ML032460781
| ML032460781 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/21/2003 |
| From: | Berry P Entergy Nuclear Northeast |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 05-333/03-301 | |
| Download: ML032460781 (250) | |
Text
Examination Outline Cross-reference: Level SRO Tier # I Partial or Complete Loss of AC I 6 Group ## 1 Ability to determine andlor interpret the following KIA # 295003 AA2.05 as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER : (CFR: 41.10 / 43.5 / 45.13)
Whether a partial or complete loss of A.C. power has occurred Importance Rating 4.2 Proposed Question: The plant is shutdown for a Refueling outage. Site electrical power is being provided from the 115 KV. The only deviation from the normal alignment is that disconnect 10017, North-South Bus Disconnect, is currently OPEN. From this condition, circuit breaker 10022, Lighthouse Hill, trips.
Which one of the following identifies the expected procedural response?
a) AOP-16, Loss of 10300 Bus and AOP-18, Loss of 10500 Bus ROlSRO b) AOP-17, Loss of 10400 Bus and AOP-19, Loss of 10600 Bus SI c) AOP-57, Recovery from Residual Bus Transfer d) AOP-49A, Station Blackout In Cold Condition Proposed Answer: b) AOP-17, Loss of 10400 Bus and AOP-19, Loss of 10600 Bus Explanation (Optional):
Technical Reference(s): OP-44, AOP-17 (Attach if not previously provided) d Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71D, EO-1.05.a, 1.06, 1.09 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
Page 7 of I 2 0 NRC Written Examination Submittal.docLast printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Complete Loss of Forced Core Flow Group # 1 1 Circulation / 1 & 4 Knowledge of facility ALARA program. KIA# 295001 2.3.2 2.3.2 (CFR: 41.12143.4 /45.9/45.10)
Importance Rating 2.5 2.9 Proposed Question: During 100% power, the Shift Manager authorizes an entry into the Steam Affected Area to find the source of a new steam leak. The Operator is given a ten (IO) minute limit for the search based on expected dose rates. Just as the Operator enters the Steam Affected Area, an announcement is made that 'A' Recirculation Pump has tripped. The dose rate in the Steam Affected Area drops to % of the pre-transient level.
The Operator should... ...
a) Double the search time that was allowed to adequately identify the leak.
ROISRO b) Double the search time that was allowed to identify additional discrepancies.
112 c) Leave the area when the ten (IO) minute time is expired.
d) Request an additional Operator to assist in the leak identification.
Proposed Answer: c) Leave the area when the ten (IO) minute time is expired.
Explanation (Optional):
W Technical Reference(s): AP-7.03 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AP-7.03, EO-28.03 (As available)
Question Source: Bank #
Modified Bank# (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis I O CFR Part 55 Content: 55.41 12 55.43 4 Comments:
Page 8 of 120 NRC Written Examination Submittal.docLast printed 6/6/2003 10:27 AM
L. Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Complete Loss of AC I6 Group # 1 1 Ability to determine andlor interpret the following WA # 295003 AA2.01 AA2.01 as they apply to PARTIAL OR COMPLETE LOSS 0FA.C. POWER: (CFR: 41.10143.5145.13)
Cause of partial or complete loss of A.C. power Importance Rating 3.4 3.7 Proposed Question: From a normal full power operating condition, a complete and instantaneous loss of bus 10500 occurs.
Which one of the following is a LIKELY cause for the occurrence?
a) Loss of DC Control Power to bus 10500 ROlSRO b) Activation of the bus 10500 Degraded Bus Voltage timer 213 c) Ground fault trip of circuit breaker 10514 d) Overcurrent condition on CRD pump A motor Proposed Answer: c) Ground fault trip of circuit breaker 10514 Explanation (Optional):
Technical Reference(s): AOP-18 (Attach if not previously provided)
AR P-09-8-2-8
~~
Proposed references to be provided to applicants during examination: None v
Learning Objective: SDLP-71E, EO-1.05.C, 1.10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 9 of 120 NRC Written Examination Submittal.docLast printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Total Loss of DC Pwr 1 6 Group # 1 1 Knowledge of the interrelations between WA# 295004 AK2.01 AK2.01 PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following: (CFR: 41.7 145.8)
Battery charger Importance Rating 3.1 3.1 Proposed Question: While at full power, Maintenance disconnects 125 VDC Station Battery A to replace a faulty cell.
WHICH ONE ( I ) of the following will be the response of the system during a large 125 VDC load emergency starting?
a) The charger will trip on high starting currents associated with emergency loads.
RO/SRO b) The charger will supply emergency loads under these conditions for one hour.
314 c) The charger will supply both normal and emergency loads for four hours.
d) The charger will supply emergency loads for four hours.
Proposed Answer: a) The charger will trip on high starting currents associated with emergency loads.
Explanation (Optional):
Technical Reference(s): OP-43A (Attach if not previously provided)
W Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71B, EO-1.05.A.2, 1.13A (As available)
Question Source: Bank # Duane Arnold 1 INPO # 7209 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/19/1996 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Comments:
Page 10 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
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u Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Total Loss of DC Pwr / 6 Group # I 1 Knowledge of the interrelations between KIA# 295004 AK2.01 AK2.01 PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following: (CFR: 41.7 / 45.8)
Battery charger Proposed Question:
"*q 'A' fu pv\-
q &d\t-L\
ROlSRO 314 d) The charger w Proposed Answer:
Explanation (Optional):
Technical Reference(s): OP-43A (Attach if not previously provided)
'v Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71B Obj. 1.05.A2 1. \ 3< (As available)
Question Source: Bank # +Qibicqp tu FG-)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam -1 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Comments:
Page 10 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44 AM h
Examination Outline Cross-reference: Level RO SRO Tier # I 1 Main Turbine Generator Trip I 3 Group # 1 1 Knowledge of the interrelations between MAIN K/A # 295005 AK2.07 AK2.07 TURBINE GENERATOR TRIP and the following:
(CFR: 41.7 145.8)
Reactor pressure control Importance Rating 3.6 3.7 Proposed Question: The following conditions exist while performing a shutdown for a refueling outage:
0 Reactor power is 27%.
Recirculation flow is at minimum.
A Main Turbine Trip occurs.
Turbine Bypass Valves respond as designed.
Which of the following is CORRECT concerning the plant and operator response?
a) Turbine Bypass Valves will be able to control Reactor pressure. Unless available steam drains are opened, it will be necessary to insert a manual Reactor SCRAM.
b) Turbine Bypass Valves will be able to control Reactor pressure. It will be ROISRO necessary to insert Control Rod Cram Groups in accordance with Reactor Analyst Instructions in RAP-7.3.16.
415 c) Turbine Bypass Valves will control Reactor pressure. It will be necessary to close all available steam line drains to assist Reactor pressure control.
d) Turbine Bypass Valves will control Reactor pressure. It will be necessary to operate Turbine Bypass Valves manually.
Proposed Answer: a) Turbine Bypass Valves will be able to control Reactor pressure. Unless available steam drains are opened, it will be necessary to insert a manual Reactor SCRAM.
Explanation (Optional):
Technical Reference(s): OP-9, AOP-2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-94C, EO-l.lO.K (As available)
Question Source: Bank # Duane Arnold 1 INPO Bank # 620 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 512511999 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Page 11 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 IO27 AM
Examination Outline Cross-reference: Level RO SRO W Tier # I 1 Main Turbine Generator Trip 13 Group # 1 I Knowledge of the interrelations between MAIN WA# 295005 AK2.07 AK2.07 TURBINE GENERATOR TRIP and the following:
(CFR: 41.7 145.8)
Reactor pressure control Importance Rating 3.6 3.7 Proposed Question:
exist:
Which of the following is% CORRECT concerning the plant and operator response to
-..._event?
this - . -. . c h_-k <2 &GiLom.I &n&& .
a) The Bvpass Valves will NOT be able to control Reactor pressua It will be I _ F ' necesiarv for oDerators to insert a manual Reactor Scram. '\.
b,tJ\
b) The Bypass Valves will NOT be able t control Reactor pressure. It will be ROISRO necessary for operators to i n s e r t d a m Broufin accordance with-hd-
- m. A4& Aa\hZst k%hLa %
yn w I .L \b 415 c) The Bypass Valves will be able to control Reactor pressure. It will be necessary for operators t o m all available steam line drains to assist Reactor pressure control. eb st t-htW,
- Proposed Answer:
d) The Bypass Valves will be able to control Reactor pressure,N operator action will be necessary h, cy&% b,pv$LL1 - L$
J ~
a) The Bypass Valves will NOT be able to control Reactor pressure. It will be 1 '
necessary for operators to insert a manual Reactor Scram Explanation (Optional):
n A
Technical Reference(s): OP-9, A O P 4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: fi44 Learning Objective: SDLP94C Obj. 1.lO.K (As available)
Question Source: Bank # Duane Arnold IINPO Bank # 620 (,* +"
Modified Bank # (Note changes or attach parent)
- Question History:
New Last NRC Exam 5/25/1999 Page 11 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8 1 4 4 ~ ~
Comments:
Page 12 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 SCRAM / 1 Group # 1 1 Knowledge of the operational implicationsof the K/A # 295006 AKI .03 AK1.03 following concepts as they apply to SCRAM :
(CFR: 41.8 to41.10)
Reactivity control Importance Rating 3.7 4.0 Proposed Question: The Control Room Supervisor orders you to insert a manual scram because power is unexpectedly rising. Which of the following responses indicates that the scram has successfully controlled reactivity under all conditions?
a) Reactor power dropping rapidly through the IRM and SRM ranges.
ROlSRO b) 6 rods indicate position 02, remaining rods indicate position 00.
5/6 c) 1 rod indicates 48, 1 rod at 10, remaining rods indicate position 00.
d) Annunciators, 09-51-13, RPS A MAN SCRAM and 09-5-1-14, RPS B MAN SCRAM are in alarm.
Proposed Answer: b) 6 rods indicate position 02, remaining rods indicate position 00.
Explanation (Optional):
Technical Reference(s): AOP-1, EP-1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP, EO-2.01, EOP2LP, EO-1.07 (As available)
Question Source: Bank # Dresden 2 INPO Bank # 6558 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 3/1VI996 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Comments:
Page I 3 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
M\ dmm-d Examination Outline Cross-reference: Level i RO SRO b
Tier # I I SCRAM I 1 Group # 1 1 Knowledge of the operational implications of the K/A # 295006 AKI .03 AK1.03 following concepts as they apply to SCRAM :
(CFR: 41.8 to 41.10)
Reactivity control &)(y :\so r cbQr61 RW-n 1 Importance Rating 3.7 4.0 c-,
Proposed Question: The.- rders you to insert a manual scram because power is unexpectedly increasing.Which of the following responses indicates that the scram has successfully controlled reactivity under all conditions?
a) Reactor power dropping rapidlySkolu;t-k
- e S&W)Ixnd S M ROISRO b) 6 rods indicate position 02, remaining rods indicate position 00.
516 1 rod indicates 48, 1 rod at 10, remaining rods indicate position 00.
_.- ---* c)
Proposed Answer: b) 6 rods indicate position 02, remaining rods indicate position 00.
Explanation (Optional):
Technical Reference(s): AOP-1 , Ep- (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: LP-AOP Obj. 2.01 (As available)
Question Source: Bank # Dresden 2 INPO Bank # 6558 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 311111996 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:
Page 13 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 I Control Room Abandonment I 7 Group # 1 1 Knowledge of the interrelations between KIA # 295016 AK2.03 AK2.03 CONTROL ROOM ABANDONMENT and the following: (CFR: 41.7 / 45.8)
Control room HVAC Importance Rating 2.9 3. I Proposed Question: A Control Room HVAC fire has resulted in a significant amount of smoke in the Control Room requiring Control Room Evacuation.
Which of the following describes the conditions under which plant control may be returned to the Control Room as directed by AOP-43, PLANT SHUTDOWN FROM OUTSIDE THE CONTROL ROOM?
a) The fire is verified to be extinguished by the Fire Brigade Leader BEFORE Remote Shutdown Panel actions have been commenced.
ROlSRO b) No Control Room damage exists as determined by the County Fire Control Coordinator and Remote Shutdown Panel actions have been completed.
617 c) The Shifi Manager has authorized transferring control to the Control Room and Remote Shutdown Panel to Control Room turnover procedures are completed.
d) The Emergency Director and Security Manager have determined the Control Room is functional and habitable.
Proposed Answer; c) The Shift Manager has authorized transferring control to the Control Room and Remote Shutdown Panel to Control Room turnover procedures are completed.
W Explanation (Optional):
Technical Reference(s): A 0 P-43 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP, EO-1.03.A (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Page 14 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Total Loss of CCW I 8 Group # 1 1 Knowledge of the operational implications of the WA # 295018 AKI .01 AKI .01 following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER :(CFR: 41.8 to 41. IO)
Effects on componenffsystemoperations Importance Rating 3.5 3.6 Proposed Question: The plant is operating at 90% power with one Reactor Building Closed Loop Cooling (RBCLC) pump tagged out of service. An electrical problem causes the two running RBCLC pumps to trip.
Operators have the ability to restore cooling via Emergency Service Water to EACH of the following EXCEPT:
a) RWCU Non- Regenerative Heat Exchanger ROISRO b) Drywell Cooling Assemblies 718 c) Recirculation Pump Seal Coolers d) Drywell Equipment Sump Cooler Proposed Answer: a) RWCU Non- Regenerative Heat Exchanger Explanation (Optional):
Technical Reference@): AOP-11 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-15, EO-1.09, SDLP-46B, EO-1.06.B (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 15 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
L Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Partial or Total Loss of Inst. Air / 8 Group # 1 1 Knowledge of refueling administrative KIA# 295019 2.2.26 2.2.26 requirements. (CFR: 43.5 / 45.13)
Importance Rating 2.5 3.7 Proposed Question: The plant is in a refueling outage. The Refuel Bridge is over the core supporting 'In-Vessel' inspections. Thirty, (30) minutes after a complete l o s s of Instrument Air occurs, an NPO calls from the refuel floor to report that Spent Fuel Pool level has risen several inches over the last hour.
Which of the below is the probable cause?
a) RWCU Blowdown Flow Control Valve (12FCV-55) failed closed.
ROlSRO b) Main Steam Line Plugs have depressurized.
8/9 c) In-service Fuel Pool Filter/ Demineralizer has isolated.
d ) Feedwater Low Flow Control Valve loss of air control signal.
Proposed Answer: a) RWCU Blowdown Flow Control Valve (12FCV-55) failed closed.
Explanation (Optional):
Technical Reference@): AOP-12 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP, EO-1.10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4,7,10 55.43 5 Comments:
Page 16 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Loss of Shutdown Cooling I 4 Group # I I Ability to operate andlor monitor the following as WA # 295021 AAI .04 AA1.04 they apply to LOSS OF SHUTDOWN COOLING :
(CFR: 41.7 145.6)
Alternate heat removal methods Importance Rating 3.7 3.7 Proposed Question: A loss of shutdown cooling has occurred. The cavity is flooded and the spent fuel pool gates are removed. The current decay heat load of the core and spent fuel pool is 1 . 8 ~ 1 0BTU/hr. ~
Which decay heat removal lineup listed below would provide sufficient decay heat removal?
a) RWCU in blowdown mode.
ROlSRO b) Fuel pool cooling system.
9110 c) Decay heat removal system.
d) RWCU in recirculation mode.
Proposed Answer: c) Decay heat removal system.
Explanation (Optional):
Technical Reference(s): AOP-30 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: AOP-30, Attachment 3 Learning 0bjective: SDLP-10, EO 1.15.a (As available)
Question Source: Bank # JAFLOR20004206B02C Rev.2 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Comments:
Page 17 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
LOSS OF SHUTDOWN COOLING AOP-30 ATTACHMENT 3 Page 1 of 1 ALTERNATE COOLING METHODS I I 1 METHOD APPROXIMATE HEAT LIMITATIONS REMOVAL CAPACITY (BTU/hr 1 Decay Heat Removal 3.00E7 Gates removed between cavity and spent fuel pool I
I I
1 I Fuel Pool Cooling 3.3036 Gates removed between caqity and spent fuel pool RBC must be available SW must be available Fuel Pool Cooling 2.40E7 RHR must be available Assist RHRSW must be available Gates removed between cavity and spent fuel pool RWCU Blowdown Mode 2.06E6 No isolation signal present W 1 pump running Makeup source must be 125 gpm blowdown flow available (see list below) 125 gpm makeup flow Main Condenser or Radwaste must be available I +
RWCU Recirc Mode
! 1.70E6 No isolation signal present I
RWCU Blowdown Mode gravity drair.
I 1.00E6 No isolation signal present Makeup source mus: be 1
50 gpm blowdown flow available (see list below) 50 gprn makeup flow Main Condenser or Radwaste must be available I I i Makeup Sources Condensate transfer keep-full using Core Spray or RHR Control Rod Drive System Condensate/Feedwater Condensate transfer to skimmer surge tanks (gates removed)
Condensate transfer to fuel pool using DHR (gates removed)
Condensate transfer using service box connections on the refuel floor (gates removed)
Fire Protectior, System water from local fire hose stations or outside sources RHR service water cross-tie Fire Water Crosstie Rev. No. 15 Page 21 of 34
L Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Refueling Acc Cooling Mode I 8 Group # 1 1 Knowledge of the reasons for the following KIA# 295023 AK3.02 AK3.02 responses as they apply to REFUELING ACCIDENTS : (CFR: 41.5 I45.6)
Interlocks associated with fuel handling equipment Importance Rating 3.4 3.8 Proposed Question: Which one (I) of the following will result in a control rod block during Refuel Floor activities in an outage?
a) Mode Switch in START/HOT STBY, Fuel Grapple loaded and Refuel Bridge near or over the Spent Fuel Pool.
ROlSRO b) Mode Switch in REFUEL, Fuel Grapple not full up and Refuel Bridge near or over the Spent Fuel Pool.
10/11 c) Mode Switch in REFUEL, a single control rod is p & full in and selection of any other control rod.
d) Mode Switch in START/HOT STBY, all control rods are full in and selection of any control rod.
Proposed Answer: c) Mode Switch in REFUEL, a single control rod is not full in and selection of any other control rod.
Explanation (Optional):
Technical Referenceb): ST-20F (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O8B, EO-1.02, 1.05.8 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Page 18 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Drywell Pressure / 5 Group # 1 1 Knowledge of the operational implications of the KIA # 295024 EK1.01 EKI .01 following concepts as they apply to HIGH DRYWELL PRESSURE : (CFR: 41.8 to 41.10)
Drywell integrity: Plant-Specific Importance Rating 4.1 4.2 Proposed Question: Following a LOCA, with severe complications, the RHRSW crosstie is being used for RPV makeup. The primary containment water level is 80 feet and rising. Primary containment pressure is 35 psig and rising.
The Primary Containment shall be vented before:
a) Torus pressure exceeds 5 6 psig.
RO/SRO b) Torus pressure reaches 62 psig.
11/12 c) Torus pressure reaches 77 psig.
d) Torus pressure reaches 85 psig.
Proposed Answer: d)Torus pressure reaches 85 psig.
Explanation (Optional):
Technical Reference(s): EOP-4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EO Ps Learning Objective: MIT-301.11Ef EO 4.07 (As available)
Question Source: Bank ## JAF LOR 20005213B05C Rev.3 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 5 Comments:
Page 19 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO L
Tier # 1 1 High Reactor Pressure 1 3 Group # 1 1 Ability to operate andlor monitor the following as K/A # 295025 EA1.07 EA1.07 they apply to HIGH REACTOR PRESSURE:
(CFR: 41.7 1 45.6)
ARI/RPTIATWS: Plant-Specific Importance Rating 4.1 4.1 Proposed Question: Which ONE of the following describes the effect a reactor vessel pressure signal of 1170 psig will have on the reactor recirculation pumps and alternate rod insertion (ARI) system?
The Recirculation motor/generator...
a) drive motor breakers will trip and the ARI solenoid valves will energize.
ROiSRO b) generator field breakers will trip and the ARI solenoid valves will energize.
12/13 c) drive motor breakers will trip and the ARI solenoid valves will de-energize.
d) generator field breakers will trip and the ARI solenoid valves will deenergize.
Proposed Answer: a) drive motor breakers will trip and the ARI solenoid valves will energize.
Explanation (Optional):
Technical Reference(s): ITS-3.3.4.1/SR-3.3.4.1.4 (Attach if not previously provided)
W Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-02H EO 1.05.C.2, SDLP-O3C E01.05.C.2 (As available)
Question Source: Bank # Quad Cities 1 INPO Bank # 16832 (Modified for JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 311611998 (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 2 Comments:
Page 20 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Reactor Pressure 1 3 Group # 1 1 Ability to operate andlor monitor the following as WA # 295025 EA1.07 EA1.07 they apply to HIGH REACTOR PRESSURE:
(CFR: 41.7 / 45.6)
AR IIRPTIATWS: PIant-S pecific Importance Rating 4.1 4.1 \\flo Proposed Question: Which ONE of the following describes the effect a reactor vessel pressure signal of KXQ psig will have on the reactor recirculation pumps and atternate rod insertion (ARI) system?
The r u b p u m p... moLlRUlk-t;j?* */y& a .
a) drive motor breake:will trip and the ARI solenoid valves will energize.
ROISRO b) generator field breaketwill trip and the ARI solenoid valves will energize.
12/13 c) drive motor breake?will trip and the ARI solenoid valves will de-energize.
d) generator field breakeiwill trip and the ARI solenoid valves will deenergize.
Proposed Answer:
Explanation (Optional):
Technical Reference(s): EOP-3 (Attach if not previously provided)
- i
\i Proposed references to be provided to applicants during examination:
Learning Objective: (As available)
Question Source: Bank # Quad Cities 1 INPO Bank# 16832 I Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 3m11 998 ii (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
I Question Cognitive Level: Memory or Fundamental Knowledge I Comprehension or Analysis I
I 10 CFR Part 55 Content: 55.41 55.43 Comments:
\
Page 20 of 129 NRC Written Examination Begin to END-docLastprinted 5/21/2003 8:44AM
L Examination Outline Cross-reference: Level SRO Tier # 1 Control Room Abandonment / 7 Group # 1 Ability to perform specific system and integrated WA # 295016 2.1.23 plant procedures during different modes of plant operation. (CFR: 45.2 I45.6)
Link to 10CFR-55.43(b)(6)
Importance Rating 4.0 Proposed Question: The Plant was operating at 100% full power with no systems out of service; all equipment was in a normal lineup. Subsequently a fire has occurred in the Control Room and the Control Room was evacuated without performing any AOP-43, Plant Shutdown from Outside the Control Room, actions.
Which of the following AOP-43 actions will ensure the Reactor is shutdown?
a) Trip RWR MG Set A & B Generator Field Breakers.
RO/SRO b) Trip RWR MG Set A & B Drive Motor Breakers.
SI4 c) Isolate and vent the scram air header at Reactor Building 272 Southwest.
d) Place RPS MG Set A Generator Output Breaker in OFF.
Proposed Answer: c) Isolate and vent the scram air header at Reactor Building 272 Southwest.
Explanation (Optional):
Technical Reference@): AOP-43 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP, EO-1.10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6,7, I O 55.43 6 Comments:
Page 21 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level SRO Tier # 1 Partial or Total Loss of Inst. Air 1 8 Group # 1 Ability to determine andlor interpret the following WA # 295019 AA2.01 as they apply to PARllAL OR COMPLETE LOSS OF INSTRUMENT AIR :(CFR: 41.10 143.5 145.13)
Instrument air system pressure Importance Rating 3.6 Proposed Question: The plant is operating at 100% power with Control Room Annunciators inoperable.
During panel walkdowns, the SNO-2 reports that all Air Compressors are running. The SNO-1 reports that Scram Air Header pressure is 60 psig and lowering and one (1) Rod Drift light is lit on the Full Core Display.
Which of the following is the correct response?
a) Enter AOP-27, Control Rod Driff, and manually SCRAM if a second rod drifts.
b) Reduce Recirculation Pumps to minimum and enter AOP-8, Loss of Reactor RO/SRO Coolant Flow.
SI5 c) Manually SCRAM the Reactor and enter AOP-1, Reactor SCRAM.
d) Trip the Main Turbine and enter AOP-2, Main Turbine Trip Without SCRAM.
Proposed Answer: c) Manually SCRAM the Reactor and enter AOP-1, Reactor SCRAM.
Explanation (Optional):
Technical Reference(s): AOP-12 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-39, EO-I .15.A, LPAOP, EO- (As available) 1.03 Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 22 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Reactor Pressure I 3 Group # 1 1 Ability to determine and/or interpret the following KIA # 295025 EA2.06 EA2.06 as they apply to HIGH REACTOR PRESSURE:
(CFR: 41.10 f43.5 f 45.13)
Reactor water level Importance Rating 3.7 3.8 Proposed Question: Following a reactor SCRAM and MSlV isolation, HPCl is injecting into the reactor. RPV level on narrow Range is 200"and rising. Reactor pressure is 800 psig and rising.
The HPCl turbine will trip ... ...
a) At a lower indicated NR level at 800 psig than at 1100 psig.
ROlSRO b) At a higher indicated NR level at 800 psig than at 1I 0 0 psig.
13/16 c) At the same indicated NR level at 800 psig and at 1100 psig.
d) When NR level indication reaches 222.5".
Proposed Answer: a) At a lower indicated NR level at 800 psig than at 1100 psig.
Explanation (Optional):
Technical Reference(s): OP-15, attachment 3 (Attach if not previously provided)
W Proposed references to be provided to applicants during examination: None
~
Learning Objective: SDLP-23, EO-1.05.C.1, 1.13 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Cornments:
Page 23 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
L Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Suppression Pool High Water Group # 1 1 Temp. I 5 Knowledge of the operational implications of the WA # 295026 EK1.01 EKI .01 following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE : (CFR: 41.8 to 41.10)
Pump NPSH Importance Rating 3.0 3.4 Proposed Question: The following plant conditions exist:
Torus Pressurel.0 psig Torus Level- 11.92 feet What is the maximum Torus water temperature that two (2) RHR Pumps can operate at 8,000 gpm each without exceeding NPSH limitations?
a) 173°F RO/SRO b) 182°F 14/17 c) 200" F d) 206°F Proposed Answer: d) 200 F Explanation (Optional):
Technical Reference(s): OP-13 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: OP-I3A, Attachment # 1 Learning Objective: SDLP-13, EO-1.13.A (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X I O CFR Part 55 Content: 55.41 8 55.43 Comments:
Page 24 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
4 EOP POSTED ATTACHMENT Page 1 of 1 (Location: Simulator & Control Room:Panel 0 9 - 3 )
FtHR NPSH AND VORTEX LIMITS RHR PUMP NPSH LIMIT 250.00 240.00 230.00 220.00 L
u l
g 210.00 Overpressure' Overpressure' 150.00 Overpressure*
140.00 130.00 0 1 2 3 4 5 6 7 8 9 10 11 SINGLE LOOP RHR - FLOW PER PUMP (gpm X 1000)
Torus Overpressure = Torus Pressure + 0.4 (Torus Water Level - 1.92) i
- VORTEX LIMIT Torus Water Level Greater Than or E q u a l to 8.92 feet OP-13A RHR - LOW PRESSURE ATTACHMENT 1 COOLANT INJECTION Rev. No. 1
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Drywell Temperature I 5 Group # 1 1 Ability to determine andlor interpretthe following WA # 295028 EA2.01 EA2.01 as they apply to HIGH DRYWELL TEMPERATURE (CFR: 41.10 / 43.5 / 45.13)
Drywell temperature Importance Rating 4.0 4.1 Proposed Question: EOP-4, PRIMARY CONTAINMENT CONTROL, has been entered due to a valid entry condition. Simultaneously, EPIC power is lost and no Control Room computer screens are available. Which one of the following describes the first preferred indicator(s) to be utilized to determine drywell temperature (Assume normal full power drywell fan configuration)?
a) Average of both drywell cooler inlet temperatures from 68TI-100 and 68TI-101 on panel 09-75.
b) Average of drywell cooling assembly air inlet and outlet ROISRO temperatures from 68TI-100 or 68TI-101 on panel 09-75.
15/18 C) Either DW TEMP A 16-1TR-108 or DW TEMP B 16-1TR-107 on panel 09-3.
d) Average of DW TEMP A 16-1TR-108 and DW TEMP B 16-1TR-107 on panel 09-3.
Proposed Answer: d) Average of DW TEMP A 16-1TR-108 and DW TEMP B 16-1TR-107 on panel 09-3.
Explanation (Optional):
Technical Reference(s): EP-1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EP-1 (excluding section 4.7)
Learning Objective: (As available)
Question Source: Bank # JAFLOR20005204B06C Rev.1 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 25 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
ENTERGY NUCLEAR OPERATIONS, INC .
JAMES A. FITZPATRICK NUCLEAR POWER PLANT L
- EOP AND SAOG SUPPORT PROCEDURE EOP ENTRY AND USE EP-1 REVISION 6 APPROVED BY:
EFFECTIVE DATE: Id 10a, FIRST ISSUE FULL REVISION LIMITED REVISION
- INFORMATIONAL USE *
- TSR *
- TECHNICAL
',, -REVISION
SUMMARY
SHEET REV. NO. CHANGE AND REASON FOR CHANGE 6 Revised the Minimum Core Flooding Interval listed on Attachment 1 and the Maximum Core Uncovery Time depicted on Attachment 2, to reflect changes from Cycle 16 core reload.
(JD-01-102 and JENG-02-0375) 5 4.2,1.A, 4th line, expanded "Steps E and F " to full step numbers for cross referencing. (4.2.1.E, 4.2.1.F)
Added para. to Subsection 4.2.3 Operator Actions / Strategies to clarify expectations for use of CS and RHR when in Alternate RPV Water Level Control. (PCR 5 , 3/31/00)
Deleted Subsection 4.4, SPDS (Safety Parameter Display System). Requirements moved to AP-12.06, Rev. 3 (PCR dated 10/5/01, originally assigned to AP-12.03)
To Subsection 5.1, added plant indications to be used to establish status of reactor shutdown. (PCR 7 , 12/21/00) 4.8.4, 5th line, changed "AOP-39" to "AOP-40" to correct typo. (EC#l, 3/16/00)
V' 4.8.4, removed the reference to "branch piping." (PCR, dated 1/24/02) 5.3, 2nd bullet, 2nd IF/THEN statement - add "temperature" after "area" in 4th line for clarification.
5.3, under Reactor Building Radiation Levels, added another means of monitoring radiation levels is by Radiation Protection survey. (PCR 6, dated 3/31/00; PCR dated 6-28-01) 5.3, under Reactor Building Radiation Levels, clarified the conditions under which the radiation levels are assumed to be greater than the maximum safe level. (PCR, dated 10/06/01)
Deleted "7.2 Validation" per AP-02.01.
Changed "RES" to "Radiation Protection, per ODSO-31.
Deleted Attachment 1, EPIC SPDS Point Status Log. Moved to AP-12.06, Rev. 3. (PCR dated 10/5/01, against AP-12.03)
Changed attachment numbers to reflect the deletion of Attachment 1.
i Rev. No. 6 Page 2 of 22
-EOP ENTRY AND USE EP-1 TABLE OF CONTENTS SECTION 1.0 PURPOSE . . . . . . . . . . . . . . . . . . . . . . . . 4 2.0 PRECAUTIONS . . . . . . . . . . . . . . . . . . . . . . 4 3.0 PREREQUISITES . . . . . . . . . . . . . . . . . . . . . 4 4.0 SPECIAL INSTRUCTIONS. . . . . . . . . . . . . . . . . . . 5 4.1 Plant Conditions and Parameters . . . . . . . . . . . 5 4.2 Use of EOPs . . . . . . . . . . . . . . . . . . . . . . 6 4.3 Manual Control of Automatic Systems . . . . . . . . . 11 4.5 Containment Instrument Nitrogen . . . . . . . . . . . 11 4.6 REIR and Core Spray Operation . . . . . . . . . . . . . 12 4.7 Reactor Shutdown Determination . . . . . . . . . . . . 13 4.8 Procedure U s e While Performing EOPs . . . . . . . . . 14 5.0 PROCEDURE . . . . . . . . . . . . . . . . . . . . . . . 15 5.1 RPVControl . . . . . . . . . . . . . . . . . . . . . . 15 5.2 Primary Containment Control . . . . . . . . . . . . . . 16 5.3 Secondary Containment Control . . . . . . . . . . . . 17 5.4 Radioactivity Release Control . . . . . . . . . . . . 18
6.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . 19 7.0 REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . 20 8.0 ATTACHMENTS . . . . . . . . . . . . . . . . . . . . . . 20
- 1. MINIMUM CORE FLOODING INTERVAL . . . . . . . . . . . 21
- 2. MAXIMUM CORE UNCOVERY TIME LIMIT . . . . . . . . . . 22 Rev. No. 6 Page 3 of 22
EOP ENTRY AND USE E'-I.
!COM7.1.1 1.0 PURPOSE %4 To provide general guidance for use of E O P s and recognition of EOP entry conditions.
This procedure applies during all plant operating modes, except when reactor coolant temperature is less than 212OF and a reactor startup or shutdown is not in progress.
2.0 PRECAUTIONS None
' 3.0 PREREQUISITES None Rev. No. 6 Page 4 of 22
4.0 SPECIAL INSTRUCTIONS 4.1 Plant Conditions and Parameters 4.1.1 Monitor the general state of the plant.
4.1.2 Monitor the following parameters using multiple indications:
Reactor power RPV water level RPV pressure Drywell temperature Drywell pressure Torus water level Torus water temperature Containment hydrogen Secondary containment temperatures Secondary containment radiation levels Secondary containment differential pressure Crescent area water levels Reactor building floor sump levels Reactor building ventilation exhaust radiation levels 4.1.3 IF local monitoring of a plant parameter is required, THEN perform the following:
A. Evaluate radiological and environmental conditions to determine accessibility.
B. IF access to Reactor Building is required, THEN follow radiation protection requirements established by Radiation Protection.
Rev. No. 6 Page 5 of 22
EOP ENTRY AND USE EP-1 A. IF an EOP entry condition occurs, THEN enter that EOP, or re-enter if that EOP has already been entered. Exceptions to this requirement are in Steps 4.2.1.E and 4 . 2 . 1 . F below.
B. IF an EOP has been entered, THEN determine whether concurrent entry into Emergency Plan is warranted.
C. WHEN an operating parameter is trending such that an EOP entry condition is imminent or inevitable, the SM or CRS may enter the applicable EOP.
D. WEIEN an EOP exit condition is satisfied, or it has been determined that an emergency no lonqer exists, enter the appropriate operating and/or abnormal operating procedures.
E. IF primary containment flooding is or was required, THEN exit the EOPs and enter the SAOGs. EOPs are not re-entered while in SAOGs, even if an EOP entry condition occurs.
F. Reactor building dP could momentarily meet the EOP-5 entry condition (at or above 0 inches of water) while manually isolating the reactor building during normal plant operation. This is an expected system response and EOP-5 entry is not required provided that dP becomes negative following isolation. Entry into EOP-5 is expected during an automatic isolation or emergency if reactor building dP meets the entry condition.
Rev. No. 6 Page 6 of 22
EOP ENTRY AND USE EP-1 4.2.2 Adequate C o r e Cooling
. I Heat removal from the reactor sufficient to prevent rupturing the fuel clad. Submergence is the preferred mechanism for cooling the core. Steam cooling is relied upon only if RPV water level cannot be restored and maintained above T A F , cannot be determined, or must be intentionally lowered below TAF. The covered portion of the core remains cooled by boiling heat transfer which generates the steam that cools the uncovered portion. Steam cooling will maintain the hottest peak clad temperature below:
Steam Cooling with injection - < 1500 OF Steam Cooling without injection - < 1800 OF
-\---
Rev. No. 6 Page 7of 22
EOP ENTRY AND USE EP-1 4.2.3 Operator Actions/Strategies A. IF conditions or actions specified by a step are not applicable or cannot be implemented, THEN the operator shall proceed to the next step.
B. The SM/CRS may direct isolation of a release path at anytime during an event, provided that isolation of the release path will not conflict with the EOPs.
C. "Anticipation of Emergency RPV Depressurization" is defined as an expectation, based upon evaluation of plant conditions, that an emergency RPV depressurization requirement will soon be reached and cannot be averted by the actions of the EOPs. Before this conclusion can be drawn, the effectiveness of the steps preceding the emergency depressurization requirement must be evaluated. The anticipatory depressurization prescribed by the override requires the MSIVs be open, with the main condenser available and the turbine bypass valves operational (bypassing or defeating the MSIV interlocks is not authorized).
Therefore, when performing EOP-2, Alternate RPV Level Control, Anticipation of Emergency RPV Depressurization is not allowed as adequate core d'
cooling exists; this time is best used attempting to align additional injection sources since the MSIVs will automatically close and render the bypass valves inoperable. If the lowering RPV level trend is reversed, the requirement for emergency depressurization will be unnecessary.
D. Anytime the EOPs direct opening 7 ADS valves, this action is performed irrespective of the resulting RPV cooldown rate.
E. WHEN performing Section ED of EOP-3, the RPV level band is established based upon whether or not level was previously intentionally lowered.
Therefore, if RPV level was intentionally lowered below 110" TAF, this upper limit still applies after the emergency depressurization.
Rev. No. 6 Page 8 of 22
EOP ENTRY Auz3 USE EP-1 F. WHEN performing S t e m Cooling, direction is
-provided to establish a stable or lowering pressure trend. It is preferred to stabilize RPV pressure, to the extent possible, at its existing value. For example: If Steam Cooling is entered and RPV pressure is 500 psig, then RPV pressure should be controlled at or below 500 psig, depending upon the systems available for use. ~f SRVs are being used for pressure control, a.band of 450-500 psig is appropriate. If RPV pressure cannot be stabilized, an alternate approach is to establish a lowering pressure trend. This preserves the assumption of the minimum zero injection RPV water level calculation but will accelerate the rate of inventory l o s s from the RPV. Therefore, the time that steam cooling can be maintained will be shortened.
G. All Available Drywell Cooling is defined as operating 3 of the 4 fans per drywell cooling assembly. Operating all 4 fans/assembly is prohibited as the drywell cooling fan motors could be overloaded causing one or more fans to trip.
H. WHEN performing R P V Flooding, the Emergency Response Organization will generate a procedure for recovery. This procedure must restore R P V level instrumentation and should consider filling reference legs, piping integrity, drywell temperature, and power supplies. In addition, this procedure shall ensure the following:
A method to intentionally lower RPV water level in order to return it on-scale.
Instrument run temperatures are below 212OF.
RPV pressure has remained at least 5 0 psig above torus pressure for the Minimum Core Flooding Interval (Attachment l), prior to terminating all RPV injection sources.
EOP-7 Shutdown Flooding shall be entered if RPV water level is not restored within the Maximum Core Uncovery Time Limit (Attachment 2).
Rev. No. 6 Page 9of 22
.. _ * . I. WHEN in Alternate RPV Water Level Control, the
-direction to maximize injection into the RPV d ensures all available normal and ECCS systems are used to restore RPV water level. If RPV pressure remains high ( > 4 5 0 psig), the low pressure ECCS systems cannot be completely lined up for injection because of interlocks. However, those low pressure ECCS systems should be lined up such that maximum flow will be delivered to the RPV as soon as RPV pressure drops below the systems shutoff head and interlock setpoints.
CS systems will be lined up and running on min flow until RPV pressure is below the shutoff head and 450 psig. If previously Terminated and Prevented, CS will be lined up per EP-5, Termination And Prevention Of RPV Injection.
RHR systems may be used f o r containment control functions as long as injection is prohibited by RPV pressure. However, once RPV pressure is within the shutoff head of the RHR pumps, the RHR system should be realigned to only the L P C I mode until RPV water level is restored. If previously Terminated and Prevented, RHR will be lined up per EP-5.
Rev. No. 6 Page 10 of 22
EOP ENTRY AND USE EP-1 4.3 Manual Control of Automatic Systems (ICOM7.1.2 4.3.1 Do not override an automatic initiation of a safety function unless one of the following conditions exist:
Adequate core cooling is assured by at least two independent indications Misoperation in automatic mode is confirmed by at least two independent indications Required by EOPs kOM7.1.3 4.3.2 IF an operator cannot be dedicated to monitor systems placed in the manual mode, THEN frequently check the system for proper operation and system response. The system is considered inoperable.
kOM7.1.3 4.3.3 WHEN manual operation is no lonser required, return systems to automatic or standby mode.
1COM7 . 1 . 3 4.3.4 Before placing controls in manual for activities which require manual control for an extended period of time, review system response and actions to be taken during potential off-normal events.
4.3.5 IF manual control of an automatic system is desired, THEN reset the initiation signal, if practicable.
This will ensure the system returns to the design setpoint if the system automatically initiates.
4.4 T o r u s Water Temperature IF torus water temperature reaches 120F, THEN depressurize RPV to LESS THAN 200 psig at normal cooldown rates unless restrained by E O P s .
4.5 Containment Instrument Nitrogen Use of containment instrument nitrogen for operation of components inside drywell, for example, MSIVs and SRVs, should take precedence over use of instrument air in order to maintain primary containment inerted during degraded plant conditions.
Rev. No. 6 Page 11 of 22
EOP ENTRY AND USE EP-1 4.6 RHR and Core Spray Operation u COM7 .l.4 4.6.1 Blockage of ECCS pump suction strainers could occur due to debris in the Torus. Within the latitude provided by EOPs to restore and maintain parameters within specified limits, potential mitigative actions may include:
Minimizing ECCS flow or removing affected ECCS pumps from service Alternating ECCS pumps from one division to another, if available Shifting ECCS pump suction to another source, if available Operation of alternate injection sources 4.6.2 Whenever RHR is in the LPCI mode, inject into RPV through RHR heat exchangers as soon as possible and establish RHRSW flow.
4.6.3 Secure RHR and core spray pumps that are not needed to support required actions of EOPs.
4.6.4 Diverting low pressure coolant injection to spray the containment should not be done unless adequate core cooling can also be maintained, or as directed per EOPs.
4.6.5 WHEN performing both EOP-2 and EOP-4, maintaining adequate core cooling normally takes precedence over maintaining containment parameters. Utilizing RHR flow for LPCI injection, containment spray, or torus cooling, singularly or in combination, is permissible provided continuous LPCI injection is not required for adequate core cooling.
Rev. No. 6 Page 12 of 22
EOP ENTRY AND USE EP-1 4.6.5 IF drywell or torus hydrogen concentration cannot be c determined to be below 58, AND drywell or torus oxygen concentraEion cannoz be determined to be below 5 % ,
THEN operate sprays irrespective of acequate c o r e cooling.
CAUTION Elevated crescent area temperature affects RHR and core spray pump motor winding temperatures and could lead E O motor failure.
4.5.7 IF an RHR or core spray pump motor winding temperature reaches alarm setpoint, AND that pump is needed to support required. accions of EOPs, THEN consider reducing pump flow rate o r using an alternate system.
4.7 Reactor Shutdown Determination Rev. No. 6 Page 13 of 22
EOP ENTRY AND USE EP-I 4.8 Procedure Use While Performing E O P s 4.8.1 AOP-1, Reactor Scram, immediate operator actions snould be performed concurrently with ir?ltial enzry into E O P - 2 .
4.8.2 AOP-1 subsequent operator actions, such as resetring the scram and balance of plant, should be performed concurrently with E O P s to aid in recovery. However, actions taken shall not contradict or subvert actions specified by the EOPs and shall not cause the l o s s or unavailability of equipment required by the EOPs.
4.8.3 AOP-39, Loss of Coolant, should be performed concurrently with applicable EOPs for events involving a loss of coolant inside the primary containment. However, actions taken per AOP-39 shall not contradict or subvert actions specified by the E O P s and shall n o t cause the l o s s or unavailability of equipment required by the EOPs.
4.8.4 AOP-40, Main Steam Line Break, should be performed concurrently with applicable EOPs for events involving a piping break in a main steam line outside the primary containment. However, actions taken per AOP-40 shall not contradict or subvert actions specified by the EOPs and shall not cause the loss or unavailability of equipment required by the EOPs.
4.8.5 Other plant procedures may be used in conjunction with EOPs to enhance emergency response and recovery. However, actions taken per other plant procedures shall not contradict or subvert actions specified by the EOPs and shall not cause the loss or unavailability of equipment required by the EOPs.
Rev. No. 6 Page 14 of 22
EOP ENTRY AND USE EP-1 5.0 PROCEDURE Monitor parameters using multiple indications specified by this procedure to determine actual status of parameter.
NOTE: Indications are listed in order of preference under each parameter.
5.1 RPV Control R W Water Level
- SPDS display
- RX WATER LVL 02-3LI-85A and 02-3LR-85B at panel 09-5
- Annunciator 09-5-1-31 RPS RX VESSEL LO LVL TRIP RPV Pressure
- SPDS display
- RX VESSEL PRESS 06PI-61A and B, and 06PR-61A and B at panel 09-3
- Annunciator 09-5-1-22 RPS HI RX PRESS TRIP Reactor Power
- SPDS display
- APRM chart recorders at panel 09-5
- APRM meters at panel 09-14
- IRM chart recorders at panel 09-5
- IRM meters at panel 09-12 Reactor Shutdown
- Full Core Display FULL-IN green lights at Panel 09-5
- SPDS Plant Display control rods full-in indication
- Four rod display notch position indication at Panel 09-5
- EPIC Full Core Rod Scan
- EPIC Rods In Monitor Program (RIMP)
Rev. No. 6 Page 15 of 22
.~
EOP ENTRY AND USE EP-1 5.2 Primary Containment Control T o m s Temperature
- SPDS display
- TORUS TEMP A 16-1TR-131A and TORUS TEMP B 16-1TR-131B at panel 09-3
- Average of bay temperatures obtained individually at 16-1TI-131A and 16-1TI-131B at MAP panel Drywell Temperature
- SPDS display
- Average of temperatures on DW TEMP A 16-1TR-108 and DW TEMP B 16-1TR-107 at panel 09-3
- Average air inlet and outlet temperature for any drywell cooling assembly that has at least one f a n operating (68TI-100 or 68TI-101 at panel 09-75).
Drywell Pressure
- SPDS display
- NR PC PRESS 27PI-115Al and 27PR-115A1, and 27PI-115B1 and 27PR-115B1 at panel 09-3
- Annunciator 09-5-1-21 RPS HI DW PRESS TRIP
- WR PC PRESS 27PI-115A2 and 27PR-115A2, and 27PI-115B2 and 27PR-115B2 at panel 09-3 Torus Level
- SPDS display
- TORUS LVL 23LI-202A and 23LR-202A, and TORUS LVL 23LI-202B and 23LR-202B at panel 09-3 Containment Hydrogen
- SPDS display
- Panel 27PCX-101A and 27PCX-101B in Relay Room
- Grab samples Rev. No. 6 Page 16 of 22
EOP EMTRY AND USE EP-1 5.3 Secondary Containment Control Differential Pressure
- SPDS display Area Temperature High
- SPDS display
- Panels 09-95, 09-96, 09-75, and 09-21, per Table 5-1 of EOP-5
- IF an area does not have remote temperature indication, THEN monitor that area locally.
IF that area is inaccessible, AND a primary system is discharging into Secondary Containment, THEN assume the area temperature is above the maximum safe level.
Reactor Building Radiation Levels
-v-
- SPDS display
- ARM at panel 09-11
- Local monitoring by Radiation Protection survey
- IF the area is accessible AND the radiation levels in the area are not available from an ARM (ARM is either out of service or is not installed in the area),
THEN locally monitor that area with a portable ARM.
IF the area is inaccessible, AND the radiation levels in the area are not available from either the installed ARM or Radiation Protection survey, THEN assume the radiation levels in that area are above the maximum safe level.
Rev. No. 6 Page 17 of 22
EOP ENTRY AND USE EP-1 reactor Building..Vent Exhaust
- SPDS display' I -.
- RX BLDG VENT RAD MON A 17RIS-452A and RX BLDG VENT RAD MON B 17RIS-452B at panel 09-12
- REFUEL FLOOR EXH RAD MONITOR 17RM-456A at panel 66HV-3A and REFUEL FLOOR EXH RAD MONITOR 17RM-456B at panel 66HV-3B (Reactor Building 272')
- Annunciators 09-3-2-29 RX BLDG VENT RAD MON HI and 09-75-1-15 REFUELING FLOOR EXH RAD MON INOP OR HI IF Reactor Building is inaccessible, THEN assume Refuel Floor exhaust monitor is GREATER THAN i o 3 counts per minute.
Reactor Building Floor Sump Level
- SPDS display
- Annunciators 25-17-1-1 RB FLR SUMP A LVL HI and 25-17-1-2 RB FLR SUMP B LVL HI at panel 25-17 in Radwaste Control Room
- Local observation IF Reactor Building is inaccessible, THEN assume reactor building floor sump level is GREATER THAN high alarm setpoint.
Crescent Area Water Level
- SPDS display
- Local observation IF Reactor Building is inaccessible, AND a primary system is discharging into Secondary Containment, TEEEN assume crescent area water level is GREATER THAN 18 inches.
5.4 Radioactivity Release Control Offsite release rates and emergency classification are determined.by Site Emergency Plan.
Rev. No. 6 Page 18 of 22
6.0 REFERENCES
6.1 Performance References 6.1.1 AOP-1, Reactor Scram 6.1.2 EOP-2, RPV Control 6.2 Developmental References 6.2.1 ODSO-28, Revision 4, EOP Entry and Use 6.2.2 EOP-2, RPV Control 6.2.3 EOP-3, Failure to Scram _-
6.2.4 EOP-4, Primary Containment Control 6.2.5 EOP-5, Secondary Containment Control 6.2.6 EOP-6, Radioactivity Release Control 6.2.7 JTS-95-0221, Operability Assessment for DER 95-0740
- 0748; Industry Notification of B-Fill Qualification Limits for Certain ITT-Barton Indicating Switches 6.2.8 JSED-95-0100, Impact of ITT Barton Industry Advisory on the Environmental Qualification of lODPIS-l25A&B, 14FIS-45A&B, and 27PS-l10A&B 6.2.9 GE Letter JAB-N8075, dated 11/2/98, MSBWP Results for FitzPatrick Cycle 14 (GE letter 262-98-172 and DRF J11-03359) 6.2.10 JENG-02-0375, dated 10/12/2002, EFFECT OF JD-01-102 (CORE RELOAD) AND DECAY HEAT CURVE CHANGES ON EOP CURVES Rev. No. 6 Page 19 of 22
EOP ENTRY AND USE EP-1 7.0 REQUIREMENTS 7.1 Cmmnitmeults 7.1.1 NRCI-94-03, JAFP-94-0175, ACTS Item 10946. Created EOP Support Procedures ( E P s ) .
7.1.2 NRCN-92-47, Intentional Bypassing of Automatic Actuation of Plant Protective Features (OER 920483, JTS 0799 )
7.1.3 ACTS Item 5899, incorporate INPO SER 87-34 (OER #870335).
7.1.4 JAFP-94-0228, Response to NRC Bulletin No. 93-02, Supplement 1, Debris Plugging of Emergency Core Cooling Suction Strainers. Added special instruction to alert operators of the potential for ECCS suction strainer clogging and to adjust flow consistent with required needs t o mitigate the clogging.
7.2 Validation Validated per .AP-02.02.
8.0 ATTACHMENTS
- 1. MINIMUM CORE FLOODING INTERVAL
- 2. MAXIMUM CORE UNCOVERY TIME LIMIT Rev. No. 6 Page 20 of 22
8 . -
EOP ENTRY AND USE EP-1 ATTACHMENT 1 Page 1 of 1 MINIMUM CORE FLOODING INTERVAL Number of Open SRVs Flooding Interval (rnin. )
7 or more 22 6 30 t
5 44 f
5 Rev. No. 6 Page 21 of 22
EOP ENTRY AND USE EP-1 ATTACHMENT 2 Page 1 of 1 MAXIMUM CORE UNCOVERY TIME LIMIT Maximum Core Uncovery Time Limit 14 12 IO 8
(;
1 2
0 0 20 40 (; 0 80 100 Time Since Reactor Shutdown [hr]
Page 22 ofi 72
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Low Suppression Pool Wtr Lvl I 5 Group # 1 1 Ability to interpret control room indicationsto WA# 295030 2.4.48 2.4.48 veriw the status and operation of system I and understand how operator actions and directives affect plant and system conditions.
(CFR: 43.5 145.12)
Importance Rating 3.5 3.8 Proposed Question: While experiencing torus water level control prc,lems, an 0peri.x opens an ADS valve with torus water level at 5.2 ft.
Opening the SRV under these conditions will result in:
a) direct suppression chamber pressurization ROISRO b) excessive hydrodynamic loading 16119 c) valve seat damage from the excessive flowrates.
d) drawing water up into the tailpipe.
Proposed Answer: a) direct suppression chamber pressurization Explanation (Optional):
Technical Reference(s): EOP-2 (Attach if not previously provided)
L, Proposed references to be provided to applicants during examination: None Learning Objective: MIT301.11E- EO 4.03 (As available)
Question Source: Bank # Dresden IINPO # 6483 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 912611998 (Optional Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 3 55.43 5 Comments:
Page 26 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Low Suppression Pool Wtr Lvi I 5 Group # 1 I Ability to interpret control room indications to K/A# 295030 2.4.48 2.4.48 verifythe status and operation of system I and understand how operator actions and directives affect plant and system conditions.
(CFR: 43.5 / 45.12)
Importance Rating 3.5 3.8 Ah Proposed Question: While experiencing
. . torus water level control problems,& operator opens an ADS valve 8 . .
I I Opening the SRV under these conditions will result in:
a) direct suppression chamber pressurization RO/SRO b) excessive hydrodynamic loading 16119 c) valve seat damage from the excessive flowrates.
d) drawing water up into the tailpipe.
I p ) &/a& su chA2A Proposed Answer:
Explanation (Optional): NO UNK to 2.4.48 was linked to 2.4.18 t
i Fc 2nz;p3lA r Technical Reference(s): fntr 3ol.\\@ 4 . b . 3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: (As available)
Question Source: Bank # Dresden 1INPO # 6483 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 912611998 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comorehension or Analvsis 10 CFR Part 55 Content: 55.41 55.43 Comments:
Page 26 of 129 NRC Written Examination Begin to ENDdocLast printed 5/21/2003 8:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Reactor Low Water Level / 2 Group # 1 1 Knowledge of the reasons for the following WA# 295031 EK3.01 EK3.01 responses as they apply to REACTOR LOW WATER LEVEL : (CFR: 41.5 / 45.6)
Automatic depressurization system actuation Importance Rating 3.9 4.2 Proposed Question: From full power and with HPCl inoperable, a SCRAM occurs from a small primary leak in the drywell simultaneous with a loss of offsite power. EDG's start and reenergize vital busses. RClC initiates, but insufficient injection results in RPV water level continuing to lower.
Which of the following is correct assuming operator action?
a) SRV's should cycle open on their automatic pressure relief setpoints and lower reactor pressure to permit level recovery injection with the Condensate system.
RO/SRO b) SRVs assigned to ADS should open when RPV level lowers to an assigned setpoint to permit level recovery injection with low pressure ECCS.
17/20 c) A residual bus transfer will result in automatic start and injection by the Condensate Booster Pumps.
d) SRV's should cycle on their automatic pressure relief setpoints and together with the reduced RClC injection will provide steam cooling with injection.
Proposed Answer: b) SRV's assigned to ADS should open when RPV level lowers to an assigned setpoint to permit level recovery injection with low pressure ECCS.
W Explanation (Optional):
Technical Reference@): 0P-68 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O2J, EO-I .01, 1.05.A, 1.05.C (As available)
Question Source: Bank ##
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis A 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Page 27 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO i
Tier # 1 1 SCRAM Condition Present and Power Above APRM Group # 1 1 Downscale or Unknown / 1 Ability to determine andor interpret the following WA # 295037 EA2.02 EA2.02 as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :
(CFR: 41.1 0 / 43.5 / 45.13)
Reactor water level Importance Rating 4.1 4.2 Proposed Question: As directed by EOP-3, the current RPV level band is -1 9 to 110 inches and being controlled at 80-100 inches with Feedwater.
Which of the following is the preferred instrumentation for maintaining the 80-100 inches band?
a) Narrow Range.
ROlSRO b) Wide Range.
18/21 c) Refuel Zone.
d) Fuel Zone.
Proposed Answer: b) Wide Range.
Explanation (Optional):
Technical Reference@): SDLP-O2B, Table IV (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O2B, EO-1.05.A.3 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 5 Comments:
Page 28 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Off-site Release Rate I 9 Group # 1 1 Knowledge of the interrelations between HIGH WA # 295038 EK2.02 EK2.02 OFFSITE RELEASE RATE and the following:
(CFR: 41.7 145.8)
Offgas system Importance Rating 3.6 3.8 Proposed Question: While operating at full power, a large fuel leak develops.
Which of the following automatic responses from a high radiation signal will occur to limit off-site release rates?
a) Condenser Vacuum Pump trip.
ROISRO b) Off gas System isolation.
19/22 c) Off gas Recombiner trip.
d) Reactor SCRAM.
Proposed Answer: b) Off gas System isolation.
Explanation (Optional):
Technical Reference@): OP-24A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-OIA, EO-1.05.C.1 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis I O CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 29 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
- - Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Plant Fire On Site I 8 Group # 1 1 Ability to operate and I or monitor the following K/A # 600000 AAI .06 AAI .06 as they apply to PLANT FIRE ON SITE:
Fire alarm Importance Rating 3.0 3.0 Proposed Question: With the Fire Protection System in a normal standby lineup, which one (1) of the following Fire Protection Panel Alarms will always be accompanied by the start of one or more Fire Pumps?
a) Heat detection actuation in the West Cable Tunnel ROlSRO b) Heat detection actuation in the North EDG Switchgear Room 20123 c) Ionization detector actuation in the Reactor Building 272 Drywell Entrance d) Ultraviolet Flame detector in the Recirculation MIG Room Proposed Answer: a) Heat detection actuation in the West Cable Tunnel Explanation (Optional):
Technical Reference(s1: OP-33 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None W Learning Objective: SDLP-76 EO 1 . 0 5 ~ (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 30 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
L Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Loss of Main Condenser Vac / 3 Group # 2 2 Ability to determine andlor interpret the following WA # 295002 AA2.01 AA2.01 as they apply to LOSS OF MAIN CONDENSER VACUUM : (CFR: 41.10 I43.5 145.13)
Condenser vacuum/absolute pressure Importance Rating 2.9 3.1 Proposed Question: Reactor power is 38%,on the APRM's when annunciator 09-6-1-29, CONDENSER VAC LOW, alarms. Condenser vacuum, as read on control room meters, indicates 24.8' and lowering slowly. If vacuum continues to lower, WHICH ONE (1) of the following automatic protective actions would occur first?
a) Reactor Feed Pump Turbine Trip RO/SRO b) Main Turbine Trip 21124 c) Bypass Valve Closure d) MSlV Closure Proposed Answer: b) Main Turbine Trip Explanation (Optional):
Technical Reference(s): OP-9, OP-2A, OP-I, AOP-31 (Attach if not previously provided) v' Proposed references to be Drovided to applicants during examination: None Learning Objective: LP-AOP, EO-1.02 (As available)
Question Source: Bank # Nine Mile Point 1 INPO # 11813 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/20/1998 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 31 of 120 NRC Written Examination SubmittaI.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO W' Tier # 1 1 Loss of Main Condenser Vac 1 3 Group # 2 2 Ability to determine andlor interpret the following,, KIA # 295002 AA2.01 AA2.01 as they apply to LOSS OF MAIN CONDENSER VACUUM :(CFR: 41.10143.5145.13)
Condenser vacuurdabsolute pressure
1 Importance Rating 2.9 3. I 4
Proposed Question: Reactor power is 38%, on the APRM's when annunciator-the following automatic protective actions would occur first?
21/24 c) Reactor Scram d) MSlV Closure
'pc\b>
Proposed Answer. b) Turbine Trip Explanation (Optional):
Technical Reference(s): 00- 9 ., OP- ; . &?
I
- ,I (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: -< .>-
I, (As available)
Question Source: Bank # Nine Mile Point 1 INPO # 11813 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/20/1998 (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:
Page 31 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Reactor Pressure / 3 Group #
Knowledge of the interrelations between HIGH KIA # 295007 AK2.04 AK2.04 REACTOR PRESSURE and the following:
(CFR: 41.7 / 45.8)
LPCS Importance Rating 3.2 3.3 Proposed Question: An Emergency Depressurization is to be performed from 700 psig to permit low pressure ECCS injection into the reactor. The only ECCS available is Core Spray System A. CS Pump A is running on minimum flow and all other components are in a normal standby condition. When SRVs are operated, only one (1) SRV responds. Reactor pressure lowers at approximately 10 psilminute.
The Core Spray Injection Valve opens when reactor pressure goes below RPV injection immediately.
a) 450 psig: occurs ROlSRO b) 450 psig: does occur 22/25 c) 310 psig: occurs d) 310 psig: does occur Proposed Answer: b) 450 psig: does occur Explanation (Optional):
J Technical Reference(s1: OP-14 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning 0bjective: SDLP-14, EO-1.13e, 1.14b (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 32 of 120 NRC Written Examination Submittaldoc Last printed 6/6/2003 1027 AM
L-Examination Outline Cross-reference: Level RO SRO Tier # 1 1 Inadvertent Reactivity Addition 1 1 Group # 2 2 Knowledge of the interrelations between KIA# 295014 AK2.07 AK2.07 INADVERTENT REACTIVITY ADDIRON and the following: (CFR: 41.7 145.8)
Reactor power Importance Rating 3.9 3.9 Proposed Question: From normal full power operation, which of the following WILL cause reactor power to rise?
a) Inadvertently isolating the Reactor Water Cleanup System.
RO/SRO b) Removing local RPS fuses for a control rod hydraulic control unit.
23/26 c) Main Condenser Circulating Water Pump Trip.
d) Closing the manual extraction steam valve for Feed Heater 6B.
Proposed Answer: d) Closing the manual extraction steam valve for Feed Heater 65.
Explanation (Optional): Explanation: The manual extraction steam valve for Feed Heater 6B closing will prevent the heating of the feedwater in the 6B heater, thereby, causing colder feedwater to enter the vessel and drive reactor power up.
Technical Reference(s): AOP-62, AOP-32, OP-3A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP EO 1.02 (As available)
Question Source: Bank # Clinton INPO # 20412 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 7/23/2001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 33 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 10:27AM
Inadvertent Reactivity Addition / 1 Group # 2 2 Knowledgeof the interrelations between W A # 295014 AK2.07 AK2.07 INADVERTENT REACTIVITY ADDITION and the following: (CFR: 41.7 / 45.8) , , 3°C fSA4b Reactor power , & nnn&
~
b - ~ f
- - Importance Rating 3.9 3.9 Proposed Question: following would cause reactor power to w?
A CGXL-- &a) . ng.
*c.u u
RO/SRO +\oJiLr;diet Valve opening.
23/26 *C) 5 closing. 1 -k-&
r-k c c w Q-& &&-$;a, d ) 3 6B Extraction Steam &Valve Proposed Answer: 96B .Extraction Steam Skd.iCValve closing.
VI /
Explanation (Optional): Explanation: The 6B Extraction Steam Shutoff Valve closing will prevent the heating of the feedwater in the 6B heater, thereby, causing colder feedwater to enter the vessel and drive reactor power up. NO KA AK 2.07exists and is tied- make a tie (was tied to 295014.aa2.01 Technical Reference(s): At,;.- ox?. 3 .q -I f40i~1;\;r, (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: Li- &bP I,0 2 (As available)
Question Source: Bank # Clinton INPO # 20412 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 7/23/2001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:
Page 33 of 129 NRC Written Examination Begin to END-docLastprinted 5/21/2003 8144 AM
Examination Outline Cross-reference: Level SRO Tier # I Refueling Acc Cooling Mode / 8 Group ## I Knowledge of the operational implications of the WA # 295023 AKI .01 following concepts as they apply to REFUELING ACCIDENTS :(CFR: 41.8 to 41.10)
Radiation exposure hazards Also IOCFR55.43(b)(4)
Importance Rating 4.1 Proposed Question: Core Alterations are in progress.
An irradiated fuel bundle being moved from the reactor cavity to the Spent Fuel Pool becomes ungrappled and falls into the reactor vessel downcomer area. (Between the vessel wall and the shroud)
Which of the below describes the person at greatest risk?
a) Mechanic working on Torus to Drywell Vacuum Breaker.
ROISRO b) Refuel SRO on the Bridge S27 c) I&C Technician at SLC Skid.
d) Mechanic working on SRVs Proposed Answer: d) Mechanic working on SRVs Explanation (Optional):
Technical Reference(s): RAP-7.1.1.04B (Attach if not previously provided)
'-4 Proposed references to be provided to applicants during examination: None Learning Objective: LP-AP, RAP-7.1.04B73.03 (As available)
Question Source: Bank # Clinton INPO # 20401 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 7/23/2001 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 9 55.43 4 Comments:
Page 34 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level Tier #
Refueling Acc Cooling Mode / 8 Group #
Knowledge of the operational implications of the WA # 295023 following concepts as they apply to REFUELING ACCIDENTS :(CFR: 41.8 to 41.10) osure hazards Importance Rating Proposed Question: Core Alterations are in progress.
An irradiated fuel bundle being moved from the reactor cavlty toSWBbecomes ungrappled and falls into the reactor vessel downcomer area. (Between the vessel wall and the shroud)
Which of the the following people would be at greatest risk of radiation overexposure?
a)
RO/SRO T_ {
\
c q Refuel SRO on the Bridge S27 c) hoqTechnician at SLC Skid.
d) Mechanic working on SRVs Proposed Answer: d) Mechanic working on SRVs Explanation (Optional):
Technical Reference(s): @f 3, 1. w& (Attach if not previously provided)
Proposed references to be provided to applicants during examination: #@Pa Learning Objective: u-,qe Qlf'7.\.648 73,03(A~avaiIable)
Question Source: Bank # Clinton INPO # 20401 b.0)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 7/23/2001 (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis
\ I 10 CFR Part 55 Content: 55.41 9 Comments:
Page 34 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44AM
i- Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Secondary Containment Group # 2 2 Area Temperature / 5 Ability to perform procedures to reduce KIA# 295032 2.3.10 2.3.10 excessive levels of radiation and guard against personnel exposure. (CFR: 43.4 / 45.10)
Importance Rating 2.9 3.3 Proposed Question: Which one of the following describes an EOP-5, "Secondary Containment Control,"
basis for isolating a primary system discharging into the secondary containment?
a) To minimize RPV inventory losses.
ROISRO b) To backup PClS automatic functions.
24/28 c) To terminate rising radiation levels.
d) To ensure Recirculation M/G Room access Proposed Answer: c) To terminate rising radiation levels.
Explanation (Optional): Secondary Containment Control does not maintain habitability for all areas. The Max Safe values are based on equipment operability and personnel access necessary for EOP actions. The Recirc MG set room is not one of the areas requiring access.
Technical Reference(s): EOP-5 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EOPSLP, EO-1.07 (As available)
Question Source: Bank # Cooper 1 INPO ## 302 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 2/12/1999 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 4 Comments:
Page 35 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Secondary Containment Group # 2 2 Area Temperature I5 Ability to perform procedures to reduce W A # 295032 2.3.10 2.3.10 excessive levels of radiation and guard against personnel exposure. (CFR: 43.4 145.10)
Importance Rating 2.9 3.3 hr?
Proposed Question. Which one of the following describes- EOP-5g "secondary Containment Control,"
basis for isolating a system discharging into the secondary containment?
f (
?
\ - - ____ f a) To minimize RPV inventory losses.
ROISRO b) To backup PClS automatic functions.
24128 c) To terminate rising tocl4geF9ktfe2 radiation I e v e l q a F .
'd) r o - R M i n t a i o ~ h e ~ a s s ~ e ~ ~ ~ ~ ~ e l .
Proposed Answer: c) To terminate rising , radiation levels, . -
Explanation (Optional): Secondary Containment Control does not maintain habitability for all areas. The Max Safe values are based on equipment operability and personnel access necessary for EOP actions. the Recirc MG set room is not one of the areas requiring access. NO KA Tie to 2.3.10- was tied to 295032.K3.03 Technical Reference(s): _--
'j .? 5- (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: t lk r 5,-1 . i\ I-
! . : ' I (As available)
Question Source: Bank ## Cooper 1 INPO # 302 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 2/12/1999 (Optional Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:
Page 35 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44 AM
L- Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Secondary Containment Group # 2 2 Area Radiation Levels 19 Ability to operate andlor monitor the following as KIA# 295033 EA1.05 EA1.05 they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS : (CFR: 41.7 145.6)
Affected systems so as to isolate damaged portions Importance Rating 3.9 4.0 Proposed Question: During power operation, the Area Radiation Monitor (ARM) for the CRD Removal Hatch area alarms, together with receipt of a Fire Protection System ionization detector alarm in the Southwest Drywell Entrance Area.
Which system@)should be considered for manual isolation?
a) HPCl and RWCU RO/SRO b) RCIC and Main Steam 25/29 c) Main Steam and RWCU d) HPCI and RClC Proposed Answer: d) HPCI and RClC Explanation (Optional):
Technical Reference@): EOP-5 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EOP5LP, EO-1.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since IO195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Coanitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 36 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
L-Examination Outline Cross-reference: Level SRO Tier # 1 High Reactor Pressure / 3 Group # I Ability to determine andlor interpret the following K/A # 295025 EA2.05 as they apply to HIGH REACTOR PRESSURE:
(CFR: 41.10/43.5 145.13)
Decay heat generation Importance Rating 3.6 Proposed Question: The plant is starting up after an extended (60 day) outage. At 15% power, a complete loss of EHC results in a manual Reactor SCRAM.
Which of the following describes the expected procedural actions?
a) Ensure one (1) Bypass Valve opens to control RPV pressure per EOP-2, RPV Control.
RO/SRO b) Startup HPCl and control RPV pressure, 900-1050 psig, per AOP-1, Reactor SCRAM.
S30 c) Reduce Secondary Plant steam loads to control RPV pressure, 900-1050 psig, per AOP-1, Reactor SCRAM.
d) Open one (1) SRV and control RPV pressure, 800-1000 psig, per EOP-2, RPV Control.
Proposed Answer: c) Reduce Secondary Plant steam loads to control RPV pressure, 900-1050 psig, per AOP-I, Reactor SCRAM.
Explanation (Optional):
-- Technical Reference(s): A 0 P- 1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning 0bjective: LP-AOP, EO-1.10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 37 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO L
Tier # 1 1 Secondary Containment High Sump/Area Water Group # 2 2 Level I5 Knowledge of the reasons for the following WA# 295036 EK3.01 EK3.01 responses as they apply to SECONDARY CONTAINMENT HIGH SUMPIAREA WATER LEVEL : (CFR: 41.5 145.6)
Emergency depressurization Importance Rating 2.6 2.8 Proposed Question: While operating at full power, an earthquake has resulted in the following:
A severe piping crack between the CSTs and the Torus results in a rapid addition of water to the Torus Room and both Crescent Areas.
A small, un-isolable leak in the RWCU Pump suction piping in the Reactor Building.
Crescent Area water levels are 19 rising 0 Highest Reactor Building Area (RB 300 Southwest) temperature is 103°F Why must an Emergency Depressurization be performed for these conditions?
a) A loss of CST inventory will result in total loss of HPCl and RClC for inventory control.
ROISRO b) Operability of equipment located in the Crescents is threatened by Crescent water level rise.
26131 c) Primary Containment integrity is threatened by Torus Room water level rise.
d) Operability of RPV Water Level instruments located on Reactor building 300 is challenged.
Proposed Answer: b) Operability of equipment located in the Crescents is threatened by Crescent water level rise.
Explanation (Optional):
Technical Reference(s): EOP-5 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOPs Learning Objective: EOPSLP, EO-1.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 5 1
Comments:
Page 38 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Page 39 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination 0utline Cross-reference: Level RO SRO Tier # 1 1 High CTMT Hydrogen Conc. 15 Group # 2 2 Ability to operate and monitor the following as KIA # 500000 EA1.06 EA1.06 they apply to HIGH CONTAINMENT HYDROGEN CONTROL: (CFR: 41.7 I 45.6)
Drywell sprays Importance Rating 3.3 3.4 Proposed Question: Which of the following requires initiation of Drywell Sprays?
DW H2 DW O2 Torus H2 Torus 0' a) 6.03 % 5.4 % 6.13 % 3.0 %
ROISRO b) 6.13 % 3.0 % 6.03 % 5.4 %
27132 c) 5.9 % 6.03 % 6.13 % 3.0 %
d) 3.0 % 6.03 % 6.13 % 5.4 %
Proposed Answer: a ) 6.03 % 5.4 % 6.13 % 3.0 %
Explanation (Optional):
Technical Reference(s): EOP-4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOP-4A, Primary Containment Gas Control Learning Objective: EOP4LP, EO-4.03 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NRN Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7, 10 55.43 5 Comments:
Page 40 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
L .
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RHRILPCI: Injection Mode Group # 1 1 Ability to predict andlor monitor changes in WA# 203000 A I .05 A I .05 parameters associated with operating the RHWLPCI INJECTION MODE (PLANT SPECIFIC) controls including: (CFR: 41.5 145.5)
Suppression pool live1 Importance Rating 3.8 3.7 Proposed Question: A Design Basis LOCA has occurred. ECCS systems are injecting into the reactor.
Suppression Pool Level is at 12.8 feet and lowering.
Which one of the following would be the expected response of the Low Pressure Coolant Injection (RHR)?
a) The RHR pumps will continue to operate regardless of Suppression Pool Level until the pumps trip on motor overload.
ROISRO b) The RHR pumps will automatically trip when Suppression Pool Level drops to the RHR Pump Vortex Limit.
28133 c) The RHR pumps will continue to operate regardless of Suppression Pool Level due to automatic bypass of all trip signals.
d) The RHR Pump Suppression Pool Suction valves will automatically close. The RHR pumps will all trip on Interlock.
Proposed Answer: a) The RHR pumps will continue to operate regardless of Suppression Pool Level until the pumps trip on motor overload.
Explanation (Optional):
\ -
Technical Reference(s): OP-13A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10, EO-1.lO.f (As available)
Question Source: Bank # Grand Gulf 1 INPO # 16342 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 4l112000 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Page 41 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RHRRPCI: Injection Mode Group # 1 1 Ability to predict andlor monitor changes in WA# 203000 Al.05 Al.05 parameters associated with operating the RHWLPCI INJECTION MODE (PLANT SPECIFIC) contrds including: (CFR: 41.5 145.5)
Suppression pool level 7---
--- --c----- -- .-- b e f i a n c e Ratipg--..-.-...- __
"-3.8 m 3.7 Proposed Question: A DBA LOCA has occurred. ECCS systems are injecting into the reactor. Suppression at "3-feet and lowering. 1 hich ongof the followingwould be the expected response of the Low Pressure Coolant Injection (RHR)?]
a) The RHR pumps will continue to operate regardless of Suppression Poolbevel until the pumps trip oncuxh41 motor overload.
call .I
- e Lw:L >
I ' 8 &-ROISRO b) The. RHR v
. pumpsi wil1h.I. trip when Suppression Pool Level drops to W
. \rbrte*L k&
t c) The RHR pumps will A**?
28/33 t n . d~9 ~ ~ ~ ~ + \ a ~ ~ + . y s
& & & c close their Suppression
~ Pool Suction
~ valves a d d t q ~ ~
e to NO suction flowpath.
/ Lvrl\ +
Proposed Answer. a) The RHR pumps will continue to operate regardless of Suppression Pool Level until the pumps trip on motor overload.
Explanation (Optional): NO KA Exists- Was tied to 203000.K1.02 Modify question to torus vs SP.
Technical Reference(s): @- r3A (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: &L/?.P I *I0.f (As available)
Question Source: Bank # Grand Gulf 1 INPO # 16342 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 41112000 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:
c 6w Page 41 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44AM
Examination Outline Cross-reference: Level SRO Tier # 1 Low Suppression Pool Wtr Lvl I 5 Group # 1 Ability to determine andlor interpret the following WA # 295030 EA2.04 as they apply to LOW SUPPRESSIONPOOL WATER LEVEL :(CFR: 41.10 143.5 145.13)
Drywell1suppression chamber differential pressure:
Mark-l&ll Importance Rating 3.7 Proposed Question: Two hours into the shift, the SNO reports that Torus water level has dropped from 14.0 ft to 13.91ft while Drywell to Torus AP has dropped from 1.8 psid to 1.6 psid and Torus pressure has remained constant at 0.0 psig. You have confirmed the indications on EPIC-LOGI.
Your actions should be... ... ....
a) Enter EOP-4, Primary Containment Control, and immediately makeup nitrogen to the Drywell restore AP.
ROISRO b) Enter EOP-4, Primary Containment Control, and immediately makeup water to the Torus to restore Torus level.
s34 c) Enter AOP-9, Loss of Primary Containment Integrity, and dispatch Operators to investigate the cause.
d) Enter AOP-9, Loss of Primary Containment Integrity, and dispatch Operators to locate the leaking Drywell to Torus Vacuum Breaker.
Proposed Answer: c) Enter AOP-9, Loss of Primary Containment Integrity, and dispatch Operators to investigate the cause.
Explanation (Optional):
Technical Referenceb): OP-37 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning 0bjective: SDLP-IGB, EO-1.09d (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NRN Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 42 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level SRO Tier # I SCRAM Condition Present and Power Above APRM Group # I Downscale or Unknown I I Ability to determine andlor interpretthe following WA # 295037 EA2.05 as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :
(CFR: 41.10143.5145.13)
Control rod position Importance Rating 4.3 Proposed Question: A failure to SCRAM occurs from a full power MSlV closure.
0 Control Rods were inserted by draining the SDIV and using several manual SCRAMS.
It is now believed that the Reactor will remain shutdown under all conditions without boron.
How can this be confirmed AND what actions will result?
a) Green Full In Lamps on Full Core Display. Secure SLC injection and enter EOP-2, RPV Control.
ROISRO b) EPIC Full Core Rod Scan. Per EOP-2, RPV Control, secure SLC injection.
s35 c) EPIC Solomon Program. Secure SLC injection per EOP-3, Failure to SCRAM, then, enter EOP-2, RPV Control.
d) Rods In Monitoring Program (RIMP). Verify Hot Shutdown Boron Weight and enter EOP-2, RPV Control.
Proposed Answer: a) Green Full In Lamps on Full Core Display. Secure SLC injection and enter EOP-2, RPV Control.
Explanation (Optional):
Technical Reference(s): EP-1, AOP-I, EOP-3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP, EO-1.03, EOP3LP, EO-1.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 43 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Page 44 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RHRILPCI: Injection Mode Group # 1 I Ability to manually operate andlor monitor in the WA # 203000 A4.11 A4.11 control room: (CFR: 41.7 145.5 to 45.8)
Indicating lights and alarms Importance Rating 3.7 3.5 Proposed Question: L P C I h a s a u t o m a t i c a l l y i n i t i a t e d due t o RPV w a t e r l e v e l l o w e r i n g below 59.5. The L P C I i n b o a r d and o u t b o a r d i n j e c t i o n valves a r e open and RPV p r e s s u r e i s 200 p s i g and l o w e r i n g . E P I C i s u n a v a i l a b l e and RHR system flow i n d i c a t i o n s a t p a n e l 09-3 a r e o u t o f service. Which o f t h e f o l l o w i n g i n d i c a t i o n s c o u l d b e used t o h e l p v e r i f y t h a t L P C I i s i n j e c t i n g water i n t o t h e RPV?
RHR PUMP RHR PUMP 1OMOV-16A (B)
MTR AMPS D I S C PRESS POSITION INDICATION a) lowering rising closed ROlSRO b) rising lowering closed 29/36 C) rising rising open d) lowering rising open Proposed Answer: b)rising lowering closed Explanation (Optional):
Technical Reference(s): OP-13A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10, EO 1 . 0 5 . a . l . b (As available)
Question Source: Bank # J A F LOR 2 0 5 0 5 0 0 1 R H R C 1 9 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 45 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Shutdown Cooling Group # 1 I Knowledge of the effect that a loss or WA# 205000 K3.04 K3.04 malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on following: (CFR: 41.7 / 45.4)
Recirculation loop temperatures Importance Rating 3.7 3.7 Proposed Question: The Plant is in Mode 4 with both Recirculation Pumps secured. Shutdown cooling is in service using RHR System A. Reactor water level is steady at 198. Coolant temperature is 140 F with a slow cool down in progress. A loss of Shutdown cooling occurs.
Which one (1) of the following responses will provide for reliable Reactor Coolant temperature indication?
a) Opening Recirculation Pumps suction and discharge valves.
ROlSRO b) Raising reactor water level to 2 234.5.
30137 c) Securing the Reactor Water Cleanup System.
d) Placing the Control Rod Hydraulic System in service.
Proposed Answer: b) Raising reactor water level to 2 234.5.
Explanation (Optional):
Technical Reference(s): OP-I3D, AOP-30, ITS Definitions (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP, EO-1.03, 1.04 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 2 Comments:
Page 46 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level SRO Tier # 1 Reactor Low Water Level 12 Group # 1 Ability to operate andlor monitor the following as WA # 295031 EA 1.08 they apply to REACTOR LOW WATER LDIEL:
(CFR: 41.7 145.6)
Alternate injection systems: Plant Specific Link to 10CFR-55.43(b)(l-6) Importance Rating 3.9 Proposed Question: A startup is in progress at 20% CTP when an RPS electrical malfunction results in the following:
HPCI/RCIC & MSIV Isolation on High Temperature Full Reactor SCRAM One (1) rod remains at position 40 and one (1) other rod is at position 02. All other rods are Full In.
RPV water level is 150 inches, slowly trending down.
RPV pressure is 1000 psig, slowly trending u p .
The correct course of action is to:
a) Enter EOP-3, stabilize RPV pressure, and maintain RPV level with Feed/Condensate.
b) Enter EOP-2, commence a normal cooldown, and maintain RPV ROlSRO level with Feed/Condensate.
S38 c) Enter EOP-3, commence a normal cooldown, and maintain RPV level with SLC/CRD.
d) Enter EOP-2, Emergency Depressurize, and maintain RPV level with SLC/CRD.
Proposed Answer: b) Enter EOP-2, commence a normal cooldown, and maintain RPV level with Feed/Condensate.
Explanation (Optional):
Technical Reference(s): EOP-2, EP-1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EOP2LP, EO-1.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 5 Comments:
Page 47 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
u Examination Outline cross-reference: Level RO SRO Tier # 2 2 HPCI Group # 1 1 Knowledge of 10 CFR: 20 and related facility KIA# 206000 2.3.1 2.3.1 radiation control requirements. (CFR: 41. I 2 143.4.
45.9 / 45.10)
Importance Rating 2.6 3.0 Proposed Question: While o p e r a t i n g a t f u l l power, which one (1) o f t h e f o l l o w i n g would lower t h e dose r a t e f o r an O p e r a t o r d u r i n g a t w e n t y ( 2 0 )
m i n u t e walk down o f t h e H P C I Pump and T u r b i n e d u r i n g a P o s t Work Test?
a) Lowering t h e i n j e c t i o n r a t e .
ROlSRO b) O p e r a t i n g HPCI from i t s Torus s u c t i o n .
3 1/39 C) O p e r a t i n g RHR System A f o r Torus c o o l i n g .
d) S t a r t i n g a d d i t i o n a l Crescent Coolers.
Proposed Answer: a ) Lowering t h e i n j e c t i o n r a t e .
Explanation (Optional):
Technical Reference(s): OP-15, Step C.2.9, AP-07.03 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-23, EO-1.13.A, LPAP-28.03 (As available)
Question Source: Bank ##
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 12 55.43 4 Comments:
Page 48 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level SRO Tier # 1 Inadvertent Reactivity Addition / 1 Group # 2 Knowledge of the operational implications of the KIA # 295014 AKI .01 following concepts as they apply to INADVERTENT REACTIVITY ADDITION :
(CFR: 41.8 to 41.10)
Prompt critical Also 10CFR-55.43(b)(6)
Importance Rating 3.8 Proposed Question: Currently the plant is in the startup mode with control rods being withdrawn to bring the Reactor critical. The selected control rod is two (2) notches from the ECP's predicted criticality when a control rod drop occurs. The control rod blade that dropped went from position 4 to 48.
Assuming no further Operator action, which of the following barriers are in place to prevent this type of event what is a potential impact?
a) ST-20A, Rod Worth Minimizer Functional Test, the Reactor will heat up until a T turns power.
ROISRO b) ST-2OA, Rod Worth Minimizer Functional Test, the Reactor will go critical until full SCRAM on IRM HI-HI trip.
S40 c) ST-23B, Control Rod Coupling Integrity Test, the Reactor will heat up until a T turns power.
d) ST-23B, Control Rod Coupling Integrity Test, the Reactor will go critical until full L SCRAM on IRM HI-HI trip.
Proposed Answer: d) ST-23B, Control Rod Coupling Integrity Test, the Reactor will go critical until full SCRAM on IRM HI-HI trip.
Explanation (Optional):
Technical Reference@): T- F P, ,4 I 4 (' ttach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O3F, EO-1.13 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 6 Comments:
Page 49 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Page 50 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
i- Examination Outline Cross-reference: Level SRO Tier # 1 Loss of CRD Pumps / 1 Group # 2 Ability to detemrine andlor interpret the following KIA# 295022 AA2.02 as they apply to LOSS OF CRD PUMPS :(CFR:
41.10/43.5/45.13)
CRD system status Importance Rating 3.4 Proposed Question: While at full power, the following alarms and indications are received:
09-5-1-9, CRD CHARGING WTR PRESS LO, is in alarm 03Pl-302, CHG WTR PRESS, indicates 0 psig.
03PDI-303, DRV WTR DlFF PRESS, indicates 0 psid.
03FI-306, CLG WTR FLOW, indicates 0 gpm.
03FI-310, CRD FLOW CNTRL, indicates 1 gpm.
Which of the following is the cause and the appropriate mitigating procedure?
a) 03CRD-56, CRD Charging Water Supply Header IsolationValve, has been closed, ARP-09-51-9, CRD ChargingWTR Press Lo.
ROISRO b) 03FCV-I9A(B), in-service CRD Drivewater Flow Control Valve, has failed closed, AOP-69, Control Rod Drive Trouble.
S41 c) 03 MOV-22, CRD Cooling Water Pressure Control Valve, has been closed, ARP-09-5-1-9, CRD Charging WTR Press Lo.
d) 03P-I6A(B), in-service CRD Drive Water Pump has failed, AOP-69, Control Rod Drive Trouble.
Proposed Answer: d) 03P-I6A(B), in-service CRD Drive Water Pump has failed, AOP-69, Control Rod Drive Trouble.
Explanation (Optional):
Technical Referenceb): OP-25, AOP-69 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O3C, EO-1.12.B (As available)
Question Source: Bank # Fermi 2 INPO # 8900 (Modified to JAF)
Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam 41611998 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 51 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 10:27 AM
Page 52 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level SRO Tier # 1 Loss of CRD Pumps 1 1 Group # 2 AA2.02 CRD system status Importance Rating 3.4 Proposed Question: The plant is operating at full power. Several annunciators have been received in the last few minutes, including 3D5, CRD Charging H20 Pressure Low, and 3D10, CRD Accumulator Trouble. The following information is available:
E4f-R609, HPCl Pump Suction Pressure Indicator, indicates 0 psig.
Differential Pmsure, indicates 0 psig.
Reactor Differential pressurehq$xtor, indicates 0 psid.
Ind, indicates 1 gpm. .'
Which one oQhe fol+wing is the keason annunciaA 3 D T r e received?
a) Cll-FO34, Charging Hebder Isolation Valve, t)as been closed.
ROISRO b) The in-service CRD Rod control valve has failed closed.
S4 I C) C1152-FO Water PCV, has been closed.
Proposed Answer.
t--- Explanation (Optional):
Technical Reference(s): (Attach if not previously provided)
Learning Objective:
Question Source: Bank #
Modified Bank #
New Proposed references to be provided to applicants during examination:
(As available)
(Note changes or attach parent)
Question History: Last NRC Exam 41611998 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:
Page 54 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44 AM
Examination Outline Cross-reference: Level SRO Tier # 1 Secondary Containment High Differential Pressure / Group # 2 5
Knowledge of the process for performing a WA# 295035 2.3.9 containment purge. (CFR: 43.4 /45. IO)
Importance Rating 3.4 Proposed Question: While a t f u l l power, a u n i s o l a b l e l e a k h a s d e v e l o p e d i n t h e RWCU s u c t i o n p i p i n g i n t h e R e a c t o r B u i l d i n g . Secondary Containment p r e s s u r e h a s r i s e n due t o t h e l e a k i n t o t h e R e a c t o r B u i l d i n g b u t is still s l i g h t l y negative.
Which of t h e f o l l o w i n g w i l l minimize t h e r a d i a t i o n h a z a r d and c o n t r o l t h e Secondary Containment p r e s s u r e ?
a ) I n i t i a t e SGT System and manually i s o l a t e R e a c t o r B u i l d i n g Ventilation.
b ) Ensure t h a t SGT s t a r t s and R e a c t o r B u i l d i n g V e n t i l a t i o n ROlSRO i s o l a t e d when High AP S e t p o i n t i s r e a c h e d .
S42 c ) P l a c e a l l C r e s c e n t Area U n i t C o o l e r s i n service.
d ) O p e r a t e RWCU i n t h e Blowdown Mode t o t h e Main Condenser.
Proposed Answer: a ) I n i t i a t e SGT System and manually i s o l a t e R e a c t o r B u i l d i n g Ventilation.
Explanation (Optional):
Technical Reference(s): OP-2O,OP-51A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-OlB, EO-1.14.E (As available)
Question Source: Bank # JAFLOR20005214BOlC Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 12 55.43 4 Comments:
Page 53 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level SRO Tier # 1 Secondary Containment High SumplArea Water Group # 2 Level I 5 Ability to determine andlor interpret the following WA# 295036 EA2.01 as they apply to SECONDARY CONTAINMENT HIGH SUMPIAREA WATER LEVEL:
(CFR: 41.10 / 43.5 / 45.13)
Operability of components within the affected area.
Importance Rating 3.2 Proposed Question: Thirty (30) minutes after an earthquake, the following conditions exist:
RPV water level is 185 inches increasing RPV Pressure is 1000 psig increasing All control rods are at position 00, except for one (1) rod at position 22 One (1) foot of water is on the Crescent Floors due to a leaking Torus drain flange.
The MSlVs are Closed.
0 Reactor Scram has been reset.
0 Torus level is 10.75 feet and slowly lowering.
Regarding the above conditions, which of the following is True?
a) EOP-3 action is based upon ensuring Reactor remains shutdown without Boron Injection.
ROlSRO b) EOP-4 action is based upon preserving HPCl Injection capability.
v s43 c) EOP-5 action is based upon a loss of the Core Spray Hold Pumps.
d) There is NO EOP entry condition. Plant is controlled by AOP-1.
Proposed Answer: c) EOP-5 action is based upon a loss of the Core Spray Hold Pumps.
Explanation (Optional):
Technical Reference(s): EOP-5, OP-14 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: MIT-301.11, EO-1.07 (As available)
Question Source: Bank # Monticello 1 INPO # 15350 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 8/23/1999 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 54 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level SRO
- 6. Tier # 1 Secondary Containment High Sump/Area Water Group # 2 Level / 5 Ability to determine andlor interpret the following WA# 295036 EA2.01 as they apply t o SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL:
(CFR: 41.10 /#a$ / 45.1 3)
Operability of components within the affected area.
Importance Rating 3.2 Proposed Question: Thirty (30) minutes after an earthquake, the following conditions exist: Reactor water level is normal.
-All control rods are at position 00. CcmcenT f b r q
- One (1) foot of water is on the 7=omdkmdue to a leaking Torus drain flange.
- The MSlVs are Closed.
- Reactor Scram has been reset.
ROlSRO s43 Proposed Answer 1
Explanation (Optional) pG -
c) d) --
4
,I I, i s t f a e
. f3mec-q t e ~ r n ~ . O 2 Technical Reference(s): (Attach if not previously provided) b P - 5 GP-\d Proposed references to be provided to applicants during examination: flolc'r Learning Objective: M X T - 301 1 I/ I , 07 (As available)
Question Source: Bank # Monticello 1 INPO # 15350 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 8/23/1999 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis Y
10 CFR Part 55 Content: 55.41 IS 55.43 3 Comments:
- .-L/
Page 57 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 HPCl Group # 1 1 Knowledge of the physical connections andlor K/A # 206000 K1.07 K1.07 cause effect relationships between HIGH PRESSURE COOLANT INJECTlON SYSTEM and the following: (CFR: 41.2 to 41.9 I45.7 to 45.8)
D.C. power: BWR-2,3,4 Importance Rating 3.7 3.8 Proposed Question: Which of the following would render HPCl incapable of accomplishing its design purpose?
a) Loss of the 10600 Bus ROlSRO b) Loss of 125 VDC Bus B 32/44 c) Loss of Condensate Storage Tank level d) Loss of Secondary Containment integrity Proposed Answer: b) Loss of 125 VDC Bus B Explanation (Optional):
Technical Reference(s): OP-15 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-23, 0-1.10. E (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 78 55.43 Comments:
Page 55 of 120 NRC Written Examination Submittaldoc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 LPCS Group # 1 1 Knowledge of tagging and clearance procedures. WA # 209001 2.2.13 2.2.13 (CFR: 41.10 145.13)
Importance Rating 3.6 3.8 Proposed Question: Which one (1) of t h e f o l l o w i n g r e q u i r e m e n t s m u s t be met i n o r d e r t o p e r m i t o p e r a t i o n of Core Spray Hold Pump A under a S t r i p e d Tag?
a) Tag Holder f o r t h e CS Hold Pump m u s t be d e s i g n a t e d by p o s i t i o n such a s E l e c t r i c a l S u p e r v i s o r .
ROlSRO b) A p r o c e d u r e o r Work Request w i t h S t e p Text m u s t e x i s t t o p r o v i d e CS Hold Pump o p e r a t i o n guidance.
33/45 C) Tag Holder f o r t h e CS Hold Pump w i t h c o n c u r r e n c e from t h e F i e l d Support S u p e r v i s o r d i r e c t s CS Hold Pump o p e r a t i o n .
d) I f t h e CS Hold Pump i s o u t of i t s p r o t e c t e d p o s i t i o n f o r
> one (1) s h i f t , Tagout c o n t r o l m u s t s h i f t t o t h e Work Week Manager.
Proposed Answer: b ) A p r o c e d u r e o r Work Request w i t h S t e p Text m u s t e x i s t t o p r o v i d e CS Hold Pump o p e r a t i o n guidance.
Explanation (Optional):
Technical Reference(s): AP-12.01 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AP-44-10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Page 56 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 LPCS continued Group # 1 1 Knowledge of the effect that a loss or K/A# 209001 K3.02 K3.02 malfunction of the LOW PRESSURE CORE SPRAY SYSTEM will have on following: (CFR:
41.7 / 45.4)
ADS logic Importance Rating 3.8 3.9 Proposed Question: While at full power, a small break LOCA with HPCl inoperable has occurred.
ADS has initiated.
The only Low Pressure ECCS in service is Core Spray B which subsequently trips.
ADS valves will ?
a) Remain open.
ROISRO b) Close immediately.
34146 c) Close after a two (2) minute delay.
d) Remain open until RPV level reaches 159.5.
Proposed Answer: b) Close immediately.
Explanation (Optional):
Technical Reference(s): OP-68,OP-14 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-14, EO-1.09 . B (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 57 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
L Examination Outline Cross-reference: Level RO SRO Tier # 2 2 SLC Group # 1 1 Ability to direct personnel activities inside the KIA# 211000 2.1.9 2. I.9 control room.(CFR: 45.5 / 45.12 I45.13)
Importance Rating 2.5 4.0 Proposed Question: A failure to SCRAM occurs from power operation:
- Reactor power is 50%
- Main Turbine/ Generator is on line
- Torus temperature is 80°Fand steady
- Feedwater/CondensateSystem is maintaining RPV level
- The SNO-1 reports that APRM power is oscillating 30%.
Which action should be directed?
a) Emergency RPV Depressurization ROlSRO b) SLC System initiation 35/47 c) Main Turbine trip d) MSlV Closure Proposed Answer: b) SLC System initiation Explanation (Optional):
i, Technical Reference(s): EOP-3 (Attach if not previously provided)
Proposed references to be provided toapplicants during examination: None Learning Objective: EOP3LP, EO-1.07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam
~ ~
(Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Comments:
Page 58 of I 2 0 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level RO SRO L-'
Tier # 2 2 RPS Group # I I Knowledge of the operational implicationsof the K/A # 212000 K5.02 K5.02 following concepts as they apply to REACTOR PROTECTION SYSTEM :(CFR: 41.5 I45.3)
Specific logic arrangements Importance Rating 3.3 3.4 Proposed Question: While a t 2 0 % power, what p o s s i b l e Reactor P r o t e c t i o n System ( R P S ) r e s p o n s e ( s ) can occur i f t h e Inboard and Outboard M S I V ' s on any two ( 2 ) Main S t e a m L i n e s a r e closed?
a) No r e s p o n s e f u l l SCRAM RO/SRO b) -
N o r e s p o n s e OR h a l f SCRAM 36/48 c) H a l f SCRAM always d) F u l l SCRAM always Proposed Answer: -
b ) N o r e s p o n s e OR h a l f SCRAM Explanation (Optional):
Technical Reference(s): ST-11, OP-I (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-29, EO 1.09.f, 1 . 1 3 . C (Asavailable)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Page 59 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 IO27 AM
Examination Outline cross-reference: Level RO SRO Tier # 2 2 IRM Group # 1 1 Ability to monitor automatic operations of the WA # 215003 A3.04 A3.04 INTERMEDIATE RANGE MONITOR (IRIIII) SYSTEM including: (CFR: 41.7 / 45.7)
Control rod block status Importance Rating 3.5 3.5 Proposed Question: An IRM HI Flux Control Rod Block is automatically bypassed when ?
a) The Reactor Mode Switch is placed in RUN.
ROlSRO b) The IRM is on Range 1.
37/49 c) The IRM's companion APRM is downscale.
d) The SRM's are fully inserted.
Proposed Answer: a) The Reactor Mode Switch is placed in RUN.
Explanation (Optional):
Technical Reference(s): OP-16 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-078, EO-1.05.C.2 (As available)
.- Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 60 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Source Range Monitor Group # I 1 Ability to monitor automatic operations of the KIA # 215004 A3.04 A3.04 SOURCE RANGE MONITOR (SRM) SYSTEM including: (CFR: 41.7 /45.7)
Control rod block status Importance Rating 3.6 3.6 Proposed Question: The following plant conditions exist:
Reactor Mode Switch is in STARTUPIHOT STBY.
Intermediate Range Monitors (IRM's) all on Range 3.
Source Range Monitor (SRM) A is reading 0.5 cps SRM's B and C are reading 8.3 x I O 4 SRM D mode switch is in STANDBY A rod block signal has been generated.
Which one of the following has caused the rod block?
a) SRM Inoperable ROlSRO b) SRM Count Circuit 38/50 c) SRM Downscale d) SRM Upscale Proposed Answer: a) SRM Inoperable Explanation (Optional):
Technical Reference(s): OP-16 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O7B, EO 1.05.b.1, EO 1 . 0 5 . C , EO 1 . 1 4 . C (As available)
Question Source: Bank # Perry 1 INPO# 21837 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/1/2001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 61 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
c Examination Outline Cross-reference:
Source Range Monitor Level Tier #
Group #
RO 2
1 SRO 2
1 Ability to monitor automatic operations of the WA # 215004 A3.04 A3.04 SOURCE RANGE MONITOR (SRM) SYSTEM including: (CFR: 41.7 / 45.7)
Control rod Mock status Importance Rating 3.6 , 3.6 Proposed Question: The following plant conditions exist: WT 9 ~ Y3, Reactor Mode Switch is in STARTUPISWWW cai Intermediate Range Monitors (IRM) pr;c*r;B;-ErendG are on Range 3;ehth+Wk Source Range Monitor (SRM) A is reading 0.5 cpsSRMs B and C are reading 8.3 x 10E4.SRM D mode switch is in STANDBY.
A rod block signal has been generated.
Which one of the following has caused the rod block?
a) SRM Inoperable ,
CnLni- C ~ C C A L ~ .,
RO/SRO b) SRM-38/50 c) SRM Downscale a) SRM Upscale Proposed Answer: a) SRM Inoperable Explanation (Optional):
Technical Referencels): OP-16 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: JA Learning Objective: SDLP-O7B, EO 1.05.a.6.b (As available)
EO 1.05.a.3.i. EO 1.14.c Question Source: Bank # Pe-rry 1 I N P O # 21837 [M4)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/ I ROO1 (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level:
10 CFR Part 55 Content:
Memory or Fundamental Knowledge Comprehension or Analysis 55.41 1 55.43 Comments:
Page 65 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44AM
W Examination Outline Cross-reference: Level RO SRO Tier # 2 2 APRM I LPRM Group # 1 1 Ability to monitor automatic operations of the KIA # 215005 A3.02 A3.02 AVERAGE POWER RANGE MONITOWLOCAL POWER RANGE MONITOR SYSTEM including:
(CFR: 41.7 145.7)
Full core display Importance Rating 3.5 3.5 Proposed Question: A r e a c t o r s t a r t u p i s b e i n g performed f o l l o w i n g a p l a n n e d o u t a g e .
A n n u n c i a t o r , 09-5-2-33, LPRM Downscale, c l e a r s .
The SNO c a n c o n f i r m t h a t t h i s i s e x p e c t e d and c o r r e c t b y verifying?
a) A l l APRM Downscale a l a r m s are c l e a r .
ROISRO b) A l l F u l l Core D i s p l a y LPRM downscale l i g h t s a r e o u t .
39/51 C) A l l IRM Range S w i t c h e s are above Range 1.
d) R e a c t o r Mode Switch i s i n RUN.
Proposed Answer: b ) A l l F u l l Core D i s p l a y LPRM downscale l i g h t s a r e o u t .
Explanation (Optional):
Technical Reference(s): OP-16, ARP- 09-5-2-33 (Attach if not previously provided)
~
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O7C, EO-1.12.D,1.05.C.l.B (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 62 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level SRO Tier # 2 sLC Group # 1 Knowledge of the effect that a loss or WA# 211000 K3.02 malfunction of the STANDBY UQUID CONTROL SYSTEM will have on following: (CFR: 41.7 145.4)
Core spray line break detection system: Plant-Specific Link to 10CFR-55.43(b)(2)
Importance Rating 3.2 Proposed Question: Engineering has just informed the Shift Manager of an industry event where flow induced vibration has breached the integrity of the in-vessel section of the Standby Liquid Control System piping.
In addition to SLC, which one (1) of the following system Technical Specification operabilities could be threatened by such an occurrence?
a) CRD System ROISRO b) LPCl System S52 c) Core Spray System d) ADS System Proposed Answer: c) Core Spray System Explanation (Optional):
Technical Reference(s): OP-14,0P-lI, ARP-09-03-1-1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learnina Obiective: SDLP-14, EO-1.07.C, 1.16, 1.05.A.13 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10I95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 2 Comments:
Page 63 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
-- Examination Outline Cross-reference: Level RO 2
SRO 2
Tier #
RCIC Group # 1 1 Knowledge of the effect that a loss or K/A # 217000 K3.01 K3.01 malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following:
(CFR: 41.7145.4)
K3.01 Reactor water level Importance Rating 3.7 3.7 Proposed Question: The Station has lost all AC electrical power. RPV level is being controlled using RCIC system operation.
Which statement below describes the effect that the loss of 'A' Station Battery will have on level control?
a) HPCI system will have to be used to control RPV level.
b) An Emergency Depressurization will be required to enable ROISRO Low Pressure Injection.
40/53 C) All injection sources will be lost. Emergency Depressurize when RPV level drops to TAF.
d) RCIC will continue to operate but must be controlled 1oca11y .
Proposed Answer: a) HPCI system will have to be used to control RPV level.
Explanation (Optional):
Technical Reference(s): AOP-45, AOP-49 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP 13, EO 1.09.A, 1.lO.B (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 64 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RCIC Group # 1 1 Ability to monitor automatic operations of the WA # 217000 A3.06 A3.06 REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) including: (CFR: 41.7 / 45.7)
Lights and alarms Importance Rating 3.5 3.4 Proposed Question: During reactor power operation with RCIC in the Standby Mode a RCIC PMP SUCT PRESS HI alarm (09-4-1-33) is received. While investigating at Panel 09-4 the operator observes the following:
RCIC Pump Minimum Flow Valve, 13MOV-27, opens.
The RCIC PMP SUCT PRESS HI alarm clears.
RCIC Pump Minimum Flow Valve, 13MOV-27, closes.
Based on these indications which one of the following actions is warranted:
a) Investigate for abnormally high RCIC pump casing and discharge piping temperatures.
b) Investigate for an erroneous RCIC automatic initiation RO/SRO signal.
41/54 c) Investigate the malfunction in valve 13MOV-27 opening logic.
d) Investigate the malfunction in valve 13MOV-27 closing logic.
Proposed Answer: a) Investigate for abnormally high RCIC pump casing and discharge piping temperatures.
Explanation (Optional):
Technical Reference@): OP-19, ARP 09-4-1-33 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-13, EO 1.05.a.10 (Asavailable)
Question Source: Bank # JAF LOR # 21701005BOlC Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 65 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Page 66 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
L.'
Examination 0utline Cross-reference: Level SRO Tier # 2 APRM / LPRM Group # 1 KIA # 215005 2&34 2.1.17 RANDOMLY =-SELECTED Ability to make accurate / clear and concise verbal reports.
(CFR: 45.12 / 45.13)
Importance Rating &3 3.6 Proposed Question: A short time after a recirculation pump trip, the SNO-1 reports that the APRM upscale alarm and rod block are in solid and that the LPRM upscale alarm is coming in and quickly clearing every few seconds. He is unable to identify the problem LPRM.
Your response to him is:
a) Manually scram the reactor, perform immediate actions of AOP-1, REACTOR SCRAM.
ROlSRO b) When identified, bypass the LPRM per OP-16, NEUTRON MONITORING.
s55 c) Identify and bypass the affected APRM per OP-16, NEUTRON MONITORING.
d) Determine if LPRM upscale alarm is valid by observing 09-14 LPRM upscale lamps.
Proposed Answer: a) Manually scram the reactor, perform AOP-1 Immediate Actions.
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 #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5, I O 55.43 5 Comments:
Page 67 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level SRO Tier ## 2 RClC Group # 1 Knowledge of electrical power supplies to the KIA # 217000 K2.04 following: (CFR: 41.7)
Gland seal compressor (vacuum pump)
Link to 10CFR-55.43(b)(2)
Importance Rating 2.6 Proposed Question: The Plant is operating at 100% power steady state with HPCl tagged out for maintenance, day three (3) of the LCO. At 02:OO am this morning, the feeder breaker to BMCC-1 opened on over-current and will not reset.
Which of the following is appropriate for this situation?
a) HPCl is the only system affected, original LCO applies.
ROlSRO b) A more restrictive LCO applies as both HPCl and ADS are affected.
S56 c) A more restrictive LCO applies as both HPCI and RClC are affected.
d) A more restrictive LCO applies due to loss of Primary Containment Isolation.
Proposed Answer: c) A more restrictive LCO applies as both HPCl and RClC are affected.
Explanation (Optional):
Technical Reference(s): OP-43,0P-19 (Attach if not previously provided) d Proposed references to be provided to applicants during examination: FE-1AH Learning Objective: SDLP-13, EO-1.04.A (As available)
Question Source: Bank ##
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 2 Comments:
Page 68 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
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Examination Outline Cross-reference: Level RO SRO Tier # 2 2 ADS Group # 1 1 Knowledge of the physical connections andlor K/A # 218000 K1.06 K1.06 cause effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following:
(CFR: 41.2 to 41.9 I45.7 to 45.8)
Safetylrelief valves Importance Rating 3.9 3.9 Proposed Question: The Plant is at 70% power; the Control Room receives annunciator, 09-4-2-37, SRV Electric Lift Initiated or Bypassed.
All SRV green lights are on.
Which of the following describes how this impacts the operation of the ADS Valve(s)?
a) Will operate normally on hydraulic overpressure.
RO/SRO b) Only operate on High RPV pressure setpoint.
42157 c) Only operate manually from Panel 02ADS-071.
d) Only operate manually from 09-4 Panel.
Proposed Answer: a) Will operate normally on hydraulic overpressure.
Explanation (Optional):
Technical Reference(s): ARP-09-4-2-37, OP-68 (Attach if not previously provided)
GE DWG 791E453 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-29, EO-1.05.A.4 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 69 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 PClSlNucIear Steam Supply Shutoff Group # 1 1 Ability to (a) predict the impacts of the following K/A # 223002 A2.04 A2.04 on the PRIMARY CONTAINMENT ISOLATION SYSTEMlNUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnomal conditions or operations: (CFR: 41.5 145.6)
Process radiation monitoringsystem failures Importance Rating 2.9 3.2 Proposed Question: While operating at normal full power with the A Standby Gas Treatment System out of service, the B Reactor Building Ventilation Exhaust Radiation Monitor fails downscale.
Which of the following describes expected Operator action(s)?
a) Using OP-51A, verify Reactor Building Ventilation isolation.
ROISRO b) Using OP-20, verify automatic start of the B Standby Gas Treatment System.
43/58 c) Using AOP-15, reset the isolation and restart Reactor Building Ventilation.
d) Using OP-20, manually start up the B Standby Gas Treatment System.
Proposed Answer: d) Using OP-20, manually start up the B Standby Gas Treatment System.
Explanation (Optional):
Technical Reference(s): JAF LER-98-001, OP-20, ITS-3.3.6.2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-015, EO-1.05.c, 1.10.a,1.14.b (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since I0195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 5 Comments:
c Page 70 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level SRO v
Tier # 2 PClSlNuclear Steam Supply Shutoff Group # 1 Knowledge of electrical power supplies to the WA# 223002 K2.01 following: (CFR: 41.7)
Logic power supplies Link to 10CFR-55.43(b)(5)
Importance Rating 2.7 Proposed Question: I n w h i c h of t h e f o l l o w i n g complete system l o s s events, would you expect t o f i n d a t l e a s t one MSIV i n each Main Steam L i n e closed?
a) AOP-59, LOSS of RPS B u s A P o w e r OR AOP-45, LOSS of DC P o w e r S y s t e m b) AOP-18, LOSS of 1 0 5 0 0 B u s RO/SRO AND AOP-45, LOSS of DC P o w e r S y s t e m s59 C) AOP-21, LOSS O f UPS AND AOP-46, LOSS of DC P o w e r S y s t e m d) AOP-19, LOSS Of 10600 Bus OR AOP-46, LOSS of DC P o w e r System Proposed Answer: b ) AOP-18, LOSS of 1 0 5 0 0 B u s AND AOP-45, Loss of DC P o w e r S y s t e m A Explanation (Optional):
Technical Reference(s): AOP-18, AOP-19, AOP-21 (Attach if not previously provided)
AOP-45, AOP-46 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-29, EO 1 . 0 4 . a , EO 1 . 0 5 . a . l . c (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
~
New New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 5 Cornments:
Page 71 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Page 72 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 PClSlNuclear Steam Supply Shutoff Group # 1 1 Knowledge of conditions and limitations in the WA # 223002 2.1.10 2.1.10 facility license.(CFR: 43.1 I 45.13)
Importance Rating 2.7 3.9 Proposed Question: During r o u t i n e f u l l power o p e r a t i o n , t h e SNO i d e n t i f i e s t h a t t h e b r e a k e r f o r 20MOV-94, Drywell Equip. Sump I s o l a t i o n Valve, h a s opened and t h e valve p o s i t i o n i s open.
Which s t a t e m e n t below d e s c r i b e s t h e r e q u i r e d p l a n t a c t i o n s ?
a) Valve may remain open as l o n g as t h e s y s t e m i s n o t operating.
RO/SRO b) An NPO must b e d i s p a t c h e d immediately t o manually c l o s e t h e valve.
44/60 C) A v a l v e i n t h a t l i n e must b e d e - a c t i v a t e d closed w i t h i n four ( 4 ) hours.
d) Valve may remain open under a d m i n i s t r a t i v e c o n t r o l indefinitely.
Proposed Answer: c) A valve i n t h a t l i n e must b e d e - a c t i v a t e d c l o s e d w i t h i n f o u r
( 4 ) hours.
Explanation (Optional):
L Technical Reference@): ST-IC, ITS- 3.6.1.3 (Attach if not previously provided)
TRM Appendix A Proposed references to be provided to applicants during examination:
Learning Objective: SDLP-16C, EO 1.13.a (As available)
Question Source: Bank # JAF L O R # 2 2 3 0 2 0 0 1 B 0 1 S Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 9 55.43 1 Comments:
Page 73 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level SRO Tier # 2 Main and Reheat Steam Group # 2 Ability to (a) predict the impacts of the following K/A # 239001 A2.01 on the MAIN AND REHEAT STEAM SYSTEM ; and (b) based on those predictions, w e procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
Malfunction of reactor turbine pressure regulating system Link to 10CFR-55.43(b)(5)
Importance Rating 3.9 Proposed Question: The r e a c t o r i s o p e r a t i n g a t 25% power w i t h r e c i r c u l a t i o n flow a t minimum.
I f a t u r b i n e t r i p o c c u r s and t h e bypass v a l v e s f a i l t o open, which of t h e f o l l o w i n g would be t h e a p p r o p r i a t e p r o c e d u r e ( s ) t o respond t o t h e e v e n t ?
a) AOP-1, Reactor Scram, AOP-6, M a l f u n c t i o n of EHC Pressure Regulator.
RO/SRO b) AOP-2, Main Turbine T r i p Without Scram, AND AOP-6, M a l f u n c t i o n of EHC P r e s s u r e R e g u l a t o r .
S61 C) EOP-2, RPV C o n t r o l , AND AOP-2, Main T u r b i n e T r i p Without Scram.
d) AOP-1, R e a c t o r Scram, AND EOP-2, RPV C o n t r o l .
Proposed Answer: d) AOP-1, Reactor Scram, AND EOP-2, RPV C o n t r o l .
Explanation (Optional): High pressure entry into EOP-2 and AOP-1 Technical Reference(s): AOP-1, EOP-2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-05, EO 1.07.a.6, 7, & 10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5, 10 55.43 5 Comments:
Page 74 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Page 75 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
L Examination Outline Cross-reference: Level RO SRO Tier # 2 2 SRVs Group # I 1 Knowledge of the physical connections andlor WA # 239002 K1.01 K1.O1 cause effect relationships between RELIEFSAFETY VALVES and the following:
(CFR: 41.2 to 41.9 I45.7 to 45.8)
Nuclear boiler Importance Rating 3.8 3.9 Proposed Question: After steady state conditions are achieved, which of the below is confirmation of an inadvertent SRV full opening while in normal full power operation?
a) Reactor Power at 100% and Main Generator Output at 875 MWe.
ROlSRO b) RPV Water level at 207 inches and Level Set at 202 inches.
45/62 c) Feed flow at 11 x I O 6 Ibm/hr and Seam flow at I O x I O 6 Ibmhr.
d) Torus water temperature trending down slowly with Torus cooling in service.
Proposed Answer: c) Feed flow at 11 x I O 6 lbmlhr and Steam flow at 10 x I O 6 Ibm/hr.
Explanation (Optional):
Technical Reference(s): AOP-36 (Attach if not previously provided)
~
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP, EO-l.02,2 .27 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam
~~
(Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 8 55.43 Comments:
Page 76 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
'v Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Reactor Water Level Control Group # 1 1 Knowledge of the physical connections andlor WA # 259002 K1.13 K1.13 cause effect relationships between REACTOR WATER LWEL CONTROL SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8)
Condensate system Importance Rating 3.2 3.2 Proposed Question: The p l a n t i s o p e r a t i n g a t 1 0 0 % power w i t h a normal Feed and Condensate a l i g n m e n t . There a r e no s y s t e m s o r components i n o p e r a b l e . The A Condensate Pump t r i p s due t o an e l e c t r i c a l fault.
Which one of t h e f o l l o w i n g i s t h e e x p e c t e d r e s u l t o f t h i s t r i p ?
a) The o p e r a t i n g pumps assume t h e a d d i t i o n a l l o a d and t h e RFPs a r e n o t a f f e c t e d . A normal power r e d u c t i o n i s required.
b) T h e A Condensate B o o s t e r Pump t r i p s on i n t e r l o c k , b u t t h e RO/SRO RFPs a r e n o t a f f e c t e d . A normal power r e d u c t i o n i s required.
46/63 C) The A Condensate B o o s t e r Pump t r i p s on i n t e r l o c k c a u s i n g RFPs t o t r i p on low s u c t i o n p r e s s u r e . A manual SCRAM i s required.
d) Condensate B o o s t e r Pump s u c t i o n p r e s s u r e d e c r e a s e s c a u s i n g RFPs t o t r i p on low s u c t i o n p r e s s u r e . A manual SCRAM i s required.
Proposed Answer: a ) The o p e r a t i n g pumps assume t h e a d d i t i o n a l l o a d and t h e RFPs a r e n o t a f f e c t e d . A normal power r e d u c t i o n i s r e q u i r e d .
Explanation (Optional):
Technical Reference(s): AOP-42,0P-3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-33, EO 1.05.b.2 6L 1 . 1 4 . C (As available)
Question Source: Bank # JAF LOR# 25601012B02C Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 77 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
.. . I .
' J Page 78 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 SGTS Group # 1 1 Knowledge of the effect that a loss or WA # 261000 K3.05 K3.05 malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: (CFR: 41.7 145.6)
Secondary containment radiation/ contamination levels Importance Rating 3.2 3.5 Proposed Question: A Station Blackout has occurred resulting in a full Reactor SCRAM with all rods in. HPCl is operating to maintain RPV water level and pressure control.
As a result of HPCI operation:
a) The AStation Battery is expected to rapidly deplete.
RO/SRO b) The Crescent Area contamination levels are expected to rise.
47/64 c) The HPCJ Turbine MUST be manually tripped on RPV high water level.
d) RPV water level is expected to slowly drop until injection overcomes decay heat losses.
Proposed Answer: b) The Crescent Area contamination levels are expected to rise.
Explanation (Optional):
Technical Reference@): AOP-49, AOP-45, AOP-46 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-018, EO-1.09.A, F, LPAOP-49, EO-1.04 (As available)
Question Source: Bank #
Modified Bank ## (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 79 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 AC Electrical Distribution Group # 1 1 Knowledge of electrical power supplies to the KIA# 262001 K2.01 K2.01 following: (CFR: 41.7)
Off-site sources of power Importance Rating 3.3 3.6 Proposed Question: The Plant is in day 8 of a refuel outage. A full core off load has just been completed.
Niagara Mohawk called to report that the Lake Road 13.2 KV line is being taken out of service immediately.
Which of the below is a priority Control Room action?
a) Dispatch an NPO to transfer DHR power to the Diesel Generator.
RO/SRO b) Transfer in house electrical distribution from Normal to Reserve Station Service.
48/65 c) Dispatch an NPO to align 115 KV control power to the alternate source.
d) Implement alternate temperature monitoring of Interim Spent Fuel Storage.
Proposed Answer: a) Dispatch an NPO to transfer DHR power to the Diesel Generator.
Explanation (Optional):
Technical Reference(s): AOP-71 (Attach if not previously provided)
L Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP, EO-1.03, SDLP-71S, EO-1.09, SDLP-32, EO-1.04,1.10.A (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 80 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination 0utline Cross-reference: Level RO SRO Tier # 2 2 UPS (ACIDC) Group # 1 1 Knowledge of tagging and clearance procedures. WA # 262002 2.2.13 2.2.13 (CFR: 41.10 145.13)
Importance Rating 3.6 3.8 Proposed Question: Operators are tagging out the UPS M/G set for bearing replacement.
Worker protection is assured by:
a) A Maintenance PTR.
ROlSRO b) A Special Condition PTR.
49166 c) A Striped PTR.
d) A Hold PTR.
Proposed Answer: b) IndependentVerification.
Explanation (0ptional):
Technical Reference(s): OP-46B, AP-12.01 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAP-EO 4 4 . 0 3 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since I0195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Page 81 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level SRO Tier # 2 Offgas Group # 2 Ability to predict andlor monitor changes in K/A# 271000 A I .06 parameters associated with operating the OFFGAS SYSTEM controls including:
(CFR: 41.5 / 45.5)
Filter differential pressure Link to IOCFR-55.43 Importance Rating 2.5 Proposed Question: The plant is operating at rated power when the following indications simultaneously occur:
09-6-1-23, OFF GAS LINE FILTER DlFF PRESS HI in and clear.
09-6 Off Gas Flow Recorder (38FR-101) drops from 120 to 70 SCFM.
0 EPIC OFFGASRAD alarm.
0 09-10 Off Gas Radiation Monitors both reading 150-175 mr/hr.
Which of the following describes the plant event and the appropriate mitigating procedure?
a) Off Gas line blockage will cause a loss of condenser vacuum. AOP-31, LOSS OF CONDENSER VACUUM.
ROlSRO b) A hydrogen fire has ignited in piping downstream of the SJAEs. AOP-5, COMBUSTION IN SJAE AFTERCONDENSER.
S67 c) Fuel failure has resulted in a large radioactive gas release from the RPV. AOP-3, HIGH ACTIVITY IN REACTOR COOLANT OR OFFGAS d) An explosion has breached the SJAE discharge piping. AOP-4, EXPLOSION IN AIR EJECTOR DISCHARGE PIPING Proposed Answer: b) A hydrogen fire has ignited in piping downstream of the SJAE's. AOP-5, COMBUSTION IN SJAE AFTERCONDENSER.
Explanation (Optional):
Technical Reference(s): ARP-0 9- 6-1-23/AOP-4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAOP-EO-1.01 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5,lO Page 82 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
55.43 5 L
Comments:
Page 83 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
W Examination Outline Cross-reference: Level RO SRO Tier # 2 2 DC Electrical Distribution Group # 1 1 Ability to (a) predict the impacts of the following K/A # 263000 A2.01 A2.01 on the D.C. ELECTRICAL DISTRIBUTION :and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
(CFR: 41.5 / 45.6)
Grounds Importance Rating 2.8 3.2 Proposed Question: The plant is operating at 100% with operators attempting to locate a ground on the "A" station battery. The next breaker to be opened is the supply for 10700 BKR Control Power (71DCA3 CrM 24).
How will the opening of this circuit affect the 4KV breakers on the 10700 bus AND which procedures will provide guidance?
a) The breakers will immediately trip, AOP-20, Loss of 10700 Bus, and AOP-22, DC Power System 'A' Ground Isolation.
b) The breakers can be tripped mechanically only, OP-46A, 4160 V & 600 V ROlSRO Normal AC Power Distribution, and AOP-22, DC Power System 'A' Ground Isolation.
50/68 c) Electrical protective trips will operate normally, OP-43A, 125 VDC Power System, and OP-46A, 4160 V & 600 V Normal AC Power Distribution.
d) Breaker position indication lights (red and green) will continue to indicate breaker positions, OP-43A, 125 VDC Power System, and AOP-23, DC Power System 'B' Ground Isolation.
Proposed Answer: b) The breakers can be tripped mechanically only, OP-46A, 4160 V & 600 V Normal AC Power Distribution, and AOP-22, DC Power System 'A' Ground Isolation.
Explanation (Optional):
Technical Reference(s): A 0 P-22 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-71B, EO-1.09.C. 1 8 (As available)
Question Source: Bank # JAF LOR# 2 0 0 0 4 2 1 1 ~ 0 1 ~
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 84 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Page 85 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 EDGs Group # 1 1 Knowledge of the effect that a loss or KIA# 264000 K6.02 K6.02 malfunction of the following will have on the EMERGENCY GENERATORS (DIESEUJET) :
(CFR: 41.7 / 45.7)
Fuel oil pumps Importance Rating 3.6 3.6 Proposed Question: A LOCA and LOOP exists. Off-site power is not expected to be returned to service for two days. All EDG equipment is operable with the exception that Fuel Oil Transfer Pumps 93P1-A1 & 2 have just tripped and cannot be started.
Based upon these events, select the expected plant response assuming Operator action.
a) EDG's "A", "B", "C",& "D" will continue to run until off- site power is restored.
b) EDG "A" will trip immediately, EDG's "B","C", & "D" will continue to run until off-RO/SRO site power is restored.
51I69 c) EDG 'A" will continue to run for up to three (3) hours then trip, EDG's "B","C", &
"D" will continue to run until off- site power is restored.
d) EDG "A" will continue to run at reduced capacity until off- site power is restored, EDG's "B",'C", & "D" will continue to run until off- site power is restored.
Proposed Answer: c) EDG "A" will continue to run for up to three (3) hours then trip, EDG's "B","C", & "D" will continue to run until off- site power is restored.
Explanation (Optional):
Technical Reference(s): OP-22 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-93, EO-1.05.A.4, l.lO.G (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 86 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
L Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of the process f o r making changes in Group ##
procedures a s described in the safety analysis report. (CFR: 43.3 / 45.13)
KIA # 2.2.6 Importance Rating 3.3 Proposed Question: Who must approve a temporary change to a Technical Specification related procedure.
a) The procedure RPO and an Operations QTR ROlSRO b) A plant management QTR and a SRO license QTR S70 c) The General Manager-PlantOperations and a plant management QTR.
d) A plant management QTR and any operations licensed QTR.
Proposed Answer: b) A plant management QTR and a SRO license QTR Explanation (Optional):
Technical Reference@): AP-2.04 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAP, EO-4.05 (As available)
Question Source: Bank #
Modified Bank # NEW (Note changes or attach parent)
New Question History: Last NRC Exam 6/5/2000 (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 3 Comments:
Page 87 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Instrument Air Group # 1 1 Knowledge of the connections and I or cause WA# 300000 K1.05 K1.05 effect relationships between INSTRUMENT AIR SYSTEM and the following:
(CFR: 41.2 to 41.9 I45.7 to 45.8)
Main Steam Isolation Valve air Importance Rating 3.1 3.2 Proposed Question: The plant is operating at 75% reactor power.
The SNO-1 depresses the TEST pushbutton for 29AOV-868, B OUTBOARD MSIV.
Which one of the following describes the response of 29AOV-86B?
a) Instrument Nitrogen bleeds off the bottom portion of the MSlV air cylinder and the top portion of the MSlV air cylinder is pressurized to stroke the MSIV closed in 3-5 seconds.
ROlSRO b) Instrument Air bleeds off the bottom portion of the MSlV air cylinder causing the MSlV to slowly close.
52/71 c) Instrument Nitrogen bleeds off the bottom portion of the MSlV air cylinder causing the MSlV to slowly close.
d) Instrument Air bleeds off the bottom portion of the MSIV air cylinder and the top portion of the MSlV air cylinder is pressurized to stroke the MSlV closed in 3-5 seconds.
Proposed Answer: b) Instrument Air bleeds off the bottom portion of the MSlV air cylinder causing the MSlV to slowly close.
Explanation (Optional):
Technical Reference@): ST-1I, OP-I (Attach if not previously provided)
~~~ ~
Proposed references to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank # Perry 1 INPO # 21861 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/1/2001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 88 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
V Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Instrument Air Group # 1 1 Knowledge of the connections and I or cause KIA# 300000 K1.05 K1.05 effect relationships between INSTRUMENT AIR SYSTEM and the following:
(CFR: 41.2 to 41.9 I45.7 to 45.8)
Main Steam Isolation Valve air - ____ WitanceRatirrg - - -- 3-1-._____ 3.2 Proposed Question: The plant is operating at 75% reactor power.
.-- . . . * . . The Control Room 2q-Y- 9bX '= MClV W . j y b i - -1Which one of the following describes e response- fo .z?law-g63 Nt2-a-a) -1nstrumentM bleeds off the bottom portion of the MSIV air cylinder and the top portion of the MSlV air cylinder is pressurized to stroke the MSlV closed in 3-5 seconds.
b) Instrument Air bleeds off the bottom portion of the MSIV air cylinder causing the RO/SRO MSIV to slowly close.
N%u-52/71 c) S&t+&M& Instrument 14iF bleeds off the bottom portion of the MSlV air cylinder causing the MSlV to slowly close.
d) Instrument Air bleeds off the bottom portion of the MSlV air cylinder and the top portion of the MSlV air cylinder is pressurized to stroke the MSlV closed in 3-5 seconds.
Proposed Answer: b) Instrument Air bleeds off the bottom portion of the MSlV air cylinder causing the MSlV to slowly close.
Explanation (Optional).
(Attach if not previously provided)
Technical Reference(s): c 1%; & 7 Proposed references to be provided to applicants during examination: No EAL Learning Objective: SDLf - z9 1106Q. ~ 1. a, (As available)
Question Source: Bank # Perry 1 INPO # 21861 (p o Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/1/2001 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis K 10 CFR Part 55 Content: 55.41 k ?
55.43 Comments:
Page 89 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44 AM
'V Examination Outline Cross-reference: Level SRO Tier # 3 Knowledgeof the effects of alterations on core Group #
configuration. (CFR: 43.6)
KIA # 2.2.32 Importance Rating 3.3 Proposed Question: The purpose of core spiral fuel un-loading is which one of the following?
a) it minimizes the possibility of flow induced vibration of nuclear instrumentation b) it precludes the formation of moderator filled cavities RO/SRO surrounded on all sides by fuel S72 c) it prevents SRM count rates from spiking when fuel is being off-loaded d) it enables the completion of a full core off-load in less time Proposed Answer: b) it precludes the formation of moderator filled cavities surrounded on all sides by fuel Explanation (Optional):
Technical Reference(s): ITS - Bases, RAP- 7.1.24, (Attach if not previously provided)
RAP-7.1.04B Section 5.10.3 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O7B, EO 1.13.E, 1.17.G (As available)
Question Source: Bank # JAF LOR # 1332 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 6 Cornments:
Page 89 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
L Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Component Cooling Water Group # I 1 Knowledge of electrical power supplies to the KIA# 400000 K2.01 K2.01 following: (CFR: 41.7)
CCW pumps Importance Rating 2.9 3.0 Proposed Question: An electrical transient has occurred and Switchgear L-14 is de-energized.
Which of the following equipment would be lost due to the degraded electrical source?
a) 12P-IA, RWCU Pump A ROlSRO b) 15P-2B, RBCLC Pump B 53/73 c) 11P-2B, SLC Pump B d) 46P-1B, Service Water Pump B Proposed Answer: b) 15P-2B, RBCLC Pump B Explanation (Optional):
Technical Reference(s): 0P-46 (Attach if not previously provided) 0P-40 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-15, EO-1.036 (As available)
L Question Source: Bank #
Modified Bank #
~
(Notechanges or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 90 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of radiation exposure limits and Group #
contamination control / including permissible levels in excess of those authorized.
(CFR: 43.4 I45.10)
KIA # 2.3.4 Importance Rating 3.1 Proposed Question: Authorization to receive radiological exposures in excess of 10CFR20 limits is the responsibility of the a) Radiation Protection Manager ROlSRO b) Emergency Director s74 c) TSC Manager d) General Manager- Plant Operations Proposed Answer: b) Emergency Director Explanation (Optional):
Technical Reference(s): EAP-15, AP-07.05 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None L Learning Objective: EP-12.5.3, EO-I .20.B (As available)
Question Source: Bank
- . # LaSalle IINPO # 19298 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 11/20/2000 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4 Comments:
Page 91 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of radiation exposure limits and Group #
contamination control / including permissible levels in excess of those authorized.
(CFR: #id / 45. I O )
KIA # 2.3.4 Importance Rating 3.1 Proposed Question:
Authorization to receive radiological exposures in excess of 10CFR20 limits is the responsibility of the a) Recovery Manager RO/SRO b) r-6 9-x y
c s74 r I~C~?MAO#-GEL 7s2 d) Radiation
.- s - a f d Proposed Answer: b) r r
Explanation (Optional): b*r_e&,, L.? -> =-A -
Technical Reference(s): GAP- (g (Attach if not previously provided)
Proposed references to be provided to applicants during examination: H 13c Learning Objective: F 1 2 . 5 . 3 1\20,b (As available)
Question Source: Bank # LaSalleIlNPO#19298 ( peL 4. g-&F\
b Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 11/20/2000 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 Comments:
Page 92 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44AM
Examination Outline Cross-reference: Level RO SRO Tier ## 2 2 CRD Hydraulic Group # 2 2 Knowledge of electrical power supplies to the KIA# 201001 K2.03 K2.03 following: (CFR: 41.7)
Backup SCRAM valve solenoids Importance Rating 3.5 3.6 Proposed Question: WHICH ONE of the following supplies power to the Backup Scram Valves?
a) 120VAC UPS ROISRO b) 125VDC 54/75 c) 24VDC d) 120VAC RPS Proposed Answer: b) 125VDC Explanation (Optional):
Technical Reference(s): OP-18 (Attach if not previously provided)
~~
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-05, EO-1.04.A (As available)
Question Source: Bank ## Oyster Creek 1 INPO # 13001 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 412911996 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 92 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 CRD Hydraulic Group # 2 2 Knowledge of electrical power supplies to the WA# 201001 K2.03 K2.03 following: (CFR: 41.7)
Backup SCRAM valve solenoids Importance Rating 3.5 3.6 Proposed Question: WHICH ONE of the following supplies power to the Backup Scram Valves?
a) 120 VAC Vital instrument ROJSRO b) 125 VDC 54ff5 c) 24VDC d) 120 VACRPS Proposed Answer: m gw t%
P q
?
S Explanation (Optional): b) \IW Lc Technical Reference(s): w - \0 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: sled*
Learning Objective: T G L ~-05 ~ \,04,a- (As available)
Question Source: Bank # Oyster Creek 1 INPO # 13001 L Modified Bank #
New (Note changes or attach parent)
Question History: Last NRC Exam 412911996 (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge K Comprehension or Analysis 10 CFR Part 55 Content: 55.41 %
! q,7 55.43 Comments:
Page 93 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44 AM
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of the requirements for reviewing and Group #
approving release permits. (CFR: 43.4 /45.10)
KIA # 2.3.6 Importance Rating 3.1 Proposed Question: Who must authorize the Liquid Radwaste Effluent Discharge Permit prior to discharge?
a) Shift Manager ROlSRO b) Control Room Supervisor S76 c) Chemistry Supervisor d) Radiation Protection Manager Proposed Answer: a) Shift Manager Explanation (Optional):
Technical Reference@): SP-1.05 Attachment # 2 (Attach if not previously provided)
OP-49 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-20, EO-I .13.B (As available)
Question Source: Bank # LaSalle 1 INPO # 19297 (Modified to JAF)
Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam 11/20/2000 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Comments:
Page 93 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of the requirements for reviewing and Group #
approving release permits. (CFR: #$$$ I45.10)
KIA # 2.3.6 Importance Rating 3.1 Proposed Question: level of station must review and approve the Supervisor RO/SRO 5 76 d) Health Phpics Manager t;.
\
Y Proposed Answer: c) Unit Supervisor Explanation (Optional):
Technical Reference(s): SF 1.0s L&-.7- (Attach if not previously provided)
OP- Y 9 Proposed references to be provided to applicants during examination: ,
Learning Objective: (As available)
Question Source: Bank ## LaSalle 1 INPO # 19297 ( pm4 3 Q I )
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1112012000 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge K Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 K 4 Comments:
Page 94 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Control Rod and Drive Mechanism Group ##
Ability to (a) predict the impacts of the following WA # 201003 A2.01 A2.01 on the CONTROL ROD AND DRIVE MECHANISM :
and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 145.6)
Stuck rod Importance Rating 3.4 3.6 Proposed Question: A normal reactor startup was in progress at 7% reactor power. Control Rod 26-35 did not move when given a withdraw signal from it's current notch position 12. Drive water differential pressure has been adjusted to 450 psid. All previous attempts to move this rod have been unsuccessful.
The operator should. . .
a) Individually $CRAM Control Rod 26-35, then disarm it electrically and hydraulically.
ROlSRO b) Attempt to move Control Rod 26-35 by performing "Double Clutching."
55/77 c) Declare Control Rod 26-35 INOPERABLE, then disarm it electrically and hydraulically.
d) Raise drive water differential pressure an additional 50 psig and attempt to withdraw Control Rod 26-35.
Proposed Answer: d) Raise drive water differential pressure an additional 50 psig and attempt to
\-- withdraw Control Rod 26-35.
Explanation (Optional):
Technical Reference(& A 0 P-24 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Obiective: SDLP-O3C, EO-1.15.C (As available)
Question Source: Bank # Quad Cities 1 INPO # 19545 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 8/13/2001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Page 94 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Control Rod and Drive Mechanism Group # 2 2 Ability to (a) predict the impacts of the following WA # 201 003 A2.01 A2.01 on the CONTROL ROD AND DRIVE MECHANISM ;
and (b) based on those predictions, use procedures to correct, control, or mitigatethe consequences of those abnormal conditions or operations: (CFR: 41.5 145.6)
Stuck rod Proposed Question:
rature. Rod H-5 did not move when given a withdraw signal from Ily and hydraulically.
b) attempt to move the d) increase drive pressure 50 psig attempt to withdraw H-5.
\
Proposed Answer: d) increase drive pressure 50 psig and reattempyo withdraw H-5.
Explanation(Optional):
Technical Reference(s): ALP.24 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: wlal Learning Objective: SrJLp- o x I&-.& (As available)
Question Source: Bank # Quad Cities 1 INPO # 19545 ( +, LJAF 1 Modified Bank # (Note changeshttach pare$)
New Question History: Last NRC Exam 811 312001 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 y 5 ? 7?
55.43 Comments:
Page 95 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44 AM
L Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RWM Group # 2 2 Ability to manually operate andlor monitor in the WA # 201006 A4.06 A4.06 control room: (CFR: 41.7 I45.5 to 45.8)
Selected rod position indication:P-Spec(Not-BWRG)
Importance Rating 3.2 3.2 Proposed Question: A manual SCRAM was i n s e r t e d b a s e d on l o w e r i n g RPV w a t e r l e v e l .
The c o n d i t i o n h a s been c o r r e c t e d and RPV l e v e l h a s been r e t u r n e d t o t h e Green Band. During t h e SCRAM, two ( 2 ) c o n t r o l r o d s f a i l e d t o f u l l y i n s e r t . The SNO h a s a t t e m p t e d t o i n s e r t c o n t r o l r o d s u s i n g t h e CRD System p e r EP-3, Backup C o n t r o l Rod I n s e r t i o n .
Which of t h e f o l l o w i n g c o n d i t i o n s c o u l d p r e v e n t manual c o n t r o l rod insertion?
a) SDIV High Level Over-ride Switch i n 'Normal'.
ROlSRO b) Rod Worth Minimizer Bypass Switch i n 'Normal' 56/78 C) A l t e r n a t e Rod I n s e r t i o n ( A R I ) NOT r e s e t .
d) R e a c t o r P r o t e c t i o n System SCRAM NOT reset.
Proposed Answer: b ) Rod Worth Minimizer Bypass Switch i n 'Normal'.
Explanation (Optional):
Technical Reference(?,) : EP-3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EOP3LP, EO 1 . 0 7 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6, 7 55.43 Comments:
Page 95 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RWCU Group # 2 2 Knowledge of REACTOR WATER CLEANUP KIA# 204000 K4.03 K4.03 SYSTEM design feature(s) andlor interlocks which provide for the following: (CFR: 41.7)
Over temperature protection for system components Importance Rating 2.9 2.9 Proposed Question: The plant is performing a reactor startup and heatup, currently at 200 psig.
Reactor water level control is via Reactor Water Cleanup (RWCU) rejecting to the main condenser hotwell Main condenser vacuum has been established with the vacuum pump The operator is cautioned to carefully monitor system parameters while rejecting.
Which of the following RWCU system tripslisolations provide protection while in this Iineup?
a) Cleanup Blowdown Flow Control Valve (12FCV-55) closure on low upstream pressure ROlSRO b) RWCU system isolation on non-regenerative heat exchanger high outlet temperature.
57/79 c) Cleanup Blowdown Flow Control Valve (12FCV-55) closure on non-regenerative heat exchanger high outlet temperature.
d) RWCU system Containment Isolation Valve closure on low upstream pressure Proposed Answer: b) RWCU system isolation on non-regenerative heat exchanger high outlet
'- temperature.
Explanation (Optional):
Technical Reference(s): OP-28, ARP 09-4-2-35 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-12, EO 1.05.C.l (As available)
Question Source: Bank # Peach Bottom 2 INPO # 18536 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question Historv: Last NRC Exam 9/19/1997 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 96 of 120 NRC Written Examination Submittaldoc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO
~-
-4' Tier # 2 2 RWCU Group # 2 2 Knowledge of REACTOR WATER CLEANUP KIA# 204000 K4.03 K4.03 SYSTEM design feature@) and/or interlocks which provide for the following: (CFR: 41.7)
Over temperature protection for system components Importance Rating 2.9 2.9 Proposed Question: - d
- --. a= reactor startup and h e a t u b d & p i p ,
-Reactor water level control is via Reactor Water Cleanup (dWCU) rejecting to the main condenser h o t w e l l
- Main condenser vacuum has been established with the vacuum pump
- The operator is cautioned to carefully monitor system parameters while rejecting Which of the following RWCU system tripshsolationsuritlprovideprotection while in this lineup?
R;f+n *.mt !=L '2fCW.5 a) C l e a n u p B w w M d e f,Control Valve (6YbB)closure on low upstream pressure ROlSRO b) RWCU system isolation on non-regenerative heat exchanger high outlet 1tGC.g-55 57/79 c) Cleanup Control Valve (6945) closure on non-regenerative heat
' exchanger high outlet iemperature. .
d) RWCU system isolation on low upstream pressure Proposed Answer: b) RWCU system isolation on non-regenerative heat exchanger high outlet temperature.
Explanation (Optional):
Technical Reference(s): OP-28, ARP 09-4-2-35 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: WONt Learning Objective: SDLP-12, EO 1.05.c.l (As available)
\
Question Source: Bank # Peach Bottom 2 INPO # 18536 b#.j Q 3 pf.)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 911911997 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge d Comprehension or Analysis 10 CFR Part 55 Content: 55.41 b( 7 55.43 Page 98 of 129 NRC Written Examination Begin to END.docLast printed 5/21/2003 8:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RPlS Group # 2 2 Ability to monitor automatic operations of the KIA # 214000 A3.02 A3.02 ROD POSITION INFORMATIONSYSTEM including: (CFR: 41.7 145.7)
Alarm and indicating lights Importance Rating 3.2 3.1 Proposed Question: During a plant startup, with reactor power at 12%, control rod 18-1Iwas selected and the following indications occur:
Annunciator 09-5-2-2, ROD WITHDRAWAL BLOCK Annunciator 09-52-1, RWM ROD BLOCK A loss of ALL rod position indications on the Four Rod Display occurred A loss of ALL red Full-Out and green Full-In indications of Full Core Display Which of the following may be the cause for these indications?
a) Loss of 120 VAC Panel 71RBACB5 ROlSRO b) Loss of Panel 71AC10 58/80 c) Loss of Reactor Protection System (RPS) Distribution Panel A d) Loss of Uninterruptible Power Supply (UPS)
Proposed Answer: d) Loss of Uninterruptible Power Supply (UPS)
Explanation (Optional):
Technical Reference(s): A 0 P-21 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-O3G, EO-1.04, LPAOP, EO-1.01 (As available)
Question Source: Bank # Fermi 2 2 INPO # 7322 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 12/11/1995 (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 97 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RPlS Group # 2 2 Ability to monitor automatic operations of the KIA# 214000 A3.02 A3.02 ROD POSITION INFORMATION SYSTEM including: (CFR: 41.7 / 45.7)
Alarm and indicating lights Importance Rating 3.2 3.1 Proposed Question: During a plant startup, with reactor power at 12%, control rod 18-11 was selected and
+$">a r%m~
- - 2-m=jse-39-s-:- ;
the following indications occur:
M,
. i2-d L37$';
RODWORTH MINIMIZER BLOCKING began alarming zdm &h at-,,\ r;'ucac*c
-A loss of ALL rod position indications on the Four Rod Display occurred
-A loss of ALL red FulI-Out and green Full-In indications of Full Core Display.
Which of the following may be the cause for these indications?
(2bV A c @
a) Loss ofMW#3-,'\\UBAC'OS b 5 9 p-4 T\hL\O.. . .
RO/SRO b) 58/80 c) Loss of Reactor Protection System (RPS) Distribution Panel A d) Loss of Uninterruptible Power Supply,(UPS) P Proposed Answer: C N n Explanation (Optional): d. LO!, :: 97s c
c- Technical Reference(s): (Attach if not previously provided)
Proposed references to be provided to applicants during examination: f4* r'c Learning Objective: (As available)
Question Source: Bank # Fermi 2 2 INPO # 7322 ( p . ~A, 5 \
Modified Bank # (Note changes ar attach parent)
New Question History: Last NRC Exam 12/11/1995 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis Y 10 CFR Part 55 Content: 55.41 o\ f 55.43 Comments:
c Page 100 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/ZOO3 8:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RBM Group # 2 2 Ability to predict andlor monitor changes in KIA # 215002 A I .01 A1.O1 parameters associated with operating the ROD BLOCK MONITOR SYSTEM controls including:
(CFR: 41.5 145.5)
Trip reference: BWR-3,4,5 Importance Rating 2.7 2.8 Proposed Question: While withdrawing control rod 26-27 at 40% power, which of the below is the probable cause of a withdraw rod block?
a) Out of sequence rod recognized by the Rod Worth Minimizer.
ROISRO b) Rod Block Monitor green Push To Setup lamp is lit.
59/81 c) Control Rod 26-27 has withdrawn more than one (1) notch beyond the other rods in that group.
d) All Detector A Bypass lamps are lit on the Four Rod Display.
Proposed Answer: b) Rod Block Monitor green Push To Setup lamp is lit.
Explanation (Optional):
Technical Reference(s): OP-16 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: RAP-7.3.16, Attachment 3 Learning Objective: SDLP-7C, EO-1.05.8.4.F (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
Page 98 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 10:27AM
APPROVED CONTROL ROD PATTERN Page 1 2:
51 47 43 39 35 31 27 23 19 15 11 7
3 2 6 10 14 18 22 26 30 34 38 42 46 50
- 1. Fully withdrawn control rods are indicated by blanks.
- 2. Immediately notify Reactor Engineer if any control rod is found out of its approved position.
Date Date RE Initial Applicable From: To:
Updated From: To:
Updated From: . To:
- This IS a Quality Record -
I 1 RAP-7.3.16 Rev. No. 36 PLANT POWER CHANGES ATTACHMENT 3 Page - 3 9 of 41 I
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Nuclear Boiler Inst. Group #
Knowledge of the effect that a loss or WA# 216000 K6.01 K6.01 malfunction of the following will have on the NUCLEAR BOILER INSTRUMENTATION :
(CFR: 41.7 145.7)
A.C. electrical distribution Importance Rating 3.1 3.3 Proposed Question: Given the following plant conditions:
- Drywell temperature - 120F
- Reactor Building temperature - 94F
- Reactor pressure- 880 psig Immediately following a loss of all AC power, WHAT is the MINIMUM reactor water level that can be monitored from the control room?
a) +14.5" ROISRO b) -150" 60182 C) -145" d) +164.5" Proposed Answer: c) -145" Explanation (Optional):
Technical Reference@): EOP- Caution #I (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOP's Learning Obiective: EOP2LP, EO-1.01 (As available)
Question Source: Bank # LaSalle 1 INPO # I1671 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1016l1995 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 99 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO
. L d Tier # 2 2 Nuclear Boiler Inst. Group # 2 2 Knowledge of the effect that a loss or WA ## 216000 K6.01 K6.01 malfunction of the following will have on the NUCLEAR BOILER INSTRUMENTATION :
(CFR: 41.7 I45.7)
AC. electrical distribution Importance Rating 3.1 3.3 Proposed Question: Following a loss of all AC power, what is the MINIMUM reactor water level that can be monitored from the control room?0PLANTCONDITIONS:ODrywelltemperature: 120F, Reactor Building temperature: 94F, Reactor pressure: 880 psig a) pinches .+./q& '
ROISRO b) -150 inches 60182 c) -Nlinches *!% e d) &inches /kc[j" Proposed Answer. a)
Pinches -tqs 'I Explanation (Optional): QUESTION IS TIED to 216000 K5.01- Make a new Tie Technical Reference(s): hf-P d A - L (Attach if not previously provided)
Proposed references to be provided to applicants during examination: %AEC Learning Objective: iul IT 7 b l , 118 113 I (As available)
Question Source: Bank # LaSalle 1 INPO # 11671 &
hJ t j P J )
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1016/1995 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis x 10 CFR Part 55 Content: 55.41 7( 7 55.43 Comments:
Page 102 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/2003 8:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RHRLPCI: Torus/Pool Spray Mode Group # 2 2 Ability to manually operate andlor monitor in the KIA # 230000 A4.02 A4.02 control room: (CFR: 41.7 / 45.5 to 45.8)
Spray valves Importance Rating 3. a 3.6 Proposed Question: During a LOCA, the A and C RHR Pumps are injecting in attempts to place the A loop of RHR in Torus Spray/ Coolin R P t W h Y I w - Torus
.vV l~slC Which one of the following will occur?
RO/SRO c 62/84 d) The valves w i NOT opeh due to a LPCl initiation s i g n a h present.
Proposed Answer: d) The valves will NOT open due to a LPCl initiation signalwpresent.
Explanation (Optional): NO TIES TO THIS KA Exist- Modify this question to test A402 Technical Reference(s): 013-13 c&h-&s (Attach if not previously provided)
Proposed references to be provided to applicants during examination: No&
Learning Objective: SIM- 13 1105.a, 7- t, (As available)
Question Source: Bank # Fermi 2 2 INPO # 8890 p& . I64bb Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 4/6/1998 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis K 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 105 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/20038:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RHRILPCI: Torus/Pool Spray Mode Group # 2 2 Ability to manually operate andlor monitor in the WA # 230000 A4.02 A4.02 control room: (CFR: 41.7 I45.5 to 45.8)
Spray valves Importance Rating 3.8 3.6 Proposed Question: During a LOCA, the A and C RHR Pumps are injecting in the LPCl Mode. An Operator attempts to place the A loop of RHR in Torus Spray as directed by the Control Room Supervisor.
Without further Operator action, design interlocks will result in which of the following when valve operation is initiated?
a) The valves may be opened but will immediately close due to a LPCl Initiation signal present.
ROlSRO b) 10MOV-39A and 34A will open, but 10MOV-38A will NOT open allowing Torus cooling mode of operation.
62/84 c) 10MOV39A and 38A will open but 10MOV-34A will NOT open allowing Torus spray mode of operation.
d) The valves will NOT open due to a LPCl initiation signal being present unless the initiation signal is first overridden.
Proposed Answer: d) The valves will NOT open due to a LPCl initiation signal being present unless the initiation signal is first overridden.
Explanation (Optional):
Technical Reference(s): OP-13 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-10, EO-I .05.A.2 (As available)
Question Source: Bank ## Fermi 2 2 INPO # 8890 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 4161199%
(Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 101 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
L Examination Outline Cross-reference: Level RO SRO Tier # 2 2 RHRILPCI: CTMT Spray Mode Group # 2 2 Knowledge of the physical connections andlor KIA # 226001 K1.12 K1.12 cause effect relationshipsbetween RHRILPCI:
CONTAINMENT SPRAY SYSTEM MODE and the following: (CFR: 41.2 to 41.9 145.7 to 45.8)
Suppression pool (spray penetration): Plant-Specific Importance Rating 3.0 3.0 Proposed Question: Why does the Torus Spray flowpath of EOP-4, Primary Containment Control, prohibit initiation of Torus Spray if Torus Level is >
26 feet?
a) Less than 95% of non-condensable gasses exist in the Torus air space.
ROlSRO b) The spray header is covered by Torus water level.
61/83 c) The DW to Torus Vent flowpath has been lost.
d) Initiation of Sprays could bring the Torus to sub-atmospheric conditions.
Proposed Answer: b) The spray header is covered by Torus water level.
Explanation (Optional):
Technical Reference(s): BWROG EPGs (Attach if not previously provided)
L EOP4LP Proposed references to be provided to applicants during examination: None Learning Objective: EOP4LP, EO-1.05 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 willgenerally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
Page 100 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Fuel Pool CoolinglCleanup Group # 2 2 Knowledge of FUEL POOL COOLING AND WA# 233000 K4.03 K4.03 CLEAN-UP design feature(s) andlor interlocks which provide for the following: (CFR: 41.7)
Maintenance of adequate pool temperature Importance Rating 2.8 3.1 Proposed Question: A design basis of the Fuel Pool Cooling and Cleanup System is to maintain the Spent Fuel Pool outlet temperature below for a peak annual refueling heat load of 10 X I O 6 BTU/Hr.
a) 155'F ROlSRO b) 145'F 63/85 c) 135°F d) 125'F Proposed Answer: c) 135 "F Explanation (Optional):
Technical Reference(s1: OP-30, AOP-53, FSAR (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None L
Learning 0bjective: SDLP-I 9, EO-1.02 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 7 Comments:
Page 102 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO L
Tier # 2 2 Reactornurbine Pressure Regulator Group # 2 2 Knowledge of the operational Implications of the WA # 241000 K5.05 K5.05 following concepts as they apply t o REACTOWTURBINE PRESSURE REGULATlNG SYSTEM : (CFR: 41.5 145.3)
Turbine inlet pressure vs. turbine load Importance Rating 2.8 2.9 Proposed Question: Reactor Power is reduced from 100% to 95% by lowering recirculation flow.
Turbine Control Valves are repositioned by EHC sensing as compared to ?
a) RPV Pressure, Pressure Setpoint.
ROISRO b) RPV Pressure, Turbine 1 Stage Pressure 64/86 c) Turbine Inlet Pressure, Turbine 1"Stage Pressure d) Turbine Inlet Pressure, Pressure Setpoint.
Proposed Answer: d) Turbine Inlet Pressure, Pressure Setpoint.
Explanation (Optional):
Technical Reference(s): SLP-74c (Attach if not previously provided)
'L Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-74C, EO-I .05.A.4 (As available)
Question Source: Bank # Dresden 2 INPO # 6524 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 3111/1996 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge A Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Comments:
c Page 103 of 120 NRC Written Examination SubmittaLdoc Last printed 6/6/2003 10:27 AM
Level
@!- Examination Outline Cross-reference: RO SRO Tier # 2 2 Reactornurbine Pressure Regulator Group # 2 2 Knowledge of the operational Implicationsof the WA # 241000 K5.05 K5.05 following concepts as they apply to REACTOWURBINE PRESSURE REGULATING SYSTEM :(CFR: 41.5 145.3)
Turbine inlet pressure vs. turbine load Importance Rating 2.8 2.9 lm Proposed Question: You reduce reactor power from 100% to 95% by 4 recirculation flow. What signals, i5;3ny, did the Electro Hydraulic Control use to move the turbine control valve?
d a) w r Pressure, Pressure Setpoint.
ROJSRO b) Reeebr Pressure, Max Combined Load Limit.
64/86 c) None; the turbine control valve did NOT move.
f4UcT d) Turbine UrrC(lr Pressure, Pressure Setpoint.
Proposed Answer: d) Turbine-* Pressure, Pressure Setpoint.
Explanation (Optional): - N tiorrwasriea IO - 1 e the tie t6441000 k595.
Technical Reference(s): (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: 3-y-P sue f,OS.&,Lt (As available)
Question Source: Bank # Dresden 2 INFO ## 6524 C & ppF Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 311111996 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis b (
10 CFR Part 55 Content: 55.41 bjs 55.43 Comments:
Page 107 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/2003 8:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 2 2 Secondary CTMT Group # 2 2 Knowledge of the effect that a loss or WA# 290001 K3.01 K3.01 malfunction of the SECONDARY CONTAINMENT will have on following: (CFR: 41.7 145.4) t0ff-site radioactive release rates Importance Rating 4.0 4.4 Proposed Question: The plant was operating at 100% power when a large steam leak occurred inside the Reactor Building. SGT Train "A"and " B are operating at rated flows. Secondary Containment pressure is +1.5" WG.
Off-Site radioactivity release rates are expected to be... ...
a) Ground releases via SGT only ROlSRO b) Ground releases via SGT and Reactor Building Ventilation 65/87 c) Elevated releases via SGT and Ground releases via Reactor Building leakage d) Elevated releases via SGT and Ground releases via Reactor Building Ventilation Proposed Answer: c) Elevated releases via SGT and Ground releases via Reactor Building leakage Explanation (Optional):
Technical Reference(s): OP-51A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-16A EO-1.09b, SDLP-66A, EO-1.05.C (Asavailable)
L-Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X 55.43 Comments:
Page 104 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
L Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Ability to explain and apply system limits and Group #
precautions. (CFR: 41.101 43.2 145.12)
WA # 2.1.32 2.1.32 Importance Rating 3.4 3.8 Proposed Question: Prior to returning to two loop operation from one loop operation which of the following LIMITS must be met and what is the REASON for that limit?
a) LIMIT - The temperature difference between the bottom head coolant and the recirc loop coolant in the loop to be started is < 145 deg F.
REASON - To prevent a violation of the RPV pressure and temperature limitationthat minimize the chances of brittle fracture from occurring.
b) LIMIT - The temperature difference between the recirc loop coolant in the loop to be started and the reactor vessel coolant is < 50 deg F.
ROISRO REASON - To prevent a violation of the RPV pressure and temperature limitation that minimize the chances of brittle fracture from occurring.
c) LIMIT - The temperature difference between the bottom head coolant and the recirc loop coolant in the loop to be started is e 145 deg F.
66188 REASON - To prevent damage to the fuel cladding that would result from the sudden increase in power due to the injection of cold water.
d) LIMIT - The temperature difference between the recirc loop coolant in the loop to be started and the reactor vessel coolant is c; 50 deg F.
REASON - To prevent damage to the fuel cladding that would result from the sudden increase in power due to the injection of cold water.
Proposed Answer: b) LIMIT - The temperature difference between the recirc loop coolant in the loop to be started and the reactor vessel coolant is < 50 deg F.
REASON - To prevent a violation of the RPV pressure and temperature limitation that minimize the chances of brittle fracture from occurring.
Explanation (Optional):
Technical Reference@): ST-26K (Attach if not previously provided)
Proposed references to be provided to applicants during examination: NONE Learning Objective: SDLP-2H, EO-1.139 (As available)
Question Source: Bank # Dresden 2 INPO # 21373 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 611412002 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X
'C Comprehension or Analysis Page 105 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Ability to explain and apply system limits and Group #
precautions. (CFR: 41.10 143.2 145.12)
KIA # 2. I.32 2.I.32 Importance Rating 3.4 3.8 Proposed Question: Prior to returning to two loop operation from one loop operation which of the following LIMITS must be met and what is the REASON for that limit?
a) LIMIT - The temperature difference between the bottom head coolant and the recirc loop coolant in the loop to be started is < 145 deg F.
REASON - To prevent a violation of the RPV pressure and temperature limitation that minimize the chances of brittle fracture from occurring.
b) LIMIT - The temperature difference between the recirc loop coolant in the loop to be started and the reactor vessel coolant is 50 deg F.
ROISRO REASON - To prevent a violation of the RPV pressure and temperature limitation that minimize the chances of brittle fracture from occurring.
c) LIMIT - The temperature difference between the bottom head coolant and the recirc loop coolant in the loop to be started is < 145 deg F.
66/88 REASON - To prevent damage to the fuel cladding that would result from the sudden increase in power due to the injection of cold water.
d) LIMIT - The temperature difference between the recirc loop coolant in the loop to be started and the reactor vessel coolant is < 50 deg F.
REASON - To prevent damage to the fuel cladding that would result from the sudden increase in power due to the injection of cold water.
Proposed Answer: b) LIMIT - The temperature difference between the recirc loop coolant in the loop to be started and the reactor vessel coolant is < 50 deg F.
REASON - To prevent a violation of the RPV pressure and temperature limitationthat minimize the chances of brittle fracture from occurring.
Explanation (Optional):
Technical Reference(s): 5 OLP - y-4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: xss Learning Objective: I \ 13\ 9 (As available)
Question Source: Bank # Dresden 2 INPO # 21373 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 6/14/2002 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC;
- failure to provide the information will necessitate a detailed review of every question.)
Page 109 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/2003 8:44 AM
10 CFR Part 55 Content: 55.41 X 55.43 Comments:
Page 106 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
L Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of how to conduct and verify valve Group #
lineups. (CFR: 41.10 145.1 145.12)
WA # 2. I .29 2.1.29 Importance Rating 3.4 3.3 Proposed Question: WHICH ONE (1) of the following conditions would allow waiving the independent verification requirements to verify a valve's position?
a) If the valve was located in the Main Stack Building ROlSRO b) If the valve was not associated with an ECCS system 67/89 c) If excessive radiation exposure would be required to verify the valve's position d) If the valve had position indication in the control room Proposed Answer: c) If excessive radiation exposure would be required to verify the valve's position Explanation (Optional):
Technical Reference@): AP-12.06 Section 7.6 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPAP-48.03 (As available)
Question Source: Bank # Pilgrim IINPO # 18369 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1011611998 (Optional - Questions validated at the facility since 10195 willgenerally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Page 107 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
.~ ~~~
Examination Outline Cross-reference: Level RO SRO h/ Tier # 3 3 Knowledge of how to conduct and verify valve G~~~~#
lineups. (CFR: 41.10 /45.1/45.12)
WA # 2.1.29 2.1.29 Importance Rating 3.4 3.3 Proposed Question: WHICH ONE ( I ) of the following conditions would allowwaiving the independent verification requirements to verify a valve's position?
%c.$ If the valve was located in the Main Stack Building ROISRO If the valve was not associated with an ECCS system 67189 4 -fI
--pxc&c&.fe~+-?-
30-FMemwould be required to venfy the valve's position 4 95) If the valve had position indication in the control room Proposed Answer: c) If it is estimated thaH%Memwould be required to verify the valve's position Explanation (Optional):
Technical Reference(s): AP IZ.Ob s.c.-?.b (Attach if not previously provided)
Proposed references to be provided to applicants during examination: L3r L k Learning Objective: (As available)
Question Source: Bank # Pilgrim IINPO # 18369 (- n,k+ j&&)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1011611398 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge bc Comprehension or Analysis 10 CFR Part 55 Content: 55.41 d lo 55.43 Comments:
Page 11Iof 129 NRC Written Examination Begin to END.doc Last printed 5/2112003 8:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Ability to perform pre-startup procedures for the Group #
facility / including operating those controls associated with Dlant equipment that could affect reactivity. (CFR: '45.I)
KIA # 2.2.1 2.2.1 Importance Rating 3.7 3.6 Proposed Question: A plant startup is in progress with the Mode Selector Switch in Startup. Control rods are being withdrawn.
The Rod Worth Minimizer (RWM) has just failed with 25% of the control rods withdrawn.
What action is required?
a) Bypass the RWM, verify all further control rod movements are in compliance using a qualified person, and continue the reactor startup.
ROlSRO b) Suspend withdrawal of the control rods, manually SCRAM the reactor, and verify operability of the RWM before commencing a reactor startup.
68/90 c) Suspend withdrawal of the control rods, verify operability of the Rod Block Monitor, and continue the reactor startup.
d) Bypass the RWM, fully insert all control rods, and verify operability of the RWM before commencing a reactor startup.
Proposed Answer: a) Bypass the RWM, ver-iy all further control rod movements are in compliance using a qualified person, and continue the reactor startup L Explanation (Optional):
Technical Reference(s): OP-64 Section E.l (Attach if not previously provided)
OP-65 Proposed references to be provided to applicants during examination: None Learning Objective: SDLP9D, EO-1.15.A, LPAP, EO-46.04 (As available)
Question Source: Bank # Quad Cities 1 INPO # 20445 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 8113/2001 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Comments:
Page 108 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Abilityto perform pre-startup procedures for the Group #
facility I including operating those controls associated with plant equipment that could affect reactivity. (CFR: 45.1)
KIA # 2.2.1 2.2.1 Importance Rating 3.7 3.6 A ' : ' * . j - ' \j\ c *,; I 4 c , z
- d ~ ~ : A , o o p ~t\<,- sar\c.- iJ . c Proposed Question: T . ' m h t - u r r U n i M n + ~in #+e STARTUP mxtmnk&ntrd nio-r an approach to criticality.$- The Rod Worth Minimizer (RWM) has just fail$ with 25% of the control rods withdrawn.- - , eIUdA4PHT STARTW+ hat isHte action that is required?
a) Bypass the RWM, verify control rod movements are in compliance using a qualified person, and continue the reactor startup.
b) Suspend withdrawal of the control rods, place the reactor mode switch in the RO/SRO SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and verify operability of the RWM before commencing a reactor startup.
68/90 c) Suspend withdrawal of the control rods, verify operability of the Rod BlockOMonitor, and continue the reactor startup.
d) Bypass the RWM, fully insert all control rods, and verify operabilky of the RWM before commencing a reactor startup.
Proposed Answer: a) Bypass the RWM, verify control rod movements are in compliance using a qualified person, and continue the reactor startup.
Explanation (Optional):
Technical Referencels): OP-64 Section E . l (Attach if not previously provided)
OP-65 Proposed references to be provided to applicants during examination: None Learning Objective: (As available)
Question Source: Bank # Quad C i e s 1 INPO # 20445 Modified Bank# (Note changes or attach parent)
New Question History: Last NRC Exam 811 32001 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
v Question Cognitive Level: Memory or Fundamental Knowledge A Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Comments:
Page 108 of 124 NRC Written Examination Begin to END.doc Last printed 5/29/2003 1 5 8 PM
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of surveillance procedures. Group #
(CFR: 41.10 145.13)
KIA # 2.2.12 2.2.12 Importance Rating 3.0 3.4 Proposed Question: The Plant is operating at 90% Reactor power. The Control Room Supervisor has ordered you to perform ST-24J, RCIC Flow Rate and Inservice Test, following maintenance.
During RCIC pump operations:
a) A manual SCRAM will be inserted if Torus water temperature exceeds 95 O F .
RO/SRO b) EHC Pressure Set will be adjusted to maintain RPV pressure 970 psig.
69/91 c) Recirculation flow will be reduced to maintain Reactor power 400%.
d) Torus cooling will be in service to prevent Torus water temperature from exceeding 105 "F.
Proposed Answer: d) Torus cooling will be in service to prevent Torus water temperature from exceeding 105 'F.
Explanation (Optional):
Technical Reference(s): ST-24J, ITS-3.6.2.1 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-13, EO-1.13.d (As available)
Question Source: Bank # LaSalle IINPO # 19132 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 11/20/2000 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Page 109 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
L Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of the process for determining the Group #
internal and external effects on core reactivity.
(CFR: 43.6)
WA # 2.2.34 2.2.34 Importance Rating 2.8 3.2 Proposed Question: A Reactor startup from Cold Shutdown is in progress. The ECP was calculated based upon the following:
e Reactor Coolant temperature at 140 "F Total Core Flow at 10 X I O 6 lbmlhr e At time of criticality, Reactor has been shutdown for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> Feedwater temperature 80 "F Which of the below will result in criticality later than the predicted ECP?
a) Criticality occurs 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after shutdown.
ROlSRO b) Feedwater temperature drops to 75 "F.
70f92 c) Total Core Flow is raised to 15 X I O 6 lbmlhr d) Reactor Coolant temperature drops to 125 O F .
Proposed Answer: a) Criticality occurs 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after shutdown.
Explanation (Optional):
L Technical Referencek): RAP-7.3.13 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LPOP-65A, EO-1.10 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
~
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 1 55.43 6 Comments:
Page 110 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
L. Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of the process for performing a Group #
containment purge.
(CFR: 43.4 145.10)
KIA ## 2.3.9 2.3.9 Importance Rating 2.5 3.4 Proposed Question: The plant is conducting a shutdown, power is currently 30% and lowering. It is desired to de-inert the Primary Containment (both the Drywell and Torus) as soon as possible to permit containment access for maintenance for a forced outage.
Which procedurally allowed flowpath and sequence would permit the most expeditious de-inerting of the Primary Containment?
a) Through the Standby Gas Treatment System with the Drywell and Torus d e inerted simultaneously.
ROlSRO b) Through the Reactor Building Ventilation System with the Drywell de-inerted first and then the Torus de-inerted.
71/93 c) Through the Reactor Building Ventilation System with the Drywell and Torus de-inerted simultaneously.
d) Through the Standby Gas Treatment System with the Drywell de-inerted first and then the Torus de-inerted.
Proposed Answer: d) Through the Standby Gas Treatment System with the Drywell de-inerted first and then the Torus de-inerted.
Explanation (Optional):
Technical Reference(s): 0P-37 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-l.O6C, EO-1.13.C (As available)
Question Source: Bank # Quad Cities IINPO # 20444 (Modified to JAF)
Modified Bank ## (Note changes or attach parent)
New Question History: Last NRC Exam 811312001 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Page I11 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of EOP implementation hierarchy and Group #
coordinationwith other support procedures.
(CFR: 41.10 I 43.5 / 45.13)
KIA ## 2.4.16 Importance Rating 4.0 I n a n emergency e v e n t t h e r e a c t o r .gh drywell Proposed Question:
pressure.
The f o l l o w i n g p l a n t c o n d i t i o n s e x i s t :
D r y w e l l t e m p e r a t u r e SPDS d i s p l a y DWT VERTICAL RUN TEMP i n d i c a t e s 300 deg F.
RPV p r e s s u r e i s 40 p s i g and e q u a l i z e d w i t h t h e d r y w e l l .
RPV water l e v e l i n d i c a t i o n s a r e v e r y e r r a t i c and do n o t c o r r e l a t e w e l l w i t h one a n o t h e r .
Under t h e s e c i r c u m s t a n c e s , t h e o p e r a t i n g crew would b e r e q u i r e d t o e x e c u t e t h e f o l l o w i n g Emergency O p e r a t i n g P r o c e d u r e ( s ) :
a ) EOP-4, Primary Containment C o n t r o l , ONLY.
ROISRO b ) EOP-2, RPV C o n t r o l , AND EOP-4, Primary Containment C o n t r o l , c o n c u r r e n t l y ONLY.
L s94 c ) I n i t i a l l y , EOP-2, RPV C o n t r o l , @ JEOP-4, J Primary Containment C o n t r o l , c o n c u r r e n t l y , Then EOP-2, RPV Control, AND EOP-7, RPV Flooding, c o n c u r r e n t l y .
d ) I n i t i a l l y EOP-2, RPV C o n t r o l , 9 EOP-4, Primary Containment C o n t r o l , c o n c u r r e n t l y , THEN EOP-7, RPV Flooding, EOP-4, Primary Containment C o n t r o l ,
concurrently.
Proposed Answer: d ) I n i t i a l l y EOP-2, RPV C o n t r o l , AND EOP-4, Primary Containment C o n t r o l , c o n c u r r e n t l y , THEN EOP-7, RPV F l o o d i n g , EOP-4, Primary Containment C o n t r o l , c o n c u r r e n t l y .
Explanation (Optional):
Technical Reference(s): EOP-2,4, 7 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOPs Learning Objective: EOP2LP, EO-1-02, 1.03, EOPQLP, EO-4.02 (As available)
Question Source: Bank # JAF LOR # 20005204B04C Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Page 112 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of surveillance procedures. Group #
(CFR: 41.10 I45.13)
K/A # 2.2.12 2.2.12 Importance Rating 3.4 L L A &- C V A ,
Proposed Question: The pkmksupervisor has ordered you to perform a a) Chemistry analysis on the ROlSRO v M w + w w 69/9 1 d) Remote Shutdown Panel B&bed%& . InstrumentationOperability Checks.
Proposed Answer: c) S u p q i p n w Temperature Monitoring w s .
6b-G Explanation (Optional):
Technical Reference(s): ST-aI 3- (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learnina Obiective: (As available)
Question Source: Bank # 'LaSalle 1 INPO # 19132 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 11/2012000 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 %.
55.43 Comments:
Page 113 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/ZOO38:44 AM
10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 113 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of 10 CFR 20 and related facility Group #
radiation control requirements.
(CFR: 41.12 143.4. 45.91 45.10)
KIA # 2.3.1 2.3.1 Importance Rating 2.6 3.0 Proposed Question: As a result of degrading plant conditions, the Shift Manager has directed you to immediately investigate an equipment problem inside a locked high radiation area. The duty RP technician is assisting workers in the plant stack and is not immediately available.
What action should be taken to expedite your entry?
a) Using the key on any NPO Duty key ring.
ROISRO b) Go to the RP office and sign out a key yourself.
72/95 C) Contact and meet the RP tech to obtain a key.
d) Obtain a radiological master key from the Shift Manager.
Proposed Answer: d) Obtain a radiological master key from the Shift Manager.
Explanation (Optional):
Technical Reference(s): AP-07.06 (Attach if not previously provided)
L Proposed references to be provided to applicants during examination: None Learning Objective: LPAP, EO-31.03.H (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10I95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 12 55.43 4 Comments:
Page 114 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1052 AM
Examination Outline Cross-reference: -Level RO SRO Tier # 3 3 Knowledge of the process for performing a Group #
containment purge.
(CFR: 43.4 I45.10)
KIA # 2.3.9 2.3.9 2.5 Proposed Question: ..f
- u+bbd&& It is desired to deinert the kH-4 Primary Containment (both the drywell andntonrs) as soon as possible to permit containment access for maintenancm . . at
,What flowpath and sequence would permit the most expeditious0de-inerting of the+" Primary Containment?
a) With the Standby Gas Treatment System with the drywell and torus de-inerted simultaneously.
RO/SRO b) Through the Reactor Building Ventilation System with the drywell de-inerted first and then the torus de-inerted.
71/93 c) Through the Reactor Building Ventilation System with the drywell and torus de-inerted simuttaneously.
d) With the Standby Gas Treatment System with the drywell de-inerted first and then the torus de-inerted.
Proposed Answer. %gh and then %Starus de-inerted.
the ReactorBuilding VentilatiokQstm witM<well
-.\
d&ineFte$first ,
- - \
Explanation (Optional):
I -p-. GL . ,-r - --\
Technical Reference(s): (2- 31 (Attach if not previously provided)
~~
Proposed references to be provided to applicants during examination:
Learning Objective: (As available)
Question Source: Bank # Quad Cities 1 INPO # 20444 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 8/13/2001 (Optional Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 o(
55.43 Comments:
Page 115 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/2003 8:44 AM
Examination Outline Cross-reference: Level SRO L
Tier # 3 Knowledge of which events related to system Group #
operationslstatusshould be reportedto outside agencies.
(CFR: 43.5 145.11)
KIA # 2.4.30 Importance Rating 3.6 Proposed Question: As a result of an error during I/C Surveillance testing at 100% power, the MSIV's inadvertently closed resulting in the following:
HPCl initiation and injection.
HPCl tripped by the Operators.
Operator control of level using Feed and Condensate.
0 Operator control of RPV pressure using SRV's.
Which of the below is the required NRC report?
a) Immediate Notification due to an Emergency Plan event declaration.
ROISRO b) One (I) Hour Notification due to a deviation from Technical Specifications authorized by 10CFR-50.54 (X).
S96 c) Four (4) Hour Notification due to ECCS discharge to Reactor Coolant System resulting from a valid signal.
d) Eight (8) Hour Notification due to a valid Containment Isolation signal affecting L
Containment Isolation Valves.
Proposed Answer: c) Four (4) Hour Notificationdue to ECCS discharge to Reactor Coolant System resulting from a valid signal.
Explanation (Optional):
Technical Reference(s): ENN-LI-102, AP-03.11 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: AP-03.11 Learning Objective: LPAP-10.06 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:
L Page 115 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Page 116 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27AM
ENTERGY NUCLEAR OPERATIONS, INC.
JAMES A. FITZPATRICR NUCLEAR POWER PLANT v ADMINISTRATIVE PROCEDURE OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03 -11 REVISION 9 APPROVED BY:
T B
RESF~ONS~S PROCE DATE $??5/.L APPROVED BY: DATE -
d GENERAL MANAGER PLANT OPERATIONS EFFECTIVE DATE: 6-392 FTRST ISSUE FULL REVISION 0 LIMITED REVISION
- INFORMATIONAL USE *
- TSR *
- ADMINISTRATIVE
- I PERIODIC mIEW D13E DATE
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03-11 REVISION
SUMMARY
SHEET
\-
REV. NO. CHANGE AND REASON FOR CHANGE 9 Where applicable, indicated steps and paragraphs which address Attachments 2, 7, and 8. ACT-02-62578 (PCR eated 1/15/02)
Changed "Director Engineering" to "Engineering Manager ( s )'
throughout procedure. ACT-02-62577 (PCR dated 1/15/02)
Included "Entergy Licensing Position #2," on Attachment 8, as Management Expectation 4.2.5.
Added cross-references where applicable.
To Section 7.0, SPECIAL INSTRUCTIONS, added information on using the attachments not already listed which only provide guidance for filling out the attachments, which are forms.
This provides a "pointer" for the procedure user to use to go to the needed attachment.
To 8.2.7, added direction to record the PLCO number on Attachment 1, providing clarification to the procedure user.
Made minor changes to 8.2.7.B to better locate the information on Contingent Operator Actions.
To 8.2.8, added clarification to complete the 10CFR50.59 Screen per MCM-4.1, using the guidance in the rest of this step and in Attachment 8.
8.2.8.C - changed "initiation of" to "implementing" for clarification.
8.4 - Added a note at the head of this subsection stating the purpose and use of Attachment 9 .
8.4.4.B - Made it easier to locate information for Contingent Operator Actions in Attachment 6.
Made minor typographical, spelling, and formatting changes throughout procedure. Rev. bars not used.
Updated procedure references. DCM-2A & 4A have been replaced by ENN-DC-126.
Deleted reference to Asst Ops Manager review requirements.
They were intended as in interim measure and are no longer desired. Maintains alignment with other EN" sites, where
~L Rev. No. 9 Page No. 2 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 the SM is the final approval authority. Deletes responsibility in Section 6, Subsection 8.5, & Attachment 'W' 1.
Made the following changes to address reportability in this procedure due to the deletion of AP-03.02 and AP-03.03:
0 Where applicable, addressed ENN-LI-102, EAP-1.1, and J A F Corrective Action Process Desk Guide.
0 Revised the procedure title, purpose statement and applicability to address immediate reportability.
0 Substituted the term "Condition Report (CR) for "
"DER" and "Corrective Action (CAI for "ACT,"
consistent with the new Paperless Condition Reporting System ( P C R S ) . Rev. bars not used.
0 Deleted Performance References AP-03.02 and AP-03.03 due to their deletion and added references to ENN-LI-102, EAP-1.1, JAF Corrective Action Process Desk Guide, AP-03.04, Technical Requirements Manual, NRC IN 89-89, NUREG-1022, and 10CFR26, Fitness for Duty Programs.
0 Added new Expectation 4 . 2 . 5 , Entergy Licensing Position #2, Evaluation and Resolution of Degraded 'd and Nonconforming Conditions.
Added definitions for Immediate Notification, Reportable Event, and Safety Function.
Modified 6.4 Operations Manager and 6.6 Shift Manager responsibilities to address reportability.
Added Director Safety Assurance, CR Screening Committee, and Regulatory Compliance Manager to 6.0 Responsibilities.
0 In 7.0 Special Instructions, added requirement to immediately notify SM for certain conditions and guidance for SM to prepare Operability Determinations and Immediate Reportability Determinations.
In 7.0, addressed the purpose and use of the attachments.
In Section 8.0, added a subsection for Operability and Reportability Review.
Rev. No. 9 -- Page No. 3 of 78
O P E M I L I T Y AND REPORTABILITY DETERMINATIONS AP-@3 - 11 0 In Section 8.0, added a new subsection 8.4 for Immediate Reportability Determinations.
Modified Section 8 . 0 to address reportability determinations.
Added a new Attachment 7 , Reportability Checklist.
Renumbered the remaining attachments. Rev. bars not used.
Changed Atts. 9 and 10 to accurately address tne operability determination process.
Rev. No. 9 Page No. 4 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AF-03. II TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE . . . . . . . . . . . . . . . . . . . . . . . . 6 2.0 APPLICABILITY . . . . . . . . . . . . . . . . . . . . . 6
3.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . 6 4.0 REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . 8 5.0 DEFINITIONS . . . . . . . . . . . . . . . . . . . . . .
6.0 RESPONSIBILITIES . . . . . . . . . . . . . . . . . . . 17 7.0 SPECIAL INSTRUCTIONS . . . . . . . . . . . . . . . . . 20 8.0 PROCEDURE . . . . . . . . . . . . . . . . . . . . . . . 22 8.1 General Requirements . . . . . . . . . . . . . . . . . 22 8.2 Operability and Immediate Reportability Reviews . . . 23 8.3 Operability D e t e d n a t i o n s and Related Actions . . . . 24 8.4 Immediate Reportability D e t e d n a t i o n s . . . . . . . . 35 8.5 Management Review of Operability . . . . . . . . . . . 36 8.6 Records . . . . . . . . . . . . . . . . . . . . . . . . 37 9.0 ATTACHMENTS . . . . . . . . . . . . . . . . . . . . . . 38
- 1. OPERABILITY DETERMINATION FORM . . . . . . . . . . . 39
- 2. ENGINEERING CONFIRMATION
SUMMARY
FORM . . . . . . - 4 2
- 3. INITIAL ENGINEERING CONFIRMATION GUIDELINES . . . . 43
- 4. DETAILED ENGINEERING CONFIRMATION GUIDELINES . . . . 48
- 5. REO/REASONABLE ASSURANCE GUIDELINES . . . . . . . - 4 9
- 6. REO/ENGINEERING OPERABILITY GUIDELINES . . . . . . . 50
- 7. IMMEDIATE REPORTABILITY CHECKLIST . . . . . . . . . 61
- 8. ENTERGY LICENSING POSITION (EVALUATION AND RESOLUTION OF DEGRADED AND NONCONFORMING CONDITIONS) . . . . .73
- 9. ENGINEERING CONFIRMATION PROCESS FLOWCHART . . . . . 77
- 10. OPERABILITY DETERMINATION PROCESS FLOWCHART . . . . 78 Rev. No. 9 Page No. 5 of 78
OPERABILITY AND REPORTABILIm DETERMINATIONS ASP-03 -11 1.0 PURPOSE 1.1 To establish a method for determining the qerability af structures, systems, or components (SSCs) lhat have b n identified as being in a degraded or n o n c d c d g c o d i t i o c .
1.2 In addition, to establish a method for detem-g the potential reportability to outside regulatmry agencies; of issues identified via the corrective actiocn prmcess.
2.0 APPLICABILITY To the performance of Operability Determiriatioms a d Immediate Reportability Determinations for E r S e m Identifications (PIDs) and Condition Report (-1 ; however, it may be used at the Shift Manager's discretitxu for performing Operability Determinations and I l m n d a t e Reportability Determinations for other c o
3.0 REFERENCES
3.1 Performance References 3.1.1 AP-01.01, Plant Operating Review Cammitte9 3-1.2 [CTS J AP-01.04, Tech Spec Relatea Reqwirements, Lists, and Tables
[ITS] T e c k i c a l Requirements Hkcr~ud 3.1.3 AP-02.08, QuaLity Assurance Reccrd I d t m t i f i c e i o n and Control 3.1.4 AP-03.04, Information Reporting &E?qci?remants 3.1.5 AP-10.01, Problem Identification jand mrk C x u - t r o l 3.1.6 EAP-1.1, Offsite Notifications 3.1.7 3.1.8 3.1.9 [CTSI O D S O - 3 4 , Technical SpecificaticM LCO a m d Maintenance R u l e Unavailability Txacklng
[ITS] AP-12-06, LCO Tracking S a f e t y Function
--..-- Determination P r o g r a m Rev. No. 9
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 3.1.10 E"-DC-126, Calculations 3.1.11 DCM-14A, Preparation and Control of Computer Generated Calculations (JAF) 3.1.12 NRC Generic Letter 90-05, Guidance for Performing Temporary Non-code Repair of ASME Code Class 1, 2 ,
and 3 Piping 3.1.13 NRC Generic Letter (GL) 91-18, Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability 3.1.14 NRC Inspection Manual, Part 9900: Technical Guidance 3.1.15 NRC IN 89-89, Emergency Notification System (ENS) 3.1.16 NUREG-1022, Event Reporting Guidelines 10CFR50.72 and 10CFR50.73 (Revision 2, October 2000) 3.1.17 JAF Updated Final Safety Analysis Report (UFSAR) 3.1.18 JAF Design Basis Documents 3.1.19 JA?? Technical Specifications I 3.1.20 E"-LI-100, Process Applicability Determination 3.2 Developmental References 3.2.1 10CFR26, Fitness For Duty Programs 3.2.2 NuAP 4.12, Resolution of Equipment Operability Concerns Related to Degraded or Nonconforming Conditions 3.2.3 NUREG-1433, Standard Technical Specifications, General Electric Plants / BWR 4s I 3.2.4 E"-LI-101, 10CFR50.59 Review Process 3.2.5 MCM-4.2, 10CFR50.59 Evaluations 3.2.6 W4.101, Waterford 3 Management Manual Procedure Operability Confirmation Process 3.2.7 Entergy Licensing Position #2, Evaluation and Resolution of Degraded and Nonconforming Conditions Rev. No. 9 Page No. 7of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-23.1:
4.0 REQUI-S 4.1 Regulations, Codes, and Standards 4.1.1 Technical Specification Section 1.0, Definizions 4.1.2 10CFR50, Appendix B, Criterion XVI, Correztive Action 4.2 Expectations 4.2.1 DER-96-0325, ACTS Item 22418, added steps to discuss and invoke the DCM process for Operability Determinations.
4.2.2 DER-97-0680, ACTS Item 26467, revised procedure to require that a Potential LCO be entered when an Operability Determination justifies continued operation, but requires any action(s) to be taken following plant shutdown.
4.2.3 DER-98-02118, ACT-98-35976, revised procedure to require an engineering peer review be obtained prior to using a previously performed calculation as a basis for operability.
L 4.2.4 DER-00-02054, ACT-00-51185, clarified the need for both engineering supervision and the SM to challenge the scope and assumptions of Engineering Confirmations prior to making operability determination. Identified the need to clarify the concept of "Timeliness Commensurate with Safety Significance".
4.2.5 Entergy Licensing Position #2, Evaluation and Resolution of Degraded and Nonconforming Conditions. I
-e-Rev. No. 9 Page No. 8 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 5.0 DEFINITIONS b-5.1 Compensatorv Actions:
Interim measures prudently taken to improve assurance that specified safety functions are being maintained during the process from initial degraded/nonconfoming condition discovery until completion of necessary corrective actions.
Compensatory measures shall address time requirements (e.g.,
duty time of RHR is 180 days post accident) established in the current licensing basis.
5.2
Dearaded Condition (See Attachment 8):
A condition of an SSC in which there has been any loss of quality or functional capability. Examples of Degraded Conditions are:
Performance trend of an IST component that indicates an ALERT or ACTION level will be reached prior to the n e x t scheduled surveillance test Leaks external to systems (e.g., steam, water oil)
Noticeable increases in parameters that are precursors to failure (e.g., vibration, noise, temperature)
Restricted flow of cooling media or process fluid to heat exchangers High resistanca electrical contacts due to pitting, oxidation, etc.
5.3 Desicrn Basis:
Information which identifies the specific functions to be performed by an SSC, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design.
5.4 Desisn Basis Event:
Those accidents and abnormal operational transients for which safety analysis are described in Chapter 14 of the UFSAR and are part of the licensing basis for the plant.
Rev. No. 9-- Page No. 9 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-C3 -11 5.5 Desian Function:
-v' UFSAR described design bases functions and other SSC functions described in the UFSfLR that support o r impact design bases functions. Maintenance Rule functions car, be considered design functions.
5.6 Enaineerins Confirmation (See Attachments 2 , 3 an2 9 ) :
The engineering evaluation process by wnich the validit)- of the Operability Determination performed by the Shift Manager is verified.
5.7 Full Oualification:
Conformance to all aspects of the current licensing basis, including codes and standards, design criteria, and commitments.
5.7.1 Qualification is the documented verification that provides assurance that an SSC or equipment has been designed, procured, tested, installed, etc., to ensure it is capable of performing its specified function under all conditions as assumed in applicable safety analyses (e.g., UFSAR, Fire Protection Programs, etc.) .
5.7.2 Qualification constitutes conformance to all aspects of the licensing basis including licensee commitments (e.g., codes and standards referenced in the UFSAR, other UFSAR commitments, and/or corrective action commitments from LERs or NRC Notices of Violation).
5.8 Immediate Notification (See Attachment 7 ) :
Notification of the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (in some cases, within 4 , 8, or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) per the Code of Federal Regulations.
5.9 Licensinq Basis:
Documents used to grant, amend, or modify the operating license and Technical Specifications and to ensure continued compliance with, and operation within, applicable NRC requirements. Licensing Basis includes, but are not limited to, the UFSAR (including documents referenced therein),
[CTSlAP-01.04 [ITSlTechnical Requirements Manual, NRC safety evaluation reports, LERs, Generic Letters, Bulletins and similar-type docketed correspondence.
Page No. 10 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 5.10 Nonconforminu Condition (See Attachment 6):
~n adverse condition affecting a safety-related, quality-related, or trip sensitive system caused by a deficiency in characteristic, documentation, or procedure which renders the quality of an item unacceptable or indeterminate. Examples of nonconforming conditions are:
Item does not conform to design/license basis.
Recurring or generic failure.
Item has a physical defect as a result of design or manufacturing process that prevents or could have prevented the component from performing its intended function.
0 Item fails testing performed to prove environmental, seismic, or design conformance.
0 Deviation from prescribed processing or inspection.
Documentation not available to confirm required inspections or tests.
M&TE out of calibration: A Condition Report is not required when the non-conforming condition is related to the calibration of M&TE and the condition is resolved through a record search, with the determination that plant hardware or system performance is not affected and no further action is required.
Missed or late preventative maintenance task required to satisfy technical specifications, environmental qualification, or station commitments.
Conditions where nuclear fuel defects exist or are suspected.
Rev. No. 9 Page No. 11 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AD-03 .II 5.11 Operable / ODeraSilitv:
'-J CCTSI A system, subsystem, train, component, or device shall be Operable or have Operability when it is capable of performing its specified functions, and when all necessary attendant instrumentation, controls, electrical power, cooling o r seal water, lubrication or other auxiliary equipment that a r e required for the system, subsystem, train, component, or device to perform its functions are also capable of performing their related support functions. Otherwise, the system, subsystem, train, component, or device is Inoperable.
ITS1 A system, subsystem, division, component, or device sha-ll be Operable or have Operability when it is capable of performing its specified safety function(s1 and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s1.
Rev. No. 9 Page No. 12 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 5.12 OPerabilitv Determination Status:
The status of an Operability Determination. Following are the types of status:
5.12.1 Active:
The Shift Manager has made an immediate determination of Operable; however, there are conditions or future actions needed to allow the determination to be completed. An Operability Determination is considered Active and a Potential LCO is entered for the following reasons :
Actions required following plant shutdown.
Actions require monitoring plant conditions.
The Operable status is conditional based on certain factors that, if changed, may render the equipment inoperable (i.e., "Operable provided lake water temperature does not exceed 79OF" or "Operable provided leakage does not exceed 5.0 GPM").
Compensatory measures needed to justify continued Operable status.
The Operable status is based on compensatory measures that are taken to ensure the equipment can be considered Operable; however, actions are being taken to restore the original design. Measures may include manual operator actions, where allowed, or establishment of administrative controls.
The Shift Manager has declared equipment Operable with a Reasonable Expectation of Operability (REO), and an Engineering Confirmation is required to develop a final basis for a completed status.
5.12.2 ComDleted:
The Shift Manager, based on clear, confirmed evidence sufficient to justify the determination,
, has made an equipment determination of Operable or Inoperable. A completed status means there are no conditional bases or compensatory measures required, and the Engineering Confirmation in support of a previous REO is sufficient to now complete the determination. ,
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Rev. No. - 9 Page No. 13 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AD-03.11 5.13 Reasonable Expectation of ODerabilitv (REO):
Technical judgment, coupled with the safety significance of a degraded or nonconforming condition, which gives a reasonable indication that an SSC is capable of performing its specifiecl functions. This constitutes a prompt determination of Operability, but indicates that further evaluation is required in support of Operability.
5.14 Reportable Event (See Attachment 7):
Event that requires a notification to the Nuclear Regulatory Commission or other regulatory bodies or a Licensee Event Report (LER) to the NRC per license requirements and 10CFR parts 20, 21, 26, 50, 50.72, 50.73, 70, 72, and 73.
5.15 Safety Function:
Those functions required to prevent the unacceptable consequences for design basis events. The unacceptable consequences for design basis events are loss of:
The integrity of the reactor coolant pressure boundary.
The capability to shut down the reactor and maintain it in a safe shutdown condition.
The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 10CFR100.
Rev. No. 9 Page No. 14 of 78
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OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 5.16 Systems, Structures, or Components ( s s c s ) : \/'
Systems, structures, or components, which are:
5.16.1 Safety-related and relied upon to remain functional during and following design basis events to:
A. Ensure the integrity of the reactor coolant pressure boundary.
B. Ensure the capability to shut down the reactor and maintain it in a safe shutdown condition.
C. Ensure the capability to prevent or mitigate the consequences of accidents that could result
~~~- in potential offsite consequences comparable to the 10CFR100 guidelines.
5.16.2 Relied on in the safety analyses or plant evaluations that are a part of the plants current licensing basis. Such analyses and evaluations include those submitted to support license amendment requests, exemption requests, or relief requests, and those submitted to demonstrate compliance with the NRC's regulations. 'J 5.16.3 Subject to 10CFR50, Appendix A (Criteria 1) or 10CFR50, Appendix B.
5.16..4 Subject to Technical Specificaticns, either explicitly or through the definition of operability, such as supporting systems.
5.16.5 Described in the Updated Final Safety Analysis Report (UFSAR).
lEXP4.2.4 5.17 Timeliness Commensurate with Safety Siqnificance (See Attachment 8 1 :
Operability evaluations must be conducted in a time frame consistent with the safety significance of the identified condition. An evaluation should be completed within the established LCO period, if the SSC has an associated LCO.
When there is no specified LCO, the safety significance should be considered based on the risk impact. Insights as to risk impact can be obtained from Planning or the PRA Group in WPO Reactor Engineering.
d Rev. No. 9 Page No. 15 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AD-03.11 5.18 UFSAR Specified Functions:
Those functions that the SSC performs that are identified t h e UFSAR. This is not limited to safety related functions.
System-specific functions are summarized in MCX-6A. The UFSAR sections referenced provide additional informazio3 that should be consulted to identify UFSAR specified funcciocs.
Rev. No. 9 -- Page No. 16 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.II 6.0 RESPONSIBILITIES 6.1 All Personnel Identify equipment of questionable conformance /
qualification as it becomes known by initiating a CR or PID and communicating this information to the Shift Manager.
6.2 General Manager Plant Operations ( G M P O )
6.2.1 Review Initial Engineering Confirmations whenever any compensatory actions are required.
6.2.2 Review Detailed Engineering Confirmations.
lEXP4.2.4 6.3 Engineering Managers 6.3.1 Review Initial Engineering Confirmations whenever any compensatory actions are required.
6.3.2 Review Detailed Engineering Confirmations.
6.4 Operations Manager NOTE : For the purpose of this procedure, Operations Manager is considered to be the Operations Manager, the Assistant Operations Manager, or a designated SRO.
6.4.1 Assist the Shift Manager in determining SSC operability.
6.4.2 Review completed Engineering Confirmations to determine operability of affected S S C s .
6.4.3 Ensure Active status Operability Determinations are presented for PORC review.
6.4.4 Provide operability and reportability review of CRs, as applicable.
6.4.5 Concur with operability and reportability determination of CRs.
Rev. No. 9 Page No. 17 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS A?-03.1:
6.5 Shift Manager (SM) 6.5.1 Monitor the operational readiness of S S C s important to the safe operation of the facllizy.
6.5.2 Determine the operability of S S C s based 02 the information available.
6.5.3 Determine Reasonable Expectation of Operability.
Notify Operations and System Engineering Manager of status and request Engineering Confirmations when needed.
6.5.4 Track active status Operability Determinations as Potential LCOs.
6.5.5 Ensure appropriate compensatory actions are taken when SSCs important to safety are degraded or Inoperable.
6.5.6 Review CRs in the SM PCRS Inbox to ensure operability and immediate reportability requirements are met.
6.5.7 Ensure notifications are made.
L 6.6 Director Safety Assurance Make notifications to NRC when on site during day shift.
6.7 System Engineering Manager 6.7.1 Assign responsibility for preparation of Engineering Confirmations to the appropriate Engineering Supervisor on site or in WPO.
6.7.2 Ensure the Engineering Supervisor has sufficient resources to perform the Engineering Confirmation as outlined in this procedure.
6.8 Engineering Supervisor 6.8.1 Provide the System Engineering Manager with results of assigned Engineering Confirmations in a timely manner.
IEXP4.2.1.
6.8.2 Ensure calculations and analyses to support Operability Determinations are performed per EN"-DC-126.
Rev. No. 9 Page No. 18 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 6.8.3 Keep the System Engineering Manager, Shifc Manager, v and NRC Resident Inspector informed of any changes to the established schedule or the results of the Engineering Confirmation, as appropriate.
6.8.4 Assign independent reviewers to assess the adequacy of the Engineering Confirmation scope, logic, and supporting technical analysis.
6.9 Plant Operating Review Committee (PORC) 6.9.1 Review Active status Operability Determinations regularly.
6.9.2 Review Initial Engineering Confirmations whenever any compensatory actions are required.
6.9.3 Review Detailed Engineering Confirmations.
6.10 Condition Report (CR) Screening Committee Review CRs and PIDs for operability and immediate reportability concerns.
6.11 Regulatory Compliance Manager Provide reportability review of CRs which are -potentially- -4 reportable. I v
Rev. No. - 9 Page No. 19 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-C3.11 7.0 SPECIAL INSTRUCTIONS
-u 7.1 A condition having an immediate impact on plant safety, involves a fire or an uncontrolled release of radioaczi\7iz>7, or poses a threat to security shall be immediately reporre?
to the Shift Manager.
7.2 The Shift Manager prepares Operability Determinations and Immediate Reportability Determinations as required and obtains assistance from other departments as necessary.
I 7.2.1 Operability and Immediate Reportability determinations and reviews are performed in accordance with ENN-LI-102, EAp-1.1, JAF Corrective Action Process Desk Guide, and this procedure.
7.2.2 SMS/SROS may issue CAS to engineering groups for support of Operability Determinations or Evaluations without the concurrence of the responsibie individual or group.
7.3 The Shift Manager ensures component and system operability continually by surveillances and formal determinations.
Operability Determinations are supplemented by:
Day-to-day facility operation
-v-Implementation of ISI/IST programs Plant walkdowns and tours Control room observations QA audits and reviews Engineering design reviews 7.4 Questions or concerns relating to SSC qualification are resolved using CRs per ENN-LI-102.
7.5 When an Operability Determination or Immediate Reportability Determination is "completed,"it is considered closed and any associated corrective actions required to restore a degraded or non-conforming condition are tracked to completion using the other processes for such actions, including work requests per AP-10.01 or an analysis or evaluation per ENN-LI-102 and the J A F Corrective Action Process Desk Guide.
I 7.6 Electronic forms are acceptable for use, as long as the forms contain all the necessary information.
L-Rev. No. Page No. 20 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 7.7 The following attachments provide information and flowcharts that are helpful in understanding the requirements of this procedure. However, the flowcharts do not contain a sufficient level of detail to fully implement the procedure.
Attachment 1 provides the form and guidelines for operability determination by the SM.
Attachment 2 provides the Engineering Confirmation Summary Form.
Attachment 3 provides Initial Engineering Confirmation guidelines.
Attachment 4 provides Detailed Engineering Confirmation guidelines.
Attachment 5 provides REO/Reasonable Assurance guidelines.
Attachment 6 provides REO/Engineering Operability guidelines.
Attachment 7 provides an Immediate Reportability Checklist.
Attachment 8 provides the Entergy Licensing Position regarding Evaluation and Resolution of Degraded and Nonconforming Conditions.
I Attachment 9 contains a flowchart overview of the Engineering Confirmation process.
Attachment 10 contains a flowchart overview of the Operability Determination process.
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'e-Rev. No. Page No. 21 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-C3.11
-- 8.0 8.1 PROCEDURE G e n e r a l Requirements 8.1.1 The Shift Manager shall ensure S S C s are considerec!
either Operable or Inoperable at all times. Ar, REO determination means an SSC is considered operable.
8.1.2 If the operability of an SSC is questionable, the Shift Manager shall make an Operability Determination and an Immediate Reportability Determination, with immediate and primary attention directed to safety concerns.
I 8.1.3 Operability Determinations and Immediate Reportability Determinations shall be made promptly, with a timeliness commensurate with the potential safety significance of the issue. Attachment 8 provides guidance for understanding the issue of timeliness.
8.1.4 As requested by the Shift Manager, others may assist in the Operability and Immediate Reportability determinations; however, the final determinations I
are made by the Shift Manager.
bEXP4.2.5 8.1.5 The Shift Manager shall review the following conditions or events which may require an Operability Determination and/or an Immediate Reportability Determination:
A. Degraded equipment condition where performance is called into question.
B. Nonconforming condition where equipment qualification is called into question, such as, non-safety related parts found in a safety related application.
C. Fitness for Duty events may require an Immediate Reportability Determination.
v Rev. No. 9 Page No. 22 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 8.2 Operability and Immediate Reportability Reviews 8.2.1 The Shift Manager shall review all CRs that are flagged as potentially affecting operability or as potentially reportable by the CR initiator. (See Attachment 1 for directions for Operability Determinations and Attachment 7 for Immediate Reportability Determinations.)
8.2.2 The CR Screening Committee shall review CRs for potential operability and potential immediate reportability considerations.
8.2.3 Operability Determinations are performed and documented in accordance with this procedure. A summary of the determination should be entered into the Operability tab in PCRS. I v
Rev. No. 2 Page No. 23 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03-11 8.3 Operability Determinations and Related Actions 8.3.1 Operability Determinations NOTE: If a deficiency is documented on both a PID and CR, only one Operability Determination is required. Preferably, a CR is used.
A. The Shift Manager shall initiate an Operability Determination using Attachment 1 for C R s and for PIDs identified as corrective maintenance, for I
the following '(Refer to Step 5.16):
SSCs which are QA Category I or M.
SSCs which are not QA Category I or M but have operability requirements specified in Technical Specifications or [CTSJAP-01-04
[ITS]Technical Requirements Manual -
SSCs described in the U F S A R that support the operability of SSCs which have operability requirements specified in Technical Specifications or ECTSJAP-01.04
[ITSITechnical Requirements Manual, B. The Shift Manager shall initiate an Immediate Reportability Determination using Attachment 7 for CRs classified as a potential operability concern.
C. The Shift Manager should consider initiating an Operability Determination if the condition involves :
Fire Protection Program requirements Seismic/Equipment Qualification Emergency Operating Procedures Accident Analysis Refuel Operations Surveillance Test Procedures Radwaste Effluent Controls Equipment that could result in a plant trip Rev. No. 9-- Page No. 24 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03 .I1 8.3.1 (cont)
D. Track each Operability Determination based on the --
associated CR or PID number. Active status determinations are numbered and tracked to completion as Potential LCOS per $?*
E.;$: T . b*
NOTE: An SSC with discrepant conditions can be considered functional or operable until available information indicates the SSC is nonfunctional or inoperable.
E. In the absence of reasonable assurance that an SSC is operable, or if evidence suggests that final evaluation will conclude the SSC can not perform its specified design or safety functions, the Shift Manager shall declare the SSC Inoperable.
F. If Engineering assistance is required to complete an Operability Determination, and the S h i d t Manager has a reasonable expectation of operability, proceed as follows:
- 1. Complete Attachment 1 as an REO.
Use Subsection 8.3.2 and Attachments 5 and 6 for guidance. _A Track as a Potential LCO per B C S C 1,
- 2. Notify the Operations Manager and Systerr.
Engineering Manager, or designees, immediately.
G. Initiate any required compensatory or corrective actions. Equipment used in compensatory actions should be controlled per AP-12.01 and AP-12.06, using the guidance in Attachment 8.
H. Document the justification for Operability Determinations on Attachment 1.
Rev. No. 9-- Page No. 25 of 78
OPERABILITY AND REPORTABILITY DETERMINATfUNS L
8.3.1 (cont) a EXP4.2.2 I:. If the Operability Determination justifies continued operation, but requires any of :he I following, enter a Potential LCO per AP-12.08, mark Attachment 1 in this procedure as Active, and record PLCO number on Attachment 1:
- 1. Actions to be taken following plant shutdown.
- 2. Conditional Operability Determinations requiring monitoring of plant conditions.
Refer to the section in Attachment 6 which provides guidance on Contingent Operator Actions.
- 3. Compensatory measures taken to justify continued equipment operable status.
J. When cornpensatom measures are used to declare SSCs Operable, the Shift Manager shall ensure a 50.59 Screen Control Form is completed per ENN-LI-101, using the following guidance and the information in Attachment 8 :
- 1. The screen may be completed as part of a procedure change or temporary modification.
- 2. The screen should focus on the effects of the compensatory measures, not the degraded condition.
- 3. The screen should be completed and reviewed prior to implementing of such measures.
- 4. If a 10CFR50.59 Evaluation is required, the Shift Manager should re-evaluate the original identified deficiency per Steps 8 . 3 . 1 . D and 8.3.1.E.
Rev. No. 9 Page No. 26 of 78
c-OPERABILITY AND RE PORTAEI L 3rY I X T m N A T IONS AP-03.12 K. The Shift Manager shall ensuzre m r a b i l i t y Determinations are sufficiemt to, address SSC capability to perform its safety functions.
Determinations may include:
0 Determining safety functions performed by the SSC by reviewing Technical W c i f ications, ICTS]AP-01-04 [ITS]Tec-can Requirements Manual, MCM-6 A Reference Dement (MCM-SA-REF), UFSAR, and De- Basis DOC-tS 0 D e t e d n i n g circumstances c f the non-camformance, includiing - s a l e failure mechaznism-0 D e t e d n i m g requirement established for the equipment and why the r e q u i m e n t may not be met.
0 D e t e d n i m g the safest plant: configuration, including effects of transition& actions.
bKypE: Successful performance of Teu3mical -
SpeciSication surveillamce requirements alone is usxzally not sufficiemt ta, determine operability when conformance? to t h e appropriate criteria in the aznrreFt licensed desigm basis is in question.
L. Describe Basis for conclusion on~Attachment 1.
The Shift Manager should u s e the! following methods to make an operability h i s i o n :
A t e s t , partial test or 0th- functional demonstration Analysis 0 past experience with operatiug events for this SSC Engineering judgment M. Probabilistic Risk Assessment ( P m ) s r probabilities of the occurrence cof accidents or events shall not be used to deteamine Operabi 1it y .
N. If, during the operability review, a CR is required, the Shift Manager shall generate a CR per ENN-LI-102 and the JAF Correaztive Action I Process D e s k Guide.
Rev. No. 9 Page Bo. 27 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AB-c'3. ;1 8.3.1 (cont)
- 0. m e n a (item) system, subsystem, t r a i r , , cor?oner?Z or device addressed in a Technical Specification LCO becomes inoperable:
- 1. Verify operability of redunaact counzer?arts.
- 2. Verify affect of inoperability on supported or supporting items.
P. Enter the results of the initial Operability Determination on a new operability record from the Operability tab in PCRS. If there are no reportability issues, the SM can sign tne PCRS record.
Rev. No. 9 Page No. 28 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.II 8.3.2 Reasonable Expectation of Operability (REO) .__,-
A. If the impact on the associated SSCs is not apparent, or the determination requires Engineering input, then the SM (or SRO/STA) shall I perform an REO evaluation.
- 1. If calculations, vendor information, etc.,
can not be obtained to substantiate the Operability Determination, then the SM (or SRO/STA), with assistance from the appropriate engineering department (System Engineering, Design Engineering, Component/Programs Engineering, etc.), will make the operability evaluation using best technical judgment.
- 2. Engineering departments that provided documented assistance should attach those documents to the assessment.
- 3. Document the bases for the REO.
- 4. The time of entry into the Engineering Confirmation process that follows the REO will be indicated.
- 5. The basis should include a statement regarding the capability of the equipment/system/train being evaluated to perform its UFSAR Specified Functions.
- 6. The Shift Manager shall either perform or review the operability evaluation bases.
B. Guidance contained in Attachments 5 and 6 should be used during performance of the REO.
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Rev. No. 9 Page No. 29 of 78
OPERABILITY AND REPORTABILITY DETERMINA?'IC)NS AP-C3. -_
- 9
'v' 8.3.3 Engineering Confirmation NOTE: Attachment 9 provides a flowchart f o r the Engineering Confirmation process and should be used as a guide for this subseczioc.
A. The responsible engineer should develop tne Initial Engineering Confirmation per Attachment 3.
- 1. If the Initial Engineering Confirmation w i l l not be completed as scheduled, then the Engineering Supervisor shall obtain,approva; for an extension from the SM. Notify the System Engineering Manager.
B. If at any time during the development of the Engineering Confirmation, it is determined, through evaluation or engineering judgmer?t, that the SSC is not Operable, then notify the SM immediately.
C. If the SSC required by Technical Specifications is determined to be Inoperable, then the SM shall comply with the Technical Specification requirements.
- 1. If necessary, the SM should consult with the GMPO, Engineering Manager(s) , and Regulatory Compliance Manager to determine if an Emergency Technical Specification Amendment or a Notice of Enforcement Discretion is appropriate.
D. If SSCs important to safety other than that required by Technical Specifications are determined to be Inoperable, then the Initial Engineering Confirmation should address if continued operation is recommended, with appropriate justification and recommendations.
- 1. The Initial Engineering Confirmation should consider administrative controls or other compensatory actions that can be taken.
- 2. If compensatory actions are recommended, perform a Process Applicability Determination per E"-LI-100 and a 50.59 Screen Control Form per ENN-LI-101. Refer to the Section in Attachment 6 which provides guidance on Contingent Operator Action(s).
Rev. No. 9 Page No. 30 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03 .I1 8.3.3 (cont) --
E. The Engineering Supervisor s h o u l d keep the System Engineering Manager, SM, and NRC Resident Inspector informed of any change or significant development as the Initial Engineerirng Evaluation is processed.
F. If a Detailed Engineering Confinmatiam is neeaed, then it should be identified in the Initial Engineering Confirmation.
- 1. The responsible engineer willl determine the approach, scope, and responsibilities for the Detailed Engineering Confirmatiom- The schedule for completion of the Detailed Engineering Confirmation s h o d d b e approved by the GMPO and Engineering m a g l e r ( S I .
- 2. The Detailed Engineering Conffirmation should be completed per the guidelimes in Attachment 4 and attached to the CR.
- 3. The Engineering Supervisor s b u l c l l notify t h e SM, System Engineering Manager, w i n e e r i n g Manager (s) and GMPO when the D e w led Engineering Confirmation is a o m p l l e t e d .
G. PORC and the -PO should review tthe following:
Initial Engineering ConfirnatLions; whenever any compensatory actions are required A l l Detailed Engineering Conffimmtions H. The responsible engineer shall emsure the Engineering Cornfirmation (InitiaR or Detailed) is provided to the SM for review and acazptance-
- 1. After the S M accepts the E n g k e e a g Confirmation, the date and t-e af t k Engineering Confirmation and exitkng of the AP-03.11 process shall be entered i n t o the station logs.
1EXP4 . 2 .1 I. Calculations or analyses required to support the completed Operability Deteminatiian s h l l be prepared per ERIN-DC-126.
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Rev. No. 9 Page No. 31 of 78
OPERABILITY AND REPORTAI~ILITYDETERMINATIONS A?-@3.11 8.3.3 (cont) lEXP4.2.3 J. If a previously performed calculaticn is use2 as a basis for operability, obtain an engineerins peer review to verify applicability of the calculation.
R. If an Engineering Confirmation is performe2 to determine system operability of through-wall leaks within the IS1 boundary, ensure the flaw evaluation is performed using NRC-GL-90-05 methodology.
L. Engineering Confirmations shall be documented using Attachment 2, following the guidance in Attachments 3 and 4, as appropriate. In addition, the guidance in NRC Inspection Manual (Part 9900) and Generic Letter 91-18, Revision 1 ,
should be used.
- i. The preparing engineer signs and dates the completed evaluation.
- 2. The independent reviewer sign and date the completed evaluation.
- 3. Independent review is performed by individuals not involved with preparing the evaluation and consists of assessing the adequacy of the Engineering Confirmation scope, logic, and supporting technical analysis.
- 4. The Engineering Supervisor reviews and signs the evaluation.
lEXP4.2.4 The intent of this review is to challenge the scope and assumptions of the evaluation prior to submitting to the SM for review and acceptance.
M. Probabilistic Risk Assessment (PRA) or probabilities of the occurrence of accidents or events shall not be used to determine
. Operability.
Rev. No. 9 Page No. 32 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 8.3.3 (cont) bEXP4.2.4 N. Completed evaluations shall be forwarded to the I
Shift Manager for review. The intent of this review is to challenge the scope and assumptions of the evaluation prior to making an operability determination.
Declare equipment Operable on Attachment 1.
Mark Attachment 1 as Complete.
Declare equipment Inoperable on Attachment 1 and initiate required actions for declaration of Inoperable.
Mark Attachment 1 as Complete.
- 3. If the evaluation supports Operable with compensatory measures , then the SM performs the following:
Declares equipment Operable on Attachment 1 and initiates required actions for declaration of Operable.
Marks Attachment 1 as Active.
Rev. No. 9
-- Page No. 33 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS A P - 0 3 -11 8.3.4 Revisions to Active Operability Determinations
'W' A. If active Operability Determinations tor. .
Engineering Confirmations) rewire r e v i s i o n before being completed, the Engineering Supervisor presents the required cnanges to tne SM for acceptance. The SM o r Engineering Supervisor (as appropriate) may perform either of the following:
- 1. Update the original documents and initial changes made.
- 2. Complete a new Operability Determination (01 Engineering Confirmation) and attach the superseded documents.
- 3. If needed, another operability record can be generated from the Operability tab in PCRS.
Rev. No. 9 Page No. 34 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 8.4 Immediate Reportability Determinations 8.4.1 Immediate Reportability Determinations are performed and documented in accordance with Attachment 7 and JAF Corrective Action Process Desk Guide. A summary of the determination should be entered into the Operability tab in PCRS. Refer to NUREG-1022, Event Reporting Guidelines 10CFR50.72 and 10CFR50.73, for additional guidance on reportability determinations.
8.4.2 If an event is determined to require an Immediate Report, Operations shall complete the Event Notification Worksheet, NRC Form 361 (a copy exists in EAP-1.1) and make the proper notifications in accordance with EAP-1.1.
8.4.3 For Independent Spent Fuel Storage Installation (ISFSI) events that are reportable under 10CFR72.75, to the extent that the information is available at the time of notification, the Event Notification Worksheet, NRC Form 361, (a copy exists in EAP-1.1) shall include a description of the quantities and physical forms of the spent fuel or high level waste involved.
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Rev. No. 9 Page No. 35 of 78
OPERABILITY AND REPORTABILXTY CETERMINATIOUS AP 1 1 I 8.5 ManagamePt Review of Operability L
1 EXPI. 2.5 8.5.1 The Engineering Xanager(s) s h a l l review Initial Engineering C o n f h a t i o n s (whenever compensatory actions are zequiired) and Detailed Engineering Confirmations to ensure timeliness is commensurate with safety signiificance and technical adequacy is consistent with nranagement expectations.
(IEXP4.2.5 8.5.2 The GMPO shall r d e v Initial Engineering Confirmations ( w m e v e r compensatory actions are required) and DeUailed Engineering Confirmations to ensure timeliness is commensurate with safety significance and technical adequacy is consistent with management expectations-8.5.3 T h e Operations Mamager shall ensure active status Operability Detenminations are presented f o r PORC review during regularly scheduled POHC meetings per AP-01.01.
L Rev. No. 9 Page No. 36 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.1:
8.6 Records .-
8.6.1 Completed Operability Determinations are considered Quality Records.
I A. Operability Determinations for CRs are retained. 1 B. Operability Determinations for PIDs are forwarded to the Operations Department General Clerk for retention as required by AP-02.08.
8.6.2 Active Operability Determinations, including REOs, being tracked as Potential LCOs, are retained with I AP-12.08 documentation until completed.
8.6.3 Engineering Confirmations are retained with the completed Operability Determination.
Rev. No. 9 Page No. 37 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-L73.11 9.0 ATTAC-S
- 1. OPERABILITY DETERMINATION FORM
- 2. ENGINEERING CONFIRMATION
SUMMARY
FORM
- 3. INITIAL ENGINEERING CONFIRMATION GUIDELINES
- 4. DETAILED ENGINEERING CONFIRMATION GUIDELINES
- 5. REO/REASONABLE ASSURANCE GUIDELINES
- 6. REO/ENGINEERING OPERABILITY GUIDELINES
- 7. IMMEDIATE REPORTABILITY CHECKLIST
- 8. ENTERGY LICENSING POSITION (EVALUATION AND RESOLUTION OF DEGRADED AND NONCONFORMING CONDITIONS)
- 9. ENGINEERING CONFIRMATION PROCESS FLOWCHART
- 10. OPERABILITY DETERMINATION PROCESS FLOWCHART Rev. No. 9 Page No. 38 of 78
OPERABILITY DETERMINATION FORM Page 1 of 3 0 ACTIVE (PLCO # 1 0 COMPLETE Describe potentially degraded or non-conforming equipment/syscems:
Describe UFSAR Specified Functions:
Effect of Condition on Owerable InoDerable REO UFSAR Specified Function Equipment 0 D O D Train 0 0 0 5 Function D D O If a x are OPERABLE, exit AP-03.11 Basis :
Review: Shift Manager If are
~ Q Y INOPERABLE, enter LCO. LCO No.
Basis :
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(Operability of redundant, supported, or supporting items considered.)
Review: Shift Manager Zontinued on next page THIS IS A QUALITY RECORD AP-03.11 OPERABILITY AND REPORTABILITY ATTACHMENT 1 Rev. No. 9 DETERMINATIONS Page 39 of 78
Basis :
Engineering CoEfi-mation: Requested:
Dare,Time Interin Review: Shift Manager:
Engineering Confirmation: Compieted:
Date,Time
(-) Operabie with no Compensatory Actions, exit AP-03.11
(-) Inoperable, enter LCO per ODSO-34 LCO No.
(-) Operable with the following Compensatory Actions, enter a Potential LCO per ODSO-34 and complete required 10CFR50 5 9 reviews :
(-1 Actions required following plant shutdown.
(-1 Actions require monitoring plant conditions.
(-) Compensatory actions taken to justify continued operable status.
LCO No. 10CFR50.59 reviews complete:
Date/Time Active Operability Determination PORC review: PORC Meeting #
Review: Shift Manager THIS IS A QUALITY RECORD AP-03-11 OPEFABILITY AND REPORTABILITY ATTACHMENT 1 Rev. No. 9 DETERMINATIONS Page 40 of 7 8
Operability Determination Guidance
- 1. Review the following documents to determine the UFSAR Specified Functions performed by SSCs identified as potentially degraded o r i non-conforming:
JAF Technical Specifications a [CTSJAP-01.04 , Technical Specification Related Requirements, Lists, and Tables
[ITS]Technical Requirements Manual MCM-6A Reference Document JAF Updated Final Safety Analysis Report JAF Design Basis Documents
- 2. Consider the following questions when performing Operability Determinations:
Will the SSC(s) be prevented from performing the design function(s)?
Could the problem affect the operability of a Technical Specification required SSC?
Could the capability of an SSC to prevent or mitigate consequences of an accident as postulated or described in the UFSAR be reduced?
Could the condition result in an SSC not meeting known design requirements contained in design documents?
Does the problem involve an INOPERABLE non-Technical Specification SSC that could functionally affect a Technical Specification S S C s ability to perform its design function?
Could the problem have adverse safety significance requiring prompt review or correction?
Could single failure design criteria have been defeated?
, 1 U
AP-03.11 OPERABILITY AND REPORTABILITY ATTACHMENT 1 Rev. No. 9 DETERMINATIONS Page 41 of 78
PID# CR# Potehtial LCO#
L .
Discrepancy is against: (Reference Document Number and Section)
Design Basis Document -
Test Results Drawing or Spec Commitment # UFSAR or Tech Specs sscs [CTSJAP-O1.04 [ITSJTRM Other (NSE, Procedure, etc)
Are there other affected SSCs? NO YES, Describe:
( O t h e r than identified on the REO)
Evaluation SUPPORTS Operability Evaluation DOES NOT SUPPORT Operability (Notify SMI Discuss logical and defensible basis for conclusion:
List supporting references: (Anaiysis, d r a w i n g s , N S E , S p e c s , etc. I
- Attached Print/Sisn Date Preparer 1 Independent Reviewer /
Engineering Supervisor /
Engineering Manager /
(or designee), if applicable GMPO (or designee), /
if applicable
- SEND COMPLETED FORM TO THE SHIFT MANAGER -
THIS IS A QUALITY RECORD AP-03.11 OPERABILITY AND REPORTABILITY ATTACHMENT 2 iu Rev. No. 9 DETERMINATIONS Page 42 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 ATTAC"T 3 Page 1 of 5 INITIAL ENGINEERING CONFIRMATION GUIDELINES - /'
PART 1 - GENERAL GUIDELINES The Initoial Engineering Confirmation is an evaluation to determine if the equipment in question is capable of performing its specified design functions.
A recommendation of Operability should be provided.
- 1. The evaluation is directed toward gathering information to confirm the operability of the equipment based on analysis, test or partial test, operating experience and technical judgment. The evaluation should conclude that reasonable assurance (refer to Attachment 5 ) does or does not exist that the equipment will perform its design function until corrective action and/or further investigation can be completed.
- 2. The magnitude of nonconforming/degraded condition should be noted for consideration. If technical judgment determines that the nonconforming/degraded condition in question has no impact on the design function, the equipment should remain operable.
- 3. A visual examination of the nonconforming/degraded equipment should be made. Any notable comparisons with similar conforming/qualified equipment should be made.
- 4. If a clearly physical problem is the basis for the nonconforming/ degraded condition, it should be so indicated. Any immediate corrective actions, such as temporary braces and/or other alternatives or "fixes" that can be quickly used to provide reasonable assurance that the equipment will --
function until corrective action can be completed should be indicated.
- 5. The evaluation should give some indication of the safety significance of the nonconforming/degraded equipment. The evaluation should conclude if the equipment is or is required to perform the design function. This includes any support function to any equipment required by Technical Specifications.
- 6. The evaluation should be independently reviewed. The independent reviewer should have had minimal involvement in the evaluation preparation. The independent reviewer's signature signifies concurrence with the evaluation and that the scope, logic, and supporting technical analysis are adequate.
Obtain and incorporate the Shift Manager's input prior to completing the independent review.
- 7. If the Operability is based on the use or availability of other equipment (e.g., LER-00-016-01), it must be verified that the equipment is capable of performing the function utilized in the evaluation (i-e.,functional testing completed, visual inspection, etc.), if plant conditions allow.
This would include any equipment used for contingency and/or having administrative controls placed on them.
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Rev. No. Page No. 43 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-23.11 ATTAC"T 3 Page 2 of 5 INITIAL ENGINEERING CONFIRWTION GUIDELINES PART 2 - FORMAT Use the following format for documentation of the Initial Engineering Confirmation.
- 1. Summary Statements Succinctly state the nonconforming/degraded condition in clear, concise terminology. Summarize the results of the evaluation, succinctly staring the operability recommendation. Underline the recommendation statement.
- 2. References List all procedures, specifications, standards, codes, calculations, drawings, regulatory documents, etc., including revision numbers that were used in the evaluation.
- 3. Detailed Problem Statements Clearly identify and discuss each item of nonconforming/degraded condition.
Describe the design function performed by the equipment.
Describe any background of events leading to the nonconforming/degraded condition, include times, dates, documents, personnel, etc. involved with related circumstances.
%L-'
Describe by what means and when the potential nonconforming/degraded condition was discovered.
Describe the failure mechanism.
If appropriate, provide a subject background summary of why the equipment/component was designed for the original application/function, i.e., summary of pertinent design basis, including any abnormality/deviation allowances of which the evaluator may be aware.
u Rev. No. 9 Page No. 44 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03. il ATTAC"T 3 Page 3 of 5 INITIAL ENGINEERING CO"ATION GUIDELINES -2
- 4. Assumptions Specifically state all assumptions made in the engineering evaluation.
- 5. Engineering Evaluation Provide an evaluation for each item in the detailed problem statements.
The evaluation summary should clearly indicate if the component can perform its design function and the basis thereof.
Describe the applicable commitments and design requirements, i.e., 10CFR, IEEE, ANSI, ASME, etc., and why they may or may be met.
If walkdowns or inspections were conducted, details should be provided here or referenced in the attachment section, including names, dates, criteria and specific results.
Describe the basis for recommending the systems operable (i.e., analysis, test or partial test, operator experience or technical judgment).
If it is determined that the nonconforming/degraded condition is OPERABLE but outside of the existing licensing basis, corrective action must either restore the nonconforming/degraded condition to the existing licensing basis, or the licensing basis must be revised to envelope the evaluated condition.
- 6. Impact on Nuclear Safety Provide a description of the impact of the nonconforming/degraded condition on nuclear safety, include an evaluation of what other equipment could be affected by a failure of the equipment determined to be inoTerable.
Specifically:
Assess the potential impact on accident response.
Include the effects of any short-term (immediate) actions in the impact assessment.
For inoperable equipment, include an evaluation of the failure effects.
Address immediate effects from the existing condition as well as possible effects from related failure.
Include the likelihood of failure.
Rev. No. 9 Page No. 45 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AF-23 -11 ATTACHMENT 3 Page 4 of 5 INITIAL ENGINEERING CONFIRMATION GUIDELINES
- 7. Immediate Actions Describe/recommend any immediate actions or alternatives, which can be taken or are needed to quickly provide reasonable assurance that the equipment in question will function until long-term correccive accion car, be completed (e.g., a temporary seismic support, temporary use of an installed spare, etc.).
If any restrictions or limitations (such as temperature, pressure, etc.)
are placed on the OPERABILITY verification, these must be clearly stated and identified for operating the plant. Provide an estimated completion date for these actions if possible.
- 8. Long Term Actions In some cases it may be possible to identify the appropriate long-term corrective action. If so, describe this and provide the status or schedule if available. As with all 10CFR50 Appendix B conditions adverse to quality, the schedule for corrective actions should be commensurate with importance to safety of the nonconforming/degraded condition. Also, identify if any further detailed engineering evaluation is required.
Describe the aspects that need further investigation. If possible, provide an estimated completion date.
If Long Term Corrective Action was previously planned for other reason(s1, then revise action (WR, CA, etc.) to reference this CR and/or Engineering L- Confirmation. Such revision provides linkage to prevent cancellation or deferral without proper review.
- 9. Signatures Provide the signatures of all preparers, independent reviewers, an Engineering Supervisor, System Engineering Manager, Design Engineering Manager (as appropriate), PORC meeting number (if necessary), and SM (concurrence). The SM shall indicate the date and time of his signature as the official time of the Operability Confirmation.
- 10. Attachments Provide any attachments necessary to substantiate the evaluation.
NOTE: If time does allow for the validation process to be completed prior to issuance of the evaluation, then the following statement (or equivalent) should be included in the evaluation:
"Our internal verification process is not yet complete for this response. The verification process will be completed as part of the Detailed Engineering Confirmation."
U' Rev. No. 9 Page No. 46 of 78
OPERABILITY ANI) REPORTABILITY DETERMINATIONS AD-03 -11 ATTACHME 3 Page 5 of 5 INITIAL ENGINEERING CONFIRMATION GUIDELINES --J GUIDELINES FOR CONFIRMATION OF OPERABILITY NOTE: The measure of reasonableshould be commensurate with importance to safety of the nonconforming/degraded condition and the magnitude of the nonconforming/degraded uncertainty problem (refer to Attachment 5 ) .
- 1. When reasonable technical judgment indicates that the nonconforming/
degraded condition is capable of performing its intended design function when required, the equipment should be declared operable.
- a. If there is reasonable assurance that the equipment is capable of performing its specified design function, and that the confirmation process will support this expectation, but there are some remaining concerns or uncertainties, the equipment can remain operable until further evaluation can resolve the concerns.
- b. If the initial engineering evaluation indicates that it can be shown that the nonconforming/degraded condition in question is irrelevant to the design function of the equipment, the equipment should remain operable.
- 2. When reasonable technical judgment indicates that the nonconforming/
degraded equipment is not capable of performing its specified design ----
function when required, the equipment should be declared inoperable.
- a. For inoperable equipment in a system not covered by the Technical Specifications, reactor operation may continue if the design function can be accomplished by other designated equipmer,t that is qualified, or if limited administrative controls can be used to ensure the design function is met.
- b. For a system covered by Technical Specifications that is capable of performing its specified function with an inoperable support system that is not covered by Technical Specifications, no additional action outside of restoring the inoperable support system is needed.
-u R e v . No. 9 Page No. 47 of 78
OPERABILITY AND REPORTABILITY DETEmINATIONS AF-c!3
- _- _-
ATTACHMENT 4 Page 1 of 1 U'
DETAILED ENGINEERING CONFIRMATION GUIDELINES NOTE: In most cases, especially when the equipment is quickly repaired or replaced, a Detailed Engineering Evaluation may not be needed.
- 1. The Detailed Engineering Confi&?nation should be a more rigorous analysis beyond the initial "engineering judgment" evaluation and typically includes more rigorous analysis if this can be done within the time allotted.
- 2. The evaluation may consider available test data to determine wnether the test conditions envelope equipment design conditions. The design conditions should ensure that the equipment will perform its design functions when called upon to mitigate the accidents for which it is needed.
- 3. The evaluation may consider a materials assessment to determine the material susceptibility to aging, peak temperature and radiation.
- 4. The evaluation may consider similarity analysis to determine that the differences between the nonconforming/degraded equipment and a conforming/qualified one would not impair the equipment design function performance.
- 5. The evaluation may consider extrapolation of available analyses to assess if the design condition would be met for the nonconforming equipment.
L 6. The evaluation should give some indication of the safety significance of the nonconforming/degraded equipment and should specify what would be a reasonable time for corrective actions before operational alternatives are taken.
u Rev. No. 9 Page No. 48 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP 1 1 ATTACHMENT 5 Page 1 of i 4'
REO/REASONABLE ASSURANCE GUIDELINES "Reasonable assurance" is a level of confidence that a particular situation or condition exists or does not exist. In making a finding of reasonable assurance, the existing facts are first gathered.
Engineering judgment is then applied to those facts resulting in a "weight of evidence" supporting a finding of reasonable assurance.
Management involvement is an essential element in the process particularly when uncertainty exists or delays in the process are excessive or unacceptable.
The concept of reasonable assurance can be applied at several stages of the deficiency evaluation process, namely in:
- 1. The initial categorization of a deficiency
- 2. The Operability Determination for safety related equipment
- 3. The development of a Justification for Continued Operation (JCO) if the affected equipment is determined to be inoperable.
Engineering judgment should be applied when it is technically appropriate and defendable to reach conclusions by this method instead of performing more rigorous analyses. The reluctance to use engineering judgment when assigning a level of confidence can contribute to delays in the deficiency evaluation process. Management attention should be applied to expedite the process and focus resources. -
Since the facts in existence may change as more information is obtained, the weight of evidence is dynamic. As the weight of evidence changes, the overall conclusion may require update. For this reason, the use of engineering judgment may require follow-up analyses, tests or inspections to confirm the validity of the conclusions reached.
Three different conclusions can be reached when making a finding of reasonable assurance:
- 1. reasonable assurance that a condition exists,
- 2. reasonable assurance that a condition does not exist, or
- 3. uncertainty as to whether a condition does or does not exist.
If uncertainty exists, management involvement is necessary. Management may decide to add to the existing facts (by directing that additional tests or analyses be conducted) or reevaluate the facts using engineering judgment in order to change the weight of evidence. The weight of evidence should be changed enough to reach a finding of reasonable assurance (either positive or negative).
4 Rev. No. 9 Page No. 49 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AF-C3 -11 ATTACHMENT 6 Page 1 cf 11 u REO/ENGINEERING OPERABILITY GUIDELINES NOTE: This information is provided as an aid. The SK maintains the responsibility to determine whether an SSC is operable c r inoperable.
GENERAL AND MISCELllLNEOUS OPERABILITY ISSUES Operability and Time of Entry into Technical Specification
[CTS]AP-O1.04 [ITS]Technical R e q u i r e m e n t s Manual Action Statements The start of a Technical Specification [CTSIAP-01.04 [ITSITechnical Requirements Manual Action Statement begins at the time the information was originally received or the event occurred.
Equipment that has been discovered to be inoperable from some previous time (known or unknown) shall be considered inoperable from the time the information was originally received for the purpose of entry into a Technical Specification Action Statement (i-e.,
Technical Specification Action Statements are not imposed retroactively.).
Items Clearly Inoperable Certain conditions clearly render equipment inoperable. In these
~u instances, the time of declaring equipment inoperable is the time of discovery.
A. If equipment is unable to perform its function due to obvious failure, damage, or malfunction or due to being removed from service (tagged out), then it is inoperable.
B. If equipment fails to start upon receipt of a valid safety signal, then it is inoperable.
C. If equipment fails to meet the quantitative requirements of Technical Specifications [CTS]AP-01,04 [ITSlTechnical Requirements Manual or of surveillances demonstrating compliance with Technical Specifications, then it is inoperable.
Examples of this are Technical Specification or surveillance required tank levels, system pressures, valve stroke times, system flow rates, etc.
Equipment exposed to operating conditions in excess of its design rating A. If equipment is exposed to operating conditions in excess of its design rating, then it is inoperable until engineering evaluation determines it to be operable.
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Rev. No. 9 Page No. 50 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 ATTACHMENT 6 Page 2 of 11 REO/ENGINEERING OPERABILITY GUIDELINES v Missed or Deficient Surveillance fCTSl 0 Technical Specification related equipment or systems are cieclared inoperable upon discovery of a missed or deficient surveillance test. At the time of discovery of the missed or deficient test, the action statement of the appropriate LCO is applicable; however, if actions are required to be performed within 2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or less, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance provided by Technical Specification 4.0.3 may be entered. If reasonable expectation of operability does not exist, immediately declare the equipment or system inoperable. If, during the testing, the surveillance results in inoperability, then the appropriate LCO action statement requirements must be applied.
[ITS1 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed from the time of discovery up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of 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.
[ITS] __-
If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
[ITS1 When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
If the missed or deficient surveillance is required by Technical Specifications [CTS]AP-01-04 [ITS]Technical Requirements Manual or ASME Section XI, then operability is unaffected.
If this is a failure to retest or perform functional verification of equipment prior to restart or return to service, then the equipment i inoperable.
I? the retest or functional verification is required by Technical Specifications [CTS]AP-01.04 [ITS]Technical Requirements Manual or ASME Section XI, then the equipment is inoperable.
Incorrect Inputs Used in a Calculation If incorrect inputs (e.g. use of the wrong response spectrum, improper cable resistance, incorrect material properties, etc.) were used in a calculation and they are determined to be non-conservative in nature, then an engineering evaluation is required to assess operability.
-J Rev. No. 9 Page No. 51 of 78
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OPERABILITY AND REPORTABILITY DETExNINATIONS AP-03. II ATTACHMENT 6 Page 3 of 11 ii REO/ENGINEERING OPERABILITY GUIDELINES Documentation Only Deficiencies If document deviations are identified that do not constitute a non-conforming condition as described in definition 5.1C, then the related S S C s are operable.
For example, if an EQ file is identified with an incorrect radiazion margin, but justification is available that envelops the plant design requirements, there is no non-conforming condition since design and licensing requirements are met.
MECHANICAL OPERABILITY ISSUES Minor process fluid leakage (packing glands, gaskets, and non-welded connections)
If the leakage clearly has no adverse impact on any SSC function, then the equipment is operable unless the leakage constitutes reactor coolant system or Containment boundary leakage. Then an engineering evaluation is required. Examples of possible adverse effects requiring engineering evaluation are: spraying water does not reach electrical equipment, contaminate oil reservoirs or result in significant spread of contamination u
Oil l e a k s If the leakage is from non safety-related pumps or motors (i.e.
equipment without an assumed long-term operating requirement in the UFSAR) and does not require reservoir replacement more than once per shift, then the equipment is operable.
If level is found to be outside the optimum range, but remains visible in the sight glass, then the equipment is operable unless the frequency of oil replacement exceeds once per shift or increases significantly.
If the leakage is from safety-related or Cat M pumps (i.e. equipment assumed to require extended operation per the UFSAR) or requires reservoir replacement more than once per shift, then an Engineering Evaluation is required.
Materials (piping, fittings, bolts, nuts, etc.)
If the material for piping, fittings, bolts and other components is discovered to be different than the design documents specify, then the equipment is operable pending Engineering Evaluation of chemical compatibility and structural strength requirements.
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Rev. No. 9 Page No. 52 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-C3.11 ATTACHMENT 6 Page 4 of 11
'4 REO/ENGINEERING OPERABILITY GUIDELINES If painting deficiencies (missing, bubbled or chipped) are found on external surfaces outside of Primary Containment that do not pose a Foreign Material Exclusion (FME) concern, then the equipment is operable.
Manual valve position If a manual valve that is required by Technical Specifications to be locked in a particular position is found in the correct position, but not locked, then the valve is inoperable until locked or until equivalent compensatory measures are taken.
0 If a manual valve that is required by a document other than Technical Specifications ([CTSIAP-01.04 [ITSITRM, UFSAR, etc.) to be locked in a particular position is found in the correct position, but not locked, then the valve is operable.
If a manual valve that is required to be in a particular position is found in an incorrect position, then the valve and its associated system are inoperable until restored to its required position.
Power operated valves If a motor operated valve is required by Technical Specifications or other licensing basis document (e.g. Appendix R analysis) to be in a specific position with its associated supply breaker open, and the breaker is not open, then the valvs and its associated system are inoperable until the breaker is opened.
If a Technical Specification required "actuated-closedpower operated valve is backseated, then the valve is inoperable until either:
A. an engineering evaluation has been performed to demonstrate that the backseating is acceptable, or B. the valve has been demonstrated to be capable of closing from its backseated position.
If a Technical Specification required "actuated-open" power operated valve is torque seated, then the valve is inoperable until either:
A. an engineeiing evaluation has been performed to demonstrate that the torque seating is acceptable, or B. the valve has been demonstrated to be capable of opening from its torque seated position.
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OPERABILITY AND REPORTABILITY DETERMINATIONS AF-03-21 ATTACHMENT 6 Page 5 of 11 L REO/ENGINEERING OPERABILITY GUIDELINES ASME Section XI qualification If ASME Section XI equipment does not meet the overall requirements of the applicable ASME Section XI specifications, then the eqtlipment is inoperable.
If a valve exceeds the IST stroke time ACTION limit, then the equipment is inoperable.
If a pump does not meet the IST ACTION criteria, then the equipment is inoperable.
ELECTRICAL/I&C OPERABILITY ISSUES Setpoint and calibration tolerance If an equipment setpoint or calibration is determined to exceed that required by Technical Specifications [CTSIAP-01.04 [ITSlTechnical Requirements Manual, then the equipment is inoperable.
If an equipment loop is determined (by test and/or calculation) to be unable to perform its intended function within its required Technical Specification limits, then the loop is inoperable.
U Equipment with automatic and manual start/stop capability If, for such equipment, the manual start/stop capability is required (by Technical Specifications CCTSIAP-01.04 [ITSlTechnical Requirements Manual, UFSAR, EOPs, etc.) to fulfill a UFSAR Specified Function and it is lost, then the equipment is inoperable.
If, for such equipment, the automatic start/stop capability is required (by Technical Specifications [CTSlAP-01.04 [ITSlTechnical Requirements Manual, UFSAR, EOPs, etc.) to fulfill a UFSAR Specified Function and it is lost, then the equipment is inoperable.
Rev. No. 9-- Page No. 54 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AD-03.11 ATTACHMENT 6 Paae 6 of 11
-i REO/ENGINEERING OPENABILITY GUIDELINES Environmental Qualification (EQ)
If equipment is installed and maintained in accordance with the J A F Environmental Qualification Program, then, from an EQ standpoint, the equipment is operable (i.e., it is environmentally qualified or has "environmental qualification") .
If, on equipment that is required to be environmentally qualified, a condition exists that obviously would not allow performance of a UFSAR Specified Function under all postulated service conditions, then the equipment is inoperable.
For example, the EQ Maintenance & Installation Requirement for an instrument transmitter may require it to be sealed against moisture/stean intrusion. If the transmitter does not have a seal installed, it is inoperable because it is obvious that it would not meet the EQ Maintenance & Installation Requirement.
If, on equipment that is required to be environmentally qualified, a condition exists that may compromise its environmental qualification, but it is not obvious whether its UFSAR Specified Function would be performed under all postulated service conditions, then the condition may require Operability Confirmation. _-
For example, the EQ Maintenance & Installation Requirement for an instrument transmitter may require it to be sealed against moisture/steam intrusion. If the transmitter has an unused conduit connection sealed anly with a plastic shipping plug, then the transmitter may be operable. This may be either because other testing has been performed for this configuration or the EQ documentation may not have differentiated between LOCA and HELB mitigation, which have different qualification requirements.
For another example, a procedural EQ Maintenance & Installation Requirement may require replacement of an instrument transmitter's O-rings at five-year intervals. If it is determined that a transmitter has exceeded this five year O-ring replacement interval, it is not obvious that performance of its UFSAR Specified Function is prevented. An evaluation using transmitter/O-ring test data, engineering analysis, etc., is required to confirm its operability, Rev. No. 9 Page No. 55 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS ;IC-93.1:
ATTACHMENT 6 Page 7 of 11 c- REO/ENGINEERING OPERABILITY GUIDELINES Electrical Breakers If an electrical breaker trips after having been reset from an earlier tripped condition, then it is inoperable.
Operability may,be re-established once the cause of the breaker trip is corrected and component operation is retested.
Emergency Diesel Generators If an Emergency Diesel Generator fails to start or load, then it is inoperable.
If an EDG trips on a non-safety trip signal that would be bypassed on receipt of an automatic start signal, then the cause of the trip requires an Engineering Evaluation to determine whether the EDG would be capable of sustaining its function after an automatic start.
EPIC Safety Parameter Display System (SPDS)
If unable to successfully restart either Cpu after failure of both
- for greater than one hour, then the SPDS is inoperable.
If unable to transmit data to either the Technical Support Center (TSC) or the Emergency Operations Facility (EOF) for a period greater than one hour, then the SPDS is inoperable.
STRUCTURAL/CIVIL OPEKABILITY ISSUES Barriers If an inadequate fire barrier exists, then it is inoperable.
If a Condition exists affecting the structural integrity of a room, building, foundation, or other structural component, then an evaluation may need to be performed to determine its operability.
If a HELB barrier is breached, then an engineering evaluation is required to confirm operability of affected SSCs.
u Rev. No. 9 Page No. 56 of 78
OPERABILITY AND REPORTAEILITY DETERMINATIONS AF-Oj. 11 ATTACHMENT 6 Page E of 11 REO/ENGINEERING OPERABILITY GUIDELINES Missing or loose nuts/fasteners If the fastener is part of one of the following joints, then the SSC is operable (provided an FME concern is not created):
A joint supporting a non-EQ cover plate for an instrument, switch, cable tray cover or other similar application, provided there are no more than two missing or loose fasteners.
A nut/bolt holding a valve handwheel, provided the stem is directed vertically upwards.
Cabinet door hinges or latches, provided the door can be opened, closed and latched.
Valve packing gland assembly, provided there is no leak at the valve stem with the system pressurized, I S T stroke times are within limits and the valve is not a Containment Isolation Valve.
A nut/bolt holding a screen on air cooling ports of equipment, provided there are sufficient bolts in place to prevent movement of the screen.
Pipe & Tubing Supports -_c If the component is obviously damaged, then an engineering evaluation is required to confirm support/system operability.
If the component is incapable of performing its function, then the component is inoperable. An engineering evaluation is required to determine system operability since such a condition may or may not render the system inoperable.
If the component is performing a function it is not designed for (e.g. pipe is binding due to inadequate gaps or supporting unauthorized/unanalyzed equipment), then an engineering evaluation is required to confirm operability.
-4 Rev. No. 9 Page No. 57 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS f L 2 - c ; -11 ATTACHMENT 6 Page 9 cf 11 REO/ENGINEERING OPERABILITY GUIDELINES Hydraulic and mechanical snubbers If a hydraulic or mechanical snubber is found to be ontside the range specified on the drawing or beyond the snubbers setting, including tolerances, then an engineering evaluation is required.
If a hydraulic snubber has "the remotest amount of fluid available in the reserve reservoir", there is enough fluid to accommodate full hydraulic response for any stroke position and it is operable.
EROSION/CORROSION If pipe wall thickness is measured to be <87.5% of the nominal wall thickness, then an engineering evaluation is required to determine if the pipe is operable. Procedure CES-7, Procedure for Structural Evaluation of Erosion/Corrosion Thinning in Carbon Steel Piping,"
provides guidance for this evaluation.
CHEXISTRY SAMPLING OPERABILITY ISSUES If a chemistry sample is outside the Technical Specification limits AND a second confirmed sample exists, OR the responsible individual concludes that the results are valid based on trend of previous analyses, then it is inoperable.
L USE OF ENGINEERING JUDGKENT FOR OPERABILITY ASSESSMENTS According to the definition of operability, an SSC must be capable of performing its specified function(s), as described in the Technical Specification Bases or the UFSAR, If a system, structure, or component is covered by the Technical Specifications, then the Technical Specifications must be followed. If an SSC is not covered by Technical Specifications, then the additional guidance of Generic Letter 91-18, can be applied.
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OPERABILITY AND REPORTABILITY DETERMINATIONS AP-C3.11 ATTACHMENT 6 Page 10 of 11
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REO/ENGINEERING OPERABILITY GUIDELINES USE OF ENGINEERING JDDGMENT FOR OPERABILITY ASSESSMENTS (COnt)
Generic Letter 91-18, and other NRC guidance acknowledge the acceptability of using engineering judgment to justify component o r system operability. The two major points that most of the guidance stresses are:
Engineering judgment should only be used to justify component or system operability when there is reasonable expectation that a detailed analysis or evaluation will prove operability of the component or system.
A sound basis for the engineering judgment conclusion must be documented.
The scope of the engineering judgment Operability Determination must be sufficient to address the capability of the equipment to perform its UFSAR Specified Functions. The determination should consider the following:
Determine what equipment is degraded.
Determine the UFSAR Specified Functions of the affected equipment. -VI Determine the extent of the degradation, including the possible failure mechanism.
Determine if the equipment is capable of performing its UFSAX Specified Function.
Determine the basis for declaring the system operable.
NOTE: If the component or system can not perform at the level credited in the accident analysis, then it should be considered inoperable unless supported by additional analysis. If the component or system can not perform at the level required by Technical Specifications [CTSIAP-01.04 [ITSITechnical Requirements Manual, then it should be considered inoperable.
The Operability Determination should discuss the considerations listed above and must document the basis for declaring the system operable.
To continue operation while a formal Operability Determination is being made, there must be a reasonable expectation that the affected safety system is operable and that the confirmation process will support that expectation. If that expectation does not exist or mounting evidence suggests that the final analysis will conclude that the equipment can not perform its UFSAR Specified Functions, then the system should be considered inoperable and appropriate actions must be taken. J Rev. No. 9 Page No. 59 of 78
OPERABILITY m REPORTABILITY DETERMINATIONS AP-C3.11 ATTACHMENT 6 Page 11 of II REO/ENGINEERING OPERABILITY GUIDELINES CONTINGENT OPERATOR ACTION ( 8 ) GUIDANCE The following criteria should be considered when operator action(s) are credited in the Operability Assessment:
- 1. Sufficient number of shift operators are available to perfoAm the required actions.
- 2. Written procedures, which outline the required actions are clear, complete, unambiguous, available and used.
3 . Operators performing the required action are properly trained.
- 4. Sufficient time is available for the operator to perform the required actions.
- 5. Locations outside of the control room at which operator actions must be performed shall be qualified to adequately protect the operator from the environmental conditions caused by the design basis event.
6 . The dose to an individual operator who is required to take actions shall not exceed 5 Rem TEDE - Limit (Internal Dose & External Dose) whole body,' for the duration of the event.
References:
Information Notice 9 7 - 7 8 and ANSI 58.8 Rev. No. 9 Page No. 60 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 ATTACHMENT 7 Page 1 of 12 IMMEDIATE REPORTABILITY CHECKLIST This checklist should be used for determining if an event is immediately reportable, or reportable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Refer to NUREG-1022, "Event Reporting Guidelines 10CFR50.72 and 10CFR50.73" (Rev. 2 ,
October 2000) for additional guidance on reportability determinations.
I. IMMEDIATE REPORTABILITY A. 10 CFR 20.1906(d) - Immediate notification to the NRC and the final,-deliverycarrier (not later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
Does the event involve the receipt of a shipping package found to be in non-compliance with external contamination or dose rate limitations? YES/NO B. 10 CFR 20.2201(a)(l)(i) - Immediate notification not later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Does the event involve loss or theft of licensed material? YES /NO C. 10 CFR 20.2202(a) - Immediate notification (not later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
Does the event involve byproduct, source or special nuclear material possessed by Entergy that may have caused or threatens to cause:
- 1. An individual to receive a total effective dose equivalent of 25 rems 01 more, an eye dose equivalent of 75 rems or more, or a shallow - dose equivalent to the skin or extremities of 250 rads. YES/NO
- 2. The release of radioactive material, inside or outside of a restricted area, so that, had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the individual could have received an intake five times the occupational annual limit on intake (does not apply to location where personnel are not normally stationed during routine operations). YES /NO i-Rev. No. Page No. 61 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-C3.11 ATTACHMENT 7 Page 2 of 1:
IMMEDIATE REPORTABILITY CHECKLIST L'
I. IMMEDIATE REPORTABILITY (cont)
D. 10 CFR 50.72(a)(l): Immediate Notification (ENS phone)
(i) Did the event result in the declaration of any of the Emergency Plan classes? (See 10 CFR 72.75(a) for ISFSI events.1 YES /NO E. 10 CFR 72.75(a):' Immediate Notification (ENS phone), followed by a written report within 30 days Did the event involve ISFSI and result in the declaration of any Emergency Plan classes? YES/NO Rev. No. 9 Page No. 62 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 ATTACHMENT 7 Page 3 of 12 IMMEDIATE REPORTABILITY CHECKLIST I
- 11. 1 HOUR REPORTABILITY A. 10CFR50.72(b) (1): 1 Hour Notification NOTE: This question may include situations that have occurred in the past 3 years, may continue to exist, or have been recently discovered.
Did the event result in any deviation from Tech Specs authorized pursuant to 10 CFR 50.54(x) (if not reported as declaration of an Emergency Class under 50.72(a) (1) above) YES /NO If the answer is YES to the above question, notify the NRC Operations Center within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B. 10CFR70.52: 1 Hour Notification (a) Does the event involve accidental criticality or any loss, other than normal operating loss, of special nuclear material? (See also 10CFR72.74(a) and 10CFR73 Appendix G (I)(a)(1)- 1 YES/NO (b) Does the event involve any loss or theft or unlawful diversion of special nuclear material or any incident in which an attempt has been made or is believed to have been made to commit a theft or unlawful diversion of such material? YES /NO --
C. 10CFR72.74(a): 1 Hour Notification Did the event involve ISFSI accidental criticality or loss of special nuclear material. (See also 10CFR70.52(a) and 1CCFR73 Appendix G (I)(a)(l).) YES/NO Rev. No. 9 Page No. 63 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 ATTACHMENT 7 Page 4 cf ;2
- 11. 1 HOUR IMMEDIATE REPORTABILITY CHECKLIST REPORTABILITY (cont)
D. 10CFR73 Appendix G (I) - 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notification followed by a written report within 30 days.
Does the event require, cause, or result in the following?
Any event in which there is reason to believe that a person has committed or caused, or attempted to commit or cause, or has made a credible threat to commit or cause:
A theft or unlawful diversion of special nuclear material. (See also 10CFR70.52 and 10CFR72.74(a).); YES/NO OR Significant physical damage to a power reactor or any facility possessing SSNM or its equipment or carrier equipment transporting nuclear fuel or spent nuclear fuel, or to a nuclear fuel or spent nuclear fuel a facility or carrier possesses; YES/NO OR Interruption of normal operation of a licensed nuclear power reactor through the unauthorized use of or tampering with its machinery, components, or controls including the security system. YES/NO An actual entry of an unauthorized person into a protected area, material access area, controlled access area, vital area, or transport. YES/NO Any failure, degradation, or the discovered vulnerability in a safeguard system that could allow unauthorized or undetected access to a protected area, material access area, controlled access area, vital area, or transport for which compensatory measures have not been employed. YES/NO The actual or attempted introduction of contraband into a protected area, material access area, vital area, or transport. YES/NO Rev. No. 9 Page No. 64 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-@3.11 ATTACHMENT 7 Page 5 of 12 IMMEDIATE REPORTABILITY CHECKLIST .-
111. 4 HOUR REPORTABILITY A. 10CFR50.72 ( b I ( 2 ) - 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Notification NOTE: These questions may include situations that have occurred in the past 3 years, may continue to exist, or have been recently discovered.
Does the event require, cause, or result in the following?
The initiation of a shutdown required by Tech Specs. YES/NO Reserved Reserved Any event that results or should have resulted in Emergency Core Cooling System (ECCS) discharge into the Reactor Coolant System as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. YES/NO Any event or condition that results in the actuation of the RPS when the reactor is critical, except when the actuation results from and is part of a pre-planned -4 sequence during testing or reactor operation. YES/NO Reserved Reserved Reserved Reserved Reserved Reserved Any event or situation, related to the health and safety of the general public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an on-site fatality or inadvertent release of radioactively contaminated materials. YES/NO If the answer is YES to any of the above statements, notify the N R C Operations Center within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
J Rev. No. 9 Page No. 65 of 78
e OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 ATTACHMENT 7 Page 6 of 1 2 u IMMEDIATE REPORTABILITY CHECKLIST 111. 4 HOUR REPORTABILITY ( c o n t )
B. 10CFR72.75(b) - 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Notification followed by a written report within 30 days.
Does the event involve ISFSI and require, cause, or result in the following:
An event that prevents immediate actions necessary to avoid exposures to radiation or radioactive materials that could exceed regulatory limits, or releases of radioactive materials that could exceed regulatory limits (e-g.,events such as fires, explosions, and toxic gas releases). YES /NO A defect in any spent fuel storage structure, system, or component which is important to safety. YES /NO A significant reduction in the effectiveness of any spent fuel storage confinement system during use. YES /NO An action taken in an emergency that departs from a condition or a technical specification contained in a license or certificate of compliance issued under 10CFR72 when the action is immediately needed to protect the public health and safety and no action consistent with license or certificate of compliance conditions or technical specifications that can provide adequate or equivalent protection is immediately apparent. YES /NO An event that requires unplanned medical treatment at an offsite medical facility of an individual with radioactive contamination on the individual's clothing or body which could cause further radioactive contamination. (See also 10CFRSO.72(b)(3)(xii)eight hour notification.) YES /NO A n unplanned fire or explosion damaging any spent fuel or high level waste, or any device, container, or equipment containing spent fuel or high level waste when the damage affects the integrity of the material or its container. YES /NO l+ - ,
Rev. No. 9 Page No. 66 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.1:
ATTACHMENT 7 Page 7 of 12 IMMEDIATE REPORTABILITY CHECKLIST IV. 8 HOUR REPORTABILITY A. 10CFR50.72(b)( 3 ) - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification NOTE : These questions may include situations that have occurred in the past 3 years, may continue to exist, or have been recently discovered.
Does the event require, cause or result in the following?
(i) Reserved (ii)(A) Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; YES/NO (ii)(B) Any event or condition that results in the nuclear power plant being in an un-analyzed condition that significantly degrades plant safety. YES/NO (iii) Reserved (iv)( A ) Any event or condition that results in valid actuation of any of the systems listed in paragraph iv) ( B ) below, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. YES/NO ~
(iv)(B) NOTE 1: The systems to which the requirements of paragraph (iv)(A) above apply are:
(1) Reactor protection (RPS) including: reactor scram and reactor trip. (See also (b)( 2 ) (iv)( B ) on page 5.)
(2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).
(3) (Not applicable to JAF)
(4) ECCS including: low-pressure core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system.
(5) Reactor core isolation cooling system (6) (Not applicable to JAFl (7) Containment heat removal and depressurization systems, including containment spray and fan cooler systems.
(8) Emergency AC electrical power systems, including emergency diesel generators (EDGs).
Rev. No. 9 Page No. 67 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS A?-C2- 7 ATTACHMENT 7 Page E of 12 IMMEDIATE REPORTABILITY CHECKLIST L,
IV. 8 HOUR REPORTABILITY (COnt)
A. 10CFR50.72(b) (3) - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification (cont)
(v) Any event or condition that at the time of discover); could have prevented the fulfillment of the safety function of structures or systems that are needed to:
(A) Shutdown the reactor and maintain it in a shutdown condition. YES/N@
(B) Remove residual heat. YES/NO (C) Control the release of radioactive material. YES/NO (D)Mitigate the consequences of an accident. YES/NO (vi) NOTE 2: Events covered in paragraph (v) above may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to paragraph (v) above if redundant equipment in the same system was operable and available to perform the required safety function.
(vii Reserved (viii Reserved (1x1 Reserved (x) Reserved (xi) Reserved (xii) Any event requiring the transport of a radioactively contaminated person to an off-site medical facility for treatment. (See also 10CFR72.75(b)( 5 ) four hour notification (ISFSI requirement).) Y E S /NO (xiii) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g.,significant portion of Control Room indication, ENS or Off-Site Notification System). YES/NO If the answer is YES to any of the above statements, notify the NRC Operations Center within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
i-.c.
Rev. No. 9 Page No. 68 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 ATTACT 7 Page 9 of 12 IMMEDIATE REPORTABILITY CHECKLIST V. 24 HOUR REPORTABILITY The following checklist should be used for determining if an event is reportable or recordable within 2 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:
A. 1 0 C F R 2 0 . 2 2 0 2 ( b ) - 2 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification Does the event involve loss of control of licensed material possessed by Entergy that may have caused or threatens to cause:
(1) An individual to receive, in a period of 2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (i), a total effective dose equivalent exceeding 5 rems (ii),an eye dose equivalent exceeding 15 rems, or (iii) a shallow-dose equivalent to the skin or extremities exceeding 50 rems. YES/NO The release of radioactive material, inside or outside of a restricted area, so that, had an individual been present for 2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the individual could have received an intake in excess of one occupational annual limit on intake (does not apply to locations where personnel are not normally stationed during routine operations).
YES/NO If the answer is YES to any of the above statements, notify the NRC Operations Center within 2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
B. 10CFR26.73 (a) - 2 4 Hour Report (Fitness for Duty)
Does the event involve a significant fitness for duty event involving:
(1) The sale, use, or possession of illegal drugs within the protected area. YES/NO Rev. No. 9 Page No. 69 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS A F - c 3 -11 ATTACHMENT 7 Page 10 of 12 IMMEDIATE REPORTABILITY CHECKLIST L e V. 24 HOUR REPORTABILITY (cont)
- 13. 10CFR26.73 (a) - 24 Hour Report (cont)
(2) Any acts by any person licensed under 10 CFR Part 55 to operate a power reactor or by any supervisory personnel assigned to perform duties:
(i) Involving the sale, use, or possession of a controlled substance YES/NO (ii) Resulting in confirmed positive tests on such persons YES/NO (iii) Involving use of alcohol within the protected area YES/NO (iv) Resulting in a determination of unfitness for scheduled work due to the consumption of alcohol. YES/NO If the answer is YES to any of the above statements, notify the NRC Operations Center as required.
C. 10CFR72.75(c) - 24 Hour Notification followed by a written report within 30 days L- Did any of the following events related to ISFSI activities involving spent nuclear fuel or high level waste occur:
(1) Any unplanned contamination event that requires access to the contaminated area by workers or the public to be restricted for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by imposing additional radiological controls or by prohibiting entry into the area. YES/NO Page No. 70 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-'23.11 ATTACHMENT 7 Page 11 of 12 IMMEDIATE REPORTABILITY CHECKLIST ./
- v. 24 HOUR REPORTABILITY (cont)
C. 10CFR72.75(c) - 2 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification followed by a written report within 30 days. (cont)
(2) An event in which safety equipment is disabled or fails to function as designed when:
(i) The equipment is required by regulation, license condition,.or certificate of compliance to be available and operable to prevent releases that could exceed regulatory limits, to prevent exposures to radiation or radioactive materials that could exceed regulatory limits, or to mitigate the consequences of an accident, and (ii) No redundant equipment was available and operable to perform the required safety function.
YES /NO D. ISFSI Certificate of Compliance No. 1014 Appendix B, Approved Contents, ( R e f . 3.2.16) Section 2 2 4 Hour Notification followed by a written report within 30 days Did the event result in violation of any fuel specification or loading condition during storage cask loading? YES/NO ..--..
E. 10CFR73 Appendix G (11) - recorded within 2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and submitted in quarterly log.
Does the event require, cause, or result in the following?
(a) Any failure, degradation, or discovered vulnerability in a safeguards system that could have allowed unauthorized or undetected access to a protected area, material access area, controlled access area, vital area, or transport had compensatory measures been employed. YES/NO Any threatened, attempted, or committed act not previously defined in 10 CFR 7 3 , Appendix G with the potential for reducing the effectiveness of the safeguards system below that committed to in a licensed physical security or contingency plan or the actual condition of such reduction in effectiveness. YES/NO If the answer is YES to any of the above statements, ensure Security/Safety Department "records" the incident.
Rev. No. -9 Page No. 71 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-C3.11 ATTACHMENT 7 Page 1 2 of :2 L--
IMMEDIATE REPORTABILITY CHECKLIST VI. 30 or 6 0 DAY REPORTABILITY Many of the events identified above, except for V . B ,
require a 30 or 60 day written notification. (See AP-03-04 for reportability requirements.)
Rev. No. 9 Page No. 72 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-03.11 ATTACHMENT 8 Page 1 of 4 ENTERGY LICENSING POSITION '-
EVALUATION AND RESOLUTION OF DEGRADED AND NONCONFORMING CONDITIONS PURPOSE This paper presents Entergy's position on evaluating and resolving degraded and nonconforming conditions, as documented on CRs per the corrective action process and PIDs,) per the Work Request process.
Included in this paper are discussions pertaining to:
Appropriate applicability of OPERABILITY Assessments Timeliness for resolving degraded and nonconforming conditions
'0 Appropriate applicability of 10CFR50.59 evaluations to degraded or nonconforming conditions.
This position paper is not intended to circumvent any actions permitted or required by plant Technical Specifications (TS) or any other regulatory requirement. Attachment 1 provides additional supporting information.
BACKGROUND On November 7, 1991, the NRC published Generic Letter (GL) 91-18 +-'
providing to licensees two new sections to its Part 9900 NRC Inspection Manual. These sections were:
- 1. Resolution of Degraded 2nd Nonconforming Conditions
- 2. Operable/Operability: Ensuring the Functional Capability of a System or Component The intent of the guidance provided in these new sections of Part 9900 was to ensure consistency in OPERABILITY assessments and in resolving degraded and nonconforming conditions.
On October 18, 1997, the NRC published Revision 1 to GL 91-18 to inform licensees of a revised section of Part 9900, Resolution of Degraded and Nonconforming Conditions. The changes to this section more explicitly discuss the role of 10CFR50.59 in resolving degraded and nonconforming conditions.
'U Rev. No. 9 Page No. 73 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS A F - 0 3 -11 ATTACHMENT 8 Page 2 of 4 b ENTERGY LICENSING POSITION ENTERGY POSITION I. APPLICABILITY OF OPERABILITY ASSESSMENTS Entergy will make every effort to operate its plants in full compliance with applicable design and licensing requirements and commitments with no degraded or nonconforming conditions.
However, when a degraded/ nonconforming condition is identified, an OPERABILITY assessment is performed to determine if the affected SSC is capable of performing its intended safety function (i.e., safety-related as defined in 10CFR50.2). Such activities are governed under the corrective action program required by 10CFR50, Appendix B, Criterion XVI. (NOTE: The level of detail provided in the OPERABILITY assessment is commensurate with the safety function of the degraded/nonconforming SSC and the complexity of the issue.)
The terms "OPERABLE" and "OPERABILITY" are defined in each plant's Technical Specifications. In addition, system OPERABILITY is governed by Technical Specifications. The OPERABILITY assessment focuses on the ability of the subject ssc to meet its specified safety functions, thereby meeting the Technical Specification definition of "OPERABLE".
In addition to clearly evaluating the condition's effect on safety function capability, the OPERABILITY assessment should include reviews to determine any compensatory actions needed to maintain OPERABILITY.
- 11. TIMELINESS OF RESOLVING DEGRADED/NONCONFORMING CONDITIONS Degraded/nonconforming conditions will be corrected as soon as practicable cammensurate with safety significance of the condition (e.g., the ability of the SSC to perform its safety function). Corrective actions should be taken to bring the degraded/nonconforming condition into compliance with the design/licensing basis at the first reasonable opportunity. As plant on-line times improve, this opportunity may be during the upcoming refueling outage. Our goal is to restore significant conditions that impact SSC OPERABILITY no later than startup from the next refueling outage unless extenuating circumstances arise which make such actions impracticable (e.g., plant conditions, parts availability, incomplete design).
u Rev. No. 9 Page No. 74 Of 78
~
OPERABILITY AND REPORTABILITY DETERMINATIONS AF-03. II ATTACHMENT 8 Page 3 of 4 ENTERGY LICENSING POSITION In cases where significant degraded/nonconforming conditions cannot be resolved prior to restart from the designated refueling outage, the Operability assessment must be re-evaluated to ensure extended, safe plant operation with the condition is justified with a revised restoration date or milestone specified (usually the next refueling outage). This situation remains under the control of the Appendix B corrective action program; application of 10CFR50.59 is not appropriate. See further discussion in Section 111, below. (Plant management may wish to keep the NRC informed of these conditions and of revised plans to correct the condition. This communication can be as simple as discussing the condition with the site resident inspector and the NRR project manager; no formal approval is required.)
GL 91-18 describes the relationship between OPERABILITY and 10CFR50, Appendix B, Criterion XVI. In brief, licensees must develop and implement a corrective action plan in parallel with the OPERABILITY assessment. The purpose of the OPERABILITY assessment is to determine if there is reasonable assurance that the degraded/nonconforming SSC will perform its safety function.
Conditions that have been identified as degraded/nonconforming should be evaluated for aggregate impact on specific safety functions. Therefore, the degraded/nonconforming condition may remain in effect while implementing a corrective action plan d intended to restore the condition to "Fully Qualified" status.
Actions to resolve a degraded/nonconforming condition must be taken promptly, commensurate with the safety significance of the adverse condition.
111. APPLICABILITY OF 10CFR50.59 EVALUATIONS TO DEGRADED/NONCONFORMING CONDITIONS 10CFR50.59 reviews and OPERABILITY assessments are mutually exclusive aspects of addressing a condition adverse to quality.
As discussed in Section I above, the OPERABILITY assessment provides reasonable assurance a degraded/nonconforming SSC is capable of performing its safety function during the time period corrective actions are being developed and implemented. The S50.59 process is only applied to certain compensatory actions that may be taken as part of the OPERABILITY assessment and to any proposed changes implemented to resolve the degraded/
nonconforming condition. These items are discussed below.
Rev. No. 9-- Page No. 75 of 78
OPERABILITY AND REPORTABILITY DETERMINATIONS A P - 0 3 . il ATTACHMENT 8 Page 4 of 4 L ENTERGY LICENSING POSITION Compensatorv Actions As discussed in Section I above, compensatory actions may be required to maintain system/component OPERABILITY, e.g., changes to procedures, temporary physical plant modifications (temporary alterations), implementing new procedures. Such actions require review for 10CFR50.59 applicability. The scope of the applicability review and any resulting evaluation is limited to the specific compensatory actions and should not include the full scope of the degraded/nonconforming condition.
Resolvinq Degraded/Nonconformina Conditions Resolving the degraded/nonconforming condition is typically addressed through one of three ways:
- 1. Restoration to current design
- 2. A physical change to the plant
- 3. A change to design and licensing bases to accept the condition as-is In cases 2 and 3 , the appropriate change process(s1 for actions to
'u be taken requires a 10CFR50.59 review. However, as specified in Section I above, the degraded/nonconforming condition itself is governed by the OPERABILITY assessment.
Failure to resolve a degraded/nonconforming condition in a timely manner is an issue with the corrective action program and not with application of 10CFR50.59.
i/
Rev. No. 9 Page No. 76 of 78
__ _ _ ~ - - ~~ __
OPERABILITY AND REPORTABILITY DETERMINATIONS AP-C3.11 ATTACHMENT 9 Page 1 o f 1 ENGINEERING CONFIRMATION PROCESS FLOWCflART I I I S M r e c e i v e s C R l P l D identifying p o t e n t i a l operability concern I
1 Notify System E n g M a n a g e r
/ \
System Engineering M a n a g e r :
I Initiate i m m e d i a t e a c t i o n s ,
a s necessary I
Brief N R C Resident Inspector Notify R e g u l a t o r y C o m p l i a n c e Manager C o n firm a tion Notify:
Shift M a n a g e r Operations Manager System Enginering M a n a g e r confirmed? Regulatory C o m p l i a n c e M a n a g e r P r o v i d e results t o :
If n e c e s s a r y , p e r f o r m D e t a i l e d E n g i n e e r i n g C o n f i r m a tion Shift M a n a g e r GMPO Engineering Manager(s)
Rev. No. 9 Page N o . 77 of 78
I I T I I I
I ON I I I I I I 1
&HVH3MO?d S S 3 I 3 O H d NOIJIVNIKH3LL3CI ALLIlIBVH3dO 01 ;LNBKH3V&JIV
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of communications procedures Group #
associated with EOP implementation.
(CFR: 41.10/45.13)
KIA # 2.4.15 2.4.15 Importance Rating 3.0 3.5 Proposed Question: The Shift Manager has implemented the Emergency Plan based on high Drywell pressure and assigned an Operator as the NRC Communicator.
WHICH ONE of the following describes when communications with the NRC may be secured?
a) Technical Support Center is activated ROlSRO b) NRC disconnects or authorizes securing line 73/97 c) Transient is over and the plant is recovering d) Once initial classification notice is provided to the NRC Proposed Answer: b) NRC disconnects or authorizes securing line Explanation (Optional):
Technical Reference(s): EAP-1.1 attachment 14 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EP-12.5.5.1, EO-2.07 (As available)
Question Source: Bank # Limerick 1 INPO # 12345 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1/20/1998 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
'k Page 117 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2QQ3 10:27AM
Examination Outline Cross-reference: Level RO SRO L
Tier # 3 3 Knowledge of facility protection requirements Group #
including fire brigade and portable fire fighting equipment usage.
(CFR: 43.5 / 45.12)
KIA # 2.4.26 2.4.26 Importance Rating 2.9 3.3 Proposed Question: Given the following conditions:
You are responding to an electrical fire as a member of the plant's fire brigade team.
You have brought a Class B/C fire extinguisher to the scene.
Other members have rigged a fire hose with a solid-stream nozzle.
Which one of the following actions should be taken?
a) Do not use the fire hose. Put the fire out with the Class B/C fire extinguisher.
b) Use the fire hose first. If it does not put out the fire, use ROlSRO the Class B / C fire extinguisher.
74/98 c ) Wait for the fire brigade member assigned to bring a Class D fire extinguisher, then use the Class D fire extinguisher.
d) Do not use the Class B/C fire extinguisher. Put the fire out with the fire hose.
Proposed Answer: a) Do not use the fire hose. Put the fire out with the Class L-B/C fire extinguisher.
Explanation (Optional):
Technical Reference(s): (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDL-76, EO-1.05.A (As available)
Question Source: Bank # LaSalle 1 INPO # 11156 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 1019/1995 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the informationwill necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
Page 118 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 1027 AM
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of the RO's responsibilities in Group #
emergency plan implementation.
(CFR: 45.11)
KIA # 2.4.39 2.4.39 Importance Rating 3.3 3.1 Proposed Question: You are a licensed Reactor Operator on dayshift, working on the FIN Team. You do not have assigned responsibilities in the Emergency Response Organization (ERO). A transient occurs that results in the declaration of an ALERT Emergency and Protected Area Evacuation. To which of the following locations do you report?
a) The Operations Support Center (OSC)
RO/SRO b) The Training Building assembly area.
75/99 c) The Technical Support Center (TSC).
d) The Offsite Assembly Area (Airport).
Proposed Answer: a) The Operations Support Center (OSC)
Explanation (Optional):
Technical Reference(s): EAP-IO (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None L Learning Objective: EP-12.5.3, EO-?. I 8 (As available)
Question Source: Bank # Duane Arnold 1 INPO # 8781 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 912011999 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Comments:
Page 1 1 9 of 120 NRC Written Examination Submitt'8l.doc Last printed 6/6/2003 10:27AM
c Examination Outline Cross-reference:
Knowledge of communications procedures Level Tier #
Group #
RO 3
SRO 3
associated with EOP implementation.
(CFR: 41.10 145.13)
KIA # 2.4.15 2.4.15 Proposed Question: The Shift Manager 3-m.7 . as the NRC Communicator.
WHICH ONE of the following describes when Izb-NFS P communications with the N R C ? m a) Technical Support Center is activated RO/SRO b) NRC disconnects or authorizes securing line 73/97 c) Transient is over and the plant is recovering d) Once initial classification is provided to the NRC 23%
Proposed Answer: b) NRC disconnects or authorizes securing line Explanation (Optional): -
Tied to G2 K1.04 make a new tie and ensure that reason for ALERT is due to EOP implementation.
Technical Reference(s): GAP \.\ & L C L lq (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: (As available)
Question Source: Bank # Limerick 1 INPO # 12345 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 112011998 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge x Comprehension or Analysis 10 CFR Part 55 Content: 55.41 3 55.43 Comments:
Page 120 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/20038:44 AM
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of the bases for prioritizingsafety Group #
functions during abnormallemergencyoperations.
(CFR: 43.5 I45.12)
KIA # 2.4.22 Importance Rating 4.0 Proposed Question: If Torus level cannot be maintained above 10.75 feet, EOP-4, Primary Containment Control, directs the operator to ensure the HPCl turbine is tripped.
Which of the following describes the bases for RClC and HPCl operation under the same EOP circumstances (ie, Torus water level cannot be maintained above 10.75 feet)?
a) RClC operation may continue ONLY if it is the last operable high pressure injection system available to provide adequate core cooling.
RO/SRO b) RClC must be secured at the same time as HPCl to minimize the containment pressure rise.
SI00 c) RClC operation may continue because the turbine exhaust energy does not exceed the vent capability of the containment.
d) RClC must be secured prior to HPCl to prevent erratic turbine operation due to exhaust back pressure fluctuation.
Proposed Answer: c) RClC operation may continue because the turbine exhaust energy does not exceed the vent capability of the containment.
'L Explanation (Optional):
Technical Reference(s): EPGl REV 2 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: EOP4LP, EO-1.05 (As available)
Question Source: Bank # Fermi 2 2 INPO# 19714 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 611412001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 5 Comments:
\
L Page 120 of 120 NRC Written Examination Submittal.doc Last printed 6/6/2003 10:27 AM
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of facility protection requirements G~~~~ #
including fire brigade and portable fire fighting equipment usage.
(CFR: 43.5 145.12)
KIA # 2.4.26 2.4.26 Importance Rating 2.9 . 3.3 Proposed Question: Given the following conditions:
-You are responding to an electrical fire as a member of the plant's fire brigade team.0-You have brought a Class WC fire extinguisher to the scene.0-Other members have rigged a fire hose with a solid-stream nozzle. OWhich one of the following actions should be taken?
a) Do not use the fire hose. Put the fire out with the Class WC fire extinguisher.
b) Use the fire hose first. If it does not put out the fire, use the Class WC fire RO/SRO extinguisher.
74/98 c) Wait for the fire brigade member assigned to bring a Class D fire extinguisher, then use the Class D fire extinguisher.
d) Do not use the Class B/C fire extinguisher. Put the fire out with the fire hose.
Proposed Answer: a) Do not use the fire hose. Put the fire out with the Class WC fire extinguisher.
Explanation (Optional): Question tied to 294001.K1.16- make a new tie Technical Reference(s): (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: S-OLP-7L l.0Z (As available)
Question Source: Bank # LaSalle IINPO # 11156 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 10/9/1995 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Y COmDrehenSiOn or Analvsis 10 CFR Part 55 Content: 55.41 K 55.43 Comments:
Page 121 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/20038:44 AM
Examination Outline Cross-reference: Level RO SRO Tier # 3 3 Knowledge of the RO's responsibilities in Group #
emergency plan implementation.
(CFR: 45.1 1)
WA # 2.4.39 2.4.39 Importance Rating Proposed Question: You are a licensed Reactor Operator on dayshift, working 6 o n h W d e r . You do not have assigned responsibilities in the Emergency Response Organization (ERO). A transient occurs that results in the declaration of an ALERT Emergency and a$mtm-ef the Evacuation&rm To which of the following locations do you report? c*-*4 &.Pa.
a) The+km+Mw+OSb 7-%: +&
RO/SRO b) T h e T ~ % e m b 1 y area.
75/99 c) The Technical Support Center (TSC).
A- +%@)
d) The O f f s i t e 4 W e a W w d Assembly taeattahfeftfttz Proposed Answer a)
OgCI Explanation (Optional):
Technical Reference(s): @ - lZ.S,3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
k Learning Objective: D I 18 (As available)
Question Source: Bank # Duane Arnold 1 INPO # 8781 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 912011999 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge $
Comprehension or Analysis 10 CFR Part 55 Content: 55.41 k 55.43 Comments:
Page 122 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/2003 8:44 AM
I -
Examination Outline Cross-reference: Level SRO Tier # 3 Knowledge of the bases for prioritizing safety Group #
functions during rgency operations.
KIA # 2.4.22 Importance Rating 4.0 Ib.7 S c
Proposed Question: If 8 i n t a i n e d above 2Y$=kbm , Primary Containment Control EOF! directs the operator to "SECURE HPCI (DISREGARD ADEQUATECORE COOLING)."dWhich of the following describes the bases for RClC and HPCI operation under the same EOP circumstances (ie, water level cannot be maintained above . )?
t w z p e r a t i o n may continue ONLY if it is the al=
'07s a high pressure injection system available to provide adequate core cooling.
b) RClC must be secured at the same time as HPCl to minimize the containment RO/SRO pressure rise.
SI00 operation due to exhaust back pressure fluctuation.
Proposed Answer: c) RCIC operation may continue because the turbine exhaust energy does not c- Explanation (Optional):
Technical Reference(s): W nwt (Attach if not previously provided)
Proposed references to be provided to applicants during examination:
Learning Objective: m\7 *!.\\E l,bc; (As available)
Question Source: Bank # Fermi 2 2 I N P W 19714 Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 6/14/2001 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Y
Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 x Comments:
Page 123 of 129 NRC Written Examination Begin to END.doc Last printed 5/21/2003 8:44 AM