidate does not recognize Recirc pump trip is LL2 (same as RWCU) would be indicated if no isolation isolation signal present.
Notes Notes From SD-60 From SD-60 E:ent Display Color Condition Condition Status Message - Code Inactive NO GROUP ISOL Green 1 No isolation solaton signal gnal CMND Safe GROUP ISOL Green t Isolation Isolation signal signa Valve
- 2. Val (nie exceeded
.... e closure time 3 At least one valve in each path closed Caution GROUP ISOL Yellow 1. Isolation signa CMND 2 Valve closure (me not exceeded Alarm NO GROUP ISOL Red 1. Isolation signal le.olation signa
- 2. Valve (me exceeded Val .... e closure time I
- 3. vave closed in aa path No val'/e Categories K/A:
KIA: 223002 A3.03 Tier / Group: T2G1 RO Rating:
RORating: 2.5 SRO Rating:
SRORating: 2.8 LP Obj: 60-4D Source: BANK Cog Level: HIGH Category 8: Y Y
Given the following small break LOCA conditions on Unit Two:
- 21. Given Drywell pressure Drywell 9.8 psig Suppression chamber pressure 8.5 psig Which one of the following identifies the response of suppression pool water level after initiating suppression pool sprays?
The suppression pool level indication will (1) slightly due to the (2) DP between the drywell and suppression pool.
A. (1) lower (2) higher B. (1) lower (2) reduced C
C~ (1) rise (2) higher D. (1) rise (2) reduced Feedback K/A: 230000 Al KIA: .06 A 1.06 Ability to predict and/or andlor monitor changes in parameters associated with operating the RHRlLPCI: RHRILPCI:
TORUSISUPPRESSION POOL SPRAY MODE controls including:
TORUS/SUPPRESSION Suppression pool level (CFR: 41.5 145.5)
/ 45.5)
RO/SRO Rating: 3.3/3.3 Objective: N/A
Reference:
none available Cog Level: High Explanation:With the SP at 8.5 psig and then sprays initiated the pressure will lower in the SP and this will cause the higher delta pressure between the DW and SP to force some water down the downcomers to slightly raise the water level in the SP due to the Higher dP. The pumps take a suction from the SP and then spray back to the SP.
Distractor Analysis:
Choice A: Plausible because a higher d/p dip would be developed from the spray initiation, but level would not lower based on dP.
Choice B: Plausible if the student has backward thinking of what is occurring with d/p. dip. Lower pressure is lowering dP.
Choice C: Correct Correct answer, see explanation Choice D: Plausible because a lower d/p dip would cause level to rise but tthe d/p dip will increase when sprays are initiated.
Notes Notes
<l a slight increase will occur as water is pushed down the downcomers into the SP.
Categories K/A:
KIA: A1.06 230000 Al.06 Tier / Group: T2G2 RORating: 3.3 SRO Rating: 3.3 SRORating:
LP Obj: NA Source: NEW Cog Level: HIGH Category 8:
- 22. Given
- 22. Given the following conditions conditions with Unit Unit One One in in Mode Mode 5:5:
A single control A control rod rod isis withdrawn to position position 48 for bladeblade removal removal RWM is Bypass RWM is in Bypass in control rod The control rod is is selected Rod Select Rod Select Power is is on Which one of the following describes the adverse consequence if Rod Select power was turned off, then back on, for uncoupling?
A. A select block will occur.
B A rod out block only will occur.
B!'"
c.
C. A rod insert block only will occur.
D. A rod insert and a rod out block will both occur.
Feedback K/A: 234000 A4.02 KJA:
Ability to manually operate and/orandlor monitor in the control room:
Control rod drive system 41.7/45.5 (CFR: 41.7 1 45.5 to 45.8)
ROISRO Rating: 3.4/3.7 Objective: CLS-LP-07 Obj. 10e 1 Oe List the conditions that will result in the following:
- e. Control Rod Block
Reference:
SO-07 SD-07 Cog Level: High Explanation:
With the Mode Switch Not in Startup or Run, the one rod out permissive must be met, all control rods must be fully inserted when select power is turned on, or the permissive is not satisfied. Once a rod is selected and withdrawn with the permissive satisfied, no other rod can be selected unless select power is turned off, then back on, but this results in a rod out block.
Distractor Oistractor Analysis:
Choice A: Plausible because it is a misconception about being able to select a different rod.
Choice B: Correct answer, see explanation Choice C: Plausible because it may be thought that since the rod is withdrawn it may give a insert block.
Choice Choice D:0: Plausible because because itit may may be be thought that that itit would prevent prevent withdrawal withdrawal of of the the inserted inserted rods and and prevent prevent insertion insertion of of the withdrawn withdrawn rod.
Notes Notes
OPEN IN START*UP' ClOSED IN REFUEL m
CLOSED WilEN ALL RODS FULL IN 1
- K4A I!.~1" ' '"' '
All.LROOS fUlLiN O
C
)
RERL REFUEL T1RT4JP SiAAT*UP MODE R£fUELMODE MODE MODE IIlODE ON~
QNQQPMJV ROO P!;RMISliIVE AUX. AU X,
FIGURE 07-15 FIGURE Rod Rod Withdrawal Block Block Circuitry (Channel (Channel A)
'OP=N ',\li"H RF SAJCGEOVER CORE
""""""1I1T C,=EtlIN <l7.'iRi"UP .'N':I M:~i:TH~~_ONE cr ROO CUT Opens on K23 cl_osao ,;:Cs=\ Rr MelDE ONE RC,=' FEAMICCPtE l:
.1.. ON IN .,,=,-UE~
O,.1::N
,.ND CHIJ'700'.'JN CHCTh t;W'i"':
OPEN WiTh ,"~ PL....TFM GFMPPL= 1_0"'0 > 4crc=
N CLSO IN AEFl.JEL AND CT,o,RTl_li'
'I "'::"': I CL "1;\'riEl*.J CLEf:' WREN RM R..r..N~E tR\' R.QE
- . OR ~\SC*.iE CN*3CL r;;r ",,.,,,,\II.
ReM'" .
~'FULL!
FLCW:-II, 'J FOWERHI.
CRfNOP ,
- w"w"" ** " ***** **************"**,, ************.*** .!.~~.. q,j BYAE O?'EN ON :::0\1 I e'(P.\3S OP5NON3CV H'!I..SIIEL OE=N=RGIZE
-To TO=1..0CI(
.OCK 80-07 1SD-07 Rev. 66 Page 57 of 57 571 Categories K/A:
KIA: 234000 A4.02 Tier / Group: T2G2 RO Rating:
RORating: 3.4 SRO SRORating: Rating: 3.7 LP LPObj:Obj: 7-bE 7-lOE Source: BANK Cog Level: HIGH mGH Category 8: Y
- 23. The following "Blue Blue Bar" Bar annunciators are received while performing OPT-11.1.2, OPT-I 1.1.2, Automatic Depressurization System and Safety Relief Valve Operability Test:
SPTMS D/V DIV /I BULK WTRWTR SETPO/NT SETPOINT TS1 TSI SPTMS D/V DIV "II BULK WTRWTR SETPO/NT SETPOINT TS1 TSI Which one of the following identifies the correct interpretation of Suppression Pool temperature as it applies to receiving the above annunciators?
Suppression Pool temperature has just reached the annunciator setpoint of:
Ak 95°F.
B 100°F C 105°F D. 110°F.
Feedback K/A: 239002 A4.04 KIA:
Ability to manually operate and/or andlor monitor in the control room:
Suppression pool temperature 41.7 / 45.5 to 45.8)
(CFR: 41.7/45.5 RO/SRO Rating: 4.3/4.3 Objective: CLS-LP-302M Obj. 1 1 Given plant conditions, determine if the following AOPs should be entered:
- c. AOP-.30 AOP-30
Reference:
APP UA-12 5-4(5-5)
Cog Level: low Explanation:
This alarm setpoint is 95°F.
Distractor Analysis:
Choice A: Correct Answer, see explanation Choice B: Plausible because this is a homogeneous setpoint distractor Choice C: Plausible because this is the setpoint for SPTMS D/VDIV I BULK WTR TEMP SETPT TMAX Choice D:
0: Plausible because this is the setpoint for Boron Injection Initiation Temperature (BIIT)
Notes
11 (f Ti 0 L 0
--. II H i-i
-o H
- 1. t
- POSSIBL~
C 1.
N :2 _
H H
[/ h
'lAPP-UA-12 H
H r : C) H H H I I) I C) P C) H 0 I F) P ( )
DE'VICE,! S~,TPOIN7S D~VICE/SETPOINTS (P P1 II Cu P Lii II) C 1I II Eli (P 1 0 LDP Lii aj 0 H rq *C) LI LI P p. H J 0 CLI P1 Ii Il II 1 Lii N II P1 fl 0 b i 0 Ci 0 H L H C) L H C I- LI H n
.PCSSIBL3: PLl'.NT 3:FFECTS PLANT EFfECTS H LII Li) II I- iLi I i i CiP Iifl0 in H
En -t0p. CL U) 0 0 HI1iIi'JCI 0 e!{.ceeds 110'=:? '"
- LI LI LIt LI LI Il0 C) U) CI Ii ru o l.I : ç LI Ci ip ii This anr.!unciator is ~equired to be operabl~ to su:ppc,rt Supp.ressi::;n LII I LILLILU I))
}fanual Reactc,r Scram P H H P H H Lii [1 II ii H CLIII LI I SPTHS DIV II BULK i\lTR TEHP SETPT Tt-IAX Ciii CD ICLI I CC, Ii I PPiIi ii I Manual Reac'tor Scram required if suppression pool t;emperature SPTHS l1icroprocessor CAC-TY-,4426-2 LI C) it p. III req~ired
- p. [s.) U)i P.
1tCf Rev. 28 H co HflO H Hi OLl It,
- p. L:y 0 II U)
(Hi L 0 0 I) C) 0 Poo o LI 00I II LII IPIPlI ID 1LILli iii UI Ilillib Lii III b o H ci P II 0 Chamber rempera!;.l;,r~ In.s-cru!':'Ie!lta-cic)n c,perabili t":.l; ,an~uncia.'Cc,,r C) 1.11 C-fl lilt 0
V < o U ()
H ip l H Ii U) LI It h1It uiiPCl ci iitIp ill III
[1,1.1 :)i:l C exceeds 110°F. if suppressic!::l pool cerr.pera:cu.re (P ,j,. ID iii I Lii Page 1 c,f 1 Lii Page 5'1 of C-)
inc,perability ro,';':,.* res'.ll;:; in a TID! Comper::sacory He.as'llre, 661 LI
From PCCP:
From PCCP:
11OF BEFORE SUPPRESSION POOL TEMP REACHES 110"F REACTOR SCRAM REQUIRED INITIATE A REACTOR SCRAM AND ENTER EOP* 01 SPIT* 09 Categories K/A:
KIA: 239002 A4.04 Tier / Group: T2Gl T2G1 RO Rating: 4.3 SRORating:
SRO Rating: 4.3 LP Obj: 302M-1C 302M-IC Source: BANK Cog Level: LOW Category 8: Y
- 24. The
- 24. DFCS control The DFCS control signal signal input input to to 2A REP has 2A RFP has been been lost.
lost. The The RO RO observes observes the the following:
following:
RFP AA CONTROL RFP CONTROL TROUBLE TROUBLE alarm alarm isis received received RFP AA Manual/DFCS RFP Manual/DFCS selectorselector switch switch isis inin DFCS DFCS DFCS Control DFCS Control light light for for RFP REP AA onon XU-1 XU-1 isis out out Which one Which one ofof the the following following describes describes how how RFP REP 2A will respond, 2A will respond, and and what what operator operator action isis required action required by 2APP-UA-13, RFP by 2APP-UA-13, RFPA CONTROL TROUBLE, A CONTROL TROUBLE, to to adjust the adjust the speed speed ofREP2A?
of RFP 2A?
REP 2A RFP 2A speed speed will will (1)
(1) operator can The operator control RFP can control RFP A A speed speed by by (2)
(2)
(1) drop A. (1) drop to to the idle speed setpoint idle speed setpoint (2) operating the RFP REP A Raise/Lower control switch B
B~ (1) remain at the last known demand (2) operating the RFP REP A Raise/Lower control switch C. (1) drop to the idle speed setpoint (2) placing the RFP REP A Speed Controller in Manual and adjusting the output demand D. (1) remain at the last known demand (2) placing the RFP A Speed Controller in Manual and adjusting the output demand
Feedback Feedback K/A: 259001 KIA: 259001 A2.06 A2.06 Ability to Ability to (a) predict the (a) predict the impacts impacts ofof the following on the following on the the REACTOR REACTOR FEEDWATER FEEDWATER SYSTEM; SYSTEM ; and and (b) based (b) based on those predictions, on those predictions, use procedures to use procedures to correct, correct, control, control, or or mitigate mitigate the the consequences consequences of those of those abnormal abnormal conditions conditions or or operations:
operations:
Loss of Loss of A.C.
A.C. electrical electrical power power (CFR: 41.5 (CFR: 41.5/145.6) 45.6)
RO/SRO Rating:
RO/SRO Rating: 3.2/3.2 3.2/3.2 Objective: CLS-LP-32.3 Objective: CLS-LP-32.3 Obj. Obj. 10j lOj Given plant Given plant conditions conditions and and one or more one or more of of the the following following events events use plant procedures use plant procedures to to determine determine thethe actions required required to control and/or mitigate mitigate the consequences of the event:
J. Loss of signal from the DFCS
Reference:
UA-13
Reference:
UA-1 3 6-56-5 Cog Level: high Explanation: UPS supplies power to the controls.
From OP-32, Section 5.7.2 (Notes)
IF RFPT B(A) MANIDFCS MAN/DFCS selector switch is in DFCS, AND the DFCS control signal subsequently drops below 2450 rpm, OR increases to greater than 5450 rpm, THEN Woodward 5009 digital controls will automatically assume RFPT speed control and maintain current speed. In this condition, the RFPT will only respond to LOWER/RAISE speed control switch commands From APP UA-13 6-5 (RFP A Control Trouble)
IF RFPT 2A DFCS CTRL light on RTGB XU-1 is NOT illuminated, THEN attempt to control RFP turbine speed as necessary using the LOWER/RAISE speed control switch Distractor Analysis:
Choice A: Plausible because the woodward manual control signal automatically tracks the DFCS output signal. An operator without this knowledge could believe the RFP REP speed would drop to minimum woodward control speed with the DFCS control signal failed Choice B: Correct answer, see explanation Choice C: Plausible because the DFCS control signal has failed. with the DFCS Control light out, the RFP is under manual control of the woodward governor and adjusting the output of the individual RFP REP Speed Controller will have no effect. An operator without understanding of the hierarchy of the REP RFP control system could believe this choice is correct.
Choice D: Plausible because the DECS DFCS control signal has failed. with the DECS DFCS Control light light out, the REP RFP is under under manual control of the woodward governor governor and adjusting adjusting the output of the individual REP RFP Speed Controller will have no effect. An operator without understanding of the hierarchy of the REP RFP control system system could could believe believe this choice is is correct.
Notes JCTIONS 1.
- 1. Info~"lll Turbine Infoi.-m Building AC)
Turbine Building AO of o£ alarm iniiaiicn initiatic([). and reqJ~est and request inveetigation in...<estigation of ala~"lll condition of alarm condition.
2.
- 2. Ncnitor Monitor reactor reactor water water level level and and feedwater feedwater flow
.fler,.; for fo.r possible possible loss loss of A _l,. RFP RFP.
3.
- 3. XE IF RFPT RFPT 2A 2A VFCS DFCS CTRL light on CTRL light on RTCB RTGB XtJ-i XO-l is is NOV NOT illuminated, TH~l illuminated, THEI attenpt attempt to to control control REPRFP turbine turbine speed speed as as neceasary necessary using using the the LOWERI_~ISE speed LONER/RAISE control swatch speed control switch.
4.
- 4. Refei-Re.fe.r to to C)AOP-23C)
O_r..OP-23.0.
2APP-UA-13 12APP-UA--13 Rev.
Rev. 43 43 Page Page 9901 1081 99 of 108
NOTE:
NOTE: IF RFPT IF RFPT B(A)B(A) MANIDFCS selector switch MAN/DFCS selector switch isis in in DFCS, DFCS, AND AND the the DFCS DFCS control control signal subsequently drops signal subsequently drops below below 2450 2450 rpm, rpm, OROR increases increases to to greater greater than than 5450 rpm, 5450 rpm, THEN Wooclward 5009 THEN Woodward 5009 cligital digital controls controls willwill automatically automatically assume assume speed control RFPT speed RFPT control and and maintain current speed.
maintain current speed. In In this this condition, condition, the the RFPT RFPT will only will only respond respond to LOWER/RAISE speed to LOWER/RAISE speed control control sljvitch switch commands commands until until the the MANlDFCS selector s\vitch MAN/DFCS selector switch is is placed placed inin MAN MAN,. DFCS DFCS CTRLCTRL RESET RESET pushbutton is pushbutton is depressed, AND the depressed, AND MAN/DFCS selector the MANIDFCS selector switch switch returned returned toto DFCS.
DFCS.
Load: 120V UPS Distribution Pane12-V10A Load: Panel 2-V1OA Location: Control Location: Control Building Bu[ldwIg 49' 49 SW Drawing
Reference:
Drawing
Reference:
F-03027 F-03027 Upstream Power Upstream Power Source:
Source: 120V UPS 120V UPS Distribution Distribution Par Par CKT DESCRIPTION LOAD DESCRIPTION I
3 Unit 2 FVVEW Control System:
RFPT A & 8lMain 8/Main Turbine High Level Trip Circuit 'W'A MV/I converters for 2-C32-TE-N006A II/lV/I 2-C32-TE-NOO6A 2-C32-TE-NOO6B and 2-C32-TE-N0068 Digital FWCS Rx Scram 8B Input Power supplies:
2-C32-ES-5782A & 8B (Digital FWCS) 2-C32-ES-5782A 2-C32-ES-5783A & 8 2-C32-ES-5783A B (Digital FWCS) 2-C32-ES-5784A & B (Digital FWCS) 2-C32-ES-5784A 2-C32-ES-5788A & 8 2-C32-ES-5786A (Digital FWCS)
B (Digita!
2-C32-K620 for 2-C32-PT2-C32-PT-N -N007 007 (Turbine Steam Flow) and 2-C32-PT-N 2-C32-PT-N008 008 (Reactor Pressure) 4.2.4 DFCS Control Signal Failure IT the 5009 control system detects that the Remote Speed Setpoint If (RSS) from thetile DFCS is outside the failure limits, an RSS signal failure condition is set and, itif the 5009 control system was in the DFCS mode, an automatic transfer to the manual mode will occur.
The RFPT speed setpoint (and (ancl hence hence RFPT speed) will be maintained at the last good "good" value and can be controlled using the Panel XU-1 RAISE IJ LOWER switch (Figure 32.3-14). The Tile MANUAL /I DFCS switch should be placed in the MANUAL position.
IISD-32.3 SD-32.3 Rev Rev 3 I Page 66 of 123 66 of 123 I Categories Categories K/A:
KIA: 259001 259001 A2.06 A2.06 Tier/Group:
Tier / Group: T2G2 T2G2 RO Rating:
RO Rating: 3.2 3.2 SRO Rating:
LP Obj: 32.3-1OJ 32.3-lOJ Source:
Source: BANK BANK Cog Level:
Cog Level: HIGH HIGH Category Category 8:8: Y Y
- 25. Unit One is operating at rated power when the Feedwater Flow B indicator has failedfaNed upscale.
Which one of the following identifies the effect this condition will have on reactor water level control with no operator actions taken?
Av A~ DFCS transfers to 1-element control and maintains current level.
B. DFCS transfers to 1-element control and RFPs reduce feedwater flow causing a X )(
reactor scram on low level.
- C. DFCS remains in 3-element control and maintains current level.
D. DFCS remains in 3-element control and RFPs reduce feedwater flow causing a \:-
reactor scram on low level.
Feedback K/A: 259002 K6.04 KIA:
Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM:
feter flow input Reactor feedw.@ter (CFR: 41.7/45.7) 41.7 / 45.7) ,
ROISRO Rating: 3.1/3.1 RO/SRO Objective: CLS-LP-32.2 Obj 7b Given plant conditions, determine the response of the DFCS to the following events:
- b. Loss of any feed flow input
Reference:
APP A-07 4-2, FW CTL SYS TROUBLE Cog Level: high Explanation:
The following signals are the permissives to operate in 3 element control:
All steam flows (4) outputs are valid (within 10% of avg)
All feed flows (2) outputs are valid (within 10% of both)
Master control station in Automatic At least one (1) Feed pump control station is in Automatic Reactor Power is > > 20%
Feed flow and Steam flow matched I element control. Level will be maintained based on level with the feed flow failure this will transfer to 1 only.
Distractor Analysis:
Choice A: Correct answer, see explanantion Choice B: Plausible because it will transfer to 1 I element control but the feed flow failure will not cause DFCS to lower the output of the feed pumps based on the upscale failure of the feed flow instrument. This is an input to the three element control signal. Which would on a slow raising start to lower but when it goes to 1 I elemeQt th~LJgent may think that since elemel?t that signal is removed. th.udent ~in~j!J:!~s ithas a high flow condition that feed flow may be reduced causing a scram on low level.
Choice C: Plausible because level will be maintained at the current setpoint but it will transfer to single element control on greater than 10% difference between feed flow signals.
Choice D: Plausible because feed flow is an input to 3 element control and the feed flow failure will not cause DFCS to lower the output of the feed pumps based on the upscale failure of the feed flow instrument, the student may think that since it has a high flow condition that feed flow may be reduced instrument.
causing a scram on low level. It will transfer to single element control.
Notes WHITE 4-2 SYSTRBL FW CTL SYS TRBL Page 11 of2 012 1.0 OPERATOR ACTIONS:
1.1 CONFIRM which vThch condition is causing this alarm by observation of the components listed under the causes section.
1.2 ,(j!,§gB!l§~~~Y~l:~gfigr~~)
OBSERVE Automatic Functions:
1.21 One or more of the following automatic actions may have occurred:
- l.2.1
- 1. Possible RFP speed locked at speed signal sensed at time of failure.
- 2. Transfer to redundant digital feedwater control channel. channel 3*t~~!~~ag~~iQgl,~t~m~iif~~fi~D
- 3. Transfer to single element control
- 4. Transfer to opposite level transmitter (C32-L T-N004A1B)
(C32-LT-NOO4AIB)
- 5. Turbine trip and possible reactor scram from high reactor water level.
5.
2.7 ;Rf1~~9ii~~~l~~r~l~iriim!ft~~iI9il§)
2.7 Reactor Feedwater Flow Transmitters A or B:
2.71 2.7.1 High (8.0 (80 Mlbs/hr) 2.T2 2.7.2 Low (0.2 Mlbs/hr)
Mlbslhr) when greater than 10% total feed flow.
How.
2.7.3 Greater than a 10% mismatch between the averaged feed flow.
2.7.4 Rate of Change 1.6 MlbslHrlSec Mlbs/Hr/Sec 1 IAPP-A-07 1APP-A-07 33 Rev. 33 Page 25 of 451 45 From SO-32.2 SD-32.2 42A 4.2.4 There are 2 normal feed flow inputs that are summed to provide an output signal to the following and dependent upon initial power level and severity seventy of failure the following may may occur:
occur Auto transfer to I element operatIon resulting from real alarm block criteria being exceeded OR individual feed flow not within
, _'.VN.,.of~~I~!I
...10% average~lm~~~~~OR feed flow OR total feed flow now <::
< 20%.
Recirc pump runback if total feedwater flow goes <:: < 16.4%
Hydrogen Water Chemistry injection solenoids trip if total feed flow 17.3%.
<17.3%.
Hydrogen Water Chemistry may trip on external setpoint step change (>5 SCFM)
With the'0[j~~~!!mltifi'9J\~i:~I~t~~uf9n the!,~Qm~~tQ~!i~~~~) the control signat is gérierated based~~cm
" ,.. ,.\,>.,;{,)"",,\:,/?;;8:;~*;;";.,
Steam and feedwater mass flow rates are not used to modify the level signal.
Categories K/A:
KIA: 259002 K6.04 Tier!/ Group: T2Gl Tier T2G1 RO Rating:
RORating: 3.1 SRORating: 3.1 SRO Rating:
LP Obj: 32.2-7B Source: NEW Cog Level: HIGH Category 8: Y
Unit One is
- 25. Unit is operating at rated power when the Feedwater rated power Feedwater Flow Flow BB indicator indicator has has failed upscale.
upscale.
Which one of the following identifies the effect this condition will have on reactor water level control with no operator actions taken?
A DFCS transfers to 1-element control and maintains current level.
A-:I B. DFCS transfers to 1-element control and RFPsREPs reduce feedwater flow causing a reactor scram on low level.
C. DFCS remains in 3-element control and maintains current level.
D. DFCS remains in 3-element control and RFPs reduce feedwater flow causing a reactor scram on low level.
Feedback Feedback K/A: 259002 KIA: 259002 K6.04 K6.04 Knowledge of Knowledge the effect of the that aa loss effect that loss or or malfunction malfunction of of the the following following will will have have on on the the REACTOR REACTOR WATER LEVEL WATER LEVEL CONTROL CONTROL SYSTEM:
SYSTEM:
Reactor feedwater Reactor feedwater flow flow input input (CFR: 41.7/45.7)
(CFR: 41.7 I 45.7)
RO/SRO Rating:
RO/SRO Rating: 3.1/3.1 3.1/3.1 Objective: CLS-LP-32.2 Objective: CLS-LP-32.2 Obj 7b Obj 7b Given plant Given conditions, determine plant conditions, determine thethe response response ofof the the DFCS DFCS to the following to the following events:
events:
- b. Loss
- b. Loss of any feed flow input input
Reference:
APP A-07 4-2, FWen FW CTL SYS TROUBLE TROUBLE Cog Level:
Cog Level: high high Explanation:
The following signals are the permissives to operate in 3 element control:
All steam flows (4) outputs are valid (within 10% of avg)
All feed flows (2) outputs are valid (within 10% of both)
Master control station in Automatic At least one (1) Feed pump control station is in Automatic Reactor Power is > 20%
Feed flow and Steam flow matched with the feed flow failure this will transfer to 1 1 element control. Level will be maintained based on level only.
Distractor Analysis:
Choice A: Correct answer, see explanantion Choice B: Plausible because it will transfer to 1 1 element control but the feed flow failure will not cause DFCS to lower the output of the feed pumps based on the upscale failure of the feed flow instrument. This is an input to the three element control signal. Which would on a slow raising start to lower but when it goes to 1 1 element that signal is removed, removed. the student may think that since it has a high flow condition that feed flow may be reduced causing a scram on low level.
Choice C: Plausible because level will be maintained at the current setpoint but it will transfer to single element control on greater than 10% difference between feed flow signals.
Choice D: Plausible because feed flow is an input to 33 element control and the feed flow failure will not cause DFCS to lower the output of the feed pumps based on the upscale upscale failure of the feed flow instrument, instrument. the student student may think that since itit has aa high flow condition that feed flow maymay be reduced causing a scram on low low level. It will transfer to single element control.
Notes Notes WHITE 4-2 FWCTLSYSTRBL FW CTl SYS TRBl Page Page 11 of2 ot2 1.0 OPERATOR 1.0 OPERATOR ACTIONS:
ACTIONS:
1.1 CONFIRM 1.1 CONFIRM which condition condition is is causing causing this alarm alarm byby observation observation ofof the the components listed components listed under under the causes causes section.
section.
1.2 OBSERVE 1.2 OBSERVE Automatic Automatic Functions:
Functions:
1 .2i One or more of the following automatic actions may have occurred:
'1.2.1
- 1. Possible RFP
- 1. RFP speed speed locked at speedspeed signal sensed at at time of failure.
- 2. Transfer to redundant redundant digital feedwater control channel.channel.
- 3. Transfer to single element control
- 4. Transfer to opposite level transmitter (C32-L (C32-LT-NOO4A/8)
T-N004A/B)
- 5. Turbine trip and possible reactor scram from high reactor water level.
2.7 Reactor Feedvmter Feedwater Flow Transmitters A or B: B:
2.7.1 High (8.0 Mlbs/hr) 2.7."1 Mlbsihr) 2.7.2 Low (0.2 (02 Mlbs/hr) when greater than "10 10%%
total feed flow.
2.73 Greater than a 10% mismatch between the averaged feed flow.
2.7.3 2.7.4 Rate of Change Chanae '1.6 Mlbs?Hr/Sec 1.6 Mlbs/Hr/8ec 1"lAPP-A-07 IAPP-A-07 Rev. 33 451 Page 25 of 45 SD-32.2 From 8D-32.2 4.2.4 Loss of Any Feed Flow Input There are 2 normal feed flow inputs that are summed to provide an output signal to the following and dependent upon initial power level and severity of failure the following follO\"Iing may occur:
Auto transfer to I1 element operation resulting from real alarm block criteria being exceeded OR individual individual feed flow not within
-10% of average feed flow OR total feed flow now << 20%.
10%
Recirc pump runback ifif total feedwater flow goes "16.4%
goes << 16.4%
Hydrogen Water Chemistry injection solenoids trip ifif total feed flow Hydrogen
<173%.
< 17.3%.
Hydiogen Hydrogen Water Chemistry may trip trip on external extemal setpoint step change (>5 change (>5 SCFM)
SCFM)
With the DFCS With the DFCS in in single element control, single element the control control, the control signal signal is is generated generated based based only on reactor water level.
only on level. Steam and feedwater mass Steam and mass flow rates are rates are not not used to modify used to modify the level signal.
the level signal.
Categories Categories KJA:
KIA: 259002 K6.04 259002 K6.04 Tier // Group:
Tier Group: T2G1 T2G1 RO Rating:
RORating: 3.1 3.1 SRO Rating:
SRORating: 3.1 3.1 LP Obj:
LP Obj: 32.2-7B 32.2-7B Source:
Source: NEW NEW Cog Level:
Cog Level: HIGH HIGH Category 8:
Category 8: Y Y
- 26. Unit Unit Two is is operating operating at rated power at rated power with Standby Standby Gas Gas Treatment Treatment (SBGT)
(SBGT) System System controls aligned controls aligned as as follows:
TrainAin Train SYSTAPREF A in SYST A PREF Train B B in in STBY STBY Drywell cooling is lost and the reactor scrams on high drywell pressure. Reactor water level drops to 130130 inches and is now rising.
Which one of the following identifies the SBGT system flows that the RO would verify on the XU-51 panel?
SBGT Train A A flow verified to be _--->.(..:..1 (1))1.-_ SCFM and SBGT Train B flow verified to be (2) SC FM.
SCFM.
(1)00 A. (1)
(2) 0 (1)0 B. (1) 0
-P3300 (2) -3300 C
C~ 3300 (1) -3300 (2) 0 3300 D. (1) -3300 3300 (2) -3300
Feedback Feedback K/A: 261000 KIA: 261000 A3.01 A3.01 Ability to Ability to monitor monitor automatic automatic operations operations ofthe of the STANDBY STANDBY GASGAS TREATMENT TREATMENT SYSTEM SYSTEM including:
including:
System flow System flow (CFR: 41.7/45.7)
(CFR: 41.7 /45.7)
ROISRO Rating:
RO/SRO Rating: 3.2/3.3 Objective: CLS-LP-10 Obj. 4 Given plant conditions determine if SBGTs should have initiated.
SD-b 1/ OP-1
Reference:
SO-10 OP-lO 0 Cog Level: high Explanation:
With the B SBGT train in STBY it will not auto start, this is more like an Off position. There is an auto start signal from high OW DW pressure and flow should be verified to be greater than 3000 scfm. The control room indicators scale is from 0 to 4500 scfm. Normal system flow is -3300 scfm.
Distractor Analysis:
Oistractor Choice A: Plausible because if there was not an initiation signal this would be correct.
Choice B: Plausible because there is an initiation signal but only one train will operate. If the examinee thinks only B only will start then it would indicate 3300 scfm.
Choice C: correct answer, see explanation Choice 0: D: Plausible because there is a initiation signal and the examinee may think that both trains would initiate. The standby position is a common misunderstanding, this is actually an OFF position.
Notes Notes From OP-10:
From OP-lO:
SBGT Systern S8GT System isis normally normally in in standby standby with with the SBGTA(8) the SBGT ArB) control control switch switch in in PREF.
PREF.
However, when However, when thethe SBGT SBGT System System is is in in operation, operation, the the foliOllvlng folloing parameters parameters andand limits should limits should be be observed:
observed:
(B) Flow Flow Greater than Greater than or or equal equal to to VA-Fl-3150-i VA-FI-3'! (Fl-3151-1) 50-'1 (FI-315'1-1) 3000 scfm, 3000 scfm.
Panel XU-5'1 Panel XU-51 VA-Fl-3150-2 (FI-3'151-2)
VA-FI-3'J50-2 (Fl-3151-2)
Local Local 3.3 3,3 the following signals will automatically start Any of the start the Standby Gas System:
Treatment System:
3.3.1 3,3,1 High drywell pressure 3.3.2 3,3.2 Low Level Level22 3.3.3 3,3.3 High radiation in in the Reactor Building exhaust ventilation duct 3.3.4 3,3.4 Reactor Building (RB) exhaust temperature high Reador 3.3.5 Main Stack Radiation Monitor High-High 3.4 The Standby Gas Treatment System wili will NOT automatically start if the control switch is in STBY, STBY.
Categories K/A:
KIA: 261000 A3.01 Tier / Group: T2G1 RO Rating:
RORating: 3.2 SRO SRORating:
Rating: 3.3 LP Obj: 10-4 Source: NEW Cog Level: HIGH Category 8: Y
- 27. Which one of the following identifies the affect on the manually initiated, automatically executed fast bus transfer capability if a trip of the feeder breaker to 125V I 25V DC panel 9A occurs?
(1)1-)_ _ if attempted for bus 11 B.
The bus transfer will _---1.(..:..1 The bus transfer will (2) if attempted for bus 1IC. C.
(1)occur A. (1) occur (2) occur B (1) occur B!'"
(2) not occur (1)notoccur C. (1) not occur (2) occur (1)notoccur D. (1) not occur (2) not occur Feedback K/A: 262001 K6.01 KIA:
Knowledge of the effect that a loss or malfunction of the following will have on the A.C.
ELECTRICAL DISTRIBUTION:
D.C. power (CFR: 41.7/45.7)
ROISRO Rating: 3.1/3.4 Objective: CLS-LP-50.1 Obj 7 Given plant conditions, predict the effect a loss of DC control power will have on the 4160 VAC System.
Reference:
01-50 Cog Level: low Explanation:
BOP Bus 1 I B has AUTO control power transfer capability where 1 1 C and 1ID D do not.
Distractor Analysis:
Choice A: Plausible because the auto transfer of control power will occur on I1B, but will not on IC 1C and D.
Recent plant mods have removed some of the auto transfer capabilities on some of the DC control power arrangements (E-busses require a manual transfer of control power).
Choice B: Correct answer, see explanation.
Choice C: Plausible because the examinee may have the logics reversed.
Choice D: Plausible because the auto bus transfer will not occur on 11C and D, but it will on I1B. Recent plant mods have removed some of the auto transfer capabilities on some of the DC control power arrangements.
Notes PANEL: BA PANEL:9A LOCATION: NORhfALSUPPLY:
NORMAL SUPPLY: ALTERNATE SUPPLY:
ALTERNATE Reference Draviing:
Reference LL-30024-12 Draair LL..3D024-12 Unit 1 Turbine BuiIdn, 20. Swtchbcard is Switchboard IS Switchboard 1A S\*litchboard IA Switch gear area (MecllanlCallll!iertOO\)
II@OC}
CKT LOJ) EfFECT EFFECT 3 .rt1;ear5us I5COIrIFO% 1. AUlemalle AucnllceuT f1oa:errIepe:.Parei2A Bus Tro:flf>:iE!r!o.a~emiitiE! po'.\W, Parter 11l1\ ~lt 11.
- . 11.
a.",", ** 'T1oo' __ ._."** 1 .... ,.," ***-1: . . . .". _ _ ** 111 ** ,," *
- 1 . . . . . ,.
5uE IC Cn: Der LO3 or
- 1. loss
- oonlrol po't.-er:)J 41>\"1 rcoIru 4KV 10;;(15 Id on r Sus Ie. C.
6 . Loss Lcss or LKV K!beaKrqr3tbr.
breaker op:r.a!IOR manual riaor.uLoii;c.
or Julnl1alc.
21 tear Sus Sv.itC!1ge.ar 5u5 10 Crro POWEr 1D Carr.rot ?r j1 Loss c( oontrCl pc",'iE!T;O coonclç ver:o 4K'/llliids 4< baI-r en ,Sus5u9 10. ID.
12 LoSS Loss c( c LKV KV breaker cprII:r man~a1 bre3er cf>2ratloli. raiia or automa':~
, ._a _.0' ... 11 ....... "",..... <_ ." ..u ____ ... n,,, ... 1""l ....... lit'! .. ~ _~_
I 1001-50
°°-° Rell.45 I Page PagcH3256 eil 321 Categories K/A:
KIA: 262001 K6.0K6.011 Tier / Group: T2G T2G11 RO Rating:
RORating: 3.1 SRO Rating: 3.4 LP Obj: 50.1-7 Source: BANK Cog Level: LOW Category 8: Y
- 28. Which one of the following correctly completes the statements below if a Loss of Offsite Power (LOOP) occurs on Unit Unit Two with DG4 under clearance?
RHR Pump 2B (1) lost power.
2-E11-FOI5B, LPCllnboard 2-E11-F015B, LPCI Inboard Injection Valve, _-.l.(.=.2)1-_lost (2) lost power.
A. (1)(1)has has (2) has B (1) has B!'"
(2) has not C. (1) has not (2) has D. (1) has not (2) has not Feedback K/A: 262001 K6.02 KIA:
Knowledge of the effect that a loss or malfunction of the following will have on the A.C.
ELECTRICAL DISTRIBUTION:
Off-site power 41.7 /45.7)
(CFR: 41.7/45.7)
ROISRO Rating: 3.6/3.9 RO/SRO Objective: CLS-LP-39 Obj 9c Describe the effects on the plant if one or more of the EDGs failed to start during the following conditions:
- c. LOOP SD-i 7
Reference:
SD-17 Cog Level: high Explanation:
B RHR Pump receives power from E4 which is feed by DG #4. since it has failed then E4 would be B
de-energized and the BB RHR Pump would have no power. The B Loop injection valves are powered from the same division, but opposite units E Bus. That would be E2, which does have power for the injection valves and D RHR pump.
Distractor Analysis:
Choice A: Plausible the B B Pump has lost power and one would logically think that the B Loop would be powered by the Div II power source, which is correct except that it is from Unit 11 Div II.
Choice B: Correct answer, see explanation Choice C: Plausible because the D D RHR has not lost power and one would logically think that the B BLoop Loop would be powered by the Div II power source, which is correct except that it is from Unit 11 Div II.
II.
Choice D: Plausible because the D D RHR has not lost power and the injection valve has has not lost power.
Notes UNIT 2 LOW PRESSURE ECCS EGGS IJ2C8A
() T
~ - - ~ __
,-_.!.,- ----~ _ _ _ .f -: L - -
r I ' - - _ , - ____ I :
,," _J I
DIU N0Th INJEClION NOTE: INJEC11ON FLOW PATH AND POWER SUPPLIES SHOWN.
OTteR R.OW LOGIC & OTtER FLOW PAlIfS PAThS NOT SHOWN.
1SO-17 SD-i7 Rev 13 Rev. 1271 ot 127 Page 99 of Categories KJA:
KIA: 262001 K6.02 Tier!/ Group: T2Gl Tier T2G1 RO Rating:
RORating: 3.6 SRO Rating: 3.9 SRORating:
LP Obj: 39-9C Source: NEW Cog Level: LOW Category 8: Y
- 29. Which one of the following identifies an instrument that is powered from UPS and is required by Technical Specification 3.3.3.1, Post Accident Monitoring Instrumentation?
A. Drywell Rad Monitor B. Rod Worth Minimizer C~
C Reactor Vessel Pressure recorder D. Shutdown Range reactor water level indicators Feedback K/A: 262002 G2.04.03 KIA: 02.04.03 Ability to identify post-accident instrumentation.
Uninterruptable Power Supply (A.C'/D.C.)
(A.C./D.C.)
41.6/45.4)
(CFR: 41.6 145.4)
R0/SRO Rating: 3.7/3.9 RO/SRO Objective: CLS-LP-01.2 Obj. 12 Given plant conditions and TS, including Bases, TRM, ODCM, and COLR, determine wether given plant conditions meet minimum TS requirements associated with Reactor Vessel Instrumentation System.
Reference:
01-50.5, TS 3.3.3.1 Cog Level: Low Explanation:
TRM 3.4 identifies the PAM instrumentation requirements. The reactor vessel pressure recorder and transmitters that feed them are in the table and are powered from UPS. DW Rad Monitor is not powered from UPS, but is in the TS table. The RWM is powered from UPS but is not in the PAM TS but is TS related. The N026s are in the TS PAM table and are not powered from UPS.
Distractor Analysis:
Choice A: Plausible because this is a TS PAM instrumentation, not powered from UPS, powered from 31AB Choice B: Plausible because this is powered from UPS and is TS related, but not PAM TS.
Choice C: Correct answer, see explanation Choice D: Plausible because this is a TS PAM instrument, but not powered from UPS.
Notes
Instrumentation PAM Instrumentation PAM 3.3.3.1 T3':1:~ 3.3.l.1-1 3.3.-I (~';!f!
>e 1I 011.1 A:c :erl ,\;'>>:f'1r.:YA"':~
Ft~ 1"::C.~1I!1i1 rIx In:'T~~n;~On 3fl
- U
- 0CNDFTICS CCNOITICt;S REFERE?-.~SO RPREN:D RE2JEC REQUIRED FROi RSQUIRED Fi.OM EUiRED FCThY4 Fur;C7.0N ~"""NELS 1-.M1L 7!CN C.!
AC7ICND.f
. Re3Ct:>>
R3I2-ie: r:Ire
!le~:el Fre1:ure E 1 ctr 5:eI Le,e
. -l4Chflb.lE0II:I?. 2 E
. C Irh: t *210 In:Ie 2 E C. D C brCh! 2 E CIaner WIr LeI 2 E
- 4. CIrtr !r 2 E
, 2 E
. DJIreure 2 E E
7, 2 E
. FOl. cItb, ~ 1='"
2rrI!crt:ene!roll:<: E 1'CWOtt,I*1'X'
.*cw E.
$, FcIJ:.
IMl U:ed.)
C. 2 F The power supplies for the Reactor Vessel Pressure Instruments are:
621-PT-N O45AC B2'I-PT-N045AIC 125
-125 VDC 1 1(2)A (2)A PilI. Pnl. 3A & 11A 1 1A (4A 4A & '!2A) 1 2A) (XU-63) 621-Pl-R6D5A B2'l-PI-R605A 125 VDC 1 i(2)A (2)A PilI. Pnl. 3A& 3A & 11A 1 1A (4A& 'l2A) 12A) (XU-63) 521-PT-N0456/D B2'J-PT-N045BfD 125 VDC 1(2}B
-125 1(2)8 Pili, Pnl. 3B & 11 IIB B (4B(46 & '12B) 128)125125 (XU-64)
B21-PI-R6D5B B2'l-PI-R605B VDC 1(2)8 PilL Pnl.. 3B & 118 (4B
& 11B (4B& & -12B) 128) (XU-64)
C32-PT-NOO5A/B UPS Panels V9A (V'lOA) (VIGA) or DC Panels 38 3B (4B)
C32-LPR-R608 UPS Panel V7 ViA A (8A)
C32-PT-3332 125
-125 VDC 1 1(2)8 (2}8 MCC 1 1(2)XDB (2)XDB C32-Pi-3332 C32-PI-3332 125
'125 VDC 1(2)8 1(2)6 MCC 1(2)XDB 1 (2)XDB C32-PT-N008 C32-PT-NOOS UPS BUS 1 1(2)APnI.
(2}A Pnl. V9A (VIDA) (Vi OA)
C32-LPR-R609 UPS BUS 1 i(2)APnI (2)A Pnl V7A (V8A) (VSA)
120\I UPS Load: '(20V Load: UPS Distribution Distribution Pane11-V9A Panel 1-V9A (HG9) (HG9)
Location: Control Location: Control Building Building 49' 49 NW NW Drawing
Reference:
Drawing
Reference:
f-90098 F-90U98 Upstream Power Source:
Upstream Power Source 120V UPS 120V UPS Distribution Distribution Panel Panel 1-1A-UPS I IA UPS CKT CKT LOAD DESCRIPTION lOAD DESCRIPTION EFFECTS ON EFFECTS ON LOSS LOSS OF OF POWER POWER
'11 SVDC and 5VDC 28VDC Power and 28VDC Power Supplies Supplies for for Loss of Loss of rod rod position position indication indication on on four-rOd four-rod Control Rod Control Rod Position Position Indication Indication System System display display panel panel and and full full core care display, display, loss loss of of (RPIS) Cabinet (RPIS) Cabinet 'I-H12-P615; i-H12-P615; Roel Worth Rod Worlh RWM, cannot RWM, cannot move move control control rods, rods, A-05-5-2 A-OS-E-2 Minimizer (RWM)
Minimizer (RWM) NUMACNUMAC Drawer Drawer will alarm alarm 1-Cl i-CNV-55l6, 1-H12-P601 l-C11-CNV-5516, 1-Hi 2-P607 (TS 3.1.3, (TS 3.1.3. 3.3.2 3.32.l
..1)
The power supplies for the Shutdown Range Instruments are:
821 -LT-N027A B2'J-l T-N027A & 120 VAC Pnl. 1(2)AB Emerg. '120 821 -LT-7468A B2'J-l T-7468A 821 -LI-R605A B2'1-1I-R605A Emerg. '120 120 'lAC
/AC Pnl. 'i(2)AB 1 (2)AB B2'I-l T-N027B &
821 -LT-N027B & 120 VAC Pnl. "I1(2)6 Emerg. "120 (2)8 821 -LT-7468B B2'I-l T-7468B Categories K/A:
KIA: 262002 G2.04.03 Tier / Group: T2G1 T2Gl RO Rating: 3.7 SRO Rating:
SRORating: 3.9 LP Obj: 1.2-12 Source: NEW Cog Level: LOW Category 8: Y
- 30. Which one of the following identifies the power supply to the Main Turbine Emergency Bearing Oil Pump (EBOP)?
A. 480V AC Division II Emergency Bus B. 480V AC Division II Emergency Bus C. 250 VDC Division I D~
D 250 VDC Division II Feedback K/A: 263000 K2.01 KIA:
Knowledge of electrical power supplies to the following:
Major D.C. loads (CFR: 41.7)
ROISRO Rating: 3.1/3.4 RO/SRO Objective: CLS-LP-26.1 Obj. 5c Identify the electrical distribution system which powers the following:
- c. Emergency Bearing Oil Pump SD-26. 1
Reference:
SD-26.1 Cog Level: Low Explanation:
The emergency bearing oil pump (EBOP) is provided to supply oil to the bearings of the main turbine when all ac power is lost. The EBOP is driven by a dc motor powered from 250 VDC 2(1)8.
2(1)6. the 480 V busses E5-E8 are powered from the DGs so could provide emergency power.
Distractor Analysis:
Choice A: Plausible because it is an emergency pump which could be suplied from an emergency source and this source under emergency conditions does get power from the DG.
Choice B: Plausible because it is an emergency pump which could be suplied from an emergency source and this source under emergency conditions does get power from the DG.
Choice C: Plausible because it is an emergency pump which would be powered by a DC source and just not the 250 VDC Div II source.
Choice D: Correct answer, see explanation Notes Equipment Power Supply Motor suction pump (MSP) 480 VAC VAG MCC MGG 1(2) 1 (2) TM Turning gear oil pump (TGOP) 480 VAC MGG 1 VAG MCC (2) TM 1(2)
Emergency Bearing Oil Pump (EBOP) 250 VDC VDG SWGR 1 1(2)B (2)B Vapor extractor 480 VAC VAG MCC MeG 1(2) 1 (2) TM 1 SD-26.i SD-26.1 Rev. 9 Rev I Page 44 of 1071 107
Categories Categories K/A:
KIA: 263000 K2.0 263000 K2.011 Tier // Group:
Tier Group: T2G1 T2G1 RU Rating:
RORating: 3.1 3.1 SRU Rating:
SRORating: 3.4
3.4 LPUbj
LP Obj: 26.1-5C 26.l-5C Source:
Source: BANK BANK Cog Level:
Cog Level: LUW LOW Category 8:
Category 8: Y Y
- 31. Which one of the following predicts the response of the Lube Oil Temperature Control Valve (TCV) as DG3 load is increased?
Initially the DG3 Lube Oil temperatures to the heat exchanger will (1) and the Lube Oil TCV will throttle (2)
A (1) rise A":I (2) closed B. (1) rise (2) open C. (1) lower (2) closed D. (1) lower (2) open Feedback K/A: 264000 A1.01 KIA: Al .01 andlor monitor changes in parameters associated with operating the Ability to predict and/or EMERGENCY GENERATORS (DIESEUJET) (DIESELIJET) controls including:
Lube oil temperature (CFR: 41.5 /45.5)
/ 45.5)
ROISRO Rating: 3.0/3.0 RO/SRO Objective: CLS-LP-18 Obj. 18 SD-39, OP-39
Reference:
SO-39, Cog Level: High (determining the effect on lube oil temperature and applying fundamentals to system knowledge)
Explanation:
Increasing load on the OG DG will cause more heat to be put into the lube oil as the OG DG does more work. The lube oil cooler bypasses oil to raise temperature of the oil. So in this case the TCV will have to close to allow more lube oil to go into the lube oil cooler.
Distractor Oistractor Analysis:
Choice A: Correct answer, see explanation Choice B: Plausible because temperatures will have to raise on the lube oil and if the TCV controlled the outlet of the oil it would have to open to cool the oil.
Choice C: Plausible because the TCV will throttle in the closed direction to allow less oil to bypass the cooler. More load correlates to more speed which would provide less for the oil to pick heat load thereby causing a temperature reduction.
Choice D:0: Plausible becasue if the TCV controlled the outlet of the oil it would have to open to cool the oil.
More load correlates to more speed which would provide less for the oil to pick heat load thereby causing a temperature reduction..
reduction ..
NotesC
-4 closing the lube oil TCV lowers the temperature of the oil; C) 0 C) CD CD 0 CD Cl) CD 3 CD 0 CD D
CO C) CD FIGURE 1 PFr I-*~--l ~----,
I ~h lit' h"'>?
lie I
~ ..
I
' .J I
ll' ~
i~ l ~ § I t1 8 ~
8
~
I
.~ ! 8 ~
p.
~
~
'. ~
~
~l ffi ~I J "I:.;r,:
---3> ~ .:-
~
8
~ ~
D~15 ~~i'5 J -~~-
~IJ' £I~
- a r 1 11 ~
~
~:....---
--E--
~
J
~ * ,~ ** ~ ****** < ** -~ ** ~~'.
1 I e q:,
}
~
! J R I 2iI J I ~
~
~ I" 0
OOP-39 -I (C Rev. 130 ID Page 34 of 233 CO CD g C) cJ)
Categories KIA: 264000 Al.Ol Tier / Group: T2G1 o..
RORating: S. csQ 3.0 0 SRORating: 3.0 LPObj: C 39-18 Source:
C BANK 4
r Cog Level: mGH Category 8: CD o
- 32. During During accident conditions on Unit Two the following sequence of events occur:
Time (seconds) Event o0 Drywell pressure rises above the scram setpoint 22 Complete Loss of Off-site Complete Off-site Power occurs 8 Reactor pressure is 400 psig 10 DGs energize their respective E Buses 15 Reactor water level drops below LL3 Which one of the following identifies the earliest time that the LPCI pumps will auto start?
A. 8 seconds B. 15 seconds C. 18 seconds D 20 seconds D!'"
Feedback K/A: 264000 K1.07 KIA:
Knowledge of the physical connections and/or cause effect relationships between EMERGENCY (DIESELIJET) and the following:
GENERATORS (DIESEUJET)
Emergency core cooling systems 41.9 / 45.7 to 45.8)
(CFR: 41.2 to 41.9/45.7 RO/SRO Rating: 3.9/4.1 LOl-CLS-LP-17 Obj 07 Objective: LOI-CLS-LP-17 Given plant conditions, determine if the RHR System should automatically initiate in the LPCI mode.
Reference:
SD-17 Cog Level: High Explanation:
The RHR System will automatically start in the LPCI mode of operation in response to either of two initiation signals: reactor vessel low level (LL 3) or drywell drywe" high pressure coincident with reactor vessel low pressure.
All A" RHR Pumps automatically start 10 seconds from receipt of the initiation signal if the Emergency busses are energized (off-site power available). If off-site power is not available, the pumps automatically start 10 seconds from the time the Emergency Diesel Generators re-energize the busses.
Distractor Analysis:
Choice A: Plausible because this is when the initiation signal is present from hi DW pressure and low reactor pressure.
Choice B: Plausible because this is when the initiation signal is present from LL3.
Choice C: Plausible because this is applying the 10 10 second time delay from when the initiation signal is present from hi DW pressure and low reactor pressure and would have started the pumps if electrical power was present.
Choice D: Correct answer, see explanation.
Notes Notes 3.2.1 3.2.1 System System Initiation Initiation The RHR The RHR System System willwill alltomatically automatically start start in in the the LPCI LPCI mode mode of of operation in operation in response response to to either either of two initiation of two initiation signals:
signals: reactor reactor vessel vessel low level low level (LL (LL 3)3) or or dryweil drywell high high pressure pressure coincident coincident withwith reactor reactor vessel vessel low pressure.
low pressure. See See Figure Figure 17-8.
17-8.
3.2.2 3.2.2 Response Response to to LPCI LPCI Initiation Initiation Signal Signal Satisfying aa system Satisfying system initiation initiation signal signal from from the the LPCI LPCI logic logic will will result result in in the following occurring for each loop:
- All RHR Pumps automatically start 10 seconds from receipt start 10 receipt of the initiation initiation signal ifif the Emergency Emergency bussesbusses are are energized (off-site (off-site power power available). IfIf off-site powerpower is is not not available, the pumps automatically start -10 10 seconds from the time the Emergency Diesel Generators re-energize the busses OR a total of 20 seconds from a loss of off-site power after a LOCA LOCA (Ref Fig 17-2C).
1 7-2C).
SD-17 1 SD-17 Rev. '1313 Page 34 of '1271 127 Categories KJA:
KIA: K1.07 264000 Kl.07 Tier!/ Group:
Tier T2G1 RO Rating:
RORating: 3.9 SRO Rating:
SRORating: 4.1 LP Obj: 17-07 Source: NEW Cog Level: HIGH Category 8: Y
Unit One
- 33. Unit One is is operating with the AOG system system bypassed.
bypassed.
AOG-HCV-102, AOG Bypass The AOG-HCV-102, Bypass Valve, controlcontrol switch switch is is in in Auto.
During performance During performance of OPT-04.1.7, OPT-04.1 .7, Main Main Condenser Condenser Air Ejector Ejector Radiation Radiation Monitor Monitor Functional Test, the operator places Functional places the A SJAE Rad Rad Monitor Monitor NUMAC NUMAC drawer drawer (D12-RM-K6OIAIB) INOP/OPER (D12-RM-K601A1B) INOP/OPER keylock keylock switches to the INOP position.
The B SJAE Rad Monitor then loses power.
Which one of the following identifies the effect that the above conditions will have on the AOG Bypass Valve?
A. closes immediately B
B~ closes after a time delay C. remains open irrespective of actual radiation conditions D. remains open but will close on an actual hi-hi radiation condition Feedback K/A: 272000 K3.05 KIA:
Knowledge of the effect that a loss or malfunction of the RADIATION MONITORING System will have on following:
Offgas system (CFR: 41 .5/45.3) 41.5/45.3)
ROISRO Rating: 3.5/3.7 RO/SRO CLS-LP-1 1.0 Obj 5c Objective: CLS-LP-11.0 Explain the effect that a loss/malfunction of the PRM System will have on the following:
- c. AOG System
Reference:
OPT-04.1.7 OPT-04.1 .7 Cog Level: High Explanation:
The permissive logic is completed when either a high-high, high-high, adownscale, or an INOP trip occur simultaneously in Channels A and B. When the logic is satisfied, Timer D12-M001 energizes to furnish the proper input signal to the off-gas-to-stack isolation valve after the predetermined delay interval and PROCESS OFF-GAS TIMER INITIATED /NITIA TED annunciator on Panel XU-3 will actuate.
Distractor Analysis:
Choice A: Plausible because the logic is made up (Inop from both A and B B logics) but is incorrect because of the time delay.
delay.
Choice Choice B:
B: Correct Correct answer, answer, see explanation.
Choice Choice C:
C: Plausible because because the the PT PT does does use use aa trip test test function. This This is is used used to to put put in in aa false signal signal to make make sure thethe instruments logic logic work work correctly.
correctly. The examinee examinee might might think that that this would override all all signals to to the the valve.
valve.
Choice Choice D:
D: Plausible Plausible because the PT because the PT does does use use aa trip test test function. This is is used used to to put put in in aa false signal signal to to make make sure sure the the instruments logic logic work work correctly.
correctly. Therefore thethe examinee examinee could could reason thatthat itit would would not not operate for the the invalid invalid signal signal
Notes Notes 3.2 3.2 The off-gas The off-gas timer timer must must be be reset reset within within ten ten minutes minutes of of initiation initiation or or aa loss loss of of ofi-gas flow off-gas flow and and subsequent subsequent loss loss of of main main condenser condenser vacuum vacuum willwill occur occur ifif the the AOG System AOG System is bypassed.
is bypassed.
QPT-04. 1.7 IOPT-04."1.7 Rev. 32 Rev. 32 Page 2 Page of lsi 2 of 18 CAUTION CAUTION IF the PROCESS OG IF OG TIMER TIMER INiTtA TED (UA-03 4-1) initiation device INITIATED device isis NOT reset within ten minutes, an isolation will occur if the AOG System NOT System is bypassed.
from the sd:Sd:
Signals from the upscale (high-high), ,Qswas.eai&, .downscale, and INOP trip circuits of the log rad monitors arEil8PlPlllEilCi'lil~~~~~~liltrsl are applied to the holdup valve control 1<D.~i~rl!0l!J:its.
logic circuits. The Air Ejector Off-Gas Radiation Monitoring System will cause the holdup outlet valve control relay to deenergize when
~t!rilfiflJimWtii7i1~flt1iiJlSIEl(lal!Dl§rm1!Ota1l1lrQ~JlIIiIth any combination of trips occur simultaneously in both cmruelB channLEls, The energized valve control timer provides the appropriate input to the isolation valve, causing it to close following the delay interval The log radiation monitor also drives four trip circuits which actuate interval.
annunciators, initiates the off-gas timer which, after 15 minutes, initiates closure of I (2)AOG-HCV-1 02 1
Categories Categories K/A:
KIA: 272000 K3.05 Tier / Group: T2G2 RO Rating:
RORating: 3.5 SRO Rating:
SRORating: 3.7 LP Obj: 11 .0-5C ll.O-5C Source: BANK Cog Level: HIGH Category 8: Y
- 34. Given the following plant conditions after a Loss of Off-Site Power to Unit One:
DG1 Running at 3575 KW Running KW load DG2 Running at 3680 KW load RB HVAC Isolated The operator is directed to restart Reactor Building HVAC using three (3) supply fans (75 KW each) and three (3) exhaust fans (45 KWeach). KW each).
Which one of the following identifies the impact of starting two supply and two exhaust fans from MCC 1IXG XG and one supply and one exhaust fan from MCC 1IXH XH on DG maximum loading?
DGI only maximum load will be exceeded.
A. DG1 B. DG2 only maximum load will be exceeded.
DGI and DG2 maximum load will be exceeded.
C. DG1 D
D~ DG1 and DG2 will remain within maximum load limits.
Feedback K/A: 288000 K1.04 KIA: Kl.04 Knowledge of the physical connections and/orandlor cause effect relationships between PLANT VENTILATION SYSTEMS and the following:
AC Electrical (CFR: 41.2 to 41.9 / 45.7 to 45.8)
RO/SRO Rating: 2.6/2.6 Objective: CLS-LP-39 Obj. 17a Given plant conditions, OP-39, OP-50.1, AOP-36.2, and/or ASSD procedures, determine the limits for the following DG parameters:
- a. Generator kW.
Reference:
AOP-36.11 Cog Level: High Explanation:
Max loading during LOOP is 110% 110% ==3850KW). 2 sets from MCC 1XG 110% of rated (3500KW x 110% (DG1) 1XG (DCI) adds 240 kw for total of 3815 KW. 1I set from MCC 1XH 1 XH (DG2) adds 120 120 KW for total of 3800 KW.
Distractor Analysis:
Choice A: Plausible if the examinee adds the wrong values for the fans, supply vs exhaust, then this answer would be correct.
Choice B: Plausible if the examinee thinks that 1XG 1XG is from DG2, then this answer would be correct.
Choice C: Plausible if the examinee considers the rated value instead instead of the emergency value for the answer.
Choice D: correct answer, see explanation.
Notes Notes CAUTION CAUTION IfIf aa diesel diesel generator generator failure failure has has occurred, occurred, power power restrictions restrictions may may prevent prevent restarting restarting all all systems required by systems required by this this section section ofof the the procedure.
procedure. The The Unit Unit SCQ SCO must must use use discretion discretion in in determining what determining what equipment equipment to to restart restart depending on existing depending on existing plant plant conditions.
conditions.
Maximum diesel Maximum diesel generator generator loading loading isis 3850 3850 KW.
KW.
(.130 KW) MCC 1XF MCC 1XF (85 (85 KW}
KW)
RBCCW Pump RBCCW Pump 1A 1A 48 RBCCW Pump RBCCW Pump 1lB B 48 RBCCW Pump RBCCW Pump 11C C 48 RWCU Pump RWCU Pump 1B lB 38 RWCU Pump 1A RWCUPump 1A 38 SBGT Train 1lB B 28 SBGT Train lA IA 28 MCC 1XH (190 KW} KW)
MCC 11XG KW)
XG (330 KW} Fuel Pool Cool Pump 'IlB B 50 Fuel Pool Cool Pump 'IA IA 50 SLC Pump 1B lB 30 SLC Pump 1A IA 30 Purge Exh Fan 1lB B 38 Purge Exh Fan 1A IA 38 RB Vent Sup Fan 'IlB B 75 RB Vent Sup Fan 'IA IA 75 RB Vent Exh Fan 1 ISB 45 RB Vent Exh Fan 1A 45 RB Vent Sup Fan 'IID D 75 RB Vent Sup Fan 'IIC C 75 RB Vent Exh Fan 10 1D 45 RB Vent Exh Fan 1IC C 45 Storage Tank Heater B SLC storage 40 OAOP-36:1 Rev. 52 94 Page 87 of 94\
Categories K/A:
KIA: 288000 K1.01 Tier / Group: T2G2 RO Rating: 2.6 SRO Rating: 2.6 LP Obj: 39-17A Source: BANK Cog Level: HIGH IDGH Category 8: Y
- 35. Unit
- 35. Unit Two Two is performing aa startup is performing startup after refueling when after refueling when aa leakleak occurs occurs on on the the reactor reactor head head inner seal o-ring.
inner seal 0-ring.
Which one Which one of of the the following following identifies identifies the the expected expected alarmalarm duedue to to this this condition?
condition?
A. SEAL LEAKAGE A. SEAL LEAKAGE FLOW FLOW DETECTION DETECTION HI HI B. DRYWELL FLR B. FLR DR DR SUMP SUMP LLVL VL HI C
C~ RPV FLANGE SEAL LEAK D. PRI CTMT CTMT HilLHI/LO0 PRESS Feedback K/A: 290002 G2.04.46 KIA: G2.04.46 Ability to verify that the alarms are consistent with the plant conditions.
Reactor Vessel Internals (CFR: 41.10/43.5/45.3/45.12)
(CFR: 41.10/43.5 / 45.3 / 45.12)
ROISRO Rating: 4.2/4.2 RO/SRO Objective: CLS-LP-01 Obj 3 Describe the plant conditions that could cause RPV Flange Seal Leak (Annunciator A-02, window 5-6) to annunciate.
SD-01 .2
Reference:
SD-01.2 Cog Level: High Explanation: The head to vessel flange seal is made up by two concentric o-rings 0-rings that are installed in grooves. A drilled passage connects the annulus between the two o-rings 0-rings to a pressure switch. If the pressure rises to 600 psig due to the inner 0-ring o-ring leaking, an annunciator, A-02 5-6 RPV FLANGE SEAL LEAK, will alarm by the pressure switch closing. Failure of both flange seals is detected by the primary containment leak detection system.
Distractor Analysis:
Choice A: Plausible because this alarm is for a seal leak, but it is for a Recirc Pump seal leak.
Choice B: Plausible because this would be the condition if there was a failure of both 0-rings.
o-rings.
Choice C: Correct Correct answer, see explanation explanation Choice Choice D:0: Plausible because because this would be aa condition would be condition of of aa failure of of both both 0-rings.
o-rings.
Notes Notes 3.7 3..7 Vessel Head Vessel Head Flange Flange Leak Leak Detection Detection The head to The head to vessel vessel flange flange seal seal is is made made up up by two concentric by two concentric o-rings 0-rings that that are are installed in grooves. A installed in grooves. drilled passage A drilled passage connects connects thethe annulus annulus between between the the two o-rings two 0-rings toto aa pressure pressure switch.
switch. IfIf the the pressure pressure rises rises to to 600 600 psigpsig due due to to the the inner 0-ring leaking, inner a-ring leaking, an annunciator, A-02 an annunciator, A-02 5-6 5-6 RPV RPV FLANGE FLANGE SEAL SEAL LEAK, LEAK, will alarm will alarm by by the the pressure pressure switch switch closing.
closing. Failure Failure of of both both flange flange seals seals isis detected by detected by the the primary primary containment containment leak leak detection detection system.
system. Figure Fgure 0'1.2-9 01.2-9 shows the shows the vessel vessel head head flange flange leak leak detection detection system.
system. The The grooves grooves for for the the O-rings are only 0-rings are only located located in in the the head.
head.
r----- ..,
I P601 paOl I luA :
...--- CON TAIHMI: NT cONTkIRuijT
[j I
1 A2 I
I l..._.., __ ...lI I
II 1
1 I
r-- - ---- -
- ...........-.-@I:
'1 101'1
........ 1'$ I
---~~+- ~ NO(l2 I F008 I I I P004 I
FOQU F006 fOOl Categories K/A:
KIA: 290002 G2.04.46 G2.04.46 Tier!
Tier / Group: T2G2 RO Rating:
RORating: 4.2 SRO Rating: 4.2 SRORating:
LP Obj: 01-3 Source: NEW Cog Cog Level: LOW Category 8:8:
- 36. Which one
- 36. Which one of of the the following alarm conditions following alarm conditions auto start the auto start the CREV CREV System?
System?
PROCESS RX A. PROCESS A. RX BLDG BLDG VENT VENT RAD HI-HI RAD HI-HI B. AREA B. AREA RAD RAD RADWASTE RAD WASTE BLDG BLDG HIGHHIGH C. REACTOR VESS VESS LO LEVEL TRIP TRIP D1 D~ PR! CTMT PRI CTMT PRESS PRESS HI HI TRIP TRIP Feedback K/A: 290003 K4.01 KIA:
Knowledge of CONTROL ROOM HVAC design of CONTROL design feature(s) and/or andlor interlocks which provide provide for the the following:
initiations/reconfiguration: Plant-Specific System initiations/reconfiguration: Plant-Specific (CFR: 41.7)
RO/SRO Rating: 3.1/3.2 Objective: CLS-LP-37 Obj. 4a Given plant conditions determine if signals exist that would cause the following to automatically start/open:
- a. Emergency Recirculation Fans
Reference:
SD-37 Cog Level: low Explanation:
An automatic start signal is initiated by any of the following:
Any one of three Area Radiation Monitors
- a. Anyone (1) Control Room (Channel 1) 1)11 mr/hr +/- .05mr
.O5mr increasing (2) Control Building Ventilation Intake (Channel 2 or 3) 7 mr/hr +/-.O5mr+/-.05mr increasing
- b. LOCA Signal detected by one of the following:
(1) Reactor Water Low Level 2 (2) Drywell Pressure - High Distractor Analysis:
Choice A: Plausible because the LOCA signal comes from the same device that initiates the Group 66 isolation isolation and this is an input into the Group 6 isolation logic.
Choice B: Plausible because because radwaste is located below the main control room and control room rad signals do do initiate initiate CREV.
CREV.
Choice C:C: Plausible since since this is is a low low level level scram setpoint setpoint and and not not the the LL2 LL2 signal that that would would initiate initiate CREV.
CREV.
Choice Choice D: Correct answer, D: Correct answer, see see explanation
Notes Notes 1.
- 1. An automatic start An automatic signal is start signal is initiated initiated by by allY any of of the the tollovl"ing:
following:
a.
- a. Any one of Anyone of three Area Radiation three Area Radiation Monitors Monitors (1)
("1) Control Room Control Room (Channel (Channel *1) 1)11 mr/hr mr/hr +/-+/- .05mr
.O5mr increasing increasing (2)
(2) Control Building Control Building Ventilation Ventilation Intake Intake (Channel (Channel 22 or or 3) mr/hr +/-
- 3) 77 mr/hr
.O5mr increasing
.05mr increasing b.
following:
(1)
("l) Reactor Water Low Low Level Level 2 (2) Drwell Pressure - High Dly'"vell -
Categories Categories K/A:
KIA: 290003 K4.01 Tier / Group:
Group: T2G2 T2G2 RO Rating:
RORating: 3.1 SRO Rating:
SRORating: 3.2 LP Obj: 37-4A Source: NEW Cog Level: LOW Category 8: Y Y
- 37. Which
- 37. Which one one of of the the following following correctly correctly completes completes thethe statement statement below?
below?
A single A single recirculation recirculation pump trip from pump trip rated power from rated will cause power will cause the the value value of of Critical Critical Power Power to (1) to (1) and and the the Critical Critical Power Power Ratio Ratio will will be (2) be (2)
A. (1 A. (1)) rise rise (2) higher higher B. (1 B. (1)) rise rise lower (2) lower (2)
C C~ (1) lower (2) higher (1)) lower D. (1 (2) lower
Feedback Feedback K/A: 295001 KIA: 295001 K1.03 K1.03 Knowledge of Knowledge of the the operational operational implications implications ofof the the following following concepts concepts as as they they apply apply to to PARTIAL PARTIAL OR OR COMPLETE LOSS COMPLETE LOSS OFOF FORCED FORCED CORE CORE FLOW FLOW CIRCULATION:
CIRCULATION:
Thermal limits Thermal limits (CFR: 41.8 (CFR: 41.8 to 41.10) to 41.10)
RO/SRO Rating:
RO/SRO Rating: 3.6/4.1 3.6/4.1 CLS-LP-1 06A*1 3B Objective: CLS-LP-106-A*13B Objective:
Describe how Describe change in how aa change each of in each of the following affects the following critical power:
affects critical power: Mass Mass flow flow rate rate
Reference:
Reference:
11 (2)OP-,
(2)OP-, Revision, Page, Section Section OPS-FUN-LP-1 04-I (Thermal Limits)
OPS-FUN-LP-104-1 LOl-CLS-LP-1 06-A LOI-CLS-LP-106-A Cog Level: low Explanation:
As actual power goes down, void fraction increases. This causes critical power to lower, although not as significantly as actual power. Since actual power drops further than critical power, the Critical Power Ratio gets larger.
CPR =
CPR=CP/AP CP 1 AP Distractor Analysis:
Choice A: Plausible because Part (1) is incorrect. As actual power goes down, void fraction increases.
This causes critical power to lower, although not as significantly as actual power. Part (2) is correct. Since actual power drops further than critical power, the Critical Power Ratio gets larger.
Choice B: Plausible because Part (1) is incorrect. As actual power goes down, void fraction increases.
This causes critical power to lower, although not as significantly as actual power. Part (2) is incorrect as stated in (a) above.
Choice C: Correct Answer Choice D: Plausible because Part (1) is correct. As the actual power goes down, the power required to cause the onset of transition boiling also goes down. Part (2) is incorrect. Although actual power goes down and critical power goes down, the power required to cause the onset of transition boiling does NOT go down as far as actual power due to the higher void fraction.
Therefore, the Critical Power Power Ratio rises.
SRO Only Basis:
Basis: N/A Notes
On Fi On. Fiire9-?, tba cente1 gure 9-7, tn.e centercur CUivecis tbecriticaI is the "cri tical The ratio (lfthe The ofthe crit1caI critical pm'.'er power 10 some se operating ouerat power (heat balance) cun'e pOWe1~ curve and it become:;
becomes power, like the one ce inFigure inFiire 9-7, 9-?, is methe critical cr1xEc!
tanet to (jlb"t tan~t (just tanches) tchea) me the *'GEXL GL COffela.tion~
correlathei power pwr ratiorik (CPR).
nesr the channel curve near chne1. <!};it.
- it, This implies that for this condition, om 0Th oClZ'.urs occrs near the top of CPR- CP >1.0 faei, bundle. For cther the fuel other flow rates or 2!.' i.1al axial, CPR>1J iUJ power shapes, the criti.cal POWe1 calbundie power may be bundle POWe1 bieher or lower.
hislle1 end om lower, and 0Th may take ula<eata place at a Where:
Whe1e:
different boj]in,g:length diffe1ent boi]inElenstk The impo ro,ait thing to imuorntthingto rember is that aJl,.l.l bundle power c:u:rr.re remembe1 crve that CPR =
= critical po",'er power rati.c rao touchei the OEXL touches GE1 {un;:
ciar.u represent.
represents the critical power for that condition POWe1 condition... CP =
= bundle PO",;;!! at ",'meh bundlepoweratwbich om OThocDioccurs*
AP = actual bundle Pm'.'e1 actnal power EqUQ/Uin Eti,,it 1-15 9-15 I
When the actual bundle power equals equals. the power, CPR critical P(lWe1, CPR. is is. equal to I.e. 1L F.:I1'Fr operating POWe1S les:s than the critical pOWe1, powers less power, CPR is ,§1eate1 fleater than LG. The CPRmnst then LO. CPRmuat always alwara eater than LO to be greater te aV.:Iid avoid OlB. 0Th. The minimum value that the CPR can ha'l,'e minimnm have anywhe1e anywhere in the cere core (MCPR) is i spe..'ifi.ed specified as an ax LCO in Technical Te hnical Spedfica1ion.s.
Sp czens. The LCO f.:lr for MCPR is i based on maintaining the MCPR eater than
,§1eate1 then LOLO?7 for any analj-zed analyzed operating operatin!
Fipire 9-S. Remembe1 transient, as shown in Fi!rore9-B.
that a MCPR ,§1eater s-eater than [0 ansure that om LO ensures 0Th p1ace. om does not take place. 0Th causes rapid the1malthermal Fi2ure9L2 CElL C.oymAlt.i'4tn Fig.ru,81-:7 GEXl. Crr2BWRHoi c.nd B~.'iI H8Q! cydin of the cladding and may lead to cycling t.:l film Eiw Crrn BQ~ CrtIlIS.5' boiling. whith resul1S in unacceptable cladding whim resi and niel fuel temperatures.
tempe1atnres.
31Ii2*.OI MiCS CfLTER9 12 nf 32 :0 .OOO NEALP]3IcS
TóLe 9-1 Factors 4ffeitLa CrIzic.9I Powtr The transients most likely to l1mit limit operation r~\.c:TOZ. ~..\,:. J:n;'!"f:U c:nt because of of~[CPR MCPR consideration.s considerations are:
- v:M:-;&; :v:M::&
iit
- II Turbine trips or generator nerator load rejections bypass valve capabmty without bypas:s capability
'*" Loss of feedwater heatin heating or inadvertent IA LY high bih pressure coolant injection II Feedvlater Feedwater controller faHure failure (maximum demand)
MAXIMUM FRACTION OF LIMniNG LIMITiNG CRITICAL POWER RATIO (MFLCPR)
(MFlCPR)
- The process computer calculates CPR e'!taluatinEcore evaluating core conditiens CPR d.ata conditions to ensure limits are data A.."Q.AL ,,:t\"i",U;
~~TI'~
- not exceeded. One efthe this data emputis of the most output is a ratio caned limiting critical power ratio useful ferms mostusefal forms of caUed the fraction fraction of ratio" (FLCPR). This
-- ratio compares the flew-adjusted to the actual bundle CPR maximum fractien fraction ef flow-adjusted operating (steady-state) maximum CPR for th=
(steady-sta1e) fuel bundle the fael CPR. From this, the of limiting critical power o ratio (J.,IFLCPR QLcPR. - pro noimced "m1ffle-s:ipp~J),
pronemlced
- miffie-sippe?),
STEADY STATE AND TRANSIENT which is the maximum fractien fraction efof limiting crit:ical critical pewer power ratio (J.,IFLCPR)
Q&LCPR) and is the ratio ratio The primaiy prima1Y design desi objec.tive objective is to 1mintainmaintain of the flow-adjusted CPR operating limit for nucleate boiling bomng: and avoid OlB. OTB, The CPR that fuel fuiel type to the bundle CPR, is developed.
thennallimH thermal limit 1;;.is settoma1n'lainadequate set to mainuinadeiate marn margin For most nuclear phn plants the ~IFLCPR CPR raterao takes tai between nuclea1e boiling and OlB.
nucleate bothnand 0Th. The steady the foHowingform:
followin!form:
state and transient MCPR thennallimits thermal limits are derived from this single sine designdesi bais ba;;1S CPRxx K, K requirem:>..nt.
requirement. NFL CPR- CPR.
MFLCPR-CPR Transients caused by single ;inIe operator error or malfanction ,shall equipment mal:ftmctim shail be limited mited :50 so that, Equad&t 9-16 EIl*llJ1.tlon9-16 consideringuncertain1ies considerinuncertaes inmonftoringthe inmonitin the core euerating operating state, more than 99..9% 9% ef of the fuel rod;;
roth are expected to avoid OlB. 0Th.
RoeClwJlallo<'l Rcroila1lc :LOOPS .ocp Operating!
per.3iç BE ;:3.1.1
..4.1 APDLICALE APL;:)LICABLE Fc AR*::V.b, For AREV, ~Il:el, 111' COLR prE<5<enls ir.2 COLR prsiL Sili/gle lrle lOOploop operauon APLhG ~nllts oWallon .,J,;?LHGR iit Z.ETY "'.NN:..
SAif,:::TY YSE YSES inn 1tITe ir formorrn 01' multlp:er !:IlallS c aa mUllIp.mer tlatl a.pplled aplli ;:1) 11i:e 1t two Iwo Il)CP Iop operaUon oprallon (CClt!nued) APLiGR APLH llnlt&
GR limits..
Tile :r3isl&n: analyses ,or CI1:apter Tre 1r*ailEilem cr3pLer 1:' 1 l)1lt1e c UFSAR UF.AR l1;a*.**ealso t3 also bl?en beci e3lu3:ed ~orslngle
!:'lI'alua.~!:d c- SIngle r.ec\:tl'Jla110n r JlIlon llOop op opera.~lCn.
cper3:n. The '!:'Il*.aluailCn e3lua:li oncie. that COndlJlliES resu or 111,!:
Itial rl?*sll'.r.s Ire trans/ell1 1rriler.1 allialysEs ar yee a&*e -e nottnflcan:ly s);,tnl'llcamly aTe1ecl b- !:Ile a~l?cteCIII:~1 eln9ie reclrcll'llatlOn t.ie slng!1e eclrcui3ion Il)ap loop operaUon..
operaDon. Tller.e There IS, I, i'lO' hoee.... S'l.'Ef, ;inan flipact 'rln Impact n ~Ile he ~Il:el cladm 11li11?grlly' uel Cladllli'lg! IrIerlLy Sl!. since
. some or since Silme or 111:e Ile uncertaln.tleS uncertaIntes for br ire p3ran1eters used 111,eparameters ue lrattle ir lIne crlUcal crIUcl pO'.'ii'er pcer ,:!eterml:1a110n delerwlnallcn *are .3re higher ininer In In single lOOP operallon. The ne~
loop opEra1lon. nei resm~
iesuii Is an IIi~r'2ase hrease In !:Ile 1ne MCPR
.aperatlillg! IImll.
Duin single rer:lrCUlaUon~aopoperailCilI,.
DUi'lngslll.gle rE rulaUon sloop operalor. m*ol:!lflcalloli TtodII1C.3Lior to 10 tne tfle ;Reactor
- eaoLor Pro1eollor system PrOlecUoli SysLern (RPS) RFSi aIJer.a.g!?
aer3ge pD\'!t'er er rang!? range monitor mon!orAPR v."'.P.~\i t Slmub:ed Thermal!
Slmula~ed Zowe-Hi, .b,lla\,'abie Therrr t?OWElT-lilgln All be Valll'eValue I'S is r.equlred required to aO:xlUIl.i aooour:
bcr the for TerenL nazed limits lJe rluerentan;il)'ZEd liflitS betll¥Een be:een t.....o-rEclrculatlon o-reoculailon (Ii'1IJe 1ve flOw row lOOp boo per-3tor and operaUon cperal10ra operauon 'wHh with onLy only l)!Ie cne lOOp.
loop. The A'?RMARM Cl1annel flanneI sJJbtracte brao1 ire J!LW tile W ,'alu!?
lue from Forr tI'Ie tie measureal rec.irciJIa1lon drl~'!?lioW rneasii-e reClw.Jlallan irle row to* o ,erfI?CU eTeollel ....etv sr1 "l1e st,r! ne limits UnIts ;ind ises =me and tlses iie adJust~
adjusiei .reC/rculiitilCn reclrcul:Ion drl\'eno'w'J'alll'e dre flow value 10' lo ielernlne ,lie
- letermlne AFRM Slmtll;i1~
fle AP'RM nulate TMrmal Tflernii pO\Ofer-HIg!11 PowerHIcIn FunctIOn Fucton trip tip selpolni.
se1pclnt Reorculallon oops opera11ng RoeC}rClJlamonlOops slsfles Crller~an operaling sa;nslles Orllerlon 2 of or ilJ CFR :"1l.36(Cn2)(Il) 10 Cj;:,R SL36( :2uII {Rer. jRer. 4J4.
LCC LCO Two T recituiailon oops are norrr311y
....o reCIk"0J1;i1Ion!Oops nOrlr~aJly requlrell requbrel to be 1.1n1 cperat!ofi peraor \'4lt1n willi melr :elr 11cs 110' ....5 naLcIne matC/n~ *,,'HI11n wllrin 1Itcelmlts i5pl?Cffi~1n ite irnlts speomeI in SR 3R 3.'::; .. 1.110
.3.4.1 110 enSlJre eneure ttlat Lial :!Ilat
- flat urlng a tOeA.
'lllJrlng OcA CiilJselii aise b)' by a break o~ttle FiPifl o~ one o he piping ore reclrClJlat!On recircuiton lOOp loop ire 3ssunlptors or:re 111-eassllmpt!oll,s LccAriaiysls O'r iZl!;E! LoeA analySIS arear!? sallered. Al~E!rna:;el)'~ wlti saUsrled. Aitern3:eiy, With only osne rec!rculaflcn
,one reclrculaUoll }Cop loop In cper:n.
opera:1IOn. rnodlfleaHons modfficaUons Lo to tie reqlJlr~ APL-t2R tile rqJlrei APLHGR rvIts L2O limits (LCD .3.2.1, 3.2.1, "AVERAGE AVEFAGE PLAR PLAN'AR LIPEARLI.N'EAR HEAT H,:::AT GENERATION G:::NERATION RATE :AP:HGRT MCFR RAn: (,t..PlHGRn, MCP'R iitrs IImi'!S ~LCO LCC 3.2.2, 'MENIMUM CRrrtCAL
.2.2, 7.1iNbMUM CftITICAL POWER RA POWE)R RATIO TIOI.iMCtPRr:,.
[MC?Rr), '_HGR LHGR IImlls ~LCO 3.l.3, Ilnits LOC 3.2.3, 'UNEAiR lINEAR HE.b, HEAT T GEEATiON GENEk=tb, nON iRATE RATE ,:LHGR}').
LKGRfl. an anal ARM A'PRM Simulated Slmu.lated Thernial Then'Tlal POI.'ii-er-Poer Igfl H,lgh Allowable AIID\,'able Value (LCO 3.3,1.~ ~, as applicable, must LCO 3.3.1.1 , applied :o nust be aplled ~o aIl allow caMinued operabcn. The COLR conllfiuealopera::tan.ihe COlLR defines deOnes adjuehrents a:llustmems .or or wodlflcaflorss m'lldmcaUoM requIred required rortfleI'Onlle AFI_HGR,
.4,P}_HGR, MPR,MCPR, and al1d LHG LHGk~ Iln1s IIm!:s or :or Ire 11le Cll'lrren'l operaIIr urer oycle.
opera111llg ,C)'cle.
{cCnUnll'ed)
IconllnLedi Brunswlctc 5run5~vlCk Urit Ulil:~ 2
- 2 E 3,4.1-3 ;Reo;lslon ReWbon No. No. 62 62
The RPT The RPT breakers eaers. are are located coated o,~ te 20' oi the 2(Y e::e'.'a'lion eevaticn ,of of lhe the Reac10r eac:or Building and Building and are are bei'~~
iee. tile the reaclor reacor recirculat'OI11 reairulat o M-G M-G Sets Sets andai the the ecrculaiion P-JfllpS.
Heorcula1ion F\imps. The The I"llain rra t tumine tibine firs.l staqe perJlli(s~\
1irs stage perrrHssve ... e is is equivalent equi o 26%
... alent 10 2% rated rate thennal termal poY/er. EOC-RPTOC-RPT ~s5 automaiic,ally auzonaicIly bypassed below bypassed belcw 26%
2e% rated rated thennal power. The s.iop ihermal power. value poSit~"l sop valve posit and and conircl vallie control vahie fast closure clcsure control conirol c pressure pressure S\',rit..:::hes.
svh:hes are are lhe the sarre sarr signats used signals. used to [n~liatea niate a rea.ctor reactor scram. The cire-tlitry circui:rys ~ rEfilollEd remod nrool from service bjol seNice by plac'ng keytock sYli~hes.
placm the key!otk svi1hes, RPT RPT SYS .A A (8J B} OUTOUT OF" OF SVC, on Panel Panel Ht2-PBOQ Hi 2-PfD and Ht2-P611 H12-P1 1 in in the INOP INOP positie:Tll.
positxL.
1lIJIih EOC-RPT noi W:ih EOC-RPT rc inn service~
seraice, penalfes penaItes are imposed iripased in ihe the caICJ:*a.tions caIction ci t.he of the MCPHMCPR limits in ihe the Core Operaiing Cperaiin limits Report "::50, Limits Report. so,ihe iheiechTech Speo MCPR Spec MCR tCO ICO OptionS.
Option 3, which alloylSalIos operatio,'l; cperatbi d'oserchiser 10 jO the MCPR MCR ihft, mus.l safety 11nit, mus iake take into account the a\lerageaverage scram iime irne of the tile coo.ir-ol ccthrol rods.
The ccnrol power iheconir*ol pcwer r'IJ.ses.
ftses fe:r for the RPT breakers reakers hiWO! have been remov-ed removed to l prevent them pre'lI'ent Them from.
iron opening openirg and the lrip frip a.l1d ard permIssi perm ssive ... e interlocks inter ks 'ii'/ithwith the MG set breakers hali'e have been bypassed_
bypassed. This was performed perfoTned io to prevent a fa~'!Ire pre'lI'ent faire *of of the interlock rofif.ay reay from frorr tripping hipping ihe the MG MG setset.
Categories K/A:
KIA: K1.03 295001 Kl.03 Tier / Group: TIG1 Tier/Group: T1G1 RO Rating:
RORating: 3.6 SRO Rating: 4.1 SRORating:
LP Obj:
LPObj: CLSLP106A*13B CLS-LP-106-A *13B Source: BANK Cog Level: LOW Category 8: Y
- 38. Both Units
- 38. Both Units were operating at were operating at rated rated power when ALL power when ALL switchyard switchyard PCB PCB position position indications turn indications turn green.
green.
Diesel Generator Diesel Generator status:
status:
DG1 DG1 Running loaded Running loaded DG2 DG2 Running Running loaded DG3 DG3 Under clearance Under DG4 DG4 Tripped on low lube Tripped lube oil pressure Which one of the following identifies the AOP(s) that Unit One and Unit Two are required to perform?
UnitOneisrequiredtoperform Unit One is required to perform (1)
Unit Two is required to perform (2)
OAOP-36. 1, Loss of Any 4160V A. (1) OAOP-36.1, 41 60V Buses or 480V E-Buses (2) OAOP-36.1, Loss of Any 4160V Buses or or48OV 480V E-Buses OAOP-36. 1, Loss of Any 4160V B. (1) OAOP-36.1, 41 60V Buses or 480V E-Buses (2) OAOP-36.2, Station Blackout C. (1) OAOP-36.2, Station Blackout (2) OAOP-36.1, Loss of Any 4160V Buses or 480V E-Buses D
D~ (1) OAOP-36.2, Station Blackout (2) OAOP-36.2, Station Blackout
Feedback Feedback K/A: 29S003 KIA: 295003 A2.0S A2.05 determine and/or Ability to determine andlor interpret interpret the the following as as they apply PARTIAL OR apply to PARTIAL OR COMPLETE COMPLETE LOSS LOSS OF A.C. POWER:
OF POWER:
Whether a partial or complete loss partial or loss of A.C. power power has has occurred (CFR: 41.10/43.5/45.13)
(CFR: 41.101 43.S 14S.13)
R0/SRO Rating: 3.9/4.2 RO/SRO L0lCLSLP303A*001 Objective: LOI-CLS-LP-303-A *001 Given plant conditions and control room indications, determine if AOP 36.2, Station Blackout Procedure, should be entered.
Reference:
OAOP-36.2, Revision 41, Page 4, Section 3.2.1 High Cog Level: High Explanation: This meets the KA because the student will have to determine that all green lights is a LOOP on BOTH Units. Then determine that Unit two is under SBO conditions and apply this to the AOP entry conditions.
Switchyard PCB green position indication shows all PCB are OPEN, which indicates Loss of ALL offsite power. Unit 2 is the only blacked out unit, but both units must enter station blackout procedure.
RO must recognize LOOP and SBO on U2.
Distractor Analysis:
Choice A: Plausible because all PCBs PCB5 open is a Loss of Offsite Power (LOOP) requiring entry into AOP-36.1. LOOP recognized without SBO on U2.
Choice B: Plausible because all PCBs open is a Loss of Offsite Power (LOOP) requiring entry into AOP-36. 1. No DGs running on Unit 2 requires SBO AOP entry on both units.
Choice C: Plausible because not knowing Loss of Offsite Power (LOOP) actions are contained in AOP-36.1. Second part is correct.
Choice D: Correct Answer SRO Only Basis: N/A Notes
tO 1.0 SYMPTOMS SYMPTOMS ii 1.1; S~ r Mdeenezed deeneTgl:zed 1.2 1.2 3us B. C, nd 0 wideriage 1.3 1.:3 us E1 Bus El and E.2E2 (E:3 E3 and E4} under underoftage
..ottage 1.4 1.4 Nodies&
- No genera1cr running diesel generators mnning and amd IO'aded oed QtIi both units OaTh one or bDth: unfts 2J 2.0 ACTIONS AUTOMATIC ACTIONS
- 2. Reaotc scram ReaotOlT 2.2 Groups 1. 1, 2, 2,5, 6, and ID isolate 10 &'Solate 2.3 Groups 3,3 andB and S isolate with Ine the DC powered outboafdoutboard isolatiDn isolation '",al'le'S ahies only.
2.4 Reacbcr Buildiing Reaotor Build HVAC trips. thps. but b dDes does NOTOT iso'late isoate until undi power is avaj;sable avable 10' 1h damper so~enoid c '!he akies.
sooid ""al'le'S.
2.5 2.,e, foIlotng nc The following DC oil pumps pwnps start on [lmllfow header pressure:
- RFPTs Re ac4or Recire Reactor Recir M-G r4i-G Se1s Seis
- 3ea! Oil Hydrogen Seag OAOP-3&2 IOAO?-3fi.2 I 41 Re'l.41 Page 2 c Of lQ 1Qfi I
3.2 3.2 SupplernenLary Supp lementary Actions Actions 3.2.1 SLatioli Blackout 3.2.1 Station Blackout Actions Actoiis NOTE:
NOTE: Priorfties are Priorities are to to restore restore .lI.C AC powei power to to the the blacked blacked ouioui unit unit battef)'
battery chargers chargers by by cross-tie of cross-tie of EE buses.
buses, alignment alinmeni *Q*f SkMA diesels.
of 8.AM,A diesels, or UAT backfeed.
or UA.T backfeed
- 1. IF .any any diesel generator blacked out hliacked generator is out unit.
unit, TH THEN EN EX~T is started started andand loaded EXIT this procedure.
loaded on procedure.
on ~hethe oQ
- 2. ENTER ihe ENTER the a.pplicable applicable Supplementary indicated in Table 1, indicated Supplementary ..A..cnon
- 1. AND EXECUTE conrurrenlly Aclion Section Section concurrentiy wilh with o
this section.
thi.s Table 11 L4DKOUT In BLACI<OUi ..ili UNIT UNIT 11 UNIT 11 UNLT.:!:
UNiT 2 UNIT :22 UN.lT UNIT I SlTE p....'a 0NE NONE NONE t~Ot~E t~ONe NONE t40N'E NONE I\'ON'E rONE AVAIUIllUTY DIESELS ONLY 003 ONl.Y 0G3 ONLY OG4-C%'IILlI' 004 ONLY CGl ONL'{ 001 ONLY CG2 003 CG3 003 6& CG4 004 OPERATII\"G SUPPL. ACnON'S SIJIPPL. AC11ONS SNIER SECT SNTER ENTER iEt~'iiER SECT ENTER seCTSECT EFTE SECT ENYTeiiISlECT ENTlER ENTSR SEC1i SECT SECTION 3.2.2 3,.2.<2 33.22
- .2.. :3 32.4
- 3.2.4 3:.2.5 3.2.5 3.2.6 SEC'iiIOt~
1Pc 6) 1P~g.e B) iPe 13p 1.P'~g.e 13 IP~Q.e 10) 2S IP;ge Pge 27) [Page P;e 34134 eLACI<OUT m*m BLACI<OUi UNIT UNIT :2 2 UNIT 1 I UNIT :2 Uf'UT 2 UNIT 1 1 &:2
& 2 snE p*...in TiTY NO\NE ONLY $T S~T #2 3 OrLY ST OI\L.Y 1 S~'ii "'1 NONE A*.....I!.IUIllUTlI*
DIESELS oro 001 CGl & 6 002 CG2 VAFMBLE
.....Af';1;:.Bl...E VARb3LE V.;:'RI':'BLE NOt,S NONE o PERATII'>G SUPPL ACTIONS SIJIPPL ACnON'S ENTER SECT:SECT ENTER Et~'iiER SECT ENTER SECT ENTEA ENTSil. SECT SECTION SECi'ION 3,2.7 3 ..2.7 L2B 3.2.8 32.5
- =.2.9 *3.1..1oill ae
!lPage421421 ,;P'rag.e Soill' lPrag.e 541 54j IP*;ge S5j Sin 1 ftOP-36.2 OAOP-36.2 Rev. 41 Page 32. of iO '1961
3.2.1 3.:2.. '1 Station Blackout Station Blackout ActionsActions NOTE:
NOTE: Unit "I and Unif and 22 Gritical Gritical Instrum~nts Instrum ts and and Gomponen~s Cornponens are are lisied lised in hi Attachment "I (page i178I and page 178) and Attachment Attachment 22 (pag~182),
ipage 162;, respectively.
respectiie1y.
3.
- 3. NOTIFY the NOTIFY the other Llnit SCO other Unit 3CC toto enter enter ihiis this procedure.
picedure. D
- 4. NOTIFY the System Systerm Dispatche.r Dpatoher Unit 1(2) 1i2; isis in Statlof:1 Station D 2 lack out 8.iaclo;out
- 5. NOTIFY Seculi!)'
NOTIFY Securfty to aciions necessary to take the ac1ions necessary fO,f for aa D Station Blackout.
Station Blackout.
- 6. the SAT was los1 IF "the lost due ~oto a fault and is unavailable, D the swc yard is energized.,
AND thes\Vrtchyard energized, THEN.
THEN ESTABLISH UAT Backfeed in accordanceaccordance with 11(2OP-5U, (2)OP-5!1. AND PERFORM CONCURRENTLY with ~hjs this procedure.
orooedure
- 7. the SAT ,,,was IF "the Ios due 10
..as los1 of peweron to loss .of rower on "the the P~ress Prcgress System, THEN PERFORM the iiollowing:
Energy Sy.stem. following:
- a. PLACE AUTO RECLOSE REGLOSE swi'lches swhches in iii MANUAL MANL4L. D
- b. PLACE transmission Ifne line PC8.
PCB SUPERVISORY D LOCAL/REMOTE swiiches LOC,4UREMOTE switches in LOCAL.
c.
c.. TRIP 01'all11 transmission line PCBs. D 1 CAOP-3O.2 Ol,OP-35.2 Rev.
ReN.41 41 1961 PaQe 44 of 16 Paga
1.0 1.0 SYM MPTOMS PTOMS Ii 1.1 LOSS of orr-fllLi.e Loes oS' OFT-ete POWel Power 1.1.1 1.1.1 SAT de>-ene-rgI2ed SAT de-energIz 1.1.2 1.1.2 Bte 5, 5l1se-s B, C, an D C, aM D 1J1l.-dervollage ureroItae 1.1.3 1.1.3 [ndIcaUor of IndlcaUoll, oIIfurdIsheit&aLors all f-our diesel 'gefi?eralors nJll,nlng mrnin 12 1.2 LoeofE5ii Loss of E BUB 1.2.1 1.2.1 Ore 4161Y'l Olle 418CV cr r a8DV 8DV E E Bus unvoIIa Ub uMlel"tOllage 1.2.2 1.2.2 Lcts cr Le,ss RPS l'JuiS of one F..PS bic t:hal:-SCi~am ha-Scrarn s~nali jnaq 1.1.3 1.2 ..3 ?aIE oz.s of i'lsmlmernlaUon PaPJlaflOss Ift1rner1allon powered pered tr,:lm rn ErnergenC)' 120 'VAC Er1irgecy 12Q IJ' ".C l.2A 1.2 .~ nlIcaUort 01
,lndlcaUorn of one ClIlesel ei generalor genermior funning rurnrng I3 1.3 Loca or Loss of One BOP 8OP Bus8ua 1.11 1.3.1 Or416DV Ollie 416C.V BOP 50P ttJS bJ undera'cltage Lndert3ge 1.3.2 ReaDLorreIrLHatn Reactor pLurp1r r~lr(lUIat!(m pll'm,p lr1p*
1 33 1.*3.3 ndIcaHor
~ndlcaUon of one Ofor two Lwo dleseID dIee qernerators ereraLors runningnn 2.0 AUTOMATIC .ACTIONS ACTIONS 21 2,.1 Loae of orr-sL1e Lose Power OtT-[ie POYil'&l 2.11 2.1.1 ReacLorcrant Reactor scram.
2.1.2 Grups 1, GrclJps 1,2,3, 2, 3,~, nd 11DDisolate,.
L 8,and
, Iso3te.. 0 2.L3 2.1.3 ".11 Al! four Illiesel ger.eralor5 s*,ar:.
!eeeI generators 0Q 2.14 2.1.~ ReaoLcr5LidIr Reactor 5UI.dlllig cC !oIaie.
nV.A.C ISOlates. 0 2.1.5 SLany Gas Slan-dby Gas. TreaLrnent WaLes.
Troeatmenl ilnllJateli. Q 2.1..5 2.1.6 CEVas eRlEV s,am 0 2.1,7 2.1.7 :H)'drogen Water ChernIsby Hydren sL3ba CMmllil1}il&Clates 0 2.1,3 2.1.;~ SW-Vi03 ar an*jj SWV1DEArjLo SW-V 1C.6 Ardto C~Dse 0 Q
IDAOP-*36.1 Rev.
Re*~. 50 Page 22 of 4 94 I
~ 3.0 30 OPERATOR ACTlO~S OPERATOR ACTIONS 3.1 11 Immediate lrnrnedtate ActiOns Actione None None 3.2 3.2 S uppremaniary Actions Supp3men1ary Actions 3.2.1 3.21 Actions ActIons lDehHmmatlon Dermination
- 1. HF
[F at any riy ;'Ime
- lrre Olllfl"'9' uir 1lT1e Ire perrom.lance perrorniance Of of r.1l!$
- 9l procedllre.
procedure, all AC pO\\\ler all AC power IS bs1 ;.a le 10$1 t elller e.fler lIhll. btl IJnlts THEN bo~1!!
unIt, THEN! nlts GOGO TO OAOP-36 TO OAOP-3.2. ..2.
NOTE:
NOTE.: S,ecllc\'ls 3.2.2 t\'IrJ)lIgt:1 Sectlcis ::'.2.2 3.2.1 D pro'lide rougT 3.2.11:' pro1de r,ec,c:1j r ery 1el7'Jac,lCns c.tIn for Ic,ssloss or or pa',\"er:~
power EOP alMj B*OP ar Emergency Ermerencf buses. usee TM Tie seque'1ce sequence ~or or l?Q.ulpmenl re:raIIon will be eq!JpnTenL res~~r.allorm Olepenp::!ermil:Jpon eperien1 upon 11!te 1re stab:Js stabis Of "Ile plan.~
oi:fle plarr al1he at lbs lime lirrie of ihe pc\\-er orihe pwer tallur.e.
f3llure.
Thererre, s1eps Therefore. sleps wlttlili wiLiln ttlese these seC'!klns esc:ins m+a)' !T3W bet :perrorme::!
perrorTe slmU1'!aneOIlS4)'
sir L:3Ieouhy or In in an)'
any order or.er neoes:s.ail"J.
rsose.s3ry. as prIOrlt!:zed prl1tzed by'!l"we LnltC3.
by:r.e IJnlt seQ.
- 2. [F at IF aL any Loss or 3riy ~Ime C-:s Po of Otl'-Sl.~e urir ilLe
- lrre IlIlIflllig ire perro rmanceof perrormarce
..",er oocurs Power of "11I:s
- nis procedllre occurs OR a :slmultanecuslos:s procedure a s-nultaneus foss of
.3 o both 11i2 (2} C and 11;2) 1(2 0D 1501P busses, THEN PERFORlM PERFORM SecUOIli 3.2.2 ,:Page Seoiior ::'.2.2 Page 6:, concurrently .,VlttI
- 8) concllrrem~1 :rl.s secJ:/.on.
with ~III:S section.
NOTE: A.1tacl'lmen'!s A1bchrners 1 hJrilL *aM 1 {unit an 2 (Unitt 2) contain (Unit :2) llsllr ct Ci1l!cal conlain a IISUng In;strumentatlon insiruwritabn an'li ara 1hethe assocla1e::!
ass ocla1e :power power SUppl.)'
supply...
3.
- .. IF at any ~Ime aL3ny :irre IlIIlfln191rve u1r ire perrorma perrornarce nceofof ::111:s
- ms procedure, any E blJSbLI uni::!ervoffa.ge urIeroiIae occurs, AND ~Ile :ne as:scClated aS5c!3ted dlesa .generator diesel enecaLor s:ar3 sJ!ams ar.i ar~::!'Ues ~Ci 1he bus.
lies :lhe TH EN bus, THEN PERFORlM lherollcfll\'lng:
PERFORM lhe foilowlig:
- a. IF necessary.
recessarv, THEN ADJUST bus.
THEiN ADJUST tis "ona,!}e voiae arndar rreque.ncy Irequency as :-CllbVlS: llows:
- BrJS VOltage 41 eusvolL3ge 410D ~i) 4200 ',f DO tc4;20D V Q
- frequenC)': 59.8 t Bus Frequency: toSQ.:2 50.2 liz Hz o Q
- b. M,'ilnltor lcaI
/cnlLor local IndlcaHors IndlcaUon.$ In accordance with Iii acor.3ne2e ydttl
!!I0P-J9 OF39 Section ,s.m. S.D.
IDAOF.36.1 DAOP-.36.1 Rev.
Re .... .50 5D Page 44 of Page o.f94 1 Categories Categories K/A:
KJA: 295003 295003 A2.05A2.05 Tier // Group:
Tier Group: T1 G1 T1G1 RO Rating:
RORating: 3.9 3.9 SRO Rating:
SRO Rating: 4.2 4.2 LP LP Obj:
Obj: LOI,CLSLP3O3A*001 LOI-CLS-LP-303-A *001 Source:
Source: NEW NEW Cog Cog Level:
Level: HIGH HIGH Category 8:
Category 8:
- 39. Following
- 39. Following aa loss loss of of feedwater feedwater onon Unit Unit Two, Two, HPCI HPCI and and RCIC RCIC are are being being used used to to restore restore Reactor water level Reactor level to to the the normal normal band.
band.
The RO The RO observes observes the the following following alarm alarm and and indications:
indications:
25OVBATTA 250V UNDERVOLTAGE BATT A UNDERVOLTAGE in Alarm in Battery Battery BusBus 2A-1 Voltage o0 Volts (XU-2)
(XU-2)
Battery Bus Battery Bus 2A-2 Voltage o0 Volts Volts (XU-2)
Battery Bus 2A-1 Voltage Battery o0 Volts (ERFIS)
Battery Bus 2A-2 Voltage o0 Volts (ERFIS)
(ERFIS)
Which one of the following correctly completes the statement below due to the conditions above?
(1) continues to operate and (2) on high RPV level.
(1)HPCI A. (1) HPCI (2) will trip (1)HPCI B. (1) HPCI (2) will not trip (1)RCIC C. (1) RCIC (2) will shutdown D
D~ (1) RCIC (2) will not shutdown
Feedback Feedback K/A: 295004 KJA: 295004 A A1.02
- 1. 02 Ability to Ability operate and/or to operate andlor monitor monitor thethe following following as as they they apply apply to PARTIAL OR to PARTIAL OR COMPLETE COMPLETE LOSS LOSS OFOF D.C. POWER:
D.C. POWER:
Systems necessary Systems necessary to assure safe to assure safe plant plant shutdown shutdown (CFR: 41.7/45.6)
(CFR: 41.7/45.6)
ROISRO Rating:
RO/SRO 3.8/4.1 Rating: 3.8/4.1 CLS-LP-1 9*26B &
Objective: CLS-LP-19*26B Objective: CLS-LP-1 6*1 5E
& CLS-LP-16*15E Given plant conditions and Given plant conditions and one of one of the following events, events, use use plant plant procedures procedures toto determine determine the the actions actions required to required control and/or to control and/or mitigate mitigate the the consequences consequences ofof the the event event: Loss Loss of of DC DC power power Given plant Given plant conditions, conditions, predict predict the RCIC RCIC System System response response toto the following conditions:DC conditions:DC power power failure
Reference:
Reference:
11 (2)OP-,
(2)OP-, Revision, Revision Page,
, Page, Section Section Cog Level: High Explanation:
Division I DC is required for HPCI start and operation. A loss of Division I DC will make the RCIC inboard isolation logic inoperable which results in failure of the steam supply valve closure on high vessel level.
A loss of Div II DC would make RCIC fail and allows for HPCI vessel high water level trip logic to be partially made up (one out of two).
Distractor Analysis:
Choice A: Plausible because HPCI and RCIC are impacted by a loss of either Division of DC. A loss of Div II DC will allow HPCI to continue to operate and will trip on high vessel water level. RCIC flow would be lost on loss of Div II DC.
Choice B: Plausible because HPCI and RCIC are impacted by a loss of either Division of DC. A loss of Div II DC will allow HPCI to continue to operate and will trip on high vessel water level. RCIC flow would be lost on loss of Div II DC.
Choice C: Plausible because HPCI and RCIC are impacted by a loss of either Division of DC. RCIC would continue to inject but would not trip on high vessel water level with a loss of Div II DC.
HPCI flow would be lost.
Choice D: Correct Answer SRO Only Basis: N/A N/A Notes
ATTACHM ENT 22 ATTACHMENT Information InfarrnahGn Page 11 of Page of 11 Use Use Plant Effects Plant Effects from from Loss Loss ofof DCDC Panel Panel 3A(4A) 3A(4A)
RCIC:
RCIC: Will not Will not shutdown shutdown on on reactor reactor high high water water level, level, inboard inboard isolation isolation logic logic inoperable (E51-F007, -F031, and -F062 will not auto close}. close). Valves E51-F005 and -F025 fail closed.
E5'!-F005 ADS: ADS Logic B is inoperable.
inoperable ADS will initiate from ADS Logic Logic A if Core Spray PumpPump B B or both RHR Loop B B pumps are running.
H PCI:
HPCI: Will not auto initiate, outboard isolation logic inoperable (E41-F003,
-F041, and -F075 will not auto close), HPCI flow controller and EGM
-F04'!,
inoperable (no flow control or indication).
indication), HPCI trip logic inoperable, E41-F053, -F054, and -F026 fail closed.
valves E4'I-F053, A Core Spray: Will not auto initiate (manual operation possible but minimum flow valve will not auto open, and injection valves can not be opened simultaneously).
OAOP-39.O IOAOP-39.0 Rev. 32 Page 2'121 of 251 25 2.0 AUTOMATIC ACTIONS 2.1 2.'1 Loss of Division II DC Power from Switchboard 1A(2A): IA(2A):
- Half scram signal Group '1 Isolation closing inboard MSIV'sMSIVs only, resulting in a reactor scram if the Mode Switch is in RUN
- IF operating, THEN a loss of DG1 DG1(DG3)
(DG3)
- IF operating, THEN a failure of HPCI 2.2 Loss of Division II DC Power from Switchboard 1'16(28): B(2B):
- Half scram signal
- Group I'1 Isolation Isolation closing outboard MSIVs MSIV's only, resulting in aa reactor scram itif the Mode Switch is in RUN
- IF operating, THEN a loss of DG2(DG4)
- IF IF operating, THEN a failure of RCIC DAOP-39.O IDAOP-39.0 Rev. 32 Rev. 32 Page 4 of Page of 2525 1
ATTACHMENT 1B ATTACHMENT 13 Page7of1 Page 7of31 PANEL: 3B LOCATION LOCATION NORMAL SUPPLY NORMAL SUPPLY ALTERNATE ALTERNATE SUPPY SUPPY Reference Drawing; LL-30024-7 Control Building Control Building 49 4Q itft North North Switchboard 11BB Switohboard N/A N/A ki Ckt.tI LOAD LOAD EFFECT EFFECT 2>3 Div IIII RHR Dill RHR Lagie cgi 1.1. RHRHR ON Div II1 In:liaiiM lntia1io Legic [nop. BE Laop Lcgiccp, iccp RnR Rv.ill \\iII auto auto initiate inif ate from D:
from D ... II RHR RHR II>,iic.
lic. intlurfng indudng ,he the El1-FO lEE opening.
El 1-FO lE8 cpwing.
. Efl-FOlEBwill
[1. E11-FOI5B notuto will not closeo auto clDse on aagroup snaI and group ;9 S"9Ilal anothelh~
Et I-F015B Ei 1-FO 153 cannot carnol te ite ma'tual:y rntuaI openedcpeed fromfrom RIGR TGS.
Civ Ii Spray 3.I. Cj'/lI Spray Lagie nop Lcgio inop LOCA Lcc'\;oo(
LOCA L&ku/f initialil),! acesIlot in iaton does 1linton rer notfuncfon the fullo'/l~~g:
rthefol1,nng;
- . RHR
. Ei Pt-fR 51';}
SW BcCSle1 Eil-FQ4EB, Eccswr IYJmps I-F04ES. El1-F017B.
pumps B&D BSD E1l-FC173Eil-F023Eli-F0l83, Eil-F028B. EH-F015B.
. EH-F0278, El1-F02BB.
E'I-F0278. Eil-F02S3, Eil-F024B.
Eil-F0243, EH-FOllBA Eli-ADI13A
. E1l-FOlEEoa Ei I-F01SB can be t.estmkeddosedrcm1heRTGB stroked dosed trem lhe RIGS
- p. El i-F003B fails clcsoo.
. El-FCE3Bfailsoeed
~. E1. i-FCE8B tl:es El i-FG6SB +/-es not r aLllo auo O'ose cose '/lith assccaze ~'.IIY:ps with asscciatoo securec, mm-ps securoo.
I. . E1-:c7E tl:es E11-FC07B +/-es rot no aLllo aico oJYen open
~. . Civt Civ II Rh~Rpusuctcnpsthncjnccn pump sucron :;r,'ps 00 not funcfcfl.
~. B&o
. 6&DRRPupinDSAreinoc.
RHR Pump in~4Jts to ./IDS A Lagie are inop.
10, Recirc valves 1D. Recire valves 832-F032B 832-31B aLllO 832-F032B 8.& 832..,""00IB dcsjre log\~
auo dOSlJre Iogc inop.
in. 'Ial',-e5 will sen valves \'Iii! s: auto ao dose close on aa LOCA on concun-ent v.ith LCCA concurrent with aa low-pressure ow-pressure signal from from elv.
Civ. ! log-c.
logo.
11.
- 11. ReceiReceive annimoar ./13-2-1
..-e annl.L~olar"::f" /2-2-7 HPO1 ON HPCI Div fI1 lsolalil),!
sols:ior Lagie Logic 1. HI?CI t-IPCI oi'lll Dlvii Isolation Isolation '/alves valves \'Iill will nDt no .lute oca& F002.
auto o*ose. F002, F042.
F42, and and F019.
5079.
. Receive annoociaOCl"
[:I. Receive annunoarA1-8-5 .11.1-6-5 HPC1 olV HPCI 1-igh Level Trip DIV IIl H"gh Tnp 1. Civil Oi'l II ftgh
!lgh revel revel thp trip seas sea\!; in on a a loss 105s of pOVter.
pOwer.
~. Hiah
. I-ugh level level trill bip wi!
wl still still foodiOli ftmciio HPC1 ON HPCI Div III low Low t.e ..-e122 !n:tla;iOII Level nciacion 1.1. ON Div IIl LawLow t.e ..-e122 initiatoo level nitacn logiclagicinop. KPCi can lnop. HPC! stil inilia"ie can still from Dill inti:e from DivcIlog-c.
logic.
RCIC Cj'/lI RelC lnibatcn LOQ-o Civ tl Initiation Logc RCIC Low 1.I. RelC Lots Lev,,!Level 22 initiation initiation lo:lic irq. Rete inio incp. ROC can can Sli11 initiate !rem stil initiate ttm Dill Div II logic.
logic.
1001-50 Rev. 45 Page 41 of 1321122
Unit Unit 22 APP-UA-23 APP-UA-23 1-'/
1-Page Page 11 of cf 33 arr A 250! BAIT 2SC'V A UNDERVOLTAGE UUDERVOLTGE AUTO ACTICNS AUTO ACflCtTS CAUSE CAUSE 1.
2A-i2A-2).
1.
~. Low DC output voltage from Low from 8attery Battery charger Charger 21..-1(21..-2).
2A-i(2A-2).
- 4. Battery charger B.attery 2A-i2A-2)
Charger 21..-1 i2A-2;' AC input breaker tripped.
5.
- 5. Circuit malfuncti.on.
Circuit malfunction.
OBSERVATLCflS OBSERVATIONS
- 1. Local .ammeter
.Local ammeter on battery ground detector unit is indicating greater than ~.O 1.0 milliamps.
- 2. Battery Bus 2A-l B.attery i2A-2 i voltage as read on 8AT-'JJ.\-?37 2A-i2A-2 BAr-VN-n7(739:(739 i on RTG8 RIGS ru-i is less than the normal range of 130 Panel XU-l 130 to 140V 14W DC.PC.
1.
- 1. Battery Charger 21..-1 B.attery 2A-1(2A-2)
(21..-2) DC output voltage as read locally e.n on BAT-vr4-6oogeoio; BAT-ilM-6009 (6()lOi is less than the normal range of cf 1313 130 te*
to H!JV 1403! DC DC..
4.*
.;1 Battery charger Battel'y Charger 21..-1 2A-i2A-2)
(2A-2) DC output current as read le.cally locally on BAr-vM-ooo6ooi; is reading more or less than the normal range e.f BAT-'.'M-6()()()i6()01) of 10 to 50 .amps.
amps.
- 5. Battery Charger 2A-l.;2A-2) 2A-i2A-2 AC Input 8reaker,Breaker, compartment Compartment COS coscoe (006) on MCC
.:m 2CA,.. is in the OFF or TRIP position.
t4CC 2CA ACPLCttS ACTICNS
- 1. tetermine which battery bus caused the annunciator.
Determine
- 2. tf the c.ause If cause of the annunciator is a gro'.md, ground, perform the DC ground isolation procedure per OP-51, 02-51, DC Electrical system.
- 1. rf the c.ause If cause of the annunciator is excessive le.ad, load, reduce all unnecessary loads on 8attery Battery 8'.1S Bus 2A-l(2A-2).
.2APP-UA-23 12APP-UA-23 Rev. 59 Page 20 of 92921 Categories KJA:
KIA: 295004 A1.02 Al.02 Tier!
Tier / Group: TIG1 T1G1 RO Rating:
RORating: 3.8 SRO Rating:
SRORating: 4.1 LP Obj: CLSLP16*15E CLS-LP-16*15E Source: BANK Cog Level: HIGH Category 8: Y
- 40. Unit One
- 40. Unit One isis operating operating at rated power at rated power with with DG1 DGI running running loaded loaded for for aa monthly monthly load load test.
test.
A fault A fault trips the Main trips the Main Generator Generator Primary Primary Lockout Lockout relay.
relay.
BOP Bus BOP Bus 1IC fails to C fails to transfer transfer on on the the generator generator lockout lockout due due to failure of to failure of the the SAT SAT supply supply breaker to breaker close.
to close.
identifies the status of the Which one of the following identifies the E1 El Bus Bus that would be be reported reported to CRS?
the CRS?
El is E1 energized from:
is energized DGI with off-site power available.
A. DG1 B. both DG1 DGI and from off-site power.
DGI with off-site power unavailable.
C. DG1 Dv D~ off-site power with DG1 DGI running unloaded.
Feedback Feedback K/A: 295005A1.07 KIA: 295005 Al.07 to operate Ability to operate and/or andlor monitor monitor the the following following asas they they apply apply to to MAIN MAIN TURBINE TURBINE GENERATOR GENERATOR TRIP:
TRIP:
A.C. electrical A.C. electrical distribution distribution (CFR: 41.7 (CFR: 41.7 /45.6)
/45.6)
ROISRO Rating:
RO/SRO Rating: 3.3/3.3 Objective:
CLSLP27*05 CLS-LP-27*05
- 5. State the effect that actuation of a main generator lockout relay will have on the Main Generator and station loads.
CLS-LP-39 /Objectives 3,7, CLS-LP-39/0bjectives 3, 7, 12 12
- 3. Given plant conditions, determine if EDGs will automatically start.
- 7. Given plant conditions, determine if:
- a. EDG output breaker will trip
- b. E Bus Master/Slave breakers will trip with the EDG in manual mode
- 12. Given plant conditions, determine if permissives are satisfied for the EDG output breaker to close (either automatically or manually).
Reference:
SD-39, Sections 3.2.4, 3.2.6, 3.2.7, and 3.2.10 Cog Level: High Explanation: Based on the conditions the RO will have to determine the status in order to report to the CRS which meets the monitoring AC electrical on a generator trip.
Generator primary lockout is a loss of off-site power signal to DG auto start logic. All four DGs will auto DGI auto start signal will trip the DG1 start. The DG1 DGI output breaker to allow the DG to transfer from the manual to auto mode of operation (governor, voltage regulator, trip circuits). The DG will then tie back onto the bus once the bus stripped interlock and bus undervoltage interlock is satisfied. Otherwise it will continue to run unloaded. Bus 1 1C C fails to transfer from UAT to SAT on the trip. This results in loss of BOP bus 1C but this bus feeds E2, not E1 El so Bus E1El will remain powered from off-site power (BOP bus).
DG2 will auto start, tie to bus E2. Bus E1 El is being powered from off-site power via BOP bus 1 1 D and the DG1 is running unloaded Distractor Analysis:
Choice A: Plausible because if the peaking relays on the E E Bus actuated and tripped the master/slave breakers. Peaking relays could actuate during a fast transfer with a DG in parallel if the turbine had tripped resulting in a backup lockout rather than a primary.
Choice B: Plausible because this would be the most likely configuration if the turbine tripped resulting in a backup generator lockout instead of a primary lockout (would not produce a LOOP signal).
Choice C: Plausible because since this would be the configuration if the BOP bus that failed to transfer to the SAT was bus 11D rather than 11C.
Choice D: Correct Answer SRO Only Basis: N/A Notes
3.2A 3.2.4 Automatic Start circuity actuates on a foss The DG auto start circuitry loss of power at designated points in the pfant paints plant electrical elecbical system and also actuates on a loss-of-coolant accident. The following is a list of the parameters or foss-of-coolant conditions which will initiate an auto start of of the EOGs.
EDGs, Each of the automac starting logic automatic logic schemes is discussed below.
SD-39 1 SD-39 Rev. 1010 35of Page 35 125 of 1251
- 2. Electrical System Faults (Figure 39-12)3912)
Loss Of Off-Site Power (LOOP)
A loss (LOOP> DG auto start signal will be generated for all four EDGs if any pg.g one of the Ibllowing following conditions exists on either unit:
- Generator Primary Lockout for either unit (Division II logic onlY) only) which is caused by:
generator overall differential generator reverse power distance relay generator output breaker failure UAT differential phase overcurrent Generator loss offield of field
- SAT Lockout (Division II logic only) for either unit.
- Generator Differential Lockout (Division IIIllogic logic only) either unit.
- Transformer Bus Differential lOCKout Lockout (Division II logic only) for either unit.
- SAT secondary side undervoltage.
3.2.10 3.2.10 Trip OfOf The The Bus dO Master Bus CIO Master Slave Slave Breaker Breaker To The The E F Bus Bus Any device that that trips the Master Master Breaker Breaker will cause cause the Slave Slave Breaker Breaker to trip, and any any device device that trips the Slave Slave Breaker Breaker will cause the the Master Master Breaker to Breaker to trip. AA trip Master/Slave breaker trip of the MasterlSlave breaker will result result in in an an E E bus bus undeioltage condition under/oltage condition allowing allowing the EDG to the EDG to auto auto start start and tie tie onto the the associated E E bus bus once EDG EDG i)reaker breaker permissives were met.
The Master and/or Slave Breakers will trip if: it
- Overcurrent is detected on any phase of master (or slave) breaker side.
supply side.
- Undervoltage exists on the 4160V AC bus feeding me the aSSOciated associated E8us.
E-Bus.
respective
- Degraded voltage is sensed on me the E F (JUs bus (after a 10 second time delay).
Master Slave Breaker Trips with EDG Paralleled (in manual mode) to the Grid With a EDG in a a manual (control room or local) mode of operation, and its output breaker closed, protective relaying at the E F bus is aligned to trip circuit of the Slave breaker to protect the EDG from an ovel1oad the trtp overload condition should the normal source of power to the E bus I)e be lost with the EDG in parallel. These relays sense E bus voltage (27 PK), bus frequency (81 PK) and directional power (32 PK) from the E bus to the BOP bus. Actuation of any of these relays With with the EDG in manual and its output breaker closed, will trip the Srave Slave (and the Master) breakers to separate the EDG EOG from the BOP bus preventing possible ovel1oad overload ofthe of the EDG. With the EDG in parallel, if the normal supply breaker to the BOP bus were opened, the EDG would try to carry the BOP bus loads. Since toads are beyond these loads bend the EDG capal)ility, capability, bus voltage and frequency would drop, and a large power lIow flow from the E E bus to the BOP bus would occur. One or more of the PK relays Will will actuate to separate the EDG from the overload, and allow the EDG EDO to carry the E F bus loads.
Note that the EDG breaker will likely trip due to the auto start signal from loss of BOP bus, and tie back onto the E bus after loads have stripped.
The EDG meanwhile would revert to the auto mode of operation.
Categories KIA: A1.07 295005 Al.07 Tier/Group:
Tier / Group: TIG1 T1GI RO Rating:
RORating: 3.3 SRO Rating:
SRORating: 3.3 LP Obj: CLSLP27*O5 CLS-LP-27*05 Source: PREV PREY Cog Level: HIGH IllGH Category 8: Y
Unit Two has
- 41. Unit has experienced an ATWS.
Which one of the following conditions satisfies the Technical Specification Shutdown Margin requirements for this condition?
Margin condition?
SLC Storage A. SLC Storage tank level reaching 20%20% following SLC SLC initiation.
initiation.
B~
B All control rods fully inserted except one control rod at position 48.
C. All control rods fully inserted except nine control rods at position 02.
D. The reactor is subcritical with reactor power below the heating range.
Feedback K/A: 295006 K1.02 KJA:
Knowledge of the operational implications Knowledge implications of of the following concepts as they apply to SCRAM:
Shutdown margin Shutdown margin (CFR: 41.8 to 41.10)
ROISRO Rating: 3.4/3.7 RO/SRO CLSLP200B*02 Objective: CLS-LP-200-B*02 Definetheterms
- 2. Define the terms listed in Section 1.1.
Reference:
Unit 2 Tech Spec Definition Cog Level: Low Explanation:
SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:
- a. The reactor is xenon free;
- b. The moderator temperature is 68°F; and
- c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
Distractor Analysis:
Choice A: Plausible because SLC tank level is below 32% (HSDBW) which allows raising reactor water level during an ATWS. 0% SLC tank level (CSDBW)is utilized to assure the reactor will remain shut down irrespective of control rod position or reactor temperature. If any amount of boron less than the CSBW has been injected into the reactor vessel, cooldown is not permitted unless it can be determined that control rod insertion alone ensures the reactor will remain shut down under all conditions. The core reactivity response from cooldown in a partially borated core is unpredictable and subsequent steps may not prescribe the correct actions for such conditions if criticality were to occur.
Choice B: Correct Answer Choice C: Plausible because this is on of the conditions in Table 5 (Shutdown Without Boron). The definition of MAXIMUM SUBCRITICAL BANKED WITHDRAWAL POSITION, The lowest control rod position to which all controls rods may be withdrawn in bank and the reactor will nonetheless remain shutdown under all conditions. This position is utilized to assure the reactor will remain shutdown irrespective of reactor water temperature. This position has recently be changed from 02 to 00.
Choice D: Plausible because this is the definition of SHUTDOWN as applied to the reactor in 2EOP-01 -LPC.
Definitions Definitions i.-I
- 1.1 1."1 1.1 Definitions Delinitions (continued)
(continued)
SHUTDOWN MARGIN SHUTDOWN MARGIN (SOM) iSDM) SOM 5DM shall shall be be the amount of reactivity reactivity by by which the reactor reactor is is sLlbcritical subcritical or would be be subcritical assuming assuming that:
a.
- a. The The reactor reactor is is xenon xenon free; free;
- b. The moderator temperature is 68°F; 68F; and
- c. All control rodsrods are fully inserted except for the single control rod of highest highest reactivity worth, which is is assumed to be fully withdrawn.
With control rods not capable of being fLllly fully inserted, the reactivity worth of these control rods mLlst must be accounted for in the determination of SOM. SDM.
Brunswick Unit 2 11.1-5
..1-5 Amendment No. 247 I SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 3:1.1 SHUTDOWN MARGIN (SOM) (SDM)
LCO 3.1:1 3.1.1 SDM shall be:
SOM
- a. :;:: 0.38%
0.35% 11k/k, ak/k, with the higl1est highest worth control rod analytically determined; or
- b. :;:: 0.28% lc1k, 11k/k, with the tile highest worth control rod determined by test.
APPLICABILITY: MODES '1,2,3,4, 1, 2, 3, 4, and 5.
5.
STEP BASES:
STEP BASES:
The Cold The Cold Shutdown Shutdown BoronBoron VI/eight Weight (CSBW)
(CSBW) isis notnot aa quantity which can quantity which can bebe measured measured by by the operator.
the operator. An An SLCSLC tank tank level level of 0% or of 0% or 6080 6080 pounds pounds of of borax borax injected injected are are equivalent equivalent to the to the CSBW.
CSBW. The The Cold Cold Shutdown Shutdown BoronBoron VVeight Weight isis defined defined toto be be the the least least weight weight of of soluble I)oran soluble boron which, which, ifif injected injected into into the the reactor reactor vessel vessel and and unifomlly unifomly mixed, mixed, will will maintain the maintain reactor shut the reactor shut down under all down under all conditions.
conditions. This This weight weight is is utilized utilized toto assure assure the reactor the reactor will remain shut will remain shut down down irrespective irrespective of of control control rod rod position position or or reactor reactor temperature.
temperature.
Cold Shutdown The Cold Shutdown BoronBoron INeight Weight is calculated as approximately is calculated approximately 126 126 pounds pounds of 47%
enriched boron enriched boron in in calculation calculation OEOP-WS-01.
OEOP-WS-01. The actual Cold Shutdown Boron Boron Weight is not used is the procedure steps. These steps used in the steps use a level level in in the SlC SLC tank and and aa weight of of borax borax asas an an equivalent for Cold Shutdown Boron for Cold Boron Weight. These values are are determined in determined calculation OEOP-WS-15.
in calculation 0EOP-WS-15. ItIt has has been been decided decided to to use use 0% 0% to to represent represent Cold Shutdown the Cold Shutdown Boron Boron Weight Weight in in the the procedure.
procedure. This This value value can can be be read read byby the operator on the indication in the Control Room. The borax concentration for Cold Shutdown Boron Weight used in the procedure is 6080 pounds.
Reactor depressurization and cooldown may not not proceed until the conditions listed in RC1P-23 through RC/P-25 Steps RCIP-23 RCIP-25 are satisfied.
001-37.5 1001-37.5 Rev. 8 Page 57 of 90 I MAXIMUM SUBCRITICAL BANKED WITHDRAWAL POSITION The lowest control rod posilionposition to which all controls rods may be withdrawn in bank and the reactor will nonetheless remain shutdown under all conditions. This position is utilized to assure the tile reactor will remain shutdown irrespective of reactor water temperature.
MINIMUM Al ALTERNATE TERN.ATE FLOODING PRESSURE The lowest reactor pressure at which steam flow through open SRVs is sufficient to preclude any clad temperature from exceeding 1500 1 500F 0 F even if the reactor core is not completely covered OEOP-01-UG IOEOP-01-UG Rev. 55 Page 69 of 151 I
STE PS RC/P-23 through RC/P-25 STEPS RC/P-2 5 (continued)
Injection of the Cold Shutdown Boron Weight into the reactor vessel also provides adequate assurance that the reactor is and wil! will remain shut down.
RC?P-23 is used to direct the proper actions. If the reactor is not shutdown, then Step RC/P-23 the pressure control actions remain in place. If lf the reactor is shutdown, then the subsequent steps can be used to detemline sul)sequent determine if the reactor cool down can proceed.
cooldown Shutdown as applied to the reactor is defined as suilcritical subcritical with reactor power below the heating range.
If no boron has been injected into the reactor vessel, depressurization and cool down cooldown may proceed as long as control rod insertion is sufficient to shut down the reactor.
Such action is permitted even though the existing margin to criticalii'l criticality may be small. A return to criticality under these conditions is acceptable because ternlination temination of the cooldown will stop the reactor power increase.
If any amount of boron less than the CSBW has been injected into the reactor vessel, cooldown is not permitted unless it can be determined that control rod insertion alone ensures the reactor 'will
,'1m remain shut down under all conditions. The core reactivity response from cooldown in a a partially borated core is unpredictable and subsequent steps may not prescribe the correct actions for such conditions if criticality were to occur.
001-37.5 1001-37.5 Rev.S Rev. 8 Page 58 of 90 I REVISION
SUMMARY
Revision 8 incorporates the Unit 2 2 MSBWP MSGWP position of 00 per EC 56472. Table 5 5 is also changed to reflect that reactor is shutdown under all conditions without boron if aonlY oniy 10 control controL rods are withdrawn to position 02 and no control rod is withdrawn beyond position 02."02. The Unit 22 value for Group 1 1 low pressure isolation has been changed from 850 to 835 per EC 50554.
001-37.5 1001-37.5 Rev.S Rev. 8 89 of 90 Page S9 I TABLE 5 SHUTDOWN WITHOUT BORON ONLY ONE ON CONTROL ROD NOT FULLY INSERTED INSERJtO NO MORE THAN 10 CONTROL RODS WITHDRAWN TO POSITION 02 AND NO CONTROL ROD WITHDRAWN BEYOND POSITION 02 AS DETERMINED BY REACTOR ENGINEERING
Categories K/A:
KIA: K1.02 295006 Kl.02 Tier/Group:
Tier / Group: T1G1 TIG1 RO Rating:
RORating: 3.4 SRO Rating:
SRORating: 3.7 LP Obj:
LPObj: CLSLP200B*02 CLS-LP-200-B*02 Source: NEW Cog Level: LOW Category 8: Y Y
- 42. Which
- 42. Which one one of of the the following following correctly correctly completes completes the the statement statement below?
below?
Excessive moisture Excessive moisture carryover carryover is is caused caused by by (1) reactor water (1) reactor water level level and and results results in in (2)
(2) steam quality exiting the reactor steam quality exiting the reactor vessel. vessel.
A. (1 A. high (1)) high (2)
(2) higher higher B (1)
B!'" (1) high high (2) lower (2) lower c.
C. (1)) low (1 low (2) higher (2)
D. (1 D. (1)) low (2) lower Feedback K/A: 295008 K1.01 KIA:
Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR WATER LEVEL:
Moisture carryover (CFR: 41.8 to 41.10)
RO/SRO Rating: 3.0/3.2 Objective: CLS-LP-01 *08 *08
- 8. With regard to moisture carryover:
- a. Define the term.
- b. Describe how it is affected by reactor water level.
- c. Describe the adverse affects.
Reference:
SD-01 SD-01, Revision 07, Page 22, Section 2.1 .14.a 2.1.14.a Cog Level: Low Explanation:
Explanation:
Moisture Moisture carryover is is defined as that moisture entrained in in the steam exiting the Reactor Pressure Pressure Vessel.
The amount of carryover is is affected by the reactor water level.level. If If the water level level is is too high, the water draining draining out ofof the separators tends to back up up resulting inin increased moisture out out the top of of the separators.
separators. TooToo much much moisture moisture will will overload overload the the steam steam dryers dryers with with aa resultant resultant decrease decrease in in steam steam quality quality exiting exiting the the reactor vessel.
vessel.
Distractor Distractor Analysis:
Analysis:
Choice Choice A: A: Plausible because (1)
Plausible because (1) is is correct correct and and (2)
(2) is is easily easily confused confused with with Moisture Moisture Content, Content, which which would would be be HIGHER.
HIGHER.
Choice Choice B: B: Correct Correct Answer Answer Choice Choice C: C: Plausible Plausible because because (1)
(1) low low reactor reactor water water level level results results inin Carryunder, Carryunder, and and (2)
(2) is is easily easily confused confused with with Moisture Moisture Content, Content, which which would would be be HIGHER.
HIGHER.
Choice Choice D: D: Plausible Plausible because (1) low because (1) low reactor reactor water water level level results results inin Carryunder, Carryunder, and and (2)
(2) isis correct.
correct.
SRO Only Basis:
SRO Only Basis: N/A N/A
Notes
- a. Moisture Carryover Moisture Carryover Mois:ure carr/over Moisture carryover is is defined as that moisture moisture entrained in in the steam exiting the the Reactor Reactor Pressure Pressure Vessel. The amount of carryover is carr/over is affected byby the reactor reactor water level. IfIf the water level level is too high, tile the water draining out of the separators tends to back up resulting in increased moisture out the top of the separators.
Too much moisture will overload the steam dryers with a resultant decrease in steam quality exiting the reactor vessel.
The amount of carryover is minimized in order to: 1) 1) increase turbine w'ear, turbine efficiency, 2) decrease [uri)ine wear, and 3}3) minimize the amount of radioactivity carried over to the balance of plant (BOP).
SD-O1 18D-01 Rev. 7 Page 22 of 81 1
- b. Carryuiider Steam Carryunder Steam canyunder cariyunder is defined as that steam entrained witll with the liquid draining to the tile downcomer from the steam separators and dryers. Carr/under Carryunder is always present to some extent, but can become excessive due to a low reactor water level condition when steam is pulled down into the bulk water region below the feedwater, The problem with an dryer skirt and mixed with feedwater.
camjunder condition is tllat excessive steam carrjUnder that this entrained steam results in a lower density fluid reaching the reactor recirculation pumps and let jet pumps and decreasing tile the available net positive suction head (NPSH). The decrease in NPSH increases the chance of racirculation recirculation pump and jet pump cavitation. Excessive steam carryunder carl'jUnder also decreases the margin to Core Thermal Limits (MCPR).
SD-01 ISD-01 Rev. 7 Page 23 of 81 1 Categories K/A:
KIA: 295008 K1.01 Kl.OI Tier/Group:
Tier / Group: T1G2 TlG2 RO Rating:
RORating: 3.0 3.0 SRO Rating:
SRORating: 3.2 3.2 LP Obj: CLSLP01*08 CLS-LP-OI *08 Source: NEW Cog Level: LOW Category 8: Y
- 43. Unit
- 43. Unit Two Two is operating at is operating at rated power when rated power when the following conditions the following conditions are are observed observed by by the the RO:
RO:
Core Thermal Core Thermal Power Power initially initially drops drops below below and and then then stabilizes stabilizes slightly slightly above above 100%.
100%.
Main Generator electrical Main Generator electrical output output (MWe)
(MWe) lowers.
lowers.
Which one Which one ofof the the following following events events caused caused the parameter changes the parameter changes observed observed above?
above?
A. A A. A single single control control rod rod drop.
drop.
B An open Safety B!" Safety Relief Valve.
C. Reactor Recirculation Pump 2A speed rising.
D. 4A Feedwater Heater Extraction Extraction steam isolation.
Feedback K/A: 295014 A2.03 KIA:
determine and/or Ability to determine andlor interpret interpret the following as they apply to INADVERTENT INADVERTENT REACTIVITY REACTIVITY ADDITION:
Cause of reactivity reactivity addition (CFR: 41.10 (CFR: 41.10/43.5 / 45.13) 143.5/45.13)
RO/SRO Rating: 4.0/4.3 CLSLP302M*01 c Objective: CLS-LP-302-M*01
- 1. Given plant conditions, determine if the following Abnormal Operating Procedure(s) (AOPs) should be entered:
- c. OAOP 30.0, Safety/Relief Valve Failures 11/09/2008 NCR# 305697, Unit 2 SRV "H" 11/09/2008 H Stuck Open SCRAM
Reference:
sd-20 Cog Level: High Explanation:
An SRV opening initially reduces reactor pressure (increasing Voids) causing reactor power to lower. The EHC system senses lower PAM pressure and reduces TCV position to restore pressure. This causes reduced steam flow to the Main Turbine and lower Generator MW output throughout. Reduced steam flow to MT causes less extraction steam flow to FWHs, causing reduction in final feedwater temperature to the reactor, which combined with pressure restoration, raises reactor power above the initial power level.
Distractor Analysis:
Choice A: Plausible because a control rod drop does provide positive reactivity addition. Generator MWe would also increase.
Choice B: Correct Answer Choice C: Plausible because 2A RR pump speed rising would provide positive reactivity addition.
Generator MWe would also increase.
Choice D: Plausible because extraction steam isolation does cause feedwater temperature reduction positive reactivity addition. Generator MWe would also increase. A FWH tube leak would look similar to the SRV opening which is different from the extraction isolation.
SRO Only Basis: N/A Notes
4.23 4.2.3 ADSISRV Failures ADSfSRV Failures Abnormal Operating Abnormal Operating Procedure Procedure AOP-30.0, AOP-30.0, Safety/Relief Safety/Rellef Valve Valve Failures, includes Failures, includes the the following following as as symptoms symptoms ofof SRV SRV failures:
failures:
- SAFETYIRELIEF V/\lVE SAFETY/RELIEF VALVE OPEN OPEN annunciates annunciates (A-03 (A-03 HO).
1-10).
Open indication on Panel P601 for the affected valve.
- Process computer prints out the affected valve number.
- Feedwater rlmv Feedwater flow is greater than steam flow due to the main steam flow detectors being located downstream of the ADSJSRVs.
ADS1SRVs.
- Generator power decreases.
- Reactor vessel level increases due to swell and then settles out at a lower level due to steam flow/feed flow mismatch.
- Suppression pool level oscillating.
- Temperature in the SRV discharge pipe is above normal on B21-TR-R614 (Panel H12-P614).
H12-P614)
- (CS-XU-73) for the affected Millivolt signal on the FFD Cabinet (CB-XU-73:1 normaL SRV higher than normal.
- Suppression pool level and temperature increases.
Automatic actions which occur as a result of an SRV failure include
- The Feedwater Control System establishing an equilibrium water level below the normal level I SD-20 SD-20 Rev 2 Rev. PAGE 28 of 61
1.0
'1.0 SYMPTOMS 1.1 SAFETWRELIEF 'v'AL SAFETY/RELIEF VALVE \lE OPEN OPEI (l\-03 (A-03 -i 0)c is in alaml.
alarm.
1.2
.2 SAFETY OR CR DEPRESS VL'v' LEAKNG (A-D31-1)
VLV LEAKI'NG IA-03 1-1) (is in alaml alarm..
1.3
.3 Open indication on Panel Ptl01 POC 1 for the affecled affected safetylrelief valve valve..
1.4
.4 Process Computer Proce.ss Con-puter .Alarm Alarm Display :indicating indicating the affected aected safetjtrelief safetyfrelief valve number..
number 1.5
.5 Steam tlowifeed flow mismatch flO\\'tfeedi ilow nismatch with flow greater than steam ow wi'th feed flO\\' flow.
18.6 Generator power decrease decrease..
1.7
.7 Reactor vessel level increase due to swell. Le~'el may seme swelL Level settle out at a lower value due to steam
'"alue steani io&feed ow mismatch.
flow/feed fl.ow 1.8 Suppression pool lever level oscillaiion.
oscillation.
1.0 1.9 Suppression pool level i.nc~ease .
level increase.
1.10
. 10 Suppression pool temperature inorease increase..
1.11
. '1 Safetreiief val Sa.fef>p'relief valve
..'e leak detection temperature.
temperature, as read on 821-TR-R614 821-Ti3.R614 at Panel P614, indicaies Pa.nel indicates higher :lhanthan nomlal normal for the afiiectedsafetyfrelJef affected saetyfrel ef ""olive.
valve.
- 12 1.12 Safetyirelief val Safetylrelief valve
..'e noise amplitude millivolt signal.
signal, as read on Fluid Flow Cabnet CB-XU-73.
Detector Cabinet C2-XU-73, indicates higher than norma:l normal for the affected safely/relief valve, s.afety)relief valve.
2.0 AUTOMATIC AUTOMATIC ACTIONS 2.1'1
- 2. IF Digital Feedwa.ter lF Feedwater level Level ControE Control Sys1em System remain,s remains in 3 element control.
confrol, the sysiem THEN 'the system wi equilibrium reactor vessel wa~er willI establish an equi;librium water level the original leveL below :lhe level.
2.2 IF Digital Feedwater level Level Control Sys~e.m shifts ~o System shffls to single sngle element control.
confrol, THEN reactor vessel level should return to to approximately the origillnallevel.
orignal level.
2.3 The EHC EHC system will rediuce reduce generatio!f generator loadi necessary to maintain reactor load as neoessary pressure.
j QP-3O .0 IOAOP-30.0 Rev 18 Page 20f 91 2 of g
4.0 GENERAL DISCUSSION Positive reactivity insertion will cause an increase in reactor thermal power. SomeSonic of the causes of cold water addition are loss Foss of feedwater heating, raising the speed of a reactor recirculation pump, control rod drop, and inadvertent HPCI or initiation. The severity of this transient is determined by how long the RCIC initialion_
abnornially high power level is sustained, especially on a abnoffilally a loss of feedwater heating.
The OPRM system provides alarms and automatic trips as applicable. If the OPRM System is inoperable, then Tech Specs require an alternate method to detect and thermal hydraulic instability oscillations in accordance with BWR Oofmer's suppress themlal Owners Group Guidelines for Stability Interim Corrective Action, June 61994_
6 1994. This requires three stability monitoring regions (Region A - manual scram, Region 8
- B immediate exit, and 5% Buffer)_
Buffer).
5.0 REFERENCES
[jJ
[]I] 5.1 NEDQ-32465-A. licensing NEDO-32465-A, Licensing Topical Report: Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applicability GE Nuclear Energy, August 1996 rnTI 5.2 SOER 84-2, Control Rod Mispositioning
[BIJ 5.3 General Electric Service Information Letter No. 251/251, Supplement 1 1 5.4 OAOP-04.4 Jet Pump Failure 6.0 ATfACHMENTS ATTACHMENTS 11 Estimated Total Core Flow vs. vs_ Core Support Plate Delta-P 2 Confirmation of Reactor Recirculation Pump Forward Confimlalion Foiward Flow 2AOP-03.0 I2AOP-03.0 Rev. 15 141 Page 10 of 14 Categories K/A:
KIA: 295014 A2.03 Tier / Group: T1G2 TlG2 RO Rating:
RORating: 4.0 SRO Rating:
SRORating: 4.3 LP Obj: CLSLP3O2M*O1C CLS-LP-302-M*0IC Source: NEW Cog Level: HIGH Category 8: Y
- 44. Which one one of the following correctly correctly identifies identifies why the RWM RWM is is bypassed bypassed lAW lAW LEP-02, LEP-02, Control Rod Alternate Control Rod Insertion?
Insertion?
Bypassing the RWM Bypassing RWM bypassesbypasses rod rod (1 (1)) Blocks to allow control rods Blocks rods to bebe (2)
(2)
A. (1 (1)) Select Select (2) selected B. (1(1)) Select (2) inserted C. (1 (1)) Insert (2) selected D
D~ (1) Insert (2) inserted Feedback K/A: 295015 K3.01 KIA:
Knowledge of the reasons for the following responses as they apply to INCOMPLETE SCRAM:
Bypassing rod insertion blocks 41.5/45.6)
(CFR: 41.5 145.6)
RO/SRO Rating: 3.4/3.7 CLS*LP007*02d Objective: CLS-LP-007*02d 2.d State the purpose(s) of the following RWM components: Bypass Switch
Reference:
LEP-02 Cog Level: Low Explanation:
LEP-02 is the procedure tht we use to insert rods that have failed to insert on a scram.
When the keylock switch is placed in the BYPASS mode, there are additional contacts that override the outputs (annunciator, insert block, etc.) from the NUMAC RWM. The BYPASS mode menus, displays and functions are identical to the OPERATE mode menus, displays and functions with the exception that the mode will be displayed as BYPASS. The NUMAC RWM will continue to calculate, display and enforce sequence conditions - however the keylock switch contacts will prevent any actual rod blocks from occurring. Placing the RWM NORMAL/BYPASS NORMAUBYPASS switch to BYPASS "BYPASS" to insert rods defeats the RWM interlocks.
Distractor Analysis:
Choice A: Plausible because the RWM does input to the RMCS to provide for Select Blocks, which if not bypassed will prevent control rod selection.
Choice B: B: Plausible Plausible because because the the RWM RWM does does input input to the the RMCS to provide for Select Select Blocks, Blocks, which which ifif not not selected selected will will prevent prevent inserting inserting control control rods.
Choice Choice C: C: Plausible because because bypassing the the RWM RWM does does bypass insert insert blocks, blocks, but but inserting inserting via via RMCS is is only possible possible ifif aa control control rod rod is is selected.
Choice D: D: Correct Correct Answer SRO SRO Only Only Basis:
Basis: N/A N/A Notes Notes
Section 55 Steo: Section Source: PSTGPSTG RC/Q-6.2 RC1Q-6.2 Justification of Difference: Section 5 provides the plant-specific steps required required to insert control rods with the Reactor Manual Control System (RMCS) defeating defeatin RWM interlocks. The plant-specific steps included in Section 5 are beyond the RWM scope of the PSTG but are required to meet the intent of the PSTG.
Discussion: The purpose of Section 5 is to insert control rods '.vith with RMCS. This method is best applied when only a few control rods cannot I)e be inserted, alternate methods are being performed perlormed which cannot cannot be performed continuously, RPS cannot be reset, or individual control rod scrams are not effective.
effective To assist in driving control rods it is possible to maximize drive pressure by starting both CRD pumps; throttling open Flow Control Valve, C11-F002A C11-FOO2A (F002S) [C12-F002A (F002B)[C12-F002 (F002B)];
(FOO2B)1; and, if necessary, throttling closed Drive Pressure Valve, Valve.
(C12-PCV-F0O3. Placing the R\I'VM C11-PCV-F003 (C12-PCV-F003). RWM NORMAUSYPASS NORMALJBYPASS switch BYPASS to insert rods defeats the RWM interlocks.
to "BYPASS" 001-37.1 1001-37.1 Rev. 13 85 Page 20 of 851 FIGURE 07.1-22 Function RWM-OD Bypass function AC Power Source SD-Oil 18D-07.1 Rev. 7 Page 125 of 125 Page 125 1251
3.4A 3.4.4 RWM Operator Display Interface The RWM-OD is interfaced to the RWM-CD to provide a system capability. A contact set provides the bypass capability which bypass capabilfty.
includes (Figure 07.1-22):
- Insert permissive bypass, closed in bypass.
- permisstve bypass, closed in bypass.
Withdraw permissive
- Rod drive block bypass, closed in bypass.
- Settle bypass, open in bypass.
8ettle
- Annunciator bypass, open in bypass.
SD-07.1 180-07.1 Rev. 7 Page 28 of 1251125 3.5.3 3.53 RWM Bypass Placing the RWM-OD keylock mode switch in BYPASS will negate all RWM output contacts. If the RWM-CD keylock keyloclc mode switch is in OPER (operate) and the RI/VM-OD RWM-OD keylock mode switch is in BYPASS. the RWM will continue to calculate, display, and enforce BYPASS, sequence conditions - however, the keylock switch contacts will prevent any actual rod blocks from occurring. Bypass capability continues to exist following loss of power to the RWM-OO.
RWM-OD.
The purpose of the RWM bypass capability is provided so that tI"\at the RWM CO CD chassis can be removed, and replaced while the bypass switch is in the bypass state, without interrupting the system function.
When the RWM is bypassed, procedure OGP-10 OGP-1O provides the only control rod movement constraints. Second operator verification of control rod select, position, pOSition, and movement is employed using OGP-1 OGP-11. 1. Bypass is provided to perform maintenance and testing on the RWM without limiting plant operation. Bypass is also provided to enable control rods to be I>e manually inserted without RWM restriction following a reactor scram.
SD-07,1 180-07.1 Rev. 7 Page 30 of 1251125
3.6.4 Chassis OPERATE Mode Bypass - Chassis
- Mode 4 RWM-CD operates as in The RWM-CO in mode mode 11 to provide provide permissive permissive and RWM-CD bypass annunciation. The RWM-CO bypass switch overrides the RWr ...l-CO RWM-CD outputs.
When the RWM is bypassed, the system provides insert and and withdraw permissive information and no annunciation. The capability to receive pemlissive data, system status, and to receive and and transmit rod position data.
record rod scram time data is not inhibited.
3.6.5 Chassis INOP Mode 5 Bypass - Chassis Vhen the RWM-OD is bypassed and the RWM-CD is in INOP, the When only insert/withdrawal insertfwithdrawal permissives are those provided by RMCS.
with this condition.
There is no annunciation associated v.'lih SD-07.1
\ SO-07.1 Rev. 7 Page 40 of 125\
125 1.0 iNTRODUCTION INTRODUCTION 1.1 System Purpose The purpose of the Reactor Manual Control System (RMCS) is to allow the operator to control core reactivity by inserting and withdrawing control rods.
The system consists of the electrical components and logic cirCUits circuits required to monitor and manipulate the control rods. The Reactor Manual Control and/or selection in response to block rod motion andlor System also acts to i)lock protective signals generated by other plant monitoring systems.
Supporting the RMCS is ihe the Rod Position Information System (RPIS) which provides the operator with a a means for determining the positions of all control rods in the core and for observing the pOSition position of a a selected rod in relation to specific adjacent rods. RPIS also provides rod position and identification data to the process computer. For the purposes of this text, RPIS will be considered as a sub-system of RMCS.
I SO-07 SD-07 Rev. 6 I. PageSSof57 Page ofJ I
3.1.3 3.1.3 Rod Motion Rod Motion Inhibits Inhibits Control rod Control rod movement movement can can bebe inhibited inhibited byby preventing preventing rod rod selection, selection.
blockina rod blocking rod withdraw'al, withdrawal, or or blocking blocking rodrod insertion.
insertion. These These actions actions can can be taken directly be taken directly by by various various RMCS RMCS circuits circuits or or in in response response toto signals signals generated by generated by other other plant plant monitoring monitoring systems.
system&
Three conditions Three conditions will prevent aa control will prevent control rod rod from from being being selected:
selected:
'" RPIS inoperable RPIS inoperable
'" Timer Malfunction Timer Malfunction Select Select Block Block
'"* Loss of Loss of 28 28 VDe VDC to the select logic looic A failure in in the RPIS RPIS cancan prevent prevent a rod rod from being being selected or deselect a rod deselect rod already already selected.
selected. Failures Failures that will cause cause the RPIS RPIS to bebe inoperative inoperative are:
are:
'"* Master Clock Failure --the clock regulates the internal functions the clock RPI&
of the RPI8.
'"* Power Supply Failure Failure - The RPIS uses power from the UPS which is converted to 24 24 'lDC
/DC for useuse in the RPIS.
- Card removed or defective - Each position indicator
- indicator provides information to an associated buffer card for processing and use in the RPIS.
1 SD-07 Rev. 6 Page 15 of 571 5J ATTACHMENT 2 Page 44of8of 8 OEOP-O1 ..LEP-02 OEOP-01-lEP-02 Alternate Control Rod Insertion Step 5 provides the instructions needed to insert any control rods which are not fully inserted. This step bypasses the Rod Worth Minimizer and inserts the control rods with the Emergency Rod In Notch Override switch. This may be required if any control rods did not fully insert to position 00 or bounced back to position 02 on the reactor scram.
001-37.1 1001-37.1 Rev. 13 13 Page 17 851 17 of 85 Categories K/A:
KIA: 295015 295015 K3.01 Tier / Group:
Tier! TlG2 T1G2 RO Rating:
RORating: 3.4 3.4 SRO Rating:
SRORating: 3.7 3.7 LP Obj:
Obj: CLSLPOO7*02D CLS-LP-007*02D Source:
Source: NEW Cog Level:
Cog Level: LOW LOW Category Category 8: 8: Y
- 45. Control Room
- 45. Control Room evacuation evacuation has has been been directed directed by the Shift by the Shift Manager Manager due due toto toxic toxic gas.gas.
Which one Which one of the following of the following correctly correctly identifies identifies the the required required procedure procedure to to be be entered entered and and the proper the proper order order ofof immediate immediate actions actions performed performed prior prior to evacuation?
to evacuation?
(1) ,is (1) , is required required to to be entered.
be entered.
Insert a Manual Scram Insert a Manual Scram followed by followed by (2)
(2)
A. (1) OASSD-02, A. (1) OASSD-02, Control Control Building Building (2) confirmation (2) confirmation that that reactor reactor power power is is less than 2%
less than 2%
B. (1) OASSD-02, Control Building (2) tripping the main turbine C. (1)(1) OAOP-32, Plant Plant Shutdown From From Outside the Control Room Room (2) confirmation that reactor power is less than 2%
D D~ (1) OAOP-32, Plant Shutdown From Outside the Control Room (2) tripping the main turbine Feedback K/A: 295016G 2.04.01 KIA:
Knowledge of EOP entry conditions and immediate action steps.
Control Room Abandonment 41.10/43.5/45.13)
(CFR: 41.10 143.5/45.13)
There is no EOP for Control Room Abandonment, AOP-32.0 AOP-32.O does have entry entiy conditions and immediate actions. KIAK/A applied to AOP.
RO/SRO Rating: 4.6/4.8 Objective: CLSLP302E*002 CLS-LP-302-E*002
- 2. List the Immediate Operator Actions required in accordance with OAOP-32.0, Plant Shutdown from Outside Control Room.
Reference:
OAOP-32.0, Rev. 47 Cog Level: High Explanation:
Control room evacuation due to:
Control 1.1 Unsafe conditions such as toxic gas, 1.1 Unsafe gas, high airborne activity, or unforeseen emergencies which requires evacuation of the Control Room.
1.2 Control Room habitability, with the assistance 1.2 assistance of aa breathing breathing apparatus, has been evaluated evaluated by the Site Emergency Coordinator, Coordinator, AND an evacuation has has been determined to be necessary.
IF IF Control Room Room evacuation was due due to fire, explosion, or similar occurrence occurrence that could result in in degradation of of the Control Control Room wiring, wiring, THEN THEN EXIT EXIT AOP-32 AOP-32 & GO TO
& GO TO OPFP-013.
OPFP-013.
OPFP-013 directsdirects reference to OASSD-01 OASSD-01 (Index)
(Index) for the appropriate appropriate procedure procedure based based onon fire fire location.
Control Control Building Building fire requires entry into into OASSD-02.
OASSD-02.
AOP-32 immediate actions actions are are to be performed to be performed in in aa specific specific order order so so as to not as to not challenge equipment.
equipment.
- 1. MANUALLY SCRAM
- 1. MANUALLY SCRAM the the reactor.
reactor.
2.
- 2. TRIP TRIP the main turbine.
the main turbine.
3.
- 3. OBSERVE OBSERVE auxiliaryauxiliary power power transferred transferred toto the the SAT.
SAT.
4.
- 4. Unit only: PLACE Unit I1 only: PLACE the the Reactor Mode Switch Reactor Mode Switch toto SHUTDOWN.
SHUTDOWN.
- 5. Unit 22 only:
- 5. Unit only: WHEN WHEN steam steam flow flow is is less less than than 33 xx 106 Ib/hr, THEN 106 lblhr, THEN PLACE PLACE thethe Reactor Mode Switch Reactor Mode Switch to to SHUTDOWN.
SHUTDOWN.
- 6. TRIP
- 6. TRIP both both Reactor Reactor Recirculation Recirculation Pumps.
Pumps.
- 7. REDUCE
- 7. REDUCE reactor reactor pressure pressure to approximately 700 to approximately 700 psig psig using using thethe bypass bypass valve valve opening opening jack.
jack.
- 8. WHEN
- 8. WHEN reactor reactor pressure pressure reaches reaches approximately approximately 700 700 psig, psig, THEN THEN PLACE PLACE the the control control switches switches forfor the the INBOARD and INBOARD and OUTBOARD OUTBOARD MSIVS MS/VS to to CLOSE.
CLOSE.
- 9. PLACE
- 9. PLACE Mode Selector Switches Mode Selector Switches for for Condensate Condensate Booster Booster Pumps Pumps in in MAN.
MAN.
- 10. PLACE
- 10. PLACE Mode Mode Selector Selector Switches Switches forfor the the Condensate Condensate Pumps Pumps in in MAN.
MAN.
- 11. GO
- 11. GO TOTO 11(2)EOP-01-RSP (2)EOP-01-RSP AND PERFORM CONCURRENTLY AND PERFORM CONCURRENTLY as as many many ofof the the actions actions as as possible possible prior to prior evacuation.
to evacuation.
Distractor Analysis:
Distractor Analysis:
Choice A:
Choice A: Plausible Plausible because because OASSD-02, OASSD-02, ControlControl Building, Building, would would bebe required required to to be be entered entered duedue to to control control room evacuation due to fire. The confirmation room evacuation confirmation ofof reactor reactor power power is less than 2% is is less is the next next step in OASSD-01 in OASSD-01 following the Unit SCO the Unit SCO determines determines alternative alternative safesafe shutdown actions are required.
required.
is also a recent This is recent change to the order in in which the RSPRSP immediate immediate actions were.
Choice B:
Choice B: Plausible Plausible because because Plausible Plausible because because OASSD-02, OASSD-02, Control Control Building, Building, would be be required required toto be be entered due entered due to to control control room room evacuation evacuation duedue to fire. Tripping to fire. Tripping the main main turbine turbine is is correct.
Choice C: Plausible because AOP-32 is correct, The confirmation of reactor power is less than 2% is the step in OASSD-01 following the Unit SCO determines alternative safe shutdown actions next step are required. This is also a recentrecent change to the order in in which the RSP immediate actions were.
Choice 0:D: Correct Answer SRO Only Basis: N/A Notes
tO 1.0 SYMPTOMS 1.1 1.1 Unsafe conditions such as toxic gas, high airborne activity, or unforeseen emergencies which requires evacuation of the Control Room.
1.2 1.2 Control Room habitability, with the assistance of a breathlng breathing apparatus, has been evaluated by the Site Emergency Coordinator, AND an evacuation has determined to be necessary.
been detemlined 2O 2.0 AUTOMATIC ACTIONS None 10 3.0 OPERATOR ACTIONS 3,1 3.1 Immediate Actions 3.1.1 explosion.
IF Control Room evacuation was due to fire, explosion, D or similar occurrence that could result in degradation of the Control Room wiring, THEN EXIT this procedure AND GO TO OPFP-013.
3.1.2 WHEN Control Room evacuation is determined to be required, THEN COMPLETE as many of the following actions as possible in the sequence listed prior to the evacuation:
- 1. MANUALLY SCRAM the reactor. D
- 2. TRIP the main turbine. D LI CAUTION Auxiliary power should automatically transfer from the :he UAT JAT to the SAt.
SAT. If aa :ransfer transfer does NO NOT occur, occur, and manual ac:ions actions are taken to restore the buses, buses. reenergizing thethe SAT may resul:
result in an an auto auto start of of plant equipment (Circulating Water Pumps.
equipment (Circulating Pumps, Condensa:e Condensate Pumps, and Condensate Booster Booster Pumps, etc.::.
etc.).
Rev. 47 47 Page Page 220f751 of 75
3.0 OPERATOR ACTIONS 4.
- 4. Jni: 1I only:
Unit on: PLACE PLACE the Reactor Reactor Mode S'Nitch Switch to D S/-?UTDOW.L SHUTDOWN.
5 Jn: 22 onlv:
Unit c::lj: WHEN steamsteam flow isis less than than 3 x 101; iO Ill/hr, Ibhr, D THEN PLACEPLACE the the Reactor Reactor Mode Mode Switch Switch to
- c SHUTDOWN.
S-1UTDOL1/2\.
3.
- 6. TRIP Iloth bo:n Reactor Reactor Recirculation Pumps.
Dumps fl D
CAUTION Lnit 2 onlv:
Unit S:ean flow only: Steam rlow must nust be maintained nantained less than
- han 33 x 11 061b/hr O 11)/hr during pressure oressure reduction to reduction :o prevent
- 0 700 psig to oreven: Group Grouo 11 Isolation.
lsoatior 7
- 7. REDUCE reactor REDUCE reac:or pressure to approximately acproxmately 700 70C psig U D
using the:1e bypass valve vale opening opening jack_
jack.
8_
WHEN reactor pressure reaches approximatelyapproxinlately 700 ps:g.
7O psig, D THEN PLACEPLACE the control co:roI switches for thethe INBOARD and ard OUTBOARD MSIVS MS/VS to CLOSE 9_
- 9. PLACE Mode Selector Selec:or Switches for Condensate Booster Gooster U D
Pumps UI9PS in iT MAN.
18.
- 10. PLACE Mode Selector PLACE SeleD:or Switches for the Condensate D Dufl ps in MAN_
Pumps M4V
- 11. 60 TO 1 GO 1(2)EOP-31-RSP (2)EOP-01-RSP AND PERFORM D CONCURRENTLY as many nany of the actions as possible prior to to evacuation.
3.1.3 :he reactor SCRAM AND MSIV closure could NOT Ile IF the be completed prior pr:or :o to evacuabon, evacuation, THEN PERFORM the following fromfrom the callie cable spread spread area:
t
- 1 RPS 4. U D
- 3. OPENE4#iRPSMGSETB.
4_
& OPENE4#3RPSMGSETB.
OPEN EPA #3 RPS MG SET B. D 5,
&6. OPEN OPEN EPA EP4 #5 #RPSRPS MG SET.4LTERN4 SET ALTERNATE YE SOURCE. Q D
Rev. 47 47 Page Page 33 of of Th 75 1
3.0 TECHNIQUES OF EOP EOP USE USE No lines shall cross or intersect on the flowchart. When more than one line goes to the same point the lines shall be combined so that a path is represented by a single line. At the point wnere where each line meets with the path line a direction arrow shall be included on the line to indicate the direction the line is going.
Connecting lines that run parallel to each other shall have directional flow direction of the arrows appropriately placed to indicate the flo'",'
individual lines.
3.3 Operator Actions 3.3.1 Control Operator Immediate Actions The control operator immediate actions are those actions which may be performed following a reactor scram prior to entering the scram procedure (EOP-01). These actions are not mandatory and shall not conflict with entering the scram procedure. All the control operator immediate actions are located in the scram procedure flowchart.
There are no control. operator immediate actions in EOP-02 through EOP-04. In the event the actions are not performed prior to enteling entering the scram procedure, the scram procedure shall tal,e take precedence.
The control operator immediate actions which should be memolized memorized by control operators, are defined as follows:
- 1. Unit 2 Only: After steam flow is less than 3 x 106 Il)lhr, x 10 lbThr, PLACE the reactor mode switch to SHUTDOWN.
Unit 1I Only: PLACE the reactor mode switch to SHUTDOWN.
- 2. IF reactor power is below 2% (APRM downscale trip), THEN TRIP the main turbine.
- 3. ENSURE the master reactor level controller setpoint is +17O+110M.
- 4. IF two reactor feed pumps are running, AND reactor vessel level is above ÷160
+160" AND rising, riSing, THEN TRIP one.
The EOP actions are those which are contained within EOP-01 through EOP-04, EOP-04. In the event the control operator immediate actions are not performed prior to entering EOP-O1, EOP-01, these actions become EOP actions.
Since the EOP actions are readily available to the control operator, there is no need to memorize memoriZe them.
OEOP-01-UG IOEOP-01-UG Rev.
Rev. 55 55 Page 22 22 of of 151 151 I
3.0 3.0 OPERATOR ACTIONS OPERATOR NOTE:
NOTE: ASSD procedure When an ,A.SSD procedure is is identified identified fOf for a given given ASSD fire area,area. entry' entrj into into procedure is that procedure is NOT made until NOT made until directed directed by by this this procedure.
procedure.
3.1 WHEN the specific location of the fire has Ileen REFER to Table 11 AND Tallie been identified, Table 2 to determine whether an ASSD oU tire area is involved AND the correct procedure.
fire 3,2 3.2 IF the fire is NOT located in an ASSD fire THEN EXITEXIT this procedure.
tire area for either either unit, oU 3.3 IF the fire is in an ASSD fire area on either unit, THEN the Unit IF Unit SCO will assess the situation considering the following:
SeQ Location, size and severity of the fire 0 Effect of the fire on ASSD equipment 0 Limiting Conditions for Operation resulting from fire tire 0U Control. room habitability 0U Effect of the fire on balance of plant 0 lnfomiatior provided in the applicable ASSD Infomlation 0 procedure 3.4 SCO determines the al)ility IF the Unit SCQ ability to confirm reactor jeopardy. THEN PERFORM the less than 2% is in jeopardy, power tess following:
3.4.1 MANUALLY SCRAM the reactor.
MANUAllY oU 3.4.2 CONFIRM reactor power is less than 2% using one of the following:
Neutron Monitoring System oU Indication System Rod Position Indlcation oU OASSD-O1 IOASSD-D1 Rev. 32 Page 6 of 1441144
3.n 3.0 OPERATOR ACTIONS OPERATOR ACTIONS NOTE:
NOTE: IF local IF local ASSD actions actions are are in in progress progress inin aa plant plant area area adjacent adjacent toto the fire the fire area. THEN area, THEN thethe Shift Shift Incident Incident Commander Commander shouldshould be be notified.
notified.
NOTE:
NOTE: A determination that A determination that safe safe shutdown shutdown actions actions are are required required implies implies that that the the Unit Unit SRO has SRQ has decided decided that useuse of of the guidance guidance within the ,.:\SSD ASSD procedures procedures is is necessary in necessary in order order to achieve achieve andand maintain maintain Cold Cold Shutdown Shutdown as as an event event mitigation strategy.
mitigation 3.5 Unit SCQ IF the Unit IF SCO determines alternative safe shutdown actions required:, THEN PERFORM the following:
are required, 3.5:1 3.5.-1 shutdown. THEN IF the affected unit is NOT shutdown, a.
a MANUALLY SCRAM the reactor. D
- b. CONFIRM reactor power power is less less than 2% by using one of the following:
Neutron Monitoring System Dfl Rod Position Indication lndicaion 8ystem System D 3.5.2 IF the fire is in the Control Building fire area, THEN PERFORM the following:
- a. MANUALLY SCRAM Unit 1 1 reactor. D
- b. PLACE Unit 1 1 MSIV control switches in CLOSE. D
- c. MANUALLY SCRAM Unit 22 reactor. D U
- d. PLACE Unit 2 MSIV control switches in CLOSE. D U
- e. OBTAIN Control Room controlled controfted copy of the D U
Plant Emergency Procedures Procedures (PEPs) manual.manuaL
- f. Both units EXIT EXIT this procedure AND ENTER D OASSD-02, OA88D-02, Control Building.
OASSD-O1 IOASSD-01 Rev.
Rev. 32 32 Page 77 of Page of 144 1441 Categories Categories K/A:
KIA: 295016G2.04.O1 295016G 2.04.01 Tier/Group:
Tier / Group: TIG1 T1G1 RO Rating:
RORating: 4.6 4.6 SRO Rating:
SRORating: 4.8 4.8 LP LP Obj:
Obj: CLSLP3O2E*OO2 CLS-LP-302-E*002 Source:
Source: NEW NEW Cog Cog Level:
Level: HIGH HIGH Category Category 8: 8: Y Y
- 46. AA reactor
- 46. reactor Scram Scram was inserted on was inserted on Unit Unit Two Two due due to to aa complete complete loss loss of of RBCCW.
Which one Which one ofof the the following following identifes identifes the the MAXIMUM MAXIMUM time time the the CRD CRD Pumps Pumps areare allowed allowed to be operated lAW OAOP-16.0, to be operated lAW OAOP-16.0, RBCCW RBCCW System System Failure?
Failure?
A. 1.5 A. 1 .5 minutes minutes B. 10 B. 10 minutes minutes C
C~ 20 minutes 20 minutes D. 30 minutes D.
Feedback K/A: 295018 KIA: 295018 A A1.02 1.02 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:
System loads System loads (CFR: 41.7/45.6) 41.7 /45.6)
ROISRO Rating: 3.3/3.4 RO/SRO CLSLP302H*012a Objective: CLS-LP-302-H*012a
- 12. Given plant conditions and entry into any of the following AOPs, explain the basis for a specific caution, note, or series of procedure steps.
- a. OAOP-16.0, RBCCW System Failure
Reference:
Revisionl9, Page 4, Section 3.2.3 NOTE OAOP-16, Revision19, Cog Level: Low Explanation:
A loss of RBCCW will result in elevated CRD pump component temperatures and could possibly lead to CRD pump failure. Both pumps should be tripped if a total loss of RBCCW occurs, but may be run for up to 20 minutes without RBCCW Cooling if directed by the SRO for rod insertions or RPV level control.
Distractor Analysis:
Choice A: Plausible because Reactor Recirculation pumps must be shutdown within 90 seconds (1.5 minutes) with a loss of seal injection and seal cooling flow.
Choice B: Plausible because Reactor Recirculation pumps are allowed continued operation for a maximum maximum of 10 10 minutes with no no RBCCW cooling flow.
Choice C: Correct Answer Choice D: Plausible because 30 30 minutes is the required Drywell cooldown cooldown time prior to RBCCW pump restart.
restart.
SRO SRO OnlyOnly Basis:
Basis: N/A N/A Notes Notes
NOTE:
NOTE: CRD pumps CRD pumps may may NOTNOT bebe operated operated forfor greater greater than than 2020 minutes minutes without without cooling water cooling water except except as directed by as directed by the the Unit Unit SCO SCO under the following under the following conditions:
conditions:
- CRD pump
- .4.A CRD pump isis available available ANDAND alternate alternate control control rod rod insertion insertion isis required required OR OR
- operation is
- CRD pump operation is required for reactor reactor vessel level control control a.
- 8. CRD pumps IF CRD OR reactor OR pumps areare NOT reactor vessel level NOT needed needed for for control control rod level control, THEN TRIP both rod insertion insertion both CRD CRD o
pumps.
pumps.
3.2.4 IF there is a partial loss of RBCCW pressure or service water, THEN PERFORM the following:
water.
1.
- 1. IF any of the following conditions exist, IF exist. THEN REFER to OAOP-18.0 or OAOP-19.0:
High temperatures on equipment cooled by RBCCW
- NSW or CSW header pressure approaching pump 0 shutoff head (approximately 90 psig)
- R8CCW HX OUTLET HDR TEMP HI RBCCW Hi (UA-03 1-3)1-3> in alarm OAOP-10.0 IOAOP-16.0 Rev. 19 Page 4 of 11 I 3.1.13 IF Recirculation Pump seal injectioninjection flow is lost, lost, THEN it should be I)e restored in in accordance with Section Section 8.6, 8.7,8.7, or 8.8 8.8 depending on the cause of the loss loss of ftow and of flow and the condition of of pump sealseal leakage.
3.1.14 3.1.14 IF seal injection flow and If seal and seal cooling water water flow are bothboth lost lost to to an an operating operating Recirculation Pump, Pump, THEN THEN the the seal seal staging staging valve valve must must bebe closed and closed and the pump shut the pump shut down down within 90 seconds.
90 seconds.
20P-02 12oP-02 Rev.
Rev. 139 139 Page 9 011561 Page9of156
3.0 3.0 OPERATOR ACTIONS OPERATOR ACTIONS Z
- 2. MONITOR recirculation MONITOR recirculation pump pump seal seal temperature temperature on on RECJRC. PUMP RECJRC. PUMP TEMP TEMP recorder.
recorder, B32-TR-F?601.
832-TR-R601.
3.
- 3. IF either of the following conditions exist, IF exist. THEN THEN SHUT SHUT DOWN the affected reactor DOWN reactor recirculation pump(s):
pump(s):
-- exchanger inlet temperature for Seal 1I or Seal heat exchanger or Seal 2 exceeds Seal2 exceeds 200"F 200°F
-- RBCCW to the recirculation pump seal heat exchangers is lost for more than 10 10 minutes.
!OAOP-16.0 OAOP-160 Rev. 19 Rev. 19 Page Page 55 of 11 11 !
AHACHMENT 6 ATT.<\CHMENT Page 11 of 11 IJ Required Drywell Cooldown Time Prior to RBCCW Requirecl RBCCW Pump Restart R Reference Rerarence Use U Be NOTE: dr.well temperature ranges given below, the peak local For the local drywell temperature must have cooled to equal to or less than 230°F for the time indicated before RBCCIlV RBCCW pumps may I)e be restarted.
NOTE: CAC-TR-4426 only: If any air temperature indication at or below the 29' 29 elevation reached or exceeded 285~F: 285F, then the Required Drywell Drjwell Cooldown Time shown in Table 11 or Table 2 will be determined using the highest indicated air temperature :::260¢F 260F currently existing at or below the 29' 29 elevation. This action takes precedence over temperature indications above 29 elevation.
the 29' NOTE: CAC-TR-778 only: If any air temperature indication at or below the 29 29' elevation reached or exceeded 284F, 284 4 F, then the Required Drywell Cooldown Time shown in Table 11 or Table 22 will be determined using the highest highest indicated air temperature :::258 258F 4 F currently existing at or below the 29 29' elevation. This action takes precedence over temperature indications above the 2929' elevation.
TABLE 1 TABLE_1
>450°F
>450°F >400°F
>400"F and >350°F
>350"F and and >300°F and
>300°F and CAC-TR-4426:
CAC-TR-4426:
450°F i450"F 400°F i400"F 350°F i350°F >260"F and
>260°F 5;300"F 300°F CAC-TR-778:
>258"F and
>255°F and 5;300"F 300°F 43 minutes 43 minutes 39 minutes 39 minutes 36 36 minutes minutes 30 minutes 30 minutes 22 minutes 22 minutes 20P-21
!20P-21 Rev. 63 Rev. 63 Page 66 Page of 671 66 of
Categories Categories K/A:
KJA: 295018AL02 295018 A1.02 Tier/Group:
Tier / Group: T1G1 TIG1 RO Rating:
RORating: 3.3 3.3 SRO Rating:
SRORating: 3.4 3.4 LP Obj:
LP Obj: CLSLP3O2H*O12A CLS-LP-302-H*0 12A Source:
Source: NEW NEW Cog Level:
Cog Level: LOW LOW Category 8:
Category 8: YY
- 47. The Unit
- 47. The Two Reactor Unit Two Reactor Instrument Instrument AirAir Non-Interruptible/Pneumatic Non-Interruptible/Pneumatic Nitrogen Nitrogen Supply Supply (RNAIPNS) header (RNAIPNS) pressure has header pressure lowered to has lowered to 6565 psig due to psig due to aa leak.
leak.
Which one Which one of of the the following following identifies identifies the the impact impact on on the the Outboard Outboard MSIVsMSIVs duedue toto the the RNAIPNS header RNAIPNS header pressure?
pressure?
The Outboard The Outboard MSIVs: MSIVs:
A. will stay stay open open duedue to the accumulators accumulators associated associated withwith the the valve valve operators.
operators.
B. will stay B. stay open due to the Backup Backup Nitrogen Nitrogen valves, SV-5481 and SV-5482, SV-5482, opening.
opening.
C. closed immediately when RNAIPNS RNAIPNS header pressure dropped below 70 psig.
D D~ eventually drift closed due to the sustained low low header header pressure.
Feedback K/A: 295019G 2.02.37 KIA:
Ability to determine operability and/or availability of safety related equipment.
Partial or Complete Loss of Instrument Air
/43.5 145.12)
(CFR: 41.7 143.5 /45.12)
ROISRO Rating: 3.6/4.6 RO/SRO CLSLP25*08b Objective: CLS-LP-25*08b
- 8. Given plant conditions, predict the effect that the following will have on the Main Steam System:
- b. Loss of Reactor Non-interruptible Air (RNA)/PNS/Backup Nitrogen.
Reference:
SD-46, Revision 10, Page 33, Section 4.2.4 Cog Level: Low Explanation:
The pneumatic sources for the outboard MSIVs are Reactor Building Non-Interruptible Air (RNA) System Division II and Division II. The pneumatic sources for the inboard MSIVs are Pneumatic Nitrogen System (PNS) Division II and Division II when operating in mode 11 and (RNA) System Division II and Division II when shutdown. Unlike the SRVs, a loss of PNS does not result in lining up the BU N2 System to the pneumatic operators. An air accumulator located between the MSIV air operator and the check valves provides backup operating air. The capacity of the accumulator is sufficient for the air operator to exercise the valve through one-half of a cycle (open-to-clos (open-to-closeded or closed-to-ope n) should the supply air to the closed-to-open) operator be interrupted.
Distractor Analysis:
Choice A: Plausible because the MSIVs have have accumulators which will eventually bleed off due to not being completely leak tight.
Choice Choice B:B: Plausible Plausible because because Backup Backup Nitrogen Nitrogen supply supply to to components components located outside and located outside and within within the the orywel I.
Drywell.
Choice Choice C:C: Plausible Plausible because because the the MSIVs not close immediately MSIV's will not immediately due to the due to the accumulators accumulators associated associated with the with the valves.
valves.
Choice Choice D:0: Correct Correct Answer Answer SRO SRO Only Only Basis:
Basis: N/ANIA Notes Notes
The most The probable failure most probable failure mode mode of of the the air air system system isis when when individual individual equipment or equipment or components components of of the the system system fail.
faiL These These types types ofof failures failures would result would result in in reduction reduction of system capacity of system capacity but but would would notnot normally normally restrict plant restrict plant operation.
operation On aa foss On of plant loss of plant air, air, the following general the following general plant plant response response should should be be expected (without expected (without operator action).
operator action).
approx, 70 approx. psig 70 psig MSIVs may MSIVs may start start drifting drifting closed closed after after aa sustained sustained loss at loss at this pressure pressure (due(due to accumulators accumulators on on the valve valve operators) operators) approx. 60 60 psig Loss of control of the condensate loss condensate and feedwater Occurs. Pump minimum fiow system occurs. flow valves start to drift open, feedwater level control control valves drift open, hotweillevel open, hotwell level control control valves drift drift open, open, heater heater drain drain pump pump discharge discharge valves driftdrift open, SULCV SULCV drifts closed.
approx. 40 psig Control rods start drifting in due to outlet scram valves opening.
SD-46 1 SO-46 Rev. 10 Page 33 of 80 1 4.3.5 Main Steam System Pneumatically operated components of the Main Steam System are supplied by the RNNPNS System (refer to Table Tal)le 46-3).
The Outboard MSIV accumulators are supplied from the RNA subsystem and have no backup source of air if RNA pressure is lost.
The Outboard Outl>oard MSIVs will drift closed following a loss of RNA.
PNS nomally nomlally provides aa source of nitrogen nitrogen to the Inboard MSIV and Safety Relief \alves Valves' (SRV) accumulators. However, itit is permissible to supply these loads from the RNA Subsystem Subsystem if Reactor power does does not not exceed exceed 15% 15%. InIn the event the normalnormal RNA!PNS RNNPNS supply is is lost, lost, the Nitrogen backup Backup System will automatically automatically provide operating air to the SRV accumulators when SRV when RNA/PNS RNAIPNS supply header pressure drops to 95 psig.
95 psig SD-46 IS0-46 Rev.
Rev. 10 10 Page36of8t3 Page 36 of 80 1
The most The most probable failure mode probable failure mode of of the the air air system system isis when when individual individual equipment equipment or or components of components of the the system system faiL faiL These These Pipes types ofof failures failures would would result result in in reduction reduction of of system capacity system capacity !Jut would not but would not nom1ally nomially restrict restrict plant plant operation.
operation.
On aa loss On loss of of plant plant air, air, the the following following general general plant plant response response should should bebe expected expected (without operator (without operator action).
action).
- approx. 70
- approx. 70 psig psig MSIVs may MSIVs may start drifting closed start drifting closed after after aa sustained sustained loss at loss at this pressure (due this pressure (due to to accumulators accumulators on on the the operators) valve operators)
- approx. 60
- approx. 60 psig psig Loss of Loss control of the of control the condensate and teed feed system.
Pump minimum flow valves Pump valves start start to drift open, open.
feedwater level control valves feedwater valves drift open, hotweli hotwell control valves drift open, heater level control heater drain pump discl1arge discharge valves drift open, SUlCV SULCV drifts drifts closed.
closed.
- approx. 50 pSig
- psig S/VIA cross-tie valves begin SAlIA begin to close.
- approx. 40 psig
- Control rods start drifting in due to outlet scram valves opening.
OAOP-20.O IOAOP-20.0 Rev. 35 Page 11 of 18181
(fj * . . . . .. . . . . . .. . . . . .. .. NITRO<1EN BACKUP SYSTUI NITROGEN BACKUP SYSTEM p RE MTOGL MEC1OR Lt*G 0
C, 0
ccii s tfJ'1'NS PM rI.*.*4 ib . . . )d--_, lfK141 z
PHS
'112 o
0 -3)
>t)
-I z0
-I I
0 a) 0 SBCOV?I8 0
Cl) * * * * * * ** * ***.*.* " .,. NITROGEN BACKUPSYSTUI NLTRCGEN BACKUP SYSTEM
ç:i IWIOi[MTlIOG(!4 r-------.......~t____l sT~TlIlfloorslC(
RaCIOR BUllDlNQ
- 0 tI)0 Q)
,.o0 ORW(Ell o.,C to
- ~~~ ~w~ ***** ~.ri *** ~~. ~ **** ~ ****** ~.. * ** ~ *** _ *********** ~ -c:
~
z o
0 z0 I
0 CD 0
0
Unit 2 APP UA-01 U.6, ..01 44 4-Page 1 of 2 Page'lof2 INSR AIR PREf:S-lQW INSTR PRESS-LOW ACTIONS (Continued) 4.
4.. Refer to OAOP-20.0.
OAOP-20.U, Pneumatic {AirlNitrogen}
(Air/Nitrogen) System Failures.
- 5. Should a total Icss Shoud loss of interruptible instrument .air air occur, backup pneumatic sources should be placed in service or verified in service for equ pntent equipment bacl up pneumatic sources inn accordance having back a.ccordal1C!! with OOP-48; OOP-46; e.g. Fuel pool pcol gates.
gates.
DEVICE?SETPOINTS DEVICEISETPOINTS Pressure Switch lA-PS-723 IA..PS..723 8-102 98.. '102 psig POSSIBLE POSSlBl;:: PLAN PLANT EFFECTS
- FFECTS 1.
1.. Loss Lcss of service seJVice air.
air.
2.
- 2. Lc'll instrument air Low air pressure may cause abnomlal operation of cause abnormal instruments, valves, valves. and controls contro!s using instrument air air and and may resu resulttin in loss of systems and
[ass of and consequent consequent loss loss of of generation.
3.
- 3. If nonintenruptible If noninterruptible instrument air air also falls in pressure, individual in pressure, individual control rods rods WI
'Hill drift drift in in followed by by main main steam steam isolation iSCllation valve osure closure and and associated associated RPS RPS Group Group 11 isolation isolation scram scram ofof all all control control rods.
rods.
REFERENCES REFERENCES 1.
1.. LL-352-LL..9352 -13 13 2.
- 2. OAOP-20.0, OAOP-20.0. Pneumatic Pneumatic (Air/Nitrogen)
(AirlNiuogen) System System Failures Failures 3.
- 3. 1APP-UA-0 1APP-UA-O'11 5-4, 5-4. Service SerJice Air Air Press-Low Pre.5s-Low 4.
- 4. 1/2 APP-UA-01 1/2 APP-UA-01 5-3, 5-3. IAPP-UA-01 1APP-U.A.-01 6-3e-3 AIR AIR ORVER DRYER TROUBLE TROUBLE 5.
- 5. SOER SOER 38-01.
88-01, Recommendation Recommendation I1 I2APP . UA-01 2APP-UA-01 Rev.
Rev. 64 84 Page Page 6262 ofof 102 1021
7.6 7.13 Nitroçjer System Pneumatic Nitrogen C OUnucus COnUnUIlUIi u**
7.6.1 7.13.1 Initial Conditions NOTE: Securiftg the Pneumalic Securing Pneumatic Nitrogen Nitnogen System without transferring loads 10 to the system will result in the follo'Mng:
compressed air sys!em following:
- drywell RNA headers.
Loss of pressure to dr/well
- PNS Division I and Dillision Division II low ow and low-low pressure annunciations.
annuncialions.
backup initiation signal from PNS-PSL-5843A
- N2 baokup
- PNS-PSL5843A and PNS-PSL-5843B.
PNS-PSL58439.
inboard MSIVs drifting
- Inboard
- diifting closed.
- i. Primary Containment is NOT required 10 to be inerted in D Specification 3.6.3.1 OR accordance with Technical Speoificalion action statement of Technical Speciiicalion aclion Specification 3.6.3.1 3.8.3.1 has been entered a.sas if the limit has been exceeded.
7.6.2 Sreps Procedural Steps CAUTION Edreme cau1ion Extreme cauti or must rjst be used!
usec when wie, operating operatir PNS-VB.
P iS-V. An inadvertent Scram on the ir savenent Scram tie coposite unit opposite wil: occur if the wrong ui: wil! cngaeiscoea:ed.
valve is operated. The Unit 11 hand\\lheei handcheel is painted yellow and! Unit :1: hanclwheel is painted blue.
- 1. IF drywell ()neumatio pneumatic loads are to be transferred to the D Compressed Air System THEN PERFORM Seoiion Com()fessed Section 8.17 AND RETURN to Step 7.6.2.2.7.8.2.2.
ISOLATION lI.4LVE,
- 3. ENSURE the following valves are locked closed:
- a. PPJS DIll.
PNS OIV II/SOLA ii ISOLATION DRYWELL 4LVE TlON TO DRYI*lIELL lI.4L VE.* r:i D
PPJS-V11 PNS-V11
PNS-V12.
- 4. COMPLETE AttachmentS.
Attachment 6. D U
GOR4S IOOP-46 140 Rev. 140 Page 45 of 280 289 I Categories K/A:
KIA: 2950 295019G 19G 2.02.37 Tier/Group:
Tier / Group: T1G1 TIGl RO Rating:
RORating: 3.6 SRO Rating:
SRORating: 4.6 LP Obj: CLSLP25*08B CLS-LP-25*08B Source: BANK Cog Level: LOW Category 8: Y
- 48. The following Suppression Pool temperatures are observed on Unit Two after a Reactor Scram and inadvertent Group 1I isolation.
Time Suppression Pool Temperature 0000 93°F 0002 97°F 0004 103°F 0006 109°F 0008 113°F Which one of the following is the latest time requiring entry into RVCP due to Suppression Pool Temperature ONLY?
A. 0002 B. 0004 C
C~ 0006 D. 0008
Feedback K/A: 295020G 2.04.01 KIA: 2.04.01 Knowledge of Knowledge of EOP EOP entry conditions conditions and and immediate immediate action steps.
Inadvertent Containment Isolation Inadvertent Isolation (CFR: 41.10 (CFR: 41.10/43.5/45.13) 143.5 145.13)
The EOPs at Brunswick do The do not have any immediate operator actions, so the the question is written written for only entiy conditions.
the EOP entry RO/SRO Rating: 4.6/4.8 LOlCLSLP3O0L*O8A Objective: LOI-CLS-LP-300-L *08A
- 8. Given the Primary Containment Control Procedure and plant conditions, determine if the following actions are required:
- a. Manual reactor scram
Reference:
001-37.4, Revision 8, Page 4, Section 3.0 Cog Level: High Explanation:
EOP-02-PCCP requires scram before SPT reaches 110°F, RSP requires entry into RVCP when scram is required by EOP-02, 03 or 04. There are no immediate operator actions for EOP entries.
Distractor Analysis:
Choice A: Plausible because SPT exceeded 95°F which is PCCP entry condition.
Choice B: Plausible because SPT exceeded 105°F which is PCCP entry condition while testing.
Choice C: Correct Answer Choice D: Plausible because SPT exceeded 110°F which is the temperature inserting a Scram is required prior to exceeding.
SRO Only Basis: N/A Notes
REACTOR VESSEL CONTROL I RVCP-1 ENTRY CONDITIONS:
- REACTOR WATER LEVEL IS LESS THAN 166 INCHES
- REACTOR PRESS IS ABOVE 1060 PSIG
- DRYWELL PRESS IS ABOVE 1.7 PSIG
EOP 02.PCCP, PCCP, EOP~
EOP- 03w 03-SCCP, SCCP, OR 04- RRCP EOP- 04*
RVCP- 2 RVCP*2
/
\c1ROL PRIMARY CONTAINMENT CONTROL PCCP-1
/* ENTRY CONDITIONS:
- SUPPRESSION POOL TEMP ABOVE 95"F QRABOVE ABOVE95°FQKABOVE 105°FWHENDUETO 10S"FWHEN DUE TO TESTING
- DRYWELL AVERAGE TEMPABOVE AIR TEMP 15OF ABOVE 1SO"F
- DRYWELL PRESS ABOVE 1.7PSIG 1.7 PSIG
- LEVEL SUPPRESSION POOL WATER LEVEL ABOVE *27 -27 INCHES
(-22 FEET & 3 INCHES)
(-
- SUPPRESSION POOL WATER LEVEL BELOW --31 INCHES 31 INCH ES
(-2
(- 2 FEET & 7 INCHES)
- PRIMARY PRIMARYCTMTH2 CTMT H2 CONCENTRATION CONCENTRATION ABOVE 1.5%
1.S%
PCCP-2
STARTAVPJLABLERHR START AVAILABLE RHR I LOOPSINSUPPRESSION LOOPS IN SUPPRESSION I POOLCOOLINGMODEAS POOL COOLING MODEAS NECESSARY TO NECESSARY TO MAINTAIN MAINTAIN TEMP BELOW 9S*F TEMP BELOW 95F (OP-i1}
(OP-I?)
SPIT-04 nrc
- CAN SUPPRESSION POOLTEMP ...YES
.. BE MAINTAINED BELOW SPIT-os (STAPTALLAVAIi.B START ALL AVAILABLE RHR RHR I LOOPS IN LOOPS IN SUPPRESSION COOLING MODE EXCEPT POOL POOL I RHRRHR PUMPS REQUIRED REQUIRED FORFOR I ADEQUATE CORE COOLING BYCONTINUOUS OPERATION IN LPCI MODE SPIT-GB 110°F SUPPRESSION POOL TEMP REACHES IIBF i10"F REACTOR SCRAM REQUIRED SPIT-B?
..-.-- REACTOR NO
.REQUIRED _
SPIT-GB
]YES INITIATEA INITIATE SCR.AM A REACTOR SCRAM ENTER EOP-OI AND ENTEREOP.01 Categories K/A:
KIA: 295020G2.04.O1 295020G 2.04.01 Tier/Group:
Tier / Group: T1G2 TlG2 RO Rating:
RORating: 4.6 SRO Rating:
SRORating: 4.8 LP Obj:
LPObj: LOICLSLP3OOL*O8A LOI-CLS-LP-300-L *08A Source: NEW Cog Level: HIGH Category 8: Y
- 49. lAW lAW OGP-05, OGP-05, Unit Unit Shutdown, Shutdown, Unit Unit Two is is in in Mode Mode 4, flooding the RPV RPV above above the Main Main Steam Lines Steam Lines prior prior to entering entering Mode Mode 5.5.
Which one of the following symptoms can can bebe used used to determine that a loss loss of Cooling flow has Shutdown Cooling has occurred?
A. Rising vessel pressure.
B. Annunciator REACTOR WATER LEVEL HIGHILOWis HIGH/LOW is in alarm.
C C~ Annunciator CORE SPRASPRAY Y OR RHR PUMPS RUNNING is flashing.
D. Loss of power to RHR Shutdown Cooling Inbd lnbd Isolation Valve, 2-E11-F009.
2-E11FOO9.
Feedback Feedback K/A: 295021 KIA: 295021 A2.02 A2.02 Ability to Ability determine and/or to determine andlor interpret interpret the the following following as as they they apply apply to to LOSS LOSS OF OF SHUTDOWN SHUTDOWN COOLING:
COOLING:
RHR/shutdown cooling RHRlshutdown cooling system system flow flow (CFR: 41.10/43.5/45.13)
(CFR: 41.10 /43.5/45.13)
RO/SRO Rating:
RO/SRO Rating: 3.4/3.4 3.4/3.4 Objective: CLSLP302L*01 Objective: CLS-LP-302-L *01 aa
- 1. Given
- 1. Given plant plant conditions, conditions, determine determine ifif the the following following Abnormal Abnormal Operating Operating Procedure(s)
Procedure(s) (AOPs)
(AOPs) should should be be entered:
entered:
- a. AOP-15.0, Loss of Shutdown AOP-1 5.0, Loss Shutdown Cooling
Reference:
Reference:
OAOP-15, Rev.
OAOP-15, Rev. 23, Page 22 - Symptoms 23, Page - Symptoms Cog Level: High Explanation: When the RHR Explanation: RHR pumps trip the student will knowknow anly from the annunciator being able to be be cleared which indicates that there is no no flow in the system.
The following prerequisites must be met for vessel flooding above the head flange to support head removal:
Reactor average temperature is less than 200°F At least one loop of RHR is in Shutdown Cooling At least one main steam line (preferably MSL A) is available to be used as steam vent While flooding above the MSLs (260 inches to commence filling), level is raised to 330 inches (bottom area of the lower flange). During this period of time annunciators for Hi Reactor Water Level and Core Spray or RHR Pumps Running are expected to be in alarm, and RPV pressure increases are also expected due to limited venting in comparison to filling capability. RHR pump discharge pressure dropping below 115 psig would cause annunciator to clear. This is recent plant specific OE.
Distractor Analysis:
Choice A: Plausible because in Mode 4, raising RPV level above the MSLs will cause pressure to rise.
Any unexplained temperature/pressure rise would be indicative of a loss of SDC.
Choice B: Plausible because a low reactor water level causes SDC isolation.
Choice C: Correct Answer Choice D: Plausible because valve position directly inputs to RHR pump trip logic. Loss of valve 480V or control power would cause this position indication, mechanical position limit switch actually inputs to pump trip logic (no suction path) and is powered through the limitorque limitorque with a source other than the 480 V Breaker or control power.
SRO Only Basis:
Basis: N/A N/A Notes
1.0 1.0 SYMPTOMS SYMPTOMS 1.1 1.1 RHR SW RHR SW PUMP PUMP 1A(2A) 1.4(24) TR1P{A-011-9)
TRIP (A-al 1-9) oror Ir:tHR RHR SW PUMP 1C SW PUMP IC (2C) TRiP (2CJ TRIP (A-0l 3-9)
(.4.-01 is inin alarm.
3-9) is alarm.
1.2 1.2 RHR SW RHR SW PUMP PUMP 1B(28) f2S) TRIPTRIP (A-03 (A-03 1-8) or RHR 1-8) or RHR SWSW PUMP PUMP 10 ID (2D) 2Ql TRIP TRiP (A-03 3-8)
(.4.-03 is in 3-8) is in alarm.
alarm.
1.3 1.3 RHR PUMP RHR PUMP 1A(2A)14(24) TRIP TRIP (A-01 (A-UI 3-8) of RHR 3-8) Of PUMP 1C(2C)
RHR PWI'IP IC(2C) TRIP TRIP (A-01 (A-al 5-8) 5-8) is in alarm.
is in alarm.
1.4 1.4 RHR PUMP RHR PUMP 18(28) 1B(28 !HIP TRiP (A-03 (403 3-7) or RHR RHR PUMPPUMP 1O(2D) 1D(2D TRiP TRIP (A-03 (A-03 5-7) 5-7) is in is in alarm.
alarm.
1.5 1.5 RHR HX AlB RHR 418 DiSCHD1SCH CLGCLG L'VTR WTR TEMP TEMP HI {A-03 (A-03 2-9) is in 2-9) is in alarm.
alarm.
1.8 1.5 RHR AlB 413 DISCH & SUCT HDR PRESS HI Hi (A-03 3-9) is in alarm.
1.7 REACTOR VESS VESS LO LEVEL LEVEL T,r:t/P TRIP (A-05 (405 2-6) is in alarm.
alarm.
1.8 1.8 CORE SPRAY CORE SPRAY OR OR RHR PUMPS .r:tUNNJNG RUNNING (A-03 (403 2-1) 2-1) is is flashing.
1.9 ERFIS valve monitoring alamling alarming on RHR valve closure 1.10 Group 8 Isolation Valves close.
1.11 Increasing Reactor Coolant Temperature and/or andior Pressure.
] 1.12 High NSINNSW or CSW header pressure approaching pump shutoff head (approximately 90 psig).
] 1.13 Unexplained changes in running RHRSW loop flow or pump discharge pressure.
( OAOP-15.0 IOAOP-15.0 Rev. 23 Page2of2l Page 2 of 21 I
Unit 22 Unit MP ISP A- n 2-1 A-fl 2-1 ad 22 Pape 11 of Page CDP.E SPRAY CORE PP.AY OR OP. P.!!R P1 PUl-IPS cuMps RUl:<IIllIG RUNUU7G AUTO ;'CTICNS AUTO ACTICtS 1.
- 1. An A.DS
.An ACS permio.:::si perminoive zignal ia.
!:: !J.ignbl in ge:n~rat.ed generated if if cne Core SE='ra~t one Core Spray or or two two P..HR im puopt io j;:Ul1ptt in one one l~F'loop are are operal:.ing.
operatin.
1.
- 1. Core Spr-ay C.::>re Spray or or REmW purrp running.
pump l."'Unning.
2.
- 2. Circuit m.!sl Cil."cuit. malfunction.
funC't ion.
OESERVATKNS OBSER'iATICNS 1.
- 1. Core Spray c,:>re Spray pump pump(n) (el or or R!IR P1 purrppump(z (<>> "re are c,n.
on.
- 2. Core Spray C,:>re Spray purrp pump diach"r9" diocharge pre""ure precoure greater greater than than 115119 pllig poig iR2l-PI-P.SOOA or (E21-PI-R600A or R~OOBlRO2) .
3.
- 3. RHR pump di"ch:!.L-ge
]JR purrp *iitcharge prea<>urepreozure gre greater
... ter thanthan 115 119 p<>ig prig (Bll-PI-R606A (Ell-PI-P.OGA or or RSCfl)
R6')6Ill.
ACTIONS ACTICNS
- 1. If a Core Spr Spray
... y or RHR P1tR purrp pump haa ban been m!l.ou...lly otarted. t"ke manually etarted, take no actitn.
.:!'ctlc,n.
CAUTION C.lU'l.'ION After an an aut<:matic automatic initiation, initiation. a" EOCS "ubaYlOt.en.
an BCCS tubnyrtem or or RCIC S~'"t.em Syntam ahall ntt ~
ohall not. be ohut down "hut down or placed planed in in """nual manual until at. at lea"t leant t.wo two independent indicat.ion" indioaticno are verified for one of of the the following condit.ion", :cntticno:
I.
1.. Adequate core Adecruat:e core coolinq noting iA eneured...
in enDured
- 2. The init.iation The Initiation !lignal nignal va" war not valid.
3.
3:.. cycten io The ayctA~'rI, funntioning properly in not funct:ioniog properly in t:.he the AutoTlat.ic autonati: mode.
mode.
- 2. If the RHR If Syotem .,,""
P1W. Syat.em Inadvertently "t.art.ed wan in"dvert.ent.ly otarted due to f"ult.yfaulty inotnimentation, *ilh.~D inatrument:4tic*n,. when t.he reactor water level an,d the react.or and dl.'}'Vell drywell precaul."e; pracoure have been verifled h,a,\,re verified nOl."lltalnormal,v ch.ut. thut dj*~D down the REP. P.3W. Svot.em Syoteet ,and rentore it:
and rectore it atanv configurat.ion to a Dt4ndby oonfiguration per OP-17. .
- 3. If the Core Spr Spray Lyotem w
..y Sy"tem wan inadvertently "t.
... o in:ldvertent.ly otarted to e
.. rt.ed due t.o ...ult.y faulty inotnimentation, ~lIheD inotrument.atic,n. when t.h.e reactor *water the r,~~ct.or drywell precl!lure
.... ater level and dl."'Y""'ell prenoure 2APP-A-03 12APp-A-03 Rev. 49 Rell.49 102 Page 22 of 1021
5.3 5.3 Reactor Shutdown Reactor Shutdown Initials NOTE:
NOTE: The follow:ng steps The following steps provide provi direction direotion for far vessel vessel flooding flooding abo above the head
... e the head flange flange to oool ine to cao1 the head head and and studs studs prior prior to to head removal. NIA head remcval. NA the remaining steps the remaining steps ifif NOT applicable.
NOT appIioable.
5.3.65 CONFIRM the follcwing following oondiflOns elst:
concit ars e::<lst:
Reactor average temperature :e:aeatjre is less than han 2C()"F 2CYF teas: one At leas! one loop boo ej o RHR RHR isSn in Shutdo'Nn
.utdo Cooling Cooing At least Al one main east one main steam line lire (preferably ipeferaoly MSL MS_ A) A:I is available to be is he used used as Sleam steam vent NOTE:
NOTE: Floodup above Floodup aoove the the head flange, f are using Jsir eRD injection flow CR0 injection fiowat at 100 gpm with minima. RWCU minima.l .RWCU reject 9cw. ,v...iIIII increase reect flow, norease vessel level at aa rate rate of appro:<imaiely approximately 0.65 incl.r 9:e. Estimated 0.55 inchlrninute. Estimated time me to flood from 220 to f acd from inches is to 415 inches is 8.5 B.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> hours at this flow at w rate.
a:e.
CAUTION Duling vessei iloacip. some D ng vesseilioodup, sonic steaming may may oceur occur at the Me vessel wals, wa s. This This will will be he vented tirough the head through heaa vents to to the Jre. Equipment tne Orywell Equiprcn: Drain Sump S.mo OR te to file me mainntain steam line if NOT f.oc.ded. IF vessel pressure NOT flooded. eaceecs 25 pSig, pressure exceeds osig. THEN the f100dup flcodup rate should sboijd be be reduced oror stopped unlil.in:i! pressure p-essure is a reduced to 25 However, a short term pressure ps:g. However, 25 psig. pressure as high as 40 psig psig is permitted permitted if Ihe the OWED OWED sump temperature is below below 140"F140F AND only the small AOV head vents, vents. .B21-FOO3 B21-F003 and FOO4, F004. are open, open.
CAUTION WHEN flooding Ihe the RPII RPV above aaove lhethe main main steam steam lines, increases iraaees in i., vessel pressure pressure are expecteo. Care should expected, should be he taken beforw concluding taker before concluding an inacvertertn,cde an inadvertent mode change ha.s nas cccured. IF vessel bWk ocrurred. au k average average temperature temp erature is maintained mantained less than 212*F, 212F. THEN a rncde change has NOT occurred.
mode 5.3.63 5.3.86 INCREASE vessel vessel level leve to :c 250 nones.
250 inches.
CGP-05 IOGP-1J5 Rev. 144 Re\'.l44 54jI Page 54 of 6B
CAUTION CAUTION WHEN flooding WHEN flooding ihethe RPVRPV above above lhethe main main steam steam lines, lines, temperature temperature stratification stratiflcation between between the upper the uppervessel vessel and and 100Ner lower vessel vessel are expected expected to to occur occur because forced circulation because forced cirm.dation cooling is hampered above 240 in. due to lhe the vessel intemais internals and nd their lalent latent heat heat.
Average bulk Average bulk water Hater temperature from RCR Of or RHR should shoutd be used below I}0 psig. psig.
5.3.7 5.3.71 INCREASE '1esselle'~ello INCREASE vessel level to 260 inches to commence commence main sieam filling main steam lines.
NOTE:
NOTE: It will It will take approximately 1I hour take a.pproximatelt hour to the main to fill the main steam steam lines lines and and seesee an sn increase in increa.se in vessel level.
level.
5.3.72 FREQUENTLY MONITOR FREQUENTLY MONITOR reactor reactor vessel vessel pressure pressure to ensure less than 25 psig.
ensure lesslhan psig.
5,3.73 5.3.73 FREQUENTLY MONITOR FREQUENTLY MONITOR dly drywell levels AND surnp levels
.... ell sump temperature.
5.3.74 REQUEST E&RC periodically monitor radiological conditions elevation conditions at 5 ft. ele'Jalion of dly diywell....ell due to head venting.
'Jenting.
5.3.75 WHEN main WHEN main steam lines are filled, THEN INCREASE vessel level vessel level to 330 inches, which co.rresponds corresponds to bottom area of the lower flange.
CAUTION DO NOT DO NOT auihorize removal of authorize removal RPV head piping of RPII piping unless RPV level has been been stabilized stabiled at 400 to 420 inches.
5.3.75 5.3.76 SLOWLY INCREASE level to between 400 bo420 SLOWLY to42D inches.
inches. whichwhich corresponds to midway on head dome.
maximum cooftng The maximum (The cooling effect will be seen if level level can be maintained al at the upper upperend end of lhe the lelevel bard.
....eI band.)
ICGP-05 CGP-D5 Rev. 144t44 Page 56 of 68 Page I 4.2.5 Loss of Shutdown Cooling Loss cJ of ccc ng in a shutdown reactor cOO~l1g reac:or can have halle serious consequences.
ccnsequences. The reactor ri!'actor coolant and RPV RF'V temperatures can use TiSi!'
dramatically dramatioally causing steam production prodlJotion and rising pressure pq-essure at a time timi!'
that containment integrity is not that maintained. OP-li not maintained. OP-17 and AOP.-15 and.';OP-15 provIde alternatives pro>i.de alternatives to the inormaF shutdovm cooling procedure.
"nomlal" shutdown procedure.
DP-17 provIdes direction OP-17 proll:des directicn for mactorvessel temperature in fer controlling reactorvessel in the RHRRH shutdown shutdOl'm cooling mode mode and alternatealternate cooling using usillg the thi!' fuel pool heathi!'at exchangers.
exchallgsrs. It It includes inclwes many precautions and !Emitatiicns many pq-ecauliOIlS rniatcns depending on the specIfic di!'pending mode of spi!'cifrc moc!e operaton to ensure forced oIoperation forcoo circulation through circulation thrcugh the reactor reaC:Or core.
MinImum Miniinum water :evel r wati!'! fevel fur SDC SOC is 2lY200"' to 220 220' or greater for natural circulation circulation duringduring a loss loss ofof SOC.
SOC.
NOTE:
NOTE: Normal level band Normallelli!'1 band is a .appmx.
approx.1S2-182 to 192".
jSD-17 Rev.
Rev. 13 13 Page 53 of 127 Pagesaof 1271
AOP-150 AOP-15.0 ,sis entered when neiltler neither RHR !oop loop can be placed in shutdown cooling and reactor 1>:lOiant coolant temperature is increasing.
inoreas.
Automatic Actions aJe are verifkajion verification of aa Group S 8 isolation signal if loss situldown coo~l1g of shutdown coang was caused by low Low Le'lel Level One or High Steam El l-FOlAB isolates on lOVi Dome Pressure. Ell-F015A{B) Level Cne Low Le',el One on'1'.
oniy.
ERFIS provides a shutdown screen to mDnitor monitor crit<<:al critcal parameters.
A0P-l.D provides comingencies AOP-15.0 contingencies f,or for lhe the fe/owing fcildwing methods methads of decay heat removal:
- RHR SW Loop Failure RHR Loop Failure Coaling Failure Condenser Coo+1ng 3 Iced CombinatiDns Feed and Bleed Combinations
,his aiso pro'l:des provldes oootingel1ties coutingencies for resloringrestoring second,","'Y secondary containment and initial emergency acfoos aofcns for shutdown fcr loss of shutdo"m cooling.
C001,{l1g.
4.3.7 AC Power Distribution lnthvidual MOV power supplies are gi'lel1 Ind:vidual given in Table 17-4.
7-4. A loss lass of the normal source of AC power to the RHR System will not affeot affect ltle the system provided the Diesel tite Emergency Diese! Generators are available. A loss of boltl both power sources is a serious proo:em prcdiern in maintaining long cooling after an accident temi eoolil't'i/
ternl Depending on the exact nature of the failure, the ross Depe!1ding loss of an Emergency Diesel Generator {EOG) (EOG) could affect the system in the ~'le following man.ner.
fellowing manner. See Figure 17-28:
RI4R and RHRSW pumps cou!d RHR inoperative iffif busses E1-E4 could be inoperali'/e El -54 lost.
are lest.
Motor operated valves could become inoperative if Buses RHR MO'!or E&E8 are IQst.
E5-:ES last.
instrumentation on P601 cou:d RHR System inslrumentation could be lost if 120 VAC distribution panels are lOst.
emergency distr:buticn bst.
4.:1.8 4.3.8 Reactor Water Le Level... el A failure of RHR in LPCI mode impacts mpacts the operators abi'ltyablity tD to reflood retlaod the core following a LOCA. A Eo'll low RPV leyel level could indicate an RHR system rupture during S1'O 510 Ccomg.
Cooling. The Group :2 a 2 and S isclations isolations occur at Low level Level One (LL l). #1).
Closure of PC CIQsure P015 IS Group 8 valves FODS. P008, F009.
F000, F015A(8)
F01SA(B) isclates isolates the preventing a significant loss of reactor coolant. A failure RHR system pre'lel1ting to dose coo:ld tQ could result in ltle the loss of coolant ooolant ififsa leak O"..ourred.
occurred.
Inadvertent olosure Inadver-ent closure while in SID CODling Cooling would temperature Vl"ou:d prevent ten-ql)erature the reactor. S!lbsequent control of ltle Subsequent heatup and pressuriuticn pressurization coold could cooling is reestablished, occur if *roc'ing not reestab~shed.
Iso-17 50-17 Rev.
Re .... 13 Page5oft27 Page 51! 0f1:271
TABLE 17-2 TABLE 17-2 Page 77of Page of 10 10 Instrument and Instrument and Control Control Setpoint Setpoint Residual Heat Residual Heat Removal Removal System System l41LMET IN:"TRliME!->T NICATCft I~D1C".TeRl s:FJJMENr1TilF Il'lSTRUIIENT FiNTiCN TRIP FliNCTICN DEl3 OE<:4C ..N"'i"ION COfiDEF
- 'ECORDER T SE'i","OINT TIiJ" E O:tTMJD AND Fll~eTl~
FU4CTYJ
['1\,,101 Re$sLre we1 Prl::eom: E11-PThq43A-El1-PTIM-IJ1P1> ... j or8-1 oe- Eli-PT El1 i-i3i
.. PTM*!->'019 27J1EIg 27Fe19 ..
- C01tiilml:nl Co1irrneiiprav Spray rl antl .. l.BE-l, A-i, A .. l,e"I, C.1 rreaJrg Increasmg PE:m:IS.$l~t-e Pemisete E1-rM-N1SC-l El1-PTM .. I\IJ1£C..11:rO rD-1 ..1 riD-1
.00 0..1 ReecLcr V~1lI Re3der i,esei Le.'!!l Lee: ILLLL: ;;<3) E1-LTS-iG3A-4 B21*t:r*':; ..!->031A-4 or oC-4 0 13:l1 .. LTI,l-1\IJ31 eai-Ln,i-N) Ctirs.
~51Ji~. ..
- IrlU:;les lriHaIeefr Rh" LPCILPDI
- nd ri A-1,E-i.C-i.
A"l.B"l,e .. 1, 7.42 rAl 17A2rrA)
EI-LT34oD..
B2R 1'3-I;03t13-4 or 0..4- s-jD-l an:lD-l eCre3s1; bEcreasiig ..
- ."J:nJnclatiOn ArrtJonI-R RhR
. $ysterfl i AIltuaL-";l system AteI Am. i-rn Fri. .....;)l
!'fl.
- ArnJqcaty, RhR ArnJnclatl:ln I cyslerl II .~ctCa"i!\I Systi:n1 iLalel M'l.
Am.
Frl.
Fri. .....
A-2 ~3 Reaclcf V~
Re;;dCf feeseI Pr:ssure Pressure ..~021A..2 cr i-PT2-4G2lA-2 E:!1-PTS cc B E-2
..2 1321 .. PTI/I*K021 E2i-PTM-iG21 el 412 FEIg 41Q ..
- P:rn:Is&'Ie PenrIs bJ RHfI.
FIHR Purrp urp slcI an:! A-i, A e-i, c-i,
..l. B"l,e-l, (1712 n 117.12 ......}
rIAl sturt tart1Lleanj LPCI *.t.;!ve E2I-PT2-N1c-2 1321-PTS ..!->0210-2 or 0..2 0-2 ncl 0..1 antl 0-1 eCre33l1; b;;crea.'1' Oper:;UCi1\\ten Operahcnat-ei In H cal1,)(J;;rce ouhu)ierce \\t!1 t H~ H 1ll)\\1lI1 rI pr;;:eoJrepreeeire AFlinip0tsge A RHR Fcmp OlEcta'!!l!
FFresSure
...:eoure Eli-El1-PS-I\IJI or llr t421A .... ---- ~ 115FEIg IIFI ..
- Pi5setoAD0 P"m:IS6!voHD IrIllaIirr Ln;tc IflU.nan ADS Locic B E E1i-PS-N02CA El1-P3-NJ2r: ....
rcrealrg Incre..'I5lng ..
- AnoJl1ciJtlOn Mninc1at Care Core 2çray er SFf3Y Cr RHR PUfl~PumpS ftinring .",nn.
RcJnrlng A1. PnL 403 .OJ Tecll peftsicn re(ale3.
IS0-17 1 80- 17 R.:v.
Rev.I3 13 ?.age 1S 0; 121 ge7c127 I TABLE 17-4 Page 2 or9 of 9 Residual Heat Removal System Valves IlZLW, Va,e Nan Name Oze. Tye Sze. Type CllrlI~1 S"ll£n C1ri sltcfl FV.it 24çy FV,RSuF1lIY Cll1t!as CollioC cm1ler CCI1l11:rts F]117A Flllll ... /nlrarnIc RHR !lInlnO) flo.v i-rar0arirG
,-' ..11CI1OrOanlrll Ell-SleA El1 ..S1A 12(Cl l(~)l'.A!Dfl 1l1re=
ThreeFoelltrr p:>SIbon !!olay to: m3!l~;a1y op~.
Maytemariialvoceneo.
~pass Vae eypasS IlZL'le lex Wedg:
flax Wedge Galf: Gale sYcI! (CLOS::-
S,I1I'l:1i CLOSE Vaie "I:n lJa;-." tti SM5-CO 2Ii3-EO i-JJrc-0?!
.. UTO-O?::N) Wit aiYTialtoa1y cp;;n WII.aUlmlalll:aiy cpei IIit now fl3 15 s lESS less ,n;lfi
- nsr 2230!Jlm 2230 fli 31:1 cr.e or avrl ere 4CJalcc
.".Ct'J3tlOf pitg rEwm Splng rah1m 10to Its a5$(daled leted purrps purrpels Is r~nnlng rJnnHg ror tY 10 &:ron~.
$eOOTUs.
'FOO7i5 F2127e EI1-G1E8 E1Ha8 l(2)XB!D!.3 l2iXE0.3 AUfO AUTOpsScn pos:aoo. WllacliX11iiJh::al:y wits CfOO!! r S)~tem II:lN ,.2<00 Icalycosersyelectlt 23O3gm. 9Ffn.
FOlllS F2Z5 LUCCI Local from 11Cm lJa;-.-e; Va1sehasa t,ed1lecl Ila\'e ,al>>'e d11led c!1lllllntaa!!!
Cii Tie lntoar0 dISC dSC 10 pre/Ern Dpl5t lk-5aer Irealer PreG$CT!?
prereus ICC~~ icctcnçj per ESR:s eotis ~0241 -zozii a1l a W-C(;166.
-coie.
Fare F2C lSDCcln2 RHR S!D cooing 2i1" 22 A')i)'"Ior ArTiorOaiTrg Da"lrg 1(2)-A71-CS-fiJi-CS- P4Trmal:
Nonml: ThrOlile val,'e.
1111018ll ale. Opar.mr cperaDrniayopert may CFEfl tireaxpressurels-430t rea@:" pressure IS .::,1::0.8 ps.Jg ard 3M UCrl lso1at SOO'Jon 150l3tiOn flex \'ieCI?- Gale fliOX WecSe Gat.: SIC S10 11,2)-XOB!BEO i2)-XOaEW Three poililon 1l1re= p:>SInon rea@:" level reaotx l;;vellS ~w Leel leLcw Le',llI Cite ene LL(LL ei;;'1 ~
is-CijttoaT IlZL*,r.-C>Jtt~ tWtt lIa1>" \\1:0 SE-3 S13..3 ASS:) Feedt
..ll AOtT2 AL Feed: eetli CLOSE SIIlI'l:li(CLOS::-
,4iatcr
."OJJ3!Cf i2)-1X0A25 1(2)-1XDAi826 NORI,l-OPEt-.')
N0iiM-0PE I\rJli) closes A,flT CIao:s 11 p~ Is It preeexe > 1J~.6 pug IS -ifl pejg Cr or level evellS",
15 LL I. ;;'1. NNo plIngrehrn spingrEwm to 10 o,~ reil'l.Jre 'E il"ta'I.tla Lecela1-vesItorn o%TeajreIsaialatle. LE,'S s:gnal llOl1es /T00i N0il,4 N~~MpaSllon. posTal. r15irurriit t40Taasoc3leJ3 IIl&'rurre"lt NOT assacl3t:ll $01 ,1m FOTP.
F039. P:JIrprrp b t1p ~"11er>>:i<
Looa(
1..0031 ticre treaier iI:OOl Treater remans reI113Jn. 3::7',10 d.l!1rg 100$
Ive.irlrg lOOZL cpera qErK.i01 (cc rer F33t FIJl6.
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ebacln X-12.
FealO F2111 RHRcroS&'l...~
Crosetesade- 2(1" AiTlTr 2 ."'::"Ior OiiJ1r:g liailrg I*U....
i,1A l...Oll"<ejc'OSed Looe.lt eaesl lI.13nu;,1 M3ILal ~
SIl'JIi)Il' utvJor 1'l::< WecSe
-x WEO?- Gale Gam Va5
'lat."
011A ffil1A Thu Heal
,OUt".He./ 4A,2sTr 4' A,:l1:lr 05rlfl 1Je1~ E11-Ca7A Ell ..S37A 1(2tXAJ0F2 1I2IXAOF2 lITE.: pos:ncl1 TlTeepcssllcll Auaalyoueeon Aulamlll:aly CbSe6 00 tPCI LPCllnr.'a':bn siaI. .191.1.
Exdlrger Crairr TxcluW Cf3In b FISa Flex W1e Weelge GateGate asic, SWn::t1 (CLOSE-T0li5 TOOJ& VaI,'e'"lIl1 ialdeAIUi ....UID-OPENj ATO-CPENI
"""fili13 10115 Tie-2:Cfldxuatr SIIB-(JQO AdJJalllr E11-S-3Th Ell ..S375 1(2)XB,OLIS il21X.OU5 Cptrgreixnro
~n~9 reIJ.rn to
....uro AlTO CSiG1 pasl!o:l
FIGURE 17-9A FIGURE 17-9A RHR Pump RHR Pump Start/Stop StartiStop Logic Logic "Au A
~
[l-i ,- ~i2 r;'{ ~~
I
- I I I
I 1l'>~i§1 ! ...;f; ,-f\ :i , ..... I~ . ~",I
" '" A'
.J SD-17 1 8D-17 I Rev 13 Rev. Page1O7of127 Page 107 of 1271
ATTACHMENT 33 ATTACHMENT Page6ofl8 Page 6 of 18 480V Substation 480V Substation E7JMCC/Panei E7IMCC1PaneI Load LoaU Summary Summary 480V Motor Contro Center 2-2XA Load: 480V Motor Control Center load: 2-2XA Location Reaclor location: Reactor Building Building 20' 20 NENE OrawThg
Reference:
Drawing
Reference:
F-03049 F-0309 Upstream Power upstream Pcver Source:
Source: 4BOV 480V Substation Substation E7 E7 COMPT COMPT LOAD DESCRIPTION LOAD DESCRIPTION EFFECTS ON EFFECTS ON lOSS LOSS OF OF POWER POWER DOS OG5 RHR Heat RHR Heat Exchanger Exchanger 2A 2A Ser/ice Service Water Water Loss of Loss of load load Discharge Valve Discharge Valve El1-POV-FOtl3A El 1-PDV-F063A (TS 3.4.7, 3.4.B, iTS 3.4.7, 3.6.2.3, 3.7.1) 3.4.8. 3.tl.2.3, 3.7.1)
DID DID Conventional Header Conventional Header To To Vital Vital Header Header Loss of Loss load of load Isolation Valve Isolation Valve SW'-Vl11 SW-Vu i (TS ITS 3.6.2.3) 3.8.2.3)
DE2 DE2 Normal Feed to HPCI Turbine N()rmal Tuthire Exhaust Less of load Loss Vacuum Breaker Va.cuum Breaker Valve E41-F075 E4i -F075 (TS3.5,l, (TS 3.6.1.3. 3.3.3.1) 3.5.1,3.6.1.3. 3,3.3.1)
D63 DE3 Service Water Header To RBCCW Heat Set'Jice Loss of load Frimarg Isol Exchangers Primary lsct '.IN V SW-Vl0tl SW-V106 (TS 3.7.2 and TRM 3.16)
ITS 3.7..2 3.18)
Normal Feed to RCIC Turbine Exhaust DE4 Loss of load Vacuum Breaker Valve valve E51-F062 251-FU82
{IS 3.8.1.3. 3.3.3.1}
(TS 3.5.3, 3.6.1.3. 3.2.3.1) 007 OG7 Exchanger Vent Valve RHR Heat Exchanger Loss of load El i-Fl 04A El'l-Fl04A 008 DG8 RKR Pump 2A and 2C Torus Suction RHR Such err Loss of load Valve El1-F020A El 1-FO2OA (TS (IS 3.4.7, 3.4.8, 3.5.1.
3.6.1, 3.5,2, 3.6.1.3, 3.8.2.3, 3.3.3.1) 3.5.2,3.6.1.3.3.8.2.3,3.3.3.1}
OH 1 DH1 Recirc Pump 2A Purge Seal Reactor Recire Loss of load Injection Valve B32-'./22 Injection B32-V22(TS 2.4.1. 3.tl.1.3, (TS 3.4.1. 3.6.1.3.
3.3.2.1) 3.3.3.1)
DH3 Peed To RHR Shutdown Cooling Normal Feed Loss of load Inboard Isolation lsolatien Valve El 61 l-FC0 t-FOO9 (TS 3.4.7.
3.4.7, 3.4.8. 3.6.1.3, 3.3.3.1) 3.4.B, 3.tl.l.3, DP6 OF6 0r,wll Spray Inboard Ol)Well Spray Inboa.rd lscladcn
!solation Valve Loss of load E1l-FO21A (TS Etl-F02*1A 3.8.1.3, 3.3.3.1)
ITS 3.6.1.3, DEl Service SeNice Water Vital Header Croestie Crosstie Loss of load VaWe Va.l'le SW-V118 S'W-V118 I OF7 DF7 RHR RHR Torus Torus Cooling Isolation E1l-F024A Ell-F024A (TS Isolation Valve
{IS 3.8.1.3, 3.6.1.3, 3.&2.3, Valve 3.6.2.3, 2,3.3.1) 3.3.3.1}
Less of Loss of load load 001-50.3 1001-50.3 Rev.
Rev. 36 36 Page 15 oJ Page 15 of 521 Categories KJA:
KIA: 295021 A2.02 295021 A2.02 Tier/Group:
Tier / Group: TIG1 T1G1 RO Rating:
RORating: 3.4 3.4 SRO Rating:
SRORating: 3.4 3.4 LP Obj:
LPObj: CLSLP302L*0lA CLS-LP-302-L *0 lA Source:
Source: NEW NEW Cog Cog Level:
Level: HIGH HIGH Category Category 8: 8: Y Y
- 50. spent fuel
- 50. AA spent fuel bundle bundle has has been dropped on been dropped on the the refuel refuel floor.
floor. The The following following alarms alarms are are received:
received:
PROCESS RX PROCESS R)( BLDG BLDG VENT VENT RADRAD HIGH HIGH PROCESS RX PROCESS RX BLDG BLDG VENT VENT RADRAD HI-HI HI-HI SCCP directs SCCP directs the the operator operator to to isolate isolate Reactor Building HVAC Reactor Building HVAC and and initiate initiate SBGT.
SBGT.
Which one Which one of of the the following is reason for this is the reason this action?
action?
maintain the Reactor To maintain Reactor Building Building pressure pressure (1)
(1) with respect respect to atmosphere and provide a(an) (2) provide release point.
(2) release (1) positive A. (1) positive (2) elevated B (1) negative B:"
(2) elevated C. (1) positive (2) ground level D. (1) negative (2) ground level
Feedback K/A: 295023 K3.03 KIA:
Knowledge of Knowledge of the reasons reasons for the following responses responses as they apply apply to REFUELING REFUELING ACCIDENTS:
isolation Ventilation isolation (CFR: 41.5 145.6)
(CFR: /45.6)
ROISRO Rating: 3.3/3.6 RO/SRO Objective: CLS-LP-1 09-A09A*01
- 01 d
- 1. Identify the following as related to a Refueling Accident:
- b. Analyzed plant response.
- d. Plant design features that mitigate the consequences of the accident.
Reference:
001-37.9 Cog Level: Low Explanation:
From 01-37.9 If the reactor building ventilation exhaust radiation level is above 4 mR/hr, then the Reactor Building HVAC should have automatically isolated. This step ensures that a required automatic function has initiated. Confirming isolation of Reactor Building HVAC subsequent to receipt of a high radiation signal or a high temperature condition terminates any further release of radioactivity to the environment from this system.
SBGT is the normal mechanism employed under post transient conditions to maintain reactor building pressure negative with respect to the atmosphere since the exhaust from this system is processed and directed to an elevated release point before being discharged to the environment.
This question requires the operator to have knowledge of the reason for ventilation isolation as related to a refueling accident therefore matches the referenced KA statement. This KA statement is cross referenced to CFR 41.5 Distractor Analysis:
Choice A: Plausible because since reactor building is maintained at a negative pressure Choice B: Correct Answer Choice C: Plausible because since reactor building is maintained at a negative pressure and SBGT discharges to the main stack which is elevated Choice D: Plausible because since SBGT discharges to the main stack which is elevated SRO Only Basis: N/A Notes
AA high reactor building high reactor differenf at pressure build ng differential pressure is s lndicati:
ndicaflve
... e of of aa potential potential loss loss ofo reacior reactor building struclura.!
building niegiiiy and struciural in~egri.iy and could could result resuI inin unconirolled uncomrolled release release of of radioacii radioaciivhy....iiy to to the the enironrrient. Arm;un*cialof em'ironment. Ann unciaor procedure procedure UA-05UA-O 6-7 -7 deals deals with with loss loss ofof negative negative pressure pressure and descr.ibes a.nd describes .actionsactions to* remedy the to remed~1 the siu.Jaiion..
situation. IfIf ihethe eevent
....ent isis not not caused caused by by aa malfunction of malfunction of the the Reactor Reactor Building Building HVAC HVAC System.
System, the the annunciator annunciator procedures procedures will will require entry require entry into*
into the the SCCP..
3CC P. IfIf an an H"l.AC HVAC malfunction malfunction hascaused1he has caused the problem problem and arid negatiie pressure negafi'o'e pressure cannot cannot be be maintained maintained b~' by ma.nipurating manipulating the the Heacler eaotor Buildirlg Building HVAC. HbC Sysiem or siarting S3GT, then the Sysiemor siarting S8GT. then 1he annunciator proceduresannunciator procedures 'hill require en~ry iiil requ.ire en:ry into into the the SDCP. This SCCP. This *,will preclude entry
..~II preclude entry into this procedure nto 11115 procedure for for problems problems wiih with Reacler eactor Building Building HVAC 'h'hich HVAC tich can can be be immediately immediately oolTecied correc:ed by by opera1or opera:or action action Of or initiation initiation *of of SBOT.
SB3T.
High reaC10f High reaco buildingbuilthng ventilla~ion entiIa1ion exhaust exhaust radiation radiation may may indicate indicate that ra&oactvity is that !'"adioact:lVity is being released to the environmen~. erwironnient whenwhen ~hethe s~'stel11 system should ha\l12 have automatica!l;y automatically isolated.
THe PROC The PROCESS ESS AX X BLDG VENIT VENT RAiDRAD HI annunciator pifocedure procedure (UAAJ3 LkA-33 4-5) will direct the ope!'"afo!'"
operator 40 tO 1he the SCCP since ~he annunciator seipoint is the annunolator5e4point is ~he the same atS as ~he the EOP ECIP enbry entry condition.
area ,radiation An alea radiation ~e level i maximum normalopen'l~ing
....el above its normal operaiing le"'el IeeI is an indication that water from a primarJ' primary system (or br *from secondary s~lsfem from a primary to secondary system leak) leake may may be discharging into the Reacto!'"
dilscharging Reactor 8UlYJding.
3uding. The The AREA RAn RAD AX X BLDG HIGH a.nnunciatorannunciator procedure {UA-03 IUA-D3 2-7} provides :for 2-7) pro.vides for entry into nto ihe the SCCP 'I.'hen wtien entry condition lev'els levels fo!'"for any areas are exceeded.
3.ny A H PCt A. PCI, RHR RHR, or Core Spray Spray room water le level
..'e:'1 above ilsits maximum normalo,pe.raling normal operating le'l'el ,:6 leeI 6 mnches) nches) is an indication that sieam seani or orwaer dscharging into ~he waier may be discharging: the Reactor Building. The annunciator procedures for HPCR, B,uildi!17lg.. HZC[, RHR" RHR, and Core Spray rooms FLOOD LEVEL HI and HI-HI {UA-12} (UA-12: will direct the operator to the SCC SCCP. P. The H HII level level annuncia:or sepoints annunCiator se~points correspond to the e&y entry conditions for this ih1S EOP.
EOP.
COl-37.
100:1-37 .. 9 Re ...... 11 Rev. Page 10 of 2
YES yes SCCP-9 SCCP-10 STEP BASES:
1 35 F is not available, the operator is directed to use Since instrumentation to read the -135=
135 F).
the annunciator (setpoint of 13S~
If the reactor building ventilation exhaust radiation level is above 4 mRfhr, niR!hr, then the Buiding HVAC should hal/e Reactor Building have automatically isolated. This step ensures that a required automatic function has initiated. Confirming isolation of or Reactor Building 8uilding HVAC subsequent to receipt of a high radiation signal or a high temperature condition HV.A.C terminates any further release of radioactivity to the environment from this system.
SBGT S8GT is the normal mechanism employed under post transient conditions to maintain reactor building pressure negative with respect to the atmosphere since the exhaust from this system is processed and directed to an elevated release point before being discharged to the environment.
environment 001-37.9 1001-37.9 Rev. I1 Page 1313 of 391 of 39
A or A or B - Isolates the B - Isolates the remaining Group 66 isoiation remaining Group isolation valves valves notnot listed listed under A under A or B above.
or B above.
4.3.12 4.3.12 Refueling Operations Refueling Operations and Accidents and Accidents Refueling operations Refueling operations require require Reactor Reactor Building Building HVAC HVAC to to be be in in operation to operation maintain aa dean to maintain clean and and relatively relatively dry dry atmosphere atmosphere on on the the refuel floor.
refuel floor. Failure Failure of of the the RB HVAC System RB HVAC System w'Ould ould cause cause the refuel refuel floor humidity floor humidity and temperature temperature to to rise. The The loss loss of air flow across the the top of the pools 'NOuld would allow contamination from evaporation and diffusion of gases to diffusion to contaminate the the refuel floor.
refueling accident A refueling accident of of sufficient size will cause the sufficient size the Reactor Reactor Building Building Exhaust Rad Exhaust Rad Monitors Monitocs to increase increase...4.t At 4 mrJhrthe mrihr the RBRB I*WAC HVAC System isolate and SBGT will start. Refuel will isolate Refuel Floor Floor monitors monitors have have nono isolation input to RB HV HVAC.
AC.
4.3.13 Primary Containment The Primary Containment and RB HVAC interface interface through the SBGT SBGT System and tile containment Purge subsystem. The only the Primary Containment HVAC tl1at failures of RB HV.4.C that would affect the Primary Containment 'ould would be mechanical failures of the duct lNOrk. work. Depending on the location the failure, of tile failure. it would either prevent tile the Containment Purge System from functioning or exhaust tile the containment into the Reactor Building.
ISD-37.1 Rev. 10 Page 40 at of 70 I
4.2.2 4.2.2 Secondary Containment Secondary Containment Isolation Isolation Mode Mode The secondary The secondary containment containment isolation isolation mode normally initiated mode isis normally initiated automatically as automatically as aa result result of ofthe SBGT System the SBGT System receiving receMng an an automatic automatic start signal.
start signal.
When this When this occurs occurs the the supply supply and and exhaust exhaust air air fans fails are are tripped, tripped, the the Reactor Building Reactor Building ventilation ventilation isolation isolation dampers dampers close, close, the the purge purge exhaust fans exhaust trip and fans trip and meir their dampers dampers closeclose ifif in in operation.
operation. Both Both SBGT SBGT trains '.viii trains 1ll start start and continue to and continue maintain the to maintain the negative negative pressure pressure inside inside the Reactor the Reactor Building.
Building.
A failure to A failure to isolate isolate when when required required could could lead lead to to aa ground ground level level release release radioactivity, of radioactivity.
of 4.2.3 4.2.3 Abnormal Operating Abnonnal Operating Conditions Conditions While not VVhile not abnormal abnormal operating operating relationships, relationships, conditions conditions maymay occur occur periodically that may periodically result in may result in abnormal operating conditions.
conditions. SomeSome are discussed below.
are discussed below.
1.
- 1. Extremely low Extremely Low Outside Temperatures Extremely 10..,.low outside temperature can cause problems wim with components located wimin within me the Reactor Building as mere there is is no available for me heating system availal)le the Reactor Building. Reactor Building temperature can lJe be controlled to an extent by varying me the flow rate of the supply and exhaust exhaust fans or limiting me the number of supply and exhaust fans running. The requirements for maintenance of a negative pressure in the Reactor Building must be maintained as required by operational conditions. If temperatures in the Reactor Building drop below 40°F, 4OF, Conduct of Ops Manual, 001-1.02, directs contacting engineering to evaluate equipment operability.
IS0-37.1 SD-37.1 I Rev. 10 I Page 33 of 70 I
Unit Unit22 APP UA-03 3-5 APP UA*Q3 3-5 Page 11 of*1 Page of I PRO CESS RX PROCESS RX BLDG BLDG \lENT VENT RADRAD HI-HI HI-HI AUTO ACTIONS AUTO ACTIONS 1.1. Reactor Building Reactor Buildiog ventilation ventilation system system trips hips and and isolates.
isolates.
2.
- 2. Standby gas Standby gas trealment treatment trains trains start start 3.
- 3. IfIf open, open, the inboard and the inboard outioard primary and OllIDoard prinlary containment containriient purge purge and and vent vent valves valves close.
close.
4.
close.
CAUSE CAUSE 1.
.1. High airborne High airborne activiDjlevel$
activity levels inin the the Reactor Reactor Building Building exhaust exhaust plenum.
plenum.
2.
- 2. Circuit malfunction.
Circuit malfunction.
OBSERVATIONS OBSERVATIONS 1.
- 1. Reactor Building Reactor Building exhaust exhaust plenum plenum rad red monitor monitor indicates indicates greater than 44 mRlhr greater than mRihr onon Panel H12*P60S.
Panel H12-P806.
- 2. PROCESS RX BLDG 'VENT VENT RAD HIGH (UA-03 4*S) 4-5) alarm.
ACTIONS I.
- 1. Verify auto actions.
2,
- 2. Refer to AOP-05.0, AOP-05.O, Radioactive Spills, SpiIs, High Radiation,and Radiation, and .4.irborne Airborne .A.ctivity.
ActMty.
- 3. EOP-03-SCCP, Seconda!"'j Enter EOP-03-SCCP, Secondary Containment Containnient Control.
- 4. EOP-24-RRCP, Radiological Release Control; enter as appropriate.
Refer to EOP-04-RRCP, 5.
S. If a circuit malfunction is suspected, sueiJected, ensure that a Trouble Tag is prepared.
DEVJCEISETPOINTS DEv'lCElSETPOINTS Rad Monitor 012*RM*K609.4JB D12-RM-K&19A18 4 niRlhr mRlhr POSSIBLE POSSIBLE PLANT EFFECTS t
'1. Possible Possible release to to enirons environs in in excess excess ofof 00CM ODCM 7.3.7.
7.3.7.
2.
- 2. This This annunciator annunciator is is required required to to be be operable operable to to support support Rx Rx Brdg B!dg Vent Vent RadRad Monitor Monitor operability; operability; annunciator annunciator inoperability inoperabilit}' will will result in in a LCO.
LCO.
REFERENC REFERENCES ES I.
- 1. LL-93S3-35 ll-9353 - 35 2.
- 2. AOP-05.G, AOP-05.0, Radioactive Radioactive Spifls, Spills, High High Radiation, Radiation, andand Airborne A.irbome Activity
.A.ctivity 3.
- 3. EDP-03.-SC EOP*03-SCCP CP 4.
- 4. EOP-fl4-RR EOP-04-RRCP CP 5.
- 5. Technical Technical Specificaons Specifications 3.3.61.
3.3.6.1, Table Table 3.3.8.1-1 3.3.6.1-1 Function Function 2d 2d and and 3.3.8.2, 3.3.6.2, Table Table 3.3.6.2-i 3.3.6.2-1 Function Function 33 6.
- 6. 00CM ODeM 7.3.7 7.3.7 2APP-UA-03 12APP-UA-03 Rev.
Rev. 4646 Pag~ 32 Pane of 33 32 oT 631
Unit Unt2 2 APP APP UA-03 UA-J3 4-5 4-5 Page 11 of Page of "1I PROCESS RX PROCESS RX BLDG8LDG \lENT VENT RADHIGH RAD HIGH AUTO ACTIONS AUTO ACTIONS NONE NONE CAUSE CAUSE I.
- 1. High airborne High airborne activity activity in in Reactor Reactor Building Building ventilation ventilation exhaust exhaust plenum.
plenum.
2.
- 2. Circuit niafunction.
Circuit malfunction.
OBSERVAT OBSERVA IONS nONS 1.
- 1. Reactor Building Reactor Building Vent Rad Rad Recorder Recorder D"12-RR-R605 D12-RR-R605 Channel Channel A A or or BB indicates highhigh radiation level.
rooiation
- 2. Reactor Building Exhaust Plenum Reactor P[enuni Rad Rad Monitor Monitor Channel A or B indicates greater mRihr on Panel H12-P606.
than 3 mRfhr HI 2-P606.
ACTIONS ACTIONS
- 1. Enter EOP-03-SCCP, Eneer EOP-03-SCCP, Secondary Containment Containnent Control.
- 2. Refer to ADP-G5.O, Radioactive Spills, AOP-OS.O, Radioacti'.*e Splils, High High Radiation, and .4.iroorne AiFoorne Activity.
3.
- 3. circuit malfunction is Slmpected, If a circuit suspected, ensure that a Trouble Tag is prepared.
DEVICEISE TPO INTS DE\~1CE/SETPO Di 2-RR-R605 red or black pen D12-RR-R605 pen niR?hr 33mRlhr POSSIBLE POSSIBLE PLANT PLANT EFFECTSEFfECTS I.
- 1. Possible Possible release to environs.
environs.
2..
- 2. airborne activity IfIf airborne activity increases increases to to 44 mRhr, mRlhr, Reactor Reactor Building Building HVAC H\lAC isdation, isolation, aa Group GfOUp 66 isolation.
isolation, drjweil purge isolation, drywell purge isolation, and and initiation the Standby of the initiation of Standby Gas Gas Treatment Treatment System System vAIl "hill occur.
occur.
REFERENC REFERENCES ES I.1. LL-9353-35 LL-9353 - 35 2.
- 33. EOP-03-SCC EOP-03-SCCP P
- 44. Plant Plant Modification Modification 85-081 85-081 2APP-UA-03 12APP-UA-03 Rev. 46 Rev. 46 Page 41 Page 41 of of 63 631
Categories Categories KIA:
KIA: 295023 K3.03 295023 K3.03 Tier // Group:
Tier Group: TIG1 TIGl RO Rating:
RORating: 3.3 3.3 SRO Rating:
LP Obj: CLSLP1O9A*01D CLS-LP-I09-A *OlD Source:
Source: BANK BANK Cog Level:
Cog Level: LOW LOW Category 8:8:
Category YY
- 51. Following aa loss
- 51. Following loss ofof feedwater, feedwater, HPCI initiated on HPCI initiated on low low reactor reactor water water level level then then tripped tripped on on high reactor high water level.
reactor water level.
Current plant Current plant conditions conditions are:
are:
Reactor water Reactor water level level 180 180 inches, inches, steady steady HPCI TURB HPCI TURB TRIPTRIP alarm is alarm is sealed inin HPCI TURB TURB TRIPTRIP SOL SQL ENER ENER alarm is alarm is sealed inin Initiation Signal/Reset seal in white light HPCI Initiation HPCI light is LIT LIT HPCI High Water Level Signal Signal Reset white light light is LIT LIT Which one of the following identifies the impact on the HPCI System System if drywell pressure rises to 3.0 psig with the above conditions present?
HPCI will initiate HPCI initiate and inject to the reactor:
reactor:
A. with no operator action.
B. only if the injection valve is manually opened.
C only if the High Water Level Signal Reset push button is depressed.
C;I D. only if the injection valve is manually opened after the High Water Level Signal Reset push button is depressed.
Feedback Feedback K/A: 295024G KIA: 295024G 2.02.37 2.02.37 Ability to Ability to determine determine operability operability and/or andlor availability availability of of safety safety related related equipment.
equipment.
High Drywell High Drywell Pressure Pressure (CFR: 41.7/43.5/45.12)
(CFR: 41.7 /43.5 /45.12)
ROISRO Rating:
RO/SRO Rating: 3.6/4.6 3.6/4.6 Objective: LOI-CLS-LP-01 9A*03M Objective: LOI-CLS-LP-019-A *03M
Reference:
Reference:
SD-19, Revision SO-19, Revision 16, 16, Page Page 33, 33, Section Section 3.5 3.5 Cog Level:
Cog High Level: High Explanation:
Explanation:
A Reactor High A Reactor High Water Water Level Level trip trip is is initiated initiated and the signal and the signal seals-in seals-in when when aa high high level level is is sensed sensed by two by two instruments. Once instruments. Once the high level the high condition clears, level condition clears, the the trip is is reset reset by by aa subsequent subsequent Reactor Reactor Low Low Level Level 22 upon depressing signal or upon depressing the PanelPanel P601 P601 High High Level Level Trip Trip Reset Reset pushbutton.
pushbutton. AA High High Orywell Drywell Pressure Pressure Initiation signal Initiation signal will will not not reset reset the High High Water Water Level Level trip.
Distractor Analysis:
Oistractor Choice A: Plausible because the high water level trip does automatically reset on LL2. If high water level is reset with an initiation signal (Hi OW DW Press), the system automatically aligns for injection without operator actions.
Choice B: Plausible because the high water level trip does automatically reset on LL2. Injection valve requires active initiation condition + Stop Valve (Va) (V8) && Steam Supply Valve (F001) (FOOl) not full closed to automatically open. Relay timing has caused HPCI initiation with injection valve not opening (LER 2-90-015).
Choice C: Correct Answer Choice D: 0: Plausible because high water level does not automatically reset due to Hi DW OW Press. Injection valve requires active initiation condition + Stop Valve (Va) (V8) & Steam Supply Valve (F001)(FOOl) not full closed to automatically open. Relay timing has caused HPCI initiation with injection valve not opening (LER 2-90-015) 2-90-01 5)
SRO Only Basis: N/A Notes Notes
3.2 3.2 HPCI System HPCI System Automaticlnitiatiol1 Automatic Initiation Control Control (Figure (Figure 19-6) 19-6)
The HPCI The HPCI System System isis automatically automaticallyinitiated initiated inin response response to to aa Reactor Reactor Low Low level Level Twoor Two High Drywell or aa High Drywell Pressure Pressure Signal, signal, asas shown showii below.
below. There There are are four four trip trip units for units each parameter for each parameter sensed, sensed, with with tv.'O two of of the trip units the trip units from from ECCS ECCS Logic Logic Division II and Division and two two from from ECCS ECCS logic Logic Division Division II.II. The The four four trip trip units units forfor each each parameter are parameter are arranged arranged inin aa one-out-of-1:\'Io-taken-twice one-out-of-two-taken-twice logic. logic.
Table 19-5 Table 19 HPCllnitiation
- HPCI Initiation Setpoints Setpoints Signal Signal I Setpoint*
Setpnr ITech Tech Spec" Spec Reactor low Reactor Level Two Low Level T 185 105" ;::101" 101 High Dryv.oell High Dryll Pressure Pressure t7 psig t7 psig 1a psig
- 1.B psig All ~lpoinls
.. All seoints shown Ehown in this SO in this SD are are '.he nominal values the nominal values of o proce.ss orocess ata: instrument nsjment actuation. Seboins are acuation. Selpoinls are as as dentied in identified in the applicable APPs the applicable APPs and the EOP and the EOF Users Users Guide.
Gude. Tolerance Tol&aree and and scaling scaling (including cncludng head head correction) ccrrectcn niomialion is information s available avaialde from from EDBS EDGS or calibraon procedures.
or calibration procedures "'Vhen When aa Tech Tech Spec Spec value value is isted. this is listed, is this is actually the Technical aotually the Tethnical Speci.iicalion Specification Allowable Allowable Value Value information.
information.
I SD-19 SD-19 Rev 16 Page 18 18 of 10B I Valve E41-F006 Valve E41 -P006 will automatically open on a HPCI System initiation signal if both the Turbine Stop Valve, Signal Valve. E41-VB, E41-V, and the Turbine Steam Supply Valve, E41-F001, are not fully closed. This valve will automatically dose close if either the Tul1>ine Turbine Stop Valve or the Turbine Steam Supply Valve is fully closed. HPCllnjection HPCI Injection Valve,Valve. E41-F006 may also be opened or closed from the Control Room using its control switch on IJe however once the valve reaches its open or closed limit Panel P601; however, switch, the valve will respond to the approprtate appropriate automatic open or close signals discussed above.
The configuration of the automatic opening circuitry circuitiy for E41-F006 can lead to a condition where the HPCI System automatically initiates but the HPCI Injection Valve does not open HPCllnjection open. 8ectrical Bectrical and hydraulic transients have occurred in the plant wlch which were sufficient sumcient enough to generate aa nornentary Reactor Low momentary low Level 2 signal. This signal automatically initiated the HPCI System bu bu~ prior to the tile Turbine Stop Valve and the Turbine Steam Supply Supply Valve Varve opening to permit pemlit E41-F006 to automatically open, the Reactor Low Low Level 2 condition cleared. This resulted in in the HPCI HPC I Turbine running on minimum minimum flow with with the injection injection valve closed. (Refer (Refer to to LER LER 2-90-015).
2-90-015).
[SD-19 1 80-19 Rev16 Page22of10.
Page 22 of 10BJ I
3.5 3.5 HPCI Turbine HPCI Turbine Trip Control (Figures Trip Control (Figures 19-2519-25 and and 26) 26)
The HPCI The HPCI turiJine turbine will will automatically automatically shutdO'.vn shutdown (Tumine (Turbine Stop Stop Valve Valve closes) upon closes upon receipt of receipt of one one ofthe of the signals signals listed listed inin Table Table 19~7, 19-7, below.
below.
Table 19-7 Table 19 HPCI
- Trips HPCI Trips Signal Signal ISetpoint ITech Tech Spec Spec Turbine Overspeed Turbine Overspeed 4600 rpm 46GO rpm :1:150 rpm(1 10%
+/-150 rpm{110% NJA N/A of original of original rated rated speed speed - -
4000 rpm) rpm)
Reactor High Reactor High Water LevelLevel 206 206" ~207" 207 HPCI Pump HPCI Pump lowLow Suction Suction 15 inches 15 inches after after 13 13 sec.
sec. N/A Pressure Pressure time delay time delay E>haust Turbine High Exhaust 157.5 psig N/A Pressure HPCI System Isolation See Table 19-8 See Table 19-8 1 9-8 Manual Trip Manual N/A N/A N/A N/A Low Steam Line low Line Pressure" Pressure 115 psig 115 :<::104psig 104 psig SD-19 1 SO-19 Rev. 16 Page 31 of 108 1 There is no seal in of the turbine trips, except for the trips caused by Reactor High Water Level and isolation, thus the operator must be he alert to conditions turbine tripping and then resetting when the trip condition that could lead to the turl)ine clears. This could be repeated until equipment damage occurs. occurs A Turbine Trip pushbutton, on Panel P601, A P001, wiuich Which also energizes the remoteoperated solenoid oil dump valve (E41-C002-SV1) remote-operated (E41-C002-SVI) as discussed above, above.
may be used to shut down the turbine turbine.
A Reactor High Water Level trip is initiated and the signal seals-in when a high by two instruments, l)Qth level is sensed IJytwo both powered from 125 VDC Bus. Once the high level condition clears, the trip is reset by a subsequent Reactor Low Level 2 signal or upon depressing the Panel P601 High Level Top Reset level Trip push button A pushbutton. A High Drywell Orywell Pressure Initiation signal will not reset the High Water Level trip. -
During a HPCI HPCI Turbine TUri)ine start, pump suction pressure could possibly drop drop below the trip initiation setpoint. For this reason, aa 13 the 13 second second time delay has been been added added to prevent spurious trips upon system system friation initiation.
I1SD-19 80 -19 I Rev. 16 16 I Page 33 of 1081 Page33of 1081
Unit rinit 22 APP APP A-Ol A-fl 3-1 3-1 PagE; Page 11 of of 44 HPCI HPCI TURB TURn TRIP TRIP AWID ACTICt:rS AUTO ACTICtIS 1.
- 1. If the If npcr turb1nE; the HPCI turbine tr1ps, trips, thE; the following following occurs:
occurs:
a.
- a. If Qpen, If open, thE; the Turb1nE; Turbine StopStop ValvE;,
Valve, E41-\,9, 541-V9, trips trips closE;d.
closed.
b.
- b. If If Qpen, open, thE; the 8PCI WcI InjE;ction Injection valvE;,
Valve, E:41-F006, 241-2006, clOSE;S closes..
c.
.~. If Qpen, If open, thE; Minimum Flow the ~!1n1mum Flow 8ypass mypass To To Torus Torus valvE;,
Valve, E:41-F012,241-2012, clQses.
closes.
2.
- 2. If If the the HPCI EPcI turbinE; turbine 1so1atE;s, isolates, thE;the following following occurs:
occurs:
a.
- a. If ':Jpen, If open, thE; Steam supply the StE;am Supply Inboard Inboard Isolatic.n Isolation val Valve,....E;, E:41-F002, 241-2002, clQses.
closes.
t..
- b. If open,.. thE; If Qpen the StE;am steam Supply Supply outboard Outboard Isolation Isolation val Valve, ....E;,
541-FOOl, closE;s.
E41-FOOl, closes.
c.
- c. If Qpen, If open, thE; the TurbinE; Turbine Stopstop ValvE;,
Valve, E41-'19, 541-Va, closE;s.
closes.
d.
- d. open, thE; If ':lpen, the 8PCI HPCX InjE;ction Injection ValvE;,
Valve, E:41-F006, 241-2006, closE;s.
closes.
- e. open, thE; If Qpen, Minimum Flow 8ypass the ~!1n1mum mvpass To Torus val ....E;, E:41-F012, Valve, 241-2012, closes.
clQses.
- f. If Qpen, open, thE; the Torus Suction valvE;, Valve, E:41-F041, 241-2041, closE;s.
closes.
- g. open, thE; If Qpen, the Torus Suct10n Suction valvE;,
Valve, E:41-F042, 241-2042, clc*sE;s.
closes.
cAUSE CAUSE
- 1. nigh reactQr High reactor vessE;l vessel watE;r water lE;vE;l level (206 (206 incnE;s).
inches).
- 2. Mechanical QverspE;E;d overspeed tr1ptrip (4600 rpm) .
- 3. nigh High turbine exhaust prE;ssurE; pressure i157.5 (iS7.5 ps1gi.
psig) 4.
- 4. tow HPCI pump suction prE;ssurE; t.QW pressure i15 (15 1nchE;s inches Elg Etg vaC'.1um vacuum aftE;rafter 13 secQnd time delav) second delay;..
S.
- 5. High turbine exhaust d1aphragmdiaphragm prE;ssurE; pressure (7 ps1gi.
psig).
- 6. High steam line diffE;rE;nt1al differential prE;ssurE;.
pressure.
7.
- 1. Low steam supply prE;ssurE; t.QW pressure illS (115 ps1g).
psig) a.
B. nacz rQom nigh HPCI High room arE;a area ambiE;nt ambient tE;mpE;raturE; temperature (165'F).
t1652)
- 9. High ncr ste.am nigh HPCI steam linE; line arE;a area amb1E;nt ambient tE;mpE;raturE; temperature (190'Fi.
(120?).
10.
le.. HPcI steam linE; High HPCI line tunnE;l tunnel tE;mpE;raturE; temperature (165'F).
(165?).
Li.
- 11. ncr steam linE; nigh HPCI High line arE;a area d1ffE;rE;ntial differential tE;mpE;raturE; temperature (47'F).(47?;.
- 12. Turbine trip push button.
TUrbine n.
- 13. circuit malfunct1on.
Circuit malfunction.
OBSERvATICtt OBSERVATICNS ROTSi NOTE. Once the turbine tripe, trips, exhaust pressure and suction prE;ssurE; pressure w111 will return to zero or a ~oeitivepositive value.
1.
- 1. Reactor ReactQr vessel water watE;r level lE;vE;l greater grE;atE;r than 206 inches 1ncnE;s (multiple (m'.11t1plE; RTG2 RTG8 1nd ic.at iQns) .
indications).
2.
- 2. TUrbine speed.
Turbine speed.
a.
- l. mrbtne TUrbine exhaust pressureprE;ssurE; greater grE;atE;r than 157.5 157.5 psig ps1g (241-PI-P.6031.
(E:41-PI-R603i.
2APP-A-O1 12APP-A-01 Rev. 55 Page 41 of of114 114 I
1.
- 1. in an tf in If an .accident a:ident status, status, utll1Z~ the RCIC utilize th~ FoD1C syet~m system p~rper 01'-16 OP-16 to to astntain reactor maintain reator v~ss~l vessel l~v~l.
level.
- 1. If the If the reactor reactor v~ss~l vessel "llfat~r level drops water l~v~l drops back back toto 11)5 inches and 1O inctl~e, and it it is desired is desired toto r~comm~nc~ pci inj~ction, recommence 91'CI injection, p~rform the following perform th~ following steps steps:
a.
- a. eset the Reset the high high "llfat~r water l~v~l level shutdo'ilIl shutdown f~atur~
feature byby d~pr~eeing depressing the High the 3]igh Wat~r Water L~v~l Level signal Reset Pueh signal R~s~t Push Button, gutton, E41-S2S.
g$1-g25.
b.
- b. Verify that Verify that th~
the Turbin~
Turbine stop Valve, E41-V9, top valv~, reopens and
.t1-V, r~op~ne and that that Gysteni r~starts EPCI syst~m HPCI restarts p~rper 01'-19.
OP-l.
L
- l. If the If reactor v~ss~l the reactor vessel "llfat~r level 1s water l~v~l stable and is stabl~ the RPCI and th~ system ie RC! syet~m is no longer no longer requ1r~d required for operation, shut for op~ration, shut down the R1'CI down th~ RPCX Syet~m System p~rper p- ig.
12APP-A-01 L2APP-A-Q1 Rev. 55 Rev. 55 Page Page 42 42 of of 1141 114
Unit Jnit 22 APP P A-Ol?-i 4-14-i pag&
Page 11 of of 22 NPCI 'I'lIRB BPCI T1JE TRIP EOL ENllR TRtP SOL ENER AUTO ACTICtiS 1.
- 1. rf .~pen, rf open, the Turbine Stop the TUrbin& Etop Valv&,
Valve, E41-VS, E4L-V, clos&s.
closes.
CAUSE 1.
- 1. Nigh reactor Bigh reactor vess&l vessel wat&r water l&v&l level i206 2C.t inch&s).
inches) 2.
- 2. 4echanical ':lversp&&d Mechanical *3verspeed trip trip (4600 4fOO rpm) rpm)..
1.
- 3. Nigh turbine Bigh turbine exhaust exhaust pr&ssur&
pressure (157.5 157.5 psigi.
psig).
4.
- 4. Low BPcr Low NPCI pumppump s'.Iction suction pr&ssur&
pressure (1515 1nch&s inches Kgg vacuum) vacuum:..
5.
- 5. Nigh turbina Bigh turbine axhaust exhaust d1aphragm diaphragm pr&ssur&
pressure (7 7 psig).
psig).
f.
- 6. Nigh staam Bigh steam lina line d1f.f&r&nt differential 1al pr&ssur&.
pressure.
7.
- 1. Low staam Low supply pr&ssur&
steam supply pressure (115 1L5 pa1gi.
psig).
P.
B. High BPcr Bigh HPCE room room ar&a area amb1&nt ambient t&mperature temperature (lGS'P).
t165P).
P.
- 9. Nigh BPcr Bigh steam line NPCI staam line area ambient temperat.~re area amb1ent temperature (190'Pi.
(1;P).
ic.
- 10. 3igh BPCr Bigh wcr staam line t'.Innel steam 11ne tunnel temperature (lGS*P;(165P:.*.
11.
- 11. Nigh BPCr Bigh i4PCt staam steam lin&
line area d1f.ferent1al differential temperature (4?'P). (47P).
12.
- 12. Turbine trip push button.
TUrbina 11.
- 13. malfunction.
Circuit malfun-:::t10n.
BERVATtCNS OBSERVATrCNS
- 1. Turbine stop TUrbina Ptop Valve, E41-VS, E41-V2, closed.
ACTrCNS
- 1. the turbine tr1pped rf tha tripped or 1s01ated, lsolated, refer to tc APP PP A-Ol A-Oi 3-1.
i-i.
- 2. tf aa circuit malfunct10n rf malfunction 1s ensure that aa 'ii'R/,-"<)
is suspected, ens'~re wP./3 is prepared.
praparad.
DEVICE /EETPDtNT DEVrCE!SETPOrNTS HPCI ?aixiliary Ralay BPcr Auxiliary Relay E41-K12 Energized Level Tl'ansmitter Lavel Transmitter J-Iaster Master Tr1p Trip Un1t Jnit 20G 206 inches B21-LTt4-MOL7E-2 and 0-2 B21-LTM-~f017B-2 D-2 Turbine spaed TUl-bina speed 4GOO 46c0 rpm Turbine TUrbine Exhaust Pressure switch Switch 157.5 157 .5 psig E41-P-C17 E41-Ps-mn7A and H01.7E N017S NPCI BPer Pump Puction Pressure Suct ion Press'.Ire switch SlIIttch 15 inches Hg Kg vacuum tafter (aft&r E41-P-tc1c E41-PS-N010 13 1 second timetim& delay:
d&lay)
TUrbine Exhaust Diaphragm Turbine D1aphragm Pressure 7 psig E41-PPN-012?
E41-PSB-~f012A thru 0 D Pte.am Flow Staam &'10111 Differential Pressure -9S.G inches of
-P5.6 water Naster Mastar Trip Unit unit E41-PDTM-t1004-1 E41-PDTM-N004-1 (includes -26.7 Cincludea in.
in. head correct1c.n;,
correction) 2PP-A-O1 12APP-A-01 Rev.
Rev. 55 55 Page 62 of 114 Page62of114I Categories Categories KJA:
KIA: 295024G2.02.37 295024G 2.02.37 Tier/Group:
Tier / Group: TIG!
T1GI RO Rating:
RORating: 3.6 3.6 SRO Rating:
SRORating: 4.6 4.6 UP LP Obj:
Obj: Source:
Source: NEW NEW Cog Cog Level:
Level: HIGH HIGH Category Category 8:8: Y
- 52. AA Loss
- 52. Loss of of Off-site Off-site Power Power (LOOP)
(LOOP) occursoccurs on Unit One on Unit One following following operation operation at at rated rated power for power for the the last last 18 18 months.
months.
The RSP The RSP directs directs the the following following step:
step:
STABILIZE PRESS STABILIZE PRESS BELOW BELOW N 1050 PSIG 1050 PSIG WITH WITH ONE ONE OR MORE OR MORE OF OF THE FOLLOWING TIlE FOLLOWING SYSTEM&
SYSTEMS:
- VALVES MAiN TURBINE BYPASS MAIN TURBINE VALVES BYPASS
- MAIN STEAM LINE MAIN STEAM LINE DRAINS DRAINS
- RCIC RCIC
- HPCI HPCI
- CONTINUOUS SRV-IFA SRV*IFA PNEUMATIC CONTINUOUS PNEUMATIC SUPPLY IS IS AVAILABLE USE OPENING USE OPENING SEQUENCE SEQUENCJ 039 039 Which one of the following identifies the system or combination of systems which will provide sufficient steam flow to stabilize reactor pressure initially (within the first 10 minutes) following the event and why 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later this decision will be different regardless of Off-site power status?
(1) have the capacity to stabilize pressure immediately following the event.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later (2) 2 A. (1) Only SRVs (2) sufficient time has been available to to allow use of MSL Drains B (1) Only SRVs B!'"
(2) decay heat generation has significantly lowered to within the capacity of HPCI C. (1) HPCI and RCIC combined (2) sufficient time has been available to allow use of MSL Drains D. (1) HPCI and RCIC combined (2) decay heat generation has significantly lowered to within the capacity of HPCI
Feedback Feedback K/A: 295025 KJA: 295025 A2.05A2.05 Ability to Ability determine and/or to determine and!or interpret interpret the the following following as as they they apply apply to to HIGH HIGH REACTOR REACTOR PRESSURE:
PRESSURE:
Decay heat Decay heat generation generation (CFR: 41.10/43.5/45.13)
(CFR: 41.10/43.5/45.13)
ROISRO Rating:
RO/SRO Rating: 3.4/3.6 3.4/3.6 CLSLP19*22c (16*16c)
Objective: CLS-LP-19*22c Objective: (1 6*1 6c)
- 22. Given
- 22. Given plant plant conditions, conditions, predict predict how how aa loss loss or or malfunction malfunction of of the the HPCI HPCI System System will will affect affect the the following:
following:
- c. Ability
- c. Ability to remove decay to remove heat decay heat
- 16. Given
- 16. Given plant plant conditions, conditions, predict predict how how aa loss loss or or malfunction malfunction of of the the RCIC RCIC System System will will affect affect the the following:
following:
- c. Ability to
- c. Ability remove decay to remove heat.
decay heat.
Reference:
Reference:
001-37.3, Revision 001-37.3, Revision 10,10, Page Page 34, 34, Section Section Cog Level:
Cog Level: HighHigh Explanation:
Explanation:
The amount The amount of of decay decay heat heat added added depends depends onon the the power power history history of of the the reactor reactor and and the the amount of time time since the reactor was shut down. The number of fissions that have occurred determines the number of fission fragments in the core. Initial Decay Heat generation is equivalent to approximately 7% (beyond the equilibrium power prior to the scram. 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> following the scram, Decay Heat capacity of HPCI) of the equilibrium generation is equivalent to approximately 1% power (within the capacity of HPCI and maybe RCIC).
Distractor Analysis:
Distractor Choice A: Plausible because only SRVs is correct and the use of MSL drains is desired but is dependent upon Off-site power availability (CWIPs needed to allow the main condenser to be available as a heat sink), Group 1I isolation signal remains due to low condenser vacuum with no 11 OP-25 or EOP guidance to bypass and reset the isolation signal. Reopening MSIVs would not be procedurally allowed due to Cond/FW and CW systems not having power.
Choice B: Correct Answer Choice C: Plausible because HPCI and RCIC combined capacity is below 7% and the use of MSL drains is desired but is dependent upon Off-site power availability (CWIP5 (CWIPs needed to allow the main condenser to be available as a heat sink), Group 11 isolation signal remains due to low condenser vacuum with no 10P-25 1 OP-25 or EOP guidance to bypass and reset the isolation signal.
Reopening MSIVs would not be procedurally allowed due due to Cond/FW and CW systems not having power.
Choice D: Plausible becausebecause HPCI and RCIC combined capacity is below 7%
is below 7% and decay decay heat generation lowering isis correct.
SRO Only Basis:
SRO Only Basis: N/A Notes Notes
STEP 039 I
,I" TABIUZE PRESS STABILIL£ PRESS BELOW BELOW ISSO PSIG lQ~ WITH ONE P510 'tilTH ONE OR !\lORE OR NORE OF OFTH FOLLOWING THlE FOLLOWING Yfl4MS:
sYSTeMS:
- I4A1N TURIiiIIE IiiYll'ASS HANTURP BYPASS VALVES VALVES
- I4AIIli L4N 51'f:AM DRAINS DRAINS:
STEAM liNe LINE
- ECIC RICI!;:
- RV-A
- SR .... -IFA CONTINUOUS PNEUMATIC CONTIIilUOUS PNEUMATIC UPPIY IS SUPPI.Y IS il,VAlLAllll.e AAUA2L )
\_ UE OPeNING ItS!;
I OPENING SEQUeNCE SEQUENCE
- 039 STEP 039 (continued)
When manual SRV actuation is required for reactor pressure control, an opening sequence is preferred which distributes heat uniformly throughout the Suppression Pool hih local pool temperatures which may result in inefficient pool cooling. The to avoid high opening sequence also uniformly distributes tile the total number of SRV actuations among the total number of SRVs.
Use of steam driven pumps (te.,(Le., HPCI, RCIC, and RFP) to augment reactor pressure control may be required. These systems do do not draw a significant amount of steam but be sufficient to control reactor pressure increases, or in conjunction with other may I)e systems may assist in controlling reactor pressure. Suction for HPCI I-f PCI and RCIC, RCIC. in the pressure control mode, is always to be aligned to the condensate storage tank (CST).
Use of auxiliary systems and lineups may be required to keep water in the CST.
001-37.3 1001-37.3 Rev. 10 10 381 Page 34 of 38
1.3 1.3 General Description General Description {Figure (Figure 16-1) 16-1)
Following aa reactor Following reactor scram, steam generation scram, steam generation will will continue continue duedue toto the the fission fission product decay product decay heat heat. Normally, Normally, the the Main Main Turbine Turbine Bypass Bypass System System will will divert divert the steam the steam to to the the main condenser, and main condenser, and the the Feedwater Feedwater SystemSystem will will supply supply thethe makeup water makeup water required required toto maintain maintain reactor reactor vessel vessel inventory.
inventory.
In the event In event the reactor reactor vessel is is isolated isolated due to MainMain Steam Isolation Isolation Valve (MSIV) closure,
{MSIV) closure, the relief relief valves will maintain pressure pressure in the vessel within limits. The isolation of acceptable limits. of the the reactor vessel will disable disable the Feedwater System Feedwater System since since the the steam steam required required for for the the operation operation ofor the the reactor reactor teed pump turbines feed pump turbines isis supplied supplied from from thethe main main steam steam lines.
lines. Due Due toto the continuous steam generation and discharge through the relief valves, '.'later water level in the reactor vessel will decrease.
decrease. To maintain reactor reactor vessel inventory, the RCIC System may may be he used for injection injection to compensate for thls this loss in makeup water. The RCIC System also helps to depressurize the vessel by using decay heat steam from the vessel to operate the RCIC RCIC Turbine and by returning cooler water to the reactor vessel.
The RCIC System consists of a 100 percent capacity steam turbine driven associated piping, valves, lnstrumentation, pump, with aSSOCiated instrumentation, controls, and accessories. The system is capable of delivering makeup water to the reactor vessel under rated pressure conditions. The RCIC System has a capacity approximately equal to the reactor water boif-off boil-off rate 15 to 20 minutes after shutdown. All components necessary for initiating operations of RCIC are completely independent of auxiliary or emergency AC power, service air, and external cooling water systems, and require only DC plant servIce power from the station batteries, therefore providing a high degree of assurance that RCIC will operate wl1en when required.
The loss of feedwater evaluation for the 105% Power Uprate relied on RCIC offeedwater operating alone with 360 gpm of makeup starting 60 seconds after initiation.
For some loss of feedwater events, indicated water level may drop below the LL3 setpoint, resulting in MSIV closure, ADS timer start, and a ll3 a low low pressure ECCS start signal. Even though ll3 LL3 actuations may occur, operators are expected to inhibit ADS and allow RCIC to restore level. The lowest expected level inside the shroud would be no less than 4.7 ft above the top of active fuel. This is considered acceptable since ADS .L\OS blowdown and low pressure ECCS injection are not required to provide adequate core cooling.
ISD-16 SO-16 Rev. 99 Page 7 of Page of 120 120 I
The horsepower demand The horsepower demand on on the the turbine turbine isisa a function function ofthe of the requirements requirements to dri'/e to the pumps.
drive the pumps. Therefore, Therefore, the the steam steam drawn drawn by by the the turbine turbine can can behe regulated by regulated by adjusting adjusting the power used the power used byby the the pumps.
pumps. ThisThis relationship relationship is used is when the used when the HPCI HPCI System System is is operated operated in in the the Test/Pressure Test/Pressure Control Control Mode to Mode remove decay to remove decay heat heat from from thethe Reactor.
Reactor. ByBy regulating regulating thethe load load on on the pumps the pumps (Le.,
(i.e., adjusting adjusting pump pump discharge discharge pressure pressure by by throttling throttling thethe test return valve retum valve andJor and/or adjusting adjusting flow),
flow), the the amount amount of of decay decay heat heat removed removed from the from the Reactor Reactor cancan be controlled. Design be controlled. Design data data for for the the turbine turbine isis shown shown in Table 19-2, in 19-2, below.
below.
19 HPCI Table 19-2 - HPCI Turbine DesignDesign Data Type speed, non Variable speed, noncondensing condensing turbine turbine Rated Speed Rated 4100 rpm rpm Rated Steam Inlet 135 psig to 1250 psig conditions 358 of °F to 575 of °F Exhaust Pressure 200 psig max 50 psig design operating 15 - 30 psig nominal 15 - nominal operating Turbine Steam flow data based on pump speeds and required power for 4250 gpm discharge flows.
Pump Head Speed Brake Horsepower Steam Flow ft 525 It 2100 rpm 750 bhp 83,0001bmfhr 83,000 Ibm/hr 2712 ft 3940 rpm 3850 bhp 3850l>hp 178,000 Ibm/hr 178,0001bm/hr 2800 ft 3995 rpm 4000 bhp 182,0001bm/hr 182,000 Ibm/hr 2823 ft 4015 rpm 4050 bhp 184,000 Ibm/hr 184,0001bm/hr 2970 ft 4100 rpm 4350 bhp 191.000 Ibm/hr 191,0001bm/hr 2.2.2 Piping and Valves (figure (Figure 19-2)
Steam to drive the turbine is supplied from Main Steam Line A, "A",
upstream of the Main Steam Isolation Isolation Valves, through the Steam Supply Inboard Inboard and and Outboard Isolation Valves, E41 Outboard Isolation E41-F002
-F002 and E41 E41-F003,
-F003, respectively. The Steam Supply Supply Isolation Isolation Valves, like the HPCI HPCI Pump Suppression Pool Suction Valves, 841 -F042 and F041, are PCIS Group E41-F042 4 Isolation Valves which receive automatic isolation isolation signals. The 1 SD-19 SD-19 Rev. 16 Rev. 16 Page Page l2of 12 of 108 1081
Figure 8-9 Figure
- 8-9 plots decay decay heat heat following following aa reactor reactor shutdown.
shutdown.
100 ...
100 111:::'-------------.....,
so 80 j
00 1-0.. 60 60
<C ....
0 11.1
- c~Q) 40 40 0::
~
20 00 0.01 0.1 1.0 1.0 10 100 1000 1000 10.000 1000 TIME AFTER SHUTDOWlI llME SHUTDOWN (SEC)
Figure 8-9 Decay Heatvs. Heat vs. Time aftel' after Shutdown B\rR 'REACTOR rnEORY DOF3C?.AflCN F.EADDR OPEP_4.TIONALPHYSICS
?.EACTOR OPEP_CON;i PHY:iDS REV 3
X.
X. DECAYHEAT DECAY HEAT PRODUCTION PRODUCTION (Figure (Figure7) 7)
Decay heat Decay heat isis themlal themtal energy energyproduced produced by bythe decayoffission the decayof fission products.
products. The energy Theenergy carried by carried the beta by the beta and and gamma gamma radiation radiationemit1ed emitted by fission products by fission products isis rapidlyoofl\ler1ed rapidlycortverted to to thermal energy thermal energy as as the the radiation radiation reacts reacts with withthe thesurrounding surrounding medium.meditm. ItItamollnts amounts to to approximately 7%
approximately 7% ofof the the energy energy produced produced ininfission.
fission.
i.
- U:*LP-107-A Rev. 5 Fage 38 of
- M. EM STUDENT HANDOUT Even though the Even the neutron neutron chain chain reaction reaction isis abruptly halled hailed folkriAng following aa scram, scram, decay decay energy continues energy continues to be be produced produced in in large large quantities quantities.
~'I9,1~ t t ' even after vellater several hours several hours the decay heat heat is produoed produced at a rate rate of about 1% 1% 0of full power.
power.
Aquantitative A quantitative estimate of decay heat'llas heatwas filStgiven Wignerartd Way in 1948.
firstgiven by Wignerand 194& Their formula follows:
It) pp (1)) S .OZt Pp [1>
O.06H It + t It +1>>
- (1)
) I rJ: P(t) is power generation due to beta and gamma rays P.(t)
P, is reactor power before shutdowll P. shutdown
- t. is time of power operation before shutdowll shutdown (sec},(sac), and is tis time elapsed since shutdown (see) shutoown (sec)
If adequate cooling is not provided following reactor shutdown, the decay heatC3l1.ses folkrMng reaclershutdo'im, heatcauses overheating and eventual melting meltirtgof the reacler of tile reactor fuel. This in turn tllm leads to the releaseof the more volatile fission products.
Decay heat varies linearlylinearlywith with power beforeshutd()'Im.
beforeshutdown. The higher the powerbeforepower before shutdown, shutdoYm, the more decayheat decay heat will Y.ill be produced at a given time after shutdown. Of course, time spent ata given power powerlevs imponant tocf If levd is importanttoof the reaclorwas Ifthe reaclerwasat at 100%
100%
powerfor power for several days. days, then then decreased decreased to 15% 15% fora bert time before shiadown, for a sshort silutdown, we must must considerthe consider the time Spellt ateach time spent at each level.
level. Creditfortirneat Credit for time at eachlevel is given each level is given through through the the !'AM!"AIJ1~WK The The amount amollnt of of decay produce:! decreases with time after decay heat produced shutdown.
.shuidowiC" "
cio 10.0 1,0 -
w I- -
z Lii -
ci Lii
~3"[]d lN3J~3d ci 0.1 :
- 1 HOUR 1 DAY 1 WK 1MON 1YR 10 YRS o
ci 0.01_ I I I I I I I I I
- o ru o ('I) - o\0 o ,.... o CX) ci 0.1 1 10 -o iü2 4 io - io6 io8 0\0
- - - l.aJ 2:
I-TIME (SEC) l.aJ CI) "U 5
io -
Categories Categories K/A:
KIA: 295025 A2.05 295025 A2.05 Tier// Group:
Tier Group: T1G1 TIGl RO Rating:
RORating: 3.4 3.4 SRO Rating:
SRORating: 3.6 3.6 LP Obj:
LPObj: CLSLP19*22C CLS-LP-19*22C Source:
Source: NEW NEW Cog Level:
Cog Level: HIGH HIGH Category 8:8:
Category YY
- 53. An ATWS
- 53. An ATWS hashas occurred occurred on Unit One on Unit One with with the the following following plant plant conditions:
conditions:
Reactor power Reactor power 3%, slowly 3%, slowly lowering lowering RPV water level RPV water level +60 inches,
+60 inches, steady steady RPV pressure RPV pressure 300 psig 300 psig Drywell pressure Drywell pressure 3.0 psig 3.0 psig Suppression pool Suppression pool temp temp 108°F 108°F Which one Which one of of the the following following identifies identifies the RHR logic the RHR logic requirement(s),
requirement(s), ifif any, any, to to place place Suppression Pool Suppression Pool Cooling Cooling in service under in service under thethe current current plant plant conditions?
conditions?
Suppression Pool Suppression Pool Cooling Cooling isis placed placed in service:
in service:
A. without the A. without the use use of of any any overrides.
overrides.
B B~ only by placing the Think Switch to Manual.
by placing Manual.
C. only by C. placing the Think Switch by placing Switch to Manual Manual followed by by bypassing bypassing the 2/3rd2/3rd core height & LPCI interlocks.
D. only by bypassing the 2/3rd core height & LPCI interlocks followed by placing the Think Switch to Manual.
Feedback Feedback K/A: 295026 KIA: 295026 A1.01 Aid Ability to Ability operate and/or to operate monitor the andlor monitor the following following asas they they apply apply to to SUPPRESSION SUPPRESSION POOL POOL HIGHHIGH WATER TEMPERATURE:
WATER TEMPERATURE:
Suppression pool Suppression pooi cooling cooling (CFR: 41.7 (CFR: 41.7 /45.6)
/ 45.6)
RO/SRO Rating:
RO/SRO Rating: 4.1/4.1 4.1/4.1 Objective: LOl-CLS-LP-01 7A*009 Objective: LOI-CLS-LP-017 -A*009 Given an Given an RHR RHR pump pump or valve, list or valve, list the interlocks, permissives the interlocks, permissives and/or and/or automatic automatic actions actions associated associated with with the the RHR pump RHR pump or valve, including or valve, including setpoints.
setpoints.
Reference:
Reference:
lOP-i 7, Revision 10P-17, Revision 97, 97, Page Page 282, 282, Attachment Attachment 88 Cog Level: High Cog Level: High Explanation:
Explanation:
Suppression Pool Suppression Pool temperature temperature and and DW DW pressure pressure are are elevated, elevated, along along with with RPV RPV water water level level above above LL3 LL3
(+45 inches).
(+45 inches). With With no LOCA signal no LOCA signal present, RHR RHR can can be be placed placed in in SPC SPC without the the use use of any logic logic overrides /I bypasses. Use of SPC Hardcard is required to place RHR RHR in SPC during EOPs.
EOPs. RO RO needs to recognize no LOCA signal present for the stated conditions and that making-up the SPC/Spray logic is not required.
Distractor Analysis:
Choice A: Plausible because if the LOCA signal was not present this would be correct.
Choice B: Correct Answer.
Choice C: Plausible because incorrect recognition of LOCA signal conditions and wrong order of Cooling/Spray logic switch manipulation.
Choice D: Plausible because incorrect recognition of LOCA signal conditions and correct order of Cooling/Spray logic switch manipulation.
Notes Notes 3.2.4 3.2.4 Containment Cooling Containment Cooling Logic, Logic, (Figures (Figures 17*11 17-11 and 17-12) and 17-12)
Containment Cooling Containment Cooling logic was modified logic was modified to to allow allow Suppression Suppression Pool Pool Cooling to be placed inin service Cooling to be placed service with with aa LPCI LPCI initiation initiation signal signal present present when Dr-ywell when Drywell pressure pressure isis below below aa 2.7 21 psig psig permissive.
permissive. This This allows allows operators to operators to perform perform Suppression Suppression Pool Pool Cooling Cooling during during an an Anticipated Anticipated Transient Without Transient Without Scram Scram (A (ATWS) event when TWS) event when level level is is lowered lowered below below the LPCI the LPCI initiation initiation setpoint setpoint.
Placing Suppression Placing Suppression Pool Pool Cooling Cooling (E1 l-FO2SA(B) and (El 'I-F028A(B) and E'11-F024A(B>>
El l-F024A(B))
in service in service with with aa LPCI LPCI initiation initiation signal signal present present requires:
requires:
- Reactor water level Reactor level above 213 core above 2/3 core height, height, OR OR
- 213 Core The 2/3 Core Height Height LPCllnitiation LPCI Initiation Override keylockkeylock switch be be placed in placed in MANUAL MANUAL OVERRIDE, OVERRIDE, AND
- The Containment Spray Valve Control (THINK) switch, S17B, SI7B, be placed in MANUAL.
FIGURE "17-1217-12 CoolinglSpray Permissive Logic Cooling/Spray CLOSED IN &
AFTER MANUAL I CLOSED IN MANUAL K69S (SEAL IN)
KOOB (818B) 2faCORE I k14B K14B CLOSED-213 CLOSED>213 CORE CORE HEIGHT HEIGHT (SEAL IN) HEIGHT K61B K61B OVRD CLOSED ON CLOSED ON LPCI LPOIINITIATION SIGNAL INITIATION SIGNAL K"'10A - 1(1 1O K110B}
K68B K111A I....-----4>I K111B K111)
CLOSED ON CLOSED ORYWELL DRYWELL ON PRESS>2.7PSIG PRESS>2.7PSIG KOOB (F0241F028) CONTAINMENT CONTAINMENT SPRAY SPRAY PERMISSIVE PERMISSIVE (F016. Ff21, (Ff18, F021, Ff27)
F027)
Categories Categories K/A:
KIA: 295026 Al.01 295026 A1.01 Tier // Group:
Tier Group: TIG1 T1G1 RORating:
RORating: 4.1 4.1 SRO Rating:
SRORating: 4.1 4.1 LP Obj:
LP Obj: LOICLSLP017A*009 LOI-CLS-LP-017-A*009 Source:
Source: NEW NEW Cog Level:
Cog Level: HIGH HIGH Category 8:
Category 8: YY
- 54. Which
- 54. Which oneone ofof the following identifies:
the following identifies:
(1) when (1) when aa reactor reactor scram scram due to Drywell due to Drywell Average Average Temperature Temperature isis required required lAW lAW PCCP PCCP and and (2) the (2) reason the the reason the reactor reactor scram scram isis required?
required?
A(1)
A'! before 300°F (1) before 300°F (2) tripping the (2) tripping the recirc recirc pumps pumps B. (1)
B. before 300°F (1) before 300°F (2) locking (2) locking out out the the drywell drywell coolers coolers C. (1)
C. cannot be (1) cannot be restored restored and maintained maintained below below 300°F 300°F (2) tripping the recirc (2) tripping recirc pumps pumps D. (1) cannot be restored and maintained below 300°F (2) locking out the drywell coolerscoolers
Feedback Feedback K/A: 295028 KIA: 295028 K3.05 K3.05 Knowledge of Knowledge of the the reasons reasons for for the the following following responses responses as as they they apply apply to to HIGH HIGH DRYWELL DRYWELL TEMPERATURE:
TEMPERATURE:
Reactor SCRAM Reactor SCRAM (CFR: 41.5 (CFR: 41.5/45.6) 1 45.6)
ROISRO Rating:
RO/SRO Rating: 3.6/3.7 3.6/3.7 Objective: CLSLP300L*005g Objective: CLS-LP-300-L *005g
- 5. Given
- 5. Given the the Primary Primary Containment Containment Control Control Procedure, Procedure, determine determine the the appropriate appropriate operator operator actions actions ifif any any of the following limits of limits are approached approached or exceeded:
- g. Drywell Design Design Temperature Limit Limit
Reference:
Reference:
001-37.8, Revision 001-37.8, Revision 4, Page Page 21, 21, STEPS STEPS DW/T-09 DW/T-09 through DW/T-17 DWIT-17 Cog Level: Low Explanation:
A reactor scram is inserted once it has been determined that drywell temperature cannot be maintained A
below 300°F and DW Spray is required. In order to spray the DW, the Reactor Recirculation Pumps and DW Coolers need to be secured. The reactor is not allowed operation at power without Recirculation Pumps in service. The scram requirement step satisfies shutting down the reactor to support tripping the Recirculation pumps.
Distractor Analysis:
Choice A: Correct Answer Choice B: Plausible because 300°F Drywell temperature is correct. Locking out the DW coolers is an action in the DW spray procedure but is not the reason for scramming the reactor reactor....
Choice C: Plausible because not being able to restore and maintain below 300°F Drywell temperature means exceeding 300°F is allowed without scram but is the step requiring emergency depressurization and tripping the recirc pumps is correct.
Choice D: Plausible because not being able to restore and maintain below 300°F Drywell temperature means exceeding 300°F is allowed without scram but is the step requiring emergency depressurization and Locking out the DW coolers is an action in the DW spray procedure but is not the reason for scramming the reactor.
SRO Only Basis: N/A Notes
OWJT 150°F St*RT ALL ?AaASLE CWLLCOOL9t. ATLO Wfl8LcOLER 1UOCKS NECESSARYR OWJT 300"F IIRW#B.L AVERAGE AIR TEMP ilEACllES 300'F IlfiI'WELLSFRAY 1lEQIJUIB) 1lW1T*11
STEPS DW/T-09 DW?T-09 through DWiT-17 (continued)
[)W!T-17 (continued) e.
- e. The sertsi:iiy of llle The sellsili~~iy the p"Jmp pump tooper.alioll to operation beyondbeyono the limit limit f.
- i. The COil The con sequences secuences cf of not no operating the pump beyond beycn the limit mmeoiaIe alld tmmediate and catastrophio catastrophic failure failure isis Ilot not expected iff a pump pump i'S5 operaled operated beyond beyonc the NPSH or NPSH or vortex vortex limit.
The initiabon of The initiation of drYVle~
drywe sprays sprays is is conditiolled concticned on on Ihe the fQllowing rollcwing restri,;:tions restrotions 00on Ihe the plallt plant.
The .recirculation The recirwlation pumps pumps and and drywell drwell cooling cco-ng fans are requred 10 are requim-ed be secured to be secured prior prior to the inhiation of iniiiation of di!)'wellsprays.
drywell sprays. TheseThese aClions actions are covered covered illin EOP-Ol-SEP-02.
EOP-O1-SEP-02. Sillce Since reactor operation at power is at power is IlQt not allowed or or desired desired without recirculation recirculation pumps pumps illin service and is reslricted restricted ill in time for temperatures alio'le E.D"F. aa step has above 1150F. has been been added to scram ihe the reactor and enter into into EOP-D EOPD1. 1.
Another restrictioll restriction 011 initiation of drY on the initia!:on drywei
....ell sprays isis for suppression pool water level level to bE'!
to be below +21 inchE'!s.
inches. This pro'idE'!s provdes protectlon protection for t.he the operaiicll operation of the suppression ohamber-co-drywell ohamber-io-drywE'!lIvaauum vacuum breakers. The The vacuum breakers will not funaton brE'!akers funct:on as designed ifif any designed any portioo the valvE'!
portion of llle valve [5s covered with water. The specified water level assures that assures that 110 rio poit'on porbon of the drywell side of the valve is submerged for any dr/well drywall be!ow wetwell befo\'i wetwell differenti.al diiferental prE'!ssure pressure less lhan than or Cr equal to the val-Ie valve opening differential pressure, Spray operatiollwith pressure. operation with lIaouum vacuum breakers inoperableinoparabe (I.e._
(i.e., wiih wth no dJ)well dryeti vacuum re!ief
',aCllum ree1 capab~ity) capab-iWl may calise cause the cOlltaillmellt containment differential pressure capability to be exceeded and is lherefore therefore Ilot not pemlitted.
pemitted.
Step DWiT-16 Step DWiT-1 assllresassures adequate core coo's'llg coong ial<.E'!S takes precedence prececence over initiating dr/well dr,well spray ill spray in this this case since catastrophilc catastropWc fa.~urefa-ure or or the primarl primary containment ~ -s not expected under the oonditions under conditions for which spray requiremellts requirements are eSlabl~hed.
establshec. The wording of the step doE'!s step perm: alternating does permit alternating between reacior reactor 'lessel vessel inject'on njecbon .and dryaell spray modes and drywell as the as the need need far for each OCC'UfS.
occurs. pro'l:ded proced adequate oore core cooling cooling can be maintained.
car, bs Drywell sprays Drywell drfneli pressure drops spray-s are secured iff drlt'lei! drops 10 to 2.5 ps'g.
psg. This ~ S a backup step to the automatic securing of the sprays during a LOCA coodition con diton when the sprayscray pemissive interlook pemlissrve interlock drops 0111.cut, This precludes air from being crawrt ill beiri drawil in through the vacuum re!'ef YaD.iUm reef system to de-inendc-inert the primary prrnary collta.nment ccnIannient and also provides a positi~-e positive margin to margin to ihe the negati-ie negative design design press'Ure pressure of the primary oontainment.
containment.
The dryvlE!~
The dry.v&I sprays actuated rt accordance sprays are aot'Uated acoordance whh EOP-D1-SP-O2.
with EOP-D1-SEP-D2.
aol-37.a 1001-37.8 Rev.44 Rev. Page 21 of 58 I Categories K/A:
KIA: 295028 K3.05 Tier!
Tier / Group: TIOl T1G1 RO Rating:
RORating: 3.6 SRO Rating:
SRORating: 3.7 LP Obj: CLSLP300L*005G CLS-LP-300-L *005G Source: NEW Cog Level: LOW Category 8: Y
- 55. The Safety
- 55. The Parameter Display Safety Parameter Display System System (SPOS)
(SPDS) Plant Plant Status Status Matrix Matrix indicates indicates Suppression Pool Suppression Pool level level isis -31.5
-31.5 inches.
inches.
Which one Which one of the following of the following identifies identifies the the color color code code displayed displayed by by SPOS SPDS duedue to to Suppression Pool Suppression Pool level?
level?
SPDS Suppression SPOS Suppression PoolPool level level color color code code is:
is:
Green A. Green A.
Yellow B. Yellow B.
Red Ce Red Cy D. Cyan O. Cyan
Feedback Feedback K/A: 295030 KIA: 295030 K2.09 K2.09 Knowledge ofthe Knowledge of the interrelations interrelations between between LOW LOW SUPPRESSION SUPPRESSION POOL POOL WATER WATER LEVELLEVEL and and thethe following:
following:
SPDS/ERFIS/CRI DSIGDS: Plant-Specific SPDS/ERFIS/CRIDS/GDS: Plant-Specific (CFR: 41.7 (CFR: 41.7 1/ 45.8) 45.8)
RO/SRO Rating:
RO/SRO Rating: 2.5/2.8 2.5/2.8 CLSLP060*002 Objective: CLS-LP-060*002 Objective:
- 02. Describe
- 02. Describe thethe basic basic operation operation ofof the ERFIS/SPDS Computer:
the ERFIS/SPDS Computer:
- e. Monitor Display
- e. Monitor Display Color Color Code Code
- 03. Describe the
- 03. Describe the information information on on the Critical Plant the Critical Matrix.
Plant Matrix.
- 04. Describe
- 04. Describe the methods methods used used to do the following do the following onon the ERFIS/SPDS ERFIS/SPDS Computer:
Computer:
Evaluate EOP
- a. Evaluate EOP entry entry conditions.
Reference:
Reference:
11 (2)OP-, Revision, Page, Page Section SD-60 Rev.2, ERFIS DATA ACQUISITION, ACQUISITION, PROCESSING,PROCESSING, AND DISPLAY Cog Level: Low Low Explanation:
Requires RO to know Tech Spec required Suppression Pool water level of =:: -31 inches and .:5.<-27 of> -27 inches and that PCCP entry condition is SP level below -31" -31 or above -27"
-27 (i.e. -31.2 or -26.8).
Knowing the specific level at which the display turns yellow just informs the operator that it is approaching the High/Low alarm (TS Limits).
SPDS display will be green when SP level is < -27.5 and>
< -27.5" and > -30.5"
-30.5 the indication will turn yellow above
-27.5 or below -30.5"
-27.5" -30.5 until the limit of -27"
-27 or -31"
-31 is reached at which time the code turns red. The red code alerts the operator of possible PCCP entry condition. AOP-14 must be exited under these conditions.
Distractor Analysis:
Choice A: Plausible because -31.5 is easily confused due to being a negative number which combined with greater than or equal signs make this value within the normal band and therefore Green.
Choice B: Plausible because for the same reason above except approaching alarm limit.
Choice C: Correct Answer Choice D: Plausible because a wrong assumption by the candidate beyond the stem of the question. All of the inputs to SPDS are operable.
SRO Only Basis: N/A Notes
PRIMARY CONTAINMENT PRIMARY CONTAINMENT
\ CONTROL CONTROL PCCP-1 PCCP-1 ENTRY CONDITIONS:
ENTRY CONDITIONS:
- SUPPRESSION POOL SUPPRESSION ABOVE POOL TEMP 95°FQABOV ABOVE 95°F .QB.ABOVE TEMP E
105°F WHEN 105°F WHEN DUE DUE TO TO TESTING TESTING
- DRYWELL AVERAGE DRYWELL AVERAGE TEMPABOV AIR TEMP ABOVE 150°F E 150"F
- SUPPRESSION POOL WATER SUPPRESSION LEVEL ABOVE -27 INCHES
(-2 FEET & 3 INCHES)
- SUPPRESSION POOL WATER SUPPRESSION LEVEL BELOW ~-31 31 INCHES
(-2 FEET & 7 INCHES)
- PRIMARYC PRIMARY TMTH2 CTMT H2 CONCENTRATION ABOVE I .5%
1.5%
PCCP-2
15.
- 15. Color Coding Color Coding The SPDS The SPOS displays displays utilize utilize several several colors colors which which indicate indicate oondition condition or or status.
status.
The following The foDtYing CRT CRT oolors colors have have been been selected selected for for use use on on the the displays.
displays.
1)
- 1) AWhite - used
'White - used for for dra'.vings, drawings, some some box box outlines, outlines, numbers numbers andand titles.
titles.
2)
- 2) Green Safe Greell- - Safe condition condition (within (within limits).
limits). Green Green isis also used to also used to indicate closed indicate closed valves valves and/or andior piping piping systems.
systems.
3)
- 3) - Caution. The parameter Yellow - Caution. parameter or condition is out of of the nomial operating norma! operating band but has not yet reached band but reached an alarm alarm condition.
4)
- 4) Magenta - Bad Magenta - Bad datafnot data!not measured measured indicated indicated parameter parameter isis magenta.
- 5) Cyan - Data not validated. This color indicates that the parameter has not not Men been validated. If there is only one output parameter or if the on-scale signals are not signal for a parameter not consistent itit will appear as a solid cyan block containing containin white test.
- 6) Red - .Alarm
- Alarm Condition - When a parameter reaches the alarm limit the display changes to red. Red also indicates inchcates an open position andfor valve pOSition andlor piping system.
- 7) Blue - First five keys on the SOFT KEY menu bar matches Iirst five the first fNe function keys on the keyboard. This makes the associated keys easier to locate.
8)
) Gold - Last five keys on the SOFT KEY menu bar matches the five function keys on the keyboard.
I 1 SD-SO SO-50 Rev.66 Rev. Page 20 of 1041 Page2Oof1U4
APPENOIX C
.APPENDIX C SS PDS PDS Interpretation Interpretation of Da.ta of Data Cobra Ccor6 Indicate COlors in iae tM slaLus of" e slalus system. TIle o system. Tne COl!Jrs and ttu=r cors and tier meallllngEi rne3iings are are 03& as 1't:lIk:~'#s:.
foi1s:
1 unse f.~ rea digital 4:l1&~
1.
- 1. An urtSafE!!
An sIe fEii Is beil1Q beiq measlJiiellj measured. A A red 'ilQ1IiII diepay Imd~ ind1cate i1I1 au lil~ MQIl'I<< ~ .canjjl1li:iA.
uneafehlqhouii condition 2
- 2. Va1IecqerL
'u'iIN'E! Is ern A IIiibL flaL an alarm cordlilon Ic beftnçt recorded.
-J Bail IaLa Data ecWed Thxn iáit(c) is be{nq sar d Is [flyalil. Arhy output liag a magenta color sibould be rejeded kDT.aSLred Forers iLe Mdb&ec iateie oataisecisousedtor n1ccmaliiaJ :e. or exampe, a p1cc wl be ccored wMte : aiow tIe Jeer 1 e quickly ofler:ed wItr the sreen n1ornaiIci. Tie rte pipe, oweer, is rt nicasured pcnL Cotorsindicae COlors Indlca'le ::he current slaiLfe C~
ne currentsla!u:s. c th!!
Inc 5)'5lem.
systen. TIlu:s.,
This, If an alarm COnllltlon Iraralarni conllton occur*soccurs andthE!n and rE::urn~ ro then reurrs [0 rormna ~l1e normal, aarm WII:
ne alarm wil pa&5 red to rrorr .red pass rrom 10 glfeE!1l green regardlE!SS regardless c,1o DpEl'aLor opera Lot '&'llelilen1k:n IiLerienlhn ImUle ir 1Fe GDP s)'s1e.mtsysien.
SymbOls Symbola Aya, A bJe-greeni p3iern cyan (blue-green) terd aaien pa~lE!m betl.&1l1J1 dispayed 'talue gIvEn displayed aiLe Iir.ica1es1r3ttre IltIlI catf:S 11i,at lti!2 '/alue aiue is nil;: \'alllllatelJl.
1& n:vaJ3te.
Asleiske fl a 1)ed indIa:e :nal ie aIa corio in s not meesired or oa. The aetereks are maer:3 611.(1: aIa Il'ld!cail:ooB Ba4 Da.ia IndcaLIonB In addmon in addl!lOn 10to Inc thE ba da1a ncaonc ball: data ist,ed in
.IrtlJllCatJonslsted In COLORS COLORSanlJlan SY1BO..S, SYMBOJLS, 1e tlrle.101IC'i\\'&1lg oiown ar1aons "'iiIlla!lOnS also al:s.a indioa:c Indica,,. bad dale:dii1a.:
A Mager:a border arJrd a te bOx BB.. V3iLedisayed1n V.alu:E! diSplayed In nlagenr.3.
magenta, {See Seeais als!) COLORS:
COLORS.
C.
C. Wt?:e WI0",. or or Magenta Magenta .. '. 4See:See alsoalso SYMCiLSi SYM50LSj o.
D, wr te or WIl1I:!! or IMagEl'lta egenta aiue value o !:If I E25 E.",25. See(SeE! also also COLORS COLORS)
I1 SO-GO
-° Rev..
I Page 59 of Page '1041
122.
122. SUPP POOL SUPP POOLWTR WTR LVLLVL (Plant Status Matrix]
tPlantStatus Matrix D[spIayCokr Status Message Code 1rai& iFEET1WNCLES) c3reen t-23Wi-27> SP water Iev& > -2 Safe Daubon FEET i-ND INCHES Yelloi - 3" (-27') > level ~ -2' 3%" (-27.5";1
-2' 25
-or or-
- 2. -2' 6W h"!O.o")
-2 6%." i-20.51 ~ revel> -2' 7" eeI :-2 7 (-31")
(-31:
Aami Alaml (FEET N4 INCHES Red Red 1 -2 3"
-2' 3 (-27'}
(-2r) ::; Ie"'el eveI
- or-
- 2. ::;-2'
- 2. 7 r (-31")
(4T).
123.
123. SUPPRESSION POOL POOL TEMPERATURES [241] [24fl values in The ~'alues ir the graphic are are drill driver by jhe en by the temperature validation paranie~rs ernperature validatiGn parameters shewn shovn on displays dispays 75 and 76D.
755 7E2.
The a.zirnu1h a2imuih leca.lion Iccation of each SRV SRVta1pipe, H PCI, and Rele tailpipe, HPCI, RCIC discharge into the suppression pool nto !he pcoi are sha.r. The display also shoo\'I1. shcvs -the aso shov.'s the 1erliperalure temperare at .seven ocatons in ihe seven iocations the pool.
SD-60 1 SO-50 Rev. 6 Rev.S Page 103 of 1041 I 0$
Suppression Pool Level Lowering Level
-27"
-27
-31 I
Categories Categories K/A:
KIA: 295030 K2.09 295030 K2.09 Tier // Group:
Tier Group: T1G1 TIGl RO Rating:
RORating: 2.5 2.5 SRO Rating:
SRORating: 2.8 2.8 LP Obj:
LP Obj: CLSLPO60*0O2 CLS-LP-060*002 Source:
Source: NEW NEW Cog Level:
Cog Level: LOW LOW Categoiy 8:
Category 8: YY
- 56. AA Design
- 56. Design BasisBasis LOCA LOCA has has occurred occurred on on Unit Unit One One with with CSCS LoopLoop BB as as the the only only available available RPV injection RPV injection source.
source.
Which one Which one of of the the following following correctly correctly completes completes the the statement statement below? below?
Maintaining reactor Maintaining reactor water water level level above above -57.5 -57.5 inches inches with with aa minimum minimum Core Core Spray Spray injection injection flowflow of of (1) (1) gpm gpm provides provides assurance assurance that that (2) (2) exists.
exists.
A. (1) 4700 A. (1) 4700 (2) adequate (2) adequate core core cooling cooling B. (1)
B. (1) 4700 4700 (2) minimum steam (2) minimum steam cooling cooling water level level C
C~ (1) 5000 (1) 5000 (2) adequate core cooling D. (1) 5000 (2) minimum steam cooling water level Feedback K/A: 295031 K3.03 KIA:
Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL:
Spray cooling
/45.6)
(CFR: 41.5 145.6)
ROISRO Rating: 4.1/4.4 RO/SRO CLSLP300B*0O8 Objective: CLS-LP-300-B*008 Define all EOP terms per the EOP definitions list in EOP-01-UG.EOP-01 -UG.
Reference:
EOP-0l EOP-01-UG -UG Cog Level: Low Cog Explanation: The reason is adequate adequate core cooling Adequate core cooling cooling exists per per EOP-UG if RPV level level is is at at the jet pump suction jet pump suction with Core Core Spray Spray injecting injecting at at @ 5000 5000 gpm. Jet pump suction Jet pump suction elevation elevation is -59", specified in is @ -59, in RVCP RVCP as as -57.5
-57.5 for instrument instrument readability.
Distractor Distractor Analysis:
Analysis:
Choice Choice A: A: Plausible Plausible because because 4700gpm 4700gpm waswas thethe old flow requirement old flow requirement prior prior to to EC#63657 EC#63657 and and ACC ACC is is correct.
correct.
Choice B:
Choice B: Plausible Plausible because because 4700gpm 4700gpm waswas thethe old flow requirement old flow requirement prior prior to to EC#63657 EC#63657 and and MSCWL MSCWL (LL4)
(LL4) is is -30
-30 inches inches (depressurize (depressurized) d) and and would would not be applicable not be applicable under under these these accident accident conditions.
conditions.
Choice Choice C: Correct Answer C: Correct Answer Choice 0:
Choice D: Plausible Plausible because because 5000 gpm isis correct 5000 gpm and MSCWL correct and MSCWL (LL4) -30 inches (LL4) isis -30 inches (depressurize (depressurized) and d) and would would not not be be applicable applicable under under these these accident accident conditions.
conditions.
Basis: N/A N/A Notes Notes
A1TACHMENT 5
,c..ITA;CHMENT Pag2of27 Page 2 of 27 Definitions Defil1itiol1s ADEQLWE CORE ADEQU.ATE CORE CooUNG COOliNG Hea1 removal froo"h Heai remo.val frc the the reaci!or reaccc suificient suffident to to prevenl rupuring the preveri rup1uring &eI elscil.
the fuel clad, Four viable Four iab!e medhanl:5fI\'S mechanms *of of adequate adequate core core cooling cooling exist exist 'Ni!:hin within the the EOPs:
EOPs:
- Core submergence Core
- Steam cooling Steam cooling 'hilh with injection injection ofat makeup makeup water water toto ihe reactor tle reactor Steam cooling Steam cooling '/.ilhout without injection injecon of makeup water to of makeup to therea.clor the reactor Reactor water fevel Reactor level at at jet pul"l1P sucticc 'Niih pump suction with a£ at East least one one core core spray pump pump injecting in'lothe injecting into the reaci!or :essel at reaccr vessel at 5000 5X0 gpm.
gpm.
AFTER AFTER FoIIcirig in FolIO'hlng in time t:rne or place place ANTICIPATED TRA.NSIENT ANTIC1P.6.TED TRANSIENT WlTHOUT WiTROUT SCRAM reactor IS The reaci!Gf is no.tshutdO'/.n nct shutdown follov.~ing fcIlcwing a scram.
ALTERNATE INJECTION SUBSYSTEMS ALTERNATE Systems whichwtiicli may be used 10. o inject water to ihethe reactor when the injection systems canno.t supply sufficient cannot sup~y injection water to ihe suFdent Injection lie *.resse~..
vesseL They are as tolIcws:
follO',l's:
- Service Water Fire PlOcection fire Protection System
- Demineraiized water via OenlineraflZed Keepfil Sy.stem
\!~ia ECCS KeepfiJl System SLC System (boron boron solution or demIneralized demineraii.zed water water)
- Heater Drains System RGDC RC[C iocaI (local manual operation)
- Emergency Diesel Makeup Pump Emergencry' IDEOP-Ol-UG OEOP-01-UG Rev. 5 Page 6262 of 11 o.f151 I
AiTAC-fMENT 6e ATTACHMENT Page 12 Page of 19 12 o,f 1 FIGURE 17 FIGURE 17 Unit 1I Reactor Unit Reactor Water Water Leve:l Level at at TAF TAF 0
-10
- 10 0
LU W
20
- 20
- t * *3;0 0
U Z
- .. 40
-40
-I j I'll!!!
FEI LEG LU W leMP' TMfr jcvF w
W -50
.... ElQ'II..
2r)J)'P
-J
.J - RI!f!UI!G Rt!I LG ci lllililP dP 4
Tr C -60 BELOW 00 OR LU W EQLIAL TO EQUAI..,O
~ -70 m10 200:'F 2flF 0
U C
ci z
Z -80
-90
=90
-100 ii i I I!j I III 1111111 I 11 w 1 1,150 roo ann oo 1 700 9001,100 uo 60 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)
WHEN REACTOR R8\CTOR PRESSURE IS IS LESS THAN 6G 60 P31G.
PSIG, USE !INDICATED INDICATED LEVEL TAF TAF IS IS -7.6
-7.5 NCKES.
INCHES.
IDEOP-Ol-UG DEOP'-01-UG I Rev. 55 I Page Page 99 of 1f1 o,i151 I
ATTACHMENi 68 AITACHMENT Pe 14 P.?!ge of 19 14 of W FIGURE 1&
FIGURE 1 Unit 1I Reactor Unit Reactoi-Watei- Level at Water level at ll-4 LL-4
{Minimum Steam
{Minimum Steam Cooling Cooling level)
Level) 0 HIUILL!
-w
-10 -
ABOVE (f) w W
- x:
- 2:0 30
-30 LL-4 00 z
=-
-J
..J
-40
- 40 LU W
- .- 50 WLU *50
-J
..J fit 60 1'IJ!!f!
FLti filM!"
AEOVE A2OVE ll!:G
!;( 1ilirF 0C.) -70 70 - REnE,;}
T8IIIP TEMP C .
-z BELOW OR CR Z 80
- 80 I!WIt.l'fO tQvM. TO NO'f!
-90
-100 1,150 300 500 700 900 1,100 i100 60 200 2:00 400 600 800 1,000 1.000 REACTOR PRESSURE (PSIG) (FSIG)
WHEN RE.4CTOR PRESSURE IS WHEfll REACTOR IS LESS THAN 80 60 P51G.
PSIG, USE USE INDICATED INDIC.t..1ED LEVEL.
lEIJiEL LL-4 lL41S -30.0 INCHES.
15-30.0 INCHES.
IDEOP-Ol-UG DEOP-01-UG Rei.
Rev. :55 I P Page 101 '101 of of 1i 151 I
ATrAcH1EN1 66 ATIACH!\1ENT le c 19 P.?(Ie1 t3 of FIGURE 19 FIGURE 1 Unit 1I Reactoll"Wated..;evel Unit Reactor Water Level at LL-5 at lL-5
{Minimum Zero (Minimum Zero Injection Injection LeveR)
Level) 00
-10
.-. m 20 2(1
~
- r: -30
-30 U0 z
Z
.-J W
..J w
-40
-40 w
w -50
~5(J
-J
....J C
W
-so 5
Q -70
-10 FI fl Z
I TlP ABOVE
-80 RF.W IL r 7L4p nrwrni
-90 m 9(1
-100
!JLiHIlIlIIJHI IHIIIWH[II liii tiso 100 3U0 500 700 [ 900 iWo 60 80 200 400 800 800 1,000 iG00 REACTOR PRESSURE (PSiG REACTO:R (PS'IG)
WHEN REACTOR REI>.CTOR PRESSURE PRESSURE IS IS LESS THAN 80 60 P51G.
PSIG. USE LISE INDICATED
[INDICATED LEVEL lEVEL lL-5 S LL-S 47.5 INCHES.
RS-47.5INCHES_
IEJEOP-Ol-UG OEOP-01-L1G Rev 55 Pace Page 102 103 Gfof if 151I I
AA1TAcHMErff rr,lliCHMENT11 Pane 55 Page 55 of 101 of 1.01 EPG Plant EPG Pant Specific Specific Technical Tectrnieal Guidelines Guidelines.
C1- IfWreactorvese1 C1-5 water level reactor \I"eSosel 'Nater level can oan be restored and e restored and maintained maintained abo'.'e:
above:
LL4 (Minimum Ll-4 iMinimurn steam steam Cooling Cooling Reac.cor Reactor Water Water Level)
Leveli or or
- -57.5 inohes
-57.5 inches {elevation eIevation of of the the jet pumps suction}
jet pumps with at suction: with at least least one one Core Core Spray pump Spray pump inieoting irec1ing in~o into the the reactor reactor vessel essel at greater than at greater than or equal to or equal to SOOD prn.
500llgprn.
establish prin1;31)1 esiablish primary coniai,nmenl containment oooling tooling require merits to mainfain requirements maintain NPSH NPSK furor the the ECGS pumps ECGS reducing LPCI pumps reduoing LPCI injeciion i1cw before injection flo'N before suppression suppression poolpcc ~emperalure temperature reaches iF.
reache.s1 5ff'F.
C1-C 1-6 IfWreactor water level reactor vessel 'I.later level c...-mnot cannot be be res.~ored restored and ma:intained maintained above -57.5 inches {elevation lelevation of the the jet pumps pumps suction) with at least one Core Spra~'
suction. with Spray pump pump injecting into the react()if injecting reactor vessel wessel at greater than or equal SOCO gpm, or it has ecual to 5000 been determined been determined '!hat that f1000ing flooding of primary primary containment containment is is required required for for long lcrg term cooling, core oooling.
PRIMARY CONTAINM;ENT GONTAINMENT FLOODING FLOODINO IS IS REQUIRED; REQUIRED enier enter the Reador Reactor Ves.sel Vessel and Pr:imary and! Primary Containment Flooding Severe Seiere Accide.nt Accident Guideline.
°°-
1001-37 Rev. 54 I Page 62 of 23 239 I
AITACHMEN1 22 AITACHMENT Page 8B Page SB ofof 128 128 EPGIPSTG Step EPGJPSTG Step Documentation Documentation CONTINGENCY #1 CONTINGENCY #1 Al ALTERNATE TERNATELEVEl LEVEL CONTROL CONTROL SECTION SECTION EPG None EPGNol1e PSTG Ci-6 PSTGC1-6 iD Ul DEVIATIONS DBllAnONS This sep pro Thissiep proide guidarir fer
....ides gu;1dan(;e for fransferr,ing transfeming to primary oonlainrnent flooding primary oonlainm2nt flooding from the emergenc~'
enie genc operating operating procedures.
procedures. The The tr,ansfer transfer is is based based uponupon n01 nd being being ableable to maintain the to Fnalntain the loog term CQre long term core cooling cooling del'ined d&ned by by the UFS UFSAR.....R.lf Wreactor reactor vessel vessel water leye!
water Ieel cannot cannot be be d1"Iainfainedabove maintained abcwe ~he the sucfion of the jef suction Gfthe jet pumps, pumps, or or ifif injection injection spray cannot be I'lOOintained from core sprayoannot maintained above 5000 O0C glOm.
gpm. then there !is is not adequate assurance or of iong term core ion feml core cooling. The basis ior for the long lang term cOe core is that cooling cis oooling that 'the the rore reflcoded to.
core has been reflooded to the elevation eleaion of the jet pump pump sucfion suction and lhat core spray injection is that the core is pIo'.*iding providing cooling cooling tG to the up'per portion of upper pGrnon onheihe core fo oore maintain :iE!liipera~ure to maintain tenctperalure lb.\'.
ow. IfIf neiiher neither ofof these conditions conditions are mel, ~hen are mel. then additional aciions aclians are required reqired to main.iain restore cooling to the cOe.
mainlain or restoreoooling care.
211 2.0 DIFFERENCES None 001-37 1001-37 Rev. 54 Page ic of P,age 196 2391 of 23
Categories Categories K/A:
KIA: 295031 K3.03 295031 K3.03 Tier / Group:
Tier/Group: TIG1 T1G1 RO Rating:
RORating: 4.1 4.1 SRO SRORating:
Rating: 4.4 4.4 LP Obj:
LP Obj: CLSLP3O0B*0O8 CLS-LP-300-B*008 Source:
Source: NEW NEW Cog Level:
Cog Level: LOW LOW Category 8:8:
Category YY
- 57. During accident
- 57. During accident conditions conditions on on Unit Unit Two Two SCCP SCCP directed directed restarting restarting Reactor Reactor Building Building HVAC lAW HVAC lAW SEP-04, SEP-04, Reactor Reactor Building Building HVAC HVAC Restart Restart Procedure.
Procedure.
Shortly following Shortly following restart restart of of the the ventilation ventilation system system the the RO RO observes observes the the following:
following:
RX BLDG RX BLDG VENT VENT TEMP TEMP HIGHHIGH in Alarm in Alarm Rx Bldg Vent Exhaust Rad Rx Bldg Vent Exhaust Monitor A Rad Monitor A indication indication 2.0 2.0 mR/hr mR/hr Rx Bldg Rx Bldg Vent Vent Exhaust Exhaust Rad Rad Monitor Monitor BB indication indication 3.5 3.5 mR/hr mR/hr Based on Based on the the current current conditions conditions which which one one of of the following actions the following actions is(are) is(are) required?
required?
Continue to A. Continue to operate operate Reactor Reactor Building Building HVAC HVAC because because SEP-04 SEP-04 bypassed bypassed all all the the isolation logic.
B. Continue B. Continue to operate operate Reactor Reactor Building Building HVAC HVAC because because the Rad Rad Monitor Monitor readings readings are are no longer reliable.
Reactor Building HVAC and ensure SBGT is running because SEP-04 C. Isolate Reactor bypassed ALL the isolation logic.
D Isolate Reactor Building HVAC and ensure SBGT is running because the Rad Dy Monitor readings are no longer reliable.
Feedback Feedback K/A: 295032 KIA: K2.02 295032 K2.02 Knowledge of Knowledge of the the interrelations interrelations between between HIGH HIGH SECONDARY SECONDARY CONTAINMENT CONTAINMENT AREA AREA TEMPERATURE and TEMPERATURE and the the following:
following:
Secondary containment Secondary containment ventilation ventilation (CFR: 41.7 (CFR: 41.7 145.8)
/45.8)
RO/SRO Rating:
RO/SRO Rating: 3.6/3.7 3.6/3.7 CLSLP3ooM(K)*o1 1 Objective: CLS-LP-300-M(K)*011 Objective:
- 11. Given
- 11. Given plant conditions involving plant conditions involving Reactor Reactor Building Building HVAC HVAC system system isolation isolation and and the the Secondary Secondary Containment Control Containment Control Procedure, Procedure, determine determine ifif the the Reactor Reactor Building Building HVAC HVAC system system should be be restarted.
restarted.
- 11. Given plant
- 11. plant conditions and OEOP-01-SEP-04, OEOP-01 -SEP-04, determine the required required operator actions ifif a high high Reactor Building Vent radiation or Reactor Building Building Vent high temperature annunciator activates when restarting Reactor restarting Reactor Building HVAC.
HVAC.
Reference:
Reference:
OEOP-01 -SEP-04, Revision 12, Page 4, Section 2.9 OEOP-01-SEP-04, Cog Level: High Explanation:
Rx Building Vent Temp Hi alarm indicates temperature in the exhaust duct :::135°F 1 35°F deg. This exceeds the EQ of the Exh rad monitors. SEP-04 defeats RPV Low level, Hi DW pressure, and Main Stack rad Hi.
Rx Bldg Vent Rad Hi-Hi and Vent Temp Hi remain active and should have isolated RBHVAC and started both SBGT trains. SEP-04 also provides verification of these actions should either condition occur.
Distractor Analysis:
Choice A: Plausible because SCCP provided guidance to "restart" restart RB HVAC which can be interpreted to mean under any conditions. Not ALL isolation logic is bypassed.
Choice B: Plausible because SCCP provided guidance to "restart" restart RB HVAC which can be interpreted to mean under any conditions. Rad monitor readings not being reliable is correct.
Choice C: Plausible because isolating RB HVAC is correct, but not ALL isolation logic is bypassed.
Choice D: Correct Answer SRO Only Basis: N/A Notes
Another input
,Another input toto isolation isolation logic logic for for the the Reactor Reactor Building Building are are the the two two temperature switches temperature switches located located inin the the exhaust exhaust plenum.
plenum Should Should plenum plenum air air temperature rise temperature rise above i35F Reactor above 135"F Reactor Building Building HVAC HVAC will will be be isolated isolated and and SBGT will SBGT will start.
start. Also, Also, Annunciator Annunciator UA-3 UA-3 6-2, 6-2, RX RX BLDG BLDG VENTVENT TEMPTEMP Hi HI will will alarm. These alarm. These temperature temperature switches switches are also po'.vered are also from the powered from the same same source source as the radiation as me radiation modules modules as they are as they are incorporated incorporated in in series series into into the the logic logic scheme with scheme with the the radiation radiation monitOring.
monitoring.
2,12 2.12 RB Ventilation COOling RB Ventilation Cooling Unit 11 (Figures Unit (Figures 10, 10, 11, 11, 12) 12)
The Unit The Unit 1I Reactor Reactor Building Building Ventilation Ventilation Cooling Cooling System System isis aa closed closed loop loop that that circulates chilled circulates chilled water from two chillers and two chillers and t'.vo to pumps pumps through through aa heatheat exchanger and exchanger and two setssets of cooling coils.
of cooling The system coils The system isis used used to to cool cool methe Unit Unit 11 Reactor Reactor Building, Building, Radwaste Radwaste Building, Building, andand the Unit Unit 11 or or Unit Unit 22 Drywell Dr/Nell Ventilation Systems. The chillers [1-VA-1A(1B)-CHU-RB]
Ventilation [1 -VA-1A(i B)-CHU-RB] and pumps pumps [1-VA-11-VA-1A(1 B)-CHU-RB-PMPJ 1A(1 B)-CHU-RB-PMP) are are located on on the Unit 11 Reactor the Unit Reactor Building Building RHR RHR HeatHeat Exchanger Room Exchanger Room roof.roof. The The Dry./Jell Ventilation Heat Dre1l Ventilation Heat Exchanger Exchanger (1-VA-DIN-(1-VA-DW-HTX) is located on the Radwaste Building loading dock.
HTX) dock The ReactorReactor Building cooling coils (1-VA-CLR-5095) are located located inin the Unit Unit 11 Reactor Building Air Intake Building Intake Plenum.
Plenum. The Radwaste Building COOling cooling coils [1-VA-
[i-VA 1A(1 B)(1 1A(1 B)(1 C){1 C)(1 D)-CHU-COll]
D)-CHU-COILJ are located in the Radwaste Radwaste Building Building Air IntakeIntake Plenum.
Isolation/manual control valves control the flow of chilled water from the Unit 11 Reactor Building Ventilation Cooling System to the Reactor Building, Radwaste Building, and Drrwell Dr,well Heat Exchanger.
Unit 11 RB Ventilation Cooling System 300 ton (1-VA-1B-CHU-RB) (1-VA-i B-CHU-RB) is a a 30X325 air cooled liquid chiller designed to produce 300 tons Carrier Model 30XI\325 of effective cooling capability. It is designed for operating wim with environmentally safe R-134a refrigerant refrigerant.
Other features include the following:
Omer
- an automatic start circuit for the chill water pump upon starting of the new chiller; me
- Interlock with the 2D RBCCW pump to provide shutdown of the pump on loss of the new chiller
- ,A.nAn interposing relay in the 2D RBCCW remote starter panel to alleviate low voltage at the pump contactors caused by the loog long loop..
control loop Flexible connections are installed in each of the chilled water supply and return piping connections to the chiller evaporator. The flex connections will ....'ill prevent the possibility of vibration vil>ration from the me piping damaging the chiller evaporator.
ISD-37.1 S0-37.1 Rev. 10 Page 20 of 70 I
REACTOR BUILDING REACTOR BUILDING HVAC HVAC RESTARTRESTART 1.0 1.0 CONDITIONS ENTRY CONDITIONS ENTRY
- As directed As directed byby Secondary Secondary Containment Containment Control Control Procedure, Procedure. EOP-03-SCCP EOP-03-SCCP OR OR As directed by As directed by Containment Containment and and Radioactivity RadioactMty Release Release Control, Control, SAMG-0.2 SAMG-02 2.0 OPERATOR ACTIONS ACTIONS NOTE:
NOTE: Manpower:
Manpower: 11 Control Control Operator Operator 11 Auxiliarl Auxiliarj Operator Operator 1I Independent Verifier Independent Verifier Special equipment: 2 jumpers (10 and 11) 11)
CO: 2.1. reactor building ventilation IF the reactor ventilation radiation radia:on monitors have beenbeen 0 high as indicated off scale high indicated on D12-RR-R605 D12-RR-R6O OR the reactor reactor building exhaust temperature building :eniperature has has exceeded 135"Fi35F (UA-03, 6-2),
THEN EXIT this this procedure.
orocedure.
CAUTION Installaion of the following Installation folloMng jumpers wil! will also inhibit the
- ie automatic autornaic start of SBGT SBST on reactor low water level and on high dry wwell 1
dr ell pressure.
pressure 2.2 INSTALL the follo'follciMng
....;ng jumpers to bypass the reactor law water wa:er level and drywell drwell high pressure interlocks:
CO: - Jumper 10.10 in Panel XU-27, Terminal Board E, light side of right E. from the Terminal 28 to the right side of Terminal 30 ofTerminal28 oE CO: - Jumper 11 in Panel XU-2B, XU-28, Terminal Board E, E. from the light side of Terminal 28 to the right side of Terminal 30 right o
jIOEOP-01-SEP-0.4 OEOP-0l-SEP-04 Rev.
Rev. 1212 Page 2of6 20f61
2O 2.0 OPERATOR ACTIONS OPERATOR ACTIONS 2.7 2.7 OPEN the OPEN following valves:
the following valves:
CO:
CO: - RB VENT/NBD/SOL RB VENTINBDJSOL VALVES, SUPPLYA-BF!
SUPPL Y A-BFIV-RBV-RB VALVES EXHAUSTC-BFIV-RB, EXHA UST c-SF!v-na o CO:
CO: - RB VENT RB VENT OTBD SUPPLY SUPPL OTBD !SOL B-SF!V-RB Y B-BFIV-RB 1SOL 11.4LVES, V4LVES, EXH.4UST EXMUSTD-BRV-RR D-BFJ'V-RB, oU NOTE:
NOTE: In order In order to to start start aa reactor reactor building building supply supply or or exhaust exhaust fan, fan, the the control control sWitch switch should be should held in be held in START START untiluntil the the discharge discharge damperis full open.
damper is full open CO:
CO: 2.8 2.8 START as START as many many reactor reactor buifding bui[ding exhaust exhaust and and supply supply fans fans asas 0U possible to possible provide maximum to provide maximum ventilation ventilation {OP-37.1}.
(OP-371).
2.9 2.9 IF PROCESS IF PROCESS RX RXBLDGBLDG VENTVENTRADH!-Hlannunciator(UA-03 RAD HI-HI annunciator (UA-03 3-5) (alarm 3-5) setpoint at (alarm setpoint rnRihr) OR at 44 mRJ'hr) OR RX BLDG BLDG VENT VENT TEMPTEMP HiGH annunciator (UA-03 HIGH 6-2) (alarm (UA-03 6-2) <alarm setpoint at 135 1 35SF)
Q F) is is received, THEN:
CO:
CO: a. ENSURE reactor building exhaust and supply fans are off. 0
- b. ENSURE the following valves are closed:
CO:
CO: -
RB VENT tNBD !NBD lSOL JSOL I/ALl,.'ES, V4LVES, EXHAUST EXMUST 0 C-BRV-RB, SUPPLY A-BFIV-RB C-BFIV-RB, A -BF!V-RB CO:
CO: - VENT OTSD 1SOL VALVES, EXHAUST RB VENTOTBD/SOL 0 D-BFIV-R& SUPPLY B-BFIV-RB D-BFIV-RB, CO:
CO: c. ENSURE the SBGT System has initiated (OP-10). (OP-b). 0 OEOP-O1-SEP-04 IOEOP-01-SEP-04 Rev.
Rev. 1212 page40f61 Page4of6
Unt2 Unit:?
APP UA-03 APP UA-03 6-2 6-2 Page 1 of Page'1 of 1I RX BLDG RX BLDG VENT VENT TEMPTEMP HIGHHIGH AUTO ACTIONS AUTO ACTIONS t
- 1. Reactor Building Reactor Building ventilation ventilaton system system trips trips and and isolates.
isolates.
2.
- 2. Standby gas Standby gas treatment treatment trains trains start.
start.
3.
- 3. IfIf open, open, the the inboard inboard and and outboard outaoard primary containment purge pnniarj containment purge and and vent vent valves valves close.
close.
4.
to torus close.
CAUSE CAUSE I.
- 1. High temperature High temperature in in tile Reactor Building the Reactor Building exhaust plenum, 13S~F.
exhaust plenum, i35F.
2.
- 2. malfunction, Circuit malfunction.
Circuit OBSERVATIONS 1.
- 1. Building air Reactor Building air temperature monitor indicates greater greater than 13S'F 135SF on on Panel Panel xU-3.
XU-3.
2.
- 2. RX BLDG RX ISOLATED (2APP-UA-DS BLDG ISOLATED (2APP-UA-05 6-10) alarm.
6-IC) alarm.
ACTIONS t
- 1. Verify auto actions.
- 2. If entry conditions are met, enter if enter DEOP-03-SCCP, OEOP-03-SCCP, Secondarl Secondary Containment Control.
- 3. If entrl If entry conditions are met, met, enter DEOP-04-RRCP, OEOP-04-RRCP, Radiological Release Control.
- 4. OAOP-05.0, Radioactive Refer to OAOP-OS.O, Radioactive Spills, HighHigh Radiation, Radiation, and and .4.irbome Airborne .A.ctivity.
Activity.
- 5. If a circuit malfunction is suspected, ensure that a DLE If CLE is submitted.
DEVICE1SETPOINTS DEV1CEISETPOINTS D12-TS-NOIOAIB Rad Monitor D12-TS-N010AfB I35F POSSI8LE PLANT EFFECTS POSSIBLE Possible release to environs in excess of ODCM limits. linrits REFERENCES 1.
- 1. LL-9353-30 LL-9353-3D
- 2. LL-93053-32
- 3. OAOP-05.0, Radioactive Spills, High Radiation and Airborne ActMP; OAOP-OS.O, Activity
- 4. OEOP-03-SCCP, OEOP-03-SCCP. Secondary Containment Control Procedure
- 5. OEOP-04-RRCP, OEOP-04-RRCP, Radiological Release Control Procedure
- 6. ODCM 7.3.7
- 7. APP UA-0S UA-OS 6-10, Rx Bldg Bldg Isolated 2APP-UA-03 12APP-UA-03 Rev. 46 Rell.46 Page 55 of 63 Page 63\
Categories Categories KJA:
KIA: 295032 K2.02 Tier!
Tier / Group: T1G2 TlG2 RO Rating:
RORating: 3.6 3.6 SRO Rating:
SRORating: 3.7 3.7 LP Obj:
Obj: CLSLP3O0K CLS-LP-300-K*011 *01 1 Source:
Source: NEW Cog Level:
Cog Level: HIGH IllGH Category Category 8:8: Y
- 58. Following aa Reactor
- 58. Following Reactor Scram Scram on Unit Two on Unit Two due to aa loss due to loss of of Off-site Off-site power power (LOOP)
(LOOP) the the following conditions exist:
plant conditions following plant exist:
AREA RAD AREA RAD RX RX BLDG BLDG HIGH HIGH In alarm In alarm SOUTH RHR SOUTH RHR RM FLOOD LEVEL RM FLOOD LEVEL HI HI In alarm In alarm SOUTH CS SOUTH CS RM FLOOD LEVEL RM FLOOD LEVEL HI HI In alarm In alarm Reactor Building Reactor Building 20' 20 Rad Level Rad Level Approaching Max Approaching Max Norm Norm Operating Operating Rad Rad Reactor Building Reactor Building 20' 20 Temperature Temperature Approaching Max Approaching Max Norm Norm Operating Operating Temp Temp Based on Based on the the conditions conditions above above which which one one of of the the following following identifies identifies the the operator operator action action required lAW required lAW SCCP?
SCCP?
A. Open seven ADS A. Open ADS valves.
B Reset B:' Reset RPS RPS lAWlAW LEP-02.
LEP-02.
C. Rapidly depressurize C. Rapidly depressurize the RPVRPV to the main main condenser.
D. Isolate the RWCU system prior to reaching Max Safe Operating Temp.
Feedback Feedback K/A: 295033 KIA: 295033 A A1.05 1.05 Ability to Ability to operate operate and/or andlor monitor monitor the the following following as as they they apply apply to to HIGH HIGH SECONDARY SECONDARY CONTAINMENT AREA CONTAINMENT AREA RADIATION RADIATION LEVELS:
LEVELS:
Affected systems Affected systems soso as as to isolate damaged to isolate damaged portions portions (CFR: 41.7/45.6)
(CFR: 41.7/45.6)
RO/SRO Rating:
RO/SRO Rating: 3.9/4.0 3.9/4.0 CLSLP300M*08a Objective: CLS-LP-300-M*08a Objective:
- 8. Given
- 8. Given plant plant conditions conditions and and the the Secondary Secondary Containment Containment Control Control Procedure, Procedure, determine determine ifif any any of of the the following are following required:
are required:
Manual reactor
- b. Consider Anticipation of of Emergency Depressurization
- c. Emergency Depressurization
Reference:
Reference:
RSP, SCCP, 001-37.9 Cog Level: High Explanation:This meets the KA due to having to reset RPS to isolate the affected system (SDV leaking) thereby closing the scram valves which are the source of the leak causing the high rad levels in the reactor building.
Reactor Scram due to LOOP providing indications of SDV rupture.
The Maximum Normal Operating Values are the highest radiation levels expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
The Maximum Safe Operating Values are the radiation levels above which personnel access necessary for the safe shutdown of the plant will be precluded. These radiation levels are utilized in establishing the conditions under which reactor depressurization is required. Separate radiation levels are provided for each Secondary Containment area.
Flood level Hi is MNOWL entry condition to SCCP. LOOP automatically provides Groups 1,2,3,6,8, 1,2,3,8,8, & 10 isolations. No RWCU (Grp 3) isolation failure provided in stem. Based upon flood level status, along with rising 20 20' temperature and radiation leads to SDV rupture.
2 areas above MNOWL with primary system discharge requires Reactor Scram, cooldown <100 deg/hr, and consideration for anticipation of ED. No areas have reached Max Safe Operating Values, Emergency Depressurization is not required.
RPS can be reset and SCCP directs isolating the primary system discharge, main condenser not available due duetoto LOOP Distractor Analysis:
Choice A: Plausible because due to areas above MSOWL with aa primary system leak requires ED.
Choice B: Correct Answer Choice Choice C:C: Plausible because because primary system leaking to secondary containment with conditions primary system conditions degrading degrading would lead would lead RO RO to consider consider anticipating anticipating ED, ED, however however with aa LOOP, LOOP, the main main condenser condenser is is unavailable unavailable and therefore not not allowed.
Choice Choice D:D: Plausible because because due due to the location location of leak could of the leak be from the RWCU system, could be system, however however incorrect incorrect to assume failure of to assume Group 33 isolation of Group isolation from from LOOP.
LOOP. Current Current conditions, conditions, normal normal cooldown cooldown to reduce the to reduce the leak leak rate is appropriate, rate is appropriate, however the main however the main condenser condenser isis not not available available and therefore and not allowed.
therefore not allowed.
SRO Only Basis:
SRO Only Basis: N/A N/A Notes Notes
TABLE 3 AREA RADIATION LIMITS z 1 Cl, I
PLAHT PLAHT LOCAlKlN MAX tIOIliI MAX SAFE
-A DESCflIIP1'IOH CHANIIEL OPER.OJING OPBlPJ'1NG VALIJIE QmRi'HR) VALIJIE QmRi'HR)
HCORE SPRAY H IXI!ESPRAY ROOM 15 2.<<1
.,. 7_
SCORE SCORESPRAY 1& 2.<<1 .,. 7000 SPRAY ROOM H_ H_
ROOM 17 2.<<1 .,. 7000 S_
SRHR ROOM lS 2.<<1 .,. lOCO HPCI HPClROOM H.'A HIA .,. lOCO HACROSS FROM TIP ROOM 19 RX DRYWELL 20 BI..DG ENTRAHCE 80 .,. 2COO 20FT REV DECOH ROOM 12 RAIILROAD DOORS
- n SAMPLE RXBI..DG 24 STPJ'1OIIi 50FT 80 .,. 2COO RXIlLDG
.,. 7_
REV 25 AIRLOCK HOFFIJIEL 27 80 STORAGE POOL RX BLDG BETWEEHRX 7_
117FT 2S 1000
&FIJIELPooL ELEV CASKW>t.5H AREA 29 90 .,. 7000 RXBLDG SPENTFIJIEL .,. lOCO lO 90 80FTELEV COOLING SYSTEM
.,. x w CONTACT E&RC TO DETERMINE IF MAX SAFE OPERATING VALIlE 15 EXCEEDED w C
<UI TABLE 1 UI UI AREA TEMPERATURE LIMITS I -J UI U, I
PLANT PLANT IoIAXIIORIt IoIAXSAFE AIFTO AREA LOCA"IK>>I OPERA'lIIIG OP&I.<mHG GROUP Il£SCIlIP"IK>>I VALIIE('F) VALIIE('F) ISOL (HOTEl)
HCOR£ HCOR£ 121) 175 SPRAY SPRAY ROOK R'A SCOR£ SCOR£ 121) 175 SPRAY SPRAY ROOK R'A R'M:U PIIIIIP ROOIolA R'M:U PIIIIIP 1411< .uS R'M:U l R00I0I6 RWCUHX ROOIoI H_ 175 295 H.'A H_
EQUIP ROOM S_
175 295 III/A EQUIP ROOM SRHR RaCEQU¥P If,5 2'i$
ROOK $
HPQ HPClEQUIIP 165 165 4 ROOK RaCS'IM liM 295 5 STEAK 1UNIiEL TUMEL HPQS'IM liM 2'i$ 4 TlIMEL 2I)FTHORlH 2I)FT 1411< 1M R'A 2I)FT SOUTH
~FTtffl
~FT 1411< 1M R'A
~FTSE IoIIH.TI'LE R£ACTOR AR£A!I AI.AIIII BlDG ANNUNaATOR SETPOINT R'A 1. 4,AHD.'OR 5
'-Gl:S.7 WSIV R£ACTOR PIT AI.AIIII BlDG ANNUNaATOR SETPOINT H.'A 1
'-066-7 HOTEl: IoIAX IIORltOPERA'lIIIG VALlIE III THE ANNUNaATCft'GROUP 1Il000'IIIlH SETPOINT _EAPPLICAI!LE
TABLE 4 AREAWATER AREA WATER LEVEL LEVEL LIMITS LIMITS
-PLI4T PIJIJIT AREA EcOEE5AV IlCOI!£SmAY MAX I4CWd MAXNOHM A14G OPl3RAl'IIIG VAUJE(lImE VALIJE 4OIE1)1) 4CRE)
';aHCIES)
MAA5AfE MAX A OI'ER.6l'IIIG VALIJE(HOTE VAUE(ICTE2)L (IIICIES)
CRE3) 12 12 SCORE RAX SCOR£SPRAY 12 6
HEIRt IIRHR 6e 12 12 SRI 5RHR 12 12 6
CI HPCI 12 66 1: RI'tI NOTE 1:
HOlE 2:
NOTE IWFLOcC 2 RI'tIFLOIlO RMFLOO LEVEL LEV HI FLOIlO LEVEL I AIIIItJlICI.OlTORlIIllfCATES LF. IIJI.
I HI MREHCLETORLHf2CAES ,6 IIICIES IR AIIIIIJIICJIO.Tat ARRUHCTOR IIIIlfCATES
_EiR ICRES WIE LEVEL LEL E4ECTE 12I11CIllESWATERLEVEL 12E4C3 WAlER LEEL
MAX NORM MAX NORM MAX MAX NORM NORM
",nmill. ...ti.'"
CGCIiNlGING1NTOTmr JlmAII!)LCEPT lSH1tIIS
- ElfERaEHCVOPII!IUaWG TQUOPlUtRmBYAN f* ***,"E"""'"
roRDJlWAGIZ'CamooL SCCp..14 F~,=,r::~~mcD t
I!J\'(ilMEIUl'GTO E'MUi'A11! IQE""'.U)PI!
I SCC~f:S ccts
mx MAX SAFE SCCp.,16
, IMTW'1t AIUACTOR SC:RAII AM)DlTmEOP~Ot
'I
> rn MORE THAN 1 ABOVE rn MAX SAFE or EQ rn R i 1EIIm.<.ZHCycc:nUUlIlJRRE:
Im 1JC REACTOR Pll!R:TJC RCI'SECnONoarEOP-Ot tlCCPl-l'G z 0 I; BNP VOL- VI OEOP SCCP C I,, 0 C Cl, C.) C) -o Cl, REVISION NO: 7 0z z 0 -4
/\ATTACHMENT TTACHMENT 11 Page 19 Page l9ofof 101 101
[PG Plant EPG Plant Specific Specific Technical Technical Guidelines Guidelines RCIP Monitor RGJP Monitor andand control control reactor reactor vessel vessel pressure pressure while executing IfIf while executing thethe follO'lving following steps:
steps:
- AA high high drywell drjwell pressure pressure EGCSECCS initiation initiation signal signal (1.7 (1.7 psig psg drwell pressure
[drywell pressure which which initiates initiates EGCS])
ECCSI) exists, exists. prevent prevent injection from those Core Spray and RHR injection RHR pumps pumps notnot required to assure adequate core cooling prior required prior to depressurizing belowbelow their maximum injection pressures.
pressures.
Emergency Depressurization
- Emergency Depressurization is is anticipated anticipated and and either either all all control rods control rods are are inserted inserted to to or or beyond beyond position 00 (Maximum position 00 (Maximum Subcritical Banked Withdrawal Position) or it has been determined that the reactor will remain shutdown under all boron. rapidly conditions without boron, rapidly depressurize the reactor vessel with the main turbine bypass valves, irrespective of the resulting cooldown rate.
resulting
- Emergency Depressurization is or has been required, enter Emergency Procedure Guideline Contingency #2.
- Reactor vessel water level cannot be determined, enter Emergency Procedure Guideline Contingency #4. #4 RCIP-1 RGlP-1 cycling. manually open SRVs until reactor vessel If any SRV is cycling, pressure drops to 950 psig (reactor vessel pressure at which all turbine bypass valves are fully open).
j1001-37 001-37 Rev. 54 Page 26 of 239 Page 2391
STEP SCCP-16 STEP SCCP-16 ISA PRMARY NO STEP BASES:
BASES:
Primary systems comprise the pipes, valves, and other equipment which connect directly to the reactor such that a reduction in reactor pressure v,111 ll effect a decrease in the flow of steam or water being discharged through an un isolated break in the system.
unisolated If a primary system is discharging into the Reactor Building when this step of the procedure is reached, one of three conditions must exist procedure exist:
- A primary system break cannot be isolated because system operation is required A primar/
cooling or shutdown the reactor.
to assure adequate core coolillg
- valves No isolation val ....es exist upstream of a primary system break, or if isolation valves do exist, they cannot be closed because of some mechanicalleleetricat(pneumatic faifure.
mechanical/electrical/pneumatic failure.
- The source of the discharge cannot be determined.
Another criteria which may be used in the case of unknown sources, source& *'Is Is the leaking coolant? If not, then no primarl water reactor coolant?" primary system is discharging. SpeCifically, Specifically, SRVs being open with a a rupture oHIle of the suppression pool is not a primarl primary system leaking to secondary containment. Suppression pool water does not fit the description of having come from aa primary system as aa reduction in reactor pressure does not result in aa reduction in the suppression pool water leak rate rate.
The subsequent steps provide instructions to shutdown, shutdn, scram, or rapidly depressurize the reactor based upon the source of heat addition to the Reactor Building. Discharge of steam or water from aa primayprimary system requires that the operator take actions in accordance with subsequent Steps. steps. IfIf the heat addition to secondary secondarJ containment is from aa source other than aa primary system discharging into an area, appropriate operator actions are directed.
001-37.9 1001-37.9 Rev.
Rev. 11 Page 31 Page 391 of 39 31 of
STEPS RC/P-05 STEPS RCIP-05 through through RC/P-11 RC/P-1 I RAPIDLY DEPRE5WR!lE tHt! R!ACTOR WItH llll! MAlI'!
10RI!IINf. 6YPI\$!; ~Al\'e$
UUI&SIlECTII/F. OF THE RESULTiNG COOLDOfIN RATE
' - - - - - - , . . . . - - - - " RC'P.1f STEP BASES:
As conditions which will require Emergency Depressurization are approached, it is appropriate to rapidly reject as much heat energy as possible from the reactor vessel to a heat sink other than the Suppression Pool. Such action preserves the heat capacity of the Suppression Pool for as long as possible, until aa requirement for Emergency Depressurization actually exists.
001-37.4 1001-37.4 Rev Rev. 88 Page 51 Page 51 of of7B 78 I
STEPS RC/P-05 STEPS RCIP-05 through through RC/P-11 RC/P-1 I (continued)
(continued)
The term The Anticipated implies term "Anticipated" implies an an expectation, expectation, based based on on an an evaluation evaluation of of plant plant conditions and conditions and extrapolation extrapolation of of parameter parameter trends, trends, that that an an Emergency Emergency Depressurization Depressurization requirement will requirement will soon soon I)e reached and be reached and cannot cannot be be aaverted
....erted by by actions actions prescribed prescribed in in the the EOPs. Before this EOPs. Before this conclusion conclusion cancan bebe drawn, however, the drawn, however, the effectiveness effectiveness of of the the steps steps preceding the preceding the depressurization depressurization requirement requirement must must be be eevaluated.
....aluated.
appropriate EOPs The appropriate EOP5 contain contain notes notes stating stating to "Consider Anticipation of Comsider AnticipaUoli of Emergency Emergency Depresurzation ....
Depressurization there has IfIf there has not not been been fuel fuel failure failure indicated indicated by an Abnormal by an Abnormal Core Core Conditions Conditions andand Core Core Damage Unusual Event EAL Damage EARL classification or a steam line break, break, and the the main condenser is is available as aa heat sink, discharging reactor steam to the main condenser through the main tumine turbine bypass valves valves is the most ....viable is the iable method method of rapidly reducing reactor pressure reactor pressure without adding adding heatheat to to the Suppression Pool.
the Suppression Pool. Other Other mechanisms mechanisms have have less heat less removal heat remo ....al capacity, take longer longer to establish establish the approprtate appropnate valve lineup, lineup. etc.
The anticipatory depressurization prescribed by' by this critical step is permitted only if the reactor will remain shutdown under all possible conditions of coolant temperature and boron concentration. Bypassing or defeating isolation interlocks is not authorized in these steps.
The depressurization is performed "irrespectiveirrespective of the resulting cooldown rate" rate since the need for rapid depressurization takes precedence over nomlal nomial cool down rate limits.
cooldown If the rapid depressurization is not performed, emergency depressurization would soon be required.
The caution detailing the restrictions on steam flow for Unit 2 which may cause a PC PCIS IS Group 11 isolation is added because it is not desirable to purposely cause an MSI'v' MSIV isolationwhen isolation* When the MSIVs are open. This equipment design does not apply on Unit 1, 1 therefore this step does not exist on the Unit 11 flow chart, resulting in step numbering differences between the Unit 1I and 2 flowcharts from Step RCfP-10 RC!P-10 through RCJP-16.
RC1P-16.
001-37.4 1001-37.4 Rev.
Rev. B8 Page 5252 of 78 781
STEP SCCP-24 STEP SCCP-24
/ ~
NQI I CONSIDERANllCIPAlIo.'4 CONSIDER AN1ICIPAHON OFCF JEMEROENCY DEPRESSURIZATIGN EMERGENCY DEPRESSURIZATION t PER RC$P seCTION PER ReI? SECTION OFQI
\ REACTOR VESSEL
- REACTOR VESSEL
\ CONTROL CONTROL PROCEOURE" PROCEDURE (EOP*01- RYe?)
CEQP-O1-RVCP)
STEP BASES:
As conditions As conditions which will require Emergency Depressurization require Emergency Depressurization are are approached approached itit is is appropriate to appropriate rapidly reject to rapidly reject asas much much heat heat energy energy asas possible possible from the reactor reactor vessel vessel to aa heat sink other than than the suppression pool. Such action preserves the heat capacity of the suppression pool for as long as possible, until until a requirement for Emergency Depressurization actually exists.
Discharging reactor steam to the main condenser through the main turbine bypass Dlscharging bypass valves isis the most viable method method of rapidly reducing reactor pressure without adding heat to the suppression pool. Other mechanisms have less heat removal capacity, take longer to establish the appropriate valve lineup, etc.
Anticipate Emergency Depressurization" In order to ~Anticipate Depre.ssurization per EOP-01-RVCP, the following conditions must I)e be met met:
- a. The MSIVs must be open.
b.
Il. The main turbine bypass valves must be operational.
- c. The main condenser must be available.
- d. The reactor must remain shutdown for all possible conditions of coolant temperature without boron.
Bypassing or defeating Bypassing defeating isolation interlocks is not authorized durin duringg ~Anhicipate Anticipate Depressurzation actions.
Emergency Depressurization" j1001-37.9 001-37.9 Rev Rev. 11 Page 36 391 36 of 39 Categories K/A:
KIA: 295033 A1.05 Al.05 Tier/Group:
Tier / Group: TlG2 T1G2 RO Rating:
RORating: 3.9 3.9 SRO Rating:
SRORating: 4.0 4.0 LP Obj: CLSLP300M*08A CLS-LP-300-M*08A Source:
Source: NEW NEW Cog Cog Level:
Level: HIGH HIGH Category Category 8:8: Y
- 59. Which one
- 59. Which one of the following of the following annunciators annunciators indicates indicates aa condition condition that that trips trips the the Reactor Reactor Building Supply Building Supply and and Exhaust Exhaust Fans Fans without without automatically automatically starting starling the the Standby Standby Gas Gas Treatment System?
Treatment System?
A. AREA A. AREA RAD RAD RX RX BLDG BLDG HIGH HIGH B. RX B. R)( BLDG BLDG VENT VENT TEMPTEMP HIGH HIGH C RX C:' RX BLDG BLDG DIFF DIFF PRESS PRESS HIGH/LOW HIGH/LOW D. PROCESS RX RK BLDG VENT VENT RAD HIGH Feedback Feedback K/A: 295035 KIA: 295035 A1.01 Ability to operate and/or andlor monitor monitor the following as as they apply apply to to SECONDARY SECONDARY CONTAINMENT CONTAINMENT HIGH DIFFERENTIAL PRESSURE:
Secondary containment ventilation system (CFR: 41.7 1/ 45.6)
RO/SRO Rating: 3.6/3.6
- 06a CLS-LP-037. 1 *06a Objective: CLS-LP-037.1
- 4. List the signals and setpoints that will cause the Reactor Building Ventilation System to automatically isolate.
- 6. List the signals that will cause the following to automatically stop:
- a. Reactor Building supply fans
- b. Reactor Building exhaust fans
Reference:
I (2)OP-, Revision, 1 Revision Page, Section Cog Level: Low Explanation:
Rx Bldg ARMs have no automatic function. Vent Temp high will isolate secondary containment and initiate SBGT. Process Rx Bldg vent rad hi is entry into SCCP but does not trip fans (Hi-Hi isolates secondary containment and initiates SBGT). Fans trip on excessive building differential pressure (+4 inches or -4 inches) but SBGT does not start (APP UA-12 3-3)
Distractor Analysis:
Choice A: Plausible because is confused with Exhaust Rad Hi which will trip RB HVAC & auto start SBGT.
Choice B: Plausible because high RB vent temperature is not required by TS, but but does trip RB HVAC &
auto start SBGT.
Choice Choice C: Correct Answer Choice Choice D:0: Plausible because Process RB Vent Rad high high is SCCP entry condition is aa SCCP condition and and is is easily easily confused with Process RB Vent with Process Vent Rad Hi-Hi which does trip which does trip RB RB HVAC HVAC && auto auto start start SBGT.
SRO SRO Only Basis: N/A Only Basis:
Notes Notes
Unit2 Unit 2 UA-03 2-7 APP UA-03 Page 11 of 2 AREA RAD RX BLDG HIGH AUTO ACTIONS NONE Unit2 Unit 2 APP UA-03 6-2 Page 11 of 11 RX BLDG VENT TEMP HIGH AUTO ACTIONS 1.
'1. Reactor Building ventilation system trips and isolates.
- 2. Standby gas treatment trains start.
- 3. If open, the inboard and outboard primary containment purge and vent valves close.
- 4. close.
PASS sample valves to torus dose.
Unit tJrit 22 APP UA-12 APE> JA-2 3-3 Page 1 Cll:
cf 1 DZFF PRESS HIGH/LOif RX BLDG DIFF HIGH/LOW
{Reaccor
{RacEor Building Diffe:renr;ial Differential Pressure High/Lo,1) f1gh/Lcw)
AtfTO AuTO ACT IOS ACTIONS
- 1. Re:actor Building supply and exhaust Reactor e:xhaust fans trip crip.
Unit2 Unit 2 APP UA-03 4-5 Page 1 1 of of'lI PROCESS PROCESS RX BLDG VENT RAD HIGH AUTO ACTIONS NONE
Categories Categories K/A:
KIA: 295035 AA1.01 295035 1.0 1 Tier/Group:
Tier / Group: T1G2 TlG2 RU Rating:
RORating: 3.6 3.6 SRO Rating:
SRORating: 3.6 3.6 LP Obj:
LP Obj: CLSLP037.1*06A CLS-LP-037.l *06A Source:
Source: BANK BANK Cog Level:
Cog Level: LOW LOW Category 8:8:
Category YY
- 60. Which one
- 60. Which one of of the the following following identifies identifies positive positive reactivity reactivity effects effects that that SLCSLC injection injection mustmust overcome during overcome during anan ATWS?
ATWS?
100% Voids A. 100%
A. Voids and and Iodine Iodine decay decay 100% Voids and Xenon decay B. 100%
B.
C. Withdrawn Control Rods Rods and IodineIodine decay D
D~ Withdrawn ControlControl Rods Rods andand Xenon decay decay Feedback Feedback K/A: 295037 KIA: 295037 K1.03 K1.03 Knowledge of the operational implications Knowledge implications of the following concepts concepts as as they apply apply to SCRAM PRESENT AND REACTOR CONDITION PRESENT REACTOR POWER POWER ABOVE APRM DOWNSCALE DOWNSCALE OR UNKNOWN: UNKNOWN:
Boron effects on reactor power (SBLC)
(CFR: 41.8 to 41.10)
RO/SRO Rating: 4.2/4.4 CLSLP005*03 Objective: CLS-LP-005*03
- 03. List the positive reactivity effects that must be overcome by SLC injection.
Reference:
SD-5, Revision, Page 7, Section 1.3 Cog Level: Low Explanation:
Requires understanding of conditions which cause positive and negative reactivity effects.
When the contents of the storage tank have been injected into the reactor vessel, a specified minimum average concentration equivalent to 720 ppm natural boron provides an adequate shutdown margin to compensate for the positive reactivity effects of xenon decay, zero percent voids, reduced Doppler effect and moderator temperature decrease to 70°F, and control rods fully withdrawn. Unless procedures direct otherwise, the total contents of the storage tank should be injected anytime the system is needed to ensure sufficient neutron absorber is injected to maintain the reactor shutdown during the cooldown.
Positive Reactivity effects overcome by SLC:
1.0%
1.0% Voids
- 2. Xe Decay
- 3. Moderator Moderator Temperature
- 4. Reduced Doppler Doppler
- 5. Control Rods not inserted Distractor Analysis:
Choice A: Plausible because because 100%
100% voids provides for negative negative reactivity reactivity effect which which isis easily easily confused with 0% voids, voids. and Iodine decay decay produces Xenon which also provides negative negative reactivity reactivity effect.
Choice Choice B: Plausible because 100% voids provides for significant because 100% significant negative negative reactivity and Xenon decay decay isis correct providing positive correct providing reactivity effect.
positive reactivity effect.
Choice Choice C:C: Plausible Plausible because because withdrawn withdrawn control control rods rods is is correct correct providing providing positive positive reactivity reactivity effect effect and and Iodine Iodine decay decay produces Xenon which produces Xenon which provides provides negative negative reactivity reactivity effect.
effect.
Choice Choice D:D: Correct Correct Answer Answer SRO SRO Only Only Basis:
Basis: N/A N/A Notes Notes
When the When the contents contents of the storage of the storage tank have been tank have been injected injected into into the the reactor reactor vessel, aa specified vessel, specified minimum minimum averageaverage concentration concentration equivalent equivalent to to 720 720 ppm ppm natural boron natural boron provides provides an an adequate adequate shutdown shutdown marginmargin to to compensate compensate for for the the positive reactiviPf positive reactivity effects effects of of xenon xenon decay, decay, zero zero percent percent voids, voids, reduced reduced Doppler effect Doppler effect and and moderator moderator temperature temperature decrease decrease to to 70 oF, and 70°F, and control control rods fully withdrawn, Figure 05-4. Unless rods fully withdrawn, Figure 05-4. Unless procedures procedures directdirect otherwise, otherwise. the the total contents total contents of of the the storage storage tanktank should should bebe injected injected anytime anytime the the system system isis needed to needed to ensure ensure sufficient sufficient neutron neutron absorber absorber is is injected injected to to maintain maintain the the reactor shutdown reactor shutdown during during thethe cooldown.
cooldown.
The SLC The boron solution SLC boron solution storage storage tank, tank. test test tank, tank, the the two two positive positive displacement pumps, displacement pumps, Squib Squib valves, valves, andand the the associated associated local local valves valves and and controls are controls are iocated located on the 80 on the 80 foot, east elevation foot, east elevation oftheof the reactor reactor building.
building.
The SLCSLC System System solution is is discharged discharged intointo the the reactor reactor vessel near near thethe bottom of the core bottom core shroud wherewhere itit mixes mixes with rising rising coolant, coolant, enters the core and absorJ)s and absorbs thermal thermal neutrons neutrons to to shutdown the the reactor reactor by by temlinating terminating the nuclear fission nuclear fission chain reaction, Figure chain reaction, Figure 05-5.
05-5.
2.0
2.0 DESCRIPTION
/DESIGN DATA COMPONENT DESCRIPTIONIDESIGN 2.1 SLC Storage Tank (Figure (Figure 05-6) (Figure (Figure 05-10)
The SLC Storage Tank is located on the 80-foot elevation of the reactor building. This storage tank provides a reservoir for preparing, storing, and maintaining the sodium pentallorate pentaborate solution in a state of constant readiness. It has a top hatch for the addition of chemicals, chernical& an air sparger for mixing of the solution, solution. two submersed electric heaters and assorted system monitoring instrumentation. The tank has a bubbler tube for level indication.
monitOring 2.2 Air Sparger An air sparger is provided in the SLC storage tank for mixing the solution in the tank. The air sparger is supplied from plant Service Air.
2.3 Heater The heater system normally maintains the solution temperature bePNeen between 66F and BO"F.
66*F 80F.
There are two electric heaters in the storage tank. The A A heater is 10KW and has both automatic and manual contro/s. controls. The B B heater is 40 KW manual control only. The BB heater is used to heat the solution in the tank during chemical addition (endothermic reaction) and to backup the A heater.
1SD-OS 80 -05 Rev. 77 Page Page 770f431 of 43
1I1t1 U \,Ulildllllllt:IIL UJHLdII ILII IL COLD SHUTDOWN BORON WEIGHT The least weight of soluble boron which, if injected into the reactor and mixed uniformly, will maintain the reactor shutdown under all conditions. This weight is utilized to assure the reactor will remain shutdown irrespective of control rod position or reactor '.vater water temperature.
CONDENSATE SYSTEM For the purpose ofthis of this EOP, the Condensate System consists of a minimum of one condensate pump capable of injecting water into the reactor.
I OEOP-01-UG OEOP-O1-UG Rev. 55 Page 64 of 151 I
-'-*r***_** .. ~.- . . ___ .. u ; ~- . .... - - .... - _ ....... _ ............... - t""-"'~"t' '.,,_. - -t' HOT SHUTDOWN BORON WEIGHT The least weight of soluble boron which, if injected into the reactor and mixed uniformly, will maintain the reactor shutdown under tinder hot standby conditions. This weight is utilized to assure the reactor will be shutdown irrespective of control rod position when reactor water level is raised to uniformly mix the injected boron.
OEOP-O1-UG IOEOP-01-UG Rev. 55 Page 66 at of 151 I Categories K/A:
KIA: 295037K1.03 295037 K1.03 Tier/Group:
Tier / Group: T1G1 TIGl RO Rating:
RORating: 4.2 SRO Rating:
SRORating: 4.4 LP Obj: CLSLPOO5*O3 CLS-LP-005*03 Source: NEW Cog Level: LOW Category 8: Y
- 61. Unit
- 61. Unit Two Two startup startup is in progress is in progress with with both Mechanical Vacuum both Mechanical Vacuum Pumps Pumps (MVPs)
(MVPs) in in service service establishing main establishing main condenser condenser vacuum.
vacuum.
Which one Which one of of the following identifies:
the following identifies:
(1) When (1) When both both MVPs MVPs willwill automatically automatically trip trip and and (2) the (2) the reason reason this action action isis required?
required?
(1) Two MSL A. (1) MSL RadRad Hi-Hi Hi-Hi conditions inin one division (2) To reduce (2) reduce off-site release rates.
B. (1) Two MSL Rad B. Rad Hi-Hi conditions in one division (2) To minimize hydrogen explosion hazards.
C C~ (1) One MSL Rad Hi-Hi condition in each division (2) To reduce off-site release rates.
D. (1) One MSL Rad Hi-Hi condition in each division (2) To minimize hydrogen explosion hazards.
Feedback Feedback K/A: 295038 KIA: 295038 K2.1K2.100 Knowledge of Knowledge of the the interrelations interrelations between between HIGH HIGH OFF OFF SITE SITE RELEASE RELEASE RATE RATE and and thethe following:
following:
Condenser air Condenser removal.
air removal.
(CFR: 41.7 (CFR: 41.7 145.8)
/45.8)
ROISRO Rating:
RO/SRO Rating: 3.2/3.4 3.2/3.4 Objective: CLSLP30*1 lb Objective: CLS-LP-30*11 b 11 .Given the 11.Given the necessary necessary plant plant conditions, conditions, describe describe the effect that the effect that aa malfunction malfunction or loss of or loss of the the Condenser Condenser Air Removal/Augmented Air Removal/Augmented Off-Gas Off-Gas System System would would have have onon the the following:
following:
- b. Radioactive
- b. Radioactive Release Release Rates Rates
Reference:
Reference:
SD-30 SO-30 Cog Level: High Cog Level: High Explanation:
Explanation:
Hi-Hi trip Hi-Hi trip on on the the MSL MSL Rad Rad channels channels will cause cause both both Mechanical Mechanical vacuum vacuum pumps pumps toto trip trip and and OG-V7 OG-V7 valve valve to to close. The logic close. logic trips the MVPs MVPs when a hi-hihi-hi condition condition is is present inin each each division.
division. The MVPsMVPs discharge discharge via the 1.8 minute holdupholdup line to the main stack. MSL MSL rad high conditions directly impactimpact off-site release rates with the MVPs rates MVPs in service due to no no discharge path processing. MVPs path processing. MVP5 are only allowed to be operated below 5% reactor power due no hydrogen explosion hazards present.
Distractor Analysis:
Oistractor Choice A: Plausible due to two MSL Rad Hi-Hi in one division only satisfies inbd-otbd logic which is easily confused with coincidence logic and reducing off-site release rates is correct.
Choice B: Plausible due to two MSL Rad Hi-Hi in one division only satisfies inbd-otbd logic which is easily confused with coincidence logic and reducing hydrogen explosion hazards is the reason MVPs are not operated at >5% >5% reactor power.
Choice C: Correct answer.
Choice D: 0: Plausible due to the MSL Hi-Hi correct and reducing hydrogen explosion hazards is the reason MVPs are not operated at >5% reactor power.
SRO Only Basis: N/A Notes
3.2.2 3.2.2 Mechanical Vacuum Mechanical Vacuum Pump Pump Assembly Assembly Control Control (Figure (Figure 30-12) 30-12)
Operation of Operation of mechanical mechanical vacuum pumps (MVP) vacuum pumps (MVP) .A.A && BB are are controlled controlled by three by three Control Control Switches Switches (CS-354, (CS-354. CS-368, CS-368, and and CS-369)
CS-369) located located on on Control Room Control Room Panel XU-2.
Panel XU-2.
Control Switch Control Switch CS-354 CS-354 isis aa two to position position {CLOSE-HOG}
CLOSE-HOG switch switch that that controls the controls the position of Hogging position of Hogging Valve Valve V7.
V7. The The HOG HOG position position opens opens VT. starts V7, starts the the MVPs MVPs and and then then the the condensate condensate return return (seal)
(seal) pump.
pump.
CAUTION If control switches If these control are held in the switches are the start start position, position, the the MVP MVP will start with OG-V7 OG-V7 closed and closed, and when released released the associated associated pump pump immediately immediately stops.
stops Control Switches CS-368 and CS-369 are three position (STOP-N-START), spring return to Neutral t~'pe type switches that control the MVPs.
The mechanical vacuum pumps are placed in service by by placing CS-354 to the HOG position. This opens Hogging Valve (V7), starts both MVPs and their associated condensate return pumps.
There are three automatic trips associated with the mechanical vacuum pumps for each BNP unit unit. They are as rollows:
follows:
- Low MVP oil pressure (time delay for 3 seconds on startup) low
- Hi-Hi/INOP Main Steam Line radiation Hi-HiIiNOP
- Hogging Valve OG-V7 fully closed If both MVPs are not required, one pump may be stopped by momentarily placing the aSSOCiated associated control switch to STOP. This action breaks the seal in circuit and stops the pump.
A Main Main Steam Lineline high-high radiation or MSLRM MSlRM INOP condition will prevent the opening of the hogging valve OG-V7 and prevent start of the mechanical vacuum pump. If already running, the signal will trip MVP, shut the Hogging Valve, OG-V7, and stop the condensate the MVP.
return pump.
SO-3~
I SD-3D Rev.
Rev. 9 Page 31 of 100 I Pae31of100
4.1.3 4.1.3 Procedural Cautions Procedural Cautions 1.
- 1. No open No open flames flames or lighted objects or lighted objects should should bebe permitted permitted near near this this system or system or its its components components at at any any time time due due toto the the presence presence of of hydrogen. Hydrogen hydrogen. Hydrogen concentrations concentrations in in excess excess ofof 4%
4% present present aa significant explosion significant explosion hazard.
hazard.
2.
- 2. IfIf recombiner recombiner operation operation is is required required with with any any Irydrogenfoxygen hydrogen/oxygen analyzers inoperative, analyzers inoperative, the the recomblner recombiner temperature shall be monitored temperature shallile monitored closely and ifif temperature shows aa large large fluctuation during steady shutdown of the affected train state operations, shutdow'n train should be considered.
considered.
3.
- 3. Due to Due to the the analysis analysis method, method, the the H2/02
!0 analyzers H
2 analyzers dodo not not give give aa "real real time output, time" output, l)lit but take take aa sample sample andand update update the the outputs outputs for for H2 2 and H and O 022 concentration about concentration about once once every every 55 minutes.
4
- 4. When the Hzl02 ?0 analyzer keylock switch is in DATA ENTRY, the H
2 analyzer program can be inadvertently altered or the analyzer operation can be inadvertently faulted by improper operation of the analyzer keyboard. Faults Faults may not be noticeable noticeable from observation of analyzer operation.
5,
- 5. Operation of a SJAE in the warm-up mode with the OPPOSite opposite SJ,I\E SJAE train in service will cause erroneous SJAE activity indication due to sample flow dilution from the opening of the idle SJAE sample valve.
- 6. Mechanical vacuum pump operation when the reactor is producing more than 5% thermal power poses a potential hazard due to hydrogen explosion.
- 7. The Hydrogen Water Chemistry System does not have the capability for adjusting oxygen injection flow into a a SJAE that is being placed into service. To prevent high hydrogen concentration in the SJAE that is being started, the HWC HWC system must I)e be removed from service I SD-30 SD-3D Rev.99 Rev. Page 43 of 100 Page43of1OO I
SD-30 oen I
~6i)'/MCC W
HFIl o
Gs 7f- es l j, l sr.",:!
-'J ~t~ LE! s:
FIGURE 30-12
, (J)
("l Ii 42X 42:<1 ;j-l!>l
- r Q)
- l 3, ?;'
- M'.'P !:!..
CLoseD i];J,_I II II Ol~ CLOSED 'lei EN Li 1i 42 II
'1,HtN PRESS 1, "'/.lGGI~G <
tA','P ell 1121 SEC VAl"vt.vr NOT FULLY Q)"
("l_
l::G)
"RESS C'.. Ci!'~D 5c Rev.9 Li
~
>3 FSIG ::::1;::0 iii "tlm Mechanical Vacuum Pump Control Circuit
(!) 4~ l::w I\ 3?
'0 .....
nt'V
~
~
iJ Q.
n (iLç MVP ~.
l::
(-
r.r,.::, TO ) -!; S6:r=.
Page90of100 10 Jr (TIl 4.75 SEC.I Et>ERGIZES liFTER 5 SEC U ,
(J)
- -8
...g, oQ
INSTRUMENTNUMBER:
INSTRUMENT NUMBER: D12-RM-Kec2A,8,C,D D12-RM-I<802A,S.C.D INSTRUMENT NAME:
INSTRUMENT NAME: Main S:eam Main Swam LmeL:ne Radiation Rad tior/ HI-HI.'1NOP I-Il-HI:INOP TS
REFERENCE:
TS
REFERENCE:
3.3.7.2; TRM 3.3.7.2; TRM T.able Table 3.2.7.2-1.1 3.2.72-1.1 TRIP CH TRIP CHANNEL:
.... NNEL: ....A1-K33A
,-K'~03A 31-1<5238 51-1{603B
....A2-K603C 2-K603C 32-KEG3D 52-1<6030 TRIP LOGIC:
TRIP LOGIC: AAl or A2 1 or and 81 AC and B 1 or 82 == Trips cr82 Tripsboth both mechanical mechanical vaouum vacuum pumps pumps and and closes closes OG-V7 OG-V7 Place channel Place channel in kipped ooncttion in tripped corcdcn by: by: Pull fuse pun fuse CHANNEL CH .... NNEL INSTRUMENT NUMBER INSTRUMENT NUMBER TRIP TRIP ACTION ACTION PANEL PANEL FUNCTION FUNCTION SETPOINT SETFCINT UNIT UNIT Al Al 012-RM-KGO2A D12-RM-K602A N/A NI.4 A7i8lEA A71B-?=19A H12-Pf08 H12-P609 Mecnanical Vacuum Trips Mechanical Trips Vacuum Pumps Pumps and ann 2.8 2.3 :(x Background Bagrourd clos",.OG-V7 closes OG-V7 AC A2 D12-PM-X303C 012-RM-K603C N/A NtA A7IB-FIBC AtIB-FleC H12-P609 H12-PeOg Trips Tops Mechanical Mechanical Vacuum Vacuum Pumps and Pumpsand 2.3 xa Background 2.8 Background closes 0OG-V7 closes G-V7 Bl Bl D12-RM-I1803B DI2-RM-K6026 N/A N(.o, A718-F1PB A7IS-FI9B H12.Pd1 I H12-P611 Trips Mechanical Trips Mechanical Vacuum Pumps and Vacuum Pumps arId 2.8 2.Sxx Background
Background
close. OG-V7 closes OG-V7 BC B2 oD12-RM-K603O 12-RM-K6030 N/A Nl.A, ATIE-F1BD
... tIS-FISD H12-P611 H12-Pdll Trips Tnps Mechanica!
Mechanical Vacuum Vacuum Pumps Pumps and and 2.8 xa Background 2.8 Background closes closes OG-V7 OG-V7 COMMENTS:
COMMENTS: ifS both both channels channels in trip system in 3a nip are inop.
sbstent are both channels mop, both channels must must be tipped to be tripped to assure assure all all required requirec functions will occur.
functions wilt occur.
REFERENCE DRAWINGS:
REFERENCE DRAWINGS: FP-e3O6 FP-00066 001-18 1001-18 Rev. 58 page470f10BI Page47oj Categories Categories KJA:
KIA: 295038 295038 K2.10 K2.l0 Tier/Group:
Tier / Group: TIGl T1G1 RO Rating:
RORating: 3.2 3.2 SRO SRORating:
Rating: 3.4 3.4 LP Obj:
LPObj: CLSLP30*11B CLS-LP-30* lIB Source:
Source: NEW NEW Cog Level:
Cog Level: HIGH HIGH Category Category 8: 8: Y
- 62. Unit Two
- 62. Unit Two isis operating operating at rated power at rated power whenwhen plugging plugging of of C12-D006A, Cl 2-DOO6A, Supply Supply Air Air Filter, Filter, causes the causes the SCRAM SCRAM VALVE VALVE PIL PIL AIR AIR HDR HDR PRESS PRESS HIILO HI/LO alarm alarm to to be be received.
received.
Which one Which one ofof the the following following identifies identifies the the impact impact of of lowering lowering scram scram air air header header pressure pressure as the as the filter filter continues continues to to plug?
plug?
The CRD The CRD Scram Scram Outlet Outlet Valves Valves will will fail fail (1) (1) on on aa low low scram scram air air header header pressure pressure and and cause the cause the OWDW Lower Lower Vent Vent Dampers Dampers to to reposition reposition to to the the (2) (2) position.
position.
A (1)
A'I (1) open open (2) MAX (2) MAX B. (1)
B. (1) open (2) MIN (2) MIN C. (1) closed (2) MAX D. (1) (I) closed (2) MIN Feedback K/A: 300000 KS.13 KIA: K5.13 Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM:
Filters 41.5 / 45.3)
(CFR: 41.5/45.3)
RO/SRO Rating: 2.9/2.9 Objective: CLS-LP-08 Obj. 7d 1 /g State the normal and fail position for the following components:
Reference:
APP-A-07 /1SD-08 SO-08 Cog Level: MemoryMemory Explanation:
Explanation:
Loss Loss of of the air supply the air supply will result in the in-service in the in-service flowflow control control valve valve closing. With With no no dirve dirve water water pressure pressure RMCS will not be able not be able to to move move rods but but they they could could still be scrammed. IfIf air still be air pressure pressure would would continue to lower lower below below 40 40 psig psig the the scram scram inlets inlets and and outlet outlet valves valves would would fail fail open open on on the the loss loss ofof air.
air. The The lower lower DW OW dampers dampers go go to the MAX to the MAX position position on on aa scram scram as as sensed sensed by by pressure switches in pressure switches in the the scram scram airair header.
Distractor Oistractor Analysis:
Analysis:
Choice Choice A: A: Correct Correct answer, answer, see see explanation.
explanation.
Choice Choice B: Plausible since B: Plausible since itit does fail open does fail open andand itit may may seem seem correct correct that that more more cooling cooling would would bebe needed needed inin the the upper upper part of the part of the DW OW since since hot air rises.
hot air rises.
Choice Choice C: C: Plausible Plausible because because some some valves valves do do fail fail closed closed (CRD (CRO flow flow conrol conrol valve) valve) andand the the dampers dampers do do go go to to the the MAX MAX position position Choice Choice D: 0: Plausible Plausible because because some some valves valves do do fail closed (CRD fail closed (CRO flow flow conrol conrol valve) valve) andand itit may may seem seem correct correct that that more more cooling cooling would would bebe needed needed inin the the upper upper part part of the DW of the OW since since hot air rises.
hot air rises.
Notes Notes From theAPP From the APP 1.3.2 IF
'1.32 IF Instrument Instrument AirAir Header Header pressure pressure is is normal normal AND AND Scram Scram Pilot Pilot Air Air Header Header pressure is pressure is less less than 65 psig, than 65 psig, THEN THEN PCV-JA-2878 PCV-JA-2878 has has failed Tailed in in the the closed closed direction OR direction OR thethe Supply Supply Air Air Filter Filter C12-D006A(B)
C12-DOO6A(B) isis dirty.
dirty.
2APP-A-07 I2APP-A-07 Rev, 32 Rev. 32 Page 31 Page 31 of 451 of 45 From SD-08 From SD-08 2.12 Scram 2.12 Scram Air Header Air Header The Scram Air Header is a normally pressurized air header supplying filtered air to supplying to the scram inlet and outlet valves and and outlet and the SDV vent and SDV and drain drain valves. Air to this header header is is supplied supplied from both both Non-Interruptible Instrument Non-Interruptible Instrument Air header header divisions. The The supplied air air maintains the scram valves dosed closed and the SOV SDV vent and drain valves open until a scram signal is received.received. (Reference Figure 08-2)
ISD-08 Rev. -1010 Page 18 681 18 of f58 Drywell Lower vent The Dryweillower Vent dampers can be positioned to either MIN or MAX sitch on Panel XU-3. Normal plant position by a two position control switch operating position for these dampers is the MIN position. Placing these dampers to MAX position during plant operation may produce extreme temperature excursions in the upper drywell regions. Low scram air header pressure will reposition these dampers to the MAX position and automatically start and idle drywell cooling fan selected for AUTO.
Categories K/A:
KIA: 300000 K5.13 Tier Tier!/ Group: T2G1 T2GI RO Rating:
RORating: 2.9 SRO Rating:
SRORating: 2.9 LP Obj:
LPObj: 08-7 Source: BANK Cog Level: LOW Category 8: Y
- 63. Which
- 63. Which one one ofof the the following following choices choices correctly correctly completes completes the the statements statements below?
below?
IfIf system system pressure pressure drops drops to to (1) psig the (1) psig the standby standby CSW CSW pumppump willwill auto auto start.
start.
If pressure remains below this setpoint If pressure remains below this setpoint for for (2) seconds the (2) seconds the SW-V3(V4),
SW-V3(V4), SW SW to to TBCCW Hxs TBCCW Hxs Otbd(lnbd)
Otbd(lnbd) Isol, Isol, will will reposition reposition toto their their throttled throttled positions.
positions.
A. (1) (1)65 65 (2) 30 (2)
B. (1) (1)65 65 (2) 70 C. (1)(1)4040 (2) 30 D
D~ (1)40 (1) 40 (2) 70 Feedback K/A: 400000 K4.01 KIA:
Knowledge of CCWS design feature(s) and or interlocks which provide for the following:
Automatic start of standby pump (CFR: 41.7)
RO/SRO Rating: 3.4/3.9 Objective: CLS-LP-43 Obj 6d Given plant conditions, predict whether any of the following pumps should start:
- d. Conventional Service Water Pumps
Reference:
SD-43 1 /AOP-19 AOP-19 Cog Level: Memory Explanation:
The CSW pumps will auto start at 40 psig, the RCC pumps start at 65 psig.
The SW-V3/4 throttle to a mid position if the low pressure exists for 70 seconds.
The DG cooling valves swap to the opposite unit after low pressure for 30 seconds.
Distractor Analysis:
Choice Choice A: A: Plausible Plausible because because the the RCC RCC pumps auto start pumps auto start at at 65 65 psig psig and and the the DG DG cooling cooling valves swap swap to to the the opposite opposite unit unit after after low low pressure pressure for 30 30 seconds.
seconds.
Choice Choice B: B: Plausible because because the RCC pumps pumps auto start atat 65 psig.
Choice C: C: Plausible because because the the DG DG cooling valves valves swap swap to the opposite unit unit after low low pressure for 30 30 seconds.
seconds.
Choice Choice D: Correct answer, D: Correct see explanation.
answer, see explanation.
Notes Notes 2.0 2.0 AUTOMATIC ACTIONS AUTOMATIC ACTIONS 2.1'1
- 2. Standby pump Standby pump selected selected to to tile the conventional conventional service service water water header starts header starts at at 40 40 psig.
psig.
2.3 2.3 IF conventional IF conventional service service water header header pressure pressure remains remains below below 40 psig for 70 psig for 70 seconds, THEN:
- SW TO SW TO TBGGW TBCCW HXSHXS OTBDOTBD ISOL, SW-V3 closes ISOL. SW-V3 closes to to aa position throttled position
- SW TO SW TO TBGGW TBCCWHXS HXS INBD ISOL, SW-V4 closes to a 1SOL, SW-V4 throttled position tllrottled OAOP-19.0 IOAOP-19.0 Rev. 18 Page 2 of 71 7
From AOP-16:
2.1 2:1 IF system pressure decreases to 65 psig, THEN the standby RBCCW pump will start.
From the SD-43:
- 3. Diesel Generator Cooling Water Supply Valves Downstream of the Diesel Generator cooling water header valve, each diesel generator has two supply Valves '12)-SW-V679 1 (2)-SW-V679 for Diesel Generator 1, '1 (2}-SW-V680 (2)-SW-V680 for Diesel Generator 2, 11 (2)-SW-V68'I (2)-SW-V681 for Diesel Generator 3, and 'I1 (2)-SW-V682 for Diesel Generator 4. One supply valve is designated as the normal supply valve and will open when the diesel generator start is initiated and the diesel speed reaches 500 rpm. Tile The other valve is the alternate supply valve. If sufficient pressure of 5.6 psig is not reached in -0. 30 seconds, the alternate supply valve will open. Once the tile alternate supply valve is full open, the normal supply valve will close. This transfer sequence is initiated anytime service water pressure is lost when aa diesel generator is operating. When the engine is shutdown and speed drops below 500 rpm, the open valve will automatically close.
Initial service water cooling to the diesel generators i.e., (Le., 10
-10 minutes)
ISD-43 SD-43 I Rev. 17 17 I Page 21 2-1 of 78 1 Categories Categories KJA:
KIA: 400000 400000 K4.01 K4.01 Tier Tier / Group:
Group: T2G1 T2G1 RO Rating:
RORating: 3.4 3.4 SRO Rating:
SRORating: 3.9 3.9 LP Obj:
Obj: 43-6D 43-6D Source:
Source: BANK BANK Cog Level:
Cog Level: LOW LOW Category Category 8:8: Y
- 64. Which
- 64. Which one one of of the following identifies the following identifies the the type of detector type of detector and and howhow the the fire fire suppression suppression system operates system operates for SBGT Train for aa SBGT Train charcoal charcoal fire?
fire?
The SBGT The SBGT Trains Trains utilize utilize (1)(1) to to indicate indicate aa fire fire in in the the filter filter bank.
bank.
The Fire The Fire Suppression Suppression System's Systems deluge deluge valve valve for for the the associated associated SBGTSBGT carbon carbon bank bank will automatically will automatically open open and and suppression suppression system system injection injection begins begins (2) (2)
A. (1)
A. temperature switches (1) temperature switches (2) ONLY (2) ONLY ifif the affected Train the affected Train is shutdown is shutdown B (1)
B!'" temperature switches (1) temperature switches (2) following local (2) following local valve valve manipulations manipulations c.
C. (1) ionization detectors (1) ionization detectors (2) ONLY if the affected Train (2) ONLY if the affected Train is shutdown is shutdown D. (1)
D. ionization detectors (1) ionization (2) following local valve manipulations manipulations
Feedback Feedback K/A: 600000 KIA: 600000 K2.01 K2.01 Knowledge of Knowledge of the the interrelations interrelations between between PLANT PLANT FIRE FIRE ONON SITE SITE and and the the following:
following:
Sensors 1I detectors Sensors detectors andand valves valves RO/SRO Rating:
ROISRO 2.6/2.7 Rating: 2.6/2.7 Objective: CLS-LP-41 Objective: CLS-LP-41 *21 *21
- 21. Given
- 21. Given plant plant conditions, conditions, predict predict the the response response of of the Fire Suppression the Fire Suppression and and Fire Fire Detection Detection Systems.
Systems.
- 07. State
- 07. State the the reason(s):
reason(s):
- a. For
- a. For obtaining obtaining SCO's SCOs permission permission to to manually manually operate operate the the SBGT SBGT deluge deluge valves.
valves.
Reference:
Reference:
SD-b, Revision SD-10, Revision 5, 5, Page Page 18, 18, Section Section 3.2.5 Cog Level:
Cog Level: Low Low Explanation:
There are two temperature switches to monitor the temperature of each Carbon Filter in each SBGT train.(TS 3/4)
Switches VA-TS-5302-1 (VA-TS-5302-2), and VA-TS-5297-1 (VA-TS-5297-2) monitor Carbon Filter Bank No. I and actuate at 210°F, rising, to indicate a fire in No.1 in the filter bank.
bank. Actuation of any switch will automatically open the Fire Fire Suppression System's Systems deluge valve for the associated carbon bank bank (note that the associated isolation valves must be opened for this system to inject) and trip the associated Fan and Heater unless compartment inlet temperature is
> 180°F. Local and remote lights indicate switch actuation.
Distractor Analysis:
Choice A: Plausible because temperature switches is correct. High temperature trips the train with inlet temp < < 180°F but suppression flow is not dependant on train status due to Local Deluge System Manual Operation lAW 11(2)OP-b0.(2)OP-1O.
Choice B: Correct Answer Choice C: Plausible because Ionization detectors detect the early products of combustion before they become visible smoke. High temperature trips the train with inlet temp < < 180°F but suppression flow is not dependant on train status due to Local Deluge System Manual Operation lAW I (2)OP-1O.
1 (2)OP-1 0.
Choice D: Plausible because Ionization detectors detect the early products of combustion before they become visible smoke and local manipulations is correct.
2.1.8 2.1.8 Deluge Valve System Deluge System Each SBGT System Each SBGT System is is equipped with a deluge system, including including two deluge valves. The purpose of the deluge system is is to extinguish a fire sensed in tl1e the carbon filter compartments. The deluge valves will autoniatical[y, as sensed by rising temperature in the filters, or open automatlcalfy, manually.
NOTE: The deluge valves are manually isolated. In order for water to tlow, flow, the isolation valves for the deluge valve must be manually opened 2.2 Standby Gas Treatment System Flowpaths 2.2.1 Normal Flow Path 1 0-1 illustrates the arrangement of components and piping for Figure 10-1 the various flow paths of the SBGT System.
The normal system intake is from the 50' 6D elevation of the Reactor Building through two motor operated intake isolation dampers (D, (0. H) and into a common inlet duct. All areas of the Reactor Building with this area. The common inlet duct splits and is communicate '.'lith routed to each Filter train through a moior motor operated, operated. train inlet isolation damper (C, G).
Each Filter train component is duplicated in each train. Flow entering the Filter train first encounters the MOisture Moisture Separator then the electric Heater. Flow then passes through the Prefilter, HEPA Filter electriC No. 1, Charcoal Filters Nos. 1 No.1, 1 and 2, and HEPA 1-IEPA Filter NO.2.
NO. 2.
Flow exiting the filters passes through a a duct to the Fan inlet. Flow from the Fan is routed through a check damper and motor operated discharge isolation damper (S, (B, E). From the discharge isolation damper flow is routed to the Plant Stack.
A penetration of the common duct downstream of the fans permits A
sampling the gas stream with the Post Accident Sampling System (PASS) prior to its entering the Plant Stack. The sample line is isolated by a a solenoid operated valve.
2.2.2 Primary Containment Purge (Vent (Vent) Flow Path The inlet to the SBGT Filter trains may be agned atigned to either the Primaiy Containment Drywell or the Suppression Chamiler Primary Chamber air space for purging operation.
1SD-lU SO-lO Rev.5 Rev. 5 Page12of38 Page 12 of 381
3.
- 3. Carbon Filter Carbon Filter Banks Banks There are There two temperature are two temperature switches switches to to monitor monitor the the temperature temperature of of each Carbon each Carbon Filter Filter inin each each SBGT SBGT train.
train.
(TS 3/4)
(TS 3/4)
Switches VA-TS-5302-1 Switches VA-TS-5302-1 (VA-TS-5302-2),
(YA-TS-5302-2), and and VA-TS-5297-1 VA-TS-5297-1 (VA-TS-5297-2) monitor (VA-TS-5297-2) monitor Carbon Carbon Filter Filter Bank Bank No.1 No. 1 and and actuate actuate at at 210 oF, rising, 21OF, rising, to to indicate indicate aa fire fire in in the the filter filter bank.
bank. Actuation Actuation of IDrl ofj switch will switch automatically open will automatically open thethe Fire Fire Suppression Suppression System's Systems deluge deluge valve for the associated carl)Qn carbon bankbank (note (note that the associated isolation valves must be opened for this system to inject) and trip trip the associated Fan Fan andand Heater Heater unless compartment inlet temperature is 1 80F. Local and
>> 180*)F. and remote remote lights indicate indicate switch switch actuation.
actuation.
(TS 5/6)
(TS 5/6)
Switches VA-TS-5303-1 tVA-TS-5303-2), and VA-TS-5298-1 (YA-TS-5298-2), monitor Carl>on (VA-TS-529S-2), carbon Filter Bank No.2 No. 2 and actuate at 210°F, rising, to indicate a fire in the #2 filter bank. Actuation of ~
automatically open the Fire Suppression System's switch will automatically Systems deluge valve for the associated carl>on carbon bank (note (note that the associated isolation valves must be opened for this system to inject) and trip the associated Fan and Heater unless compartment inlet temperature is is 180°F. Local and remote lights indicate switch actuation.
> 180"F.
- 4. HEPA Filter No.2 No. 2 Compartment Switches TSL-3456 (3455) provide annunciation of SBGT Filter train A/B Hi humidity.
..vB 3.2.6 Automatic
- 1. Upon receipt of an automatic initiation signal both trains of SBGT will start.
Unit 1I ONLY The dampers associated with Unit 1I SBGT System will receive automatic open signals when an initiation signal is received EXCEPT for the train inlet and outlet outret dampers, (BFVS-1(GFVs-1 B, 1C,1E,and 1C,1 E,and 1G).
10).
Should these nomally nom1ally openopen dampers be manually closed locally via their CLOSE/OPEN CLOSE/OPEN pushbuttons, pushbuUons, they will NOT automatically reopen andand the associated SBGT will not automatically start.
I SD-lU SO-lO Rev 5 Page 118 of 3838 I Categories Categories K/A:
KIA: 600000 600000 K2.01 K2.01 Tier / Group:
Tier Group: TIGl T1G1 RO Rating:
RORating: 2.6 2.6 SRO Rating:
SRORating: 2.7 2.7 LP LP Obj:
Obj: CLSLP41*21 CLS-LP-41 *21 Source:
Source: NEW NEW Cog Cog Level:
Level: LOW LOW Category Category 8:8: YY
- 65. Unit
- 65. Unit Two Two operating operating at rated power at rated power with with Main Main Generator Generator MVARMVAR loading loading at at +300
+300 MVARs.
Which one Which one ofof the following correctly the following correctly completes completes the statement below the statement below based based onon these these conditions?
conditions?
The Main The Main Generator Generator component component that that would overheat isis the would overheat the (1)
(1) andand lAW lAW 20P-27, 20P-27, Generator and Exciter System Operating Generator and Exciter System Operating Procedure,Procedure, MVARs MVARs must must be be lowered lowered to to less less than (2) than (2) .
(1) armature A. (1)
A. armature (2) +70 (2) +70 B. (1)
B. (1) field (2) +70 (2) +70 C. (1) armature (2) +170 D (1) field D:'
(2) +170
Feedback Feedback K/A: 700000 KIA: K1.02 700000 K1.02 Knowledge of Knowledge of the the operational operational implications implications of of the following concepts the following concepts asas they they apply apply to to GENERATOR VOLTAGE GENERATOR VOLTAGE AND AND ELECTRIC ELECTRIC GRID:GRID:
Over-excitation Over-excitation (CFR: 41.4, (CFR: 41.4,41.5,41.7,41.10/45.8) 41.5, 41.7, 41.10/45.8)
RO/SRO Rating:
RO/SRO Rating: 3.3/3.4 3.3/3.4 CLSLP27*1 if Objective: CLS-LP-27*11f Objective:
- 11. Given
- 11. Given plant plant conditions, conditions, describe the effect describe the effect that that aa loss loss or or malfunction malfunction of of the the following following may may have have onon the the Main Generator:
Main Generator:
regulator (including
- f. Voltage regulator (including Under Under and Over-Excitation)
Reference:
Reference:
SD-27, Revision SO-27, Revision 14,14, Page Page 22, 22, Section Section 2.17 2.17 Cog Level: High Explanation:
placed on lagging MVARs The limitation placed MVARs ((+ + MVARs) of the estimated capability curve, limits operation because of excessive heating that occurs in the generator field. Since the generator is is operating in in an overexcited condition, a larger field current is is necessary to produce the extra KVARKVAR being supplied to the system. with the conditions given, the generator is operating outside the capabilities curve. The minimum gross MVAR requirement is 70 (positive) while the maximum gross MVAR requirement is 170 (positive).
Distractor Analysis:
Oistractor Choice A: Plausible because the limitation placed on leading MVARs ((-- MVARs) of the estimated capability curve are less effected by Hydrogen pressure. We see that the curves come together sharply. As the system is required to supply more reactive power to the generator field, the flux in the air gap between the field and stator becomes more distorted. The distortion results in the exposed ends of the stator coils becoming overheated. As field strength is reduced, this heating accelerates. The +70 is the lower end of the operating band for the generator VARS.
Choice B: Plausible because overheating of the field is correct and the +70 is the lower end of the operating band for the generator VARS.
Choice C: Plausible because the limitation placed on leading MVARs ((-- MVARs) MVAR5) of the estimated capability curve are less effected by Hydrogen pressure. We see that the curves come together sharply. As the system is required to supply more reactive power to the generator field, the flux in the air gap between the field and stator becomes more distorted. The distortion results in the exposed ends of the stator coils becoming overheated. As field strength is reduced, this heating accelerates. The +170 is the high end of the operating band for the generator VARS.
Choice D:0: Correct Answer SRO Only Basis:
Basis: N/A N/A Notes
GENERATOR GENERATOR REACTIVE REACTIVE CAPABILITY CAPABILITY CURVECURVE ATB 44 POLE ATB POLE 1.039.0.0.0 1039000 KVA 1800 RPM RVA 18.0.0 RPM 24.0.0.0 24000 VOLTS VOLTS 0.964PF 0964PF
.0,53 SCR, 60 PSIG 0,53 SCR,60 PSIG HYDROGEN PRESSuRE. see HYDROGEN PRESSURE. 500 VOLTS VOLTS EXCITATION EXCITATION 1200 122121 iooø 60 PS]G 50 PSG :
600j 46 48 45 PSIG PSIG 2 -
P5KG PSIG .90 30 PSIC a 4X
.95 15 PSIG 2 -
. 8.98 5
0 (F)
Vl 0::
~ 200 0Q -200
--J
--J
~
g} .-4OO B.95
-400
$9 3.
600
-600 o
~
-J
--J 800
-600 -
200 400 600 1300 800 000 1200 1000 1.0'00 KILOWATTS OP-27 Rev. 58 Rev. Page 39 481 39 of 48
From the From the SO:
SD:
The limitation The limitation placed placed on on lagging lagging MVARs MVARs (( ++ MVARs)
MVARs) of of the estimated the estimated capability curve, limits operation because capability curve, limits operation because ofof excessive excessive heating heating thatthat occurs in occurs in the the generator generator field.
field. Since Since the generator isis operating the generator operating in in an an overexcited condition, overexcited condition, aa larger larger field field current is necessary current is necessary to produce to produce the extra the extra KV KVAR being supplied AR being supplied to to the the system.
system. This This results results in greater in greater heating of heating of the the rotor windings.
rotor windings.
I SO-27 SD-27 Rev. '14 Rev. 14 Page 48 of 1271 Page 127 From the From the OP:
OP:
3.1 Maintain generator I'vlaintain generator loading loading within the limits limits of of Figure Figure '11 and the following:
NOTE: The System Operations Load Dispatcher should be contacted if the MVAR loading can NOT be maintained within prescribed limits. limit&
3.1.1 The minimum gross MVAR requirement is 70 (positive) as read on the main generator terminals.
3.1.2 The maximum gross MVAR requirement is -170 170 (positive) as read on the main generator terminals.
20P-27 120P-27 Rev. 58 48 Page 4 of 481 Categories K!A:
KIA: 700000 K1.02 Tier!
Tier / Group: T1G1 TIG1 RO Rating:
RORating: 3.3 SRO Rating: 3.4 SRORating:
LP Obj:
LPObj: CLSLP27*11F CLS-LP-27*11F Source: NEW Cog Level: HIGH mGH Category 8:
- 66. Which
- 66. Which one one ofof the the following following issthe purpose of the purpose of the the High High Pressure Pressure Coolant Coolant Injection Injection (HPCI) System?
(HPCI) System?
HPCI isis designed HPCI designed to provide sufficient to provide sufficient coolant coolant injection injection to maintain the to maintain the Reactor Reactor core core covered during covered during aa (1) Loss-Of-Coolant-Accident to (1) Loss-Of-Coolant-Accident to maintain maintain fuelfuel cladding cladding temperatures below temperatures below (2)
(2)
A. (1 A. (1)) small small break break (2) 1800°F (2) 1800°F B
B~ (1) small (1) small break break (2) 2200°F (2) 2200°F C. (1)) large C. (1 large break break 1800°F (2) 1800°F (1) large break D. (1)
D.
(2) 2200°F Feedback K/A: G2.01.27 KJA: G2.01.27 Conduct of Operations Knowledge of system purpose and/or andlor function.
(CFR: 41.7)
(CFR:
RO/SRO Rating: 3.9/4.0 RO/SRO CLS-LP-01 9*01 Objective: CLS-LP-019*01
- 1. State
- 1. State the purpose of the High Pressure Coolant Injection (HPCI) System.
Reference:
SD-19, SD-19, Revision 16, 16, Page 6, Section 1.2 1.2 Cog Level: Low Explanation:
Explanation:
The High High Pressure Coolant Injection (HPCI) System System was designed designed to provide sufficient coolant injection to maintain maintain the Reactor core covered during during a small line break Loss-Of-Coolant-Accident (LOCA) which does does not result in rapid vessel depressurizat depressurization,ion, thus maintaining fuel cladding temperatures temperatures below below 2200°F.
2200°F. TheThe original original design design basis of the basis of the HPCI HPCI System System was was to provide partpart of of the the Emergency Core Core Cooling System System (ECCS)
(ECCS) function. HPCI HPCI system system operation operation mitigated mitigated small small break break LOCAs LOCAs where where the the depressurizat depressurizationion function [Automatic
[Automatic Depressuriza tion System Depressurization (ADS) I1SRVs]
System (ADS) SRVs] waswas assumed assumed to to fail.
Distractor Distractor Analysis:
Analysis:
Choice Choice A: A: Plausible Plausible because because small small break break is is correct, and 1800°F correct, and 1800°F isis the the number number forfor ifif adequate adequate core core cooling cooling can can not not bebe maintained maintained by by core core submergence submergence..
Choice Choice B: B: Correct Correct Answer Answer Choice Choice C: C: Plausible Plausible because because HPCI HPCI isis aa high high capacity, capacity, high high pressure pressure injection injection system system which which is is easily easily mistaken mistaken forfor large large break break LOCA LOCA makeupmakeup requirements requirements,, and and 1800°F 1800°F isis the the number number forfor ifif adequate core cooling adequate core can not cooling can not be be maintained maintained byby core core submergence submergence..
Choice Choice D: D: Plausible Plausible because because HPCIHPCI isis aa high high capacity, high pressure capacity, high pressure injection injection system system which which is is easily easily mistaken mistaken forfor large large break break LOCA LOCA makeupmakeup requirements requirements,, and and 2200°F 2200°F isis the the temperature temperature that that cladding cladding will will not not exceed exceed with with core core submergence submergence. .
SRO SRO Only Only Basis:
Basis: N/A N/A
Notes Notes 1.0
1.0 INTRODUCTION
INTRODUCTION 1.1 1.1 System Purpose System Purpose The High The High Pressure Pressure Coolant Coolant Injection Injection (HPCI)
(H PCI) System System was was designed designed toto provide provide sufficient coolant sufficient coolant injection injection to to maintain maintain the Reactor core the Reactor core covered covered during during aa small small line break line break Loss-Of-Coolant-Accident Loss-Of-Coolant-Accident (LOCA) (LOCA) which which does does not not result result in rapid in rapid vessel depressurization, vessel depressurization, thus thus maintaining maintaining fuelfuel cladding cladding temperatures temperatures below below 2200°F.
2200"F.
SD-19 1 SD-19 Rev. 17 Rev. 17 Page Page 6l ofof 1081 108 MINIMUM ZERO-INJECTION MINIMUM ZERO-INJECTION REACTORREACTOR WATER WATER LEVELLEVEL The lowest reactor water level at which tile the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of of the core from exceeding -1800°F.
1800°F. This water level is lIsedused by the Steam Cooling Procedure to preclude significant fuel damage and hydrogen generation for as long as possible (Unit -I1 only: Figure '19;19; Unit 2 onlv:
only: Figure 19A).
Categories KJA:
KIA: G2.O1.27 G2.01.27 Tier/Group:
Tier / Group: T3 RO Rating:
RORating: 3.9 SRO Rating: 4.0 SRORating:
LP Obj: CLSLP019*O1 CLS-LP-019*01 Source: NEW Cog Level: LOW Category 8:
- 67. LOCA
- 67. LOCA conditions conditions exist exist on on Unit One. CREV Unit One. CREV has has failed failed to to auto auto start start and and the the CRS CRS has has ordered CREV ordered CREV to to be manually started be manually started per per the the Hard Hard Card.
Card.
Which one Which of the one of the following following identifies:
identifies:
(1) the action(s)
(1) the action(s) required required to start the to start the CB CB Emerg Emerg Recirc Recirc Fan Fan and and (2) the (2) the desired desired position position indication indication for for the CB Emerg the CB Emerg Recirc Recirc Damper Damper (VA-2J-D-CB)?
(VA-2J-D-CB)?
(1) Place A. (1) Place the the CB CB Emerg Emerg Recirc Recirc Fan Fan control control switch switch onon Unit Unit Two Two XU-3 XU-3 panel panel to to On On Green (2) Green (2)
B (1)
B!'" (1) Place Place the CBCB Emerg Emerg Recirc Recirc Fan Fan control switch on Unit Unit Two XU-3 panel panel to On On (2) Red (2) Red (1) Simultaneously C. (1) Simultaneously place place both Units' Units CB Emerg Emerg Recirc Fan control control switches onon their respective XU-3 panel respective panel to On (2) Green Simultaneously place both Units' D. (1) Simultaneously Units CB Emerg Recirc Fan control switches on their respective XU-3 panel to On (2) Red
Feedback Feedback K/A: G2.01.31 KIA: G2.01.31 Conduct of Conduct of Operations Operations Ability to Ability to locate locate control control room room switches, switches, controls, controls, and and indications, indications, and and to to determine determine that that they they correctly reflect correctly reflect the the desired desired plant plant lineup.
lineup.
(CFR: 41.10/45.12)
(CFR: 41.10/45.12)
ROISRO Rating:
RO/SRO Rating: 4.6/4.3 4.6/4.3 Objective: CLS-LP-37, Objective: CLS-LP-37, Obj Obj 12d 1 2d Explain the Explain the following:
following:
- d. How
- d. How toto place place the the Control Control Room Room Ventilation Ventilation system system inin Recirculation Recirculation Mode.
Mode.
Reference:
Reference:
OOP-37 OOP-37 Cog Level: Comprehensive Cog Level: Comprehensive Explanation:
The controls for the CREV system are on U2 only. Indication for the CREV System is on both units. The emergency recirc damper will open when the fan is started and the open indication is red. The normal makeup damper closes on starting the fan in which the closed indication is green.
The Control Building Mechanical Equipment Room Vent Fans can only be stopped by simultaneously Units control switches in OFF.
placing both Units' Distractor Analysis:
Choice A: Plausible because the control switch is located on U2, but the recirc damper will open which is red. The normal makeup damper closes which is a green indication.
Choice B: Correct answer, see explanation.
Choice C: Plausible because the CB Mechanical Equipment Room Vent Fans can only be stopped by Units control switches and the normal makeup damper closes simultaneously operating both Units' which is a green indication.
Choice D: Plausible because the CB Mechanical Equipment Room Vent Fans can only be stopped by simultaneously operating both Units Units' control switches and the recirc damper does open which is a red indication.
SRO Only Basis:
Basis: N/A N/A
Notes Notes 53.2 5.3.2 Procedural Steps Procedural Steps NOTE:
NOTE: Indications for Indications for the Control Building the Control Building Ventilation Ventilation System System areare located located on on Panels Panels XU-3 on XU-3 both units.
on both units.
NOTE:
NOTE: Controls for Controls for the the Mechanical Mechanical Equipment Equipment RoomRoom Ventilation Ventilation Fans Fans and and the the Control Building Control Building Wash Wash Room Room Exhaust Exhaust FanFan are are on on XU-3 XU-3 onon Units Units 'II and and 2.
2.
NOTE:
NOTE: Controls for Controls for the the Control Control Building Building Emergency Emergency Recirculation Recirculation Fans Fans are are on on Panel Panel XU-3 on Unit 2.
on Unit 2.
NOTE:
NOTE: Controls for Controls for the the Cable Cable Spread Spread Room Room ventilation ventilation fans fans are are on Panel XU-3 on Panel XU-3 for for respective unit.
the respective unit.
OOP-37
\ OOP-37 Rev. 57 Page 15 15 of 69\
69 PLACE ONE OF THE CB EMERG EMERO RECIRC FANS, 2A(8)-ERF-CB, 2A(B)-ERF-CB, IN ON.
ENSURE CTL RM NORM MU AIR MR DMPR, 2L-D-CB, 2LD-CB, CLOSES.
ENSURE CB EMERG RECIRC DAMPER, VA-2J-D-CB, VA-2J-D-CB, OPENS.
NOTE:
NOTE: The Control Building Mechanical Equipment Room Vent Fans can be stopped only by simultaneously simultaneously placing both Units'Units control switches in OFF.
Categories Categories K/A:
KIA: G2.O1.31 G2.01.31 Tier/Group:
Tier / Group: T3 T3 RO Rating:
RORating: 4.6 4.6 SRO Rating: 4.3 SRORating: 4.3 LP LP Obj:
Obj: 37-12 37-12 Source:
Source: NEW NEW Cog Level:
Cog Level: HIGH HIGH Category Category 8:
8:
er.
f r o m 6 600%%ppoowwer.
trip from he Sccraramm t i o nn PPuum mpp trip in t he S c t o r RR eecciircula nd oppeerraatitoionn isis in t r c u l a t io a r rriieen cceedd aa R Reeacteoen eenntteerreedd aand o s e x p p e e n ,, hhaass bbeenw maapp.. 4 to
. U n n i itt OO n n e e haas ex ore FFllo h
o w C o r e o ww e r t o ffl o lo w m f r o m ppoossititioionn224 to 6688. U OP--0 04.0 L C
.0,, Lowon o e poow w e r to rods from AOP n4ce Regiio 1IA off th p g f o u u rr ccoonntrtrool l rods idaance n nsseerrttiing f roveemo t.
Avvooid A c o m m m ennddss iin e i m pp r o v meennt. 2 ionss
, , OOppeerarat tion
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O r g g a a n i izzaatiioon UUnnitit Or s?
SSttaanndda arrdds
.0 2: p onn 1188. .
oossititiio 1 .0
. 0 1 1 . 0 e d t o to p of 00 y iinnseerrtted 188. .
A t t t a acchh meent 1I m c o n tti n in uoouusslly n c ee t o t o ppoossititioionn 1 A to be diis sttaan c t rol rods t h ee eennttiirree d too a l ll looww ss tthh e c o n i n g l lee n otc h eedd t h
, tthh eenn nnoottcchheedd t AA.. a be s in onn 2200, c o n t r o l ro ds to s e rrtteedd ttoo ppoossititiio e y iin n e uirireess tthhe nuouusslly BB!"" rreeqqu e c ontiin sttss. .
t r o l rro ds to b
c o on n onn eexxiis ddititiio c
c o n n ccyy c wwss tthhee CC.. aalllloo onn 1188. c ee aann eem meerrggeen ve d s s i inn ppoossititio i
ay be w aiiv aann c e m ing gguuid id dd pp oossit ionin itio DD.. ro r o
Feed b ack 2..0O11..3 399
/A: G2 KIA ractices. s.
o d u c t o f OOppe rations e d e c is is io io nn makingg p CC o n nseerv rvaatiti v Knnoow K wle leddge of co s 5.1 .122))
0 /4 3 .5 /4I 4 5 (CFR: 4 1 .1 ISRO Raattin ing: 3.6/4.3 roggraramm::
RO/S y M an a n ag a g emen em entt PPro 244ff iv eactctiv it ity LS-LP-201 01D -D**2 -NG GG GC C-1-13066 Rea conntrtrooll ro ds. s.
Objectivee:: C e follo g re g ar a rdding OPSS-N o n in g in te rm ed e d te co iate ia ll oww in p o si ti ti
- 4. E xplain ce th meen ireem nts for 224 ural requir
- f. T Thhe p roced annagagem emen ent,t, in n g g an a nd d R ea e actctivivitityy Ma nce: atiivve De ecciissiioon Maakki Reeffeerreen R ti ioo n 5 .1 0 Consseervat Seecct 001-0 1.02 3 0 6 Se on 9.1.4 eccttiio
-N G G C -1 OOP S se e rv atiivvee ddec rvat eciissiioonns. s.
co c on s n o f Attaacchm more entt 1 en 1 a a re re rmed edia te ppo oon iate o si siti ti CCog LLeev veell:: High e g u id an a n ce of ce re to to an a n interm in ti o tio nn th th en en ulations w ithhiinn th fourr notches es orr in tended ed fifinnaall ppoosisi ro d m an a n ip d is m o v ed e d ef o fo re re th e e sh oo u u ld ld b b e e
on: Coon ntrolol n a contro l ro at the notc h notches h be ss,, th thee ro rodd sh Exp E xpla lanaattiio ir ir eeccts tha wheen att w u ld ld be stopped ed at re e es oorr less rodd sh ro shoouuld ld bbee 1 -0 1 .0 2 , d o l ro d sh o u g m o v ed e d th er ef re o fo re re th th e e cy PP e r 0 0 0 co n tr is b ei e n in otc e h es s th e eme gen er rg ency en 002 2 aan n d 46) the ti tioonn. W Wheenn a rod to 188 is 1 is three n on o n ly dduuririnngg an ly an em (beettw (b weee naal po p o si froomm 2 4 w ai a v iv ed e d ea e act ct oorr tchheedd to th the finfi Moov viinng a rod me en nts can bee min ined ed by by th thee RR ssiinngle nnootc th th e e en e n tire move ve.. M se s e rv rv a at iv ti v e rod move er em al a l li m lim it ss ass det eteerrm al a l co co n n ddititiioonn..
tchh e ed d eess for cco o n late th e rmm abnn oorm rm ssiinngle nnootc Th he g u id id el e in lin g w ou o uld v io ergg en e n cy c y b uutt an ab an utess a a
tchheedd. T ngle notchin nnoot cco ute anan em emer const stiittu ssiinngle nnootc T WW S) o r w S whheen n ssiin o n do onststiittu tt a reacacto scra torr sc ram m co a s a n an AA T e q u es e st io ti p re v en e n (such a (s onditio ionns ooff th Maan naag ge emmenent to ineee er. T Thhee cco Re ctiivvity M eaact EEnnggin m e n en ts ooff ingg th thee re requirireem Waaiivvin W
a tivee deeccisio n
nnoonn-c -coonnse ativ s e rv fullll oouutt bu o r fu butt e
ei er fu itthher in or fullll in inse se rt rt io n to ctor A Annaallyysi sis:
s: continuous DDisistrtraac h m e ennt 1 1 a lows co alllo a
at tt a ta c ch le bbeeccaauusse e since CChhooic icee AA:: PPlausib r th e ggiv iv e enn condit onns.
itiio s.
nnoot uunnddeer the priioorr to to thee n s w sw e err se se rt rt io n to the notchh pr d is only is on ly Coorr rreecct A A n ontinuous in s the ro CChhooic icee BB:: C h m eenn t 1 1 a alllo lo ws cco In th e g iv e enn condditition e aatttta acch s.
usse e ssiinncce ore notchees.
e CC :: PP la la usible bbeeccaau heen n m oovv in g four oorr m ic CChhooice intetennddeed d ppoosititioonn w wh aiivve edd duuriringg an an in o v e d ed 3 3 nn ootc tchheess..
id eellin in e ess m ay be wa ay bbeeiinngg m mo v veem me ennt g u a tt taa c ch hme ennt 1I rod mo v ausse in e ssin ccee at icee DD:: PPla lauusisible bbeeccau CChhooic meerrggeency n c y .
eem sis: N N/A
/A SSR RO OO Onnllyy BBaasis:
Notess Note nt Progra gramm for the NGG by out outlini ngg linin ure def ines ine s the Rea Rea ctiv ctivity ity Ma nageme em ent This pro ced ed inin reac reactiv ty man tiviity ageem ma nag welll as those entt as wel men ons ibil itie litie s of of key pos itio itionnss inv invo olv lved ration operati on the resp respon sibi a conserv ativvee philosophy servati of reac phy of torr ope rea cto ms esta esta blis blis hed whi ch imp lem ent progra [Ri3 - SOER 94-1, redd.. [R1 tha t the integrit suchh that grityy of the reac torr core is assu rea cto assure Re Rec om mendatio com tionn 1]1]
ncyy conditi rgeenc ons ditio ns idan cee of Ste ps C.3 .i.1 2 & '13 is waived during eme 13 is em erg NOTE: The gLl guid anc cur ren ce, ce, has cto r Engineer; r, wit withh Shiftft Manager con or when the Reactor in ther lt in mall lim the rma ts challenges.
limiits tha t notch operati nedd that rmiine on will resu res ult dete det erm e, stop morre, sto pss one notch 12)
-12) When movin vingg a control rod four not ches or mo con trol rod is goi goinngg positio n unle ess unl ss the n single not ches to finall pos ition rt, then short, positio to pos ition 48" (full out n "48 out))..
less;, performs separat ara tee
-13)
- 13) When movin vingg a control trol rod three notch or less single notch mo ves.
Rev. 68 Page 27 of 45 451
-011.0022 0011-0 100 gori Catego Cate es ries Tier/G Tier / Gro p: T3 rouup:
K/A:
KIA 02.0 1.39 G2.O1.39 SRO Rati SRO Rating: 4.3 g:
atinng:
RO Rati ROR 3.6 D*224F 4F ce:
Sourrce:
Sou PREV Y bj: CLSLP CLS 2O1I-D*
-LP-20 LP Obj:
LPO Cate gory 8: Y l:
Leveel:
Cog Lev HIGH
- 69. Which one
- 69. Which one ofof the the following following identifies identifies where where the the Electric Electric Fire Fire Pump Pump can can be be started started from?
from?
The Electric The Electric Fire Fire Pump Pump can can be be started started locally:
locally:
ONLY.
A. ONLY.
A.
B and at the Unit B!'" Unit One RTGB RTGB ONLY.
c.
C. and at and at the Unit Unit Two RTGB RTGB ONLY.
D. and at either Unit One or Unit Two RTGBs.
Feedback K/A: G2.02.04 KIA: G2.02.04 EQUIPMENT CONTROL EQUIPMENT CONTROL boardlcontrol room layouts, systems, (multi-unit license) Ability to explain the variations in control board/control instrumentation, and procedural actions between units at a facility.
41.6/41.7 / 41.10 /45.1/45.13)
(CFR: 41.6/41.7/41.10/45.1/45.13)
ROISRO Rating: 3.6/3.6 Objective: None
Reference:
OOP-41, Revision 101, Page 18, Section 8.3 SD-41, Revision 8, Page 28, Section 3.2.2 S0-41, Cog Level: Low Explanation:
Ability to explain the variation is identifing the difference in location of the controls.
The Electric Fire Pump can be started
- 1. Automatically (Low system pressure) 1.
- 2. Manually from the Control Room (Panel XU-69 which is only located on Unit One)
- 3. Manually from local control panel.
Distractor Oistractor Analysis:
Choice A: Plausible because the majority of License Operator simulator training is performed on U2 U2 simulator which does not have Panel XU-69.
Choice B: Correct Answer Choice C: Plausible because Choice because controls controls would exist on Unit exist on Unit Two however however they dodo not.
not. This is is a difference difference between between Units Units Choice Choice D: 0: Plausible because controls would exist on Unit Two however however they dodo not.
not. This is aa difference difference between between Units Units SRO SRO Only Basis: N/A Only Basis: NIA
Notes Notes 12.2 3.2.2 Electric Fire Electric Fire Pump Pump Control power Control power for for the the Electric Electric Fire Fire Pump Pump local local panel panel isis provided provided from from thethe 4160 4160 VV feed feed E-2 or (E-2 or E-4) through transformers.
E-4) through transformers. When When the the control control panelpanel isis energized, energized, the the pump pump isis in the in the AUTO AUTO mode mode of of operation operation andand will start when will start when fire fire system system pressure pressure isis approximately 105 psig.
approximate~1105 psig. Once Once thethe pump pump starts, starts, itit will will continue continue to to run run until until manually manually stopped at stopped the local at the local control control panel.
panel. A local "manual A local manual start"start button button will will also also start start the the pump. Once pump. Once started, started, the the pump pump willwill run run until until manually manually stopped.
stopped.
There isis aa safety There safety interlock interlock in in the the Electric Electric Fire Fire Pump Pump local local panel panel that that 'Nill will not not aliow allow power to the Electric power Electric Fire Fire Pump Pump if either of of the the front upper upper or or lower lower high high voltage access panels access panels are are open or improperly improperly secured secured closed.
A pilot, lamp on A pilot on the control control panel panel and and oneone in thethe control control room room lights whenever whenever voltage voltage appears on appears on the the load load side side of of the the circuit circuit breaker breaker and and indicates indicates that that power power isis available available at at the controller.
tile controller.
The electric Tile electric fire tire pump pump may also be started started from the Fire Protection panel XU-59 XU-69 in the Control Room. The The remote remote start pusll push button startsstarts the pump without supervision of the pressure Switch tile switch and the pump will run until manually stopped at the local panel.
Alarms are provided
,A.larms provided on Annunciator Panel Panel UA-37:
UA-37:
- MINTMWT BLDG ElEC ELEC FIRE PUMP RUNNING - initiated by (PS-1871 at 105 psig.!.
- psig ÷ 10 psig and lowering)
- ELEC FIRE PUMP FAil ElEC FAIL TO RUN Electrical power for the electric fire pump P-2 issupplied from bus E-2 (CB-AHl) (CB-AH7) and from Ilus bus E-4 (CB-Al3).
(CBAL3). The two tco feeders terminate at Transfer SWitch Switch (lG-5)
(LG-5) located adjacent to the Water Treatment Building. Upon loss of power from the nom1al nomial source (E-2), a manual transfer switch is available to transfer to the alternate source (E-4).
3.2.3 Diesel Fire Pump The Diesel Fire Pump, Pump. P-1, P-i. local control panel is equipped with aa five-position selector switch (OFF. l\o1ANUAl Switch (OFF, MANUAL START A, MANW\l MANUAL START B, S. TEST, AUTO) and to two manual push buttons (M.O\NUAl (MANUAL ST.O\RT START and RESET).
With the selector switch in AUTO, the pump will start when system pressure drops to approximately 90 §0 psig. Once started, the engine will continue to run until unfil manually manualj stopped or automacalIy automaticafly stopped by engine overspeed. There are no other automac automatic shutdowns.
The pump may also be started by placing the control switch to MANUAL A or MANUAL B B and depressing the MANUAL START pushbutton. The selection of
',MNUAl A or MANUAL BB selects the battery that will crank the diesel unl MANUAL until it fires.
fifes.
ISD-41 ISD-41 Rev 8a 28 of aD 28of80J I Categories Categories K/A:
KIA: G2.02.04 Tier Tier / Group:
Group: T3 T3 RO Rating:
RORating: 3.6 3.6 SRO Rating: 3.6 SRORating: 3.6 LP Obj:
LPObj: NONE NONE Source:
Source: BANK BANK Cog Cog Level:
Level: LOW LOW Category Category 8: 8:
- 70. Which one
- 70. Which of the one of the following following describes describes the the bases bases for for the the Minimum Minimum Critical Critical Power Power Ratio Ratio (MCPR) Safety (MCPR) Safety Limit?
Limit?
The MCPR The MCPR Safety Limit ensures Safety Limit ensures thatthat _ _ _ duringduring normal normal operation operation and and during during Anticipated Operational Anticipated Operational Occurrences.
Occurrences.
A. 17% cladding A. 17% cladding oxidation oxidation doesdoes notnot occur occur B. coolable core B. aa coolable core geometry geometry is is maintained maintained C. cladding plastic C. cladding plastic strain strain remains remains lessless than than 11%%
Dv at DY' least 99.9%
at least 99.9% of the fuel rods of the rods dodo not not experience experience transition boilingboiling Feedback Feedback K/A: G2.02.25 KIA: G2.02.25 EQUIPMENT CONTROL EQUIPMENT CONTROL Knowledge of Knowledge of the the bases bases inin Technical Technical Specifications Specifications for for limiting limiting conditions conditions for for operations operations and and safety limits.
/41.7/43.2)
(CFR: 41.5 141.7 1 43.2)
RO/SRO Rating: 3.2/4.2 CLSLP200B*O3 Objective: CLS-LP-200-B*03
- 03. State each TS Safety Limit and discuss the basis for each of the Safety Limits.
Reference:
Reference:
U2 TS Bases Cog Level: Low Explanation:
Requires knowledge of TS Safety Limit Bases and the ability to distinguish between Safety Limits and Operating Limits. The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core do not experience transition boiling.
Distractor Analysis:
Choice A: Plausible because sincesince this is is ECCS acceptance criteria.
Choice B:
B: Plausible because since this is ECCSECCS acceptance criteria.
criteria.
Choice C: Plausible because Choice C: because since since this is is the the basis basis for the the LHGR limit.
limit.
Choice Choice D:
D: Correct Correct Answer.
Answer.
SRO Only Basis:
SRO Only Basis: N/A N/A Notes Notes
Reactor Feactor Core Core SLs SLs B52.1.1 2:1.1 B8 2.0 2.0 SA.FETY SAFETY LIMITS UMITS (SLS)
(SLS)
B8 2.'1.1 2.1.1 eactor Core Reactor Core SLs SLe BASES BASES BACKGROUND BACKGROUND SLs ensure SLs ensure thatthat specified specified acceptable acceptable fuel fuel design design limits limits are are not not exceeded exceeded during steady state during steady state operation, operation. nom1al normal operational operational transients, transients. and and anticipated operational anticipated operational occurrences occurrences (AOOs).(AOOs).
The fuel The cladding integrity fuel cladding integrity SL SL isis set set such such that that no no fuel fuel damage damage isis calculated to calculated to occur occur ifif the the limit limit isis not not violated.
violated. Because Because fuel fuel damage damage isis not not directly observable, directly observable, aa stepback stepback approach approach isis used used toto establish establish anan SL, such SL, such that the MCPR that the MCP is is not less than not less than thethe limit limit specified specified in Specification 2.1.1.2.
in Specification 2.1.1.2.
MCPR greater MCPR greater than than the specified limit the specified limit represents represents aa conservative conservative margin margin relative to relative to the the conditions conditions required required to to maintain fuel cladding maintain fuel cladding integrity.
integrity.
The fuel The cladding is fuel cladding is one of of the physical physical barriers barriers that that separate separate the radioactive materials radioactive materials from the environs. environs. The integrity integrity ofthis of this cladding cladding barrier is related barrier related to its relative freedom from perforations or cracking.
Although somesonic corrosion or use related cracking may occur during during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding cumUlative cladding perforations, however, can result from thermal stresses, which occur from reactor significantly above design conditions.
operation Significantly While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration deterioration.. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce produce onset of transition boiling (i.&. (i.e.,
1.00). These conditions represent MCPR == 1.00). repres*ent a significant departure from the condition intended intended by by design design for planned operation. The MCPR MCPR fuel cladding integrity SL ensures that during normal cladding integrity nonmal operation operation and duringduring AOOs, at least least 99.9%
99.9% of the fuel rods in in the core core do notnot experience transition transition boiling.
Icontinued) icontinued)
Brunswick Brunswick Unit Unit 22 5B 2.1 .11 2:1.1-1 Revision No.
Revision No. 30 30 I
lHGR LHGR B3.2.3 B 3.2.3 B5 3.2 POWER DISTRIBUTION 3.2 POWER DISTRIBUTION LIMITSLIMITS 3.2.3 B5 3.2.3 LINEAR HEAT LINEAR HEAT GENERATION GENERATION RATE RATE (LHGR)
(LHGR)
BASES BASES BACKGROUND BACKGROUND The lHGR The LI-IGR is measure of is aa measure of the the heat heat generation generation rate rate of of aa fuel fuel rod rod in in aa fuel fuel assenthly at assembly at any axial location. limits any a:(iallocation. Limits on on iIle the LHGR LHGR are are specified specified to to ensure that ensure that fuel fuel design design limits limits are not exceeded are not exceeded anywhere anywhere in in the the core core during normal during normal operation, operation, including including anticipated anticipated operational operational occurrences occurrences tACOs). Exceeding (AOOs). Exceeding the the lHGR LHGR limit limit could could potentially potentially result result in in fuel fuel damage damage and subsequent ami subsequent release reease of of radioactive radioactive materials.
materials. FuelFuel design design limitslimits areare specified to specified to ensure ensure that that fuel fuel system system damage, damage, fuel fuel rod rod railure failure or or inability inability to to cool the cool the fuel fuel does does not not occur occur during during iIle anticipated operating the anticipated operating conditions conditions identified in identified References 11 and in References and 2.
For GNF (uel, For fuel, lCO LCO 3.2.1 AVEARGE PLANAR 3.2.1 *AIJEARGE PLANAR LINEAR LINEAR HEAT GENERATION RATE (APlHGR)"
GENERATION (APLHGRy ensures that the rue! fuel design limits are not exceeded during normal operation and anticipated operational occurrences.
APPLICABLE APPLICABLE analy&al methods and assumptions used in evaluating the fuel The analytical SAFETY ANALYSES system design are presented in References 11 and 2. The fuel assembly SAFETY is designed to ensure (in in conjunction with iIle the core nuclear and thermal hydraulic design, plant equipment, instrumentation,instrumentation, and protection system) that the fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20. 20, 50, and 50.67.
The mechanisms that could cause fuel damage during normal normal operations and operational transients and that are considered in fuel evaluations are:
- a. Rupture of of the fuel rod cladding caused by strain from the relative expansion of of the UO U02 pellet; and and b.
- b. Severe overheating of the the fuel fllel rod rod cladding cladding caused caused by by inadequate inadequate cooling cooling A value value ofof 1%
'1% plastic plastic strain strain of of the the fuel cladding cladding has has been been defined defined as as the limit limit below below which which fuel fuel damage damage caused caused by by overstraining overstraining of of the the fuel fuel cladding cladding is not expected is not expected to to occur occur (Ref.
(Ref. 3). 3).
Fuel Fliel design design evaluations evaluations have have been been performed performed and and demonstrate demonstrate that that thethe 1% fuel 1% fuel cladding cladding plastic plastic strain strain design design limit limit isis not not exceeded exceeded duringduring continuous continuous operation operation with with LHGRs I..HGRs up to the liP to the operating operating limit limit specified specified in in the the COLR.
COlR. The The analysis also includes analysis also includes allowances allowances for for short short term term transient transient operation operation above above the the operating operating limit limit toto account account for for AOOs.
AOOs.
(continued)
Gnjnswick Brunswick Unit Unit 22 BB3.2.3-1 3.2.3-1 Revision Na.
Revision No. 62 62
PURPOSE PURPOSE RADIOACTIVE RELEASE FROM THE PLANT WITHIN LIMITS FAILURE FAILURE MECHANISM MECHANISM CAUSE OF CAUSE OF FAI LURE FAILURE LIMITING LIMITING CONDITION CONDITION ITEM ITEM MEASURED MEASURED LIMITING LIMITING OPERATION OPERATION LIMITING LIMITING PARAMETE PARAMETER R CALCULAT CALCULATED ED PARAMETE PARAMETER R Categories Categories K/A:
KIA: G2.02.25 G2.02.25 Tier Tier // Group:
Group: T3 T3 RO Rating:
RORating: 3.2 3.2 SRO Rating: 4.2 SRORating: 4.2 LP LP Obj:
Obj: CLSLP2OOB *O3 CLS-LP-200-B*03 Source:
Source: NEW NEW Cog Cog Level:
Level: LOW LOW Category Category 8:8:
- 71. Unit
- 71. UnitTwo Twostartup startupisisininprogress progresswhen whenthe theoperating operatingCRD CRDPumpPumptrips.
trips.
Thefollowing The followingplant plantconditions conditionsexist:
exist:
CRD pumps CRD pumps Unavailable Unavailable ReactorPressure Reactor Pressure 850psig 850 psig Chargingheader Charging headerpressure pressure 875psig 875 psig CRD ACCU M CRD ACCUM LO PRESS/HILO PRESS/HILEVEL LEVEL InIn alarm alarm Contro l Rod Control Rod 18-19 18-19 Full Fullcore coredisplay display AmberAccumulator Amber Accumulatorlight lightisislitlit Control Rod Control Rod 18-19 18-19 Full Full core coredisplay display Red Red Full Full Out Outlight light isis litlit Which one Which one ofofthe the following following describes describes the the required required action?
action?
A':I Immediately insert A Immediately insert aa manual manual reactor reactor scram.
B. Wait B. Wait 2020 minutes minutes withwith no no CRD CRD Pump Pump inin service service andand then then insert insert aa manualmanual reactorreactor scram.
C. Wait C. Wait for for AO AO confirmation confirmation thatthat the the accumulator accumulator alarm alarm isis due due to to low low pressure pressure and and then insert a manua then insert a manuall reactorreactor scram.
D. Immediately D. Immediately insert insert a manual manual reactor scram when aa second second accumulator accumulator alarm alarm isis receive d on received on the the full core display.
core display.
Feedback Feedback G2.02.39 K/A: G2.02.39 KIA:
EQUIPMENT CONTROL EQUIPMENT CONTROL Knowledge of Knowledge of less less than than oror equal equal toto one one hour hour Technical Technical Specification Specification action action statements statements forfor systems.
systems.
(CFR: 41.7/41.10/43.2/45.13)
(CFR: 41.7 /41.10 / 43.2 / 45.13)
ROISRO Rating:
RO/SRO Rating: 3.9/4.5 3.9/4.5 CLSLP008B*1 0 Objective: CLS-LP-008-B*10 Objective:
- 10. Given
- 10. Given plant plant conditions, conditions, determine determine proper proper operator operator actions actions ifif no no CRD CRD pumps pumps are are operating.
operating.
Reference:
Reference:
Unit Tech Spec Unit 22 Tech Spec 3.1.5 3.1.5 (Control (Control Rod Rod Scram Scram Accumulators),
Accumulators), Condition Condition D D Cog Level: High Cog Level: High Explanation:
Explanation:
Immediate scram Immediate scram required required byby Tech Tech Spec Spec 3.1.5, 3.1.5, Conditions Conditions C C&& D.
D.
The reactor The reactor must must bebe immediately immediately scrammed scrammed ifif either either the the Required Required Action Action and and associated associated Completion Completion Time associated associated with loss loss of of the CRD CRD charging charging pump pump (Required (Required Actions B.1 B.1 andand C.1)
C.1) cannot cannot be be met.
met. This ensures that that all insertable insertable control rodsrods are inserted and that that the reactor reactor is is in in aa condition that does not not require the require the active function function (Le.,
(i.e., scram) of the control rods.rods.
supplemental actions of OAOP-02.0.
Scram also required by supplemental Distractor Analysis:
Distractor Choice A:
Choice A: Correct Answer Choice B: Plausible because with reactor pressure.::: pressure > 950 psig combined with charging header pressure
<<940940 psig, 20 minutes is allowed for restoration of charging water header pressure.
Choice C: Plausible because accumulator alarm could be due to High water level which would still provide for sufficient accumulator pressure to fully insert the control rod. Revision 12 of 2AOP-02.0 provided guidance on time frame to IMMEDIATELY IMMEDIATELY insert a manual scram upon receipt of the first HCU low pressure alarm (A-07 6-1, confirmed by amber light on Full Core Display).
Choice D: Plausible because waiting for the second accumulator is applicable to reactor pressures
> 950 psig for restoration of charging water header pressure.
SRO SRO Only Only Basis:
Basis: N/A Notes Notes a.
- a. IF reactor pressure IF during during startup pressure is startup or is less or shutdown less than than 950 shutdown evolutions),
950 psig psig (e.g.,
evolutions), AND (e.g.,
AND CRD GRD o
pressure pressure CANNOT CANNOT be be restored restored to to greater greater than than or or equal to 940 psig with equal to 940 psig with either either CRD CRD Pump,Pump, THENTHEN upon upon receipt receipt ofof the the first first HCU HCU low low pressure pressure alarm alarm (A-07 (A-07 6-i, 6-°', confirmed confirmed by by amber amber lightlight on on Full Full Core Core Display)
Display) IMMEDIAT IMMEDIATELY ELY INSERT INSERT aa manualmanual reactor reactor SCRAM SCRAM immediately immediately..
b.
- b. IF reactor IF psig, AND psig, AND two pressure is reactor pressure two or is greater or more greater than more HCU HCU [ow than oror equal low pressure equal to to 950 pressure alarms alarms 950 o (A-07 (A-07 6-i)6-°1) are are received received (confirmed (confirmed by by amber amber light light on on Full Full Core Display), THEN Core Display), THEN ENSURE ENSU.RE CRD CRD pressure pressure isis restored restored toto greater greaterthanthan oror equal equal to to 940 940 psig psig within within 20 20 minutes.
minutes.
QAOP-02.O
\ OAOP-02.0 Rev. *18 Rev. 18 Page 44 Of9\
Page of 9 Control Rod Control Rod Scram Scram Accumulators Accumulators 3.1.5 3.1.5 ACTIONS (continued)
ACTIONS (continuedi CONDITION CONDITION REQUIRED ACTION REQUIRED ACTION COMPLETION COMPLETION TIME TIME 8.
B. Twa or Two or more more control control rod rod 5.1 6.1 Restore charging Restore charging water water 20 20 minutes from minutes from scram accumulators scram accumulators header pressure header pressure to to discovery of discovery of inoperable with inoperable with reactor reactor a 940 psig.
- 940 psig. Condition 65 Condition steam dome steam dome pressure pressure concurrent concurrent with with a 950 psig.
2:950 psig. charging charging water water header header pressure pressure
< 940
<: 940 psig psig AND 5.2.1 6.2.1 NOTE
NOTE-------------
Only applicable if the associated control roo rod scram time was within the of Table 3.1.4-1 limits ofTable dunng the last scram time during Surveillance Surveillance.
the associated Declare tile 1 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> control conlrol rod scram time slow.
"slow."
OR 6.2.2 6.2.2 Declare thetile associated 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> control control rod rod inoperable.
(continued)
(continued)
Biun wick Unit Brunswick Unit22 3.1-15 3.1-16 AmendmentNo.
Amendment No.233 233
Control Rod Control ScramAccumulators RodScram Accumulators 3.1.5 3.1.5 ACTIONS (continued)
ACTIONS (continuedr CONDITION CONDITION REQUIRED ACTION REQUIRED ACTION COMPLETION COMPLETION TIME TIME C. One C. One oror more more control control rod rod Ci C., Verily all Verify all control control rods rods Immediately Immediately upon upon scramaccumulators scram accumulators associated with associated with inoperable moperable discovery discover) ofof inoperable with inoperable with reactor reactor accumulators are accumulators arefuny fully charging charging water water steam dome steam dome pressure pressure inseried.
inserted. header header pressure pressure 950 psig.
<<< 950 psig. <<< 940 940 psig psig AND AND C.2 C.2 Declare the Declare the associated associated -I1 hour hour control rod control rod inoperable.
D, D. Required Action Required Action B.l Bi or orCi C:l Di 0.1 -----.-----NOTE-------------
NOTE------------
and associated and associated Completion Completion Not applicab!e Not applicable ifif all all Time not Time met.
not met. inoperable control inoperable control rod rod scram accumulators scram accumulators are are associated with associated with fully fully inserted control inserted control rods.
rods.
---~-------------------------
Manually scram Manually scram the reactor. Immediately the reaclor. Immediately SURVEILLANCE REQUIREMENTS SUR\lElllJ>.NCE REQUIREMENTS SURVEILLANCE SURVEILLANCE FREQUENCY FREQUENCY SR 3.1.5.1 SR 3.1.5.1 Verify each control rod scram accumulator VeriPj accumuiator pressure is 7 days 940
- . 940 psig.
Brunswick Brunswick Unit Unit 22 3.1-17 3.1-'17 Amendment No.
Amendment No. 233 233 Categories Categories K/A:
KIA: G2.02.39 G2.02.39 Tier Tier // Group:
Group: T3 T3 RORating:
RORating: 3.9 3.9 SRO Rating:
SRORating: 4.5 4.5 LP LP Obj:
Obj: CLSLPOO8B CLS-LP-008-B *10
- 1O Source:
Source: BANK BANK Cog CogLevel:
Level: HIGH HIGH Category Category 8:8: YY
- 72. The H2
- 72. The Controller, HWCH-FIC-5713, Flow Controller, H2 Flow HWCH-FIC-5713, fails fails high high on on Unit Unit Two Two while while operating operating atat 55% power.
55% power.
Which one Which one of of the following identifies the following identifies the the impact impact this failure will this failure will have have onon plant plant radiation radiation levels (assume levels (assume HWC HWC remains remains inin service) service) andand thethe reason reason forfor this this impact?
impact?
The Main The Main Steam Steam Line Line Radiation Radiation Monitors Monitors will will indicate indicate (1)
(1) due due toto the the (2)
(2)
A. (1)
A. lower (1) lower (2) reduced production of (2) reduced production Nitrates (N03) of Nitrates (N03)
B. (1)
B. lower (1) lower (2) reduced production of (2) reduced production Ammonia (NH3) of Ammonia (NH3)
C. (1)
C. higher (1) higher (2) increased production (2) increased production of Nitrates Nitrates (N03)
(N03)
Dv D~ (1) higher (2) increased production (2) increased production of Ammonia (NH3) (NH3)
Feedback Feedback K/A: G2.03.14 KIA: G2.03.14 Radiation Control Radiation Control Knowledgeof Knowledge ofradiation radiation or orcontamination contamination hazards hazardsthat thatmay mayarise ariseduring during normal, normal, abnormal, abnormal, or or emergency conditions emergency conditions or oractivities.
activities.
(CFR: 41.12/43.4/45.10)
(CFR: 41.12/43.4/45.10)
RO/SRO Rating:
RO/SRO Rating: 3.4/3.8 3.4/3.8 CLSLP59*14 Objective: CLS-LP-59*14 Objective:
- 14. Explain
- 14. Explain whywhy background background radiation radiation levels levels outside outside primary primary containment containment increase increase when when the the HWC HWC System isis placed System placed inin service.
service.
- 15. State
- 15. State the the parameter parameter used used for for the reactor power the reactor power levellevel reference reference input input to to the the hydrogen hydrogen injection flow controller, injection flow controller, and and explain explain why why itit isis used.
used.
Reference:
Reference:
SD-59, Revision 14, SO-59, Revision 14, Page Page 8, 8, Section Section 1.3.21.3.2 Cog Level:
Cog Level: Low Low Explanation:
Explanation:
The implementation of The implementation of Hydrogen Hydrogen Water Chemistry Chemistry (H2 (H2 injection) injection) alters alters the the Nitrogen-16 Nitrogen-16 carryover carryover ratio.
ratio.
The net The net production production of of Nitrogen-16 Nitrogen-16 is is not not influenced influenced by Hydrogen injection.
by Hydrogen injection. The excess excess Hydrogen Hydrogen injected injected into the into the reactor reactor coolant coolant creates creates the driving force the driving force to to shift shift the the Nitrogen-16 Nitrogen-16 distribution ratio, resulting in in aa larger fraction larger fraction of of the Nitrogen-16 forming volatile Ammonia and aa smaller smaller fraction forming Nitrites Nitrites and and Nitrates. This Nitrates. This additional additional volatile Ammonia is then carried over in the reactor reactor steam resulting in higher higher background radiation background radiation levels. Any increase in Hydrogen injection rates will result in a proportional increase in background in background radiation radiation levels and vice-versa.
Distractor Analysis:
Oistractor Analysis:
Choice A: Plausible because lower reactor power level would normally lower MSL rad monitor and at Choice A: Plausible at lower reactor lower reactor power levels, production of Nitrates (N03) does reduce but does not impact MSL MSL levels.
rad levels.
rad Choice Choice B: B: Plausible Plausible because lower reactor power level would normally normally lower MSL rad monitor monitor and at lower lower reactor power levels, production of of Ammonia (NH3) (NH3) does does reduce but the excess H2 H2 shifts shifts to produce to produce more.
Choice Choice C: C: Plausible Plausible because because MSL MSL rad monitors monitors will indicate higher and and atat lower lower reactor power power levels, levels, production production of of Ammonia Ammonia (NH3) (NH3) does does reduce reduce but but the the excess excess H2 H2 shifts shifts toto produce produce more.
more.
Choice Choice D: 0: Correct Correct Answer Answer SRO SRO Only Only Basis:
Basis: N/AN/A Notes Notes s ieriiuveu ieu i,u, seric.
1.3.2 1.3.2 Radiological Radiological Implications ImplicationsOf OfHydrogen Hydrogen WaterWater Chemistry Chemistry Control Control The primary The primary source source of ofbackground radiation levels background radiation levelsdtring during reactor reactor operation, operation, nearnearsteam steam lines linesoutside outside the the Primary PrimaryContainment Containment, isis attributed attributedtoto thethedecay decayof ofNitrogen-16 Nitrogen-16(N (N 15). N 1N16 has hasaahalf-hfe half-fifeof of 7.1 seconds and decays 7.1 seconds and decays withthe with theemission emissionof ofaa high-energy high-energygamma gamma (13.1 (B.1 Mev).
Mev).
I so-s S0-59 Rev. 14 Rev. 14 Page770f5S1 Page of 55
The major The majorsources sourcesof Nitrogeninin aa B\I'v'R ofNitrogen 8WR are are from from Oxygen-16 Oxygen-16 and and from from the leakage the leakage of nitrogen based ofnitrogen based chemical chemical compounds compounds from from the the RWCU RWCU and Condensate and Condensate deminerafizers.
demineralizers. Oxygen-16 Oxygen-i 6 formsforms Nitrogen-16 Nitrogen-i 6 via via aa neutron-proton reaction.
neutron-proton reaction.
(Qib + 16 + p)
(Oi6 + 1)-->> N N + p)
When using When using normal normal water water chemistry chemistry methods methods (Le., (Le.. without without H22 injection),
H injection).
aa major major portion portion of of the the Nitrogen-16 Nitrogen-i 6 present present inin the the reactor reactor coolant coolant combines '.'lith combines the free with the free Oxygen Oxygen to form water-soluble to form water-soluble Nitrites Nitrites (NOz)
) and 2
(NO and Nitrates (NOJ).
Nitrates ). These 3
(NO These compounds compounds are are circulated circulated through through thethe reactor reactor coolant systems coolant systems and and are are uflimately uhmately removed removed fly by the the RWCU RWCU System.
Systerm A A smaller fraction smaller fraction of of the the Nitrogen-'16 Nitrogen-i 6 isis carried carried over over in in the the steam steam inin the the form of form of Nitrogen Nitrogen gasgas (N2)
) and 2
(N and Ammonia Ammonia (NH3) ) and 3
(NH and isis the the predominate predominate contributor to contributor to background background radiation radiaon levels.
levels.
The implementation implementation of HydrogenHydrogen Water Chemistry Chemistiy (H 22 injection) alters (H
Nitrogen-i 6 carryover the Nitrogen-16 carryover ratio. The net production production of Nitrogen-16 Nitrogen-i6 is not not influenced by inifuenced by Hydrogen Hydnxen injection.
injection.
The excess Hydrogen injected into the reactor coolant creates the Nitrogen-lB distribution ratiO, driving force to shift the Nitrogen-16 ratio, resulting in a larger fraclion frachon of the Nitrogen-16 Nitrogen-lB forming volatile Ammonia and a smaller fraction forming Nitrites and Nitrates. Nitrates This additional volatile Ammonia is then carried over in the reactor steam resulting in higher background radiation levels. Any increase in Hydrogen injection rates will result in a proportional increase in background radiation levels and wiU vise-versa.
High injeclion injection rates during normal operation have impacted BNP's BNPs accumulated radiation exposure to such an extent that thaI shielding has been installed to minimize the shine "shine" from the Turbine Building 70 70' elevation to outiying outlying buildings. Essentially, aa roof has been built built over each MSRMSR using the the outside Turbine Building wall and the wall waft between the the High High Pressure Pressure Turbine Turbine andand the the Main Main Generator Generator for supporisupport Structural structural steel steel between between the Moisture SeparatorSeparator Reheaters (MSR) and the turbines serves serves as interior interior support to this roof roof, which covers the the MSRs MSRs and and the the crossover crossover piping.
piping. Carbon Carbon steelsteel plates plates makemake upup the the roof roof and and areare removable, removable, allowing allowing access access forfor maintenance maintenance.. Additionally Additionally,, aa 4
4' masonry masonry wall wall (originally (oliginaf[y temporary temporary shielding) shielding) has has been been installed installed onon the the East East and and West West walls waifs outboard outboard thethe MSRs MSRs wherewhere the the elevation elevation of ofthe the crossover crossover piping piping exceeds exceeds the the concrete concrete shielding shielding wall.wall. TheThe MSRs, MSRs, Main Main Turbines Turbines and and crossover crossover piping piping account account for for tip up to to 80%
80% of of the the occupational occupational radiation radiation exposure exposure due due toto Hydrogen Hydrogen WaterWater Chemistry.
Chemistry.
SD-59 lSD-59 Rev.
Rev. 1414 Page 88 of Page of55 55 I Categories Categories K/A:
KIA: G2.03.14 G2.03.14 Tier/Group:
Tier / Group: T3 T3 RO Rating:
RORating: 3.4 3.4 SRO SRORating:
Rating: 3.8 3.8 LP Obj:
LPObj: CLSLP59*1 CLS-LP-59* 14 4 Source:
Source: NEW NEW Cog CogLevel:
Level: LOW LOW Category Category 8:8: YY
73.
73.
Which one of the following identifies the current Drywell radiation level?
-2O R/h A. -20 B
B~ RIh
-200 R/h c.
C. -I000R/h
-1000 R/h D. 10000R/h
-10000 R/h
Feedback Feedback K/A: G2.03.15 KJA: G2.03.15 Radiation Control Radiation Control Knowledgeof Knowledge ofradiation radiation monitoring monitoringsystems, systems,suchsuch asasfixed fixed radiation radiation monitors, monitors, portable portablesurvey survey instruments, personnel instruments, personnel monitoring monitoring equipment, equipment, etc.etc.
(CFR: 41.12/43.4/45.9)
(CFR: 41.12 /43.4/45.9)
ROISRO Rating:
RO/SRO Rating: 2.9/3.1 2.9/3.1 Objective: CLS-LP-1 1 .1*03a Objective: CLS-LP-11.1 *03a
- 3. Describe
- 3. Describe the function/operation of the function/operation the following:
of the following:
a.a. Drywell Drywell High High Range Range Radiation Radiation Monitors Monitors
Reference:
Reference:
SD-11.1 Section 2.5 SD-11.1 Section 2.5 Cog Level:
Cog Level: Low Low Explanation:
Explanation:
Drywell high Drywell high range range area area monitors monitors provide provide indications indications of of gross gross fuel fuel failure failure and and are are used used to to determine determine emergency plan emergency plan emergency emergency action action level level associated associated with with abnormal abnormal core core conditions.
conditions. With With the the function function switch in switch in the the E1-E4, E1-E4, meter meter readings readings are are taken taken from from the the lower lower scale scale between between 1010 - 10000
- 10000 R/h.
R/h. Current Current indication of indication of 200 R/hRJh Distractor Analysis:
Distractor Analysis:
Choice A:
Choice A: Plausible Plausible ifif function switch is is not not taken into into account would be be 20 R/h.
R/h.
Choice B:
Choice Correct answer B: Correct Choice C:
Choice C: Plausible Plausible if read directly off the upper scale Choice D:
Choice D: Plausible if read off the upper scale and adjusted by a factor of 10 for function switch position.
Basis: N/A
Notes Notes FIGURE 1",1-FIGURE 11.1-5 5
DRYWELL HIGH DRY\NELL HIGH RANGE RANGE RADIATION RADIATION MONITOR MONITOR CONTROUTRIP CONTROL/TRIP UNIT UNIT 10"-
1O- 10" 1O 10'-10' 1o-1ct G 1O 10" 10 2 10" '10" 1<1'-10' 10'-10*
- 1 O*-1 1I 44
, 10 44 68
':?, .. (> a 43
.. Q
-tOn
- )ot a
1iI iJ
@ ALL (RED)
(REO) ...
1-10~
1-1o TEST TEST R/hr R/hr otjcic SUICK ALERT ALERT HIOH HIGH lies cs SAfE SAFE HNflL ClfAiNNeL RESET RESET TEST Categories Categories K/A:
KIA: G2.03.15 G2.03.lS Tier/Group:
Tier / Group: T3 RO Rating:
RORating: 2.9
2.9 Rating
3.1 SRORating:
Obj: CLSLP59*1 CLS-LP-S9*14 4 Source:
Source: NEW NEW Cog Level:
Cog Level: LOW LOW Category Category 8:8:
- 74. Which
- 74. Which oneoneof ofthe thefollowing following Reactor ReactorBuilding Buildingradiation radiationannunciators annunciators requiresrequiresentry entryinto into RRCP, Radioactivity RRCP, RadioactivityRelease ReleaseControlControl Procedure?
Procedure?
A. AREA RAD A. AREA RADRX RXBLDG BLDGHIGH HIGH By R)(BLDG B RX BLDG ROOF ROOF VENT VENTRAD RAD HIGHHIGH C. PROCESS C. PROCESS RX RXBLDG BLDG VENT VENTRAD RAD HIGH HIGH D. PROCESS D. PROCESS RX R)(BLDG BLDG VENT VENTRAD RAD HI-HI HI-HI Feedback Feedback K/A: G2.04.01 KIA: G2.04.0l Emergency Procedures Emergency Procedures II Plan Plan Knowledge of Knowledge EOP entry of EOP entry conditions conditions and and immediate immediate action action steps.
steps.
(CFR: 41.10/43.5/45.13)
(CFR: 41.10/43.5145.13)
RO/SRO Rating:
RO/SRO Rating: 4.6/4.8 4.6/4.8 LOlCLSLP300N*002 Objective: LOI-CLS-LP-300-N*002 Objective:
- 2. Given
- 2. Given plant plant conditions, conditions, determine determine ifif OEOP-04-RRCP OEOP-04-RRCP should should bebe entered.
entered.
Reference:
Reference:
RRCP RRCP Cog Level:
Cog Level: LowLow Explanation: Brunswick Explanation: Brunswick does does not have any immediate operator actions in any EOP.
Annunciator requires Annunciator requires immediate immediate operator action of entry into RRCP. RRCP provides guidance to to the the operator for operator for minimizing minimizing off-site off-site radioactivity releases up to and including events involving substantial degradation degradation of of all all of of the fission product barriers (e.g., fuel, fuel clad, reactor vessel pressure boundary, primary containment, primary containment, and secondary containment) containment)..
RRCP RRCP andand SCCP SCCP areare used used concurrently to control releases from primary systems. This procedure controls controls non-primary non-primary system releases through through actions actions incorporated incorporated in in the the non-PSTG non-PSTG legs legs of of the the procedure.
procedure.
Distractor Distractor Analysis:
Analysis:
Choice Choice A:
A: Plausible Plausible because because SCCP SCCP entry entry would would be be appropriate.
appropriate.
Choice Choice B:B: Correct Correct Answer Answer Choice Choice C:C: Plausible Plausible because because isis easily easily confused confused with with the the roof roof vent vent alarm alarm and and isis aa SCCP SCCP entry entry condition.
condition.
Choice Choice D:0: Plausible Plausible because this annunciator because this annunciator provides provides indication indication of of Secondary Secondary Containment Containment auto auto isolation isolation setpoint setpoint and and isis easily easily confused confused with with the the roof roof vent vent alarm.
alarm.
SRO SRO Only Only Basis:
Basis: N/A N/A Notes Notes
RADIOA CTlVllY RELEASE CONTROL RR-1 RR-1 ENTRY CONDITIONS:
ENTRY CONDITI ONS:
- MAIN STEAM LINE RAD HI RAD HIANNUN ANNUN SETPOIN SETPOINT T EXCEEDED EXCEED ED (UA-23,2 (UA-23,2-- 6)
- PROCES PROCESS HIANNUN HI ANNUN OFF-GAS S OFF-GA S RAD SETPOINT SETPOIN T EXCEEDED EXCEED (UA-03,s* 2)
- HIGHAN RX BLDG ROOF VENT HIGHANNUN IENT RAD NUN SETPOIN SETPOINT T EXCEEDED EXCEED ED (UA-03,2 (UA-03,2--3)3)
- TURB BLDG VENT RAD HIGHANNIJN HIGHAN NUN SETPOIN SETPOINT T EXCEEDED EXCEED (UA-03,3-- 3)
ED (UA-03,3
- PROCES PROCESSOGS OG VENT PIPE RAD HIANNUN HIANNU N SETPOIN SETPOINT T EXCEEDED EXCEED UA-03,6-- 4)
ED (UA-03,6 (STACK)
- SERVICE EFFLUENT WTR EFFLUE RAD HIGH ANNUN NT SETPOINT SETPOIN EXCEEDED T EXCEED ED
{UA- 03,5-(UA- 03,5-5) 5)
- ANY UNMONI UNMONITOREDTORED OFF- SITE OFF-SIT E RADIOA RADIOACTIVfly CTIVllY RELEASE RELEAS E
- CALCULCALCULATED LIMITOF ATED DOSE RATE LlMITOF "NOBLE NOBLE GAS INSTANTANEOUS INSTANT ANEOUS RELEASE
\ \ RATE DETERM DETERMINATION INATION"
\ (E&RC-202o)
. (E&RC-2 EXCEEDED 020) EXCEED ED /
RR-2
/SEcONDARY CONTMNMENT\
( CONTROL CONTROL PROCEDURE PROCEDURE SCCP1 ENTRYCONDITIONS ENTRY CONDITIONS
- AREATEMPABOVE
- AREA TEMP ABOVE THEMAX THE MAX NORM NORM OPERATING VALUE OPERATING VALUE TABLE 1I TABLE
- ARE DIFFERENTIAL ARE DIFFERENTIAL TEMPABOVETHE TEMP ABOVE THE MAX NORM MAX NORM OPERATING OPERATING VALUE TASLE2 VALUE TA8LE2
- REACTOR BLDG REACTOR BLDG VENTILATION EXHAUST VENTILATION EXKAUST RADIATION LEVEL RADIATION LEVEL EXCEEDS 33 mR/HR EXCEEDS mRIHR
- AREA AREARADIATION ABOVE LEVEL RADIATION LEVEL ABOVE THE THE MAX NORM OPERATING NORM OPERATING VALUETABLE3 VALUE TABLE 3
- HPCI,RHR,OR HPCI, RHR, OR CORESPRAY CORE SPRAY ROOM ROOM WATER WATER LEVEL LEVEL EXCEEOS 86 INCHES EXCEEDS INCHES ABOVE FLOOR ABOVE FLOOR TABLE 4 CP2
Unit 2 Unit2 APP APPUA-03 UA-032-32-3 Paget Page 1 ofof1I RXBLDG RX BLDGROOF ROOFVENT VENTRAD RADHIGHHIGH AUTO ACTIONS
.A.UTO ACTIONS NONE NONE CAUSE
'1.I. High noble High nobe gasgas concentration concentration inin the the Reactor ReactorBuilding Butding ventvent exhaust.
exhaust.
2.
- 2. Circuit malfunction.
Circuit malftincton.
OBSERVATIONS OBSERVATIONS 1.
- 1. The Reactor The Reactor Building BuiIdng Roof Roof Vent Vent Radiation Radiation Monitor, Monitor, CAC-.A.QH-1264-3, CAC-AOH-1264-3, on on Panel Panel XU-55, isis in XU-55, in alarm.
aarm.
2.
- 2. The 1264 The 1264 chart chart recorder recorder isis trending trending upward.
upward.
ACTIONS ACTIONS 1.
- 1. Enter EOP-04-RRCP, Radioactivity Enter EOP-04-RRCP, RadioactMty Release Release Control, Control, and and execute execute concurrently concurrently with this with this procedure.
procedure.
2.
- 2. IfIf steam steam leaks leaks inin the Reactor BuUding the Reactor 8uiding are are causing causing local local area area radiation radiation levels levels or or ambient temperatures ambient temperatures to to increase, increase, enter enter EOP-03-SCCP, EOP-03-SCCP, SecondarjSecondary Containment Confrot, as Control, appropriate.
as appropriate.
3.
- 3. Refer to Refer to AOP-05.0, AOP-05.O, Radioactive Radioactive Spills, Spills. High High Radiation, Radiation, andan Airborne Acti'Jity.
ActMty.
4.
- 4. circuit malfunction is suspected, ensure that a Trouble Tag is prepared.
If a circuit DEVICE/S ETPOINTS DEVICEfSETPOINTS Rad Monitor CAC-AQH-1264-3 Rad CAC-AQH-1264-3 Variable (contact E&RC for current setpoint)
POSSIBLE PL4.NT POSSIBLE PLANT EFFECTS EFFECTS I.
- 1. Possible release to environs.
2.
- 2. This This annunciator annunciator is required to be operab!eoperabl*e to support reactor building ventilation radiation radiation monitor monitor operabWty; operability; annunciator inoperability inoperability iIlwill result in aa Required Cornpensato Compensatory rj Measure.
Measure.
REFERENC REFERENCES ES 1.
- 1. LL-9353-31 LL-9353-31 2.
- 2. EOP-04-RRC EOP-04-RRCP, P, RadioactMty Radioacti.ity Release Release Control Control 3.
- 3. AOP-05.0, AOP-05.0, Radioactive Radioactive Sphls, SpDls, High High Radiation, Radiation, and and Airborne
.A.irborne Activity Activity 4.
- 4. EOP-03-SCc EOP-03-SCCP, P. Secondary Secondary Containment Containment Control Control 5.
- 6. ODCM ODCM 7.3.2 7.3.2 and and 7.3.7 7.3.7 2APP-UA-03 12APP-UA-03 Rev. 46 Rev. 46 . Page 19 Page 19 of 631 of 63
AREA RAD
.AREA RADRXRX BLDGBLDGHIGHHIGH AUTOACTIONS A.UTO ACTIONS NONE NONE CAUSE
'I.1. High radiation High one or level inin one radiation level or more more ofofthe the following following areas:
areas:
- a. a. Core Spray Core Spray Pump Pump Room Room 2A 2A (channel (channel 'IS).
iS:.
- b. b. Core Spray Core Spray Pump Pump Room Room 2B 28 (channel (channel 16).
16).
c.c. RHR Heat RHR Heat Exchanger Exchanger and and Pump Pump Room Room 2A 2A (channel (channe 17).
17>.
- d. d. RHR Heat RHR Heat Exchanger Exchanger and and Pump Pump Room Room 2B 28 (channel (channel 18).
iS).
- e. e. Reactor Building Reactor Building air air lock lock (channel (channel 19}.
19).
f.f. Oryweil entrance Driwell entrance (channel (channel 20).
20).
g.g. Decontamination room Decontamination room {channel (channel 22}.
22:1.
h.h. Eguipnent entry Equipment enby (channel (channei 23).23).
- 1. Reactor Building Reactor Building sample sample station station (channel (channel 24).
24).
j.j. Reactor Building Reactor Building air air lock lock {E1 (El 50'}
- 50) (channel (channel 25).
25).
Iek Spent fuel Spent fuel pool pool cooling cooling system system (channel (channel 30).
30).
2.
- 2. Transfer of Transfer of either either units units RWCU RWCU backwash backwash receiving receiving tank tank to to Radwaste Radste (Channel (Channel 25) 25) 3.
- 3. Circuit malfunction.
Circuit malfunction OBSERVATIONS OBSERVATIONS 1.
- 1. Affected ARM indicator and trip until Upscale Affected Upsca!e light illuminated on Panel Panel Hl2P60O.
H12-P600.
ACTIONS A.CTIONS 1.
- 1. Refer to EOP-03-SCCP, Refer EOP-03-SCCP, Table 3; enter EOP-03-SCCP EOP-03-SCCP as appropriate.
2.
- 2. Refer to Refer to AOP-05.0, Radioactive Spils. Spills, High Radiation, and Airborne ActivW/.
Activity.
3.
- 3. circuit malfunction is suspected, ensure that a Troub!e If a circuit Troube Tag is prepared.
DEVICE/S ETPOINTS DEVICE/SETPOINTS Channel Channel IS Relay 15 K2 Relay mRlhr 20 mR/hr Channel Channel 16 K2 Relay 16 1<2 Re!ay 20mRlhr 20 mR/hr Channel Channel 17 K2 Relay 17 1<2 Relay 20 mR/hr 20 mRlhr Channel Channel 18 K2 Relay 18 1<2 20 mR/hr 20 mRlhr Channel Channel 19 19 1<2 K2 Relay Relay 20 mR/hr 20 mRlhr Channel Channel 20 K2 Relay 20 1<2 Relay 20 mR/hr 20 mRlhr Channel Channel 22 K2 Relay 22 K2 Relay 20mRlhr 20 mR/hr Channel Channel 23 K2 Relay 23 1<2 Relay 20mRlhr 20 mR/hr Channel Channel 24 24 1<2 K2 Relay Relay 20 mR/hr 20 mRlhr Channel Channel 25 K2 Relay 25 1<2 Relay 20 20 mR/hr mRlhr Channel Channel 30 30 1<2 K2 Relay Relay 50mRlhr 50 mR/hr 2APP-UA-03 12APP-UA-03 Rev.
Rev. 4646 Page 23 Page of 63 23 of 631
Unit Unit 22 APP APPUA-03 UA-033-5 3-5 Pagel Page 1 of*l of I PROCESS RX PROCESS RX BLDGBLDG VENTVENT RAD RAD HI-HI HI-HI AUTO ACTIONS AUTO ACTIONS
- 1.I. Reactor Building Reactor Building ventilation ventilation system system trips trips and isolates.
and isolates.
2.
- 2. Standby gas Standby trains start treatment trains gas treatment start.
3.
- 3. IfIf open, open. the inboard and the inboard and outboard outboard primary primarj containment containment purge purge and and vent vent valves valves close.
close.
4.
close.
CAUSE CAUSE I.
- 1. High airborne High airborne activity activity levels levels in the Reactor in the Reactor Building Building exhaust exhaust plenum.
plenum.
2.
- 2. Circuit malfunction.
Circuit malfunction.
OBSERVATIONS OBSERVATIONS I.
- 1. Reactor Building Reactor Building exhaust exhaust plenum plenum rad rad monitor monttor indicates indicates greater than 44 mRfhr greater than niRfhr onon H12-P656.
Panel H12-P606.
Panel
- 2. PROCESS RX BLDG BLDG VENT RAD HIGH HIGH (U.A.-03 (UA-g3 4-5) alamt 4-5 alarm.
ACTIONS ACTIONS 1.
- 1. Vertty auto actions.
Vern,!
2.
- 2. Refer to AOP-05.0, Refer AOP-G5.O, Radioactive Spills, Spills, High Radiation, and Airbome Airborne .A.ctillity.
ActMty.
- 3. Enter EOP-03-SCCP, Enter EOP-03-SCCP, SecondarySecondarl Containment Control.
4.
- 4. EOP-04-RRCP, Radiological Release Control; enter as appropriate.
Refer to EOP-04-RRCP, 5.
- 5. IfIf a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.
DEVICEiSETPOINTS DEVICE/SETPOINTS 012-RM-K609N8 Rad Monitor D12-RM-KS09A1B mRhr 4 mRlhr POSSIBLE PLANT EFFECTS POSSIBLE PL4.NT 1.
- 1. Possible release to environs environs in in excess of ODCM 7.3.7. 7.3.7.
2.
- 2. This This annunciator annunciator is required to to be operable operable to support Rx Bldg Vent Rad Monitor Monitor operability; annunciator inoperability inoperabilitjr will result in aa LCO.
REFERENC REFERENCES ES 1.
- 1. LL-9353-35 LL-9353 - 35 2.
- 2. AOP-05.O.
AOP-05.0, Radioactive Radioactive Spills, Spills, High High Radiation, Radiation, and and Airborne Airborne Activity Activity 3.
- 3. EOP-03-SCC EOP-03-SCCP P 4.
- 4. EOP-04-RRC EOP-04-RRCP P 5.
- 5. Technical Technical Specincation Specifications s 3.3.6.1, 3.3.S.1, Table Table 3.3.6.1-1 3.3.6.1-1 Function flmction 2d 2d and 3.3.6.2, Table and 3.3.6.2. Table 3.3.6.2-1 3.3.6.2-1 Function function 33 6.
- 6. ODCM aDCM 7.31 7.3.7 2APP-UA-03 12APP-uA-03 Rev.
Rev. 46 46 Page 32 Page 32 of 631 of 63
Unit2 Unit 2 APP APPUA-D3 UA-034-54-5 Page Page11 of'l of I PROCESS RX PROCESS RX BLDGBLDG VENTVENTRAD RAD HIGHHIGH AUTO ACTIONS
.A.UTO ACTIONS NONE NONE CAUSE 1.
- 1. High airborne activity High airborne activity inin Reactor Building ventilation Reactor Building ventilation exhaust exhaust plenum.
plenum.
2.
- 2. Circuit malfunction.
Circuit malfunction.
OBSERVATIONS OBSERVATIONS 1.
- 1. Reactor Building Reactor Building Vent Vent RadRad Recorder Recorder D12-RR-R6DS DI2-RR-R6DS Channel Channel AA or or B B indicates indicates high high radiation level.
radiation tCCi.
2.
- 2. Reactor Building Reactor Building Exhaust Exhaust Plenum Plenum RadRad Monitor Monitor Channel Channel A oror B B indicates indicates greater than 3 mRlhr than mRihr on Panel Panel H12-P6D6 H12-P606..
ACTIONS
.A.CTIONS I.
- 1. Enter EOP-03-SCCP.
Enter EOP-03-SCCP. Secondary Containment Control. ControL 2.
- 2. Refer to AOP-OS.O, Refer AOP-0S.O. Radioactive Spills, Spills. High Radiation, and Airborne .A.ctivity.
ActivIty.
3.
- 3. circuit malfunction is suspected, ensure that a Trouble Tag is prepared.
If a circuit If OEVICE/SETPOINTS DEVICEfSETPOINTS D12-RR-R605 red or black D'12-RR-R6D5 back pen mRlllr 3 mRThr POSSIBLE POSSIBLE PLANT PLAI'JT EFFECTS 1.
- 1. Possible Possible release to environs.
2.
- 2. IfIf airborne airborne activity activity increases increases to to 44 rnRihr, mRfhr, Reactor Reactor Building Building HVAC HVAC isolation, isolation, aa Group Group 66 isolation, isolation, drywell dfljwell purge isolation, isolation, and and initiation of the Standby initiation of Standby Gas Gas Treatment Treatment System System will will occur.
occur.
REFERENC REFERENCES ES I.
- 1. LL-9353 LL-9353 35 2.
- 3. EOP-03-SCC EOP-03-SCCP P 4.
- 4. Plant Plant Modincation Modification 35-081 85-081 2APP-UA-03 I2APP-UA-03 Rev.
Rev. 46 46 Page41 Page 41 of 63 1 of63 Categories Categories K/A:
KIA: G2.04.O1 G2.04.01 Tier/Group:
Tier / Group: T3 T3 RO Rating:
RORating: 4.6 4.6 SRO SRORating:
Rating: 4.8 4.8 LP Obj:
LPObj: LOICLSLP3 LOI-CLS-LP-300-N*002OON*0O2 Source:
Source: BANK BANK Cog CogLevel:
Level: LOW LOW Category Category8:8: YY
- 75. An ATWS has occurred on Unit One with the following plant conditions:
Reactoorr Water Level React 130 inches (stable)
Injection Syste System mss CRD Reactoorr Powe React Powerr APRM downscale lights are illumin illuminated ated Control Rods 19 rods failed to insert SRVs All closed Suppre Suppr ssion Pool Temp ession Temp.. 92°0 F 92 Which one of the followfollowing choices correc correcttly ly compl completes etes the statement below lAW LPC?
Reacto React orr Recirculation pump pumpss (1) require requir edd to be tripped and the SLC Pump Pumpss (2) require requir d to be started.
ed A(1 A'! (1)) are not (2) are not B. (1) are not (2) are C. (1) are (2) are not D. (1) are (2) are
Feedback Feedback K/A: G2.04.09 KIA: G2.04.09 EmergencyProcedures Emergency ProceduresI IPlan Plan Knowledgeof oflow lowpowerlshutdown powerlshutdown implications implicationsininaccident accident(e.g.,
(e.g.,loss lossof ofcoolant coolantaccident accidentor Knowledge orloss loss ofofresidual residual heat heatremoval) removal)mitigation mitigationstrategies.
strategies.
(CFR:41.10 (CFR: 41.10143.5 I 43.5145.13)
/45.13) e.g. is e.g. is an anabbreviation abbreviationfor forthe thelatin latinphrase phraseexempli exempligratia.
gratia. When Whenyou youmeanmean "that thatis"isuse usei.e.
i.e. Since Sincee.g.e.g.
indicatesaapartial indicates partiallist, list, ititwould wouldbe beredundant redundanttotoadd addetc.
etc. at theend atthe endofofaalist listintroduced introducedby bythis thisabbreviation.
abbreviation.
(fromanswerbag.com)
(from answerbag.com) The Therecirc recircpumps pumpsand andSLC SLCpumps pumpsare areused usedfor mitigationof formitigation ofthe theAATWS.
TWS.
RO/SRO Rating:
RO/SRO Rating: 3.8/4.2 3.8/4.2 CLSLP300E*01 7 Objective: CLS-LP-300-E*017 Objective:
- 17. Compare
- 17. Compare and and contrast contrast the the operator operator actions actions for for emergency emergency depressurization depressurization with with an an ATWS ATWS condition condition present versus present versus those those withwith all all control control rods rods inserted.
inserted.
Reference:
Reference:
11 (2)OP-,
(2)OP-, Revision, Revision, Page, Page, Section Section Cog Level:
Cog Level: HighHigh Explanation:
Explanation:
Reactor recirc Reactor recirc pumps pumps are are evaluated evaluated during during the the ATWS ATWS and and are are tripped tripped toto reduce reduce corecore flow flow thereby thereby reducing reactor reducing reactor power.
power. IfIf power power is is less less than than 2%2% then then tripping tripping the the pumps pumps is is not not required.
required. SLC SLC is is injected injected into the into the reactor reactor to provide an alternate means of shutting down the reactor. ItIt is to provide is required to to be be injected injected prior to prior exceeding 110° to exceeding 1100 FF in the torus. With SRVs SRVs closed and and temperature temperature at at 92 92 with no no additional heat heat load load to to the the torus torus SLC SLC injection injection is not required.
Distractor Analysis:
Distractor Analysis:
Choice A:
Choice A: Correct Correct answer, answer, see explanation.
Choice B:
Choice Plausible because B: Plausible because recirc pumps are not required to be shutdown is correct and if the torus was was in jepordy in jepordy of reaching 110 then this would be corect.
Choice Choice C: C: Plausible Plausible because if power was greater greater than 2% the recirc pumps pumps are required to be shutdown shutdown is is correct correct andand the torus is not in in jepordy jepordy of reaching 110 110 soso this this is is corect.
Choice Choice D: D: Plausible Plausible because because ifif powerpower waswas above above 2%
2% andand the the torus torus temperature temperature in in danger danger of of reaching reaching 110 110 then then these these actions actions wouldwould be be correct.
SRO SRO Only Only Basis:
Basis: N/AN/A Notes Notes
From001-37.5:
From 001-37.5:
STEPS RetQ-06 STEPS RC/Q-06 and and RC/Q-07 RC/Q-07
...----.....--~AN
...---- REACTOR POWER r---,Y=ES,,-< OE DETERMINED TO BE LESS "rHAN _.-'"
2'" .-'"
.....--~
RCIQ-07 STEP BASES:
STEP BASES:
IfIf reactor reactor power power is less than is less than 2%, 2% the the operator operator is is routed routed aroundaround the the step step which which directs directs trippin g both recircu lation tripping both recirculation pumps. pumps .
Tripping the Tripping the recirculation recirculation pumps pumps from high reactor power effects a prompt reduction in prompt reduction in power. IfIf boron power. boron injection injection is is later later required, required, three-dimensional three-dimensional model tests have demonstrated that demonstrated that forced forced recirculation recirculation need not be maintained maintained because because natural circula tion flow provid circulation flow provides adequatees adequa te boron mixing mixing.. Trippi Tripping ng the recirculation recirculation pumps is is allowed in allowed in this this step step becaus because e the operator operator will have already run the speeds back, ifif necess necessary, aiy, and and the the resulti resulting ng change changes s will be signifi significantlycantly reduce reduced. d.
IfIf reactor reactor power power is is below below the the APRM downs downscale cale trip setpoin setpoint, t, trippin tripping the recircu g the recirculation lation pumps results in little, if any, reducti on pumps results in little, if any, reduction in reactor power in reactor power since since power power is is already already near near thethe decay decay heatheat level.
level. In In this this case, case, forced forced recircu recirculation lation flow flow is is permit permitted to contin ted to continue for the ue for the purpos purpose e ofof maxim maximizingizing boron boron mixing mixing should should boron boron injecti injection on later later be be require required.
d.
STEPS RC/Q-08 STEPS RCIQ-08through throuah RC/Q-10 RCIQ-1O
.....~~
...............- CAN ~
REACTOR
.---- REACTOR POlliER POWER NO
~ BE BEDETERMINED DETERMO4EDTO TOBE BE LESSTHAN LESS THAN /".--...-
Z3 .....~
2"~~
__ RC.'O-06 RCQ O RCfQ*10 RCJQ. tO STEP BASES:
STEP BASES:
If reactor If reactor power power isis above above 2%, 2%, the operator operator is directed directed to inject boron. This is is aa conservative action conservative action because because with power above 2%, Suppression Suppression Pool temperature temperature will will steadil y increas e toward steadily increase towards '110°F.s 110°F. This also allows sufficient time for the Hot Shutdown sufficient Shutdown Boron Weigh Boron Weightt of boron to of boron to be be injecte injected.d. The extra time may be needed since the the alterna te system s used alternate systems used for boron for boron injection require significantly more time to inject boron boron should should the the SLC SLC System System fail. fail. The SLC system is initiated to shut down the reactor reactor..
The The Boron Boron Injection Injection Initiation Initiation Tempe Temperaturerature is is defined defined to to be be the greater greater of:
of:
a.
- a. The The Suppre ssion Pool Suppression Pool temper temperature ature at at which which initiation initiation ofof aa reactor reactor scram scram is is require d by Techni required by Technical cal Specif ications; or Specifications, or b.
- b. The The highes highest t Suppre Suppressionssion PoolPool temper temperature ature at at which which initiation initiation of of boron boron injection injection using SLC will result in injection using SLC will result in injection of of the the Hot Hot Shutdo Shutdown wn Boron Boron Weigh Weight of boron t of boron before before Suppre ssion Pool Suppression Pool temper temperature ature exceed exceeds s the the Heat Heat Capaci Capacity Temperature ty Tempe rature Limit.
Limit Categories Categories K/A:
KIA: G2.04.09 G2.04.09 Tier Tier/ /Group:
Group: T3 T3 RO Rating:
RORating: 3.8 3.8 SRO SRORating:
Rating: 4.24.2 LP LPObj:
Obj: CLSLP3OOE CLS-LP-300-E*O *O17 17 Source: E14rJ K.
Source:
Cog CogLevel:
Level: HIGH HIGH Category Category8:8: