ML063380414
| ML063380414 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 09/07/2006 |
| From: | Entergy Nuclear Operations |
| To: | David Silk Operations Branch I |
| Sykes, Marvin D. | |
| References | |
| Download: ML063380414 (371) | |
Text
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet l /'
Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#1 Group # 1 WA # ~~
000007EA1.04 OK Importance Rating 3.6 Ability to operate and monitor RCP operation and flow rates as they apply to a reactor trip Proposed Question: Common 1 Given the following conditions:
0 The plant is operating at 25% power 0 The high pressure tap to RCS flow instrument FT-416A on loop 31 develops a large leak What is the resulting plant condition, if NO operator action is taken?
-../
'~
A. All loop 31 flow indicators will read low, but the reactor trip is NOT generated on RCS loop low flow.
B. All loop 31 flow indicators will read low, but the reactor trip is generated on low PRZR pressure.
C. Only FI-416A RCS flow indication will read low, but the reactor trip is NOT generated on RCS loop low flow.
D. Only FI-416A RCS flow indication will read low, but the reactor trip is generated on low PRZR pressure.
Proposed Answer:
A. All loop 31 flow indicators will read low, but the reactor trip is NOT generated on RCS loop low flow.
Explanation (Optional):
W
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): SD-1.1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-RCS-001 0 0 0 2 ~ (As available)
Question Source: Bank # INPO Modified Bank # 19403 (Note changes or attach parent)
New Question History: 12/11/2000 Kewaunee, Unit 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2,3,7,14 55.43 Comments:
System Description 1.1 Reactor Coolant System 443 connected to the loop 34 hot leg via a common root line isolated by manual valve RC-512.
PI-475 is a locally indicating, drag type pressure indicator with a range of 0-2500 psig. It is mounted adjacent to steam generator 34 on the 46-foot elevation in containment.
PT-403 is a wide range transmitter that supplies a pressure signal to:
the plant computer, 0 recorder PR-402/3 (range 0-3000 psig) on panel FDF, 0 indicator PI-403 (range 0-3000 psig) on panel FDF, e the RCS saturation monitor, 0 RHR suction valve RH-MOV-731 motor operator.
+ The signal to RHR suction valve RH-MOV-731 is an interlock signal. MOV-731 cannot be opened until PT-403 is less than 450 psig and the valve auto closes if PT-403 exceeds 550 psig. (RHR Suction Valves, RH-MOV-730 and RH-MOV-731 are located in series so that closure of either valve isolates the RHR system from the RCS.)
PT-443 provides pressure input signals to the OPS, the plant computer, and to PI-443K (0-15OOpsig range) on panel SFF.
LOOP 33 The loop 33 pressure transmitters are shown on Figure 1.1-25 and include PT-433 and PI-1400.
PT-433, which taps off the loop 33 hot leg sample line via manual isolation valve SP-1026. PT-433 supplies M-433K (0-1500 psig range) on panel SFF, and provides pressure input signals to the OPS and to the plant computer.
Figure 1.1-25 also shows PI-1400, a 0-3000 psig locally indicating pressure gauge. PI-1400 is installed on the combined Loop 31 and 33 hot leg sample line providing flow to the Gross Failed Fuel Sample Cooler. The gauge is located on the 67 elevation of the PAB adjacent to the gross failed fuel detector. This gauge can be manually isolated via valves SP-650A and/or SP-650B.
2.4.6 Loop Flow Instruments Three flow transmitters are installed in the intermediate leg of each RCS loop. Their primary function is to indicate whether a reduction in loop flow rate has occurred rather than to provide an accurate measure of loop flow rate.
u Rev3, 10/16/2000 - Page 26 -
System Description 1.l Reactor Coolant System The flow transmitters shown in Figure 1.1-12 are differential pressure cells that develop a signal proportional to the flow rate. Unlike typical flow detectors that use a flow-restricting device to produce a pressure drop, the RCS loop flow transmitters are installed on the elbow in RCS piping at the RCP suction. As the coolant flows through the elbow, it exerts a higher pressure on the outer curvature of the elbow than it does on the inner curvature. This pressure difference is a result of the difference in centrifugal force on the fluid between the inner and outer curvature. The expected accuracy of this method of flow detection is
+IO%. Density compensation is not required because of the minimal variations in the intermediate leg coolant density as a function of the unit power.
Three low pressure root lines are used to ensure that a single failure cannot prevent the loop flow instrumentation from generating a loop low flow signal. If a low pressure root line ruptures, the pressure in the line decreases. This would cause the differential pressure to increase, indicating a high loop flow rate. This is a non-conservative failure. With three independent low pressure root lines, failure of one line leaves two other transmitters available to detect an actual low flow condition.
--Only one high pressure root line is required because its failure I (rupture) would cause all three differential pressure detectors to sense d a low flow. Thus, the loop flow instruments fail safe (produce a loop low flow signal) if: the single high pressure root line fails. Using only one high pressure line reduces the number of penetrations in the RCS.
Each ff ow tap is provided with a manual root isolation valve (HPand all 3 LP taps). For Loop 31, these are RC-513 (HP tap) and RC-514A, RC-524B, and RC514C (LP taps).
Figure 1.1-26 presents the three flow instruments for RCS loop 31 and is typical for a27 loops. Each flow transmitter supplies A flow indicator (FI-414, FI-415, or FI-416) on panel SAF in the control room (range 0-120%),
A computer input (CI), and A low flow comparator bistable (FC-414, FC-415, or FC-416).
+ At 93% of normal flow on one of the three sensors, the comparator generates a low flow trip signal and an alarm on panel SAF.
- 1) If two of the three sensors detect 93% of normal flow, the coincidence Iogic gate transmits a loop low flow signal. This
~d Rev3, 10/16/2000 - Page 27 -
System Description 28.0 Overall Unit Protection 2.2.2 P-7 Permissive (Figure 28-15)
The P-7 permissive is used to block the high pressurizer level, low pressurizer pressure reactor trips, reactor coolant low flow and undervoltage reactor trip signals to the reactor protection system. The P-7 permissive is activated by a bistable circuit indicating less than 10%
power as measured by both turbine first stage pressure detectors and 3/4 power range channels. The power range input is supplied by the P-10 permissive. A white "POWER BELOW P-7 lamp illuminates on the control room FBF panel while the P-7permissive is active and extinguishes when reactor power and/or turbine power are >lo%.
2.2.3 P-8 Permissive (Figure 28-14)
The P-8 permissive blocks the automatic reactor trip on low flow in one loop if power is below 35% at the time one RCP is lost. The permissive also blocks a reactor trip due to a turbine trip when power is below 35%. A white "POWER BELOW P-8"lamp illuminates on the control room FBF panel when the P-8 permissive is active and extinguishes when reactor power is above 35%.
2.2.4 P-10Permissive (Figure 28-14)
The P-10 permissive blocks the intermediate range channel and low power range channel trips during an approach to power. It is also used to backup the P-6 permissive to block the Source Range instrumentation and is one of the inputs to the P-7permissive.
When 2 of 4 power range channels indicate greater than 8.5%power the P-10 permissive is activated and a white "POWER ABOVE P-10" lamp illuminates. Once the P-10 lamp is lit, the low power and intermediate range hi flu trips may be manually blocked as described in the sections for those trips.
The P-10permissive and associated manual blocks are automatically reinstated if power falls below 8.5%on 3/4 Power Range channels.
2.2.5 Low Power Auto Rod Withdrawal Block Automatic control rod withdrawal is blocked until turbine power, as sensed by PT-4124 (turbine first stage pressure), is greater than 15%.
Automatic control rod insertion is not blocked. A white "LOW P W R AUTO ROD WITHDRAWAL BLOCK" light, on the FBF panel, is illuminated when the permissive is active and extinguishes when turbine power is >IS%.
d' Rev. 3,06/29/2000 - Page 23 -
System Description 28.0 Overall Unit Protection
/ Power > P 4 Reactor Coolant Pumps
.-l Power > P 3 Reactor Coolant Pumps Power e P no pumps are required All reactor coolant flow trips are blocked below 10% power by permissive P-7 because natural circulation is capable of cooling the core below 10% power. Administratively all four reactor coolant pumps are required to be in operation any time the reactor trip breakers are closed (T.S.is less restrictive).
Three flow measuring circuits monitor each reactor coolant loop and are arranged in a two out of three logic. A low flow trip will occw when any two of three channels indicate that flow has decreased to 93% of normal full flow. This trip signal is used in two logic circuits.
The first circuit is used to trip the reactor if power is > P-8 and flow is lost in any one loop. P-8 is activated when reactor power increases to 35%. The second logic circuit is used to trip the reactor if power is greater than or equal to 10%and flow is low in 2 or more loops. P-7 is activated when power is less than 10% blocking the low flow trips.
A single channel tripping would annunciate the "REACTOR COOLANT LOOP 31 (32,33,34) LOW FLOW CHANNEL TRIP" Alarm on the SAF panel in the Control Room.
The RCP breaker open signal is derived from a breaker auxiliary "a" contact to actuate a low flow trip signal. This trip signal is provided to anticipate probable plant transients and to avoid the resultant thermal transient. The trip signal is sent to a one out of two logic device that senses breaker position and loop flow. A breaker open signal will initiate a signal that is sent to two logic circuits to compare the flow combination with the reactor power level. These circuits are the ones described above. With power above 35% and P-8 inactive one RCP breaker open signal causes a trip; with power below P-8 and >lo%
AND P-7 inactive it takes two RCP tip signals to cause the trip.
The RCP undervoltage trip signal provides protection following a complete loss of power. The RCP's are powered from 6.9KV buses 1 through 4. A decrease in voltage to 75% of nominal voltage, as sensed by the undervoltage relays/ initiates the signal. The four bus undervoltage relay signals are sent to a two out of four logic circuit. It requires two bus undervoltage signals to generate the undervoltage reactor trip. This signal indicates a probable loss of power and protects the core from DNB. Permissive P-7blocks this trip below 10%power.
A sustained undervol tage condition for .5 seconds, overcurrent, underfrequency on two out of four 6.9 kV busses (1-4) or a ground ij condition trips the RCP breaker using a one out of four logic. These Rev. 3, 06/29/2000 - Page 14 -
- The plant is operating at 18% power
- T h e high pressure tap to RCS flow instrument FT-411 on loop A fails hat is the resulting plant condition, if NO operator action is taken?
IAll loop A flow indicators will read low, and the reactor trip is generated on RCS loop low flow.
]generated on low PRZR pressure.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4 Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#2 Group # 1 KIA # 000008AA2.20 OK Importance Rating 3.4 Ability to determine and interpret ti le effect of an open P O R l on code safety, based on observation of plant parameters as they apply to the Pressurizer Vapor Space Accident Proposed Question: Common 2 The following conditions are observed about 3 minutes after an automatic safety injection:
Core exit T/Cs 540°F Pressurizer level 58% and increasing 0 RCS pressure 1100 psig and decreasing 0 Containment pressure 0 psig Based on these indications, it is likely that a; A. small break LOCA has occurred outside the containment.
6 . steam line break has occurred outside containment.
C. steam generator tube rupture has occurred.
D. a pressurizer PORV is stuck open.
Proposed Answer:
D. pressurizer PORV is stuck open.
Explanation (Optional):
-.-/ Technical Reference(s): SD-1.4 (Attach if not previously
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-RCSPZR E6a (As available)
Question Source: Bank # INPO 26069 Modified Bank # (Note changes or attach parent)
New Question History: 9/01/2003 Prairie Island 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3,7,14 55.43 5 Comments:
System Description I.4 Pressurizer & Pressurizer Relief Tank 0 if the RCS pressure increases to greater than 2335, then the bperator must ensure the PORVs are unisolated (block valves open) and have opened.
Pressure decreases below the normal control band if L- I -_
0 a PRZR spray valve fails open (or fails to close),
0X-P lose) 0 all PRZR heaters fail off 0 the auxiliary spray valve fails open.
The first two possible causes (spray valve or PORV failure) can cause a rapid reduction in pressure. mahout rapid operator response, a reactor trip and a safety injection on low pressurizer pressure occur.
If the operator has time to respond, ONOP-RCS-2 directs the operator to check for failed components based on their relative effect on PRZR pressure:
ensure the PORVs are closed. If a PORV does not close, then its block valve is closed and the fuses to the PORVs solenoid circuit are removed. Removing the PORV solenoid circuit fuses causes the PORV to fail closed; 0 ensure both spray valves are closed. If a spray valve is open in AUTO and it should be closed, it is closed in MANUAL. If it cannot be closed in MANUAL, then the valve controllers power fuses are removed. Removing the power fuses should cause the valve to fail closed. If a spray valve is still open, the RCPs (which provide the driving force for spray) must be tripped. Tripping RCPs requires that the Reactor be tripped first; If pressure is less than 2185 psig (and the PORVs and spray valves are not the cause of the event, or have been closed),
ensure all the backup heaters have energized to restore RCS pressure. If the backup heaters can not be energized from the control room, then the 31 backup heaters can be operated from the Local Pressurizer Pressure and Level Control Panel in the PAB.
0 Ensure that auxiliary spray valve AOV-212 is closed and is not leaking.
( Li Rev. 3, 06/02/2000 - Page 61 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet W
Examination 0utIine Cross-reference: Level RO SRO Tier # 1 WS#3 Group # 1 WA # 000009EK2.03 OK Importance Rating 3 .O Knowledge of the interrelations between the small break LOCA and the SGs Proposed Question: Common 3 The following plant conditions exist:
0 A reactor trip with SI has occurred.
0 The crew transitioned from E-0, Reactor Trip or Safety Injection, to FR-H.1, Loss of Secondary Heat Sink, based on valid red path condition on the heat sink CSF.
0 RCS pressure is 700 psig and slowly decreasing.
0 All S/G pressures are approximately 950 psig and stable.
u Which of the following summarizes plant conditions and what procedure should be implemented?
A. Because S/Gs are the sole heat sink, a transition to E-I , Loss of Reactor or Secondary Coolant, is made to minimize coolant loss and restore S/G levels to normal band.
B. Heat transfer in the RCS during this event is such that the S/Gs are currently NOT functioning as a heat sink and therefore NOT required. Return to E-0 then transition to E-I , Loss of Reactor or Secondary Coolant.
C. Heat transfer in the RCS during this event is such that the S/Gs are currently NOT functioning as a heat sink. Remain in FR-H.l to restore S/G levels to normal band.
D. Remain in FR-H.l until feed is restored then transition to E-1 where a depressurization of the secondary is prescribed to increase the heat transfer between the RCS and S/Gs.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
B. Heat transfer in the RCS during this casualty is such that the S/Gs are currently NOT functioning as a heat sink and therefore NOT required. Return to E-0 then transition to E-I , Loss of Reactor or Secondary Coolant.
Explanation (Optional):
Technical Reference(s): FR-H. 1 bases, step 1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOPFRH 7 (As available)
Question Source: Bank # INPO 2471 7 Modified Bank # (Note changes or attach parent)
New Question History: 5/30/2003 Seabrook 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,7 55.43 Comments:
('
L// . . -- -
SI e: Check Secondary Heat Sink Is Required PURPOSF: T o check i f a secondary (SG) heat sink i s required f o r heat removal BAS IS ;
Before implementing actions t o restore flow t o the steam generators, the operator s h o u l d check i f secondary heat s i n k is required. For larger LOCA break sizes, t h e RCS will depressurize below the intact steam generator pressures. The steam generators no longer function as a heat s i n k and the core decay heat i s removed by the RCS break flow. For this range o f LOCA break sizes, the secondary heat s i n k i s not required and actions t o restore secondary heat s i n k are not necessary. For these cases, the operator returns t o the guideline and step i n effect.
(
1
-. Since Step 8 directs the operator t o return t o S t e p 1 i f the loss o f secondary heat s i n k parameters are not exceeded, break sizes that take longer t o d *-
I depressurize the RCS will be detected on Subsequent passes through Step 1.
I f RCS temperature i s low enough t o place the RHR System i n service, then the RHR System i s an alternate heat sink t o the secondary system. Therefore, an attempt i s made t o place the RHR System i n service i n parallel t o the attempts t o reestablish feedwater flow. RCS pressure must be below normal RHR System pressure 1imi ts.
ACT IONS :
o Determine i f RCS pressure i s greater than any nonfaulted SG pressure o Determine i f RCS temperature i s greater than (F.06)'F [(F.O7)OF for I adverse containment]
o Determine i f adequate cooling i s established w i t h RHR System o Try t o place RHR System i n service while continuing w i t h guideline o Return t o guideline and step i n effect I NSTRUMFNTATION:
o RCS pressure indication o SG pressure indication o RCS 'temperature indication o Plant specific RHR System instrumentation i n c l u d i n g valve position and pump status indications FR-H. 1 LFRHl
Number :
Title:
R e v i s i o n Number:
E-0 REACTOR TRIP OR SAFETY INJECTION 21 4
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 5. CHECK AFW S t a t u s :
a . V E R I F Y t o t a l AFW flow - a . PERFORM t h e f o l l o w i n g :
GREATER THAN 365 GPM 1 ) Manually START available pump( s 1 .
I 2 ) A L I G N valves as required.
- 3) cutback c o n t r o l l e r i s malfunctioning, THEN ATTEMPT m a n u a l c o n t r o l .
- 4) a l l SG NR l e v e l s a r e l e s s t h a n 9% C14%1AND t o t a l AFW flow can NOT be maintained g r e a t e r t h a n 365 gpm. THEN GO To F R - H . l , RESPONSE TO LOSS OF SECONDARY H E A T S I N K .
I b . CONTROL feed f l o w t o I
m a i n t a i n SG NR l e v e l s between 9% C14%1 a n d 50%
CAUTION S T A R T I N G OF EOUIPMENT MUST B E COORDINATED W I T H A L L CONTROL ROOM OPERATORS TO ENSURE THAT TWO COMPONENTS ARE NOT S T A R T E D A T THE SAME T I M E ON THE SAME POWER S U P P L Y .
- 6.
- D I R E C T BOP ODerator t o PERFORM *
- R O - 1 , BOP OPERATOR A C T I O N S T *
- D U R I N G USE OF EOPS *
- Page 6 o f 4 0
Number :
Title:
Revision Number:
FR-H. 1 RESPONSE TO LOSS OF SECONDARY HEAT SINK 18 RESPONSE NOT OBTAINED CAUTION I F TOTAL FEEDFLOW I S LESS THAN 365 GPM DUE TO OPERATOR A C T I O N ,
T H I S PROCEDURE SHOULD NOT BE PERFORMED.
0 FEEDFLOW SHOULD NOT BE R E E S T A B L I S H E D TO ANY F A U L T E D SG A NON-FAULTED SG I S A V A I L A B L E .
- 1. DETERMINE If Secondary Heat Sink Is Required:
a . CHECK RCS p r e s s u r e - a . RETURN To P r o c e d u r e and GREATER THAN ANY Step i n e f f e c t .
NON-FAULTED SG PRESSURE b . CHECK RCS h o t l e g b. PERFORM t h e f o l l o w i n g :
t e m p e r a t u r e s - ANY GREATER THAN 350°F 1) TRY t o p l a c e RHR c o o l i n g i n service while continuing with t h i s procedure:
REFER TO S O P - R H R - 1 ,
R E S I D U A L H E A T REMOVAL SYSTEM.
2 ) WHEN RCS t e m p e r a t u r e i s s t a b l e o r d e c r e a s i n g on RHR c o o l i n g . THEN RETURN T o P r o c e d u r e and Step i n effect.
Page 2 of 47
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
,d Examination Outline Cross-reference: LeveI RO SRO Tier ## 1 WS#5 Group ## 1 WA # ~
000015/17G2.1.28 OK Importance Rating 3.2 Knowledge of the purpose and function of major system components and controls Proposed Question: Common 4 When a RCP is stopped IAW 3-AOP-RCP-I , RCP MALFUNCTION, due to high #Seal I leakoff flow, the respective Seal Leakoff Isolation Valve, is closed after stopping the pump.
Which one of the following correctly describes the reason for closing the respective Seal Leakoff Isolation Valve?
A. Prevent excessive back pressure from interfering with leakoff from the operating
'4 RCP's.
B. Prevent over-pressurization of the return line and a possible LOCA Outside Containment.
C. Prevent thermal shock to the in-service CCW Heat Exchanger.
D. Reduce RCS inventory loss by directing all #seal I leakoff to #2 Seal.
Proposed Answer:
D. Reduce RCS inventory loss by directing all #I seal leakoff to #2 Seal.
Explanation (Optional):
Technical Reference(s): 3-AOP-RCP-1 step 13 bases (Attach if not previously provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet, ,
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-RCSRCP H (As available)
I3LP-ILO-AOPRCP K Question Source: Bank # INPO 23133 Modified Bank # (Note changes or attach parent)
New Question History: Salem Unit 1 11/4/2002 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
La5 %-
INSTRUCTOR LESSON PLAN
- 2) Before performance of E-0 IOAs C. INITIATE E-0
- 1) Procedure remains open and in effect
- 2) Additional actions may be required to mitigate the event a) Close seal return valve (1) Backpressure on #2 seal (2) Stop flow of hot water to VCT (3) Stop flow of hot water over thermal barrier Boiling in CCW system heat exchanger (4) Stop Flow of hot water over #1 & #2 seals Prevent hrther degradation
- 3. IAAT affected RCP has stopped rotating Step 4.13 Step 4.23 RCPs operating and flow 20 - 30 % affected loop
. All RCPs secured and flow 0%
- a. Close Seal return isolation valve (261A-D)
- 1) Backpressure on #2 seal
- 2) Stop flow of hot water to VCT
- 3) Stop flow of hot water over thermal barrier heat Boiling in CCW system exchanger
- 4) Stop Flow of hot water over #1 & #2 seals Prevent further degradation
- b. Close affected spray valve if 33 or 34 RCP
- 1) Prevent short cycling spray flow from unaffected loop
- 4. IATT #1 seal return flow is < 0.84 gpm check conditions to Step 4.17 t i p reactor and stop affected RCP 46
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#8 Group # 1 KIA # 000027AK2.03 OK Importance Rating 2.6 Knowledge of the interrelations between the Pressurizer Pressure Control Malfunction and controllers and positioners Proposed Question: Common 5 The plant is operating at 100% power with all systems and controls in a normal lineup.
Which ONE of the following could cause a Reactor Trip AND Safety Injection actuation?
(Assume NO operator action.)
A. TE-433B, Loop 34 Cold Leg (Channel 4), fails LOW.
B. PT-455, Pressurizer Pressure (Channel I), fails HIGH C. A trip of both Main Feed Pumps.
D. Direct Trip from Buchanan.
Proposed Answer:
B. PT-455, Pressurizer Pressure (Channel I), fails HIGH Explanation (Optional):
Technical Reference(s): SD 1.4 (Attach if not previously provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-ICPZPC E-5.a (As available) 13LP-ILO-AOPINT I.a.
Question Source: Bank # INPO 21540 Modified Bank # (Note changes or attach parent)
New Question History: 9/06/2002 Kewaunee, Unit 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
System Description 1.4 Pressurizer & Pressurizer Relief Tank high level reactor trip bistable. Once these actions are completed, maintenance repairs the instrument.
3.2.2 PRZR Pressure Channel Failure PRZR pressure instrument failures are addressed in ONOP-RPC-1, Instrument Failures. The following discussion assumes all systems are aligned for normal operations.
If any pressurizer pressure channel, PT-455,lT-456, PT-457, or PT-474, fails high, the affected channel indicates full scale and the Pressurizer High Pressure Channel Trip alarm is triggered on panel SAF (See Figure 1.422). If the failed channel is being used for alarm train (PT.456 or PT474):
PCV-456 (PORV) receives an open signal, however it does not open because the open permissive interlock is not satisfied (PT-457 is not greater than 2335 psig), and Pressurizer High Pressure alarm is triggered on panel SAF.
If the failed channel is being used as the input to the control train (PT-455 or PT-457), pressure controller PC-455K senses a large deviation above the setpoint pressure and outputs a large PERROR
! signal:
0 Pressurizer High Pressure alarm panel SAF, Control Heater Group are fully de-energized, both spray valves go full open causing a rapid lowering of actual P E R pressure, RC-PCV455C (PORV) receives an open signal, however it does not open because permissive interlock is not satisfied (PT-474 pressure is not above 2335 psig),
If operator action is not taken, the decrease in pressure causes protective actions:
a low pressure reactor trip when two of the remaining three pressure channels decrease to the trip setpoint; a safety injection actuation when two the operable pressure SI input channels reach the 1720 psig setpoint.
If a PRZR pressure channel fails low, the affected channel indicates down scale. The Pressurizer Low Pressure Channel Trip and Pressurizer Low Pressure (SI) Channel Trip alarms are triggered on panel SAF (Part of Reactor Protection and Safeguards not shown on L i Rev. 3, 06/02/2000 - Page 58 -
System Description 1.4 Pressurizer & Pressurizer Relief Tank n
Rev. 4, 12/14/2005
- Page 101 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'il Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#9 Group # 1 K/A # 000029EK1.01 OK Importance Rating 2.8 Knowledge of the operational implications of reactor nucleonics and thermo-hydraulics behavior as they apply to the ATWS Proposed Question: Common 6 During an ATWS event, the fuel cladding fission product barrier is severely challenged.
Which ONE of the following conditions is the mechanism which causes the fuelkladding challenge?
A. Heat generated from the Zr-H20 reaction.
B. Excessive radial flux distribution.
C. High RCS pressure caused by high temperature.
D. Fuel overheating from DNBR limits being exceeded.
Proposed Answer:
D. Fuel overheating from DNBR limits being exceeded.
Explanation (Optional):
Technical Reference(s): EOP FR-S. 1 Bases (Attach if not previously provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-EOPFRS 3 (As available)
Question Source: Bank # INPO 20212 Modified Bank # (Note changes or attach parent)
New Question History: 9/10/2001 Cook 1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 1,7,14 55.43 6 Comments:
'4
2.3.. A W_-- .f v e n t s =- c ATWS events are postulated - to be initiated from Condition I1 transients.
Commonly termed "Anticipated Transients to distinguish them from the more severe, lower probability Condition 111 and I V transients,' they include:
- 1) Uncontrolled RCCA Bank Withdrawal
- 2) Uncontrolled RCCA Bank Misalignment '
- 3) Partial Loss o f Forced Reactor Coolant Flow
- 4) Loss of Load and/or Turbine Trip
- 5) Loss of Normal Feedwater
- 6) Station Blackout (Loss o f Offsite Power)
- 7) Accidental RCS Depressurization The basic design criteria for Condition 11 events require that they be tolerated with, at most, a shutdown o f the reactor with the plant capable of returning to operation after corrective action. Fue.1 damage is not expected f o r Conditon I1 events, although a small number of fuel rods may experience limited damage. These are within the capdbility o f the plant clean-up systems.
A common characteristic of these events is a power generation-power removal mismatch l e a d i n g to temperature excursions of the RCS. Some are characterized by increasing RCS pressures and others by RCS depressurization. It is usual to evaluate the core performance i n terms o f changes in the DNB ratio (or DNBR). The design duty cycle and SAR analyses report the results o f these events i n terms of DNBR changes. These results are used, in part, to e s t a b l i s h the set points for the reactor protection system.
ATWS events are postulated to initiate from the Condition I1 transients, except that the reactor protection system is assumed t o malfunction i n a manner to preclude rod drop into the core. Several sets of analyses have been I
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'd Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#11 Group # 1 K/A # WlE1262.2.34 OK Importance Rating 2.8 Knowledge of the process for determining the internal and external effects on core reactivity Proposed Question: Common 7 Given the fo Ilowing:
0 Operation at 100% power at EOL 0 Unit 3 experienced a large steam line break on the common steam header.
0 A Reactor Trip and Safety Injection occurred.
0 NONE of the MSIV's closed automatically or manually.
4 Which one of the following describes the response of core reactivity following the Reactor Trip and Safety Injection?
A. Reactivity DECREASED due to the RCS cooldown then INCREASED due to the Safety Injection.
B. Reactivity INCREASED due to the RCS cooldown then DECREASED due to the Safety Injection.
C. Reactivity DECREASED throughout the whole transient.
D. Reactivity INCREASED throughout the whole transient.
Proposed Answer:
B. Reactivity INCREASED due to the RCS cooldown then DECREASED due to the Safety Injection.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
-4 Explanation (Optional):
Technical Reference(s): EOP ECA-2.1 Bases (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOPE2O 2 (As available)
Question Source: Bank # INPO 22527 Modified Bank # (Note changes or attach parent)
New Question History: 10/1/2002 Diablo Canyon 1
~4Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 1 55.43 6 Comments:
4"'
Intermediate Size Break A break i s assumed t h a t will r e s u l t i n a l l loops experiencing decreasing steam pressure and increasing steam load f o r which control systems a r e unable t o compensate. Steam generator water level and primary average temDerature w i l l slowly decrease and the control rods will commence stepping out o f t h e core i n a n attempt t o maintain nominal primary system average temperature. A l s o , due t o the decreasing temperature, a primary pressure decrease occurs. This t r e n d will continue u n t i l e i t h e r t h e operator manually t r i p s t h e reactor o r a t r i p s e t p o i n t is reached on overpower AT, low steamline pressure, o r low pressurizer pressure. In any case, s i g n a l s would eventually be generated f o r r e a c t o r t r i p , turbine t r i p , safety i n j e c t i o n , feedwater i s o l a t i o n , steamline i s o l a t i o n , and a u x i l i a r y feedwater i n i t i a t i o n . W i t h a f a i l u r e t o r e s t o r e i n t e g r i t y t o any steam generator ( e . g . , a l l MSIVs f a i l t o c l o s e ) , a l l steam generators would continue t o blowdown t o atmospheric pressure resulting i n a continued decrease i n primary system temperature and pressure. As t h e primary I system temperature drops, t h e heat t r a n s f e r t o the steam generators and'the
'.>' primary system cooldown r a t e will be redu,ced. T h i s trend will continue t o t h e point wnere the p r i m a r y system water v'olumk shrinkage (caused by the cooldown)
!5 overcome by the s a f e t y i n j e c t i o n system flowrate. This r e s u l t s i n the c v i m a r y system pressure and pressurizer 'level r e s t o r a t i o n . Feed flow i s t h e n al:us:ed t G control RCS t e m e r a t u r e s .
- OF t n e double-enaed oreak, ap. immediate decrease i n steamline pressure t o t h e low steamline pressure s e t p o i n t (0.5-10 seconds) r e s u l t s i n s a f e t y i n j e c t i o n ,
f e e d l i n e i s o l a t i o n , reactor t r i p , turbine t r i p and a u x i l i a r y feedwater i n i t 7 a t i o n signals. W i t h the f a i l u r e t o r e s t o r e secondary pressure boundary i n t e g r i t y t o any steam generator ( i . e . , common f a i l u r e of the main steam 1 s o l a t i o r valves), a r a p i d , extensive primary system cooldown and aepressu-izaton occurs. As the primary system temperature d r o p s , the heat t r a n s f e r t o the steam generators and the primary system cooldown r a t e w i l l be ECA- 2 . 1 13 LP-Rev. 1 0125V
reduced. Thfs trend w i l l continue t o t h e point where the primary system water
'4 vofurne%hankage (caused by the cooldown) -isovercome by t h e s a f e t y i n j e c t i o n system flowrate. Thfs r e s u l t s i n t h e primary system pressure and p r e s s u r i z e r
'level r e s t o r a t i o n . Depending on the i n . i t i a 1 conditions o f t h e systems and the size and location of the break, one o f two conditions w i l l be reached on the blowdown. The f i r s t is when the steam generator blowdowns a r e e s s e n t i a l l y completed and f u r t h e r cooldown of t h e primary system i s c o n t r o l l e d by the a u x i l i a r y feedwater flow. The second i s when the primary temperature i s reduced so f a r t h a t the heat t r a n s f e r t o t h e steam generators matches t h e heat generation i n the primary system which results i n a s t a b i l i z e d primary temperature.
ECA-2.1 14 0125V LP-Rev. 1
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'4 Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#4 Group # 1 W A# 00001 1 EA2.14 OK Importance Rating 3.6 Ability to determine or interpret the actions to be taken if limits for PTS are violated as they apply to the Large Break LOCA Proposed Question: Common 8 Given the following conditions:
0 A LOCA had occurred from HOT SHUTDOWN conditions 30 minutes ago 0 RCS pressure is 125 psig 0 RCS Core Exit TCs read 380°F 0 RCS Cold Leg temperatures are all 220°F 0 31 SI Pump is running providing 325 gpm flow 0 31 RHR Pump is running providing 1150 gpm flow What is the appropriate action taken in response to the above conditions?
Entry into FR-P.1, Response to Pressurized Thermal Shock Condition, is ...
A. NOT required since RCS pressure is below 350 psig.
B. made but NO actions are implemented before returning to procedure in effect.
C. made and a RCS temperature soak for a ONE hour period will be completed.
D. made and cooldown will continue within a limit of 50°F in any 60 minute period.
Proposed Answer:
B. made but NO actions are implemented before returning to procedure in effect.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
%d Technical Reference(s): FR-P. 1, step 1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-EOPFRP 12 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X 4
Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3,5,7 55.43 5 Comments:
Number:
Title:
R e v i s i o n Number:
FR-P.1 RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITIONS 14
- 1. CHECK RCS P r e s s u r e - GREATER -
I F RHR f l o w i s g r e a t e r than THAN 325 P S I G 1650 P S I G l 300 qpm on each o f two i n d i c a t o r s , THEN RETURN To P r o c e d u r e and S t e p in e f f e c t .
I I F THE TURBINE-DRIVEN AFW PUMP I S THE ONLY AVAILABLE SOURCE OF FEED FLOW, THEN THE S T E A M SUPPLY TO THE TURBINE-DRIVEN AFW PUMP MUST BE MAINTAINED FROM ONE SG.
NOTE A f a u l t e d SG i s any S G . t h a t i s d e p r e s s u r i z i n g i n an u n c o n t r o l l e d manner o r c o m p l e t e l y d e p r e s s u r i z e d .
TemDeratures - S T A B L E OR INCREASING a . VERIFY SG a t m o s p h e r i c s a r e closed.
b . V E R I F Y condenser steam dump valves are closed.
(STEP 2 CONTINUED ON NEXT PAGE)
Page 2 of 3 4
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination Outline Cross-reference: Level RO SRO Tier # 1 W S # 12 Group # 1 WA # 000054AK1.01 OK Importance Rating 4.1 Knowledge of the operational implications of the MFW line break depressurizes the SG as they apply to the Loss of Main Feedwater Proposed Question: Common 9 Given the following plant conditions on Unit 3:
Reactor power is 90%.
RCS Tave is stable at 565°F on all 4 loops RCS pressure is stable at 2235 psig Containment Pressure is INCREASING 33 SG Feed Flow is pegged HIGH 33 SG Main Fw Reg Valve is full OPEN.
33 SG pressure is STABLE 33 SG level is DECREASING Which of the following events is in progress?
A. Main Feed Pump trip B. Feed Flow Indicator failed HIGH.
C. Feed Line Break INSIDE Containment.
D. Main FW Reg Valve failed OPEN.
Proposed Answer:
%d
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet C. Feed Line Break INSIDE Containment.
Explanation (Optional):
Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOPE20 2 (As available)
Question Source: Bank # INPO 19254 Modified Bank # (Note changes or attach parent)
New Question History: 10/20/2000 Braidwood 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8,lO 55.43 Cornments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet L.J Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#13 Group # 1 KIA # 000055EK3.02 OK Importance Rating 4.3 Knowledge of the reasons for actions contained in EOP for loss of offsite and onsite power as they apply to the Station Blackout Proposed Question: Common 10 The following plant conditions exist:
Unit 3 has experienced a loss of all AC Power due to severe weather conditions and failure of emergency diesel generators to start and supply safeguard buses.
The operating crew is carrying out actions of ECA-0.0, Loss of All AC Power.
Immediate actions have been completed and steps to restore power are in progress.
The operators are at a point where they are to commence cooldown and depressurization of the steam generators.
Based on these conditions, which ONE of the following statements describes the reason why a secondary depressurization is directed?
A. To prevent a challenge to the Integrity Critical Safety Function Status Tree which is being monitored for implementation.
B. To minimize RCS inventory loss through the RCP seals which maximizes time to core uncovery.
C. To ensure the reactor remains subcritical and does not result in a restart accident.
D. To remove all the stored energy in the steam generators to prevent a secondary safety valve from lifting.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
.<d Proposed Answer:
B. To minimize RCS inventory loss through the RCP seals which maximizes time to core uncovery.
Explanation (Optional):
Technical Reference(s): ECA-0.0 Bases (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOPCOO 8 5804 (As available)
Question Source: Bank # INPO 20572 Modified Bank # (Note changes or attach parent) 4 New Question History: 2/2/2002 Point Beach 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,lO 55.43 5 Comments:
STFP DFSCRI PTION TABLE FOR ECA-0 .O Step 16 STEP: Depressurize I n t a c t SGs To (0.08) PSIG PURPOSE: T o depressurize the i n t a c t steam generators BASIS:
Step 16 depressurizes the i n t a c t SGs, thereby reducing RCS temperature and pressure t o reduce RCP seal leakage and minimize RCS inventory l o s s . The advantages t o performing this action, as well as r e s t r i c t i o n s t h a t apply d u r i n g the action, are detailed i n Subsection 2.3.
During SG depressurization, SG level must be maintained above the top of the SG U-tubes i n a t l e a s t one SG. Maintaining the U-tubes covered i n a t l e a s t one SG will ensure t h a t s u f f i c i e n t heat t r a n s f e r c a p a b i l i t y exists t o remove heat from t h e RCS via either natural c i r c u l a t i o n o r reflux boiling a f t e r the RCS saturates. Step 16a requires that SG level be i n the narrow range i n a t l e a s t one SG before SG depressurization i s i n i t i a t e d i n Step 16b. I f level i s n o t i n the narrow range i n a t l e a s t one SG, RNO 16a instructs the operator t o maintain maximum AFW flow u n t i l narrow range level i s established i n one SG.
When narrow range 1 eve1 i s establ i shed;. SG"depressurization can be s t a r t e d or continued via Step 16b.
Step 16b i n s t r u c t s the operator t o reduce SG pressures by depressurizing the i n t a c t SGs. Depressurization should be accomplished by opening the PORVs on the i n t a c t SGs t o e s t a b l i s h a maximum steam dump rate, consistent w i t h plant s p e c i f i c c o n s t r a i n t s . The step i s structured assuming t h a t the operator can open and control SG PORVs from the control room. T h i s s t r u c t u r e assumes t h a t the PORVs are air-operated and have dc control power and pneumatic power
( i . e . , either a i r reservoirs o r nitrogen b o t t l e s ) available. Some plants may n o t have the c a p a b i l i t y t o open the SG PORVs from the control room. These plants s h o u l d evaluate their c a p a b i l i t y t o accomplish this s t e p l o c a l l y via PORV handwheels. Such an evaluation should consider accessibility and comnunications necessary t o accomplish local PORV operation.
Once depressurization i s i n i t i a t e d , maintenance of a specified r a t e i s not c r i t i c a l . The depressurization r a t e should be s u f f i c i e n t l y f a s t t o expeditiously reduce SG pressures, b u t not s o f a s t t h a t SG pressures cannot be controlled. I t i s important t h a t the depressurization not reduce SG pressures i n an uncontro-lJed manner t h a t undershoots the pressure l i m i t , t h u s p e r m i t t i n g potenti=? introduction of nitrogen from the accumulators i n t o the RCS.
ECA-0 .O 117 LP-Rev. 1C LECAOO
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examinat ion 0utline Cross-reference: Level RO SRO Tier # 1 ws # 101 Group # 1 KIA # 000026AA2.01 Importance Rating 2.9 Ability to determine and interpret the location of a leak in the CCWS as they apply to the Loss of Component Cooling Water Proposed Question: Common 11 The operators are responding to a CCW leak IAW 3-AOP-CCW-I , Loss of Component Cooling Water. Makeup to 31 and 32 CCW Surge Tanks has been established. Both surge tanks continue to decrease slowly. After splitting CCW Headers, 31CCW Surge Tank is stable and 32 CCW Surge Tank continues to decrease. Which of the following sets of components could be the source of the CCW leak?
A. RCP Motor cooling, 31 SI Pump or Spent Fuel Pit Heat Exchanger W
B. 32 Waste Gas Compressor, Charging Pumps or 31 Aux Component Cooling Pump C. CVCS Non-reg Heat Exchanger, RCP Motor cooling or Seal Water Return Heat Exchanger D. Seal Water Return Heat Exchanger, 31 RHR Heat Exchanger or Reactor Vessel Support Pads Proposed Answer:
C. CVCS Non-reg Heat Exchanger, RCP Motor cooling or Seal Water Return Heat Exchanger Explanation (Optional):
A. 31 SI Pump and SFP HX on 31 Header B. 32 WGC on 31 Header C. Correct D. 31 RHR HX on 31 Header
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d Technical Reference(s): SD-4.1 Figure 4.1.1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-ccwoo1 0001 (As available)
Question Source: Bank#
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis d
10 CFR Part 55 Content: 55.41 55.43 5 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
c RCV CCW SYSTEM 31 VENT START STBY PUMP
'1 I
31 RHR PUMP 31 SI PUMP RECIRC 31,32 A.C.C. PUMPS f--
! c-- .
31,32, WASTE GAS COMP.
F.E. PRODUCT COOLER R-37 SAMPLE COOLER f-Pw MAKEUP 32 SUPPLY HEADER U L
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 14 Group # 1 KIA # 000056AK1.01 OK Importance Rating 3.7 Knowledge of the operational implications of the principles of cooling by natural convection as they apply to Loss of Offsite Power Proposed Question: Common 12 Unit 3 was operating at 100% power when a reactor trip occurred due to a loss of offsite power. The operators completed the actions of ES-0.1, Reactor Trip Response and have transitioned to ES-0.2, Natural Circulation Cooldown, where they are initiating a natural circulation cooldown.
- At the onset of the natural circulation cooldown, which ONE of the following processes will remove the MOST heat from the Reactor Vessel HEAD?
d' A. The 25"F/hr natural circulation cooldown of the RCS.
B. Heat losses to ambient.
C. All CRDM fans running.
D. Upper head bypass flow.
Proposed Answer:
C. All CRDM fans running.
Explanation (Optional):
Technical Reference(s): ES-0.2 Bases (Attach if not previously d provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet , ,
Proposed References to be provided to applicants
. . during examination: NONE Learning Objective: 13LP-ILO-EOPEOO 7 (As available)
Question Source: Bank # INPO 20211 Modified Bank # (Note changes or attach parent)
New Question History: 9/10/2001 Cook 1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis I O CFR Part 55 Content: 55.41 8.10.14 55.43 Comments:
i initial upper head temperature for these plants was conservatively chosen as 7 .
plants operate with sufficient flow from the upper
' HOT .--Orher Westinghouse downcomer to the upper head region to make the upper head fluid temperature equal to the cold leg fluid temperature (TCoLD). Both types of plants were analyzed .
Another parameter which affects void formation in the upper head region is the cooiaown rate o f the primary system. Natura1,circulation cooloown rates o f
,75OF./hr and SO°F/hr were analyzed for THOT and T~~~~ plants, respectively.
A finai parameter important in the formation o f voids in the upper head.is the near, removal rate from the upper head. The two primary means o f heat loss are ambient neat l'osses and heat removal by the control rod drive mechanism (CRDM) fans. Tne effect of ambient heat losses tnrough the reactor vessel on upper nead temperature is small compared to the effect o f the CRDM fans. The cooloff rate of the upper head due to ambient heat losses is less than 1°F/hr and was neglected in the analysis. However, metal heat addition to the-upper I neac area from the reactor vessel and upper internals was taken into account.
Tno CRDM cooli'ng system consists o f fans which maintain a suitable atmosphere
~ - t r i :ne r CROP snrouc to proiect and prolong the life of the CRDM motors.
Tne system induzes cooler containment air into the CRDM shroud and exhausts harmec air tnrough the fans. The CRDM fans remove 8 kwidrive train at full Dowe' Lor a 4-loop olant witk 57 full-length and 8 part-length rods, the C R Y fars remove E kw x (57 - 8 ) = 520 kw. For a 3 or 2 - l o o p plant, multiply tne zctal number o f fuli-iength plus part-length rods by 8 kw/rod to obtain tne heat removal comparable to tne 520 kw for 4-loop plants. The ratio of neat removal by the CRDF! fans t o the upper head total energy (or upper head volume) i s essentially the same for 2 and 3-loop plants as for 4-loop plants.
Tnus. tne head cocldown rates determined for 4-loop plants (21°F/hr at 6 O O O F an=: l!OF/hr at 350OF) are applicable to 2 and 3-100p plants.
I ES-0.2 4 LP-Rev. 1 0090V : lb
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination 0utIine Cross-reference: Level RO SRO Tier # 1 WS # 15 Group # 1 WA # 000065AA1.05 OK Importance Rating 3.3 Ability to operate and / or monitor the RPS as it applies to the Loss of Instrument Air Proposed Question: Common 13 Unit 3 is at 100% power. IF a rupture occurs in the Instrument Air system, you should monitor plant conditions and initiate a manual reactor trip if plant conditions approach any automatic reactor trip setpoint.
Which plant parameter is going to reach its automatic reactor trip setpoint FIRST for this event?
A. Steam generator level B. Pressurizer level C. Pressurizer pressure D. RCSloop AT (OTAT)
Proposed Answer:
A. Steam generator level Explanation (Optional):
Technical Reference(s): 3-AOP-AIR-1 (Attach if not previously provided)
ES-407 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-IA001 7 1799 (As available)
Question Source: Bank # INPO 26084 Modified Bank # (Note changes or attach parent)
New Question History: 9/1/2003 Prairie Island 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Air Systems Malfunction 3-AOP-AIR-1 Rev. 02 Page 37 of 61 Attachment 1 Valves of Immediate Concern Page 7 of 17 CONDENSATE SYSTEM 1 I VALVE I FUNCTION FAIL P O S l T l O N I I HOTWELL MAKE-UP - 7
~~ ~
rCD-LCV-1128 Closed 1 cT-Lcv-l CONDENSATE STORAGE TANK LO LEVEL ISOLATION VALVE 1 Closed I cT-Lcv-l 58-2 CONDENSATE STORAGE TANK LO LEVEL ISOLATION VALVE 1 Closed I
MAIN FEEDWATER SYSTEM VALVE I FUNCTION 1 FAIL POSITION
~~
IBFD-FCV-417 I 31 S/G MAIN FW REG I I BFD-FCV-427 I 32 S/G MAIN FW REG I Closed I BFD-FCV-437 I 33 S/G MAIN FW REG I BFD-FCV-447 34 S/G MAIN FW REG BFD-FCV417L 31 SG BYPASS FW REG BFD-FCV-427L 32 SG BYPASS FW REG Closed BFD-FCV437L 33 SG BYPASS FW REG BFD-FCV-447L 34 SG BYPASS FW REG BFR-FCV-1115 31 MBFP RECIRC VALVE Open BFR-FCV-1116 32 MBFP RECIRC VALVE SERVICE WATER SYSTEM VALVE I FUNCTION I FAIL POSITION CONTAINMENT TEMPERATURE CONTROL Open I SWN-TCV-1105 I p 1 SWN-TCV-1113 1 31 & 32 IACC HEAT EXCHANGERS OUTLET TEMPERATURE CONTROL VALVE I Open
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet i/
Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#6 Group # 1 KIA # 000022AK1.01 OK Importance Rating 2.8 Knowledge of the operational implications of consequences of thermal shock to RCP seals as it applies to Loss of Reactor Coolant Pump Makeup Proposed Question: Common 14 Unit 3 has entered 3-AOP-CVCS-1, Chemical and Volume Control System Malfunction due to a loss of all charging. The procedure cautions the operator not to re-establish seal injection if seal injection temperature has reached 225°F until the plant has been placed in Mode 5.
Which ONE of the following is the reason for this action?
'd A. To minimize thermal shock of the RCP seals.
B. To prevent thermal barrier heat exchanger failure.
C. To minimize the pressurizer control transient.
D. To prevent cocking the RCP seal causing excessive leakage Proposed Answer:
A. To minimize thermal shock of the RCP seals.
Explanation (Optional):
Technical Reference(s): 3AOP-CVCS-1, Note prior to (Attach if not previously step 4.37 provided)
'L.1
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet ,
I3LP-ILO-EOPCOO 5976 SD 1.3 Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-CVCOOI 8.h. (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8,lO 55.43
..4' Comments:
Chemical and Volume Control System 3-AOP-CVCS-I Rev. 1 MaIfunction Page 79 of 63 i'
I ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED 1 4.36 Dispatch an operator to close the following:
-CH-241A (RCP 31 Seal Injection Flow Control Valve)
-CH-2416 (RCP 32 Seal Injection Flow Control Valve)
-CH-241C (RCP 33 Seal Injection Flow Control Valve)
-CH-241D (RCP 34 Seal Injection Flow Control Valve)
NOTE I Ifseal inlet temperature reached 225"F,seat injection will not be restored until plant is in MODE 5.
A 4.37 -Determine and correct cause for
- l loss of charging 4.38 -PERFORM
--t and seal injection.
Attachment I(Charging and Letdown Restoration) (Page 45) to restore charging and letdown.
4.39 -RETURN effect.
to procedure and step in I
I I_-L_ --_-
System Description 1.3 Reactor Coolant Pump affected pump. The result of a #3 Seal failure is a standpipe low level for the affected pump.
3.2.1 Emergency Shutdown Immediate pump shutdown is required if any of the following conditions occur:
0 A loss of BOTH seal injection AND CCW cooling to any RCP.
0 The pump #1 Seal Delta P is less than 200 psig.
0 RCP seal return flow is less than 0.2 gpm.
0 RCP seal in temperature exceeds 225 OF.
RCP seal outlet temperature is greater than 235°F.
RCP vibration is greater than 5 mils Frame or 20 mils shaft.
RCP Motor Winding temperatures exceed 250 "E 0 Either motor bearing temperature reaches 200°F. In the event CCW is lost, temperatures will exceed 200°F in one to two minutes.
The reactor is to be placed in hot shutdown via normal operating procedures and the affected RCP secured if any of the following conditions occur:
Seal injection temperature reaches 150°F.
Motor bearing oil reservoir level alarms accompanied by abnormal pump indications.
Restoration of seal injection after a loss of both CCW and seal injection must be done very slowly (l"F/min) to reduce thermal stresses on the RCP internds. A seal package inspection and evaluation should be completed prior to any restart of an affected RCP.
,u' Rev. 3, 5/24/99 - Page 21 -
- a. Hot RCS fluid flows up shaft and is not cooled by thermal barrier.
L v55, n/ pld f 3 r ,p . - m ,-&Vc oo
- b. The seals are not designed for 500 deg F + water.
- c. Rubber "0Ring" deforms, causing loss of sealing and increased leakage.
- d. A volume of cool injection water exists in the RCP seal area which can sustain cooling for a few minutes.
- e. Maximum leakage is postulated to be 300 gpm per RCP.
(1) This is based on full delta P across the RCP labyrinth seal of 2200 psid.
- f. Tests on seal packages indicate that expected leakage is much less than the maximum, and on the order of 16 gpm per RCP.
- g. It is impossible to predict how long the seal will last on a loss of all seal cooling.
- h. New seal materials increase time to seal failure.
(1) Seal coating - silicon nitrate.
(2) New High Temperature 0-Rings
- 3. Ask the students what is the best indicator available of RCP seal response. (Answer: after action to isolate the RCS (ECA-0.0 step 3), PZR level is best indicator.)
- 4. Seal cooling restoration.
- a. Following restoration of AC power, it is desirable to TP-6 keep the RCP seal cooling isolated to prevent thermal shock of the RCP seals.
- b. The RCP should only be started if an extreme (red level) or severe (orange level) challenge to a Critical Safety Function is diagnosed via Status Tree monitoring and the operator is instructed to start an RCP in the associated Function Restoration Procedure.
Page 8 of 40
W
. --. RCS temperature reduces the thermal degradation o f materials and Reduciag thermal expansion effects that tend t o degrade the seal systm sealing capability and sealing: l i f e . Consequently, any actions t o reduce RCS pressure and temperature during a loss of a l l ac power event are consistent w i t h m i n i m i z i n g RCS inventory loss and maximizing time t o core .uncovery.
I I
F O lowing
~ the restoration o f ac power, the operator w i 11 have the capabi 1i t y I t o restore seal cooling by reestablishing seal injection flow o r 1 reestablishing t h e n a l barrier cooling using the component cooling water I system. Restoring seal cooling may have several benefits such as reducing I seal leakage and preventing further damage t o the seal components. However, I Westinghouse has not perfonned an analysis o f how the RCP seal package will I react as the seals cool, f i t s contract, the shaft moves, etc., possibly with I partially extruded O-rings. There may be a potential t o make seal leakage I worse by restoring seal cooling, depending on how i t is done. I c/ L I
The RCP Vendor Manual identifies l i m i t : for reestablishing seal cooling t o a 1 hot seal package t o prevent further damage due t o t h e m 1 shock and t o prevent I warping of the RCP shaft due t o uneven cooling. These limits are only I intended for a loss of seal cooling of short enough duration that the seal I package heatup i s limited. Although t h e limits have been extrapolated f o r an I extended loss of seal cooling event i n the past, they have not been validated I f o r such an event that i s beyond t h e design basis o f t h e RCP. Therefore, no I specific conclusions may be taken from the RCP vendor manual guidance f o r I teestabl i shi ng seal cool ing fol 1owi ng an extended 1oss o f seal cooling event. I The following provides a qualitative assessment that determines the most I appropriate method o f restoring seal cooling following an extended loss of a l l I ac power event: I I
To minimize the potential f o r thermal shock of the seals and shaft warping, 1 :
component cool!ng water can be established t o the thermal barrier heat I '
I-exchanger before seal injection is established. Note that since the loss o f I LJ' a l l ac power event i s beyond the design basis of the p l a n t , the perqonnanse of I t h e CCW system has never been analyzed under these conditions. Establishing I ECA-0.0 8 I trcPn
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 16 Group # 1 WA # W/E04EA1.2 OK Importance Rating 3.6 Ability to operate and / or monitor operating behavior characteristics of the facility as they apply to a LOCA Outside Containment Proposed Question: Common 15 Unit 3 is in MODE 4 cooling down on RHR with the following plant conditions:
0 RCS Temperature 340°F slowly lowering 0 RCS pressure 300 psig lowering 0 PZRlevel 42% lowering 0 CNMT temperature 100°F 0 R-27, Wide Range Plant Vent Gas Activity Monitor, went into ALARM SG levels 42% (31) 40% (32) 43% (33) 40% (34) 0 SG pressures 115 psig (31) 115 psig (32) 115 psig (33) 115 psig (34)
What event is taking place?
A. A steam leak has occurred inside CNMT B. The Cold Overpressure system has actuated.
C. A LOCA has occurred on the suction of the RHR pump.
D. Letdown line pressure control valve, PCV-135, has failed open.
Proposed Answer:
C. A LOCA has occurred on the suction of the RHR pump.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3SG-ILO-AOPRHR 1 (As available)
Question Source: Bank # INPO 19269 Modified Bank # (Note changes or attach parent)
New Question History: 10/20/2000 Braidwood 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X I O CFR Part 55 Content: 55.41 5,lO 55.43 5 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'4' Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#17 Group # 1 KIA # W/E11EA2.1 OK Importance Rating 3.4 Ability to determine and interpret the facility conditions and selection of appropriate procedures during abnormal and emergency operations as they apply to Loss of Emergency Coolant Recirculation Proposed Question: Common 16 During a LOCA, emergency coolant recirculation capability was lost, and ECA-1.I, Loss of Emergency Coolant Recirculation, is currently in progress. A RED path is identified on the CONTAINMENT status tree, and transition to FR-Z.1, Response to High Containment Pressure, is performed.
d What procedure should be used to operate the containment spray pumps, and why?
A. ECA-1.I, because it provides for REDUCED containment spray.
B. ECA-1.I, because an ECA should be completed prior to transferring to a Function Restoration Procedure.
C. FR-Z.l because it takes precedence over ECA-1.I.
D. FR-Z.1, because it provides for GREATER containment spray.
Proposed Answer:
A. ECA-1. I , because it provides for REDUCED containment spray.
Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): ECA-1.1, step 3 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOPFRz 7 (As available)
Question Source: Bank # INPO 22433 Modified Bank # (Note changes or attach parent)
New Question History: 10/1/2002 Diablo Canyon 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
/.
Number:
Title:
Revision Number:
FR-Z.l RESPONSE TO HIGH CONTAINMENT PRESSURE 10 i
i ACTION/EXPECTE
- 3. DETERMINE Which Procedure
/
SDrav:
a . CHECK E C A - 1 . 1 , LOSS OF a . GO To S t e p 4 .
EMERGENCY COOLANT RECIRCULATION - rti EFFECT b . GO To S t e p 5 Page 4 of 9
/
Number:
Title:
R e v i s i o n Number: -
F R - Z ,1 RESPONSE TO HIGH CONTAINMENT PRESSURE 10 RESPONSE NOT OBTAINED
- 4. DETERMINE I f Containment S o r a v Is Reauired:
a . CHECK c o n t a i n m e n t p r e s s u r e a . RETURN To P r o c e d u r e and
- HAS I N C R E A S E D TO GREATER Step i n e f f e c t .
THAN 22 P S I G b . C H E C K s p r a y system - b. aligned for A L I G N E D FOR INJECTION r e c i r c u l a t i o n , THEN PERFORM t h e f o l l owing:
- 1) ENSURE a t l e a s t one spray r e c i r c stop valve i s open:
889A 8898 2 ) GO To S t e p 4 . f .
I :,
L c . CHECK s p r a y pumps d i s c h a r g e c. Manually OPEN v a l v e ( s 1 .
v a l v e s - OPEN 866A 8666 d . CHECK s p r a y a d d i t i o n t a n k d . E NaOH a d d i t i o n i s d i s c h a r g e v a l v e s - OPEN d e s i r e d , THEN OPEN v a l v e ( s ) .
0 876A 0 8766
~ _ _
e . CHECK s p r a y pumps - R U N N I N G e. START s p r a y pump(s).
~
f . CHECK c o n t a i n m e n t i s o l a t i o n f . M a n u a l l y CLOSE v a l v e ( s 1 Phase B v a l v e s - CLOSED
- g. STOP a l l RCPs Page 5 o f 9
Number:
Title:
Revi s i on Number:
F R - Z .1 RESPONSE TO HIGH CONTAINMENT PRESSURE 10
,(-
RESPONSE NOT OBTAINED
- 5. V E R I F Y C o n t a i n m e n t FCU S t a t u s :
a . C H E C K FCUs - ALL R U N N I N G a . M a n u a l l y START FCU(s1.
b . PLACE FCU damper c o n t r o l s w i t c h e s i n - INCIDENT MODE position c . CHECK FCU dampers f o r a l l FCUS - I N INCIDENT MODE POSITION Dampers A / B - CLOSED (in l e t dampers 1 0 Damper C - CLOSED
( b y p a s s damper) 0 Damper D - OPEN
( f i l t e r o u t l e t damper) d . PLACE c o n t r o l s w i t c h e s f o r 1104 AND 1105 t o OPEN
- e. CHECK S e r v Wtr Cont C l g v a l v e s - OPEN 0 1104 0 1105 Page 6 o f 9
i'
.J . .. - --
- -- - STFP DESCRIPTION TABLE -
- FOR FR 7 . 1 Step 3 4
- CAUT70N mUTIOF(: If ECA-1.1;- LOSS OF EMERGENCY COOLANT RECIRCULATION, i s i n effect, containment spray should be operated as directed i n ECA-1.1 rather than step 3 below.
PURPOSF: To ensure containment spray pumps are operated as directed in ECA-1 . 1 instead o f t h i s guideline, i f ECA-1.1 i s i n effect This caution warns the operator that the operation o f the containment spray pumps indicated i n guideline ECA-1.1 takes precedence over that noted in Step 3 of this guideline. This guideline specifies maximum available heat removal system operabi I ity in order to reduce containment pressure. Gui del i ne ECA-I. 1 uses a less restrictive criteria, which pennits reduced spray pump operation depending on RMST level, containment pressure and number of emergency fan coolers operating. The less restrictive criteria for containment spray operation is used in guideline ECA-1.1 since recirculation flow to the RCS is 1
not available and it i s very important t o conserve'RWST water, if possible, by
-1 s toppi ng containment spray pumps. --.- ..
ACTIONS :
Determine i f ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION, i s i n e f f e c t JNSTRUMFNTATXON:
CONTROI fF0UI PMFNT :
N/A KNOWLFDGE:
N/A PI ANT-SPFCIFIC INFORMATION:
N/A
>4
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'LJ Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#18 Group # 1 KIA # W/E05G2.4.20 OK Importance Rating 3.3 Knowledge of operational implications of EOP warnings, cautions and notes Proposed Question: Common 17 A Unit 3 Reactor Trip occurred after a 200 day continuous run at 100% power.
Following the trip, all AFW flow was lost and the Team transitioned to FR-H.1, Loss of Secondary Heat Sink. Due to distractions caused by a pressure channel failure, bleed and feed steps were not initiated until WR S/G levels were all 4 0 % .
Which one of the following correctly describes the general consequence of the delay?
A. Core uncovery will be more severe only if the PRT rupture disk fails, increasing the loss of mass, while ECCS flow is limited by RCS pressure.
B. Core uncovery will NOT occur as long as one PZR PORV is open and one charging pump is injecting prior to SG dryout.
C. Core uncovery will NOT occur as long as both PZR PORVs are open and two charging pumps are injecting prior to SG dryout.
D. Core uncovery will be more severe because RCS pressure will remain at a higher value for a longer time, limiting ECCS flow.
Proposed Answer:
D. Core uncovery will be more severe because RCS pressure will remain at a higher value for a longer time, limiting ECCS flow.
Explanation (Optional):
.W' Technical Reference(s): FR-H. 1 Bases (Attach if not previously
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet.
provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOPFRH 2 (As available)
Question Source: Bank #
Modified Bank ## (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
- 2. DESCRIPTION 4
A loss o f secondary heat sink can occur as a result o f several different initiating events. Po-ssibilities are a. l o s s o f main feedwater during power operation, a loss of offsite power, or any other scenario for which main feedwater i s isolated or lost when the steam generators provide the main heat removal path. For these initiating transients a failure of the auxiliary feedwater (AW) system to inject or a loss of AFW early in the cooldown, before RHR System operation can be established, could lead to a loss of secondary heat sink.
A loss of all feedwater transient is characterized by a depletion o f secondary inventory and eventual degradation o f secondary heat transfer capability. As secondary heat transfer capability degrades, a loss of secondary heat sink results and core decay heat generation will increase RCS temperature and pressure unti 1 the pressurizer power operated re1 ief valves (PORV) or pressurizer safety valves open to relieve the increasing RCS pressure. At d
in a loss of RCS i?ventory similar in-natbre t h i s point the opening and closing o f the PORVs or safety valves will result to a small break l o s s of coolant accident. If operator action is not taken, the pressurizer PORVs or safety valves will continue to cycle open and closed at the valve setpoint pressure removing RCS inventory and a limited amount of core decay heat until eventually enough inventory will be lost to result i n core uncovery..
The plant status upon entering this guideline will be a function of the initiating event. If the initiating event is a loss of main feedwater during power operation with AFW flow unavailable, or from any other anticipated transient resulting in reactor trip and main feedwater isolation or failure with AFW flow unavailable, the transient may not result in an automatic SI actuation. If the initiating event has resulted in a reactor trip due to primary depressurization (i . e . , small LOCA, secondary break or steam generator tube rupture) with AFW flow unavailable, then SI should have been automatically -initiated. However, the status o f SI upon entering the guideline i s not important to the actions that will be taken. Should it become necessary to establish a bleed and feed heat removal path (actuating SI
FR-H.
1 in n -
- STEP DESCRIPTION TABLE FOR FA lid L Step 2 - CAUTION 1 KNOWLEDGE:
0 The importance of e s t a b l i s h i n g bleed and feed as an a l t e r n a t i v e heat sink t o prevent core uncovery and inadequate core cooling.
0 If PORV block valves a r e closed, they should be opened a t t h i s time unless they are closed t o i s o l a t e a f a u l t y PORV.
o When t h e RCPs a r e stopped due t o l o s s of heat s i n k , RCS pressure and temperature are expected t o increase s l i g h t l y and s t a b i l i z e below the PRZR PORV setpoint. RCS pressure and temperature will continue t o be r e l a t i v e l y constant u n t i l SG dryout occurs (approximately 20 - 30 minutes) . A t this p o i n t , the primary-to-secondary heat t r a n s f e r rate degrades and the RCS begins t o heat up and repressurize and will eventually r e s u l t i n the opening o f the P U R PORVs.
This should not be confused w i t h the onset of natural c i r c u l a t i o n i n w h i c h the RCS pressure continues t o increase a f t e r the RCPs are stopped and may reach the PUR PORV setpoint. The key t o determining i f the RCS pressure rise i s due t o loss of heat s i n k or natural c i r c u l a t i o n i s the 1oop del ta-T. The loop delta-T i s expected t o be large for natural c i r c u l a t i o n and small f o r a l o s s of heat s i n k since there i s no heat
.4 t r a n s f e r t o the secondary. -
Therefore, veri f y i ng a sl owly i ncreasi ng RCS pressure and temperature trend p l u s a large loop delta-T prior t o the PORV opening confirms natural c i r c u l a t i o n whereas a r e l a t i v e l y s t a b l e temperature and pressure and a small loop delta-T combined w i t h SG wide range low level p r i o r t o t h e PORV opening confirms a l o s s of heat s i n k .
PLANT-SPFCIFIC IN FORMATION:
o (X.01) Parameter and s e t p o i n t f o r diagnosing l o s s of secondary heat sink, I including a1 lowances for normal channel accuracy. Refer t o Background Document f o r guideline FR-H.l.
o (X.02) Parameter and s e t p o i n t f o r diagnosing loss o f secondary heat s i n k , l i ncl u d i ng a1 1owances f o r normal channel accuracy and post accident t r a n s m i t t e r errors. Refer t o Background Document f o r guideline FR-H.l.
o Parameters (X.01) and (X.02) a r e described in subsection 2.2.4, Pl ant -SDecif i C - SYTl)DtOmS f o r l o s s of Heat S ink, of t h i s background document. Parameters and setpoint; i n this CAUTION should be consistent w i t h S t e p 8 .
FR-H. 1 66 LP-Rev. 1C LFRHl
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
.4 Examination Outline Cross-reference: Level RO SRO Tier # 1 WS#19 Group # 2 WA # 000001AK1.06 OK Importance Rating 4.0 Knowledge of the operational implications of the relationship of reactivity and reactor power to rod movement as they apply to the continuous rod withdrawal Proposed Question: Common 18 With Unit 3 operating at 88% power, the following symptoms occur:
Reactor power INCREASING.
Tave GREATER THAN Tref.
Pressurizer Pressure INCREASING.
0 Pressurizer Level INCREASING.
4 Which ONE of the following would cause the above symptoms to occur INITIALLY?
A. First Stage Turbine Pressure transmitter, PT-412A, Failed LOW.
B. Power range channel N-43 fails high.
C. Failed OPEN SG safety valve.
D. Uncontrolled rod withdrawal.
Proposed Answer:
D. Uncontrolled rod withdrawal.
Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): (Attach if not previously u provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-AOPROD 3 (As available)
Question Source: Bank # INPO 20764 Modified Bank # (Note changes or attach parent)
New Question History: 10/29/2001 Braidwood 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X I 0 CFR Part 55 Content: 55.41 8,lO u
55.43 6 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d Examination Outline Cross-reference: Level RO SRO Tier # 1 ws # 20 Group # 2 KIA # 000028AK2.02 OK Importance Rating 2.6 Knowledge of the interrelations between the Pressurizer Level Control Malfunctions and sensors and detectors Proposed Question: Common 19 The plant is operating at 100% power with all control systems operating normally. The controlling channel, LT-460, reference leg of Pressurizer Level has just developed a leak where the reference leg connects to the D/P cell. Which one of the following best describes the immediate plant response from this leak?
A. LT-460 - indication will decrease, LT-459 indication will INCREASE, LT-461 -
indication will INCREASE, charging flow will INCREASE.
B. LT-460 - indication will INCREASE, LT-459 indication will DECREASE, LT-461 -
indication will DECREASE, charging flow will DECREASE.
C. LT-460 - indication will INCREASE, LT-459 indication will decrease, LT-461 indication will DECREASE, backup heaters will de-energize.
D. LT-460 - indication will DECREASE, LT-459 indication will DECREASE, LT-461 indication will DECREASE, backup heaters will energize.
Proposed Answer:
B. LT-460 - indication will INCREASE, LT-459 indication will DECREASE, LT-461 -
indication will DECREASE, charging flow will DECREASE.
Explanation (Optional):
9 Technical Reference(s): SD- 1.4 (Attach if not previously
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-ICPZLV E-5 (As available)
Question Source: Bank # INPO 246 12 Modified Bank # (Note changes or attach parent)
New Question History: 5/30/2003 Seabrook 1 Question Cognitive Level: Memorv or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
System Description 1.4 Pressurizer & Pressurizer Relief Tank Figure 1.4-20:- . -Ptessunhr Level Block Diagram (LOG-13A)
PRESSURIZER LEVEL BLOCK DIAGRAM PRE8BuR(zER Hn+lLE'YEI-LOW CHUKI(N0 FLOWAUWM Rev. 4, 12/14/2005 - Page 99 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'd Examination Outline Cross-reference: Level RO SRO Tier # 1 ws # 21 Group # 2 KIA # 000033AK1.01 OK Importance Rating 2.7 Knowledge of the operational implications of the effects of voltage changes on performance as they apply to Loss of Intermediate Range Nuclear Instrumentation Proposed Question: Common 20 Unit 3 is performing a plant shutdown. Power is 3% when alarm "Intermediate Range N36 Loss of Compensate Voltage," comes in.
How does this affect the Nuclear Instrumentation?
A. N36 reading would immediately drop about 1 decade. During the subsequent shutdown, SR Nls will energize automatically when N35 drops below P-6
'4 setpoint.
B. N36 reading would immediately rise about 1 decade. During the subsequent shutdown, SR Nls will NOT energize automatically because N36 reading will remain above the P-6 setpoint.
C. N36 reading would NOT immediately change. During the subsequent shutdown, SR Nls will NOT energize automatically because N36 reading will remain above the P-6 setpoint.
D. N36 reading would NOT immediately change. During the subsequent shutdown, SR Nls will energize automatically when N35 drops below P-6 setpoint.
Proposed Answer:
C. N36 reading would NOT immediately change. During the subsequent shutdown, SR Nls will NOT energize automatically because N36 reading will remain above the P-6 setpoint.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
Technical Reference(s): 3-AOP-NI- 1 (Attach if not previously SD-13 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-ICEXC E-3 (As available)
Question Source: Bank # rNP0 26088 Modified Bank # (Note changes or attach parent)
New Question History: 9/1/2003 Prairie Island 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8,lO 55.43 Comments:
Nuclear Instrument Failure 3-AOP-NI-I Rev. 1 Page 11 of 25
.W I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 4.13 -Has a failure in Intermediate - GO TO Step 4.19.
Range Nls occurred?
4 . 1 4 -1s remaining Intermediate Range - GO TO Step 4.16.
channel inoperable?
4.1 5 -Do at least 2 of 4 Power Range - IF NOT already being met, channels indicate > lo%? THEN meet requirements of TS 3.3.1.
CAUTION Failure of either Intermediate Range channel in the high direction will block automatic reenergizing of Source Range High Voltage. Simultaneous operation of both Intermediate Range Permissive Defeat Pushbuttons (when remaining Intermediate Range channel indicates < 3.6E-1 I) will be required to obtain Source Range operation.
4.1 6 -INITIATE applicable section of SOP-Nl-001 (Excore Nuclear Instrumentation System Operation) to remove affected channel from
, service.
',- 4.17 Notify the following of status of NI failures:
-Site Operations Manager
-Unit Operations Manager
-Reactor Engineer
. I ..
4.18 -RETURN to procedure and step in I
effect.
13.0 Excore Nuclear Instrumentation the inner and outer cylinders), the neutrons react with the boron causing ionization.
5 B 1 0 + 0 N 1 ~ ( 5 B 1 1 ) * ~ 3 L i 7 + + + + 2 a 4 + ++energy The lithium and alpha particle resulting from this reaction cause secondary ionization in the outer can. The electrons produced by the ionization are collected on the outer can wall. This produces a signal that is proportional to the neutron flux. Electrons are also collected on the outer can wall from the gamma radiation, which interacts with the outer gas volume. This additional signal is proportional to the gamma flux and is additive to the neutron flux signal. The outer can operates in the Ionization Region; thus, all the charged particles produced in the initial ionizing events are collected on the electrodes.
In the inner can, the gamma flux also reacts with the N, gas, producing a signal proportional to the gamma radiation. The inner can is operated in the Recombination Region to permit adjustment of the output current by varying its applied voltage. The inner can voltage is called the compensating voltage. If the compensation voltage is set properly, the outer can signal due to gamma plus neutron flux, will interact with the inner can gamma flux only signal. The gamma signals cancel out leaving the neutron only signal which is then amplified before it is displayed on the meter or sent to the protection f and control circuitry.
2.10.2 Gamma Compensation in the Intermediate Range It is necessary to define the term compensation and the effects of under-compensation and over-compensation to clearly understand the process of neutron detection in the intermediate range.
Compensation is a term applied to the negative voltage signal applied to the inner can of the CIC which cancels or compensates for the current signal produced by the gamma radiation interacting within the outer can of the detector. This becomes very important to the operator because an incorrect setting of compensating voltage, i.e. over-compensation or under-compensation, would cause an erroneous neutron level indication on the meters, as shown on Figure 13-22.
Over-compensation occurs when the compensation voltage is set to high. This results in a higher current due to gamma flux in the inner can than is being generated in the outer can due to the same gamma flux. The results of this mismatch is that part of the current due to the neutron flux is also cancelled, causing the indicated current level to be less than actual.
Rev. 4, 08/03/2004 - Page 41 -
13.0 Excore Nuclear Instrumentation Under-compensation occurs when the compensation voltage is less than that required. The current due to gamma from the inner can is now smaller than the current due to gamma from the outer can. This allows some current due to gamma in the outer can to remain and add to the neutron flux resulting in increased output current from the detector causing it to indicate above actual levels.
To obtain this true neutron-only signal, the two opposing g a m a signals must be cancelled exactly. Since it is physically impossible for both the inner and outer cans to be manufactured identically sensitive to the gamma flux present under all operating conditions, the problem of how to ensure exact compensation arises. By grooving the inner electrode and applying a variable negative voltage, the size of the inner can is adjusted electrically. The inner can of the CIC operates in the recombination region of the detector characteristic curve and, by adjusting the compensating voltage, only a fraction of the total ionization is collected.
The IR drawer monitors reactor power over a range of eight decades between lo-*and 10 ion chamber amperes. Indications of level and startup rate (SUR)are provided at the NIS cabinets, and on panel FCF.
Because neutron events are occurring at a high rate, no signal conditioning is necessary prior to the log current amplifiers. A block diagram of the intermediate range in provided as Figure 13-23.
2.10.3 Log Current Amplifier This assembly receives current from the detector in the range between lo- and lo5 amperes. The assembly provides a logarithmic voltage output, 0 to 10 VDC, proportional to a linear input current. With the use of the log amplifier, the wide range current input is compressed logarithmically to a usable voltage suitable for metering and the generation of trip signals. (Figure 13-23 provides a block diagram of the intermediate range.) The output from the log amplifier is simultaneously coupled to an isolation amplifier and four bistable relay drivers. The output is also displayed on the neutron level meter calibrated in amperes between lo- and 10-3.amps.
Internal switches and potentiometers are provided for setting and adjusting the log current amplifiers. Both fixed and variable signals can be injected into the log amplifiers for testing and calibration purposes. This is accomplished by the use of switches located on the front panel of the drawer and a calibrate module located inside of the IR drawer assembly.
Wl Rev. 4,08/03/2004 - Page 42 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'4 Examination Outline Cross-reference: Level RO SRO Tier # 1 ws # 22 Group # 2 K/A # 000037AA2.12 OK Importance Rating 3.3 Ability to determine and interpret the flow rate of leak as it applies to the Steam Generator Tube Leak Proposed Question: Common 21 A SG tube leak is in progress. Plant conditions just before the leak were steady state with no evolutions in progress. Some time later, the following conditions exist:
CVCS charging flow rate = 63 gpm 0 CVCS letdown flow = 75 gpm 0 Total RCP seal injection = 32 gpm 0 Total RCP seal leakoff flow = 12 gpm
'4 RCS temperature at no load Tave and steady PZR Press and Level are stable Based on the above indications, what is the approximate RCS SG leak rate?
A. 1 gpm B. 8gpm C. 20gpm D. 28gpm Proposed Answer:
B. 8gpm
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
X d '
Technical Reference(s): SD-3.O (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-AOPSGI A (As available)
Question Source: Bank # INPO 20216 Modified Bank # (Note changes or attach parent)
New Question History: 9/10/2001 Cook 1 Question Cognitive Level: Memory or Fundamental Knowledge d
Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'4 Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 23 Group # 2 KIA # 000051G2.2.12 OK Importance Rating 3.O Knowledge of the surveillance procedures Proposed Question: Common 22 Unit 3 is running 3-PT-V089, Online Turbine Mechanical Trip Features Test.
The "Test Handle" has just been placed in the TEST position in preparation for doing the Low Vacuum Trip Test (Simulated).
The control room reports that actual condenser vacuum has dropped to the turbine trip setpoint.
'W' With no operator action, which one of the following will occur?
A. The turbine low vacuum trip device will NOT actuate, and the turbine will NOT trip.
B. The turbine low vacuum trip device will actuate, and the turbine will trip.
C. The turbine low vacuum trip device will NOT actuate, but the turbine will trip.
D. The turbine low vacuum trip device will actuate, but the turbine will NOT trip.
Proposed Answer:
D. The turbine low vacuum trip device will actuate, but the turbine will NOT trip.
Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): I3LP-ILO-MTGOO1 Page 30,3 1 (Attach if not previously d' provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-MTG001 4 (1583) (As available)
Question Source: Bank # INPO 2251 8 Modified Bank # (Note changes or attach parent)
New Question History: 10/1/2002 Diablo Canyon 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.4 1 4,lO 55.43 Comments:
(/asAJ3
- 4. Hydraulic inducqd trip dumps high pressure autostop oil pressure off of the overspeed trip valves by operation of a pivot plate arrangement
- a. 20 AST solenoid trip
- b. Low bearing oil pressure trip
- c. Low vacuum trip
- d. Thrust bearing wear trip
- e. In addition, a manual trip from the front standard and the mechanical overspeed trip device also cause a hydraulic trip through the overspeed trip device
- f. Once HP autostop oil dumps, low pressure autostop oil dumps and the trip process continues the same as for a solenoid trip
- 5. Autostop oil header is monitored by three pressure switches Logic Diagram (63/AST 2,3,4). They are arranged into three redundant 565D172 Sht 2 2/3 logic circuits
- a. Two 2/3 logics are used to trip the reactor above P-8
&35% Rx Pwr)
- b. One 2/3 logic trips the generator upon initiation of a turbine trip
- c. If 1/3 pressure switches senses AST oil pressure
<45 psig, alarms at panel FAF (normal pressure 120 P m
- 6. Low bearing oil pressure trip TP-14
- a. Prevents overheating and possible bearing damage
- b. Pressure of 6.5 psig at No. 1 bearing initiates trip
- c. Pressure switch (63LBO) in the bearing supply actuates the low bearing pressure alarm on panel FAF
- 7. Low vacuum trip E.0.4 d
I3LP-EO-MTGOO 1.DOC Page 30 of 40
~
- a. Prevents excessive heating of low pressure turbine i exhaust hood and last row of turbine blades
- b. Turbine is tripped if condenser vacuum falls below 18" mercury
- c. Trip feature may be bypassed locally during startup and automatically engages when vacuum reaches
>18 inches mercury or >3psig back pressure
- d. Pressure switch 63LV provides first out alarm at panel FAF
- e. A special drip leg (trap) on the vacuum sensing line keeps any oil leakage from being drawn into the main condenser
- 8. Thrust bearing wear trip
- a. It protects the turbine from mechanical damage when excessive thrust bearing wear on shoes or excessive axial movement is detected
- b. One detector (1) Located on governor end and senses thrust in both directions using a back press signal (2) Setpoint 43 psig, alarms on panel FAF as sensed by PS-63/TB
- 9. Manual Trip
- a. Lever located on front standard to manually dump autostop oil
- b. Button on CR flight panel opens both solenoid valves to dump autostop oil
- c. Either manual trip actuates the first out MANUAL TURBINE TRIP alarm at panel FAF
- 10. Generator Trip
- a. Prevents excessive overspeed conditions
.d I3LP-LO-MTG001.DOC Page 31 of 40
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 24 Group # 2 K/A # 000068AA2.06 OK Importance Rating 4.1 Ability to determine and interpret RCS pressure as it applies to the Control Room Evacuation Proposed Question: Common 23 The control room has been evacuated due to a fire with heavy smoke. Offsite power has been lost. All three EDGs are supplying their respective buses. Which of the following describes the primary method that will be used to control RCS pressure?
A. Manually opening and closing one Pressurizer PORV.
B. Adjusting 31 Charging Pump speed to control level and thus pressure.
C. Energizing and de-energizing 31 Pressurizer backup heaters.
D. Local operation of Auxiliary Spray.
Proposed Answer:
C. Energizing and de-energizing 31 Pressurizer backup heaters.
Explanation (Optional):
Technical Reference(s): 3-AOP-SSD-1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet W' Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5,lO 55.43 5 Comments:
Control Room Inaccessibility Safe u35-q 3-AOP-SSD-1 Rev 05 Shutdown Control Page 23 of 191 P
L../
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 4.27 -Are an^ Containment Spray - GO TO Step 4.29.
Pump(s) running?
4.28 -Remove control power fuses 1.- Secure associated 480V Bus(es) by open 480V breakers for running opening bus supply breaker or shutting Containment Spray Pumps. down associated EDG.
2.IF pump(s) are still running, THEN dispatch NPO to close the following valves: (PAB 68ft., Piping Pen Area):
- SI-869A (31 SPRAY PUMP DISCHARGE LINE ISOLATION)
- SI-869B (32 SPRAY PUMP DISCHARGE LINE ISOLATION) 4.29 -Are any RHR pumps running that - GO TO Step 4.31.
were not previously running?
4.30 -Remove control power fuses and - Secure associated 480V Bus(es) by P open 480V breaker for running RHR Pumps.
opening bus supply breaker or shutting down associated EDG.
W 4.31 -Are any HHSl pumps running? - GO TO Step 4.34.
4.32 -Is PRZR level > 5% and - GO TO Step 4.34.
- Secure associated 480V Bus(es) by open 480V breaker for running opening bus supply breaker or HHSl Pumps. shutting down associated EDG.
on essential header? (CB, safeguards buses energized? SOP-EL-1 (Diesel Generator Operation) to r\ -2N3A attempt power restoration to at least i
-.-J - 5A -
two safeguards buses.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
.4 Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 25 Group # 2 WA # W/E14AK3.01 OK Importance Rating 3.8 Knowledge of the reasons for guidance contained in EOP for loss of containment integrity Proposed Question: Common 24 Given the following plant conditions:
0 The Unit has experienced a fault on 31 Steam Generator inside containment.
0 The crew has transitioned from E-0 to E-2, Faulted Steam Generator Isolation.
0 Containment pressure is currently at 28 psig and slowly rising.
0 Both Containment Spray Pumps are NOT operating.
~4 Which ONE of the following indicates the correct action for the crew to take?
A. Continue in E-2, Faulted Steam Generator Isolation and transition to FR-2.1, Response To High Containment Pressure if containment pressure exceeds 46 psig.
B. Continue in E-2, then use ES-0.0, Rediagnosis (if necessary) to transition to the correct procedure C. Immediately transition to FR-2.1, Response To High Containment Pressure.
D. Go to E-0, Reactor Trip or Safety Injection and revalidate SI automatic actions to ensure Containment Fan Cooler Units are operating properly.
Proposed Answer:
C. Immediately transition to FR-2.1, Response To High Containment Pressure.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
Technical Reference(s) : F-0.5 Containment Status Tree (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-EOPFRZ 8 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,lO
~
55.43 5 Comments:
Number: iltle:
Revision F-0.5 CON TA I NMENT 8 CONTAINMENT PRESSURE THAN 47 PSlG CONTAINMENT NO PRESSURE LESS THAN 22 PSlG y~s ORANG GO TO
$Jf@# mm FR-z.2 CONTAINMENT e""
LML NO LESS THAN 49 FEET YES 8 INCHES YELLOW 0 0 0 0 0
I I I 0 I CONTAINMENT NO RADlAllON LESS THAN 3 R/HR YES CSF SAT L-----yLJ
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d,
Examination Outline Cross-reference: Level RO SRO Tier # 1 WS ## 26 Group # 2 WA # 000076AA1.04 OK Importance Rating 3.2 Ability to operate and / or monitor the failed fuel-monitoring equipment as they apply to the High Reactor Coolant Activity Proposed Question: Common 25 In accordance with the abnormal operating procedure 3-AOP-HIACT-I , High Activity, what should the operators do once Chemistry verifies the high activity condition?
A. Place Excess letdown in service in addition to normal letdown.
6 . Divert letdown to CVCS HUT and maximize makeup.
W C. Remove cation demineralizer from service.
D. Maximize letdown flow.
Proposed Answer:
D. Maximize letdown flow.
Explanation (Optional):
Technical Reference(s): 3-AOP-HIACT-1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Learning Objective: 13LP-ILO-AOPACT 3 (As available)
L/
Question Source: Bank # INPO 24663 Modified Bank # (Note changes or attach parent)
New Question History: 5/30/2003 Seabrook 1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Cornments:
High RCS Activity AOP-HIACT-1 Page 7 of 17 i 4. SUBSEQUENT ACTIONS ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I
CAUTION 0 Large or sudden changes in RCS temperature or flow could cause a crud burst.
If RCS activity is 2 TO 25% OFTS 3.4.16 limits, the plant is required to be in MODE 3 with Tave 500°F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
4.1 -IAAT RCS activity is 2 TO 25% OF TS 3.4.1 6 limits AND reactor is in MODE 1 or MODE 2, THEN INITIATE plant shutdown according to applicable POP.
4.2 -MAT the following conditions exist:
- RCS activity 2 TO 25% OFTS 3.4.1 6 limits i - Reactor is in MODE 3
- Tave 2 500°F THEN INITIATE POP-3.3 (Plant Cooldown - Hot To Cold Shutdown) to reduce Tave 500°F.
NOTE If activity on R-63 is > 5 yCi/cc but 50 pCi/cc, then HP should assist Chemistry in primary sampling.
If activity on R-63 is > 50 pCi/cc, then consideration should be given to using the Post-Accident Sampling Systern.
4.3 -Notify Chemistry to periodically sample for RCS activity trend resuIts.
4.4 -Is letdown aligned through CVCS - PERFORM applicable section of demineralizers? SOP-CVCS-004 (Placing the CVCS Demineralizers In or Out of Service) to place mixed bed demin in service.
4.5 -INITIATE applicable section of s0P-cvcs-002(Charging, Seal I \J Water, and Letdown Control) to increase letdown flow up to a maximum of 120 gpm.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d
Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 27 Group # 2 WA # WlE08G2.1.23 OK Importance Rating 3.9 Ability to perform specific system and integrated plant procedures during all modes of plant operation Proposed Question: Common 26 On the current outage schedule, the RCS cooldown (from 340°F to <2OO0F)is supposed to occur in the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. You have just finished placing RHR in service for cooldown but during this evolution you determined that 31 RHR HX had a tube leak. 31 RHR HX is currently isolated. How will this impact the RCS cooldown on the outage schedule?
A. The cooldown will be completed on schedule because the RHR system has two 100% redundant trains.
- 8. The RCS cooldown will be completed but it will take longer but not more than twice as long as scheduled.
C. The cooldown will be completed on schedule because the SGs will do most of the cooling until RCS temperature is below 212°F.
D. The RCS cooldown can not be completed until decay heat level drops below the capacity of the single RHR train.
Proposed Answer:
B. The RCS cooldown will be completed but it will take longer but not more than twice as long as scheduled.
Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-RHR001 6 (As available)
Question Source: Bank # INPO 26098 Modified Bank # (Note changes or attach parent)
New Question History: 9/1/2003 Prairie Island 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet i_/
Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 28 Group ## 1 WA # 003A4.07 OK Importance Rating 2.6 Ability to manually operate and / or monitor in the control room RCP seal bypass Proposed Question: Common 27 Given the following conditions:
Plant cooldown is in progress.
RCS temperature is 220°F.
RCS pressure is 375 psig.
VCT pressure is 25 psig.
RCPs are operating with all No.1 Seal Leakoff valves open.
All RCP seal injection flows are 8 gpm RCP seal discharge valves 261A-D are open 32 RCP # I seal leakoff flow indicates 0.8 gpm and slowly decreasing.
32 RCP lower radial bearing temperature is 195°F and slowly rising Which ONE (1) of the following actions is required?
A. OPEN RCP Seal Bypass Valve, 246, to increase seal leakoff flow.
B. CLOSE HCV-142, Charging Line Flow Control Valve, to increase seal injection flow.
C. Trip operating RCPs and isolate seal leakoff due to insufficient seal DP.
D. Isolate # I seal leakoff for 32 RCP to increase #seal I DP.
Proposed Answer:
d A. OPEN RCP Seal Bypass Valve, 246, to increase seal leakoff flow.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
Technical Reference(s): 3-SOP-RCS-001 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-RCSRCP B (As available)
Question Source: Bank # rNP0 28048 Modified Bank # (Note changes or attach parent)
New Question History: 9/27/2004 Robinson 2 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
No:3-SOP-RCS-001 Rev: 34 REACTOR COOLANT PUMP OPERATION Page 17 of 41 r - -
u' 4.3.3 securing 34 RCP, THEN PLACE RC-PCV-455A1Pressurizer Spray
- Pressure Control Valve, in MANUAL and CLOSE.
MAINTAIN CCW flow to motor bearing oil coolers for at least 30 minutes
- 4.3.4 after shutdown.
MAINTAIN CCW flow to RCP thermal barrier until RCS temp is less than
- 4.3.5 150°F MAINTAIN seal injection between 6-12 gprn.
MAINTAIN RCP seal injection flow until RCS pressure is less than
- 4.3.6 100 psig.
4.3.7 LOG RCP shutdown in Unit Log.
4.4 Operation of RCP Seal Bvpass Valve 1
I Bypass line allows additional seal injection flow through pump bearing for cooling. I
~ __
4.4.1 RCP lower radial bearing temp, as monitored via RCP seal inlet temp r indicator TI-155, TI-154, TI-153, or TI-152 exceeds 190°F (Ref. 5.2.12)
OR No. 1 seal outlet temp indicator TI-148, TI-146, TI-132, or TI-125 d' exceeds 200°F,
-THEN:
- 4.4.1 .I VERIFY all of the following conditions are met:
- Continued operation of applicable RCP is required.
0
- 0 RCS pressure is greater than 100 psig and less than 1000 psig.
- RCP seal discharge valves are OPEN:
- 0 31 VLV261A
- 0 32 VLV 2616
- 0 33 VLV261C 0
34 VLV 261D
- 0 RCP seal return flow as indicated on FR-758, FR-159, r FR-156, or FR-157 is less than 1 gpm (Ref. 5.2.12).
d - 0 Seal injection flow rate to each pump is 6 to 12 gpm.
- 4.4.1.2 OPEN RCP Seal Bypass Valve No. 246.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 30 Group # 1 KIA # 004K4.01 OK Importance Rating 2.8 Knowledge of CVCS design feature and / or interlock which provide for oxygen control in the RCS Proposed Question: Common 28 Hydrogen is supplied to the Volume Control Tank (VCT) via an automatic pressure regulator.
This design feature of the CVCS system is provided to A. lower iodine levels in the RCS.
B. minimize oxygen in the RCS.
C. control the pH in the RCS.
D. maintain corrosion product solubility in the RCS.
Proposed Answer:
B. minimize oxygen in the RCS.
Explanation (Optional):
Technical Reference(s): SD 3.0 (Attach if not previously provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet L Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-CVCOOI 3.a. (As available)
Question Source: Bank # INPO 21 571 Modified Bank # (Note changes or attach parent)
New Question History: 7/17/2002 Braidwood 1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
d
GX .JJs 56 System Description 3.0 Chemical and Volume Control System downstream of TCV-149 and the demineralizers in order to ensure that whichever flowpath the letdown stream takes it must pass through the filter. Figure 3.0-1 shows the location of the filter. Physically, the reactor coolant filter is located in the Primary Auxiliary Building at the 73 foot elevation. Flow enters the filter near the top, passes down through filter cartridges and out near the bottom. The filter can be bypassed when needed for filter cartridge replacement. The filter cartridges are replaced when the pressure drop across them reaches approximately 20 psid or when the filter housing reads 5 to 10 Rem/hr on contact or in accordance with applicable RES department procedures. Vent and drain connections are provided for use in filter cartridge replacement.
2.4 Volume Control Tank (Figure 3.0-17)
Letdown flow continues to the VCT level control valve (LCV-112A) which directs the flow to the VCT or to the CVCS hold up tanks (Hut's). LCV-112A is normally aligned to the VCT, however, it automatically shifts to the Hut's on a high VCT level of 92% (reset 83%). Level control is discussed in detail in section 2.4.1. A normal-divert switch on the CVCS supervisory panel SFF in the control room allows the operator to manually divert flow to the HUT regardless of VCT level. Upon loss of air pressure or electrical power, LCV-112A fails to the VCT position.
The VCT provides an additional surge capacity for the reactor coolant system that is not accommodated by the pressurizer following load transients. It provides a convenient point for adding hydrogen to the coolant for RCS Oxygen control during operation and a means of degassing the RCS during shutdown and cooldown. The VCT also provides a suction head for the charging pumps and the backpressure for the Reactor Coolant Pump (RCP) seals. The VCT has a capacity of 3000 gallons and is located in the Primary Auxiliary Building at the 73 foot elevation.
The letdown flow enters the VCT through a spray nozzle located at the top of the tank. The spray nozzle enhances coolant hydrogen absorption by increasing the water to gas blanket contact area. The vapor space is predominantly hydrogen. It also provides a scrubbing action for the removal of fission product and other non-condensable gases. A hydrogen gas blanket is maintained on the VCT to control the RCS hydrogen concentration between 25 and 35 cc/kg. This ensures that sufficient hydrogen concentration is available for reaction with oxygen and subsequently assists with corrosion control. Adjustments in the hydrogen concentration are made by changing the setpoint of d
Rev. 6, 12/18/2002 - Page 13 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d
Examination Outline Cross-reference: Level RO SRO Tier # 2 WS#31 Group # 1 WA # 005K5.09
~
OK Importance Rating 3.2 Knowledge of the operational implications of dilution and boration considerations Proposed Question: Common 29 Given the following conditions:
The plant is being cooled down to 140°F for maintenance which will NOT require the RCS be opened. The crew is in the process of placing the first Residual Heat Removal (RHR) train in service for RCS cooling. Current RCS temperature is 345°F.
Current boron concentrations are as follows:
.LJ 0 RHR (train to be placed in service) boron 1020 ppm Required Shutdown Margin at 300°F boron 1750 ppm Required Shutdown Margin at 68°F boron 1800 ppm 0 RCS boron 2025 ppm 0 Refueling boron 2050 ppm Before the RHR train can be placed in service for RCS cooling, RHR boron concentration must be increased by a MINIMUM of ...
A. 730 B. 780 C. 1005 E. 1030
-J' Proposed Answer:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet B. 780 Explanation (Optional):
Technical Reference(s): Graph RCS-4A (Attach if not previously 3-SOP-RHR-001 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-RHR001 8.a. (I 170) (As available)
Question Source: Bank # INPO 27468 Modified Bank # 27468 (Note changes or attach parent)
New Question History: 3/24/2004 Harris I Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
4 No: 3-SOP-RHR-001 Rev: 35 RESIDUAL HEAT REMOVAL SYSTEM Page 4 of 47 I.1 This procedure establishes requirements for RHR System operation.
1.2 This procedure applies to the RHR System.
2.0 PRECAUTIONS AND LIMITATIONS 2.1 WHEN RCS pressure is greater than equal to 400 psis RCS temp is greater than OR equal to 350°F, THEN RHR SHALL NOT be in service (Ref 5.2.8).
2.2 RCS pressure increases to greater than or equal to 550 psig, THEN AC-MOV-730, RHR Loop Suction Isolation, and AC-MOV-731, RHR Loop Suction Isolation, will automatically close: (Ref 5.2.9) 0 -
IF AC-MOV-730 AC-MOV-731 close, THEN RCS pressure must be lowered to less than 450 psig to enable valve reopening.
e WHEN AC-MOV-730 AC-MOV-731 start to close, THEN valves can NOT be reopened until fully closed.
2.3 RCS and RHR boron concentrations SHALL be greater than or equal to minimum required concentration for shutdown per Graphs RCS4A and RCS4B, Minimum Required Boron for Shutdown.
2.4 WHEN RHR warmup is complete, THEN RHR flow SHALL be maintained as follows:
0 Total RHR flow (recirc + miniflow + core flow) greater than 1170 gpm to ensure minimum flow for continuous RHR pump operation
{Ref. S.l.'l}.
0 Core flow greater than 1000 gpm to ensure adequate RCS mixing 2.5 To prevent equipment damage, flow through a single RHR pump:(Ref 5.2.10) with 1 RHR HX in service flow SHALL NOT exceed 3000 gpm.
0 with 1 RHR Pump in service flow SHALL NOT exceed 4500 gpm.
No: 3-SOP-RHR-001 Rev: 35 RESIDUAL HEAT REMOVAL SYSTEM Page 15 of 47 L&
CAUTION 0 WHEN RHR pump flow is greater than 300 gpm AND less than 1170 gpm THEN RHR pump operation should be limited to 3 hrs or less in a 24-hr period (Ref 5.2.1 1) 0 RHR pump minimum flow SHALL be maintained greater than or equal to 100 gpm during startlstop pump operation, which is consider to be 30 min or less (Ref 5.2.1 1)
- 4.2.2.4 START desired RHR pump.
4.2.2.5 WAIT at least 10 min (Ref 5.2.16) to allow for proper mixing.
- 4.2.2.6 REQUEST Chemistry to sample boron concentration.
- 4.2.2.7 WHEN sample has been obtained, THEN STOP desired RHR pump.
4.2.2.8 -IF sample results indicate RHR loop boron concentration is equal to or greater than the 68°F (Ref 5.2.17) Cold Shutdown boron concentration, L - 4.2.2.9 THEN GO TO Step 4.2.3.
IF sample results indicate RHR loop boron concentration is less than RCS boron concentration AND RX Trip breakers are open, THEN:
REVIEW Unit Log to ensure motor starting requirements of 3-SOP-EL-O04A, Electric Motor Operation, will be met.
START RHR Pump.
ENSURE letdown pressure as indicated on PI-135, LP Letdown Press, is less than RHR pump discharge pressure as indicated locally on Pi-635, RHR Pump Discharge Pressure Indicator, by adjusting PCV-135, Low Pressure Letdown Line Backpressure Control Valve.
ENSURE a CVCS HUT is aligned to receive CVCS letdown per 3-SOP-CVCS-001, CVCS Holdup Tank Operation.
ENSURE CH-LCV-112A, VCT Inlet Diversion, is in DIVERT.
0
- 0 0
00 0
.0 o Fc 0
.8(D 0
0 0
v) 0 0
- 0 0
0 0
m 0
0 0
N 0
0 0
c-0
Questionld [ "'T
' ExamType ExamDate AbbrevLocName Harris 1 NSSSVendor -1 NSSSType Given the following conditions:
QuestionStem The plant is being cooled down to 140oF for maintenance which will NOT require the RCS be opened.
The crew is in the process of placing the first Residual Heat Removal (RHR) train in service for RCS cc Current boron concentrations are as follows:
RHR (train to be placed in service) boron1021 ppm Required Shutdown Margin boron1200 pprn RCS boron1341 pprn
- Cold Shutdown boron1750 ppm Refueling boron2261 ppm I Before the RHR train can be placed in service for RCS cooling, RHR boron concentration must be incrc MINIMUM of QuestionCommen n
CognitiveLevel ExamLevel Rematerial ParentQuestionld I
Answer
~d Distract1 320 ppm. Distractl Co Distract2 '729 ppm.
Distract3
f KaNumber
'L' KaSegmentl Y
KaSegment2 KaSegmenD loling.
KaSegment4 KaSegment5 KaRevision sased by a Reference Req'd (Y/N) i
'4
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
..d Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 32 Group # 1 WA # ~
006A1.18
~
OK Importance Rating 4.0 Ability to predict and I or monitor changes in parameters to prevent exceeding design limits associated with operating the ECCS controls including PZR level and pressure Proposed Question: Common 30 Given the following conditions:
0 A LOCA has occurred.
The crew is petforming actions of ES-1.2, Post LOCA Cooldown And Depressurization.
0 Pressurizer level is stable at 58%.
0 RCS pressure is stable at 1280 psig.
L/
0 The CRS determines that a SI pump can be stopped in accordance with ES-1.2.
When the RO stops the SI Pump, which one of the following describes the Pressurizer level response?
A. PRZR level will remain at its current value.
B. PRZR level will rise until charging is realigned to the VCT.
C. PRZR level will drop until normal charging and letdown are restored D. PRZR level will drop until RCS pressure stabilizes at a lower value, then will stabilize.
Proposed Answer:
D. PRZR level will drop until RCS pressure stabilizes at a lower value, then will
\e stabilize.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d Explanation (Optional):
Technical Reference(s): ES-1.2, Bases (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOPEI0 6 (As available)
Question Source: Bank # TNPO 24940 Modified Bank # (Note changes or attach parent)
New Question History: 12/1/2002 Beaver Valley 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
As noted in"Tab1e 2 , _-the subcooling criteria f o r stopping t h e high-head Si
'J . -- - .--
pumps are given i n two ways. For the case "w/o throttling," maxirnum'charging flow from the two charging pumps i s assumed at a l l times, before and after the SI f l o w reduction. For the case "with throttling," the charging f l o w is adjusted and t h e following sequence of actions i s performed m e n one high-nead S I oump i s stopoad:
- 1) Reduce charging flow to a minimum (zero) while maintaining a: leasc a minimum subcooling (errors).
- 2) A l l o w RCS subcooling to increase to the required criterion for stopping the n e x t SI pump. The subcooling will increase a s the RCS coolaown continuer.
- 3) Stop SI pump and immediately increase charging f l o w to a maximum.
5y throttling tire cnarging flow i n t h i s manner, the injection f l o w reduction wnen s:cFgin$ a n Si pump w i l l be smaller than the f l o w reduction for the
" w j o tnrot:linS" case. The subcooling criterion is a l s o reduced when
- ?rotcling i s used.
1:: t n transient
~ a n a i y s i s presented nore, the f i r s t (and second) hioh-hea2 SI dj.,,r
- . I * - 'r*zs j t S p D 2 6 birnou; ihrottlinS tne c n a r g i n g f l o w , consistent with the way
- ?e
-c . n z 3 - 1 .;uicisline
~ i s srructured. Since cnarging f l o w i s a small percencage
" - ; n e :a=?:
? - i c j e z z i o n f l o w , tne subcooling c r i t e r i a with and w i t h o u t
-_m. .y.-,,,:ins
--, are no: significantly different, 5 5 O F v - r s u s 45'; for stopping t h e
. .. . c c - n i s h - n e a t SI pump.
A: t= 62:3C, 2C.S subcooling had increased to 5 5 O F and PRZR level was 46% and
- ncre?;inS. Tne f i r s t high-head S I Dump was then stopDed (Step 11). Over tire n e x t -:ve minuzes, RCS D r e s s w e aecreased (from 1220 p s i g ) and stabilized a t
_"q . .lYnC a s j s . P E R level cecressed to 4C% and continued tc decrease slowly as
- ?e csolcawr: concinued.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet u
Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 33 Group # 1 WA # 007G2.1.32 OK Importance Rating 3.4 Ability to explain and apply all system limits and precautions Proposed Question: Common 31 Which of the following describes the adverse affects of NOT maintaining the Pressurizer Relief Tank (PRT) within design level band?
A. If the level is too high, the tank will overflow to CNMT sump causing possible false indications of RCS leakage to CNMT.
B. If the level is too high, the sparger pipe will be too far underwater rendering the cooling affect of makeup water ineffective.
C. If the level is too low, there would be insufficient water volume to absorb and condense a design discharge of PRZR safety leading to possible over temperature and overpressure of the PRT.
D. If the level is too low the radioactive gases that leak from the top of the PRZR would not be adequately scrubbed, thus causing subsequent elevated gaseous activity levels inside CNMT.
Proposed Answer:
C. If the level is too low, there would be insufficient water volume to absorb and condense a design discharge of PRZR safety leading to possible over temperature and overpressure of the PRT.
Explanation (Optional):
Technical Reference(s): 3-SOP-RCS-007 (Attach if not previously Ll
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet SD-1.4 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-RCSPZR E-8 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10
'-d 55.43 5 Comments:
0 s11-33 System Description 1.4 Pressurizer & Pressurizer Relief Tank
(.
2.3.5 Acoustical Monitoring of Code Safeties and PORVs The safety and relief valve acoustical monitor system provides control room indications representative of valve position. The parameter actually monitored is flow.
One piezoelectric accelerometer is clamped to the outside of each code safety valve and PORV tail pipe, as shown in Figure 1.4-14. Flow through the tail pipe, which constitutes positive indication that the valve is open, causes flow noise and pipe vibration (acoustical acceleration). The accelerometer produces a piezoelectric charge proportional to the acceleration; the charge is converted to a voltage by a remote charge converter mounted on the west wall of the pressurizer shield at the 107-foot elevation. The voltage is applied to the TEC valve flow monitor module located in the control room-above the east access door to the rear of the supervisory panels. (TEC is the manufacturer's acronym.) The valve flow monitor module processes the voltage signal and indicates the flow on a lighted bar graph display calibrated in 10 increments of full flow. Full flow is 1.0 on the indicator. The monitor module contains a signal processing channel and display for each monitored valve: PCV-455C, PCV-456, PCV-464, I
PCV-466, and PCV-468.
If any one of the five monitor channels detects a flow signal greater than 25% of the full flow signal, it triggers a common Pressurizer PORV and Safety Acoustic Monitoring alarm on control room panel SAF.
2.4 Pressurizer Relief Tank If a PORV or code safety valve lifts, the steam, water, hydrogen and other gases in the PRZR flow to the pressurizer relief tank (PRT) through a common 12 inch discharge line that serves all five valves.
The piping is arranged as shown on Figure 1.4-15. This line is connected to the 12 inch perforated sprarger that is installed just above the bottom of the tank's bottom.
Normally, the tank is partially filled with water at or near containment ambient temperature and contains a predominantly nitrogen atmosphere maintained at a pressure of 0.5 to 3 psig. The nitrogen is maintained at this pressure by a nitrogen pressure regulator and is designed to eliminate air in-leakage. Sparging nozzles located beneath the water surface, discharge steam into the water volume. The mixing that results, condenses and cools the discharged steam. A %-inch vent d hole is drilled into the relief valve discharge line inside the PRT toward Rev. 3, 06/02/2000 - Page 28 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
\d Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 34 Group # 1 KIA # 008K1.04 OK Importance Rating 3.3 Knowledge of the physical connections and / or cause-effect relationship between the CCWS and the RCS, in order to determine sources(s) of RCS leakage into the CCWS Proposed Question: Common 32 Unit 3 is at 100% power when the following events occur in the order shown:
0 PRMS channels R-I 7A/17B,Component Cooling Water Activity Monitors, IN alarm.
CCW Surge TANK levels increasing rapidly.
Annunciator, RCP THERMAL BARRIER COOLING RETURN HIGH TEMP, in alarm.
Pressurizer level decreases and the running charging pump speed goes to maximum.
0 Annunciator, PRESSURIZER LOW LEVEL in alarm.
Which ONE of the following describes the event that has occurred?
A. A CVCS letdown non-regenerative tube has burst and LCV-459/460, High Press LID lsol Valves, have failed to close.
B. A CVCS letdown non-regenerative tube has burst and protective functions have responded as designed.
C. A RCP thermal barrier leak has occurred and protective functions have responded as designed.
D. A RCP thermal barrier leak has occurred and MOV-625, RCP Thermal Barrier Outlet Valve, has failed to close.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
.LJ Proposed Answer:
D. An RCP thermal barrier leak has occurred and MOV-625, RCP Thermal Barrier Outlet Valves, has failed to close.
Explanation (Optional):
Technical Reference(s): SD-4.1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-CCWOOI 0004,0006 (Asavailable)
Question Source: Bank # rNP0 26954 Modified Bank # (Note changes or attach parent)
New Question History: 12/15/2003 Turkey Point 3 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2,9 55.43 Comments:
$/I57 ' 1 P.
System Description 4.1 Component Cooling Water System 813E for compressor 32 using local flow and temperature indicators FI-666 and 668 and TI-667 and 669. CCW is isolated to the waste gas compressors unless the compressor is in service. Flow is manually throttled at 25 to 30 gpm.
CCW also provides normal makeup to the seal tank and is automatically controlled by level control valves on the cooler inlets after CCW has been unisolated.
Relief valves 821C and D provide overpressure protection at 150 psig.
2.7.5 RCPs and Vessel Support Pads (Figure 4.1-6)
(Header 32)
One 6 inch supply line provides CCW flow to all four RCPs and the four Reactor Vessel Support Blocks. Normally 20 gprn total is sent to the blocks and about 720 gpm to the RCPs. Each RCP requires approximately 5 gpm to the lower motor bearing cooler, 150 gpm to the upper motor bearing cooler, and 25 gprn to the thermal barrier. The throttle valves and flow indicators used to set these flowrates are aIl inside Containment.
There are two common return headers from the RCPs and the support blocks. One return header is for the upper and lower RCP bearing oil coolers and the support blocks. The other header is for the thermal barriers of all four RCPs.
The thermal barriers are segregated for leak considerations since a rupture at this point in the system would result in RCS inIeakage to CCW, The portion of the CCW System from the inlet check valves to the containment isolation valves is rated for 2500 psig and 650°F.
Leak protection is provided by FCV-625 and FE-625. If return flow increases to 175 gpm, FCV-625 wiI1 close and isolate the thermal barriers.
DP switch FIC-625 has a mechanical internal dampening feature set for 8 to 22 seconds. The dampening prevents spurious closure of FCV-625 due to the transients associated with sudden pressure surges when a CCW pump is secured or started, but allows closure of FCV-625 under sustained high flow conditions.
Overpressure protection is provided for each thermal barrier return line.
These relief valves are set at 2485 psig.
In addition to leak protection, the thermal barrier return line flow transmitter provides Control Room indication and a low flow alarm at 100 gprn (ThermalBarrier CCW Header Low Flow on SGF). This alarm will actuate if FCV-625 closes. TIC-624, also on the thermd barrier d
Rev. 3, 08/21/2001 - Page 14 -
System Description 4.1 Component Cooling Water System I' return line, actuates a high temperature alarm at 14OOF (RCP Thermal Barrier Cooling Return High Temp on SGF).
4 One temperature detector in the combined bearing cooler return actuates a high temperature alarm on Panel SGF at 13OOF (RCP Bearings Cooling Water Return High Temp).
2.7.6 Excess Letdown Heat Exchanger (Figure 4.1-7)
(Header 32)
All valves for the excess letdown heat exchanger are located just outside Containment. Normally, no flow is supplied to the heat exchanger.
When it is to be placed in service, only the Containment isolation valves have to be opened. These are controlled from the Control Room. When in service, approximately 240 gpm will be supplied.
2.7.7 RCP Seal Water Return Heat Exchanger (Figure 4.1-8)
(Header 32) 210 gprn is supplied to the RCP Seal Water Heat Exchanger. This is set using valve 808, a manual throttle valve. Outlet flow and temperature indicators, FI-605 and TI-606, are available locally.
2.7.8 CVCS Non-Regen Heat Exchanger (Figure 4.1-9)
'4 (Header 32)
The flowrate through the NonLRegen Heat Exchanger (NRHX) is automatically controlled by TCV-130, a 6 inch ball valve. Letdown temperature is fed to temperature controller TIC-130 in the Control Room, which controls TCV-130 automatically. CCW outlet flow and temperature indication is available on FI-607and TI-608.
Relief valve AC-812 relieves around TCV-130 if pressure exceeds 150 psig.
2.7.9 Gross Failed Fuel Detector (Figure 4.1-10)
(Header 32) 14 gpm is supplied to the detector. The flowrate is controlled manually with valve 1899B. Local flow indication is provided on FI-657.
d Rev. 3, 08/21/2001 - Page 15 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'-4' Examination 0utIine Cross-reference: LeveI RO SRO Tier # 2 ws # 35 Group # 1 KIA # 010A2.02 OK Importance Rating 3.9 Ability to predict the impacts of a spray valve failure on PZR PCS and on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations Proposed Question: Common 33 Given the fo Ilowing:
0 Unit 3 reactor power is 30%.
0 RCS pressure is 2075 psig and slowly lowering.
0 All Pressurizer heaters are energized.
0 You notice that PCV-455B (PZR spray) is failed OPEN.
\-.-,
0 When placed in manual PCV-455B will NOT close.
0 Operators removed control power fuses for PCV-455B but valve will not close Which ONE of the following is the proper sequence of actions to stop the pressure reduction?
A. Manually trip the reactor.
Trip 33 and 34 RCPs.
Go to E-0, Reactor Trip Or Safety Injection.
B. Trip 33 RCP.
The RCP trip will NOT cause a reactor trip at this power.
Dispatch an NPO to locally isolate Spray Valve PCV-455B.
C. Reduce Power to ~ 2 5 % so a RCP trip will NOT cause a reactor trip.
Trip 33 RCP.
Dispatch an NPO to locally isolate Spray Valve PCV-455B.
D. Trip 33 and 34 RCPs.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet The reactor will trip when the RCPs are tripped.
Go to E-0, Reactor Trip Or Safety Injection.
Proposed Answer:
A. Manually trip the reactor.
Trip 33 and 34RCPs.
Go to E-0, Reactor Trip Or Safety Injection.
Explanation (Optional):
Technical Reference(s): 3-ARP-003, page 14 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE
>d Learning Objective: (As available)
Question Source: Bank # INPO 22876 Modified Bank # (Note changes or attach parent)
New Question History: 12/9/2002 Cook 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
NO: 3-ARP-003 Rev. 43 PANEL SAF - REACTOR COOLANT SYSTEM Page 14 of 59 3.4.2.1 CLOSE PORV(s).
3.4.2.2 E PORV(s) do NOT close, THEN:
a) CLOSE the affected PORV Block Valve.
b) REMOVE power fuses from the PORV solenoid to fail valve closed:
RC-PCV-455C SOLENOID FUSES (Panel FCR, FU-398 & 399)
RC-PCV-456 SOLENOID FUSES (Panel FBR, FU-347 & 348) e) Dispatch NPO to DE-ENERGIZE the affected PORV Block Valve:
RC-MOV-536 at MCC 36A 0 RC-MOV-535 at MCC 366 d) PRZR PORV(s) are NOT closed, THEN GO.TO 3-AOP-LEAK-1, Sudden Increase in Reactor Coolant System Leakage.
3.4.3 PRZR Spray Valve(s) are NOT closed, THEN:
3.4.3.1 CLOSE PRZR Spray Valve in manual.
3.4.3.2 E PRZR Spray Valve does NOT close, THEN CLOSE PRZR Spray Valve by removing power fuse from back of Foxboro controller for affecfed PRZR Spray Valve.
3.4.3.3 E PRZR Spray VaIve(s) can NOT be closed, THEN:
a) TRIP the Reactor.
b) TRIP 33 and 34 RCPs.
c) GO TO E-0, Reacior Trip or Safety Injection.
(CONTINUED ON THE NEXT PAGE)
System Description 1.4 Pressurizer & Pressurizer Relief Tank The pressure indicator, PI-455 or PI-457, of the channel supplying the control train is illuminated on the control board.
2.7.3 PRZR Pressure Alarm Train The alarm train consists of a dual high/low pressure comparator PC-456 F/G which compares the signal representing actual pressure to fixed high/low pressure setpoints, 2335 psig and 2185 psig, respectively. At 2335 psig, PC456F triggers the pressurizer high-pressure alarm on control room panel SAF. This alarm function constitutes a backup for a component in the control train, PC-4551, which triggers the same alarm at 2310 psig.
At 2335 psig, PC-456F also sends a signal to the opening logic for RC-PCV-456. If Pc-457F also senses a 2335 psig from interlocking channel PT-457, and the control switch for RC-PCV-456 is in AUTO, relief valve RC-PCV-456 opens.
Pressure bistable PC-456G triggers the Pressurizer Low Pressure alarm on panel SAF, at 2185 psig. This same alarm can be triggered by bistables PC455J in the control train, also at 2185 psig.
2.7.4 PRZR Pressure Control Train The PRZR pressure control train is composed of:
pressure controller PC-455K, control heater group controller PC-455L, spray valve RC-PCV-455A controller, PC-455G, spray valve RC-PCV-455B controller, PC-455H, 0 dual high/low pressure deviation comparator PC455 I/ J, and trip bistable for PORV RC-PCV-455C, PC-455F.
2.7.5 PRZR Pressure Controller The control signal development is shown schematically in Figure 1.4-23. Actual pressure, designated PACT,is the input to the PRZR pressure controller PC-455K, located on panel FBF in the control room.
A manually adjustable setpoint signal, designated Pm, is set by means of a dial on the face of PC-455K. During normal operations, the setpoint is set to 2235 psig, which corresponds to a PC-455K setting of 66.8% (Refer to Attachment A for detailed discussion on PC-455K operation).
4 Controller PC-455K is a MD (Proportional, Integral, and Derivative) controller. The derivative function is set to zero. The proportional (P)
Rev. 3, 06/02/2000 - Page 41 -
System Description 1.4 Pressurizer & Pressurizer Relief Tank f
component of PC-455K is functionally a summer with gain and bias.
W The summer receives PREP as a negative input and PACTas a positive input; it determines the offset. It then imposes a gain of 2 on the offset, that is, the output signal is twice the size of the input signal (the offset).
The proportional component also applies a bias that produces a 30 milliamp (ma) output signal when PACT = PREF.Thus, the proportional component of PC-455K, PERROR, varies above or below 30 ma as PACT varies above or below PREF.
The integral component of PK-455K adjusts the controller signal based on the length of time PACTdiffers from PREF. In particular, the magnitude of the output signal increases as long as the offset is not zero.
This compensated PERROR signal is fed to:
0 both spray valve controllers, 0 the PRZR heater control group (modulating)controller and 0 the deviation comparator (PC-455 I/J), which controls the backup heaters, and I
0 the trip bistable (PC-455F) for PORV RC-PCV-455C.
The operation of these components as part of the overall pressure control scheme is shown graphically in Figure 1.4-24.This scheme is based on:
PFEFset to 2235 psig, all groups of backup heaters in AUTO, and both spray valves being set to operate at the same setpoints.
The assumption the PREF = 2235 psig is useful because this is normal operating pressure. However, this can be changed on PC-455K, therefore the offset values are given in parentheses.
In addition, Figure 1.4-24,and the discussion below, do not include the effects of the integral (I) component of PC-455K. The actual output signal is greater than the arithmetic difference between PACTand PREF, if that difference has existed (with the same sign) for a length of time.
This means the system responses may occur at higher or a lower pressures than listed.
2.7.6 PRZR Heater Control Group Controller The PRZR Heater Control Group (modulating) controller PC-455L, 4
located in instrument rack (Foxboro) B-6 is a proportional controller with automatic and manual control modes. In automatic, the Rev. 3, 06/02/2000 - Page 42 -
System Description 1.4 Pressurizer & Pressurizer Relief Tank f '
controller uses the compensated PERROR signal to generate the control d signal for the control group heater SCRs. When pressure is at PREF the output of PC-455L is at 50% of maximum and the thermal output of the control group is 50% of maximum. By the time pressure rises to 2250 psig (PREF + 15 psi) the controller output is zero and the control group heaters are off. By the time pressure falls to 2220 psig (Pm - 15 psi) the modulating heaters are fully on.
In manual, the compensated PERROR signal is disconnected from the controller and the control signal to the SCRs is adjusted using the manual operation bar on the controller.
2.7.7 P U R Spray Valve Controllers The spray valve controllers PC-455G and PC-455H, located on panel FBF, are proportional (P) controllers with automatic and manual control modes. In automatic, they use the compensated PERROR to generate the control signal for positioning the spray valves. They generate no output until pressure rises to 2260 psig (PREF+ 25 psi).
Then they ramp open the spray valves in a linear fashion until they are full open at 2310 psig ( P m + 75 psi).
In manual, the compensated PERROR signal is disconnected from the spray controller. The signal to the spray valves is adjusted using the manual operation bar on the controller.
The compensated PERROR signal is applied to a high-pressure comparator, PC-455F, which at 2335 psig (PREF + 100 psi) supplies a trip signal to the opening logic for PORV,RC-PCV-455C. The PORV opens if the interlock signal from PC-474B is also present and its control switch is in the auto position. The open permissive input to RC-PCV-455C ensures that the integrating (I) function of PC-455K does not cause RC-PCV-455C to open at a pressure less than 2335 psig.
However, the integrating function may delay opening of PCV-455C until pressure is greater than 2335 psig.
Finally, the compensated PERROR signal is applied to the dual high/low pressure comparator PC-455 I/J. AT 2310 psig (PREF + 75 psi), PC-4551 triggers the Pressurizer High Pressure annunciator on panel SAF.
(PC-456F in the alarm train triggers this same alarm window at 2335 psig fixed setpoint.) At 2185 psig (PREF - 50 psi), PC-455J triggers the Pressurizer Low Pressure annunciator on panel SAF. This is the same 2185 psig alarm (fixed setpoint) that PC-456G annunciates. Also, at 2185 psig (PREF - 50 psi) PC-455J energizes the backup heaters that are selected to AUTO. The backup heaters remain on until pressure is
'L.l Rev. 3, 06/02/2000 - Page 43 -
System Description 1.4 Pressurizer & Pressurizer Relief Tank restored to 2200 psig ( P w - 35 psi), which resets PC-455J and turning the backup heaters off.
The following is a table of error signal/control functions based on the output of the master controller (PC-455K) (without the integral component).
PRZR Functions as a Function of PC-455K Output Pressure Deviation PC-455K Output Function from setpoint (psi) ("4 Deviation Low Press -50 37.50 Alarm Backup Heaters On I -50 1 37.50 Backup Heaters Off -35 41.25 Control Heaters Full On -15 46.25 Pressure Setpoint (Pw) 0 50.00
'V Control Heaters full Off I +15 I 53.75
~
Spray Initiation +25 56.25 spray Full On +75 68.75 High Pressure Alarm +75 68.75
?ORV Closes I +85 I 72.50
?ORV Opens 1 +loo I 75.00 2.7.8 PRZR Pressure Control System Operation The current mode of operation of the PRZR pressure control system differs from that described above.
The first difference is that one spray valve controller, either PC-455G or PC-455H, is biased such that its associated valve opens before (or leads) the other. This is done to produce a smoother spray initiation.
The second difference is that one set of backup heaters is kept on.
Continuously energizing one set of backup heaters increases the heat input to liquid space, raising the pressure. The pressure increase u causes:
Rev. 3, 06/02/2000 - Page 44 -
System Description 1.4 Pressurizer & Pressurizer Relief Tank 0 Control heater group thermal output to decrease (Off at 2250 psig (PREF + 15 psi)), and 0 Spray valves to open (2260 psig ( P w + 25 psi)).
The pressure increase caused by the backup heaters stops when it is balanced by the condensation caused by the additional spray flow.
However, the system cannot remain at this pressure (above 2260 psig) because this produces a non-zero offset in PC-455C. As long as a non-zero offset exists, the integral (reset) component of PC-455C causes the magnitude of the control signal to continue to increase. Thus, the spray valves receive a signal to open more. The additional spray flow causes pressure to decrease toward 2235 psig. However, if the offset-plus-reset signal decreases to the point that spray valves close, the energized backup heaters raise pressure above 2235 psig again, the offset-plus-reset signal increases and a spray valve eventually reopens.
The system can only achieve stability (a relatively constant pressure) when:
0 the pressure increase caused by the heat input of the backup heaters I
equals u ' 0 the pressure reduction caused by the total spray flow (bypass flow plus spray valve(s) flow).
Since the bypass spray flow is insufficient to compensate for the energized backup heater group, the spray valve@)remain open (or continually cycle open) a small amount. This requires a relatively continuous open demand signal from the PRZR pressure controller.
The demand signal is due to the integration (reset) signal developed from of a small offset above 2235 psig. Therefore, with one set of backup heaters ON, the RCS pressure stabilizes with a spray valve open a small amount, at a pressure above 2235 psig, but less than 2260 psig.
2.8 Over-pressurization Protection System The Over-pressurization Protection System (OB) is designed to prevent over-pressurization of the reactor vessel when the RCS is at low temperatures. At low temperatures, the reactor vessel is less ductile and more susceptible to brittle fracture. Therefore, stress on the vessel from all sources must be reduced as the vessel temperature
\d decreases. In particular, the maximum allowable RCS pressure must be reduced, as a function of temperature, to limit the pressure-induced Rev. 3, 06/02/2000 - Page 45 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 36 Group # 1 WA # 012A1.01 OK Importance Rating 2.9 Ability to predict and / or monitor changes in parameters to prevent exceeding design limits associated with operating the RPS controls including trip setpoint adjustments Proposed Question: Common 34 During the performance of an NIS power range heat balance at 90% power, an operator uses a feedwater temperature 30°F lower than actual. Would the calculated value of power be HIGHER or LOWER than actual power, and would an adjustment of the NIS power range channels, based on this value, be CONSERVATIVE or NON CONSERVATIVE with respect to High Power Reactor Trip protection setpoints?
Calculated Power Setpoints would be..
'4 A. Higher - Non Conservative B. Higher - Conservative C. Lower - Non Conservative D. .Lower - Conservative Proposed Answer:
B. Higher - Conservative Explanation (Optional):
Technical Reference(s): (Attach if not previously d
'~
provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # INPO 27669 Modified Bank # (Note changes or attach parent)
New Question History: 10/5/2004 Cook 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 37 Group # 1 K/A ## 013A1.09 OK Importance Rating 3.4 Ability to predict and I or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ESFAS controls including T-hot.
Proposed Question: Common 35 Initial Conditions:
0 The plant is at 100% power, beginning of life.
0 Rod Control is in MANUAL.
0 Tave is on program.
0 The Reactor Engineer has requested the crew to slowly withdraw control bank I'D" rods to full out after Moderator Temperature Coefficient (MTC) testing.
0 The crew is to allow MTC to control reactor power, without borating during the rod withdrawal.
The RO slowly withdraws control bank "D" rods, resulting in the following:
RCS Narrow Range Thot increases by 4°F.
PZR pressure control system maintains RCS pressure stable.
0 Delta Flux remains in the program band. .
How do the OTAT and OPAT trip setpoints respond?
A. OTAT setpoint DECREASES.
OPAT setpoint DECREASES.
B. OTAT setpoint DECREASES.
OPAT setpoint DOES NOT CHANGE.
C. OTAT setpoint DOES NOT CHANGE.
OPAT setpoint DECREASES.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet "d D. OTAT setpoint DOES NOT CHANGE.
OPAT setpoint DOES NOT CHANGE.
Proposed Answer:
A. OTAT setpoint DECREASES.
OPAT setpoint DECREASES.
Explanation (Optional):
Technical Reference(s): ITS 3.3.1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE u
Learning Objective: 13LP-ILO-ICRXR E l (As available)
Question Source: Bank # INPO 27099 Modified Bank # (Note changes or attach parent)
New Question History: 7/16/2004 Millstone 3 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
ut's 3 7 J
t RPS Instrumentation 3.3.1 (7 Table 3.3.1-1 (page 7 of 8)
Reactor Protection System Instrumentation i /
Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 2.8% of AT span :
Where: AT is measured RCS AT, OF.
ATo is the indicated AT at RTP, O F .
s is the Laplace transform operator, sec-'.
T is the measured RCS average temperature, OF.
T' is the nominal Tavg at RTP, I [
- 1°F.
P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, s [ * ] psig K1 < [
- I K2 2 [ * ]/OF K3 2 [ * ]/psig 2 [ * ] sec z2 -< [ * ] sec f
e f,(AI) = - - qb)}
[ * ] {[
- J -I(qt when 41 - qb I- [
-[
- 1 {(St - qb) - [
- I} when qi - qb > [
- 1% RTP ..
Where qt and qb are percent RTP in the upper and lower halves of the core, I
respectively, and qt + qb is the total THERMAL POWER in percent RTP.
The values denoted with [ * ] are specified in the COLR.
INDIAN POINT 3 3.3.1-1 9 Amendment 225
f RPS Instrumentation 3.3.1 f Table 3.3.1-1 (page 8 of 8 )
Reactor Protection System Instrumentation Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.8% of AT span:
Where: AT is measured RCS AT, OF.
ATa is the indicated AT at RTP, O F .
s is the Laplace transform operator, sed.
T is the measured RCS average temperature, OF.
T is the nominal Tavg at RTP, s [
- 1°F.
&s [*I I K,r [*]/°FforincreasingT,,g, K6> [*]/°FwhenT>T
[ ] /OF for decreasing Taq [ * ] / O F when T I T 2 [ * ] sec f2(W = [
- 1 L *The values denoted with [ ] are specified in the COLR.
INDIAN POINT 3 3.3.1-20 Amendment 225
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 39 Group # 1 KIA # 022K2.01 OK Importance Rating 3 .O Knowledge of power supplies to the containment cooling fans Proposed Question: Common 36 A small break LOCA occurred causing a Reactor Trip AND Safety Injection. All safeguards equipment operated as designed. During the performance of E-0, Reactor Trip or Safety Injection, Offsite Power was lost.
Given the following:
0 31 EDG failed to start All remaining equipment operated as designed All equipment that should have automatically started has started on their respective buses. With no operator action, what will be the configuration for the Containment Fan Cooler Units (FCU)?
A. 31, 32, 33 and 35 FCUs running B. 31,33 and 35 FCUs running C. 33,34 and 35 FCUs running D. 32, 33,34 and 35 FCUs running Proposed Answer:
- 8. 31,33 and 35 FCUs running Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): SD-10.3 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-VCCARC 0003 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
System Description 27.3 us 39.
J Emergency Diesels W
J n c3 n
?-
,a' n
W Z
I I
W Z
0
,d Figure 27.3-I : Emergency Diesels One-Line Diagram (EDS-01)
Rev. 2,03/01/2001 - Page 48 -
System Description 10.3 Containment Air Recirculation Cooling And Filtration System Related Power Supplies Component Power Supply Comments 31 Fan Cooler Unit 480 V Bus 5A 32 Fan Cooler Unit 480 V Bus 2A 33 Fan Cooler Unit 480 V Bus 5A 34 Fan Cooler Unit 480 V Bus 3A 35 Fan Cooler Unit 480 V B u s 6A Rev. 3, 08/30/1999 - Page 23 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet i_/ Examination Outline Cross-reference: Level RO SRO Tier # 2 WS#40 Group # 1 KIA # 026K3.01 OK Importance Rating 3.9 Knowledge of the effect that a loss or malfunction of the CSS will have on the CCS Proposed Question: Common 37 Given the following conditions:
0 Unit 3 is operating at 100% power.
0 31 Containment Spray pump has been declared INOPERABLE due to an oil leak.
0 32 and 34 Fan cooler Units (FCU) are INOPERABLE and isolated due to service water leaks.
0 All other ECCS equipment is OPERABLE.
u With the plant in this configuration, which of the following describes if the plant is being operated within the Design Basis for containment cooling, and the BASES for your answer?
A. No,two (2) Containment Spray pumps and five (5) FCU's are required to be OPERABLE to meet the design basis for containment cooling.
B. No, one containment Spray pump and four (4) FCU's are required to be OPERABLE to meet the design basis for containment cooling.
C. Yes, one (1) OPERABLE Containment Spray pump combined with three (3)
OPERABLE FCU's meets the design basis for containment cooling.
D. Yes, a single OPERABLE Containment Spray pump meets the design basis for containment cooling.
Proposed Answer:
'd C. Yes, one (I) OPERABLE Containment Spray pump combined with three (3)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet OPERABLE FCU's meets the design basis for containment cooling.
'4 Explanation (Optional):
Technical Reference(s): SD-10.3 (Attach if not previously ITS 3.6.6 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-VCCARC 0004, 0006 (As available)
Question Source: Bank # INPO 24076 Modified Bank # (Note changes or attach parent)
New Question History: 5/5/2003 Salem Unit 1
%d Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Cornments:
System Description 10.3 Containment Air Recirculation Cooling And Filtration System Specification 4.5.A.4 specifies testing requirements for the containment air filtration system.
3.5 Improved Technical Specifications TS 3.6.5 (Containment Air Temperature) limits containment average air temperature during normal operation to preserve the initial conditions assumed in the accident analyses for a Loss of Coolant Accident (LOCA) or Steam Line Break (SLB). The temperature limit is used to establish the environmental qualification operating envelope for containment. The maximum peak containment air temperature was calculated to exceed the containment design temperature for only a few seconds during the transient and has been determined to be acceptable for the DBA LOCA or SLB. The bases of the containment design temperature, however, is to ensure the performance of safety
._ related equipment inside containment.
TS 3.6.6 (Containment Spray System and Containment Fan Cooler System) sets limits on FCU and containment spray trains. Accident Analysis assumptions regarding containment air cooling and iodine removal are met by any of the following configurations:
- a. Two containment spray trains; or,
- b. T h e e fan cooler trains (i.e.., five fan cooler units); or
- c. One containment spray train and any two fan cooler trains (i.e., at least three fan cooler units).
The last configuration, one containment spray train and two fan cooler trains, is the minimum configuration available following the loss of any safeguards power train (e.g., diesel failure).
TS 3.4.15 (RCS Leakage Detection Instrumentation) requires instruments of diverse monitoring principles operable to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition when RCS leakage indicates possible Reactor Coolant Pressure Boundary (RCPB) degradation.
iw Rev. 3,08/30/1999 -Page 21 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
.4 Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 29 Group # 1 KIA # 003K6.04 OK Importance Rating 2.8 Knowledge of the effects of a loss or malfunction on the containment isolation valves affecting RCP operation will have on the RCPs Proposed Question: Common 38 Unit 3 was operating at 100% power when an inadvertent Phase B isolation occurred.
Which of the following describes the required actions for the Phase B isolation and the reason for performing those actions?
A. Immediately trip the Reactor, trip all four RCPs and enter E-0, Reactor Trip due to loss of RCP seal return flow.
B. Immediately trip the Reactor, trip all four RCPs and enter E-0, Reactor Trip due to loss of CCW to all four RCPs.
C. If Phase B CCW supply and return valves for RCP motor cooling are not opened within 2 minutes then trip the Reactor, trip all four RCPs and enter E-0, due to loss of RCP motor cooling.
D. If Phase B CCW supply and return valves for RCP motor cooling are not opened within 2 minutes then trip the Reactor, trip all four RCPs and enter E-0, due to loss of RCP seal cooling.
Proposed Answer:
C. If Phase 8 CCW supply and return valves for RCP motor cooling are not opened within 2 minutes then trip the Reactor, trip all four RCPs and enter E-0, due to loss of RCP motor cooling.
Explanation (Optional):
~d
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4 Technical Reference(s): 3-AOP-CCW-1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # ~
(Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
W
Loss Of Component Cooling Water 3-AOP-CCW-1Rev. 02 Page 7 of 59
- - -_-_.-..-----_L_-_ ~ ~ ~ ~
4.2 Do any of the following conditions - RETURN to procedure and step in exist? effect.
-System leak
-l o w system flow/pressure
-High system temperature __--_____ - --.__-_-I-__
4.3 -lAAT either of the following - GO TO Step 4.8.
conditions exist:
-CCW flow to any RCP is lost for 2 2 minutes
-RCP motor bearing temperature exceeds 200°F THEN perform Steps 4.4 - 4.7.
4.4 -Are the Reactor Trip Breakers 1. -Trip affected RCPs.
closed?
W -- _-__I__ ~ -
__I___- __ 2. -GO TO Step 4.8.
4.5 -Trip the reactor.
--_l_____l__~----_-I ________ ~
System leak 4.9 II Low system flow/Pressure 1 -4741 High system 4.108 temperature
ES-40I Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
\--.//
Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 42 Group # 1 KIA # 039K5.08 OK Importance Rating 3.6 Knowledge of the operational implications of the effect of steam removal on reactivity as it applies to the MRSS Proposed Question: Common 39 Given the following conditions:
A Unit startup is in progress following a mid-cycle outage.
0 The reactor is critical at 1E-8 amps.
0 A condenser steam dump valve fails partially open.
L/ Assuming NO action by the operating crew, which one of the following describes the immediate effect on the plant?
A. RCS Temperature INCREASES; Power INCREASES.
B. RCS Temperature INCREASES; Power DECREASES.
C. RCS Temperature DECREASES; Power DECREASES D. RCS Temperature DECREASES; Power INCREASES.
Proposed Answer:
D. RCS Temperature DECREASES; Power INCREASES.
Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s) : (Attach if not previously
'd provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: TAA-C-005 2504 (As available)
TAA-C-0 11 02532 Question Source: Bank # INPO 24963 Modified Bank # (Note changes or attach parent)
New Question History: 12/1/2002 Beaver Valley 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 44 Group # 1 KIA # 059K4.08 OK Importance Rating 2.5 Knowledge of MFW design features(s) and / or interlock(s) which provide for feedwater regulatory valve operation (on basis of steam flow, feed flow mismatch)
Proposed Question: Common 40 Given the following conditions:
A plant startup is in progress The Unit is at 30% power.
All Main Feed Regulating Valves are in AUTO 31 Feed Flow channel FT-418B is selected for control of 31 SG.
e 31 Feed Flow transmitter PT-418B fails 10% high.
L--,,
Assuming NO operator action, which of the following statements describes the response of 31 Main Feed Reg (MFR) Valve?
A. 31 MFR valve will initially throttle in the CLOSE direction and then over time will return to it's original position.
B. 31 MFR valve will initially throttle in the OPEN direction and then over time will return to it's original position.
C. 31 MFR valve will CLOSE and then over time the Reactor will trip on Low SG leveI.
D. 31 MFR valve will OPEN and then over time the Turbine will trip on High SG level.
4 Proposed Answer:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet A. 31 MFR valve will initially throttle in the CLOSE direction and then over time will return to it's original position.
Explanation (Optional):
Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-ICLOVE E (As available)
I3LP-ILO-ICSGL 2.0 Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
3-ARP-005 Rev. 30 Page 20 of 51 W
INPUT DEVICE: LC-417A LC-417c STEAM GEN # 31 LC-417E HIGH LEVEL SETPOINT: 75% of span 1.0 CAUSES 1.1 High level on 1 of 3 SG level channels.
2.0 AUTOMATIC ACTIONS 2.1 level is equal to or greater than 75% on 2 of 3 level channels, THEN:
0 Turbine trips iJ MBFP discharge valves close 3.0 SUBSEQUENT ACTIONS 3.1 VERIFY alarm by observing all SG level indicators and trip status lights.
3.2 E channel failure has occurred, THEN GO TO ONOP-RPC-1 I Instrument Failures.
3.3 level is high on only 1 of 3 level channels, THEN:
3.3.1 TRANSFER SG level control to manual as required.
3.3.2 RETURN level to programmed value.
(CONTINUED ON THE NEXT PAGE)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
.L/
Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 45 Group # 1 WA # 06 1K5.01 OK Importance Rating 3.6 Knowledge of the operational implications of the relationship between AFW flow and RCS heat transfer as it applies to the AFW Proposed Question: Common 41 Given the following plant conditions:
0 A reactor trip has occurred from 100% power 0 The operators have not operated any controls post-trip.
0 The crew has just entered ES-0.1 Reactor Trip Response.
0 PZR level is 25% and slowly decreasing.
0 Steam Generator pressures are approximately 990 psig and slowly decreasing.
0 Tave is 545°F and slowly decreasing.
0 RCS pressure is 2020 psia and slowly decreasing.
What action must be taken by the crew per ES-0.1 to address the cooldown?
A. Commence immediate boration.
B. Throttle Auxiliary Feedwater flow.
C. Initiate SI and return to step Iof E-0.
D. Close the MSlVs and MSlV bypass valves.
Proposed Answer:
B. Throttle Auxiliary Feedwater flow.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
TechnicaI Reference(s): ES-0.1, Step 1 RNO (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # INPO 27090 Modified Bank # (Note changes or attach parent)
New Question History: 7/16/2004 Millstone 3 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2.9 55.43 Comments:
Number:
Title:
Revision Number:
ES-0.1 REACTOR TRIP RESPONSE 18 4
ACTION/EXPE
- 9
- 1.
- C H E C K RCS Averase TemDerature:
- PERFORM t h e f o l l o w i n g : 9
- 3
- AVERAGE T E M P E R A T U R E S T A B L E 5 4 7 ° F AND decreasing, T H E N j;
- AT T R E N D I N G T O 547°F PERFORM the following:
- OR 1) S T O P d u m p i n g steam. *
- LEG TEMPERATURES S T A B L E A T -
T H E N PERFORM the
- OR T R E N D I N G T O 5 4 7 ° F following: *
- a ) C O N T R O L t o t a l feed *
- flow.
- i
- b ) M A I N T A I N greater t h a n *
- 365 gpm f l o w u n t i l a t *
- l e a s t one S G NR level *
- greater t h a n 9%. *
- 3 ) IF cooldown continues, *
- T H E N PERFORM t h e *
- fol 1 o w i n g : *
- a > CLOSE a l l M S I V s . *
- b ) E M S I V ( s ) can NOT be *
- closed, T H E N D I S P A T C H *
- NPO t o l o c a l l y close *
- M S I V s per S O P - E S P - 1 . *
- c ) D I S P A T C H NPO t o *
- ensure a l l M S I V *
- bypass valves a r e *
- c l osed, i f requi red. *
(STE 3NTINUED ON NEXT PAGE)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
>W' Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 38 Group # 1 WA # 013K3.03 OK Importance Rating 4.3 Knowledge of the effects that a loss or malfunction of the ESFAS will have on the containment Proposed Question: Common 42 Given the following conditions:
0 A Large Break LOCA has occurred.
0 Train B ECCS has failed to actuate.
0 All other actuations actuate and Train A ECCS equipment is running as required.
i/'
Assuming no action by the crew, which ONE (1) of the following describes the effect on the plant?
A. Containment Isolation Phase A will actuate. Phase B will NOT actuate.
B. Containment Isolation Phase A will NOT actuate. Phase B will actuate.
C. Containment Isolation Phase A will NOT actuate. Phase B will NOT actuate.
D. Containment Isolation Phase A and B will actuate.
Proposed Answer:
D. Containment Isolation Phase A and B will actuate.
Explanation (Optional):
W'
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided tu applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # INPO 28065 Modified Bank # (Note changes or attach parent)
New Question History: 9/27/2004 Robinson 2 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
System Description 10.2 Containment Spray System The Containment pressure can be maintained less than the designed pressure of 47 psig and 271OF following the LOCA by:
AI1 five Containment Recirculation Fan Cooler Units, or, Both Containment Spray Pumps, Or Three Containment Recirculation Fan Cooler Units and one Containment Spray Pump.
The spray additive tank contains a minimum of 4,000 gallons of solution with a sodium hydroxide concentration not less than 35% and no greater than 38% by weight. This volume, when mixed with the water from the RWST, accumulators and the Reactor Coolant System, wiIl result in a solution in the Containment sump with a pH greater than 8.3. This allows continued iodine removal by spray during the recirculation phase of operation.
The RWST contains a minimum of 342,200 (35.4') gallons of water with a boron concentration >2400 but <2600 PPM. Approximately 167,000 gallons is used by the Containment Spray System, on a design basis accident.
In the first twenty minutes or so following the maximum L E A , the Containment Spray Pumps operate to meet the design heat removal capacity for Containment. The total heat absorption capability of each spray pump delivering 2500 gpm is 2.18 x 108 BTU/HR based on addition of 100°F RWST water.
14 L./
Rev. 4, 06/03/1999 - Page 3 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
~..d Examination Outline Cross-reference: Level RO SRO Tier # 2 WS ## 41 Group # 1 WA ## 026A3 .O 1 OK Importance Rating 4.3 Ability to monitor automatic operation of the CSS including pump starts and correct MOV positioning Proposed Question: Common 43 Given the following plant conditions:
0 A Large Break LOCA occurred with SI actuation 0 15 seconds after the SI actuation containment pressure rises to 24 psig Which one of the following sets of pumps/valves receives a stadopen signal on the automatic Containment Spray Actuation?
.4 A. Spray pumps 31 and 32 after -34 second time delay, MOV 866A and B, CNMT Spray Pump Discharge Valves immediately and AOV 876A and B, CNMT Spray NaOH addition after -2 minute time delay.
B. Spray pumps 31 and 32 after -34 second time delay, 869A and B, Containment Spray Header Isolation Valves immediately, and MOV 866A and B, CNMT Spray Pump Discharge Valves immediately.
C. Spray pumps 31 and 32 after -2 minute time delay, 869A and B, Containment Spray Header Isolation Valves, and 880A-K, CNMT Spray Charcoal Filter Douse Valves.
D. Spray pumps 31 and 32 after -2 minute time delay, AOV 876A and B, CNMT Spray NaOH addition and 880A-K, CNMT Spray Charcoal Filter Douse Valves.
Proposed Answer:
A. Spray pumps 31 and 32 after -34 second time delay, MOV 866A and B, CNMT 4 Spray Pump Discharge Valves immediately and AOV 876A and B, CNMT Spray
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet NaOH addition after -2 minute time delay.
W Explanation (Optional):
Technical Reference(s): SD-10.2 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-CSOO1 E-I (As available)
Question Source: Bank # INPO 27300 Modified Bank # (Note changes or attach parent)
New Question History: 4/27/2004 Ginna 1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
Id5 Y /
System Description 10.2 Containment Spray System contents of the RWST. At 1.5 ft in the RWST the operator secures the spray pump and shuts the discharge valve 866A.
The Containment Spray Headers are supplied by the recirculation pumps during the recirculation phase. This water is from the recirculation sump, cooled to approx. 134OF through the residual heat exchanger, and results in a heat removal rate of 1.63 x 10s BTU/HR.
This equals the core decay heat after 5000 seconds. The recirculation phase of Containment Spray continues following the LOCA to ensure the complete removal of iodine from the Containment atmosphere.
The recirculation mode of operation is detailed in System Description 10.1 Safety Injection.
The spray system is automatically actuated on a high high Containment pressure of approx. 22 psig. This signal is a coincident signal ( 2 / 3 ) from two separate channels. Each channel consists of three pressure instruments. One transmitter from each channel comes from a common containment sensing line. Three sensing lines feed six transmitters. One channel is feed from PI-948A, B, C and the other channel from PI-949A, B, C. The actuation of either channel, energized to trip, causes an alarm on the safeguards panel (SBF-2) and indicates on the trip status panel (SOF) which channel has tripped. Both r channels (2/2)must trip to actuate the system. Redundant logics are used in the formation of the high-high Containment pressure signal to prevent actuation of the Containment Spray System on a spurious signal or loss of power. This signal acts as a back up to the Safety Injection signal and results in a phase "B'I Containment isolation, a Containment Ventilation isolation, and a Steam Line isolation.
The system can be manually actuated by simultaneously depressing two red Spray Initiation pushbuttons on safeguards panel SBF-1 in the control room. This manual action also initiates a Containment Ventilation isolation and a Containment phase "B" isolation. The manual signal does not activate a Safety Injection signal or the Steam Line isolation signal. The Containment Spray actuation logics are detailed in System Description 28, Overall Unit Protection.
When the Containment Spray System is actuated, each motor operated pump discharge valve (866A and 866B) opens and the spray pumps start. A portion of each pump discharge is directed through two liquid jet eductors which add the sodium hydroxide solution contained in the spray additive tank into the spray pump suction.
The spray additive tank remains isolated for two minutes after a Containment Spray actuation signal. This two minute time delay gives f" the operators time to assess plant conditions and to cancel the Sodium d'
Rev. 4, 06/03/1999 - Page 5 -
System Description 10.2 Containment Spray System Hydroxide addition. Isolation valves (876A and 8768) for spray i
additive tank are normally shut, fail open, air operated diaphragm L/ valves. The isolation valves can be opened by switches located on SBF-1 panel in control room. Automatic opening of the spray additive tank isolation valves on Containment Spray signal can be canceled by pushing the red NaOH cancel pushbutton on SBF-1 panel in control room. (Currently, there is no procedural guidance to cancel NaOH addition even if the signal is inadvertent.)
The Containment Spray actuation signal is reset by the use of two pushbuttons, one for each logic channel, located on the safeguards panel SBF-1. Depressing both pushbuttons will not stop any running equipment, but it will permit the operator to take manual action to secure the spray pumps and shut the motor operated discharge valves.
If Containment pressure again reaches the actuation pressure of approx. 22 psig, the Containment Spray System will again actuate.
The system also supplies fire-dousing water to the five Containment Recirculation Fan Cooler Unit charcoal filter banks. The water-dousing supply is used in the unlikely event of a fire resulting from the heat generated by the decay of iodine absorbed in the charcoal filters. The filter banks of each fan unit are supplied through two, normally shut motor operated valves. These valves are operated with individual, two-position switches, and individual valve position is indicated with red (open) and green (shut) lights all located on panel SMF in the control room.
2.2 Containment Spray Pumps 31 and 32 The Containment Spray Pumps are located at elevation 41 foot in the Primary Auxiliary Building. The pumps are single stage horizontal centrifugal pumps rated for 2600 gpm at a 427 foot discharge head.
The pumps are equipped with mechanical seals to eliminate leakage and require no external means of cooling. They are driven by 400 HP electric motors powered from 480V buses 5A and 6A, respectively.
The spray pumps are designed to deliver their rated flow with the RWST at a level of 1.5 feet (empty) against a head equal to the sum of the design pressure of Containment, the head to the uppermost spray nozzles, and the line and nozzle pressure losses. Each spray pump provides 50% of the design flow to the Containment Spray Headers necessary to maintain Containment pressure below 47 psig.
The pumps are controlled using individual, four-position switches located on the SBF-1 panel. The switch positions are PULL-OUT, Rev. 4, 06/03/1999 - Page 6 -
1 3 - 4 01 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet u
Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 46 Group # 1 WA # 061K6.02 OK Importance Rating 2.6 Knowledge of the effect of a loss or malfunction of the pumps will have on the AFW components Proposed Question: Common 44 Given the following plant conditions:
0 Natural Circulation C/D in progress at 2O"Flhr 0 S/G Atmospheric Steam dumps in manual for C/D 32 ABFP supplying 125 gpm to each of the four S/Gs 0 31 and 33 ABFPs are shutdown in AUTO 0 All S/G levels being maintained at 45%
What would be the effect on the AFW System should 32 ABFP trip on overspeed?
(Assume NO operator actions)
A. No AFW Pump would start causing all four SGs to eventually dry out.
B. Both motor driven AFW Pumps would immediately start and commence feeding all four SGs causing S/Glevels to continually increase.
C. Both motor driven AFW Pumps would start when any one of the four S/G levels decreased to 8% causing SG levels to continually increase.
D. Both motor driven AFW Pumps would start when any one of the four S/G levels decreased to 8% causing SG levels to go to and automatically maintain program value.
Proposed Answer:
.d C. Both motor driven AFW Pumps would start when any one of the four S/G levels
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet decreased to 8%causing S G levels to continually increase.
Explanation (Optional):
Technical Reference(s): SD-21.2 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-AFWOOI 0005 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
, /
7;s :/6 System Description 21.2 Auxiliary Feedwater The ABFP motors are limited to 400 HP. The nominal voltage of their emergency power system limits the motor size to 400 HP. If flow increases to greater than 415 gpm the pump will trip on overcurrent.
(Refer to DBD-303, section 3.1.3.7) 2.4.1 Auto Starts The motor-driven ABFPs start automatically on the following signals (Refer to Figure 21.2-9):
Loss of voltage on the 480V buses 6A (ABFP 33) or bus 3A (ABFP 31). The ABFPs will start 28 seconds after the undervoltage condition is clear. (Time delay relays in the starting circuits provide this time delay.)
Low-Low level (8%)on 2/3 detectors on any steam generator.
Automatic trip of either Main Boiler Feedwater Pump A Safety Injection Signal AMSAC Actuation Signal generated.
If the pump is turned off with an auto start signal present, it will restart in 28 seconds unless in the pull out position.
2.4.2 Motor-drivenABFP Trips ABFP #31 and #33 breakers will trip on the following conditions:
Overcurrent Low Suction flow when recirc valves control switch is in auto.
4 If flow is less than 40 gpm a flow switch will cause the recirculation valve for the effected pump to open. If flow is not greater than 75 gprn within 15 seconds after recirc valve is given open signal, the pump will trip. This trip is disabled when the recirculation valve control switch is in "Open" or "Close".
BUS Strip conditions (Refer to Figure 21.2-8)
+ ABFPs #31 & #33 would be deenergized on undervoltage to their respective buses.
2.4.3 Motor Driven ABFP Controls Each motor-driven ABFP has a control switch on panel SCF in the Control Room. The switch positions are:
START - AFW pump will start if power is available
\u' Rev. 3, 12/02/03 - Page 12 -
System Description 21.2 Auxiliary Feedwater Turning the supply (upper) valve 180 degrees bypasses the positioner and directs the actuation air to the lower valve.
0 Operating the bypass (lower) valve will port control pressure to the valve diaphragm.
FCV-406A has a replacement positioner. It lacks the capability to manually apply I/A pressure to the valve diaphragm. Although it has the "rabbit ears" on the positioner, they are not fully functional.
All the Auxiliary Feed Regulating Valves may also be operated locally by failing the air and using the attached handwheel.
All eight valves have nitrogen backup to instrument air.
Each Auxiliary Feed Regulating Valve has a locked open inlet and outlet isolation valve and a non return check valve between the outlet isolation valve and the Auxiliary Feed Regulating Valve.
2.6.1 Motor-driven pump runout protection The Auxiliary Feed Flow Regulating Valves in the motor-driven pump headers, FCV-406A, B, C, D are provided with a feature that prevents a motor-driven pump runout condition.
Runout can occur when the pumps are supplying feed to steam generators at any pressure. Operating at SG pressure reduced to as low as 110 psig presents the most limiting pump conditions. Runout could result in overheating and damage to the pump motors. High flow rates require high power output from the ABFW pump motors.
Power greater than 400 hp reduces the service factor of the pump, decreasing its reliability. Limitations of the 480 VAC power supplies and their backup diesel generators require strict pump motor current limitations on the motor driven pump.
Pressure transmitter PT-406A(B) signal, at the discharge of each individual pump , is fed to its corresponding Pressure (cutback}
Controller PC-406A(B). Setpoint of this controller is set biased on the total flow output signals from two SG ABFW supply line flow transmitters FT-1200 & 1201 (FT-1202 & 1203) Currently, the Pressure Controller setpoints are adjusted to obtain an ABFW flow rate of approximately 370 to 380 gpm per each pump.
This flow range was selected to provide a margin above minimum required AFWS flow of 345 gpm at the lowest bus voltage conditions.
If the Pressure Controller input signal is <setpoint, the controller output to the HI Signal Selectors increases.
The Hi Selectors PM-406A(B) select the highest signal of either Pressure (cutback) Controller or the operators manual Hand u'
Rev. 3, 12/02/03 - Page 23 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'V Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 47 Group # 1 WA # 062K2.01 OK Importance Rating 3.3 Knowledge of bus power supplies to major system loads Proposed Question: Common 45 The Main Generator just tripped due to a pilot wire transfer trip from Buchanan. The 25x1 sync check relay which ensures synchronization between 6.9 KV buses 5 and 1 for the auto transfer has failed. All other circuits are intact. Which of the following describes the affect to the Reactor Coolant Pumps (RCP)?
A. Only 32, 33, and 34 RCPs will be operating B. Only 32 and 33 RCPs will be operating C. Only 33 and 34 RCPs will be operating D. Only 31, 32 and 33 RCPs will be operating Proposed Answer:
B. Only 32 and 33 RCPs will be operating Explanation (Optional):
6.9 KV bus tie 2-5 utilizes sync check relay between bus 1 and 5 also (25x1). Bus two will not transfer of offsite power. Bus 1 - 3 1 RCP. Bus 2 - 34 RCP.
Technical Reference(s): (Attach if not previously provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-RCSRCP 0 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
System Description 1.3 Reactor Coolant Pump The pumps are operated from CCR panel SAF. RCPs #31, #32, #33 &
- 34 are powered from 6.9 kV buses 1,4,3 and 2 respectively.
RCPs are tripped on underfrequency and undervoltage conditions.
Their setpoints and logic are found in System Description #28, Overall Unit Protection.
2.3.8 Vibration and Noise Indication Two vibration detectors located in the lower motor section are routed to a remote readout panel in the lower electrical tunnel where a measuring device is plugged in to provide measurement. Maximum displacement is 0.002 inch (2 mils), peak to peak.
A complete system is also employed to continuously monitor motor vibrations of both shaft and frame. Two proximity probes per pump are mounted 90 degrees apart near the shaft. One mounted vertically in line with the pump discharge and the other mounted horizontally which is perpendicular to the discharge. Two frame mounted velocity seismoprobes mounted 90 degrees apart are also installed for frame monitoring. The proximitors condition the probes electrical input as well adjusting the return signal to provide an output proportional to probe-shaft gap. The velocity to displacement converters linearize the output from the seismoprobes.
A third probe (Keyphazer) is configured midway between the proximity probes. Sensing a notch in the coupling, it supplies a reference signal used to balance the pump.
Two meters per pump are located next to CCR rack C10. One meter reads shaft vibration (0 to 30 mils) while the other reads frame vibration displacement (0 to 10 mils). The meters read the highest probe and each has a switch enabling the selection of either probe.
The monitors contain alert and danger alarm relays adjustable over the full range of scale and annunciate a common alarm on CCR panel SKF (RCP high vibration). A self-checking circuit provides a green light when the monitor is operable. An analog output jack is available at the back of the monitor as well as output jacks for frequency filters, oscilloscopes and other devices on the face of the panel.
A sixteen point recorder (0 to 30 mils) located below the monitoring panel records all shaft and frame outputs, while two trend recorders can be selected to any RCP to record it's frame and shaft vibration trends.
- /.J Rev. 3, 5/24/99 - Page 16 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'W' Examination 0utI ine Cross-reference: Level RO SRO Tier # 2 WS # 51 Group # 1 WA # 073K1.01 OK Importance Rating 3.6 Knowledge of PRM system design features and / or interlocks which provide for release termination when radiation exceeds setpoints Proposed Question: Common 46 Given the following plant conditions:
0 A plant heatup is in progress in Mode 3 at 450°F.
0 A steam generator tube leak developed.
0 SG activity increases to the SG blowdown isolation setpoint for R-19, S/G Blowdown Liquid Activity Monitor.
0 SG blowdown isolates.
0 Chemistry is requested to verify the Steam Generator activity level.
What is required to allow the Chemist to sample the Steam Generators for activity?
A. Remove the high radiation close signal by pulling the R-'l9 fuses at Bantam 1 I cabinet.
B. Place the SG Blowdown Sample Isolation Valve switches in the PAB Sample Room to the Rad BYP position for the SG to be sampled.
C. Hold the SG Blowdown Isolation Valve control switch on panel SCF in the open position using a device made for this purpose for the SG to be sampled.
D. Only by raising the R-19 alarm setpoint to a value which is above the leaking SG activity.
~-d Proposed Answer:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet B. Place the SG Blowdown Sample Isolation Valve switches in the PAB Sample Room to the Rad BYP position for the SG to be sampled.
Explanation (Optional):
Technical Reference(s): SD-7 (Attach if not previously 3-SOP-SG-1, ARP-040 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-MFW002 0005 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X
.u Question History: 10/1/2002 Diablo Canyon 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
System Description 7 Steam Generator Blowdown a SG tube leak is high activity in the blowdown fluid, as indicated by the R-19 blowdown monitor or by a grab sample. R-19 is common to all SG blowdown lines. If more than one SG is providing blowdown sample flow, then R-19 does not indicate which SG has the tube leak. If R-19 reaches its alarm setpoint, the following occurs:
Blowdown sample containment isolation valves close, 0 Blowdown containment isolation valves close, and 0 City water to blowdown tank isolation valves closes.
Note that closing the blowdown sample containment isolation valves stops flow to the blowdown sampling system and to R-19. If a SG sample is necessary the containment isolation valve control switch is placed in RAD BYPASS. Sampling may be required to:
Identify the affected SG Quantify the magnitude of the leak.
If blowdown is not isolated automatically by an R-19 alarm, the operators isolate blowdown from the leaking SG.If the leaking SG has not been identified, blowdown is secured from all SG. Isolating blowdown reduces the radiological release and reduces the amount of contamination in the blowdown purification system.
The remainder of the actions taken in the SG Tube Leak procedure involve:
Determining whether the magnitude of the leak requires unit shutdown-for example, leak is in excess of the Technical Specificationlimits for primary to secondary SG leakage).
Determining Emergency Classification based on Ieak rate Determining leak rate and shutdown criteria.
Isolating the leaking SG to the greatest degree possible to minimize the radiological release.
If and when the turbine is shutdown, the leaking SG is isolated completely (shut MSIV and bypass).
If and when the reactor is shutdown, the RCS temperature and pressure are reduced to:
+ ToblockSI
+ RCS pressure is further reduced until it is equalized with SG pressure equals reduce SG leakage bd
+ Maintain an adequate minimum subcooling margin.
Rev. 4, 02/27/2001 - Page 21 -
3-ARP-040 Rev. 1 Page 8 of 3$
INPUT DEVICE: 3RM019 R19 S.G.
BLOWN SETPOINT: Variable 9.0 CAUSES I.1 Steam generator blowdown high activity (SG tube leak) 2.0 AUTOMATIC ACTIONS 2.1 SG blowdown and sample valves close.
ex 3.0 SUBSEQUENT ACTIONS 3.1 IF:a Steam Generator tube leak is suspected, THEN GO TO 3-AOP-SG-1, Steam Generator Tube Leak.
3.2 GO TO ONOP-RM-2, High Activity - Radiation Monitoring System.
4.0 REFERENCES
- 4. f 3-AOP-SG-1, Steam Generator Tube Leak.
4.2 ONOP-RM-2, High Activity - Radiation Monitoring System.
I' I
No:3-SOP-SG-001 Rev: 34 STEAM GENERATOR BLOWDOWN SYSTEM .
Page 20 of 74 L ~
d CAUTION The Rad Bypass position SHALL NOT be used if the unit is in Modes 1 OR 2.
NOTE In accordance with Con Edison - Memorandum Of Understanding No. 15, the Plant Manager (currently titled General Manager Operations) of IP3 SHALL authorize use of emergency surge waste space (Le., SGBD capacity) at Con Edison's Unit No. 1 waste collection tanks.
4.5 Establishing SGBD to Unit 1 4.5.1 COMPLETE SGBD initial lineup per Step 4.1 for SGs selected for blowdown to Unit I.
4.5.2 REQUEST Unit 2 Shift Manager to verify the following:
- P-MT-122, Pressure Test of Secondary Boiler Blowdown Purification System, has been performed.
- Unit 3 to Unit 1 Secondary Boiler Blowdown Purification System (SBBPS) is available.
4.5.3 PLACE Containment Isolation valves control switches for affected SG(s) in RAD. BYPASS.
31 SG:
PCV-1214:
PCV-1214A:
Blowdown lsol Vlv 1 31 Steam Generator Blowdown lsol Vlv 2 31 Steam Generator 32 SG:
PCV-1215:
PCV-1215A:
Blowdown lsol Vlv I32 Steam Generator Blowdown lsol Vlv 2 32 Steam Generator 33 SG:
- PCV-1216: Blowdown lsol Vlv I33 Steam Generator
- PCV-1216A: Blowdown Is01 Vlv 2 33 Steam Generator 34 SG:
PCV-1217:
PCV-1217A:
Blowdown lsol Vlv 1 34 Steam Generator Blowdown Is01Vlv 2 34 Steam Generator 4.5.4 OPEN BD-23, Steam Generator Blowdown To Unit 1 Header Isolation.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet e
Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 52 Group # 1 KIA # 076K2.01 OK Importance Rating 2.7 Knowledge of bus power supplies to the service water Proposed Question: Common 47 Service Water Pumps 34, 35 and 36 are lined up to the essential header, their Zurn Strainers are directly energized from ...
A. 5A, 2A and 6A
- 6. 5A, 3A and 6A C. MCC312A D. MCC 36A and 366 Proposed Answer:
D. MCC 36A and 36B Explanation (Optional):
Technical Reference(s): 3-COL-EL-001 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE u
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Learning Objective: 13LP-ILO-SW001 0004 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
J 6900 AND 480 VOLT AC DISTRIBUTION NO 13-COL-EL-I Rev: 38 i I Page 22 of 62 J f"
Des.
Oper Actual Pos. Pos. Initial Verifier 3.2 -
P A 6 55' Elev.
3.2.1 MCC-36A 31 Electrical Tunnel Exhaust Fan On Charging Pumps Seal Leakage Collection Tank Pumps A 8, B Off 31 33 35 SW Pump Strainers &
Throw-Over Switch for 37 38 39 SW Pump Strainers On RC-MOV-536 Pressurizer PORV PCV-456 isolation On AC-MOV-822A 31 RHR Heat Exchanger CCW Outlet Isolation On AC-MOV-745B 32 RHR Heat Exchanger Inlet Isolation On r' SI-MOV-746 32 RHR Hx Outlet Injection isolation AC-MOV-730 RHR Loop Suction On Isolation Off AC-MOV-797 RCP CCW Supply kolation On AC-MOV-784 RCP CCW Bearing Return Isolation On AC-FCV-625 RCP CCW Thermal Barrier Return Isolation On AC-MOV-744 RHR Pumps Locked Discharge Isolation Off SI-MOV-866A 31 Spray Pump Discharge isolation On SI-MOV-856G Loop 31 Hot Leg High Head injection Line BIT Locked Header Isolation Off SI-MOV-880A 31 FCU Charcoal Filter Dousing Isolation On SI-MOV-880C 32 FCU Charcoal Filter Dousing Isolation On Si-MOV-880E 33 FCU Charcoal Filter Dousing Isolation On
1' 6900 AND 480 VOLT AC DISTRIBUTION N0 :3-COL-EL-I Rev: 38 Page 25 of 62 Des.
Oper Actual Pos. Pos. Initial Verifier BFD-MOV-2-31 31 Main Boiler Feed Pump Discharge Isolation On SI-MOV-856E Loop 31 Cold Leg High Head Injection Line BIT Header Isolation On SI-HCV-640 32 RHR Hx Outlet Flow Control Valve On Spare Off 33 Control Building Exhaust Fan On Spare Off Spare Off Spare Off Spare Off Spare Off Spare Off 3.2.2 - .MCC=36B 32 Electrical Tunnel Exhaust Fan On Backup Feed to Instrument Busses 34 & 34A On 32 34 36 SW Pump Strainers &
Throw-Over Switch for 37 38 39 SW Pump Strainers On RC-MOV-535 Pressurizer PORV PCV-455C Isolation On AC-MOV-8220 32 RHR Heat Exchanger CCW Outlet Isolation On AC-MOV-745A 32 RHR Heat Exchanger Inlet Isolation On SI-MOV-899A 32 RHR Hx Outlet Injection Isolation On AC-MOV-731 RHR Loop Suction Isolation Off AC-MOV-769 RCP CCW Supply Isolation On l AC-MOV-786 RCP CCW Bearing Return Isolation On
INSTRUCTOR LESSON PLAN Presentation Data Activity/Notes
'd b) Annunciated by:
(1) High strainer AP:
i (a) > 7 psid for 30 seconds (2) Loss of power to control circuit:
(a) Strainer motor overload (b) Blown fuses to control circuit (c) Loss of 480VAC to strainer motor
- 2) BACKUP SERV. WATER STRAINERS TROUBLE:
a) CR panel SJF b) Same conditions as main SWS strainer trouble alarm
- 3) If AP reaches 9 psid, System Engineering evaluation required I-SOP-RW-005 1.4.2.2 i
- g. Powered Supplies: :O 0004
- 1) 3 1, 33 and 35:
a) MCC-36A
- 2) 32,34 and 36:
a) MCC-36B
- 3) 37,38 and 39:
a) Either MCC-36A or MCC-36B b) Throw switch in 480V switchgear room
- h. Instrumentation: 0 0002
- 1) 0-10 psid indication for each Zurn strainer [OD 91-3-128 a) Normal AP = 4 psid
- 2) Amber light - overload
- 3) Zurn Strainer Pit Temperature Monitor: 'OD 93-3-376 a) Alarms at 45'F
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d'
Examination Outline Cross-reference: Level RO SRO Tier # 2 ws ## 53 Group ## 1 WA # 07862.4.4 OK Importance Rating 4.0 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.
Proposed Question: Common 48 Given the following:
75%power 0 PRZR Pressure 2265 slowly increasing 0 PRZR spray valves closed 0 Letdown Orifice Valves closed 0 Charging flow increasing 0 PRZR Level 54% and increasing Which one of the following malfunctions would cause these indications?
A. Controlling Pressurizer Pressure Channel Failed Low B. Controlling Pressurizer Level Failed Low C. Pressurizer Pressure Master Controller Failure D. Loss of Instrument Air Proposed Answer:
D. Loss of Instrument Air Explanation (Optional):
L.l
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 3-AOP-AIR-1 (Attach if not previously SD-3.0 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-IA001 7 (1799) (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X
...- . l 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
Air Systems Malfunction 3-AOP-AIR4 Rev. 02 Page 31 of 61 Attachment 1 r Valves of Immediate Concern W Page I of 17 CHEMICAL AND VOLUME CONTROL SYSTEM VALVE FUNCTION FAIL POSITION CH-LCV-459 LETDOWN ISOLATION VALVE Closed CH-LCV-460 LETDOWN ISOLATION VALVE closed CH-AOV-ZOOA LETDOWN CONTROL VALVE Closed CH-AOV-2008 LETDOWN CONTROL VALVE Closed CH-AOV-2OOC LETDOWN CONTROL VALVE Closed CH-AOV-201 LETDOWN ISOLATION VALVE Closed CH-AOV-202 LETDOWN ISOLATION VALVE Closed I CH-FCV-11OA I BORIC ACID BLENDER BORIC ACID FLOW CONTROL VALVE 1 I
I CH-FCV-1106 I VCT MAKEUP VALVE 1 Closed r
1 1
CH-FCV-1 11A MAKEUP H20 TO BORIC ACID BLENDER Closed CH-FCV-1 I1B VCT MAKEUP VALVE Closed i
d CH-HCV-142 CHG. LINE FLOW CONTROL VALVE Closed CH-AOV-261A 31 SEAL RETURN ISOIATION VALVE Open CH-AOV-261 B 32 SEAL RETURN ISOLATION VALVE Open CH-AOV-261C 33 SEAL RETURN ISOLATION VALVE Open CH-AOV-261D 34 SEAL RETURN ISOIATION VALVE Open CH-AOV-246 RCPS NO.l SEAL BYPASS Closed CH-HCV-123 EXCESS LTDN FLOW CONTROL Closed
~-~
CH-HCV-133 RESID HR LP BYPASS TO DEMIN Closed CH-AOV-204A LOOP 32 ALTERNATE CHARGING ISOLATION Open CH-AOV-2048 LOOP 37 NORMAL CHARGING ISOIATION Open BATCHING TANK AUXILIARY STEAM CH-TCV-lOO Open TEMPERATURE CONTROL VALVE
.J Air Systems Malfunction 3-AOP-AIR-1 Rev. 02 Page 33 of 61 Attachment 1 Valves of Immediate Concern Page 3 of 17 VALVE FUNCTION FAfL POSITION RC-PCV-455A PRZR SPRAY VLV LOOP 34 Closed RC-PCV-455B PRZR SPRAY VLV LOOP 33 Closed NNE-AOV-863 ACCUMUIATOR N2 SUPPLY VALVE Closed PRIMARY WATER CONTAINMENT ISOLATION RC-AOV-5 19 Closed VALVE 1 RC-AOV-552 PRfMARY WATER CONTAINMENT ISOLATION
/VALVE I Closed I I RC-AOV-548 I PRESSURIZER RELIEF TANK TO GAS ANALYZER ISOLATION Closed PRESSURIZER RELlEF TANK TO GAS RC-A0V-549 ANALYZER ISOLATION Closed I
RC-AOV-550 I N2 TO PRT I Closed A 4 SAFETY JNJECTION SYSTEM VALVE I FUNCTION FAIL POSITION I
1 S1-A0V-'851A SI-AOV-1851B I BIT RECIRCULATION VALVES I
SI-AOV-876A SPRAY ADDITIVE TANK DISCHARGE SI-AOV-8766 ISOLATION Open
System Description 3.0 Chemical and Volume Control System from the alternate source, only local operation of the pump breaker at MCC-312A is available. Normal speed control is available if instrument buses are energized.
Charging pump speed and flow rate is automatically controlled as a function of pressurizer level. The pressurizer level programmer (TC-412N in control room rack D-8) produces an output linearly proportional to T, therefore a constant RCS mass is maintained during load changes and steady state operation. This output is compared, by LC459D in rack B-6, to actual pressurizer level (see graph RCS-10) and the resulting output represents the error signal.
The output is transmitted from the master level controller to the three charging pump speed controllers located on flight panel FBF. When the master auto-manual controller is in the AUTO position, it passes the pressurizer level error signal to the three pump speed controllers and controls which ever pump is in AUTO. When the master auto-manual controller is in MANUAL, the pressurizer error signal is blocked and the operator can control the pump speed by adjusting the master auto-manual controller. The output of the individual pump controller is used to drive the positioner to properly position the scoop tube in the fluid drive coupling. The level programmer (TC-412N) is a proportional controller (reset is dialed out), the master controller (LC-i 459D) is a proportional-integral-derivative controller (derivative is set
, Ll to zero), and the charging pump speed controllers are proportional controllers. Each pump has an associated run time meter which displays (run time in hours) inside the flight panel.
Each charging pump has an adjustable low speed stop on the scoop tube to ensure the reactor coolant pumps have seal water flow at all times. This stop is adjustable to allow a higher flow to be set if mechanical seal leakage increases. If the electrical signal from the control room speed controller is lost, the pumps will go to minimum speed. When this occurs, the pump speed can be controlled from the local panel on the 55 foot PAB elevation by placing the controllers in MANUAL. If instrument air is lost due to a slow leak, 31, 32 and 33 charging pumps will fail to maximum speed. If the loss of instrument air is a sudden catastrophic loss the simultaneous loss of pressure on the upper and lower chambers of the piston will cause the actuator to fail as is. In this situation, pump speed can be controlled by isolating and venting the air to the scoop tube positioner and manually positioning it. This is listed as a subsequent action in ONOP-IA-1, Loss of Instrument Air.
A Low Charging Flow alarm on CVCS supervisory panel SFF is derived from the magnitude of the signal to the pump speed U
controllers. If the pressurizer level is above its setpoint for an Rev. 6, 12/18/2002 - Page 23 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet L/
Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 54 Group # 1 KIA # 103A3 .O1 OK Importance Rating 3.9 Ability to monitor automatic operation of the containment system including containment isolation Proposed Question: Common 49 Given the following conditions:
0 A Large Break LOCA has occurred 0 Containment pressure is 27 psig Containment Spray failed to AUTO actuate 0 Manual Containment Spray actuation was successful 4
What affect does this condition have on Containment Phase B isolation?
A. Phase B will automatically actuate by the Manual Containment Spray signal.
- 6. Phase B must be manually actuated using 1 of 2 pushbuttons.
C. Phase B must be manually actuated using 2 of 2 pushbuttons.
D. The Phase B valves must be manually closed using individual control switches in the control room.
Proposed Answer:
A. Phase B will automatically actuate by the Manual Containment Spray signal.
Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): SD-10.0 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-I LO-SIso01 c (As available)
Question Source: Bank ##
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
System Description 10.0 Engineered Safeguards Systems I (Second Omortunitv ~
I 32 CS Pump Immediately With Containment Spray and No SI signal (SI reset) 2.3.10 Containment Spray (Figure 10.0-6 & 7)
The Containment Spray system reduces containment pressure and removes fission products from containment atmosphere. Pressure is reduced by condensing steam present inside the containment. Fission products (particularly 1-131) are removed by absorption in the spray water. Sodium Hydroxide is added to the spray water to create a more soluble form of Iodine in the containment and recirculation sumps.
Containment pressure is monitored by 6 pressure detectors arranged into two groups (948A, 948B, & 948C and 949A, 949B, & 949C). When two of three pressure channels in both groups reach 22 psig, an automatic Containment Spray Actuation occurs.
Manual Containment Spray is actuated using 2 of 2 pushbuttons. This electrical arrangement prevents inadvertent actuation of Containment Spray.
An Automatic or Manual Containment Spray Signal will cause:
0 Actuation of Containment Spray Containment Phase B Isolation Containment Ventilation Isolation As with other actuation signals, it may be necessary to manually control components while containment conditions still require a containment spray (i.e./ pressure > 22 psig). The containment spray signal can be reset by depressing train specific reset pushbuttons. The containment spray signal is "blocked" and components that receive containment spray signals can be manually operated. If containment pressure decreases to less than 22 psig, and the containment spray actuation signal is removed, and the re-initiation "block" is removed.
This means that if containment pressure subsequently increases to greater than 22 psig, containment spray will re-actuate.
A Containment Spray actuation signal will:
Send an open signal to spray pump motor operated discharge valves 866A and 866B, and Start the 31 (32) Containment Spray Pumps under one of the three following conditions:
Rev. 3, 08/07/2000 - Page 25 -
I I NOTES
- 1. SPRAY ACTUATION 81-STABLES ARE ENEROZED TO AClUATE.
- 2 PRESSURE El-STABLES SET TO A NATE AT APPRO% 22 PSC (CONTAINMENT HIGH-HIM PRESS!!.
LOGIC DIAGRAMS oc'0'- SAFEGUARDS ACTUATION llEv SHEET-1 11 5651072
- 12
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 50 Group ## 1 KIA ## 064K2.03 OK Importance Rating 3.2 Knowledge of bus power supplies to the control power Proposed Question: Common 50 An electrical short caused a loss of 33 DC Power Panel. Subsequently an inadvertent Safety Injection signal was generated. What would be the configuration of the Emergency Diesel Generators following the SI?
A. ALL three EDGs would be running.
B. Only 31 and 32 EDGs would be running.
C. Only 32 and 33 EDGs would be running.
D. Only 31 and 33 EDGs would be running.
Proposed Answer:
C. Only 32 and 33 EDGs would be running.
Explanation (Optional):
Technical Reference(s): 3-AOP-DC-1, Attachment 9 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE
~~
Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Learning Objective: 13LP-ILO-EDSEDG 3 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Loss Of A 125V DC Panel 3-AOP-DC-1 Rev. 1 Page 49 of 65 Attachment 9
, Loads on 33 DC Power Panel
'W Page 1 of 'l Ckt Load Comment
' 480 V SWGR 31 Bus 2A Bkr Cont. & Bus ZA, 3A Safeguards r 2 I 480 v SWGR 32 BUS 3~ Bkr Control 3 33 - 34 DC Power Panel Tie Breaker 4 Diesel Gen 31 Control Panel Supervisory Panel SC - Aux. BFP 31 Recirc Valve (SOV-1321) 6 34 Cont. Recirc Fan SOV I 7 I 32 Cont. Recirc Fan SOV Supervisory Panel SB Independent Indication I 9 1 Service Water Mode Selector Switch I I IO 1 Spare I ~~
11 Spare
~4 I
12 Spare I W'
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet u
Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 55 Group # 1 WA # 103A4.04 OK Importance Rating 3.5 Ability to manually operate and / or monitor in the control room Phase A and Phase B resets Proposed Question: Common 51 The plant was operating in the normal full power lineup. An inadvertent Safety Injection and Phase A Isolation occurred. Safety Injection has been RESET. What action will allow Letdown Isolation Valves, 201 and 202, to be opened?
A. Depress both Phase A master Reset pushbuttons and the valves will reopen.
B. Depress both Phase A master Reset pushbuttons then put 201 and 202 valves
%e switches to close and then back to open and the valves will reopen.
C. Depress both individual Reset pushbuttons then put 201 and 202 valves switches to close and then back to open and the valves will reopen.
D. Depress both Phase A master Reset pushbuttons then depress the individual valve Reset pushbuttons and the valves will reopen.
Proposed Answer:
D. Depress both Phase A master Reset pushbuttons then depress the individual valve Reset pushbuttons and the valves will reopen.
Explanation (Optional):
Technical Reference(s): 3-ES-1.l, Attachment 3 (Attach if not previously provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-CVCOOI 5.0 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
I Number:
Title:
Revi s i on Number :
t ES-1.1 SI T E R M I N A T I O N 17 L '
L v
ATTACHMENT 3 (Attachment page 14 of 19)
RE-ESTABLISHING OPERATOR CONTROL OF VALVES FOLLOWING PHASE A RESET
( S E C T I O N I 1 c o n t i n u e d from p r e v i o u s page)
2 CA-PCV-1229 - SJAE t o V . C .
3 AC-AOV-798 - CCW l n l e t t o Excess.
L t d n . Hx 4 AC-AOV-791 - CCW I n l e t t o Excess.
t W Ltdn. Hx 5 AC-AOV-796 - CCW O u t l e t f r o m Excess L t d n . Hx 6 AC-AOV-793 - CCW O u t l e t from Excess L t d n . Hx 7 RC-AOV-548 - PRT t o Gas A n a l y z e r 8 RC-AOV-519 - PW Cont. 1501.
9 VS-PCV-1235 & 1237 - V.C. Rad. M o n i t o r PS-PCV-1239 & 1241 - WCCPP t o VC Rad M o n i t o r 10 VS-PCV-1234 & 1236 - V.C. Rad. M o n i t o r PS-PCV-1238 & 1240 - WCCPP t o VC Rad M o n i t o r t I d 11 12 31 33
( S E C T I O N 11 CONTINUED ON NEXT PAGE)
CH-AOV-202 CH-AOV-201 RC-AOV-552 RC-AOV-549 Page 48 o f 54 Letdown Letdown PW Cont. I S O l .
PRT t o Gas A n a l y z e r b
ES-40 I Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet e Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 56 Group # 2 WA # 00 1K3.02 OK Importance Rating 3.4 Knowledge of the effect that a loss or malfunction of the CRDS will have on the RCS Proposed Question: Common 52 The following conditions exist on Unit 3:
0 The plant was operating at 100% power when a Main Turbine Control Oil leak caused a very slow turbine load rejection to 90% power 0 Prior to and during the load rejection Rod Control was in Manual Tave increased from 567°F to 575°F 0 The CRS orders Rod Control be placed in MANUAL J
'- When Rod Control is placed from MANUAL to AUTO, Control Rod Speed should go from steps per minute (SPM) in MANUAL to SPM in AUTO?
A. 72 SPM in MANUAL, 66 SPM in AUTO B. 66 SPM in MANUAL, 72 SPM in AUTO C. 72 SPM in MANUAL, 8 SPM in AUTO D. 66 SPM in MANUAL, 8 SPM in AUTO Proposed Answer:
B. 66 SPM in MANUAL, 72 SPM in AUTO Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): SD-16.1 page 28 and Figure (Attach if not previously
-' 16.1-28 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-ICROD G (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: 7/17/2002 Braidwood 1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 57 Group # 2 WA # 002K6.02 OK Importance Rating 3.6 Knowledge of the effect or a loss or malfunction of the RCP on the RCS Proposed Question: Common 53 Unit 3 was operating at 28% power when 31 Reactor Coolant Pump (RCP) tripped on overcurrent.
Which of the following describes the unit's initial response? (Assume NO operator action AND NO rod motion.)
A. A reactor trip occurs and unaffected loops TAVEdecreases.
B. A reactor trip occurs and unaffected loops TAVEincreases.
C. A reactor trip will NOT occur and unaffected loops TAVEdecreases.
D. A reactor trip will NOT occur and unaffected loops TAVEincreases.
Proposed Answer:
D. A reactor trip will NOT occur and unaffected loops TAVEincreases.
Explanation (Optional):
Technical Reference(s): SD-28 (Attach if not previously 3-AM-002 page 4 provided)
Simulator
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-ICRXP E-3 (As available)
Question Source: Bank # INPO 20589 Modified Bank # (Note changes or attach parent)
New Question History: 6/29/2000 Byron 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
ARP-002 Rev. 14 i Page 4 of 42 d
INPUT DEVICE: Loop Flow RCP Breakers FC-4 14 52b RCP-31x FC-415 LF-1X LOSS OF RCP31 FC-4 16 FLOW SINGLE LOOP FC-424 52b RCP-32~
FC-425 LF-2X RCP32 FC-426 FC-434 52b RCP-33~
FC-435 LF-3X RCP33 FC-436 FC-444 52b RCP-34~
FC-445 LF-4X RCP34 FC446 4
NOTE This trip and alarm are bypassed below P-8 permissive (approximately 35% power).
I.o CAUSES I.1 Low flow in any 1 of 4 loops as indicated by 2 out of 3 flow transmitters in loop when power is greater than P-8 setpoint.
1.2 Any Reactor Coolant Pump breaker open when power is greater than P-8 setpoint.
2.0 AUTOMATIC ACTIONS 2.1 If power is greater than P-8 the reactor trips.
(CONTlNUED ON THE NEXT PAGE)
System Description 28.0 Overall Unit Protection
[' 2.2.2 P-7 Permissive (Figure 28-15)
The P-7 permissive is used to block the high pressurizer level, low d
pressurizer pressure reactor trips, reactor coolant low flow and undervoltage reactor trip signals to the reactor protection system. The P-7permissive is activated by a bistable circuit indicating less than 10%
power as measured by both turbine first stage pressure detectors and 3/4 power range channels. The power range input is supplied by the P-10 permissive. A white "POWER BELOW P-7 lamp illuminates on the control room FBF panel while the P-7 permissive is active and extinguishes when reactor power and/or turbine power are >lo%.
2.2.3 P-8 Permissive (Figure 28-14)
The P-8 permissive blocks the automatic reactor trip on low flow in one loop if power is below 35% at the time one RCP is lost. The permissive also blocks a reactor trip due to a turbine trip when power is below 35%. A white "POWER BELOW P-8" lamp illuminates on the control room FBF panel when the P-8 permissive is active and extinguishes when reactor power is above 35%.
2.2.4 P-10Permissive (Figure 28-14)
The P-10 permissive blocks the intermediate range channel and low power range channel trips during an approach to power. It is also used to backup the P-6 permissive to block the Source Range instrumentation and is one of the inputs to the P-7 permissive.
When 2 of 4 power range channels indicate greater than 8.5%power the P-10 permissive is activated and a white "POWER ABOVE P-10" lamp illuminates. Once the P-10 lamp is lit, the low power and intermediate range hi flux trips may be manually blocked as described in the sections for those trips.
The P-10 permissive and associated manual blocks are automatically reinstated if power falls below 8.5%on 3/4 Power Range channels.
2.2.5 Low Power Auto Rod Withdrawal Block Automatic control rod withdrawal is blocked until turbine power, as sensed by PT-4124 (turbine first stage pressure), is greater than 15%.
Automatic control rod insertion is not blocked. A white "LOW PWR AUTO ROD WITHDRAWAL BLOCK" light, on the FBF panel, is illuminated when the permissive is active and extinguishes when turbine power is >15%.
4 Rev. 3, 06/29/2000 - Page 23 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 48 Group # 1 WA # 063A3.01 OK Importance Rating 2.7 Ability to monitor automatic operation of the DC electrical system meters, annunciators, dials, recorders and indicating lights Proposed Question: Common 54 Given the following plant conditions:
0 A BATTERY CHARGE TROUBLE category alarm was received in the control room e The conventional NPO reports the RED light for + (positive) Ground Detection for 31 Battery Charger is LIT
~ L J What actions are required (if any) to clear the alarm in the control room making it available to alarm on any future alarm condition?
A. Push the Acknowledge and Reset pushbuttons in the control room.
B. Place the NormaVBypass switch on 31 Battery Charger in Bypass position.
C. Open 31 Battery Charger output breaker.
D. Cannot be cleared until the + ground condition is corrected.
Proposed Answer:
B. Place the Normal/Bypass switch on 31 Battery Charger in Bypass position.
Explanation (Optional):
~u' Technical Reference(s): SD-27.5 (Attach if not previously
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EDSI25 E-7 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 L./
55.43 Comments:
'W'
I' System Description 27.5 Electrical Systems: Low Voltage 125 VDC, 118 VAC & Plant Lighting clear the input to the common alarm on supervisory panel SHF, and reset the HV S / D circuit for the battery charger.
FLOAT ADJUST--screw type adjust potentiometer used to adjust the nominal float voltage that the charger will maintain on the battery.
EQUALIZE ADJUST--screw type adjust potentiometer used to adjust the equalizing voltage used when performing an equalizing charge on the battery.
EQUALIZEFLOAT SELECT TOGGLE S WITCH--this toggle switch has two positions, equalize and float, and is used to select which type of charge is being performed on the battery.
ALARM BYPASS TOGGLE SWITCHES--Six toggle switches are provided under a plexiglass cover on the charger control panel to bypass individual alarm functions in the event of a spurious alarm circuit or for maintenance. The switches have two positions, normal and bypass. In NORMAL, each of the alarm functions input to the common "BATTERY CHARGER TROUBLE" alarm on supervisory panel SHF. In BYPASS, the inputs to the common alarm are disabled and an amber light above the switch lights. This switch will not affect the local alarm buzzer on the charger control panel. The following alarms are inputs to the common alarm on SHF Erom 3 1 and 32 battery chargers:
NOTE: The local alarms for 35 battery charger are not wired into the supervisory alarm on SHF LOW DC VOLTAGE GROUNDS POS AND NEG ACFAILURE BATTERY DISCHARGE 0 OVERTEMPERATURE CHARGER SHUTDOWN COMMON ALARM BUZZER ON/OFF TOGGLE S WITCH--this switch is used to turn on and off the common alarm buzzer at the battery charger. An amber light above the switch will light when the switch is in the off position. This switch disables the local buzzer for the following alarms:
0 LOW DC VOLTAGE 0 GROUNDS POS AND NEG ACFAILURE Rev. 3, 05/03/2005 - Page 10 -
System Description 27.5 Electrical Systems: Low Voltage 125 VDC, 118 VAC & Plant Lighting 0 BATTERY DISCHARGE 4 0 OVERTEMPERATURE AC VOLTAGE SELECTOR SWITCH-this switch is used to select which voltage will be read on the AC voltmeter. It is a four position switch, Ll-L2, L2-L3, L3-L1 and OFF.
AC AMPERAGE SELECTOR SWITCW--this switch is used to select which of the phases will be read on the AC ammeter. It's a (3) position switch, 1,2 and 3.
2.1.3.3 Indications 3 1, 32, and 35 Chargers (Figure 27.5-4)
AC ON--a green light on the control panel indicating that AC power is available at the charger @e. the AC input breaker is closed and power is available from the MCC).
AC FAILURE--a red light indicating that AC power to the charger has been lost.
DC VOLTAGE--a 0-200 volt meter that indicates the DC voltage out of the charger.
DC AMPS--a 0-500 amp meter that indicates the DC load on the battery charger.
GROUND DETECTION--Two red lights are provided for ground detection, 1 for (+) ground and 1 for (-) ground which monitor the
(+) and (-) terminal voltage to chassis ground. With no ground present, both lights are de-energized. If a ground condition occurs, the light will illuminate. Once the relay has been energized and the light lit for this type of ground detection circuit, the relay must be reset. This is done with the ground detection relay reset switch, HIGH VOLTAGE SHUTDOWN-this is a red light on the control panel that will light if the charger has been shutdown due to high voltage. The setting is 146 volts DC. Once the charger is shutdown on high voltage the reset toggle is used to reset the charger to operation and clear the local and common category alarm on SHF. This alarm input to the common battery trouble alarm on SHF cannot be bypassed.
LOW VOLTAGE--a red light on the control panel to indicate that the DC output of the charger is below 116 volts .
OVERTEMP--a red light on the control panel to indicate an over temperature condition exists Le., >160°F.
BATTERY DISCHARGE--an amber light on the control panel to indicate that the battery is no longer receiving charging current.
This is done by monitoring the direction of current flow in a DC Rev. 3, 05/03/2005 -Page 11 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination 0utline Cross-reference: Level RO SRO Tier # 2 WS # 58 Group # 2 KIA ## 01 1K5.15 OK Importance Rating 3.6 Knowledge of the operational implications of PZR level indication when RCS is saturated as it applies to the PZR LCS Proposed Question: Common 55 Given the following conditions:
0 A Loss of Offsite Power has occurred.
0 RCS cooldown is being performed in accordance with ES-0.2, Natural Circulation Cooldown.
0 RCPs cannot yet be started.
0 The RO is depressurizing using auxiliary spray.
d 0 PRZR level rapidly rises from 24% to 66%.
Which one of the following describes the reason for the Pressurizer level increase?
A. Loss of Secondary Heat Sink.
B. Charging flow is refilling the Pressurizer as RCS pressure drops.
C. Cooldown rate is not high enough to maintain Pressurizer level with auxiliary spray in service.
D. Portions of the RCS have reached saturation temperature.
Proposed Answer:
D. Portions of the RCS have reached saturation temperature.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
Technical Reference(s): ES-0.2 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOPE00 G (As available)
Question Source: Bank # INPO 24956 Modified Bank # (Note changes or attach parent)
New Question History: 12/1/2002 Beaver Valley 1 Question Cognitive level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Number:
Title:
R e v i s i o n Number:
ES-0.2 NATURAL CIRCULATION COOLDOWN 17
\.J
- 14. C O N T I N U E RCS C o o l d o w n And DeDressurization:
a . M A I N T A I N cooldown r a t e i n RCS c o l d l e g s - L E S S T H A N 25" F I H R b . M A I N T A I N RCS s u b c o o l i n g b . STOP d e p r e s s u r i z a t i o n a n d b a s e d on q u a l i f i e d c o r e REESTABLISH subcooling.
e x i t T C s - GREATER T H A N 9 0 ° F c . M A I N T A I N RCS c o l d l e g t e m p e r a t u r e and p r e s s u r e -
W I T H I N L I M I T S OF O P E R A T I O N S GRAPH R C S - 1 6
- 15. VERIFY R e a c t o r V e s s e l - FREE PERFORM t h e f o l l o w i n g :
OF STEAM VOIDS a . R e - P R E S S U R I Z E RCS w i t h i n 0 P R Z R Level - NO U N E X P E C T E D t h e l i m i t s of LARGE V A R I A T I O N S O p e r a t i o n s Graph R C S - 1 B t o collapse potential voids 0 R V L I S F u l l Range - G R E A T E R and CONTINUE cooldown.
T H A N 100%
- b. RCS d e p r e s s u r i z a t i o n must c o n t i n u e , T H E N GO T o ES-0.3, NATURAL C I R C U L A T I O N COOLDOWN W I T H S T E A M V O I D I N VESSEL ( W I T H R V L I S ) .
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'.d Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 59 Group # 2 WA # 014K1.02 OK Importance Rating 3.0 Knowledge of the physical connections and / or cause effect relationships between the RPlS and the NIS Proposed Question: Common 56 The following conditions exist on Unit 3:
Reactor power is holding steady at 1E-8 amps during a normal reactor startup 0 Individual and group position indicators show all control bank D rods at 120 steps withdrawn When the ATC begins to withdraw control rods to raise reactor power, both IR NIS 4
indications suddenly drops by 1/3 decade and continues to decrease at a negative (-)
0.25 DPM. There is no significant change in RCS TA"~.The control bank D step counters now read 121 steps for groups 1 and 2. IRPl indicators for rods P-IO, K-2 and H-8 (CBD Group 2) indicate 0 steps. All other rod position indicators (IRPls) are unchanged.
Which of the following has occurred based on these indications?
A. The control bank D group 2 step counter has failed; it should also read 0 steps if the rods in this group are fully inserted.
B. The individual rod position indicators appear to have failed, more than a single dropped rod would have resulted in a reactor trip.
C. The control bank step counters and associated IRPl indicators, along with the NIS indications are consistent with multiple dropped rods.
D. Either the control bank D group step counter or 3 IRPl indicators have failed, not enough information is provided to determine which.
ES-40I Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
C. The control bank step counters and associated IRPl indicators, along with the NIS indications are consistent with multiple dropped rods.
Explanation (Optional):
Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-ICROD C (As available)
Question Source: Bank # INPO 21426 Modified Bank # (Note changes or attach parent)
New Question History: 7/17/2002 Braidwood 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3,9 55.43 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet,
'4 Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 49 Group # 1 KIA ## 063K3.02 OK Importance Rating 3.5 Knowledge of the effects that a loss or malfunction of the DC electrical system will have on components using DC control power Proposed Question: Common 57 Given the following:
The plant is at 100% power.
0 Safeguards Train B, DC Power has failed.
Which ONE of the following describes the response of AFW Pump 33 to a safeguards actuation?
L l A. Pump will auto start but does NOT supply water to any S/G.
B. Pump will auto start and supplies water to only 34 S/G.
C. Pump will auto start and supplies water to 33 and 34 S/Gs.
D. Pump will NOT auto start.
Proposed Answer:
D. Pump will NOT auto start.
Explanation (Optional):
Technical Reference(s): SD-10.0 (Attach if not previously
~.4' provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-EDSI25 E-5 (As available)
Question Source: Bank# INPO 28414 Modified Bank # (Note changes or attach parent)
New Question History: 11/15/2004 Kewaunee, Unit 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
System Description 10.0 Engineered Safeguards Systems I 2.4.5 Containment Spray Reset (Figure 10.0-6& 7)
The Containment Spray reset signal is reset from CR panel SBF-1. Two
'4 black pushbuttons are provided, one for each safeguards logic channel.
The spray signal can be reset with the Hi-Hi VC Pressure present, however, automatic Containment Spray reactuation will be prohibited until the High-High Containment Pressure signal has cleared. Manual actuation of the Containment Spray Signal can be initiated at any time.
Engineered Safeguards Power Supplies Power for ESS equipment is derived either directly from the four 480 safeguards volt buses or via the safeguards motor controI centers.
These buses can be powered from one of two sources, either the 6.9 kV buses (via transformers) or from the emergency diesel generators, The arrangement of the equipment on the 480 V buses ensures that minimum safeguards equipment are available with a single active failure. Accommodations for testing of components are also provided.
Any loss of a single component or power supply still satisfies the minimum safeguards equipment.
Two logic circuits (Train A and Train B) exist for safeguards actuation.
The logic relays associated with each actuation signal are powered
' 4 from DC Distribution Panel 31 (Train A) and DC Distribution Panel 32 (Train B). The sequence signals associated with Train A and Train B provide signals in a manner that each would satisfy the minimum safeguards equipment. Train A feeds buses 5A, 2A, and 3A. Train B feeds buses 6A, ZA, and 3A. The Bus Sequencing relays are powered as follows: DC Power Panel 31: Bus 5A., DC Power Panel 33: Buses 2A and 3A., and DC Power Panel 32: Bus 6A. These circuits also serve as control power for the respective bus except Bus 3A (control power is from DC Power Panel 33 Circuit 2). Failure of control power is indicated by the alarm SAFE-GUARDS INITIATION RACK On 480 V SWGR. SEQ. DC POWER FAILURE on panel SBF-1.
2.5.1 SI With Off-Site Power Available During a safety injection signal with offsite power available, the station auxiliary transformer, via bus 5 and 6 tie breakers, provides power to 6.9 kV buses 1-4. This tie is accomplished automatically on a unit trip when no bus faults are present and the breakers are not in the pullout position. Buses 2,3,5, and 6 provide power to 480 volt buses 2A, 3A, 5A, and 6A via Station Service Transformers. (Refer to System Description 27.4, Electrical Systems - Medium Voltage for details.)
L l Rev. 3, 08/07/2000 - Page 29 -
System Description 10.0 Engineered Safeguards Systems While this is occurring, the safety injection signal strips the 480 volt buses of all non-required safeguards equipment (Non-Essential Load Shed). It also sends an open signal to the 480 volt bus tie breakers between buses 2A and 5A, and buses 3A and 6A. Motor Control Centers 36A - 36E remain energized during a Safety Injection. These MCCs service equipment needed by the Safeguards System. Refer to Figure 10.0-8for EDG starting and 480V bus clearing logic diagram.
The safety injection signal starts the diesel generators in anticipation of a loss of off-site power. The SI signal also blocks the EDG overcurrent, and reverse power trips. This adds a margin of safety in case of a subsequent blackout by having power readily available.
The SI signal (with voltage available to the safeguards bus) energizes the following equipment:
Bus SA Bus 2A Bus 3A Bus 6A 31 S W Pump if 32 SW Pump if 33 S W Pump if selected selected selected 34 SW Pump if 35 SW Pump if 36 S W Pump if selected selected selected 31 AFW Pump 33 A W Pump I I 1 32 AFW Pump 1 32 AFW Pump I I 31 FCU I 32FCU I34FCU 1 35FCU I I33FCU I I I 31 SI Pump 32 SI Pump I 33 SI Pump 31 CS Pump if spray signal 2.5.2 SI with a Loss of Off-site Power Condition When a loss of offsite power to the 480 V buses occurs with a safety injection signal (SI Blackout), the following occurs:
0 The 480 volt buses are isolated from the 6.9 kV buses by tripping breakers 52/2A, 52/3A, 52/5A and 52/6A. These bus supply breakers are tripped when an undervoltage condition is sensed on the respective bus. 480 volt tie breakers 2AT5A and 3AT6A (normally open and racked to test position) receive a trip signal to prevent fault transfer. The equipment on the 480 volt buses trips off due to the undervoltage, and aII auto-starting non-safeguards equipment is locked out. The diesels start and come up to the idle condition. When the diesels are up to voltage and the buses are isolated, the breakers are closed
'd Rev. 3, 08/07/2000 - Page 30 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination 0utIine Cross-reference: Level RO SRO Tier # 2 WS # 60 Group # 2 KIA # 0 15K5.02 OK Importance Rating 2.7 Knowledge of the operational implications of discriminator/compensation operation concepts as they apply to the NIS Proposed Question: Common 58 Which one of the following contains BOTH conditions that will result in indicated reactor power being LOWER than actual reactor power?
A. Source Range pulse height discrimination set too HIGH.
Intermediate Range compensating voltage set too HIGH.
B. Source Range pulse height discrimination set too LOW.
Intermediate Range compensating voltage set too LOW.
C. Source Range pulse height discrimination set too LOW.
Intermediate Range compensating voltage set too HIGH.
D. Source Range pulse height discrimination set too HIGH.
Intermediate Range compensating voltage set too LOW.
Proposed Answer:
A. Source Range pulse height discrimination set too HIGH.
Intermediate Range compensating voltage set too HIGH Explanation (Optional):
Technical Reference(s): SD-13 (Attach if not previously
,>d provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-ICEXC E-3 (As available)
Question Source: Bank # INPO 24930 Modified Bank # (Note changes or attach parent)
New Question History: 12/1/2002 Beaver Valley 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
13.0 Excore Nuclear Instrumentation attenuator selector switch, located inside the SR drawer assembly. The attenuator is preset to compensate for the input cable attenuation of the pulses.
After the pulses have been amplified, they are passed through a discriminator unit. The discriminator electrically gates or discriminates out pulses or heights below certain levels, so only those desired are read. The level below which the input pulses are discriminated is determined by adjusting the discriminator adjust potentiometer located inside the SR drawer. In this manner, the pulses caused by gamma events and electrical noise are removed. This produces a noise-free signal that can be accurately used to determine reactor neutron flux while in the SR.
Figure 13-13 shows the effect of changing the discriminator setting. At the minimum setting, the discriminator is not effective and the pulse amplifier will pass every event, gamma and neutron, onto the pulse counter. As the discriminator setting is increased, the pulse amplifier begins to ignore the lower energy, g a m a events. As the lower graph shows, there is a gap between the energy level of a gamma event and the energy level of a neutron event. When the discriminator setting corresponds to this energy gap, the amplifier will pass all neutron events to the pulse counter, while ignoring all gamma events. If the discriminator is set too high, the pulse amplifier will begin to ignore the less energetic neutrons until finally the maximum setting will cause the amplifier to ignore all events regardless of energy level.
Discrimination is important since the g a m a rays from fission products, which accumulate in the fuel as reactor operations progresses, would mask the low neutron field during the initial phases of startup.
The pulse amplifier also provides an isolated output to drive audio circuits for generating an audible signal proportional to the count rate.
A test-calibrate module, similar to that of the preamplifier, is provided to insert signals of 60, lo3, 105and lo6pulses per second into the pulse amplifier. Switching operations are controlled from the front of the Source Range drawer assembly.
2.6.4 Pulse Shaper The neutron pulses coming from the pulse amplifier are further processed by the pulse shaper. The processing performed by the pulse shaper is designed to prepare the Source Range neutron pulses for final processing by the log pulse integrator.
P Pulse shaping is a necessary intermediate step, so that the log pulse integrator can function in conjunction with the pulse shaper module to L l Rev. 4, 08/03/2004 - Page 22 -
9J5 6 0 13.0 Excore Nuclear Instrumentation v)
P W W 0z E 5
W i 3
0 U
4 I
0 4
CI z
0 4
a W
d I fr I I Ln 3RU l l N f l / S l N 3 3 NIS-14.0WG REV. 0 (8/13/96)
Figure 13-13: SR Discriminator Setting (NIS- 14)
Rev. 4, 08/03/2004 -Page 711 -
13.0 Excore Nuclear Instrumentation the inner and outer cylinders), the neutrons react with the boron causing ionization.
5 6 1O.t 0 N 1 -+ (5 B 11>. -+3 Li 7+++ + 2 a4+++ energy The lithium and alpha particle resulting from this reaction cause secondary ionization in the outer can. The electrons produced by the ionization are collected on the outer can wall. This produces a signal that is proportional to the neutron flux. Electrons are also collected on the outer can wall from the gamma radiation, which interacts with the outer gas volume. This additional signal is proportional to the gamma flux and is additive to the neutron flux signal. The outer can operates in the Ionization Region; thus, all the charged particles produced in the initial ionizing events are collected on the electrodes.
In the inner can, the g a m a flux also reacts with the Nzgas, producing a signal proportional to the gamma radiation. The inner can is operated in the Recombination Region to permit adjustment of the output current by varying its applied voltage. The inner can voltage is called the compensating voltage. If the compensation voltage is set properly, the outer can signal due to gamma plus neutron flux, will interact with the inner can gamma flux only signal. The gamma signals cancel out leaving the neutron only signal which is then amplified before it is displayed on the meter or sent to the protection and control circuitry.
2.1 0.2 Gamma Compensation in the Intermediate Range It is necessary to define the term compensation and the effects of "under-compensation" and "over-compensation"to clearly understand the process of neutron detection in the intermediate range.
Compensation is a term applied to the negative voltage signal applied to the inner can of the CIC which cancels or compensates for the current signal produced by the gamma radiation interacting within the outer can of the detector. This becomes very important to the operator because an incorrect setting of compensating voltage, i.e. over-compensation or under-compensation, would cause an erroneous neutron Ievel indication on the meters, as shown on Figure 13-22.
Over-compensation occurs when the compensation voltage is set to high. This results in a higher current due to g a m a flux in the inner can than is being generated in the outer can due to the same gamma flux. The results of this mismatch is that part of the current due to the neutron flux is also cancelled, causing the indicated current level to be less than actual.
Rev. 4, 08/03/2004 - Page 41 -
13.0 Excore Nuclear Instrumentation Under-compensation occurs when the compensation voltage is less than that required. The current due to gamma from the L
inner can is now smaller than the current due to gamma from the outer can. This allows some current due to gamma in the outer can to remain and add to the neutron flux resulting in increased output current from the detector causing it to indicate above actual levels.
To obtain this true neutron-only signal, the two opposing gamma signals must be cancelled exactly. Since it is physically impossible for both the inner and outer cans to be manufactured identically sensitive to the g a m a flux present under all operating conditions, the problem of how to ensure exact compensation arises. By grooving the inner electrode and applying a variable negative voltage, the size of the inner can is adjusted electrically. The inner can of the CIC operates in the recombination region of the detector characteristic curve and, by adjusting the compensating voltage, only a fraction of the total ionization is collected.
The IR drawer monitors reactor power over a range of eight decades between lo- and lo5 ion chamber amperes. Indications of level and startup rate (SUR) are provided at the NIS cabinets, and on panel FCF.
Because neutron events are occurring at a high rate, no signal conditioning is necessary prior to the log current amplifiers. A block diagram of the intermediate range in provided as Figure 13-23.
2.10.3 Log Current Amplifier This assembly receives current from the detector in the range between lo- and lo- amperes. The assembly provides a logarithmic voltage output, 0 to 10 VDC, proportional to a linear input current. With the use of the log amplifier, the wide range current input is compressed logarithmically to a usable voltage suitable for metering and the generation of trip signals. (Figure 13-23 provides a block diagram of the intermediate range.) The output from the log amplifier is simultaneously coupled to an isolation amplifier and four bistable relay drivers. The output is also displayed on the neutron level meter calibrated in amperes between lo- and IO-.amps.
Internal switches and potentiometers are provided for setting and adjusting the log current amplifiers. Both fixed and variable signals can be injected into the log amplifiers for testing and calibration purposes. This is accomphhed by the use of switches located on the front panel of the drawer and a calibrate module located inside of the IR drawer assembly.
Rev. 4, 08/03/2004 - Page 42 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 61 Group # 2 WA # 028A2.03 OK Importance Rating 3.4 Malfunctions or operations on the HRPS; and based on those predictions use procedures to correct, control or mitigate the consequences of the hydrogen air concentration in excess of limit flame propagation or detonation with resulting
.~
equipment damage in containment Proposed Question: Common 59 Given the following plant conditions:
0 From full power, a Large Break LOCA occurred.
0 Containment hydrogen concentration is at 2%.
31, 32 and 34 Fan Cooler Units started automatic< n the Safety Injection 4
0 31 and 32 Containment Spray Pumps automatically started Which one of the following actions should be taken to address these conditions?
A. Start 33 and 35 Containment Fan Cooler Units to ensure adequate mixing of Containment atmosphere.
6 . Operate at least one of the Hydrogen Recombiners, thereby minimizing the potential for a hydrogen burn.
C. Initiate a containment purge to reduce hydrogen below 1YO,thereby minimizing the potential for a hydrogen burn.
D. Allow Containment Spray to continue to run for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, then resample to see if spray flow has reduced Hydrogen concentration to u Which ONE of the following requirements and/or limitations applies concerning the Containment Purge during this evolution?
A. Notification of the NRC is required.
B. A gaseous discharge permit is required.
C. At least one train of Auxiliary Building Ventilation System shall be in operation.
D. A Containment Purge Exhaust Fan must be started prior to a Containment Vent Exhaust Fan.
Proposed Answer:
B. A gaseous discharge permit is required.
Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): 3-SOP-CB-003 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-VCVCB 6 (As available)
Question Source: Bank # INPO 28436 Modified Bank # (Note changes or attach parent)
New Question History: 11/15/2004 Kewaunee, Unit 1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 9 Comments:
CONTAINMENT PRESSURE RELIEF AND Rev: 29 PURGE SYSTEMS OPERATION Page 8 of 17 1
CB Purge System SHALL NOT be operated in Modes 1,2,3, or 4.
WHEN in Mode 6 AND Containment Closure is set, THEN starting and stopping Purge Supply Fan may affect VC pressure:
o WHEN changing purge Supply Fan status, THEN exercise caution to preclude transferring Refueling Cavity water from VC to FSB due to atmospheric APs.
(Reference 5.2.2)
I I Only section(s) of 3-PT-Q075AYChannel Functional Test RM 11/12 for testing auto closure of CB purge isolation valves is required to be performed. I 4.2 CB Purge System 4.2.1 Startup
- 4.2.1.1 required,
-THEN PERFORM 3-SOP-WDS-013, Gaseous Waste Release Permits.
4.2.1.2 auto closure of VC purge isolation valves has NOT been tested in last 90 days, THEN PERFORM 3-PT-Q075AYChannel Functional Test RM 11/12 prior to entering Mode 6.
I 4.2.1.3 ENSURE the following valve Individual Air Supply Valves are unlocked and open:
- Air Supply Isolation Valve to FCV-I 170 Operator (inside VC under 95 grating at purge valves)
- 0 Air Supply Isolation Valve to FCV-1171 Operator (at purge valves outside containment)
- Air Supply Isolation Valve to FCV-I 172 Operator (inside VC under 95 grating at purge valves)
- 0 Air Supply Isolation Valve to FCV-I 773 Operator (at purge valves outside containment)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'u' Examination Outline Cross-reference: Level RO SRO Tier # G ws # 73 Group # 4 KIA # G2.4.18 OK Importance Rating 2.7 Knowledge of the specific bases for EOPs Proposed Question: Common 72 The following conditions exist:
0 A reactor trip has occurred due to a loss of MFW.
0 FR-H.l, Response to a Loss of Secondary Heat Sink is in progress.
0 The RCS is in a Bleed-and-Feed condition with RCS Temperature stable at 570°F.
0 31 and 32 SG WR levels are off scale low 0 33 and 34 SG WR levels are 10%
'4 0 The operators restore a feedwater source and prepare to feed the S/Gs which are dry.
0 The CRS directs the operator to establish feed water to only one S/G.
Which one of the following describes the reason for feeding only one S/G under these cond itions?
A. To ensure that if a S/G failure occurs due to excessive stresses, the failure is isolated to one S/G.
B. To prevent a rapid cooldown of the RCS that could lead to a pressurized thermal shock condition.
C. To demonstrate the reliability of the FW source before filling all of the steam generators.
D. To determine if one S/G is capable of maintaining adequate heat sink so that RCS bleed-and-feed can be terminated.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
A. To ensure that if a S/G failure occurs due to excessive stresses, the failure is isolated to one S/G.
Explanation (Optional):
Technical Reference(s): FR-H. 1, Foldout Page (Attach if not previously FR-H. 1, Bases provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: I3LP-ILO-EOPFRH 7 (As available)
LJ Question Source: Bank# INPO 24409 Modified Bank # (Note changes or attach parent)
New Question History: 9/17/2002 Summer 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X I O CFR Part 55 Content: 55.41 10 55.43 Comments:
FOLDOUT PAGE
- 1. BLEED AND F E E D CRITERION:
7I F t h e f o l l o w i n g c o n d i t i o n s e x i s t AND b l e e d and feed has NOT been L./
i n i t i a t e d , THEN i m m e d i a t e l y GO To F R - H . 1 . RESPONSE TO LOSS OF SECONDARY HEAT S I N K , Procedure S e c t i o n , Step 9 , t o i n i t i a t e b l e e d and feed :
o Average o f t h e 3 l o w e s t WR SG l e v e l s i s l e s s t h a n 25% C30%1 RCS p r e s s u r e i s g r e a t e r t h a n h i g h e s t SG p r e s s u r e
- 2. SG FEEDFLOW LIMITATIONS:
I F any WR SG l e v e l i s l e s s t h a n 12% C16%J RCS h o t l e g t e m p e r a t u r e i s g r e a t e r t h a n 5 5 O O F . THEN PERFORM t h e f o l l o w i n g f o r a f f e c t e d SG(s):
IF RCS b l e e d and feed has been i n i t i a t e d&A RCS temperatures a r e i n c r e a s i n g , THEN LIMIT use o f a f f e c t e d SGs t o o n l y one and feed t h a t SG a t maximum r a t e .
_IF R C S t e m p e r a t u r e s a r e s t a b l e Q& d e c r e a s i n g , THEN L I M I T f e e d f l o w t o a f f e c t e d SG(s) t o l e s s t h a n o r equal t o 100 gpm and a l l f l o w changes must be made s l o w l y .
IF t h e a f f e c t e d SG(s) w i t h f e e d f l o w has a f a u l t o r r u p t u r e AND any o t h e r SG i s a v a i l a b l e , THEN I S O L A T E f a u l t e d o r r u p t u r e d SG.
i
f damage. Therefore, based on t h e above arguments, f e e d and bleed i s not
-d recomm&?ie8 t o p r o v i d e an a l t e r n a t i v e h e a t removal method d u r i n g a l o s s of secondary h e a t s i n k condition.
2.4 Feeding of a Hot, Dry Steam Generator During r e s t o r a t i o n of secondary h e a t s i n k , i t may become n e c e s s a r y t o e s t a b l i s h feedwater t o a h o t , d r y steam g e n e r a t o r . A h o t , d r y steam g e n e r a t o r i s d e f i n e d a s a steam g e n e r a t o r i n which t h e primary s i d e o f the steam g e n e r a t o r tubes i s above SSOOF and t h e secondary s i d e has no l i q u i d i n v e n t o r y . The primary s i d e SG tube temperature is determined from hot l e g t e m p e r a t u r e r e a d i n g s . Reesta-blishment o f feedwater i s t h e more d e s i r a b l e mode o f recovery from a loss o f secondary h e a t sink than remaining on b l e e d and f e e d and e s t a b l i s h i n g c o l d l e g r e c i r c u l a t i o n f o r long term c o o l i n g . However, care must be taken i f feedwater i s t o be r e e s t a b l i s h e d t o a h o t , d r y steam g e n e r a t o r .
Since t h e h e a t removal c a p a b i l i t y o f one steam g e n e r a t o r i s always g r e a t e r than decay h e a t , i t i s a d v i s a b l e t o r e e s t a b l i s h feedwater t o o n l y one steam L./
g e n e r a t o r r e g a r d l e s 3 of the s i z e of thEi-plant o r number o f loops. T h u s , i f a f a i l u r e i n an SG o c c u r s due t o e x c e s s i v e thermal stresses, the f a i l u r e i s i s o l a t e d t o one steam g e n e r a t o r .
5 5 O O F i s a temperature e v a l u a t e d t o be low enough t h a t thermal s t r e s s would n o t l e a d t o a f a i l u r e when feedwater i s e s t a b l i s h e d t o any remaining d r y steam g e n e r a t o r .
FR-H. 1 51 I n-n-A . --.
ES-40 1 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet u
Examination Outline Cross-reference: Level RO SRO Tier # G ws # 74 Group # 4 KIA # G2.4.1 OK Importance Rating 4.3 Knowledge of EOP entry conditions and immediate action steps Proposed Question: Common 73 Given the following:
0 The plant is at 100% power.
0 A total loss of 480V power occurs.
Which ONE of the following describes the correct procedure and immediate operator action?
-d A. E-0, Rx Trip or Safety Injection, ensure Rx is tripped by all Rod bottom lights lit B. ECA-0.0, Loss of All AC Power, ensure Rx tripped by neutron flux decreasing and close all MSIVs.
C. ES-0.1, Reactor Trip Response, when SI is NOT required, attempt to restore 480V power.
D. ECA-0.0, Loss of All AC Power, ensure Rx tripped by all Rod bottom lights lit and manually trip the turbine.
Proposed Answer:
B. ECA-0.0, Loss of All AC Power, ensure R tripped by neutron flu: decreasing.
Explanation (Optional):
Technical Reference(s): AOP-12 (Attach if not previously
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet ECA-0.0 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOPROU 8 (As available) 13LP-ILO-EOPC00 6 Question Source: 28437 Modified Bank # (Note changes or attach parent)
New Question History: 11/15/2004 Kewaunee, Unit 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 u
55.43 5 Comments:
r No:OAP-OI 2 Rev: I EOP USERS GUIDE Page 6 of 34 4.2.5 Several procedures contain steps which are designated
- as Immediate Action steps.
4.2.5.1 The operator is required to memorize all immediate actions AND SHALL be capable of performing all immediate action steps from memory.
4.2.5.2 The following procedures contain immediate action steps:
. E-0 REACTOR TRIP OR SAFETY INJECTION (Steps 1-4) 9 FR-S.1 RESPONSE TO NUCLEAR GENERATION/ATWS (Steps 1-2)
ECA-0.0 LOSS OF ALL AC POWER (Steps 1-2) 4.2.5.3 While operators are performing immediate actions from memory, the CRS SHhLL obtain the procedure and restart the procedure performance with the team at Step 1.
4.2.5.4 Transitions OR unexpected conditions SHALL be called out during performance.
4.2.5.5 As a general rule, operators should refrain from taking any additional actions, prudent or not, until the immediate actions are complete. The only exceptions are those actions specifically directed to be performed by AOPdONOPs.
4.2.6 Action steps are written such that an operator would normally proceed directly down the left-hand column only. The column contains all the expected conditions, actions, and checks required to accomplish the stated purpose.
4.2.7 The left-hand column contains a high-level statement which describes the task to be performed. This is column is titled "ACTION/EXPECTED RESPONSE". It is called the AER column.
4.2.8 -
IF sequence of performance is important, THEN the sub-tasks are designated by letters (or numbers if finer detail is provided). Step sequence SHALL be strictly adhered to.
4.2.8.1 E sequence of performance is not important, THEN the sub-tasks are designated by bullets.
- Number:
Title:
Revision Number:
ECA-0.0 LOSS OF ALL AC POWER 16 r
NOTE 0 C S F S t a t u s Trees should be monitored f o r information only. FRPs s h o u l d NOT be implemented.
0 Normal communication channels may be unavailable without A C power. R a d i o s should be used by watch personnel o u t s i d e t h e Control Room.
- 1. VERIFY Reactor T r i o : PERFORM t h e following:
Reactor t r i p a n d bypass a . M a n u a l l y TRIP Reactor.
breakers - OPEN
! b. Reactor w i l l NOT t r i p .
-d' 0 Neutron f l u x - DECREASING -
THEN PERFORM t h e following:
- 1) INSERT c o n t r o l rods i n auto o r m a n u a l while continuing performance of t h i s procedure.
2 ) DISPATCH NPO t o t r i p r e a c t o r i n accordance w i t h posted o p e r a t o r a i d .
Page 2 of 49
Number:
Title:
R e v i s i o n Number:
ECA-0.0 LOSS OF ALL AC POWER 16 ACTION/EXPECTED RESPONSE
- 2. ISOLATE Main S t e a m :
a . Manually CLOSE a l l M S I V s a . PERFORM t h e following:
1 ) V E R I F Y a l l turbine s t o p valves a r e closed.
2 ) DISPATCH NPO to close MSIVs per SOP-ESP-1.
- b. CHECK MSIV b y p a s s valves - b. DISPATCH NPO t o close all CLOSED MSIV bypass valves, i f requi red.
i I
Page 3 of 4 9
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'L/'
Examination 0utIine Cross-reference: Level RO SRO Tier # G ws # 75 Group # 4 KIA # G2.4.7 OK Importance Rating 3.1 Knowledge of event based EOP mitigation strategies Proposed Question: Common 74 Given the following conditions:
0 A cooldown and depressurization of the RCS is in progress as directed by ES-0.3, Natural Circulation Cooldown with Steam Void in Vessel.
0 A Yellow path is noted for Inventory that directs the crew to FR-1.3, Response to Voids in Reactor Vessel.
0 The decision is made to continue with the actions of ES-0.3 and NOT transition to FR-1.3, Response to Voids in Reactor Vessel u
Why would a transition to FR-1.3 NOT be made?
A. FR-1.3 would only be entered prior to performing a cooldown and depressurization.
B. FR-1.3 addresses voids resulting from non condensable gas evolution, NOT from steam void formation.
C. Upper head steam voiding is expected in these conditions and accounted for in ES-0.3.
D. The Status Trees are monitored "for information only" in these conditions.
Proposed Answer:
C. Upper head steam voiding is expected in these conditions and accounted for in ES-0.3.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
Technical Reference(s): FR-1.3 (Attach if not previously ES-0.3 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # rNP0 26894 Modified Bank # (Note changes or attach parent)
New Question History: 2/2/2004 Kewaunee, Unit 1 W
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
-Number:
Title:
R e v i s i o n Number:
FR-I.3 RESPONSE TO VOIDS IN REACTOR VESSEL 12 W'
~
CAUTION I F A C O N T R O L L E D N A T U R A L C I R C U L A T I O N COOLDOWN I S I N PROGRESS AND A V O I D IN T H E REACTOR V E S S E L UPPER HEAD I S EXPECTED, T H I S PROCEDURE SHOULD NOT BE PERFORMED.
- 1. CHECK HHSI PurnDs - A L L STOPPED RETURN T o P r o c e d u r e and S t e p i n effect.
Page 2 o f 30
Number:
Title:
R e v i s i o n Number:
N A T U R A L C I R C U L A T I O N COOLDOWN WITH STEAM V O I D I N ES-0.3 12 VESSEL (WITH R V L I S )
A. PURPOSE This p r o c e d u r e p r o v i d e s a c t i o n s t o c o n t i n u e p l a n t cooldown and d e p r e s s u r i z a t i o n t o c o l d shutdown w i t h no a c c i d e n t i n p r o g r e s s . under c o n d i t i o n s t h a t a l l o w f o r t h e p o t e n t i a l formation of a void i n t h e upper head r e g i o n w i t h a v e s s e l l e v e l s y s t e m a v a i l a b l e t o . m o n i t o r v o i d growth.
B. ENTRY CONDITIONS T h i s procedure i s entered from:
- 1. ES-0.2, NATURAL CIRCULATION COOLDOWN, a f t e r c o m p l e t i n g t h e f i r s t e l even s t e p s ,
- 2. ES-0.2, NATURAL CIRCULATION COOLDOWN, WHEN steam v o i d s have f o r m e d i n t h e R e a c t o r v e s s e l and d e p r e s s u r i z a t i o n must c o n t i n u e .
Procedure S e c t i o n , S t e p 15 OR L
ATTACHMENT 1 , COOLDOWN WITHOUT CRDM FANS, S t e p 7 OR ATTACHMENT 1 , COOLDOWN WITHOUT CRDM FANS, S t e p 16
- 3. ES-0.4, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL (WITHOUT RVLIS), S t e p 1 . WHEN RVLIS is a v a i l a b l e .
~~ -
d Page 1 o f 1 6
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet u'
Examinat ion 0ut1ine Cross-reference: LeveI RO SRO Tier # 1 WS#10 Group # 1 WA ?Y 000038EA 1.16 OK Importance Rating 4.4 Ability to operate and monitor SG atmospheric relief valve and secondary PORV controllers and indicators as they apply to a SGTR Proposed Question: Common 75 Given the following conditions:
0 A Steam generator Tube Rupture has occurred on the 31 SG.
0 All equipment is operating as designed.
0 31 SG has been isolated.
The following indications exist:
.J 0 31 SG pressure is 1000 psig and trending UP.
0 31 SG NR level is 55% and trending UP.
Which one of the following describes how pressure will be controlled on 31 SG prior to completion of the RCS depressurization?
A. Automatically at the first SG safety valve setpoint.
B. Manually at the condenser steam dump pressure setpoint.
C. Automatically with the SG atmospheric dump valve controller set at 1040 psig.
D. Manually by performing secondary depressurization to cool down the RCS below initial target temperature.
Proposed Answer:
u C. Automatically with the SG atmospheric dump valve controller set at 1040 psig.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):
Technical Reference(s): E-3, step 3 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-EOP30 7 (As available)
Question Source: Bank # TNPO 25007 Modified Bank # (Note changes or attach parent)
New Question History: 12/1/2002 Beaver Valley 1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4,5,7 55.43 5 Comments:
Number:
Title:
R e v i s i o n Number:
E-3 STEAM GENERATOR TUBE RUPTURE 20 J
,t
+ -
0 CAUTION THE T U R B I N E - D R I V E N AFW PUMP I S THE ONLY A V A I L A B L E SOURCE OF I
FEED FLOW, STEAM SUPPLY TO T H E T U R B I N E - D R I V E N AFW PUMP SHOULD B E M A I N T A I N E D FROM A T L E A S T ONE SG.
0 A T L E A S T ONE SG MUST B E M A I N T A I N E D A V A I L A B L E FOR RCS COOLDOWN.
- 3. I S O L A T E F l o w From RuDtured SG(s):
a . ADJUST r u p t u r e d S G ( s )
atmospheric c o n t r o l l e r t o m a i n t a i n 1040 p s i g b . CHECK r u p t u r e d S G ( s ) b . WHEN r u p t u r e d SG p r e s s u r e a t m o s p h e r i c dump v a l v e ( s ) - i s l e s s t h a n 1040 p s i g ,
CLOSED THEN PERFORM t h e f o l l o w i n g :
- 1) V E R I F Y a t m o s p h e r i c dump Val v e ( s 1 c l o s e d .
2 ) I F NOT, THEN P L A C E a f f e c t e d atmospheric dump v a l v e ( s 1 i n MANUAL and CLOSE.
- 3) a t m o s p h e r i c dump v a l v e ( s 1 can NOT be c l o s e d , THEN D I S P A T C H NPO t o c l o s e p e r SOP-ESP-1, LOCAL O P E R A T I O N OF S A F E SHUTDOWN E Q U I P M E N T .
(STEP 3 CONTINUED ON NEXT PAGE)
I Page 5 o f 5 8
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet i/
Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 76 Group # 1 WA # 025AA2.07 OK Importance Rating 3.7 Ability to determine and interpret pump cavitation as it applies to the Loss of Residual Heat Removal System.
Proposed Question: SRO ONLY 76 Given the following plant conditions:
0 The plant is in CSD following a refueling outage.
A vacuum has been drawn on the RCS in preparation for vacuum filling of the RCS.
0 Vacuum is currently 26Hg 0 RCS level is 626 0 RHR flow prior to drawing the vacuum was 2500 gpm.
0 RCS temperature is 125°F 0 RHR Flow Indicator (FI-640) starts fluctuating from 1500 gpm to 2000 gpm with the 31 RHR pump running.
What has caused the reduction in RHR flow?
A. Letdown flow has been lost.
B. 31 RHR pump is cavitating due to high RCS temperature.
C. RCS inventory is being lost due to lifting of RHR discharge header relief valve733A or 7338 due to high discharge pressure.
D. 31 RHR Pump is vortexing due to low RCS inventory from draining below mid-loop.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
d B. 31 RHR pump is cavitating due to high RCS temperature.
Explanation (Optional):
Technical Reference(s): 3-SOP-RCS-017 (Attach if not previously 3-POP-4.2 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-POP-004 A (As available)
Question Source: Bank # INPO Modified Bank # 28389 (Note changes or attach parent)
New
\d Question History: 11/15/2004 Kewaunee, Unit 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:
f REACTOR VESSEL VACUUM REFILL AND No:3-SOP-RCS-017 Rev: 6 d MANSELL LEVEL MONITORING SYSTEM OPERATION Page 30 of I 2 2 4.5.7.4 INITIATE Aux Spray per 3-SOP-CVCS-002, Charging Seal Water And Letdown Control (both spray valves OPEN to minimize flow to PZR).
4.5.8 ISOLATE RCP #I Seal leakoff stop valves by CLOSING the following:
- CH-AOV-261A 31 RCP Seal Return Isolation Valve
- CH-AOV-261B 32 RCP Seal Return Isolation Valve CH-AOV-261C 33 RCP Seal Return Isolation Valve CH-AOV-261D 34 RCP Seal Return Isolation Valve NOTE RCS Temperature should be maintained near the temperature used to determine target RCS vacuum until RCS Level is greater than 66' (See P&L 2.3.1)
-IF achievable vacuum is less than 20" Hg THEN vacuum SHALL be broken only by steam bubble in PRZR (e.g. 21" Hg)
Attachment 9, Vacuum Pump Skid Diagram provides clarification for operation of the Vacuum Pump.
Vacuum SHALL NOT be drawn on the RCS until the Reactor Head has been fully tensioned and all Con0 Seals are installed.
CTOs may be applied to the vacuum pump, to ensure vacuum pump power and seal water are maintained during the vacuum evolution.
o Temporary Power Supply o Seal Water
~ ~~
4.5.9 RCS Evacuation 4.5.9.1 DETERMINE MAXIMUM RCS vacuum using Attachment 1, Maximum Allowable Vacuum.
4.5.9.2 DETERMINE target RCS Vacuum between 20"Hg and maximum determined in previous step (typically 23").
4.5.9.3 DETERMINE RCS Evacuation Time using Attachment 2, Vacuum Refill Performance, Figure RCS-17-2 (page 2 of 4)
OR Figure RCS-17-3 (page 3 of 4) for PRT not drained.
i REACTOR VESSEL VACUUM REFILL AND W MANSELL LEVEL MONITORING SYSTEM 0PERATION Page 10 of 122 2.3 Temperature Related Precautions and Limitations 2.3.1 RCS temperature SHALL be maintained in accordance with the following table in order to ensure adequate NPSH to the RHR pump while the RCS is under vacuum conditions. (Graph in Attachment I Maximum Allowable I
Vacuum provides pressure and flow guidance)
RCS Level Allowable RCS Temperature Less than 66 100°F to 115°F (Stable) 66 to 67 UP TO 120°F 67 to 68 UP TO 122°F 68 to 69 1 UP TO 124°F I 69 to 70 1 UP TO 126°F 70 to 71 UP TO 128°F 71 to 72 1 UP TO 130°F I 72 to 73 1 UP TO 132°F 73to 74 I UP TO 134°F I
74 to 75 UP TO 136°F 5 to 76 UP TO 138°F 6 to 77 UP TO 140°F 7 to 78(Z 10% PRZR level) UP TO 142°F
~
78 - 2°F for every foot of water level added to the RCS.
REACTOR VESSEL VACUUM REFILL AND No:3-SOP-RCS-017 Rev: 6 d MANSELL LEVEL MONITORING SYSTEM 0PE RATIO N Page 66 of 122 Attachment 1 Maximum Allowable Vacuum (Attachment Page 1 of I)
I I 170-1600 gprn rhr flov 27 (max vacuum limited b) m RCS temperature) 3 26 -2000 gpm RHR flow 25 s
s 24
- 2500 gpm RHR flow 3
23
- 3000 gprn RHR flow 22 21 NOTE IF target vacuum is less 20 .
than 23" Hg THEN vacuum 90 95 100 105 110 115 120 SHALL be broken only by steam bubble in PRZR (e.9.
RHR Inlet Temperature, F 21" Hg)
NO:3-POP-4.2 Rev: 22 OPERATION BELOW 10% PRZR LEVEL WITH FUEL IN THE REACTOR Page 54 of 64 RHR Flow vs. RCS Level To Prevent Air Entrainmentl Vortexing n
Y I000 1200 1400 1600 1800 2000 2200 2400 2600 2800 3000 3200 RHR Flow (gpm)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination Outline Cross-reference: Level RO SRO Tier # 1 ws # 77 Group # 1 KIA # W/E12EA2.1 OK Importance Rating 4.0 Ability to determine and interpret the facility conditions and selection of appropriate procedures during abnormal and emergency operations as they apply to the Uncontrolled Depressurization of all Steam Generators Proposed Question: SRO ONLY 77 Given the following plant conditions:
0 The Unit has sustained a main team line break ffecting all 4 SGs.
0 The crew is currently performing ECA 2.1, Uncontrolled Depressurization Of All Steam Generators.
The crew has throttled AFW flow to 100 gpm to each SG to minimize the RCS cooldown. Safety Injection Termination Criteria is NOT met.
SG Level Pressure TREND 31 SG 19% WR 320 psig SLOWLY DECREASING 32 SG 18% WR 310 psig SLOWLY DECREASING 33 SG 26% WR 380 psig SLOWLY INCREASING 34 SG 18% WR 310 psig SLOWLY DECREASING Which one of the following describes the required action and the reason for the action?
A. Continue with ECA 2.1, Uncontrolled Depressurization Of All Steam Generators, because Safety Injection termination is not complete.
B. Transition to FR-H.1, Loss Of Secondary Heat Sink because there is a RED condition on the Heat Sink Status Tree.
C. Transition to E-2, Faulted Steam Generator Isolation because there is an intact SG available.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet D. Transition to E-3, Steam Generator Tube Rupture because there is an 4 unexplained increase in SG level.
Proposed Answer:
C. Transition to E-2, Faulted Steam Generator Isolation because there is an intact SG available.
Explanation (Optional):
Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE
~4 Learning Objective: (As available)
Question Source: Bank # INPO 25026 Modified Bank # (Note changes or attach parent)
New Question History: 12/1/2002 Beaver Valley 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:
- - - TION
- 1. . . JNTROnUC Guideline ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS, provides procedural guidance t o recover from an event where a71 steam generators a r e depressurizing i n an uncontrolled manner. Due t o t h e low probabi 1 i t y of an uncontrol led depressurization of a1 1 steam generators occurri ng , additional acci dents (e. g ., subsequent SGTRs) were n o t addressed in the development of guideline ECA-2.1.
T h i s guideline i s entered from E-2, FAULTED STEAM GENERATOR ISOLATION, Step 2, when an uncontrolled depressurization of a l l steam generators occurs.
Potential i n i t i a t i n g events f o r this contingency could include steamline breaks, stuck open r e l i e f o r safety valves, o r any combination o f conditions that would a f f e c t a l l steam generators. Failure of all main steamline isolation valves i n conjunction w i t h a single break o r stuck open valve could also lead t o an uncontrolled depressurization of a l l SGs i n plants that do not have SG non-return valves. Guideline ECA-2.1 is always entered a t Step 1.
Guideline ECA-2.1 i s exited whenever any steam generator pressure boundary i s reestablished as indicated by an increase i n t h e associated steam generator pressure indication. In t h i s case, the operator transfers t o guideline E-2 f o r further recovery actions. I t i s noted that this t r a n s i t i o n only applies when t h e S I termination c r i t e r i a are not yet met o r a f t e r SI termination has already been completed i n guideline ECA-2.1. Guideline ECA-2.1 i s also exited a t Steps 6 , 7 , 9 , o r 42. Step 6 d i r e c t s a t r a n s i t i o n t o E-3, STEAM GENERATOR 1 TUBE RUPTURE, i f abnormal radiation e x i s t s i n a steam generator. Step 7 directs a transition t o E-1, LOSS OF REACTOR OR SECONDARY COOLANT, i f RCS pressure i s less than the low-head SI pump shutoff head, f o r further recovery actions. S t e p 9 directs a transition t o ES-1.3, TRAVSFER TO COLD LEG RECIRCULATION, if RUST inventory i s low. S t e p 43 i s t h e final step of the guideline. A decision a t t h i s p o i n t i s made by the plant engineering s t a f f concerning future actions with the plant.
ECA-2.1 1 LP-Rev. 1C LECAP 1
f FOLDOUT PAGE
- 1. S I REINITIATION CRITERIA:
-I F EITHER c o n d i t i o n l i s t e d below occurs, THEN manually START S I pumps: i "d
PRZR l e v e l - CAN NOT BE M A I N T A I N E D GREATER THAN 14% [32%1 RCS s u b c o o l i n g based on q u a l i f i e d core e x i t TCs - LESS THAN 4 0 ° F [ S E E TABLE BELOW3 RCS PRESSURE RCS SUBCOOLING
>1900 p s i g C 63"FI
>lo00 p s i g [ 78"FJ
~ 1 0 0 0p s i g C112" F3
- 2. E - 2 TRANSITION CRITERTA:
IF any SG pressure i n c r e a s e s a t any t i m e (EXCEPT w h i l e p e r f o r m i n g S I t e r m i n a t i o n i n E C A - 2 . 1 , UNCONTROLLED DEPRESSURIZATION OF A L L STEAM GENERATORS, Procedure S e c t i o n , Step 10, through Procedure S e c t i o n ,
Step 191, THEN GO To E - 2 , FAULTED STEAM GENERATOR I S O L A T I O N .
- 3. E-3 T R A N S I T I O N C R I T E R I A :
Manually s t a r t S I pumps as r e q u i r e d , AND Go T o E - 3 , STEAM GENERATOR TUBE RUPTURE i f any SG l e v e l i n c r e a s e s i n an u n c o n t r o l l e d manner o r i
/-.-\ any SG has abnormal r a d i a t i o n .
i
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet WI Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 78 Group # 1 WA # 05562.4.30 OK Importance Rating 3.6 Knowledge of which events related to system operations/status should be reported to outside agencies Proposed Question: SRO ONLY 78 A Loss of Offsite and Onsite power occurred.
The following conditions exist on Unit 3:
0 A loss of all AC power occurred 20 minutes ago 0 The Emergency Director has classified the event in progress as a Site Emergency 0 All State and NRC initial notifications have been made as required 0 Maintenance now estimates 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to restore AC power to any 480V bus 0 The Emergency Director has upgraded the classification to a General Emergency 0 The time now is 01:15 The State of New York must be notified of this change in emergency plan classification NO LATER THAN:
A. 01:30 B. 02:OO C. 0215 D. 0230 Proposed Answer:
LJ A. 01:30
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet L/ Explanation (Optional):
Technical Reference(s): IP-EP-130 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: IOLP-ILO-ERT004 (As available)
Question Source: Bank # rNP0 21488 Modified Bank # (Note changes or attach parent)
New Question History: 7/17/2002 Braidwood 1
',d Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Cornments:
t fi 4
' L '
IP-EC '
-=2- A?%!H@ IMPLEMENTING PROCEDURES
__ - N~N-QUAU-V_.REMT-EQ PROCEDURE REFERENCE USE
- --tf).w.130 Page
~-RevTs7Dn
-7 of 8 4 -
I I
- - I 5.4 Alert / SAE / GE UpgradeNpdate Notifications - CCWEOF Communicator 5.4.1 Upgradelupdate notifications are made for EAL upgrades and for periodic updates during an Alert,, Site Area Emergency (SAE) or General Emergency (GE).
5.4.2 Use an Upgradelupdate Notification AlerVSAE/GE Checklist, (IP-EP-115 Form EP-5) to make and document the emergency classification upgrade or update notifications.
5.4.3 Obtain the completed Radiological Emergency Data Form Part I (IP-EP-115 Form EP-1) from the Shift ManagerEmergency Director AN0 notify State and Counties within 15 minutes of any emergency classification change or approximately every 30 minutes otherwise. Time intervals may be lengthened with concurrence of offsite agencies.
I 6.0 INTERFACES 6.1 SOP-CG-7-1, "Notification During Nuclear Emergency Involving IP No. 2" 6.2 IP-EP-115, "Emergency Plan Forms" I P-EP-250, "Emergency Operations Facility" f- 6-3
\J 6-4 IP-EP-210, "Control Room" ._
7.0 RECORDS All Logs, Completed Forms and other records generated during an actual emergency shall be considered quality records and maintained for the life of the plant.
8.0 REQUIREMENTS AND COMMITTMENTS NONE 9.0 ATTACHMENTS 9.1 Local Government Radio System Locations & Call Letters
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet L.l Examination Outline Cross-reference: Level RO SRO Tier # 1 ws # 79 Group # 1 KIA I# ~~
057A2.19 Importance Rating 4.3 Ability to determine and interpret plant automatic actions that will occur on a loss of a vital ac electrical instrument bus as they apply to the loss of a vital instrument bus Proposed Question: SRO ONLY 79 Given the following plant conditions:
0 Unit 3 is at 100% power 0 8Steam Flow and Feed Flow are in control 0 PRZR pressure channel 1 is in control and 2 is in alarm positions 0 PRZR level channel 2 is in control and 1 is in alarm positions All systems are in their normal full power lineup 0 A loss of 31 Instrument Bus occurs Which of the below describes the plant response with no operator actions?
A. Letdown isolates, all PRZR heaters de-energize, 31 and 33 Main feed Reg valves go open.
B. Letdown isolates, all PRZR heaters de-energize, ALL four Main feed Reg valves remain at their initial position.
C. Letdown remains in service, all PRZR heaters remain in their initial condition, 31 and 33 Main feed Reg valves go closed.
D. Letdown remains in service, all PRZR heaters remain in their initial condition, ALL four Main feed Reg valves remain at their initial position.
Proposed Answer:
L/
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet B. Letdown isolates, all PRZR heaters de-energize, ALL four Main feed Reg valves 4 remain at their initial position.
Explanation (Optional):
LCV-459 closes isolating L/D and PRZR heaters de-energize due to low P U R level and MFW Reg valves have SF and FF matched (0 lbm/hr)
Technical Reference(s): 3-AOP-IB-1, step 4.4, (Attach if not previously Attachment 5 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (AS available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:
Loss Of Power To An 3-AOP-IB-1 Rev. I Instrument Bus Attachment 5 Page 91 of 131 Loads on 3land 31A Instrument Buses Page 9 of 11 C kt 31A Instrument Bus Load Comment 1 RPS - Analog - Ch. II (Rack A8)
Rad. Mon. Control Panel PA6 Microprocessor 1A Rack (RG 1.97) & Sampling Assy Flow Meter &
Temp. Control - Power Supply Ckt. 1A mounted above Ckt. amplifier Ch. ,
2 -
Cont. Parameters Recorder Ch. It (Cab. J02) I Rad. Mon. Control Panel PA6 Microprocessor Rack (RG 1.97) & Sampling Assy Flow Meter &
Temp. Control:
R-56 (Sewage Treatment) 2A Ckt. 2A mounted above R-67 (41' PAB) Ckt. 2 amplifier Ch. II R-68 (15' PAB)
R-69 (55' Pipe Pen)
R-70 (80'PAB) 3 X-Core Neutron Detector N-39 (LOC.90")
4 -
QCS - Analog Ch. II (Rack A7) 5 Spare I 6 3FMS Iso. (Cab. II) 3SPDS Train A
+i- 71 lSPDS Train A ADFP 31 & Lovejoy Speed Controller System CVCS Aux. (Rack C9)
Panel FAF Panel SAF
~~
12 RVLIS JrainA
~ ~ ~-
13 PAB & Trench Temp. Monitoring Cabinet Rad. Mon. Rack 14 PAB Microprocessor Rack R-26 (Cont. Bldg. DHRRM)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet e
Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 80 Group # 1 K/A # 05862.1.30 OK Importance Rating 3.4 Ability to locate and operate components, including local controls.
Proposed Question: SRO ONLY 80 Given the following conditions:
0 The plant is at 100% power.
0 34 and 36 Service Water Pumps (SWPs) are in service supplying the essential header.
0 125 VDC control power to the 34 SWP is lost.
Which ONE (I) of the following describes the effect on the operation of 34 SWP and
%-4 what actions, if any, would be required should 34 SWP need to be secured?
A. Breaker indication in CCR is lost, CCR breaker control is lost, pump remains running - Remove control power fuses, press trip plate on cubicle door to secure the pump.
B. Breaker indication in CCR is available, CCR breaker control is lost, pump will remain running - Remove control power fuses, press trip plate on front of breaker inside cubicle to secure the pump.
C. Breaker indication is available, CCR breaker control is lost, pump will trip - no further action required to secure the pump.
D. Breaker indication in CCR is lost, CCR breaker control is lost, pump will trip - no further action required to secure the pump.
Proposed Answer:
d
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet A. Breaker indication in CCR is lost, CCR breaker control is lost, pump remains running - Remove control power fuses, press trip plate on cubicle door to secure the pump Explanation (Optional):
Technical Reference(s): 3-AOP-DC- 1, Attachment 12 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes
- or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Loss Of A 125V DC Panel 3-AOP-DC-1 Rev. 1 Page 11 of 65
[ ACTlON/EXPECTED RESPONSE 1 RESPONSE NOT OBTAINED I I -
NOTE I
I Valves in the following step have valve stem locking devices (screws).
I 4.10 IAAT CRS/SM desires to throttle charging pump recirc valve@) to aid inventory control, THEN dispatch an operator to throttle the following:
31 (31 CHARGING PUMF REClRCULATION ISOLATION1 CH-408 (32 CHARGING PUMF 32 RECIRCULATION ISOLATION)
CH-409 (33 CHARGING PUMP 33 d REC1RCU LATlON ISOLATION 4.1 1 -INITIATE investigation of fault.
4.12 PERFORM Attachment 12 (Local Operation of 6900V and 480V Breakers)
(Page 55) as necessary to operate the following:
7 6900V breakers on Buses 1,2, or 5
-480V breakers on Bus 5A 4.13 IAAT operation of 52/UT3ST6 or 52/UT4ST6 is required, THEN PERFORM applicable section of SOP-EL-005 (Operation of On-Site Power Sources) to operate breakers.
4.14 -1s 31 DC Power Panel - WHEN affected DC Distribution Panel de-energized? has been re-energized, THEN GO TO Step 4.21.
Loss Of A 125V DC Panel 3-AOP-OC-I Rev. 1 Page 57 of 65 Attachment 12 Local Operation of 6900V and 480V Breakers Page 3 of 5 Operating 480V Breakers Task I Steps I
- 1. Remove breaker control power fuses.
Trip load, feed or tie breaker 2. Press trip plate on cubicle door or trip plate on front of breaker (inside cubicle).
- 1. Remove breaker control power fuses.
'lose load' feed Or 2. Charge spring until spring status plate indicates "Spring tie breaker (See Charged".
notes below)
- 3. Press Push-To-Close button.
Breakers 521312,52/313,and 312-313 tie have a key interlock system that prevents closing of all three breakers at the same time. Normally, two supply breakers are closed and tie breaker is locked in open position by interlock. To close tie breaker, either supply breaker must first be locked open (instructions are inside cubicle door).
Feed breakers for motor-driven fire pump (Buses 312 and 5A) and the associated manual transfer switch (C6 15' el., Northwest corner) have a Kirk key interlock system that prevents operation of the transfer switch under energized conditions. To operate the transfer switch, both feed breakers must first be locked open (instructions are inside cubicle door for each feed breaker).
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination Outline Cross-reference: Level RO SRO Tier # 1 ws # 81 Group # 1 WA # W/E05EAZ.l OK Importance Rating 4.4 Ability to determine and interpret facility conditions and selection of appropriate procedures during abnormal and emergency operations Proposed Question: SRO ONLY 81 Given the following:
0 Unit 3 has had a loss of both Feedwater Pumps from 100% power.
0 SG LOW level annunciators alarm and the Reactor failed to trip.
0 Actions of FR-S.1, Response to Nuclear Power Generation / ATWS are being performed.
0 All AFW pumps failed to start and cannot be started.
LJ 0 Reactor Power has just been verified to be 5%, with a negative start up rate.
Which one of the following procedures should the SRO transition to?
A. Immediately enter FR-H.1, Response to Loss of Secondary Heat Sink.
B. Complete all actions in FR-S.l, then transition to FR-H.1, Response to Loss of Secondary Heat Sink.
C. Re-enter E-0, Reactor Trip or Safety Injection at the beginning and transition to ES-0.1 Reactor Trip Response when directed by E-0.
D. Re-enter E-0, Reactor Trip or Safety Injection at step 1, complete immediate operator actions and then transition to FR-H.1, Response to Loss of Secondary Heat Sink.
4 Proposed Answer:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet B. Complete all actions in FR-S.l, then transition to FR-H.l, Response to Loss of
'L/ Secondary Heat Sink.
Explanation (Optional):
Technical Reference(s): OAP-12 (Attach if not previously FR-S. 1 provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # ~
rNP0 Modified Bank # 258 15 (Note changes or attach parent)
New
'.4 Question History: 3/14/2003 surry 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:
No:OAP-O12 Rev: 1 EOP USERS GUIDE Page 12 of 34 "4
4.3.9 during the performance of any RED condition Function Restoration Procedure (FRP), a RED condition of higher priority occurs, THEN the higher priority RED condition should be addressed first, AND the lower RED condition FRP suspended.
4.3.10 E any ORANGE terminus is encountered, THEN the operator is expected to monitor ALL of the remaining trees, AND THEN, E NO RED is encountered, SUSPEND any E-Set procedure in progress AND transition to the FRP required by the ORANGE terminus.
4.3.1 1 E during performance of an ORANGE condition FRP, any RED condition or higher priority ORANGE condition arises, THEN the RED or higher priority ORANGE condition is addressed first, AND the original ORANGE condition FRP suspended. E a FRP specifically states that higher priority condition should be addressed, THEN this requirement does NOT apply.
4.3.12 Once a FRP is entered due to a RED or ORANGE condition, that FRP is performed to corntietion, unless that FRP is preempted by a higher priority condition.
4.3.13 It is expected that the actions in the FRP will clear the RED or ORANGE condition before all operator actiohs are complete. However, these procedures should be performed to the point of the defined transition to a specific procedure or to the return to "procedure and step in effect".
4.3.14 Status Tree monitoring should be CONTINUOUS any ORANGE or RED condition is found to exist. no condition more serious than YELLOW is encountered, THEN monitoring frequency may be reduced to 10-20 minutes, UNLESS some significant change in plant status OCCUTS.
4.3.15 A YELLOW terminus does not require immediate operator attention.
Frequently it is indicative of an off-normal and/or temporary condition which will be restored to normal status by actions already in progress. In other cases, the YELLOW status might provide an early indication of a developing RED or ORANGE condition. Following FRP implementation, a YELLOW might indicate a residual off-normal condition. The operator is allowed to decide whether or not to implement any YELLOW condition FRP.
4.3.16 The operator should be familiar with all instrumentation used in the status trees. The only unique application is the use of core exit thermocouples in the Core Cooling Status Tree, where the implementation of FRPs as a result of core exit temperature should be based on at least five (5)thermocouples reading greater than the action temperature level. Therefore, FRP implementation is based on the fifth highest core exit thermocouple reading, when used in the status tree.
L 1
Number:
Title:
R e v i s i o n Number:
FR-S.l RESPONSE TO NUCLEAR POWER GENERATION/ATWS 14 4
- 14. CHECK Core E x i t TCs - LESS -I F core . e x i t temperatures a r e THAN 1200°F g r e a t e r t h a n 1200°F AND i n c r e a s i n g , THEN GO To SACRG-I, Severe A c c i d e n t C o n t r o l Room G u i d e l i n e I n i t i a l Response.
- 15. V E R I F Y Reactor S u b c r i t i c a l :
a . CHECK R e a c t o r Power and SUR: a . PERFORM t h e f o l l o w i n g :
0 Power range channels - 1) C O N T I N U E t o b o r a t e .
LESS THAN 5%
- 2) IF b o r a t i o n i s NOT 0 I n t e r m e d i a t e range a v a i l a b l e , THEN ALLOW channels - RCS t o h e a t up.
Zero o r N e g a t i v e s t a r t u p rate 3 ) PERFORM a c t i o n s o f o t h e r Function Restoration Procedures i n e f f e c t which do J 0-J c o o l down OR o t h e r w i s e add positive reactivity t o the core.
- 4) RETURN To Step 4 , Page 5.
- b. CHECK a71 r o d s - LESS THAN b . CONTINUE t o i n s e r t rods 20 STEPS m a n u a l l y u n t i l a l l rods a r e LESS THAN 20 STEPS.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 1 ws # 82 Group # 2 KIA # 00005AA2.03 OK Importance Rating 4.4 Ability to determine and interpret the required actions if more than one rod is stuck or inoperable as they apply to the inoperable / Stuck Rod Proposed Question: SRO ONLY 82 Given the following plant conditions:
The unit is at 88% power 0 Control Bank D is at 210 steps 0 A single rod misalignment has been discovered (14 steps out of alignment).
0 Core peaking factors have been determined to be within limits.
0 Attempts to realign the rod have failed and the rod has been determined to be mechanically stuck.
W Subsequently, a second control rod falls out of alignment criteria (13 steps out of alignment). Attempts to realign this rod have also failed and this rod has been determined to be mechanically stuck. Core peaking factors are within limits with both rods misaligned.
Using the attached Technical Specification reference, which ONE of the following indicates the course of action the operating crew should now take?
A. Reduce reactor power to less than 50% because rod misalignment criteria does not apply at power levels less than 50%
B. Power can be maintained at 88% since core peaking factors are within the limit C. Commence a plant shutdown because the shutdown margin requirements are no longer met.
D. Reduce reactor power to less than 85% in order to increase the available rod misalignment allowance from 12 to 24 steps.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
C. Commence the standard shutdown sequence because the shutdown margin requirements are no longer met Explanation (Optional):
Technical Reference(s): ITS 3.1.4 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: ITS 3.1.4 Learning Objective: (As available)
Question Source: Bank # rNP0 21615 Modified Bank # (Note changes or attach parent)
New Question History: 9/6/2002 Kewaunee, Unit 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
w 5 $2 Rod Group Alignment Limits 3.1.4
\ 3.1 REACTIVITY CONTROL SYSTEMS W'
3.1.4 Rod Group Alignment Limits LCO 3.1.4 A l l shutdown and control rods shall be OPERABLE, w i t h rod group alignment l i m i t s as follows:
- a. When THERMAL POWER i s > 85% RTP, the difference between each individual indicated rod position and i t s group step counter demand position shall be w i t h i n the l i m i t s specified i n Table 3.1.4-1 f o r the group step counter demand position: and I
- b. When MERMAL POWER i s s 85% RTP, the difference between each individual indicated rod position and i t s group step counter demand position shall be w i t h i n 24 steps, APPLICABILITY: MODES 1 and 2.
I CONDITION REQUIRED ACTION COMPLETION TIME L./
A. One or more rod(s) A.l.l Verify SDM i s within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> untrippable. l i m i t s specified i n the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 6 hours (continued)
\4 INDIAN POINT 3 3.1.4-1 Amendment 205
Rod Group Alignment Limits 3.1.4
\
ACTIONS (continued) ~ - ~ ~~
L CONDITION REQUIRED ACTION COMPLETION TIME B. One rod not within 6.1 Restore.rod t o w i t h i n 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a1ignment 1im i t s . a1 ignment 1i m i t s .
!B B.2.1.1 Verify SDM i s w i t h i n the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> l i m i t s specified i n the COLR .
8.2.1.2 I n i t i a t e boration t o 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore SDM t o w i t h i n limit.
A8.Q 8.2.2 Reduce THERMAL POWER t o 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> i 75% RTP.
AN2 8.2.3 Verify SDM i s w i t h i n the Once per l i m i t s specified i n the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> COLR .
I Am 3.2.4 Perform SR 3.2.1.1. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> w
3.2.5 Perform SR 3.2.2.1. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AN2 (continued)
INDIAN POINT 3 3.1.4-2 Amendment 205
Rod Group A1 ignment Limits 3.1.4 ACTIONS CONDI T 1ON REQUIRED ACTION COMPLETION TIME w
B (continued) B .2.6 Re-evaluate ,safety 5 days analyses and confim r results remain v a l i d f o r duration o f operation under these conditions.
C. Required Action and c.1 .Be i n MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associ ated Completion Time o f Condition B not met,
~~
D. More than one, rod not D.l.l Verify SDM i s within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with n alignment i m i t . l i m i t s specified i n the COLR .
ai D.1.2 I n i t i a t e boration t o 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore required SDM t o within l i m i t .
AN2 D. 2 Be i n MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> INDIAN POINT 3 3.1.4-3 Amendment 205
Rod Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 -.....--..----.-.- NOTE --.--.-..-.----..--...--
Not required t o be met f o r individual control rods u n t i l 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a f t e r completion o f control rod movement.
.-----.--.-----._..-_--1---.-..-.--.-.-.--.-.....--
Verify individual rod positions w i t h i n a1 ignment 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> limit.
SR 3.1.4.2 Verify rod freedom o f movement ( t r i p p a b i l i t y ) by 92 days moving each rod not f u l l y inserted i n the core
> 10 steps i n one direction.
SR 3.1.4.3 Verify rod drop time o f each rod, from the f u l l y P r i o r t o reactor withdrawn position, i s s 1.8 seconds from the c r i t i c a l it y a f t e r loss o f stationary gripper c o i l voltage t o each removal o f dashpot entry, with: the reactor head
- a. Tavs2 500°F; and
- b. A l l reactor coolant pumps operating.
INDIAN POINT 3 3.1.4-4 Amendment 205
Rod Group Alignment Limits 3.1.4 Table 3.1.4-1 Maximum Permissible Rod M i s a l ignment (Indicated Rod Position minus Group Step Counter Demand Position)
When > 85 % RTP Step Counter Demand Maximum Permissible Devi a t i ons Position (IRPI Position minus Step Counter Demand Position)
(steps) (steps) s 212 2 -12 and s +12 213 t o 225 2 -12 and i +17 226 2 -13 and i +17 227 2 -14 and i +17 228 2 -15 and i +17 229 r: -16 and s +17
- ? 230 2 -17 and 5 +17 INDIAN POINT 3 3.1.4-5 Amendment 205
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet L/
Examination Outline Cross-reference: Level RO SRO Tier # 1 W S # 83 Group # 2 WA # 000036G2.2.27 OK Importance Rating 3.5 Knowledge of the refueling process Proposed Question: SRO ONLY 83 Given the following:
0 Refueling operations are in progress Irradiated fuel is being moved in the Manipulator Crane from the core to the Containment upender for transfer to the spent fuel pool.
0 Decreasing Spent Fuel Pool (SFP) water level has been reported.
Identify the responsibility of the SRO in containment assigned to the fuel shuffle during d this event.
A. Direct personnel in FSB to isolate the SFP from the cavity per 3-AOP-FH-I .
B. Locate the Manipulator Crane to the south end of the Reactor Cavity.
C. Evacuate ALL personnel from containment.
D. Evacuate ALL personnel from the Fuel Storage Building.
Proposed Answer:
A. Direct personnel in FSB to isolate the SFP from the cavity per 3-AOP-FH-I .
Explanation (Optional):
%.d Technical Reference(s): 3-AOP-SF-1 (Attach if not previously
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41
'4 55.43 6 Comments:
Fuel Damage OR Loss of SFP/Refueling 3-AOP-FH-I Rev. I Cavity Level Page 13 of 95 ACTlON/EXPECTED RESPONSE RESPONSE NOT OBTAINED 1 Unit Status A loss of level has occurred in SFP or Refueling Cavity. !
4.24. -Notify personnel in FSB to perform Attachment 1 (FSB Actions for Loss of Level) (Page 41).
4.25. -Notify personnel in VC to perform Attachment 2 (VCActions for Loss of Level) (Page 89).
4.26. -Evacuate &Inonessential personnel from FSB and VC.
4.27. -Not@ HP to report to FSB.
4.28. -Dispatch at least one operator to determine cause of level decrease.
4.29. -WHEN notified of SFP to refueling cavity isolation status, THEN continue in this procedure.
4.30. -1s refueling cavity isolated from - GO TO Step 4.42.
SFP?
4.31. _Is level decreasing in SFP? - GO TO Step 4.42.
NOTE All further actions for an SFP level decrease are controlled locally by Attachment 1.
4.32. -RETURN to procedure and step in effect.
Fuel Damage OR Loss of SFPRefueling 3-AOP-FH-1 Rev. I Cavity Level Page 41 of 95 Attachment 1 FSB Actions for Loss of Level Page 1 of 47 I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I 1.I.-Is fuel transfer canal gate closed I.- Close and latch fuel transfer canal
-and latched? gate.
2.- IF fuel transfer canal gate CANNOT be closed and latched, THEN GO TO Step 1.4.
1.2. -Is fuel transfer canal gate seal I.- Inflate fuel transfer canal gate seal to inflated? 5 - 25 psig.
2.- IF fuel transfer canal gate has been successfully inflated, THEN notify CCR that SFP is isolated from refueling cavity via fuel transfer canal gate.
3.- GO TO Step 1.4.
I %.d 1.3. -Notify CCR that SFP is isolated from refueling cavity via fuel transfer canal gate. ~ ~~ ~-~ ~
I.4. -Is fuel transfer car in refueling Move fuel transfer car to refueling cavity cavity? using one of the following methods:
I Electrically
- Using manual hand-wheel on Transfer Motor
- Using emergency withdrawal cable I.5. -1s fuel transfer tube gate valve 1.- Close fuel transfer tube gate valve.
closed? 2.- IF fuel transfer tube gate valve is closed, THEN notify CCR that SFP is isolated from refueling cavity via fuel transfer tube gate valve.
3.- GO TO Step 1.7.
1.6. -Notify CCR that SFP is isolated from refueling cavity via fuel L) transfer tube gate valve.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet e
Examination Outline Cross-reference: Level RO SRO Tier # 1 WS # 84 Group # 2 KIA # WlE06G2.4.4
_________~
OK Importance Rating 4.3 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.
Proposed Question: SRO ONLY 84 The Reactor has tripped with a loss of offsite power. SI has actuated. The crew is performing actions in E-0, Reactor Trip or Safety Injection. Given the following conditions:
RCS pressure 1700 psig and trending up 31, 32, 34 SG pressures = 1015 psig stable 33 SG pressure = 700 psig and trending down CETs 750°F and trending up SG Narrow Range level off scale Low Maximum available AFW flow of approximately 75 gpm to each SG PRZR level 15% and trending down CNMT pressure 5 psig and trending up Power is 2% in the PR and IR SUR is slightly negative RVLIS level 45%
Which ONE of the following describes the first procedure transition from E-O?
A. E-2, Faulted Steam Generator Isolation B. FR-S.1, Response to Reactor RestarVATWS C. FR-C.l, Response to Inadequate Core Cooling D. FR-H.l, Response to Loss of Secondary Heat Sink
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'U Proposed Answer:
D. FR-H.1, Response to Loss of Secondary Heat Sink Explanation (Optional):
Technical Reference(s): E-0, step 5 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # INPO 28255 Modified Bank # X (Note changes or attach parent)
New Question History: 11/1/2004 Ginna 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 2 Comments:
Number :
Title:
R e v i s i o n Number:
E-0 REACTOR T R I P OR SAFETY I N J E C T I O N 21
- 5. CHECK AFW S t a t u s :
a . V E R I F Y t o t a l AFW flow - a . PERFORM the following:
GREATER THAN 365 GPM
- 1) Manually START a v a i l a b l e pump(s 1 .
2 ) A L I G N valves as required.
- 3) IF cutback controller i s malfunctioning, T H E N ATTEMPT m a n u a l c o n t r o l .
4 ) _IF a l l SG NR l e v e l s a r e l e s s t h a n 9% C14%1 t o t a l AFW flow can NOT be maintained g r e a t e r t h a n 365 gpm, T H E N GO T o F R - H . l , RESPONSE TO LOSS OF SECONDARY HEAT S I N K .
b . CONTROL feed f l o w t o m a i n t a i n SG NR levels between 9% [14%3 a n d 50%
CAUTION S T A R T I N G OF EOUIPMENT MUST B E COORDINATED W I T H A L L CONTROL ROOM OPERATORS TO ENSURE THAT TWO COMPONENTS ARE NOT STARTED A T THE SAME T I M E ON THE SAME POWER SUPPLY.
- 6,
- D I R E C T BOP Ooerator t o PERFORM *
- R O - 1 . BOP OPERATOR A C T I O N S *
- D U R I N G USE OF EOPS *
- Page 6 of 40
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet LJ Examination Outline Cross-reference: Level RO SRO Tier # 1 WS ## 85 Group # 2 K/A # WlE15G2.1.7 OK Importance Rating 4.4 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation Proposed Question: SRO ONLY 85 Unit 3 experienced a Safety Injection and Containment Spray actuation due to a large break LOCA. E 1, Loss of Reactor or Secondary Coolant, is being performed following a transition from E 0, Reactor Trip or Safety Injection. The STA has just made his initial scan of the Status Trees. The following conditions exist 0 Pressurizer level is 0%
0 Cnmt pressure is 2.8 psig 0 Containment Rad Monitors, R-25 and R-26, have just gone into ALARM 0 Containment Sump Level is 51 ft.
Which of the following procedures must be entered to address the above conditions?
A. FR 1.2, Response to Low Pressurizer Level
- 8. FR 2.3, Response to High Containment Radiation Level C. FR Z. 1, Response to High Containment Pressure D. FR 2.2, Response to Containment Flooding Proposed Answer:
D. FR 2.2, Response to Containment Flooding d Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): F-0.5, F-0.6 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X d
10 CFR Part 55 Content: 55.41 55.43 5 Comments:
Numb= Tius:
Revision
'ii F-0.5 CONTAINMENT 8 CONTAINMENT PRESSURE LESS MAN 22 PSlG ORANGE I I CONTAINMENT Lfvu NO LESS THAN 49 FEFr YES 8 INCHES YELLOW 0 0 0 CONTAINMENT I 0 0
RADIATION LESS THAN 3 R/HR
i Numb-
Title:
Revision F-0.6 INVENTORY 10
- RRANGE W S FULL GREATER THAN 100%
NO YES
- 0 NO R W S DYNAMIC ENTER PRESSURIZER HEAD RANGE
+LNELLEss GREATERTHAN No mm 90% YES
- - 69% ---
loox MCP 3RCP oooa-48% 2RCP YES 37% 1RCP 0
0 Kuow FR-1.1 00000000000000000 yaLoyT fnMm FR-1.2 0
A V
W m
0 0 -(
00 )$-E 0
R W S DYNAMIC H E M RANGE
- GREATERTHAN loox 4RCP No 48% --
69% 3UCP 2RcP YES 37% 1RCP
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 86 Group # 1 K/A # 004A2.32 OK Importance Rating 3.9 Ability to predict the impacts of expected reactivity changes after valving in a new mixed-bed demineralizer that has not been pre-borated and based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations Proposed Question: SRO ONLY 86 The following plant conditions exist:
0 Unit 3 is in a normal full power lineup at MOL 0 A fresh CVCS purification demineralizer has been placed in service.
0 The resin has NOT been boron saturated.
Which of the following is the expected plant response, with no operator action and what operator actions should be taken to mitigate this event?
A. Power level will go up, commence emergency boration using MOV-333 to maintain power 5 100% and TAVGat program value.
B. Power level will go up, commence a normal boration to maintain power 5 100%
and TAVGat program value.
C. RCS pressure will decrease energize Pressurizer Backup Heaters to maintain RCS pressure at 2235 psig.
D. RCS pressure will increase; trip the Reactor, initiate E-0, Reactor Trip or Safety Injection.
Proposed Answer:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet B. Power level will go up, commence a normal boration to maintain power 5 100%
and TAVGat program value.
Explanation (Optional):
Technical Reference(s): SD-3.0 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # _
_ _ _ ~
Modified Bank # 18456 (Note changes or attach parent)
New
-d Question History: 1/1/2000 San Onofre 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X I O CFR Part 55 Content: 55.41 5 55.43 5 Comments:
System Description 3.0 Chemical and Volume Control System Replacement - CVCS Demineralizers, contains all the required steps for resin replacement.
2.2.1 Mixed Bed Demineralizers General purification of the letdown flow takes place in the resin bed of one of the two full flow mixed bed demineralizers. The resin bed performs ion exchange and mechanical filtration of particles greater than 25 microns. The resin bed is lithium hydroxide and the basic ion exchange process releases Li' and O H ions from the demineralizer effluent. Due to the relatively high concentration of boric acid in the coolant, the BO," present rapidly exchanges with the anion resin forming H,BO,J (di-hydrogen borate) when the resin is initially placed in service. The reduction in boron adds positive reactivity to the RCS.
The decrease in boron concentration in the RCS that occurs when a fresh resin bed is placed in service is about 100 ppm at BOL and about 40 ppm at EOL. The O H ions released in the effluent will increase the system pH. After the bed reaches equilibrium, impurities exchange with the anion resin to release BO, ions and the mixed beds have little effect on pH. The 66% anion and 33% cation mixture ensures that the effluent will be neutral (pH = 7) because the Li cation has a tendency to create a strong basic solution.
In addition to removing general ionic impurities, the resin bed is designed to reduce the concentration of isotopes in the purification stream by a minimum decontamination factor (ratio of inlet to outlet activity) of 10, with the exception of cesium. Each demineralizer is sized to accommodate a maximum letdown flow of 120 gpm and has a sufficient capacity to reduce RCS activity to refueling concentrations after operating for one core cycle with one percent fuel defects. The mixed bed will remove hydrazine (added for oxygen scavenging at plant startup) and should be bypassed during hydrazine additions to prevent early depletion of the resin bed and unwanted removal of hydrazine.
2.2.2 Cation Bed Demineralizer (Figures 3.0-5 and 6)
The cation bed demineralizer is located downstream of the mixed bed demineralizers and is normally bypassed. It is not a full flow demineralizer and will only accommodate up to 42 gpm (limited to 40 gpm by procedure). The resin is a hydrogen form cation, and is used intermittently- to maintain cesium 137 activity concentration in the coolant below 1.0 yci/cc (with one percent fuel defects) and to control the concentration of lithium-7.
Li'+H'R-(cation resin) = Li+R(Liresin)+H' Ll Rev. 6, i2,'18,/2002 - Page 11 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet i/
Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 87 Group # 1 WA ## 01262.2.25 OK Importance Rating 3.7 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits Proposed Question: SRO ONLY 87 Given the following plant conditions:
0 Unit 3 is at 100% power.
0 PT-4126, First Stage Turbine Pressure, has just failed low.
What action is required by Technical Specifications and why?
d A. Within 30 minutes verify the P-7 interlock relay is de-energized to ensure the PZR pressure LOW, PZR water level HIGH, RCS flow LOW (2 Loops), RCP breaker open (2 Loops), RCP undervoltage and RCP under frequency Reactor Trips are enabled.
B. Within Ihour verify the P-7 interlock relay is de-energized to ensure the PZR pressure LOW, PZR water level HIGH, RCS flow LOW (2 Loops), RCP breaker open (2 Loops), RCP undervoltage and RCP under frequency Reactor Trips are enabled.
C. Within 30 minutes verify the P-IO interlock relay is energized to allow ensure the PZR pressure LOW, PZR water level HIGH, RCS flow LOW (2 Loops), RCP breaker open (2 Loops), RCP undervoltage and RCP under frequency Reactor Trips are enabled.
D. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> verify the P-10 interlock relay is energized to ensure the PZR pressure LOW, PZR water level HIGH, RCS flow LOW (2 Loops), RCP breaker open (2 Loops), RCP undervoltage and RCP under frequency Reactor Trips are enabled.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
B. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> verify the P-7 interlock relay is de-energized to ensure the PZR pressure LOW, PZR water level HIGH, RCS flow LOW (2 Loops), RCP breaker open (2 Loops),
RCP undervoltage and RCP under frequency Reactor Trips are enabled.
Explanation (Optional):
Technical Reference(s): TS 3.3.1 Condition N and bases (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available) l/ Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:
RPS Instrumentation 3.3.1 rl ACTIONS (continued)
L/
CONDITION REQUIRED ACTION , COMPLETION TIME N. Otie or more channels N.l Verify interlock i s i n 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable. required state for existing unit conditions.
QE N.2 Be i n MODE 2. 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />
- 0. One t r i p mechanism I. 1 Restore inoperable t r i p 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable for one RTB . mechanism to OPERABLE status.
rp1 0.2. Be i n MODE 3. 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br />
-4 INDIAN POINT 3 3.3.1-7 Amendment 205
RPS Instrumentation 3.3.1 n Table 3.3.1-1 (page 5 o f 8) i Reactor Protection System Instrumentation
'L' APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE
, FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALVE
- 16. Safety I n j e c t i o n 1.2 2 trains K SR 3.3.1.14 NA (SI) Input from Engineered Safety Feature Actuation System (ESFAS)
- 17. Reactor T r i p System Interlocks
- b. Low Power 1 2 trains N SR 3.3.1.11 NA I( Reactor Trips SR 3.3.1.13 Block. P-7
~d
- e. Turbine F i r s t 1 2 N SR 3.3.1.1 NA Stage Pressure, SR 3.3.1.10 P-7 Input SR 3.3.1.13 (continued)
(d) Below the P-6 (Intermediate Range Neutron F l u ) interlocks, (h) Above the P-8 (Power Range Neutron Flux) interlock.
d INDIAN POINT 3 3.3.1- 17 Amendment 205
RPS I nstrumentat ion B 3.3.1 BASES LJ APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
Above the P-6 interlock setpoint, the NIS Source Range Neutron Flux reactor t r i p w i l l be blocked, and t h i s Function w i l l no longer be necessary.
I n MODE 3, 4, 5, o r 6, the P-6 interlock does not have t o be OPERABLE because the NIS Source Range i s providing core protection i f required.
The Allowable Value i s NA f o r t h i s function because there i s no corresponding analytical l i m i t modeled i n the accident analysis. The surveillance acceptance c r i t e r i o n used f o r t h i s Function i s 23.1E-11 Amps.
- b. Low Power Reactor T r i m Block. P-7 The Low Power Reactor Trips Block, P-7 interlock, i s actuated by input f r o m either the Power Range Neutron Flux, P-10, o r the Turbine F i r s t Stage Pressure. The LCO requirement f o r the P-7 interlock ensures that the following Functions are performed:
(1) on increasing power, the P-7 interlock (i.e., 2 o f 4 Power Range channels increasing above the P-10 (Function 17.d) setpoint or 1 o f 2 Turbine F i r s t Stage Pressure (Function 17.e) setpoint) automatically enables reactor t r i p s on the f ol 1owi ng Functi ons :
e Pressurizer Pressure- Low:
Pressurizer Water Level -High; Reactor Coolant Flow-Low (Two Loops):
0 RCPs Breaker Open (Two Loops):
0 Undervoltage RCPs; and Underf requency RCPs (continued)
L INDIAN POINT 3 B 3.3.1-32 Revision 2
RPS Instrumentation B 3.3.1 BASES u
APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
These reactor t r i p s are only required when operating above the P-7 setpoint (approximately 10%
power). The reactor t r i p s provide protection against violating the DNBR l i m i t . Below the P-7 setpoint, the RCS i s capable o f providing s u f f i c i e n t natural c i r c u l a t i o n without any RCP running .
(2) on decreasing power, the P-7 interlock (i.e., 3 o f 4 Power Range channels decreasing below the P-10 (Function 17.d) setpoint and 2 o f 2 Turbine F i r s t Stage Pressure channels decreasing below the Turbine F i r s t Stage Pressure (Function 17.e) setpoint) automatically blocks reactor t r i p s on the f o l l owing Functions:
Pressurizer Pressure-Low; 0 Pressurizer Water Level -High; Reactor Coolant Flow-Low (TWO Loops);
0 RCP Breaker Position (Two Loops);
e Undervoltage RCPs; and e Underf requency RCPs An Allowable Value i s not applicable t o the P-7 interlock because i t i s a l o g i c Function. The P-10 interlock (Function 17.d) governs input from the Power Range instruments and the Turbine F i r s t Stage Pressure interlock (Function 17.e) governs input f o r turbine power.
The P-7 interlock i s a l o g i c Function with t r a i n and not channel identity. Therefore, the LCO requires one channel per t r a i n (i.e., two trains) o f Low Power Reactor Trips Block, P-7 interlock t o be OPERABLE i n MODE 1.
(conti nued1 INDIAN POINT 3 B 3.3.1-33 Revision 2
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet i.d Examinat ion OutIine Cross-reference: LeveI RO SRO Tier # 2 WS # 88 Group # 1 IUA # 059A2.07 OK Importance Rating 3.3 Ability to predict the impacts of a trip of MFW pump turbine on the MFW and based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations Proposed Question: SRO ONLY 88 Given the following plant conditions; 0 Unit 3 is performing a power ascension and is currently at 77% power 0 31 and 33 Condensate Pumps are in service 0 32 Condensate Pump is secured but available 0 Both Heater Drain Tank Pumps are in service 0 Both MBFPs are in service A problem develops with 32 MBFP Thrust Bearing causing the pump to TRIP.
What is the appropriate course of action for the above conditions?
A. Perform the immediate operator actions of 3-AOP-FW-I, Loss of Feedwater, reduce load to approximately 700 MWE, adjust speed on 31 MBFP as necessary to maintain suction pressure >350 psig and discharge pressure ~ 1 3 9 0 psig and then start 32 Condensate Pump.
B. Commence a rapid load reduction to 500 MWE, perform the immediate operator actions of 3-AOP-FW-1, start 32 Condensate Pump and then adjust speed on 31 MBFP as necessary to maintain suction pressure >350 psig and discharge pressure e1390 psi.
C. Perform the immediate operator actions of 3-AOP-FW-1, trip the Reactor and enter E-0, Reactor Trip or Safety injection.
D. Trip the Reactor and enter E-0.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
A. Perform the immediate operator actions of 3-AOP-FW-I , Loss of Feedwater, reduce load to approximately 700 MWE, adjust speed on 31 MBFP as necessary to maintain suction pressure >350 psig and discharge pressure <I390 psig and then start 32 Condensate Pump.
Explanation (Optional):
Technical Reference(s): 3-AOP-FW- 1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
'. - . . Loss of Feedwater 3-AOP-FW-I Rev. 04 I Page 5 of 59 64 I . PURPOSE d To respond to a reduction of or total loss of feedwater flow due to loss of MBFPs, condensate pumps, heater drain pumps, feedwater valve failure, or feed line break.
- 2. ENTRY CONDITIONS Any unanticipated reduction in feedwater flow.
- 3. IMMEDIATE ACTIONS 3.1 -Is any MBFP operating? 1.- IF reactor power is > 4%,
THEN trip the reactor and GO TO E-0.
2.- GO TO Step 4.1.
3.2 -Are both MBFPs operating? - IF reactor power is > 80%,
THEN trip the reactor and GO TO E-0.
3.3 -GO TO Step 4.1.
Loss of Feedwater 3-AOP-FW-1 Rev. 04 Page 7 of 59
- 4. SUBSEQUENT ACTIONS ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED P+
d 2.- Adjust running MBFP speed as necessary to accomplish the 0 Match steam flow and feed flow 0 Maintain MBFP suction pressure
c Loss of Feedwater 3-AOP-FW-1 Rev. 04 Page 35 of 59 Attachment 1 Approximate Unit Load With Various Pump Configurations Main Boiler Feed Condensate 'OndensatePumps Booster Heater Drain Approximate Pumps") Pumps Allowable MWE Pumps (Ifl/s)'2' 2 3 2 2 Full Load 2 2 2 2 900 MWE'3) 2 3 2 1 700 MWE'3' I 2 2 2 700 MWE@'
1 2 2 1 550 MWE@'
2 2 2 1 550 MWE(3) 2 2 2 0 400 MWE@)
(1) Each MBFP is rated at 15300 gpm at 4740 rpm with a discharge pressure of 970 psig. A single MBFP can supply a maximum flowrate of 18900 gpm calculated on a head curve at 4875 rpm (expected to occur at approximately 700 MWE).
ll (2) Operation of condensate booster pumps at less than 400 MWE will be dictated by
.J running pump current and total condensate flow. Condensate booster pump running current should be maintained less than 53 amps.
(3) Load will be limited by MBFP suction pressure. Per 3-SOP-FW-001, the suction pressure lower limit is 350 psig during normal operation. At 310 psig, the Standby Condensate Booster Pumps will automatically start (if in AUTO) and CD-AOV-521 I
(POLISHER VESSELS AND POST FILTERS BYPASS) automatically opens. An automatic low suction pressure cutback actuates at 265 psig to lower MBFP speed.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
%d Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 89 Group # 1 KIA # 061G2.1.12 Importance Rating 4.0 Ability to apply technical specifications for a system Proposed Question: SRO ONLY 89 Given the following conditions with Unit 3 operating at 100% power:
0 31 Auxiliary Feedwater (AFW) Pump is out of service for repairs. Repairs will take at least 24 more hours.
0 A routine QA Audit of completed surveillance procedures has determined the quarterly surveillance performed on 33 AFW Pump 35 days ago was NOT properly completed.
In accordance with Technical Specifications, which one of the following actions is
~d correct for this situation?
A. Enter T.S. LCO 3.0.3 and IAW T.S. SR 3.0.4, re-perform the surveillance on 33 AFW Pump within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> B. Enter T.S. LCO 3.0.3 but the required actions can be delayed for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> IAW T.S. SR 3.0.3.
C. Enter T.S. 3.7.5 Condition C, i.e., 2 AFW Trains inoperable, but the required actions can be delayed for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> IAW T.S. SR 3.0.4.
D. Continue T.S 3.7.5 Condition 6, Le., 1 AFW Train inoperable. Re-perform the surveillance on 33 AFW Pump within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> IAW T.S. SR 3.0.3.
Proposed Answer:
D. Continue T.S 3.7.5 Condition B, i.e., 1 AFW Train inoperable. Re-perform the surveillance on 33 AFW Pump within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> IAW T.S. SR 3.0.3.
4
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d Explanation (Optional):
Technical Reference(s): T.S. 3.7.5 & SR 3.0.3 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # INPO 23173 Modified Bank # (Note changes or attach parent)
New Question History: 11/4/2002 Salem Unit 1 u Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2, 5 Comments:
SR Appl i cabi 1 i t y 3.0 r"
Y 3 .O SURVEILLANCE REQUIREMENT (SR) APPLICABILITY L..l SR 3.0.1 SRs shall be met during the MODES or sother specified conditions i n the Applicability for individual LCOs, unless otherwise stated in the SR. Failure t o meet a Surveillance, whether such failure i s experienced during the performance of the Survei 11 ance or between performances of the Surveillance, shall be f a i l u r e t o meet the LCO.
Failure t o perform a Surveillance within the specified Frequency shall be failure t o meet the LCO except as provided in SR 3.0.3.
Surveillances do not have t o be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2 The specified Frequency for each SR is met i f the Surveillance i s performed within 1.25 times the interval specified i n the Frequency, as measured from tfie previous performance or as measured from the time a specified condition of the Frequency i s met.
F o r Frequencies specified as "once," the above interval extension does n o t apply.
If a Completion Time requires periodic performance on a "once per .. . I ' basis, the above Frequency extension applies t o each performance a f t e r the i n i t i a l performance.
Exceptions t o this Specification are stated i n the individual Specifications.
SR 3.0.3 If i t i s discovered t h a t a Surveillance was not performed within its specified Frequency, then compliance w i t h the requirement t o I
declare the LCO not met may be delayed, from the time of discovery, up t o 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> o r up to the limit o f the specified Frequency, whichever i s greater. This delay period i s permitted t o allow performance o f the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.
If the Surveillance is not performed within the delay period, the LCO must immediately be declared n o t met, and the applicable Condi tion(s) must be entered.
When the Surveillance i s performed within the delay period and the Surveillance i s not met, the LCO must immediately be declared n o t met, and the applicable Condition(s) must be entered.
h d (continued)
INDIAN POINT 3 3.0 -4 Amendment 212
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 90 Group # 1 WA # 076A2.0 1 OK Importance Rating 3.7 Ability to predict the impacts of loss of service water on the SWS and based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions Proposed Question: SRO ONLY 90 Given the following plant conditions:
0 A loss of ALL normal Service Water Pumps has occurred due large amount of debris on the screens 0 No Circ Water Pumps are available due to the debris 4
Which of the following describes the required operator action for the above condition?
A. Trip the Reactor and initiate E-0, Reactor Trip or Safety injection B. Trip the Reactor, shut the MSlVs and initiate E-0, Reactor Trip or Safety injection C. Trip the Reactor initiate manual Safety Injection and initiate E-0, Reactor Trip or Safety injection D. Commence a rapid plant shutdown as long as temperatures remain below the trip setpoint Proposed Answer:
B. Trip the Reactor, shut the MSlVs and initiate E-0, Reactor Trip or Safety injection Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): AOP-SW-1 step 4.50 (Attach if not previously provided)
ProDosed References to be Drovided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
Service Water Malfunction 3-AOP-SW-1 Rev. I Page 9 of 43 I ACTIONEXPECTED RESPONSE RESPONSE NOT OBTAINED A- - GO TO 4.1 -
-._-- Is essential header affected?
Step 4.4.
-- - --1---1-__ -__--_._____ -
4.2 -Are Backup SW Pumps available?
...l_l_----_-..
- GO TO Step
-_I_____ -_.___
4.4.
4.3 -Start Backup SW Pumps as necessary to maintain 75 to 110 psig.
-_______- -- - - - _ _ - I - -
4.4 IAAT SW Header pressure can NOT be maintained > 50 psig, THEN PERFORM one of the following based on systems ability to provide cooling:
-Trip the reactor and INITIATE E-0.
-INITIATE plant shutdown to MODE 3 using of the following as applicable:
0 POP-2.1 (Operation At Greater Than 45% Power) d 0 POP-3.1 (Plant Shutdown from 45%
Power)
According to accident analysis, only a loss of the intake structure would cause a loss of all normal SW Pumps.
header from pressure indications and individual component temperatures.
- ~
NO: 3-ARP-008 Rev. 46 PANEL SEF - TURBINE START-UP Page 43 of 107 3.3.1 DISPATCH OPERATOR to check TCV-I 102 (service water control valve on outlet of main oil coolers) for proper operation.
3.3.2 IF TCV-1102 is NOT functioning properly, THEN DIRECT OPERATOR throttle open SWT-6, TCV-I 102 Bypass Isolation, to return temperature to normal.
3.3.3 bearing oil drain temperature reaches 180°F, THEN:
3.3.3.1 E P-8 is NOT illuminated, THEN:
3.3.3.1.1 TRIP the Reactor.
3.3.3.1.2 GO TO E-0, Reactor Trip or Safety Injection.
3.3.3.2 P-8 is illuminated, THEN:
3.3.3.2.1 TRIP the Turbine.
3.3.3.2.2 GO TO 3AOP-TURB-1, Main Turbine Trip Without a Reactor Trip.
3.3.4 bearing metal temperature on TIR-1101 or TIR-1102 reaches 225OF, THEN:
3.3.4.1 E P-8 is NOT illuminated, THEN:
3.3.4.1.1 TRIP the Reactor.
3.3.4.1.2 GO TO G O , Reactor Trip or Safety Injection.
3.3.4.2 P-8 is illuminated, THEN:
3.3.4.2.1 TRIP the Turbine.
3.3.4.2.2 GO TO 3AOP-TURB-I, Main Turbine Trip Without a Reactor Trip.
(CONTINUED ON THE NEXT PAGE)
NO:3-ARP-008 Rev. 46 PANEL SEF -
- . -- . TURBINE
- START-UP Page 44 of 107 i/ -
NOTE WHEN placing MBFP in service on condensate recirculation flow (wind milling),
THEN bearing monitor alarm for thrust bearing temperature may occur due to uneven loading.
3.4 E alarm is from an instrument numbered from 16 through 34 (Boiler Feed Pump No. 31 and 32), THEN:
3.4.1 DIRECT OPERATOR to throttle open SWT-16-1, (SWT-16-2), 31(32)
MBFP Cooler Inlet Isolation, to return temperature to normal.
3.4.2 temperature continues to increase, THEN:
3.4.2.1 REDUCE main turbine load to less than 700 MWe per POP-2. I, Operation at Greater Than 45% Power.
3.4.2.2 REMOVE affected MBFP from service per SOP-Nv-001, Main Feedwater System Operation.
3.4.3 either of the following occurs: THEN 0 Oil temperature from the MBFP bearings increases to 190°F 0 Turbine thrust pad metal temperature increases to 220°F 3.4.3.1 TRIP the affected MBFP.
3.4.3.2 GO TO 3AOP-FW-1, Loss of Feedwater.
3.4.4 both MBFPs are tripped, THEN GO TO E-0, Reactor Trip or Safety Injection.
3.5 - IF alarm is from an instrument numbered from 35 through 52 (Circulating Water Pump No. 31 through 36), THEN:
3.5.1 motor bearing temperature increases to 19B0F,THEN REMOVE affected circulating water pump from service per SOP-RW-001, Circulating Water System Operation.
(CONTINUED ON THE NEXT PAGE)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 WS#91 Group # 2 KIA # 034K4.02 OK Importance Rating 3.3 Knowledge of design feature(s) and / or interlock(s) which provide for fuel movement Proposed Question: SRO ONLY 91 Which of the following describes what occurs when the Manipulator Crane INTERLOCK OVERRIDE Keyswitch is engaged?
A. Hoist Load Interlocks are bypassed except overload.
6 . Gripper Interlocks are NOT bypassed to prevent dropping a fuel assembly.
C. Directly connects bridge, trolley and hoist controls to joystick; speeds are limited to 10 fpm.
D. Boundary Zone Interlocks are bypassed and Bridge/Trolley speed is limited to 30 fpm.
Proposed Answer:
C. Directly connects bridge, trolley and hoist controls to joystick; speeds are limited to 10 fpm.
Explanation (Optional):
Technical Reference(s): SD-17, page 34 (Attach if not previously provided)
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
'd Proposed References to be provided to applicants during examination: NONE Learning Obiective: 13LP-ILO-FHD001 C (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 5.43 Comments:
System Description 17.0 Fuel and Core Component Handling 0 Press for database info - accesses the database screen. The database contains the bridge and trolley coordinates for requested locations.
0 Press for utilities - accesses the utility screen. Displays encoder selection information and allows changing the active encoder.
0 Press for operation - returns the operator to the main menu.
0 Lo~off- allows the operator to log off from the system.
0 Program manager - provides access to the Windows program manager. Requires SRO or system administrator access level.
2.2.5.8 Bypasses and Keyswitches Depressing the BOUNDARY ZONE BYPASS pushbutton allows movement outside the n o d boundary zone. Depressing the pushbutton for ten seconds seals in the interlock until the manipulator crane is returned to the boundary zone. Bridge and trolley speed is limited to a maximum of ten fpm. The pushbutton is backlighted yellow and "BOUNDARYLIMIT INTERLOCK" is visible and blinking when this bypass is active.
Depressing the HOIST LOAD BYPASS pushbutton allows operation of the hoist with an underload condition. The pushbutton is backlighted red and "LOAD BYPASS OVERRIDE ACTIVE" is visible and blinking when this bypass is active.
The INTERLOCK OVERRIDE keyswitch allows manipulator crane operation when it is necessary to place a fuel assembly or the crane in a safe condition. It bypasses alI PLC control functions and directly couples the bridge, trolley, and hoist to their joysticks. The maximum speed is ten feet per minute. Gripper operation is allowed at any time, but the mechanical interlock prevents a loaded gripper from unlatching. "PLC OVERRIDE DETECTED" is visible and blinking when this interlock is active.
2.2.6 RCCA Change Fixture (Figure 17.047)
The RCCA change fixture is designed to remove rod cluster control and spider mounted secondary source assemblies from spent fuel assemblies and insert them into new or partially spent fuel assemblies.
The RCCA change fixture is not normally used during a full core off-load. All insert moves are performed in the SFP.
The fixture consists of two main components: (1) a guide tube, permanently mounted to the reactor cavity wall, for containing and LJ Rev. 3, 10/04/2005 - Page 34 -
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d
Examination Outline Cross-reference: Level RO SRO Tier # 2 WS # 92 Group # 2 KIA # 035A2.03 OK Importance Rating 3.6 Ability to predict the impacts of pressure/level transmitter failure on the S/G and based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations Proposed Question: SRO ONLY 92 Given the following:
0 Unit 3 operating at 100% power 0 B channels of steam flow and feed flow in control 0 Main Feed Regulating Valves in AUTO 0 Steam Generator Pressure Channel PT-419B fails HIGH Which of the below statements describes the plant response and the required actions to stabilize the plant?
A. 31 S/G controlling steam flow indication would increase, 31 S/G actual feed flow and level would increase, 31WG level error would return level to program level but the operator should swap Steam Flow and Feed Flow for 31 S/G to A channel.
B. 31 S/G controlling steam flow indication would increase, 31 S/G actual feed flow and level would increase, the operator must swap Steam Flow for 31 S/G to A channel to prevent a Turbine Trip.
C. 31 S/G controlling steam flow indication would decrease, 31 S/G actual feed flow and level would decrease, 31S/G level error would return level to program level but the operator should swap Steam Flow and Feed Flow for 31 S/G to A channel.
D. 31 S/G controlling steam flow indication would decrease, 31 S/G actual feed flow and level would decrease, the operator must swap Steam Flow for 31 S/G to A d channel to prevent a Turbine Trip.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
A. 31 S/Gcontrolling steam flow indication would increase, 31 S/Gactual feed flow and level would increase, 31WG level error would return level to program level but the operator should swap Steam Flow and Feed Flow for 31 S/Gto A channel.
Explanation (Optional):
Pressure compensation for steam flow will cause indicated flow to increase. The SF/FF mismatch is small enough for level error to compensate and return S/G level back to program thus prevent and Turbine Trip on High S/G level.
Technical Reference(s): 3-AOP-INSTR-1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE d Learning Objective: 13LP-ILO-ICSGL 5.0 (As available)
Question Source: Bank ##
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
Instrument/Controller Failures 3-AOP-INST-1 Rev. 03 Page 23 of 169
.e t ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED 1 4.41 -Has Channel Cfailed? - GO TO Step 4.43.
I NOTE Affected atmospheric steam dump must remain in manual control until Channel C instrument is restored to service.
4.42 -Place affected atmospheric steam dump in manual and return to appropriate position for plant conditions.
~-
- Select both SG Transfer Switches to
~
4.43 -Are both SG Transfer Switches selected to the m-affected the non-affected channel.
channel?
4.44 Refer to the following TS tables for required actions:
-3.3.1-1
-3.3.2-1
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet u
Examination Outline Cross-reference: Level RO SRO Tier # 2 ws # 93 Group # 2 KIA # 068A2.04 OK Importance Rating 3.3 Ability to predict the impacts of failure of automatic isolation on the Liquid Radwaste System and based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations Proposed Question: SRO ONLY 93 A liquid release of 32 Monitor Tank is in progress. A release permit was generated for the release and was approved. The following annunicatorskonditions are received in the control room:
e R18, LIQUID EFF e CHANNEL FAILURE u e R-18, Liquid Waste Effluent monitor is alarming e The discharge remains in progress Which one of the following describes the effect on the plant and the actions required?
A. The release should have automatically terminated. Stop the release and direct chemistry to sample the 32 Monitor Tank then re-calculate allowable release rate to determine if release may continue.
B. The release should have automatically terminated. Stop the release and re-verify the release permit calculations, release may resume provided calculations were correct.
C. R-18 monitor has failed. Request HP recheck calculations for liquid release and recommend corrective action that will be required per the ODCM.
D. R-I 8 monitor has failed. The release may continue provided two independent samples are taken to validate the release permit.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
A. The release should have automatically terminated. Stop the release and direct chemistry to sample the 32 Monitor Tank then re-calculate allowable release rate to determine if release may continue.
Explanation (Optional):
Technical Reference(s): ONOP-RM-2 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: 13LP-ILO-RMSPRM E (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:
LJ -3 73 Number:
0N OP- R M-2 HIGH ACTIVITY - RADIATION MONITORING SYSTEM ATTACHMENT 1 RADlATlON MONITOR AUTOMATIC ACTIONS
- 2. The "SJAE BLOWER" starts.
- 3. CA-PCV-1229 & 1230 "SJAE EFFLUENT R-15 ISOLATION" OPEN
- 4. MS-PCV-1133 "Main Steam to hoggers Pressure Regulator" closes
- 5. MS-MOV-19 "Main Steam to Aux. Steam and n Valve" closes xtraction Steam to Reboiler Steam Generator Blowdown Monitor
- 3. PAB exhaust will be diverted through the charcoal filters
-NOTE ST-7 can be operated by a switch on the north wall of the sewage holding tank building R-56 N B / C 1 Sewage Effluent Monitor pumped to the 35,000 gallon-sewage holding tank IF R 56C alarms, THEN the upper lift station will shutdown
- 1. WDL-AOV-32 and WDL-AOV-37 HTDS and R-61 CPF Regen Release LTDS discharge valves will close Monitor 2. WDL-AOV-31 and WDL-AOV-39 Recirculation Page 10 of 30 ONOP-RM-2
0NOP -RM-2 HIGH ACTIVITY - RADIATION MONITORING SYSTEM ATTACHMENT 6 HIGH RADIATION ALARM ON R-18 Page 1 of 1
. -IF R-18 has ALARMED, THEN PERFORM the following:
- a. Place the permissive switch located on the Radiation Monitoring Control Panel in the CCR to the "Blocked" position to prevent inadvertent opening of RCV-018.
- b. Request the Watch Chemist sample the Monitor Tank that was being released and analyze for radioactivity.
- c. WHEN the Chemist sample results are obtained, THEN re-calculate allowable release rate and determine whether to release the tank or reprocess the tank through the Waste Disposal Facility.
- d. To release the Monitor Tank perform the following:
- 1) Depress the "reset" pushbutton for R-18.
- 2) Place permissive switch for R-18 to the "Unblocked" position.
- 3) Have the Nuclear NPO open RCV-018.
-END OF ATTACHMENT-Page 18 of 30 ONOP-RM-2
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination Outline Cross-reference: Level RO SRO Tier # G ws # 94 Group # 1 WA # 2.1.11 OK Importance Rating 3.8 Knowledge of less than one hour technical specifications for a system Proposed Question: SRO ONLY 94 A plant heatupktartup is in progress with RCS average temperature at 280°F. The following plant conditions develop:
0 31 and 32 RHR pumps become inoperable e 31 and 32 SI pumps become inoperable Which one of the following describes the Technical Specification Actions?
A. Restore only the SI pumps to OPERABLE status before reaching 350°F
C. Immediately initiate action to restore one of the RHR pumps to OPERABLE status.
Proposed Answer:
C. Immediately initiate action to restore one of the RHR pumps to OPERABLE status.
Explanation (Optional):
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Technical Reference(s): TS 3.5.3 Condition A (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X
'..J 10 CFR Part 55 Content: 55.41 55.43 2,5 Comments:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet d
Examination Outline Cross-reference: Level RO SRO Tier # G ws # 95 Group # 1 KIA # 2.1.33 Importance Rating 4.0 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications Proposed Question: SRO ONLY 95 Given the fo Ilowing:
0 The plant is at 75% power.
0 31 CCW heat Exchanger is isolated and the appropriate Condition of TS 3.7.8 is entered 0 Engineering reports that the CCW flow through 32 CCW Heat Exchanger is inadequate to supply appropriate cooling to both RHR Heat Exchangers.
4 0 As such, the RHR trains are inoperable.
0 Operations has isolated CCW to 31 RHR Heat Exchanger and documented proper flow capability to 32 RHR Heat Exchanger.
Per the attached Technical Specifications, do Technical Specifications require entry into Condition A of TS 3.5.2, including why?
A. Yes. LCO 3.0.1 requires all LCOs to be met including the 72-hour LCO of TS 3.5.2.
B. Yes. LCO 3.0.2 requires entry into TS 3.5.2 ACTIONS since LCO 3.5.2is NOT met.
C. No. LCO 3.0.5 waives the requirement to enter TS 3.5.2 ACTIONS during OPERABILITY determinations.
D. No. LCO 3.0.6 waives the requirement to enter TS 3.5.2 ACTIONS provided the safety function is maintained.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Proposed Answer:
D. No. LCO 3.0.6 waives the requirement to enter TS 3.5.2 ACTIONS provided the safety function is maintained.
Explanation (Optional):
Technical Reference(s): TS 3.0.6 (Attach if not previously TS 3.5.2 provided)
~ ~ ~~
Proposed References to be provided to applicants during examination: TS 3.5.2 Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History: 5/10/2004 Davis-Besse 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2.3 Comments:
LCO Appl icabi 15 t y 3.0 3.0 LCO APPLICABILITY (continued)
~.4 LCO 3.0.6 When a supported system LCO i s not met solely due t o a support system LCO not being met, the Conditions and Required Actions associated with t h i s supported system are not required t o be entered. Only the support system LCO ACTIONS are required t o be entered. This i s an exception t o LCO 3.0.2 f o r the supported system. I n t h i s event, an evaluation s h a l l be performed i n accordance w i t h Specification 5.5.14, "Safety Function Determination Program (SFDP) ." I f a loss o f safety function i s determined t o e x i s t by t h i s program, the appropriate Conditions and Required Actions o f the LCO i n which the loss o f s a f e t y function e x i s t s are required t o be entered.
When a support system's Required Action d i r e c t s a supported system t o be declared inoperable o r d i r e c t s entry i n t o Conditions and Required Actions f o r a supported system, the applicable Conditions and Required Actions s h a l l be entered i n accordance with LCO 3.0.2.
LCO 3.0.7 Test Exception LCOs, such as 3.1.8, allow specified Technical Specification (TS) requirements t o be changed t o permit performance o f special t e s t s and operations. Unless otherwise specified, a l l other TS requirements remain unchanged. Compliance w i t h Test Exception LCOs i s optional. When a Test Exception LCO i s desired t o be met but i s not met, the ACTIONS o f the Test Exception LCO s h a l l be met. When a Test Exception LCO i s not desired t o be met, entry i n t o a MODE o r other specified condition i n the A p p l i c a b i l i t y s h a l l be made i n accordance w i t h the other applicable Specifications.
LCO 3.0.8 When one or more required snubbers are unable t o perform t h e i r associated support function(s) , any affected supported LCO(s) are not required t o be declared not met s o l e l y for t h i s reason i f r i s k i s assessed and managed, and:
- a. the snubbers not able t o perform t h e i r associated support function(s) are associated with only one t r a i n o r subsystem o f a m u l t i p l e t r a i n or subsystem supported system o r are associated with a s i n g l e t r a i n o r subsystem supported system and are able t o perform t h e i r associated support function w i t h i n 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; o r
- b. the snubbers not able t o perform t h e i r associated support function(s) are associated with more than one t r a i n o r subsystem o f a m u l t i p l e t r a i n o r subsystem supported system and are able t o perform t h e i r associated support function w i t h i n 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
A t the end o f the specified period the required snubbers must be able t o perform t h e i r associated support function(s), o r t h e affected supported system LCO(s) shall be declared not met.
INDIAN POINT 3 3.0 -3 Amendment 229
ECCS -Operating 3.5.2 L- 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) c u 3.5.2 ECCS - Operating LCO 3.5.2 Three ECCS t r a i n s shall be OPERABLE.
.--.-.._..-._.-_-.----... NOTES..--.-..--.-.-.-..--.-.--.-.---.--.
1 7.. unnr 9 L-IL ii11e-7
~
flow paths may be isol ated by,closing the
~p t o 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> t o perform pressure
&w9fl ig per SR 3.4.14.1.
Aff,,nfle,,t t h HHSI pumps made
!, "Low Temperature incapable o f i n j e c t i n g Overpressure
- allowed for up t o 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> o r u n t i l the i cold legs exceeds 375°F. whichever comes APPLICAB I Lrn
- MODES 1,' 2, and 3.
d CONDITION REQUIRED ACTION COMPLETION TIME A. One or more t r a i n s A.l Restore train(s1 t o 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperabl e. OPERABLE status, AND Two HHSI pumps, one. RHR pump and one Containment Reci rcul ation pump are OPERABLE.
-/
INDIAN POINT 3 3.5.2-1 Amendment 205
ECCS - Operating 3.5.2 CONDITION REQUIRED ACTION , COMPLETION TIME L./'
B. Required Action and B.l Be i n MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.
8.2 Be i n MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> I
'SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are i n the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l i s t e d position with power t o the valve operator removed.
NumberPosition FunctSon f4 i SI - 8568 Cl osed HHSl Loop 33 Hot Leg
'd Injection Stop Va7 ve S I 8566 C1osed HHSI Loop 31 Hot Leg Injection Stop Valve SI-1810 Open RWST outlet isolation AC-744 Open Common discharge isol a t i on for RHR pumps SI-882 Open Comon RWST suction isolation for RHR pumps SI-842 Open HHSI pump minimum flow line isolation SI-843 Open HHSI pump minimum flow l i n e isol a t i on SI- 883 Closed RHR pump return t o RWST is01ation AC-1870 Open RHR pump minimum flow l i n e isolation AC-743 Open RHR pump minimum flow l i n e isol ation (continued)
%J INDIAN POINT 3 3.5.2-2 Ahendment 205
ECCS - Operating 3.5.2 I'
SURVEILLANCE REQUIREMENTS (continued) d SURVEILLANCE FREQUENCY I
SR 3.5.2.2 V e r i f y t h a t each ECCS manual, power operated, 31 days and automatic valve i n the flow path, t h a t i s n o t locked, sealed, o r otherwise secured i n p o s i t i o n , i s i n the correct position.
SR 3.5.2.3 V e r i f y each ECCS pump's developed head a t the In accordance t e s t f l o w p o i n t i s greater than or equal t o the w i t h the requ ired devel oped head. I n s e r v i ce Test ing Program
~-
SR 3.5.2.4 V e r i f y each ECCS automatic valve in the f l o w 24 months path t h a t i s not locked, sealed, o r otherwise secured i n position, actuates t o the c o r r e c t p o s i t i o n on an actual o r simulated actuation signal.
SR 3.5.2.5 V e r i f y each ECCS pump s t a r t s automatically on 24 months an actual o r simulated actuation signal.
SR 3.5.2.6 V e r i f y , f o r each ECCS t h r o t t l e valve 24 months l i s t e d below, each p o s i t i o n stop i s i n t h e correct position.
Valve Numbers SI-8566 SI-856G SI-2165 SI-2170 SI-856C SI-856H SI-2166 SI-2171 SI-856D SI-856J SI-2168 SI-2172 SI-856E SI-856K SI -2169 (continued)
INDIAN POINT 3 3.5.2-3 Amendment 230
ECCS - Operating 3.5.2 C
SURVEILLANCE FREQUENCY
. SR 3.5.2.7 Verify, by visual inspection, each ECCS train 24 months containment sump suction i n l et and reci rcul ation sump suction inlet is not restricted by debris and the suction inlet screens show no evidence of structural distress or abnormal corrosion.
I INDIAN POINT 3 3.5.2-4 Amendment 205
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # G WS # 96 Group # 2 WA # 2.2.26 OK Importance Rating 3.7 Knowledge of refueling administrative requirements Proposed Question: SRO ONLY 96 Which of the below is the first evolution that the Refueling Senior Reactor Operator must be stationed in the Vapor Containment Building?
A. Reactor Head Detensioning B. Reactor Head Lift C. Upper lnternals Lift D. Only during Fuel movement Proposed Answer:
- 6. Reactor Head Lift Explanation (Optional):
Technical Reference(s): RP- 1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 5 Comments:
f REACTOR AUXILIARY EQUIPMENT DISASSEMBLY AND REASSEMBLY RP-1 REV. 5 - SECTION 1 I I
A I Y u
A. RSRO is in charge of all refueling activities and SHALL directly supervise all core alterations in accordance with Technical Specifications. RSRO has complete control and authorization over all refueling activities and all personnel involved with any aspect of refueling process. RSRO has complete authorization to stop any activity affecting refueling process.
B. RSRO SHALL NOT have any additional duties that conflict with his refueling responsibilities.
C. RSRO is responsible for all refueling activities occurring on his shift. As SRO in charge of refueling activities, it is RSROs responsibility to ensure all Technical Specification requirements and procedural prerequisites are satisfied.
D. RSRO SHALL ensure that refueling process is performed in accordance with approved refueling
_- procedures.
E. RSRO SHALL ensure implementation of any emer ency procedure required as a result of B
refue ing operation and SHALL communicate condition to Control Room Supervisor (CRS).
F. RSRO SHALL be present and in a location that allows direct observation of following refueling activities:
- Any movement of RV Head
- Any movement of Upper lnternals
- Control Rod latching, unlatching and testing
- Any movement of fuel into or out of RV G. During fuel movement into or out of the RV, RSRO will normall osition himself in Containment and position an k8, Reactor Engineer, or NPO in the Fuel Storage Building. IF RSRO must leave Containment, THEN fuel movement SHALL stop.
H. RSRO SHALl be only approval authority for use of any mani ulator crane interlock bypass not P
specifical y allowed by refueling procedure.
I AND REASSEMBLY
-e I
L___-
Page 9 of 40 M. RSRO SHALL maintain a refueling logbook as per 00-5, Narrative Log Keeping.
N. RSRO SHALL verify that each fuel assembly inserted or withdrawn is at the correct core location.
1.4.1.3. R e f i r e l l n g i s m s (RMS)
A. R M S is in char disassembly and auxiliary equipment as and RP-2 . RMS has any activity affecting B. RMS SHALL NOT have an additional duties that Y
conflict with his disassemb yheassembiy responsibilities.
C. RMS SHALL ensure that reactor and auxiliary equipment disassembly/reassembfy is performed in accordance with approved refueling procedures.
D. R M S is responsible for ensuring the RSRO, CRS or SM is cognizant of all ongoing reactor and auxiliary equipment disassembly and reassembly.
E. RMS is responsible for notifying the SM or CRS of any significant change in plant conditions as a d result of activities under his supervision or of any unexpected problems or anomalies of a significant nature.
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet
\J Examination Outline Cross-reference: Level RO SRO Tier # G ws # 97 Group # 2 WA # 2.2.24 Importance Rating 3.8 Ability to analyze the affects of maintenance activities on LCO status Proposed Question: SRO ONLY 97 The following plant conditions exist:
0 100 %power 31 EDG is out of service for preventative maintenance The maintenance supervisor requests a work permit for 33 Safety Injection (SI) pump.
Should 33 SI pump be taken out of service? Select the proper action to be taken with 4 the justification for your choice.
A. Yes. Technical Specifications allow for one SI pump to be out of service, provided it is returned to service within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the remaining 2 SI pumps are demonstrated operable.
B. Yes. There are no restrictions applicable to the current plant conditions concerning the removing from service 31 SI pump.
C. No. Technical Specifications require 3 SI pumps together with their associated piping and valves to be operable for all conditions except low power physics testing.
D. No. Technical Specifications state that if one EDG is out of service then the other 2 EDGs and their associated safeguards equipment must be operable within 4 hrs.
4 Proposed Answer:
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet D. No. Technical Specifications state that if one EDG is out of service then the d other 2 EDG's and their associated safeguards equipment must be operable within 4 hrs.
Explanation (Optional):
Technical Reference(s): TS 3.8.1, Bases (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X I O CFR Part 55 Content: 55.41 55.43 Comments:
AC Sources - Operating 6 3.8.1 BASES ACT IONS -
8.2 (cont i nued) safeguards power train {and DG). However, if a required safety feature is supported by an inoperable DG and the redundant safety feature that is powered from a different safeguards power train i s also inoperable, then a loss of offsite power will result in the loss of a safety function. Required Action 8.2 ensures that appropriate compensatory measures are taken for a Condition where the loss of offsite power could result in the loss of a safety function when a DG is not OPERABLE.
The turbine driven auxiliary feedwater pump is not required to be considered a redundant required feature, and, therefore, not required to be determined OPERABLE by this Required Action, because the design is such that the remaining OPERABLE motor driven auxiliary feedwater pumps is capable (without any reliance on the motor driven auxi 1 iary feedwater pump powered by the emergency bus associated with the inoperable diesel generator) of providing 100%
of the auxiliary feedwater flow assumed in the safety analysis.
The Completion Time for Required Action 8.2 i s intended to allow the operator time to evaluate and repair any discovered inoperabilities.
This completion Time also allows for an exception to the normal time zero for beginning the allowed outage time clock. In this Required Action, the Completion Time only begins on discovery that both:
- a. An inoperable DG exists; and
- b. A required feature powered from another safeguards power train i s inoperable.
If at any time during the existence of this Condition (one DG inoperable) a required feature subsequently becomes inoperable, this Completion Time would begin to be tracked.
Discovering one required DG inoperable coincident with one or more inoperable required support or supported features, or both, that are associated with either OPERABLE DG, results in starting the Completion Time for the Required Action. A COMPLETION TIME of four hours from the discovery of these events existing concurrently is Acceptable because it minimizes risk while allowing time for restoration before subjecting the unit t o transients associated with shutdown.
(continued) d INDIAN POINT 3 6 3.8.1 -13 Revision 3
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination 0utIine Cross-reference: Level RO SRO Tier # G WS # 98 Group # 3 KIA # 2.3.6 OK Importance Rating 3.1 Knowledge of the requirements for reviewing and approving release permits Proposed Question: SRO ONLY 98 Which ONE of the following can provide final authorization for a Liquid Rad Waste release?
A. Only the Shift Manager B. Only the Shift Manager or Control Room Supervisor C. Only the Shift Manager or Chemistry Supervisor D. Only the Shift Manager or HP Supervisor Proposed Answer:
B. Only the Shift Manager or Control Room Supervisor Explanation (Optional):
Technical Reference(s): 3-SOP-WDS-014, Attachment 1 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Learning Objective: (As available)
Question Source: Bank # rNP0 19844 Modified Bank # (Note changes or attach parent)
New Question History: 10/29/2001 Braidwood 1 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Comments:
LIQUID WASTE RELEASES I No: 3-SOP-WDS-014 Rev: 23 1 Page 34 of 49 ATTACHMENT 1 LIQUID RADIOACTIVE WASTE RELEASE PERMIT FORM (Page 1 of I)
R e l e a s e ID: IVolume: e;ll. or Flow rate IPermit #:
~-
Recirc. 1 Min. Recirc. I Earliest Sample:
Started -Date: nm: ITime (T): h. I Date: Time:
Radiation In Service (Circle one): Rad Monitor Source Check: Available dilution flow for release (B)
Monitor #: YES NO SAT UNSAT gpn Chemistry analysis Sample #: Dale: Tim: Total Gamma Boron ppm: Boron pounds: Activity (e):
Wowable Diluted Concentration Permissible Chemistry
[ADC) in discharge canal: pci/d Discharge Rate (Dc): Em
'ermissible Radioactive Most Restrictive From: Waste Chemistry
)ischarge Rate (DO: wm Discharge Rate (D): gpm (Check one) 0 Radioactivity Alert Setpoint: uci/ml I Calculated Alarm Setpoint (R): p ~ I
/
Actual Alarm Setpoint:
~ pwnl Monitor taken out of service on: 2nd sample obtained and analyzed by:
(Max. 30 days) Date: Time:
Release calculations performed by: Release calculations verified by:
(1) Release valve alignment reviewed ( i applicable)
(2) Release calcula,tions reviewed (3) Release authorized by: (SMICRS)
Release Rad Monitor reading nitiated: Dale: lime: during release:
- IF flow meter is OOS, lischarge flow meter operable: -
THEN RECORD estimate approx 1hr into release and then every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (required by IP-Circle one)
SMM-CY-001, Radioactive Effluents Control Program):
YES NO*
1 hr 4 hr 8 hr
'ost-Release Release Actual volume iection terminated: Date: Time: released:
dert and Alarm setpoints reset per 3-SOP-RM-010: YES Performed By: (initials)
N/A COMMENTS:
'1
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Examination 0utI ine Cross-reference: LeveI RO SRO Tier # G ws # 99 Group # 4 KIA # 2.4.41 OK Importance Rating 4.1 Knowledge of emergency action level thresholds and classifications Proposed Question: SRO ONLY 99 A Steam Generator Tube Rupture occurred on 34 Steam Generator. Prior to RCS depressurization and SI termination, 34 SG Atmospheric Steam Dump valve was periodically lifting due to high SG pressure. What Emergency Action Level declaration should be made for this event?
A. NUE B. Alert C. SAE D. GE Proposed Answer:
B. Alert Explanation (Optional):
Technical Reference(s): IP-EP- 120 (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet Learning Objective: (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New X Question History:
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:
NON-QUALITY RELATED PROCEDURE IMPLEMENTING
===-E n l w PROCEDURES REFERENCE USE 9.1 Emergency Action Levels CATEGORY 3.0 REACTOR coa ANT SYSTEM Category General Site Area Alert Unusual Event RCS Leakage 1.1.3 {>2OO0F, 5 200°F }
ZVLlS cannot be maintained [Unit 21 3.1.2 p200F)
Primary system leakage exceeding 3.1.1 9200OF)
Unidentified or pressure I
41% [Unit 31 > 33% with no RCPs running capacity (> 75 gpm) of single charging boundary leakage OR Pump 10 gpm OR c
M I the reactor vessel head removed, it is eported that water level in the Reactor Identified leakage llessel is dropping in an uncontrolled manner 25gpm and core uncovery is likely
- ~~
3.2 3.2.2 {*2000F) 3.2.1 (>20O0F)
Primary to Unisolable release of secondary Unisolable release of secondary side to Secondary Leakage side to atmosphere from the atmosphere from the affected steam generator@)with primary to secondary affected steam generator(s) with leakage exceeding capadty (* 75 gpm) of a primary to secondary leakage single charging pump Technical Specifications limit in any steam generator 3.2.3 {>2OO0Q Uniaolablerelease of secondary side to atmosphere from the affeded steam generator(s) with primary to secondary leakage > Technical Spedfication limit in any steam generator AND Coolant adivii 300 pCUcc of 1-131 equivalent 3.3 3.3.1 c*2oooF)
I RCS Subcooling I RCS subcooling SI initiation setpoint i due to RCS leakage
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet 4
Examination Outline Cross-reference: Level RO SRO Tier # G ws # 100 Group # 4 WA # 2.4.28 OK Importance Rating 4.3 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.
Proposed Question: SRO ONLY 100 Assume the operators have just started a depressurization of the intact S/Gs per Step 11 of FR-C.l, Response To Inadequate Core Cooling with the following indications:
0 Core exit TCs at 1250°F and decreasing 0 SG pressures 900 psig and decreasing 0 RWST level just decreased to 11.5 ft.
.-d Select the appropriate action for the above conditions.
A. Continue in FR-C.l until directed to E-I, Loss of Reactor or Secondary Coolant, then Transfer to ES-I .3, Transfer to Cold Leg Recirculation.
B. Complete step 11 i.e.; SG<125 psig and Th less than 350°F, then transfer to ES-1.3, Transfer to Cold Leg Recirculation. Initiate cold leg recirculation, then return to FR-C.1.
C. Immediately transfer to ES-I .3, Transfer to Cold Leg Recirculation while continuing SG depressurization. Initiate cold leg recirculation, then return to FR-C.l step 11.
D. Transfer to ES-1.3, Transfer to Cold Leg Recirculation as soon as core exit TC less than 1200°F. Initiate cold leg recirculation, then return to FR-C.l step 11.
Proposed Answer:
.d C. Immediately transfer to ES-1.3, Transfer to Cold Leg Recirculation while
ES-401 Indian Point Unit 3 Written Examination Form ES-401-5 Question Worksheet continuing SG depressurization. Initiate cold leg recirculation, then return to FR-
'4 C.l step 11.
Explanation (Optional):
Technical Reference(s): (Attach if not previously provided)
Proposed References to be provided to applicants during examination: NONE Learning Objective: (As available)
Question Source: Bank # rNP0 Modified Bank # 27336 (Note changes or attach parent)
New Question History: 4/27/2004 Ginna 1 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 2 Comments:
No:OAP-OI 2 Rev:1 '
EOP USERS GUIDE I
Page 13 of 34 4.3.1 7 Certain contingency EOPs take priority over FRPs due to specific initiating events. These procedures are identified by a note at the beginning of the EOP:
ECA-0.0 LOSS OF ALL AC POWER ECA-0.1 LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED ECA-0.2 LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED ES-1.3 TRANSFER TO COLD LEG RECIRCULATION 4.3.78 In general, an ORANGE or RED condition develops and subsequently clears during execution of a procedure, THEN it is not necessary to enter the FRP for the ORANGE or RED condition at conclusion of current procedure. The exception is the INTEGRITY FRP, which should be entered to establish appropriate soak time, UNLESS it can be determined that the alarm cleared within 25 minutes of initiation.
4.3.19 the CFMS is available for monitoring, THEN it is acceptable to use the CFMS to monitor the status trees. (IP3) 4.4 CONTROL ROOM USAGE OF THE EOP NETWORK 4.4.1 Entry into the EOPs is limited to the following conditions:
WHEN the reactor is in Hot Standby or greater AND any Reactor Trip or Safety Injection occurs =is required, THEN E-O, REACTOR TRIP OR SAFETY INJECTION, SHALL be entered, unless the Control Room has been evacuated a complete loss of all AC Safeguards buses has occurred.
WHEN the reactor is in Hot Shutdown or greater AND a complete loss of power on all AC Safeguards buses occurs, THEN ECA-O.0, LOSS OF ALL AC POWER, SHALL be entered, unless the Control Room has been evacuated. This entry condition also applies during performance of ANY other EOP.
E the Control Room has been evacuated, THEN AOP-SSD-7, CONTROL ROOM INACCESSIBILITY SAFE SHUTDOWN CONTROL, SHALL take priority over all EOPs.
AJTACHMENT 7, EOP APPLICABILTY, presents the overall mode applicabilty for each EOP.
c m:If RWST level decreases t o less than (U.O2), the SI System should be I aligned f o r cold leg recirculation usfng ES-1.3, TRANSFER TO COLD LEG RECIRCULATION.
PURPOSE: To guarantee coolant flow t o the core by switching t o cold leg recirculation i f the RWST level decreases below the switchover setpoi n t BASIS:
If the switchover level i n t h e RWST is reached, which could happen a t any time d u r i n g the course o f guideline FR-C.l depending upon the amount o f RCS inventory losses , the operator should inmediately go t o ES-1.3, TUNSFFR TQ a m t o maintain coolant flow t o the core. When RWST level decreases t o (U.02), there should be sufficient water avai7able i n the i recirculation sump t o switch the suction supply t o the SI pumps. The remainder o f RWST water i s reserved for spray pump usage.
d ACTIO&:
Determine i f RWST Ievef decreases t o less than (U.02)
INSTRUMFNTATION:
RWST 1eve1 in d i cat i on CONTROI /FOUI PMEU:
N/A u'
(U.02) RWST switchover setpoint i n plant specific u n i t s . I FR-C. 1 10 LP-Rev. 1C LFRCl