ML101650244
| ML101650244 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/07/2010 |
| From: | NRC/RGN-II |
| To: | Southern Nuclear Operating Co |
| References | |
| 50-348/10-301, 50-364/10-301 | |
| Download: ML101650244 (558) | |
Text
1. 001AK2.06 OO1AK2.06 001/FNP OO1/FNP BANK/RO/C/A 2.9/3.2IY 2.913.2/Y 2007/N/2ICVRIY 2007/N/2/CVR/Y Unit 11 is at 74% power and stable, and the following conditions occurred:
At 1000:
- Rod control is in AUTO.
Tl-408A, Tavg - Tref deviation, indicates O°F
- TI-408A, - 0°F and stable.
- Pressurizer level is stable.
- Reactor Power is approximately 75% and stable.
- Control Bank 0 D step counters are at 144 steps.
At 1002:
Tl-408A, Tavg - Tref deviation, indicates approximately +2°F and
- TI-408A, -
rising.
- Pressurizer level is slowly rising.
- Pressurizer spray valves have throttled open.
- Reactor Power is approximately 76% and slowly rising.
- Control Bank 0 D step counters are at 150 steps and rising at 8 steps per minute.
- There is no load change in progress.
Which one of the following describes:
- 1) the event in progress and
- 2) the NEXT action that must be performed lAW AOP-19.0, Malfunction of Rod Control System?
A. 1) Inadvertent RCS boration;
- 2) Trip the reactor and enter EEP-O, EEP-0, Reactor Trip or Safety Injection.
B. 1) Inadvertent RCS boration;
- 2) Place the rod control mode selector switch to MANUAL and match Tavg with Tref by inserting rods.
C. 1) Uncontrolled Continuous Rod Withdrawal; EEP-0, Reactor Trip or Safety Injection.
- 2) Trip the reactor and enter EEP-O, D~
D 1) Uncontrolled Continuous Rod Withdrawal;
- 2) Place the rod control mode selector switch to MANUAL and verify that rod motion stops.
Page: 1 c:I 200 Page:1c1200 1211412009 12/1412009
A - Incorrect.
A - Incorrect. The The first first part part is is incorrect, incorrect, since since for for anan inadvertent inadvertent boration, boration, TavglTref Tavg/Tref mismatch would mismatch would be be less less than than -1.5
-1 .5 (with (with rods rods toto be be moving moving outward) outward) and and power power would be would be less less than than 75%75% instead instead of of 76%.
76%. Plausible, Plausible, since since rods would be moving rods would be moving outout and Tavg/Tref mismatch and TavglTref mismatch couldcould be increasing (which be increasing (which would would cause cause Przr Przr level level to to rise rise and spray and spray valves valves to throttle open) to throttle open) with with an an inadvertent inadvertent boration.
boration. The The second second partpart is is incorrect, but incorrect, plausible. The but plausible. The stated action is is the RNO RNO if rods do not cease moving if rods do not cease moving have been once they have placed in been placed manual lAW in manual lAW AOP-19.
AOP-19. Also, a conservative action may be chosen to to trip the reactor, reactor, but would not but this would not be be in in accordance accordance with AOP-19.0 for this AOP-19.0 this situation, situation, nor nor would itit be necessary.
be necessary.
B - The first part is incorrect (see A). Second part is
- is correct lAW AOP-19 AOP-1 9 for a continuous rod withdrawal (see D).
Incorrect. The first part C - Incorrect.
- part is correct (see D). The second part part is incorrect (see A).
oD - Correct. A CRW is taking place as indicated by the Tavg/Tref meter value going up above +1.5 and continuing to increase. This shows rods should actually be moving to lower the high temperature, and the action is to place rods in Manual if they are stepping while in AUTO.
Technical
Reference:
AOP-19 Malfunction of Rod Control, Version Verelon 26.0 Previous NRC exam history if any: FNP 2007 NRC exam, but with different distractors (changed from inadvertent dilution to inadvertent boration in A & B). This is the only question in the bank that comes close to meeting this k/a (searched BOTH "KA" KA and second KA" "second contains 001AK").
KA on "contains OOIAK).
OOIAK2.06 001AK2.06 001 Continuous Conti nuous Rod Withdrawal AK2. Kncmledge Kno.vledge of the interrelations between the Continuous Rod Withdrawal and the follcming:
follo.ving: (CFR 41.7
/45.7) AK2.06 T-ae.Iret. deion ma:er ......................................... 3.0*
T-ave./rEf. deviation 3.0* 3.1 Match justification: This question presents conditions indicating a Continuous Rod Withdrawal, and the Tavg/Tref meter value and trend is provided.
To obtain the correct answer, aa knowledge of the relationship between the CRW and the Tavg/Tref meter response is required.
Objective:
- 4. EVALUATE plant conditions and DETERM II NE NE if any system components need need to be operated while performing AOP-19, Malfunction of Rod Control System.
(OPS-52520S06)
(OPS-52520SOO)
Page: 2of Page: 2 of 200 200 1211412009 12114/2009
10/27/09 10:06:24 10/27/09 10:06:24 ...' i ':d~.i...
FNP-1-AOP-19.0 MALFUNCTION OF ROD CONTROL SYSTEM Version 26.0 Step Action/Expected Response
~ Response Not Obtained I I [
NOTE: Steps 11 and 2 are IMMEDIATE OPERAOPERATORTOR actions.
- 1 Verify NO load change in progress. 1I Check for cause of load change.
[YES[ 1.1 IF load rejection in progress or has occurred, THEN go to FNP-1-AOP-17.0, FNP-l-AOP-17.0, RAPID LOAD REDUCTION.
1.2 IF secondary leakage is indicated, THEN go to FNP-1-AOP-14.0, SECONDARY SYSTEM LEAKAGE.
2 IF unexplained rod motion occurring, THEN stop rod motion.
CORRRECT 2.1 IF rod control in AUTO, 2.1 IF rod control in MANUAL, THEN place rod control in MANUAL. THEN place rod control in AUTO NOTE: In AUTO rod control, rods will step OUT ifTAVG if TAVG less than TREF by at least 1.5 degrees, and TAVG Rods will step IN if T A VG greater than TREF by at least 1.5 degrees.
2.1.1 IF AUTO rod motion due to TAVG/TREF TA VG/TREF mismatch, THEN verify rod motion stops when TAVG is within 1 ofTREF I degree of TREF 2.2 IF unexplained rod motion NQINOT stopped, THEN perform the following.
2.2.1 Trip the reactor ~
A&C IA&C INCORRRECT IINCORRRECT I
2.2.2 Go to FNP-1-EEP-0, REACTOR TRIP OR SAFETY SAFETY INJECTION Page Completed Page 22 of of99
/ ",,~.-~ ~-'-.\
- 2. 001K2.02
- 2. 001K2.02 00l1MOJ:;!JRO/MEM 3.613.7IINI2ICVRIVER 55 EDITORIAL OO11MOPJO/MEM 3.6/3.71IN/2ICVRNER EDITORIAL onéf Which one- of the following correctly correctly describes describes components components in in the power power flow path path toto Reactor Trip Breakers?
the Reactor Breakers?
600V (1)
The 600V (1) CRDM MG supply the CRDM MG set supply supply breakers, breakers, then the (2)
(2) ,then
, then Reactor trip breakers.
the Reactor breakers.
(1))
(1 (2)
A. LCCs 0D and E Motor Generator Sets, then the Power Cabinets B. MCCs A and B Motor Generator Sets, then the Power Cabinets C
C~ LCCs 0D and E £ Motor Generator Sets D. MCCs A and B Motor Generator Sets A. Incorrect. Plausible because all parts of the correct answer are listed, but in an incorrect order. The power cabinet is actually downstream of the Reactor trip breaker.
B. Incorrect. MCC A & B are incorrect, but Plausible; these are also safety related 600V switchgear. See A for second part.
C. Correct. Per load list and FSD on Reactor Protection, A181007, Figure F-1. F-i.
D. Incorrect. MCC A & &BB are incorrect, but Plausible; these are also safety related 600V switchgear.
£ach MG set consists Each consistsof of a600V AC, 150 hp induction motor, a astainlesssteel stainless steel flywheel, and a 260V AC, 3 phase synchronous generator located I ocated in the non-rad Aux bldg 139' 139 elev.
el ei. The motors for the M MG G sets are powered from two 600V load centers (LCs) (M (MGG A is powered from LC D; M LCD; MGG B is powered from L LC £). These motor supply breakers can be operated from the M C E). MGG set control panels located in the rod control room (Aux bldg 121' 121 elei.)
elev.) or from the MCB.
Page:
Page: 33 c of 200 200 1211412009 12/1412009
Previous NRC Previous NRC examexam history any: 2005 history ifif any: 2005 Vogtle Vogtle NRC NRC exam under 001 exam under 001 K2.01 K2.01 (power (power supplies to supplies to MG MG SETS)
SETS) 001 K2.02 001 K2.02 001 Control 001 Control RodRod Drive Drive System System K2 KnoNledgeof K2 Kncmledgecl buspo.ver bus power supplies aippliestothefdkming: (CFR: 41.7) to the foilONi ng: (CFR: 41.7)
One-I ne di ran of K2.02 One-linediaaran K2.02 of power power supply suppl yto trip brEEKers to trip brers ..................... 3.63.7 3.6 3.7 Match justification: The Match The power power flow flow to to the the Reactor trip breakers Reactor trip breakers is is examined examined here here including the Busses including the Busses that that supply the CRDM the CRDM MG. MG. This This power power supply supply flow must must be be understood to correctly answer answer this question.
Objective:
Objective:
1.
- 1. NAME NAM DENTI FY the power supply for the following cabinets E AND IIDENTIFY cabins associated associat with Reactor Protection the Reactor Protection System (RPS) to include include those items found on Figure 12, 120 (OPS-52201 104).
VAC Distribution (OPS-52201104).
Page:4d200 Page: 4 of 200 1211412009 1211412009
c I;:'
CONTROL BOARD { MASTER AND ACTUATE TRAIN ACTUATE B TRAIN 'B' SWITCHES OUTPUT SLAVE RELAYS SAFEGUARDS SAFEGUARDS
!r
<TRAIN 'B')
)~----------~----4-------~
b\~
PROTECTION ANALOG ANALOG PROTECTION PROTECTION SYSTEM SYSTEM SYSTEM LOGIC SOLID STATE LOGIC
<'tI.
(W 7300 7300 SYSTEM)
SYSTEM) TRAIN B NUCLEAR NUCLEAR INSTRUMENTATION SYSTEM INSTRUMENTATION SYSTEM ROD FIELD CONTACTS OR FlELD;oNTACTS DR { CONTROL ~C' SYSTEM
/ " INPUT PROCESS PROCESS CHANNEL CHANNEL L __ -+__~____~__L-__4-__~__~__
SENSOR SENSOR COMPUTER 1\-
r (1 IV*}----/
,TRIP I,.- 'OR' CABLE
( BRKR.
"B' I BYPASS BRKR. 'B' V ~
Il)
JJ ~
CONTROL BOARD C"" BRKR.
, 'A' (BYPASS BRKR. 'A' BISTABLE
~J._---LJ-_--l...l _ ___ L~-, CONTROL PROTECTION I I I BOARD SYSTEM I SOLID STATE LOGIC I DEMUX TRAIN A _ _ _ -1 CABINET MASTER AND 1"- ISOLA nON R~ II
'"~
CONTROL SLAVE RELAYS BOARD A TRAIN'A' ACTUATE TRAIN CONTROL CONTROL SWITCHES
<TRAIN 'A')
Urf---------t-- ACTUATE SAFE GUARDS SAFEGUARDS M-G SET A M-G SETB>
)
,It{J L{~ &£'0 I-vI.- L--
REACTOR PROTECTION SYSTEM BOUNDARIES fiGURE f-l
FNP UNIT 1I FNPUNIT LOAD LIST A-506250 A-S062S0
~----
DF03 10 600V LOAD CENT~
-~
- 139 D177010 BKR TPNS TPNS DESCRIPTION DESCRIPTION SEE PAGE PAGE EDO1 Q1R16BIKI IPT COMPARTMENT EDO2 Q1R11B0004A liD 4160/600V SST (NORMAL) <<< DFO3 c
EDO) TiiMOObiA[1A CRDM MG SET EO4 Q1R42E0001A-A 1ATTERY CHAER >>> 1A 125V DC F-3 F 3 SWITCHGEAR ( Q1R42B0001A-A)>>>
EDO5 QSR17B0006-A iF 600/208V McC >> F-79 EDO6 Q1E22M0001A A REACTOR CAVITY ui. T.f1TTflhJ FAN EDO8 Q1R16B0002-A 1A600V LOAD CENTER (ALTERNATE - ENERG) D-46 0-46 I
>>> EAO9 >>>
ED09 Q1R42E0001C-AB SW Q1R18B0001A-A >>>
600V AC DISC SWQ1Rl >>> iC 1C BAT.T BATT F- F33 CHARGER* (A TRAIN SUPPLY) >>>
CHARGER (ATRAINStJPPLY) >>> i2SVDC 125V DC DISC SWQiR18B0002A-A SW Q1R18B0002A-A >>>.lA12SVDC
EDiO ED1O Q1R17B0001-A iA 600 lA 600/208V MCC >>> F-92 F 92 EDII ED11 NIT47MOOOIB-A N1T47M0001B-A IB 15 CRDM COOLER FAN ED12 Q1R16B0008-AB iF 600V LOAD CENTER (ALTERNATE) <<< <<< EF06 EFO6 ED13 Q1R17B0509A is 600/208V MCC (NORMAL) >>> F-98 ED14 Q1R17B0008A 1U 600/208V MCC >>> F-I04 F104 ED15 QIE12MOOOIA-A Q1E12M0001AA IA 1A CONTAINMENT COOLER (EMERG. (EMERG./ LOW SPEED)
ED16 QIE12MOOOIB-A Q1E12M0001B-A IB lB CONTAINMENT COOLER (EMERG./ LOW SPEED) 1 sectf.doc lsectf.doc Page F --22 Ver. 47.0
FNP UNIT 11 FNPUNIT LOAD LIST A-506250 A-S062S0
.--~~"-"--"~'-~-----,~ --'""
/DGO3 DG03 ~~
1E 600V LOAD CENT~
IE CENTER ')
) AB -121'
- 121 D177011 j
BKR TPNS DESCRIPTION SEE PAGE Q1R16B000733 IE 600V LOAD CENTER EEOI Q1RI6BKREEO1 Q I R I6BKREEO I PT COMPARTMENT EEO2 Q1RllB0005-B Q1R11B0005-B 1E 416W(i(O 1 SST (NORMAL) <<< DGO3 NICIIMOOOIB-N NIC1 IM000 lB-N lB CRDM MG SET EE05 Q1R4213000113-fl 1111XTTERY CHARGER QIR42E0001B-B>>> 113 12SVDC G-3 SWGR>>>
EEO6 Q1R42B0001B-B 600V AC DISC SW Q1R18B0001B-B>>> >>>lCBATT 1C I3ATT G-3 CHARGER Q1R42E0001C.AB(B Q1R42E0001C-AB(B TRAIN*SUPPLY)
TRAIN SUPPLY)>>> >>> 125V DCDISe DC DISC SWQ1R18B0002B':B SW Q1R1SU0002B-B>>> >>> IB125V 113 12W DCSWGR >>>
DC SWGR>>>
EEO7 Q1R16B0005 Q1RI6B0005-B ..B le 1C 600V LOAD CENTER (ALTERNATE-EMERG>>>>
(ALTERNATE-EMERG)>>> E-5 EC10>>>
EC1O>>>
EEO8 EEOS QIEI2M0001C-B QIEI2MOOOIC-B IC CONTAINMENT COOLER (EMERG./ LOW SPEED)
EEO9 EE09 QIE22MOOOIB-B Q1E22M000 lB-B IB lB REACTOR CAVITY DILUTION FAN EE1O EE10 Q1R17130002-fl 113 6001208V MCC>>> G-76 EE11 EEll QSR17130007-B 1G 600/208V MCC>>>
MCC >>> G-S4 G-84 EE12 Q1R16B0008-AB lF iF 600V LOAD CENTER (ALTERNATE) <<< EFO8
<<< EF08 EE13 NIT47M000 NI T47MOOOIA-BlA-B 1A CRDM COOLER FAN IA EEI4 EE14 Q1R17BO51O-B 1iT 600/208V Mee T600/208V MCC>>>>>> G-99 EEl5 EEI5 Q1R17B0009-B 1lv 600/208V MCe V600/208V MCC>>>>>> G-105 II EEI6 QIE12M000ID-B QIEI2MOOOID-B ID lD CONTAINMENT COOLER (EMERG./ LOW SPEED) 1 sectg.doc lsectg.doc Page G - 2
- Rev. 23
- 1. ROD CONT-40204I04 OO4/ITI/M (LEVEL 1) SYS/001K2.011111 CONT-40204104 004IHLT//M SYS/001K2.O1////
001 K2.02 Which ONE of the following correctly states the order of components through which power flows to the Control Rod Drive Mechanisms?
A. MG Supply Breakers, then 600v LCC D D and E, then Motor Generator Sets, then Power Cabinets, then Reactor Trip Breakers.
B. MCC A and B, Power Cabinets, then MG Supply Breakers, then Motor Generator Sets, then Reactor Trip Breakers.
C 600v LCC D and E, then MG Supply Breakers, then Motor Generator Sets, then C':'"
Reactor Trip Breakers, then Power Cabinets.
D. MCC A and B, Motor Generator Sets, then MG Supply Breakers, then Reactor Trip Breakers, then Power Cabinets.
A. Incorrect. Plausible because all parts of the correct answer are listed, but in an incorrect order.
B. Incorrect. Plausible because all parts of the correct answer are listed, but in an incorrect order.
C. Correct. See Reference 1, Page 9.
D. Incorrect. Plausible because all parts of the correct answer are listed, but in an incorrect order. This choice would be correct if "Motor Motor Breakers" Breakers were Generator Breakers".
replaced with "Generator Breakers.
Each MG set consists of a 600V AC, 150 hp induction motor, a stainless steel flywheel, and a 260V AC, 33 phase synchronous generator located in the non-rad Aux bldg 139'139 elev. The motors for the MG sets are powered from two 600V load centers (LCs) (MG A is powered from LC D; MG B is powered from LC E). These motor supply breakers can be operated from the MG set control panels located in the rod control room (Aux bldg 121' 121 elev.) or from the MCB.
2005 VNP nrc exam K/A KIA 001 Control Rod Drive K2.01 Knowledge of bus power supplies to the following: One-line diagram of power supply to MIGM/G sets.
K/A MATCH ANALYSIS KIA Question tests knowledge of the power supplies to the MIG M/G Sets at the memory level.
Page: 11 10/26/2009 10126/2009
001(N~~ko/C/A 2.7/3.1/N/N/4/CVRIY
- 3. 003A2.03 OO1(N(AO/C/A 2.713.1ININI4ICVRIY Unit 1 I is at 25% power and the following conditions occurred:
At 1000:
- 1 IAA RCP amps and motor winding temperature were observed to be rising while 1A IA RCS LOOP flow was decreasing.
At 1002:
- . 1A RCP Handswitch indicating Green and Amber lights are LIT, the Red light is NOT LIT.
- AOP-4.O, AOP-4.0, Loss Of Reactor Coolant Flow, immediate actions have been completed.
RCS Temperatures are:
- IA RCS LOOP Tavg 1A is 53rF.
537°F.
- 18 1 B RCS LOOP Tavg 553° F.
is 553°F.
- 1C 1 C RCS LOOP Tavg is 553°F.
Which one of the following correctly describes the CAUSE of these indications and the ACTION required lAW AOP-4.0?
AOP-4.O?
CAUSE ACTION A'I Seized motor bearing A Trip the Reactor B. Sheared shaft Trip the Reactor C. Seized motor bearing Commence Normal Reactor Shutdown D. Sheared shaft Commence Normal Reactor Shutdown Page: 5 5 of 200 1211412009
This is on the RO level, since TS 3.4.2 requiresrequires Mode 3 in 30 minutes for Tavg below Minimum Temperature for Criticality. AOP-4.0 requires a reactor trip in this situation at step 3 after immediate action steps 11 & & 2. This would be SRO QnIyJmQwl~g~~jLil only knowledgJf it were
~_ot _~I~~.~.~.~..~~.~~rI§ §!fl!QIL.tQJ~e. il}.m9(:t~.9.J;?IiQIIQ:g.~I!iii6iEu.Q.£Qmt!:!f1.~.D.~!!!!~£~-*
not also a < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS .. action to be in modfiTTo being.able to conduct a normal
~wn.
ytPwn. 4 A - Correct. As the motor bearing starts to seize, the amps go up and flow goes down until the breaker trips on overcurrent. Then, AOP-4.0 entry conditions are met (This procedure is entered when forced RCS flow is lost in one or more loops and
("This required.) AOP-4.0 requires a reactor trip if any Tavg is << 541°F, no reactor trip is required.")
and the stem gives 539° F for A Loop.
539°F B - Incorrect. The first part is incorrect, but plausible. RCS flow would go down if the shaft sheared, but current would also go down instead of up. The second part is correct (see A).
C - Incorrect. The first part is correct (see A). The second part is incorrect, but plausible. It would be correct at this power level <<30%, (<30%, P-8) if Tavg was not less than the Minimum Temperature for Criticality.
D - Incorrect. The first part is incorrect (see B). The second part is incorrect (see C).
AOP-4, Version 18.0 TS 3.4.2 Page: 66 of 200 1211412009 12/1412009
Previous NRC exam history if any:
003A2.03 003 Reactor Cool Coolant ant Pump System A2 Ability Ability to (a) predict predict the the impacts impacts of thefoilONing the fdlcming malfunctions or operations on the RepS; RCPS; and (b) based on those on those predictions, proceduresto predidion use procedures contrd, or mitigate the conSlquences to correct, control, consequences of those malfunctions operations (CFR: 41.5/43.5/45.3/45/13) or operations:
A2.03 Problemsassxicted Problem scierJ with RCP motors, including failty fruity motors and current, a1d end winding and booring bri ng tempercture problem teniperure probl ems ....... 2.7 3.1 Match justification: This question requires knowledge of what type of motor malfunction would give the indications in the stem. The indications were given in the stem and the applicant is required to analyze and diagnose what malfunction would cause these indications to avoid backwards logic. This order was also required to allow choosing actions which are based on the indications. At FNP, Motor bearing temperature indication is not available, but motor winding temperatures are.
Winding temperatures would go up due to the motor shaft and bearing seizing, so is included in the stem. The second part of the question and choices require knowledge of what action is required for the given set of indications.
Objective:
- 6. DEFINE DEFI NE AND EVALUATE the operational implications of normal I/ abnormal plant or equipment conditions associated with the safe operation of the Reactor Coolant Pumps (RCPS) components and equi equipment, pment, to iinclude following ncl ude the foil (OPS4O3O1 007):
owi ng (OPS-40301 D07):
- Normal Control Methods
- Abnormal and Emergency Control Methods
- Protective Protecti ve iisolations sol ati ons such as hi gh flow, low pressure, low Ileiel high eve! iincluding setpoint ncl udi ng setpoi nt
- Fast dead bus transfer
- Automatic actuation A utomati c actuati on iincluding setpoints ncl udi ng setpoi nts
- Actions A cti ons needed to mi mititi gate the consequence of the abnormal abnormalityi ty 7of Page: 7 of 200 1211412009
05/12/09 12:22:58 FNP-I-AOP-4.0 FNP-1-AOP-4.0 LOSS OF REACTOR COOLANT FLOW Version 16.0\
Step Action/Expected Response Response Not Obtained NOTE: Step 11 and 2 are IMMEDIATE OPERATOR actions.
1 Maintain SG narrow range level stable at 1 IF SG level rise cannot be controlled, approximately 65% using: THEN close the affected SG Main Feedwater Stop Valve(s) OR Auxiliary Feedwater Stop
[]
[J Main Feedwater Regulating Valves valve(s).
va I ves-
[]
[1 Main Feedwater Bypass Regulating Valves.
[]
[1 Auxiliary Feedwater Control Valves.
((1] IA 1A SG QIN21MOV3232A Q1N21M0V3232A
((1] IB lB SG QIN21MOV3232B Q1N21MOV3232B
((1] lC IC SG QIN21MOV3232C QIN2IMOV3232C OR
(()) lA 1A SG QIN23MOV3350A Q1N23M0V3350A
(()) 1B lB SG QIN23MOV3350B Q1N23M0V3350B
((1] lC SG SC QIN23MOV3350C Q1N23MOV335OC 2 Check lA 1A and IBlB RCPs - RUNNING.
- 2 Manually close pressurizer spray valve for affected RCP.
(()) lA1A RCS loop spray valve PK-444C
((j] 1BlB RCS loop spray valve PK-444D
( 3 Monitor Tavg for all three RCS loops 2::
541°F. (TS 3.4.2) 541°F.(TSIL2) 33 following....
Perform the following 1 IF the main generator is ON LINE> )? <
W (J THEN trip the reactor and go to THENtripthereactorandgoto FNP-I-EEP-O, FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION J (2
I 3.2 IF the main generator is OFF LINE, jf THEN raise Tavg 2:: 541°F within 30 minutes 3.2.1 Adjust steam dumps to reduce secondary power demand as necessary 3.2.2 Verify rod control in MANUAL o Step 3 continued on next page Page Completed Page 2 of 11
05112/09 05/12/09 12:22:58 ' , ) L.~i JLJL FNP-I-AOP-4.0 FNP-1-AOP-4.0 LOSS OF REACTOR COOLANT FLOW Version 16.0
~ Step Action/Expected Response Response Not Obtained I I I 3.2.3 Stabilize Tavg in the idle loop(s)
> 541°F 54 1°F while maintaining the running loop(s) < 554°F by adjusting rod
<554°F position and/or boron concentration 3.2.4 IF unable to restore Tavg ~ 541°F, jf 54 1°F, THEN trip the reactor and go to FNP-1-EEP-0, REACTOR TRIP OR SAFETY INJECTION 4 Maintain PRZR pressure 2200-2300 psig.
4.1 Control PRZR heaters as required.
4.2 IA and IB IF lA ffl lB RCPs running, 4.2 Perform the following THEN, control pressurizer pressure with both normal spray valves. 4.2.1 IF 1B lB RCP running, THEN control pressurizer pressure with PK-444D.
NOTE: Running lAIA and lC 1C RCPs will be required to provide adequate spray flow through the 1A RCS loop spray valve.
4.2.2 IF 1A & lCIC RCPs are running, THEN control pressurizer pressure with PK-444C.
4.2.3 IF spray flow is adequate, IE THEN proceed to step 5.
4.2.4 IF no spray valves are available, THEN proceed to step 4.4.
4.3 Proceed to step 5.
o Step 4 continued on next page Page Completed Page 33 of 11
05/12/09 12:22:58 05112/0912:22:58 FNP-1-AOP-4.0 LOSS OF REACTOR COOLANT FLOW Version 16.0 Step Action/Expected Response Response Not Obtained 5.2 IF letdown has isolated due to a plant transient, THEN establish normal letdown using ATTACHMENT 1, RESTORING LETDOWN.
5.3 IF a letdown isolated due to a system malfunction, THEN perform the following:
[1 Attempt to restore any letdown flow using
[]
FNP-1-AOP-16.0, CVCS MALFUNCTION.
[] Continue with applicable steps of this
[]
procedure.
5.4 WHEN normal letdown restored AND IF required, THEN return to step 4.4 to establish auxiliary spray.
6 Maintain PRZR level at approximately 2_---
/ 77 Within six hours of the loss of ReS complete the following:
RCS flow 7.1 IF the unit is in Mode 11 or 2, y d Qt THEN place unit in Mode 33 using the following procedures:
[]
[] FNP-1-UOP-3.1, POWER OPERATION FNP-1-UOP-3.1,POWEROPETION
[] FNP-1-UOP-2.1, SHUTDOWN OF UNIT
[j FNP-1-UOP-2.l, FROM MINIMUM LOAD TO HOT STANDBY o Step 7 continued on next page Page Completed Page 6 of 11
RCS Minimum Temperature for Criticality 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Minimum Temperature for Criticality LCO 3.4.2 Each RCS loop average temperature (Tavg) shall be ~ 541°F.
APPLICABILITY: MODEl, MODE 1, kelt ~ 1.0.
MODE 2 with keff ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
~ Tavg in one or ~;;RCs~ A.1 loops not within limit. /
~------------
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify RCS Tavg in each loop ~ 541°F. NOTE
N 0 TE--------
Only required if low low T Tavg avg alarm not reset and any RCS loop Tavg
< 547°F 30 minutes thereafter Farley Units 11 and 2 3.4.2-1 Amendment No. 146 (Unit 1)
Amendment No. 137 (Unit 2)
- 4. 004G2.4.21 001/NEW/RO/C/A OO1/NEW/RO/C/A 4.0/4.6/N/N/2ICVRIY 4.014.6ININI2ICVR/Y A LOCA and LOSP has occurred on Unit 1, and the following conditions occurred:
- FRP-C.2, Response To Degraded Core Cooling, is in progress.
FRP-C.2,
- All charging pumps have tripped.
- RCPs are secured.
All RCP's
- The five hottest CETCs are; 773°F, 779°F, 1023°F, 1252°F, 1508°F and all stable.
- All SG pressures are at 1000 psig.
- Off-Site Power is available.
Which one of the following states:
- 1) the FRP that must be in effect for the conditions given (FRP-C.2 Response To Degraded Core Cooling, OR FRP-C.1 Response To Inadequate Core Cooling),
and
- 2) whether the RCPs will be started or not?
A. Enter FRP-C.1 RCPs will be started B. Enter FRP-C.1 RCPs will NOT be started C. Remain in FRP-C.2 RCPs will be started D Remain in FRP-C.2 Dy RCPs will NOT be started Page: 8of 8 c:i 200 1211412009 12/1412009
A - Incorrect. The first part is incorrect, but is plausible. The fifth hottest core Core Exit Thermo Couple is not higherhigherthan than 1200°F, so FRP-C.1 is not entered, but the highest three are >1200°F. Confusion may exist as to which of the highest
>1200°F prior to entry into FRP-C.1. IfIf the first part was thermocouples have to be >120QoF correct, the second part would be correct also. A major difference between FRP-C.1 & & C.2 is that in FRP-C.1, RCPs are started as a last resort even with no support conditions. In FRP-C.2 a RCP is started only if all support conditions are met.
B - Incorrect. The first part is incorrect (see A). The second part is incorrect, but plausible, since for FRP-C.2 and all other procedures it is correct. Confusion may exist as to whether or not FRP-C.1 directs starting the RCPs without support conditions when FRP-C.2, and all other procedures do not.
C - Incorrect. The first part is correct (see D). The second part is incorrect (see A).
D Correct. The fifth hottest Core Exit TC 0- - TO is >
> 700°F but <1200°F. Therefore, FRP-C.2 is still in effect and FRP-C.1 is not entered. FRP-C.2 does not direct starting a RCP without support conditions, but FRP-C.1 does.
CSF-O, Critical Safety Function Status Trees, Revision 17 FNP-1-FRP-C.1, Response To Inadequate Core Cooling, Revision 17 FNP-1-FRP-C.2, Response To Degraded Core Cooling, Revision 17 Page: 99 c of 200 1211412009 1211412009
Previous NRC exam history history if any:
004G2.4.21 Chemical 004 Chemi Volume cal and Vol ume Control System 2.4.21 KKncmledge ncmledge eX ci the parameters and logicIogc used uead to assess the status eX assesathe ci safety functions, fundions ccntrd, core cooling such as reactivity control, coding and heat removal, remcwal, reactor coolant codant system sjstem integrity, inteity, conditioris radioactivity release containment conditions, rdease control, contrd, etc.
etc (CFR: 41.7/43.5/45.12) 41.7 / 43.5/45.12) RO 4.0 SRO4.6 SR04.6 Match justification: RO level knowledge of the entry condition parameters and logic used to assess Red and Orange path (and logic) of the Critical Safety Functions is required to answer this question correctly. The Charging pumps in the CVCS system (which are also HHSI pumps during a LOCA) are provided as tripped which prevent all RCP support conditions from being met (along with a loss of CCW to the RCP Thermal barriers which is also listed in the stem). The second part of the question directly addresses the effect of the CVCS system on the procedure directions concerning starting or not starting RCPs.
Objective:
- 1. EVALUATE plant conditions and DETERM II NE if entry into (1) FRP-C.1, Responseto Response to II nooequate Cooling; nalequate Core Cool Degraded Core Cool i ng; or (2) FRP-C.2, Response to Degrooed Cooling; i ng; or (3)
FRP-C.3, Response to Saturated Core Cool Cooling required.
i ng is requi red. (OPS-52533C02)
- 2. EVALUATE plant conditions and DETERM I NE if any system Wstem components need to be operated whi while performing Ie performi FRP-C. 1 Response to IInadequate ng (1) FRP-C.1, Cooling; nooequate Core Cool i ng; (2)
Degraded Core Cool FRP-C.2, Response to Degrooed Cooling; i ng; (3) FRP-C.3, Response to Saturated Core Cooling.
Cool i ng. (OPS-52533C06)
Page: 10 10 of 200 1211412009
8/29/2007 08:33 8/2912007 08:3 3 TJNTT I FNP-1-CSF-O.2 FNP-1-CSF-O.2 CORE COOLING Revision 17 Revision 17
/ 1 P-c) 1O1EST I
NO p ....
CQEEXIT t
iC LESN THAN FIFTH / NO 1200 F
0 j
) YES
) HOTTEST (%
--~ CORECORE EXIT TC LESS 70000 THAN 700 YES (4J2 RCS SUBCOOLING NO GO TO FROM CORE FRP-C.3 oC TCS EXIT TC'S GRTR THAN GRTRTHAN YES 16°0 F {45° F}
16 I fb CSF SAT 1 of 1 Page 1 I
FNP-1-FR)
FNP-1-FR RESPONSE RESPONSE TO TO INADEQUATE INADEQUATE CORE CORE COOLING COOLING Revision Revision 17 17 Step Step Action/Expected Response Action/Expected Response Response Response NOT NOT Obtained Obtained n I I I 19 19 Check core Check core cooling.
cooling.
19.1 19.1 Check core exit T/Cs - LESS 19.1 19.1 Proceed to Step 21. OBSERVE THAN 1200°F.
1200°F. NOTE PRIOR TO STEP 21.
19.2 19.2 Check at least two RCS hot leg 19.2 Return to step 17.
temperatures - LESS THAN 350° F.
RCS HOT LEG TEMP
[1 TR 413
[]
19.3 Check REACTOR VESSEL LEVEL 19.3 Return to step 17.
indication - GREATER THAN 0%
UPPER PLENUM.
20 Go to FNP-1-EEP-1. LOSS OF REACTOR OR SECONDARY COOLANT.
step 14.
NOTE: ~
Normal conditions are desired but not required for starting RCPs. ----------J
=<
21 Check if RCPs should be I Il ((e started.
H C \Ld 21.1 Check core exit TiCs T/Cs - GREATER
- 21.1 Proceed to step 22.
THAN 1200°F.
Step 21 Step 21 continued continued on on next next page.
page.
_Page Completed Page Completed Page 22 Page 22 ofof 33 33
FNP-l RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n I I CAUTION: Further degradation of core cooling can occur if any running RCP is stopped before being directed by this procedure even if normal support conditions are lost.
1 Monitor RWST level.
RWST LVL p*;2 /2
[1 LI 4075A
[]
[1 LI 4075B
[]
1.1 [CA] WHEN RWST level less than rCA]
12.5 ft, c-LcL / /-
THEN go to FNP-I-ESP-l.3, FNP-1-ESP-1.3, TRANSFER TO COLD LEG RECIRCULATION.
62 1>
2 Verify proper SI valve alignment using ATTACHMENT 2. 2, SI VALVE ALIGNMENT FOR COLD LEG INJECTION.
I 2 f+
-ç1
___Page Completed Page 22 of 22
it.: \ \.1 V !.
FNP-I-FRP-C.2 FNP-1--FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 17 Step Action/Expected Response Response NOT Obtained n I I I 3 Check any HHSI HEIST flow - GREATER 33 Perform the following.
THAN 0 gpm.
3.1 Verify all available charging A
A TRN pumps started.
HHSI FLOW E] FI
[] Fl 943 CHG PUMP
[]
[I lA amps >> 0 HHSI []
El lB amps >
IB > 0 B
B TRN RECIRC []
El 1C amps >
lC > 0 0
FLOW
[] FI Fl 940 3.2 Verify charging pump MOV disconnects closed using ATTACHMENT 3, CHARGING PUMP MOV DISCONNECTS.
3.3 Verify proper SI alignment.
CHG PUMPS TO REGENERATIVE HX
[]
[1 QIE21MOV8107 Q1E21MOV81O7 closed
[]
[1 QIE21MOV8108 Q1E21MOV81O8 closed RWSTT RWS TO CHG PUMP
[]
El QIE21LCVllSB O1E21LCV115B open
[]
[1 QIE21LCVllSD Q1E21LCV115D open VCT OUTLET ISO
[]
[I QIE21LCVllSC Q1E21LCV115C closed
[] QIE21LCVllSE Q1E21LCV115E closed HHSI TO RCS CL ISO
[] QIE21MOV8803A open El Q1E21MOV88O3A
[] QIE21MOV8803B open El Q1E21MOV88O3B CHG PUMP SUCTION HDR ISO El QIE21MOV8130A open
[ ] O1E21MOV813OA
[] QIE21MOV8130B open El Q1E21MOV813OB El QIE21MOV8131A open
[ ] O1E21MOV8131A El QIE21MOV8131B open
[ ] O1E21MOV8131B CHG PUMP DISCH HDR ISO
[]
El QIE21MOV8132A O1E21MOV8132A open
[]
El QIE21MOV8132B O1E21MOV8132B open
[]
El QIE21MOV8133A O1E21MOV8133A open
[]
El QIE21MOV8133B O1E21MOV8133B open Step 33 continued continued on next page.
_Page Page Completed Page 33 of 22
l, . c.' L\\I\'. !'.
FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n
3.4 IF HHSI flow now established, THEN proceed to step 4, IF NOT, establish HHSI bypass SI flow.
CHG PMP RECIRC TO RCS COLD LEGS
[]
El Q1E21MOV8885 01E21M0V8885 open HHSI TO RCS CL ISO
[]
El Q1E21MOV8803A Q1E21MOV88O3A closed
[]
El Q1E21MOV8803B Q1E21MOV88O3B closed 3.5 IF HHSI flow now established, THEN proceed to step 4, IF NOT, perform the following.
3.5.1 Open HHSI isolation valves.
[]
El Q1E21MOV8803A O1E21MOV88O3A
[]
El Q1E21MOV8803B Q1E21MOV88O3B 3.5.2 Align charging pump suction header isolation valves based on 1B lB charging pump status.
1B lB Charging Pump Aligned As Aligned As Status A A Train pump B B Train pump CHG PUMP SUCTION HDR ISO Q1E21MOV [El] 8130A 813OA []
El 8130A 813OA open closed
[El] 8130B 813OB [El] 8130B 813OB open closed
[]
El 8131A []
[I 8131A closed open El 8131B
[] []
El 8131B closed open Step 33 continued on next page.
_Page Page Completed Page 44 of 22
" T\',i ';1__,:'---_ _ _ _ _ _ _--,._ _ _ _ _ _ _ _-.
~;,);~j,'! ,[
FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n I I 3.5.3 Align charging pump discharge header isolation valves based on 1B lB charging pump status.
1B lB Charging Pump Aligned As Aligned As Status A Train pump A BB Train pump CHG PUMP DISCH HDR ISO Q1E21MOV ((1] 8132A ((1] 8132A open closed
[]
[1 8132B ((1] 8132B open closed
((1] 8133A (()) 8133A closed open
[] 8133B ((1] 8133B closed open NOTE: Continuing efforts to establish SI flow should not interfere with performance of the remainder of this procedure.
3.6 If HHSI flow NOT established, THEN Continue efforts to establish SI flow.
- . HHSI flow I
- . LHSI flow I
- . Any form of RCS injection. I
_Page Page Completed Page 55 of 22
~ __________- r__________________~K~~T~'~~~7!*0**_****\ .*~*_\_~)~______________- ,______________- ,
- L;<,i):'.L .1 FNP-l-FRP-C.2 FNP-1-ERP-C.2 RESPONSE TO DEGRADED CORE RESPONSE CORE COOLING COOLING Revision 17 17 Step Step Response Action/Expected Response Obtained Response NOT Obtained n I I I CAUTION: Pump damage may occur if RHR pumps are operated operated on miniflow for longer than three hours with no CCW supplied to the RHR heat exchangers.
44 Check LHSI status.
4.1 Verify CCW flow to RHR heat exchangers - ESTABLISHED.
exchangers -
CCW TO lA(lB) RHR fiX 1A(1B) HX
[] QlP17MOV3l85A open
[1 O1P17MOV3185A
[] QlP17MOV3l85B O1P17MOV31S5B open 4.2 Check RCS pressure - LESS THAN
- 4.2 Proceed to step 5.
psig(435 psig).
275 psigt435 psig}.
lC (lA) LP 1C(1A)
RCS NR PRESS
[]
Ii PI 4O2B P1 402B
[]
[] PI P1 403B 4O3B Step 4 continued on next page.
___Page Completed Page 66 of 22
'c) I."!!. i .i FNP-1-FRP-C.2 FNP-l-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n I I 4.3 Check both RHR flows - GREATER
- 4.3 Verify LHSI properly aligned.
33 gpm.
THAN 1.Sx10 l.5xlO RHR PMP 1A(lB) 1A(1B) [1 1A amps >
[] > 0 0
E] 1B lE amps > > 0 0
FLOW
[] FI Fl 60SA 605A 1A(lB) RHR HX TO RCS
[]
[1 FI Fl 60SB 6O5E COLD LEGS ISO
[]
E] Q1E11MOV8888A O1E11MOV8888A open
[]
El Q1E11MOV8888B O1E11MOV8888B open RWST TO 1A(lB) lA(1E) RHR PUMP
[El] Q1EllMOV8809A O1E11MOV88O9A open
[]
El Q1EllMOV8809B O1E11MOV88O9B open 1A(lB) RHR HX DISCH VLV
[El] HIK 603A open
((I] HIK 603B 6O3B open 1A (lB) RHR HX 1A(1B)
[]
[I FK 60SA 605A closed
[]
[1 FK 60SB 605B closed 1A(lB) 1A(lE) RHR TO RCS HOT LEGS XC ON XCON
[]
[I Q1EllMOV8887 O1E11MOV8887A A open
[]
El Q1EllMOV8887B Q1E11MOV8887B open
_Page Page Completed Page 77 of 22
".,) L\; "i FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n I I I 5 Check RCS vent paths.
5.1 Check any PRZR PORV ISO - - 5.1 Restore power to PRZR PORV ISO POWER AVAILABLE.
AVAILABLE, valves unless de-energized for inoperable PORVs not capable of being manually cycled.
5.2 Verify both PRZR PORVs - - 5.2 Perform the following.
CLOSED.
5.2.1 Close PRZR PORVs.
5.2.2 IF any valve can NOT be closed.
THEN close its PORV ISO valve.
NOTE: The purpose of the following step is to establish an available PORV flowpath for mitigation of overpressure conditions.
conditions, without relying on the PRZR code safety valves. A A failed open PORV must not be unisolated. A A leaking PORV which is isolated with power available to the isolation valve should remain isolated until needed to reduce RCS pressure or mitigate an RCS overpressure condition. Any leaking PORV should be re-isolated when not in use.
5.3 Check at least one PRZR PORV 5.3 upen any PRZR PORV ISO not ISO - OPEN.
- required to isolate an open or leaking PORV.
5.4 Verify reactor vessel head vent valves - CLOSED.
RX VESSEL HEAD VENT OUTER ISO
[] Q1B13SV2213A
[] Q1B13SV2213B RX VESSEL HEAD VENT INNER ISO
[] Q1B13SV2214A
[1 01B13SV2214A
[] Q1B13SV2214B
[1 01B13SV2214B
___Page Completed Page 88 of 22
it) . i i . : '
FNP-1-FRP-C.2 FNP-1-FRP-C.2 RESPONSE TO RESPONSE TO DEGRADED DEGRADED CORECORE COOLING COOLING Revision Revision 17 17 Step Step Action/Expected Response Action/Expected Response Response NOT Obtained Response Obtained n I I I 66 Check RCP status.
Check Status.
6.1 Check at least one RCP - - 6.1 Proceed to Step 8.
STARTED.
CAUTION: To prevent potential seal damage, neither seal injection nor CCW cooling should be restored to a RCP which has lost both seal injection and CCW cooling.
NOTE: Normal support conditions for running RCPs are desired, however, RCP operation must continue even if support conditions cannot be maintained.
6.2 No. 1 seal support Verify No.1 conditions established.
6.2.1 [CA] Maintain seal rCA]
injection flow - GREATER THAN 66 gpm.
6.2.2 Verify No.No.11 seal leakoff flow - WITHIN FIGURE 11 LIMITS.
6.2.3 Verify No.No.11 seal differential pressure - -
GREATER THAN 200 psid.
6.3 Verify CCW - ALIGNED.
[1[] Q1P17HV3O45 Q1P17HV3045 open
[1[] 01P17HV3184 Q1P17HV3184 open Step 66 continued Step continued on on next next page.
page.
Page
___ Page Completed Completed Page 99 of Page of 22 22
!.\) .l v>: 1. :
FNP-l-FRP-C.2 FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n II II II I 6.4 Check RCP thermal barrier - - 6.4 Verify CCW flow isolated.
INTACT.
CCW FROM RCP RCP THRM BARR THRM TERM BARR []
[1 QlP17HV3045 Q1P17HV3O45 closed CCW FLOW []
[1 QlP17HV3l84 QlPl7HV3l84 closed HI
((1] Annunciator DD2 clear 6.5 Check CCW to RCP oil coolers - - 6.5 Verify CCW - ALIGNED.
SUFFICIENT.
CCW TO RCP CLRS CCW FLOW [] QlP17MOV3052 O1P17MOV3O52 open FROM RCP OIL CLRS CCW FROM RCP LO OIL CLRS
((1] Annunciator DD3 clear []
[1 QlP17MOV3046 O1P17MOV3O46 open
[]
[1 QlP17MOV3l82 O1P17MOV3182 open 6.6 Check RCP oil level - -
SUFFICIENT.
RCP lA(lB,lC) 1A(lB,1C) BRG UPPER/LOWER OIL RES LO LVL
(()) Annunciator HHlHIll clear
(()) Annunciator HH2 clear
[]
[1 Annunciator HH3 clear NOTE: Since RCP damage may occur when operating RCPs without normal support conditions established or under highly voided RCS conditions, the intent of the following step is to save one RCP (which provides the best pressurizer spray capability) for future use, if all three RCPs are running.
77 Check if one RCP should be stopped.
7.1 Check ALL RCPs - STARTED
- 7.1 Proceed to Step 9.
7.2 Stop RCP lB.
7.3 Proceed to Step 9.
_Page Page Completed Page 10 of 22
- \......t**" ' E FNP-I-FRP-C.2 FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 17 Step Step Action/Expected Response Response NOT Obtained n
8 Check core cooling.
S.l 8.1 Check REACTOR VESSEL LEVEL S.l 8.1 IF SI established, indication - GREATER THAN 0%
- THEN return to step 2, 2.
UPPER PLENUM. IF NOT, proceed to step 9.
S.2 8.2 Check core exit TiCs T/Cs - LESS
- S.2 8.2 IF core exit TiCs T/Cs falling, THAN 700°F. THEN return to step 2, IF NOT, proceed to step 9.
S.3 8.3 Go to procedure and step in effect.
9 Check SI accumulator discharge valve status.
9.1 Check power to discharge 9.1 Close accumulator discharge valves - AVAILABLE.
- valve disconnects using ATTACHMENT 1.
lA(1B,lC) 1A(1B,1C) ACCUM DISCH ISO
[] QIE21MOVSSOSA Q1E21MOV88O8A
[1 QIE21MOVSSOSB
[] Q1E21MOV88O8B
[1 QIE21MOVSSOSC
[] O1E21MOV88O8C 9.2 Check discharge valves - OPEN.
- 9.2 IF accumulators have NOT discharged, lA(1B,lC) 1A(1B,1C) ACCUM THEN open discharge valves.
IN DISCH ISO
[1 QIE21MOVSSOSA
[] Q1E21MOV88O8A lA(1B,lC) 1A(1B,1C) ACCUM
[1 QIE21MOVSSOSB
[] Q1E21MOV88O8B DISCH 015CR ISO
[] QIE21MOVS808C 01E21M0V8808C [] QIE21MOV8808A 01E21M0V8808A
[] QIE21MOV8808B O1E21MOV88OSB
[] QIE21MOV8808C Q1E21MOV88O8C 10 Monitor CST level.
10.1 rCA] Check CST level greater
[CA] 10.1 Align AFW pumps suction to SW than 5.3 ft. FNP-I-S0P-22.0, using FNP-1-SOP-22.O, AUXILIARY FEEDWATER SYSTEM.
SYSTEM, CST LVL
[] LI 4132A
[] LI 4132B 10.2 10.2 Align makeup to the CST from water treatment plant OR demin water system using FNP-1-S0P-5.0, DEMINERALIZED FNP-1-SOP-5.O, DEMINERALIZED MAKEUP WATER SYSTEM, as necessary.
_Page Page Completed Page 11 11 ofof 22
,~,) ;\! .1.; .,.
FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 17 Step Action/Expected Response Response NOT Obtained n I I I CAUTION: To prevent potential release of radioactive material to the atmosphere, a faulted or ruptured SG should only be used if no intact SG is available.
11 Check intact SG levels.
11.1 Check narrow range levels - - 11.1 Verify total AFW flow to GREATER THAN 31%{48%}.
31%(48%}. intact SGs greater than 395 gpm.
AFW FLOW TO 1A(1B,lC) 1A(1E,1C) SG
[]
[1 FI Fl 3229A
[]
[1 FI Fl 3229B 3229E
[]
[1 FI Fl 3229C AFW TOTAL FLOW
[]
[1 FIFl 3229 11 continued on next page.
Step 11
___Page Completed Page 12 Page 12 of 22
,l....i i.':'
FNP-I-FRP-C.2 FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n
11.2 [CAl (CA] WHEN intact SG narrow range level 31%-65%{48%-65%},
31%-65%(48%-65%},
THEN maintain intact SG narrow range level 31%-65%{48%-65%}.
31%-65%f48%-65%}.
11.2.1 Control MDAFWP flow.
[] AA TRN reset
[]
j BB TRN reset MDAFWP TO lA/IB/IC lA/lB/iC SG SC B
B TRN
[]
[1 FCV 3227 in MOD Intact SGSC lA 1A IB lB lC 1C MDAFWP TO lA(lB,IC) 1A(1BiC) SG QIN23HV Q1N23HV ((1] 3227A ((1] 3227B (()) 3227C in MOD in MOD in MOD MDAFWP TO lA(lB,IC) 1A(1B,1C) SG SC FLOW CONT HIC []3227AA [] 3227BA []3227CA
[]3227BA adjusted adjusted adjusted Step 11 continued on next page.
_Page Page Completed Page 13 of 22
\ . ,J .i.e! I,!: i.
FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n II I1 II 11.2.2 Control TDAFWP flow.
TDAFWP FCV 3228
[] RESET reset TDAFWP SPEED CONT
((1] SIC 3405 adjusted Intact SG 1A 1B lB 1C TDAFWP TO 1A(1B,lC) 1A(1B,1C) SG Q1N23HV 01N23HV []
[1 3228A [] 3228B []
[1 3228C in MOD in MOD in MOD TDAFWP TO 1A(1B,lC) 1A(1B,1C) SG FLOW CONT HIC [] 3228AA
[]3228AA [] 3228BA []
[]3228BA 3228CA
[]3228CA adjusted adjusted adjusted
_Page Page Completed Page 14 of 22
.---------r-----------....;:,..--;::.(~~-;---',..---------..__--------...,
FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING 17 Revision 17 Step Action/Expected Response Response NOT Obtained n I I CAUTION: Performance of step 12 will cause accumulator injection which may result in a red path on the INTEGRITY status tree. This procedure FNP-1-FRP-P.1. RESPONSE TO should be completed before transition to FNP-1-FRP-P.1, IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS.
NOTE: After the low steam line pressure SI is blocked,blocked. excessive opening of steam dumps can cause a high steam flow LO-LO TAVG main steam isolation signal.
12 Reduce pressure in all intact SGs to 100 psig.
12.1 WHEN P-12 light lit (543°F).
(543°F),
THEN perform the following.
12.1.1 Block low steam line pressure SI.
STM LINE PRESS SI BLOCK - RESET
[] A A TRN to BLOCK
[]
[1 B B TRN to BLOCK 12.1.2 Verify blocked indication.
BYP && PERMISSIVE STM LINE ISOL.
SAFETY INJ.
[]
[1 TRAIN A A BLOCKED light lit
[] TRAIN B B BLOCKED light lit 12.1.3 Bypass the steam dump interlock.
STM DUMP INTERLOCK
[]
[]
[1 B B TRN to BYP INTLK Step 12 continued on next page.
PagePage Completed Page 15 of 22
!,; ,'<:v. ..:I FNP-1-FRP-C.2 FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE RESPONSE CORE COOLING Revision 1717 Step Action/Expected Response Response NOT Obtained n 1 I 12.1.4 12.1.4 Adjust steam header pressure controller to pressure control cooldown rate.
[] PK 464 adjusted
[1 12.2 12.2 [CAl Maintain
[CA] Maintain RCS cold cold leg cooldown rate - LESS THAN 100°F IN ANY 60 MINUTE PERIOD.
100°F PERIOD.
12.3 available.
IF condenser available, 12.3 Dump steam to atmosphere.
THEN dump steam to condenser from intact SGs. 12.3.1 Direct counting room to FNP-0-CCP-64S. MAIN perform FNP-O-CCP-645, BYP && PERMISSIVE STEAM ABNORMAL COND ENVIRONMENTAL RELEASE.
AVAIL
[]
[1 C-g C-9 status light lit 12.3.2 IF normal air available.
available, THEN control atmospheric
))N STM DUMP relief valves to dump steam
[] MODE SEL A-B TRN in STM PRESS from intact SGs.
SGs, NOT. dump steam using IF NOT, STM DUMP FNP-1-S0P-62.0.
FNP-l-SOP-62.0, EMERGENCY INTERLOCK AIR SYSTEM.
[]
[1 A A TRN in ON
[]
[1 B B TRN in ON 1A(lB.1C) 1A(1B.1C) MS ATMOS REL VLV STM HDR [] PC 3371A adjusted PRESS []
[1 PC 3371B adjusted
[] PK 464 adjusted []
[1 PC 3371C adjusted 12.3.3 II IF no source of air available.
available, THEN locally control SG atmospheric relief valves with handwheel to dump steam from intact SGs.
(127 ft. AUX BLDG main steam valve room)
Intact SG 1A 1B lB 1C Q1N11PCV [] [1 3371A []
[1 3371B [][1 3371C 12.4 Check all intact SG pressures 12.4 Return to Step 11. OBSERVE
- LESS THAN 100 psig.
CAUTION PRIOR TO STEP 11.
Step 12 continued on next page.
_Page Page Completed Page 16 of 22
X) .l,\i X E :
FNP-I-FRP-C.2 FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE RESPONSE CORE COOLING Revision 17 17 Step Step Action/Expected Response Response NOT NOT Obtained n I I I 12.5 Check at least two RCS hot leg 12.5 12.5 Return to Step Step 11.11. OBSERVE temperatures - LESS THAN
- CAUTION PRIOR TO STEP 11. 11.
350 0 F.
350°F.
RCS HOT LEG TEMP Ii[] TR 413 12.6 12.6 Stop SGSG pressure reduction.
[1[] PK 464 adjusted lA(lB,lC) 1A(1B,1C) MS ATMOS REL VLV
[]
[1 PC 3371A adjusted
[]
[1 PC 3371B adjusted
[]
[1 PC 3371C adjusted OR lA(lB,lC) 1A(1B,JC) MS ATMOS REL VLV
[] QINIIPCV3371A Q1N11PCV3371A closed
[]
[1 QINIIPCV3371B Q1N11PCV3371B closed
[] QINI1PCV3371C Q1N11PCV3371C closed CAUTION: Pump damage may occur if RHR pumps are operated on miniflow for longer than 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> with no CCW supplied to the RHR heat exchangers.
13 Verify RHR RIIR pumps - STARTED.
RHR PUMP
[1 1A amps >
[] > 00
[]
[1 IB lB amps > > 00
_Page Page Completed Page 17 of 22
~h,) I '\i pel, .t1 FNP-I-FRP-C.2 FNP-1-FRP--C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n I I 14 [CAl
[CA] Check if SI accumulators should be isolated.
NOTE: Step 14.1 is a continuing action.
14.1 [CAl
[CA] Check at least two RCS 14.1 Perform the following.
hot leg temperatures - LESS THAN 350°F. 14.1.1 WHEN at least two RCS hot leg temperatures are less RCS HOT LEG TEMP than 350°F,
[]
[1 TR 413 THEN perform steps 14.2 and 14.3 to isolate accumulators.
14.1.2 Proceed to step 15.
OBSERVE CAUTION PRIOR TO STEP 15.
14.2 Reset SI. 14.2 IF any train will NOT reset j
using the MCB SI RESET
[] MLB-l MLB-1 1-1 not lit (A TRN) pushbuttons,
[] MLB-l 11-1 not lit (B TRN) THEN place the affected train S821 RESET switch to RESET.
(SSPS TEST CAB.)
14.3 Close all SI accumulator 14.3 Perform the following.
discharge valves.
14.3.1 Vent any SI accumulator lA(lB,IC) 1A(lB,1C) ACCUM that cannot be isolated.
DISCH ISO
[] QIE21MOV8808A Q1E21MOV88O8A ACCUM
[] QIE21MOV8808B Q1E21MOV88O8B N2 VENT
[] QIE21MOV8808C O1E21MOV88O8C [] HIK 936 open SI ACCUM lA 1A IB lB lC lA(lB,IC) 1A(lB,1C) ACCUM N2 SUPP SUPP/VT IVT ISO QIE21HV Q1E21HV []
[1 8875A [] 8875B [] 8875C open open open 14.3.2 IF an accumulator can NOT be isolated or vented, THEN consult the TSC staff to determine contingency actions.
_Page Page Completed Page 18 of 22
'~} t'z; 1 f. :1.
FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n I I CAUTION: Core cooling may degrade during subsequent steps. FNP-1-CSF-0.2.
FNP1-CSF-O.2, CORE COOLING status tree should be closely monitored.
15 Stop all Reps.
RCPs.
[]
[1 lA1A
[]
[1 1BlB
[]
[1 1C 16 Reduce pressure in all intact SGs to atmospheric pressure.
16.1 Maintain RCS cold leg cooldown rate - LESS THAN 100°F
- 100° F IN ANY 60 MINUTE PERIOD.
16.2 IF condenser available.
available, 16.2 Dump steam to atmosphere.
THEN dump steam to condenser from intact SGs. 16.2.1 Direct counting room to perform FNP-0-CCP-64S.
FNP-O-CCP-645, MAIN BYP && PERMISSIVE STEAM ABNORMAL COND ENVIRONMENTAL RELEASE.
AVAIL
[]
[1 C-9 status light lit 16.2.2 IF normal air available.
available, THEN control atmospheric STM DUMP relief valves to dump steam
[]
[1 MODE SEL A-B TRN in STM PRESS from intact SGs.
SGs, IF NOT. dump steam using STM DUMP FNP-1-S0P-62.0.
FNP-1-SOP-62.O, EMERGENCY INTERLOCK AIR SYSTEM.
[]
j A A TRN in ON
[]
[1 BB TRN in ON 1A(lB.1C) 1A(1B.1C) MS ATMOS REL VLV STM HDR []
[1 PC 3371A adjusted PRESS ((1] PC 3371B adjusted
[]
[1 PK 464 ((1] PC 3371C adjusted Step 16 continued on next page.
_Page Page Completed Page 19 of 22
i./ }',! ,i...~ ,\
FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n I 16.2.3 16,2.3 IF no source of air available.
available, THEN locally control SG atmospheric relief valves with handwheel to dump steam from intact SGs.
(127 ft. AUX BLDG main steam valve room)
Intact SG Q1NllPCV O1N11PCV 1A
[] 3371A []
[1 1B lB 3371B I []
[1 1C 3371C 3371C 3371A1 3371B1 17 Verify any SI flow established. 17 Perform the following.
A A TRN
[]
[1 FI Fl 943
- Any form of RCS injection.
B B TRN RECIRC FLOW 17.2 Return to Step 16.
[]
[1 FI Fl 940 OR
- Verify any LHSI flow - -
33 gpm.
GREATER THAN 1.5x10 1.5x1O 1A(lB) 1A(1B)
[] FI Fl 60SA 6O5A
[] FI Fl 60SB 605B
_Page Page Completed Page 20 of 22
t) I. ',i Ii j:v FNP-I-FRP-C.2 FNP-1-FRP-C.2 RESPONSE TO DEGRADED CORE COOLING Revision 17 Step Action/Expected Response Response NOT Obtained n I I 18 Check core cooling. 18 Return to step 16 16..
- REACTOR VESSEL LEVEL indication - GREATER THAN 0%
UPPER PLENUM.
- At least two RCS hot leg temperatures - LESS THAN 350 350°0 F.
RCS HOT LEG TEMP
[]
[1 TR 413 19 Go to FNP-I-EEP-l.
FNP-1--EEP-1, LOSS OF REACTOR OR SECONDARY COOLANT.
step 14.
-END-
-END Page 21 of 22
- 5. 004K1.04 OQI1NEW/ROICIA 3.4/3/8/N/N/2JCVRIY 004K 1.04 OQ1/NE\iY/RO/C/A 3.4I3I8ININI2ICVRIY Unit 1 a'CfbO%, and the following conditions occurred:
I is atiOO%,
- One Letdown orifice is on service.
- LK-459F, PRZR LVL, controller demand has failed high.
Which one of the following describes the effect on Charging Flow and RCP Seal Injection flows, with no operator actions?
Charging Flow Seal injection iniection Flows A'I A Go up Go Down B. Goup Go up Goup Go up C. Go Down Goup Go up D. Go Down Go Down Page: 11 eX d 200 1211412009
A - Correct. When FK-122 fails high, charging flow increases. This robs flow from the Seal injection lines and the Seal Injection flows go down. When Seal Injection flows go down, #2 seal flow and leakoff flow also goes down, since it is supplied by by Seal Injection flow. When charging flow goes up, and letdown is unchanged, VCT level Injection goes down. VCT pressure goes down due to expansion of the gas volume in the seal leakoff flow to the VCT goes VCT. When pressure in the VCT goes down, #1 sealleakoff up due to less back pressure.
B - Incorrect. The first part is correct (see A). The second part is incorrect, due to the immediate effect of Seal Injection decreasing due to charging flow being in parallel with Seal inj. Flow. Charging flow increasing robs flow from seal injection flow.
Plausible, since the VCT pressure goes down as VCT level goes down and the Number 1I seal leak off does go up eventually due to the VCT pressure drop, but the seal injection flow does not go up.
C - Incorrect. Charging flow goes up due to the direct relationship between the master LK-459 Pressurizer level controller and the slave FK-122 controller, and the valve FCV-I 22. Plausible, since some of the MCB master slave controllers position of FCV-122.
have an inverse relationship, such as PK-444A, PRZR PRESS REFERENCE controller and PK-444C & D, 1A & 1B 1 B LOOP SPRAY VLV controllers. The SPRAY VLV controller demands go up when the REFERENCE controller demand goes up.
The second part is incorrect (see B). Plausible, since if the first part were correct, the second part would be correct also.
D - Incorrect. The first part is incorrect (see C). The second part is correct (see A).
Plausible, since physical connections and the cause/effect relationships between the CVCS system and the RCPS may be misunderstood and confusion could exist as to the inverse relationship between the two flows.
CVCSIHHSI/ACCUMULATOR/RMWSA-181009 FSD: CVCS'HHSI/A CCU MULA TORlRMWS A-181 009 P1 D 175039 SH 6, eves PI CVCS chg & seal injection REACTOR COOLANT PUMPS, OPS-621 OPS-62101D, OPS-5210ID, 01 D, OPS-521 OPS-40301D, 01 D, OPS-40301 D, STUDENT TEXT Page: 12 12 of 200 1211412009 12/14/2009
Previous NRC Previous NRC examexam history history ifif any:
004K1.04 Chemical 004 Chemi Volume cal and Vol Control System ume Control System KI1 Knowledge K KncMlledge of of the the physical connections and/or phycaI connections andlor cause-effect caus-effect relationships rdationips between the eves bween the CVCS and and the the fdIeiing sjem (CFR: 41.2 to 41.9 follcming ~sl:ems: 41.9/45.7
/ 45.7 to 45.8) injection flows ................................ 3.4 3.8 K1.04 RCPS, including seal injEdion Match justification: The RCP #2 seal flow and the Seal Injection flows are both affected by the CVCS system during a Charging flow and/or VCT level/pressure transient. To correctly answer this question, knowledge of this relationship as well as the physical connections is required.
Objective:
- 4. EVALUATE plant conditions and DETERM I NE if any system components need to be operated while performing AOP-100, Instrument Malfunction. (OPS-S2S21Q06). (OPS-52521Q06).
- 2. RELATE AND IDENTIFY the operational characteristics including design features, cacities and protective interlocks for the components associated with the Chemical and capacities Volume Control System, to include the components found on Figure 3, Chemical and Volume Vol ume Control System and Fi Figure RCP-Seal IInjection gure 4, RCP-Sea1 nj ecti on System (OPS-40301 F02).
Page:
Page: 13 13 of of 200 200 1211412009 12/14/2009
Date:
Date: 10/7/2009 10/7/2009 -
l-8404A 4048 Time Time:11:08:41AM 11 :08:41 AM jI 3/4T78 178 2CS250TR WHO 941 441 WH 3/4 -CS2501R 3/4 CO 2501R F 2-CS-2501R002568 QV259A 181096 M 1-8109A tl -175039 175039 SN.1 SR 8-12 2TU78FISA 2_TM 78FNW eHG COO PUMP PUMP BYPASS BYPASS l
COOS CR60018 01111CC NO ORIFICE 90.11 C'lCS 6SR.1 CHARGING 1 F22 ITEM ITEM. ORCP SRCP j
........ 110E386 SN.1 LOC(i-1) FO 70 6004 6084 (HOlE NOSE 4) 4)
00405.4
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SOS RAD 143 55 RAT 0VS66A 005660 3 COB 3CS2S018-2CS25O1R DR 0V575D 0 75A i_.4__/
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110E398 00.1 L0C C DV3206 3/4W P0 428 P0142A [}0 002590 060 PUMP BYPASS 185 IT 181320
/
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REF.
121974 ANNUN C
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/ SlEETY CI.A8S2A 0 005 )A0W 4) Y 4_178 56
,-----1;~~:~----, 073 ILEOI ISA 003248 003268 181306 94 808927898 080328 4 3/4 C l3/lEREO
%_GM 7 FNAU]f RAT 004956 FENT 004959 T I84799 3/4-178 8049205W LOOP SEAL 00459
$ 003248 3/4
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~~~~o TEST II I IES2501R l-CS-2S01R 094858
/
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I 3 078 78 1.4 I 3/4176 CONNECTION CONNECTION x 3ID 7 \ / 6ESISTR 00485
- ....L. I 001286 841 I 183854 004949 SV494A QV494B CO L4 I 3/4 178 3/4 3/'- 00 ~I-.... 4 t--l 144818 N0TEI 001828 I I---:' ....L._
065668 3C78 / I 30225018 OVS7SE 55759 / 7I/0TE 6 65.92 I 901 27C DVT27A 004936 004938 005660 OR I14 I 00130% / 003278 001648 183848 I 2CS-25OIR 4382% 4 e43/4 VENT 1-81336 3/4W 2 CC818 1-8511
/ 2176 I 2178 / / 001298 40878F89 DAWN 2ES2SOIR 3/4-178 I 1-83885 CR0 PUMP I SEAt INjECTION F ORIFICE 60.3 FILTER I 3/4CS2SOIR OEM: ORCP I 0A128C SlEETY CLASS-2A SAFETY CLASS 28 WHT 01E21FO048 004898 I 18386C 002S9C 70 6006 )ROTE 4) 0032S% I
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/ DAWN :
0 175039 SH.l 0-175039 SAl J-9 J 9 Li!i_f eveG !XAL INJ 001300 000300 +7T 101204 00128 00330 <.ZiSTH1C-8I 11OE86 SH.l SR. LOC (a-9) \S 1838491 (5 8388 806186 / \ SIS SOS RHRS RARS HX HA 110E86 2T761 IV 3078 3 887 8 00 / \ I 110E389 LOC G-6 110[389 06 A 00 REDI 0ON - 3$C8-2o OV263B 000838 \FlY
\ r I 1-81168 181168 'b--_l'--ZO-175037 D 175037 SH.2(C 50.0)01!
-I 001290 SlEETY ELISS2A 96 3/4 3/4 RA72SAB ~
3/4-RV72SWB t Res ACS PRT PAT 5 0048980097 OTE2IFOD4B IIOE3A7 SH.2 110[.387 SH2 lOC 2CC8 IN 3/4-CCD24T 3/4"-CCD-240 9012S 06557 UI DA DH N1E21Y136 H1E2lV136 19388 004906 4 2T78 AS
-/
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Title:
C:\Reference Disk\Exam Reference Disk\Drawings\D1 Disk\Drawings\D17S039-0006.cal 75039-0006.cal
REACTOR COOLANT PUMPS OpsRcp012 OpsRcpOl 2 RCP SEAL STANDPIPE
___ ~.;r __ _
ABOVE #3 SEAL TO
RMVV SYSTEM 8539
'-.l > ~~DT FROM FROM HX eves CHG. PUMP 81 SEAL INJECTION DISCHARGE REACTOR COOLANT OUT TO PRT 8121 8121 1SS PSIG)
(150PSIG)
- 1
- 1 SEAL SEAL BYPASS BYPASS REACTOR COOLANT REACTOR COOLANT IN IN Figure Figure 10 10 - RCP
- Rep - Seal
- Seal Injection Injection System System OPS-62101D-52101D140301DIESP-52101D- Version 11 OPS-62101D-52101D!40301D/ESP-52101D-
/"' '\
6 005A
- 6. 005A11.07 07 001/NEyy/RO/C/A OOi/NEW/RO/C/A 2.5/3.1/N/N/3/CVRIY 25/3 1/N/N/3/CVR/Y stroke of Q1 A time strole Eli MOV8889, Qi E11 M0V8889, RHR TO RCS HOT LEGS ISO, in the open STP-l 1 .6, Residual Heat Removal Valves Inservice direction has been performed per STP-11.6, Test.
Open direction ACCEPTABLE STROKE TIME RANGE is 9.96 to 13.47 Sec.
Open direction MAXIMUM ALLOWABLE TIME is 16 Sec.
Stroke times obtained were as follows:
- At 1000 First time stroke: 15.35 Secs
- At 1005 Second time stroke: 15.52 Secs Which one of the following describes MOV-8889 OPERABILITY lAW Technical Specifications and what the CR will require for these results lAW STP-11.6?
STP-1 1 .6?
A'I*
A MOV-8889 is OPERABLE
- Analysis of the time stroke results within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to determine if new stroke time is acceptable.
B.
- B.* Declare MOV-8889 INOPERABLE
- Analysis of the time stroke results within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to determine if new stroke time is acceptable.
C.*
C. MOV-8889 is OPERABLE
- Repair or replacement of MOV8889.
D.
- D.* Declare MOV-8889 INOPERABLE
- M0V8889.
Repair or replacement of MOV8889.
A - Correct. Tech Specs requires the time stroke to be less than the Maximum stroke STP-1 1 .6, Step 5.3.3.4 & 5.3.3.5, but time. This is stated as acceptance criteria in STP-11.6, outside of the Acceptable Stroke Time Range the valve must have a retest and an analysis if the stroke time is still outside of the Acceptable range but less than the STP-1 1.6, Step 5.4.2.
maximum. The second part is correct per STP-11.6, B - Incorrect. The first part is incorrect, but plausible. The tech spec limit is the same B -
as the maximum time for the valve stroke. A stroke time above the maximum does not meet acceptance criteria and requires declaring the valve inoperable, but above the acceptable range AND below the Max time meets acceptance criteria.
Acceptable range may be confused with acceptance criteria. Not meeting acceptance criteria indicates inoperability due to TS requirements not being met.
Also, if either of the tests were greater than the maximum, or if no retest was possible this choice would be correct. Second part is correct (see A).
C - Incorrect. First part is correct. Second part is incorrect but plausible. Writing a CR is required, but requiring repair or replacement of the valve is only required for a Page: 14 of 200 1211412009
valve that hashas been been required required to be be declared declared inoperable inoperable (no (no analysis in in 96 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, hours, greater than MAX greater MAX stroke stroke time, or or nono retest retest possible possible and outside outside of the acceptable acceptable range). Analysis and range). and possible possible resetting resetting the baseline baseline of the valve stroke time is is required by allowed and required by the STP.
D - Incorrect.
D - Incorrect. First First part part isis incorrect incorrect (see (see B).
B). Second part part is is incorrect incorrect (see(see C)C) but but plausible. If the first part was correct, then this would be correct per STP-11.6, STP-1 1 .6, AND tech specs would not be met until the valve was repaired to allow time stroking in less than the Max allowed time.
STP-1 1.6 step 5.4 Version 36 STP-11.6 compareActual 5.4 In Table 1, compare Actual Stroke TimesTimesto AllowableTimes to Maximum Allowable Times and to the Acceptable Acceptabl e Stroke Ti me Range and perform the foil following applicable:
owi ng as appl i cabl e:
theActual 5.4.1 IF the Actual Stroke Time for a valve exceeds exceedsthe the Maximum Allowable Time, THEN perform the following:
- 1. Decl are the val valve ve iinoperable.
noperabl e.
appropriate
- 2. Check the appropri ate TTechnical Specifications, echni cal Speci Technical fi cati ons, T echni cal Requi rements Requirements Manual, and Fourth 10-Year Interval 1ST Program for corrective action requirements requi rements.
5.4.2 IIF F the Actual Stroke Ti me for a valve is outside the Acceptable Stroke Ti me Range AND does NOT exceed the Maximum Allowable Time, THEN perform the following:
- 1. IImmediately mmedi ately retest the valve.
possibleto
- 2. IF it is NOT possible thevalve, to retest the valve, THEN declaredeclarethevalve the valve inoperable.
- 3. IIFF the thevalveis val ve is retested AND the thesecond second set of data is aI so outsi isalso outsidethe de the Acceptable Stroke Time Ti me Range, THEN perform the following:
- a. Submit a CR to have the data analyzed withi withinn 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to verify that the new nw stroke time represents acceptable valve operation.
- b. Enter the CR number in T Table abl e 1.
- c. Initiate an Admin LCO to declare declarethevalve the valve inoperable if not analyzed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
- 4. IF IFthevalve the valve is retested AND the second set of data iswithin the Acceptable Stroke Ti me Range, THEN analyze the cause of the initial deii Time ati on and submit deviation aa CR to have the results resul ts documented in the Record of Tests.
5.4.3 IF any valve is declared inoperable, THEN perform the following: followi ng:
- 1. Resolve the unacceptable condition by performing one of the following:
- Repair
- Repai r the valve.
val ve.
- RepI
- ace the valve.
Replace
- Analyze
- Anal yze the associated associ ated valve val ve stroke data to determine determi ne the cause of the deviation devi ati on and whether valve operation is acceptable as is.
- 2. Prior to returning any valveto valve to servicefollowing service following repair, replacement, or analysis, writeaCRto requestthat analysis, write a CR to request that ESissuen, ES issue new baselinedata baseline data.
Page:
Page: 15 15 of of 200 200 1211412009 1211412009
Previous NRC Previous NRC exam exam history history ifif any:
any:
005A1 .07 005A1.07 Residual Heat 005 Residual 005 Heat Removal Removal System System Al Ability A1 Ability to predict and/or to predict monitor changes andlor monitor changes in in parameters paramers(to preient exceeding (to prevent desgn limit~
exceeding design limit aSBX:iated adated with operating the RH RS controls contrds induding: (CFR: 41.5/45.5) 41.5/ 45.5)
AA1.07 Derminion of 1.07 Determina:ion of test a::ceptcbility aDctalHity by comparison of rocorda::l r&ordarl valve response times with Te:::h-Spe:::
ronsotimeswith Th-Sp 2.53.1*
requirenents .............. 2.53.1*
r&:luiranents Match justification: Recorded values of valve response times are given and the applicant is required to assess whether or not Tech Specs are met on the RO level of knowledge. The STPs are the mechanism with which ROs assess operability of valves per their stroke times. The "Tech-Spec Tech-Spec requirements" requirements are assessed in the valve stroke STPs with acceptance criteria being met or not met.
This question provides a stroke time with a retest (directed by the procedure in this case) and the applicant must assess "Tech-Spec Tech-Spec requirements" requirements as to declaring inoperable or not (in the first part of the answers), and further actions per the STP (in the second part of the answers).
Objective:
I 1 RECALL AND APPLY the LCO and APPLICABILITY for Technical Specifications (TS) or TRM requirements, and the REQUI RED ACTIONS for 11 HR or less TS or TRM requirements, requi rements, and the rei releiant portions evant porti BASES ons of BA SES that 0 DEFI NE EFI N OPERABILITY E the OPERA B I L ITY and APPLICABILITY of the LCO associated with the Residual Heat Removal System components and attendant equipment alignment, to include the following (OPS-521 01 K01):
(OPS-52101 KOl):
- 3.4.3, RCS Pressure and Temperature (PIT) Limits
- Loops MODE 4 3.4.6, RCS Loops-
- Loops MODE 5, Loops Filled 3.4.7, RCS Loops- -
- 3.4.8, RCS Loops- MODE 5, Loops Not Filled
- 3.5.2, ECCSOperating ECCS - Operati ng
- 3.5.3, ECCS - Shutdown
- 3.9.4, Residual Heat Removal (RHR) and Coolant Circulation - High Water Leiel - Level
- 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation - Low Water LeieI - Level
- 13.5.1, 13.5.1, Emergency Emergency Core Cooling Cool i ng System (ECCS)
Page: 16 of Page: 16 of 200 200 1211412009 1211412009
05/12/09 12:38:23 05112/09 FNP-I-STP-ll.6 FNP-1-STP-1 1.6 5.3 Q1E1 1M0V8889 Exercise Test is IF the RHR TO RCS HOT LEGS ISO, QIEIIMOV8889 required, THEN perform the following:
5.3.1 Open the following valves:
5.3.1.1 RHR TO RCS HOT LEGS HDR TEST CONN ROOT, Q1E1 1V048A.
QIEll V048A.
5.3.1.2 RHR TO RCS HOT LEGS HDR TEST CONN ROOT, Q1E1 1V048E.
QIEll V048E.
5.3.2 Record the pressure on NINl EIIPI2262.
El 1P12262.
PRESSURE _ _ _ _ _ _ PSI.
PRESSURE 5.3.3 IF the pressure indicated on N1E1 jf 1P12262 is less than Q.
NIEI1PI2262 OR equal to 550 psig, THEN perform the following:
5.3.3.1 [f required to isolate RHR TO RCS HOT LEGS ISO, IF Q1EI 1M0V8889 from an operating RHR train, THEN QIEIIMOV8889 close RHR TO RCS HOT LEGS XCONN, Q1E1 1M0V8887A.
QIEIIMOV8887A.
5.3.3.2 IF required to isolate RHR TO RCS HOT LEGS ISO, Q1E1 1M0V8889 from an operating RHR train, THEN QIEI1MOV8889 close RHR TO RCS HOT LEGS XCONIST,XCONN, Q1E1 1MOV8887B.
QIEIIMOV8887B.
5.3.3.3 Unlock and close disconnect switch QIRI8B036-B.
Q1R18BO36-B.
4
- 5334
- 5.3.3.4 ro~R (pen)RHR TO RCS HOT LEGS ISO QIEI "~?5' Q1E1 LOV8889) g ~d1ecord i~crBAL htl4ecord time required for valve opening inih-ACEUAL STROKE TIME column of Table Tabk I.1.
/IACCEPTANCE CRITERIA:
LEPTANCE Stroke times are less than or equal to Maximum Maximum\
\L__---------"--~ Allowable Times listed in Table 1.
~
3
- *535
- 5.3.3.5 Clos~ RHR TO RCS HOT LEGS ISO QIEll Q1E11O ~OV888D V888 a record time required for valve closing in t~~l:::JAL thA+UAL STROKE TIME column of Table 1.
/'~
'\
I
(
ACCEPTANCE CRITERIA: Stroke times are less than or equal to the Maximum ) .
Allowable Times listed in Table 1. I Version 35.0
05/12/09 12:38:23 05112/09 FNP-1-STP-1 1.6 FNP-I-STP-ll.6 5.4 1, compare Actual Stroke Times to Maximum Allowable Times and to In Table 1, the Acceptable Stroke Time Range and perform the following as applicable:
5.4.1 IF the Actual Stroke Time for aavalve valve exceeds the Maximum Allowable Time, THEN perform the following:
Allowable CCI 1. Declare the
- 2. Check the appropriate Technical Specifications, Technical Requirements Manual, and Fourth 110-Year 0- Year Interval 1ST Program for corrective action requirements.
~4 5.4.2 IF ff the Actual Stroke Time for a valve is outside the Acceptable Stroke Time Range AND does NOT exceed the Maximum Allowable Time,
_THEN perform the following:
- 1. Immediately retest the val~_ valy
,,+ _
~~--"---.--.--
2.
,;i)2. IF it is NOT possible
!~_itiS.NOIJl~. ssible_!~!~tes~:~~~alve, to retest the valve, THEN THEN d~lar~~~
declare the A
" { \" ~~t.!'~ valve ~~£erable.
valve inoperable. ______________ - __.
I ~~-i/
2 c4 !l~--j,- .!Eili~ IF the val~e valve i~ted is retested AND the second s~t set of d~ta data i; is al;;-----'
also outside the Acceptable Stroke Time Range, THEN perform the Ei /. 4" following:
~ following:
C,*~,,/,
"ftcJ/{3> -,~ a. Submit a CR to have the d~..illY.zed data anajyzed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to
-? ~'\J~:>t __ ~erIThat1fiei1ew stro~epresent~ceptable iihitThktime represents acceptable valve vafTe operatIOn. --
- b. Enter the CR number in Table 1.
- c. Initiate an Admin LCO to declare the valve inoperable if not analyzed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
- 4. IF the valve is retested AND the second set of data is within the Acceptable Stroke Time Range, THEN analyze the cause of the initial deviation and submit a CR to have the results documented in the Record of Tests.
5.4.3 IF any valve is declared inoperable, THEN flE prform perf.Qrm the following:
,..--- ---=::=,;;;;"""'= ....----:::=. -------.
fL /1 . Rfi!l'~lolv.e~
Resolve c()!2~!~_n by unacceptable condition the unacceptable performing ~ne b)-'1>erforming o~!!
one ouil*
( following:
o owmg:
- Repairthe Repair the valve. ______
- Replace the valve-r------___ valve.
~----.-.-~
Cl .
- Analyze the associated valve stroke data to determine the cause of the deviation and whether valve operation is acceptable as is.
- 2. Prior to returning any valve to service following repair, replacement, or analysis, write a CR to request that ES issue new baseline data.
Version 35.0
/*******1 005K4.03 OO1/NEy?fROICIA
- 7. 005K4.0300;VNE IRO/C/A 2.9/3.21N/N/4/CVRIY 2.9/3.2JNINI4ICVRIY Unit 1I is p'elforming pilorming a plant cooldown using the A Train RHR system, and the following conditions occurred:
At 1000:
- HIK-603A demand setting is increased to 60%.
At 1005:
- RHR system flow is stable.
At 1010:
- Instrument Air is lost to FCV-605A due to an air supply line break.
Which one of the following describes, with no operator actions:
- 1) RH R flow indicated on FI-605A Fl-605A at 1005, and
- 2) the position of FCV-605A at 101 1010?
O?
At1005 At 1005 AtlOlO At1010 FI-605A, RHR HDR FLOW FCV-605A, 1A RHR HX BYP FLOW A. > > 3100 gpm Closed B. 3100 gpm Open C. > > 3100 gpm Open D~
D 3100 gpm Closed Page: 17 ci of 200 1211412009
AA - Incorrect.
- Incorrect. This This first first part part is is incorrect, incorrect, since since even even thought thought the the flow does does initially initiaNy go go up, up, the FT the FT senses senses this this and and the the HX HX BYP BYP FKFK demands demands the HX HX BYP BYP FCV FCV to close close down down to maintain the 3100 gpm initial maintain initial flow. Plausible, Plausible, since flow does does go go up initially. Also, ifif up initially.
BYP FCV the BYP FCV isis in in manual manual which itit normally normally is, is, this this choice choice would bebe correct.
correct. TheThe second part second part is is correct (see (see D). D).
B - Incorrect. The first part is correct (see D). The second part is incorrect, but plausible. The valve fails closed on loss of air to maximize flow through the HX during a LOCA, but this valve could be confused with the HX discharge valve which fails open for the same reason.
C - Incorrect. The first part is incorrect (see A). The second part is incorrect (see B).
D - Correct. The design for the RHR HX BYP FCV is to operate in auto to maintain the total system flow rate constant while flow through the HX is adjusted with the potentiometer for the HX DISCH valve. The fail position of the valve is closed.
Remcwal-Lcm Head Safety I njection Functional System FSD: A181002, Residual Heat Removal-LON Syetem Description De&:ription 3.15 RHR HEAT EXCHANGER DISCHARGE VALVES 5.1 RHR HEAT EXCHANGER BYPASS FLOW CONTROL Page:
Page: 18 18 of of 200 200 1211412009 1211412009
Previous NRC Previous NRC exam history if any:
005K4 .03 005K4.03 Heat Removal 005 Residual Heat Removal System K4 Knowledge K4 Kncmledge cl RH RS ci RH RS design feature( and/or desgn feature(s) andlor interlock(s) prcwide or the following:
inter Iock( which prOllide follcming:
(CFR: 41.7) h exchooger K4.03 RHR he:t echaige bypassflow byps flow control ............................ 2.93.2 2.9 3.2 Match justification: The design features of normal cooldown operation of the RHR HX BYP FCV (in auto adjusting to maintain constant total flow vice adjusting to maintain constant valve position-first part of choices) and the design fail position of the valve (closed-second part of choices) must be understood to correctly answer both parts of this question.
Objective:
- 7. DEFI NE AND EVALUATE the operational implications of normal I abnormal plant or DEFINE equipment conditions associated Equipment sociated with the safe operation of the Residual Heat Removal System components and Equipment, include the following (OPS-40301K07):
equipment, to includethefollowing (OPS-40301 K07):
- Normal Control Methods
- Abnormal A bnormal and Emergency Control Methods (Changes in system flow rates, Loss of control from the control room)
- setpoi nts (examples-Automatic actuation including setpoints (examples Reactor Trip, SI, PhaseA,
- Phese A, LOSPIIoss of all AC power)
LOSP/Iossof
- Protectiveisolationssuchashighflow, Protecti le,el iincluding ve i sol ati ons such as hi gh flow, low pressure, low level setpoint ncl udi ng setpoi nt
- Protective interlocks
- Actions needed to mitigate the consequence of the abnormal abnormalityity Page: 19 cl 200 19 of 1211412009
FNP Units 1I &
&2 RESIDUAL HEAT HEAT REMOVAL A-181002 5.0 NON-CRITICAL COMPONENT FUNCTIONAL DESIGN REQUIREMENTS This section presents the functional design requirements for those components which are not critical to system function as defined in Section 3.0. These components enhance system performance, facilitate system maintenance or provide additional redundancy to components discussed in Section 3.0.
5.1 RHR HEAT EXCHANGER BYPASS FLOW CONJ:R~ COIJRQL Valves:
~CV~_?OS~-iQl.!~ILY~~~:&~=~)
(CV-05A1QLEHV033A l-RHR-FCV:605B (Q1E1 1V033B) l-RHR-FCV-60SB (QIEIIV033B) 2-RHR-FCV-605A (Q2Ell 2-RHR-FCV-60SA (Q2E1 1V033A)
V033A) 2-RHR-FCV-605B (Q2Ell 2-RHR-FCV-60SB (Q2E1 1V033B)
V033B)
Flow Transmitters:
Q1EI IFT6O5A QIEIIFT60SA Q2E1 1FT6O5A Q2EIIFT60SA Q1E1 1FT6O5B QIEIIFT60SB Q2E1 1FT6O5B Q2EIIFT60SB 5.1.1 Basic Functions /7 These air-operated butterfly valves povid/to maintain return flow to the RCS. The valves are normal y closed 1
i:recositioiing during safety injec ion. An orifice type flow 1
transmitter, FT-605A/B, is provided in each ftheLHS1IRHR headers, ddWiiW to 1ow inout for normal coo 5.1.2 Functional Requirements 5.1.2.1 The valves must be designed for pressure and temperature conditions of 600 psig and 400°F. (References 6.4.12, 6.5.6 and 6.S.13) 6.5.5, 6.S.6 6.S.S, 6.5.13) 5.1.2.2 Valve thermal design transient requirements are summarized in Table T-14. (References 6.5.56.S.S and 6.5.6) 6.S.6) 5.1.2.3 Maximum allowable valve Cv equals 10S0 1050 at 600 full open.
(Reference 6.4.12) 5.1.2.4 The design stroke time for opening or closing this air-operated valve is less than or equal to 10 seconds.
5-1 S-1 Rev. 15 IS I
8 006K6.13
- 8. 006K6 13 o01/IT!Sie BAN~/RO/MEM 2.8/3.1/N/N/4/CVRIY OO1/FNP BANK/RO/MEM 28/3 1ININ/4/CVR/Y A Small BreaKTOCA BrékLOCA has occurred on Unit 1, and the following conditions occurred:
- A reactor trip and safety injection is in progress.
- IA 1A Charging Pump failed to auto start.
- IC 1C Charging Pump is the only charging pump running.
- RCS pressure is 1000 psig.
Which one of the following states the Safety Injection flow indication on FI-943, A TRN HHSI FLOW, with no operator action?
approximately__________
Safety Injection flow is approximately _ _ __
A. Ogpm B. 150 gpm Cy C 450 gpm D. 800 gpm Page: 20 of 200 1211412009
Incorrect. See C. Plausible since the meter is labeled "A A -Incorrect.
- A train",
train, and during cold leg recirc this meter indicates only A train flow which is 0 gpm with no A train charging pump running. However, the trains are cross connected during the injection phase, and the B B train pump flow is also indicated by this meter during the injection phase.
Normal Charging flow indicated by FCV-122 indicates 0 for this condition.
B - Incorrect. See C. Plausible, since this is the approximate flow at normal RCS pressure with one charging pump. Also, it is the maximum charging indicated flow through the normal charging flow path at normal RCS pressure, but the normal charging flow path is isolated by the SI signal. Since the RCS pressure is less than NOP, the flow is greater than 150 gpm.
C - Correct. At -1000
- 1000 psig RCS Pressure, one HHSI (Charging) pump can provide lC-73 SBLOCA from 100%
about 450 gpm of flow. [Verified on simulator laptop, IC-73 power, 10,000 gpm leak. With RCS pressure at 1013 psig and one HHSI Pump tripped, HHSI flow on FI-943 Fl-943 was 440 gpm].
gpmj. A knowledge of the exact value of charging flow from one pump at an RCS pressure of 1000 psig is not required to answer this question correctly. A knowledge of the characteristic pump curve for a centrifiugal pump and Charging pump capacity/capability at minimum is required.
Also, knowledge of the system configuration in the injection phase of the LOCA (cross connected trains and both trains flow past the "AA train" train flow indicator).
?OOC 3000 2530 2000 203 1500 1000 500 0
500 1000 0
-500 graph shows single pump curve, parrellel pump (2 pumps) curve, and a generic System characteristic curves for SBLOCA and LBLOCA.
o -Incorrect.
D - Incorrect. See C. Plausible, Two charging pumps could deliver 800 gpm into the RCS during aa Large break LOCA if the RCS was at minimum pressure. This value is that which might be generally recalled from simulator observations for different conditions..
conditions ..
Page: 21 of 200 1211412009 12/14/2009
Previous NRC exam history if any:
006K6.1 3 006K6.13 006 Emergency Core Cool Coolingi ng System Kncmledge of the effect K6 Knowledge offect of a loesor follcwing will have on the ECCS:
losor malfunction on the following 41.7/ 45.7)
(CFR: 41.7/45.7)
K6.13 Pumps ......................................................... 2.83.1 2.8 3.1 Match justification: This question requires knowledge of the effect on the ECCS system flow rate with one pump (HHSI) tripped.
Objective:
- 2. RELATE AND IDENTIFY the operational characteristics including design features, cacities and protective interlocks for the components associated with the Emergency capacities Core Cooling System, to include the components found on Figure 2, Accumulators, Store Tank, and Figure4, Emergency Core Cooling System Figure 3, Refueling Water Storage (OPS-40302C02)..
(OPS-40302C02)
Page: 22 of 200 1211412009
- 9. 007A3.01 007A3.O1 001/FNP OO1/FNP BANK/RO/C/A 2.7/2.9/N/N/3/CVRIY 2712.9ININI3ICVRIY Unit 2 is in Mode 5, and the following conditions occurred:
- P1-472, PRT PRESS, reads 7.5 psig and is rising slowly.
PI-472,
- Ll-470, PRT LVL, reads 78% and is rising slowly.
LI-470,
- RCS Pressure is 225 psig.
- A Train RHR is aligned in the RCS Cooldown Operation lAW SOP-7.0, Residual Heat Removal System.
- It has been determined that V8708A, A Train RHR Pump suction relief valve, is leaking by the seat.
Which one of the following correctly states the impact on the PRT with no operator SOP-i .2, action and the required procedure actions to mitigate this condition per SOP-1.2, Reactor Coolant Pressure Relief System?
The PRT Pressure will reach a maximum pressure of (1) psig, and to prevent reaching the PRT maximum pressure, the operator will be directed to (2) SOP-i .2, Reactor Coolant Pressure Relief System.
per SOP-1.2, A. (1) 150 psig.
(2) pump down the PRT with the RCDT pump and vent the PRT to #7 WGDT, if necessary.
B. (1) 150 psig.
(2) gravity drain the PRT to the WHT.
C. (1) 100 psig.
(2) gravity drain the PRT to the WHT.
D~ (1) 100 psig.
D(l)
(2) pump down the PRT with the RCDT pump and vent the PRT to #7 WGDT, if necessary.
Page: 23 c:I of 200 1211412009
A - Incorrect. Part
- Part 1I incorrect, but plausible, plausible, since RCDT relief is set at 150 psig. Part 2 is correct (see D). D).
Incorrect. Part one is incorrect (see A). Part B - Incorrect.
- Part 2 is incorrect, but plausible, since it would be correct IF the RCDT pumps were inoperable per SOP-1.2 SOP-i .2 step 4.3.3.
Venting should not be necessary in this case due to the low energy of the RCS in
(<200°F), but the procedure does not address lowering pressure by just mode 5 <<200°F),
lowering level. Pressure is high because of level only, lowering level will also lower pressure and could be preferred, but lowering level by gravity draining should only be used if the RCDT pumps are inoperable.
Incorrect. Part 11 is correct (see D). Part 2 is incorrect (see B).
C -Incorrect.
D --Correct.
Correct. Both parts correct. RHR pump suction pressure is approx. the same as RCS pressure in this lineup: 225 psig per the stem. This makes it credible in that it could actually cause the rupture disc to break at it's its setpoint of 100 psig per SOP-i .2 Step 3.5 "PRT SOP-1.2 PRT pressure should be maintained << 100 psig to prevent preíent rupture disc blowout.
blowout."
The PRT has a N2 pressure established of approx. 0.5 to 3 psig to prevent formation of explosive gasses. This bubble will compress as level rises.
FNP-2-SOP-I .2, REACTOR COOLANT PRESSURE RELIEF SYSTEM, Version 30.0 FNP-2-S0P-1.2, 4.4 Reducing PRT Pressure 4.4.1 Have Chemistry verify gas addition to the shutdown gas decay tank to be used for PRT venting (#7 or #8) is acceptable (e.g. H2 < <4%
4% and 02 < 1% per CCP-203).
RCDT Relief valve pressure is 150 psig Per U259507.
PRT Rupture disks blows at 100 psig per SOP-1.2 SOP-i .2 STEP 3.5.
SOP-i .2:
Per SOP-1.2:
4.3.3 Gravity Draining PRT to WHT NOTES: ** This method of draining the PRT should only be used if RCDT pumps are inoperable.
2-SOP-7.0, 2-S0P-7.0, Residual Heat Removal System, Version 79.0 2-SOP-I 2-S0P-1.2, .2, Reactor Coolant Pressure Relief System, Version 31 leiel should be 68-78% during normal operation.
3.4 PRT level 3.5 PRT pressure should maintained shoul d be mai ntai ned < 100 psi g to prevent preient rupture disc di sc blowout.
4.3.2 Draining the PRT Using an RCDT Pump [Normal preferred method]
4.3.3 Gravity Draining Drai ni ng PRT to WHT NOTES:** This method d NOTES: of draining the PRT should only be ud drainingthePRT used if RCDT pumpsare noperabIe iinoperable.
Page: 24 ci 200 1211412009
Previous NRC exam history if any: n/a 007A3.01 007 Pressurizer Relief Tank / Quench Tank System A3 Ability to monitor automatic operation of the PRTS, induding: (CFR: 41.7 / 45.5) 41.7/45.5)
A3.01 Components A3.O1 which diochcrgetothePRT ............................. 2.7*
ComponentswhichdisthagetothePRT 2.7* 2.9 Match justification: This question requires monitoring MCB MOB PRT pressure indication and to know the pressure for automatic operation of the PRTS (at 100 psig the rupture disks ruptures). In this question, a component is discharging into the PRT (RHR Suction relief), and to answer this question, knowledge is required of what pressure will be indicated on the MCB prior to the PRT rupture disk automatically rupturing to relieve the pressure.
Objective:
- 2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Pressurizer System, to include the components found on Figure 3, Pressurizer and Pressurizer Relief (OPS-40301 E02).
Tank (OPS-40301E02).
Page: 25 of 200 1211412009
07/02/09 6:42:21 07/02/096:42:21 FNP-2-S0P-l.2 FNP-2-SOP-1 .2 3.0 3.0 Precautions and Precautions and Limitations Limitations 3.1 3.1 PRT temperature PRT temperature should should not not exceed exceed 120°F 120°F during during normal normal plant plant operation.
operation.
3.2 3.2 PRT nitrogen PRT nitrogen overpressure overpressure of 0.5 to of 0.5 psig should to 33 psig should be be maintained maintained to to prevent prevent formation of formation of an explosive hydrogen an explosive hydrogen - oxygen
- oxygen mixture.
mixture.
3.3 PRT PR pressure should T pressure should not not exceed exceed 66 psig psig during during normal normal plant plant operation.
operation.
level should be 68-78% during normal operation.
PRT level PRT pressure should be maintained <<100 100 psig to prevent rupture disc blowoiN 3.6 At least one of the two reactor vessel head vent system paths, vv",.,,,,,,u consisting f two valves in series powered from the Auxiliary Building DC Distribution System, shall be OPERABLE and closed at all times when in Modes 1-4. 1-4. (TR 13.4.3) 13.4.3) 3.7 While stroking the upstream valve (Q2B13SV2214A or Q2B13SV2214B), MCB closed indication could be momentarily lost on the downstream valve (Q2B13SV2213A or Q2B13SV2213B) due to minor water hammer. This phenomenon is common and documented for Plant Farley and for other plants, and has been evaluated to have no detrimental impact. {CR 200710311420071031 14 3.8 (Q2B13SV2213A or Q2B13SV2213B), While stroking the downstream valve (Q2BI3SV2213A Q2B135V2213B), MCB closed indication could be momentarily lost on the upstream valve (Q2B135V2214A or Q2B13SV2214B) due to rapid depressurization across the (Q2B13SV2214A upstream valve. This phenomenon is documented for Plant Farley and has been evaluated to have no detrimental impact. {CR 200710312007103114 14}} Version 30.0 Version 30.0
07/02/09 6:42:21 FNP-2-S0P-1.2 FNP-2-SOP-1.2 4.3.2.18 RCDT was aligned to WHT, THEN perform the IF RCDTwas following: 1.1. Close Close RCDT DISCH TO WHT, Q2G21 V009 Q2G21V009 (2-LWP-V-7 137). (2-LWP-V-7137).
- 2. Open RCDT PUMP DISCH TO RHT ISO, Q2E21V315 (2 CVC V 8551).
4.3.2.19 Verify closed the following:
** PRTN2 PRT N2 SUPPLY ISO Q2B31HV8047
** PRT N2 SUPPLY ISO Q2B31HV8033
** Nitrogen supply from bulk storage to PRT valve 2-GWD-V-7920 (Q2G22V215) (121' (121 PPR) 4.3.2.20 Align RCDT system as desired per FNP-2-S0P-50.0, FNP-2-SOP-50.0, LIQUID WASTE PROCESSING SYSTEM.
4.3.3 Gravity Draining PRT to NOTES: ** This method of draining the PRT should only be used if ReDT RCDT '~~)' pumps are inoperable. __
"'--------~-. -
The bottom of the PRT sparger is 12" 12 == ~500 500 gallons == ~5%. 5%. The
,. ç)j 12 perforated pipe that sits 12" sparger is a 12" 12 off the bottom of the PRT. The top of the sparger is at 24" PRT. 24 = ~1400
-4400 gallons =
= ~16%.
-46%.
doesnt have to be below the bottom of the sparger The level doesn't A2 1 b1/ > because the pipe is perforated on all sides, but it may be desirable. 4.3.3.1 Verify closed the following:
** (Q2B 1 3V064), 121'.
PRT vent to GDT 2-RC-V-8025 (Q2B13V064), 121.
** S/D Gas Decay Tanks 2-GWD-V-7935 PRT vent to SID (Q2G22V237), 83 (Q2G22V237),83' 4.3.3.2 Verify closed nitrogen/hydrogen supply to SID GDT's S/D GDTs 2-GWD-V-7849 isolation valve 2-GWD-V -7849 (Q2G22V040).
4.3.3.3 Open nitrogen supply from bulk storage to PRT valve 2-GWD-V-7920 (Q2G22V215). 4.3.3.4 Verify PRT regulator 2-RC-PCV-8034 (Q2B13V042) adjusted to 33 psig. 4.3.3.5 Open the following PRT N2 SUPPLY ISO valves (MCB):
- Q2B31HV8047
- Q2B31HV8033 Version 30.0
- 10. 007EK2.02 001/NEW/RO/C/A 001/N EW/RO/C/A 2.6/2.8/N/N/4/CVRIY 2.6/2.8/NIN/4/CVR/Y Unit 1 1 was at 100% power, and the following conditions occurred:
- FRP-S.1, Response To Nuclear Power Generation - ATWT, is in progress.
- MOB.
The Main Turbine was unable to be tripped from the MCB.
- A Safety Injection (SI) has NOT occurred.
- Tavg is 563°F.
Which one of the following describes the immediate effects if the Reactor Trip Breakers are opened locally at this time? A.. A.
- The Block of an Auto SI will be allowed.
- The Feed Water Reg Valves will trip closed.
B.* B.
- The Block of an Auto SI will be allowed.
- The Steam Flow high setpoint will be reset.
C.* C.
- The Main Turbine will trip.
- The Feed Water Reg Valves will trip closed.
D
- D'!"'* The Main Turbine will trip.
- The Steam Flow high setpoint will be reset.
Page: 26 of 200 12/1412009 1211412009
A - Incorrect. The A - Incorrect. The first first part part isis incorrect, incorrect, since since aa block block of of SI SI isis not not an an effect effect of of opening opening the RT the RT bkrsbkrs unless unless the the SI SI hashas already already initiated. initiated. Plausible, Plausible, sincesince ifif an SI had an SI had initiated this initiated this would would be correct, and be correct, and in in many many cases cases with with anan ATWT ATWT and and NONO Turbine Turbine trip an trip an SI SI occurs, occurs, but the stem but the stem states states that that an an SI SI has has NOT occurred. The second NOT occurred. The second part is part incorrect also, is incorrect since The also, since The FeedFeed waterwater Regulating Regulating ValvesValves are are onlyonly tripped tripped closed by closed opening RT by opening RT bkrs bkrs (P-4)(P-4) in in coincidence coincidence with with aa LowLow Tavg Tavg signal signal ofof 554°F. 554°F. Since Tavg is Since is still still above above 554°F,554°F, aa FWIS FWIS will not not occur occur immediately. immediately. Plausible, Plausible, since since most reactor on most reactor trips, a Low Low Tavg occurs occurs due to steam steam dumpdump operation very quickly after the Trip, but after but in this case in this case Tavg is high high due to to the ATWT. ATWT. BB - Incorrect.
- Incorrect. The first part is is incorrect incorrect (see A). The second part is is correct (see D).
C - Incorrect. The first part
- part is is correct (see (see D). The second part is incorrect (see (see A).
oD - Correct.
- P-4 will trip the Main Turbine regardless of any other plant condition or parameter(s), and this will occur immediately when the Reactor Trip breakers are open. The Steam Flow high setpoint will be reset by P-4 immediately when the Reactor trip breakers are open regardless of any other plant parameter (the actuation of the steam flow MSIV isolation signal requires a Low Low Tavg: P-12, but resetting the signal occurs regardless of Tavg on a reactor trip. These are two of the several functions of P-4. The MCB handswitch which trips the Turbine directly did not work in this scenario (it operates the 20 AST-2 relay). P-4 operates 2OAST-1 and 20ET the 20AST-1 2OET relays which open the interface valve and bleed EH fluid off of Throttle valves & Reheat Stop valves to trip main turbine per Figure 19 in the Student text for Main Turbine.
FNP-0-SOP-0.3, OPERA OPERATI TI ONS REFERENCE IINFORMATION, NFORMATI ON, APPENDIX APPENDI X G, OPERATIONAL PERMISSIVESAND CONTROL INTERLOCKS, Veron Version 39.0 Per mive Permissive P-4Reactor
- 1. P-4 Reactor Trip Interlock Source Reactor Trip and Bypass Breers Breakers Sdpd Setpoint nt Breakers Open Breers Coinddence&
Coincidence & LiitStatus LightStatus RTA& RTA & BYAOpenor BY A Open or RTB & BYB Open No No Light Function Prevents aa ri Reients rapi dd cool down of pri may ma-y system after aa reactor trip. tri p. 1.1. Trips Turbine Turbi ne
- 2. Trips F.W. R
- 2. TripsF.W. Valves on Reg Valves on Low Low Tavg
- 3. Seals in
- 3. Seals F.W. R in F.W. Valve Tri ps from S.I.
Reg ValveTripsfrom S.I. and and S/G S/G Hi Hi Hi Hi Leid Level 4.
- 4. Allows Allows S. I. ssignal S.I. gnal to to be ock after be bIblocked after S. niti ati on I. initiation S.I.
- 5. Resets Hi Stm
- 5. ResHi Stm Flow Flow Sstpoint Setpoint 6.
- 6. Arms Arms stnsteam dump dump system, enables p1 system, enables ait trip plant controller aid trip controller and disables loss of di sabl es loss of loaJ load controller.
control Ier. Page: Page: 27 27 d cJ 200 200 1211412009 12/1412009
007EK2.02 007EK2.02 007 Reactor 007 Reactor Trip Tn p EK2 KniIedge ci the EK2 KnOlNleclgeof the interrelations bween aa reactor interrdationsbetween reactor trip trip and and thefollcming: the fdlcming: (CFR (CFR 41.7 41.7 // 45.7) 45.7) EK2.02 Bree:i<ers, EK2.02 Bres, relaysrdisend dionnects................................... 2.62.8 aid dis::onnocts 2.6 2.8 Match justification: Match justification: To answer this To answer this question question the applicant must the applicant must know know thethe normal normal relationship between relationship between the P-4 interlock the P-4 interlock and and the the reactor reactor triptrip breakers breakers and the and the various various functions functions itit accomplishes accomplishes during during aa reactor reactor trip. trip. Objective: Objective: 1.
- 1. RECALL AND RECALL AND DESCRI DESCRIBE the operation BE the operation and function of and function of the following reactor thefollowing reactor trip trip permissives, control interlocks, signals, permissives, interlocks, and engineered safeguards actuation signals associated with the associated the Reactor Reactor Protection Protection System System (RPS)(RPS) and and Engineered Engineered Safeguards Safeguards Features (ESF)
Features (ESF) to iinclude setpoi nt, ncl ude setpoi nt, coi coi ncidence, dence, raterate functi ons (if functions (if any), any), reset reset features, features, and and the potential consequences for improper improper conditions to incl ude those items include items in the following followi ng tables (OPS-52201107): (OPS-52201 107):
- Table 1, Reactor Trip Signals
- Table T abl e 2, Engineered Engi neered Safeguards Features A Actuation Signals ctuati on Si gnal s
- Table 5, Permissives
- Table 6, Control interlocks
- 5. DEFINE DEFI NE AND EVALUATE the operational impl ications of abnormal plant or implications equipment conditions associated with the operation of the Reactor Protection System (RPS) components and equi pment to iinclude equipment following ncl ude the foil owi ng (OPS-522011 (OPS-52201 109). 09).
- Normal Control Methods
- Abnormal and Emergency Control Methods
- Automatic actuation including setpoint (( erample example SI, PhaseA, Phase B, MSLIAS, L asp, SG Ilevel)
LOSP, eve!)
- Actions A cti ons needed to mitigate mi ti gate the consequence of the abnormality abnormal ity Page: 28 Page: 28 ci of 200 200 1211412009 1211412009
13: 17 :30 04/03/09 13:17:30 FNP-0-SOP-0.3 FNP-O-SOP-O.3 APPENDIX APPENDIXG G APPENDIXG APPENDIX G OPERATIONAL PERMISSIVES AND CONTROL IIJTERLOCKS INTERLOCKS PERMISSIVES
~(J Coincidence & & Light Permissive Source Setpoint Status Function /
/
~
7_ Reactor Trip ReactorTrip I {I. Reactor* \\ and Bypass (1. P-4 Reactor Breakers Open RTA & BYA BYA Open or systemlafter a reactor trip. ~.f\ Prevents a rapid cooldown of primary systemkfter Trip Interlock \ Breakers RTB&BYB RTB & BYB Open ~ / / No Light DripsF.w.~valve~ ~ f}- c
/
- 3. Seals in F.W. Reg Valve Trips from S.I. S.1. and S/G Hi Hi Level I/ ;7
,3.. Allows S.1. signal to be blocke¥fter S.1. initi~
D C- S. 5jReFlrStTWF1OWetpoint
- 5. Resets Hi Stm FIDWSetp~
tnTDUiij
- 6. Arms steam steTh,enthles dump system, enables plant trip controller and disables loss of load controller.
- 2. P-6 I1R hR Power NIS 35 and 36 10 amps 10-10 10- 1/2>
112 > Setpoint Allows power escalation into the IR by turning j Both Train A && Escalation Lit> Setpoint B Source Range Block switches to Block. Above setpoint Permissive Permission to Block 1. Blocks SR Hi 00 Reactor Trip Source Range 2. Turns off Hi volt to SR Instr. Below setpoint Auto reinstates Hi volt to SR lnstr. Instr. Page 11 of7 of 7 Version 38.0
- 11. OOBAA2.26 008AA2.26 001/NEW/RO/C/A OO1INEWIROICIA 3.1/3.4/N/N/2ICVR/Y 3.113.4ININI2JCVR/Y Unit 1I has manually initiated a Safety Injection due to rapidly falling pressurizer pressure, and the following conditions occurred:
At 1000:
- Pressurizer level 35% and rising.
- RCS pressure 1700 psig and falling.
- PRT level is 73% and pressure is 5 psig and stable.
Tl-453, PORV downstream temperature, is 11
- TI-453, 117°F.
rF. Tl-453, Safety Valve downstream temperature, is 101°F.
- TI-453, Tl-453, Safety Valve downstream temperature, is 101°F.
- TI-453, 101 °F.
Tl-453, Safety Valve downstream temperature, is 102°F.
- TI-453,
- Containment Pressure 0.2 psig and slowly rising.
- R-2, 7, 11 and 12 are in alarm.
At 1015: EEP-1 .0, Loss of Reactor or Secondary Coolant, and the Transition is made to EEP-1.0, following conditions exist:
- Pressurizer level 99% and rising.
- RCS pressure 1400 psig and rising.
- PRT level is 73% and pressure is 5 psig and stable.
- Tl-453, PORV downstream temperature, is 138°F and rising.
TI-453,
- Tl-455, Safety Valve downstream temperature, is 125°F and rising.
TI-455,
- Tl-457, Safety Valve downstream temperature, is 125°F and rising.
TI-457,
- Tl-459, Safety Valve downstream temperature, is 126°F and rising.
TI-459,
- Containment Pressure 0.96 psig and rising.
- Containment sump level is rising slowly.
** 7, 11 and 12 are in alarm.
R-2, 7,11 Which one of the following states only potential sources of the RCS leak indicated by the given conditions? A. PORV leakby Safety valve leakby B. PORV leakby PRZR Level upper tap break C. PRZR Steam Space sample line break Safety valve leakby D~ PRZR Steam Space sample line break D PRZR Level upper tap break Page: 29 of cj 200 1211412009
A - Incorrect. Both are incorrect, since the PRT parameters are unchanged after the event has been in progress for 15 minutes. If the PORVs or the Safeties had leaked by, the PRT parameters would be higher than they initially would. Plausible, since the downstream temperatures are higher than they were, but only slightly due to elevated ctmt ambient temp in the vicinity of the steam space break. If either of the PORVs or Safeties were leaking by, the tailpiece temperatures would be much higher than this. B - Incorrect. The first part is incorrect (see A). The second part is correct (see D). C - Incorrect. The first part is correct (see D). The second part is incorrect (see A). oD - Correct.
- These are both correct, since per the indications (przr level high and pressure low and rising due to going solid on SI flow) there is a steam space break.
This choice has parts which are similar to incorrect choices which would be Przr Liquid Space sample and Przr Level lower tap, steam space sample and upper tap are both steam space penetrations. Ran a 400 gpm steam space break from 100% power (IC-73) on the simulator laptop to validate these numbers. D-175037 SH 2 Drawing PID: 0-175037 Previous NRC exam history if any: 008AA2.26 00BAA2.26 008 Pressurizer Vapor Space OOB Spae Accident AA2. Ability to determine and interpret the following fdlcming asasthey Preirizer Vapor SpaceAccident: they apply to the Pres;urizer Space Acddent: 43.5145.13) (CFR: 43.5/45.13) AA2.26 ProbIe PZR AA2.26 Probctlle PZR stean spa:e spae le9<:aJe Ie paths phs other th PORV or code safety oth- than .... 3.1 3.4 safety.... Match justification: This question presents a scenario with symptoms given of a steam space break. The indications have similarities to either a PORV or code safety leaking by OR another leakage path other than the PORV or code safeties, and some differences. In this case the applicant must correctly identify the potential sources of a a steam space break which for these symptoms must be other than aa PORV or aa code safety leaking by. Objective: I 1 LABEL AND ILLUSTRATE I LLUSTRATE the Pressurizer System flow paths pathsto to includethe components found on Figure 3, Pressurizer and Pressurizer Relief Tank (OPS-40301 E05). (OPS.40301 Page: 30 of 200 1211412009 1211412009
Date: 10/7/2009 Time: 01 :37:54 PM I C
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Date: 10/7/2009 Date: Time: 01 :38:43 :38:43 PM NO. NO. 1C1C TUB UPPER UPPER S
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TYPICAL FOR HV.3179C, HV3180C, HV3179C, HV318OC, HV3181C SEE MATCf-MATCI REV. NO. _ _ DATE ___ REV. NO. _ _ DATE ___ REV. NO. _ _ DATE ___ REV. NO. _ _ DATE ___ REV. NO. _ _ DATE _ __ REV. NO. 11 2 3 4 5
Title:
C:\Reference C:\Reference Disk\Exam Disk\Exam Reference Disk\Drawings\D1 Disk\Drawings\D17S009-0002.cal 75009-0002.cal
- 12. 008K3.01 008K3.O1 001/FNP OO1/FNP BANK/RO/C/A 3.4/3.5/N/N/3/CVRIY 3.413.5ININI3ICVRIY Unit 1 I is at 100% power and the following occurred:
- TK-144, L LTDN TON HX OUTLET TEMP controller, demand failed high.
Which one of the following describes the impact on the Letdown System Temperature, and the required action? A.. A. Higher Letdown temperature.
- Isolate Letdown and place Excess Letdown in service.
By* B* Higher Letdown temperature.
- Place TK-144 in manual and adjust flow.
C.* C.
- Lower Letdown temperature.
- Isolate Letdown and place Excess Letdown in service.
D.* D. Lower Letdown temperature.
- Place TK-144 in manual and adjust flow.
A - Incorrect. The first part is correct (see B). The second part is incorrect (see B). Plausible, since "IF IF letdown flow cannot be reduced [in manual control], THEN this would be correct, but it is not the first strategy prior to manual control of temperature. B - Correct. This controller demand goes up to raise temperature, which at 100% B - demand, sends a full closed signal to the CCW to the Letdown HX valve. The first alarm to come in would be: ARP-1.4,ARP-1 .4, DF1, DEl, L LTDN TON TO DEMIN DIVERTED TEMP HI 135°F],, and that ARP states to: [at 135°F] Take manual control of L "Take LTDN TON HX Outlet Temp TK-144 and attempt to increase CCW flow to the Letdown Heat Exchanger." Exchanger. Skill of the craft" The OPS "Skill craft policy also states that this is appropriate prior to attempting anything else. The ARP also states:
- 6. Adjust charging or letdown flow as required to reduce the letdown flow "6.
temperature. AND: 5. IF letdown temperature can NOT be reduced, THEN close LTDN ORIF ISO 45 (60) GPM Q1 QIE2IHV8I49A, E21 HV8149A, B, and C." C. C - Incorrect. This is incorrect since demand failing high causes the CCW to the Letdown HX valve to close (to raise Letdown Temperature). Plausible, since many valves open when demand goes to 100%. Also, if the valve did go open, it would cause boron absorption in the Mixed Bed demineralizer due to the cooler Letdown temperature. The second part is incorrect, but plausible. If the TK-144 valve could not be controlled in Manual, this would correct. oD - Incorrect.
- The first part is incorrect (see C). The second part is correct. Per OPS craft policy, placing a controller in MAN from AUTO when necessary to skill of the craft" "skill Page: 31 dof 200 1211412009
control parameters is always appropriate. One skill of the craft item, which may be performed as necessary without procedure guidance, is: "?? ?? Adjusting pots and controllers, including transfer between AUTO and MANUAL, to maintain parameters within log spec or procedural specs." specs. The other procedure guidance for a Low Letdown temperature is the requirement to maintain Reactor power <100% at all times. Ran on Simulator Laptop (IC-73) (10-73) AT 100%. DF1 DFI was the first alarm to come in (less than 30 secs). ARP-1.4, VERSION 48, DF1, DEl, LTDN TO DEMIN DIVERTED TEMP TEMPHI HI Tale manual control of LLTDN
- 3. Take TON HX Outlet Temp TK-144 and attempt to increase CCWCOW flow to the Letdown Heat Exchanger.
- 4. Adjust charging or letdown flow as required to reduce the letdown flow temperature.
- 5. IIF eleiated F cause for the eI THEN evated temperature has been corrected, TH EN refer to FNP-1-SOP-2.1, CHEMICAL AND VOLUME CONTROL SYSTEM PLANT STARTUP AND OPERATION to return TCV143 TCVI43 to DEMIN.
- 6. IF letdown temperature can NOT be reduced, THEN close L LTDN TON ORIF ISO 45 (60) GPM Q1 Q1E21HV8149A, E21 HV8149A, B, and C.
NOTE: T Transientsthat r ansi ents that will require boration or dilution should be avoided avded if letdONn Ietdcwn has been secured. remp is in progress, THEN place turbine load on HOLD
- 7. IF a ramp
- 8. Go to FNP-1-AOP-16.0, CVCS MALFUNCTION to address the loss of letdown flow.
ARP-l.4, VERSION ARP-1.4, VERSI ON 48, DF5,DF5,VCTVeT TTEMP HI, EM PHI,
- 4. Adjust charging or letdown flow as required to reduce the Letdown Flow Temperature.
- 5. Adjust LLTDN Temperature<< 111°F.
TON HX Outlet Temperature D175039 SH 2 DRAWING D175039SH of 200 Page: 32 eX 1211412009
Previous NRC exam history if any: 008K3.01 008 Component Cool Cooling Water System i ng Wo1er K3 Knowledge Kncmledge of the effect
&fect that a lcesar loor malfunction of of the CCWS will have have on the foilONing:
fdkming: K3.01 LoascooI&ibyCCWS K3.O1 Loa:Is cooled by CCWS ........................................... 3.4 3.5 Match justification: This question presents a specific type of malfunction of the CCW system (Failure of the CCW to the Letdown HX control valve controller). To answer this question correctly, knowledge of the effect of this malfunction of the CCW system on the Load (Letdown) cooled by CCW is required. The effect is that LETDOWN temperature goes up due to the controller failure causing ca using the CCW valve to the load (letdown) to go closed. The second part of the question and answers were added to gain 3 plausible but incorrect distractors. Objective:
- 7. DEFINE DEFI NE AND EVALUATE the opero1ional operational implico1ions implications of normal /I abnormal plant or associated with the safe opero1ion equipment conditions associo1ed operation of the CCW System components includethefollowing and equipment, to include the following (OPS-40204A07):
mcthods
** Normal control methods
** Abnormal and Emergency Control Methods Mathods
** Automatic A actuation utomo1i c actuo1i on iincluding satpoint(anpleSl, ncl udi ng setpoi PhaseA, nt (exampl e 81, Phase A, Phase B, High Radiation, LOSP)
Radio1ion,
- Protective isolo1ions isolations such as high flow, low pressure, low level including setpoint
- Protective interlocks
- Actions needed to mitigo1e mitigate the consequence of the abnormal abnormalityity Page: 33 of 200 1211412009 12/1412009
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13 :21 :40 04/03/09 13:21:40 04/03/09 FNP.. 1 -ARP- 1.4 FNP-I-ARP-l.4 LOCATION LOCATION DF1 DFI SETPOINT: SETPOINT: 135 D F 135°F jJ LTDNTO LTDNTO DEMIN DEMIN ORIGIN: 1-TY-143X ORIGIN: I-TY-143X Auxiliary Relay actuated by DIVERTED DIVERTED-Temperature Bistable (Ni Temperature E2 1 TB (NIE21 143) TB143) TEMP HI HI PROBABLE CAUSE 1.
- 1. Low or Loss of CCW Flow to the Letdown Heat Exchanger.
- 2. Letdown Flow greater than Charging Flow.
AUTOMATIC ACTION
- 1. Letdown High Temperature Divert Valve Qi E2 1 TCV QIE21 143 diverts Letdown TCV143 V CT. {CMT 0008644}
Flow to the VCT. 0008644 } OPERATOR OPERATOR ACTION
- 1. QIE21 TCV143 has diverted letdown flow to VCT to bypass Verify Q1E21TCV143 demins
- 2. Monitor charging and letdown flows and temperatures.
4 3. TI(144 and U"'~'HII-'L of LTDN HX Outlet Temp TK-144 Take manual control ofLTDN attempt to
//
(2 rd increase CCW flow to the Letdown Heat Exchanger. LA
- 4. Adjust charging or letdown flow as required to reduce the letdown flow temperature.
PTn,,,,," temperature can NOT be reduced, THEN close L j letdown LTDN TDN ORIF Q1E21HV8149A, B, and C. ISO 45 (60) GPM QIE21HV8149A, NOTE: Transients that will require boration or dilution should be avoided if letdown has been secured.
- 6. IF jf a ramp is in progress, THEN place turbine load on HOLD
- 7. Go to FNP-I-AOP-16.0, FNP-1-AOP..16.0, CVCS MALFUNCTION to address the loss of letdown flow.
References:
A-177100, Sh. 206; D-175039, Sh.2; D-I77091; D-177375; U-175997; PLS D-177091; D-I77375; Document Page 11 of 11 Version 45.0
- 13. 008K4.09 001/MOD/RO/C/A OO1/MOD/RO/C/A - 2.7/2.91IN/2JCVRIY
- 2.712.91/N/2ICVRIY Unit 1 1 was operating at 100% power, and the following conditions occurred:
At 1000:
- A Train is the "On On Service" Service train.
- 1 1B B CCW COW pump is running and supplying loads in the on-service train.
- 1A IA CCW COW pump is running to support charging pump operations.
- 1 IC COW pump is aligned and OPERABLE.
C CCW At 1005:
- A Safety Injection and LOSP occurred simultaneously.
Which one of the following combinations of CCW COW pumps will be running following the operation of the ESF sequencers, with no operator actions? A 1A A'! IA and 1IC COW pumps ONLY. C CCW B. 1 lB Bandand 11000Wpump5ONLY. C CCW pumps ONLY. C. 1A IA and 1lB COW pumps ONLY. B CCW D. 1A IA and 1BlB and 1C10 CCW COW pumps. COW pump is running on A train but will trip on Load shed. Then, the A - Correct. 11 B CCW auto start circuitry starts up the non-swing, train related 1A IA & 1C 10 pumps per CCWCOW FSD Appendix A step 3.1.2.2, LOSP. B - Incorrect. 1
- COW pump is running on A train and if there was no LOSP signal, the I B CCW SI auto start circuitry would leave the 1 I B running and not start the pump on the same train. The opposite train pump is 1A, 10. This is plausible since this IA, and not 1C.
is the opposite train and CCW COW has backward logic. logic, normally 1IA A pump would be assigned to A train, but CCW COW is an exception to this general rule. O - Incorrect. Plausible, since this would be correct with an SI and no LOSP. C - D Incorrect. Plausible, since the SW pumps would have all pumps including the swing 0- - running in the event of an SI if the swing pump was running to start with. For this COW system alignment: 1I B CCW COW pump is running on A train and 1A B CCW IA CCW COW pump is running on B train to start with. For an SI alone, 1 I Band B and 1A would be left running. For an LOSP, 1A and 1 10 C would be started. However, the LOSP sequencer secures 1 B prior to starting 110. 1 C. FNP-1-SOP-23.0, Version 83.0 COW is normally lined up so that 3.2 CCW
- One CCW COW pump and one CCW COW heat exchanger is in operation supplying the on-service train exchangers and the secondary heat exchangers.
- The remaining pump and heat exchanger are valved into a closed loop with the redundant safety off-servi ce trai safy train. The off-service 14 supplying the trainn is normally in operation in modes 1-4 operati oDerating ng pump, with charging ng chargi non-operating wi th the non-operati ng SFP HX flowpath aIaligned COW to the i gned and CCW 201)
Page: 34 of 200 1211412009
RHR HX isolated. (Reference RER 1080944901) A-I 81 000 FSD A-181 Appendix A 3.1.1.3 SIAS In the event of a SIAS with offsite power available, the on-service pump shall continue to operate, and the off-service (redundant) train-dedicated pump shall automatically start. Swing pump B shall continue to provide backup in the event of a fault trip of the dedicated pump in the train to which swing pump B is aligned, as above (Reference 6.7.11). 3.1.2 Swing Pump B on-Service, Dedicated Pump Available (Possible Alternate Alignment To Equalize Pump Wear) 3.1.2.1 On-Service Pump Trips
- a. During normal plant operation, if on-service pump B B trips due to a fault, the dedicated pump in the on-service (operational) train shall automatically start and supply component cooling water to the on-service component cooling heat exchanger (Reference 6.7.11).
3.1.2.2 LOSP In the event of a LOSP with or without a SIAS, on-service pump B shall be shed and the two train dedicated pumps, C and A, shall be automatically sequenced onto the diesel generators (Reference 6.7.11). Swing pump B B shall provide backup in the event of a fault trip of the dedicated pump in the train to which swing pump B is aligned (Reference 6.7.11). 3.1.2.3 SIAS In the event of a SIAS with offsite power available, on-service pump B B shall continue to operate and the off-service train-dedicated pump shall automatically start. The dedicated pump in the on-service (operational) train shall continue to provide backup in the event of a fault trip of swing pump B B (Reference 6.7.11). Page: 35 of 200 1211412009
Previous NRC Previous NRC exam historyhistory ifif any: 008 K4 .09 008K4.09 008 Component Component CoolCooling i ng Water System System K4 KncvIedge of CCWS K4 Knowledge of CCWS design de9gn feature(s) feature( and/or andlor interlock(s) interIock( which which provide pride for for the the follONing: fdlcwing: (CFR: 41.7) K4.09 The "staldby" K4.09 adby feEturefor fure for theCCW pumps ............................ 2.72.9 2.7 2.9 Match justification: Objective:
- 2. RELATE AND IIDENTIFY DENTI FY the operational characteristics including design features, capacities and protective interlocks for the components associated with the CCW System, to include the components found on Figure 2, Component Cooling Water System, Figure 3, Secondary Heat Exchanger Header, and Figure 5, RCP-CCW & 8W SW System (OPS-40204A02)..
(OPS-40204A02)
- 7. DEFI NE AND EVALUATE the operational implications of normal /I abnormal plant or DEFINE equipment conditions associated with the safe operation of the CCW System components and equi equipment, pment, to iinclude ncl ude the foilfollowing (OPS-40204A07):
owi ng (OPS-40204A 07):
- Normal control methods
- Abnormal and Emergency Control Methods
- A Automatic actuation utomati c actuati on iincluding setpoint ncl udi ng setpoi (example nt (exampl SI, Phase e 81, PhaseA,A, Phase B, High Radiation, LOSP)
- Protective isolations such as high flow, low pressure, low level l,el including setpoint
- Protective interlocks
- Actions A cti ons needed to mi mitigate abnormality ti gate the consequence of the abnormal ity Page: 36 of Page: 36 of 200 200 1211412009 1211412009
(A) (A) (A) (A) (A)
*l I1 3.1 3.0 2.0 1.0 0
I1 3.1.1 H 3.1.1.2 3.1.1.1 0 7597\A-181000.SD II WrrH :- c C) b a.Pi 0 Cl) 0 C) b[rtfl li-i dCDr rti- ti H Dh d0 CD Cr I:- C) 013 CDCDOfl I >13CD1313 Id PJLQP) Cf CD E W 1i-3 CtCDI.QCDCD I RI
-H-ft I-I. frwCi0PiFdrt l Fj I FNP UNITS 1 & 2 CD C) -3 -3 Ii CDH-ShHCD13d 10 ixjH-Mi 10 dHHAdO F H-C)CD ItJ P1C)CnhP)0 IC,)
CCW pumps C and A shall be train dedicated and aligned
\t CD O PJd -1 c C)C)d PURPOSE C)1C) CD Ct CDP) QCQct id HrtOCDH liii
~ The following provides the functional requirements for OID(DCD H C)QC) C)rtrtS i CD I-I. txI H-H13 3 HPFd IC) H- H-H 1 CD dOdCD JCDOtIH- (i txj <HP)CD En I3 IQrtrtCDOO rtH-CD Cr -IH-- 9ifr 3 CD CD< Cl) 11-4 3 13CD 3 HEC) CD biO Ct 13 P1rtCQ II I I H<C) CD><rtEnH HbPlrtk< 10 INTRODUCTION 3\\3Pic-t t13 trtcI CD I-I CD Once the (swing) pump B becomes the on-service F-.
to the 4 kV buses F (train A) and G (train B) , c1crrr H-Ct13CD CDOH-H-13 0 CtP1CDH-0. bPi 13 CflQ During normal plant operation, if the on service (DIOCDO 3 CD h QQ Ii. EI)HCD13 rtCD H-C) IC) U-h rttrt . rtP1Cnh CD C) ii CD W Cn 13C1 d <113 IC) I-One Train-Dedicated Pump On-Service, Swinq PumP B in rtKt CD rtOCDObCfl .13H Lii dCD WH-C)CD * çiH] i- H-13rCtdO There are no new Oj13dPi Mi CDH1Pi Cr O<rtflCfl I the various CCW pump alignments. 13 -3 CD CDP1PIbCDH- CD k<H Cr CD D-13 iCD13c<1 Iti CCW PUMP ALIGNMENTS FdI\$ I- t-<CDPi13 HiO CD 0 I:)) Cl) CD H- H CD respectively. i En CCW motor pump B, when available, shall CD H- H- CI hI9dO CD CDH Q 013W d I-13<1C)I-i 13 I pump, the breaker of the pump with the fault is On-Service Pump Trips O-Cfl[f) 13 OCflfrCD t- I rtrtH D 0
- rtCDO 0P) CD CD ID C) CD13PJh CDWPIOPI rJ 00rtrtH 13FPICn I H-LW) CD 3 PJC) ti 13CDh 13 CD Lxi I-iMi0frH rCtddH I pump trips due to a fault, the standby pump shall A-1 H I\D1n CrCrHQ C)frrtP1rt rtH-CD I-P1 W0dH-rt It 313Cflrt WCDC13 H Cl)
CDIF-1( H- 3HCDM CD1 0 H-b IH Fj H-P1MiO d 1 1>< (r pirt OCD w CD rt 3 CD MiCD Iii CD 10 requirements in this Appendix in addition to those in Cli be aligned to either of the vital 4 kV buses, F (train bQ1313 J Mi 3 Ct I:- 00CDE APPENDIX A CCW SYSTEM ALL THREE CCW PUMPS OPERABLE P)CD CDPCD Cl) Lxi rt* 13 Srt 00H-CDH, I C) 0 Q PlrtCD - CD P-<d-- Lii required to be racked out immediately, or the (ii 130 rt b C)WctP) -C i H-H- I Cr13 I j F3 dCD H0ç rt H--13ctWPIH- rtrtH-CflC) I-3 Deleted (Reference 6.7.057) CCW PUMP ALIGNMENTS dpi) CD CD H- C) H ti H f 4 1313QPi 1313 1:j-1:5*13* Ct IC)
\I (EnH- Oh drt0 C) CD H P OH- H automatically start and supply component cooling A) or G (Train B), corresponding to whichever train the HI0\13rt Cl) hSZjS 0H1313 CD Mi Ct COP) 0 Standby (Aligned to the On-Service Train), One Train-Ii13 Id0CDCD CD 3-<CD- t013QCD Ixj 2-13 the FSD text. The purpose of this Appendix is to Hcni-0 dEnPCn CD -3 0 CD rtj* H CDCDH-H- 3C)EnH- 1 d13H- OdH-CDH 131 lock-out relay shall not be reset by the operator, hr-tçvrtct rtortM, i13 FtC) drtF CD b CDP) )13 Pl F-I H- P1P1 CDH-CD Cl) P1 HirtbCD CD C)d13Cr dC)O<1PICt B pump has been valved into. The B pump shall be 1130 CD CD CDH-<Hrt H- JP1H-CD H-13PiQ
< H-Pi Cl) h1 0 < J0 013 tYCD C)CnCDCflH-13p. OH- t-CD H CD13rt- CD13 HCD< 1313<1CDH 13>rtCDH-water to the on-service component cooling heat 13( 13 j I H-13 0 0 PiCDEnPiPi 0 -1 extract the requirements from the FSD which pertain to Ct Cl) C) 0 CD 0 Hi Cfl 3 rrtd 3 Ii H-H b13 AtiH 13CD d- CDMi 1P)CD CD rtH HCl CDWrt0 in order to allow the proper operation of swing oc:pi 0 ow CrXICD 11 13 CD physically valved into the train from which it is set Mi rtrt 0WdrrH 130dCD -3 H-bh P1 CrCr01313 d
H 1 -13 CD13rtH- 00 I- -C COOP) H PJOCflCDCt i p1 \f)CD JH-CD C) PIHCI)< H-CtCl) H- H- CDCI) H CD Cli 3 CD H- CD Ct H- 13 H fJ Cfl 1313 (1) 13 CrFrt HCOrt CO 13piC) CD PIP) 13 H- Mi H Dedicated Pump Off-Service (Nor.mal System Alignment)
- H 0 CQHCD Cr CtH-H CD Cr 3 0 0 exchanger (Reference 6.7.11). pump alignments.
H-\<0 II H 1313H Cl 0 l- 0 pump B in the event of an LOSP (Reference 6.7.11) to receive its power. A-181000 01 0 Rev. 5 d V CD
C-K c : i~
" W W F3j I- I-ur~ 3.1. 2
\: 3.1.2.3 3.1.2.2 3.1.2.1 3.1.1.3
)
7597\A-181000,SD
..fo£A 7Th4
+/- c- OC-t-4thtOO H c,w OH b. a. P)P)OCtH WO OrlIfUii HP CDCD 0 t HCDI-CiF-hi H Hi
-Dd)IItI H-H- CI I 0f H-O.OrtMiCD SlAS H
. Wrt LOSP CF Cl)ct CDh(Ifltflrt CI:iHi ti W COH-OI Cf FNP UNITS 1 & 2 HDfrc11WIICD (1c-t(.Q HCDO CDi - CD U iOH-WOi 3 (lIn the event of a SIAS with offsite power available, In the HJ*DljCD frCD Li CDP) dO 1 CDP)OP)CDiCD P)CD rjct In the event of a SIAS with offsite power available, CDc I-j[ C)çt CD PCi iCi < b çc-rCDrt H HOct CD1-L H-<CJ)CD event of a LOSP with or without a SIAS,
- iF-- jJH-CD CDCD H-CD f- CD H-Cl rtO-jH-V0 NH-S CD 3CDO dH- 0 CD Clb0H-CD< I-OCD iJ10.ACDCD CDCD rF PiiO 3 CD P1 P)P)OCD OPJCD\CD/l i 3 CD3 CVit1iI.O WdOHCD<i CDH hhH \Jdrt Qj CF Cl COctct I t IlW CiH H-Ct 1O Swing pump B shall provide backup in the event of a During norm~lant operation, if on-service ctc Hi CDOH-Cnrti rt Ci-<---C) P1
ç-)
on-service urn B shall continue to .2J2erate_and the P) CDfrU$O H- Cl) F OCDfrO (DO tYdtL1 frCDO Ct the on-service pump shall continue to operate, and the rt wrt Cl) if rtP)FH-F ti I-(11 0 CflCD I-h dCD
\ P)Pdfrh I-h1HH-l JP) rt
-- CD H<d I H-H-rtCld CDCl P)P)OtE1P) P)CDH wP) I-h CDHHU)P1 P)C11 CDiiP)CiCiP) 1h t CD CD k<IC) H I-( rt(D CDd
-9n-service 12ump B sh~ be shed and ~ two train-Cl)Ii I-L MiO CD F- CD CFCFCFCldCc CD CDiC1) Ci < 3P) H CDS On-Service Pump Trips 3 I- I rtCDH -
0 Piw m CDCDCD iCn pump B trips yet~9 a faulh tjie__dedicated 12~n HH (12
-service train-deaicated pum12 shall a~tomatically tJ HHCI)
P)Fd0H. CDH-O Hd C) CDWP)O P1 I-0 I-h CFiCI) off-service (redundant) train-dedicated pump shall C) A-2 p,c)1 Cl) CD iCDFdrti I--(D CDCFCDCIDPi CDH Swing Pump B on-Service, Dedicated Pump Available M ç H-HiP) P)H- CD O1-CFCDOCF H 3
\--- L FHctOH (-t F- J CD- h CDP1CDH-ctHH-of the dedicated pump in the train to which swing pump
- dCDOrt JCDCD
- HPIPIPIO iH-3h CF CF fault trip of the dedicated pump in the train to which 00 C1 (D 3 (30ICt d (D(D oictCOploi CCW SYSTEM sN_ HiOH CF Mi b CD * *CD H-F-hCD CD H-O *CD
( CDFZiP) ipi 0
. tare.
dH-O 2 OP1hL rrF- CFOO H dedicated pumps, C and A, shall be automatically
\
Tne(credlcated pump in the on-service~9--~ ~he on-serVlce (operational) train shall -- wrt iHi CDH U] 0Cl)iCiP) 0 J0HiCi I ctH, Hrt automatically start. Swing pump B shall continue to DS 3 HOCiPJHCF t-3Ø) SlH-H, Deleted (Reference 6.7.057) H-i CDCI) 3d Pi HdHrtH- Oct dCDiCl) P1 C-iiCD H- C) d rt d- 0 CD iH- H 0CD cnrtrt CDH- OCD . 0 3 LXJP HH-M,WH-CDrt 3 Oi-OCD iH iI-<rtCF- HOP) C) CD drtiPl o 3 CflPi CDj F-LY 3EflPiCt CF OIHjOd *ct iI0P1CDH- HCF0d 0
*~tional) train shall continue to provide backup in CRHPdQ -JYCt rt0H- Mi Cl)CtP1CD 0 provide backup in the event of a fault trip of the
>* ddCD (D *CDi iCl I-- HClO sequenced onto the diesel enerators (Referenoe~ }UEomatically staEt and suppI~on~,t_~~__
Nç FdCD swing pump B is aligned (Reference 6.7.11). fril-CD H CD CD Od CD0 N H-rtH cWO<Pifr-L Hrt F-hP) 00nCl3 ih-L dCDF-L H-
- <H-ctct IICD CDCF 0 i 3H- I CO H-C) Ci I-L C) r- H-OOCDPJ PJ HICDPJOCI) t0P)P) dP-(D H-CD CDC) H-!iHPICD - d JZ3CF CDCD P)P)Pi iIrFHrrF- 0rt CDPI Cl) the event of a fault trip of swing pump B (Reference Hi rtiH- CF OH CD< F-hH-Cn- H dedicated pump in the train to which swing pump B is N CO CDb /H-H CF CDH C) ClH- d H H water to the on-serVlce component cooli~t P) 10 Pi 00 i0 C) I cFCiPIP1P1 3 f CDOP)ctb NI -h CD0 CD mm WiCDHib CO 6.7.11) .
;ii i1HiH P1H Ci CD HClH H CD Oi(HCDCD CFH- II H- CF CD (Possible Alternate Alignment To Egualize Pump Wear) m H 3 z5 Cl) 0 Ct Ci H 0 CO 0 aligned, as above (Reference 6.7.11).
B is aligned (Reference 6.7.11). 0 SlAS H H- CD A-181000 0 Rev. 5 6.7.11). U] I2:Kchanger (Reference 6.7.11)ji
FNP Units 11 & 2 SERVICE WATER SYSTEM A-181001 are required between pumps 1I Band B and 1IC C and between pumps 11 C and 1ID D to separate trains for fire protection purposes. (Reference 6.1.010) 3.1.4.4 Each pair of train oriented Service Water pumps along with the swing SW pump shall be provided with a minimum flow bypass valve (See Section 3.4 for the required flow rates) to recirculate service water to the service water intake structure wet pit. (References 6.4.014 and 6.4.018) 3.1.4.5 Each Service Water pump motor shall be equipped with bearing temperature monitoring devices. (Reference 6.5.003) 3.1.5 I & C Requirements 3.1.5.1 The Service Water pumps shall be automatically started by a signal from the LOSP or ESS sequencer. The Service Water swing pump shall be automatically startedb;ra-
"'"Slgi1aIfrom the LOSPOr~-SS sequencer-when in service _
.-replacing one of the train oriented pumps. (References ---- -t-_**-~ t LL.Y1,\
;f f-ce{e -6.7.039 and 6.1.00J3) v\ / (67039and6
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3.1.5.2 Key interlocking of power supply breakers, bteakers, disconnect , - switches, and SW header cross-connect valves shall be f/ c.t. A f\ used to ensure alignment of the Service Water swing pump to one train only. (Reference 6.1.006) 3.1.5.3 Annunciation shall be provided in the Control Room to alert the operator when a Service Water pump breaker trips. 6.4.104 (References 6.4.1 04 through 6.4.115, 6.1.007) 3.1.5.4 Monitor lights shall be provided in the control room to allow quick verification of the status of Service Water Pumps A, B, D, and E following a safety injection signal. (Reference 6.7.124) 3.1.5.5 The Service Water (SW) System shall have redundant level instrumentation to monitor and control the Storage Pond/Service Water Pump Wet Pit level. (Reference 6.4.075) Service Train TPNS Nos. SW Pump Wet Pit A N1(2)P25L14066A Nl(2)P25LI4066A SW Pump Wet Pit B N1(2)P25L14066B Nl(2)P25LI4066B 3-5 Rev. 44 Rev.44 I
1.
- 1. CCW-40204A07 013IHLT/IMEM CCW-40204A07 013/HLT//MEM 2.712.9/00SK4.09////00SA3.0S 2.7/2.9/008K4.09////008A3.08 008K4.09 008K4.09 Unit 1I is Unit is operating operating atat 100%
100% power power with the following conditions: conditions:
** "A" A Train is is the "On On Service" Service train.
- 118B CCW pump is running and supplying loads in the on-service train.
- 1AIA CCW pump is running to support charging pump operations.
- 110 COW pump is aligned and OPERABLE.
C CCW A Safety Injection occurs at this time. Which one of the following combinations of CCW COW pumps will be running following the operation of the ESF sequencers? (Assume no operator action is taken) A. 1AIA and 110 COW pumps ONLY C CCW B. 1lBBandand 11000W C CCW pumps ONLY C 1A and 1lB C'!' COW pumps ONLY B CCW D. 1AlAand and 1B 18 and 1C 10 CCW COW pumps Page: Page: 11 of of33 10/26/2009 10/26/2009
008 K4. 09 008K4.09 008 Component Cooling Water System (CCWS) K4 Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) K4.09 The "standby" standby feature for the CCW pumps ....... 2.72.9 2.7 2.9 A. Incorrect, 11 B CCW pump is running on A train therefore, 1IC C CCW pump will not start. This is plausible since on a load shed and LOSP sequencer operation this would occur. B. Incorrect, 1I B CCW pump is running on A train therefore, 1IC C CCW pump will not IA pump will start on B train. This is plausible since this is the opposite train and start, 1A CCW has backward logic. C. Correct, 1 1 B CCW pump is running on A train therefore, 11 C CCW pump will not start. IA CCW pump is running and does not recieve trip signal and it remains running. 1A D. Incorrect, 1 lBB CCW pump is running on A train and 1A IA CCW pump is running and would recieve a start signal from B train sequencer. The 1 ICC CCW pump will not start since the 1 lBB CCW pump is running. This is plausible since the SW pumps would have all pumps running in this situation. Lesson plan ops-52102G The three CCW pumps (Figures 6 & 7) can be operated from the MCB or locally at the HSP by a three-position handswitch (STOP/AUTO/START, spring return to AUTO). A two-position selector switch (LOCAL/REMOTE) at the HSPs determines which station has control of the pumps. The dedicated S-signal or a loss of offsite power (LOSP) pumps (A and C) will automatically start on receiving an "S"-signal signal, provided that the local remote selector switch is in the REMOTE position and the MCB handswitch is in the AUTO position. The swing pump (B) acts as a backup for the dedicated pumps by being mechanically and electrically aligned to the same train as one or the other of the dedicated pumps. The swing pump will automatically start when: (1) the dedicated pump it is backing up trips on overload; (2) the selector switch is in the REMOTE position; (3) the MCB hand switch is in the AUTO position. The swing pump also receives start signals from the safety injection (SI) sequencer and the LOSP sequencer. However, it will only start if the selector switch is in REMOTE, the MCB switch is in AUTO, the dedicated pump A or C C (depending on which train it is lined up to) has tripped on overload, or its supply breaker has been racked out. FSD A-181000 A-i 81000 3.1.5.2 During normal plant operation, with all pumps operational, if the operating pump power supply breaker trips, the standby pump shall automatically start and supply CCW to the CCW heat exchanger in operation. The breaker of the pump with the fault shall be racked out immediately, or the lockout relay shall not be reset by the operator in order to allow the proper operation of the standby pump in the event of a loss of offsite power (LOSP). The CCW pump overload trip shall be alarmed in the MCR to alert the operator 6.1.01, 6.4.15, 6.4.16, 6.4.17). (References 6.1.01,6.4.15,6.4.16,6.4.17). 2 of 3 Page: 20f3 10/26/2009
2008 NRC 2008 NRC examexam Technical
Reference:
Technical
Reference:
FSD FSD A-181000 A-i 81000 Learning Objective: Learning Objective: 40204A07 40204A07 List the List the automatic automatic actions actions associated associated with the the Component Component Cooling Cooling Water System System components components and equipment and equipment during during normal normal and and abnormal abnormal operations operations including including (OPS40204A07): (0PS40204A07):
- Normal control methods
- Automatic actuation including setpoint (example SI, Phase A, Phase B, High Radiation, LOSP)
- Protective isolations such as high flow, low pressure, low level including setpoint
- Protective interlocks 52i02G02 also 52102G02 Comments: This meets the KA since it tests the standby feature of the standby pump for the train it is aligned to. This is the standby feature for the main pump (ie., 1C IC or 1A IA CCW pump).
Our SW pumps do not have a feature where the standby pump looks to see if the other pump is running before starting or not starting the other pump in that train for an SI signal. In that case there would be 5 SW pumps running. For CCW, if the swing pump is running, then the other pump in that train will not start on the SI signal. All distracters are plausible since our trains are not set up in a logical way and C CCW pump is A train and A CCW pump is B Train. Most other components are configured correctly and differently. Had to change the stem to take into account the new CCW and charging pump line up. CCW-52i02G02 05 FNP BANK: CCW-52102G02 Page: Page: 33 of of33 10/2612009 10/26/2009
- 14. 009EG2.1.23 001/NEW/RO/C/A 001/NEW/RD/C/A 4.3/4.4/N/N/4/CVRNER 4.3/4.4/N/N/4/CVR/VER 5 EDITORIAL Unit 1 1 has experienced a Small Break LOCA, and the following conditions occurred:
At 1000:
- ESP-i .2, Post LOCA Cooldown and Depressurization, is in progress.
ESP-1.2,
- Normal Charging has been established.
At 1010:
- CTMT Pressure is 6 psig and rising.
- Subcooling is 24°F and decreasing.
- PRZR Level is 28% and decreasing.
Which one of the following is the required action lAW ESP-1.2? ES P-i .2? A. FK-122 must be adjusted to raise Przr level. B. Place the SI ACTUATION switch to ACTUATE. C. FK-122 must be adjusted to maintain current Przr level. D HHSI flow must be established and additional CHG PUMPs started. Dy Subcoohng A - Incorrect. The Fold Out Page requires reinitiating HHSI flow due to both Subcooling and PRZR Level being too low with adverse numbers "16°F{45°F} 16°F{45°F} & 13%{43%}. Plausible, since this would be correct if the procedure step for maintaining pressurizer level was initiated with adverse numbers and 50% PRZR level was required, while forgetting about the FOP requirement to re-establish HHSI flow. B - Incorrect. (see D). Plausible, since HHSI flow is needed, and it may seem more B - convenient to turn the SI switch vice going to the attachment to manipulate each component, but the FOP requires the attachment be used. This ensures that only the SI equipment and Phase A components desired are manipulated. C - Incorrect. (see A). Plausible, since if Adverse numbers were not taken into ESP-1 .2. account, this would be correct per step 20.2.1 of ESP-1.2. D - Correct. The FOP requires this for these Subcooling and Przr level values. RO D - knowledge requires knowing the FOP requirements. ESP-1.2, Revision 23 Page: 37 of 200 1211412009
Previous NRC exam history Previous NRC history ifif any: 009EG2.1 .23 009EG2.1.23 009 Small Break LOCA 2.1.23 Ability to 2.1.23 perform specific to perform spedfic system syem and integrated plant and integrated plant procedures procedures during all all of plant operation. (CFR: 41.10/43.5/45.2/45.6) RO 4.3 SRO 4.4 modes c:I Match justification: This question requires knowledge of specific system and integrated (ESP-1 .2 Fold out Page) during a SBLOCA to plant procedures (ESP-1.2 answer correctly. Objective:
- 6. EVALUATE plant conditions and DETERM I NE if any system components need to be operated while performing ESP-1.2, Post LOCA Cooldown and Depressurization.
(OPS-52531 F06) of 200 Page: 38 d 1211412009 12/1412009
FNP-1-ESP-l.2 FNP-I-ESP-l.2 POST LOCA COOLDOWN AND DEPRESSURIZATION Revision 23 Step Step Action/Expected Response Response NOT Obtained n I I I NOTE: During a LOCA, a full or rlslng rising pressurizer level may indicate a steam space LOCA exists. In that event, step 20.2, RNO provides gUidance guidance if the RCS must be operated water solid, and charging used to to maintain subcooling instead of pressurizer level. 7 ""':'--------~ 20.2 [CA] Maintain pressurize~ [CAl pressurizeN 20.2 IF solid plant operation level greater than 25%{50%}. 25%t5O%}. required, THEN perform the following. 20.2.1 IF charging flow path aligned, a) Maintain SUBCOOLED MARGIN THEN control charging flow. I+/-N MONITOR indication greater than 26°F{55°F}. 26°F{55°F) CHG FLOW [1 FK 122 adjusted b) Control charging flow to stabilize subcooling at 1\'\ ) existing value. CHG FLOW {C( ((11 FK 122 adjusted CAUTION: To prevent potential seal damage, neither seal injection nor CCW cooling should be restored to a RCP which has lost both seal injection and CCW cooling. _21 21 [CA] Check if RCP(s) should be [CAl reconfigured to optimize RCS flow and pressurizer spray performance. 21.1 Check RCP lB IB - STOPPED.
- 21.1 Perform the following.
21.1.1 Verify RCPs 1A lA A]AND lC 1C -- STOPPED. RCP [] [l 1A lA [] [l 1C lC 21.1.2 Proceed to step 22. 22. OBSERVE OBSERVE CAUTION AND NOTE PRIOR TO 22. STEP 22. Step 21 continued on next page. ___Page Completed Page 25 of 50
r"""? ....--------.-----------'"""'rh.:t,,:)-i:rir,,-- . , 7 - - - - - - - - - - . , . - - - - - - - - - - - . FObEiOU1~ FOLDOAGE :eAGE FNP-1-ESP-1.2 POST LOCA COOLDOWN AND DEPRESSURIZATION DEPRESSTJRIZATION Revision 23 Step Action/Expected Response Response NOT Obtained n II II II __ t L;;,C'(C-e c~
/'_
I 1 Monitor SI reinitiation criteria following HHSI ifilSI isolation. '~ 16°FQ~ ~'1 ;sta~li~HHsI--flo;_-~~~' i_ 1.1 1.1 Greater than l6°F CE~~and subcooled in CET,C - 1""..J./ e and PRZR
/ -
Establish HHSI flow, and additional CHG PUMPs PUMPS as level above 13~ 43%1) l30/J. -. required using ATTACHMENT 5. 5, ' 2 PRE-ESTABLISHING RE-ESTABLISHING HHSI FLOW. \ I ) 2 Monitor FNP-I-EEP-2 FNP-1-EEP-2 and FNP-I-EEP-3 FNP-1-EEP-3 branch criter1a.criteria. 2.1 No SG pressure falling in an 2.1 IF affected SG NOT previously j uncontrolled manner or less isolated. isolated, than 50 psig. THEN go to FNP-l-EEP-2. FNP-1-EEP-2. 2.2 No high secondary radiation or 2.2 Establish HHSI flow. flow, and start SG level rising uncontrolled, uncontrolled. additional CHG PUMPs PUMPS as required using ATTACHMENT 5. 5, RE-ESTABLISHING HHSI FLOW THEN go to FNP-1-EEP-3. 33 Monitor switchover criteria. 3.1 RWST level greater than 3.1 Go to FNP-1-ESP-1.3. FNP-l-ESP-1.3. 12.5 ft. 3.2 CST level greater than 5.3 ft. 3.2 Align AFW pumps suction to SW using FNP-1-S0P-22.0. FNP-1-SOP-22.O. 44 Monitor charging miniflow criteria (during SIlo SI). 4.1 RCS pressure less than 4.1 Verify miniflow valves open. 1900 psig. 4.2 RCS pressure greater than 4.2 Verify miniflow valves closed. l300 1300 psig. 55 Monitor adverse containment contaimnent criteria. 5.1 CTMT pressure less than 44 psig 5.1 Utilize bracketed adverse CTMT and radiation less than condition numbers. l05 R/hr. 10
- 15. OIOK103
- 15. 001/NEW/RO/C/A - 3.613.7ININI4ICVRIY 010K1.03 OO1/NEW/RO/C/A - 3.6/3.7/N/N/4/CVRN Unit I1 was Unit was atat 28%
28% power and the power and following conditions the following conditions occurred: occurred:
** All All PRZR PRZR Backup Heaters are Backup Heaters are in in AUTO.
** A CVCS CVCS Malfunction has occurred.
Malfunction has occurred.
** FK-122, FK-122, CHGCHG FLOW, FLOW, has been placed has been placed in in manual.
manual.
** PRZR level is at 36% and rising.
Which one of the following describes the operation of the Backup Heaters and the Spray valves with no operator actions? All PRZR Backup Heaters will be (1) and BOTH PRZR Spray Valves will be _---1.!:(2:..t..)(2) _ _ (1) Backup Heaters (2) Spray Valves A'! A ON Opening B. ON Closing C. OFF Opening D. OFF Closing Page: 39 of 200 1211412009 1211412009
A - Correct. First part: The CVCS Malfunction caused an insurge, which caused the PRZR level in increase. PRZR level program is 21.4-50.2% level from 547-573°F Tavg, so at 28% power, program level is 29.5% przr level. There is a 6.5% level deviation (>5%). Przr level >5% above the program level turns on all BU heaters which cause the pressure to go up more (after the water reaches the new higher saturation temperature). Second part: The insurge caused the PRZR steam space to be compressed, which causes the Pressure go up. The pressure controller opens both spray valves until pressure stabilizes. While pressurizer level is increasing and all backup heaters are on, pressure will continue to increase and spray valves will continue to open. B - Incorrect. First part correct (see A). B - Second part incorrect. Plausible, since subcooled water has insurged into the pressurizer, and a cooler steam space would cause pressure to decrease, but the compression of the steam space in the pressurizer due to the increasing level raises pressure and overrides the cooler temperature of the pressurizer liquid which would tend to lower pressure. C - Incorrect. First part incorrect. Plausible, due to the pressure going up. This automatically turns off all Backup heaters unless there is a >
> 5% high level deviation as in this case.
Second part is correct (see A). D - Incorrect. First and second parts incorrect. Plausible, since an error in the second part (thinking that the subcooled water would drop pressure in the pressurizer) combined with a miscalculation of program level, or using the 100% value of program level, would indicate a PRZR level less than program and pressure low. These errors would cause this choice to be selected. Second part is incorrect (see B). ARP-1.8, Version 33.0 Drawing 0175037 Sheet 2 lC-38, 27% Power, when PRZR level Ran this malfunction on the simulator laptop: IC-38, increased to 6% above program level all Backup heaters were on and spray valve demand had increased from the initial value of 6.6% to 20% open. The spray valves continued to open further for several more minutes. Page: 40 of 200 1211412009
Previous NRC NRC exam exam history if any: 010K1.03 O1OK1 .03 010 Pressurizer Pressure Control System System K1 K KI KnOllVleclge nvIedge of the phycaI of the physical connections connectionsandlor cause-effect rdationips and/or cau-effect between the relationships between the PZR pes and the PZR PCS the foilOlNing sjetems fdlawing 41.9 / 45.7 to 45.8) systems: (CFR: 41.2 to 41.9/45.7 K1.03 RCS ........................................................... 3.63.7 K1.O3RCS Match justification: A CVCS malfuction in this question causes excess mass in the RCS which causes an insurge into the pressurizer. This causes aa level deviation in the pressurizer which energizes the pressurizer heaters even though pressure is high, and opens spray valves due to the steam space compression and rising pressure, even though the insurge water is subcooled. Objective:
- 11. a set of plant conditions, LIST AND DESCRI BE theactionseffectsthat Given asst theactions'effects that will occur following aa CVCS Malfunction with no operator action (OPS-52201H15).
Page: 41 of 200 1211412009
11/25/08 7:43:37 11125/087:43:37 FNP-1-ARP-1 .8 FNP-I-ARP-I.8 LOCATION HA2
---.----~ .. ..
SETPOIN~%_~!~!,~;~-L~~~~~~-=> SETPOIN% of Span above Level PRZR LVL PRZRLVL ORIGIN: Level Bistable LB-459D from Level Transmitter DEV HI DEVHI or LT-461 and TY-408 median TAVG. LT-459 orLT-461 B/U HTRS ON BIUHTRS PROBABLE CAUSE
- 1. Pressurizer Level Instrument or Control System malfunction.
- 2. Plant Transient while in manual rod control.
- 3. Rod Control System malfunction.
- 4. Charging or Letdown System malfunction.
2 AUTOMATIC ACTION Pressurizer Backup Heaters energize. "~ OPERATOR ACTION
- 1. Place turbine load on HOLD.
- 2. Check pressurizer level indications and determine the actual level deviation.
- 3. IF an instrument failure has occurred, THEN go to FNP-1-AOP-100, FNP-I-AOP-I00, INSTRUMENT MALFUNCTION.
- 4. Ensure that the pressurizer backup heaters are energized.
- 5. IF required, THEN take manual control ofCHG ff of CHG FLOW FK 122 and decrease charging flow to return pressurizer level to the program band.
- 6. Determine the cause of the level deviation by checking:
5.1 Charging flow 5.2 Letdown flow 5.3 BTRS flow 5.4 Charging pump status
- 7. IF the alarm was caused by a plant transient, THEN control the transient and return Pressurizer Level to normal.
- 8. IF a charging Q.
OR letdown system malfunction exists, THEN go to FNP-I-AOP-16.0, FNP- 1 -AOP-. 16.0, CVCS MALFUNCTION.
References:
A-177100,Sh.357;U-260610;D-177109;D-177111;D-177112;D-177113; A-I77100, Sh. 357; U-260610; D-177109; D-177111; D-177112; D-177113; U-26664 U-2666477 PLS Document; Technical Specifications Page 11 of 11 Version 32.0
- 16. 011K
- 16. 011 K 5.05 001/NEW/RO/C/A 2.8/31/N/N/2ICVRIY 5.05 001/NEW/RO/C/A 2.8/3.1/N/N/2ICVRIY Unit 11 is Unit is at 100% power, at 100% power, and and hashas experienced Pressurizer Level experienced aa Pressurizer Level Control Control Malfunction Malfunction due to due to the the controlling controlling pressurizer pressurizer level level transmitter transmitter failing.
The following conditions The conditions exist: exist:
** PRZR LVL PRZR CONT CH, LVL CONT LS/459Z, is CH, L51459Z, is in in the the I/Il LT459/60" position.
"1/11 LT459160 position.
** Tavg is Tavg is 573.0° 573.0°F.F.
** AOP-100, Instrumentation Instrumentation Malfunction, Malfunction, is in progress.
is in progress.
** level is Przr level Przr is 40%
40% andand rising.
** Przr level control is in Manual.
** Charging flow is 125 125 gpm.
gpm.
** Letdown flow is 130 130 gpm.
1A 1A lB 1B IC 1C
** Injection flows are:
Seal Injection are: 8.1 gpm gpm 7.9 gpmgpm 8.0 gpm gpm
** Seal Leakoff Flows are: 2.9 gpm 3.0 gpm 3.1 gpm.
Which one of the following is the:
- 1) approximate time that it will take for the Pressurizer level to get to program level at 1) the current rate in Manual control, and
- 2) the correct switch position for PRZR LVL CONT CH LS/459Z lAW AOP-1 AOP-100? 00?
Time Switch position A. 56 Minutes I/Ill, LLT459/61 lIlli, T459/61 B. 94 Minutes I/Ill, LT459/61 lIlli, Cy C 56 Minutes 111111, Ill/Il, L LT461/60 T461/60 D. 94 Minutes 111/11, Ill/Il, LT461/60 A - Incorrect. The time is correct (see C). The second part is incorrect, since LLT-459
- T-459 was the controlling channel, and it needs to be selected completely out by selecting 111111, Ill/Il, 111/11, T461/60. This will place the remaining two operable LLTs Ill/Il, LLT461/60. Ts in service for Pressurizer level control. Plausible, Plausible, since confusion may may exist as to which of the two selected channels controls pressurizer level performs other control level and which performs functions in relationship functions relationship to to the position. For the switch position. For example, ifif LLT-460 T-460 waswas the the failure this would be be correct.
B B - Incorrect.
- Incorrect. The The time is is incorrect, incorrect, since level of since the level pressurizer in of the pressurizer in percent percent isis a volume affected affected by by the specific specific volume at at normal normal Operating Operating Pressure Pressure Pressurizer Pressurizer Temperature Temperature (about (about 648°F). since the Plausible, since 648°F). Plausible, the pressurizer lists the curve lists pressurizer curve change the change in in level level from from 40-50%
40-50% at at 93 gals II %, 93 gals but charging
%, but charging 100°F water of 100°F water 56 gal of 56 gal volume will volume will expand to 93 gallons expand to 93 gallons for a 1% rise.for a 1 % rise. The The second second part part is is incorrect incorrect (see (see A).
A). Page: Page: 42 c 200 42 of 200 12114/2009 1211412009
C - Correct. The time is correct, with a 100% program przr level of 50.2%, since a properly performed flow balance calculation shows that there is 10 gpm more charging into the RCS (in Charging and Seal inj mi minus the seal leakoff) than is leaving (in Letdown). Due to the specific volume of water at charging system temperature (about 100°F), which expands to pressurizer temperature (about 650°F), 56.3 gallons of charging water will equal 1 1%
% in the pressurizer (Per STP-9.0, RCS Leakrate determination).
min/lO gals) ** (56.3 gals/%) ** (10%) = (1 min/10 = 56.3 mins This gallons/% level relationship is also verified by steam table Specific Volume calculation (see below). D - Incorrect. The time is incorrect (see B). The second part is correct (see C). ft3/lbm per the steam tables, and it At 100°F, the specific volume of water is 0.016130 ft3/1bm
@ pressurizer temperature of 648°F. The charging ft3/lbm @
would expand to 0.02657 ft3/1bm gals/mm will expand to 93.5 gals/min water of 56 gals/min gals/mm at PRZR temperature (which is 11% % in the Pressurizer per Tank Curve 42). In approximately 56 minutes, at 10 gpm net Charging flow into the RCS, the pressurizer level will rise 10%. Tank Curve: Unit 1 1 Volume II Curve 42 (Hot Calibrated) 40% level=4403.04 gals 50% level=5336 gals (650° F) 93.5 gals/% PRZR level Hot calibrated (650°F) Per Steam Table: 0.016130 ft 3 3/1bm @1
/ftlbm OO°F
@100°F 0.02249 ft 3/1bm @
/lbm 3
ft @ 575°F 0.02657 ft 3 ft3/1bm @648°F
/lbm (Charging water expands in the RCS, which causes a pressurizer insurge, which in turn expands further in the pressurizer).
PRESSURIZER PRESSURE AND LEVEL CONTROL, OPS-62201 OPS-62201H, OPS-52201H, H, OPS-52201 H, ESP-52201H, ESP-52201 H, Student Text, Figure 8 Page: 43 dof 200 1211412009
LEVEL /
/
LEVEL
//
21.4 21.4 , 547 547 573,0 MEDIAN TAVG MEDIAN Previous NRC NRC exam history if any: 011 K 5.05 011K 011 Pressurizer Leiel Level Control System K5 Knowledge of of the operational implications foilaNing concepts asthey of the fdlcming implications of as they apply apply to the PZR LeS: (CFR: PZR LCS: (CFR: 41.5/45.7) K5.05 ndi caterl chagi I nterrati on of iindicaro K5,05Interr!ciion ng flow rate with volume of water rui chcrging requirro leII! back ran to bring PZR leid bed< to progrcrnmro leeI progranman leII! hot/cold ................. 2.8 3.1 2.83.1 Match justification: This question requires knowledge of determining what the net charging flow into the RCS is, and then determining the time for the pressurizer level return to program setpoint. The pressurizer level program value has been provided to ensure it is clear which program level is being used in this question. Program level changes each cycle and with changes in Tavg throughout each operating cycle, so it is provided. To obtain 3 plausible but incorrect distractors, a second part was added to test the system knowledge of the Pressurizer control system selector switch. Objective:
- 5. DEFI DEFINE NE AND EVALUATE the operational implications of abnormal plant or equipment conditions associated with the operation of the Pressurizer Pressure and Level Lel/el Control System components and equi pment to iinclude equipment following ncl ude the foil owi ng (OPS-52201 H07): H 07):
- Normal Control Methods
- Abnormal and Emergency Control Methods
- Automatic actuation including setpoint, if applicable
- Protective Interlocks A cti ons needed to mi Actions mitigate abnormality ti gate the consequence of the abnormal i ty Page: 44 of 200 Page: 200 1211412009
05/12/09 12:23:30 FNP-1-AOP-100 FNP-1-AOP-l00 INSTRUMENTATION MALFUNCTION Version 7.0 SECTION 1.2 Figure 11 OpsPcsl 13 OpsPcs113 6~ T T 59 462 LI LI LI 459Z 462 HSDP COLD-CALIBRATED COLD-CALIBRATED [/A LER LI ROOM 459B LI 459A MCB MEDIAN T AVG HI PRZR LVL RX TRIP LR459 LK 459 CHARGING DEVIATION DEVIATION PRZR FLOW LO LVL HI LVL HI LVL CONTROL ALARM ALARM; ALARM FCV 122 B/U HTRS ON ISOLATE 8149ABC; LCV 8149A,B,C; TURN OFF ALL HTRS; MCB ALARM MCBALARM PRESSURIZER LEVEL PROTECTION AND CONTROL Page 11 of 1I
(Q1B31 Pressurizer (QI Pressurizer B31KOOl) Capacity vs. K001) Capacity vs. % % Level level Volume HnCurve Unit I1 Volume Unit J Curve 42 42 Hot Cahbrated* Date-#- Hot Calibrateq* Approved Approved Date___________ Rev,No.~ Rev, No. 3. ES Manr 100 . . ... . 90 -1OO1283 90 80 . **. ..... . ..._.;
-767542 70
....c 60 %Leve Gaflons
%level@.allQns CD o 663.18 u
(.) 50 > I 10 1598.15
'- 20 20 2533,11 2533.11 Q)
- 11. 40 .
25 25 30o59 3000.59 --5-""3- '? f, ~!) 04 30 3468J!8 4403.04 . . 17CC 3pt?r~ 3 30 ... . . . . . . 60 338.00 6272.97 62n97 :..
~c;)Q 7 20 3OOO.59 . 70 7207.94 7207.94 2
. . 75 7675.42 7675.42
-. 93::;;g:~
e
/
80 8142.90 10 10 . - . 90 8142:90 9077.87 9071.87 --// 100 1001283
/('2t" ___
10012.83 0 0 663.18 2000 4000 6000 8000 1000Y 93. Sf; 1000 od,'"
-~
Gallons Gallons 1% 8ase1 on "Based on saturated saturated liQuid Hgud temperature temperature at at 2235 psig 2235 psig CacuLation Ref. Calculation SJ-O02290-003 No, SJ-OO-2200..Q03 Ref. No.
OpsPrsOl3 OpsPrs013 50.2 --- ------- ---------- ------------------------------------------- 50.2 PRZR PROGRAM LEVEL 21.4 547 573.0 MEDIAN TTAVG AVG Pressurizer Program Level Figure 8 H/52, OPS-6220 H/52:
- 17. 012K2O1
- 17. 012K2.01 OO1/FNP001/FNP BANK/RO/C/A BANK/RO/C/A 3.3/3.7/N/N/3/CVR/Y 3.3/3.7/N/N/3/CVRIY A loss loss of of A Train Building I125V Auxiliary Building Train Auxiliary 25V DC DC BusBus has occurred on has occurred on Unit Unit 1.
1. IfIf the the plant experienced aa problem plant experienced problem which which required manually tripping the required manually the reactor, reactor, which which one of one of the describes the the following describes effect (on the effect closed Reactor any closed (on any Trip and/or Reactor Trip and/or Bypass Bypass breakers) of breakers) of placing placing the RX TRIP the RX ACTUATION switch TRIP ACTUATION switch on on the MCB to the MCB to TRIP? TRIP? Placing the Placing the MCB MCB handswitch handswitch in would ____ ifif they were closed. in TRIP would closed. A'I open ALL reactor trip and bypass breakers. A. breakers. B. ONLY open the A breaker and
'A' reactor trip breaker and the B 'B' reactor trip bypass breaker.
C. ONLY open the B breaker and
'B' reactor trip breaker and the A 'A' reactor trip bypass breaker.
breaker. D. open BOTH reactor trip breakers but NOT open either reactor trip bypass breaker. A - Correct. Aux Building DC power is not required to trip open breakers, as long as the UV coils are deenergized by Solid State (SSPS). Voltage from SSPS feeds the 48V UV coils that will allow the trip breakers to open when power is removed (a trip signal deenergizes the UV coils). Loss of "A" A train AB DC would prevent the closure of the A RTB & B B RTBYP breakers, AND would prevent the shunt trip coils on the A RT & &B B BYP breakers from being energized to provide an additional trip signal. SSPS power is from the inverters which supply power from the Regulated AC, bypassing the inverters, if AB DC is lost. B B -Incorrect.
- Incorrect. See A. Plausible, since the Reactor Trip and Bypass breakers are operated and tripped by opposite trains. However, these two breakers are both operated by the B train aux building DC, and not the A train. Also, the Shunt trip coils operate to trip these breakers and the coils get power from AB DC (B train).
However, the UV coils can still deenergize if needed and trip all of the reactor trip breakers. Confusion may exist as to which train of breaker is operated by which train of DC, AND as to which type of DC is needed to trip the breaker (UV coil 48V or Shunt Trip coil 125 V). C -Incorrect. See A. Plausible, since the Reactor Trip and Bypass breakers are operated and tripped by opposite trains, AND these two breakers are both operated by the A train aux building DC. Also, the Shunt trip coils operate to trip these breakers and the coils get power from A train AB DC. However, However, the UV UV coils can still deenergize if needed and trip all of the reactor trip breakers. Confusion may exist as to which type of DC DC is needed to trip the breaker (UV is needed coil 48V deenergizing (UV coil48V or Shunt Trip coil 125 125 V energizing). D D -Incorrect. See See A. Plausible, Plausible, since the AB AB DCDC does supply the shunt trip coils, and only only the local local pushbutton pushbutton energizes energizes the Shunt Shunt trip coil to trip the Bypass breakers, Bypass breakers, so so there is a difference is difference in in the the way the trip breakers breakers and and the bypass bypass breakers breakers work for for loss loss of of AB AB DC. DC. However, However, the the UVUV coil coil will all RT trip all still trip will still RT && BYP breakers if BYP breakers if aa manual manual triptrip is is called called for. for. Confusion Confusion may may exist as to exist as the redundant to the using methods using redundant methods the UV and Shunt the UV and Shunt Trip Trip coils coils to to trip trip the the reactor. reactor. Page: Page: 45 45 c:JCl 200 200 1211412009 1211412009
Reactor Protection Functional System Diagram (FSD) Al 81 007, A181007, section 3.3.2 Eoch circuit breaker shall be equipped with a48 volt DC instantaneous undervoIte Each trip device undervoltaJetrip and a 125 Vdc shunt trip device. (Reference 6.4.086) The Shunt Trip Attachment Attochment coil shall 125 Vdc operate on 125 undervoltaJe trip functi on as aa backup for the undervoltage V dc and function tri p device. devi ceo The first method of tripping the breaker (i.e., reactor trip or bypass breakers) is by a loss or drop of rated voltage to the Undervoltage Relay (UV). The relay is normally energized from the 48 volt DC from the RPS. When the voltage is removed by an automatic reactor trip signal, the relay is de-energized and releases the UV trip lever, which actuates the trip shaft, causing the breaker to unlatch from the closed position. The second method of tripping the trip shaft is by the shunt trip lever when the normally de-energized shunt trip (SHTR) coil is energized. When energized, the SHTR coil is powered from the 125 volt DC system used to close the reactor trip and bypass breaker closing circuits. For the reactor trip bypass breaker, the SHTR relay is energized only by a manual pushbutton. After the reactor trip bypass breaker is opened, then a contact in series with the SHTR relay opens to de-energize the coil. Thus, the SHTR relay is only momentarily energized. reactor tri For the reoctor rday is energized only by a manual tripp bypass breaker, the SH T R relay reactor trip bypass breaker is opened, then a contact in series with the pushbutton. After the reoctor de-energize the coil. Thus, the SHTR relay is only momentarily energized. SHTR relay opens to d&energizethecoil. Train A of the reactor protection system powers the UV and Shunt Trip coils for RTA and BYB, and train B powers the UV and Shunt Trip coils for RTB and BYA per Reactor Protection Functional System Diagram (FSD) A A181007, 181007, Figure F-1.F-i. Page: 46 of 200 Page:46of200 1211412009
Previous NRC exam history if any: 012 K2 .01 012K2.01 Reedor Protection 012 Reactor Protecti on System KnONIedgeof K2 Knowledge suppliestothefoiloNing: (CFR: 41.7) of bus power aippliestothefdlcming: K2.O1 channels, components, end K2.01 RPS chennets, and interconnections interconnej:ions ..................... 3.3 3.7 3.33.7 Match justification: The 125V 1 25V Aux Building DC busses supply the Reactor Trip Breakers and Bypass Breakers (RPS components). They provide power to the Reactor trip breakers for both closing power and one of the sources of power for tripping the breakers. To correctly answer this question, the power supplies to the Reactor Trip breakers must be understood, including the A train 125V Aux Building DC bus. Objective:
- 2. RELATE AND DESCRIBE the operation of the Reedor Reactor Trip Breakers and Reactor Trip Bypass Breakers to include the operation of the following :(OPS-40302F02):
Shunt Trip Coils Undervoltage Coils
- 1. RELATE AND IIDENTIFY DENTI FY the operational characteristics including design features, capacities and protective interlocks for the following components associated with the Reedor Protection Reactor Protecti (OPS-52201 102):
on System (RPS) (OPS-522011 02):
- Solid state protection system (SSPS) cabinets cabi nets (A train/B train)
- IInput relay nput rei cabinets ay cabi nets
- Logiccabinets Logi c cabi nets
- relay Output rei cabinets ay cabi nets
- Safeguards test cabi cabinets nets
- Reactor tri Reedor tripp breakers
- Reactor tritripp bypass breakers Page: 47 of 200 12/14/2009 1211412009
FNP Units 1I &
& 22 REACTOR PROTECTION SYSTEM SYSTEM A18 1007 A181007 3.2.7 Interface Requirements Interface The STC The STC shall shall interface with the SSPS SSPS and and shall shall be supplied supplied by qualified Class IE 1E power from the 120 120 Vac vital power cabinets.
cabinets. (References 6.7.014, 6.4.059, 6.4.060, 6.4.084) 6.7.014,6.4.059,6.4.060,6.4.084) 3.3 REACTOR TRIP SWITCHGEAR TPNS Nos. Service QC1 1EOO4A-AB QC11E004A-AB (RTA, BYB) QC11EOO4B-AB QC11E004B-AB (RTB, BYA) 3.3.1 Basic Functions The reactor trip switchgear functions to switch power to or remove power from the control rod positioning equipment. The switchgear opens the reactor trip and bypass breakers A and B on reactor trip causing the control rods to fall by gravity into the reactor core. 3.3.2 Functional Requirements The switchgear assembly shall consist of two low voltage metal enclosed switchgear sections. One section will contain two series connected reactor trip circuit breakers. The second will contain two bypass circuit breakers connected so that a bypass breaker parallels each reactor trip breaker. The bypass circuit breaker is used to bypass the reactor trip breaker for on-line testing of the latter with the reactor in operation. The system also includes two 260 volt line to line identical three phase Motor-Generator sets rated at 400 KVA,KVA, reverse current relay, generator output circuit breaker, a synchronizer, and a common ground relay. Each circuit breaker shall have provisions for locking it in the Test "Test" and Disconnected "Disconnected" draw-out positions. The circuit breaker also includes positions for Connected "Connected" and "Remove." (Reference 6.4.077) Remove. Interposing relays shall be used to isolate Train A from Train B B wiring where it is necessary to parallel these circuits into a single output. Each circuit breaker shall undervolta~~e undervoltage e shall be equipped with aa 48 volt DC ingppçous trip device a a 125125 Vdc V dc shunt iIl.§1antaneous_ trip device. (Reference shunttrip 3-9 3-9 Rev. 00
FNP Units 1I &
&2 REACTOR PROTECTION SYSTEM A 181007 A181007 6.4.086) The Shunt Trip Attachment coil shall operate on 125 125 V Vdcdc and function as a backup for the undervoltage trip device.
The first method of tripping the breaker (i.e., reactor trip or bypass 4 breakers) is by a loss or drop of rated voltage to the Undervoltage Relay (UV). Ihe relay is no~mally energiz~fro~~lt DC from th~~ When the voltage is removed by an automatic reactor trip signal, the relay is de-energized and releases the UV trip lever, which actuates the trip shaft, causing the breaker to unlatch from the closed position. The second method of tripping the trip shaft is by the shunt trip lever when the normally de-energized shunt trip (SHTR) coil is energized. When energized, the SHTR coil is powered from the 125 system used to) 125 volt DC systel!!. to , close the reactor trip and bypass breaker closing circuits. j1 I 5U I-i letfJ-/ For the reactor trip bypass breaker, the SHTR relay is energized only by a manual pushbutton. After the reactor trip bypass breaker is opened, then a contact in series with the SHTR relay opens to de-energize the coil. Thus, the SHTR relay is only momentarily energized. For the reactor trip breaker, the SHTR relay is energized by the closing of a contact associated with a shunt trip attachment relay (STA for 52/RTA and STB for 52/RTB). STA (STB) is energized from the RPS voltage to the UV trip coil of the 52/RTA 52!RTA (52/RTB). When the voltage is removed by an automatic reactor trip signal, the relay will de-energize, closing its contact to energize the shunt trip coil of 52/RTA 52!RTA (52/RTB). (52IRTB). After the reactor trip breaker is opened, then a contact in series with the SHTR relay opens to de-energize the coil. Thus, the SHTR relay is only momentarily energized. 3.3.3 Design Transients The ambient design conditions are: 95% relative humidity and 40 deg. F to 120 deg. F temperature. (Reference 6.4.090) Also see Protection Features 3.3.7. 3-10 Rev. 0 Rev.0
CONTROL BOARD { MASTER AND ACTUATE TRAIN 'B' S~ITCHES OUTPUT SLAVE RELAYS SAFEGUARDS
<TRAIN 'B") ISOLA TION TO ROD DRIVE MECHANISM
/r----------~----~------~ 2v
~ ~
PROTECTION ! ANALOG ANALUG PROTECTION PROTECTION SYSTEM SYSTEM SYSTEM -- LOGIC SOLID STATE LOGIC ('J. ( 73007300 SYSTEM) SYSTEM) TRAIN B NUCLEAR NUCLEAR INSTRUMENTATION INSTRUMENTATION SYSTEM SYSTEM ROD DR OR FIELD CONTACTS FIELD CONTACTS { CONTROL SYSTEM
/ 1\ " INPUT PROCESS SENSOR CHANNEL CHANNEL_r -L-------------~------~~::~::::::=+::::::~~::::~tt---~ COMPUTER IVl-----/
CHANNEL 1ll)1----~
---1 [ill-{ j,- "DR" CABLE CHANNEL INPUT ;1 II )-)- - - - - - -
RELAYS
'i @)--{ <TYP.) TRIP REACTDR TRIP REACTOR SWITCHGEAR S~ITCHGEAR CONTROL BOARD I ,__~
r {
/" -<",
1\ BISTABL~ INPUT ( It IBYPASS
~
RKR . 'A' I CONTROL PROTECTION SYSTEM TRAIN A LOGIC
~:"":::'::":===-=t=---t----=:1
\
L----------f__==~~:__ ISOLATION i'1 C!~~~:T~. rrn IJ r---f1
' ¥\
ROD CONTROL 1 ROD CONTROL ACTUATE TRAIN 'A' M-G M-G SAFEGUARDS SET SET REACTOR PROTECTION SYSTEM BOUNDARIES FIGURE F-l
FNP UNIT 11 FNPUNIT LOAD LIST A- 5 0 625 0 A-S062S0 DFO3 E 1/) b V U VIX?
~
DF03 - I EDO4 ED04 LA13 lB 125V DC DIST PNL IB AB-139' AB-139 D177082 (CONTD) (CONT'D) BKR BKR - - TPNS DESCRIPTION SEE PAGE lB-ll 1311 N1R15GOOO1C-N N1R15G0001CN lF iF 4160V BUS BREAKER TEST CABINET lB-12 13-12 ... ~lN21LOOO1A""'A Q1N21L0001A-A lA 1A HOT SHUTDOWN PANEL .AUX.*RELAY AUX RELAY CABINET F-56 lB-13 1B-13 Q1H21EOOO4-A Q1H21E0004-A lF iF 4160V BUS LOCAL CONTROL PANEL DIFFERENTIAL LOCKOUT RELAY CONTROL CIRCUIT lB-14 1B-14 Q1R43EOOO1A-A Q1R43E0001A-A lF iF BUS LOADING SEQUENCER CONTROL POWER TO: LOSP SEQ, LOAD SHEDDING, BKR CLOSE FAILURE && SEQ LOCAL ANNUN lB-15 1B-15 Q1H21E0504-A Q1H21EO5O4-A lH 1H 4160V BUS LOCAL CONTROL PANEL DIFFERENTIAL PROTECTION
~{ECTION CONTROL CIRCUIT
\
lB-16 1B-16 Q1CllEOOO4B-AB Q1C11E0004B-AB "~REACTOR A)REACTOR TRIP SWITCHGEAR CONTROL POWER TO
.~. PASS BYPASS BREAKER && REACTOR TRIP BREAKER lB.-J.7 1B-17 Q1H21NBAFP26O5A lA Q1H21NBAFP2605A 1A LOCAL HOT SHUTDOWN PANEL >>> >>> F-57 F- 57
...-A A
Q1H21NBAFl?2.~05G. lG Q1H21NBAFP26O5G HOT SHUTDOWN PANEL >>> 1G LOCAL HO.T >>> F-58 F- 58
-A lB-18 iB-18 Q1R43E0501A-A Q1R4_0501A-A lH SFrNCFR 1H BUS LOADING SEQUENCER ui CONTROL CON JIrc.cjj POWER TO iO LOAD SHEDDING CONTROL CIRCUIT lB':"'19 Q1H25LOOO4-A .. 4A TERMINATION ..CABINET PANEL 4. REAR >>> F-59 F59 lB-20 1B-20 N1R15AOOO3-N N1R15A0003-N lC 1C 4160V E BUS UNDERVOLTAGE
*uaOLTL AND UNDERFREQUENCY PROTECTIVE RELAYING lB-21*
13-21 Q1H25LOOO6..;.A Q1H25L0006-A 6A TERMINATION CABINET PANEL 1 >>> I FRONT >>> F-60 lB-22 1B22 ------------ SPARE 1Isectf.doc sectf.doc Page F - 54
- Rev. 15IS
- 18. 012K6.03 001/FNP OO1/FNP BANK/RO/MEM 3.1/3.5/N/N/3/CVRIY 3.113.5ININI3ICVRIY Unit 2 is at 100% power, and the following conditions occurred:
- PT-455, PRZR PRESS, has failed off-scale HIGH.
- NO Operator action has been taken.
Which one of the following identifies the MINIMUM additional channels required to meet the RPS and ESF actuation logic to initiate any reactor trip or any safety injection on Pressurizer Pressure? Reactor Trip Iniection Safety Injection A. I 1 I 1 B~ B 1 1 2 C. 2 1 1 D. 2 2 A - Incorrect. The first part is correct. The second part is incorrect, but plausible since if the failed instrument tripped all bistables in the fail safe condition it would be correct. It would also be correct after the applicable TS and procedure directed assume no actions were complete (tripping all bistables), but the question specifies "assume actions. Pressure SI is only on low pressure, and the instrument failing operator actions". high does not automatically trip the low pressure bistable. The Reactor trip is on low or high pressure at this power level, and the high pressure condition would need only one more bistable in to cause a reactor trip. B - Correct. One more bistable on high pressure would cause a reactor trip, but the SI is actuated on low pressure only, so two more in the low pressure condition are required for an SI. C - Incorrect. First part is incorrect, but correct for a low pressure reactor trip. However, the High pressure reactor trip has one channel already tripped, and one additional channel will give a reactor trip signal. The second part is incorrect (see A). Both parts together are also plausible since confusion could cause choosing the exact opposite of the correct answer. oD - Incorrect.
- The first part is incorrect, but plausible since for a low pressure reactor trip it is correct (see C). The second part is correct since the only SI PRZR pressure actuation is low pressure (see B).
Page: 48 of 200 1211412009
Previous NRC exam history if any: 01 2K6.03 012K6.03 012 Reactor Protection System Knvledge d KG Knowledge etfect of a loss of the effect malfundion d lo or malfunction of the fdlcming following will have on the RPS: (CFR: 41.7/45/7) K6.O3Triplogiccircuits ................................................ 3.13.5 K6.03Triplogiccircuits 3.1 3.5 Match justification: A channel failure in one direction (failing high) causes a loss of the potential for meeting coincidence in the opposite direction (failing low) from that channel. This is one way of losing a a trip logic circuit for one of the channels. This question presents a a scenario where one of the 3 required trip logic coincidence circuits for Pressurizer pressure is lost, and knowledge of the effect on the RPS system is required to answer the question. Objective:
- 1. RECALL AND OESCRI DESCRIBE theoperation BE the operation and function of the thefollowing following reactor trip signals, permissives, control interlocks, and engineered safeguards actuation signals associated with the Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) to include setpoint, coincidence, rate ratefunctions(if functions (if any), reset features, and potential the potenti aI consequences for improper condiconditions ti ons to iinclude ncl ude those items in the taljles(OPS-52201107):
following tables (OPS-52201107):
- Taljlel, Table 1, Reactor Trip Signals
- Tle2, Engineered Safeguards Features Actuation Signals Table2,
- TalDle5, Table 5, Permissives
- Table 6, Control interlocks Taljle6, Page: 49 d of 200 1211412009
10/8/2009 09:51 10/81200909:51 FNP-I-EEP-O FNP-1-EEP-0 REACTOR TRIP OR SAFETY INJECTION Revision 37 B. Symptoms I. The following are symptoms that require a reactor trip.trip, if one has not occurred: Reactor Trip Instrumentation Setpoint Coincidence
- 1. Source Range NI - 31.,32 NI-31 32 cps 112 1/2 High Flux (TSLB3 1-1.1-2) 1-1,1-2)
(If not blocked)
- 2. Intermediate NI-35.36 NI-35 ,36 Reference 1/2 Range High Flux (TSLB32-1,2-2)
(TSLB3 2-1,2-2) Surveillance (If not blocked) Test Data Book for current S.P.
- 3. Power Range NI-41,42.43.44 NI-4l,42,43,44 25% Rx Pwr 2/4 High Flux.
Flux, Low (TSLB3 6-1.6-2.6-3.6-4) 6-1,6-2,6-3,6-4) Setpoint (If not blocked)
- 4. Power Range NI-41.42.43.44 NI-41,42,43,44 109% Rx Pwr 2/4 High Flux.
Flux, (TSLB2 11-1.11-2.11-3. 11-1,11-2,11-3, High Setpoint 11-4)
- 5. Power Range NI Cabinets +5%/2 sec. 2/4 High Positive (TSLB2 12-1.12-2.12-3.
12-1,12-2,12-3, Flux Rate 12-4)
+credits
- 6. OTlIT OTAT TI-412C.422C.432C 117% 2/3 (TSLB2 7-1.7-2.7-3) 7-1,7-2,7-3) -penalties
- 7. OPlIT 0PT TI-412B.422B.432B TI-412B,422B. 432B 110%-penalties 2/3 (TSLB2 (TLB2 8-1.8-2.8-3) 8-1 8-2 8-3)
~~:r-dPI-45;J4~b.q57 4b,L57 2 19-1.19-2 19-3 19-1,19-2,19-3) 1865 psig (rate compensated)
Gii) (Rx Pwr >
> 10%)
- 9. Pressurizer. 5,456,457 r -
2385 psig Pressure 20 1,20-2,20-3)
??( // gci L
/
C -- (Le A ) c:-- I 2 /fr C( ( 4 /. I
/
Page 2 2 of 37
10/8/2009 10: 10:18 18 FNP-1-EEP-0 FNP-l-EEP-O REACTOR TRIP OR SAFETY INJECTION Revision 37 II. The following are symptoms of a reactor trip:
- a. Any reactor trip annunciator lit.lit
- b. Rapid decrease in neutron level indicated by nuclear instrumentation.
- c. All shutdown and control rods are fully inserted. Rod bottom lights are lit.
III. The following are symptoms that require safety injection. injection, if one has not occurred:
/
~'"
( SI Signv Instrumentation Setpoint Coincidence r-..... (TSLB) C
~ ~~
lO r
~"~'/\"~
- 1. pressuri~ v~
Pressurizecç ( PI-455 456.457 1850 psig 2/3 pressure 1 pressure(lf ) ~ 2 17-1.17-2.17-3) c ed) ,---- (If not 15bT3ed)
- 2. Steam Line PI-474.484.494.
P1-474,484,494, 100 psid 11 steam line Differential PI-475.485.495. P1-475.485,495, 100 psig less pressure PI-476.486.496 P1-476.486.496 than other two (TSLB4 10-2.10-3.10-4. 10-2,10-3,10-4, on 2/3 protection 11-2.11-3.11-4. 11-2.11-3,11-4, sets 12-2.12-3.12-4. 12-2,12-3,12-4, 13-2.13-3.13-4. 13-2,13-3,13-4. 14-2.14-3.14-4. 14-2,14-3,14-4, 15-2.15-3.15-4) 15-2,15-3,15-4)
- 3. Low Steam Line PI-474.485.496 P1-474.485,496 585 psig 2/3 pressure (TSLB4 19-2,19-3,19-4) (rate (TSLE4 19-2.19-3.19-4)
(If not blocked) compensated)
- 4. Containment PI-951.952.953 P1-951.952.953 44 psig 2/3 pressure high (TSLBI 1-2.1-3.1-4)
(TSLB1 1-2,1-3,1-4)
- 5. Manual N/A N/A 112 1/2 IV. The following are symptoms of a safety injection:
- a. Any SI annunciator lit.
- b. BYP && PERMISSIVE SAFETY INJECTION ACTUATED status light lit
- c. 1-1 or MLB-1 MLB-1 1-1 MLB-l 11-1 11-1 lit
- d. HHSI flow greater than 00 gpm.
Z- /v1X c /) v - (7 -vl
)) L L~ r ?
Page 44 of 37
QUESTIONS REPORT for RD RO 2010 2010 NRCNRC EXAM SUBMITTAL 12-15-0912-15-09
- 19. 013K2.OI
- 19. 3.6/3.8JNJN/3/CVRlVER 55 EDITORIAL 002JNEW/RO/MEM 16/3.8/N/N/3/CVR/VER 013K2.01 002/NEW/RO/MEM A loss of BB Train Auxiliary Building 1125V 25V DC Bus has occurred on Unit 1.1.
Which one of the following is is the correct impact impact on BB Train ESF Equipment control? The BB Train SI actuated MOVs (1) automatically stroke upon an SI actuation, and BB Train ESF pumps (2) be started in LOCAL at the HSP. (1 ) (1) (2) A. will can B~ B will can NOT C. will NOT can D. will NOT can NOT Tuesday, December 15, 2009 6:38:08 AM 50
A - Incorrect. The first part is correct (see B), however The second part is
- is not correct, these breakers receive control power from B B train DC and although there is an alternate control power that is placed into the circuit when in LOCAL at the HSP, it is also from B train DC.
"LOCAL" Plausible, since B train has alternate control power and SOP-36.6, CIRCUIT BREAKER RACKING PROCEDURE, has numerous cautions about an additional Control Power source for the B train ESF Pumps.
The existance of Alternate Control power may cause confusion as to the ultimate source of the alternate control power. B - Correct.
- The B Train SI actuated (that would normally stroke upon an SI actuation signal) MOVs can be operated with or without DC power (a separate DC power source is provided to some of these valves if equipped with a disconnect for position indication only-- and these valves do not stroke automatically following an SI actuation because of normal position of that disconnect--ie MOV8808B).
the "normal" M0V8808B). The control power for operation comes from the 600V AC supply for each MOV via transformer. The control power for operation of the B Train ESF breakers is supplied from B Train 125V Aux Bldg DC. Although equipped with an alternate control power source, that power is also supplied form B B train DC on another breaker with a different cable run (for Appendix R concerns). C - Incorrect. The first part is incorrect (see B). Plausible, since some of these MOVs are equipped with a DC power (M0V8808B). Further plausibility is provided from supply for indication (MOV8808B). many solenoid operated valves auto stroke after SI, and they usually require DC power (although for the opening) The second part is also incorrect (see A). DD - Incorrect. The first part is incorrect (see C). The second part is correct (see B). Page: 51 of 200 1211412009
Previous NRC Previous NRC exam exam history history ifif any: any: 013 K2 .01 013K2.01 Engineered Safety 013 Engineered Safety Features Features Actuation System System K2 Knowledge K2 Knowledge of bus bus power power supplies to the following: (CFR: (CFR: 41.7) K2.01 ESFAS/safeguards K2.01 ESFAS/safeguards equipment control .............................. . 3.6* 3.8 3.6* Match justification: ESF equipment (pump) control requires DC for Pump breaker operation and breaker indication, even though the components themselves are powered from AC. ESF MOVs are powered from the 600V MCC AC and get control power for valve position indication from the same MCC AC source. To answer this question correctly knowledge of the power supplies for these ESF control functions is required. Wrote this question to intentionally stay away from 120V vital AC Instrumentation power due to potential overlap with other questions on this exam. Objective: 1I NAME AND IDENTIFY the Bus power supplies, for those electrical components associated with the Emergency Core Cooling System, to include those items in Table 4- Power Supplies (OPS-40302C04).
- 2. RELATE AND DESCRIBE the effect(s) on the Emergency Core Cooling System for a loss of an AC or DC bus, or a malfunction of the Instrument Air System (OPS-40302C06).
Page: Page: 52 52 of 200 c:l200 1211412009 1211412009
2wP 7 07/01109 15:27:43 07/01/09 15:27:43 FN?P-36.6 CAUTION: CAUTION: Failure to Failure deenergize all to deenergize sources of all sources of DC DC control control power power while while engaged engaged in breaker in breaker racking racking could could result result in in equipment equipment damage damage AND AND severe severe injury OR personal injury personal 213 death. death. Refer to Refer to Table Table 11 to determine IF to determine IF breaker breaker has has alternate alternate DC DC control control power, power, AND AND IF IF additional action additional action isis required. required. A breaker in an ESF 4160V bus, 1I2F, 1/2F, 1I2G, 112G, 1I2K, 1/2K, 1I2L, 1/2L, 1I2J 1/2J and 1I2H, 1/2H, cannot be left in the TEST position if the switchgear switchgear is required to be operable in Modes 1-4, 1-4, unless the seismic modification has been implemented on both the breaker and the cubicle. In In Modes Modes 5, 6, and defueIed, defueled, the breaker may be left in the TEST position and the switchgear will still be operable. (See Precautions and Limitations.) 4.7 PRIOR to racking 4160V 41 60V Circuit Breaker, perform a visual inspection of visible pi control wires on the cubicle door to ensure:
- no significant insulation damage is present.
- visible wires are landed at terminations.
4.8 Removing Breaker to TEST Position 4.8.1 Insert racking lever on left side of fulcrum plate and attach to mating hole on breaker. (Refer to Fig. 2.) 4.8.2 Depress breaker release lever, this releases the breaker from the interlock bar. A. At the same time push down on the racking lever to disengage the main line contacts. B. When the plunger is out of the guide rail notch, the breaker release lever can be released. NOTE: It is possible to use the racking lever to pull the breaker to the TEST position, It but caution should be used to ensure that the operators operator's hands do not slip off the lever, or that the handle does not slip off the lever. 4.8.3 Remove the racking lever from the breaker compartment. Version 55.0 Version 55.0
4P ¶ 07/01/09 15:27:43 07/01109 15:27:43 FNP-0-SOP-36.6 FNP-0SOP-36.6 TABLE 11 TABLE 1I BREAKERS WITH ALTERNATE DC CONTROL POWER CAU'l7ION: CAUTION: . deenergize all sources of Failure to deenergizeaU of DC control power. power while engaged in breaker racking could in. breaker ..ackingcould result in equipment damage AND severe personal injury OR death.
- 1. Breakers in this table have two separate DC control power supplies. One DC supply feeds the breaker through the DC supply in the breaker cubicle. An alternate DC supply has been installed to provide a separate supply for components capable of being operated from the Hot Shutdown Panel (HSP).
- 2. Alternate DC power supplies are only energized when the handswitch on the HSP is place in LOCAL.
- 3. Remote DC control power fuses will be removed (replaced) from (in) breakers in the table in conjunction with securing (restoring) local DC control power QE OR the appropriate HSP control switch will be verified in the REMOTE position.
REMOTE CONTROL COMPONENT BREAKER FUSE LOCATION Charging/HHSI Pump 1B/2B (B Train) DGO7 DG07 HSP-C Charging/HHSI Pump I1 CI2C CharginglHHSI C/2C DGO6 DG06 HSP-C CCW Pump 1A/2A DGO4 DG04 HSP.C HSP-C CCW Pump 1B/2B (B Train) DGO5 DG05 HSP-C MDAFW Pump 1B/2B IB/2B DG1O DGIO HSP-C Pressurizer Heater Group 1B/2B ECu EC11 HSP-C Page 1I of 11 Version 55.0
~~-
FNP UNIT 11 FNPUNIT LOAD LIST LOAD LIST ~ A-S062S0 06250 1G416OVBU IG 4160VBUS S AB AB -121'
- 121 D177006 Dl 77006 BKR TPNS TPNS DESCRIPTION DESCRIPTION SEE SEE PAGE PAGE Q1R1SA0007-B 1G 4160V BUS DGO1 NIR11AO5O1-N IA 1A STARTUP TRANSFORMER (ALTERNATE) <<<
STARTUP TRANSFORMER <<< DGO2 Q1R15A0506-B BUS>>> 4160V BUS>>> 1L4160V L-I L-1 DGO3 Q1R1IB0005-B 1E .4160/600VSST
.1E 4160/600V SST (NQRMAL)
(NORMAL)>>> >>>EE02 EEO2 >>>
>> G-2 DGO4 DG04 Q1P17M000IA-B QIPI7MOOOIA-B IA CCW PUMP IACCWPUMP DGO5 DG05 QIP17M0001B-AB Q 1P 17MOOO lB-AB lB CCW PUMP DISC SWITCH QIR18A00004B-B IB QIR18A00004B.-B >>>
>>> lB CCW SUPPLY)
PUMP (B TRAIN SUPPL Y) DGO6 DG06 QIE2IM000IC-B QIE21MOOOIC-B IC CHARGING/HHSI PUMP DGO7 DG07 Q1E2IM000IB-AB Q I E21 MOOO lB-AB lB QIRI8A000IB-B >>> CHARGING/HHSI PUMP DISC SWITCH QIRI8AOOOlB-B lB CHARGINGIHHSI >>> lB CHARGING/HHSI PUMP (B TRAIN SUPPLY) DGO8 DG08 Q1R43A0502-B QIR43A0502-B lB DIESEL GENERATOR (EMERG) <<< IB <<< DGO9 DG09 Q1EIIM0001B-B QIEIIMOOOlB-B 1lB RHR/LHSI PUMP B RHRlLHSI DGIO DGlO QIN23M000IB-B QIN23MOOOIB-B lB AFW PUMP IBAFWPUMP DGII Q1E13M0001B-B QIE13MOOOIB-B 1BCTMTSPRAYPUMP lB CTMT SPRAY PUMP DG12 Q1R11B0006-AB QIRl1BOOO6-AB IF iF 4160/600V SST SWITCH QIR18A0003B~B SST DISC SWITCH Q1R18A0003B-B>>>>>> IFiF F F--113 113 I 4160/600V SST >>> IF SST>>> iF LOAD CENTER CENTER (B TRAIN TRAiN SUPPLY) Si DG13 Q1R15A0504-B 4160V BUS>>:> 1J 4160V J-l DGI4 Q1RI5BKRDG14 QIRl5BKRDGI4 PT COMPARTMENT DG1S DGl5 N1R11AO5O2-N NIRIIA0502-N IB STARTUP TRANSFORMER (NORMAL) <<< lB STARTUP <<<
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sectg.doc 1lsectg.doc Page G-lI Page G - Rev. 13 Rev. 13
FNP UNIT FNP UNIT 11 LOAD LIST LOAD LIST A-506250 A-506250 DG03 DG03 EEO5 EE05
~~------- 72 j)P
~5V (E25VDC1M PN~) STPNL) AB-121' AB-121 D177083 D177083
-- DC DIST TPNS DESCRIPTION SEE SEE PAGE 1E 125V DC DISTRIBUTION PANEL <<< tB07 1E-Ol Q1R15A0007-B 1G 4160V BUS DC CONTROL POWER FOR INC BREAKERS DGO1, DGO8 & DG15 1E-02 lE-02 Q1R16B0005-B QIR16B0005-B 1C 600V LOAD CENTER DC CONTROL POWER FOR INC lC BREAKERS EC02, ECO2, EC07, ECO7, EC08 ECO8 & & ECIO EdO 1E-03 lE-03 Q1R15A0007-B QIR15A0007-B 1G 4160V IG 4160V BUS BUS DC CONT~O OWER FO~ER CONTROJPPOWER FORE DER BREAKERS ,-,>-,-_~,-,02, GO2, DG03, D~ DG05,~)
DGO3,(DGO DGO5, DG07, DGO7, G1O DGll, DG11, DGIDG12& DG13 IG 1G 4160V BUS U/F TRIP AUX RELAYS (TRIP DG DC BKR DGO8) DG08) 1E-04 lE-04 Q1R1630005-B QIR16B0005-B 1C 600V LOAD CENTER DC CONTROL POWER FOR FOR lC FDR BREAKERS EC03, ECO3, EC04, EC05, ECO6, ECO4, ECO5, EC06, ECO9, EC09, ECu, ECll, EC12, EC13 & & EC14 1E-05 iE-OS Q1H21NBL27O2S-B TI3T TRAIN
... """'~....... PENETRATION PANEL >>> >>>
PENETRATION ROOMROOMIi:)v..l~-,I; ISOLATION G-51 II I 1E-06 lE-06 Q1R16B000 QIR16 7-B 07-B lB IE 600V LOAD CENTER DC CONTROL POWER FOR INC BREAKERS EEO2, EE02, EEO7 EE07 && EE12 1E-07 lE-07 Q1R15A050 4-B QIR15A0504-B IJ 4160V BUS DC CONTROL lJ CONTROL POWER FOR BREAKERS DJOl, DJO2, DJO1, DJ02, DJO3, DJ03, DJO4, DJ04, DJO6 DJ06 && DJO7 DJ07 IJ 4l6OV 1J 4160V BUS U/F U/F TRIP AUX RELAYS (TRIP DG BKR (TRIP BKR DJO6) DJ06) N1R15A050 9-N NIR15A0509-N IJ 4160V 1J 4160V BUS BUS BREAKER BREAKER TEST TEST CABINET 1E-08 lE-08 Q1R16B000 7-B QIR16B0007-B IE 600V 1E 600V LOAD LOAD CENTER CENTER DCDC CONTROL CONTROL POWER POWER FOR FOR FEEDER FEEDER BREAKERS EEO3, BREAKERS EE03, EEO5, EE05, EEO6, EE06, EEO8, EE08, EEO9, EE09, EE1O, EEIO, EEll, EE13, EE11, EE13, EE14, EE14, EE15 EE15 CC( - sectg.doc 1lsectg.doc Page Page GG - 49
- 49 Rev 66 Rev
FNP UNIT 11 FNPUNIT DGO3 DG03 LOAD LIST LOAD LIST c
?06250
?A5O625O EEO5 EE05
__ _ , Q 5 LB~4 7 0 F l~CDiSTRP~ II//) (iFi2wnC1iii rIuiNr AB421 D177083 BKR TPNS I DESCRIPTION DESCRIPTION SEE SEE PAGE I 1F02 SPARE 1F03 1F-03 SPARE 1F-04 1F04 SPARE 1F-05 1F05 G22NAHR261 B- 1B N1G22NAHR261OB- lB CATALYTIC H2 RECOMBINER DC CONTROL PANEL - - N N ANNUNCIATORS & & ECV-1119, TCV-1114, ECV-1112 SOLENOIDS SF-06 IF-06 N1G21NGWEP2~02-N1G21NGWE?2602- WASTE ;ENCAPSULATION ENCAPSULATION SYSTEM CONTROL PANEL G-71 TG-71 II N >>> 1F-07 04 B1TB0004 JUNCTION BOX FOR MSVR FLOODING SENSOR RELAYS 49-1, 492, 491, 49-2, 493 49-3 && LSX (2-3A (23A FUSES) 1F-08 SPARE
~1~F~-~O~9~fq~2:U~Uij~~~LOCiL:HQo~s~~~<~rn)ml~W-EP~ANE~iL~"ilBBulI->>>>>~------lIcG~-07~2~1 1F-09 Q1H2iNBAFP26O5B LOCAL HOT SHUTDOWN PA11EL 1B >>> G-72
-B I 1F-iG Q1P15NFSS26O7B- B TRAIN. SAMPLE ISOLATION VALVE CONTROL "Bn .. TP.AIN G-73 II B PANEL >>> >>>
1F-11 iF-li SPARE CABINET G-74 1F12 Q1N21L0001BB 15 ROT SHUTDOWN PANEL PANEL AUK AUX RELAY CABINET G- 74 II II ( 1F-14 , 1H21NBAFP26O5C 1C LOCAL
-B SWITCH BOX >>>
HOTSBUTDOWN
.L.IV\,....n..&.I HOT SHU1 PANEL SELECTOR CTOR G-75
/ \j 2) C/7 -2Z L)J, t7 sectg.doc 1lsectg.doc Page G - 69 Page G - 69 Rev.
Rev. 66
FNP UNIT 11 FNPUNIT LOAD LIST LOAD LIST t A-506250 DGO3 DG03 EEO5 EE05 LB14 (TF~ SCAL )CCOCAL HOT SHUTDOWN PNL SEL SWITCH BOX AB-121' AB-121 D181664 FUSE TPNS DESCRIPTION Q1H21NBAP26O5C 1C LOCAL HOT SHUTDOWN PANEL SELECTOR SWITCH B BOX <<< iF-id Fl Q1H22L0004-B RELAYS TR1-TR6 TR1TR6 ON TRC- TRC3 FOR: RX VSL HEAD VENT SVs 2213B-B 2213BB & & 2214B-B; 2214BB; PRZR PWR REL SVs 0444BA-04445A B B &
& BB-B AND ISO MOV 8000B-B; RWST MOV 0115D-B; CCW HX MOV NOV 3047-B; "HSP HSP SEL SW IN LOCAL" LOCAL ALARM F8 52-DG06 52-DGO6 LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO CHG/HI HEAD SAFETY INJECTION PMP 1C lC t F9 52-DG07 52-DGO7 LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO 7 / /2 ORG/HI HEAD SAFETY INJECTION PMP 1B CHG/HI lB FlO 52-DG04 52-DGO4 LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO CCW Fil F11 F12 52-DG05 52-DGO5 52-DG10 52-DG1O PMP 1A LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO CCW PMP 1B lB LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO AFW j
PMP 1B lB F13 52-EC11 52-ECu LOCAL MODE DC CONTROL PWR FOR FEEDER BKR TO PRZR HTR BACKUP GROUP 1B lB yp 1)
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- 20. 013K5.01 013K5.O1 001/NEW/RO/C/A OO1/NEW/RO/C/A - 2.8/3.21N/N/3/CVRN
- 2.8/3.21N/N/3/CVRIY Unit 2 was operating at 100% power, and the following conditions have occurred:
- PT-950, CTMT PRESS, has failed.
- PT-950, HI-3 Hl-3 bistable, is in the BYPASS condition.
- Subsequently, the 20 2D vital panel has become de-energized.
If a Large Break LOCA occurs, which one of the following describes:
- 1) the number of channels of Hi-3 bistables which will be actuated and
- 2) the number of trains of containment spray (CS) that actuate automatically?
A A~ 1) Two channels ONLY will be actuated.
- 2) One train ONLY will actuate.
B. 1) Two channels ONLY will be actuated.
- 2) Two trains will actuate.
C. 1) Three channels will be actuated.
- 2) One train ONLY will actuate.
o. D. 1) Three channels will be actuated.
- 2) Two trains will actuate.
Page: 53 of 200 1211412009 1211412009
A - Correct. The channel I bistable is bypassed and won't
- wont actuate. The channel IV wont actuate since it is deenergized by the loss of 1ID Bistable won't D vital 120V AC, and it is an energize to actuate bistable. Even though the coincidence for CS actuation would still be met with channel II & & III, Ill, and both trains of SSPS would get the signal to actuate both trains of CS, the slave relays are deenergized in SSPS train B due to the loss of 11 D vital 120V AC panel. This would prevent Train B CS from actuating.
B - Incorrect. The first part is correct (see A). The second part is incorrect, but plausible since both trains of SSPS get a signal to initiate CS, even with a loss of 11 D 120V vital AC panel. The master relays call for an actuation on Both trains, but on B train the slave relays don't dont have power to start the loads and operate the valves for the ESF actuations. C - Incorrect. The first part is incorrect, but plausible. For most bistables, when they lose power they deenergize to actuate. Containment spray is an exception to this general rule. Channel IV is thus deenergized and will NOT actuate. Examinee could also not realize that the bypass function (which prevents the bistable from actuating) is opposite the usual trip bistable function (which causes the bistable to trip) for a loop in maintenence. This would cause this choice to be selected. The second part is correct (see A). D - Incorrect. Both parts are incorrect (See C & B). FSD: A181007 REACTOR PROTECTION SYSTEM 2.2.2 The RPS system is housed in two physically phycally and electrically independent equipment trains(Train trains (Train A" A II and Train B), II B"), typically the Solid State Protection System asthe referred to as (SSPS) cabinets. (Reference 6.7.003) Z23 Any singlefailurewithin the RPS system (sensor channel or actuation train) shall not 2.2.3 prient the redundant system actuation. On loss of channel prevent channd or train poNerpcmer the bistable biabIe shall be tripped. T The he only exception to tothelossof channd or the loss of channel w train poNer pcwer causing causrng the bistable biabIeto to trip is isfor for Containment Spray and Containment Phase B IIsolation SJlation where the biabIesmu bistables must enerçjze energize to actuate. (References 6.1.002, 6.1.41, 6.7.014,) 2.Z6 nstrument channdsshall 2.2.61I nrument channels shall be poNered pcmered from four separate independent AC inrument instrument diribution distribution panels panels. These pandsshall panels shall be fed from four separate and independent ClasslE Class 1E inverters inverters. (References 6.1.023, 6.4.081, 6.4.091, 6.7.014, 6.7.016) 6.1.023,6.4.081,6.4.091,6.7.014,6.7.016) 2.2.15 Except as noted below, bel ow, all reactor triptri p and safeguards actuation actuati on channds channels shall be placed pi aced in the trip mode when the channel is out of service for any reason. The reactor trip and safeguards actuation drcuitscircuits noted bdow belON be adminirativdy administratively bypassed for maintenance mai ntenance on a single si ngl e channel. 1.
- 1. Source range high hi gh neutron flux fl ux trip tri p Intermediate
- 2. I ntermedi ate range high hi gh neutron trip tri p
- 3. High 33 containment contai nment pressure actuation actuati on of containment contai nment spr spray LOAD LIST: A-506250, Page G-39, ID PageG-39, 120V Vital AC Di 1D 120V Dist PNL.
Page: 54 of 200 1211412009 12/14/2009
Previous NRC exam history if any: 01 3K5.01 013K5.01 FeaturesActuation 013 Engineered Safety Features Actuation System K5 Knowledge K5 Kncmledge of the operational implicationsofthefoilONing implications of the fdlowing concepts asasthey to the ESFAS: they apply totheESFAS: 41.5 /45.7) (CFR: 41.5/45.7) K5.O1 Detinitionsof K5.01 D&initionsof sctety sy tran end chain ........................... 2.83.2 aid ESF chennel 2.8 3.2 Match justification: To answer this question correctly, knowledge is required of what constitutes a safety train, an ESF channel, and the operational implications of each must be understood. Objective:
- 1. RECALL AND DESCRI DESCRIBE BE the operation and function of the thefollowing followi ng reactor trip signals, permissives, control interlocks, and engineered safeguards actuation signals associated with the Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) to iinclude setpoint, ncl ude setpoi coincidence, nt, coi nci dence, rate functi ons (if any), reset features, and functions(if the potenti aI consequences for improper condi potential conditions ti ons to iinclude ncl ude those items in the following followi (OPS-52201107):
ng tables (OP&52201107):
- Table 1, Reactor Trip Signals
- Table T abl e 2, Engineered Engi neered Safeguards Features FeaturesActuation A ctuati on SiSignals gnal s
- Table 5, Permissives
- Table 6, Control interlocks Page: 55 of 200 1211412009 12114/2009
FNP Units FNP Units 1I & 22 REACTOR REACTOR PROTECTION PROTECTION SYSTEMSYSTEM A- 18 1007 A-18lO07 TABLE T-4 TABLE T ENGINEERED
- ENGINEERED SAFEGUARDS ACTUATION ACTUATION SIGNALS SIGNALS CONTAINMENT CONTAINMENT SPRAY SPRAY ACTUATION CONTAINMENT SPRAY SPRAY . JJ SETPOINT SETPOINT COINCIDENCE INTERLOCKS INTERLOCKS
& BLOCKS BLOCKS PROTECTION PROVIDED PROVIDED FOR FOR MODES MODES OF OPERATION OF FSD FSD SECTION SECTION High-3 High-? ~7 {/'i\ <2--q,~V 2/4 2/4 None None Protects Protects containment containment for 1,2,3 1,2,3 2.4 2.4 contamme~ High-3 contaimnent containment aa loss of of coolant or or steam steam 2.7.1 2.7.1 signals 2:
pressure signals line line break break inside 2.7.2 2.7.2 setpoint setpoint containment containment Fig. Fig. 22 e Slit. Sht. 88 ~
&( Prevents over over pressurization pressurization ofof P__,
by JP )2
)P i?C'/'v *V~ k / ,f~ containment containment structure and and subsequent structure subsequent building building rupture rupture Manual Manual N/A NIA -2/4
-214 switches switches
-Consist
-Consist ofof four
-Containment
-Containment spray actuation spray is manually actuation is manually Operator discretion f 1,2, 3,4 1,2,3,4 2.7.1 2.7.1 Fig.
Fig. 22 momentary momentary reset by by depressing depressing Sht. Sht. 88 switches switches inin two both both train A and and train train groups B push buttons B reset push buttons on MCB onMCB T4-3 T4-3 Rev. Rev. 00
ESFAS ESFAS Instrumentation Instrumentation BB 3.3.2 3.3.2 BASES BASES APPLICABLE APPLICABLE b.
- b. Containment Spray Containment Spray - Automatic Automatic Actuation Actuation Logic Logic and and SAFETY ANALYSES, SAFETY ANALYSES, Actuation Relays Actuation (continued)
Relays (continued) LCO, and LCO, and APPLICABILITY APPLICABI L1TY the use the use of of the the Manual Manual Initiation Initiation Switches. Switches. Automatic Automatic Actuation Logic Actuation and Actuation Logic and Actuation Relays Relays must must be be OPERABLE OPERABLE in MODE 4 to support system level in level Manual Manual Initiation. Initiation. In In MODES 5 and 6, there is insufficient 6, there insufficient energy in the primary and secondary systems to result in containment overpressure. In overpressure. MODES 55 and In MODES and 6, 6, there is is also also adequate adequate operators to evaluate time for the operators evaluate unit unit conditions conditions andand respond, to mitigate the consequences of abnormal conditions by manually starting individual components.
- c. Containment Spray - Containment Pressure - High 3 This signal provides protection against a LOCA or an SLB inside containment. The transmitters (dip (d/p cells) and electronics are located outside of containment with the sensing line (high pressure side of the transmitter) located inside containment. Thus, the transmitters will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties.
This Function requires the bistable output to energize to perform its required action. It is not desirable to have a loss of power actuate containment spray, since the consequences of an inadvertent actuation of containment spray could be serious. Note that this Function also has the inoperable channel placed inEYpass ini ~ rather than probabi1TfTf trip to decrease the probability of an inadvertent actuation. The Containment Pressure High 33 instrument Function consists of aa two-out-of-four logic configuration. Since containment pressure is not used for control, this arrangement exceeds the minimum redundancy requirements. Additional redundancy is warranted warranted because this Function Function is energize to trip. Containment Pressure - High 33 must be OPERABLE in MODES 1, 1, 2, and 33 when there isis sufficient energy in in the primary and secondary sides to pressurize pressurize the containment following aa pipe break. In (continued) (continued) Farley Units 11 and Farley Units and 22 BB 3.3.2-14 3.3.2-14 Revision 00 Revision
FNP Units FNP Units 11 && 22 REACTOR PROTECTION REACTOR PROTECTION SYSTEM SYSTEM A-181007 A-181007 injection isis required injection required and and containment containment sprayspray may may bebe necessary necessary to to ensure the ensure the integrity integrity of the containment of the containment building building and and limit limit offsite offsite dosage. Actuation dosage. Actuation signals signals for for these these protective protective functions functions areare provided by provided by containment containment pressure pressure transmitters. transmitters. If If 2/3 2/3 high-l high-I containment pressure containment pressure channels channels sense pressure greater sense aa pressure greater than than 44 psig, psig, then safety then safety injection injection will will be be initiated. initiated. (References (References 6.1.002, 6.1.002, 6.4.007, 6.4.015, 6.4.036) 6.4.007,6.4.015,6.4.036)
- 2. Containment High-2 Pressure Pressure If containment pressure increases to 16.2 psig on 2/3 containment 16.2 psig pressure channels, a main steam steam line isolation signal is generated.
This signal is installed to isolate a leaking steam steam line and prevent prevent further pressure increases in containment. (References 6.1.002, 6.4.007, 6.4.0 15, 6.4.036) 6.4.007,6.4.015,6.4.036)
- 3. Containment High-3 Pressure The high-3 containment building pressure detection circuitry functions to initiate a Phase B containment isolation and actuate the containment spray system. Spraying is necessary to prevent overpressurization of the containment structure and maintain structural integrity to limit the offsite dosage.
The high-3 pressure setpoint is 27 psig. An actuation signal occurs when 2/4 containment pressure channels exceed their setpoints. These are the only circuits which are energized to actuate, because inadvertent spray actuation is not desirable. The containment spray actuation shall utilize all four pressure channels in 2/4 coincidence logic. Only one channel should be bypassed at a time. An additional protection channel degraded by single failure criteria should not render the spray actuation inoperable. The 2/4 coincidence logic which is utilized for containment spray ensures the availability of the minimum protection channels for actuation. It meets the requirement of IEEE-279. Moreover, the reliability of the circuit is established by periodic testing. (References 6.1.002, 6.4.007, 6.4.015, 6.4.036)
- 4. Low Pressurizer Pressure A steam break accident or a LOCA A LOCA will cause a decrease in in pressurizer pressure due to the outsurge of water from the pressurizer. The outsurge outs urge will be caused caused by either the water loss from the LOCA or the the rapid cooldown produced by by the steam steam break. Safety Safety injection will be be initiated by low pressurizer pressure of of 1850 1850 psig, psig, asas sensed sensed byby 2/3 2/3 pressure pressure detectors.
detectors. This This meets meets thethe 2-29 2-29 Rev. 10 Rev. 10 I
FNP Units 1I &
&2 REACTOR PROTECTION SYSTEM A-lSlO07 A-18 1007 TABLE T ENGINEERED SAFEGUARDS ACTUATION SIGNALS SAFETY INJECTION ACTUATION (CONTINUED)
ACTUATION INTERLOCKS PROTECTION MODES OF FSD SIGNAL SETPOINT COINCIDENCE & BLOCKS PROVIDED FOR OPERATION SECTION High-i High-l containment 4 psig High-l 2/3 High-i None Loss of coolant or 1,2,3 1, 2, 3 2.4 pressure pressure signals ::: steam line break within 2.7.1 PB951B . setpoint containment 2.7.2 PB952B Fig. 2 PB953B ) . Sht. 8S Manual N/A 1/2 112 Momentary None Operator discretion 1,2,3,4 1, 2, 3, 4 2.7.1 switches Fig. 22
!!1;~
Sht. 8S Sht. T4-2 T4-2 Rev. 00 Rev.
10/8/2009 13: 10 13:10 FNP-2-EEP-0 FNP-2-EEP-O REACTOR TRIP OR SAFETY INJECTION Revision 33 II. The following are symptoms of a reactor trip:
- a. Any reactor trip annunciator lit.
lit
- b. Rapid decrease in neutron level indicated by nuclear instrumentation.
- c. All shutdown and control rods are fully inserted. Rod bottom lights are lit.
III. The following are symptoms that require safety injection. injection, if one has not occurred:
~
SI Signa:-) Signal Instrumentation Setpoint Coincidence (TSLB) 1.
- l. Pressurizer PI-455.456.457 P1-455,456,457 1850 psig 2/3 pressure low (TSLB2 (TSLE2 17-1.17-2.17-3) 17-1,17-2,17-3)
(If not blocked)
- 2. Steam Line PI-474.484.494.
P1-474,484,494, 100 psid 11 steam line Differential PI-475.485.495. P1-475,485,495, 100 psig less pressure PI-476.486.496 P1-476,486,496 than other two (TSLB4 10-2.10-3.10-4. 10-2,10-3,10-4, on 2/3 protection 11-2.11-3.11-4. 11-2,11-3,11-4, sets 12-2.12-3.12-4. 12-2,12-3.12-4, 13-2.13-3.13-4. 13-2,13-3.13-4, 14-2.14-3.14-4. 14-2,14-3,14-4, 15-2.15-3.15-4) 15-2,15-3,15-4)
- 3. Low Steam Line PI-474.485.496 P1-474,485,496 585 psig 2/3 pressure (TSLB4 (TSLE4 19-2.19-3.19-4) 19-2,19-3,19-4) (rate (If not blocked) compensated)
\
- 4. Containment PI-951.952.953 P1-951,952,953 44 psig 2/3 pressure high (TSLB1 (TSLE1 1-2.1-3.1-4)
,- 1-2,1-3,1-4)
- 5. Manual NIA N/A N/A 1/2 IV. The following are symptoms of a safety injection:
- a. Any SI annunciator lit.
- b. BYP && PERMISSIVE SAFETY INJECTION ACTUATED status light lit
- c. MLB-1 MLB-l 1-1 or MLB-1 11-1 lit
- d. HHSI flow greater than 00 gpm.
)() C C Page 4 of 38
FNP Units FNP Units 11 && 22 REACTOR PROTECTION REACTOR PROTECTION SYSTEM SYSTEM A-181007 A- 181007 Trip/Actuation Accuracy Trip/Actuation Accuracy - This
- This definition definition includes includes comparator comparator accuracy, accuracy, channel channel accuracy for accuracy for each input, and each input, and rack environmental effects.
rack environmental effects. This This is the tolerance expressed is the tolerance expressed in process in process terms terms (or (or percent percent ofof span) span) within within which which thethe complete complete channel channel shall shall perform perform its its intended trip/actuation function. intended function. This includes all instrument errors includes all errors but no process process effects effects such as streaming. streaming. 2.2 2.2 GENERAL FUNCTIONAL GENERAL FUNCTIONAL REQUIREMENTS-REACTOR REQUIREMENTS-REACTOR TRIP TRIP SYSTEM, SYSTEM, ENGINEERED SAFETY SAFETY FEATURES ACTUATION SYSTEM. SYSTEM. 2.2.1 The Reactor Protection System acts to limit the consequences of ANSI Condition II II events (e.g. loss offeedwater, of feedwater, etc.). Minimum departure from nucleate boiling ratio (DNBR) shall not be less than the designed limit DNBR as a result of any anticipated transient or malfunction for Condition II events and as such shall meet 95/95 criterion. That is, departure from nucleate boiling will not occur on at least 95 percent of the limiting fuel rods at 95 percent confidence. (DNBR design limit for vantage 5 fuel is 1.2411.23 1.24/1.23 for the typical and 1.25/1.24 for LOPAR fuel for the typical and thimble thimble cells and 1.2511.24 cells respectively). Rod linear Power density shall not exceed the rated design value (22.4 KW/ft.) and the stress limits of the reactor coolant system as specified for Condition II events (2735 psig) shall not be violated. Release of radioactive material for any Condition III fault shall not be sufficient to interrupt or restrict public use of those areas beyond the exclusion radius and shall not exceed the guidelines of 10 CFR 20. For any Condition IV TV fault, release of radioactive material shall not result in an undue risk to public health and safety nor shall it exceed the guidelines of 10 CFR 100, Reactor Site Criteria.
"Reactor Criteria."
The Engineered Safety Feature Actuation System, in addition to Reactor Trip, limits the consequences of ANSI Condition III events and mitigates ANSI Condition IV events. Table T-I T-l summarizes the ANS classification of faults and safety analysis approach as outlined in FNP-FSAR-Chapter
- 15. Most of the analyzed events/transients can
- 15. can be detected by one or more protection functions. The various Safety Analyses take credit for most of protection functions. Those functions for which no credit is taken in the protection the analyses are stillstill required to be operable to enhance the overall reliability reliability and diversity diversity of the protection system. This of the This design design shall shall meet meet the requirements the requirements of of IEEE-279-IEEE-279-1971.1971.
(References 6.1.008, 6.1.031, (References 6.1.031, 6.7.0 13, 6.7.039) 6.7.013, 6.7.039) 2-3 2-3 Rev. 66 Rev.
FNP Units FNP Units 1I && 22 REACTOR PROTECTION REACTOR PROTECTION SYSTEM SYSTEM A-181007 A-18 1007 2.2.2 2.2.2 The RPS The system isis housed RPS system housed inin two two physically physically and and electrically electrically independent independent equipment trains (Train "A" equipment trains (Train Ar and Train "B"), and Train B), typically t typically referred referred to to as as the the Solid State Solid State Protection Protection System System (SSPS) (SSPS) cabinets. cabinets. (Reference (Reference 6.7.003) 6.7.003) 2.2.3 single failure Any single within the RPS system (sensor failure within (sensor channel or or actuation prevent the redundant system actuation. On loss train) shall not prevent loss of channel or or train power the the bistable shall be tripped. shall be tripped. The The only only exception exception to to the the loss loss of channel of channel or or train power causing causing the bistable to to trip is is for for Containment Containment Spray and Containment Phase B B Isolation Isolation where the bistables must 6.1.002,6.1.41,6.7.014,) energize to actuate. (References 6.1.002, 6.1.41, 6.7.014,) 2.2.4 Actuation shall be automatic when the limits of the monitored parameters are exceeded. While one train is in test, the redundant train shall be Trip/ESFAS capable of performing Reactor Trip/ESF AS actuation. (References 6.1.002, 6.7.014, 6.7.015) 6.1.002,6.7.014,6.7.015) 2.2.5 The Reactor Protection System shall have provisions in the control room for manually initiating the reactor trip or actuating engineered safety features systems. (References 6.1.002, 6.7.014) 2.2.6 Instrument channels shall be powered from four separate independent AC instrument distribution panels. These panels shall be fed from four separate and independent Class IE 1E inverters. (References 6.1.023, 6.4.081, 6.4.091, 6.7.0 14, 6.7.016) 6.4.081,6.4.091,6.7.014,6.7.016) 2.2.7 Auxiliary devices that are required to operate on an ESF ESFASAS actuation to support train-related functions, shall be supplied from the same distribution panel to prevent the loss of electric power in one protection set from causing the loss of equipment in the redundant protection set. 6.4.08 1, 6.4.091) (References 6.1.023, 6.4.081, 2.2.8 Each distribution panel shall have access to its respective inverter and standby power supply. (References 6.1.023, 6.4.081, 6.4.091) 2.2.9 A protective action at the system level, once initiated, shall go to completion. Actuation is sealed-in until manually removed from operation. (Reference 6.7.014) 2.2.10 Each RPS actuation shall be alarmed and annunciated in the control room. (Reference 6.7.0 14) 6.7.014) 2.2.11 2.2.11 Protection Protection interlocks and bypasses shall interlocks and be designed shall be designed toto meet meet the the requirements of IEEE 279-1971 ofIEEE 279-1971 Sections Sections 4.12 through through 4.14. 4.14. (Reference (Reference 6.7.0 14) 6.7.014) 2-4 2-4 Rev. 4 Rev.4
FNP Units FNP Units 11 && 22 REACTOR PROTECTION REACTOR PROTECTION SYSTEM SYSTEM A-18!007 A- 181007 2.2.12 2.2.12 The RPS The shall be RPS shall capable of be capable of testing testing at at power power and and shall shall follow follow the the guide guide lines of Regulatory lines of Regulatory Guide Guide 1.22, 1.22, IEEE IEEE 338-1971, 338-1971, and and IEEE IEEE 279-1971 279-1971 Section 4.10. Section 4.10. (References (References 6.7.014, 6.7.014, 6.7.015, 6.7.015, 6.7.032) 6.7.032) 2.2.13 shall be tested either in "GO ESF shall GO TEST" TEST mode or in "BLOCK" BLOCK test mode. In "GOGO TEST" TEST mode mode the ESF device device is operated, or equipment alignments for special for special operation operation areare performed. performed. In In the "BLOCK" BLOCK testtest mode, where where the endend device device test would cause cause plant upset, the the end end device device actuation actuation is is ACTUATION signal is verified by continuity check. blocked while the "ACTUATION" Typical examples are Feedwater control valves, steam line isolation valves, RCP breaker test, etc. (References 6.7.003, 6.7.011, 6.7.032) 2.2.14 Electrical or mechanical interlocks and bypasses on safety-related equipment, when initiated manually or automatically, shall be continuously indicated in the main control room. No more than one train or channel may be bypassed at one time. (Reference 6.7.014) 6.7.0 14) 2.2.15 Except as noted below, all reactor trip and safeguards actuation channels shall be placed in the trip mode when the channel is out of service for any reason. The reactor trip and safeguards actuation circuits noted below may be administratively bypassed for maintenance on a single channel.
- 1. Source range high neutron flux trip
- 2. Intermediate range high neutron trip
- 3. High 33 containment pressure actuation of containment spray 6.1.004, 6.1.011, 6.1.022, 6.4.007, 6.7.014, 6.7.045, 6.7.046)
(References 6.1.004,6.1.011,6.1.022,6.4.007,6.7.014,6.7.045,6.7.046) 2.2.16 Channel independence shall be required throughout the system, extending from the sensor through the devices actuating the protective function. Redundant logic system cabinets shall be maintained and separated from the analog charmels. channels. Reactor trip and ESFAS ESF AS analog circuits may be routed in the same raceway if the circuits have the same power supply and sensor protection channel set. Cabinet separation criteria shall be verified by tests. (References 6.1.011, 6.1.042, 6.1.046, 6.4.081, 6.4.091, 6.7.001, 6.7.002, 6.7.003, 6.7.0 14, 6.7.054) 6.7.002,6.7.003,6.7.014,6.7.054) 2.2.17 2.2.17 The system The system shall shall have functional diversity. As an example, example, for a loss-of loss-of-coolant accident, safety injection signal accident, aa safety signal can can be obtained manually be obtained manually oror by by automatic initiation from two automatic two diverse parameter parameter measurements. These These are: are: 1.1. Low pressurizer pressure pressure 2-5 2-5 Rev. 4 Rev.4
FNP UNIT 1I FNPUNIT LOAD LIST LOAD A-506250 DGO3 DG03 EEO5 EE05 LBO6 LB06 1D 120V VITAL AC DIST PNL ID AB-121' AB-121 D177025 BKR TPNS DESCRIPTION SEE PAGE Q121LOOO1D-4 It) 120V VITAL AC INST DISTRIBUTION PANEL CR 4 1D 12LEQOO9D-4 MAIN BREAKER <<< INVEP.ER it) <<< LBQ6 MAIN (PREFERRED) OR BKR iR4 (ALT) ON 1R 208/120v REG AC DIST PANEL iD-Cl Q1H11NGNIS25O3D-4 it) NUCLEAR INSTRUMENTATION SYSTEM CABINET CHANNEL 4 1D-02 10-02 Q1H11NGNIS2503D-4 Q1H11NGNIS25O3D4 10 it) NUCLEAR INSTRUMENTATION SYSTEM CABINET CHANNEL 44 Q1H11NGPIC25O5D-4 PROCESS I&C PROTECTION CAB #4 CHANNEL 4 SD-03 Q1H11NGPIC25050-4 10-03 G-40 IG-40 10.,...04 10-05 1D-06 10 N1C56LOO01B-N 1D-05 N1C56L0001BN Q1H11NGPIC25O5H-4 PROCESS* 1D-04 Q1H11NGPl:C2505H-4. PROCESS I&C CONTROL.CAB#8 1B lB INCORE T CONTROL CAB #8 CHANNEL 4 >>> T/C COLD REF CONN BOX N1G22NBWPP26O3C-N WASTE GAS PROCESSING PANELS >>>
.. 06 N1.G22NBWPP2603C-N >>>
>>> G-41 G-43 II 10-07 1D-07 Q1H11NGSSP2506K-A RAIN A Q1H11NGSSP25O6K-A TRAIN A SOLID STATE PROT SYSTEM INPUT CAB CH 4 4 10-08 lD-08 Q1H11NGSSP2506G-B Q1H11NGSSP25O6G-B TRAIN B B SOLID STATE PROT SYSTEM INPUT CAB CH 4 4 1D-09 ------------
10-09 SPARE 10-10 N1H11NGAR2506E':'B 2506E-B RELAY RACK TRAIN B >>> AUX .RELAY >>> G-44 CB25OOA- MAIN CONTROL . BOARD .SECTION. A>>> SECTION A G-45
-N 1A CONTROL BOARD DEMULTIPLEXER CABINE CABINET I 1D-13 Q1H11NGASC25O6C-B TRAIN TRAmB B AUX SAFEGUARD CAB >>> G-46 N1G12NBWPP26O3A-N BORON WASTE GAS PROCESSING r.n.l: * .I:a.L.I >>>
PANEL >>> G-47
- 603B-N LIQUID WASTE PROCESSING PANEL >>> >>> G-48 SPARE 1lsectg.doc sectg.doc Page Page G - 38
- Rev 66
FNPUNIT 1 H C HI LOAD LIST HI Cl) H A-S062S0 DG03 EE05 ~~~~~~-;~ id CM z AB-l2l' D177025 BKR TPNS Z CM DESCRIPTION C z
/
F LU t~--
~-~=:~~~
i- LU rfl) r i c Q1H11NGSSP25061 TRAIN B SOLID STATE PROTECTION SYS OUTP~' 10-16 ci 0 U) z (I: U) 0-i 0 H H z U) 0 HI 0 H i:t: U) U-i H H U) FH 0 z U) > U) 0 0H 0 H 0
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U) U) F:I U) U) H 0 0 U) 10-17ID i- r
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w I I I I I I I I I I I iI Ci 10-18ID Q1H11NGSSP2506J TRAIN B SOLID STATE PROTECTION SYS SAFEGUARDS ci U) z U) Lf) 0 LU HH I-HH U) 0-i 0 U) HI 0 H U) R1 U) H 0-i H 0 U) H 0 z U) U) U) U) 0 R U) 0
-B TEST CAB 0 U) f 1 ,+
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1sectg.doc 0 -q C) Page G - 39 Rev 6
- 21. 015AK2.10 015AK2.1O 001/NEW/RO/C/A OO1/NEW/RO/C/A - 2.Bl2.8/N/N/2ICVRIY
- 2.812.8/N/N/2ICVR/Y Unit 1I is at 20% power and conditions are as follows:
At 1000: IA 1A lB 1B IC 1C
- RCP amps: 670 680 690 At 1005:
1A lB 1.!2 IC 1C
- RCP amps: 670 680 0
- EF3, 1 1C C RCS LOOP FLOW LO OR 1 1C C RCP BKR OPEN, is in alarm.
Which one of the following describes the expected indications on 1A IA RCS LOOP and IC 1 C RCS LOOP flow rates at 101 1010? O? IA 1 A RCS LOOP Flow rate 1IC C RCS LOOP Flow rate A. 105% and stable (-- 0% and stable B~ B 105% and stable 10% and stable C. 100% and stable 10% and stable D. 100% and stable 0% and stable Page: 56 d d 200 12/1412009 1211412009
A - Incorrect. The first part is correct (See B). The second part is incorrect, but plausible, since the RCP amps are 0, and it is tripped as indicated by the bkr light werent for reverse flow caused by the discharge pressure of the and amps. If it weren't other two pumps 0% would be correct. BB - Correct. Each of the two loops with forced flow provide some backflow through the tripped pump (approx 5% each for a total of 10%). The tripped pump has an indicated flow (approximately 10%) due to the flow indicator sensing a positive value of flow >0, even though the direction of flow is reversed. C - Incorrect. The flow in the 1A
- IA & 1lB B loops is greater than 100%, but this is plausible since the amps in the 1A IA & 1I B pumps are unchanged. The increased flow is due to decreased resistance to flow downstream of these pumps which is why flow increases with no increased amps to the pump motors. The second part is correct (see B).
oD - Incorrect.
- Both parts incorrect but plausible. This would seem to be indicated by the amps and the fact that the 1 ICC pump is not pumping any flow due to being tripped by annunciator indication. However the piping system of the RCS allows back flow into the C loop in this condition and it is indicated on all three loop flowmeters.
Ran on simulator laptop to verify flows. No technical document was found which stated this characteristic in writing. Loss Of Reactor Coolant Flow, OPS-62520D OPS-52520D, Student Text- TextVersion Version 2, listed this. Cvr cvr 8-4-09 Previous NRC exam history if any: 015AK2.10 015 Reactor Coolant Pump Malfunctions AK2. Kncmledge of the interrelations interrdations between bween the Reador Reactor Coolant Codant Pump MMalfunctions (Los of RC Flc:m) alfundions (Lex:;s FIcw) the fdkming: (CFR 41.7/45.7) and thefollc:ming: 41.7145.7) A K2.1 0 RCP iindicors AK2.1O ndi cators end and control 2.8* 2.8 controlss . . . . . . . . . . ............................. 2.8* Match justification: This question provides indications which accompany a RCP trip and must be recognized as such. The knowledge of the interrelations between the RCP loss of flow and the flow indicators (and what to expect the RCP flow indicators to read after a RCP trip) must be used to obtain the correct answer. Objective:
- 2. RELATE AND IDENTIFY the operational characteristics including design features, cacities and protective interlocks for the components associated with the Reactor capacities Coolant System (RCS) to include the components found on Figure 1, Reactor Coolant System (OPS-40301A02).
Page: 57 of 200 12114/2009 1211412009
will effectively will effectively increase increase thethe unaffected unaffected SG'sSGs steaming steaming rate rate and and power power output output to to compensate compensate for for the reduced the reduced steaming steaming rate from the rate from affected SG. the affected SG. During the transient, the affected SG SG main feedfeed regulating or bypass feed feed regulating valve controller(s) valve controller(s) isis in automatic. Initially, in automatic. Initially, the the regulating valve valve will will go go shut shut in in response response to the the reduced steam reduced steam flow flow from from the affected affected SG. SG. This This causes causes SG SG level level in in the affected affected SGSG to begin to fall. Contributing to the drop in SG level is the phenomenon of SG SG shrink. Shrink is caused by the density change in the tube section of the SG, which occurs as a result of the drop in temperature on the primary side of the affected SG. The drop in indicated level causes the feed valves to reopen fully in an attempt to bring SG level back up. Overfeeding of the affected SG can occur, which could lead to a turbine trip and SG feed pump (SGFP) trips followed by a reactor trip. The turbine trip and SGFP trip occur at 82% SG level (P-14). To minimize the effects a loss of coolant flow has on the affected SG level, the operator is instructed to take manual control of the affected SGs feed regulating valves. Another concern on any loss of coolant flow situation is pressurizer pressure control. Automatic control of pressurizer pressure will be affected due to the loss of spray flow if the loss of coolant flow occurs in loops A and/or B. If only one loop is involved, the affected loop's ioops spray valve controller should be placed in MANUAL and the valve closed to prevent spray flow from the unaffected ioop loop bypassing the pressurizer. If both A and B B loops are affected, auxiliary spray flow should be utilized if normal letdown is available. The loop flow indications observed by the operators would be as follows: For the affected loop, flow would slowly decrease to 00 and then return to approximately 10%; for the unaffected loops, the flow should increase to approximately 1105% 05% (each loop). The flow indication in the idle ioop loop occurs as flow stops and then begins again in the reverse direction. Since flow rates in the RCS loops are derived from the differential pressure felt in an elbow in each ioop, loop, any flow at all will be be indicated, regardless of the direction. The indication indication observed in the two ioops loops with the running pumps is due simply simply to the pumps in those loops picking up a small portion of small portion of the flow lost lost in in the idle loop. the idle loop. 22 OPS-6252001525200 - Version OPS-62520D/52520D Version 22
- 22. 022A 1.01 001/NEW/RO/MEM 001/NEW/RO/M EM 3.6/3.7/N/N/3/CVRIY 3.6/3.7ININI3ICVRIY 100%, and the following conditions occurred:
Unit 2 is at 100%, At 1000: 1000:
- The Containment Cooling system is in the normal mode of operation per SOP-i 2.1, Containment Air Cooling System.
SOP-12.1,
- Containment temperature is slowly rising.
At 1100:
- The crew has configured the containment cooling system per SOP-12.1.
SOP-i2.1.
- The emergency service water from CTMT coolers; MOVs 3024A, B, C and 0Dare SOP-I 2.1.
are OPEN lAW SOP-12.1. Which one of the following identifies the:
- 1) MINIMUM temperature at which a Technical Specification action statement must be entered for Tech Spec 3.6.5, Containment Temperature, and
- 2) the speed of the Containment Cooling Fans lAW SOP-12.1?
(1)) (1 (2) A. 110°F FAST B. 110°F SLOW C C~ 120°F FAST D. 120°F SLOW Page: 58 of 200 1211412009 1211412009
AA - Incorrect.
- Incorrect. The 110°F 110°F is is incorrect incorrect for for Ctmt Ctmt temp temp limit, limit, but but is is plausible, plausible, since since 110°F 110°F isis the temperature per per the the CTMT CTMT HI HI TEMP TEMP alarm alarm ARP-1.2, ARP-1 .2, BB3, BB3, to to start start all all ctmt ctmt dome dome recirc fans in recirc in fast speed. The second part is is correct per per SOP-12.1 SOP-12.1 step step 4.1.6 4.1.6 note.
note. BB - Incorrect.
- Incorrect. The first part part is is incorrect incorrect (See (See A). TheThe second part part isis incorrect incorrect butbut plausible, since Slow is the speed that the fans automatically shift to in a very high temperature LOCA environment. However, it is due to the humidity rather than the heat that they are shifted to slow to protect the fans. Under normal conditions, the lower humidity allows fast speed operation to remove more heat from containment.
C - Correct. 120 degrees F is the TS 3.6.5 limit. Fans must be operated in fast per note prior to step 4.1.6 normally to maintain ctmt less than this limit per SOP-12.1, Step 4.1.9-4.1.11, ver. 37.0. oD - Incorrect.
- The first part is correct (See C). The second part is incorrect (See B).
ARP-12, 883, ARP-1.2, BB3, CTMT AI Al R TEM TEMP HI, Version PHI, Verson 44.0 440 OPERATOR OPERA TOR ACTION containment
- 5. IF contai a/ere ai nment average airr temperature is greater than 110° 110°F, THEN F, TH verify EN veri containment fy contai nment dome recirc fans in service on fast speed.
Previous NRC exam history if any: 022A1 .01 022A1.01 Containment 022 Contai Cooling nment Cool i ng System Al A1 Ability to predict and/or monitor changes andlor chancjesin paramders(to in parameters preient exceeding design (to prevent degn limits) Iimit asariated adated contrds induding: (CFR: 41.5/45.5) with operating the CCS controls A1.O1 Contanment tnpure ......................................... 3.63.7 A1.01 Containmenttemperaure 3.6 3.7 Match justification: Question asks what the TS containment temperature limit is, and which controls of the Containment cooling system must be operated (fast or slow speed fans) to prevent exceeding the containment temperature limit. Objective:
- 7. DEFINE AND EVALUATE the operational implications of normal I abnormal plant or normal/abnormal equipment conditionsassociated conditions associated with thesafeoperation the safe operation of theContainment the Containment Ventilation and Purge System components and equipment, equi pment, to include i ncl ude the following foil owi ng (OPS-40304A07):
(OPS-40304A07) :
- Normal control methods Abnormal and Emergency Control Methods Automatic actuation including setpoint setpoint (example SI, SI, PhaseA, Phase A, Phase B, MSLIAS, MSLlAS, LOSP, SG leIeI)
SO level) Protective Protecti ve isolations hi gh flow, low i sol ati ons such as high low pressure, low Ilevelevel including i ncl udi ng setpoi nt nt Protective Protecti ve interlocks i nterl ocks Page: Page: 59 59 of ci 200 200 1211412009 1211412009
Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be :<::; 120°F. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air A.1 Restore containment 8 hours temperature not within average air temperature limit, limit. to within limit. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. ~ B.2 Be in MODE 5. 36 hours SURVEILLANCE_REQUIREMENTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within 24 hours limit. Farley Units 1I and 22 3.6.5-1 Amendment No. 146 146 (Unit 1) 1) Amendment No. 137 137 (Unit 2)
09/24/06 13:00:16 09/24/06 13:00:16 FNP-2-ARP-l.2 FNP-2-ARP..1.2 LOCATION LOCATION BB3 BB3 OPERATOR ACTION OPERATOR ACTION (continued) (continued) 4.
- 4. IF possible, THEN fl possible, THEN start start additional additional fan fan coolers coolers to to insure insure that that the the containment average containment average air air temperature temperature does does not not exceed exceed 120°F.
120°F. 5.
- 5. IF containment IF containment average average airair temperature temperature isis greater greater than than 110°F, 110°F, THEN THEN verify Containment Containment Dome Recirc Fans in service on fast speed.
- 6. FNP-2.SOP-12.1, CONTAINMENT AIR COOLING Refer to FNP-2-S0P-12.1, COOLESIG SYSTEM, Q2P16M0V3024A (B, C, D), EMERG SW for guidance on operating Q2P16MOV3024A FROM 2A (B, C, D) CTMT CLR.
- 7. Refer to the Technical Specifications, LCO 3.6.5, for checking Containment average air temperature and LCO Requirements.
References:
A-207100, A-207100, Sh. Sh. 98; 98; B-205968; B-205968; D-205010, D-205010, Sh.Sh. 1; 1; Technical Technical Specifications; Specifications; PCN PCN B-862-3923; B-86-2-3923; MDFD MDFD 88-1907; 88-1907; PCN PCN B91-2-7619; B91-2-7619; PCN PCN B91-2-7574, B91-2-7574, U-280362 (Vendor (Vendor manual for N2T12TRSH3 N2T12TRSH3188) 188) Page Page 22 of 2 of2 Version 31.0 Version 31.0
07/02/09 06:48:57 07/02/09 06:48:57 FNP-2-S0P-12.1 FNP-2-SOP-12.1 NOTE: NOTE: 2A CTMT CLR 2A CTMT CLR fan fan fast fast speed speed breaker EA-1O-2 is breaker EA-10-2 is interlocked interlocked with with tie tie breaker breaker EAO8-2 to EA08-2 prevent starting to prevent starting ifif the the emergency emergency section section of of 2A 2A 600V 600V LC LC isis aligned aligned to to 2D 600V LC 2D 600V (EAO8-2 open). LC (EA08-2 DCP B87 open). DCP B87-2-4592 4592 4.1.6 Start 2A, 2B, Start 2B, 2C, 2C, and containment coolers and 2D containment coolers in in FAST (SLOW) (SLOW) speed. speed.
** 2A2A containment cooler
** 2B2B containment containment cooler cooler
** 2C containment cooler 2C containment cooler
** 2D containment cooler 4.1.7 Check CTMT CLR 2A(2B,2C,2D) DISCH 3186A(B,C,D) OPEN light illuminated.
** CTMT CLR2A CLR 2A DISCH 3186A 31 86A
** CTMT CLR 2B DISCH 3186B 31 86B
** CTMT CLR 2C DISCH 3186C
** CTMT CLR 2D DISCH 3186D 4.1.8 Place 2A, 2B, 2C, and 2D CTMT DOME RECIRC FANS in HIGH (LOW) speed.
** 2A CTMT DOME RECIRC FAN
- 2B CTMT DOME RECIRC FAN
- 2C CTMT DOME RECIRC FAN
- 2D CTMT DOME RECIRC FAN 4.1.9 Operate the containment dome recirculation fans and containmçj containment coolers asasnecessmain necessary to maintain tain containment temperature below 120°F._(See "i20°F. (See section 4.7 for shifting containment cooler fan speeds.)
4.1.10 Open 2A and 2B RX CAV CAV CLG DMPRS
- 2A RX CAV CA V CLG DMPR Q2E12HV3999A
- 2B RX CAV CA V CLG DMPR Q2E12HV3999B 44 Version 24.0 Version 24.0
07/02/09 06:48:57 07/0210906:48:57 FNP-2-S0P-12.1 FNP-2-SOP-12.1 CAUTION: :tl1MERG.SW AUTION! EMERG SW FROM FROM 2A(2B,2C,2D) 2A(2B,2C,2D) CTMT CTMT CLRvalves CLR valves
- , .. Q2P16M0V3024A(B,C,D) [Q2P16V043A(B,C,D)1arenormally Q2P16MOV3024A(B,C,D) [Q2P16V043A(B,C,D)] are normally maintained maintained closed, BUT maybe cl()s~,BUT may be opened opened fortemperatnre for temperature control.
control. However,operation However, operation with these witli these valves valves open should be open s.hould be minimizedtorednce minimized to reduce the the potential.for potential for long long t~rm term degradation to the containment coolers from from the higher higher fl()w flow rates rates..
.Tl1er~f(}~~,t~eOperations Therefore, the Operations Shift Shift Manager should .be consulted prior prior toto opening these opening these valves.
valves. 4.1.11 IF necessary to IF to maintain containment temperature below 120°F, 120°F, THEN open and and caution tag EMERG SW FROM 2A(2B,2C,2D) CTMT CLR valves, as desired.
** EMERG SW FROM 2A CTMT CLR VALVES Q2P16MOV3024A Q2P16M0V3024A
** EMERG SW FROM 2B CTMT CLR VALVES Q2P16MOV3024B Q2P16M0V3024B
** EMERG SW FROM 2C CTMT CLR VALVES Q2P16MOV3024C Q2P16M0V3024C
** EMERG SW FROM 2D CTMT CLR VALVES Q2P16MOV3024D Q2P16M0V3024D 4.1 .12 4.1.12 WHEN additional service water flow through the containment coolers is no longer necessary for temperature control, THEN remove caution tag(s) and close EMERG SW FROM 2A(2B,2C,2D) CTMT CLR valves:
** EMERG SW FROM 2A CTMT CLR valves Q2P16MOV3024A Q2P16MOV3O24A
** EMERG SW FROM 2B CTMT CLR valves Q2P16MOV3024B Q2P16M0V3024B
- EMERG SW FROM 2C CTMT CLR valves Q2P16MOV3024C Q2PI6MOV3O24C
** EMERG SW FROM 2D CTMT CLR valves Q2P16MOV3024D Q2P16MOV3O24D 55 Version 24.0 Version 24.0
- 23. 022AK3.06
- 23. 022AK3.06 001/NEW/RO/MEM OO1/NEW/RO/M EM 3.213.3/N/N/4/CVRIY 3.213.3ININI4ICVR/Y Unit Unit 1I Reactor Reactor has has just tripped, and and the the following following conditions conditions occurred:
occurred:
** RCPs have All three RCPs have just tripped.
** Charging has All Charging has been been lost.
lost. Which one of the following correctly states the reason for maintaining CCW cooling flow to the Thermal Barrier HX in this condition? Maintaining CCW cooling flow to the Thermal Barrier HX will prevent the RCP (1) from starting to degrade due to overheating in as early as (2) (1 (1)) (2) A. #1 seal 2 minutes B B~ #1 seal 13 minutes C. lower radial bearing 2 minutes D. lower radial bearing 13 minutes A - Incorrect. Incorrect, since the seal area takes time to void of the cooler water prior to allowing the hotter RCS water to the seal area. The WOG background document for ECP-O.O ECP-0.0 gives 13 minutes for the time to degrade the RCP #1 seal after losing CCW to the thermal barrier and seal injection. The #1 seal degrading is the correct concern and reason, but the 2 minutes is incorrect. Plausible, because with loss of CCW to the motor oil coolers the upper and lower MOTOR bearing (but not the radial bearing) can overheat in a maximum of 2 minutes per UOP-1.1 Step 3.10 (P
& L). And, the lower RADIAL bearing would heat up in the event that both Seal Injection and CCW Thermal Barrier cooling were lost, but that is not the limiting concern or reason.
B B - Correct. The WOG background document for ECP-0.0
- ECP-O.O gives 13 13 minutes for the time to degrade the RCP #1 seal after losing CCW to the thermal barrier and seal injection.
C C - Incorrect. The bearing is incorrect, since it is the #1 seal that is the limiting concern. Plausible, since the bearing will heat up in the condition given without CCW cooling to the Thermal Bearing, but the #1 seal is the limiting condition. The time is incorrect also, since the 13 13 minutes is given in the WOG background document for ECP-O.O. Plausible, since the 22 minutes would apply to an overheat RCP manual ECP-0.0. trip criteria in 22 minutes or less for aa lowerlower motor bearing if motor oil CCW cooling is lost (but not for lower RADIAL bearing). D - Incorrect. The first part is incorrect D - incorrect (see C). The second part is is correct (see B). B). WOG Background Document FNP-O-ECB-O.O, LOSS OF Background Document OF ALL AC POWER, POWER, Plant Spedfic Specific Background Information (pgs39 & 40 40 of ci 88). 88). Page: 60 of Page: 60 200 cl200 1211412009 1211412009
Isolating the loolating the RCPthermai RCP thermal ba-rier barier CCW CCW raurn rsturn outside outside containment containment ioolation isolation valve valve prepares prepares thethe plant plant for for recovery whi recovery while protecting Ie protecti the CCW ng the CCW system system from stearn steam formati formation due to on due to RCP RCP thermal barrier thermal ba-ri er heati heating.ng. Followingtheloss Foil owi ng the loss of of all all ac ac power, power, hot hot reactor reactor cool coolant willII grcduall ant wi graiually replacethe y repl ace the normal normally Iy cool cool sealseal injection water in injection in the the RCP RCP sealseal a-ea. area. As Asthe hot reactor the hot reactor coolant coolant leaks leaks up up the shaft, the the shaft, water in thewater in thethe thermal ba-rier barrier will heatheat up up and and potentially potentially form stearn steam in inthethermal barrier and the thermal ba-rier and in in the CCW lines theCCW lines aij acent to the thermal ba-ri cdj barn er. SubSffluent Subuent automati automaticc start of the CCW COW pump woul would d del d iver iver CCW flow to the thermal ba-ri barrier, er, flflushing steam into the CCW system. IIff abnormal RCP seal leakage hcd ushi ng the stearn had deial oped in a pump, the abnormally high Iffi<age developed leakage rate could exceed ceai the cooling capacity of the CCW COW flow to that pump thermal ba-ri barrier er and tend to generate more stearn steam in the RCP thermal ba-ri er CCW barrier COW rsturn lines. 1001 raurn Isolating th lines prevents ati ng these preients the potenti potential aI iintroduction ntroducti on of thi thiss stearn steam into the mai mainn porti on of portion the CCW system upon CCW COW pump start. Thi Thiss keeps the ma mann porti portion on of the CCW COW system em avai aiai II abl ablee for cooling cool equipment i ng qui necessary for recoveri pment necessa-y recovering ng the piplant ant when ac power is restored.rorer1. Knowledge: 1. RCP KnOlNledge: RCPseal intrity concernsfollowing seal integrity concemnsfollowing loss lossof of ac power (See(SeeSubsection Subsoction 2.1).
- 2. Analyses Analysesof of RCPseaI performancefollowing aalossof loss of all seal cooling esti estimatethat mate that increased seal leakage may begin as early as 13 minutes due to seal degradation at high fluid temperatures. temperatures It is establish important to establ suffident ish suffi backpressurein ci ent backpressure in the seal Ilealcoff ffi<off line by iisolating rsturn line 001 ati ng the seal raurn degradation occurs in order to limit RCP seal leakage. Thetirne before seal degrcdation The time of 13 minuteswas minutes was dsterminal in WCA daermined WCAP-10541 P-1 0541 as thetime when "...the asthetimewhen internalsvdumewill lower pump internals
... theIOlNer volume will be compIely purged and the seal area water temperature will be approaching the 5500F canpletely 550°F reactor codant temperature.
coolant temperature." FSD, CVCSlHHSl/ACCUMULATORlRMWS, CVCHHSI/ACCUMULATORIRMWS, A-181009 2.2.3.2 capability This capabil satisfiesthe ity satisfies requirement the seal water requi rement for the RCP No. II seal. A portion of the seal injection flow (nominally 5 gpm per pump) enters the RCS through the labyrinth seals and the thermal barrier. This in-leakage precludes leakage of reactor coolant through the No. II seal during duri operation. ng normal operati on. The remai nder of the seal iinjection remainder (nominally nj ecti on flow (nomi nally 3 gpm) flows up the pump shaft, cooling the pump lower bearing and the No. II seal. The 5-micron filtration requirement is based upon RCP minimum seal clearances. (Reference 6.2.44) FNP-1-SOP-1.1, Veron Version 40.0 CCW flow to the RCPmotor 3.10 IF CCWfiow RCP motor bearing oil coolers is lost, THEN pump operation may be continued conti nued untilunti I the motor upper or lower bearing beari ng temperature reaches 195 195°0 FF (approximately 22 minutes after cooling water flow stops). 61 c Page: 61 Page: c:I 200 12/14/2009 1211412009
Previous NRC Previous NRC exam exam history history ifif any: any: 022AK3.06 022AK3.06 022 Loss 022 Loss of of Reactor Rctor Coolant Coolant Makeup Makeup AK3. KnCNVIedge AK3. Knor4edge of of the the reaSXlsfor reansfor thefollCMIing thefdlcming responses responsasth, apply to as they apply the Los; to the Lo ofof Reador Reactor Coolant Coolant M akeup: (CFR Makeup: (CFR 41.5, 41.10! 45.6! 45.13) 41.5, 41.10/45.6/45.13) AK3.06 RCPthermai AK3.06 RCPtherm bcrrier cooling ................................ 3.2 barier cooling 3.2 3.3 3.3 Match justification: To Match To correctly answer this correctly answer question, knowledge this question, knowledge is is required required ofof the the reason for requiring reason for requiring RCP RCP thermal thermal barrier barrier cooling cooling during during aa loss loss of of all all RCS RCS makeup makeup (which would include a loss of seal injection). Objective:
- 2. RELATE AND IDENTIFY RELATE IDENTIFY the operational characteristics including design features, characteristics including cadties and protective interlocks for the components associated with the Reactor capacities Cool ant System (RCS) to include the components found on Figure Coolant Figure 1, Reactor Coolant Cool ant (OPS-40301A02).
System (OP&40301A02).
- 4. LABEL AND IILLUSTRATE LLUSTRATE Reactor Coolant System (RCS) flow paths pathsto to include the components found on Figure 1, Reactor Coolant System (OPS-40301A05).
Page: 62 Page: of 200 62 of 200 12!14(2009 1211412009
06/27/07 16:11:08 06/27/0716:11:08 FNP-O-ECB-O.O FNP-0-ECB-0.0 LOSS OF ALL AC POWER Plant Specific Background Information Plant Section: Procedure Unit 11 ERP Step: 8 Unit 2 ERP Step: 8 ERG Step No: 8 ERP StepText: Isolate RCP seals using ATTACHMENT 3. Step Text: ERG StepText: Dispatch Personnel To Locally Close Valves To Isolate RCP Seals And Place Valve Valve Switches In CLOSED Position
Purpose:
To isolate the RCP seals Basis: This step groups three actions, with different purposes, aimed at isolating the RCP seals. The actions are grouped since all require an auxiliary operator, dispatched from the control room, to locally close containment isolation valves (the reference plant utilizes motor operated valves for the RCP seal return, RCP thermal barrier CCW return lines and RCP seal injection lines). This grouping assumes that the subject valves are located in the same penetration room area and that they are accessible. Concurrent with dispatching the auxiliary operator, the control room operator should place the valve switches for the motor operated valves in the closed position so that the valves remain closed when ac power is restored. Isolating the seal return line prevents seal leakage from filling the volume control tank (VCT) (via seal return relief valve outside containment) and subsequent transfer to other auxiliary building holdup tanks (via VCT relief valve) with the potential for radioactive release within the auxiliary building. Such a release, without auxiliary building ventilation available, could limit personnel access for local operations. Isolating the seal return line also enables pressure in the number 11 sealleakoff seal leakoff line to increase up to the relief valve setpoint of 150 psig. seal leakoff line of at least 150 psig enables development Maintaining a backpressure in the sealleakoff seal leakoff cavity with a steady-state seal leakage rate of high pressure in the number 11 sealleakoff established due to the self-limiting leakage characteristic of the number 11 seal. Under these conditions, with the number 11 seal functioning as expected and the number 2 seal remaining gpmlpump. This is consistent with the steady-closed, the expected leakage flow rate is 21.1 gprn/pump. state pressure distribution and seal leakage determined in the WCAP-10541 analysis and used in the latest RCP seal leakage PRA model in WCAP-15603. Isolating the RCP seal injection lines prepares the plant for recovery while protecting the RCPs from seal and shaft damage that may occur when a charging/SI pump is started as part of the recovery. With the RCP seal STEP DESCRIPTION TABLE FOR ECA-0.OStep8 injection lines isolated, a charging/SI pump can be started in the normal charging mode without concern for cold seal injection flow thermally shocking the RCPs. Seal injection can subsequently by established
~
toJhe..RCP-consistenLwi1h.appropriate-plant.specific.prGGedw:es( Isolating the RCP thermal to rarrier CCW return outside containment isolation valve prepares the plant for recovery while (barrier protecting the CCW system from steam formation due to RCP thermal barrier heating.
\ !rotecting Following the loss of all ac power, hot reactor coolant will gradually replace the normally cool seal injection water in the RCP seal area. As the hot reactor coolant leaks up the shaft, the water in the thermal barrier will heat up and potentially form steam in the thermal barrier and in the CCW lines adjacent to the thermal barrier. Subsequent automatic start of the CCW pump would deliver CCW flow to the thermal barrier, flushing the steam into the CCW system. If abnormal RCP seal leakage had developed in a pump, the abnormally high leakage rate could exceed the cooling capacity of the CCW flow to that pump thermal barrier and tend to generate more steam in the RCP thermal barrier CCW return lines. Isolating these 39 of 88 Version: 1.0
06/27/07 16:11:08 06/27/0716:11:08 FNP-O-ECB-O.O FNP-0-ECB-0.0 LOSS OF LOSS OF ALL ALL AC AC POWER POWER Plant Specific Plant Specific Background Background Information Information Section: Procedure lines prevents the potential introduction of this steam into the main portion of the CCW L system upon CCW pump start. This keeps the main portion of the CCW system available for cooling equipment necessary for recovering the plant when ac power is restored. I 1) Knowledge: _ . RCP seal inte ritnt concerns followin loss of ac power (See Subsection 2.1). 2. Analyses of R seal per o~ance ormance following a loss of all seal cooling es ~sed seal 1 a e tlTfiniised estiTate leakage may begin as early as 13 minutes due to seal degradation at high'ffi:iid1eiiij?efatures. It
'lSTriijJ6rtant establish sufficient backpressure in the se~off Thiiiiortant to establiSh seal leakoff line by isolating the seal leakage._The return line before seal degradation occurs in order to limit RCP seal leakage. The time of 13
,_minutes was determined in WCAP-1 0541 as the time when" ... the lower pump internals volume will be completely purged and the seal area water temperature will be app~_
the 550°F reactor coolant temperature." .). Time-critical actions of this step (for example, local seal return isolation) may be located earlier in the guideline if necessary to meet individual plant capabilities and requirements. Note that the Step Sequence Requirements allow interchangeability between Steps 6, 7 and 8 of this guideline.
References:
Justification of Differences: 1 Placed actions in an Attachment to allow an extra operator to perform required actions outside of the control room without interfering with the flow of the procedure. 40 of88 400f88 Version: 1.0 1.0
01/03/08 12:40:36 01/03/08 12:40:36 FNP-1-SOP-1.1 FNP-I-S0P-l.l 3.6 3.6 12Q NOT attempt DO attempt to to start start aa RCP RCP unless its oil unless its oil lift lift pump pump has has been been delivering delivering oiloil to to the upper the thrust shoes upper thrust shoes for at least for at least two minutes. Observe two minutes. Observe the the oil oil lift lift pumps pumps indicating lights indicating lights to to verify correct oil verify correct pump motor oil pump motor operation operation and and oil oil pressure. pressure. TheThe oil lift pumps should oil lift pumps should run at least run at least 11 minute minute after after the RCPs are the RCP's are started. started. An An interlock interlock starting aa RCP until 600 will prevent starting 600 psig oil pressure is is established. established. 3.7 3.7 Shift Supervisor's Shift Supervisors approval approval must must be obtained prior prior to to removing removing any any seal wires or changing the changing the position position of any throttle of any valves. throttle valves. 3.8 RCP seal water injection flow flow of 6 gpm or CCW to the RCP thermal barrier must continuously supplied be continuously supplied when RCS temperature exceeds exceeds 150°F. 150°F. 3.9 Maintain RCP CCW and seal injection water supply temperature less than 105°F 105°F f 7( and 130°F respectively. f CCW flow to the RCP motor bearing oil coolers is lost, THEN pump operation IF may be continued until the motor upper or lower bearing temperature reaches V 195°F (approximately 2 minutes after cooling water flow stops). 3.11 For RCP operations, a minimum pressure differential of200 of 200 psid must be maintained across RCP No. 1I seals. 3.12 The following precautions apply in the case of a RCP ##11 seal failure. 3.12.1 DO NOT restart an RCP with an indicated No. No.11 seal failure. 3.12.2 FNP-1-ARP-1.4, MAIN CONTROL BOARD Refer to FNP-I-ARP-l.4, ANNUNCIATOR PANEL "D", D, for guidance if No. 1I sealleakoffflow seal leakoff flow is abnormally low (Ann. DCI) DC 1) or abnormally high (Ann. DC2). 3.13 The No. I1 seal bypass valve should NOT be opened unless either the pump bearing temperature (seal inlet temperature) or the No. 1I seal sealleakofftemperature leakoff temperature approaches its alarm level. The No. 1I seal bypass valve should then be opened only if all of the following conditions are met: 3.13.1 Reactor coolant system pressure is greater than 100 PSIG AND less than 1000 1000 PSIG. 3.13.2 3.13.2 No. II seal leakoff valve is open. sealleakoffvalve 3.13.3 3.13.3 No. No. 1I seal leakoffflowrate sealleakoff flowrate isis less less than oneone GPM. GPM. 3.13.4 3.13.4 Seal Seal injection injection water water flow raterate to to each each pump pump is is greater greater than than six six GPM. GPM. 3.14 3.14 For For RCP operations, operations, the the required minimum minimum back back pressure pressure of of 15 15 psig psig on on the RCP No. No.1I seals is ensured seals is ensured by by maintaining maintaining aa pressure pressure of of at at least least 18 18 psig psig inin the the VCT. VCT. Version 40.0 Version 40.0
WESTINGHOUSE PROPRIETARY WESTINGHOUSE PROPRIETARY CLASS CLASS 22 FNP Units FNP Units 11 && 22 CVCS/HHSI/ACCUMULATOR/RMWS CVCSIHHSIIACCUMULATORlRMWS A-181009 A-181009 These requirements These requirements are are to to be be satisfied satisfied assuming assuming the the nominal nominal excess letdown excess letdown flow flow rate rate (12,400 (12,400 lbmlhr), lbmlhr), normal normal RCP RCP No. No. II seal seal leakage leakage (3 (3 gpm) gpm) from from each each RCP, 60 gpm recirculation RCP, 60 gpm recirculation flow from flow from one charging pump, one charging pump, andand normal normal VCT VCT pressure. pressure. (Reference 6.2.1) (Reference 6.2.1) 2.2.2.15 The CVCS The CVCS is is required toto makeup makeup for for shrinkage shrinkage during during aa 100 degree F/hr 100 F/hr cooldown coo idown of the RCSRCS from hot zero power considered to be an original design basis 350°F. This is considered to 350°F. function of the CVCS. 2.2.3 Seal Injection and Leakoff 2.2.3.1 The CVCS is required to cool excess letdown, RCP No. I1 seal leakage, and recirculation flow from at least one charging pump to 115°F prior to returning the flow to the charging pump suction. The cooling function is performed by the excess letdown heat exchanger and the seal water return heat exchanger. The 115°F temperature is based upon the RCP seal injection temperature limit of 130°F. (References 6.2.44, 6.2.7 and 6.2.8) 2.2.3.2 The CVCS is required to provide a seal water injection flow rate adjustable over a normal range of 88 to 13 gpm to each RCP No. 1I seal. It is required that suspended solid particles larger than 55 microns in size be removed from the injection stream. This capability satisfies the seal water requirement for the RCP No. 1I seal. A portion of the seal injection flow (nominally 55 gpm per pump) enters the RCS through the labyrinth seals and the thermal barrier. This in-leakage precludes leakage of reactor coolant through the No. II seal during normal operation. The remainder of the seal injection flow (nominally 33 gpm) flows up the pump shaft, cooling the pump lower bearing and the No. 1I seal. The 5-micron filtration requirement is is based upon RCP minimum seal clearances. clearances. (Reference (Reference 6.2.44) 2.2.3.3 The CVCS CVCS is required to provide aa means for cooling the RCP lower bearing under low RCS pressure pressure conditions conditions when when thethe RCP RCP No.No. 1I seal leakoff flow isis less sealleakoff than 1.0 less than 1.0 gpm. gpm. 2-12 2-12 Rev. 18 Rev. 18
- 24. 026AA
- 24. 026AA 1.01 101 001/NEW/RO/C/A 001/N EW/RO/C/A 3.1/3.1/N/N/2ICVR/Y 3.1/3. 1/N/N/2ICVR/Y Unit Unit 1I is is at at 100%
100% power, power, and and thethe following conditions conditions occurred: occurred: 1000: At 1000:
** B B Train CCW COW is is on on service, service, andand aligned aligned for split split train operation.
operation.
** A A loss loss of of 1IL L 4160V Bus Bus hashas occurred.
At 1015:
- RCP motor bearing temperatures are:
- IA: 172°F 1A: 172°F and rising
- 11B:
B: 165°F and rising
- 1IC: 197°F and rising C: 19rF At 1020:
- RCP bearing temperatures are:
- IA: 230°F and rising 1A:
- 11B:
B: 221°F and rising
- 1IC:
C: 240°F and rising Which one of the following states: I) the procedure(s) that must be entered, 1) and
- 2) the EARLIEST time that a reactor trip is required based on RCP Bearing Temperatures?
T emperatu res? (1)) (1 (2) A. AOP-10.0 ONLY 1015 B. AOP-10.0 ONLY 1020 C C~ Both AOP-9.0 AND 10.0 10.0 1015 D. Both AOP-9.0 AND 10.0 1020 Page: Page: 63 63 d of 200 200 1211412009 1211412009
AA - IIncorrect.
- ncorrect. First First part part isis incorrect incorrect since since entry entry conditions conditions are are met met for for both both AOP-1 AOP-10 0& &
AOP-9.0. They would be They would be done done in in parallel. parallel. AOP-10 AOP-10 would direct direct AOP-9 AOP-9 to to be be entered ifif itit was entered was notnot entered entered directly. directly. AOP-1 AOP-1 00 ONLY ONLY isis plausible, plausible, since since the the loss loss of of SW is the initiating event and AOP-10 SW is the initiating event and AOP-1 0 would deal would deal with with it. it. Second Second part part is is correct correct since the since the temperature temperature for for requiring requiring aa reactor reactor trip trip on on high high RCP RCP Motor Motor bearing bearing temperatures (195°F) temperatures (195°F) per per step step 22 ofof AOP-9.0 AOP-9.0 have have been been exceeded. exceeded. B - Incorrect. B - Incorrect. The first part part is is incorrect incorrect (see(see A). A). The second second part part is is incorrect incorrect (see (see A). Plausible, since per Plausible, per AOP-4.1 AOP-4.1,, Abnormal Reactor Reactor Coolant Pump Pump SealSeal Leakage, Leakage, there is a RCP RCP trip criteria for "CHECK CHECK RCP RCP lower seal water bearing and seal seal water outlet temperatures stabilize less than 225°F". 225°F. Confusion may exist between the two temperatures at which to trip the Reactor Reactor and RCP: RCP: 195195 & 225. C - Correct. The AOP-1 0 entry is required per the loss of the B train SW Buss, and loss of B train SW, per the entry conditions. AOP-10 AOP-1 0 will also direct entry into AOP-9 further into the procedure, and the entry conditions for AOP-9, Loss of CCW, COW, directs entry into AOP-9.0 for a loss of SW supplying an operating CCW train. Second part is correct since 1015 is the earliest that 195°F RCP bearing temperature is exceeded, and this is the requirement for a reactor trip per the continuing action STEP 2 and associated note in AOP-9.0. Note in AOP-9 also states that if flow is not adequate to maintain temperatures, trip the reactor. D - Incorrect. The first part is correct (see C). The second part is incorrect (see A). AOP-9.0 Ver. 22 AOP-10.0 AOP-1 0.0 Ver. 15 FNP-1-AOP-4.1, Abnormal Reactor Coolant Codant Pump Seal Leakage, Veron Version 5.0 _7CHECK RCPlowerssd _ 7 CH ECK RCP lower seal water beering becri n9 aid seal water outlet and outl temperatures atatilizea ess than 225°F. stabi Iizes Iless 0 225 F. FNP-1-AOP-9.0, Los Canponent Cooling Water, Veron Loss Of Component Version 22.0
- Chk Check RCP motor becri n9 temperatures Iless bearing ess than 195 195°F.
0 F. Page: Page: 64d 64 of 200 200 1211412009 1211412009
Previous NRC Previous NRC exam exam history history ifif any: any: 026AA1 .01 026AA1.01 026 Loss 026 Loss ofof Component Component Cooling Cooling Water AAI. Ability AA1. to operate Ability to operate and and!loror monitor monitor the the following fdlowing as asth apply to they apply to the the Loss Lo ofof Component Coding Component Cooling Water: (CFR 41.7/45.5/45.6) 41.7! 45.5145.6) AAI1.01 AA CCW temperGture
.01 CCW tenperure indicaions I ndicions ...................................... 3.1 3.1 Match justification: A loss of CCW is provided in the question, and rising temperature values are given. Knowledge is required to answer the question of how to operate as a result of the temperatures and monitor the temperatures (i.e which procedure(s) is(are) entered for this condition, and at which temperatures a reactor trip is required).
Objective:
- 2. EVALUATE plant conditions and DETERMINE if entry into AOP-10.0, Loss of Service Water is required. (OPS-52520J02)
- 4. LIST AND DESCRI DESCRIBE BE the sequence of major actions associated with AOP-10.0, Loss ServiceWater.
of Service Water. (OPS-52520J04).
- 5. EVALUATE plant conditions and DETERM I NE if any system components need to be operated while performing AOP-10.0, Loss of Service Water. (OPS-52520J06).
- 2. EVALUATE plant conditions and DETERMINE DETERM I NE if entry into AOP-9.0, Loss of Component Cooling Water is required. (OPS-52520102)
(OPS-52520l02)
- 6. EVALUATE plant conditions and DETERM II NE if any system components need to be operated while performing AOP-9.0, Loss of Component Cooling Water.
(OPS-525201 06). Page: 65 of Page: 65 of 200 200 1211412009 1211412009
09/02/08 10:43:52 09/02/08 10:43:52 tt"J .L\?i;L . "L;! FNP-1-AOP-10.0 LOSS OF OF SERVICE SERVICE WATER WATER Version 14.0 14.0 A. A. Purpose This procedure provides actions for This for response to a loss of one or both trains of service water. This procedure is applicable at all times. B. Symptoms or Entry Conditions I. This procedure is entered when a loss of either train of service water is indicated by any of the following:
- a. Actuation of SW PRESS A TRN LO annunciator AD4 or SW PRESS B TRN LO annunciator AD5 (60 psig)
- b. Actuation of SW PUMP TRIPPED annunciator AE4
- c. Actuation of SW TO AUX BLDG HDR PRESS A OR B TRN LO annunciator AE5 (50 psig)
- d. Trip of any operating SW PUMP
- e. Rising temperatures on components supplied by service water
- f. Loss of power to one or both S SWW 4160 V busses 11K K or 11L L
(ec Page 1 1 of 16
11/25/087:42:14 FNP-1-AOP-9.0 LOSS OF COMPONENT COOLING WATER Version 22.0 A. Purpose This procedure provides actions for response to a loss of an operating component cooling water train. This procedure is applicable at all times. B. Symptoms or Entry Conditions I. This procedure is entered when a loss of component cooling water is indicated by any of the following:
- a. Trip of any operating CCW PUMP
- b. Loss of SW supply to an operating CCW train
.7, Page 11 of 11 oflI
11125/087:42:14 11/25/087:42:14 FNP-1-AOP-9.0 LOSS OF COMPONENT COOLING WATER Version 22.0 Step Action/Expected Response Response Not Obtained NOTE:
- Step 2 is a continuing action step.
*. IF RCP motor bearing temperatures exceed 195°F, THEN the ON SERVICE train is affected.
*. Adequate CCW flow means sufficient cooling is available to maintain acceptable temperatures.(i.e. charging pumps, RHR cooling, SFP cooling, RCP's RCPs etc.)
- Indications of pump cavitation are: Abnormal CCW flow oscillations or cavitation noise reported at the pump.
2 Check CCW system adequate for 2 Perform the following: continued plant support.
)\0
- Check CCW flow adequate in 2.1 IF the ON SERVICE train is affected, affected train. THEN perform the following:
\
i-j 1 ** Check RC~ RCP motor bearing temperatures less than 195°F. 2.1.1 IF the reactor is critical, çt. cç 9 THEN trip the reactor and perform,
- Check CCW pump not FNP-I-EEP-O, FNP-1-EEP-0, REACTOR TRIP OR cavitating. Stop any cavitating SAFETY INJECTION, while CCW pump.
CCWpump. continuing with this procedure.
- CCW Surge tank level being maintained at or above 13 2.1.2 Verify all Reactor Coolant pumps inches. stopped.
2.1.3 IF in Mode 3 or 4, llinMode3or4, THEN perform FNP-I-AOP-4.0, FNP-l-AOP-4.0, LOSS OF REACTOR COOLANT FLOW while continuing with procedure. NOTE: Indications of CCW pump cavitation will be abnormal CCW flow oscillations or cavitation noise reported at the pump. 2.2 IF evidence of CCW pump cavitation exists, THEN stop affected CCW pump. o Step 2 continued on next page _Page Completed Page 33 of 11
11/25/08 7:41:37 11125/087:41 :37 Ii)",,,) "?:; S .2L i. FNP-1-AOP-4.1 ABNORMAL REACTOR COOLANT PUMP SEAL Version 5.0 LEAKAGE ~ Step Action/Expected Response Response Not Obtained I I I CAUTION: To prevent potential seal damage, neither seal injection nor CCW cooling should be restored to a RCP which has lost both seal injection and CCW cooling. NOTE: Refer to the Integrated Plant Computer page lRCP, 1 RCP, RCP Temperature Summary, for RCP seal water temperatures. 7 CHECK RCP lower seal water bearing 7 Perform the following: and seal water outlet temperatures stabilizes less than 225°F stabiIlzes 225°F.. 7.1 Shutdown the affected reactor coolant Monitor the following computer points for pump as follows: the affected pump. ((1] TE 132 RCP A SEAL WATER OUTLET TEMP 7.1.1 Manually trip the reactor, AIJP AND go to (()) TEOI31 RCP A LOWER SEAL WATER TE0131 FNP-1-EEP-O, REACTOR TRIP OR FNP-I-EEP-O, BRG TEMP BRGTEMP SAFETY INJECTION. (()) TEO 129 RCP B SEAL WATER OUTLET 0003 908} {CMT 0003908} TEMP (()) TE 128 RCP B LOWER SEAL WATER BRG TEMP BRGTEMP 7.1.2 WHEN the reactor is shutdown, ((j] TEO 126 RCP C SEAL WATER OUTLET THEN stop the affected RCP(s). TEMP {CMT 0003908} ((1] TEO 125 RCP C LOWER SEAL WATER TE0125 BRG TEMP BRGTEMP 7.2 IA OR IB IF 1A lB RCP is secured, THEN close the pressurizer spray valve for the affected RCP. ((j] PK444C for lA 1A RCP ((j] PK444D for IB lB RCP
, Step 7 continued on next page Page Completed Page 8 of24 of 24
- 25. 026K1.01 026K 1.01 001/NEW/RO/MEM 001/NEW/RO/M EM 4.214.21N/N/3/CVRIY 4.2J4.2JNINI3ICVR/Y A Unit 1I Safety Injection is in progress due to a Large Break LOCA.
Which one of the following describes the connection(s) between the RWST, A Train CS and ECCS pumps suction, and the operation of MOV-8827 MOV-8827A A and MOV-8826A, CTMT SUMP TO 1A IA CS PUMP valves? A Train CS Pump, A Train HHSI Pump, and the A Train RHR Pump have (1) suction header(s) penetrating the RWST, and the CS Sump suction valves (2) automatically open on a LO-LO RWST condition. (1 (1)) (2) A. separate will NOT B. one common will C. separate will D 0:' one common will NOT A -. Incorrect. The first part is incorrect, but plausible since most of the safety related equipment has physical train separation for piping. The RWST is designed to minimize tank penetrations, and uses only one penetrations for suctions to all CS pumps, RHR pumps, and CVCS/HHSI pumps. The second part is correct. B - Incorrect. The first part is correct. The second part is incorrect, but plausible since this would be correct for the RHR sump suctions which have the auto function described. C - Incorrect. The first part is incorrect (See A). The Second part is incorrect (See B). oD - Correct.
- The RWST is designed to minimize tank penetrations, and uses only one penetrations for suctions to all CS pumps, RHR pumps, and CVCS/HHSI pumps.
The CS Sump suction valves do not have the auto open feature, but the RHR sump suctions do. Page: 66 of 200 12/1412009 1211412009
Previous NRC exam history if any: n/a 026K1 .01 026K1.01 Containment 026 Contai nment Spray System KI Knowledge K1 Kncmledgeofof the physical connections and/or cause-effect phycaI connectionsandlor cau-effect relationships rdationips between the CSS and the foiIONing~stems: fdlcming sjetems (CFR: 41.2 to 41.9 / 45.7 to 45.8) 41.9/45.7 K1.O1 ECCS .......................................................... 4.24.2 K1.01 4.2 4.2 Match justification: The only physical connection between the CS system and the ECCS system is at the RWST suction of the pumps, which is tested in the first part of each choice. The second part of the distractor contrasts the design of the CSS with the ECCS system sump suction valves to provide symmetry and three plausible but incorrect distractors. Objective: I 1 LABEL AND II LLUSTRATE the Emergency Core Cooling System to include the components found on the following foIl owl ng figures (OPS-40302C05):
- Figure 2, Accumulators
- Figure 3, Refueling Water Storage Tank and Figure 4, Emergency Core Cooling System
- The flow paths found on Figure 14, ECCS Injection Phase, Figure 15, ECCS Cold Lag Re:::irculation, Leg Recirculation, Figure 16, ECCS Simultaneous Hot & Cold Leg Lag Re:::irculation Recirculation Normal, and Figure 17, ECCS Simultaneous Hot & Cold Leg Lag Re:::irculation Recirculation Alternate.
Page: 67 of 200 1211412009 12/14/2009
Date: 10/21/2009 10/21/2009 Time Time:: 07:43:34 AM 8 I 9 I 10 I 11 I 12 12 I 13 2" HCD97 2 HCD-97 r - - - - - - - - - - - - - - - - - - - - <<D D 175039 175039 SH. SH, 77 (H3)< (H 3l< I CCD 5 CVCS BA BLEND
~ fig~3g~ ~~E~DLOC.
I 1OE386 SH 2 LOC. E6 E-6 . 1----------------------------fT-f£~~J QV090 1 8928 D-175038 SH.2 (E-ll) 11OE388SH3 110E388 SH.3 LO LOC. J-4 - I > I
> 0-170118 D-170118 >i )-,
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ç YJ,7 12. 1 &r; V " t..';'~'1
,f I 3/4T58 SIS. ACCUMCCUM TST. ,tiCQ-J~'"'D::---;1:;;75"'0"'43;--("'B--6"')-<7:
75043 (8-6) II ~V ,"t' : 3/4-S1-2501R 3/4SI25D1R 110E388 SH.2 LOC.F2
?E88SH2 . LOC. F-2
~1\ SFPCS I I I ,,/I U HYDRO TEST LV
/S) 11OE4DO 110E400 [0 C J8 LOC. J-8 : : tw'\.'
CONN. CONN. 18869 3/4-Rv58SW HCO-96 HCD-96 I L REFUEUNGR L ~L......;'-:-. . . .---..;.;;:,::.....:;"---_ilf_ "N:171'::-18-'- - -- STORAGETANK 0-3 NV086 501R NOTE 2 IN 1TI<82 / SURGE FILL CONN NOTE l ATORS51 1" 1 2-SH51R HCD-96 HCD96 l-SH51R 1SI151R 1516 Ni Fl 6TA41 N1F16TA4145A 45A
/ F SAFCLASS2A SEE D-5D6647 S 7:1 i A, 6TA4145B
ONINJECTION 9 SURGE TANK 3/4-SI-25D1R 3/4-SI-2501R r1 ITEM: ATST 8ASAETEV CLASS NNS
/1 BB Q1E21TDO5 NOTE 1C NOTE 7 Sl-S1-2501R I_________
2 rlSI Ow I IN.503_69t IN.5D3 IN.503-70 NOTE 2 J 7 LO. 0 N1FT6TE5D9 N1F16TE51D N1F16TE510 QVO21 IN.61079 IN.61080 18934 18933 UQ771B NOTE 6 Q1F16Th6511 Q1F16512 3/4X42D 3/4-X42 18930 NOTE 8 NOTE 1 T58 BORON INJ DH LOCAL SAMP LOCAL SAMPLE LE 2X42D 2 2 HCB79 < Q1P44V506 V0024 FLUSH ORIFICE I 3/4" DR. f----tlM 3/4 4 3/4 3/4" VV BRP. ITEM: ORFL HI I OV903 QV9D3 SAMPLE POINT QV900 0175043 (F1O)4 I.1" -[::rL-~=.:.l--""'I--I--t~ FO6007 QV9043/4" v V 4V505 QV022 1-8935 8935 QV9D4 1l-HCV-947.../"" HCV947 .
. SFPCS. PURIF PMP 11OE400 LOC. 03
\2 HCD93 151R L /4HCB42 cC l-T58 1758 l-RA58D lRA58D l-S1-2501 R.../ FO 1SI25D1R 01E13 12"12 HCB-152 HCB152 1
1 1/2" 1/2 XX 1" 1 RED--./ RED \8932 3/4 0 N N 3/4 3/4" XX 1/2" 1/2 RED RED\Z 1~~~~~ QV542 D-175038 SH.3 E-12 I 18979 CD to 3/4 X42D 3/4-X42D C SIS SPRAY PUMPSPUMP 0 11OE388 SH3 110E388 SH.3 LOC. F 1 LOC F-l 0 SEE NOTE 15
;., 1/2 X 2 RED WHT 3
WHT OH DH 14" HCB-43 D-175038 SH.2 (F -11)
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=
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-11
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NOTES:
~~~~~~~~ ~OfJ:TAi~~6L~U~ ~6~C~~~0f;fg*
1A. ELECTRICAL POWER SUPPLY IS OISCONNECTED. lA. EQUIPMENT IS INSTALLED BUT NOT REQUIRED A ,'~'z. JM"l/'s
'-' eI Li> '-' eI I r M, 8 ","'
i!2 1i' 13 to ro :r:'....j
=u
~ 18.
1 FOR SYSTEM FOR SYSTEM OPERATION. OPERATION. B. AIR AND ELECTRICAL POWER REMOVED. 7894) 4
~f -{MY [M]--
S
- i '-',lei I 6 15l:: 2 ~~~~~JNg~~~~o~UT VALVE INSTALLED BUT NOT REQUIRED FOR SYSTEM OPERATION. NOT REQUIRED FOR EC I r-: :--'11--,
*O__
QV026 1 1 C1C. . PIPING AND COMPONENTS OF THE BIT/BIT 80/BIT I I I I 1-8926 ~:ul RECIRCULATION SYSTEM RECIRCULATION SYSTEM WERE WERE INSTALLED INSTALLED
-~
4" X X 3" 3 RED REDI1
...1') I "r I
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'T' C r-.J-l><}-.J ...
I I I
-!..-------.1-----..lL"L I r - -./ / ?
'\ "
I I IA8" HCB-142 HCO-142 S SPEC. AMENDMENT 30. SAFETI RELIEF VALVE GAGGED DURING INITIAL
- 2. SAFETY HYDROTEST AND SET TO 700 7DD PSIG FOR 78 v , , ,--"--------:><-----i VEMS/ rM'l/ NORMAL PLANT OPERATION.
--' '------l __t ____j
'2.. JM'
- I rrM,1 11 L
J
'-'lei II
'-j NO. 11
" I L...
L - L -C': '-'lei 3. LOCATE TAPS 2'-0" 20 APART.
- 4. LOCATE CONNECTION ABOVE WATER LEVEL.
4, Li' 45 "eI LM.J I ,'-
- .A I rr-i, \~ 5, FLANGES FOR FLOW METERING ORIFICE I r-t.
r r 1 i
- -'Ir-, HIGH HIGH HEAD HEAD M
,-"rL';,
L1/5 TO VERIFY FLOW DURING PRE-OPERATIONALPREOPERATIONAL
-'- I I SAFTEY INJECTION I 8-SH51R 8SI151R TES11NG.
TESTING.
)t -{MJ : -~
"M"-}~ 6. EQUIPMENT INSIDE BOX IN OTHERS SCOPE. SCOPE.
TIT L.M.J I I "r / (CHARGING)
,,,,S PUMPS F 1 45 LI L\
r QV543 6.
- 7. 3 7.
EQUIPMENT 3 IN. IN. 150 WITH 3 150 LB. 3 IN. INSIDE LB. R.F. BOX IN R.F. STUDDING IN, BLIND FLANGE. OTHERS STUDDING OUTLET OUTLET F 3/4 0 D I
- ----l><}-.J r-i-r-v ...
I : I r-!.-------.1-----..lL{ 1.
'\ "
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1--------- --, I 3/4" I
- 8. 1
- 9. A 1 1/2 IN. 150 LB. R.F. STUDDING OUTLET.
A 3/8" 3/8 FLOW RESTRICTOR IS REQUIRED AS NOTED ON LEGEND.
> '2..JM' ri I rFM]
M , \_.-
, _ / ,,:1 ,~ I 10. ALL VALVE & LINE NUMBERS ON THIS DWG.
.... L.' çLI 4
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L "lei NO. 2 NO.2 75 EM} LM.r 7, I ARE PREFIXED BY Q1E21 EXCEPT WHERE
~
2
- 0
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- "':~
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- 11. ~g~~D T~~~~R~~I~'S NOTED OTHERWISE.
- 11. WORK THESE P810s WITH DWG.
THRU 0 175372 0175372. DWG. 0-175364 D175364
-~ \~r.A ~
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<< 4EM] -{M I7 DWG (jf) fl0E386 SH DWG 110E386 SH. 22 " :1 :- DWG. 0-175044
- 12. SEE owe. 0175044 FOR WESTINGHOUSE
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LM.T r.:l/s:r: LM.J ro I I SYSTEM LEGEND. SYSTEM LEGEND.
- 13. FOR PIPING CLASS
SUMMARY
SHEETS SEE i5- I I II r - - I I I SPEC. SS-1109-1. 5511091. r-l><I_--v ...r---------.1----- ii "L, _~_..JL ____ .;><I_----,L- ~,: J 14. FOR VALVE INFORMATION SEE MASTER 25dB I I L - - - ..
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45 15. (SS110239) VALVE LIST (SS-1102-39) IS. TEMPORARY STRAINER IS PLACED IN SPOOL PIECE DURING INffIAL INITIAL FLUSHING OPERATIONS STRAINER MUST BE REMOVED BEFORE PLANT STARTUP. STARTUP.
- 16. ADJUST VALVES IN FIELD DURING PRE-OPER- PREOPER G
I '-' eI LI, ATIONAL TESTING TO BALANCE INJECTION FLOW
,J, c ,0, 22. FLOW LIMITING ORIFICE INSTALLED TO ACHIEVE ~ 4 00 AND LIMIT PUMP
~~~I~b~T POSITION.
PUMP RUN RUN OUT, OUT, THENTHEN LOCK INTO LOCK INTO
~-L) ~ PROPER INJECTION PROPER FLOW.
INJECTION FLOW. ~ 'l' ~ 17. DRAIN ALL DRAIN CONNECTIONS TO LOCAL I 23. LINE 23, HCC195 HAS LINE HCC-195 HAS BEENBEEN DOWN GRADED FROM DOWN GRADED FROM ~ oS EQUIPMENT DRAINS. L, /1t g L,':
\/
15 I I ASME CLASS 3 SEISMIC CATEGORY 11 TO NNS SEISMIC CATAGORY II BY FSAR AMENDMENT 73 VALVE AND LINE NUMBERS HAVE NOT CHANGED.
- i:I ~to
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- 16. FOR COMPLETE INSTRUMENT NUMBERS REFER
- 18. ~gRIN~~~0~rJT 19.
TO INSTRUMENT I~NS6:J~E~i7~~~1.ERS
- 19. DELETED.
DELETED. INDEX 8175803. REFER
~ FLAG (6)
FLAG () DESIGNATION DESIGNATION NNS NNS CLASIFICATlON, CLASIFICATION. ~ u ~ r- '" 20 MULTIPLE LOOP PWR SUPPLY (CABINET MTD) I - - - oJ U262852BOP PROCESS INSTRUMENT AND CONTROLS,
- 24. U262852-BOP to,5
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- TRAIN A A DWG. 7408023, 74D8023, TRAIN B B DWG.DWG, 7408045. :I: ~ '" t3 MULTIPLE LOOP PWR. PWR SUPPLY SUPPLY CONTROL
~ ~ ~ ~",' 21. M(NUOLTTIPSLHEOWLONO.)P CONTROL BOARDBOARD
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- 25. TE'S,
- 25. TES, AS NOTED ARE f'VI"'1 I~I" U(\lIITTf'oOlt..lf' CLAMPED ON RTD'S RTDS FOR THERMAL I U
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0 21. (NOT SHOWN.)
Title:
C:\Reference Disk\Exam Reference Disk\Drawings\D175038-0001.cal Disk\Drawings\D1 75038-000 1 .cal
FNP Units 11 &
&2 RESIDUAL HEAT REMOVAL A-18 1002 A-181002 3.3.3.2 The tank is designed according to the requirements of the ASME Boiler and Pressure Vessel Code Section III, Class
- 2. (Reference 6.1.29) 3.3.3.3 Level Indication complies with the requirements of IEEE-279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations. (Reference 6.1.33) 3.3.3.4 Level Indication complies with the requirements of IEEE-308-1969, Criteria for Class IE 1E Power Systems for Nuclear Power Generating Stations. (Reference 6.1.33) 3.3.4 I&C Requirements 3.3.4.1 The R RWST WST must be provided with 2 level transmitters.
One channel provides high, Technical Specification minimum, low, and low-low level alarms. The other channel provides low and low-low alarms. These channels provide the following alarm functions (References 6.2.14, 6.4.2, 6.7.79 and 6.7.80): 6.4.2,6.7.79
- a. High Level Alarm - This alarm alerts the operator to a high level which could lead to tank overflow.
- b. Technical Specification Minimum Level Alarm - -
Indicates that the RWST level is less than that required by the Technical Specifications during normal operations.
- c. Lo Level Alarm - Alerts the operator that the ECCS pumps should be manually aligned for recirculation phase operation.
- d. Lo-Lo Level Alarm - Alerts the operator that the CS
/ , pumps should be manually aligned for the z recirculation phase. If an SI signal is present, this alarm will automatically open RHRlLHSI RHRILHSI pump
(/i sump suction valves 8811A & BandB and 8812A & B. _ (References 6.7.43 [Section 6.3.2] and 6.7.42 Lr [Section 6.3.2.4])
/ ,
3.3.4.2 One of the two redundant level indicators in the control room must be operable for Post Accident Monitoring. (References 6.1.18, 6.7.15 and 6.7.16) 3-10 Rev.25 I
- 26. 027AKl.02
- 26. 027AK1.02 OOlINEW/RO/MEM 001/NEW/RO/MEM 2.8/3.1/N/N/3/CVR/VER 2.8/3.1/N/N/3/CVR/VER 55 EDITORIAL EDITORIAL Unit Unit 22 is is at at 50% power, 50% power, and and PT-444, PT-444, PRZR PRESS, pressure PRZR PRESS, pressure transmitter transmitter has has failed failed to to the 2230 the 2230 psig psig position.
position. Which one Which one ofof the the following following describes describes the the effects effects on on PK-444A, PK-444A, PRZR PRZR PRESS PRESS REFERENCE controller, REFERENCE controller, and and the the pressurizer pressurizer liquid liquid density density due due to to this this malfunction? malfunction? PK-444A controller controller demand goes (1) (1) and the density of the pressurizer liquid goes (2) (1)) (1 (2) A. down up B. down down C. up up D D~ up down A - Incorrect. The first part is incorrect (see D). Plausible, since if the PT had failed 6 psig higher (above 2235 psig), the proportional integral controller would integrate the error signal down until the PORV 444B opened and the sprays opened. Also, the spray valve controllers are controlled by the "master" master controller and when the pressure must be increased, the demand goes down. Confusion could exist as which controller function is being described. The second part is incorrect. Plausible, since the spray valve controllers are controlled by the "master" master controller and when their demand goes up pressure goes down and the liquid density goes up. Also, steam space density does go up in this condition, and the liquid specific volume goes up (and specific volume, not density, is the value given in the steam table for the property of the liquid). BB - Incorrect. The first part is incorrect (see A). The second part is correct (see D). C - Incorrect. The first part is correct (see D). The second part is incorrect (see A). C - D0- Correct. The Proportional/Integral PRZR
- PRZR PRESS PRESS controller senses aa low low pressure and the demand starts integrating higher and and higher. This first causes the spray valves to close and the proportional heaters increase output. Then, the backup heaters energize. The pressurizer liquid heats up and expands (density goes down) due to the increased due increased heat heat input input into the pressurizer into the pressurizer liquid.
liquid. The integral integral part part of of the the controller continues continues toto add add toto the the error error signal signal and and PORV-445A PORV-445A opensopens due due to to actual actual pressure pressure increasing increasing toto 2235 2235 on on PT 445. The PT 445. The pressure cycles around pressure cycles around the the setpoint setpoint ofof the the PORV PORV at at 2235 2235 psig psig with higher pressurizer with aa higher pressurizer liquid liquid temperature. Page: Page: 68 68 ofof 209 200 12/14/2009 12/14/2009
Previous NRC exam history if any: 027AK1 .02 027AK1.02 027 Pressurizer Pressure Control System Malfunction AK1. Knowledge ofthe AKI. of the operational implications of the following concepts as they apply to Pressurizer Pressure Control Malfunctions: (CFR 41.8 1I 41.10 1I 45.3) AKI.02 Expansion ofliquids AKl.02 of liquids as temperature increases ........................ 2.8 3.1 Match justification: To answer this question correctly, it must be recognized that for this particular malfunction of the PRZR Press control system, the pressurizer liquid heats up and expands due to pressurizer heaters energizing and sprays closing. The operational implications must also be understood in that this causes controller demand to go up (which would cause actual pressure go up until a PORV-445A). PORV will lift: PORV-44SA). Objective:
- 2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the following components associated with the Pressurizer Pressure and Level Control System to include those items found on Figure 2, Pressurizer and Pressure Relief Tank, Figure 3, Pressurizer Pressure Protection and (OPS-52201H02).
Control, and Figure 7, Pressurizer Level Protection and Control (OPS-5220IH02).
- 5. DEFINE AND EVALUATE the operational implications of abnormal plant or equipment conditions associated with the operation of the Pressurizer Pressure and Level (OPS-52201H07):
Control System components and equipment to include the following (OPS-5220IH07):
- Normal Control Methods
- Abnormal and Emergency Control Methods
- Automatic actuation including setpoint, if applicable
- Protective Interlocks Actions needed to mitigate the consequence of the abnormality Page: 69 of 200 12/14/2009
A62 -y t<<-;)dh l-'J-?tl.(.-e-f,?~ >;y>->~/ PT-'/ rn ~ ~ ~ -It;,~,~ ~ r 1c..-i'lfl1 t/crv // f -c 1ftYt-t-Y ~~
Table 2: Tabe 2: Saturated Saturated Steam: Steam: PressurePressure TaWe Table
------- Specific Volume Specific
~------
Volume Enthalpy Enthalpy
-~-.----------------------.-----
Entropy Entropy Abs Press. Abs Press. Temp Temp Sat. Sat. Sat Sat. Sat. Sat. Sat. Sat. Sat. Sat. Sat. Sat. Abs Abs Press. Press. LbiSq In. LbfSq In. Fahr Fahr Liquid Liquid Euap Evap Vapor Vapor Liquid liquid Evap Evap Vapor Vapor Liquid Liquid Eoap E.. ap Vapor Vapor lb/Sq In. Lb/Sq In. Pp 1t vi ¶.ig V tg Vg Yg hr hf hg h fg hhg s sj 5 1Sgfg g Sg PP n.08m 8.00065 32,018 32.018 (},01602Z 0.016022 3302.4 3302.4 3302,4 3302.4 0,0003 0.0003 1075.5 1075,5 1075.5 W75.S 0.0000 0.0000 2.1872 2,1872 2.1872 2.1872 8.08085 0.08865 0,25 0.25 59.323 59.323 0,016032 0.016032 1235.5 1235.5 1235.5 1235.5 27.382 27.382 1063.1 1060.1 1087.4 1087.4 0.0542 0.0542 2.0425 2,0425 2.0967 2,0967 0.25 0,25 0,50 0.50 79,586 79.586 0.016071 0.016071 641.5 641.5 641.6 641.5 47,623 47.623 1048,6 1048.6 1096.3 1096.3 0.0925 0,0925 1.9446 1.9446 2.0370 2,0370 8.58 0.50 1.0 1.0 101.74 101,74 0,015135 0.016138 333.59 333.59 333.60 333.60 59.73 69.73 1036.1 1036.1 1105.8 1105,8 0.1326 0.1325 1.8455 1.8455 1.9781 1.9781 1.0 1,0 5.0 5.0 162,24 162.24 0.016407 0,016407 73.515 73.515 73.632 73.532 130.20 130.20 1000,9 1000.9 1131.1 1l3Ll 0.2349 0,2349 1.6094 1.6094 1.8443 1.8443 5.0 5.0 In.O 10.0 193,21 193.21 0,0!6592 0.016592 38,404 38.404 38,420 38.420 161.26 161.26 982.1 982.1 2143.3 1143.3 0.2836 0,2836 1.5043 1.5043 1.7879 L7879 10.0 10,0 14.696 14,696 212,00 212.00 0,016719 0.026719 26.782 26.782 26.799 25.799 180.17 180.17 970.3 970.3 1150.5 1150.5 0.3121 0,3121 1.4447 1.4447 1.7568 1.7568 14,090 14.696 15.0 15.0 213.03 213.03 0,016726 0.016726 26,274 26.274 26.290 26.290 161.21 181.21 969.7 969.7 1150.9 1150,9 0.3137 0,3J37 1.4415 1.4415 1.7552 1.7552 25.0 15.0 20.0 20.0 227.96 227.96 0.016834 0.016834 20.070 20.070 20.067 20,087 196,27 196.27 960.1 960.1 llS6,3 1156.3 0.3358 0.3358 1.3962 1.3962 1.7320 U320 28,0 20.0 30.0 30.0 250.34 250.34 0.017009 0.017009 13.7266 13.7266 13.7436 13.7436 218.9 218,9 945.2 945.2 1164.1 1164.1 0.3682 0,3682 1.3313 i.3313 1,6995 1.6995 30.0 30,0 40.0 40.0 267.25 267.25 0,017151 0.017151 10.4794 10.6794 10,4965 10.4965 236.1 236.1 933.6 933,6 1169.8 lJ69.8 0.3921 0.3921 1.2844 1.2844 1.6765 1.6765 40.8 40.0 50.0 50.0 281.02 281.02 0,017274 0.817274 8.4957 8.4967 8,5140 8.5140 250.2 250.2 923.9 9:23.9 1174.1 il74.1 6.4112 DAli2 1.2474 1.2474 1.6586 1.6586 50.0 50.n SO.O 60.0 292.71 292.71 0,017383 0,017383 7.1562 7.1562 7.1736 7.1736 262.2 262.2 915.4 915,4 1177.6
!177.6 0.4773 0,4273 1.2167 1.2167 1,6440 1.6440 68.8 60.0 70,0 70.0 302.93 302.93 0,01,482 0.01482 6.1875 6.1875 6,2050 6.2050 272.7 272.7 907.8 907.8 ilBO,6 1100.6 0.4411 0.4411 1.1905
!.l905 1.6316 1.6316 78.0 10.0 BO,O 00.0 312,04 312.04 0,017573 0.017573 5,4536 5.4536 5.4711 5A7ll 282.2 282.i 900,9 900.9 1183.1 1183.1 0.4534 0.4534 1.1675
!.l675 1.6206 1.6208 80.0 88,0 90.0 90.0 320,28 320.28 0.017659 0.017659 4.8779 3.8779 4,8953 4.8953 290.7 290.7 894.6 894.6 1185.3 lJB5.3 6.4643 0.4643 1.1470 U470 1,6113 1.6113 90.0 90.0 100.0 100.0 327.82 327.82 0.017740 0.017740 4.4133 4,4133 4.4310 4.4310 298.5 298.5 888.6 888,6 1167.2 1187,2 0.4743 0.4743 1.1284 1.1284 1.6027 1.6027 180.8 100.0 110.0 118.0 334.79 334.79 0,01782 0.01782 4,0306 4.0306 4.0484 4.0483 305.8 305.8 883.1 883.1 1188,9 1188.9 0.4834 0.4834 1.1115 Ul15 1.5960 1.5950 108.0 110.0 120.0 128.8 341.27 341.27 0,01789 0.01789 3.7097 3.7097 3.7275 3.7275 312.6 312.6 677.8 877.8 1190.4 1190.4 0.4919 0,4919 1.0960 1.0960 1.5879 1.5879 120,0 120,0 139.0 138.0 347.33 347.33 0.01796 3.4364 3.4364 3.4544 3,4544 319.0 319.0 672.8 872.8 1191.7 1191.7 0.4998 0.4998 1,0815 1,0815 1.5813 1.5813 130.0 130.0 140.0 149.8 353,04 353.04 0.01803 now 3.2010 3.2190 3.2190 325,0 325.0 868,0 868.0 1193.0 1193.0 0.5071 0,5071 1.0681 1.0681 1.5752 1.5752 140.0 140,0 150,0 150.0 358.43 358.43 0.01809 0.01809 2.9958 2.9958 3.0139 2.0139 330,6 330.6 863,4 863.4 1194.1 1194,1 0.5141 0,5141 1.0554 1.0554 2.5695 1.5695 150.0 150.0 160.0 168.8 363.55 363.ss 0,01815 0.01815 2,8155 2.8155 2.8336 2.0336 336.1 336,1 859,Q 859.0 1195.1 1195.1 0.5206 0.5206 , 1,0435
.1.0435 1.5641 1.5641 160.0 160.0 110,0 110.0 368.42 368.42 0.01821 0,01821 2.6556 2,6738 2.6736 341,2 341.2 854.8 854,8 1196,0 1196.0 0.5269 0.5269 1.0322 LOm 1.5591 1.5591 170.0 170.0 180,n 180.0 373.08 373.08 O.Ql827 0.01827 2,5129 2.5129 2,5312 2.5312 346.2 346.2 850.7 850.7 1196.9
!l96,9 0.5328 0,5328 1.0215 1.0215 1.5543 1.5543 100.0 1BO.O 190.0 198.0 377.53 377.53 0,01833 0.01833 2.3847 2.3847 2.4030 350,9 350.9 846.7 846.7 1197.6 1197,6 0,5384 0,5384 1.0113 1.0113 1.5498 1.5498 188.0 190,0 20B,0 208.0 381.80 0.01839 0,01839 2,2689 2.2889 2.2873 355.5 355,5 842.0 842,8 1198.3 1198.3 0,5431 0,5438 1.0026 L0016 1.5454 1.5454 208.0 200,0 210.0 385,91 385.91 0.01844 2.16373 2.18217 2,18217 359,9 359.9 830.1 839.1 1199.0 1)99.0 0.5490 0,5490 0.9923 0,9923 1.5413 1.5413 210.0 210,0 220.0 389,88 389.88 0.01850 0,01850 2.06779 2.08629 2.08629 364.2 835.4 1199.6 1199,6 0.5540 0,5540 0.9834 0,9834 1.5374 1.5374 220.0 230.0 230.8 393,70 393.70 0,01855 0.01055 1.97991 1,97991 1.99846 1.99846 368.3 831.8 831.8 1200.1 1200.1 0.5588 0.9748 1.5336 230.8 230.0 240.0 240.8 397.39 0.01860 1.89909 1.91769 1.91769 372.3 372.3 828.4 1200.6 1200.6 0.5634 0,5634 0.9665 1.5299 1.5299 240.0 250.0 400.97 0,01865 1.82452 1.82452 1.84317 1.84317 376.1 376.i 825.0 1201.1 12GlJ 0.5679 0,5679 0.9585 0,9585 1.5264 250.0 250,0 260,0 260.0 404.44 0.01870 U5548 1.75548 L77418 1,77416 379.9 379,9 821.6 1201.5 1201.5 0.5722 0.9508 0,9508 1.5230 268.0 260,0 270.0 407.80 407.80 0,01875 0,01075 1.69137 1.71013 1.71013 383,6 818.3 818.3 1202.9 1201.9 0.5764 0.9433 0,9433 1.5197 1.5197 278.8 270.0 280,0 280,6 4]1.07 411.07 Q.n1880 0.01880 1.63169 1.55049 2,65049 387,1 387.2 825.1 815.1 1202,3 1202,3 0.5805 0,5805 0.9361 0,9361 1.5166 208.0 280,0 290.0 414.25 424.25 0,01885 0.01885 1.57597 1.57507 1.59482 1.59482 390,6 390.6 812,0 812.0 1202,6 1202.6 0.5844 0.9202 0,9291 1,5135 1.5J35 280.0 2BD.O 300.0 417.35 0,01889 0.81889 1.52384 1.54274 394,0 006.9 808.9 1202.9 0.5682 0,5882 0.9223 1.5105 306.0 300,0 350.0 431.73 0.01912 0,02922 1.30642 1.36642 L 1.32554 409,8 409.8 794.2 1204.0 J204.0 0.6059 0,6059 0.8909 1.4968 350.0 444,60 0.01934 L14162 4gC:~9 408.0
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'(1:8318 t93 8 1.4847 1439 lAng 408,8 400.0
. 4580 45tHr 50a.o 500.0 467,01 467.01 0.01975 0.90787 0,90787 lm62 0.92762 449.5 755.1 2204.3 1204.7 0.6498 0,6490 0.8148 1.3639 1.4639 580.0 560.0 550.0 990.0 476,94 476.94 0,OJ994 0.01994 0.82183 0.82083 0,84177 0.84177 460.9 743.3 1204.3 1204,3 0.6611 0,6611 0.7936 0]936 1,4547 1.4547 950.0 550,0 600,0 688.0 486,20 486.20 0,02013 0.74962 0.76975 471.7 732.0 1203.7 0.6723 0,6723 0.7736 0,7738 1.4460 1.4461 680.0 SOO,O 650,D 650.0 494,89 494.89 0.02032 0,68811 0.68811 0,70843 0.70843 481.9 720.9 720,9 1202,8 1202.8 0.6028 0,6828 0.7552 Q.7552 1.4381 650.8 650.0 100.0 503.08 0.02050 0.02058 0.63505 0,63505 0,55556 0.65556 492.6 491.6 710,2 710.2 1201.8 0.6928 0,6928 0.7377 1.4304 708.0 700.0 150,0 758.0 510,84 510.84 0,02069 0.02069 0,58880 0.58880 0,60949 0.60949 500,9 500.9 699.8 0200,7 1200.7 0.7022 0.7210 1.4232 750.0 150.0 800.0 880.0 518.21 0,02087 0.02087 0,54809 0.54809 0.56896 0.568% 509.8 689.6 689,6 1199.4 il99.4 0.7111 0,7051 0.7051 1.4163 080.0 800,0 850.0 525.24 Q,02l0S 0.02105 O.SU97 0.51197 0.53302 518.4 579,5 679.5 1198.0 0.7197 0.6899 1.4096 L4096 050.0 850.0 900.0 908.0 531.95 0.02123 0,02123 0.47968 0.47966 0.50091 526,7 526.7 669.7 669,7 1196,4 1196.4 0.7279 0.6753 1.4032 988.0 906.0 950.0 538,39 538.39 0.02141 0,45054 0.45064 0,47205 0.47205 534.7 860.0 660,0 1194.7 0.7358 0,7358 0.6612 0,6612 1,3970 1.3970 958.0 950.0 1000.0 1088.0 544.58 a,02i5S 0.02259 0,42436 042436 0.44595 0.44596 542,6 542.6 650.4 1192.9 0,7434 0)434 0.6476 1,3910 1.3910 1000.0 1000,0 1050.0 1090.0 550.53 0,02177 0.40047 0.42224 550.1 640.9 1192.0 119LO 0,7507 0.7507 0.6344 0,6344 1.3851 1850.8 1050.0 nao.o 1180.0 556.28 0,02195 0.37863 0.40058 557,5 557.5 631,5 631.5 1189.1
!l89.1 0.7578 0.6216 1.3794 2100.0 1100.0 1150.0 561.82 0,02214 0.02214 0,35859 0.38073 564,8 564.8 622.2 1187.0 1187,Q 0.7647 0,6092 0,6091 1.3738 L3738 1150.0 1150,0 1200.0 1208,0 557.19 567.19 0.02232 0.340.13 0.34013 0,36245 0.36245 571.9 613.0 1184.8 1184,8 0.7714 0.5969 1.3683 2288.0 1200,0 1250,0 1258.0 572.38 572.30 0,02250 0.32306 0.32308 0,34556 0.34556 578.8 603.8 1182,6 H82,6 0.7700 0.7780 0.5850 0,5850 1.3630 1.3630 1250.0 1300,0 1300.0 577.42 577.32 0,02269 0.02269 0.30722 0,32991 0.32992 585.6 584.6 594,6 1180,2 1180.2 0,7843 0.7843 0.5733 0,5733 1.3577 1.3577 1300.0 130M 1350.0 1350.0 58232 582.32 0.02288 0,29250 0.29250 0.31537 592.3 585.4 1177.8 1177.8 0.7906 0,7906 0.5620 1.3525 1.3525 1350.8 1350,6 1400,0 1400.0 587.Q7 587,07 0.02307 0,02307 027871 Q,30178 0.30178 998.8 598.8 576.5 H75,3 1175.3 0.7966 0.5507 0,5507 1,3474 1.3474 1400,8 1400.0 1450.0 1450.0 591.70 0,02327 0.02327 0.26584 0,26584 028911 0.289 it 605,3 605.3 567.4 1172.8 1172.8 0.1028 0,8026 0.s397 0,5397 1.3423 1.3423 1450.0 1450,6 15QO,0 1500.0 596.20 0,02346 0.02346 0.25372 0.27719 611.7 5;;8.4 558.4 1270.1 1170.1 0.8085 0,8085 0.5288 0,5288 1.3373 1.3373 1500.0 1500,0 1550.0 1550.0 600.59 0,02366 0,24235 0.24235 0.26601 028601 618,0 618.0 549.4 1167,8 1167.4 0.8142 0.5182 1.3324 1.3324 1558,0 1550.0 1600.6 1600.0 604.87 Q,02387 0.02387 0.23159 0,23158 {),25545 0,25545 624,2 624.2 540,3 1164,5 1164.5 0.0199 0.3199 0.5076 1.3274 1.327.4 1600,0 1600.0 1650,0 1650.0 509.05 609.05 0,02407 0.02407 0.22143 0,22141 0,24551 0.24551 630.4 531.3 1161.6 1161.6 0.8254 0,8254 0.4971 0.497J 1.3225 1.3225 1850.8 165n,0 1700,0 1700.0 613.13 Q,02428 0.02428 0,21178 0.21178 0.23607 0,23607 636.5 522.2 1158.6
!l58.6 0.8309 0,8309 0.4867 .1.3176
.1.3176 1700.8 1100.0 1150.0 1750.0 1800.0 617.12 617.12 621.02 0,02450 0.02450 0.20263 0.20263 0,19390 0.22713 0.22713 0.21861 0.21861 642,5 642.5 648.5 513.1 513.1 503,8 503.8
]iSS,S 1155.6 1152.3 1152.3 0.8363 0,8363 0.8417 0.4765 0,4765 0.4662 1.3128 1.3128 1.3079 1750.0 1750.8 '
1800,0 1000.0 621.02 0.02472 0.02472 0.19390 648.5 0,8417 0.4662 1.3079 1800.0 1850,D 1858.0 624,83 624.83 {),02495 0.02495 0.18558 0.18558 0,21052 0.21052 654.5 654.5 494.6 494.6 1149.0 1l49,Q 0.8470 0,8470 0.4562 0.4561 1,3030 1.3030 1850.0 1650,0 1900.0 1900.0 528.56 628.56 0.02517 0.02517 0.17761 0,17761 0.20278 0.20276 660.4 660.4 486.2 485.2 1145,6 1145.6 0.8522 0,8522 0.4459 0.4459 1.2981 1.2981 1868.0 1900.0 1950.0 1950.0 632.22 632.22 Q,{)Z541 0.02541 0.16999 0.16999 0.10540 0.19540 666,3 666.3 475.6 475.8 1142,{) 1242.0 0.8574 0,8574 04350 {),4358 1.2931 1.2931 1958.0 1950.0 200U 2008.8 535.80 635.80 0,02565 0.02565 0.16266 0.16266 0.18831 0.10831 672.1 672.1 466.2 4662 0138,3 1138.3 {),8625 0.8625 0.4256 0.4256 1.288u 1.288i 2800.0 2000.0 2100.0 642.76 0.14885 0.17501 0,17501 683,8 446.7 1130,5 1138.5 0.8727 0.4053 1.2780 1.2780 2188.0 2108.0 2200,0 2200.0 642.76 649,45 649.45 0.02 0.0261i.
" 0.14885 0.13603 0.13683 0.16272 0.16272 683.8 695.9 695,~
446.7 426.7 426.7 1122,2 1222.2 0,8727 0.8828 0,8828 0.4053 0.3848 0.3848 1.2678 1.2676 210n,0 2280.8 2200,0
~
2308.0 2300.8 655,89 655.89 02727 0.12406 0.12406 0.15133 0.15133 707.2 707.2 406,0 406,0 liI3,2 1213.2 0.8929 0.8929 0.3640 0,3640 1.2569 1.2569 2300.0 2300,0 2400.0 2488.8 66Z.1l 662.11 . 0.11287 0.81207 0.14076 9.14076 719,0 719.6 388.8 384,8 1103.7 li03.? 0.9031 0,9031 0.3430 0.3430 1.2360 1.2460 2400.0 24DIl.0 2500.0 2500.8 668.11 668.11 0,02859 0.02859 0.10209 0.10209 0.13068 0.13060 731.7 73U 361.6 361.6 1003.3 1093.3 0.9139 0,9139 0.3206 0.3206 1.2345 1.2345 2500.8 2500.0 2600.0 2600.8 673,91 673.91 0,02938 0.02938 0.09172 0.09172 0.12110 0.12110 744,5 744.5 337.6 337.6 1082.0 1082,0 0.9247 0,9247 0.2977 0,2977 1.2225 1.2225 2688.0 2600.0 2100.0 2100.0 679.53 679,53 0,03029 0.03029 0.08165 0.06165 0.1lJ94 0.12194 757.3 75i.3 312.3 312.3 20691 1069.7 0.9356 0.9356 0.2741 0,2741 1.2097 1.2097 2708.0 2100.0 2800.0 2000.0 684,96 684.96 0,03134 0.03134 0,0717i 0,07171 O.103()5 0.10305 770,7 770.7 285.8 285.1 1055.8 1055,S 0.9368 0.9468 02491 0.2491 1.1958 1.1958 2080.8 2800.0 2900.6 2000.6 69{),22 690.22 0.03262 0.03262 0.06158 0.06258 0.09420 0.09420 785,1 785.1 254.7 254.7 1039.8 1039.8 0.9588 0,9588 0.2215 0.2215 1.1803 1,1803 2908.8 2900.0 3000.0 3080.0 695.33 695.33 0,03428 0.03426 Q,050i3 0.05073 Q.085 0.88560 801.8 801.8 210.4 218A 1020,3 1020.3 0.9728 0,9728 0.1891 0.1891 1.1629 U619 3008.8 3000.0 3100,0 3108.8 700.28 700.28 0,03681 0.03601 0,Q3i7l 0.83771 0.07452 0.07452 824,0 824.0 169.3 169,3 953.3 993.3 0.9914 0.9914 0.1460 0.1460 1.1373 1.1373 3180.8 3100.0 3200,0 3280.0 705.08 705.08 0.04472 0.04472 0,01191 0.01191 0,05663 0,05663 1l1!i.s 079.5 56.2 56.1 931.6 93L6 1.0351 1.0351 0.0482 0,0482 1.0832 1.0832 3200.8 3280,0 3208.2" 3288.2W 705,47 705.47 0.05078 0.05078 0.00000 0.00000 0.05870 0.05078 906.0 906,0 0,0 0.0 906.0 906.0 1,0612 1.0612 0.0000 0.0000 1,0612 1.0612 3288.2 3208,2"
~ ~G.tfFL-- t3p)~ f1 ~7 :z;r ~~( t:-j
~
cL- +7 :j;; I\J
*Gritical pressure
'Critlcal pressure C- fr
- ;:;: :,;-;:_ .,:.:-.-~",." "- - "'. ~-"- ."' ',.).~.-~- ' - ; ~.- -"' - --'-"'~'"'""-""~To.1J£~:>J / , r _
,,1)8:;-_
Table 3. Table 3. Superheated Superheated Steam Steam Abs Press Abs Press. LbiSq In. LbiSq In Sat. Sat Sat Sat. Temporature - Degrees Temperature Degrees fahrenheit Fahrenhm( (Sat. Temp) (Sat Terop) Water Steam Water Steam 200 200 250 250 300 300 350 350 400 400 450 450 500 500 600 600 100 700 800 808 900 900 1000 1080 1100 1108 1200 1200 Sb Sh 98.26 98.26 148.26 148.26 198.26 198.26 248.26 24826 298.26 298.26 348.26 348.25 398.26 398.26 49826 49825 598.26 598.25 69826 598.20 198.26 798.26 898.26 998.26 99826 998.20 1098.26 1098.26 11.4 0.01614 333.6 392.5 422.4 452.3 482.1 511.9 541.7 (l0174) v 0.01614 333.6 392.5 422.4 452.3 482.1 511.9 541.7 571.5 571.5 63U 631,1 690} 6907 750.3 750.3 809.8 809.1 809.4 809.4 9290 929.0 988.5 908.6
- h 69.73 69.73 1185.8 1105.8 1150.2 1150.2 1172.9 1172.9 1195.7 l!95.7 1218.7 1218.7 1241.8 1241.8 1265.1 1265.1 1288.6 1288.6 1336.1 1336.1 1384.5 1384.5 1433.7 1433.7. 1483.8 1413.8 1534.9 1534.9 1586.8 1506.0 l639.7 1819,7 o 0.1326 0.1326 1.9781 1.9781 2.0509 2.0509 2.0841 2.0841 2.1152 2.1152 2.1445 2.1445 2,1722 2.1722 2.1985 2.1985 2.2237 2.2237 2.2708 22708 2.3144 2.3044 2.3551 2.3551 2.3934 2.3934 24296 24296 2.4640 2.4540 2.4969 2.4969 Sb Sh 37.76 37.76 87.76 137.76 87.76 137.76 107.70 187.76 237.76 237.76 287.76 337.76 287.76 337.76 437.76 437.76 537.76 537.76 637.76 637.76 737.76 737.76 837.76 837.76 937.76 1037.76 937.76 1037.76 5 . v 0.01641 O.OI64J 73.53 73.53 78.14 78.14 8421 84.2190.24 90.24 96.25 96.25 (02.24 102.24 100.23 114.2]
108.23 126.15 114,21 126.15 138.08 138,08 150.01 150.01 161.94 161.94 173.86 173.86 185.78 185.78 197.70 197.70 (62.24; 062.24) 1 1131.1 130.20 113U 1048.6 1148.6 1171.7 li94.8
!l7L7 1(94.8 1218.0 1218.0 1241.3 1241.3 1264.7 1288.2 1264.7 1288.2 1335.9 1384.3 1384.3 1433.6 1483.7 1534.7 1586.7 130.20 1335.9 1433.6 1483.7 1534,7 1586.7 1639.6 1639.6 s 0.2349 1.8443 0.2349 1.8443 1.8716 1.8716 1.9054 1.9369 1.9054 1.9369 1.9664 1.9664 1.0043 1.9943 20208 2.0460 20208 2.0460 2.0932 2.0132 2.1369 2.1369 2.1776 2.1776 2.2159 2.2159 2.2521 2.2521 2.2866 2.2866 2.3194 2.3194 78 Sb Sh 6.79 6.79 56.78 56.79 006.79 106.79 156.79 156.79 206.79 206.79 256,79 256.79 306.79 306.79 406.79 406.79 506.79 506.79 606.79 606.79 706.79 706.79 806.79 806.79 906.79 906.79 1006.79 1006.79 16 v 0,01659 0.01659 38.42 38.42 38.84 38.84 41.93 41.93 44.98 44.98 46.02 48.02 51.03 51.03 54.04 54.04 57.04 57.04 63.03 63.03 69.00 74.98 80.94 86.91 92.87 98.84 (193.21) 69.00 74.98 80.94 86.91 92.87 98.84 h 161.26 161.26 1143.3 1143.3 1146.6 1146.6 1170.2 mO.2 1193.7 1193.7 1217.1 1217.1 1240.6 1240.6 1264.1 1264.1 1287.8 1287.8 1335.5 1335.5 1384.0 1384.0 1433.4 1433.4 1483.5 1483.5 1534.6 1634.6 1586.6 1586.6 1639.5 1639.5 o 0.2836 0.2836 1.7879 1.7879 1.7928 L7928 1.8273 1.8273 1.8593 L8593 1.8892 1.8892 1.9173 1.9173 1.9439 1.9439 1.9692 1.%92 2.0166 2.0166 2.0603 2.0603 2.1011 2,1011 2.1394 2,1394 2.1757 2.1757 2.2101 2.2101 2.2430 2.2430 1.4.69 Sb Sh 38.06 38.00 88.01 88.00 138.00 130.00 108.00 lS8.00 238.00 230.00 288.00 288.00 388.00 108.00 488.00 488.09 588.GO 906 Ut 688.00 088.10 788.00 768,00 8go.00 800.00 988.00 968.80
, 14.696
,212.OW (212.00) v .ow
.0167 26 799 20.199 20.42 28.42 30.52 39.52 32.60 3260 34.67 34.67 35.72 3&.72 30.77 38.77 42.86 42.86 4693 46.93 5l.0D 5(00 55.06 55.06 59.13 59.13 63.19 63.19 51.25 67.25 0 100.17 180.17 11595 1150.5 1168.4 1168.8 1192.6 1192.6 l2lU 1216.3 1238.9 1239.9 1261.6 1263.6 1287.4 1261.4 1335.2 1335.2 1383.8 1383.8 1433.2 1433.2 1483.4 1483.4 1534.5 1534.5 1586.5 1586.5 1639.4 1636.4 s .3121
.3121 1.7500 1i,68 (.7833 1.7833 1.8158 1.8188 1.8459 1.8459 1,8743 1.8743 1.8010 1.9010 1.9205 1.9265 1.9739 1.9739 2.0177 2.0177 2.0585 2.0585 2.0969 2.0989 2.1332 2.1332 2.1676 2.1878 2.2005 2.2005 Sb Sh 36.97 88.97 86.97 136.97 130.97 186.97 i86.97 236.97 286.97 288.97 386.97 386.07 48697 486.97 586.97 686.97 786.97 706.97 88697 066.97 985.97 15 15 0.01673 26.290 27.937 33.962 t2L.03(
27837 29.899 31.939 33.963 35.977 37.985 41.986 41,986 45.978 49.984 49.964 53.946 57.926 6L905 61.905 55.882 65.882 (213.03) 4 101.21 181.21 1150.9 1150.9 1166.7 11687 11915 1192.5 1216.2 1215.2 1239.9 1239.9 1263.6 1263.6 1287.3 (287.3 13352 1335.2 1383.8 1383.8 14312 1433.2 1483.4 14814 1534.5 1534.5 1586.5 1639.4 1586.5 1639,6 o 0.3137 1.7552 i7552 1.7809
!.7809 18134 1 8134 1,0437 1.8437 1.8720 1.8720 1.8988 1.8988 1.9242 1,9262 1.9717 1.9717 2.0155 2.0563 2.0946 2.0066 21309 2.1309 2.l653 2.1653 2.1982 Sb Sh 22,04 22.04 72.04 122.04 172.04 172.04 222.04 272.04 372.04 37204 472.04 572.04 672.04 772.04 872.04 872.34 972.04 28 20 , 6.01683 0.01683 20.087 20081 227.96) 20.788 22355 22.356 23,900 23.900 25428 26.945 26.948 28.457 28,457 31.466 31.456 34.465 37.458 40.447 43435 43.435 46.420 46.470 49.405 19,405 i227.961 b 19627 196.27 0156,3 1156.3 !1167.1 167.1 1191.4 1191.4 0215.4 1215.4 (2392
,2392 1263.0 1263,0 1286.9 1266.9 1334.9 1383.5 1383.5 14329 14329 1483.2 1403.2 1534.3 1534,3 1586.3 1639.3 1506.3 1638.3 s 0.3358 1.7320 L7320 1,7475 1.7475 1.7805 Ll80S 1 8111 l.8111 1 8397 L8397 1.8666 L8666 1.8921 1 8921 1.9397 1.9836 1.9836 2.0244 2.0628 2,0626 2.0991 2.!336 2.1336 2.1665 Sb Sh 9.93 59.83 59.93 109.93 155.33 159.93 201.83 209.93 259.93 258.93 359.93 459.93 559.93 659.93 659.03 759.93 759,13 859.93 859,93 95993 25 V 0.01693 0.01593 16.301 16.558 (7.828 17.829 18.176 19.076 20.307 21.527 22.740 25.153 27.557 29.954 29.956 32.348 34.740 36.740 37130 32.130 39.518
,240.0. (
(240.Q7) h 206.52 208.52 1160.6 mO.6 1165.6 1190.2 1190.2 12105 1214.5 1238.5 1262.5 1262.5 1286.4 1285.4 1334.6 1383.3 1432.7 1483.0 14810 1534.2 15862 1506.2 1639.2 S 0.3535 1.7141 1.7212 1.7547 L7547 1.7858 1.7856 1,8145 1.8145 1 8415 1.8415 1.8672 1.0872 1.9149 1.9588 1.9580 1.9997 2.0381 2.0744 21089 2.1089 2.1418 2 1418 Sb Sh 49.66 89.66 99.66 149.66 199.66 249.66 249,66 349.66 349,06 449.66 44906 549.66 649.66 749.66 741.66 84966 649.66 949.66 30 , i o.oi7m 0.01701 13.744 14.010 14.810 15.359 15.859 16.892 17.914 18.929 20.945 22.951 24.952 26.949 28.943 28943 30935 30.036 32.927 250.34 (250.34) h 216.93 218.93 1(64.1 1164.i 1199.0 1189.0 1213.6 1237.6 1237.8 1261.9 1286.0 1288.0 1334.2 i383.0 13830 i432.5 1432.5 1482.8 1534.0 1534.6 1586.1 1639.0 s 0.3682 0.3582 1.6965 1.6995 1.7334 1.7647 1.7937 1.8210 1.8467 1.8946 i.9386 1.9386 1.9795 1.9765 2.0179 20543 2.0543 2.0888 2.0680 2.!217 2,1217 61, Sh 48 AO.711 907 90.71 407 140.71 190]1 199,1 240.71 240,1 34071 340]) 440.71 440 1 54Q.71 SOil 640.71 640,1 740)1 40/ 840)1 84071 940.71 94071 35 u 001708 Ml108 11.896 12.654 13.562 14.453 15.336 15.334 16.207 17.939 19.662 21.379 23.092 24.803 26.512 28.220 259 (259.29) 1167.1 228.03 !l6ll 1187.8 1212.7 1237.1 1261.3 1285.5 1333.9 13319 1382.8 1432.3 1482.7 1533.9 1586.0 1638.9 1538.9 0.3809 1.6872 1.7152
!.715;> 1.7468 1.7761 1.8035 1.8294 1.8774 1.9214 1.9624 2.0009 2.0372 2.0717 2.1046 48 Sh 32.75 82,75 82.75 132.75 l82.75 232.75 182.75 33275 432.75 43275 53215 53275 632.75 732.75 832.75 93V5 932.75 46 , 0.01715 0.497 10.497 11.035 lL036 11.339 1l.838 12.624 13.398 14.165 15685 15.685 17195 17.195 18.699 20199 20 199 21697 23194 23.194 24.689 (267.25, (26725) 236.14 1160.8 1169.8 1186.6
!la6.6 1211.7 121i.7 1236.4 1236.4 1260.8 1285.0 1260.8 13336 13336 1382.5 1432.1 1432.1 1482.5 1482.5 15337 J5337 15858 16388 0.3921 0.3921 1.6765 1,6992 1.6992 1.7312 1.7312 1 7608 1.7608 1.7883 1.8143 1,8624 1.8624 1.9065 1.8065 1.9476 1.9860 1.9860 2.0224 2.0569 2.0899 20899 45 Sb Sh 25.56 75.56 125.56 125.56 175.56 225.56 325.~6 325.56 425.56 525.56 625.56 725.56 825.56 925.56 (27444) V -3.01721 0.01721 9.399 9.777 10.497 11.201 11.201 11.892 12.577 13.932 15.276 16.614 16.614 17.950 19.282 19.262 20.613 20.513 21.943 243.49 1172.1 1172.l 1185.4 1185.4 121004 1210.4 1235.7 1260.2 1284.6 1333.3 1382.3 1382.3 1431.9 1482.3 1533.6 1585}
1585.7 1638} 1638.7 S 0.4021 1.6671 1.6671 1.6849 1.7173 J.7173 1.7471 U471 1.7748 1,8010 1.8010 1.8492 1.8934 1.9345 1.9730 2.0093 2.0439 2.0768 59 Sb Sh 18.98 18.98 68.98 118.98 168.98 218.98 2i8.98 31898 31898 418.98 518.98 51898 618.98 718.98 818.98 918,98 918.98 (281.02) u 0.01727 8.514 8.769 9.424 10.062 10.688 11.306 lL306 12.529 13.741 13741 14.947 16150 16.150 17358 17.350 18.549 19746 19746 250.21 250.21 1174,1 1I74.l 1184.1 li84.i 1209.9 1209.9 1234.9 1234.9 1259.6 1259.6 1284.1 1284.1 1332.9 1332.9 1382.0 1382.0 1431.7 1431.7 1482.2 1482.2 1533,4 1533.4 1585 S 15856 1638.6 1638.6 S 0.4112 OAll2 1.6586 1.6586 1.6720 1.6720 1.7048 1.7048 1.7349 1.7349 1.7628 L7628 1.7890 1.7890 1.8374 1.8374 1.8816 1.8816 1.9227 1.9227 1.9613 1.9613 1,9977 1.9977 2.0322 2.0322 2.0652 2.0652 55 - Sb . 12.93 12.93 62,93 62.93 112.93 112.93 162.93 162.93 212.93 m.93 312.93 312.93 412.93 512,93 512.93 612,93 612.93 712.93 812.93 812.93 912.93 912,93 (281.01) (287.011 Shv 0.01733 0,91733 li87 7,945 7.945 8.546 8.546 9.130 9.130 9.762 9.702 10.267 10.267 11.381 11.381 12.485 12.485 13.583 13.583 14,677 14.677 15,769 15769 16.859 15.859 17.948 17.948 0 256.43 256.43 i,,76.0 illS.O 1182.9 1182.9 1208.9 1208.9 1234.2 1234.2 1259.1 1259.1 1283.6 1283.6 1332.6 1332,6 1381.8 1381.0 1431.5 1431.5 1482.0 1482,0 1533.3 1533.3 1585.5 1585.5 1638.5 1638,5 o 0.4196 0.419& 1,0510 1.6510 1.6601 1.6601 1,6933 1.6933 1,7237 1.7237 1.7518 1.7518 9.7781 t7781 1,8266 1.8266 1.8710 1.8710 1,9121 1.9121 1.9507 1.9507 1.987 1.987 2.022 2.02? 2.055 2.055 68 60 Sb Sh 7.29 729 57.29 57.29 107.29 107.29 157,29 157.29 20729 20729 30729 30729 407,29 407.29 50729 507.29 607.29 607.29 707.20... 707.29 ... 807.25 807.29 907.29 987.29 (292.71) vv 0.01738 0.01738 7.174 7.174 7.257 7.257 7.815 7.815 8.354 8.354 8.881 8.881 9.400 9.400 10.425 10.425 11.438 11.438 12.446 J2.446 13.450 13.450 14.452 14.452 15.452 15.452 16.450 16.450 h 262.21 262.21 1177.6 1177.6 1181.6 1181.6 1208.0 1208.0 1233.5 1233.5 1258.5 1258.5 1283.2 1283.2 1332.3 1332.3 1381.5 1381.5 1431.3 1431.3 1481.8 1481.8 l!i33.2 1585.3 1033.2 1585.3 1638.4 1638.4 0.4273 0.4273 1.6440 1.6440 1.6492 1.6492 s.6829 1.6829 1.7134 1.7134 1.7417 1.7417 1,7681 1.7681 1.8168 1.8168 1.8612 1.8612 1.9024 1.9024 1.9410 1.9410 1.9774 2.0120 1.9774 2.0120 2;045.0 2W45,0 Sb Sh 2.02 2.02 52.02 52.02 102.02 102.02 152.02 152.02 202.02 202.02 302.02 302.02 402.82 402.02 502.02 502.02 682.02 602.02 702.02 802.02 702.02 802.02 902.02 902.02 65 0.01743 0.01743 6.653 6.653 6.675 6.675 7,195 7.195 7.697 7.697 8.186 8.186 8.667 8.667 9.615 9.615 10,552 10.552 11,484 11.484 12.412 12.412 13337 14.261 13.337 14.261 15.183 15.183 (297,98) 1297.98) h 267.63 267.63 1179.1 1179.l 1180.3 1180.3 1207.0 1207.0 1232.7 1232.7 1257.9 1257.9 1282.7 1282.7 1331.9 1381.3 14.31.1 1331.9 1381.3 143U 1481.6 1481.6 1533.0 15852 1533.0 1585.2 1638.3 1638.3
$ 0,4344 0.4344 1.6375 1.6375 1.6390 1.6390 1.6731 1.6731 1.7040 Ll040 1.7324 1.7324 1.7580 Li590 1.8077 1.8077 1.8522 1.8522 1.8935 1.8935 1.9321 1.9321 2.0031 1.%85 2.0031 1.9685 2.0361 2.0361 Sb Sh 47,07 47.07 97.07 97.07 14/.07 141.07 197.07 197.07 297.07 397.07 497,07 297.07 397.07 497.07 597.07 597.07 697.07 697.07 797.Q7 797.07 897.07 897.07 78 10 v 6.01748 0.01748 6.205 6.205 6.664 6.564 7.133 7.133 7.590 7.590 8.039 8.039 8.922 8.922 9.793 10.659 9.793 10.659 11.522 11.522 12.382 12.382 13.240 13.240 14.097 14,097 (302.931 (302.93) 8h 272.74 272]4 1186.6 1180.6 1206.0 1206.0 1232.0 1232.0 1257.3 1257.3 1282.2 1282.2 1331.6 1381.0 1438.9 1331.6 1381.0 1430.9 1481.5 1481.5 1532.9 1532.9 1585.1 1585.1 1638.2 1635.2 o 0.4411 O.4411 1.6316 1.6316 1.6640 1.6640 1.6951 1.6951 1.7237 1.7237 1.7504 L7504 1.7993 1.8439 1.8852 1.7993 1.8439 1.8852 1.9238 1.9238 1.9603 1.9603 1.9949 1.9949 2.0279 2.0279 Sb Sh 4239 4239 92.39 92.39 142.39 142.39 192.39 192.39 292.39 292.39 392.39 392.39 492.39 492.39 592.39 592.39 692.39 592.39 792.39 892.39 792.30 892.39 75.
15 v 0.01753 0.01753 5,814 5.814 6.204 6.204 6.645 6.645 7.074 7.074 7.494 7.494 8.320 8.320 9.135 9.135 9.9459.945 10.750 10.750 11.553 11.553 12.355 13.155 12.355 13.155 (307,ol) (307.61) 1 277.56 277.56 1181.9 118J.9 1205.0 1205.0 1231.2 1231.2 1256.7 1256.7 128L7 1281.7 1331.3 1380.7 14.30.7 1331.3 1380.7 1430} 1481.3 1481.3 1532.7 1532.7 1585.0 1638.1 1585.0 1638.1 0.4474 0.4474 1.6260 1.6260 1.6554 1.6554 1.6068 1.6868 1.7156 1.7156 1.7424 1.7424 1.7015 1.7915 1.8361 l.8361 (.8774 1.8774 1.9161 1.9161 1.9526 1.9526 1.9872 L9872 2.0202 2.0202 superheat. FF Sh ="" superheat. Sh enthalpy, Btu hh == enthalpy, Btu per per lbIb vv == speoific specific volume, volume, Cu cu ftIt per per lbIb ss = entropy, entropy, BtuBtu perper RR per per lbJb
- ~f' -, 1
- 27. 027Kl.Ol
- 27. 027K 1.01 OOllNEW/RO/MEM 001 /NEW/RO/MEM 3.4/3.7!N!N12/CVRIY 3.4/3 .7fN/N/2/CVRIY Which one Which one of theof the following following correctly correctly states states how the Containment how the Containment Spray Spray System System reduces radioactive reduces radioactive iodine iodine inin the the Containment Containment atmosphere atmosphere during during aa LOCA?
LOCA? To enhance To enhance absorption absorption of Iodine from of Iodine from the the Containment Containment atmosphere, atmosphere, thethe Containment Containment Spray System Spray sprays water System sprays water from from the the (1) (1) at aa pH at pH of approximately (2) of approximately (2) (1 (1)) (2) (2) A. containment sump 4.5 4.5 B. RWST 7.5 C~ C containment sump 7.5 D. RWST 4.5 Incorrect. First part correct, see C. A - Incorrect. second part - 4.5 pH is incorrect. Plausible, since the Borated water from the RWST in the injection phase is a pH of approx. 4.5 due to the 2300-2500 ppm borated water. However, the recirc phase begins the spray of the sump water which has the Tn-Sodium Phosphate in it, and the pH of that water is higher at 7.5 to dissolved Tri-Sodium 10.5. B - Incorrect. The 7.5 is correct, but the RWST is acidic at a pH of 4.5 due to the high concentration of boric acid in. The low pH is not conducive to absorbing the iodine. The Iodine is absorbed during the recirc phase when the CS takes a suction on the Containment sump after the TSP has dissolved and raised the pH of the Spray water. Plausible, if confusion exists as to the need for the pH to be higher in order to absorb the Iodine out of the containment atmosphere. C - Correct. The TSP in the Containment Sump dissolves in the Containment sump C water during the injection phase, and raises the pH from about 4.5 to aa range of 7.5-10.5. During the Containment Spray recirc phase, this causes the iodine in the containment atmosphere to be absorbed in the spray water and convert to aa non-volatile form. Then, it stays in the sump water, and does not leak out of containment via any ctmt atmosphere leakage paths. Even though some iodine would be absorbed by the mechanical action of the spray water in the containment atmosphere, the higher pH enhances the effect, and the retention of the iodine in the sump water is due to the higher pH. D D - Incorrect. Both parts are incorrect (see A &
- & C). Plausible, since the pH is is correct for the RWST source, but this pH is not conducive to removing iodine iodine from the the containment atmosphere. Confusion Confusion may exist as to the exact mechanism mechanism of iodine iodine removal by the CS system.
TS TS B3.5.6 83.5.6 FSD FSD Al 81 008, CS A 181008, CS System System Page: 70 Page: 70 of of 200 200 12/1412009 12/14/2009
2.0 SYSTEM FUNCTIONAL REQUIREMENTS The safety-related function of the CSS is to reduce the containment building pressure and temperature following a LOCA or high-energy line rupture and to reduce airborne fission products in the containment atmosphere following a LOCA. During the injection phase, the CSS pumps are aligned to take suction off the RWST. When the RWST reaches low-low level, the spray pumps operate in the recirculation mode from the containment sump. Operator action to perform realignment of the CSS pumps to sump recirculation must be completed within 130 seconds of reaching the RWST low-low level setpoint. Completion of this operator action in 130 seconds ensures sufficient volume remains in the RWST to ensure adequate pump NPSH is available and 6.3.020, 6.7.039). Trisodium phosphate to prevent vortexing in the RWST (References 6.3.020,6.7.039). (TSP) filled baskets in the recirculation area of containment provide iodine absorption 6.2.00 1, 6.3.001, 6.7.001). and retention in the containment sump solution (References 6.2.001, 6.7.00 1). As the RCS inventory combined with Eecs ECCS solution accumulates in the recirculation sump, the rising water level dissolves the TSP crystals in the baskets (References 6.7.033 and 6.7.034). The spray water is maintained at a pH level of approximately 4.5 during injection. During recirculation, a pH of approximately 7.5 enhances the absorption of the airborne fission product iodine, retains the iodine in the containment sump solution, and minimizes potential for chloride induced stress corrosion cracking (References 6.1.001, 6.2.001, 6.3.001, 6.3.017). 6.2.001,6.3.001,6.3.017). The development of the iodine removal coefficient is a function of the characteristics of the CSS. The design value of the iodine removal coefficient is 10 hr -1.-i. This coefficient is based on one CSS pump operating at a flow rate of2,200 of 2,200 gpm, and a spray fall height of 110 ft (References 6.3.018, 6.7.003). 3.1 CSS PUMPS 3.1.1 Basic Functions Post-LOCA, the CSS pumps shall deliver borated water from the RWST during the injection mode, water from the containment sump and trisodium phosphate from the TSP baskets during the recirculation mode, to the containment spray ring headers (References 6.2.00 1, 6.7.033, 6.2.001, 6.7.034). Page: 71 of 200 12/14/2009 12/1412009
Previous NRC Previous NRC examexam history history ifif any: any: Wrote Wrote aa new new question question and and intentionally intentionally stayed stayed away from away the 2008 from the 2008 nrc nrc exam exam question question on on k/a k/a 027G2.1.27, 027G2.1 .27, CS&COOL-40302D02 CS&COOL-40302D02 17 17 to to prevent going prevent over the going over limit of the limit of 44 RORO questions questions from from thethe previous 2 NRC exams. previous 2 NRC exams. 027K1 .01 027K1.01 027 Containment Iodine Removal System 027 System KI Knowledge Kl Knowledge of of the the physical physical connections connections and/or andlor cause-effect cause-effect relationships relationships between between the the CIRS CIRS and and the the following systems: (CFR: following systems: (CFR: 41.2 to 41.9 41.2 to 41.9 1/ 45.7 to 45.8) KLO1 CSS K1.01 CSS ........................................................... 3.4* 34* 3.7* 37* Match justification: To answer this question correctly, the physical connections to the Match Iodine Removal and the CSS (only (only connected during the CS recirc phase taking a suction from the Sump instead of the RWST) RWST),, and the knowledge knowledge of the TSP (iodine removal) cause-effect on the CSS of adjusting the pH FROM FROM 4.5 TO 7.5 or greater is required. Objective: 11 LABEL AND ILLUSTRATE the Emergency Core Cooling System to include the components found on the following figures (OPS-40302C05):
- Figure 2, Accumulators
- Figure 3, Refueling Water Storage Tank and Figure 4, Emergency Core Cooling System
- The flow paths found on Figure 14, ECCS Injection Phase, Figure 15, ECCS Cold Leg Recirculation, Figure 16, ECCS Simultaneous Hot & & Cold Leg Recirculation Normal, and Figure 17, ECCS Simultaneous Hot & & Cold Leg Recirculation Alternate.
2 LABEL AND ILLUSTRATE the Containment Spray and Cooling System flow paths, to include the components found on Figure 2, Containment Cooling System, Figure 3, Containment Spray System and Figure 4, Service Water to Containment Coolers (OPS-40302D05).
- 2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Containment Spray and Cooling System to include the components found on Figure 2, Containment Cooling System, Figure 3, Containment Spray System and Figure 4, Service Water to Containment Containment Coolers and the following (OPS-40302D02):
- Containment Cooler Service Water Inlet Isolation Valves (MOV-3019A, B, C, and D)
- Trisodium Phosphate Baskets Baskets Page: 72 of Page: 72 200 of200 12/14/2009 12/14/2009
ECCS Recirculation ECCS Recirculation Fluid Fluid pHpH Control Control System System B B 3.5.6 3.5.6 B 3.5 B 3.5 EMERGENCY EMERGENCY CORE CORE COOLING COOLING SYSTEMSSYSTEMS (ECCS) (ECCS) B 3.5.6 B ECCS Recirculation 3.5.6 ECCS Recirculation Fluid pH Control Fluid pH Control System System BASES BASES BACKGROUND BACKGROUND The Recirculation The Recirculation Fluid Fluid pH Control System pH Control System isis aa passive passive system system designed to to raise the long long term pH of the solution in in the containment sump following a Design Basis Accident Accident (DBA). The Recirculation Fluid Fluid pH Control System consists Control System consists of of three storage storage baskets containing containing trisodium phosphate phosphate (TSP) (TSP) as as Na P3O 3 Na 4 P04 ** 12H 0 2 12H 2 0 ** Y-tNaOH. 1/4NaOH. An equivalent amount of equivalent of trisodium phosphate phosphate compound compound with aa different chemical formula may be used. When equivalent compounds are used, the allowable weights/volumes may be be different; however, the equivalent amount of trisodium phosphate comp' comp raise the pH of the recirculating solution into the range 0o 7.5 to 10.5. 10.5. In the event of a loss of coolant accident (LOCA), the SP contained in the storage baskets will be dissolved in the Reactor Coolant S stem RCS (a"ildFefueling Water Storage Tank (RWS Inventories inventories lost through fT1 pipe tJ:ie'"pTpe break. The resulting increase in the ecirculation so u ion H into the range of 7.5 to 10.5 assures that iodine is retalne retaine In in solution and that chloride induced stress corrosion on mechanical systems and cI Recirculalion Fluid pH Control components is minimized (Ref. 1). The Recirculation System performs no function during normal plant operation. Radioiodine in its various forms is the fission product of primary concern in the evaluation of a DBA. Fuel damage following a DBA will cause iodine to be released into the reactor coolant and containment r,,-s::'/A' 'i-:; '<""--f;; -¥ \ I7vJ 7') \.J ~ 0- L
~ i o¥~ atmosphere. Iodine released to the containment atmosphere is absorbed by the containment spray and washed into the containment sump. Since the ECCS water is borated for reactivity control, the recircuiation solu . .
with aj2fapiEiateIy pH of approximately 4.5. containment sump will initiall be acidic
- 5. In aa low pp (acidic) solution, some of
() H::.J/. .7 the dissolved iodine will be converted to aa volatile form and evolve out t3eC0nverted lG '}. ~ of solution into the containment atmosphere. In order to reduce the
- yrp.
~ v- lP "
_A potential for elemental iodine evolution, the ECCS recirculaton is adjusted (buffered) to achieve aa long recirculation solution long term alkaline pF-I of no les~ pH ot fc ,than 7~. than 7.5. An alkaline pH promotes iodine hydrolysis, in which iodine is
'1.- ~ ,..converted to nonvolatile nonvolatile forms. In addition to ensuring iodine is retained in solution, an alkaline recirculation solution will minimize chloride induced induced stress corrosion cracking of austenitic stainless steel (continued)
(continued) Farley Units 11 and Farley Units and 22 BB 3.5.6-1 3.5.6-1 Revision 00 Revision
- 28. 029EAl.06
- 28. 029EA 1.06 OOllMOD/ROICIA 00 1/MOD/RO/C/A 3.2/3.11N1N12ICVRIY 3.2/3. 1/N/N/2/CVRIY Unit 2 has experienced an Anticipated Transient Unit 2 has experienced an Anticipated Transient Without Without Trip Trip (ATWT)
(ATWT) and and the the following plant following plant conditions conditions occurred: occurred:
** Charging Flow Charging Flow is 68 gpm.
is 68 gpm.
** 2A Boric Acid 2A Boric Transfer Pump Acid Transfer Pump isis tagged tagged out.
out.
** Injection has Safety Injection has NOT NOT actuated at at this time.
** lAW FRP-S.1, Response lAW FRP-S.1, Response to Nuclear Nuclear Power Power Generation - ATWT, the UO UO is is Emergency Boration.
establishing Emergency Boration.
** 2B Boric Acid 2B Boric Transfer Pump Acid Transfer tripped when Pump tripped when itit was was started.
started. Which one of the following states: I) the required
- 1) required actions to establish an emergency boration boration flow path, path, and
- 2) the MINIMUM required action for FK-122, CHG FLOW controller, lAW FRP-S.1?
A. 1) Open V185, MAN EMERG BORATION valve, AND open FCV-113A, BORIC ACID TO BLENDER valve.
- 2) Place FK-122 in MAN ONLY.
B~ B I) Open LCV-115B
- 1) LCV-I 15B and 0, D, RWST TO CHG PUMP valves, AND close LCV-115C LCV-1 150 and E, VCT OUTLET ISO valves.
- 2) Place FK-122 in MAN AND raise demand.
C. 1) Open V185, VI 85, MAN EMERG BORATION valve, AND open FCV-113A, FCV-1 1 3A, BORIC ACID TO BLENDER valve.
- 2) Place FK-122 in MAN AND raise demand.
D. 1) 1) Open LCV-115B LCV-l 1 5B and D, 0, RWST TO CHG PUMP valves, AND close LCV-l LCV-115CI 5C and E, VCT OUTLET ISO valves.
- 2) Place FK-122 in MAN ONLY.
Page: 73 Page: 73 of of 200 200 12/14/2009 12/14/2009
AA - Incorrect.
- Incorrect. The The first part isis incorrect, first part incorrect, since since the Manual Emergency the Manual Emergency Borate Borate flowpath flowpath will not will work in not work in this this situation.
situation. Neither Neither BATBAT pump pump is is available, available, and and at at least least one one isis required to required to use use either either the the normal normal emergency emergency or or manual manual emergency emergency borate borate flowpath. flowpath. Plausible, since Plausible, since in other situations in other situations withwith aa loss loss ofof the the normal normal emergency emergency borate borate flowpath, the MANUAL flowpath, the MANUAL emergency emergency borate borate flowpath flowpath would would bebe used used per per FRP-S.1 FRP-S.i Step 4.3 Step 4.3 RNO. RNO. The The second second part is incorrect part is incorrect duedue toto the the charging charging flow flow being being less less required (92 than required (92 gpm) with the RWST RWST boration boration flowpath aligned, aligned, and the charging will have to bebe increased increased by by adjusting FK-122 FK-122 in in the raise raise direction per per FRP-S.1, FRP-S.1, 4.6 & 4.7.3. Plausible, Step 4.6 Plausible, since since the charging charging flow is greater than requiredrequired for the normal emergency normal emergency or or manual emergency borate manual emergency borate flowpath flowpath (40 (40 gpm) gpm) per per FRP-S.1 FRP-S.1,, step 4.6. B - Correct. Per FRP-S.1 Steps 4.2.1 RNO,
- RNO, the RWST boration boration flow path path will be aligned due to the inability to start either BAT BAT pump. The flow from the RWST RWST to RCS is required to be >> 92 gpm per step 4.6, thus the charging demand must be the RCS raised in manual.
C - Incorrect. The first part is incorrect (see A). The second part is correct (see B). D Incorrect. The first part is correct (see B). The second part is incorrect (see A). 0- - FRP-S.1 Revision 25 Previous NRC exam history if any: 029EA1 .06 029EA1.06 029 Anticipated Transient Without Scram (A (ATWS) TWS) EAI Ability to operate and monitor the following as they apply to a ATWS: (CFR 41.7 145.5/45.6) EA1 /45.5 /45.6) EA1.06 Operating switches for normal charging header isolation valves .......... 3.2* 3.1 Match justification: To answer this question the applicant must know what to do with the operating switch for normal charging header isolation valve (Operate & monitor: FCV-122 Controlled by FK-122 controller), in response to the flow indication on Fl-i FI-122 22 in the given situation during an ATWS. Objective:
- 6. EVALUATE plant plant conditions conditions andand DETERMINE if any system components components need to be operated operated while performing performing (1) FRP-S.1, FRP-S.l, Response to Nuclear Power Generation/ATWT; Generation/ATWT; (2) FRP-S.2, Response to Loss of Core Shutdown. (OPS-52533A06)
(OPS-52533A06) Page: 74 Page: 74 of 200 of200 12/14/2009 12/14/2009
\\,) j\.i i 1 "
FNP-1-FRP-S.1 FNP-1-FRP-S.1 RESPONSE RESPONSE TO TO NUCLEAR POWER GENERATION/ATWT GENERATION/ATWT Revision 25 25 Step Step Action/Expected Response Action/Expected Response Response Response NOT NOT Obtained Obtained n I I 33 Verify AFW A1?W pumps - RUNNING.
- RUNNING.
3.1 MDAFWPs - RUNNING [1 1A amps > [] > 0 [1 1B [] lB amps > > 0 3.2 TDAFWP - RUNNING IF NECESSARY
** TDAFWP TDAFWP STM SUPP FROM 1B (IC) SG 1B(1C)
[I MLB-4 1-3 lit [] El MLB-4 2-3 lit [] [1 MLB-4 3-3 lit []
** TDAFWP SPEED TDAFWP SPEED El
[] SI 3411A > > 3900 rpm
** TDAFWP TDAFWP SPEED CONT El SIC 3405 at 100%
[] NOTE:
- 2500 gallons of emergency boration is required for each control rod not fully inserted, up to a maximum of 17,309 gallons.
* [CA]
rCA] Emergency boration should continue until an adequate shutdown margin is established. 44 Initiate Emergency Boration of the RCS. ReS. 4.1 Verify at least one CHG PUMP - - RUNNING. Step 44 continued Step continued on on next next page. page. Page Completed _Page Completed Page 55 of Page of 19 19
FNP-1-FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT CENERATION/ATWT Revision 25 Step Action/Expected Response Response NOT Obtained n 4.2 Start a boric acid transfer 4.2 Perform the following. pump. pump. ;J 4.2.1 Align charging pump suction Suction [] BAT BATP ~ El 1A .---;v
.7-/2 lit? }~ to RWST.
[]El 1B~T\I'f',f-£o 1B_ Fc a> RWST RWS T TO CRG CRC PUMP ((1] Q1E2ILCV1l5 Q1E21LCV115B B open [][1 Q1E2ILCV1l5D Q1E21LCV115D open VCT OUTLET ISO (()) Q1E21LCV1l5C Q1E21LCV115C closed [][1 Q1E21LCV1l5E Q1E21LCV115E closed 4.2.2 Proceed to step 4. 4.3 Align normal emergency Perform the following. boration. ~
- Align charging pump suction EMERG BORATE to RWST.
TO CRG CRC PUMP SUCT [] Q1E21MOV8104 O1E21MOV81O4 open RWST TO CRG CRC PUMP [] (1 Q1E21LCV1l5B Q1E21LCV115B open [] El Q1E21LCV1l5D Q1E21LCV115D open VCT OUTLET ISO ((1] Q1E21LCV1l5C Q1E21LCV115C closed ((1] Q1E21LCV1l5E Q1E21LCV115E closed
- Align manual emergency boration flow path.
BORIC ACID TO BLENDER [] El Q1E21FCVl13A O1E21FCV113A open MAN EMERG BORATION [] Q1E21V185 open Ii 01E21V185 (100 ft. AUX BLDG rad-side chemical mixing tank area) Step 44 continued on next pagepage. _Page Completed Page Page 66 of 19
~)(. "",.!.1; 1'1 FNP-1-FRP-S.1 FNP-l-FRP-S.l RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 25 Step Action/Expected Response Response NOT Obtained n I I I 4.4 Establish adequate letdown.
4.4.1 Verify 45 GPM letdown orifice - IN SERVICE. LTDN ORIF ISO 45 GPM [] Q1E21HV8149A 01E21HV8149A open 4.4.2 Verify one 60 GPM letdown orifice - IN SERVICE. LTDN ORIF ISO 60 GPM [] [1 Q1E21HV8149B 01E21HV8149E open [] [1 QlE21HV8149C Q1E21HV8149C open 4.5 Check pressurizer pressure 4.5 Verify PRZR PORVs and PRZR LESS THAN 2335 psig. PORV ISOs - OPEN.
- IF NOT, THEN open PRZR PORVs and PORV ISOs as necessary until pressurizer pressure less than 2135 psig.
4.6 Establish adequate charging flow flow..
- IF boration is from boricbor~c '(/~
acid ac i storage tank, fJ
~HEN verify chargingg flow -
TBEWTTFin ~
.f~
I It-GREATER THAN _40 40 GPM GPM... ( R- q-- 4- {) l. j\.. J. 42 OR
- IF boration is from th~ the RWST, THEN verify charging GREATER THAN 92 GPM.
Step 44 continued on next page. _Page Page Completed Page 77 of 19
~<) !.'.; fi. .I.
FNP-1-FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWT Revision 25 Step Action/Expected Response Response NOT Obtained n 4.7 Verify emergency boration flow adequate. 4.7.1 IF normal emergency boration flow path aligned, THEN check emergency 2 ft4_-/ 3k-boration flow greater than 30 GPM. GPM
~CID~
BORIC ACID EMERG BORATE EMERG BORATE [] [1 FI Fl 110 4.7.2 IF manual emergency boration flow path aligned, THEN check boric acid flow greater than 30 GPM. MAKEUP FLOW TO CHG/VCT CHGIVCT [] [1 BA FI Fl 113 4.7.3 IF boration is from the RWST, THEN verify charging flow - JIIN - GREATER THAN 92 GPM. Page Completed Page Page 8 of 19
- 1. 029EAl.05
- 1. 029EA1.05 OOIlI/l/ATWT OO1/1/l/ATWT - BIT
- BIT OUTLET OUTLET SW/CIA SW/C/A - 3.7/MODIFIEDIRlNRC
- 3.7/MODIFIED/RC RO/TNT RO/TNT 1/ RLM RLM 029EA1.O 029EA1.0 The following The following plant plant conditions conditions exist:
exist: Unit 22 has
- Unit
- has experienced experienced an Anticipated Transient an Anticipated Transient Without Without Trip Trip (ATWT)
(ATWT) and and has has implemented FRP-1 9211, Response implemented FRP-19211, Response to to Nuclear Nuclear Power Power Generation Generation ATWT. ATWT.
- A
- A Charging Charging Pump Pump is is running.
running.
- Boric
- Transfer Pump Boric Acid Transfer Pump # 11 isis tagged tagged out.
out.
- Boric Acid Transfer Pump # 2 trips on start.
has NOT
- SI has
- NOT actuated at this time.
- The SS has directed the RO to establish Emergency Boration in accordance with SOP-i 3009, "CVCS SOP-13009, CVCS Reactor Makeup Makeup Control System".
System. Which ONE of the following actions would establish a CORRECT emergency boration flow path in accordance with the SOP? (Assume 12 gpm seal return flow) A. 1) Open HV-8104 EMERGENCY BORATE Valve.
- 2) Adjust charging flow controller FIC-0121 to obtain> obtain > 42 gpm flow through the Normal Charging flow path.
LV-01 12D and LV-0112E B. 1) Open LV-0112D LV-01 12E RWST TO CHARGING PUMP SUCT valves.
- 2) Adjust charging flow controller FIC-0121 to obtain> obtain > 42 gpm flow through the Normal Charging flow path.
C. 1)1) Open FV-I1OA FV-110A BAto BA to Blenderand Blender and FV-1IOB FV-110B BLENDER OUTLETTO OUTLET TO CHARGING PUMPS SUCT.
- 2) Adjust charging flow controller FIC-Oi2i FIC-0121 to obtain> 100 gpm flow through the Normal Charging Path.
D~ 1) D 1) Open LV-01 LV-0112D12D and LV-01 12E RWST TO CHARGING PUMP SUCT valves LV-0112E and HV-880iA and HV-8801A and HV-8801B BIT BIT DISCHARGE ISOLATION valves.
- 2) Verify BIT BIT flow (Fl-0917A),
(FI-0917 A), plus total seal injection flow, minus total seal seal return flow is> 100 gpm. is > 100 gpm. Page: of 3 Page: 1lof3 10/26/2009 10/26/2009
KIA KIA 029 Anticipated 029 Anticipated Transient Transient Without Without Scram Scram (ATWS): (ATWS): EAI .05 Ability EA1.05 Ability to to operate operate and and monitor monitor the the following following as as they they apply apply to to an an ATWS. ATWS. BIT outlet valve BIT outlet valve switches. switches. MATCH ANALYSIS KIA MATCH KIA Question gives aa plausible Question gives plausible scenario scenario with anan ATWT in in progress. progress. Neither Neither Boric Boric Acid Acid Transfer Pump is available. Candidate must choose a correct emergency emergency boration flow path that would achieve Emergency Boration Flow. ANSWER I DISTRACTOR ANALYSIS A. Incorrect. Without BA Transfer Pumps available there would be no flow through HV-8104. HV-81 04. Plausible the candidate may not realize BA Transfer Pump impact on flow path. Flow rates given would satisfy the flow path if BA Transfer Pumps available. LV-1 12D and LV-112E B. Incorrect. LV-112D LV-1 12E would satisfy the flow path requirements but the minimum flow requirement via this path would be 100 gpm. Plausible candidate could recognize a correct flow path but confuse the flow rate requirements. FV-01 1 OA C. Incorrect. FV-011 FV-01 1 OB would not have flow through this path without 0A and FV-011 the Boric Acid Transfer Pumps available. Plausible candidate may not realize BA Transfer Pump impact on the flow path and confuse the flow rate requirements. LV-1 12D and LV-112E D. Correct. Opening LV-112D LV-1 12E would establish boration flow from RWST and flow requirements would be satisfied with 100 gpm to BIT. 100 gpm used to sound more like choice B to make question symetrical with choices and NOT be a NOT question. REFERENCES 19211-C, Nuclear Power Generation ATWT page 4 19211-C, 13009-1/2, CVCS Makeup Control System section 4.9 for Emergency Boration pages 38 through 41. LO-PP-09300-06-001, LO-PP-09300-06-001, 003, and 004 from Vogtle LO Active Exam Bank VEGP learning objectives: LO-PP-09300-06, LO-PP-09300-06, Describe all all emergency flow path
- a. borated water
- a. borated water source source and and discharge flow path discharge flow path b.
- b. minimum minimum flow flow requirements requirements Page: 220f3 Page: of 3 10/26/2009 10/26/2009
Page: 30f3 3 of 3 10/26/2009 10/26/2009
- 29. 032AK3.02
- 29. 032AK3 .02 OOlINEW/RO/C/A 00 1/NEW/RO!C/A 3.7/4.1/N/N/3/CVRIY 3.7/4. 1ININ/3/CVRJY Unit 11 is Unit is at at 100%,
100%, andand the following conditions the following conditions occurred: occurred:
** Intermediate Range Intermediate Range Channel Channel N-35N-35 lost compensating voltage.
lost compensating voltage.
** l&C is I&C called to is called investigate.
to investigate.
** Prior to Prior to any any action action by l&C, aa reactor by I&C, reactor trip trip occurs.
occurs. Which one of the following describes the Source Source Range Range NI NI detectors detectors response response after trip, and the the trip, the required actions lAW required actions ESP-0.1, Reactor lAW ESP-0.1, Reactor Trip Response? Response? Source Range Source Range Instruments Instruments will (1) ;; and they must be manually (2) A. (1) automatically energize prematurely
-te--prevent4amagc1o1he.
(2) de-energized until approximately 5 minutes post-trip -t-e-prevent-damage,-tQ...the,-_ detectors
~'aetector:s-B. (1) automatically energize prematurely (2) de-energized until approximately 15 minutes post-trip-te-\O>FeveFlt-G1ama~e-te-the-post-trip- prevent-damage-1o-t-he
-detectors---
--aeteeter C. (1) NOT automatically energize when required (2) energized approximately 5 minutes post-trip-topreventa post-trip-to-prevent alossof--reactorpower--
loss 0heactorpnwer-
---4ndication----
,-il'ldieatiefl~'
D~ D (1) NOT automatically energize when required (2) energized approximately 15 minutes post-trip ,te-pfevent'a-Iossofreactorpower'
-to--prevent a loss of reactor power
,"i-indic-ation--
nGI ic~tie Fl-~ Page: 75 of Page: 75 of 200 200 12/1412009 12/14/2009
A - Incorrect. Plausible, since examinee may believe loss of compensating voltage will make IR power read lower than actual and energize the Source range Nls NIs above power level that they are normally operated at. UOP-1.2 the power UOP-1 .2 (step 5.18) & UOP-1.3 UOP-1 .3 direct the Source Range Nls NIs deenergized above P-6, IR>1 lR>1OE-10 OE-1 0 amps, and applicant may believe there is similar guidance in ESP-0.1 for a premature energizing of the Source Range Nls. The second part is incorrect, since the decay into the Source range is at -1/3 1/3 dpm for about 6 decades, and thus takes about 15 minutes. Plausible, since confusion may exist between the decay into the source range from the power range and the limit on power ascension rate of 11 DPM from procedures to travel the required five decades. B - Incorrect. The first part is incorrect (see A). The second part is correct (see D). C - Incorrect. The first part is correct (see D). The second part is incorrect (see A). D - Correct. Loss of compensation is under compensated, which means that IR power will read higher than actual. SR automatic energization requires 2 of 2 IR detectors
<P-6.
< P-6. Power drops immediately after a trip approximately one decade from 100%
to approximately 7% (even though Nls NIs indicate 0%). Then, it decays at approximately 1/3 dpm for 5 decades to 10E-10 1OE-10 amps in the IR. Assuming -1/3 DPM SUR for about 5 decades, -15 15 minutes post trip is when the Source Range is required, and automatically energized if both IR NI detectors are working properly. Per ESP-0.1, Step 12; since P-6 interlock should have reinstated the SR NI high verify source range detectors energized" voltage power, "verify energized is directed. "Verify" Verify means take action to accomplish it if it didn'tdidnt already happen. At this time, the other Intermediate range that is reading correctly will indicate that the power level is below P-6, and the source range Nls NIs must be energized manually. ESP-O.1, Reactor Trip Response, Revision 29 12 Monitor nuclear instrumentation. [CA] WHEN intermediate range 12.1 [CAl 12.1 IF no source range detector indication less than 10-10 amps energized, THEN within one hour OR BYP & PERMISSIVE verify adequate shutdown margin P-6 light off, THEN verify source range using FNP-1-STP-29.1, detectors - ENERGIZED SHUTDOWN MARGIN CALCULATION (TAVG 547??F), or FNP-1 -STP-29.2, SHUTDOWN MARGIN CALCULATION FNP-1-STP-29.2, (TAVG <547F OR BEFORE THE INITIAL CRITICALITY FOLLOWING REFUELING). Page: 76 of 200 12/14/2009
Previous NRC exam history if any: 032AK3.02 032 Loss of Source Range Nuclear Instrumentation AK3. Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR 41.5,41.10 /I 45.6 1/ 45.13) instrumentation... 3.7* AK3.02 Guidance contained in EOP for loss of source-range nuclear instrumentation 3.7 4.1 Match justification: This has a IR channel malfunction that causes the SR instruments to be de-energized at a time they should be energized and the procedural guidance and time when the SR instruments will be energized by the operator. Objective:
- 6. EVALUATE plant conditions and DETERMINE DETER1VIINE if any system components need to be ESP-0. 1, Reactor Trip Response. (OPS-52531B06) operated while performing ESP-O.l, (OPS-5253 1 B06)
Page: 77 of 200 of200 12/14/2009
04/03109 04/03/09 13 13:42:27
- 42:27 FNP- 1 -UOP-2. 1 FNP-1-UOP-2.1 5.21 WHEN the Source Range Permissive P-6 light goes off (2/2 intermediate ranges 101010 less than 1I x 10- amps), THEN check the following:
5.21.1 Check SOURCE RANGE TRAIN A TRIP BLOCED/HIGH VOLT OFF status light is OFF. 5.2 1.1 5.21.1 Check SOURCE RANGE TRAIN B TRIP BLOCED/HIGH VOLT OFF status light is OFF. trf~
,(' 521 3 Check SR LOSS OF DETECTOR VOLAGE annunciator FA3 is clear.
L '~~ eY". rr l .... 5.21*~;;~source
~J automatically range channel hi h voltage is NOT automaticall energIzed ue to an under compensate r malfunctioning if lJI\1- JP It/:J:::-. ¥- r intermediate range channel, cnannei, THEN manually RESET the affected
~ V-,yLAl source range high voltage when the operable intermediate range
~~1' ~V~lllV'" channel channel indicates less than indicates less than 55 xx 10-11 10-11 amps.
amps. __ \1--'
~[
r: . # ( 5.2 1.4.1 5.21.4.1 IF required, THEN take SOURCE RANGE BLOCK-RESET A TRAIN to RESET. __ ~.t [~} 5.2 1.4.2 5.21.4.2 IF required, THEN take SOURCE RANGE W( BLOCK-RESET B TRAIN to RESET. lY- .22 WHEN the Source Range Nuclear Instruments are energized, THEN perform the following: 5.22.1 Verify Verifv the scaler timer aligned and operating properly. 5.22.2 Verify audio count rate amplifier aligned and operating properly. 5.23 Within one hour after P-6 is reached perform the following: 5.23.1 Perform SR 3.3. 3.3.1.1(channel
- 1. 1(channel check) for the Source Range Nuclear
_1- Instruments. 5.23.2 Document the channel check in FNP-1-STP-l.0, FNP-1-STP-1.0, OPERATIONS DAILY DAIL Y AND SHIFT SURVEILLANCE REQUIREMENTS. 5.23.3 IF the channel check cannot be performed, THEN verify adequate _1- shutdown margin using FNP-1-STP-29.1, SHUTDOWN MARGIN CALCULATION (TAVG 547°F), or FNP-1-STP-29.2, (TAVG < < 547°F OR BEFORE THE INITIAL CRITICALITY FOLLOWING REFUELING) Version 63.0
05112109 05/12/09 12:45:53 FNP- 1 -UOP- 1.2 FNP-I-UOP-1.2 NOTE: The rod position corresponding to ECC 0.5% SK/K AKJK is determined in FNP-1-STP-29.6, Appendix 1, CALCULATION OF ESTIMATED CRITICAL CONDITION. 5.16 IF criticality has NOT been achieved with the rods withdrawn to 0.5% ~KlK AK/K _1_/ (500 pcm) past the estimated critical position, THEN performJhe perform the following: 5.16.1 Insert all control banks to the bottom of the core.
/
5.16.2 Direct Chemistry to sample the RCS for boron concentration. 5.16.3 Re-calculate the ECC. 5.16.4 Determine and correct any discrepancy in the ECC. 5.16.5 IF no error can be found in the ECC or the RCS boron concentration, THEN contact the Reactor Engineering for assistance with the ECC. 5.16.6 IF required, THEN establish the correct critical boron concentration. 5.16.7 WHEN all errors have been corrected, THEN using the Inverse _1-/ Count Rate Ratio Plot procedure per Appendix 1, withdraw the control rods in MANUAL to establish reactor criticality. tJcff!J, J ~ f'-c~ 5.17 Establish a startup rate of approximately 3/4 decade per minute. - . . .Y ~ JtA~~
/ aC f'~ ~
1 f WHEN the Source Range Permissive P-6 light is on (112
-,-,-,,=c...:.the (1/2 intermediate ranges \ d~-ea fl
ç 10 10b0 greater than 10- amps), THEN perform the following:
/
5.18.1 Block the Source Range High Flux Reactor Trip. [] [] Source Range BLOCK-RESET A TRN taken to BLOCK [] Source Range BLOCK-RESET B TRN taken to BLOCK []
)
5.18.2 On the Bypass and Permissive Panel, verify that the following
/ windows are illuminated.
{] SOURCE RANGE TRAIN A BLOCKED HI VOLTS OFF [] [] SOURCE RANGE TRAIN B BLOCKED HI VOLTS OFF [] 5.18.3 Verify SR LOSS OF DET VOLTAGE annunciator F FA3 A3 is illuminated.
/
5.18.4 Verify the Source Range NI drawers indicate zero voltage. 5.18.5 18.5 Ensure the Scaler-Timer is shutdown per FNP-I-S0P-39.0, FNP-1-SOP-39.0, INSTRUMENTATION NUCLEAR INSTRUMENT A nON SYSTEM. Version 92.0
G/v:/ 2_ 1 REACTOR SHUTDOWN RESPONSE TO A AND RCS COOLDOWN REACTOR TRIP A normal plant shutdown and cooldown are The actions taken by reactor operators following performed periodically for refueling or a reactor trip are dictated by approved station maintenance. Power reduction is performed by procedures. These procedures ensure that the decreasing the external load on the turbine reactor is shut down, the turbine is tripped, generator in conjunction with a boration of the normal and/or emergency power sources are RCS. This maintains control rod position and available, and the plant response is as expected. satisfies axial flux difference requirements and If needed, compensatory actions are taken in rod insertion limits. As power is decreased accordance with the procedures. below 15%, the rods are put in manual control and the reactor operator manipulates the rods as Figure 8-24 shows the behavior of a reactor necessary to control RCS temperature. When power drop following a reactor trip. the turbine generator load has been decreased to approximately 50 MW the turbine is tripped and 100% A the control rods are positioned to maintain approximately 2% reactor power. After power has been stabilized at ffi approximately 2%, the reactor operator records ~ the information required for Reference ~ YI).. d fM Reactivity Data (RRD): power level, rod ~ B ,/ Y position, and actual boron concentration. This ~ B d.# I shutdown_______c:w=-__-I--~ data will be used for calculation of shutdownc;.-___ ~ margins and for subsequent ECPs. Once the.l3e~ the ,- C RRD data has been recorded, the reactor is .y>&"we-r ...v ____ . 0 shutdown by fully inserting all control banks. ~ 0% L...-_ _ _ _ _ _ _ _ _ _ _ gz_p, --.......;;._ _
..., TIME AFTER TRIP ~
Before starting the reactor cooldown, the RCS is borated to achieve the xenon-free shutdown Figure 8-24 Reactor Power Drop Following a margin required by technical specifications for r _ /) Reactor Trip RCS temperature below 200°F (typically -1,000 *1~2J.;~~ k Ak/k). Once this boration is completed and the i1k1k). Thfission rate decreases to below the power Th1lfission RCS boron concentration has been verified by range immediately upon insertion of the control chemical analysis, the cooldown is performed. and shutdown rods. This rapid reactivity When cold shutdown conditions are reached, the insertion is denoted by the neutron flux trace shutdown margin is re-verified. If adequate, the (power drop) from A to B in the figure. This is shutdown banks are fully inserted and the referred to as prompt drop following the reactor reactor trip breakers are opened. trip. During the period from B to C, the neutron population is dominated by the appearance of delayed neutrons from shorter- and intermediate-lived delayed neutron precursors. PWR / REACTOR THEORY / CHAPTER 88 45 of 70 450[70 © 2007 GENERAL PHYSICS CORPORATION I/ REACTOR OPERATIONAL PHYSICS REV 4 ur ~gpwunu w IUt::.lOUIll VY v '15J:'
.p vvui lu vv IUt::.lOUIll
,vllU iuc.Lviii
These precursors, which were formed when the reactor was at 100% power, decay within a few A reactor that has been operating at steady-minutes. Once the shorter-lived precursors have state 100% power trips, dropping rods effectively all decayed, neutron population is %~k/k (10,000 worth 1100 %Ak!k (l0,000 pcm) into the controlled by the appearance of delayed core. This causes an immediate prompt neutrons from the longest-lived precursors. drop in reactor power to approximately ___ %, followed by a slower decrease. From C to D, 0, power falls at a constant 80 -80 second period based on the mean life of the longest lived delayed neutron precursor, bromine-87 (half-life of about 56 seconds). The
-80 second period is eguivalent to about a 80
-1/3 decade per minute (OPM) startup rate (SUR). This continues until neutron popi:i1ai1Oil piT6E rate- I;? df?-M cl.ev~h 18 low ~ough for the effect of source neutrons ~ r--d eC Z5
-c.-- :::;>
tobe to be seen and ançcriticalilibri_ a subcritical eguilibrium is ;- reached. reached. Core thermal power remains high for several ~7~/Sr<'-1"'v.kP seconds after the trip (as shown by points B to C). There is a time lag of a few seconds for the heat generated in the fuel to be conducted into the coolant, and the decay heat immediately fbllowingtheproniFaropisappro1rnatT7 following the prompt drop is approxImately 7% Of of rated thermal power (R (RTP), assumin a trip TP), assuming 'trom from eqi ibriurn equilibrium full power operation. (This (Thlli """occurs occurs at about point C.)
- RCS temperature is reduced by the steam dump system and stabilizes at no-load Tave.
- Ten seconds after the trip, decay heat is still approximately 5% RTP, and it decreases to about 1% RTP in a little less than three hours (between points C to 0). D).
Example 8-25 PWR / REACTOR THEORY / CHAPTER 8 46 of 70 460[70 © 2007 GENERAL PHYSICS CORPORATION
/ REACTOR OPERATIONAL PHYSICS REV 4 .Ji luwluc.l..v1II vv vv vy .p vui IUVV iUc.4U111
05/12/09 12:45:53 05112/09 FNP-1 -UOP-1 .2 FNP-1-UOP-1.2 Figure 11 II io-3 10- 120% D P NN 0 T E 100% w---"- M M R R 10-4 E I d.e- c "-ce E E R t' lo\ ~ f-t:v.A ~ ~~ oD E I R-ctP R J-\t7P A 105 10- 50% A A L T N N E R G A 1066 10- E N N G ---- --- 0%_ 0% -l-t, ~ E iO-7 10-1 "2 sS 1066 - - . - - - 10 10-88 10-L'73 o0 109 10-1 i U U 10 io5 --1--- - -- R C C E 104 - - - 10 10b0 10- 10 1 R 10h1 10- 11 A A N N i03 - - - 10 G E 1022 - - - 10 1011 - - - 10 10° Page 11 of 11 Version 92.0
05/12/09 12:45:53 05112/09 FNP-1-UOP-1.2 FNP-I-UOP-l.2 5.11 WHEN the reactor is critical, THEN perform one of the following: 5.11.1 Verify the low low Tavg alarm reset (RX COOLANT LOOPS lA, 1A, IB lB
/
_1- 1C TAVG LO-LO annunciator HF4) AND all RCS Loop Tavg or lC greater than or equal to 547°F. OR 5.11.2 ioop Tavg greater than or equal to 541°F Verify each reactor coolant loop 54 1°F at _1-/ least once ever 30 minutes per FNP-I-STP-35.l, FNP-1-STP-35.l, UNIT STARTUP TECHNICAL SPECIFICATION VERIFICATION. (Technical Specification 3.4.2) CAUTIONS: ** During all rod withdrawals, monitor nuclear instrumentation. time the control rods are being Criticality shall be anticipated any time. withdrawn... Additionally, during any approach to criticality withdrawn monitor all pertinent instrumentation to allow errors in the ECC or problems with other instrumentation to be dete.cted detected early. Consider the use of audio count rate speakers as an aid to determine increasing flux rate. (SOER 88-02)
- Do not exceed a sustained startup rate of one decade per peiminute.
mmute. I 5.12 Using the Inverse Count Rate Ratio procedure per Appendix 11 withdraw the control rod banks in MANUAL to establish reactor criticality. {CMT-0008411} 5.13 Verify proper overlap ofl.R. of I.R. (See Fig 1) Note S.R. count when I.R. starts to
/
_1- come on scale. (CR 1-2000-148) ib 5.13.1 Record Source Range Counts. ____ (N31) 27 / _ _ _ (N32) 5.13.2 Record SR counts in Reactor Operators Log. 5.13.3 Record SR counts in Surveillance Test Data Book. Version 92.0
FNP Units 11 &
&2 REACTOR PROTECTION SYSTEM A-181007 A-18 1007 2.8.1 Reactor Permissives
- 1. Power Escalation The overpower protection provided by the excore nuclear instrumentation shall consist of three discrete, but overlapping levels. Continuation of startup operation of power increase shall require a permissive signal from the higher range instrumentation channels before the lower range level trips can be manually blocked by the operator.
A one out of two intermediate range permissive signal P-6 (set at lxi 0 10 amp) is required prior to source range level trip blocking and source range detector high voltage cutoff. Source range level trips are automatically reactivated and high voltage restored when both intermediate range aiekre beIe the permissive (P-6) leve Ihere shall be apual reset switchor administratively reactivating e source range Veripi aetector high vTf w en ween the er e P-6 and P- 10 (set at 10% rated thermal power) level if required. Source range level trip block and high voltage cutoff shall always be active when above the permissive P-1P-b 0 level. (t The intermediate range reactor trip and power-range (low setpoint) reactor trip shall only be blocked after satisfactory operation and permissive information are obtained from two of four power range channels which indicates P-1P-b. O. Individual blocking switches shall be provided so that the low setpoint power range trip and intermediate range trip can be independently blocked. Moreover P-i 0 allows the operator to manually block the intermediate range P-I0 C-i rod stop. These trips are aut9matically C-l automatically reactivated when any three of the four power range channels are below the permissive (P- 10) level, thus ensuring automatic activation to more restrictive (P-I0) trip protection. See Table T-3 for a comprehensive list of Reactor Protection System permissives. (References 6.4.007, 6.4.011, 6.7.012)
- 2. Blocks of Reactor Trips at Low Power Permissive P-7 shall prevent unnecessary at power reactor trips during low power by auto blocking the following reactor trips:
- Low reactor coolant flow in any two loops
- RCP breaker trip
- Undervoltage condition on RCP electrical buses
- Underfrequency condition on RCP electrical buses 2-32 Rev. 10 I
- 30. 035K3.02 005lNEW/ROICIA 005/NEW/RO/C/A 4.0/4.3/N/N/4/HOWARDSIVER 4.O/4.3/N/N/4/HOWARDS/VER 55 EDITORIAL Unit 11 has experienced a Loss of Offsite Power and a Tube Rupture on the 1A SG, and the following conditions exist:
- RCS cooldown at the maximum obtainable rate is in progress lAW EEP-3, Steam Generator Tube Rupture.
- INTEGRITY Critical Safety Function Status Tree has turned ORANGE due to the IA RCS LOOP cold leg temperature dropping rapidly.
1A Which one of the following describes the reason the 1 IA A RCS LOOP cold leg temperature has dropped rapidly? 1A IA RCS Loop flow has _ _ _ _ _ _ _ _ __ A. increased, moving the cold 1A IA SG U-tube water past the T COLD instruments. TCOLD B. restarted, causing a sudden rise then rapid drop in temperature as the stagnant water from the hot leg is flushed through the loop. C. reversed, causing the cold water from 1 lB and 1IC Band C loops to pass over the T TCQLD COLD instruments in 1A loop. D~ D stopped, allowing the cold Safety Injection water to pass over the T COLD TCOLD instruments. Page: 78 of 200 12/14/2009 12/1412009
A - Incorrect. During E-3 max rate cooldown under natural circ conditions, the ruptured SG loop flow will stagnate and may reverse, but NOT increase since the ruptured SG is not steamed, the differential temperature causing the Thermal Driving Head will be lost in that loop. Plausible: performing a cooldown increases the TDH for the intact loops and the cooldown for the intact loops would result in colder SG U-tube water to pass the T TCOLD COLD instruments. B - Incorrect. SEE A. Loop flow is expected to stall not restart. Plausible: Initiating the cooldown, would restart or improve the intact SG loop flows. Also, after the termination of the cooldown/depressurization there is expected to be a minor restoration of flow in A RCS loop during the recovery actions and subsequent stabilization procedures. C - Incorrect. 1A RCS LOOP flow will stop, however T
- cold of the active loops will not be Tcold sufficiently low to cause integrity to be challenged in the inactive loop, otherwise the Integrity status tree would be VALID and thier temperatures would ALSO result in an ORANGE INTEGRITY condition.
Plausible: a flow reversal is discussed in the occurrance of this condition and the 1 I B & 11 C loops temperatures are lower than the 1A RCS loop. oD - Correct.
- CAUTION-i warns the operator to not enter The basis for step 6.4 CAUTION-1 FRP-P.i if caused from the LOOP with the Ruptured SG. This is FRP-P.1 because SI flow reversal will likely occur in the ruptured Loop and "result result in the indicated cold leg temperature (due to the location of the cold leg RTD) to decrease. The flow stagnation in the 1A RCS loop, combined RTD)"
with the SI Flow into the loop, and a leak in the 1 1A A SG would cause an accumulation of the Cold SI (RWST) water to accumulate between the SI thermal sleeve and the SG, causing a >100°F/hr cooldown (which is in all loops due to operator action) AND <250°F (285°F unit 2)-- an ORANGE path condition on FRP-P. EEB-3.0, EEB-3.O, ver 1, pg 31: Basis for step 6.4 CAUTION-1 CAUTION-i If the RCS is being cooled down on natural circulation during a steam generator tube rupture event, "If reverse flow through the ruptured loop during the cooldown or when the pressurizer PORV is opened to depressurize the ReS RCS is possible and could cause the SI flow path in the ruptured loop to change. This change in the SI flow path could result in an indicated cold leg temperature (due to the location fo the cold leg RTD) that decreases to the point that the symptoms for FR-P.1 would occur. This false indication would only be seen in the ruptured loop since it is essentially stagnant while th either loops are circulated by natural circulation. When the PORV is closed, the flow paths are expected to change and the indicated cold leg temperature should increase resulting in the symptoms disappearing. When SI is terminated, the indicated cold leg temperature would increase if it did not do so FR-P.i no longer being present. This is an expected condition and earlier resulting in the symptoms for FR-P.1 purposes. the operator should only monitor the F-O.4, Integrity Status Tree for information purposes." Page: 79 of 200 12/14/2009 12/1412009
Previous NRC Previous NRC examexam history history ifif any: any: NoneNone 035K3.02 035K3.02 Steam Generator 035 Steam 035 Generator System System K3 Knowledge K3 Knowledge of the effect of the effect that that aa loss or malfunction loss or mahunction ofthe of the S/GS S/GS will viI1 have have on on the the following: following: (CFR: 41.7 (CFR: 41.7 145.6)
/ 45.6)
K3.02 ECCS K3.02 ECCS .......................................................... 4.04.3 4.0 4.3 Match justification: the Match the effect effect of of ECCS ECCS flow into the flow into the loop ioop due due toto implementing implementing E-3 E-3 is is aa stagnation/flow reversal stagnation/flow reversal in loop, which in A loop, which is indicated by is indicated by aa rapid rapid drop drop in in indicated indicated loop cold leg temperature. This rapid drop in temperature results in FRP-P.1 being ORANGE, and and understanding the mitigation strategy IMPACT on ECCS EGGS flowpath ensures that the cooldown cooldown would not not bebe erroneously terminating. Terminating the cooldown due to the indications indications presented here, here, would complicate stabilization of the plant. Objective: OPS-52530D03; State and Explain the basis for all Cautions, Notes and Actions associated with EEP-3 ((...]. ... ]. Page: 80 Page: 80 of of 200 200 12/14/2009 12/14/2009
( ) j'\j .\i!! ~. FNP-1-EEP-3 STEAM GENERATOR TUBE RUPTURE Revision 24 Step Action/Expected Response Response NOT Obtained n I I CAUTION: With all RCPs secured RCS cooldown may cause a false FNP-1-CSF-0.4 FNP-1-CSF-O.4 Integrity Status Tree indication for the ruptured loop. Disregard ruptured loop cold leg temperature until completion of step 30. NOTE:
- The steam dumps will be interlocked closed when RCS TAVG reaches P-12 (543°F). This interlock may be bypassed for A A and E E steam dumps with the STM DUMP INTERLOCK switches switches..
- Excessive opening of steam dumps can cause a high steam flow LO-LO TAVG main steam line isolation signal.
6.4 IF condenser available. available, 6.4 Dump steam to atmosphere. THEN dump steam to condenser from intact SGs at maximum 6.4.1 Direct counting room to attainable rate. perform FNP-0-CCP-645. FNP-O-CCP-645, MAIN STEAM ABNORMAL BYP && PERMISSIVE ENVIRONMENTAL RELEASE. COND AVAIL 6.4.2 IF normal air available. available, [] El C-9 light lit THEN control atmospheric relief valves to dump steam STM DUMP from intact SGs at maximum [] [1 MODE SEL A-B TRN in STM PRESS attainable rate, rate. NOT. dump steam using IF NOT, STM DUMP FNP-1-S0P-62.0. FNP-1-SOP-62.O, EMERGENCY INTERLOCK AIR SYSTEM. [] [I A A TRN in ON [] [I B B TRN in ON 1A(1B.1C) 1A(1B,1C) MS ATMOS REL VLV STM HDR [] [I PC 3371A adjusted PRESS [] El PC 3371B adjusted [] [I PK 464 adjusted [] [I PC 3371C adjusted Step 66 continued on next page. Page Completed Page 16 of 57
06/27/07 16:10:46 06127/0716:10:46 FNP-0-EEB-3.0 STEAM GENERATOR TUBE RUPTURE Plant Specific Background Information Section: Procedure Unit 11 ERP Step: 6.4 CAUTION-1 CAUTION-i Unit 2 ERP Step: 6.4 CAUTION-1 CAUTION-i ERG Step No: 6 CAUTION-I CAUTION-] ERP StepText: With all RCPs secured RCS cooldown may cause a false FNP-2-CSF-0.4 Integrity Status Tree indication for the ruptured loop. Disregard ruptured loop cold leg temperature until completion of step 30. ERG Step Text: If RCPs are not running, the following steps may cause a false F-O.4, Integrity Status Tree afalse indication for the ruptured loop. Disregard the ruptured loop T-cold indication until after performing Step 29.
Purpose:
To alert the operator that during a natural circulation cooldown cooidown a false symptom of a red or orange path condition in F-O.4, F-0.4, Integrity Status Tree is possible due to the redirection of SI flow in the ruptured loop. Basis: If the RCS is being cooled down on natural circulation during a steam generator tube rupture event, reverse flow through the ruptured loop during the cooldown or when the pressurizer PORV is opened to depressurize the RCS is possible and could cause the SI flow path in the ruptured loop ioop to change. This change in the SI flow path could result in an indicated cold leg temperature (due to the location of the cold leg RTD) that decreases to the point that the symptoms for FR-P.1 FR-P. 1 would occur. This false indication would only be seen in the ruptured loop ioop since it is essentially stagnant while the other loops are circulating by natural circulation. When the PORV is closed, the flow paths are expected to change and the indicated cold leg temperature should increase resulting in the symptoms disappearing. When SI is terminated, the indicated cold leg temperature would increase if it did not do so earlier resulting in the symptoms for FR-P.1 FR-P. 1 no longer being present. This is an expected condition F-0.4, Integrity Status Tree for information and the operator should only monitor the F-O.4, purposes. After the cooldown and depressurization is completed and SI is terminated, the operator should monitor the F-O.4, F-0.4, Integrity Status Tree to determine if a red or orange path still exists and FR-P.1 should be implemented. His decision should be based on the FR-P. 1 symptoms existing after SI is terminated. If a multiple or subsequent accident occurs, the operator could transfer out of E-3 prior to terminating SI. For that case he should monitor the F-O.4, Integrity Status Tree when he makes the transition out of E-3 to determine if at that F-0.4, time a red or orange path exists and FR-P.1 FR-P. 1 should be implemented. STEP DESCRIPTION TABLE FOR E-3Step 66-- CAUTION Knowledge: If a multiple or subsequent accident occurs, the operator could transfer out of E-3 prior to F-0.4, Integrity Status Tree when he terminating SI. For that case he should monitor the F-O.4, makes the transition out of E-3 to determine if at that time a red or orange path exists and FR-P. 1 should be implemented. FR-P.1
References:
DW 028 DW-96-028 31 of 119 Version: 1.0
- 31. 037AA2.0S 037AA2.08 OOlINEW/RO/C/A OO1/NEW/RO/C/A 2.S/3.3/N/N/2/CVRlY 2.8/3.3/N/N/2/CVR/Y Unit 1 1 is at 12% power, and the following conditions exist:
- R-15A, SJAE EXH, has failed.
- IA 1A SG has developed a 10 gpm tube leak.
- IA SG safeties is leaking by.
One of the 1A Which one of the following radiation monitors will provide the EARLIEST indication of the 1A IA SG Tube leak? A'I R-19, SGBD SAMPLE, alarm. A B. R-23B, SGBD TO DILUTION, alarm. C. R-70A, 1A SG TUBE LEAK DET, alarm. D. R-60A, 1A IA STEAM GENERATOR, alarm. A - Correct. R-19 is continuously monitoring the SGBD system sample stream and will be the first indication of an alarm considering only the 4 choices given. BB - Incorrect. R-23B will only alarm after the SGBD surge tank starts filling with the contaminated SGBD water. The tank is maintained 50% full, and there will be a diluting effect at first. Downstream of this tank is R-23B in a flow stream going to the environment. R-19 samples undiluted SGBD water continuously, and would alarm sooner than R-23B. Plausible, since R-23A samples blowdown water at the inlet of the Surge Tank and is undiluted SGBD water. Due to the higher flowrate of SGBD (about 130 gpm) than the sample stream, it alarms sooner than the R-19 R-1 9 alarm for a particular SGTL event. Confusion may exist as to the difference between the choice for R-23B and R-23A which would alarm prior to R-19 R-1 9 (as seen on the simulator during SGTL events). C - Incorrect. R-70A alarm setpoints are not valid at this power level. The R-70s shift automatically from the gpd Mode to the ME mode below 20% power, and the alarm functions are set for gpd. Plausible, since it is on the Steam line at the outlet of the SG, and alarms first before any other Radiation monitor in the event of a SGTL above 20% reactor power. D - Incorrect. R-60 is a high range monitor that does not upscale in the event of SGTL D - with no fuel failure, even with an open safety or SG SC Atmospheric relief. Plausible, R-1 9 since it is monitoring the since if it was a lower scale, it would alarm prior to R-19 steam coming from the 1A IA SG SC with the tube leak. Also, it would alarm if the dose from the SG was high enough (such as due to a SGTR and fuel element failure). The SGTL combined with the safety leakby allows it to monitor the actual1Aactual 1A SG contaminated steam as it escapes through the safety. FNP-1-AOP-2.O, Steam Generator Tube Leakage, Version 33.0 FNP-I-AOP-2.0, B. Symptoms or Entry Conditions I. Enter this procedure when RCS tube leakage is indicated by high secondary activity on any of of 200 Page: 81 of200 12/14/2009 12/1412009
the following radiation monitors or by sample results.
- a. R-15 SJAE EXH [listed here to show the nomenclature in the procedure]
- b. R-15B or R-15C TURB BLDG VNTL
- c. R-19 SGBD SAMPLE [listed here to show the nomenclature in the procedure]d.
procedurejd. R-23A SGBD HX HX OUTLET
- e. R-23B SGBD TO DILUTION [listed here to show the nomenclature in the procedure]
1A(1B,IC) SG
- f. R-70A, R-70B or R-70C IA(IB,IC) SC TUBE LEAK DET
- g. SG sample results indicate primary to secondary leakage for any SG SC greater than or equal to the FG 1, SG normal alarm setpoint for annunciator FGI, SC TUBE LEAK ABOVE SETPT.
FNP-1-ARP-1.6, FHl FNP-l-ARP-l.6, FHI AUTOMATIC ACTIONS (cont)
- 2. ARDA will automatically start for the following conditions:
2.1 ARDA will automatically start when any of the following monitors go into alarm for two polis one minute apart on either unit and use the latest 15 minute average consecutive system polls monitor value to perform the calculations: Plant Vent Stack Monitors R29 (SPING) Noble Gas 4.44e-4 clml dm1 I .20e-6 clml Iodine 1.20e-6 c/mi dm1 4.OOe-5 clml Particulate 4.00e-5 Ri 5C 27 mrlhr Steam Jet air Ejector R15C mr/hr TDAFW Exhaust R60D R6OD 38 mrlhr mr/hr Steam Generator A R60A R6OA 38 mrlhr mr/hr [listed here to show the nomenclature in the procedure] Steam Generator B R60B R6OB 38 mrlhr mr/hr Steam Generator C R60C R6OC 38 mrlhr mr/hr Page: 82 of 200 12/14/2009 12/1412009
Previous NRC exam history if any: Previous 037AA2.08 037 Steam Generator Tube Leak AA2. Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak:
/45.13)
(CFR: 43.5 145.13) AA2.08 Failure of Condensate air ejector exhaust monitor ..................... 2.83.3 8 3.3 Match justification: A scenario is given with a SGTL and a failed Condensate air ejector exhaust monitor (R-15A), which is normally the first indication of a SGTL. To answer this question correctly, determining how the failed R-15 R-1 5 applies to the SGTL is required. I. E., since it is no longer the first indicaiton of a SGTL, which is the first indication? Objective:
- 2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Radiation Monitoring System to include those items in Table 4- Remote and Local Indications and Controls (OPS-40305A02).
- 5. DEFINE AND EVALUATE the operational implications of normal/abnormal normal / abnormal plant or equipment conditions associated with the safe operation of the Radiation Monitoring System components and equipment, to include the following (OPS-40305A07):
- Normal control methods
- Abnormal and Emergency Control Methods
- Automatic actuation
- Protective isolations
- Protective interlocks
- Actions needed to mitigate the consequence of the abnormality Page: 83 of 200 12/1412009 12/14/2009
03112/03 11:24:25 03/12/03 11 :24:25 FNP-1-SOP-69.0 FNP-I-SOP-69.0 FARLEY NUCLEAR FARLEY NUCLEAR PLANT PLANT UNIT UNIT 11 SYSTEM OPERATING PROCEDURE SYSTEM PROCEDURE SOP-69.0 SOP-69.0 N-I6 PRIMARY TO SECONDARY N-16 SECONDARY LEAK DETECTION SYSTEM 1.0 1.0 Purpose Secondary Leak Detection To provide guidance for operation of the Primary to Secondary Detection System. 2.0 Initial Conditions 2.l 2.1 120V Regulated Instrumentation Panel PaneliBlB is energized per FNP-1-SOP-36.4, FNP-I-SOP-36.4, 120V A.C. DISTRIBUTION SYSTEMS. 3.0 Precautions and Limitations 3.1 The system receives a reactor power input from power range channel N-43. IF
~ . (~-43 T-43 fails OR is in Test OR is less than 20% power, THEN the system cannot C 1 \\&7 f A V mode, and the indicators will display accurately estimate a leak rate in the AV oJ/7 <20%. If desired, the Counting Room can configure the N-PN <20%".
"PN 16 system in the N-16 L ')0/ !<~ ME counts per second (CIS) FNP-0-CCP-3 1, LEAK RATE (C/S) mode using FNP-0-CCP-31, Q ~ DETERMINATION. While not able to provide a leak rate determination, this
/\Jt' mode can be used to indicate if leakage is increasing based on the indication trending up. The A AVV mode is the preferred mode of operation above 20% reactor power. The ME mode should only be utilized below 20% reactor power. 3.2 The N 16 Leak Detection System cannot determine the location of a leak within a N--16 specific Steam Generator. The system can however provide a more accurate leak rate determination if the location of the leak is known to be in one of the following locations: Cold Leg - CB, Hot Leg - HB or U-Bend region - BE WHEN a leak location is selected (CB, HB or BE), THEN the processor displays a leak rate that assumes the leak is at the location you have selected. The AAVV mode is essentially the average of the three leak rates at the specific locations. 3.3 The N-16 system is limited to an upward range of 1,000 gallons per day. Version 5.0
FNP Units 11 & &2 RADIATION RADIA TION MONITORING SYSTEM A-18 1015 A-181015 increasing radiation to initiate the RMS High Radiation annunciator on the main control board. In addition, each function shall actuate an indicating light on the ratemeter front panel defining the alarm condition (References 6.4.034, 6.4.2 14, and 6.4.249). The ratemeter operation 6.4.034,6.4.214, selector switch shall actuate the RMS CH Test annunciator on the main control board when placed in any position other than OPERATE (References 6.4.034 and 6.4.249). 3.2.5.3.4 Normally, no contaminated leakage is expected into the Service Water system. Accordingly, the monitor setpoint should be set approximately one half decade above the detectors normal response (Reference 6.7.062 and detector's 6.7.080). 3.2.5.4 Interface Requirements The instrument power supply for the RMS system panel NIHIINGRM N1H11NGRM 2502A, B, and C is 120 VAC distribution panel IB, 1B, breaker number 2 (Reference 6.4.219). The control power supply for the RMS system N1H1 INGRM 2502A, B, and Cis panel NIHIINGRM C is 2081120 208/120 VAC control power iN, breaker number 6 (Reference 6.4.107). The instrument power panel IN, supply for the RMS system panel N2H1 1NGRM 2502A, B, and Cis N2HIINGRM C is 120 VAC distribution panel 2B, breaker number 2 (Reference 6.4.345). panel2B, The control power supply for the RMS system panel N2HI 1NGRM N2HIINGRM 208/120 VAC control power panel 2N, breaker 2502A, B, and C is 2081120 number 6 (Reference 6.4.106). 3.2.5.5 Shielding Design In addition to the shielding provided by the monitor housing, additional shielding was added surrounding the high voltage electronics housing above the detector. The purpose of the added shielding is to reduce the influence of background radiation, that caused spiking of the monitors and isolation of the process system and to improve monitor sensitivity (References 6.7.033, 6.7.035, and 6.7.080). 3.2.6 Steam Generator Blowclown Blowdown Service TPNS Nos. SG Blowdown to Processing System ND1 NDllREiRE 0023A SG Blowdown Discharge NDI NDllREiRE 0023B 23162.3431A-181015.RM 334 3-34 Rev. 00
FNP Units 11 & &2 RADIATION MONITORING SYSTEM A-181015 3.2.6.1 Basic Function 3.2.6.1.1 Radiation detector RE 0023A, located in the steam generator blowdown discharge line upstream of the steam generator blowdown surge tank, monitors for an increase in radioactivity in the secondary system. To minimize contamination of the processing system and potential inadvertent release of radioactive gases through the surge tank vent, steam generator blowdown is automatically terminated by closing valve NB2 NB211 FCV 1152 when the activity exceeds the setpoint. An increase in radioactivity in the secondary system would be indicative of a steam generator tube rupture accident (References 6.7.084 and 6.4.366). 3.2.6.1.2 Radiation monitor RE 0023B is located downstream of the steam generator blowdown discharge pumps prior to discharging to the dilution discharge on the service water system. This detector monitors the discharge stream to comply with GDCs 60 and 64. On an increase in radioactivity, the discharge is isolated by automatically closing NB2 1 RCVO23B (References 6.7.084 and 6.4.366). NB21RCV023B 3.2.6.2 Functional Requirements 3.2.6.2.1 An in-line liquid monitor shall be provided to directly monitor the process medium. The use of this type of monitor provides the fastest response time and easiest decontamination (References 6.4.366 and 6.7.080). 3.2.6.2.2 RE 0023B shall alarm and isolate the effluent discharge often prior to exceeding the limits of ten times the concentrations stated in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the II.A design objectives of Appendix I, 10 CFR 50, Section ILA for a member of the public, and (2) the limits of 10 CFR 20.130 1 for the population. The limiting concentration of 20.1301 1 04 !-LCi/ml dissolved and entrained noble gases is 11 x 10- .tCi/mI 6.7.08 1). (Reference 6.7.078, 6.7.081). Setpoints are based on ensuring the discharge limits presented in Section 2.1.2 of the ODCM are not exceeded. 23162343\A-181015.RM 23162.3431A-181015.RM 335 3-35 Rev. 0
FNP Units 11 & 2 RADIA nON MONITORING SYSTEM RADIATION A-18 1015 A-181015 3.3.16.5 Interface Requirements 3.3.16.5.1 The 120 VAC power supply for RMS panel QSD1 QSD11REiRE 2081120 VAC control power panel 1R, 0035A is 208/120 lR, breaker number 15 (Reference 6.4.108). VAC pump power 6.4.1 08). The 208 VAC supply for RMS panel QSD1QSD11REiRE 0035A is 208/120 2081120 VAC control power panel 1R, 1 R, breaker number 12 (Reference 6.4.108). The 120 VAC power supply for RMS panel QSD11RE QSD1 2081120 VAC control power panel 1iS, iRE 0035B is 208/120 S, breaker number 15 (Reference 6.4.108). The 208 VAC VAC pump power supply for RMS panel QSD1 iRE 0035B is QSD lIRE 2081120 VAC control power panel iS, 208/120 IS, breaker number 12 (Reference 6.4.108). 3.3.16.5.2 The instrument air system shall provide a dry, filtered air source for monitor purge air (Reference 6.7.080). 3.3.17 Main Steam Safety and Atmospheric Relief and TDAFW Pump Exhaust Noble Gas Monitors Service TPNS Nos. Safety/Atmospheric Steam Generator A Safety!Atmospheric ND liRE 0060A NDllRE Relief Valve Exhaust Safety/Atmospheric Steam Generator B Safety! Atmospheric ND1 ND11REiRE 0060B Relief Valve Exhaust Safety/Atmospheric Steam Generator C Safety! Atmospheric liRE ND 11 RE 0060C Relief Valve Exhaust Turbine Driven AFW Pump Exhaust ND111iRE ND RE 0060D 3.3.17.1 Basic Function 3.3.17.1.1 The monitors provide post-accident effluent monitoring for the Steam Generator Safety Valve and Turbine Driven Auxiliary Feedwater Pump turbine exhaust points in compliance with RG 1.97 (References 6.4.051 and 6.4.350). 3.3.17.1.2 These monitors may be used for RG 1.21 effluent activity tracking (References 6.7.078 and 6.7.060). 3.3.17.2 Functional Requirements 3.3.17.2.1 The monitors shall provide continuous indication over a lOl1 to 10 range of 10- 103 microcuries per cubic centimeter 23162343\A-181015RM 23162.3431A-18101S.RM 393 3-93 Rev. 0
FNP Units I1&& 22 FNP Units RADIATION MONITORIN RADIATION MONITORING G SYSTEM SYSTEM A-181015 A-181015 (llCilcc). The (p.Cilcc). monitors shall The monitors provide an shall provide an approximatel approximately y linear response for linear response gamma energies for gamma energies between between 0.5 and 33 MeV 0.5 and MeV (References 6.4.051 and (References 6.4.051 6.4.350, and and 6.4.350, 6.7.003). NUREG and 6.7.003). NUREG Clarification Item 0737, Clarification 0737, Il.F.1, Attachment Item II.F.l, requires that Attachment 1,1, requires that gas radiation noble gas noble monitors be provided for radiation monitors be provided for effluent points effluent points monitor from which monitor which from normal operating levels normal operating levels toto aa maximum maximum of 105 p.Ci/cc of calibration) for (Xenon-133 calibration) llCi/cc (Xenon-133 for undiluted undiluted effluents and containment effluents containment and 101 lO-1 to 103 llCi/cc to iü llCi/cc forfor buildings buildings with systems with containing primary systems containing primary coolant such as the coolant such as the auxiliary building. The auxiliary building. The present plant configuration present plant configuration monitors for area monitors provides area provides for these parameters used these parameters used in in aa process monitoring process monitoring application. application. The The meters meters for for these these read in monitors read monitors R/hr and in RJhr conversion charts and conversion have been charts have been provided to provided convert the to convert meter readings to the meter to jiCi/cc. llCi/cc. The The indicate aa range monitors indicate present monitors present range ofof I10- to 106 022 to 106 mR/hr, mR/hr, corresponds to 110-which corresponds 05 to 1.4. 04 p.Ci/cc 1.4. xx 110 llCi/cc (References (References 6.3.00 and 6.7.005). 6.3.0011 and 6.7.005). 3.3.17.2.2 Each radiation monitor is located to view its respective steam generator safety valve plumes and atmospheric relief valve plume. Each monitor is located on the auxiliary building roof and oriented to minimize the effects of radiation shine from the containment following a design basis LOCA. The monitors are located as far from the safety valve and main steam atmospheric vent valve vent stacks as permitted by the containment wall while keeping all monitors the same distance from the centerline of their respective main steam atmospheric vent valve. The monitor for the "B" B steam generator limits this distance to 17.17 feet. Therefore, the monitors are located on a 17.17-foot circle around the main steam atmospheric vent valve stack and in a configuration not facing the containment, but steam generator safety valve have the steam valve plumes and the main steam atmospheric vent valve plume for that particular steam generator in steam generator full view. in full The The viewing angle in viewing angle the horizontal plane in the plane was was determined based based onon the location of the location of the monitors and the monitors and was determined was determined to to be be 55°. 55°. A viewing angle A viewing angle of monitoring of allows monitoring 55° allows of 55° all of all plumes of a plumes of a particular particular steam steam generator generator while while excluding excluding the the field field of of view most of from most view from other steam the other ofthe generator steam generator plumes, plumes, andand not containment. facing containment. not facing For For the the vertical plane, the vertical plane, the monitor angle must viewing angle monitor viewing must be be small small enough enough to prevent the to prevent monitor from the monitor from being influenced being influenced from radiation shine from radiation from the shine from containment and the containment and large large 3-94 3-94 Rev. 1I Rev.
- 32. 038EA2.07 OOllNEW/ROICIA 00 1/NEW/RO/C/A 4.4/4.8ININ/3/CVRlY 4.414.81N/N/3/CVRIY Unit 1 1 has manually Tripped and Safety Injected from 14% power lAW AOP-2, Steam Generator Tube Leakage. The following conditions exist:
- DA-07, 1A IA 4160V 4I60V BUS SUPP FROM 1A IA S/U XFMR, breaker tripped open.
- AFW Flows were maintained matched to all 3 SGs until securing AFW Flow.
- AFW Flow has been secured to all SGs.
- SG Pressures:
- SG NR levels:
- IA 980 psig and stable
- 1A 61%
- 1A 61
- % and stable
-1
- IBB 980 psig and stable lB 61%
- 1B 61 % and rising
-1
- ICC 980 psig and stable -1
- 1C C 50% and lowering Which one of the following correctly describes the event in progress based on the MCB indications?
A'I SGTR on 1lBB SG ONLY. A B. SGTR on 1A AND 1I B B SGs ONLY. C. SGTR on 1 lB B SG AND a Steam Leak on 1 1C C SG ONLY. D. SGTR on 1A AND 1 lBB SG AND a Steam Leak on 1C SG. 8G. A - Correct. The level rising in 1
- 1 B SG with no AFW flow, ,=,rE?_s;"s;!!fl~~~JJ~y~~I".QtQEELlJg Pressurizer level dropping with max chg and no letdown and all SG pressures stable indicates this answer to be correct. C SG Ivl lvi is decreasing due to being the only SG steaming with no AFW flow. This is normal indication for this condition, and will require AFW flow to maintain C SG Ivl. IvI.
B - Incorrect. The SGTR on 1
- lB B SG is correct, but the SGTR on 1 1A A is incorrect.
Plausible, since with the AFW flows matched to all SGs, and 1A IA SG level higher than the other intact SG by 10% NR Lvi, Lvl, a SG tube leak would be indicated if not for the tripped RCP in that loop, and the Ivl IvI being stable instead of rising with no AFW flow AND no steaming. If the RCP was not tripped, this would be a correct answer, since with no AFW flow and decay heat removal level staying constant would indicate a SGTR. C - Incorrect. The SGTR on 1
- 1B B SG is correct, but the Steam Leak on 11 C SG is incorrect due to the pressure being stable at 980 psig on all three SGs. Plausible, since 1 ICC Level is dropping with the 1 IA A SG level stable and the 1 IC C SG dropping and no AFW flow to any of the SGs.
D D - Incorrect. SGTR on 1A
- IA & 1lBB incorrect (see B). Steam leak on 1 1C C incorrect (see C).
Page: 84 of 200 12/14/2009 12114/2009
Previous NRC Previous NRC exam exam history history ifif any: any: 038 EA2 .07 038EA2.07 038 Steam 038 Steam Generator Generator Tube Tube Rupture Rupture EA2 Ability EA2 to determine Ability to determine or or interpret interpret the the following following as as they apply to they apply to aa SGTR: SGTR: (CFR (CFR 43.5/45.13) 43.5 / 45.13) EA2.07 Plant EA2.07 Plant conditions, conditions, from from survey survey of control room of control room indications indications .............. 4.4 4.4 4.8 4.8 Match justification: Control Match Control Room Room indications indications are are given given which which could could indicate indicate aa SGTR SGTR in in two SGs two SGs and and aa Steam Steam leakleak in in one one SG under slightly SG under slightly different different conditions conditions thanthan given. With given. With one one tripped tripped RCP RCP one one SG level is SG level is high high due not steaming due to not steaming and and not not due to a SGTR. One SG SG is is high due to a SGTR. SGTR. One SG level is low and dropping due to being the only intact SG producing steam, steam, even though the other intact SG Ivl intact IvI is is stable (due (due to the tripped RCP). RCP). The applicant must must correctly correctly indications and diagnose evaluate all these indications diagnose the event to be be a SGTR in in one SGSG only. Objective:
- 3. LIST AND DESCRIBE the sequence of major actions, when and how continuous actions will be implemented, associated with (1) EEP-O, EEP-0, Reactor Trip or Safety Injection ESP-0.0, Rediagnosis. (OPS-52530A04) and (2) ESP-O.O, Page: 85 of Page: 85 200 of200 12/14/2009 12/1412009
- 33. 039K4.05 OOl/FNP 00 1/FNP BANK/RO/MEM 3.7/3.7/N/N/2/CVRIY 3.7/3 .7/N/N/2/CVR!Y Which one of the following adequately describes the setpoint of the steam line flow for the High Main Steam Line Flow with Low-Low T avg MSIV isolation?
Tavg A. Increases linearly from 40% to 110% steam flow as power increases from 0% to 100%. B. Increases linearly from 20% to 110% steam flow as power increases from 0% to 100%. C. Constant 20% steam flow up to 10% power; then increases linearly to 110% flow as power increases from 10% to 100%. D~ Constant 40% steam flow up to 20% power; then increases linearly to 110% flow as D power increases from 20% to 100%. A - Incorrect. Plausible, since the numbers are the same as the values for the correct setpoint, but the Constant value of 40% steam flow limit from 1 1%-
%- 20% is left out of this choice.
B - Incorrect. Plausible, since the numbers are the same as the values for the correct setpoint, but the Constant value of 40% steam flow limit from 1 1%-
%- 20% is left out of this choice.
C - Incorrect. Plausible, since this choice correctly states that there is a constant value of Steam Flow setpoint up to a certain power, but the constant value and associated power level are incorrect. The setpoint at 100% power is correct. 0-D Correct. OPS-52201 K OPS-52201K A higher than expected steam flow from the steam generators, along with a decreasing T avg, is another Tavg indication of a steam break that will shut the MSIVs. The high steam flow set point is varied with turbine power by Pimp. imp The set point is 40 percent steam flow from 0 percent to 20 percent turbine power. It then increases linearly from 20 percent turbine power to 100 percent turbine power where the set point is 110 percent steam flow. 2/3 steam lines reaching the set point and T avg below the P-12 set point will shut the MSIVs. It Tavg requires only one of the two, density-compensated steam flow detectors per steam line to reach the set point to actuate the MSIV closure with a one second time delay. This main steam line isolation is not able to be blocked or bypassed. Page: 86 of 200 12/14/2009 12/1412009
Previous NRC Previous NRC exam exam history history ifif any: any: 039 K4 .05 039K4.05 039 Main and Reheat Steam 039 Steam System System K4 Knowledge K4 Knowledge ofMRSS of MRSS design design feature(s) feature(s) andlor and/or interlock(s) interlock(s) which which provide provide for for the the following: following: (CFR: 41.7) K4.05 Automatic isolation of steam line ................................... 3.7 3.7 Match justification: Other MSIAS are on exam so this type of question was selected to avoid double jeopardy. Objective:
- 6. DEFINE AND EVALUATE the operational implications of normal/abnormal normal / abnormal plant or equipment conditions associated with the safe operation of the Main and Reheat Steam System components and equipment, to include the following (OPS-40201A07):
** Normal control methods
** Abnormal and Emergency Control Methods
** Automatic actuation including setpoint (example SI, Phase A, Phase B, MSLIAS, LOSP, SO SG level)
** Protective isolations such as high flow, low pressure, low level including setpoint
** Protective interlocks
** Actions needed to mitigate the consequence of the abnormality Page:
Page: 87 87 of of 200 200 12/14/2009 12/14/2009
5.
- 5. High-i containment High-1 containment pressure pressure (PB-951B, PB-952B, (pB-95IB, PB-952B, PB-953B)
PB-953B) 4.0 4.0 psig psig 6.
- 6. Time delay Time delay on on SI SI manual manual reset reset 11 minute minute B.
B. Steam line Steam line isolation isolation la. 1a. High steam High line flow steam line flow in in coincidence coincidence with with b-b T..-a 10-10 T dip d/p corresponding corresponding to to (FB-474A, FB-475A, FB-484A, FB-485A, 110% offull 110% of full steam steam flow flow FB-494A, FB-495A) FB-494A, FB-495A) at full load at full load 40% of full steam flow between 0% and 20% load dip d/p setpoint linear with turbine first stage pressure between 200~ 20% load and full load lb. Lo-boT.P-12 1..0-10 Tava (P-12))
çrB-412E, TB-422E, TB-432B)
(TB-412E, TB-432E) 543 0 F 543°F ic. Ie. Filter Lag Time Constant (FY-474B, FY-475B, FY-484B, 0oSeconds Seconds* FY-485B, FY-485B, FY-494B, FY-494B, FY-495B) FY-495B)
*ysetupto
*May be set up to 1.5 1.5 seconds
- 2. Low steam line pressure (PB-474A, (pB-474A, PB-485A, PB-496A) psis SIS psig 585 Lead Lead timetime constant constant (PY-474B, (pY-474B, PY-485B, PY-485B. PY-496B)
PY-496B) seoonds SO seconds 50 Lag Lag me tUne constant constant (PY-474B, (pY-474B. PY-485B, PY-485B, PY-496B) PY-496B) seccmds 5S seconds 3.
- 3. Hlgh-2 High-2 containment containment pressure pressure (PB-951B, (pB-95IB, PB-952B, PB-952B, PB-953B)
PB-953B) 1.6.2 psig 1.6.2 psig Rev.A7 Rev. A7 11 11
- 34. 041K4.14 04 1K4. 14 00l/NEW/ROIC/A 00 1!NEW/RO/C/A 2.S/2.8/N/N/4/CVRN 2.5/2.8/N/NI4ICVR/Y Unit 11 was at 26% power and 180 MWe, and the following conditions occurred:
- The reactor tripped.
- The "A"A Reactor Trip Breaker failed to open.
Which one of the following correctly states the arming signal for the Steam Dumps, and the RCS temperature maintained by the Steam Dumps? The Steam Dumps are armed due to the (1) and RCS temperature will be controlled at (2) (1 (1)) (2) A. P-4 signal 547° F 54rF B. P-4 signal 551°F CY' C Loss of Load signal 547°F 54rF D. Loss of Load signal 551°F A - Incorrect. The first part is incorrect but plausible, since the reactor did trip and if A RT bkr would have opened this would be correct. Most functions of the P-4 Permissive come from both trains, but this function comes only from A train. The second part is correct, and is the result of the B B train P-4 signal which is present as normal with the B train RT bkr open. B - Incorrect. The first and second part are incorrect but plausible, since this choice would be correct for a B train RT bkr failing to open. A train P-4 would arm the Steam Dumps and the LOL controller would stay in the circuit (since the B train P-4 did not shift controllers to the Plant Trip mode) to control Tavg 4°F higher than no load Tavg (547+4=551). C - Correct. The A train P-4 did not arm the steam dumps, and the Loss of Load did (the loss of load was 20% instantaneously, and thus greater than the LOL arming setpoint of 15% with a 120 secono jjc.. time constant). The B train P-4 shifted the 9 sec controllers from the LOL to the(PIa-nrfrfp"'c~~t~OII~~ 1.9 the Plant Trip controller which maintains a constant no load Tavg of load Tavg of 54rF. 547°F. "--~-.--.-----------"--- D - Incorrect. The first part is correct (see C). The second part is incorrect, but plausible, since it would be correct for a loss of load or reactor trip with the B B train RT bkr open instead of the A. Page: 88 of 200 12/14/2009 12/1412009
Previous NRC exam history if any: None 041 K4.14 041 Steam Dump System and Turbine Bypass Control K4 Knowledge of SDS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) 2.5** 2.8 K4.14 Operation of loss-of-load bistable taps upon turbine load loss ........... 2.5 Match justification: The loss-of-bistable arms the steam dump when the loss of load magnitude is greater than the variable setpoint in a given time. The Reactor trip overrides the arming of the loss of load due the A RT bkr P-4 signal for a normal reactor trip. This question requires knowledge the loss of load setpoint and the times that it does and does not arm the Steam dumps. It also requires knowledge of what controller (Plant Trip or Loss of Load) is in the circuit under different conditions than normal. Several design features and interlocks relating to and affecting the Loss of Load interlock must be understood to correctly answer this question. Objective:
- 5. DEFINE AND EVALUATE the operational implications of abnormal plant or equipment conditions associated with the operation of the Steam Dump System components and equipment to include the following (OPS-5220 (OPS-52201G07):
1G07):
- Normal Control Methods (Steam dump valves)
- Abnormal and Emergency Control Methods (Steam dump valves, Steam dump system solenoid-operated three-way valves)
- Automatic actuation including setpoint (High-1 (High-i and High-2 trip bistables)
Protective isolations (Plant trip controller, Loss of load controller, C-7)
- Protective Interlocks (Condenser available, C-9, Low-Low TAVG TA VG signal, P-I 2)
P-12) Actions needed to mitigate the consequence of the abnormality Page: 89 of 200 12/14/2009 12/1412009
Date: 1OI Date: 10r 2009 '0009
,- Time Time: "\44:26 PM 44:26 1 PM
~/
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'-------,\: TUR\INE PRESSURE STHERS REDUNDANT r-REDUNDANT 41
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MEDIAN MEDIAN TT RVS AVG
- -- - -----L - - - - I I CONDENSER AVAILA%LE CONDENSER AVAILA1LE SIGNALS SIGNALS <SHEET MI (SHEET 9)
HEADER HEADER STEAM DUMP IL 1/ A , I REF ERENCE REFERENCE TT AVG AVG PRESSURE PRESSURE STEAM STEAM LINE LINE PRESSURE PRESSURE I INTERLOCK SELECTOR INTERLOCK SELECTOR II CIRCULATION WATER CIRCULATION \lATER INTERNAL INTERNAL S'w'lTCH (NOTE (NOTE 313) CONDENSER PUMP REACTOR TRIP I I iI LOOP2 LOOP3 SWITCH I I CONDENSER PUMP SETPDINT SETPOINT STEAM STEAM DUMP CONTROL MODE DUMP CONTROL SELECTOR SWITCH MODE SELECTOR S,,",JTCH PRESSURE SAl PRESSURE S'w'ITCH CA BREAKERS BREA CLOSED ERS CLOSED (SHEET PlO TRAIN A RAIN I STEAM HERDER J STEAM GENERRTO STEAM SENCRATO STEAM GENERATOR STEAM PRESS. 2E) [EPTRR3 I r
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IOES CONTROLLER CSNTRQLLER
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,,r ATMOSPHERIC ATMOSPHERIC ATMOSPHERIC ATMOSPHERIC ATMOSPHERIC ATMOSPHERIC S= H f NOTES: 1. STEAM STEAM DUMP RELIEF _V~LVE RE~~F~~E_ RELIEF _VALVE
[FVALVERELIEFVALVERELIEFVBLVE DUMP IS BLOCKED BY IS BLOCKED BY BLOCKING BLOCKING AIR AIR TO TO THETHE
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BY .OTHERS OTHERS DUMP DUMP VALVESVALVES AND AND VENTINO VENTING THE DIAPHRAGM. THE THE DIAPHRADPT. THE REDUNDANT [DOTE REDUNDANT LOGIC OUTPUT OPERATES 22 SOLENOID OUTPUT OPERATES SOLENOID VENT VENT VALVES VALVES IN TO REDUNDANTLYREDUNDANTLY INTERLEIEK INTERLOCK THE THE AIRAIR LINE LINE BETWEEN
,r IN SERIES EACH SERIES TO EACH VALVE VALVE DIAPHRAGM DIAPHRAGM AIlS AND ITS ASSOCIATED PUSTTIONER, ITS ASSEETATED POSITIONER.
BET\.IEEN
,, THE SOlENDlD VALVES THE SOLENOID VALVES ARE ARE SEENEROIZED DE -ENERGIZED TO TO VENT VENT ENJOINS CAUSING lL~P~
IN FIVE FIVE SECONDS.
,r THE MAIN THE MAIN DUMP DUMP VALVE VALVE CLOSE IN TO CLSSE TO SECONDS, REDUNDANT EXCEPT
,, 2. CIRCUITRY UN
- 2. CIRCUITAT
\.IHERE INDICATED WHERE ON THIS INDICATED REDUNDANY.
THIS SHEET SHEET IS REDUNDANT. IS NDTNOT REDLNDANT EXCEPT
, ) ,r SELECTOR SWITCH
- 3. SELECTOR
- 3. S\.IITCH WITH \.lITH THE FDLLO\.lING 33 POSITIONS THE FOLLOWING POSITIONS, 1\;\1~7J ON- STEAM ON STEAM DUMP DUMP IS IS PERPITTED PERMITTED tJ ~I[LI~I~
V BYPASS- TT AVG BTPASS AVG. INTER,DDK INTERLOCK IS IS BYPASSED BYPASSED 5R FOR LOW LO\.l REDUNDANT T.AVS, T.AVG. SPRING SPRING RETURNRETURN TO ON POSITIEN. TO 0.5 POSITION. orr - STEAM OFF STEAM DUMP DUMP IS IS NOT PERMITTED AND NOT PERM:TYED AND RESETRESET
---~
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SAyS, T.AVG, BYPASS.
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THE REDUNDANT INTERLCC.K THE REJJNDANT INTERLOCK SELECTORSELECTOR SWITCH S\.IITCH EONSISTS CONSISTS If)-' BY I (l) BY OF OF TWO CONTROLS ON T\.IO CONTROLS ON THE CONTROL BOARD, THE CONTROL BOARD, ONE ONE FOR FOR
-JACH TRAIN, 2°
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MODULAT~H~~Ot~7r'w'I~~L¥[5uE~CCgRDING TO S. TO
- 5. THE TWO ANALOG SISNAL PRESSURE MUST PRESSURE MEET THE TO MEET THE CONDENSER MUST CORE THE SINGLE CONDENSER AVAILABLE COME FROM SINGLE FAILURE INPUTS COMING SIGNAL INPUTS FROM DIFFERENT FAILURE CRITERION.
AVAILABLE SIGNAL COMING FROM DIFFERENT PRESSURE CRITERION. SIGNAL LOGIC LOGIC IS FROM TURBINE PRESSURE TAPS TURBINE IS TYPICAL TYPICAL, TAPS o
",,0 6° IBY IOTHERs ",,2 VALVES MODULATED OPEN DR ACTUAL ALL TEMPERATURE 6, ALL U.
IMPLEMENTATION MAY ACTUAL IMPLEMENTATION TEMPERATURE BISTABLES MAY BE BIST ABLES SN BE DIFFERENT. DIFFERENT. ON THIS SHEET AND THIS SHEET AND TURBINE
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BLOCK NTERM OURP TO BLOCK STERM DUMP TO TRIP OPE 1FPE( I/P CLOSED <ZERO TO FULL OPEN) IMPULSE CHAMBER IMPULSE CHAMBER PRESSURE PRESSURE BOOTABLE BISTABLE RPB447A TURBINE OD N1N36V501 N1N36V501E tlPB-447A 00..1 VALVENEMCFCR TAMP VALVES NINX6V5IIA,ECC DUMP VRLVES NINX6VSIIB,FR,H NIN36V501~: NIN36V501G NIN36V501B, NIN36V501F 7. TNERGIZE TO ARE ENERGIZE ARE LIGHTS SHOULD
- 7. LIGHTS SHOULD BE TO ACTUATE.
ACTUATE.6 BE PROVIDED PROVIDED IN IN THE CONTROL ROOM THE CONTROL ROOM FOR FOR Zc..o G~ zw
~tO :::>f-(NOTE II NIN36V50lD, NIN36V501H EACH DUMP EACH DUMP VALVE VALVE TO INDICATE WHEN TO INDICATE 'WHEN THE THE VALVE VALVE IS IS
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Title:
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FNP Units 11 & 2 REACTOR PROTECTION SYSTEM A-18 1007 A-181007 range channels exceeds a current equivalent to 20 percent reactor power. The rod stop may be manually blocked when above the P-l P-100 setpoint, but is automatically reinstated below P-I0 P-10 (3/4). This can be manually bypassed atatNIS NIS racks. (References 6.4.007, 6.4.011, 6.4.016) C-2 Interlock. The C-2 overpower rod stop blocks automatic and manual control rod withdrawal. The block action occurs when 114 1/4 power range channels exceeds 103 percent reactor power. Each power range channel may be manually bypassed at the NIS racks. All power range channels cannot be bypassed at the same time. Only two power range channels may be blocked at one time using two switches located on the NIS Miscellaneous Control and Indication Drawer. (References 6.4.007, 6.4.011, 6.4.016) 6.4.011,6.4.016) C-3 Interlock. The C-3 control interlock is generated by the OTDT circuitry. The setpoint is 33 percent below the variable OTDT reactor trip setpoint. C-3 generates a block of automatic and manual rod withdrawal, when 2/3 loop delta Ts exceed their setpoint. The function of the rod block is to eliminate the cause of the impending trip, thereby preventing it. Since relatively slow transients are typical of those requiring OTOT OTDT protection, there is sufficient time for a load reduction to correct the situation. (References 6.2.003, 6.4.007, 6.4.012, 6.4.016) C-4 Interlock. The C-4 control interlock is generated by the OPDT circuitry. The setpoint is 33 percent below the variable OPOT OPDT reactor trip setpoint. C-4 generates a block of automatic and manual rod withdrawal, when 2/3 loop delta Ts exceed their setpoint. The function of the rod block is to eliminate the cause of the impending trip, thereby preventing it. (References 6.2.003, 6.4.007, 6.4.012, 6.4.016) C-5 Interlock. The C-5 interlock ensures that automatic rod withdrawal system is prevented when less than 15 percent power. It also prevents automatic rod withdrawal when power falls below 15 percent. The setpoint is 15 percent power as detected by turbine first stage impulse 6.4.007, 6.4.016, 6.4.022) pressure. (References 6.4.007,6.4.016,6.4.022) C-7 Interlock. This control interlock arms the steam dumps upon a load rejection (when in coincidence with C-9). The steam dump demand interlock (C-7) is actuated when turbine load is reduced by greater than 15 percent with a 120 second time constant. Rate differentiation of the first stage turbine impulse chamber pressure signal provides the equivalent turbine load signal. It must be manually reset. (Only PT-447 provides 6.4.0 17) input to C-7.) (References 6.4.007, 6.4.017) C-9 Interlock. C-9 is the condenser-available interlock. This interlock allows the steam dump valves to be armed if the condenser is available. It also prevents an overpressure condition which could damage the 2-36 Rev. 10 II
FNP Units 11 & 2 REACTOR PROTECTION SYSTEM A- 181007 A-I8I007
- Pressurizer low pressure
- Pressurizer high level Permissive P-7 shall block the above listed reactor trips below 10 percent of full power. The low power signal is derived from the power range neutron flux channels (P-I (P- 10)
- 0) and the turbine impulse chamber pressure channels (P-13). The blocking feature occurs during the absence of P-7, meaning that P-7 is not active. This occurs when 3/4 power range neutron flux channels are below setpoint and 2/2 turbine impulse pressure channels are below setpoint. The P-7 Permissive is active when 2/4 power range neutron flux channels or 1/2 turbine impulse pressure channels are above setpoint. When P-7 is active all previous blocked reactor trips are reinstated.
The P-8 Permissive blocks a reactor trip when the plant is below 30 percent of full power on a low reactor coolant flow or a RCP any one loop (113 breaker open signal in anyone (1/3 coincidence). The block action (absence of the P-8 Permissive signal) occurs when three out of four neutron flux power range signals are below the 30% setpoint. Thus, below the P-8 setpoint, the reactor will be allowed to operate with one inactive reactor coolant loop and trip will not occur until two loops are indicating low flow. Permissive P-9 shall block a reactor trip following a turbine trip below 35 percent power which is based on the ability of the rod control system and steam dump system to adequately control Tavg on a 50% load rejection (FNP Tech Spec Section B2-8). If2/4 If 2/4 power range nuclear instruments are above 35 percent power, a turbine trip will cause a reactor trip. If 3/4 power range nuclear instruments are below the P-9 setpoint, a turbine trip will not cause a reactor trip. See Table T-3 for a comprehensive list of the protection system blocks. (References 6.4.007, 6.4.011, 6.4.022, 6.7.012) 2.8.2 ESF Permissives
- 1. P-4 Permissive The P-4 permissive is generated when both the reactor trip breaker and the bypass breaker, which physically bypasses it, are open.
Train A of the reactor protection system uses RTA and BYB, and train Buses B uses RTB and BYA. The following are functions ofP-4: of P-4:
- a. Causes a turbine trip 2-33 Rev. 10 I
FNP Units 11 & 2 REACTOR PROTECTION SYSTEM A-181007
- b. Main Feedwater Isolation - Closes main feedwater regulating valves and feedwater bypass valves if low Tavg Tavg (554 OF) is also present and requires a manual reset.
(554°F)
- c. Seals in feedwater isolation signal from safety injection or steam generator high-high water level
- d. Resets high steam flow setpoint to 40 percent
- e. Allows operator reset of the safety injection signal after a 60 second time delay This feature ensures that the reactor is tripped and that all emergency core coolant system (ECCS) loads are started before the operator overrides what could be a spurious actuation signal. The block does not prevent the operator from reinitiating safety injection through use of either manual safety injection actuation switch
- f. Arms steam dumps on a plant trip, defeats the output of the load rejection controller, and places the plant trip controller into control.
6.7.0 12 6.7.059, 6.7.060) (References 6.4.007, 6.4.009, 6.7.012
- 2. P-i 1 Permissive P-11 P-il allows manual block of low pressurizer The permissive P-11 pressure safety injection actuation. This permits normal plant cooldown and depressurization.
When 2/3 pressurizer pressure bistables sense less than 2000 psig, the low pressurizer pressure safety injection signal may be manually blocked. Placing the train A and train B pressurizer safety injection block switches to BLOCK will now prevent low pressurizer pressure (1850 psig) from initiating safety injection. Each switch will initiate the block function in its respective P-il setpoint on 2/3 protective train. If pressure rises above the P-11 channels, the block automatically clears. If 2/3 channels exceed P-i 1 setpoint, and power is available, any shut ECCS the P-11 accumulator isolation valves will automatically open. In addition, P-i 1 setpoint blocks the 2/3 pressurizer bistables below the P-ll automatic opening of the pressurizer power operated relief valves 6.7.0 12, 6.4.017) (PORVs). (References 6.4.007, 6.4.13, 6.7.012, 6.4.0 17) 2-34 Rev. 10 I
6.6. speedgain speed gain 32 steps/minute/°F 7.7. manual rod manual rodspeed speed control rods control rods 48 stepslminute(l} 48 steps/minute shutdown rods shutdown rods 62 stepslminute(l) 62 steps/rn nutem 2.
- 2. Steam Dump Steam Dump Control Control A.
A. Impulse unit Impulse unit time time constant constant ofloss of loss ofload of load interlock channel interlock channel (PY-447C) (pY-447C) 120 sec. 120 sec. B. B. Sudden load Sudden load lou loss setpoint (C-i) setpoiut (C-7) (PB-447A) (PB-447A) 15% of full load. 15% load C. Proportional gain in percent oftota! of total dump capacity per of OF
** Setpoint for full load Im = 577
= .ZeF 577.2°F Ja = = 567.2°F 1
of load controller (TY Loss ofload -408]) (TY-408J) 9.0o/JOF 9.00/d0F ) 16. 3cy'JoF1) 16.3°/d°F 1 Plant trip controller (TY (TY-408L)
-408L) 3.3o/JoF 3.30/J0F) S.O%,oF1)
** For other full load T T&VI between 567.2°F and 577.2tF, 577.2eF, the proportional gain setpoints should be calculated as follows.
(a) Loss of load controller (TY-408J), cy'JGF (TY-40SJ). %/°F 100 100
- [(Full
[(Full Load Load T - T _Ioed) T &VI - T 0 1 12] - Deadband
) /21)- Deadband (TY- (TY-408 408 J)J)
(b) (b) Plant Plant trip controller (TY-408L), trip controller cy'cJoF (TY-40SL), %l°F 100 100
- (Full Load (Full Load T T&VI -T,..i-Deadba
- T_...) - Deadband (TY- 408L) nd (TY-408L)
There There are are 88 condenser condenser dumpdump valves. valves. The should be controlJen should The controllers such that adjusted. such be ad4usted that the the dump dump capacity linear with capacity isis linear the output with the ofTY-408J, output of TY-408J, TY4OL, TY-40aL, or PC-464 (below). QI' PC-464 (bcIGw).
- That That is, is, the the second second bank bank does does not 29 29 not begin begin to to modulate modulate open the fir
\IDtil tha opeD until first bank bank has Rev.A7 Rev. A7 has
receivedaa signal received signalto modulatefull to modulate full open; etc. The open; etc. sequence for The sequence for modulating modulatingthe the valves valves closedisis the closed the reverse reverse ofofthethe opening opening sequence; sequence; i.e., the fourth i.e., the fourth bank bank to the first qen isis the to opeD. first bank to bank close, and to close, and the the third third bank bank starts starts to close after to close aIer the bank bas fourth bank the fourth has received received aa signal to signal close; etc. to close; etc. TheThe first first four four valves valves to modulate open to modulate open areare also also the the first first four four valves to valves to be be tripped tripped open. open. The The last four valves last four valves to modulate open to modulate open areare the the last last four four valves to valves to trip trip open. open. TheThe two two valves valves inin the bank are first bank the first designated as are designated as the the cooldown cooldown dwnp valves. The dump valves. The input input ranges ranges for for modulation modulation (full closed to (full closed to full open) and full open) and the the order order for modulating the for modulating the dump valves Ql2m dump valves Q are: are: BA14K iN / OUT IN/OUT IN I OUT IN/OUT vALVES VALVES (Fully closed (Fully closed to to fully fully open) open) First bank First 4.0 - 8.0
- 8.0 rna ma => . 0.0 - 2.5 v=>4
- v 4 --20 20 ma NIN36VSOIA(1)
NiN36V50iA> (Cooldown (N1C24TY-4 (NIC241Y-408S) 08S) (Ni C24TY408N) (NIC241Y-4QS.N) :NIN36VSOU:(1) N1N36V501E valves) 1 Second bank - 12.0 ma => 2.S 8.0 -12.0 . 2.5 - 5.0 v=>4
- v -lDma 4-20 ma N;lN36VSOIC N1N36V501C) 1 (N1C24TY-408I)
(NIC241Y-408'I) (N1C24TY-4tP) (NIC241Y-4.0SP) N1N36VSOla NIN36V5O1(t) Third bank 12.0 - 16.0 rna 12.0* ma => S.O - 7.5 v=>4 5.0 - 7.S v 4 --20 ma 20 rna NIN36V501B(1) NiN36V50iB NIN36VSOIF 1 (Ni C24TY-408U (Nl C241Y -408U)) (N1C24TY-4 (NIC241Y -408Q) 08Q) N1N36V5O iF°) Fourth bank Fourthbank 16.0 20.0 ma . 7.5-10.0v=> 16.0-20.0rna=>
- 7.5 10.0 v
- 44-20rna
-20 ma NIN36V501D(1)
NiN36V50iD l (Ni C24TY-408V) (NIC241Y-408V) (N1C24TY-408R) (NIC241Y-408R) NIN36V50nf N1N36V5O1 H°>) D. D. Lead time constant (TY-408D) (1Y-408D) 55 seconds seconds(l) E. E. Lag time constant (rY-408D) (1Y-408D) 55 seconds seconds(l) 0 F. F. Deadband Deadband steam steam dump dump controller controller for loss loss of load ofload (TY-408J) (1Y-408J) 4e
)
1 F 4o 1) 0. G. Deadband Deadband steam steam dumpdump controller controller forfor plant plant trip trip (TY-408L) (1Y-408L) OOF 0°F
- 30 30 Rev.A7 Rev. A7
H. H. High (fravg-Tr) wa - T.ae) *Setpomt
*Spoifor forfull fullload load
çrB-4osF)
(TB-408F) 1== 577 T T!YI 577.2°F
.2°F T!!,I = 567.2°F I=567.2°F First and First and second secondbankbanktriptrip open open (High (High 1)1) 9.5°F\)
95°F 7.00pO) 7.0°F° Third and Third and fourth fourthbank banktriptrip open open (High (High2)2) 15.l oFt ) 15.1°F 10.1 0 1"1)
- 10. PF° I. High(r.-T)
High (f&VI - TDD-Iood) *Spojnt for
*Setpomt for full full load load (TB-4083)
(TB-408J) 15 = 577.2°F
= 577.2°F T!Ya = 567.2 567.2°F G F
First and First and second bank trip second bank trip open (High 1) open (High 1) IS.l o 15.1°FF) l IO.lGF 10.1°F) 1 oP<l) 1 Third and Third and fourth bank trip fourth bank trip open open (High (High 2)2) 30.2 30.2°F 20.:zoF 20.2°F)
- For other full load 567.20F,
- For other full load Twa between 577.20F T. between 577.2°F and and 567.2 0 the setpoi.ut.s F, the setpoints should should be be calculated as calculated as follows:
follows: HighT High CI5 - T!!f) Setpoints (fB-408Fl (TB-4OF) (a) çr. Tnoe) High 11 (f&VI- - T.c) valve trip open (TB-408F), of OF (First and second bank trip open)
= [b below - deadband (1Y
= - -408J)]12 +
(TY-408J)]/2 + deadband (IT -408J) (TY-4083) (b) High 2 (f (T &VI - T
- Tc) (TB-408F), OF noe) valve trip open (TB-40SF), °F (Third and fourth bank trip open)
= (Full Load T
== T~)12 T &VI - Ti..i)/2 High CI5 Hiah Tj - T1IO=!ood) Setpoints Semoints CTB-408fl CIB-40SD (a) High 11 (T, (f.... - T.)
- T..-.) valve trip open (TB-408J),
(fB-40SJ), °F of (First and (First and second second bankbank trip trip open) open)
= [b
= [b below below - deadband
- (TY-408L)]J2 deadband (TY-408L)]/ 2++ deadband (TY-408L)
(IT-4OSL) (b) (b) High High 22 (T.(T.... - T 1 valve trip
.i) valve TDO-Iood.) open (TB-4083),
trip open (fB-408l). °F of (Third and (Third and fourth fourth bankbank trip trip open) open)
= (Full
= (Full Load Load T T~)
Twa - Tii-ioi)
- 31 31 Rev.A7 Rev. A7
y7 LOADReJECTION LOAD CONTROLLER(LOW REJECTIONCONTROLLER (LOWTAVG) TAVG) 50 4S 40 3$ c-
! J lao
'Ii t25 II
/
~ 20 J JI 15
/
V 10
/
o I o I8 ~10 n12 <<14 " " 20 T.n. En<'MIl* T_1EmIr (Ii,g -Ti() begFF TI9I) o.g fOF oL~p£ F 2 ~ je
.s Y? Y? Tt y~.:::: 5::::> 1& F (7
PLANT TRIP PLAHT TRW CONTROLLER (LOWTAYG) (LOW TAVO) . I V
/
I 10
/
./
/
/
o /
~ " 20 25 ao ti. Enor(Thvg -Teo4eed)
T_IEmlr<,ng* T....-)beg o.gFF
\J 1 242 -vI
- Steam SteamDump Dump Control Control System 32 32 System (Low (LowT,.) T*..J Rev.A7 Rev. A7
LOAD REJCTON CONTROLLER LOAD REJECTION cONTROLLER (HIGH *GH TAVG)TAVG) W 50
-45 40---
40
- 15 V
/.
35 I 130 15 lo25 :zzzz
/
lao / J 15 / 10 ___/_____._
/
/v V
o0 2 .-4* *0 T_ a1 T. EmIr w 10 En. (Tavg u12 TNt) Dog (Tavg ~.INI) Dog fF U 14 R 10 R 10 ao 40 y~F yop J&£fJ~ y7_tP rY~ ::::::?~t7,r-S:-Y7 PLANT TRIP PUNT ThW CONTROlLER CONTBOUIII (HIGH (t*OH TAVG)
- zz 40
- Ez /'
/' '"
i:EEE i/' V zz 40-------_ 15 ./' l0_7Z_ to
/'"
,/
5 7 V 0 z_ 5$ - w 10 u ao 25 30 25 T_:;_(Tavg.T~DogF To.Enit(T.og .Tr4.ed) Dig F
\
1 d 2
,J c? ed cl~
Jd ji ~o-ItC7 S - '17°F Steam Dump Control System (High T, Tavel 1 ) Rev.A7 Rev. A7 33
J.3. Headerpressure Header pressurecontroller controller (PC-464) (pC-464) setpressure set pressure 1005 psig(l) 1005 pjgO) proportional proportionalbandband(based (basedon total ontotal condenserdump condenser capacity) dumpcapacity) 200 200p$i(l) 0 psi resettime reset timeconstant constant 300 300sec.(l) sec. K. K. Steamgenerator Steam generatorrelief reliefvalve valve controllers controllers proportionalband proportional band (valve full stroke) (valve full stroke) 250 psi(l) 250 psi reset resettime time constant constant 8.5 8.5 sec.(l) 1 sec. set pressure set pressure 1035 psig 1035 psig 3.
- 3. Pressurizer Pressure Pressurizer Control (Refer Pressure Control (Refer to following figure) to followina figure)
A. A. Pressurizer pressure Pressurizer pressure controller controller (PC-444A) (PC444A) proportional proportional gain gain 0.5% controller 0.5% output/psi reset time constant 600 600 sec.(l) constant rate time canstant rate o0 sec51) pressure setpoint, Pref 2235 psia psig (42.S% (42.5% controller outputt~) 4 t) B. B. Spray valve valve controllers (PC-444C, (pC-444C, PC-444D) proportional proportional gain gain in percent spray valve valve 4o/JOie controller 4°%/% lift per psi liftperpsi OUtput(l) 0 output setpoint setpoint where spray is where spray is initiated initiated on on competed compensated pressure pressure signal signal from from PC-444A PC-444A controller output 55% controller 55% OUtput(l) C. C. Variable Variable heat heat controller controller proportional proportional gaingain in inpercent percent heating heating power powerper perpsi psi -6.67%f% controller output setpoint setpoint where where proportional proportional heatingheating isis full full on, on, on onsignal from PC-444A signal from PC-444A OU1p1K(l) CQIltrollerontput° 3S%controller 35%
- 34 34 Rev.A7 Rev. A7
------------------ ---- ------.------------~.
FNP Units 11 & & 22 REACTOR PROTECTION SYSTEM SYSTEM A-18 1007 A-181007 TABLE T REACTOR PERMISSIVES AND ESF PERMISSIVES TABLE - MEASURED COINCIDENCE & MODES OF FSD PERMISSIVE PARAMETER(S) PARAMETERfID SETPOINT *LIGHT
*LIOHT STATUS FUNCTION OPERATION SECTION P-4 Reactor Reactor trip and bypass Breakers Open - Either reactor trip breaker
- Prevents a rapid cooldown of the primary 1,2,3 and 4 l,2,3and4 2.8.2 trip interlock breakers RTA and BYA and its associated bypass system after a reactor trip Fig. 2,
@RTB and 9RTB BYB at;dIWB breaker are open breaker open - Trips turbine
- Sht. 8,
- No control board
- - T avg' initiates
- In coincidence with low Tavg, 10, 13, 10,13, indication indication other than closure of main feedwater reg valves and 15 reactor trip and bypass bypass valves breaker position indication - Prevents opening of main
- main feedwater on the reactor control control isolation valves if closed on safety panel panel SO high-high water level injection or SG signal Prevents reactuation of automatic safety
- Prevents injection after safety injection manual reset
- Resets high steam flow setpoint to 40%
- Allows reset of safety injection signal after a time delay
- Defeats the output of
- the Load Rejection ofthe Controller condenser steam dump valves
- Arms condenser P-6 Intermediate Intermediate NIS intermediate range 10b0 10- 10 amps - Allows
- Allows manual manual block ofof Allows power escalation into the 2belowthe 2 below the 2.8.1 range neutron neutron neutron flux channels source range reactor trip Intermediate Power Range by turning P-6 setpoint 2.6.1 flux power NC35D and NC36D on 1/2 intermediate range both MCB TramTrain A and B source range Fig. 22 escalation neutron flux channels BLOCK above setpoint block switches to BLOCK setpoint Sht. 4 permissive > setpoint
> - Allows manual blocking of source range
- Auto reinstates source
- high neutron flux reactor trip range reactor trip on 212 on 2/2 - Actuating manual block handswitches intermediate range setpoint deenergizes source range instruments instruments neutron flux below below - 2/2
- intermediate range channels below 212 intermediate below setpoint setpoint setpoint voltage to setpoint auto reinstates high voltage
- Permissive status light is
- source range detectors and reinstates lit when 112 1/2 channels source range high neutron flux reactor trip
> setpoint
- Designates Bypass Bypass and Permissive Light Box Status Status B-11 T3- Rev. 55
- 35. 045Al.05 045A1 .05 00lINEWIROICIA 3.8/4. 1/N/N/2/CVR/Y OO1/NEW/RO/C/A 3.8/4.1ININI2/CVRIY Unit 11 was at 30%, and the following conditions occurred:
- Control Rods are in Manual.
- Main Turbine was manually tripped.
The Main Which one of the following is the initial response of RCS Tavg and RCS Pressure, with no operator actions? Tavg (1) and RCS Pressure (2) (1)) (1 (2) A increases A'! increases B. increases decreases C. decreases increases D. decreases decreases Page: 90 of 200 12/1412009 12/14/2009
A - Correct. Steam pressure goes up, causing Tcold to go up. This causes Tavg to go przr insurge which compresses the steam space and RCS up. This causes a przr pressure goes up. B - Incorrect. The first part is correct (see A). The second part is incorrect (see A). Plausible, since the insurge is subcooled water from the RCS, but the steam space is compressed which causes pressure to go up. Once the Steam Dumps and/or SG Atmospherics open to reduce SG pressure, the RCS Tavg goes down and the outsurge causes pressure to go down to less than it was initially due to the subcooling of the przr liquid and steam space. C - Incorrect. The first part is incorrect (see A). Plausible, since the Steam Dumps and/or SG Atmospheric relief valves will open to reduce SG pressure and Tavg, but prior to the Steam Dumps and/or SG Atmospherics opening, Tavg will go up. The second part is correct (see A). D - Incorrect. The first part is incorrect (see C). The second part is incorrect (see B). This choice would be the correct response after the initial response. Ran on simulator Laptop (Ie (IC 47) Initial values: Rods in Manual 29% Reactor power 2243 psig 554°F Tripped turbine: Initially: temp, press, went up, power went down. Values 2 minutes later: 16.5% Reactor power stable NI (18% Delta T) 2209 psig 562.5° F (lower than peak of approx. 569°F) 562.5°F 569° F) Page: 91 91 of 200 12/14/2009 12/1412009
Previous NRC Previous NRC exam exam history history ifif any: any: 045A1 .05 045A1.05 045 Main 045 Turbine Generator Main Turbine Generator System System Al Ability Al to predict Ability to predict and/or and/or monitor changes in monitor changes in parameters parameters (to (to prevent prevent exceeding exceeding design design limits) limits) associated associated with operating with operating the the MT/G MT/G system controls including: system controls including: (CFR: 41.5 145.5) (CFR: 41.5 / 45.5) Al .05 Expected AI.05 Expected response response of of primary plant parameters primary plant parameters (temperature (temperature and and pressure) pressure) following following T/G T/G trip3.8 trip3.8 4.1 4.1 Match justification: This Match This question question requires requires knowledge knowledge of of the the expected expected response response (predicting the (predicting the changes changes in in the the parameters) parameters) of of primary primary plant plant Temperature Temperature and and following the Turbine Trip. Written to preclude steam dump operation pressure following due to testing steam dump operation elsewhere in exam. Objective: Objective: 15.
- 15. Describe the operation of the reactor's reactors inherent control systems as they function to re-establish a steady-state condition for the following transients: (OPS52701A (0PS52701A15) 15) 10% step load increase
- a. 10%
- b. 50% load rejection
- c. 10% ramp increase
- d. Entry into the power range
- 18. Determine the final condition of the plant for various transients assuming no operator (0P552701A1 9).
response (OPS52701AI9). Page: 92 Page: 92 of of 200 200 12/14/2009 12/14/2009
- 36. 051AA1.04
- 36. 051 AA1 .04 OOllBANKIROIMEM 00 1/BANK/RO/MEM 2.5/2.5/N/N/2/CVRIY 2.5/2.5/N/N/2/CVR!Y Unit 11 was Unit was at 100% power at 100% power and the following and the following conditions conditions exist:
exist:
** AOP-8.0, Partial AOP-B.O, Partial Loss Loss Of Of Condenser Condenser Vacuum, Vacuum, is is in in progress.
progress.
** A rapid A power reduction rapid power reduction per per AOP AOP 17.0, 17.0, Rapid Rapid Load Load Rejection, Rejection, was was completed.
completed.
** Condenser Vacuum Condenser Vacuum isis stable.
stable.
** FEI, CONT FE1, CONT RODROD BANK BANK POSITION POSITION LO, LO, is is in in alarm.
alarm. Which one Which one of the following of the following states: states:
- 1) whether or
- 1) or not not SS SS permission permission is is required required prior prior to to the Control Control Rod insertion during Rod insertion during the downpower lAW NMP-OS-001, Reactivity Management Management Program, and Boration required
- 2) the type of Boration required lAW the ARP for FE1? FE1?
Control Rod Insertion Boration A. SS permission is NOT required. Emergency boration ONLY. B SS permission is NOT required. B:t Normal OR Emergency boration. C. SS permission is required. Emergency boration ONLY. D. SS permission is required. Normal OR Emergency boration. Page: 93 of Page: 93 of 200 200 1211412009 12/14/2009
- Incorrect. Permission A - Incorrect. Permission not not required required onon insertion.
insertion. Emergency Emergency Boration Boration required required until until Rod Bank the Rod Bank LO-LO LO-LO Limit Limit alarm alarm clears. clears. With Rod Rod Bank Bank LO LO Limit Limit alarm alarm present, present, a normal boration normal boration is is required required until until the LO LO Limit Limit alarm clears clears so this is is plausible plausible for the student to confuse the the two requirements requirements for normal normal or or emergency emergency boration. boration. NMP-OS-OO1 and ARP-1.6 B - Correct. Per NMP-OS-001
- ARP-1 .6 FE1 FEI & & FE2. (See below)
C - Incorrect. Permission not required on insertion. Plausible, since it is always required to get SS permission for all positive reactivity additions, and it is expected to get permission when there is time to do so even for negative reactivity additions. However, for responding to a transient to stabilize the plant no permission is required to insert negative reactivity of any type. Emergency Boration is required until the Rod Bank LO-LO Limit alarm clears. With Rod Bank LO Limit alarm present, a normal boration is required until the LO Limit alarm clears so this is plausible for the student to confuse this. D - Incorrect. Permission not required on insertion. Emergency Boration required until the Rod Bank LO-LO Limit alarm clears. With Rod Bank LO Limit alarm present, a normal boration is required until the LO Limit alarm clears so this is plausible for the student to confuse this. NMP-OS-0O1, Reactivity Management Program, Version 13.0 NMP-OS-OOl, 6.3.8.1 During transient conditions that require a rapid reduction in reactor power, operators may take actions to insert negative reactivity that are outside the amounts discussed in the reactivity brief and without SS concurrence. ARP-1.6, FE1 ARP-l.6, FEl Annunciator: CONT ROD BANK POSITION LO, Version 64.0
- 5. Borate [NORMAL BORATION] the Control Bank "OUT" OUT as necessary using the Boron Addition Nomographs. {CMT 0008900}
ARP-1.6, FE2 Annunciator: CONT ROD BANK POSITION LO-LO, Version 64.0
- 2. Emergency borate the reactor coolant system in accordance with FNP-1-AOP-27.O, FNP-l-AOP-27.0, EMERGENCY BORATION.
BORATION. {CMTs 0008555, 0008900} 0008900) Page: Page: 94 94 of of 200 200 12/14/2009 12/14/2009
Previous NRC Previous NRC exam exam history history ifif any: any: 051 AAI .04 051AA1.04 051 Loss of of Condenser Vacuum AAI. Ability AAI. Ability to to operate operate and and 1/ or or monitor monitor the the following following as as they they apply apply to to the the Loss Loss of of Condenser Condenser Vacuum: Vacuum: (CFR (CFR 41.7 / 45.5 I 45.6) 41.7/45.5/45.6) AA1.04 Rod position .................................................... 2.5* 2.5* AAl.04 Match justification: The question presents a plausible scenario where a rapid power reduction is in progress in accordance with AOP-17. The examinee has to determine the correct procedural guidance given for control rod operation during insertions and withdrawals. The question pertains to whether SS permission is required for insertions and the predicted rod position when Emergency Boration may be terminated. Objective:
- 6. State the actions that the UO and/or OA OATCTC have the authority to perform in addition to being responsible to the Shift Supervisor (OPS52303HIO).
(0PS52303H10). Page: Page: 95 of 200 95 of 200 12114/2009 12/14/2009
07/02/09 06:30:42 07/02/09 06:30:42 FNP-I-ARP-I.6 FNP-1-ARP-1 .6 LOCATION FEl FE1 SETPOINT: SETPOINT: 10 Steps Variable; 10 Steps Greater than LO-LO Alarm El Setpoint. Setpoint. CONTROD CONT ROD ZLO = ZLOL0 + K4
= ZLO-LO 4 K BANK Where ~ 4=
K = 10 Steps (6.25 inches) POSITION LO ORIGIN: Rod Insertion Limit Computer PROBABLE CAUSE NOTE:
- Zinc Addition System injection will result in a continuous RCS dilution of as much as 1.7 gph, which may result in a reduction in shutdown margin if compensated for by inward rod motion instead of boration.
- This annunciator has REFLASH capability.
Reactor Coolant System Boric Acid Concentration too low for Reactor Power Level due to: A. Plant Transient B. Xenon Transient C. Dilution of RCS AUTOMATIC AUTOMA TIC ACTION NONE OPERATOR ACTION
- 1. Check indications and determine that actual control bank rod position is at low insertion limit.
1.1 Click on Rod Supervision button on Applications Menu. 1.2 Click on Rod Insertion Limits button. 1.3 Determine if low insertion limit exceeded.
- 2. IF reactor coolant system dilution is in progress, cI THEN stop dilution.
V3. IF a plant transient is in progress, THEN place the turbine load on HOLD."HOLD". I 4. 5. Refer to FNP-1-UOP-3.1, FNP-I-UOP-3.1, POWER OPERATIONS. Borate the Control Bank OUT"OUT" as necessary using the Boron Addition Nomographs. {CMT 0008900} 4>1 6. Refer to the Technical Specifications Specifications section section on Reactivity Reactivity Control. !f'- f}.j--' C f M f.7 ?Kec.-(~~
~ J..- J- 42
References:
A-177100, A-I77100, Sh. U-26061O; U266647 PLS Sh. 291; U-260610; PLS Document; Document; Technical Technical Specifications DCP {CMTs 0008554, 0008887} DCP 93-1-8587; {CMTs 0008887} Page Page 11 of of 11 Version 61.0 Version 61.0
07/02/09 06:30:42 07/02/09 06:30:42 FNP- 1 -ARP- 1.6 FNP-I-ARP-I.6 LOCATION LOCATION FE2 SETPOTNT: SETPOINT: Variable with Reactor Power as measured by E2 L\T and
.6.T and TAVG. CONT ROD CONTROD BANK ORIGIN: Rod Insertion Limit Computer POSITION LO-LO PROBABLE CAUSE NOTE: ** Zinc Addition System injection will result in a continuous RCS dilution of as much as 1.7 gph, which may result in a reduction in shutdown margin if compensated for by inward rod motion instead of boration.
- This annunciator has REFLASH capability.
- 1. Reactor Coolant System Boric Acid Concentration too low to ensure Reactor Protection under Accident conditions due to; A. Plant Transient B. Xenon Transient C. Dilution of RCS AUTOMATIC AUTOMA TIC ACTION NONE OPERATOR OPERA TOR ACTION
- 1. Check indications and determine that actual control bank rod position is at the low-low insertion limit.
~ 1.1 Click on Rod Supervision button on Applications Menu.
Acr~~J "'\ t- UJ ,.\ /)
. . r<-
v: lL!~;~~:!!:~:~1:,~t~:\:~;;~~~~e~;~~rdance wil i 2. 1.2 Click on Rod Insertion Limits button. 1.3 Determine if low-low insertion limit exceeded. Emergency borate the reactor coolant system in accordance with FNP- 1 -AOP-27.0, EMERGENCY BORATTON. FNP-I-AOP-27.0, {CMTs 0008555, 0008900) 0008900} BORATION . o 1cv'\ft,~D ~ 3. IF a plant transient is in progress,
?(,.{:- THEN place turbine load on HOLD.
"HOLD".
L
'1(\
wfl ! 'o~ 4. Refer to FNP-1-UOP-3.1, FNP-I-UOP-3.l, POWER OPERATIONS.
~;} ~ ~ A :5. Refer ~:~~:~
Control.
)hnical Specifications section on Reactivity to the Technical Specifications section on Reactivity t~ .p v-X f I" I 't-'~
References:
A-177100, A-I77IOO, Sh. Sh. 292; U-260610; U-26061O; U266647 PLS Document; Technical Technical Specifications; Specifications; DCP DCP 93-1-8587; {CMT {CMT 0008887) 0008887} Page Page 1I of of 11 Version 61.0 Version 61.0
Southern Nuclear Southern Nuclear Operating Operating Company Company I Nuclear Nuclear I I NMP-OS-001 NMP-OS-001 I I SOUTHERNA SOUIHERN'\ COMPANY Management Management I Reactivity Management Reactivity Management Program Program Version 13.0 Version 13.0 I COMPANY Erurgy to S""t Your WD,/J~ Procedure Procedure I Page Page 1212 of of 29 29 I 6.3.2 6.3.2 minimum, the As aa minimum, the Specific Specific Reactivity Reactivity Management Management Practices Practices contained contained in in Attachment be followed. 2 will be 6.3.3 The SS shall maintain direct supervisory oversight of reactivity manipulations. This {} I yt;k ~eans p,eans the SS will approve each reactivity manipulation, with the exception of c¥ ~ ~ u::;r('~onditfons described in step 6.3.8. DUring times where frequent reactivity oo ~ ,oJ"- Ilj manipulations are necessary, the SS can assign a reactivity management SRO to V V it? ,7 (0 3P < "P perform this function while the SS maintains oversight. d~ (:J;J [, 1 ~ 1 6.3.4 6.3.4 A reactivity brief shall take place at the beginning of each shift in modes 11 and 2. The reactivity brief should include expected reactivity manipulations during the shift needed to maintain current plant conditions or in the case of planned startups, shutdowns or power maneuvers the brief should include a discussion of reactivity changes that would be required to execute these power changes. In addition to this, the reactivity brief should include a discussion of pertinent current core reactivity parameters and any planned work activities that could potentially affect reactivity. The reactivity briefing sheet or OATC turnover sheet shall contain a list of degraded or out of service reactivity manipulation equipment. 6.3.5 When power reduction is necessary, only steam flow adjustments will be effective in reducing and maintaining reactor power below limits. While control rod insertion may appear to provide some immediate relief from high power conditions, the effects are temporary without reducing total steam flow and will only reduce nuclear instrument accuracy due to the resultant cooldown. Turbine load adjustments must be made to reduce and control reactor power, with control rods used primarily to maintain Tave on program during the power reduction. (PWR Only) 6.3.6 Peer checks will be used for reactivity changes, with the exception of conditions described in step 6.3.8. 6.3.7 During some plant operations, one or more of the various indications of reactor power may not be accurate. Therefore, control room operators should always monitor all indications of reactor power and maintain it within licensed limits. 6.3.8 Transient Conditions 6.3.8.1 During transient conditions that require aa rapid reduction in reactor power, operators ma take actions to insert ne ative reactivit that are outside the amounts discussed in the reactivity brief an without
'l{lthout SS concurrence. The operator peer checR m eac ivity requirement to have another licensed operatoEjirRTh IVlty manipulations under these conditions is also not required since it is unlikely that 1 other licensed operators would be available during the manipulation.
manipulation. The SS shall be briefed as soon as possible on the amount of negative reactivity added (number of steps steps of rod insertion, amount of boron added (PWR Only),Only), Recirculation Pump speed adjustments Pump speed adjustments (BWR (BWR Only), Only), etc. etc. I,1
- 37. OS4AG2.1.7
- 37. 054AG2. 1.700 1/NEW/RO/MEM 4.4/4.7/N/N/3/CVRlY OOlINEW/ROIMEM 4.4/4.7/N/N/3/CVR/Y A Unit Unit 11 SGFP trip has has occurred occurred from 100% 100% power, power, and and the the following following conditions conditions exist:
exist:
** Condensate And Feedwater AOP-13.0, Condensate Feedwater Malfunction, Malfunction, is is in in progress.
progress.
- The operator isis at the step to "Verify Verify automatic operation operation of of the Feedwater Feedwater Regulating Valves adequate".
Regulating adequate.
- SG NR levels are as follows:
- 1A 34% Rising
- 1I B 33% Rising
- 11 C 36% Rising Which one of the following is the correct method of controlling the Main Feed (MFRV5) during this transient lAW AOP-13.0?
Regulating valves (MFRVs) AOP-1 3.0? Place each MFRV controller in manual at (1) SG NR Level, (2) - (1)) (1 (2) A A'I 55% match steam flow and feed flow, and then place the controller back in automatic. B. 55% and then immediately place the controller back in automatic. C. 65% match steam flow and feed flow, and then place the controller back in automatic. D. 65% and then immediately place the controller back in automatic. Page: Page: 96 96 of of 200 200 12/14/2009 12114/2009
A - Correct. Per step 1.8
- 1 .8 and the "D.D. Operational Concern" Concern note of AOP-1 3.0. The AOP-13.0.
swell and the response time of the MFRV controller and valve necessitates taking the controller to manual at 55% (before 65% - program level) and matching feed and steam flows prior to taking it back to automatic. This prevents a high high level Turbine trip at 82% level which would occur due to the large Feed Flow Steam flow mismatch if feed flow is not reduced prior to 65%. B - Incorrect. The first part is correct (see A), but the second part is incorrect (see D). B - C - Incorrect. First part is incorrect, since manual control must be taken at 55% level instead of 65%. Plausible, since doing this at 65% would seem adequate, since that is the level desired to maintain. This could be chosen if the magnitude of the effects of swell and the response time of the controllers is not taken into account. By waiting until 65% to place the controller in manual, the feed flow is high enough that the time to reduce it combined with the expansion of the cooler feed water after getting to the SG can cause excessively high SG levels and a high high SG level trip of the Turbine and SGFPs and a FWIS. The second part is correct (see A). oD - Incorrect.
- The first part is incorrect (see C). The second part is incorrect, since placing in AUTO after taking to manual would not correct the high feed flow and lower the feed flow/steam flow mismatch quickly enough to prevent excessively high SG levels. Plausible, since taking the controller to manual will reset the windup and decrease the controller response time to a level transient, and this is an important part of the procedure guidance reason to go to manual. Also, AUTO control is preferred to manual when adequate for the magnitude of the transient. In smaller SG level transients, going to manual to reset the windup and then allowing AUTO to control the SG level is preferred.
FNP-1-AOP-13.0, Condensate And Feedwater Malfunction, Version 29.0 FNP-I-AOP-13.0, D. Operational Concerns II In the SO SG level recovery phase, the SOSG level will start increasing due to the feedwater flow being higher than steam flow and due to swell. If manual action is not taken before the SO SG reaches normal operating level, the combined effect of swell and additional feed flow may result in SO SGFP trip and Feedwater Isolation. Taking manual SG Hi-Hi Level Turbine and SOFP control and reducing the demand resets the level controller and flow controller integration circuits (i.e. windup) and makes the flow controller output track the associated driver card output. 1.8 Closely monitor steam generator narrow range levels. [{)) WHEN a SO SG narrow range level recovers verify its main feedwater to approximately 55%, THEN verifY regulating valve controllers in MANUAL. ((1] Match feed flow with steam flow. ((1] Return feedwater regulating valves to AUTO. Page: 97 of 200 12/14/2009 12/1412009
Previous NRC Previous NRC exam exam history history ifif any: any: 054AG2.1 .7 054AG2.1.7 054 Loss 054 Loss of Main Feedwater of Main Feedwater 2.1.7 Ability 2.1.7 Ability to evaluate plant to evaluate plant performance performance and and make make operational operational judgments judgments based based on on operating characteristics, reactor operating characteristics, reactor behavior, behavior, and and instrument instrument interpretation. interpretation. (CFR: 41.5 (CFR: 41.5 /43.5/45.12/45.13)
/43.5 /45.12 /45.13) RO RO 4.4 4.4 SRO SRO 4.7 4.7 Match justification: to Match to answer answer this this question question correctly, correctly, evaluation evaluation of of plant plant performance performance and operational judgement of how to operate the Main Feed Regulating Valves in given transient condition is required. Instrument the given Instrument interpretation is also included SG level in that with the SG level asas low low in the Narrow range range asas they are, MFRV controllers maximum output by will windup to maximum by the time the SG levellevel is normal level is at normal level of 65%,
65%, and this operating characteristic must must be be taken into account to get the correct answer. Objective:
- 4. EVALUATE plant conditions and DETERMINE ifany if any system components need to be AOP- 100, Instrument Malfunction. (OPS-52521Q06).
operated while performing AOP-IOO, (OPS-5252 1Q06). Page: 98 Page: 98 of 200 of200 12/14/2009 12114/2009
04/03/09 13 04/03/09 13:21:19
- 21: 19 FNP-1-AOP-13.0 FNP-1-AOP-13.0 CONDENSATE AND CONDENSATE AND FEEDWA FEEDWATER MALFUNCTION TER MALFUNCTION Version 29.0 Version 29.0 C.
C. Automatic Actions I Both MDAFWPs will automatically start on a trip of both SGFPs. 2 The TDAFWP will automatically start at 28% narrow range level in two steam generators. 3 A reactor trip will occur if any SG level decreases to 28% narrow range. 4 A turbine trip and feedwater isolation will occur if any SG level increases to 82% narrow range. 5 A reactor trip will occur if either intermediate range high flux trip bistable (TSLB3-2.1 (TSLB3-2. 1 or TSLB 3-2.2) does not reset before the reactor power is reduced below 10%. 10%. 6 Rapidly reducing turbine load may cause the steam dump system to operate in automatic. This will prevent further reduction in total steam flow. Dump operation in steam pressure mode at pot settings other than specified for 1005 psig will affect RCS cooldown rate. D. Operational Concerns 1 I the SG level recovery In recover phaseJ.he hase the SG level will start increasing increasin due to the feedwater flow swel. If manual action is not taken before the SG being higher than steam flow and due to swell. effe of swell and additional feed flow may
,.reaches normal operating lev,S(l, the combined effect result in SG Hi-Hi Level Turbine and SGFP trip and Feedwater Isolation. Taking manual anrudugth control an r* cli and resets the level controller and flow controller integrat~
contiMi?1iiegiatii I.e. windup) and makes the flow controller output track the associated driver card circuits1t. cIrcuits output.
/tt 1 p
Page Page 22 of 23 of23
04/03/09 13:21:19 04/03/0913:21:19 i~,,") .J. ~~l J:;.. JL I.L FNP-1-AOP-13.0 FNP-I-AOP-13.0 CONDENSATE AND CONDENSATE FEEDWATER AND FEEDWA MALFUNCTION TER MALFUNCTION Version Version 29.0 29.0 Step Action/Expected Response Response Response Response NotNot Obtained ~ Action/Expected Obtained I I I NOTE: NOTE: ** Steps through 1.3 Steps 1I through 1.3 AND through 2.1 are AND 2 through IMMEDIATE ACTION are IMMEDIATE ACTION steps. steps.
** This procedure This procedure steps through probable steps through probable condensate condensate and and feedwater feedwater system system malfunctions malfunctions in in aa systematic diagnostic manner. If the cause of of the condensate and feed malfunction malfunction is known, THEN the associated procedure section (step) may be implemented immediately.
- 1. Single SGFP Single SGFP trip - step
- step 1I
/74/ j Both SGFPs tripped - step 2 - step 2 4 7 3. SGFP malfunction - step 3 OBSERVE CAUTION prior to step 3 r 4. Main feedwater regulating valve malfunction - step 4 -
r ,
- 5. Loss of feedwater heater - step 5 -
- 6. SGFP low suction pressure - step 6 -
1 Check onlyoniy one SGFP - RUNNING
- I Proceed to step 2.
1.1 Check generator load GREATER THAN 1.1 Proceed to step 3 OBSERVE CAUTION 540 MW 540MW prior to step 3. 1.2 Check rapid turbine load reduction 1.2 IF rapid turbine load reduction NQI IE NOT required. required, THEN reduce turbine load using normal DEH controls as required AND proceed to step 3. 1.3 Check DEHC in OPERA OPERATOR TOR AUTO 1.3 Perform the following 1.3.1 Depress the SGFP SETPOINT button a) IF required by generator load on the DEHC keypad
- THEN press FAST ACTION AND 1.3.2 Press the P8 "P8" key GV CLOSE pushbuttons 1.3.3 1.3.3 On the SGFP SETPOINT screen verify On the
- Release the FAST
- FAST ACTION ACTION and GV GY the the following appear: CLOSE pushbuttons at::::
CLOSE pushbuttons at 730 MWe 730 MWe indicated on as indicated on the digital DEHC [] A TARGET of [1 540 MW of"540" display [1 [] AA RATE of"1200" MW/Mm RATE of1200 MWlMin b) IF b) IF not not required required by by generator generator load, load, 1.3.4 1.3.4 Depress the Depress "GO" pushbutton the GO pushbutton THEN press THEN press the the GV GV CLOSE CLOSE pushbutton as pushbutton as required required to to reduce reduce load load o Step 11 continued Step continued on on next next page page Page Page Completed Completed Page Page 33 of 23 of23
04/03/09 13 04/03/09 13:21:19
- 21: 19 "i,,,) .,tt;i J:
FNP-1-AOP-13.0 FNP-1-AOP-13.0 CONDENSATE CONDENSA AND FEEDW TE AND FEEDWATERA TER MALFUNCTION MALFUNCTION Version Version 29.0 29.0 Step Action/Expected Response
Response
~ Action/Expected Response Response Not Obtained Obtained I I I 1.4 1.4 Monitor for correct DEHC system response NOTE: A boration of 11 GAL per reduced MW will limit rod insertion and assist in maintaining Delta I. r 1.5 Reduce reactor power to match turbine
~
power using control rods and boron. 1.5.1 Verify rod control in MANUAL. 1.5.2 Adjust control rods in MANUAL to reduce reactor power and control RCS TAVG. V Nj
- Manual Rod Control
- Manual boration per FNP-1-S0P-2.3, FNP-1-SOP-2.3, CHEMICAL AND VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM.
** Emergency boration per FNP-1-SOP-2.3, CHEMICAL AND FNP-1-S0P-2.3, VOLUME CONTROL SYSTEM REACTOR MAKEUP CONTROL SYSTEM Figure 6.
1.6 1.6 Check proper operation of the steam dumps. 1.7 1.7 Verify automatic operation of the 1.7 1.7 Take manual control of the feedwater feedwater regulating valves adequate. regulating valves to control SG level. [] IASGFWFLOW 1ASGFWFLOW FK478 [] 1BSGFWFLOW 18 SGFWFLOW FK 478 FK478 [] ICSGFWFLOW 1C SG FW FLOW FK 498 FK498 o Step I1 continued on Step on next page Page Page Completed Completed Page 44 of Page 23 of23
04/03/09 13 04/03/09 13:21:19
- 21: 19 FNP-1-AOP-13.0 FNP-I-AOP-I3.0 CONDENSATE CONDENSA AND FEEDW TE AND FEEDWATER MALFUNCTION A TER MALFUNCTION Version Version 29.0 29.0 Step Action/Expected Response Action/Expected Response Response Not Response Not Obtained Obtained I I 1.8 1.8 Closely monitor steam generator narrow 1.8 1.8 IE SG narrow range levels NOT IF NQI maintained range levels. greater than 28%,
THEN trip the reactor and go to [1 WHEN a SG narrow range level recovers [] FNP-1-EEP-0, REACTOR TRIP OR FNP-I-EEP-O, to approximately 55%, THEN verifY verify its main feed feedwater SAFETY INJECTION. water regulating valve controllers MANU.AL. ~ controller in MANUAL. E I1 tec [1 Match feed flow with steam w. c-_* []
/
((]nfwater
] regulating_valves eturn feed water regulating to ~ '"">
valves t.£. AUTO. :.....> 1.9 Monitor feedwater flow and steam flow. 1.10 Verify that feedwater and steam flow trend to approximately equal values for the target, turbine load. 1.11 Maintain SG narrow range level approximately 65%. CAUTION: The LOSS OF LOAD INTERLOCK C 7 7A A should not be reset in the event of a failure of PT-447 PT-44 7 which actuates C-7A C-7A without consultation with the Operations manager. 1.12 Check LOSS OF LOAD INTERLOCK 1.12
- 1. 12 IF C-7 C-7AA is to be reset, C-7A C-7 A on the BYP & & PERMISSIVES THEN perform the following panel NOT illuminated.
1.12.1 1.12.1 Verify VerifY that all steam dump valves indicate closed. 1.12.2 Verify 00 demand on STM HDR PRESS controller PK 464 and STM DUMP DEMAND TI408 1.12.3 1.12.3 Place STM DUMP INTLK TRAIN A A and TRAIN B B to OFF RESET 1.12.4 1.12.4 Place STM DUMP Place DUMP MODE SEL SEL TRAINS A-B to RESET and then spring return to TAVG. release to spring TA VG. o Step Step I1 continued continued on on next next page page _Page Completed Page Page 55 of of2323
- 38. 001/MOD/RO/MEM 4.3/4.6/N/NI2/CVRN 055EK3.02 OOlIMOD/ROIMEM 4.3/4.6/N/NI2ICVR/Y ECP-O.O, Loss of All AC Power, directs the operator to:
- Dump steam from intact SGs at maximum controllable rate.
Which one of the following describes the primary reason for the step which directs dumping steam from intact SGs at maximum controllable rate? A. To minimize potential for SGTR. B B~ To minimize RCS inventory loss. C. To maximize TDAFW pump flow. D. To prevent steam voiding in the reactor vessel upper head. A - Incorrect. This is plausible, since cooling down and reducing the volume of the RCS would reduce RCS pressure, and reducing SG tube dip d/p is desirable, but this is not the reason for the depressurization. B - Correct. This is the Background Document basis for this step: 16.4 in ECB-O.O. (See below). ERG Step Text: The SGs should be depressurized at maximum rate to minimize ReS RCS inventory loss.
Purpose:
To inform the operator of the desired rate for depressurization of steam generators Basis: The intact steam generators should be depressurized as quickly as possible, to minimize RCS inventory loss ReS loss... C - Incorrect. Plausible, since the ECP-O.O note prior to step 4 does remind of the 2 hour limit on air accumulator supply and UPS power supply, and the heat sink provided by the TDAFW pump is the main source of core cooling with no AC power. Long term, the need to remove decay heat would extend beyond the 2 hours, and a lower SG pressure would allow more water to be pumped to the SG's SGs in the initial 2 hour period. However, the background document requires only SG level of >31% >31 % narrow range on one SG for an adequate heat sink. D - Incorrect. Plausible, since cooling the RCS would eventually cool the vessel head, D - and without CRDM fans running would be the main cooling for the head. However, the short term as stated in a note in the procedure is that head voiding may be caused by the depressurization. ECP-O.O, Loss Of All Ac Power, Revision 22 CAUTION: The TDAFWP will become unreliable within 22 hours following aa loss of all AC power, unless power is restored. This will occur due unless due to aa loss of air to the steam supply valves and aa loss of control power from the UPS. 44 Verify total AFW flow GREATER 44 Verify proper AFW alignment. THAN 395 gpm.gpm. 16.4 16.4 Dump steam from intactintact SGs at Page: 99 of 200 of200 12/14/2009 12/1412009
maximum controllable maximum controllable rate. rate. FNP-O-ECB-O.O FNP-O-EC8-0.0 Section: Procedure 16.4 Unit 2 ERP Unit 11 ERP Step: 16.4 16.4 ERG Step No: 16 ERP Step: 16.4 16 NOTE-1 NOTE-i Dump steam StepText: Dump ERP StepText: from intact steam from intact SGs SGs at maximum controllable at maximum controllable rate. rate. ERG Step Text: The SGs should be depressurized at maximum rate to minimize RCS inventory loss.
Purpose:
To inform the operator of the desired rate for depressurization of steam generators Basis: The intact steam generators should be depressurized as quickly as possible, to minimize ReS RCS inventory loss, but within the constraint of controllability. Controllability is required to ensure that steam generator pressures do not undershoot the specified limit. For the reference plant, the operator can control the secondary depressurization from the control room. In this case, maximum rate means steam generator PORVs full open. For plants that must control the secondary depressurization by local actions, maximum rate must be determined by the control room and local operators based on plant conditions and available communications. A slower rate is acceptable for locally controlled secondary depressurization. See Subsection FN P-O-ECB-O.O FNP-O-EC8-0.0 Section: Procedure Unit 11 ERP Step: 3 Unit 2 ERP Step: 3 ERG Step No: 3 ERP StepText: Verify RCS isolated. ERG Step Text: Check If StepText: IfRCS ReS Is Isolated
Purpose:
To ensure all RCS outflow paths are isolated Basis: A check for RCS isolation is performed to ensure that RCS inventory loss is minimized. The valves itemized are those in major RCS outflow lines that could contribute to rapid depletion of RCS inventory ofRCS inventory....
.... Following completion of this step, the only ReSRCS inventory leakage path should RCP controlled leakage seals be the Rep seals....
.... The secondary depressurization in Step 16 will minimize RCS inventory loss by reducing RCS pressure which will terminate or minimize relief valve flow. For example, reducing RCS pressure to 400 psig would permit the letdown line relief valve to close and would minimize flow through the excess letdown relief valve.
Knowledge: Need to minimize RCS inventory depletion during loss of all ac power event to maximize time to core uncovery. Page: 100 of 200 12/14/2009 12/1412009
Previous NRC exam history if any: 055EK3.02 055 Station Blackout EK3 Knowledge of the reasons for the following responses as the apply to the Station Station Blackout: Blackout: (CFR41.5 (CFR /41.10 41.5 141.1 / 45.6 /45.13) 0/45.6/45.13) EK3.02 Actions contained in EOP for loss of offsite and onsite power .......... 4.3 4.6 Match justification: This question requires knowledge of a response in the EOP (ECP-O.O) which is required to minimize inventory loss during a Station Blackout. Objective:
- 3. STATE AND EXPLAIN the basis for all Cautions, Notes, and Actions associated with (1) ECP-O.O, Loss of All AC Power; (2) ECP-O.l, Loss of All AC Power Recovery, Without SI Required; (3) ECP-O.2, Loss of All AC Power Recovery, With SI Required.
(OPS-52532A03) Page: 101 of 200 12/14/2009 12/1412009
06/27/07 16:11:08 06127/07 16:11:08 FNP-O-ECB-O.O FNP-0-ECB-0.0 LOSS OFOF ALL ACAC POWER Plant Specific Plant Specific Background Information Section: Procedure Unit 11 ERP Step: 16.4 Unit 2 ERP Step: 16.4 ERG Step No: 16 NOTE-1 NOTE-i ERP Step Text: StepText: Dump steam from intact SGs at maximum controllable rate. StepText: ERG Step Text: maximum rate to minimize ReS The SGs should be depressurized at maximum RCS inventory loss.
Purpose:
To inform the operator of the desired rate for depressurization of steam generators Basis:
.? The intact steam generators should be depressurized as quickly as possible, to minimize RCS inventory loss, but within the constraint of controllability. Controllability is required to \7 (jJ'" .
J: // ensure that steam generator pressures do not undershoot the specified limit. For the reference plant, the operator can control the secondary depressurization from the control room. In this case, maximum rate means steam generator PORVs full open. For plants that must control the secondary depressurization by local actions, maximum rate must be determined by the control room and local operators based on plant conditions and available communications. A slower rate is acceptable for locally controlled secondary depressurization. See Subsection
2.3. Knowledge
SG depressurization should proceed as quickly as possible and should not be limited by the 100°F/hr. Technical Specification RCS cooldown limit of 100°Flhr.
References:
Justification of Differences:
.Justification 1 Since the ERG Note contains a directed action it was incorporated into the ERP step for clarification and to enhance procedure flow.
56 of 88 Version: 1.0
06/27/07 16:11:08 06127/07 16:11:08 FNP-O-ECB-O.O FNP-0-ECB-0.0 LOSS OF LOSS OF ALL ALL AC AC POWER POWER Plant Specific Background Plant Specific Background Information Information Section: Procedure Unit 1 ERP Step: 3 Unit 2 ERP Step: 3 ERG Step No: 3 ERP StepText: Verify RCS isolated. StepText: ERG Step Text: Check If RCS Is Isolated
Purpose:
To ensure all RCS outflow paths are isolated Basis: A check for RCS isolation is performed to ensure that RCS inventory loss is minimized. The valves itemized are those in major RCS outflow lines that could contribute to rapid depletion of RCS inventory. This step is written for plants which utilize air operated valves (AOV (AOVs) s) in the itemized locations. The step structure assumes that the AOV AOVss fail closed on loss of all ac power (i.e., loss of air supply). The operator, therefore, checks that the valves are closed. If any valve is open, the operator should attempt to close the valve. Reasons for a valve remaining open are plant specific, for example the valves may have legitimate or spurious open signals and air pressure could be available due to air receivers or air bottles located in the air supply system. Plants with air receivers may take up to 30 minutes to lose air pressure. If nitrogen bottles are provided for specific valves, such as PORV PORVs, s, pneumatic pressure may be available for more than 30 minutes. The sequence for checking valves is based on capacity of the outflow lines and potential for RCS inventory loss: 1) The pressurizer PORVs are checked first. Since the turbine_driven AFW pump should be running, the secondary side is removing decay heat and RCS pressure should be under the pressurizer PORV setpoint. 2) The letdown line isolation valves adjacent to the RCS loop are checked second. These valves are normally open and receive a low pressurizer level isolation signal. If these valves, in conjunction with the letdown orifice isolation valves, remain open, a leak path to the pressurizer relief tank (PRT) via the letdown line relief valve may exist. These valves, including the letdown orifice isolation valves, if necessary, should be manually closed as soon as possible to isolate the letdown line and minimize RCS inventory loss prior to automatic isolation on low pressurizer level. Note that isolating the letdown line at the containment penetration will not isolate the letdown relief valve leak path to the PRT. STEP DESCRIPTION TABLE FOR ECA-0.OStep3 3) The excess letdown line isolation valves adjacent to the RCS loop are checked third. These valves are normally closed and do not receive a low pressurizer level isolation signal. If If these valves are open a leak path to the PRT via the RCP seal return relief valve may exist. These valves should be closed to isolate the excess letdown line. Note that isolating the seal return line at the containment penetration will not isolate excess letdown inventory loss to the PRT via the seal return relief valve. 4) Any additional plant specific RCS outflow lines should be included. Following completion of this step, the only RCS inventory leakage path should be the RCP controlled leakage seals. Plants which utilize motor operated valves (MOVs) for letdown or excess letdown isolation will not be able not be able to remotely remotely close close these valves to isolate isolate these RCS RCS outflow paths. paths. These plants should isolate at containment. The isolate these lines at The secondary secondary depressurization in in Step Step 1616 ~ will minimize RCS inventory inventory loss loss by by reducing reducing RCS RCS pressure pressure which which will terminate or or ';-> ~ minimize minimize relief valve valve flow. For For example, example, reducing RCSRCS pressure pressure to to 400 400 psig psig would permit permit thethe Ce:; 1e( ({t::::.eJ{ letdown letdown line line relief valve valve to close close and and would minimize minimize flow through through the the excess excess letdown relief relief valve. An alternative valve. An alternative which which can be evaluated can be based on evaluated based on plant plant specific specific considerations considerations isis to dispatch personnel personnel inside inside containment containment to manually manually close close the the subject isolation valves. valves. 11 of 88 11 of 88 Version: 1.0 Version: 1.0
16:11:08 06/27/07 16: 11 :08 FNP-O-ECB-O.O FNP-0-ECB-0.0 LOSS OF ALL AC POWER Plant Specific Background Information Section: Procedure Knowledge: Need to minimize RCS inventory depletion during loss of all ac power event to maximize time to core uncovery
References:
Justification of Differences: 1 Changed to make plant specific. 2 verify vice "check" Used action verb "verify" check since the intent of the step is to ensure the RCS is isolated. 12 of 88 1.0 Version: 1.0
,t) ; ~',l t,~ j FNP-l-ECP-O.O FNP-1-ECP-0.0 LOSS LOSS OF OF ALL ALL AC AC POWER POWER Revision Revision 22 22 Step Step Action/Expected Response Action/Expected Response Response Response NOT NOT Obtained Obtained n I I I 3.4 3.4 Verify all Verify all reactor reactor vessel vessel head head vent vent valves valves - CLOSED.
CLOSED. RX VESSEL RX VESSEL HEAD HEAD VENT OUTER VENT OUTER ISO ISO [] Q1B13SV2213A [] 01B13SV2213A [1 Q1B13SV2213B [] Q1B13SV2213B RX VESSEL HEAD VENT INNER ISO [1 Q1B13SV2214A [] 01B13SV2214A [1 Q1B13SV2214B [] CAUTION: The TDAFWP will become unreliable within 22 hours following a loss of all AC power, unless power is restored. This will occur due to a loss of air to the steam supply valves and a loss of control power from the UPS. 4 Verify total AFW flow GREATER 4 4 Verify proper AFW alignment. THAN 395 gpm. 4.1 Verify TDAFWP running. AFW FLOW TO 1A(1B,1C) SG 1A(1B,lC) SC TDAFWP STM SUPP [1 [] Fl FI 3229A FROM 1B SG (1C) SC 1B(1C) [] Fl FI 3229B [] [1 MLB-4 1-3 lit [] Fl FI 3229C [] MLB-4 2-3 lit [] MLB-4 3-3 lit AFW TOTAL FLOW FLOW TDAFWP SPEED [1[] FlFI 3229 3229 [1[] SI SI 3411A 3411A > 3900 rpm
> 3900 TDAFWP SPEED CONT SPEED CONT
[1[] SIC SIC 3405 3405 adjusted adjusted to to 100% 100% 4.2 4.2 IF TDAFWP j TDAFWP NOT NOT running, running, THEN locally locally verify verify TDAFWP TDAFWP TRIP THROTTLE TRIP THROTTLE VLV VLV Q1N12MOV34O6 Q1N12MOV3406 open. (100 open. (100 ft. AUX BLDG ft, AUX BLDG TDAFWP room) TDAFWP room) Step 44 continued Step continued on on next next page. page. Page Completed _Page Completed Page 44 of Page of 40 40
'lJ t:i J 1; 1 FNP4-ECP0.0 FNP-1-ECP-0.O LOSS LOSS OF OF ALLALL AC AC POWER POWER Revision Revision 22 22 Step Step Action/Expected Response Action/Expected Response Response Response NOT NOT Obtained Obtained n I I I CAUTION:
CAUTION: Accumulator Accumulator nitrogen nitrogen injection injection into into the the RCS RCS may may result result fromfrom reduction reduction of of SG SC pressure pressure to to less less than than 100100 psig. psig. NOTE: Reduction of of intact SGs SOs pressure should should continue continue eveneven ifif pressurizer lost or level is lost or reactor vessel head voiding occurs.~ occurs. j)
, .----- P /l-\~CY'iV/-
;k 16 Reduce intact SGs pressure to 200 psig.
16.1 Check at least one intact SG SC 16.1 Perform the following: narrow range level - GREATER THAN 31% (48%) . 31%(48%}. 16.1.1 Maintain maximum AFW flow to intact SGs SOs until narrow range SG SC level greater than 31%(48%) 31%R8%} in at least one SG. SC. TDAFWP SPEED CONT [] SIC 3405 adjusted to 100% 16.1.2 WHEN narrow range level in at least one intact SG SC is greater than 31%(48%), 31%{48%). THEN perform steps 16.2 through 16.7. 16.1.3 Proceed to step 17. Step 16 Step 16 continued continued on on next next page. page. ___Page Page Completed Completed Page 29 Page 29 ofof 4040
- 1. EC-O.OI.l.2-52532A03
- 1. EC-0.0I. 1 .2-52532A03 003iHL 003/HLTIIMEM 4.314.6IEPEO55EK3 .021111 T/IMEM 4.3/4.6/EPE055EK3.021111 055 EK3 .02 055EK3.02 ECP-0.0, Loss ECP-O.O, Loss ofof All All AC AC Power, Power, directs directs the the operator operator to:to:
Reduce intact
* "Reduce intact SGs to 200 SGs to 200 psig:"
psig: Which ONE of of the following correctly describes the reasonreason for for stopping the SG pressure reduction at 200 psig? A. To To prevent losing losing Pressurizer Pressurizer level. level. B. To minimize RCS inventory loss out of the RCP seals. C. To prevent steam voiding in the reactor vessel upper head. D D~ To prevent injection of SI Accumulator nitrogen into the RCS. Page: 11 of Page: of33 10126/2009 10/26/2009
EP E055 E K3 .02 EPE055EK3.02 055 Loss of Offsite and Onsite Power (Station Blackout) Knowledge of the reasons for the following responses as the apply to the Station EK3 Knowledge Blackout: (CFR 41.5/41.10/45.6/45.13) EK3.02 Actions contained in EOP for loss of offsite and onsite power power... 4.3 4.6 A INCORRECT Plausible, since PRZR level may be lost as stated in ECP-O.O, ECP-0.0, and is an undesirable condition. Also the note below the caution says the SG pressure reduction should continue if voiding occurs or if Przr level was lost. For this event, depressurization of SGs should be continued even if this does occur so this answer is incorrect. B INCORRECT This is given as the reason why the pressure reduction is being done, not why the pressure reduction is only to 200 psig. This distractor is plausible in that it is closely related to the FINAL PRESSURE to which the SGs are to be depressurized and describes the reason why the depressurization is done, not why it is stopped at 200 pisg. In other procedures, the accumulator MOVs are isolated and and pressure reduction is continued to 100 psig, but with no power, this procedure has pressure maintained at 200 psig. Seal Injection & CCW to thermal Barriers are lost in a LOSP with Loss of all AC power. Seals are a major concern in this procedure. C INCORRECT Plausible, since voiding may occur as stated in ECP-O.O. ECP-0.0. For this event, depressurization of SGs should be continued even if this does occur so this answer is incorrect. Also the note below the caution in ECP-O.OECP-0.0 says the SG pressure reduction should continue if voiding occurs or if Przr level was lost. oD CORRECT This is given as the basis.
REFERENCES:
ECP-0.0, Loss of All AC Power, Step 28 and NOTES and CAUTION on page 28.
- 1. ECP-O.O, CAUTION: Accumulator nitrogen injection into the RCS may result from reduction of SG pressure to less than 100 psig.
Finally, the operator should be aware of the limiting low pressure necessary to prevent introduction of noncondensibles from the accumulators. Understanding these considerations, the ofnoncondensibles operator will be able to depressurize and control secondary pressure to minimize RCS ReS inventory loss while minimizing the possibility of introducing nitrogen into the RCS Res and returning the reactor core to a critical condition. Reduce Intact SG Pressure to 200 psig SOs are to be depressurized in the this step to maximize delivery (into the RCS) SGs ReS) of the water contained in the SI accumulators while minimizing introducing nitrogen (into the RCS). ReS). Introducing nitrogen into the RCS Res could impede natural circulation by collecting in high points ReS piping (e.g., SO U-tubes). An ideal gas expansion calculation of the RCS calculation was used to determine the pressure in the RCS Res following a complete discharge of the contents in the accumulators. This ReS pressure is then correlated to the SG RCS SO pressure (100 psig), which would be indicated when the accumulators have fully discharged. Page: 220f3 of 3 10/26/2009 10/2612009
2008 NRC 2008 exam NRC exam Technical
Reference:
Technical
Reference:
ECP-0 rev ECP-O rev 22 22 and and and and the the background background documents documents for for ECP-O, ECP-0, FNP-0-ECB-0.0 Loss FNP-O-ECB-O.O Loss ofof ALL ALL AC AC Learning Objective: Learning Objective: State the State basis for the basis for all all cautions, cautions, notes, notes, and and actions actions associated associated with with EEP-3 EEP-3 (0P552530A03) (OPS52530A03) Comments: Comments: This question matches the KJA
- This
- K/A in in that itit asks the applicant to describe the reason for a particular action contained in the EOPEOP for this event.
Page: 33 of Page: of33 10/2612009 10/26/2009
- 39. 057AG2.4.49
- 39. 057AG2.4.49 OOlINEW/RO/C/A 00 1/NEW/RO/C/A 4.6/4.4/N/N/3/CVRIY 4.614.4/N/N/3/CVRJY Unit 11 is Unit is ramping ramping down down for an an outage outage at MW/mm. The at 2 MW/min. The following conditions conditions occurred:
occurred: At 1000: 1000:
** Reactor Reactor power power is is 25%.
25%.
** The Reactor Reactor Makeup Makeup Control Control System System is is aligned for repetitive repetitive batch batch borations.
borations.
- A boration is NOT currently in progress.
- LK-112, LLTDN TDN TO VCT FLOW, has been adjusted to maintain 45% level in the VCT.
- LT-112B LT-1 12B and LT-115, LT-1 15, VCT LVL, meters both indicate 45%.
At 1001 the following occurs:
- 1IA A 120V Vital AC Instrumentation Panel is de-energized due to an electrical fa u It.
fault. Which one of the following is the correct operator response to these conditions? A. Secure BOTH Reactor Makeup Water Pumps. B. Realign the Reactor Makeup system to AUTO. LCV-1 15A, VCT HI LVL DIVERT VLV, to the VCT. C Realign LCV-115A, C'!" D. Increase the ramp rate to control Tavg within the limits of AOP-17, Rapid Load Reduction. Page: Page: 102 102 of of 290 200 12/14/2009 12/1412009
A - Incorrect. The RMW pumps and Boric Acid Transfer pumps don't
- dont start unless the RMW control is in Auto when the 1A 1A 120V 120V bus fails. In the conditions given, the system is aligned per repetitive batch borations, which is common during a ramp, and the failure of the 1A1A 120V 120V bus will not give an auto makeup. Plausible, since if the RMU system was in Auto, OR if an applicant believed that AUTO makeup occurred immediately in the borate mode when the 1A 120V vital panel failed, this choice would be selected.
B - Incorrect. Aligning to AUTO is incorrect since it would cause an AUTO makeup to occur regardless of VCT level, and it would not automatically stop. Also,in the repetitive batch Boration alignment, no BAT pump starts and no valves open to cause a boration. Plausible, since confusion between the effects of the 120V vital panel failure in each switch position may exist. In AUTO, the failure causes an Auto makeup to commence. However, in BORATE, an automatic Boration does NOT commence. Failures of other 3 vital instrumentation panels (1 B, 1 I C, 1I D) do not cause auto makeups with the control switch in AUTO, so the effects of each of the four 120V Vital panels may be misunderstood. This choice would be selected if an applicant thought that a Boration from the BAT occurred immediately with the switch in BORATE, but that an auto makeup did not occur with the switch in the AUTO position. C - Correct. The LCV115A
- LCVI 1 5A does divert letdown to the RHT, and will not automatically divert back to the VCT regardless of VCT level per ARP for WD1 WDI::
- VCT Hi Lvi Divert Valve Q1E21LCV1
- QIE21LCV115A
- 15A diverts to the RHT if in auto.
In addition, an auto makeup will not occur with the switch in BORATE. Prompt action to realign LETDOWN to the VCT must occur to stop the letdown diversion prior to realignment of the RWST to Chg pump suction valves which would cause a significant boration and reactivity event and reactor power transient at the end of life occurred. if it occu rred. D - Incorrect. This is incorrect since Boration from the RWST does not occur immediately requiring this action unless LT-112 is out of service when the 1A 120V vital panel fails. Plausible, since it could occur if LT-112 LT-1 12 had also failed or was out of service per the FNP-O-ARP-2.2 WDI: WD1:
". If
- LT 112 VCT level is out of service, RWST to Chg Pump Suction Valves IfLT Q2E2ILCVI
-Q2E2ILCVI15B 15B & D open.
open." Ran on SIMULATOR from 100% 100% (1C73): (IC73): IN repetitive boration: No BAT pump starts, the B B RMW pump does not start (A RMW pump is normally running all the time), and no valves open (113A, 113B, 113B, 114A, & 114B) 114B) In AUTO: 0/S O/S BAT PUMP STARTS, BB RMW PUMP STARTS, 1114B 14B RMW TO THE BLENDER modulates open, 11 3A BORIC ACID TO BLENDER modulates open, 11 113A 3B MKUP TO 113B CHG PUMP SUCTION HDR opens. Page: 103 103 of 200 12/14/2009
Previous NRC Previous NRC exam history history ifif any: any: 057AG2.4.49 057 Loss of Vital AC Electrical Instrument Bus Bus 2.4.49 Ability to perform without reference to procedures those actions actions that require immediate operation of system components and controls. (CFR: 41.10/43.2/45.6) 41.10/43.2 / 45.6) RO 4.6 SRO 4.4 Match justification: The Chief examiner was consulted about the difficulty in meeting this k/a with a discriminatory question on the very few immediate actions that are committed to memory. Most "immediateimmediate actions" actions that are performed "without without procedures have been eliminated in recent years. He recommended reference to procedures" using actions that must be performed promptly to avoid adverse effects to the plant, whether a procedure directs them or not. This question requires knowledge of the immediate effects of this loss of vital AC Electrical Instrument Bus that would require actions very quickly to mitigate or prevent undesired plant effects. Also, to answer this question correctly, it requires knowledge from memory without a reference being provided such as immediate operator actions. This question fits these criteria, and thus matches the original intent of the KIA.K/A. A set of conditions in which a loss of a 120V Vital AC electrical Instrument Bus is given for which prompt actions must be taken to prevent a large boration which would cause a large undesired transient at EOL core conditions. A significant transient, and likely a manual trip requirement, would occur if actions are not taken. Objective:
- 2. STATE ST ATE AND EXPLAIN any special considerations such as safety hazards and plant condition changes that apply to the 120 Volt AC Distribution System (OPS-52103D04).
Page: 104 Page: 104 of 200 of200 12/14/2009 12/1412009
07/01109 15:26:52 07/01/09 FNP-0-ARP-2.2 LOCATION WD1 SETPOINT: 1. Battery near exhaustion 107V DC.
- 2. Inverter output undervoltage 108V AC 1A INV 1AINV FAULT ORIGIN: 1. Battery near exhaustion X7 Voltage sense board
- 2. Inverter current limit A3 Ammerter Relay
- 3. Inverter output undervoltage K3 Relay via X8 voltage sense board
- 4. Inverter overheating X10 relay board
- 5. Out of sync X12 relay board
- 6. Fan failure
- 7. Bypass source supplying load PROBABLE CAUSE
- 1. Bypass source carrying load.
- 2. Inverter out of sync with bypass supply.
AUTOMATIC ACTION
- 1. IF DC input voltage drops to 103 V DC, THEN inverter transfers to bypass source.
- 2. IF inverter fails, THEN the bypass source should carry the load.
- 3. An inverter fault when the bypass source is not available resulting in a loss of power to 1A 120 VAC Vital Instrument Panel will be indicated by the following:
A. Source Range Channel 31 will be de-energized. B. Intermediate Range Channel 35 will be de-energized. CAUTION: <)Iltwardrodmotionis Outward rod motion is blocked by the High Power Rod Rod Stop Bistable being t~ipped. tripped. . .. C. NI -41 will be de-energized with associated alarms and indications. NI-41 D. Annunciators FD3, FD4, DF1 and DK3 will alarm. E. No amperage indication on 1A Inverter ammeter. F. IF RCP breaker indication is lost> 35% power, the reactor will trip. of 3 Page 11 of3 Version 26.0
15:26:52 07/01/09 15 :26:52 FNP-0-ARP-2.2 LOCATION WDI WD1 OPERATOR OPERA TOR ACTION NOTE: The following controls are affected if 1A 120 V AC Vital Instrument Panel is VAC Dc-energized: De-energized:
- A TRN SSPS output relay power is lost.
c-
;!O.* VCT Hi LvI Divert Valve - Q1E21LCV115A diverts to the RHT ifin auto._ ) t
'\; t 1 pi_ I
~)
~.* LTDN Hi Temp Divert Valve - Q1E21TCV143 bypasses the demineralizers.
(.iA1A &
.1A 1A &
& 1B
& 1B lB BAT pumps start.
lB Reactor makeup water pumps start.
~ RMW to Blender - Q1E21FCV114B and Boric Acid to blender -
C
?>
M
":-l ~\....Q1E21FCV113A Q1E21FCV113A opens opens1.x S
if M/IJ Control M/u Cotitro it System is in auto.
/;) VA l./"'.
- LT 11 VCT level is out of service, RWST to Chg Pump Suction Valves -
1E21LCV115B & & D open. ~ ji) ?~Lo-l Q1E21LCV46O will not close on PZR low level.
- Q1E21LCV460
- Annunciator KG4, TURB TV closed alert, will be in alarm and bistable TSLB2, 14-1 will be lit.
- Annunciator KH5, TURB Auto/Stop oil press low, will be in alarm and bistable TSLB2, 13-1 will be lit.
- 1. IF 1A 120 VAC VITAL INSTRUMENT PANEL is de-energized, THEN immediately perform the following:
A. If FNP- I -EEP-0, REACTOR IF a reactor trip occurs, THEN refer to FNP-1-EEP-0, TRIP OR SAFETY INJECTION. B. Attempt to restore power from the bypass source by performing the following: IF the "BYPASS
- 1. If BYPASS SOURCE AVAILABLE A VAILABLE" lamp is illuminated 1A INVERTER MANUAL on the inverter, THEN transfer lA BYPASS SOURCE TO LOAD" BYPASS SWITCH to the "BYPASS LOAD position.
IF the "BYPASS
- 2. If BYPASS SOURCE AVAILABLE AVAILABLE" lamp is NOT illuminated on the inverter, THEN perform the following:
- Verify IA 1A MCC Energized
- Verify Closed Q Q1R17BKRFAF5L 1R17BKRF AF5L Supply to 11G 208V/120V G 208V 1120V REGULATED AC DISTRIBUTION PANEL
- Verify closed Breaker #8 IN IG1 G REGULATED AC DISTRIBUTION PANEL
- Transfer lA 1A INVERTER MANUAL BYPASS SWITCH to the BYPASS SOURCE TO LOAD position.
of 3 Page 2 of3 Version 26.0
07/01109 15:26:52 07/01/09 15:26:52 FNP-0-ARP-2.2 LOCATION WD WD11 OPERATOR ACTION cont'd OPERATOR contd
- 2. Notify appropriate personnel to determine the cause and correct.
NOTE: Per Table 3 of FNP-O-ACP-52.l, Guidelines for Scheduling of On-Line Maintenance, A, B, C, D B or F Inverters on bypass source are considered to be unavailable due to being status Al for the Maintenance Rule. This unavailability should be logged for tracking purposes.
- 2. Refer to Technical Specification 3.8.9 and 3.8.10.
- 3. IF 1A 120 VAC VITAL INSTRUMENT jj lA iNSTRUMENT PANEL was de-energized, THEN perform the following when it is re-energized:
A. Verify proper operation of Pressurizer level control and heaters. B. Reset hi flux positive rate trip signal on NI -41 and verify proper NI-41 operation ofNI-41. C. Verify the following CVCS components are correctly aligned for current plant conditions:
- Q1E21LCVI15A QIE21LCVl15A
- Q1E21LCV115C&E QIE21LCV115C & E
- Q1E21LCV115B&D QIE21LCVl15B & D
- Q1E21TCV143 QIE21TCV143 D. Verify the following Reactor Makeup Control System components are correctly aligned for current plant conditions:
- Q1E21FCV113A QIE21FCVI13A
- Q1E21FCV114B QIE21FCV114B E. Stop unnecessary Reactor Makeup Water and BAT pumps.
F. Verify all other MCB controls and indications have returned to normal.
- 4. Verify control systems outside the control room have returned to normal.
- 5. WHEN the cause of the fault has been determined AND corrected, THEN return 1A lA Inverter to service.
0009705} {CMT {CMT 0009705) {CMT 0005094) 0005094} applies to entire annunciator D-I77024; U-279610; PCN
References:
D-177024; PCN B87-1-2899; B87-1-2899; D-177218, Sh. Sh. 2; 2; D-177214; {CMT {CMT 0009705} {CMT 0005094} {CMT 0005094} Page Page 33 of of33 Version 26.0 Version 26.0
- 40. 00 1/MOD/RO/C/A 2.9/3.4!N!N13/CYRlY 059A2.04 OOl/MOD/ROICIA 2.9/3 .4ININ/3/CVR/Y Unit 11 was at 100%, and the following conditions occurred:
- The reactor was tripped on simultaneous loss of BOTH SGFPs.
- All AFW was subsequently lost.
- RCS Bleed and Feed is in progress in accordance with FRP-H.1, Response To Loss Of Secondary Heat Sink.
- Core Exit Thermocouples have reached 575°F and are falling.
- IA SGFP has been started.
1A
- SG Wide Range Levels are:
- 1A=
- 1A=8%
8%
- 1B=
- IB=8%
8%
- 1C=
- IC=10%
10% Which ONE of the following describes:
- 1) the Feedwater flow rate required and
- 2) the Main Feed System flowpath required for feeding the SGs lAW FRP-H.1?
A. 1) Feed ALL SGs at a minimum total flow of 395 gpm.
- 2) Use the Main Feedwater Regulating Valves.
B. 1) Feed ALL SGs at a minimum total flow of 395 gpm.
- 2) Use the Main Feedwater Regulating Bypass Valves.
20-1 00 gpm. C. 1) Feed ONE SG at a time at a flow limited to between 20-100
- 2) Use the Main Feedwater Regulating Valves.
D~ D 20-1 00 gpm.
- 1) Feed ONE SG at a time at a flow limited to between 20-100
- 2) Use the Main Feedwater Regulating Bypass Valves.
Page: 105 of200 of 200 12/14/2009 12/1412009
A - Incorrect. First part incorrect but plausible, since this would be correct if CETCs were less than 550°F, OR if CETCs were their given value and rising instead of falling. Second part is plausible, since the MFRVs are the valves used most often with the SGFPs operating, and are used exclusively on a shutdown until AFW is supplying the SGs. Also, this would be correct per FRP-H.1 Step 7.21 RNO if any of the MFRB valves would not open. B - Incorrect. First part incorrect (see A). Second part correct (see D). B - C - Incorrect. The first part is correct (see D). Second part is incorrect (see A). C - D Correct. Note prior to FRP-H.1, step 5 states: "IF 0- - IF it is necessary to feed a hot, temperature > 550°F AND SG wide range level dry SG(s) [RCS hot leg temperature>
<12%{31%}], THEN it (they) should be fed one at a time at a flow rate of 20 gpm to 100 gpm until RCS hot leg temperature falls to less than 550°F. IF bleed and feed is imminent OR bleed and feed is in progress and RCS temperatures are rising, THEN there is no limit on the feed flow rate." rate. FRP-H.1 Control feedwater regulating bypass valves to supply main STEP 7.21 states: "Control feedwater to intact SGs."SGs.
FRP-H.1, Revision 26 Previous NRC exam history if any: 059A2.04 059 Main Feedwater (MFW) System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5/43.5/45.3/45.l3) 41.5 /43.5 / 45.3 / 45.13) A2.04 Feeding a dry S/Q S/G ................................................ 2.9* 34* 9* 3.4* Match justification: The only affects on the MFW system for feeding a hot and dry SG are in a loss of heat sink where the MFW system is required to feed the SG (FRP-H.1). This question requires knowledge of the flow path in the MFW system (impacts of feeding a dry S/G SIG on the MFW system), and the required valve to use and flowrates required in the given condition per the procedure. Objective:
- 6. EVALUATE plant conditions and DETERMINE DETER1VIINE if any system components need to be FRP-H. 1, Response to Loss of Secondary Heat Sink; operated while performing (1) FRP-H.1, S ink; (2)
FRP-H.2, Response to SO SG Overpressure; (3) FRP-H.3, Response to SO SG High Level; (4) FRP-H.4, Response to Loss of Normal Steam Release Capabilities; (5) FRP-H.5, Response to SOSG Low Level. (OPS-52533F06) Page: 106 of200 of 200 12/14/2009 12114/2009
8/8/2007 08:27 8/8/200708:27 Ii]NJ iT 1. FNP-1-FRP-H.1 FNP-l-FRP-H.l RESPONSE TO LOSS OF SECONDARY HEAT SINK Revision 26 Step Action/Expected Response Response NOT Obtained n I I I 4 Monitor CST level. 4.1 [CAl [CA] Check CST level greater 4.1 Align AFW pumps suction to SW than 5.3 ft. using FNP-1-S0P-22.0, FNP-l-SOP-22.O, AUXILIARY FEEDWATER SYSTEM. CST LVL [] LI 4132A [] LI 4132B 4l32B 4.2 Align makeup to the CST from water treatment plant OR demin water system using FNP-1-S0P-5.0, FNP- 1-SOP- 5 .0, DEMINERALIZED MAKEUP WATER SYSTEM, as necessary. NOTE:
- IF some form of secondary feed flow becomes imminent and normal charging is in service, THEN raising charging flow will reduce the potential subsequent loss of pressurizer level due to cooldown shrinkage shrinkage..
C4P
- IF it is necessary to feed a hot, dry SC(s) [RCS hot leg temperature> 550° F ANQ temperature > 550°F AND SG SC wide range level < l2%{3l%}] , THEN it (they) should be fed one at a time at a flow rate of 20 gpm to 10 gpm unti ot eg tempera essta t an . .
bleed and feed is imminent OR bleed and feed is in progre~~R~ inpsERCS temper a tur es ar temperatures e ~1'1 ThEN are(ri3iT TE}e is no limit on the feed tro;'
- !~E;;;:N;;;-"t""h-e-r-e--'i""s-n-o-;l:-:i""'m""'i""t-o-n""""t"";"h-e-""-;f::-eed row
-rate. ~
" H-+-6
,9-4-i2 ;. t-J) 55 [CAl
[CA] Try to establish AFW flow to at least one SG. )V\.Co-c4 5.1
- 5. 1 Verify blowdown from all SGs SOs --
[+4fj:~. -iX ) ISOLATED. lA(lB,lC) 1A(lB,lC) SGBD SCBD
\;7(\~J ISO
(()) QIG24HV7614A 01G24HV7614A closed ((1] QIG24HV7614B Q1C24HV7614B closed [] QIG24HV7614C Q1G24HV7614C closed Step 55 continued on next page. ___Page Completed Page 44 of 49
TT T1\. T TTrrl TT I 8/8/2007 08:27 8/8/200708:27 U iI,ii.1 Ul~l.l FNP-l-FRP-H.l FNP-l-FRP-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Revision 26 Step Action/Expected Response Response NOT Obtained n I I I 7.20 Adjust master speed controller to raise feedwater discharge header pressure to 50 psi greater than steam header pressure. FW HDR PRESS 1 ((1] PIP1 508 STM S TM y-.t ffi J ItA1 f"lt"~ 17 cY? ;}- o'- tf--{-':~ 7
~p If-y,~' vJ~l V I~;' ,,!L HDR (
PRESS 0 - ttrf. ((1] PIP1 464A t/- ,Y'- 0~ vJ 7.21 Control feedwater regulating 7.21 Locally unlock and control \ c£ipass)valves to supply main maii main feedwater regulating ~~ feedwater fter to intact SGs. SOs. val ves with handwheel valves s . (127 handwheels. ft. ft, AUX BLDG main steam valve room) Intact SG lA 1A lB lC 1C lA(1B.lC) SG 1A(1B,1C) SC FW BYP FLOW Intact SG SC lA 1A lB lC 1C FK (()) 479 [] l] 489 (()) 499 adjusted adjusted adjusted lA(1B .1C) SG 1A(1B,1C) SC FW FLOW QlC22FCV Q1C22FCV [] 478 [] [1 488 [] [1 498 Key Z-12l Z-121 Z-120 Z-12O Z-119 7.22 WHEN P-12 light lit. THEN perform the following. 7.22.1 Block low steam line pressure SI. STM LINE PRESS SI BLOCK - RESET [1 AA TRN to BLOCK [] [1 BB TRN to BLOCK [] 7.22.2 indication. Verify blocked indication. BYP && PERMISSIVE STM LINE ISOL. SAFETY INJ. [1[] TRAIN AA BLOCKED BLOCKED light lit [1[] TRAIN BB BLOCKED light lit _Page Page Completed Page 15 15 of 49
- 1. FRP-H-52533F03 019IHLTIIC/A
- 1. O19TIICIA 3.7/4.3/w/E05EA2.21111 3.7/4.3/W/EO5EA2.2/III 059A2.04 Given the following:
- The reactor was tripped on simultaneous loss of both Steam Generator Feed Pumps.
- All AFW was subsequently lost.
- RCS Bleed and Feed is in progress in accordance with FRP-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.
- Core Exit Thermocouples have reached 595 of °F and are still rising.
- SGFP "A" A has been started.
- SG Wide Range Levels are:
- A= A=8% 8%
- B= B=8%8%
- C=C=14%14%
Which ONE of the following describes the method and rate of establishing feedwater flow at this time? A. Feed rate to A and B SGs is limited to between 20-100 20-1 00 gpm. No limit on the feed rate to C SG. Use the respective Feedwater Control Bypass Valves. B There is no limit on the feed rate to any of the SGs. Use the respective Feedwater B:t Control Bypass Valve. C. Feed rate to A and B B SGs is limited to between 20-100 20-1 00 gpm. No limit on the feed rate to C SG. Use the Feedwater Control Valves. D. There is no limit on the feed rate to any of the SGs. Use the respective Feedwater Control Valves. See note in FRP-H.1 regarding feeding of hot dry SGs. 2008 Harris NRC Exam Ability to determine and interpret the following as they apply to the (Loss of Secondary Heat Sink) Adherence to appropriate procedures and operation within the limitations in facility's license and amendments. the facilitys Technical
Reference:
FRP-H.1, Attachment 1 I Page: 11 10/26/2009 10126/2009
- 41. OOI/NEWIRO/C/A 3.9/3.8/N/N12/CVRIY 059G2.1.19 OO1/NEW/RO/C/A 3.9/3.8/N/N/2/CVRIY 100%, and the following conditions exist:
Unit I1 is at 100%,
** A Computer Alarm for 6A FW HTR Dump valve position is in) in.I
** On the IPC, the following indications are observed: /
- 6A FW Heater level indicates 12" 12 and stable.
- V910A, 6A FW Heater level Dump valve, symbol indicates solid red on the V91OA, system mimic.
- V909A, 6A FW Heater level Drain valve, symbol indicates solid green on the system mimic.
Which one of the following explains the given indications, and states the isolation signal which will isolate the 5A FW HTR extraction steam valve, V502A, 5A FW HTR ES ISO?
- 1) 6A FW HTR has a (1) and
- 2) 5A FW HTR Extraction Steam will isolate on high (2)
(1 (1)) (2) A. tube leak Level B. tube leak Pressure C'!" C failed open dump valve controller Level O. D. failed open dump valve controller Pressure A - Incorrect. The first part is incorrect, since the FW HTR is actually empty. Below 14"
- 14 tubes are uncovered and the htr is actually empty with steam blowing by to the HDT. Plausible, since indication is at 12" HOT. 12 and the level indication is steady. A tube leak would cause this indication in other heaters (except for 6A & & B) except for the drain valve being closed. If a tube leak were causing this indication, the drain would be open also. The second part is correct. Plausible even when combined with the incorrect first part, since it is correct for a HTR malfunction which causes steam flow through the 6A FW Htr to the HOT, HDT, but not for high liquid flow to the HOT. HDT. Often with a transient in the FW HTR strings, other heaters are affected. Confusion may exist as exactly how 6A FW HTR will affect the HOT HDT and 5A FW HTR to cause the extraction valve to close.
B - Incorrect. The first part is incorrect (see A). The second part is incorrect. B - Plausible, since the number 6 heater can cause a high level in the HOT HDT which would cause a high pressure in the 5A FW heater, however, there is no high pressure isolation in the 5A HTR, only a high level. Also, the 5A heater does not normally have any level, and the incorrect assumption could be made that it does not have a isolation The 6A FW heater has a dIp high level isolation. . signal, and confusion d/p isolation Signal, could exist as to the 5A isolation signals. of 200 Page: 107 of200 12/14/2009 12114/2009
C - Correct. Cautions in ARP-1.1
- ARP-1.1O,0, for HTR HHII LEVEL and in SOP-20 state that if the 6A HTR level drops below 14" 14 (18-19" (18-19 is normal level controlled by the drain valve),
then the tubes will be uncovered and the tank will empty. Even Even when empty, the
- 12. The dump being open below the normal level of 18-19" tank will indicate 12". 18-19 and the drain being closed is indication that the dump valve controller is failed to demand full open, or at least controlling at too low of a level. The same cautions state that with the 6A FW HTR empty, steam will blow by to the HDT, pressurize the HDT AND the 5A FW HTR, and extraction steam will isolate on high 5A FW HTR level.
D - Incorrect. The first part is correct (see C). The second part is incorrect (see 8). B). Plausible, since it may be understood that below 14" 14 the heater empties and blows steam to the HDT instead of subcooled liquid, but confusion may still exist as to the relationship between the HDT pressure going up, causing the 5A FW HTR pressure to go up, and the 5A FW HTR extraction steam isolating on high level. Also, the 5A heater does not normally have a level, and the incorrect assumption could be made that it does not have a high level isolation either, but could have a pressure isolation. ARP-1.10, KC4, FW HTR ORDRN OR DRN TK LVL HI, Version 64.0 AUTOMATIC ACTION If level is not stabilized, extraction steam to the HP and LP heaters will automatically isolate. See Table 11 for information next page. CAUTION: DO NOT let #6 FW HTR level trend below 14". 14. If the tubes are uncovered level will indicate 12",12, but steam is actually blowing by and pressurizing the HDT.HUT. This results in level increase in the #5 FW HTR due to its [IN]ability jIN]ability to drain to the HDT. On high level, the #5 FW HTR extraction steam will close. (AI (Al 2008202332) FNP-1-SOP-20.0, FEED WATER HEATER EXTRACTION, VENT AND DRAIN SYSTEM, FNP-1-S0P-20.0, Version 53.0 3.7 If the tubes in #6 FW HTR are uncovered level will indicate 12", 12, but steam is actually blowing by and pressurizing the HDT. This results in a level increase in the #5 FW HTR due to its inability to drain to the HDT. On high level, the #5 FW HTR extraction steam will (AI 2008202330) close. (Al Page: 108 108 of 200 of200 12/1412009 12/14/2009
Previous NRC exam history if any: 059G2.1 .19 059G2.1.19 059 Main Feedwater System 2.1.19 Ability to use plant computers to evaluate system or component status. 41.10 I 45.12) RO 3.9 SRO 3.8 (CFR: 41.10/45.12) Match justification: This question requires the evaluation of plant computer points for some parameters for a FW system component (6A FW Heater) to determine component status of two FW system components: 6A FW HTR actual level and 5A FW HTR extraction valve resulting status. Objective:
- 6. normal / abnormal plant or DEFINE AND EVALUATE the operational implications of normal/abnormal equipment conditions associated with the safe operation of the Main and Reheat Steam System components and equipment, to include the following (OPS-40201A07):
** Normal control methods
- Abnormal and Emergency Control Methods
- Automatic actuation including setpoint (example SI, Phase A, Phase B, MSLIAS, LOSP, SG level)
** Protective isolations such as high flow, low pressure, low level including setpoint
- Protective interlocks
- Actions needed to mitigate the consequence of the abnormality Page: 109 109 of 200 12/14/2009 12/1412009
07/02/09 06:30:47 07/02/0906:30:47 FNP-I-ARP-I.IO FNP- 1 -ARP- 1.10 LOCATION KC4 SETPOINT: Heater Drain Tanks: 4.25 inches above center
- 1. Heater
- 1. C4 I tank.
of tank. FW HTR FWHTR
- 2. L.P. Heaters: 3.0 inches above normal liquid ORDRNTK level.
- 3. H.P. Heaters: 3.0 inches above normal liquid LVL level. HI APPROXIMATE MCB LEVEL HEATER HEATER: 1A&B IA&B 13 l3" HTRDRNTK HTRDRNTK: 5175 A 51.75" 2A&B 9 9" 51.75 B 51.75" 3A&B 13 l3" 4A&B 8 8"
6A&B 21" 21 ORIGIN: The following Level Switches: a) IA HDT N1N26L1502 lAHDTNlN26LI502 e) 4AHeaterNlN2lLI538 4AHeaterNIN21LI538 I) 2B Heater NIN21Ll532 N1N21L1532 b) b) 1A Heater lA Heater NIN21LI526 N1N21L1526 f) 6A Heater NIN21LI542 N1N21L1542 j) j) 3B Heater NIN21Ll536 N1N21L1536 c) Heater NIN21LI530 2A Heater N1N21L1530 g) lB HDT NIN21LI504 1BHDTNIN21LI5O4 k) 44B B Heater N 1N21 Ll540 N1N21L1540 d) Heater NIN21LI534 3A Heater N1N21L1534 h) lB 18 Heater NIN21LI528 N1N21L1528 1) 6B Heater N1N21LI544 N1N21L1544 PROBABLE CAUSE I.
- 1. Heater tube leak or rupture.
- 2. Malfunction of a Level Controller on a Low Pressure or High Pressure Heater or Heater Drain Tank.
AUTOMATIC ACTION If level is not stabilized, extraction steam to the HP and LP heaters will automatically isolate. See Table 11 for information next page. CAUTION: DO NOT let #6 FWHTRleveltrendbelow CAUTION:. FW IITR level trend below 14". 14. If the tubes are u uncovered
.:::;;;;;....-.-~~
level will indicate 12, but steam is actually blowing by and ssurizin he
.diju..athjiDTj&jgeve,e WHTR extraction steam will close.
?
/ OPERATOR OPERATOR ACTION
- 1. Determine which heater or drain tank level is high.
- 2. Verify that the dump to condenser valve of the affected heater or drain
/ , tank is open. 46 4. IF a heater tube leak OR rupture is indicated, THEN isolate the affected heater. t 5. IF a feedwater heater malfunction is indicated, THEN go to FNP-l-AOP-13.0, FNP-I-AOP-l3.0, CONDENSATE AND FEEDWATER MALFUNCTION
- 6. Refer to FNP-l-SOP-21.0 FNP-I-SOP-2I.0 for limitations on Turbine operation with one or more Feedwater Heaters isolated.
- 6. Monitor Feedwater Heater and Drain Tank Levels to ensure that they are returning to normal.
Page 11 of2 Version 62.0
07/02/09 06:3 0:47 07/02/0906:30:47 FNP-I-ARP-1.l0 FNP- 1 -ARP- 1.10 7.
- 7. Notify appropriate Notify appropriate personnel personnel to to determine determine and and correct correct the the cause cause of of the the alarm.
alarm. Page Page 22 of of22 Version 62.0 Version 62.0
07/02/09 06:30:47 07/02/0906:30:47 FNP-I-ARP-l.10 FNP-1-ARP-1.i0 TABLE 11 TABLE Setpoint for Setpoint Heate Extraction Steam Extraction Stm Stm Level Switch Supply Breaker rr MOV MOV Isolation 1A lA 18 above HLL 18" HLL N1N39LS5O2A NIN39LS502A N1N35V517A-N NIN35V517A-N NIRI7BKRHBBB3 N1R17BKRHBBB3
-III lB IB 18 above HLL 18" 18 above HLL 18" 18 above HLL 18"
';' ,/ .
N1N39LS5O2B NIN39LS502B
.eN'//; >; ,;;
N1N39LS5O2A NIN39LS502A N 1N39LS502B NIN39LS502B N1N35V518A-N NIN35V518A-N
;/ ~".;;;
N1N35V517B-N NIN35V517B-N
.. *;*7 N 1N35V5 1 8B-N NIN35V518B-N 7
NIR17BKRHBBB4 1<;;: NIR17BKRHBBB3 N1R17BKRHBBB3 NIR17BKRHBBB4 Ni Ri 7BKRHBBB4 2A 18 above HLL 18" N1N39LS5O5A NIN39LS505A N1N35V519A-N NIN35V519A-N NIR17BKRHBBC4 N1R17BKRHBBC4 2B 18 above HLL 18" N1N39LS5O5B NIN39LS505B N1N35V519B-N NIN35V519B-N NIR17BKRHBBC5 N1R17BKRHBBC5 I 18 y,." above NIN39LS508A 3A LoIe N1N35V5O6A-N NIN35V506A-N NIR17BKRHAAA4 NIRI7BKRHAAA4 HLL I 4aove 18 y,." above NIN39LS508B 3B N1N35V506B-N NIN35V506B-N NIR17BKRHAAA5 HLL 4A 14 above HLL 14" N1N39LS51OA NIN39LS510A N1N35V507A-N NIN35V507A-N NIR17BKRHBBA6 N1R17BKRHBBA6 4B 14 above HLL 14" N1N39LS5 NIN39LS510B 1011 N1N35V507B-N NIN35V507B-N NIR17BKRHBBB2 N1R17BKRHBBB2
~5A 5A 18 above HLL 18" N1N39LS512A NIN39LS512A N1N35V502A-N NIN35V502A-N NIR17BKRHAAB2 N1RI7BKRHAAB2
.f1C-5B 5B 18 above HLL 18" N1N39LS512B NIN39LS512B N1N35V502B-N NIN35V502B-N NIR17BRKHAAB3 N1R17BRKHAAB3 I ~ 18 above HLL 18" N1N39LS515A NIN39LS515A .~ A f 7 OR OR 6A N1N35V503A-N NIN35V503A-N NIR17BRKHAAA6 N1R17BRKHAAA6 ~ @ 0-2/+0 closes @ N1N35PDS547A NIN35PDS547A OR ORB B opens@ 66 +/- 22
~ AP ~P 7 .---- 18" above HLL 18 N1N39LS515B NIN39LS515B 7 OR OR
( 6B N1N35V503B-N NIN35V503B-N NIR17BKRHAAB5 ( closes @ @ 0-2/+0 0-2/+0 N1N35PDS547A NIN35PDS547A OR B B
\\ opens@6+/-2 opens@6+/-2
\ L\P A ~P
~s:
References:
A-177100/459; D-172579/1&2; A-170750/88&89; D-172570 A-I77100/459; A-170750/88&89; D-I7257911&2; B-175968; D-I72577; D-172578/1&2; D-I72570 thru D-172577; B-170058/97&98; A-170256; B-175968; B-170058/97&98; A-170256; A-170257 D-I7257811 &2: A-170257 C 2 d / Page 33 of2 Page of2 Version 62.0 Version 62.0
- 42. 061G2.2.37 061 G2.2.37 OOllFNP 00 1IFNP BANKIROICIA BANK/RO/C/A 3.6/4.61N1N13/CVRIY 3 .614.61N/NI3ICVRIY 1000 the following plant At 1000 plant conditions conditions exist exist on on Unit Unit 1:
1:
- A TECH SPEC required required shutdown was in in progress progress due to BOTH to BOTH IA and 1lB 1A SW pumps B SW pumps inoperable inoperable and unavailable unavailable (not (not running).
running).
** 11C SW pump C SW pump is is aligned to B B Train.
At 1005 the following events occur:
- A seismic event caused a loss of BOTH SGFPs, a leak in the A Train SW header and a tear in the CST at the bottom of the tank.
- CST level is at 5 ft. and decreasing.
Which one of the following describes the purpose of the actions directed by SOP-22.0, SOP-22.O, Auxiliary Feedwater System? To establish availability of _ _ _ _ _ __ A. 1A and 11 B MDAFW pumps with SW valve alignments made from the main control room ONLY. B. the TDAFW and the 11 B MDAFW pumps with SW valve alignments made from the main control room ONLY. C. ALL AFW pumps with SW valve alignments made from in the plant AND from the main control room. D the TDAFW pump with SW valve alignments made from in the plant AND from the Dy main control room, and the 1 1B B MDAFW pump with SW valve alignments made from the main control room ONLY. Page: Page: 110 110 of of 200 200 12/1412009 12/14/2009
A - Incorrect. A - Correct for Incorrect. Correct for BB MDAFW MDAFW pump only, and pump only, and incorrect incorrect for for A MDAFW pump. A MDAFW pump. BB has aa source has source from from B B train train SW SW with MCRMCR valve valve alignments alignments only. only. Even Even though though B B MDAFW pump MDAFW pump hashas aa suction suction source from BB train source from train SW, SW, A A MDAFW MDAFW pumppump hashas no no procedurally allowed suction procedurally allowed suction source source from from A A train train SWSW per per SOP-22. SOP-22. However, However, the the system would system would allow allow cross cross connecting connecting trains trains to to supply supply both both A and/or BB MDAFW A and/or MDAFW from from BB train train SW. Plausible, since SW. Plausible, since in in this this emergency, emergency, itit could could be incorrectly assumed be incorrectly assumed thatthat maximum flexibility would be written into into the procedures procedures to allow this this option to cool SGs and Core. the SGs Core. BB - Incorrect.
- Incorrect. Correct for B MDAFW Pump. Pump. TDAFW has a source of suction from A train only aligned to enable supplying with only MCR MCR MOVMOV operations. Plausible, Plausible, since the TDAFW suction can be supplied from B B train AFW with manual valve alignments outside of the Control Room. Room.
C - Incorrect. Correct for B MDAFW and the TDAFW Pumps. Incorrect for A MDAFW pump. Plausible, since the system is versatile enough to allow cross connecting trains to supply both A and/or B MDAFW from B train SW. In this emergency, it could be incorrectly assumed that maximum flexibility would be written into the procedures to allow this option to cool the SGs and Core. oD - Correct.
- Procedurally, B MDAFW (with MCR valve alignments only) and TDAFW pump (with in-plant and MCR valve alignments required) are both able to use B train SW as an auxiliary suction source.
SOP-22, AUXILIARY FEEDWATER SYSTEM, Version 64.0 Page: 111 Page: III of of 200 200 12/14/2009 12/14/2009
Previous NRC exam history if any: 061 G2.2.37 061G2.2.37 061 Auxiliary /I Emergency Feedwater System 2.2.37 Ability to determine operability and/or availability of safety related equipment. 41.7 / 43.5 / 45.12) RO 3.6 SRO 4.6 (CFR: 41.7/43.5/45.12) Match justification: Determining availability is more on the RO level than determining operability (other than determinations of low discriminatory value such as pump trips on overcurrent, no power to start pumps, etc.). This question requires knowledge of which of the AFW pumps are available from only one train of their alternate suction source of Service Water when their primary suction source, the CST, is not available. It also requires knowledge of how the pumps will become available from their alternate suction source procedurally and by location of the valve manipulations. Objective:
- 7. normal / abnormal plant or DEFINE AND EVALUATE the operational implications of normal/abnormal equipment conditions associated with the safe operation of AFW System components and equipment to include the following (OPS-40201D07):
- Normal Control Methods
- Abnormal and Emergency Control Methods
- Automatic actuation including setpoints (examples - SI, Phase A, Phase B, MSLIAS, LOSP or SG level)
- Actions needed to mitigate the consequence of the abnormality Page: 112 112 of 209 200 12/14/2009
07/02/09 06:34:16 07/02/09 06:34: 16 FNP-I-S0P-22.0 FNP-1-SOP-22.0 4.7 4.7 Aligning Service Aligning Service Water Water to to the the AFW AFW System. System. CAUTION: CAUTION: Service water Service water does does not not meet meet secondary secondary makeup makeup specifications specifications and and should only should be used oniy be used when when required required byby emergency emergency conditions. conditions. NOTES: ** Refer to Tech. Specs. section 3.7.6 Condition A.
- 2 BOP keys from the Shift Support Support Supervisor's Supervisors office will be required in the following steps.
steps. 4.7.1 Notify Shift Chemist that SW will be added to the SO's. SGs. NOTE: FNP-1-SOP-24.0, SERVICE WATER SYSTEM, requires FNP-I-SOP-24.0, starting/stopping SW pumps as necessary to maintain pressure in each header greater than 70 psig but less than 130 psig as indicated on PI-3001AA and PI-3001BA or by adjusting CCW HX DISCH FCV HIC 3009A(B, 3009 A(B, C). 4.7.2 Verify service water is in operation per FNP-I-S0P-24.0, FNP-1-SOP-24.0, SERVICE WATER SYSTEM maintaining proper SW pressure. 4.7.3 Open MDAFWP SW SUPP: (BOP key operated switches)
- QIN23MOV3209A Q1N23M0V3209A It A- {-rV\
?-
- Q1N23M0V3209B QIN23MOV3209B irv 4.7.4 Open: (BOP)
- MDAFWP SW SUPP QIN23MOV32IOA QIN23MOV3210A
- MDAFWP SW SW SUPP Q1N23MOV321OB QIN23MOV3210B
- TDAFWP SW SW SUPP Q1N23M0V3216.
QIN23MOV3216. Version 61.0 Version 61.0
07/02/09 06:34:16 07/02/09 06:34: 16 FNP-I-S0P-22.0 FNP-1-SOP-22.0 4.7.4.1 4.7.4.1 IF necessary IF necessary to align TDAFWP to align TDAFWP suction suction from from B B Train Train service water, service water, THEN THEN perform perform the the following. following. NOTE: following two steps During the following service water trains will be temporarily steps service cross-connected. 4.7.4.1 .1 4.7.4.1.1 Unlock and open~ervice Unlock openjin service water to TDAFWP suction: Q1N23VO15D (in IB
- QIN23V015D lB MDAFWP Room) *-
L QIN23VO15C (above MDAFWP Room) ~
- QIN23V015C f 4.7.4.1.2 Unlock and close{A c1ose TralDlservice service water to TDAFWP suction:
Q1N23VO15B (above MDAFWP Rooms)
- Q1N23V015B +-
Q1N23VO15A (in 1A MDAFWP Room)~
- Q1N23V015A Room)4 4.7.5 IF required, THEN place AFW system in operation per sectl~*on f section 4.1 or a..!)
. rna IA. (.A 4.3 of 4.3 of thIS this SOP.
SOP. ~X t\ -J(! ,//vtt 1 / \)e ;; 4.8 Condensate Storage Tank Feed and Bleed. 4.8.1 Verify CST LVL LI 4005B 2: 12.5 ft and CST LVL LI 4l32B 4132B is reading maximum indication. 4.8.2 Verify with Chemistry that water is within specifications to be drained to the yard drains. 4.8.3 Remove blind flange and throttle open CST drain valve Q1P11V5O8. Q1P11V508. 4.8.4 Verify CST LVL L VL LI 4005B E 2: 12.5 ft and CST LVLL VL LI 4132B is reading its maximum indication at least once once every 3030 minutes while while feed and bleed bleed operation is in progress. 4.8.5 IF either level indication indication criteria is NOT met, THEN close CST CST drain valve (Q1P11V5O8) (Q1P11V508) and restore CST level per FNP-1-SOP-21.0, FNP-I-S0P-21.0, CONDENSATE AND AND FEEDWATER SYSTEM. SYSTEM. 4.8.6 4.8.6 WHEN WHEN feed and and bleed bleed operation operation is is complete, complete, THEN THEN close close CST CST drain drain valve valve (Q1P1 1V508) and (QIP11V508) and replace replace blind blind flange. 4.8.7 4.8.7 Verify Verify CST CST LVLLVL LI 4005B 2: 12.5 LI 4005B 12.5 FT and CST FT and CST LVL LVL LILI 4132B 4132B isis reading its maximum reading its maximum indication. indication. Version 61.0 Version 61.0
Date: ior 2OO9 Date: 10r\:2009 Time: Time: ~")51 :36 AM 51:36AM 11 I 22 I 33 :4 4 I 55 I 66 I 77 4- ;fL-P /0 \O PUMP AUTO-SEE NOTE 8
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Title:
Title: C:\Reference C:\Reference Disk\Exam Disk\Exam Reference Reference Disk\Drawings\D175007 Disk\Drawings\D1 75007-000 -0001.cal 1 .cal
- 43. 062Al.Ol 062A1.O1 OOllNEW/ROICIA OO1INEW/RO/C/A 3.4/3.8ININI4/CVRlVER 3.4/3.8/N/N/4/CVRIVER 55 EDITORIAL Unit 1 I has had a LOSP followed by a SBLOCA, and the following conditions exist:
- The l-2A DG is tagged out.
1-2A
- The lB 1 B DG is tripped.
- The IC 1 C DG is supplying power to the A Train busses.
- The load on the 1 IC C DG is 2.860 MW.
Which one of the following describes whether or not the 2000 amp hour rating on the 1C 1C DG will be exceeded if the 1 lB B PRZR HTR GROUP BACKUP is energized? lE Hthethe PRZR HTR GROUP BACKUP is energized, THEN the 1 IC C DG 2000 hour load rating (1) (I) be exceeded, and energizing the 1 I B PRZR HTR GROUP BACKUP (2) allowed lAW EEP-1, EEP-l, Loss of Reactor or Secondary Coolant. (1)) (1 (2) A. will is B~ B will is NOT C. will NOT is D. will NOT is NOT Note: in this alignment the 1 ICC DG has been manually aligned to the 1 1F F bus. A - Incorrect. The first part is correct (see B). The Second part is incorrect (see B). Plausible since the procedure does state that in an emergency, the design of the electrical system has determined that a slight overload may exist after a LOSP and a LOCA. This is acceptable as long as the 2000 hour rating is not exceeded. Also, the Basis of TS 3.8.3 allows the 2000 hour limit to be exceeded for up to 300 hour per year but EEP-1 forbits exceeding the 2000 hour limit. Confusion may exist as to which one or if both the continuous and/or the 2000 ratings may be exceeded for a period of time. Also, the continuous load limit and the 2000 hour rating values may not be remembered properly. However, the procedure states that MANUAL loading above EITHER the continuous or 2000 rating is not allowed. Confusion could exist as to what the 2000 hour load allows, i.e. it does not allow overloading above the limit for any period of time as as . the continuous load limit does
~;?
does..
- continuous Correct The continuous'16a'd B - Correct. "~~
~MW,
..,,-<: load limit is 2MW
-~~ . and the 2000 hour load limit is
~
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.1-10,,9 MW. 2.860~~Q3=3.16MW >
MW 2 860+0303 16MW> continuous load limit EEP-1 APP 44, Step 1I limit. caution states Do NOT manually load diesel generators above 2000 hr. load limit. Per EEP-1, Att. 4: "...continuous
... continuous load rating limit (i.e. 2.85 MW for small DGs, 4.075 MW for large DGs). Under these circumstances, diesel generator loading may be raised not to exceed the 2000 hour load rating limit (i.e. 3.1 MW for small DGs, 4.353 MW for large DGs DGs...).
... ).
Page: 113 of 200 12/14/2009
Co - Incorrect.
- Incorrect. The The first partpart isis incorrect.
incorrect. Plausible, Plausible, since since the continuous continuous load load limit limit and and 2000 hour the 2000 hour rating rating may may notnot bebe remembered remembered properly, properly, and/orand/or the load load in in MW MW of the the przr heaters przr heaters may may be confused with other smaller loads. The large DGs large DGs havehave a higher load limitlimit ofof 4.075 & & 4.353 MW MW for continuous continuous and and 2000 hr hr rating rating respectively. The second respectively. second part part isis incorrect incorrect (see (see A). D Incorrect. The first part is incorrect (see C). The second part is correct (see 8). 0- - B). Plausible, since the continuous load liming is already exceeded, and manual loading above the continuous load is not allowed, even though automatic loading above the continuous load limit is allowed in an emergency. FSD, Diesel Generator System 3.1.6 Interface Requirements The only time during operation (other than design basis accidents) that the diesel is intentionally loaded above its continuous rating is during Technical Specification surveillance testing when the diesels are loaded to their 2000h ratings (4353 KW for the large diesels and 3100 KW for the small diesels). APPENDIX B APPENDIXB STATIC LOADING OF THE DIESEL GENERATORS B.2.O INTRODUCTION B.2.0 During some design basic events, diesel generator 11C C is loaded above its continuous rating by less than 5%. However, this calculated loading above the continuous rating is acceptable since the diesel loading still meets the criteria contained in Position C.2 of Safety Guide 9 9 (Reference 6.7.028). G.4.3 Potential Diesel Generator Overload The potential exists for DG overload if the LOSP is followed by a LOCA after step 2 of the LOSP sequencer has been energized. In those cases, the DG will be loaded with the Reactor Cavity Cooling Fan (13 1Kw) Kw) and the CRDM fan (84 Kw) in addition to the ESS loads, and the operator may have to remove selected loads if the DG is loaded above its rated capacity. This situation does not constitute a concern given the existing guidance in the plant procedures which provides the operator with guidance for reducing the DG loading if it is above rated capacity. EEP-1 LOSS OF REACTOR OR SECONDARY COOLANT Revision 29 ATTACHMENT 4 VERIFYING 4160 V BUSSES ENERGIZED I1 Verify 4160 V busses energized. CAUTION: IF IF aa DG DG is already operating above its is already its continuous load load rating, THEN additional additional manual manual loads loads should should not not be be added. added. Unanticipated Unanticipated plant plant emergency conditions may may dictate dictate the the need need to load load the the emergency diesel diesel generators above the continuous load load rating limit limit (i.e. (i.e. 2.85 2.85 MWMW for small DGs, DGs, 4.075 4.075 MW MW for large large DGs). DGs). Under these circumstances, diesel Under these diesel generator generator loading loading maymay be be raised notnot to exceed the the 2000 2000 hour hour load load rating limit limit (i.e. (Le. 3.1 3.1 MW MW for small DGs, DGs, 4.353 MW for large 4.353 MW large DGs). DGs). Diesel Diesel loading loading should should be be reduced reduced within within the the diesel diesel generator continuous load load rating limit limit as as soon soon as as plant plant conditions allow. conditions allow. Page: Page: 114 114 of 200 of200 12/1412009 12/14/2009
CAUTION: To CAUTION: To prevent prevent diesel generator overloading, diesel generator overloading, at least 0.3 at least MW of 0.3 MW of diesel diesel generator generator capacity capacity must must bebe available prior available prior to to energizing energizing aa group group ofof pressurizer pressurizer heaters. heaters. 1.7.4 RNO 1.7.4 RNO Energize Energize pressurizer pressurizer heater heater group group 11 BB as as required. required. Previous NRC Previous NRC exam historyhistory ifif any: any: 062A1 .01 062A1.01 062 A.C. 062 A.C. Electrical Electrical Distribution Distribution Al Ability Al Ability to predict and/or to predict and/or monitor monitor changes changes inin parameters (to (to prevent exceeding exceeding design design limits) limits) associated associated with operating the ac distribution system controls including: (CFR: 4l.5 with 41.5 145.5)
/ 45.5)
A1.01 Significance of DIG Al.Ol load limits .................................... 3.4 3.8 D/G load Match justification: In In this question parameters parameters are provided which must be evaluated to predict if the DG will be overloaded if a load is manually started. The size of the load in MWs must be known and the load limit must be know to predict if the load may be started and if it will exceed the design limits. The significance of the load limits must be understood, since the continuous limit may be exceeded in an emergency without expected damage to the DG, but the 2000 hour load limit may not be exceeded for any reason. Objective:
- 2. RELATE AND IDENTIFY the operational characteristics including design features, capacities and protective interlocks for the components associated with the Diesel Generator and Auxiliaries System, to include the following (OPS-40 102C02):
(OPS-40102C02):
- a. PC2 Diesel, including capacity
- b. FM Diesel, including capacity Page: 115 of Page: 115 of 200 200 12/1412009 12/14/2009
08:19 8/8/2007 08: 8/8/2007 19 I._.iJ\TTi .11 UJ~ll FNP1EEP--1 FNP-1-EEP-1 LOSS LOSS OF OF REACTOR REACTOR OROR SECONDARY SECONDARY COOLANT COOLANT Revision 2929 Step Step Response Action/Expected Response Response NOT Obtained Response Obtained n I I I ATTACHMENT ATTACHMENT 4 1.5 1.5 Verify BKR DG02 DGO2 (lG (1G 4160 VV bus 1.5 1.5 IF diesel generator cooling IE tie to 1L 4160 V V bus) -- NOT supplied. N supplied, CLOSED. THEN secure 1BlB diesel generator using ATTACHMENT 5. 5, SECURING A A DIESEL GENERATOR WITH A A SAFETY INJECTION SIGNAL PRESENT. 1.6 l.6 Verify all RCP busses - - ENERGIZED. ((]1 1A 4160 V V bus ((]1 1B 18 4160 VV bus ((]1 1C 4160 V V bus 1.7 l.7 1E 4160 V Check IE V bus - - 1.7 Establish power to 1C 600 V V LC ENERGIZED, ENERGIZED. emergency section loads. 1.7.1 Place handswitch for pressurizer heater group 1B lB in OFF. 1.7.2 Open BKR ECOS-1. ECO81. 1.7.3 Close BKRs EE07-1 EEO7-1 and EClO-l. EC1O-1. CAUTION: To prevent diesel generator overloading. overloading, at least 0.3 MW of diesel generator capacity must be available prior to energizing a group of pressurizer heaters. 2.
- -3 / *bO YL.jw
~ ~
1.7.4 1.7.4 Energize pressurizer heater 1B as required. group lB 1.S 1.8 Check 1D 4160 VV bus - - 1.S 1.8 1.10. Proceed to step 1.10. ENERGIZED. 1.9 1.9 IF 1D IF 1D 4160 4160 VV bus bus energized. THEN THEN return to PROCEDURE PROCEDURE STEPS, step 13. STEPS. step 13. Step 11 continued Step continued on on next next page. page. Page Completed _Page Completed Page 22 of Page of 66
8/8/2007 08:19 8/8/200708:19 1.JN Ii 1. FNP-1-EEP-1 FNP-1-EEP-l LOSS OF REACTOR OR SECONDARY COOLANT Revision 29 Step Action/Expected Response Response NOT Obtained n I I I ATTACHMENT 4 VERIFYING 4160 V V BUSSES ENERGIZED 1 Verify 4160 V V busses energized. 1/4 4--C- ?- 'A- d-i6--T+-~ , Zc rec 412 '?- ~ G..'+;P - 4e-s .f~f::-? (-r Cc CAUTION: IF a DG is already operating above its continuous load rating. rating, THEN additional manual loads should not be added. Unanticipated plant emergency conditions may dictate the need to load the emergency diesel generators above the continuous load rating limit (i.e. DGs, 4.075 MW for large DGs). Under th~e~s~e~ 2.85 MW for small DGs. these ____~__ ~ cir ces enerator 10 lo n ma be raised not to exceed t 0 0 hour load rati i.e. 3.1 M for smal W
~~;=~~~~;;==~~~r=~~~~~~~~~~duced for arge . iesel oa ing s o duced within within the the diesel diesel generator continuous load rating limit as s on as plant conditions allow.
>11/ C,OH\W> 3) 11\
- * ****************** *************** *.:t;* * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *
- A -I-4- f3. I 7 .r~t $> L t>rf~
C 4: i> I) ~-f;- i'-C V s-c- )j \4...C.q--f" ~ NOTE: Plant conditions may dictate establishment of contingency electrical lineups. FNP-1-AOP-5.1. FNP-1-AOP-5.1, CONTINGENCY ELECTRICAL ALIGNMENTS provides gUidance guidance for establishing those lineups. 1.1 Check offsite power - - 1.1 Request Shift Manager AVAILABLE. AVAILABLE, coordinate efforts to restore offsite power. 1.2 DF01 (lA startup Check BKR DFO1 1.2 Verify lF IF 4160 VV bus energized transformer to iF IF 4160 VV bus) by 1-2A or 1C lC diesel
- CLOSED.
- generator.
1.3 DF02 (iF Verify BKR DFO2 (IF 4160 VV bus 1.3 jIF diesel generator cooling tie to 1KlK 4160 VV bus) -- NOT supplied from Unit 2, 2. CLOSED. THEN secure 1-2A l-2A and/or 1C diesel generator using ATTACHMENT 5, 5. SECURING AA DIESEL GENERATOR WITH A A SAFETY INJECTION SIGNAL PRESENT. 1.4 1.4 BKR DG15 (lB startup Check BKR startup 1.4 1.4 Verify 1G 1G 4160 VV bus energized transformer to 1G 1G 4160 VV bus) by lB 1B diesel generator.
- CLOSED.
Step 11 continued continued on next page. ___Page Completed Completed Page 11 of 66
- 44. 063A3.01 063A3.O1 OOI/FNP OO1/FNP BANKIROICIA BANK/RU/C/A 2.7/3.11Y 2.7/3.1/Y 2007/N/4/CVRlVER 2007/N/4/CVRIVER 5 EDITORIAL Unit 11 is at 100% power with the following conditions:
- 1A Battery Charger is on service.
- EM personnel are doing preventative maintenance on the 1A 1A battery.
The following indications and alarms are received:
- The UNIT 11 AUX BLDG DC BUS - A TRN GROUND DET white light comes on momentarily and then goes OFF.
- WC3, 1A IA 125V DC BUS BATT BKR 72-LA05 72-LAO5 TRIPPED
- WC2, 1A IA 125V DC BUS UV OR GND alarms and clears.
Which ONE of the following describes the status of the indications on the EPB for the 1A DC BUS and the 1A and 1I B Inverters? IA DC BUS VOLTAGE reads approximately 1A (1) IA and 1I B INVERTER AMPERES are reading approximately 1A (2) A. (1) 0 DC VOLTS. (2) 25 amps and being powered from the bypass source. B. (1)(1)ODC 0 DC VOLTS. (2) 0 amps and being powered from the normal source. C. (1) 125 DC VOLTS. (2) 0 amps and being powered from the bypass source. D D~ (1) 125 DC VOLTS. (2) 25 amps and being powered from the normal source. Page: 116 116 of 200 of200 12/14/2009 12/1412009
explanation explanation Battery output When the Battery output breaker breaker is is opened, opened, LA-05, LA-05, WC3 will come into into alarm alarm due due to the bb contact contact from breaker breaker LA05. LAO5. WC2 shows either a low low voltage condition or aa ground. In ground. In this case itit would be be aa ground. battery output The battery output breaker breaker has has opened due to a ground on the battery battery and when itit opens WC2 clears. The annunciators provide indication that the breaker opened and the white light provides indication of the ground. For this set of circumstances, the battery is no longer aligned to the bus and the battery charger is carrying the load. The indications will remain normal and the inverters will have normal indications. The inverters will not swap to the bypass source and will still be powered from the BC. A - Incorrect. 0 DC volts on the 1A DC bus indicates the bus is de-energized. The bus still has power from the Batt. chger. The inverters will be powered from the BC or the normal supply and will indicate 25 amps. If it were to swap to the bypass source, it would still have amp readings, but if the manual bypass switch were to be placed in the bypass position, then the amps would be 0 amps. B - Incorrect. 0 is not correct for both. Normal is correct. C - Incorrect. 125 is correct. 0 is not correct and bypass is not correct. D - Correct. 125 is correct and 25 is correct from the normal source. DWNG: D177082 sheet 1 I Page: 117 Page: 117 of of 200 200 12/14/2009 12/1412009
Previous NRC Previous NRC exam exam history history ifif any: any: 2007 2007 FNPFNP NRCNRC exam, exam, this this question question is is the the only only one one in the bank in bank tied tied to this K/A this KIA 063A3.01 063A3.01 063 D.C. Electrical Distribution A3 Ability to monitor automatic operation of the DC electrical system, including: (CFR: 41.7/45.5) 41.7 /45.5) A3.O1 Meters, annunciators, dials, recorders, and indicating lights ............. 2.73.1 A3.01 2.7 3.1 Match justification: It meets the KA in that it tests the ability to determine the proper readings on the EPB for an abnormal condition based on the indications and alarms received (white light and annunciators). The automatic portion of the KA is the breaker opening on an overcurrent condition. Objective:
- 6. DEFINE AND EVALUATE the operational implications of normal / abnormal plant or equipment conditions associated with the safe operation of the DC Distribution System components and equipment, to include the following (OPS-40204E07):
** Normal control methods
** Abnormal and Emergency Control Methods
** Automatic actuation including setpoint
** Protective isolations
- Protective interlocks
- Actions needed to mitigate the consequence of the abnormality Page: 118 of Page: 118 of 200 200 12/14/2009 12/14/2009
Date: 10r\Q009 Date: I OI 2OO9 Time:O2:18PM Time: "\02: 18 PM I --1-------------------------- ,~----------------------~~~ 1 I 22 I 33 I 4 I 55 66 I I TO 600V TO 600V LOAD LOAD CENTER 10 CENTER lB 1-5/c -I/O AA 2- lie !II? 11) (NOTE 11)
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-(NOTE. ') 1) AUTOMATIC .g ~ ~ AUTOMATIC 07 TR,\NSFER SWITCH SWITCH (NOTE (NOTE 10) ID) ~
N TRANSFER SWITCH SWITCH (NOTE 10) (NOTE I0 EE U0 4: lLIL DIESEL GENE.RATOR DIE5E.L GENEIA1DR Ie IC -2.0 (( r:s: DI!,:5f_L -GE.NE.RATOA 1-'2.A CONTP.OL CONTRa PNEL L PANE.L *'* '¢
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r-7> Ii7, PANELS PANELS EXCEPT EXCEPT THOSETHOSE NOTED NOTED ARE ARE ID D D~"'TP.Ie,UTION IL MOLDED MOLDED CASE, CASE, TWO POLE 100 TWO POLE, 00 AMP AMP FRAME FRAME IZSV DC. DC. DISTRIBUTION Pt..NEL 10,000 AMP Ie 0,000 ANIP IC e& 1'1-'5 I2. VOC VDC ], { I,,,,V PANEL 1c.. 1 QIR4ILOOIC-A IC.,QIR4ILOOIC-A
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(NOTE CONI. A - SOA .:lOA, QI PI SI-lF';;SZ.c"O'A-A QIP SL1FSS2C09A-A '"S "'-DE.\o.IOTE"" A-DENOTES E:MER()ENC'I Il-I M", IN N CONtROL. NIAIN EMERCzEF4C/ PDVJER CO>.ITIZOL "'OOM. ROOM. PDINEI2 BOARD 5OAD io) S PA B S. ----'~,..... ,.....~ SPARE SPAIZ.E. 30A '!lOA 5 PA R E 7.7- 1-.t!C-~12: REMOIE P.EMOIE WHITE-a I-4IC-I2 2 CONDUC.TOR<:. CONDUCTORS U'5EO WHITE. L.IG.kT LIG4iT AND USED FOR FOR 2 CONDUCTOR.S AND 'C.Ol-lOUc..TOIl."" IEE DIESEL ~GE.NERATOR DIE.Sf_L GENERATOR I (-ZA. -'Z.I>. 5PARE SPA RE ~. ,.....~ SPI'-RE SPARE U~ED USED FOR FOR REMOTE. R.EMOTE VOLTME.TER. VOLTMETER.
"""II ~ ~12rv'. ~'RE FIRE PROT. FIL.1. UNIT i-21c-""t'. '"
PROT. Fin-UNIT /SA CONTEOL PANoE.L"'" CONT120L PANEL fiRE /12-Yc#f" FIRE PROT. FILT. PROT. FILT. UtllTS UNITS 8, REVISED 1'I>.IP S REVISED IRIP COIL RATINO ,.0 COIL RA.TINCI TO A.C;RE.E. WITH RC,REE WITH CTMT, P CC CTMT, p.J\CC.v£NT.FltJ.~~ ~ - Arv ~ CONT. CR111 RM. FICTER UNITS RM.F,LTER UNITS FIELD FIELD CH"",C;E CHANGE TO TO BE MADE, SE MI>.DE., I.IIAQS7AII2SO7Lj 1J'c'1.*'i~1807-IJ I~ Gl~N49TAHz.e03 QW4IrAHa8o3 SPACE SPACE 14 SPACE SPACE 9. PANEL Ie, S. PA~IE.\' IC I':> OF4Q IS \-lOt,j - Q IS 1(0 SWITCH NON-
- 10. SWITCH
- 10. NON- S>E.LECTIVEL'I SELECTIVELY T-RAIJSFER T.NAIJSFER BETWEEN BETWEEt-I SPACE SPACE _! _L_ SPACE SPACE THE TWO THE TWO 12.51/.
IZSV. DC DC ~OURCES SOURCES WHEN WHEN EITHER EITHER 177 18 IS SPACE SPACE SPACE SPACE SOURCE FALLS SOURCE BELOW 90% FALLS BEL..OW 9O7 OF NOMINAL OF NOMINAL SPACE SPACE ~ IS
....~ 10
- SPACE SPACE VOLTAGE VOLTAGE. _ (2-PDLE (2-POLE 125V 125V DC DC 150150 AMPS)
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- 11. 1-4/C #16
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.ITION Pt.-NE-L. JT'ON PANEL Ie. lB SPACE SPACE SPACE SPACE 12. REFER REFER TOTO FUSE FUSE MANUAL MANUAL A-181987 A181987 FDR FUR FUSE FUSE SIZE SIZE AND AND TYPE. TYPE. 12.
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'23 23 24
'21; SPACE SPACE Fr C:\Reference Disk\Exam
Title:
C:\Reference
Title:
Disk\Exam Reference Reference Disk\Drawings\D177082-0001.cal Disk\Drawings\D1 77082-0001 .cal
Date: 1O/ Date: 10r -'f~~,o~ 2009 Time Time: \03:29 03:29 PM 1\ 1 ! I !I
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