ML090360415

Draft - RO & SRO Written Exam (Folder 2)

ML090360415
Person / Time
Site:	Hope Creek
Issue date:	12/19/2008
From:	Public Service Enterprise Group
To:	D'Antonio J M Operations Branch I
Hansell S
Shared Package
ML082600269	List: IR 05000354/2009301 ML090270205 ML090360400 ML090360406 ML090360415 ML090360438 ML090680623 ML090720629 ML090720842 ML090720939 ML091140106 ML091310431
References
50-354/09-301, ES-401, ES-401-5, TAC U01693 50-354/09-301
Download: ML090360415 (355)

Text

ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 WA # 205000 K1.04 I m porta rice Rating Group # 1 2.7 Knowledge of the physical connections and/or cause- effect relationships between SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and the following:

Fuel pool cooling assist: Plant-Specific Proposed Question: Common 1 Which ONE of the following describes the Fuel Pool Cooling Assist flowpath through RHR and the effect of a lowering Skimmer Surge Tank level. A. Flow is through any RHR Loop returning to the spent fuel pool. As Skimmer Surge Tank level lowers, NPSH would be diminished for the RHR pump. B. Flow is through ONLY RHR Loops A or B returning to the spent fuel pool. As Skimmer Surge Tank level lowers, NPSH would be diminished for the RHR pump.

C. Flow is through ONLY RHR Loops A or B returning to the spent fuel pool. As Skimmer Surge Tank level lowers, RHR pump NPSH would be unaffected.

D. Flow is through any RHR Loop returning to the spent fuel pool. As Skimmer Surge Tank level lowers, RHR pump NPSH would be unaffected.

Proposed Answer: B Explanation (Optional):

B. Correct Per FPCCS Lesson Plan, Page 15, Section lll.B.7 Skimmer Surge Tanks Provide net positive suction head (NPSH) for the FPCCS pumps and a RHR pump when operating in the augmented fuel pool cooling mode. Per FPCCS Lesson Plan, Page 24,Section III.C.l .c,,2). Flow is from the skimmer surge tank outlet to the Rl-IR System (either Loop A or B). Return from the RHR System is to the spent fuel storage po~ol only via a set of dedicated diffusers.

B. Correct. A. C. D. Incorrect. Only RHR Loop A

& B is used Incorrect. RHR pump NPSH is also affected Incorrect. Only RHR Loop A

& B may be used.

RHR pump NPSH is also affected Page 1 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Tech n ica I Proposed Reference(s) LP- NOH01 FPCCOO-05 (Attach if not previously provided)

~~ references to be provided to applicants during examination:

NONE Learning Objective:

Question Source: Question History: Question Cognitive Level: 10 CFR Part 55 Content: RHRSYSE003 (As available)

~ Bank # Modified Bank # - (Note changes or attach parent)

New X _I Last NRC Exam Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 X Comments:

I Page 2 of 205

-pj kF LESSON NAME:NOHOI FPCCOO-05 FUEL POOL COOLING AND CLEANUP SYSTEM - 07/11/06 I SOER 85-1 I Leak detection is provided by a drain line located at a low point outside of the seal.

The leakage would fill a level/flow switch identical to the type used on the spent fuel storage pool gates. The switches will then overflow to the drywell sump; which adds to the IDENTIFIED leak rate. The level switch causes the same overhead alarm as spent fuel storage pool gate leakage and drywell to reactor building seal leakage (Window Dl-B5). The level switch also alarms an individual cornputer point. IOBJ. 3d I 7. Skimmer Surge Tanks a. Purpose 1) Pravide net positive suction head (NPSH) for the FPCCS pumps and a RHR pump when operating in the augmented fuel pool cooling mode. The skimmer surge tanks also act as a surge volume to accommodate water displaced by the largest piece of equipment that will be set in or removed from the storage pools. 2) b. Characteristics There are two skimrner surge tanks, each having a capacity of 3750 gallons. The tanks are connected by an equalizing line which ensures the levels are maintained the same in both tanks. Each tank has a removable debris screen which prevents large objects from entering the FPCCS. The tanks are sealed in individual concrete rooms with access via plugs on the Refuel Floor. IOBJ. 8a, b I c. Normal make-up to ,the spent fuel storage pool is via a level control valve (LV-4660) from the Condensate Storage and Transfer System (AP). In the event of an emergency, make- up to the spent fuel storage pool can be supplied by the SSWS, Fire Water System, or RHR System. There are service boxes on the Refuel Floor for demineralized water and from the Condensate Storage and Transfer System that can also be used as sources for make- UP- Page 15 of 39 LESSON NAME:NOHOI FPCCOO-05

$\ FUEL POOL COOLING AND CLEANUP SYSTEM - 07/11/06 ~~~ ~ ~~ ~ ~ ~ ~ ~~~ ~ 0 Spent Fuel Storage Pool to Fuel Cask Storage Pit gate area 0 Fuel Cask Storage Pit b) Discharge connections 0 Condensate Storage and Transfer System 0 Reactor Cask Storage Pit Fulel Cask Storage Pit c. RHR System Augmented Fuel Pool Cooling Mode [M-51, Sheet 1, Nofe 25 1 1) Whenever the RHR System is operated in the augmented fuel pool cooling mode, flow element FE-4462A(B) (containment spray flow) must be removed and replaced with a special flow restricting orifice specially designed for this mode of operation.

This is due to the design capacity of the skimmer surge tank weirs being less than rated RHR pump flow. Fbw is from the skiinmer surge tank outlet to the m- owpkm@ f&&umfrom the RHR System is to the spent fuel storage pool only via a set of dedicated diffusers.

These diffusers are equipped with vacuum breakers.

2& IOBJ. 7b -7 2. Torus Water Cleanup The Torus Water Cleanup pump takes a suction from the torus through a strainer to keep large material from entering the pump suction. The water then passes through two containment isolation valves (HV-4680 and HV-4681) to the pump. The pump discharge is directed to the filter-demiineralizers through Reactor Building isolation valve HV-4656. From the filter-demineralizers, water flows back into the Reactor Building and through isolation valve HV-4663. Next, flow is directed back to the torus through two containment isolation valves (HV-4679 and HV-4652).

I OBJ. 8a, b -i 3. Spent Fuel Storage Pool Level Page 24 of 39 I , ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 KIA # 218000 K1.02 I m portarice Rating 4.0 Knowledge of the physical connections and/or cause-effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following:

Low pressure Core Spray Proposed Question: Common 2 Which one of the following describes the ADS response to the ECCS pump discharge pressure permissive contact opening during the period of time the 105 second Delay Timer is timing out? (Assume all other permissives continue to be met and NO operator actions are taken). The Delay Timer: A. stops and ADS actuation will NOT occur even if the ECCS pump discharge pressure permissive contact is closed. B. continues to time out and the ADS actuation will immediately occur when the ECCS pump discharge pressure permissive contact is closed. C. de-energizes, resets to zero, then when the ECCS pump discharge pressure permissive is met, the Delay Timer starts a second cycle. D. stops until the ECCS pump discharge pressure permissive contact is closed at which time ADS will initiate after the Delay Timer completes the cycle. Proposed Answer:

B Explanation (Optional): IAW SN-0001:

3.3.1. The following signals (both Sub Channels B and F (or D and H) need to be energized) auto initiates ADS 1. Drywell Pressure (1.68 psig) (Seal-In)

OR 5 minute timer times out (for line breaks outside drywell)

AND 2. Level 1 (-129 inches)

AND 3. Confirmatory Level 3 (12.5 inches) AND 4. 105 second time delay AND 5. Core Spray discharge pressure 145 psig OR RHR Pumps discharge pressure 125 psig. Once the 105 second timer times out Core Spray and RHR pump discharge pressure is evaluated.

ADS is actuated if the discharge pressure permissives are met. Page 3 of :205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

9. Correct - LP ECCS status is evaluated after the 105 sec timer times out. A. C. D. Incorrect - the 105 sec timer does not stop Incorrect - there is no second cycle for the 105 second timer Incorrect - the 105 sec timer does not stop Technical Reference(s)

HC.OP-SO.SN-0001 section (Attach if not previously provided)

3.3.1 Proposed

references to be provided to applicants during examination:

NONE Learning Objective: ADSSYSE002 (As available) 53287 Question Source: Bank

1. - Modified Bank # New (Note changes or attach parent)

-. - Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 4 of 205 I 3.2.1 1. 3.2.12. 3.2.13. 3.2.14. 3.2.15. HC . OP-S0.S N -000 1 (a) OBSERVE the Safety Relief Valve Acoustic Monitor operability requirements IAW TIS 3.4.2 1 OBSERVE the ADS requirements (of ECCS) IAW T/S 3.5.1. -. -. OBSERVE the SafetylRelief Valve and SafetyiRelief Valve Low-Low Set Function operability requirements IAW TIS 3.4.2.2. OBSERVE the Safety/Relief Valve position indication requirements IAW T/S 3.3.7.4 and 3.3.7.5. SRV Low-Low Set Function setpoints:

PSV-FO13H PSV-FO13P PSV-FO13H Open at 1047 psig; Close at 905 psig Open at 1047 psig; Close at 935 psig Subsequent opening 101 7 psig 3.3 Interlocks 3.3.). The following signals (both Sub Channels B and F (or D and H) need to be energized) auto initiates ADS: A. For Channel B (or D) 1. Drywell Pressure (1.68 psig) (Seal-In) - OR 5 minute timer times out (for line breaks outside d rywell) AND - 2. Level I (-129 inches) AND Confirmatory Level

3 (12.5 inches) AND 3. 4. 105 second time delay

/-\ND Core Spray discharge pressure 145 psig - OR RHR Pumps discharge pressure 125 psig. 5. (continued on next page) Hope Creek Page 4 of 22 Rev. 7 I 1 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 WA # 215004 K2.01 I m porta rt ce Rating Group # 1 2.6 Knowledge of electrical power supplies to the following:

SRM channelsr'detectors.

Proposed Question:

Common 3 Given the following conditions:

0 The plant is performing a startup 0 All equipment is operable 0 The RPS shorting links are installed Then, the SRM 'A drawer loses power.

Which one of the following describes what supplies the SRM

'A drawer and what is the plant response?

A. 125VDC A Reactor Scram and rod block will occur. B. +24VDC.

A Reactor Scram and rod block will occur. C. +24VDC. A rod block will occur. A Reactor Scram will NOT occur. D. 125 VDC A rod block will occur. A Reactor Scram will NOT occur Proposed Answer: C Explanation (Optional):

IAW RPS Lesson Plan NOH01 RPSOOC An SRM INOP condition will not initiate a Scram signal. [CR 960218138]

IAW SRM Lesson Plan - Section VII.B.2.a. - The +24 VDC supplies the detector HVPS.

Loss of this power will result in a loss of the HVPS and generate a channel INOPERATIVE trip. Page 5 of 205 ES-401 Sample Written Eiamination Form ES-401-5 Question Worksheet Loss of 224 VDC - The +24 VDC supplies the detector HVPS. Loss of this power will result in a loss of the HVPS and generate a channel INOPERATIVE trip. Loss of 1 BJ484 - Results in a loss of control and indications for the SRMS on 1 OC651. The SELECT, POSITION STATUS, POWER ON, DRIVING IN/OUT circuitry is lost. IAW Section V.A.2.- f 24 VDC Power System - supplies detector polarizing voltage and SRM logic modules. C. Correct. - the loss of the 24 VDC will cause a loss of SRM logic modules and an inop trip and a withdraw block. No scram signal will occur A. B. D. Incorrect. - +24 VDC is the power supply to the SRM HVPS ,A scram signal will not occur, Incorrect. - A scram signal will not occur Incorrect. - +24 VDC is the power supply to the SRM HVPS. Technical Reference(s)

LP - NOH01 RPSOOC-05 (RPS) (Attach if not previously provided) LP- (SRM) Proposed references to be provided to applicants during examination:

None Learning Objective:

SRMSYSEOI 3 (As available) Question Source: Bank

1. Modified Bank

1. New (Note changes or attach parent)

X -. -. Question History: Last NRC Exam Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 6 of 2105 LESSON NAME: SOURCE RANGE MONITORING (SRNI) SYSTEM, ,NOH04SRMSYSC-00.-

.07/18/06, b Fuel loading chambers used are substituted for normal in core SRM detectors.

1) Mounted inside dummy fuel assemblies to facilitate their movement by the refueling bridge grapple

2) Output signal connected to SRM pre-amplifier in place of normal detector signal

3) When sufficient fuel assemblies have been loaded into the core to produce a background count rate of at least 3 cps, the normal SRM delectors may be reconnected and the fuel loading chambers removed (Obj. 11 I 8. AbnormaVErnergency Operations

1. Review the following procedure relative to SRMS operation:

OP-AB.IC-0004, Neutron Monitoring Ensure symptoms (system response) and immediate operator actions are discussed.

2. SRM Malfunctions IObj. 13 - a. Loss of 524 Vecl The +24 VDC supplies the detector HVPS.

Loss of this power will result in a loss of the HVPS and generate a channel INOPERATIVE trip. Refer to Incident Report 86-067 in the Plant and Industry Event section. b. Loss of 10Y202 The detector cannot be moved because there is no power to the drive motor.

c. Loss of 1 BJ484 Results in a loss of control and indications for the SRMS on 10C651. The SELECT, POSITION STATUS, POWER ON, DRIVING IN/OUT circuitry is lost. The detector cannot be moved. C. Plant and Industry Events Page 29 of 38- C \Documents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\Refen?nces\LESSON PLANS\Source Ranqe Monitonnq\Master\Lesson Plans\NOH04SRMSYSC-OO Source Rancle Monitorinq System doc, ___ __~_ - - Deleted: SOURCE RAP 3E MONlTORlNG (SRM) s' STEM 1 ___-___--~-

__ _ - Deleted: SOURCE RAE ;E 1 MONITORING (SRM) S' STEM Deleted: NOH04SRMS SC-00 j Deleted: 07/18/06 Formatted:

Font 12 ut I _______ _________ - -___J _____ I . ________ --_ ________._

____ - J Formatted:

Font: 12 pt _____ ___ - ___ Formatted:

Indent: Left 1.Y, Numbered + Level: 1 + N ,mbering Style: Bullet + Start at: 1 Alignment.

Left + Aligned rt: 3" + Tab after: 0 + Indent at 3.25" - ___. _- ___ -- ~ Deleted: 36 Deleted: S \VDrive\TRAIh NG DOCUMENTS\Operations Training\Hope Creek\Plant Technology\Systems\Sourr;e Range Monitoring\Master\Lesson Plans\NOH04SRMSYSC-Ot~

Source Range Monitoring System *oc Deleted: S \UDnve\TRAll\

NG DOCUMENTS - WORKlNC FOLDERS\Operations\Hop, Creek\Plant Technology\System\Sourcc Range Monitoring\NOHO4SRMSY:

2-00 Source Range Monitoring System doc I LESSON NAME: NOH01 RPSOOC-05 I REACTOR PROTECTION SYSTEM - 01/10/2008 1 NOTE An SRM INOP condition will not initiafe a Scram signal. [CR 960278738]

1 e) SRM high flux scram setpoint:

2x1 O5 CPS 2) Intermediate Range Monitor: Protects against rapid increases in core neutron flux levels between the startuF range and the power range during reactor startup and shutdown operations a) 8 IRM channels (A thru H) provide upscale and inoperative scram inputs b) IRM channels A, C, E & G input to RPS Trip Systern "A"; Channels B, D, F & H input to RPS Trip System "Bll c) Logic iis 1 -out-of-4-twice 0 1 -out-of-3-twice is minimum requirement (per technical specifications) 0 Allows single IRM input to each RPS trip system to be bypassed d) Trip inputs and setpoints IRM Hi-Hi (1201125 of scale) 0 IRM INOP: mode switch not in OPERATE, detector high voltage low, module unplugged

3) Average Power Range Monitor: Initiate a scram when reactor core average power reaches or exceeds a limitinc safety system setting. a) Six APRM channels provide upscale and inoperative scram inputs b) APRM channels A, C, E input to RPS A. Channel: B, D, F input to RPS B c) Logic is 1 -out-of-3-twice 0 '1 -out-of-2-twice is minimum requirement Page 27 of 48 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008-2009\References\LESSON ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 KIA # 239002 K2.01 Importarice Rating 2.8 Knowledge of electrical power supplies to the following:

SRV solenoids. Proposed Question: Common 4 The plant is operating at rated power. Which one of the following describes the effect of a loss of 125 VDC BD417, on the Automatic Depressurization System (ADS), if the required actuation setpoints were reached? NOTE: Assume all other equipment is operable AND NO operator actions were taken to inhibit ADS A. "A ADS SRV pilot solenoids have lost power. An ADS actuation would occur. B. "A ADS SRV pilot solenoids have lost power. An ADS actuation would NOT occur. C. "B" ADS SRV pilot solenoids have lost power. An ADS actuation would occur. D. "B" ADS SRV pilot solenoids have lost poweir. An ADS actuation would NOT occur. Proposed Answer: A Explanation (Optional):

A. Correct IAW ADS Lesson Plan NOHOIADSSYSC-03 Section V.A.3 - 125 VDC Class 1E Distribution System - The 125 VDC Class 1 E Distribution System supplies electrical power to the ADS SRV pilot solenoids

&the ADS logic channels. ADS Channel B logic & the A ADS SRV pilot solenoids are powered from 1 BD417.

ADS Channel D logic & the B ADS SRV pilot solenoids are powered from 1 DD417. IAW Section III.B.l .a.2)b) - Satisfaction of either ADS Channel B ADS Channel D will result in the actuation of the ADS and the depressurization 01 the RPV. This ensures that a single failure will neither initiate nor inhibit the ADS function.

Page 7 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet B. C. D. incorrect. - Channel D would cause an actuation Incorrect. - DD417 supplies the B ADS SRV stolenoids Incorrect. - DD417 supplies B ADS SRV solenoids, Channel D would cause an actuation Technical Reference(s)

LP NOHOIADSSYSC-03 (Attach if not previously provided) (ADS) Proposed references to be provided to applicants during examination:

none Learning Objective: Question Source:

ADSSYSE007 (As available)

Bank # Modified Bank

1. -. (Note changes or attach parent)

New X Question History: Last NRC Exam _. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 8 of 205 I l Y k-j LESSON NAME: NOHOlADSSYSC-03 AUTOMATIC DEPRESSURIZATION SYSTEM - 01/22/2008 V. SYSTEM INTERRELATIONS A. Support Systems ~OSJ. 7a .I 1. RHR and Core Spray Systems

a. Pressure switches located in the RHR and/or Core Spray pump discharge headers provide permissive signals to the ADS (automatic) initiation logic

!;ubchannels.

1) Pumps are determined to be operating as sensed by pump discharge pressure being above the setpoint.

a) b) RHR pumps: 125 psig The pump discharge pressure permissive logic requires a minimum of one Core Spray loop (two operating pumps) one RHR pump. Core Spray System logic channel trip units provide RPV level logic. Core Spray pumps: 145 psig

2) b. drywell pressure trip outputs to the ADS initiation IOBJ. 76 = 2. PClG System The PClG System supplies compressed (nitrogen) gas for pnuematic operation of the SRV pilot valves. lO5J. 7c I 3. 125 VDC Class 1 E Distribution System The 125 VDC Class I E Distribution System supplies electrical power to the ADS SRV pilot solenoids

a the ADS logic channels. a. ADS Channel B logic and the A ADS SRV pilot solenoids are powered from 1 BD417. b. ADS Channel D logic are powered from 1 DD417. the B ADS SRV pilot solenoids

-3 Formatted:

Bullets and N mbertng 1 ~ Page 27 of 34 CADocumqnts and Settingskdennis\My Docurnents\Hope Creek 20082009\References\LESSON PLANSWutomatic Depressurization\Master\Lesson Plans\NOHOIADSSYSC-O3 Automatic Depressurization System.doc LESSON NAME: NOHOIADSSYSC-03 AUTOMATIC DEPRESSURIZATION SYSTEM - 01/22/2008 source of makeup water is available to the RPV prior to depressurization.

The ADS energizes the pilot solenoids of 5 of the 14 Nuclear Pressure Relief System Safety Relief Valves (SRVs) to depressurize the RPV to the suppression pool.

4. Fig. 7 I OBJ. 3 a. The SRVs associated with the ADS are PSV-F013A,B,C,D and E. 1) 2) 3) The safety lift setpoint of 'each ADS SRV is 1130 psig, which is the highest lift setpoint of the 14 SRVs. The other 9 SRVs have a safety lift setpoint of either 11 08 psig or 1120 psig. The ADS SRVs have the highest safety lift setpoint to minimize valve actuations., thereby ensuring maximum reliability and availability for the ADS function.

b. When the ADS energizes the SRV pilot solenoids, Primary Containment Instrument Gas (PCIG) pressure is admitted to the pneumatic actuator of the SRV pilot assembly, causing the valve to open. NOTE: the function, description and operation of the SRVs and the supporting and interfacing components and systems are contained in the Main Steam System lesson plan NOHOlMSTEAMC 1 B. Major Subsystems/Components

1. ADS Logic Design a. Two divisions (Channels B and [I) of ADS logic are provided:

1) Channel B is comprised of subchannels B F. 2) Channel D is comprised of subchannels D and H a) Both subchannels (B and F D and H) within an ADS channel (division) must be satisfied for the associated ADS channel to initiate depressurization of the RPV. This logic arrangement prevents an inadvertent actuation of the ADS. Channel D will result in the actuation of the b) Satisfaction of either ADS Channel B ADS Page 9 of 34 C:\Documents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\References\l.ESSON PLANSMutornatic Depressurization\Master\Lesson Plans\NOHOIADSSYSC-O3 Automatic Depressurizatior, Systern.doc LESSON NAME:

NOHOIADSSYSC-03 AUTOMATIC DEPRESSURIZATION SYSTEM - 01/22/2008 ADS and the depressurization of the RPV. This ensures that a single failure will neither initiate nor inhibit ihe ADS function.

b. Channels A and C are not use0 by ADS 1) 2) Channels A and C are used by the HPCl System initiation logic.

The ADS is electrically separated from the HPCl System, including sensors and logic circuitry, thereby ensuring that no single failure affecting the HPCl System will impact the ADS, and vice versa 2. The majority of the ADS logic is mounted in two panels located in the Lower Relay Room: Panel 10C628 (Division 2, subchannels B and F) and Panel 10C631 (Division 4, subchannels D and H). a. b. c. d. Associated ADS channel initiation logic and power fuses.

Low-Low Set relief logic (see Main Steam System Lesson Plan NOHOIMSTEAMC). SRV logic and control power fuses. Not included are the process input trip units, which are mounted on the associated ECCS control panels:

I) 2) 3. Drywell pressure input Panel 10C618 (Division 2, subchannels 6 and F). Panel 10C640 (Division 4, subchannels D and H) OBJ. 2 P I Fig. 7 .- a. Each subchannel (B,F,D,H) contains a single high drywell pressure input, which will energize relay K1 (B, F, D, H). 1) ChannelB a) b) Subchannel B: trarismitter N094B, trip unit N694B on Panel 1 OC618. Subchannel F: transmitter N094F, trip unit N694F on Panel 1 OC618. Page 10 of 34 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008-2009\References\LESSON PLANSMutornatic Depressurization\Master\Lesson Plans\NOHOIADSSYSC-O3 Automatic Depressurization System.doc

@!?? LESSON NAME: NOHOIADSSYSC-03

\ AUTOMATIC DEPRESSURIZATION SYSTEM - 01/22/2008 V. SYSTEM INTERRELATIONS A. Support Systems IOBJ. 7a I 1. RHR and Core Spray Systems a. Pressure switches located in the RHR and/or Core Spray pump discharge headers provide permissive signals to the ADS (automatic) initiation logic subchannels.

I) Pumps are determined to be operating as sensed by pump discharge pressure being above the setpoint.

a) b) RHR pumps: 125 psig The pump discharge pressure permissive logic requires a minimum of one Core Spray loop (two operating pumps) gone RHR pump. Core Spray System logic channel trip units provide RPV level and drywell pressure trip outputs to the ADS initiation logic. Core Spray pumps: 145 psig 2) b. IOBJ. 76 1 2. PClG System The PClG System supplies compressed (nitrogen) gas for pnuematic operation of the SRV pilot valves. h '\\ / = jOBJ.7C / ' '\, 5 VDC Class 1 E Distribution System e 125 VDC Class 1 E Distribution System supplies electrical wer to the ADS SRV pilot solenoids 2nd the ADS logic channels.

the A ADS SRV pilot solenoids are . ADS Channel B logic powered from 1 BD417.

b. ADS Channel D logic and the B ADS SRV pilot solenoids are powered from 1 DD417. - 7 [Formatted:

Bullets and Nu nberlng Page 27 of 34 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008-2009\References\LEiSSON PLANSMutomatic Depressurization\Master\Lesson PIans\NOHOlADSSYSC-O3 Automatic Depressurization System.doc ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Out I i ne Cross-reference:

Level RO SRO Tier # 2 KIA # 217000 K3.03 Group # 1 I m porta rice Rating

3.5 Knowledge

of the effect that a loss or malfunction of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) will have on following: Decay Heat Removal.

Proposed Question: Common 5 Following an automatic scram from 100%

RTP due to closure of the MSIVs, a steam leak developed on the RClC steam line in the pipe routing area (pipe chase) and temperature is currently 162 degrees and rising at 2 degrees per minute. Which one of the following identifies the method(s) available to reduce RPV pressure and commence reactor cooldown 30 minutes from now? A. SRVs ONLY.

B. RClC & HPCl ONLY.

C. HPCl and SRVs ONLY D. SRVs, RCIC & HPCI. Proposed Answer: C Explanation (Optional): IAW Lesson Plan (RCIC)

Section IV.E.2.a.7) - auto isolation - RClC Steam Pipe Area High High Temperature;

>160°F for 30 minutes.

Section IV.E.5.a.2) - RClC Trip is caused by any RCIC lsoaltion C. Correct - HPCl & RClC steam line supplies are separate.

SRVs are unaffected and available.

RClC will trip on isolation signal of >I60 degrees for 30 minutes A. B. D. Incorrect - HPCl would also be available Incorrect - SRVs would also be available. RCIC would have tripped on the isolation signal Incorrect - RClC would have tripped on the isolation signal Technical Reference(s)

RClC Lesson Plan (Attach if not previously provided)

' Page 9 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question ___. Worksheet NOH04RCIC00-05 Proposed references to be provided to applicants during examination:

NONE Learning Objective: Question Source:

Question History: Question Cognitive Level:

10 CFR Part 55 Content: RCICOOE02 1 (As available)

Bank # INPO Bank, Susq 2002 Modified Bank # -. (Note changes or attach parent)

New Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Comments: Page 10 of 205

.LESSON PLAN: Nh)M0.4RCICOO-Q5, RClC Room Ventilation High Differential Temperature; >70"F. quipment Area High Temperature; >160°F.

b Torus Room High Temperature; 2128°F for 30 minutes. 7) RClC Steam Pipe Area High High Temperature;

>16OoF for 30 minutes. At Panel 621 for Channel B (640 for D), switch S5B (S5D) in the bypass position removes all temperature inputs to the RClC isolation.

Pushbutton SIB for Channel B (SI D for D channel) when depressed can bypass the 30 minute time delay. These are normally used for test purposes. Automatic Isolation Logic Train Actuations are arranged as follows: 1) b. c. d. Isolation Logic Train "B", when actuated, performs the following:

a) Closes HV-F008 (Outboard Steam Supply Isolation Valve) b) Trips closed HV-4282 (Turbine Trip Throttle Valve) Isolation Logic Train "D", when actuated, performs the following:

a) Closes HV-F007 (Inboard Steam Supply Isolation Valve) b) Closes HV-F076 (Inboard Steam Supply Line Warmup Valve) c) Trips closed HV-42182 (Turbine Trip Throttle Valve) When the isolation condition is cleared, the logic must be reset by taking the NORM/RESET keylock switch to the reset position.

The switches for F007 and F008 must be matched with the valve position before valve movement is allowed. 2) e. 3. RClC System Manual Isolation

a. Manual Isolation of the RClC System is accomplished by depressing the "Isolation Logic

'El' Trip" Pushbutton

~ ~ j Deleted: 01117106 I Deleted: lZ20106 I ____ __ - ~- _ 1 Formatted: Bullets and N rnbering 1 - _-

REACTOR CORE lSOLATiODd COOL.1NG - -q012910T

1) Low steam supply line pressure 64.5 psig

2) Isolation Logic Train 'B' actuation closes HV-F062 (Outboard Vacuum Breaker Isolation Valve) Isolation Logic Train

'D' actuation closes HV-F084 (Inboard Vacuum Breaker Isolation Valve) High drywell pressure 1.68 psig

b. c. 1 "Obj. x. Fia. ?*3 5 RClC Turbine Trip

a. The RClC Turbine Trip Signals are as follows Turbine Trip Pushbutton RCIC Isolation (Man or Auto). Pump Suction Low Pressure (15" Hg Vac). Turbine Exhaust High Pressure (apsig). Mechanical Overspeed Q2J% speed). High Reactor Water LeveULevel8

(+ 54 inches) Upon a turbine trip, either the turbine trip throttle valve 4282 closes (items 1-5 above) or I the steam supply valve F045 closes (for item 6)., -- __ __ [Ged: Bullets and lumbering 1 _________-- .

ES-40 1 Sample Written Examination Form ES-40 1 -5 Question ___. Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 WA # 206000 K3.03 Importance Rating

3.4 Knowledge

of the effect that a loss or malfunction of the HIGH PRESSLJRE COOLANT INJECTION SYSTEM will have on following:

Suppression pool level control: BWR-2,3,4. Proposed Question: Common 6 HPCl is in full flow test for RPV pressure control following a reactor scram. RPV level has recovered to 35 inches and is stable. All other containment parameters are normal. Then, logic control power is lost to valve BJ-HV-F04.2 "PMP SUCT FROM SUPP CHB" Which one of the following describes the effect of the loss? A. CST level would be adversely affected because the valve would NOT auto-open on a HIGH CST level signal.

B. CST level would be adversely affected because the valve would NOT auto-close on a LOW CST level signal.

C. Suppression Pool level would be adversely affected because the valve would NOT auto- close on a LOW Suppression Pool level signal. D. Suppression Pool level would be adversely affected because the valve would NOT auto- open on a HIGH Suppression Pool level signal. Proposed Answer:

D Explanation (Optional): IAW HC.OP-BJ-0001 - BJ-HV-F042 PMP SUCT FROM SUPP CHB-Auto closes on HPCl Div 1 Isolation signal (K51A). Auto opens on CST low level OR Suppression Chamber high level (K42), IF BJ-HV-F042 handswitch is not in AUTO OPEN OVRD AND no HPCl Div 1 Isolation signal (K51A).

Opens manually IF no HPCl Div 1 Isolation signal (K51A). D. Correct - valve would not auto open on SP level high A. B. Incorrect - valve auto open on low CST level Incorrect - valve only auto closes on an isolation signal Page 11 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet C. Incorrect - valve does not auto close on low SP level Technical Reference(s) HC.OP-BJ-0001 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

NONE Learning Objective:

HPCIOOE012 (As available) Question Source:

Bank # 53735 -. (Note changes or attach parent)

-. Modified Bank

1. New __. Question History: Last NRC Exam

__ Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments: Page 12 of 205 I HC.OP-SO.BJ-000 1 (Q) 3.3.4. HPCI valve interlocks are as followv: e e e e FD-HV-F002 INBD HPCI STbI ISLN MOV Auto Closes on HPCI Div 3 Isolation signal (K50C). Opens manually if no HPCI Div 3 Isolation signal (K50C). E Isolation signal was sealed in (K92) with FD-HV-F002 handswitch in OPEN THEN handswitch must be placed in CLOSE THEN OPEN. __ FD-HV-F003 OUTB HPCI SThJ ISLN MOV Auto closes on HPCI Div 1 Isolation signal (K50A). Opens manually if no HPCl Div 1 Isolation signal (K50A). E Isolation signal was sealed in (K93) with FD-HV-F003 handswitch in OPEN THEN handswitch must be placed in CLOSE THEN OPEN. ___ FD-HV-F001 HPCI TURB STR'I SPLY Auto opens on HPCl Initiation (K33), E FD-HV-F071 TURB EXH ISLN is full open. Opens manually E FD-HV-F07 1 TURB EXH ISLN is full open. FD-HV-Fl 00 HPCI W/U VLV Auto closes on HPCI Div 3 Isolation signal (K50C & K52C). Opens manually no HPCI Div 3 Isolation signal (K50C). - BJ-HV-F004 PMP SUCT FROhil CST Auto closes BJ-HV-F042 PMP SUCT FROM SlJPP CHB is full open (K43). Auto opens BJ-HV-F042 PMP SUCT FROM SUPP CHI3 is not full open (K43)

__ ___ Auto closes on HPCI CST low level LSuppression Chamber high level (K42), BJ-HV-F042 handswitch no *PCI Div 1 Isolation signal (K5 1 A). is not in AUTO OPEN OVRD AND no HPCI Div 1 Isolation signal (K51A). Opens manually

___ e BJ-HV-F007 PMP DSCH ISLN Auto closes E Test Selector switch is in HV-F007 position AND Test Jack installed (K66). Auto opens on HPCI Initiation (K32) AND not in test (K66). - BJ-HV-FO06 PMP DSCH TO CY ISLN Auto closes E FD-HV-FOOl HPCI TURB STM SPLY is full closed (K44) Q& Test Selector switch is in HV-F006/HV-8278 (HV-F105) (AND Test Jack installed) (K65) on HPCI hitiation (K90) FD-FV4880 HPCI TURB STOP VLV is full closed. Auto opens FD-FV4880 HPCI TURB STOP VLV is not full closed (K39) FD-HV-F001 HFCl TURF: STM SPLY is not full closed (K44) AND not in test (K65). Continued on next page Hope Creek Page 12 of 57 Rev. 34

, I ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 300000 K4.02 I m porta rice Rating

3.0 Knowledge

of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following: Cross-over to other air systems Proposed Question: Common 7 The plant is operating at rated power when the following annunciators are received:

A2-AI, INST AIR HEADER A PRESSURE LO A2-A2, INST AIR HEADER B PRESSURE LO A2-B1, COMPRESSED AIR SYSTEM TROUBLE A2-B2, COMPRESSED AIR PANEL OOCl88 Current air pressures are: 0 Service Air pressure is 87 psig. 0 Instrument air pressure at the Emergency Instrument Air Receiver is 82 psig.

0 Instrument air pressure at the Instrument Air Receivers is 77 psig. What is the configuration of the Service and Instrument Air System? A. Instrument Air Dryer 1AF104 Isolation Valve, tiV-I 1416, will be open. The Standby Service Air Compressor will be running. B. The Service Air Supply Header Isolation Valve, HV-7595, will be closed. The Standby Service Air Compressor will be running.

C. Instrument Air Dryer 1AF104 Isolation Valve, HV-11416, will be closed. The Standby Service Air Compressor will NOT be running. D. The Service Air Supply Header Isolation Valve, HV-7595, will be closed. The Standby Service Air Compressor will NOT be running. Proposed Answer: A Explanation (Optional): IAW HC.OP-AB.COMP-0001

& Lesson Plan NOH01 INSAIR-02 (Instrument air) - Section IV.C.l .g. - page 57 - As the loss of air event starts, there is very little effect on plant operation. As air pressure begins to decrease, some automatic actions will occur that will attempt to stop the loss of air. Page 13 of 2105 I I ES-401 Sample Written Examination Form ES-40 1 -5 Question Worksheet 0 0 0 Standby Service Air Compressor starts (92 psig) Emergency Instrument Air Compressor starts (85 psig)

Instrument Air Dryer 1AF104 isolation valve (HV-11416) opens (85 psig) Service Air Header Isolation Valve (HV-7595) cl'oses (70 psig Instrument Air pressure)

A. Correct. The Emergency Instrument Air Compressor starts at 85#, HV-11416 opens at 85# B. C. D. Incorrect - The Service Air Supply Header Isolation Valve, HV-7595 will be open Incorrect - Instrument Air Dryer 1AF104 Isolation Valve, HV-11416 opens at 85# Incorrect - The Service Air Supply Header Isolation Valve, HV-7595 closes at 70#, the Standby Service Air Compressor starts at 92#. Technical Reference(s) HC.OP-AB.COMP-000'1 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: INSAIRE006 (As available)

Question Source: Bank # 56927 Modified Bank

1. -. (Note changes or attach parent)

New -. Question History: Last NRC Exam ___ Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 14 of 205 I HC.OP-AB.COMP-O@Ol(Q)

INSTRUMENT AND/OR SERVICE AIR IMMEDIATE OPERATOR ACTIONS Trip of the Inservice Service Air Compressor PLACE the out of service Service Air Compressor in service. (AB.ZZ-0001)

I DatelTime: I I I < 92 psig Service Air Pressure.

I Standby Service Air Compressor Auto Start 1 ~~ ~ 5 85 psig Emergency Instrument Air Receiver Pressure I EUiCAutoStart e RACS Demineralizers Isolate 5 85 psig Instrument Air Pressure - OR Loss of Power to 1A-F-104 I-KBHV-11416 OPENS 1 I 70 psig Instrument Air Pressure I 1-KARV-7595 AUTO CLOSES I Page 3 of 20 Rev. 3 I I ~. ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 KIA # 262001 K4.05 I m porta rice Rating

-~ 3.4 Knowledge of A.C. ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the following: Paralleling of A.C. sources (synchroscope). Proposed Question:

Common 8 A unit startup is in progress, the main generator is being synchronized to the grid. The following indications are observed:

0 Keylocked SYNC SCOPE switch in the ON position Sync Scope and voltages indicate the first main generator output breaker is ready for closure The operator depresses the BS2-6 CLOSE pushbutton, but the breaker fails to close.

Which condition prevented breaker closure? A. The 52x60 Generator Disconnect is open.

6. The main generator exciter field breaker is open. C. The SYNC CHECK ON pushbutton was NOT held depressed before depressing the CLOSE pushbutton.

D. The SYNC CHECK OFF pushbutton was NOT held depressed before depressing the CLOSE pushbutton.

Proposed Answer: D

Reference:

HC.OP-SO.MA-0001 - Section 5.2.16. PERFORM the following to synchronize the Main Generator using Manual Load Control:

D. PERFORM the following (with Steps 2 thru 5 being performed in rapid succession): 1. WHEN SYNCHROSCOPE Pointer is at 5 minutes before 12 O'clock position, PRESS AND HOLD SYNCH CHECK OFF push-button.

2. WHEN SYNCHROSCOPE Pointer is at 2 minutes before 12 O'clock position, CLOSE BS 6-5 (BS-2-6) Breaker.

Page 15 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet

3. RELEASE the SYNCH CHECK OFF push-button.

D. A. 6. C. Correct - The SYNC CHECK OFF pushbutton was not held depressed before depressing the CLOSE pushbutton. The OFF button is required to be held IAW HC.OP-SO.MA-0001.

Incorrect - The 52x60 Generator Disconnect is open.

For the sync scope and voltages to indicate that the machine is properly synchronized the exciter field breaker and 52x60 must both be closed. Incorrect - The SYNC CHECK ON pushbutton was not held depressed before depressing the CLOSE pushbutton. The OFF button is required to be held IAW HC.OP-SO.MA-0001.

Incorrect - The main generator exciter field breaker is open. For the sync scope and voltages to indicate that the machine is properly synchronized the exciter field breaker and 52x60 must both be closed. Technical Reference(s)

HC. OP-SO. MA-000 1 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: Question Source:

Question History: Question Cognitive Level: MNPWROEO16 (As available)

Bank # 56833 Modified Bank # (Note changes or attach parent)

-. New Last NRC Exam - Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments: Page 16 of 205 HC.OP-SO.MA-0001 (Q) CAUTION After synchronization and closing a Generator Breaker, Load Setpoint should be increased to 30% IMMEDIATELY.

If the following steps are not completed or expected indications received, the CRS should consider tripping the Main Turbine and ensuring the Generator is isolated from the 500 Kv System and the Turbine is coasting down. A 5.2.1 6. PERFORM the following to synchronize the Main Generator using Manual Load Control: A. SELECT -1, )f!ed-Laad/

NOTE Ramp Rate will indicate 20 O/O/min until breaker is closed.

Upon breaker closure, ramp rate will be 60%/min far 2 seconds (2 Yo load). Ramp Rate will then go to setpoint entered.

B. SELECT Load Set, AND ENTER 10 %/min C. SELECT Load Set, Manual Adj.:

OR Fl until pointer on SYNCHROSCOPE is moving slowly in the FAST direction. (I OC651D) Continued on next page Hope Creek Page 15 of 41 Rev. 47 I HC.OP-SO.MA-0001 (a) 5.2.16 (continued) 5.2.1 7. 5.2.18. 5.2.19. 5.2.20. 5.2.21. Hope creek D PERFORM the following (with Steps 2 thru 5 being performed in rapid succession):

1. - WHEN SYNCHROSCOPE Pointer is at 5 minutes before 12 O'clock position, PRESS AND HOLD SYNCH CHECK OFF push-button. - WHEN SYNCHROSCOPE Pointer is at 2 minutes before 12 O'clock position, CLOSE BS 6-5 (BS-2-6) Breaker.

RELEASE the SYNCH CHECK OFF push-button.

2. 3. 4. IMMEDIATELY SELECT Load Set, : I Setpoind AND ENTER 30 %. 5. QBSERVE Bypass Valves close as Generator loads to 30 % Load Set. 6. IMMEDIATELY OBSERVE the following:

0 Generator phase current increases 0 MW load and MVAR load increases Synchroscope pointer steady at 12 o'clock position 7. ADJUST Generator MVARs to within the limits of Excitation Limit Curve of Attachment

1. PLACE the 500 KV BUS BS 6-5 (BS-2-6) SYNCHROSCOPE in OFF using Control Room Bezel Key. PLACE the 500 KV BUS BS 2-6 (BS 6-5) SYNCHROSCOPE in ON using Control Room Bezel Key. CHECK SYNCHROSCOPE Pointer remains steady at 12 o'clock position.

PRESS AND HOLD BS 2-6 (BS 6-5) SYNCH CHECK OFF push-button.

CLOSE BS2-6 (BS 6-5) AND RELEASE SYNCH CHECK OFF push-button.

Page 16 of 41 - Rev. 47 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 WA # 264000 K5.06 Group # 1 Importance Rating 3.4 Knowledge of the operational implications of the following concepts as they apply to EMERGENCY GENERATORS (DIESELIJET) Load Sequencing.

Proposed Question: Common 9 The plant experienced a loss of 10A401. The bus was subsequently restored to the normal lineup.

However, before the load sequencer was reset, all power was again lost to 10A401. How will the EDG and electrical distribution system respond to this event?

A The EDG will automatically start. The loads on bus 10A401 will sequence on after the output breaker is closed.

B. The EDG will automatically start. The loads on bus 10A401 will NOT sequence on after the output breaker is closed.

C. The EDG will require a manual start. The loads on bus 10A401 will sequence on after the output breaker is closed.

D. The EDG will require a manual start. The loads on bus 10A401 will NOT sequence on after the output breaker is closed. Proposed Answer:

B Explanation (Optional):

B. Correct - The EDG will start on the loss of power to the bus. Without the sequencer being reset, no loads will sequence onto the bus. A. Incorrect - The loads will not sequence on. C. Incorrect - The EDG will auto start when the bus loses power.

D. Incorrect - The EDG will auto start when the bus loses power. Without the sequencer being reset, no loads will sequence onto the bus.

Page 17 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s) AB-ZZ-135 (Attach if not previously provided) 1 EACOOE007 Proposed references to be provided to applicants during examination:

None Learning Objective: (As available) Question Source:

Bank # INPO Bank Modified Bank # - (Note changes or attach parent)

New 25444 -. Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 18 of 205 LISER RESPONSIBLE FOR VERIFYING RLVISION.

STATU: AND CH4NGES VTD PJ81@Q-0@97 00; 3 PRINTED 3C081209 . .. I . ah energize signal to the thirty (30:) BLOCK relays. , I .. . . .. . . . - . - . . . - - . . - . - . .. . -.

, I I LISEF? RESPONSIBLE FOE VERIFYING REVISION.

STATSJ3 AND CHANGES VTD PL1810Q-O097 001 3 PRINTED 20061209 0. LOP during. LOCA Seauencina - Shm:fi an LOP signal the LOCA =flrn- When bus ratoz, the

.._..- -nd the LOCA. nu af the sequence. - .--..-a*ly reset at ' TM9N98 VOLUME I' -----_- arrive when the LOdA is-siquencing, sequencing stops and resets to time ____. power returns via the standby Diesel Gene LOCA signal overrides the LOP sequen-r a sequencer restarts from the beginnil., __ The Emergency Load Sequencer must be m=-l* the end of the sequence..

.. * . E. LOCA after LCL. Sequencing is Completed - The LOCA /I / signal sheds the non-LE loads (not part of this -6ystem) and initiates aoqunncod signals to load?. required *0.: the sefe shutdown of the reactor. EmeLgency Load Sequencer must ba reset at the und of / , /, The the sequence.

_.-- - -- F. LOP after LOCA Sequencing is Completed - If the LOCA eignal is still. present at the system; the LOCA train Breaker closes, overrides the LOP sequence and safe shutdown of the reactor. Sequencer must be manually reset at tht end of the sequence. . I, ,'. / , i:.L-- . is reinitiated as soun as the Standby Diesel Generater initiatks sequenced signals to loads required for the .t, I. The Emergency Load I .. . G. LOCA with Simultaneous LnD - *ha ?fin* -A"--* L-.--- . precedence.

As soon as Breaker closes, the LOCb sequence, initiates the 1 sequenced signals to !.>a1 must be manuallv resat at tka nnaa ck- - . shutdown of the reactor. ..._ a-..D.=lS"bJ YU4 ". - ... = y~b.n u~yKrU+ caKes the Standby Diesel Generator signal overrides the LOP LOCA train.and provides 3s required for the safe Note that in all of the situations listed above, the PSIS (Process Start Inhibit Sianalt =inn31 is generated to --=----I -*3... prevent automatic pro&a-related, 01 starts of Rnerqencv toad SeuiionFnr t c maintained contact a ---- -- =--..-.-* System controlled equipment befoie receiving the programmed train start signal. completed.

PSXS is reset when the system sequence is manually resat. The PSIS remains effective until the sequencing is Upon completion of the Sequencing cycle, the H.' Manual Test Operation - The vari-- =--I-----

-+ this system are fully detziled i.. UTrLAUaI &.J. I. LEA during Mancal Test - A LOCA n-r*-r+**-

A*-;-- - manual test start sequence takes automatic LaCa st-n t-=in an-*.-- , ..--U&LIII~

uuLrrry a precedence and --- ---.. ---r *.CULb4 s=:yuellcing commences to provide signals to.1oads required for the safe shutdown of the reac+,>y-r(rha ~mar-s--.=

---a - - ----..--.

-..= -...=LY=SILY WUCJ acquencer must De manuallv seset at ths DTIA ne ---..----

PSEG internal Use Onlv HC.OP-AB.ZZ-O135(Q) 4.24 WHEN HPCI/RCIC Room temperatures return to normal THEN PERFORM the following:

[CD-675F]

4.24.1. ENSURE the following the HPCI Room Fire Dampers are open: 0 IFP-GUD292 IFP-GUD293 4.24.2. PLACE HPCI AND RClC high temperature isolation trip switches to NORMAL at the following panels: (Key 172, PA2235 in Work Control Key Cabinet for all switches)

A. HPCl at Panel P-620 Switch B21B-S6A Panel P-641 Switch B21 B-S6C B. RClC at Panel P-621 Switch B21B-S5B Panel P-640 Switch B21 B-S5D 4.25 When directed by the CRS, PERFORM the following:

4.25.1. ENSURE all Load Sequencer breakers A (B, C, D) C428 are closed. NOTE Emergency load sequencer keys in Work Control Key Cabinet: Key 86, GEK594 or EBOOOI (Sequencer A) Key 87, CAT60 or EB0001 (Sequencer B) Key 88, GEK592 or EBOOOI (Sequencer C:i Key 89, GEK591 or EBOOOI (Sequencer D) 4.25.2. 4.25.3. RESET LOP Emergency Load sequencers using WCC keys. RESET LOCA sequencers and VERIFY "Logic A (B, C, D) -7 Initiation" lamps are extinguished. - Hope Creek Page 19 of 46 Rev 29 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 WA # 212000 K5.02 lmportarice Rating 3.3 Knowledge of the operational implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM : Specific logic arrangements. Proposed Question: Common 10 The reactor is operating at 100% power.

The operator observes the Logic A Normal light under the Group 1 Solenoids is NOT illuminated. The cause is NOT a bad light bulb.

Given this condition, if a half-scram condition occurs on the: (Chose answer based on logic function ONLY) A. 'BI' logic; 112 of the control rods will scram. B. 'A2 logic; 112 of the control rods will scram. C. D. 'AI' logic; 114 of the control rods will scram. 'B2' logic; 1/4 of the control rods will scram Page 19 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed Answer:

D a. D. A. B. C. Explanation (Optional): IAW RPS Lesson Plan NOH01 RPSOOC-05 Section ll.C.4.c. - Energized trip systems maintain the scram pilot valve solenoids energized

1) a) b) 2) 3) Two contacts in series are maintained closed to energize the A (B) solenoid One contact will open with a scram signal in trip channel AI (BI) The other contact will open with a scram signal in trip channel A2 (82) Either AI or A2 trip channel will de-energize the A solenoids for a half-scram.

Either B1 or 82 trip channel will de-energize the B solenoids for a half-scram.

Correct. ANY 'B' side RPS trip will de-energize the '6' scram pilot solenoid valves for the GPI rods, resulting in their scramming in. Each group comprises approximately 1/4 of the rods. Incorrect.

The 'A' scram pilot solenoid valves lor the GPI rods are already de-energized.

Incorrect.

The 'A' scram pilot solenoid valves for the GPI rods are already de-energized Incorrect.

ANY 'B' side RPS trip will de-energize the 'B' scram pilot solenoid valves for the GPI rods, resulting in their scramming in. Each group comprises approximately 1/4 of the rods. Technical Reference(s) Lesson Plan NOH01 RPSOOC- 05 Prints PNI-C71-1020-006 Sheets 7, 13 ,I4 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective:

RPSOOOEOI 7 (As available) Question Source:

Bank # 62627 -. Modified Bank

1. -. (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 20 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Quest ion Worksheet Page 21 of 205

' "LESSON NAME: NOH01 RPSOOC-05 REACTOR PROTECTION SYSTEM - 01/10/2008 I) Each trip channel has dedicated sensors, trip units, relays, and manual switches.

2) Trip channels measuring parameters at scram setpoints or beyond de-energize its trip system. 3) Trip channels and trip systems are powered from associated RPS Bus I TABLE 7 - J 4) See Table 1 for scram parameters 1 FIG 5,s J c. Energized trip systems maintain the scram pilot valve solenoids energized

1) Two contacts in series are maintained closed to energize the A (B) solenoid a) One contact will open with a scram signal in trip channel AI (BI) b) The other contact will open with a scram signal in trip channel A2 (B2) 2) Either AI or A2 trip channel will de-energize the A solenoids for a half-scram.

3) Either B1 or E32 trip channel will de-energize the B solenoids for a half-scram.

I OBJ I4 I 4) (Either AI or ,42) and (either B1 or B2) will produce a full scram; considered a 1 -out-of-2-taken-twice logic.

5) The de-deenergization of both solenoids allows the air to be vented off the operators of the scram inlet and outlet valves failing them open. This in turn admits pressurized water from the HCU to the under piston side of the contrcl ride drive mechanism resulting in the rapid insertion (scram) of the control rod.

Page 16 of 48 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008-2009\References\LESSON ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 WA # 263000 K6.01 I m porta rt ce Rating 3.2 Knowledge of the effect that a loss or malfunction of the following will have on the D.C. ELECTRICAL DISTRIBUTION:

A.C. electrical distribution. Proposed Question:

Common 11 The plant is operating at full power when a LOCA arid a loss of offsite power (LOP) occur.

Emergency diesel generators respond as designed, WHICH one of the following describes the status of the 1 E and Non-I E 125 VDC Battery Chargers 30 seconds after the event? A. 1 E battery chargers are in service and the Non-1 E battery chargers are load shed and CANNOT be returned to service. 6. 1 E and Non-I E battery chargers are load shed and both are automatically restored at the same time by load sequencing.

C. 1 E battery chargers are in service and the Non-I E battery chargers are load shed and can be manually restored by overriding the load shed and re-energizing the MCC's.

D. 1 E and Non-I E battery chargers are load shed; the 1 E battery chargers are automatically restored by load sequencing and the Non-I E battery chargers will be restored 2 minutes after the sequencer starts. Proposed Answer: C Explanation (Optional):

IAW DC Electrical Lesson Plan section X.C.l .b. - Upon a LOCA, the MCCs that supply the battery chargers (excluding the guardhouse battery charger 10D514) are shed from the Class 1 E 480 VAC Unit Substations that normally supply their power. Shedding of the MCCs places the 125 VDC (non- 1 E) power requirements on the respective batteries. The LOCA signal for the MCC feeder breakers can be overridden in the control room at 10C650 C. Correct - The 1 E battery chargers supply breakers are not load shed. The Non-I E battery chargers can be restored manually.

Page 22 of 2105 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet A. B. D. Incorrect - The Non 1 E chargers are not automatically restored after a LOCA. Incorrect - The 1 E chargers are not load shed. Non-I E battery chargers are not automatically restored.

Incorrect - The 1 E battery chargers are not load shed. Non-I E battery chargers are not automatically restored Technical Reference(s) OP-SO-SM-0001, Table SM- (Attach if not previously provided) 020 Proposed references to be provided to applicants during examination:

None Learning Objective: Question Source:

Question History: Question Cognitive Level:

DCELECEOI 5 (As available)

Bank # 54243 Modified Bank # (Note changes or attach parent)

New -. Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 23 of 205

.I LESSON NAME: DC ELECTRICAL DISTRIBUTION NOH01 DCELEC 07/31/06 2) The MCC for the guardhouse 125 VDC switchgear panel has a standby diesel generator (00G503) as its backup power source. Loss of Coolant Accident (LOCA) Upon a LOCA, the MCCs that supply the battery chargers (excluding the guardhouse battery charger 1OD514) are shed from the Class 1 E 4810 VAC Unit Substations that normally supply their power. I) 2) Shedding of the MCCs places the 125 VDC (non- 1E) power requirements on the respective batteries. The LOCA signal for the MCC feeder breakers can be overridden in the control room at 1OC650 (this will be covered in detail during the 1 E AC Distribution System lecture).

[ Obj. 17&27 1 c. Using the below listed procedures discuss the following:

1) HC.OP-AB.ZZ-0147, DC System Grounds a) Symptoms b) Subsequent operator actions: c) Discussion section 2) HC.OP-AB.ZZ-0149, 250 VDC Malfunction a) Symptoms b) Subsequent operator actions: c) Discussion section 3) HC.OP-AB.ZZ-0150, 125 VDC Malfunction a) Symptoms b) Subsequent operator actions: c) Discussion section 4) HC.OP-AB.ZZ-0151, +24 VDC Malfunction a) Symptorns b) Automatic Actions c) Subsequent operator actions: d) Discussion section Page 47 of '77 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008-20r)9\References\LESSON PLANS\DC Electrical Distribution\Master\Lesson Plans\NOHOl DCELEC-01 DC Electrical Distribution.doc I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 WA # 21 5003 K6.05 Importance Rating

3.1 Group

1. 1 Knowledge of the effect that a loss or malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM: Trip units. Proposed Question: Common 12 Given the following conditions:

A plant startup is in progress.

IRM "G" is inoperable and bypassed All other IRMs are indicating on range 8 Then, a power failure occurs on the trip unit for IRM "E" Which one of the following describes the effects of this power failure on the IRM trip units?

A. With the mode switch in RUN, a half scram and rod block would occur. B. With the mode switch in RUN, ONLY a rod block would occur. C. With the mode switch in STARTUP, a half scram and rod block would occur. D. With the mode switch in STARTUP, ONLY a rod block would occur. Proposed Answer: C Explanation (Optional):

IAW LP NOH01 IRMSYS-02 page 17 & 18 C. Correct. A loss of power A. B. D. Incorrect. All scrams and rod blocks are bypassed with mode switch in RUN Incorrect. All scrams and rod blocks are bypassed with mode switch in RUN to the trip units would cause a rod block and half scram with the mode switch in Startup Incorrect.

A half scram would also occur Technical Reference(s) NOH01 IRMSYS-02 (Attach if not previously provided)

Page 24 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

___. Proposed references to be provided to applicants during examination: none Learning Objective:

I RMSYSEO 12 (As available) Question Source:

Question History: Question Cognitive Level: 10 CFR Part 55 Content: Bank # Modified Bank # -. (Note changes or attach parent)

New X Last NRC Exam _. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Comments: Page 25 of ;!05 I .P??mEr - '~X?FS,LrPQXYZT-Z?2~~<vI?BX3~~~

%'-P? -&-*E &cz, -27.- LESSOM f4AKE: rdoi-1m IRMSYS-02 SYSTEFA " 05/05/05 P uaMFE2. SxITv d %sYa. *--=5QEmE--

-- " - a) Certain parameters are nmniiored which provide an indication of IFIM channel inopei-ability.

These parameters are: = Detector nign voltage power supply (LOW) Drawer module unplugged Channei mode switch (not IR OPERATE) .- \ 1 OBJ 5 -1 b) This rod block is bypassed if either: 0 The associated IRM is bypassed __._ OR a The reactor mode switch is in RUN 4) IRM Downscale The downscale rod block setpoint of 5/125 of scale indicates a possible malfunction within the IRM channel. A control rod withdrawal block is initiated to inhibit reasior power escalation under conditions oi' IRM inoperability.

This rod block is bypassed if either: e The associated IRM is bypassed e The reactor mode switch is in RUN e Associated IRM range switch selected to Range 1 (enables reactor startup) I FIG 8 I 4. Reactor Scram (Inputs into the reactor protection system - RPS) a. The IRMS provide inputs into the RPS under conditions which indicate excessive reactor power for existing conditions and/or IRM channel inoperability.

b. ) Loss of power will result in the scram signal being generated by the IRMS (fail-safe).

The assignment of IRMS to the RPS logic is as follows: IRMS RPS TRIP SYSTEM A&E A (Logic AI) " 4 Page 18 of 35 S:\VDrive\TRAINING DOCUMENTS\Operations Training\liope Creek\Plant Technology\Systerns\lntermediate Range Monitoring\Masier\Lesson Plans\NOHOl IRMSYS-02 intermediate Range Monitoring System.doc 1 I C. Detector select pushbuttons (A, !3 ,C, 3,E,F,G, H) This pushbutton control is used to enabieicikable the detector positioning circuitry for the selected detector. (i.e. it selectddeselect:

the detector for movement.)

3. Rod Blocks (Inputs into the reactor manual control system).

a. The IRMS provide inputs into the reactor manual control system (RMCS) to inhibit control rod withdrawal under conditions which coulci result in fuel cladding darnage (overpower condition) or which indicai: IRM channel inoperability.

OBJ 5 FIG 7 b. Loss of power will result in a rod block being generated by the IRMS (fail safe). There are two trip systems within the RMCS with four IRM assigned to each trip system. A single lRNl input into either RMCS tri j system will initiate a rod block. -7 I) Detector Not Full In Rod Block. Accurate core power monitoring cannot be achieved during startup if the IRM detectors are not fully inserted. Therefore, it would be imprudent to allow rod withdrawal under conditions when the IRM is required to be functional. This rod block is bypassed if either: a The associated IRM is bypassed 6 The reactor mode switch is in RUN 2) IRM Upscale Rod Block The upscale rod block setpoint of 108/? 25 of scale indicates reactor power is high for existing conditions.

The control rod withdrawal block prevents further power escalation with control rods. This rod block is bypassed if either:

10 The associated IFtM is bypassed 8 The reactor mode switch is in RUN Page 17 of 35 S:\VDrive\TRAII\IING DOCUMENTS\Operations Tratning\Hope Creek\Plant Technology\Systerns\Interniediate Range Moni toring\Master\Lesson Plans\NOHOl IRMSYS-02 Intermediate Range Monitoring System.doc ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 WA # 203000 AI .01 Importance Rating

4.2 Group

1. 1 Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) controls including: Reactor water level Proposed Question: Common 13 Given the following conditions:

0 0 0 0 The reactor is shutdown and is being cooled down RPV level is being held relatively constarit at (-30) inches using Startup Level Control Reactor pressure is about 390 psig All rods are fully inserted Then, 'C' RHR Loop initiation logic is inadvertently initiated in the LPCl mode, RHR LPCl Injection Valve BC-HV-FO17C opens. Which of the following describes the operational effect of this condition?

A. RPV level will rise and torus level will lower.

B. RPV level will lower and torus level will rise.

C. RPV level will rise and torus level will rise.

D. RPV level and torus level will remain relatively constant.

Proposed Answer: D Explanation (Optional):

IAW RHR LP NOH01 RHRSYSC-06 D. Correct - The shut-off head of the RHR pumps is about 366 psig, the min flow valve will remain open. RPV level will remain relatively constant.

A. B. C. Incorrect - RHR will not be injecting so levels will remain relatively constant Incorrect-Torus level will not be significantly affected when the RHR pump is at shutoff head Incorrect-The shut-off head of the RHR pumps is about 366 psig. Levels should remain relatively unchanged because the RHR pump will not inject at shutoff head. Page 26 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s) RHR LP NOH01 RHRS'ISC-06 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

RHRSYSEOI 1 (As available) Question Source:

Bank # 5641 2 Modified Bank

1. -. (Note changes or attach parent)

New Question History: Question Cognitive Level: 10 CFR Part 55 Content: Last NRC Exam Memory or Fundamental Knowledge Comprehension or Anaiysis X 55.41 X Comments:

Page 27 of 205

-1 LESSON NAME: NOHOIRHRSYSC-06 RESIDUAL HEAT REMOVAL SYSTEM - 02/27/08 - I 6. Fuel Pool Cooling Assist RHR Loops A and/or B can be manually aligned to provide cooling for the spent fuel pool if additional cooling capacity is required.

IObj. E003h -7 7. Alternate Injection IF other systemslsources are a available for RPV level restoration or flooding, SSW can be supplied to the RPV as a last resort.

Key lock switches on 10C650A and 10C651A allow positioning of isolation valves tc connect SSW and RHR Loop B. B . Major Subs ystems/Com pone nts IObj. E007a _I 1. RHR Pumps (AP-DP202)

a. Purpose The RHR pumps develop the required discharge head and flow rate to support all modes of RHR System operation.

b. Characteristics I) Each pump is located in individual pump rooms on the 54 ft elev, west side of the Rx Bldg. The pumps are four stage, centrifugal, deep well pumps and each can provide a flow rate of 10,000 gpm at a discharge head of approx. 171 psig (Note Technical Specifications requires LPCl mode flow of at least 10,000 gpm against a test line pressure corresponding to a reactor vessel to containment differential pressure of >20 psic .) ----> Pump shutoff head is approx. 366 psiq (875 ft'). -r I Seal Plow approx. 3-7 gprn through separator 7 2) A small portion of pump discharge flow is routed through a cyclone separator and heliflow HX to provide flushing and cooling of the pump mechanical seal. The cyclone separator ensures particle free water (centrifugal force separates suspended solids) for flushing the seal areas.

SACS flow through the heliflow HX provides cooling for the seal water.

The discharge of both the cyclone separator and the pump mechanical seal is routed back to the RHR pump suction. Page 15 of 85 C \Documents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\References\LESSON PLANS\Residual Heat Removal\Master\Lessc n -. .-.^..^.^..-^.I^^

^^ - . . .. - '.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 K/A # Importance Rating

3.0 Group

1. 1 261 000 AI .04 Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including:

Secondary contain ment differential pressure Proposed Question: Common 14 The plant is at full power. All systems are operable.

Reactor Building Ventilation (RBVS) is operating in a normal alignment Which one the following describes the actions required when swapping over to FRVS IAW HC.OP-SO.GU-0001 "FRVS Operation" and their affect on Secondary Containment DIP throughout the evolution? (Actions are listed in the order performed)

A. B. C. D. Remove the RBVS exhaust fan from service Remove the RBVS supply fan from service Start an FRVS vent fan Start the FRVS recirc fans DIP will remain negative Remove the RBVS exhaust fan from service Remove the RBVS supply fan from service Start an FRVS vent fan Start the FRVS recirc fans Initially D/P will go positive but then return to negative Start an FRVS vent fan Remove the RBVS supply fan from service Remove the RBVS exhaust fan from service Start the FRVS recirc fans D/P will remain negative Start an FRVS vent fan Remove the RBVS supply fan from service Remove the RBVS exhaust fan from service Start the FRVS recirc fans Initially DIP will go positive but then return to negative Proposed Answer: C Explanation (Optional):

IAW HC.OP-SO.GU-0001, Section 5.3.4 - The FRVS vent fan is first Page 28 of 205 I I ES-40 1 Sam p I e Written Exam i nation Question Worksheet Form ES-40 1-5 action taken in the sequence. This will ensure a negative pressure in the RB when the RBVS exhaust fans are removed from service in the next sequenced step C. Correct. A. Incorrect. The FRVS vent fan must be started first B. Incorrect.

The FRVS vent fan must be started first D. Incorrect - D/P will remain negative Technical Reference(s) HC.OP-SO.GU-0001 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

RBVENTEOOS (As available)

-. Question Source: Bank

1. Modified Bank # New (Note changes or attach parent)

X -. Question History: Last NRC Exam Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 29 of 205 o/ y PSEG Internal Use Onlv HC.OP-SO.GU-OOOl(Q) - NOTE 5.3.3 Normal operating FRVS Vent System lineup is with one FRVS Vent Fan operating and the other FRVS Vent Fan in AUTO. 5.3.3 START 1 FRVS Vent Fan as follows: A. Momentarily PRESS the MAN push-button for A(B)V206 FRVS Vent Fan. OBSERVE alarm BOP SAFETY SYS OUT OF SVCE. B. C. Momentarily PRESS the A(B)V206 START push-button for A(B)V206 FRVS Vent Fan. D. OBSERVE an indicated flow of approximately 9000 cfin on FR-9426AQ3) for the running FRVS Vent Fan.

E. PRESS the AUTO LEAD PB for the inservice FRVS Vent Fan AND INITIAL Attachment

1. Hope Creek F, PRESS the AUTO PB for the out of service FRVS Vent Fan AND IMTIAL Attachment

1. Page 12 of 27 Rev. 23 13.1 bL P EG Internal Use Only - NOTE 5.3.4 Step 5.3.4 is performed from Aux Bldg HVAC Panel 'l OC382 unless otherwise specified.

Performance of Step 5.3.4 will result in the Reactor Building Ventilation System being unavailable in the event of a LOP. In the event of a LOP, manual restoration of Reactor Building Ventilation will be required OR an FRVS Vent Fan placed in service to maintain a negative pressure in the Reactor Building.

5.3,4 PERFORM the following to remove the Reactor Building Ventilation System from service: A. PERFORM the following at. Aux Building HVAC Panel 10C382: 1, PLACE Control Switch for the idle Reactor Bldg Supply Fan C(A,B)VHSOO in STOP. STOP B(C,A)VH-300, Reactor Bldg Supply Fan. STOP A(B,C)VH-300, Reactor Bldg Supply Fan. PLACE Control Switch for the idle Reactor Bldg Exhaust Fan C(A,B)V301 in STOP. STOP B(C,A)V301, Reactor Bldg Exhaust Fan, _- 6. STOP A(J3,C)V301, Reactor Bldg Exhaust Fan. 2. 3. 4, 5. >? 2-y 4 ? .-- B. CLOSE the following dampers at Panel 10C651E: e EID9370A, Reactor Bld,g Outbd Sply 0 HD9414A, Reactor Bldg Inbd Exch HD9370B, Reactor Bldg Inbd Sply HD9414B, Reactor Bldg Outbd Exch Hope Creek Page 13 of 27 Rev. 23 19 L[ PSEG Internal Use Only HC,QP-SO.GU-OOOl(Q) 5.3.5 5.3.6 START 4-FRVS Recirculation Fans as follows: A. -- MOMENTARILY PRESS 14(B,C,D,E,F)V213 START PB for the desired fans. OBSERVE an indicated flow of approximately 30,000 cfm on FR-9377 for each of the runriing fans. B. C. MONITOR the following:

PDR-9377 FRVS DIFF PRESS for indication of 5 6.0" W.G. for each operating Air Filter Unit PDR-9425A(B) FRVS-VENT System Filter D/P 5 3.0" W.G. PERFORM independent verification that the system is aligned IAW Attachment

1. 5.4 Removing FRVS from Service Observe the Filtration, Recirculation and Ventilation System operabillty requirements of Technical Specification 3.6.5.3. 5.4.1 ENSURE all Prerequisites have bet- satisfied IAW Section 2.4. 5.4.2 REMOVE running FRVS Recirc Fans from service as follows: A. restoring from an AUTO START: 1. MOMENTARILY PRESS the STOP PB for: Continued next page Hope Creek

• e EV2 13 FRVS W-CIRC FAN FV2 13 FRVS RECIRC FAN 2. PRESS the AUTO/LOCKOUT PB to select Lockout for: EV213 FRVS RECIRC FAN 0 FV2 13 FRVS RECIRC FAN Page 14 of 27 - Rev. 23 I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Proposed Answer:

B Explanation (Optional):

IAW ARP C1-Fl & HC.OP-SO.SA-0001 B. Correct. IAW with ARP, If both pumps are not running with a valid signal present, turn the key lock on and start the pumps A. Incorrect.

timed out C. Incorrect.

timed out D. Incorrect. Both pumps should be running with a manual initiation and the 230 second timer Both pumps should be running with a manual initiation and the 230 second timer Key lock must be turned on prior to starting pump Technical Reference(s)

ARP C1 -Fl & HC.OP-SO.SA-000 1 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

RRCSOOE005 (As available)

-. Question Source: Bank # Modified Bank # -. (Note changes or attach parent)

New X -. Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 31 of 205 3.3.4 3.3.5 3.3.6 3.3.7 3.3.8 HC-OP-SO.SA-0001 (Q) RRCS - Standby Liquid Control System injection will initiate after receipt of the following signals: A. High Reactor Steam Dome Pressure (1071 PSIG) sealed in I or - B. Low Low Reactor Water Level (Level 2, -38") not sealed in I or - C. Manual Initiation RRCS - and D. 230 second time delay - and E. Reactor Power goJ downscale

(> 4% or APRM INOP) The following Reactor Water Clean-up System Valves will is SBLC System initiation:

late rpon a BG-HV-FOOI 5G-HV-F004 RRCS - ARI can be reset 30 seconds after initiation, if the initiating condition is clear. RRCS - Recirc Trip can be reset 13 minutes 50 seconds after initiation.

RRCS - Standby Liquid Control System actuation and RWCU System isolation can be reset 10 minutes after initiation.

4.0 EQUIPMENT

REQUIRED None Mope Creek Page 4 of 7 Rev. 4 -

HC.OP-AR.ZZ-0008( Q) ATTACHMENT F1 Nomenclatu relCond it ion SLCIRRCS A INITIATION SLC/RRCS INITIATION FAILURE Automatic Action Alarm only Window Location

@I-F1 FAILURE OPERATOR ACTION:

1. DETERMINE E a valid RRCS initiation signal is present, - IF signal is valid, VERIFY that both SLC Pumps are running.

2. E a valid RRCS initiation signal is present AND a SLC Pump is not running:

a. TURN the non-running SLC Pump KEY-LOCK Switch to ON. b. PRESS the START pushbutton for the failed pump. REQUEST the CRS to initiate corrective action

3. INPUTS Indication FAILURE

REFERENCES:

J-48-0, Sht. 5 E-6768-0, Sht. 2 Hope Creek Page 35 of I71 Rev. 35 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 KIA # 223002 K1.03 Importance Rating

3.0 Knowledge

of the physical connections and/or cause effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEMlNUCLEAR STEAM SUPPLY SHUT-OFF and the following: Plant Ventilation Proposed Question:

Common 16 The plant is in OPCON 4 with the following conditions:

0 0 0 0 Primary Containment has been de-inerted CACS is aligned to purge the drywell and suppression chamber The IB' and IC' Reactor Building ventilation Supply and Exhaust Fans are running The 'A Reactor Building Ventilation S~pply and Exhaust Fans are in AUTO Then, an operator arms and depresses the 'D' Channel PClS Manual Initiation pushbutton on 1 OC651 C for a surveillance test. Two minutes later plant condition(s) stabilize. Which one of the following describes the final status of the containment purge lineup and the Reactor Building Ventilation System (RBVS)?

A. The containment purge lineup will isolate. The Reactor Building Ventilation Fans will be unaffected.

B. The containment purge lineup will isolate.

NO Reactor Building Ventilation fans will be running. C. The containment purge lineup will NOT be affected.

The Reactor Building Ventilation Fans will be unaffected.

D. The containment purge lineup will NOT be affected.

NO Reactor Building Ventilation fans will be running. Page 32 of 2!05 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed Answer: B Explanation (Optional):

B. A C. D. Correct. Manual initiation of the 'Dl Channel PClS closes the GS-HV-4950, 4962, 4979, and 4980. These valves isolate the purge supply and exhaust lines. While the 'Dl channel does not directly trip the running RBVS supply and exhaust fans, it will close the GU-HD-9414B and 9370B. These valves isolate the Reactor Building Ventilation System supply and exhaust lines, which will result in all running fans tripping on low flow after a 90 second time delay. The 'A RBVS supply and exhaust fans are directly tripped (load shed) by the 'D' Channel PClS signal. Incorrect - NO RBVS fans will be running Incorrect - containment purge supply and exhaust will be isolated.

NO RBVS fans will be running. Incorrect - containment purge supply and exhaust will be isolated Technical Reference(s)

HC-0P.SO.SM-0001 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

I NERTOEO I 2 (As available)

Question Source: Bank # 62574 Question History: Question Cognitive Level:

Modified Bank # -. (Note changes or attach parent)

__. New Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 33 of 205 OBSERVE the Group 12 Valves listed in Table SM-012 have closed under the Manual or Automatic isolation Signals specified.

I - - 7 NOMENCLATURE CONT ATMOSPHERE CPCS RTN TO CNTMT I------ VALVE NO. MANUAL AUTO ISOLATIONS ISOLATION A B C D E F G H I *ACPJTMT X X X i "GS-HV4956

-71 "GS-HV4979 "GS-HV4951 17 "GS-HV4950 "GS-HV4952

-7 "GS-HV4980 "GS -HV4958 "GS-HV4963 "GS-HV4962 3 "GS-HV4964 1 I " G S - HV4 9 74 TABLE SM-012 I SUPP CHMBR TO CPCS *D CN'TMT X X X X X X d/U SPLY ISLN I SO LATlO N SETPOINT A - REACTOR VESSEL WATER LEVEL 2 -38" B - DRYWELL PRESSURE - HIGH C - REACTOR BUILDING EXHAUST RADlATlOlV - HIGH * - Can receive a Half Isolation from the corresponding NSSSS Manual Isolation " - Isolation can be bypassed by ISLN OVRD PB 1.68 psig 1 X uCi/cc Hope Creek Page 19 of 42 Rev. 16 H C .O P-SO . S M-000 I (Q) OBSERVE the Group 19 Dampers listed in Table SM-019 have closed under the Manual or Automatic Isolation Signals and other Actions have occurred for Equipment listed as specified.

TABLE SM-019 (I Of 3) ISLN INBD EXH 4 -7 REACTOR BLDG SPLYlEXH IS0 LATl ON SETPOI MT A - REACTOR VESSEL WATER LEVEL 2 B - DRYWELL PRESSURE - HIGH C - REFUEL FLOOR EXHAUST RADIATION - HIGH D - REACTOR BUILDING EXHAUST RADIATION - HIGH -38' 1.68 psig 2 x 1 0" uCi/cc 1 X I o-~ uCi/cc # - Group 19 Dampers * - Can receive a Half Isolation Signal from the corresponding NSSSS Manual Isolation Hope Creek Page 27 of 42 Rev. 16 I I ES-40 1 Sam p I e Written Exam i nation Question Worksheet Form ES-40 1-5 Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 262002 A3.01 -~ Importance Rating

2.8 Ability

to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including: Transfer from preferred to alternate source. Proposed Question: Common 17 The Manual Bypass Control Switch on a NON IE inverter has been placed in the "Bypass-to Alternate" position for testing on a faulty Static Switch.

EXHIBIT 2 TPS POWER CONTTWL (TXTICAL)

Psge 1 of 1 Which of the following describes the design response if a LOP occurs? The input to the Static Inverter section will be from.. . Page 34 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet A. B. C. D. 125 VDC but supply to the system loads will be lost. 125 VDC and maintaining system loads. 480 VAC and maintaining system loads. 480 VAC but supply to the system loads will be lost. Proposed Answer: A Explanation (Optional): See attached diagram from HCOP-SO.PN-0001 (Q) Exhibit 3.

A. B. C. D. Correct. The input of the static inverter will be supplied by 125 VDC power but supply to the system loads will be lost - Placing the switch in Bypass to Alternate closes contacts 1,2 and 5 meaning supply to system loads is off the Alternate/Backup supply which is not 1 E supplied so loads are lost. Incorrect.

The input of the static inverter will be supplied by 125 VDC power and maintaining system loads - contacts 1,2 and 5 meaning supply to system loads is off the Alternate/Backup supply Incorrect.

The input of the static inverter will be supplied by 480 VAC power and maintaining system loads - Alternate/Backup supply which is not 1 E supplied so loads are lost. Incorrect.

The input of the static inverter will be supplied by 480 VAC power but supply to the system loads will be lost - Placing the switch in Bypass to Alternate closes contactsd 1,2 and 5 meaning supply to system loads is off the Alternate/Backup supply. Technical Reference(s)

HC. OP-SO. PN-000 1 (Q) (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective:

NONI EOE003 (As available) Question Source:

Bank # 56822 Modified Bank # New (Note changes or attach parent)

Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X Page 35 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet 10 CFR Part 55 Content: 55.41 X Comments:

Page 36 of 205

I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Exam inat ion Outline Cross-reference: Level RO SRO Tier # 2 Group ki 1 KIA # 209001 A3.03 Importance Rating

3.5 Ability

to monitor automatic operations of the LOW PRESSURE CORE: SPRAY SYSTEM including: System pressure. Proposed Question: Common 18 An event has occurred at the plant.

All Core Spray Loops are injecting to the RPV when the Core Spray Min Flow Valve BE-HV-F031A inadvertently strokes open due to a malfunction of FISH-N651A.

How will this affect Core Spray discharge pressure and total indicated flow in the "A Loop in the control room? A. Discharge pressure will increase. Indicated flow will decrease.

B. Discharge pressure will increase. Indicated flow will increase.

C. Discharge pressure will decrease. Indicated flow will decrease.

D. Discharge pressure will decrease. Indicated flow will increase. Proposed Answer: C Explanation (Optional):

C. Correct. Due to where the Flow indicator and rnin flow taps off in the system A. B. D. Incorrect. Discharge pressure will decrease Incorrect. Discharge pressure will decrease, indicated flow will decrease Incorrect. indicated flow will decrease Technical Reference(s)

M-52 (Attach if not previously provided)

Page 37 of 205 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

M-52 Learning Objective:

CSSYSE004 (As available) Question Source:

Bank # Modified Bank # _. (Note changes or attach parent)

New X Question History: Question Cognitive Level:

10 CFR Part 55 Content: Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Comments:

Page 38 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO -___ Tier # 2 K/A # 21 5005 A4.03 Importance Rating

3.2 Group

1. 1 Ability to manually operate and/or monitor in the control room:

APRM back panel switches, meters and indicating lights. Proposed Question: Common 19 The plant is operating at rated power. APRM

'E' has 5 LPRMs bypassed, APRM

'F' has 5 LPRMs bypassed. All other LPRM inputs are operable. Which one of the following describes how to use the APRM back panel drawer controls to determine the number of non-bypassed LPRMs and what would their respective meters indicate?

A. B. C. D. Place the respective APRM Meter Function Switch to COUNT APRM 'E' meter will indicate 80 on the 0-125% scale. APRM 'F' meter will indicate 85 on the 0-125% scale. Place the respective APRM Meter Function Switch to COUNT APRM 'E' meter will indicate 85 on the 0-125'%0 scale. APRM 'F' meter will indicate 85 on the 0-125% scale. Place the respective APRM Mode Switch to STANDBY then Place the Meter Function Switch to COUNT. APRM 'E' meter will indicate 85 on the 0-125% scale.

APRM 'F' meter will indicate 85 on the 0-125% scale. Place the respective APRM Mode Switch to STANDBY then Place the Meter Function Switch to COUNT. APRM 'E' meter will indicate 85 on the 0-125% scale. APRM 'F' meter will indicate 80 on the 0-1 2596 scale. Page 39 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Proposed Answer:

A Explanation (Optional): IAW APRM LP NOH04APRIUOOC-04 To determine the

1. of nonbypassed LPRMs on the meter function switch is moved to count APRM 'E' has 21 LPRM inputs, if 5 are bypassed then 16 remain. Each nonbypassed LPRM is = to 5% on the meter when it is taken to count. Therefore the meter would indicate 80 for APRM 'E'. APRM 'F' has 22 LPRM inputs, if 5 are bypassed then 17 remain. Each nonbypassed LPRM is = to 5% on the meter when it is taken to count. Therefore the meter would indicate 85 for APRM IF'. A. Correct. B. C. D. Incorrect.

E would indicate 80, F-85. Incorrect. Only the meter function switch must be moved Incorrect. Only the meter function switch must be moved.

E would indicate 80, F-85. Technical Reference(s)

NOH04APRMOOC-04 (Attach if not previously provided) Proposed references to be provided to applicants during examination: none Learning Objective:

LPRMOOE005 (As available) Question Source: Question History: Question Cognitive Level:

10 CFR Part 55 Content: Bank # Modified Bank

1. New X (Note changes or attach parent)

Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Comments:

Page 40 of 205 I I - LESSON NAME: NOH04APRMOOC-04 AVERAGE POWER RANGE MONITORING (APRM) SYSTEM-1/19/2008 ~~~~ ~~ ~~~~ _____ 2) NOTE: That flow unit is removed from the comparator circuit. Bypassing the flow unit does not remove its output fror 1 the APRM or RBM channels that receive it. For examrde: If a flow unit fails low while at power, the flow biased reference signals developed will be very low and will re sult in trips of those APRMs & RBM channels. Bypassing "he flow unit will only allow clearing the rod blocks generatr$d by the flow unit comparator trips. EBJ. 2a; b; c; d A 2. 1 OC608 Controls a. Meter Function Switch I FIG. 12(A V1997m), FIG. 13(A V1997k)

I 1) The meter function switch is used to control the parameters indicated on the associated meter. The following positions ar.3 available:

AVERAGE The AVERAGE position enables the % power reading from the associated APRM channel (0 125% scale), to be displayed.

FLOW The FLOW position enables the % flow signal from the associated APRM flow unit to be displayed (0-125% scale). ~ COUNT The COUNT position enables the meter indication for Vie number of LPRM inputs into the associated APRM (or LPRM group) channel. The 0-125% scale is used and the i reading is divided by 5%. The resultant number indicales I 1 / LPRMs that have their mode switch in ~e OPERATE the number of "non-bypassed" LPRM inputs (i-e:-all.-4 r_-. - -- --- . __ - - posit ion) Ex: \\ A reading of 105% = 109- 5 = 21 LPRM inputs. d) A, B, c, D The alphabetical positions are used in conjunction with the LPRM selector switch to monitor a specific LPRM outpi t. The A, B, C, D position indicates the associated detecttzr level. The x-y detector coordinate is determined from ttie operator aid located at the top of the local panel.

Page 24 of 40 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008-2009\References\LESSON PLANSMverage Power Range Monitoring ,nnr,~mtn~--~-ii ni--\rinuninnnmrnnn n~ n ..-_-I- n _..,__ n __-- U~-..:L-:..-

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, I *- LESSON NAME: NOH04APRMOOC-04 AVERAGE POWER RANGE MONITORING (APRM)

SYSTEM- 1/19/2008 Neutron monitoring system power range panel 1OC608 is a contra1 room b xk panel consisting of five (5) bays. Together the five bays house the electro iic components associated with the six (6) APRM and two (2) RBM System channels, the two (2) LPRM groups, the four (4) Recirc flow units and the 13ur (4) power range monitoring system power supply modules. 1 FIG. 3(AV1563)

-3 2. APRM Averaging Amplifier

a. Purpose The amplifier produces an output signal that is equivalent to the average of all the LPRM inputs it receives. This output signal represents avel-age core power. I FIG. 4(A V2666c)

b. The LPRM detector outputs are appropriately assigned such that each APRM channel produces an output that is representative of average core power. In order to accomplish this, LPRM detectors are chose 1 from appropriate radial and axial core locations such that APRM channels A, C & E receive 21 inputs and APRM channels B, D & F receive 22 inputs (typically only one detector per string and 4-6 detectors per level will provide an input to each channel).

I FIG. 3fAV1563)

c. The averaging amplifier will produce an output signal of 0 to +I OVD( ;. This DC signal corresponds to a core thermal power of 0 to 125% ot rated. Each LPRM will provide an input to the assigned averaging a np if the individual LPRM amplifier card mode switch is in the operate position. With the switch in any other position, that detector will be removed as an input and the averaging amp will produce an output signal based on the remaining LPRM inputs. The averaging amp oufput is representative of the average of all LPRM inputs regardless of hoLJ many LPRMs are bypassed.

1 FIG. 5(AV0247d) - 1 3. Recirc Flow Units a. Purpose Produce a signal representing total recirc system flow for use in determining the flow biased rod block and upscale thermal power saam trip functions.

Page 10 of 40 C:\Docurnents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\References\LESSON PLANSWverage Power Range Monitoring ,nnnBmtma--&--\o

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I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 2 KIA # 400000 A4.01 Importance Rating 3.1 Group # 1 Ability to manually operate and/or monitor in the control room: CCW indications and control. Proposed Question: Common 20 Plant conditions are as follows: 0 Poweris87%

0 "A SACS Loop is supplying TACS.

0 "D" SACS pump is operating in the "B" SACS Loop.

0 "B SACS pump is in AUTO. 0 "A, "C" and "D" Service Water pumps are running. 0 "B Service Water pump is in AUTO. I&C testing causes an inadvertent LOCA signal generation on "C Core Spray logic. At the same time, an infeed undervoltage trip of the 40308 breaker occurs. All systems responded as designed.

With NO operator action, what will be the final alignment of SACS, TACS and Service Water? A. "A SACS Loop is supplying TACS. "A, "C" and "D" Service Water pumps are running.

B. "A" SACS Loop is supplying TACS. "A, "B" and "D" Service Water pumps are running.

C. "B" SACS Loop is supplying TACS. "A, "C" and "D" Service Water pumps are running. D. "B" SACS Loop is supplying TACS. "A, "B and "D Service Water pumps are running. Proposed Answer: C Explanation (Optional): IAW HC.OP-SO.EG-0001 section 3.3.8, HC.OP-SO.EA-0001 section 3.3.1, Loss of Normal supply to 1 OA403 bus requires transfer to alternate supply 40301 from 1 BX501 transformer, this is a dead bus transfer "C" SSW will trip and will start on the LOCA sequencer. Page 41 of 205 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet C. Correct - Loss of Normal supply to 1 OA403 bus requires transfer to alternate supply 40301 from 1 BX501 transformer, this is a dead bus transfer "C" SSW will trip and will re-start on the LOCA sequencer. Loss of "C" SACS pump will cause a swap of TACS to the B loop. HC.OP-SO.EG-0001, section 3.3.8 "C" channel TACS isolations 2522C/2496C close on LOCA Level 1 signal. This causes a swap to E3 loop for supply to TACS on low flow. A. B. Incorrect - TACS isolations 2522C/2496C close on LOCA Level 1 signal.

This causes a swap to B loop for supply to TACS on low flow. Incorrect - TACS isolations 2522C/2496C close on LOCA Level 1 signal. This causes a swap to opposite loop for supply to TACS on low flow. The "B" SW pump will NOT start except for a low flow signal in the associated loop, which did NOT occur.

Incorrect - The "B" SW pump will not start except for a low flow signal in the associated loop, which did not occur.

D. Technical Reference(s) HC.OP-SO.EG-0001 section (Attach if not previously provided) 3.3.8, HC.OP-SO.EA-0001 section 3.3.1 Proposed references to be provided to applicants during examination:

none Learning Objective: Question Source:

SERWATEOO6 (As available)

Bank # 56192 1 __. Modified Bank # __. (Note changes or attach parent)

New __. Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 42 of 205

, HC.OP-SO.EA-0001 (Q) 3.2.14. The lateral sluice gates may be open or closed as system demand dictates.

3.2.15. When placing a Standby SSWS Pump in service, it may be necessary to secure from Standby SACS Loop Temperature Control prior to starting the Service Water Pump.

A momentary low flow auto start signal could be generated in the opposite loop when the standby pump is placed in service.

Since starting the standby pump results in two SSWS pumps discharging thru one Hx, more flow is diverted thru the RACS cross-tie causing lower flow in the opposite loop.

3.3 Interlocks

3.3.1. 3.3.2. 3.3.3. When in AUTO with NO Process Start Inhibit signal, the Station Service Water Pumps auto start on any of the following signals (B AND D Pumps will NOT auto start if control is transferred to the Remote Shutdown Panel): Associated loop low flow c 13,475 gpm (< 1 .O psid across pump strainer). Reactor Water low level (< -38 inches) 0 Dryweil high pressure

(> 1.68 psig) 0 Reactor Building high Radiation

(> 1x10-3 uCi/cc) Refueling Floor high Radiation

(> 2x10-3 uCi/cc) 0 Containment Manual Initiation Station Service Water Pumps auto start on Sequence Timer initiated by LOCA or LOP whether in AUTO or MAN. When in AUTO, HV-2371A(B), SACS HX OUTLET VALVE, or HV-2355A(B), SACS HX 0UTLE:T VALVE, will open 10 seconds after the associated Service Water Pump starts and will CLOSE when the associated pump is stopped. The Valve will OPEN on LOCA or LOP, whether in AUTO' or MAN, if associated pump has been running at least 10 seconds (HV-2371 B and HV-2355B will NOT auto open or close if control is transferred to the RSP). Hope Creek Page 9 of 64 Rev. 33 I I HC.OP-SO.EG-0001 (Q) 3.3.8. HV-2522A-D, TACS INBD(OUTl3D) SPLYIRTN VLVS, AND HV-2496A-D, TACS INBD(0UTRD)

SPLY/RTN VLVS, AUTO CLOSE on the following signals 0 LOCA Level 1 LOP 0 Lo-Lo-Lo Expansion Tank level 0 Respective Pump trip*

• This can be overridderi by placing the respective pump STOP INPUT OVERRIDE switch in the ON position 3.3.9. HV-2522E-F, SACSRACS OUTBD (INBD) SPLY ISOL VLVS, AUTO CLOSE on the following signals:

0 Lo-Lo pressure (2/3 logic) in the TACS Supply Accumulator 0 Lo-Lo pressure (2/3 logic) in Ihe TACS Return Accutnulator 0 Loss of Power 3.3.10. EG-HV-2457NB are tripped shut at 88"combined Hx outlet temperature by TSH-2457NB. EG-HV-2517NB are tripped shut at 88°F combined Hx outlet temperature by TSH-2535NB and will not respond to TIC-2517NB until reset.

E one of these valves is known or suspected to have tripped shut on high temperature, the valve(s) can be reset when temperature has returned to < 88 "F in accordance with Section 5.14.

4.0 EQUIPMENT

REQUIRED Wrench for removing pipe caps.

Hope Creek Page 11 of 90 Rev. 39 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 KIA # 21 8000 G2.2.25 Importance Rating

3.2 Group

1. 1 -~ Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (ADS) Proposed Question: Common 21 The ADS Technical Specification LCO is based on ensuring the ability of ADS to provide a backup to -(I)- during a -(2)- break LOCA. A. (1) RClC (2) Small B. (1) HPCl (2) Small C. (1) RClC (2) Large D. (1) HPCl (2) Large Proposed Answer:

B Explanation (Optional):

B. Correct - TS bases 3.5.F. ADS serves as a backup to HPCl during a small break LOCA accident.

A. C. D. Incorrect - HPCl not RCIC. Incorrect - HPCl not RCIC, Small not Large.

Incorrect - Small not Large Break LOCA. Technical Reference(s)

TS Bases 3.5.1 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

ADSSYSEOOG (As available) Question Source: Bank

1. WTS 2006 Modified Bank # (Note changes or attach parent)

Page 43 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Question History: Question Cognitive Level:

10 CFR Part 55 Content: Comments:

-. New Last NRC Exam -. Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 X Page 44 of 205

, I 3/4.5 EMERGENCY CORE COOLING SYSTE:M - BASES 3/4.5.1 and 3/4.5.2 ECC5- - OPERATING and SHUTDOWN ~___ The core spray system (CSS), together with the LPCI mode of the RHP system, is provided to assure that the core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and iricluding the double-ended reactor recirculation line break, and for smaller breaks followi.ng depressurization by tne ADS. The CSS is a primary source c'f emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining. The surveillance requirements provide adequate assurance that the CSS will be OPERABLE when required. Although all active components are testable and full flow can be der~onstrated by recirculation through a test loop during reactor operation, a cor,plete functional test requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage to piping and to start coo1.ing at the earliest moment. The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled fol.lowing a loss-of-coolant accident.

Four subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS: The surveill.ance requirements provide adequate assurance that the LPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a conplete functional test requires reactor shutdown. The pump discharge piping is mai-ntained full to prevent water hammer damage to piping and to start cooling at t.he earliest moment. Verification every 31 days that ea'zh RHR System cross tie valve on the discharge side of the RHR pumps is closed ,and power to its operator, if any, is disconnected ensures that each LPCI sub,system remains independent and a failure in the flow path in one subsystem will not affect the flow path of the other LPCI subsystem. Acceptable methods (of removing power to the operator include de-energizing breaker control p0we.r or racking out or removing the breaker. For the valves in high radiation areas, verification may consist of verifying that no work activity was perforined in the area of the valve since the last verification was performed.

If one of the RHR System cross tie valves is open or power has not been removed from the valve operator, both associated LPCI subsystems must be c:onside.red inoperable. The 31 day frequency is acceptable, considering that these valves are under strict administrative controls that will ensure that the valves continue to remain closed with either control or motive power removed. The high pre_p&ee camkpt hieetiae Cr) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperatuse in the event of a small break in the reactor coalant system and loss of coolant which does not result in rapid depressurization of the reactor vessel.

The IlPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.

The HPCI system continues to operate until reactor vessel pressure is below the pressure at which CSS operation or LPCI mode of the RHR system operation maintains core cooling.

HOPE CREEK B 3/4 !5-1 Amendment No. 109 I I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 WA # 239002 G2.4.31 I m portarice Rating 4.2 Emergency Procedures I Plan: Knowledge of annunciator alarms, indimtions, or response procedures. (SRV) Proposed Question:

Common 22 The plant is operating at 80% power, with the following:

0 0 Reactor water level is +35 inches Then, an SRV inadvertently opens Which one of the following describes the initial Reactor Water level response and what actions are required IAW AB.RPV-0006 "Safety Relief Valve? A. RPV level will initially rise.

If the SRV CANNOT be closed, within two (2) minutes trip the recirculation pumps and lock the mode switch in Shutdown.

B. RPV level will initially rise.

If the SRV CANNOT be closed, within two (2) minutes reduce the recirculation pumps to minimum and lock the mode switch in Shutdown.

C. RPV level will initially lower.

If the SRV CANNOT be closed, within two (2) minutes trip the recirculation pumps and lock the mode switch in Shutdown.

D. RPV level will initially lower.

If the SRV CANNOT be closed, within two (2) minutes reduce the recirculation pumps to minimum and lock the mode switch in Shutdown.

Proposed Answer: B Explanation (Optional):

B. Correct - IAW AB.RPV-0006 - IF within 2 minutes the SRV fails to close, reduce the recirculation pumps to minimum and lock the mode switch in Shutdown.

RPV Swells up on the RPV pressure reduction when the SRV initially opens. A. C. D. Incorrect. RPV Level will rise Incorrect.

The recirc pumps are to be reduced to minimum, not tripped.

Incorrect.

RPV Level will rise. The recirc pumps are to be reduced to minimum, not tripped. Page 45 of 205 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s) AB.RPV-0006 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: ABRPV4E004 (As available) Question Source:

Bank # Modified Bank

1. -. 22077 (Note changes or attach parent)

New Question History: Last NRC Exam -. 2005 Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Anallisis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 46 of 205

,. PSEG Internal IJse OnIy HC.OP-AB.

RPV-O004(Q)

SAFETY RELIEF VALVE I RETAINMENT OVERRIDE 1 CONDITION 1 ACTION I. Open SRV Fails to Close Time: 1 WITHlN two minutes, perform the following:

I P a. F~EDUCE recirc pumps to minimum speed. P b. LOCK the Mode Switch in SHUTDOWN.

[T/S 3.4.2.1.b, CD-912X9 CD-220Cl Hope Creek Page 2 of 15 Rev. 1

, I I The plant is operating at 90% power, with the 'oliowing. - Reactor water level is 35 inches - Then, an SRV inadvertently opens With NO operator action, which one of the follciwing describes Reactor mater level response?

Reactor Water level will:

A. lower and then return to 35 inc#hes 8. C. rise and then return to 35 inches rise and remain above 35 inches D. lower and remain below 35 indies I 1 I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 KIA # 203000 A4.02 I m porta ri ce Rating

4.1 Ability

to manually operate andlor monitor in the control room (RHFULPCI Injection Mode): System valves. Proposed Question: Common 23 Given the following conditions:

0 Reactor power 85% 0 "A" RHR loop in full flow test mode at 10450 gpm flow Then, a LOCA occurs resulting in the following:

0 Drywell pressure is 4.5 psig increasing 0 5 minutes has elapsed since the LOCA Reactor pressure is now 500 psig and lowering Assuming NO operator action, what is the current status of the following "A RHR Loop valves? 0 LPCl injection valve (BC-HV-FO17A) 0 Test Valve (BC-HV-F024A) 0 HX Bypass valve (BC-HV-F048A)

A. (BC-HV-FO17A) - open (BC-HV-F024A) - open (BC-HV-F048A) - closed B. (BC-HV-F017A) - open (BC-HV-F024A) - closed (BC-HV-F048A) - closed C. (BC-HV-FO17A) - closed (BC-HV-F024A) - open (BC-HV-FO48A) - open D. (BC-HV-FO17A) - closed (BC-HV-F024A) - closed (BC-HV-F048A) - open Page 47 of 205 I I ___. ES-40 1 Sam p I e Written Examination Form ES-401-5 Question Worksheet Proposed Answer:

D Explanation (Optional):

IAW HC.OP-SO.BC-0001 When a LPCl signal is received, the system aligns for RPV injection.

F048 - Receives an open signal for 3 min F017 - Opens when RPV pressure drops below 450 psig F024 - Receives a close signal but can be overridden D. Correct. A. B. C. Incorrect.

F024A will be closed Incorrect.

F017A will still be closed, F024A will be closed, F048A will be open Incorrect.

F017A will still be closed, F048A will be open Technical Reference(s) HC.OP-SO.BC-0001 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: RHRSYSE014 (As available) 56220 -. Question Source: Bank

1. Modified Bank # -. (Note changes or attach parent)

New Question History: Last NRC Exam Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 48 of 205 7 HC.OP-SO.BC-OOOl(Q) .c TABLE 1 VALVE TABLE Valve Power Auto Actuations Interlocks FO17A B C D 10B212 10B222 1 OB232 1 OB242 Auto open on WGX hit. if following condition exist: (1) LP*T init. present in respective RHR loop logic (?)Power is avail. on respective pump bus (3) Reach- press. less than 450 wig (my- - be overridden by "AUTO OPEN OVW") Rx press. must be less than 450 psig to open valve either MAN or AUTO. Placing Ch. B RSP to EMERG will close F017B & inhibit all associated automatic

& OVLD protection Features.

The LPCI in-jection valve must be 100% closed (in the respective loop $\ith a LPCI initiation signal present) to OPEN: F027A(B).

F024A(B), FO 16, F02 1 (LPCI Injection)

F047A lOB212 None Placing Ch. B RSP to EMERG transfers F047B (HX Inlet) F003 A 10B2 12 Throttle Valve Placing Ch. B RSP to EMERG transfers F003B (HX Inlet) B 10B222 to RSP B 1 OB222 to RSP F048A 10B212 OPEN upon receipt.of a LPCI initiation signal Valve operation is inhibited for 3 minutes upon + B 1 OB222 receipt of hPCl init. signal, valve interlocked (HX Bypass) open Ch. B RSP to EMERG transfers F048B to the RSP. OVLD protection can be bypassed in the OPEN direction (BYP IN OPEN)

TABLE 1 VALVE TABLE Valve Power Auto Actuations Interlocks Q 1 \s: ~ ~~ FOI 5A,B 1 OB48 1 (S/D cooling return 10B242 F008 (S/D cooling suction) 1 OB242 NS4 logic (D) must be reset following isolation.

First 3 are throttle valves, Channel D RSP transfer SW to EMERG > F015B. and F008 Overload B/P with "OVLD BYP in OPEN (CLOSE)" Switch.

F008 can be opened remotely only if MCC breaker Key Lock switch is in ARM position.

(1) LPCI initiation logic must be reset to allow operation. Placing Channel D RSP to EMERG will close FOIOB. ("OVLD BYP in OPEN (CLOSE)")

Initiation signal must be reset to disable manual OVRD condition. Placing Channel B RSP to EMERG rransier controi OfFU24B to RSP (and B/P all automatic control features and OVLD Close on NS4 signals 12.5" or 82 psig reactor pressure.

These vafves off NS4 channels C and D manual initiation off Channel D. Close on either auto 3 man LPCI init. signal L,oop "C" logic for 10A Loop "D" logic 10B FO 1 OA B 10B232 10B242 F024A B 10B212 10B222 Auto close on LPCl init. signal (auto or man) A logic for 24A; B logic for 24B. Can be c\rprrjCl,&n n.heg a!! ofthe f~~jgyy~~g exist concurrently:

(1) LPCI init. signal present (2) OVRD p/b on 650 panel) Auto close on LPCI init. logic "A" for 27A B conditions exist concurrently (1) LPCI init. present (2) DW pres. (3) respective injection valve FO 17A(B) 100% closed) (RHR A&B Full F017A(B) 100% CLOSED. (AUTO CLOSE protection)F024A(B) . Flow Test) F027A IOB212 OVERRIDDEN and illuminates when auto-close bypassing all auto control and OVLD protection.

Man OVRD in effect till init. signal reset F027A(B) must be 100% closed to open FOOGA(B) B 1 OB222 logic 27B (man OVRD when following signal B/P. B RSPiEMERG closes F027B (Torus Spray) Hope Creek P .- i\c I. tr I , I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 KIA # 209001 K4.09 Importance Rating 3.3 -___ Knowledge of LOW PRESSURE CORE SPRAY SYSTEM design feature(s) and/or interlocks which provide for the following: Load Sequencing Proposed Question:

Common 24 Given the following conditions:

0 Drywell pressure increased to 2 psig 0 Off-site power is lost. Which one of the following describes the start sequence for the core spray systems after off-site power was lost? A. Core Spray pumps "A', "B'l, "C", and 'ID" start six seconds after the diesel generator output breaker is closed. 6. Core Spray pumps "A, "B", "C", and "D start immediately after the diesel generator output breakers are closed. C. Core Spray pumps "A and "C" start immediately after the diesel generator output breaker is closed. Core Spray pumps "B" and "D" start six seconds after the diesel generator output breakers are closed. D. Core Spray pumps "A and "B" start immediately after the diesel generator output breakers are closed. Core Spray pumps "C" and "D" start six seconds after the diesel generator output breakers are closed. Proposed Answer: A Explanation (Optional): IAW HC.OP-SO.BE-0001 A. Correct - Core Spray pumps "A', "B", "C", and "D" start six seconds after the diesel generator output breaker is closed. - With a LOP, all pumps start 6 seconds after the edg output breaker closed. B. C. D. Incorrect - They all start 6 seconds after diesel generator output breaker closes. Incorrect - They all start 6 seconds after diesel generator output breaker closes. Incorrect - They all start 6 seconds after diesel generator output breaker closes. Page 49 of 205 I I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s) HC.OP-SO.BE-0001 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

CSSYSOE005 (As available) Question Source:

Bank # 80663 Modified Bank # -. (Note changes or attach parent)

New Question History: Last NRC Exam ~ Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 50 of 205 I HC.OP-SO.BE-OOOl(Q) 5.3 Core Surav Svstem Automatic Initiation/Observation 5.3.1. WHEN A(B,C,D) INIT AND SEALED-IN is on, OBSERVE the following:

7=7A+ B. C. D. Core Spray Pumps start after a 10 second time delay (6 seconds after Diesel Generator breaker closes fbllowing LOP as indicated by: 0 PUMP A(B,C,D) RUNNING is on. - 0 AI-6357A(B,CYD)

CORE SPEUY PUMP A(B,C,D) MOTOR AMPS indicates 5 85.5 amps. - PI-R600A(B)-E21 CORE SPRAY SYSTEM A(B) PRESS indicates

< 475 psig. - - ENSURE HV-F015A(B)

TEST RETURN VALVE is CLOSED. WHEN PI-R605-C32 REACTOR PRESSURE < 461 psig, HV-FOOSA(B)

LOOP A(B) INBD INJECT VLV Will open. - OBSERVE FI-RGOlA(B)-E21 CORE SPRAY SYSTEM A(B) FLOW for the following:

0 WHEN flow is 775 gpm, HVF031A(B)

LOOP A(B) MINIMUM FLOW ISLN VLV closes. 0 Flow increases to 6350 gpm. - Hope Creek Page 9 of 21 Rev. 10 I , ES-40 1 Sam p le Written E:kam i nation Question Worksheet Form ES-401-5 Examination Outline Cross-reference:

Level RO SRO -~ Tier # 2 Group # 1 KIA # 259002 K5.03 Importance Rating 3.1 Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER LEVEL CONTROL SYSTEM : Water level measurement. Proposed Question: Common 25 Given the following:

0 The plant is at 85% power 0 All three Reactor Feed Pumps are in Auto 0 RPV Narrow Range Level instruments indicate:

0 N004A = 34 inches 0 N004B = 35 inches 0 N004C = 35.5 inches Which of the following describes the plant response to actual Reactor water level if a slow leak developed through the N004B detector equalizing valve eventually causing a gross failure of N004B? Actual Reactor water level would.. . A. B. C. lower 1 inch, then rise 0.5 inches. rise 1 inch, then lower 0.5 inches. lower 0.5 inch, then rise 1.5 inches.

D. rise 0.5 inch, then lower 1.5 inches. Proposed Answer:

C Explanation (Optional): C.

Correct C. CORRECT - lower 0.5 inch, then rise 1.5 inches. initially N004B is selected since DFCS selects the MEDIAN RPV level signal when three good signals are available.

With a leak through the N004B equalizing valve, N004B INDICATED level would begin to rise, resulting in a lowering of ACTUAL RPV level.

As soon as N004B exceeded 35.5 inches INDICATED, N004C would become the MEDIAN RPV level signal. ACTUAL RPV level would have lowered 1/2 inch durinq this transitim When N004B gross fails, N004A (the lowest of the two remaining signals) will become the controlling level signal. RPV water Page 51 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet level will then rise since INDICATED level on N004A is 34 inches. This is a 1.5 inch rise from the previous level. IAW FW Control Lesson Plan, Page 18 - On a failure of a narrow range Rosemount Level Detector (PDT-N004A, B or C) with the feedwater system in automatic three-element control from the Master Level Controller, level stays near its setpoint since the level signal is now the lower of the two good remaining level inputs.

If another level transmitter were to fail, the remaining signal is now the controlling signal.

A. 6. D. INCORRECT - lower 1 inch, then rise 0.5 inches. Level initially lowers by 112 inch. INCORRECT - rise 1 inch, then lower 0.5 inches. Level initially lowers.

INCORRECT - rise 0.5 inch, then lower 1.5 inches. Level initially lowers. Technical Reference(s) Engineering Drawing H-I -AE- FW Control LP - (Attach if not previously provided)

ECS-0128-0 NOH04FWCONTC-04 Proposed references to be provided to applicants during examination:

none Learning Objective: Question Source:

Question History: Question Cognitive Level: 10 CFR Part 55 Content: Comments:

FWCONTEOOI (As available) 53240 -. Bank # Modified Bank # -. (Note changes or attach parent)

New -. Last NRC Exam -. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Page 52 of 205

- - LESSON NAME: NOH04FWCONTC FEEDWATER CONTROL SYSTEM - 10/29/2007

[OW. 77 I d. On a loss of any Feed Flow signal (high or Low) the control circbiit will transfer to single element control. The operator can detect *:his failure by noting that the Total Feed Flow signal on the FR-R6O- recorder on panel 1OC65OC and by the DFCS TROUBLE overh 3ad alarm. Additionally, the Operator Display screen(s) will digitally display the failed detector(s).

I Fig. 3 - 3 e. On a failure of a narrow range Rosemount Level Detector (PDT N004A, B or C) with the feedwater system in automatic three- element control from the Master Level Controller, level stays nea its setpoint since the level signal is now the lower of the two gocd remaining level inputs. If another level transmitter were to fail, t \e remaining signal is now the controlling signal.

I) The operator can detect the level transmitter has failed h gh because the DFCS TROUBLE overhead alarm will annunciate, the appropriate screen will display the failed transmitter output on the Operator Display Station, and indicate upscale on the respective LI-R603A, B, or C on panel 1 OC650C. The operator can detect the level transmitter has failed ION because the DFCS TROUBLE overhead alarm will annunciate, the appropriate screen will display the failed transmitter output on the Operator Display Station, and indicate downscale on the appropriate LI-R603A, B, or C m panel 1 OC65OC. 2) /Fig. 7 - I f. If the RFP control is in DP Control, RFP control would shift to Manual on a failure of the Reactor Pressure or Reactor Feed Punp Discharge Header Pressure transmitter.

The operator can deteci this failure from the DFCS TROUBLE and RFP TURBINE AUTC XFR TO MANUAL alarms, and the CRT display screen.

IV. INSTRUMENTATION ALARMS AND CONTROL (OBJ. 5a-d, 6 I A. Instrumentation

1. Panel 1 OC650C - Main Control Room a. I nd ica t ions 1) Individual RFF'T LOW PRESS CONTROL VLVS position indication:

PI-1794A1, 61, Cl (0-10 x 10% scale) Page 18 of 41 C:\Documents and Settings\sdennis\Mv Documents\Hope Creek 2008-2009\References\LESSON PLANS\Feedwater ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 K/A # 261 000 A3.02 I m porta vi ce Rating 3.2 Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including:

Fan start Proposed Question:

Common 26 Given the following:

0 0 0 0 0 0 The plant is operating at full power A Loss of Offsite power occurs Drywell pressure is 3.3 psig and rising "A Emergency Diesel Generator fails to start "A FRVS Vent Fan is in Auto Lead "B" FRVS Vent Fan is in Auto Which one of the following describes the status of FRVS 2 minutes after the LOCA sequencers actuate? (Assume NO operator action)

A. Only 3 Recirc Fans and NO Vent Fans start B. Only 3 Recirc Fans and ONE Vent Fan start C. Only 4 Recirc Fans and NO Vent Fans start D. Only 4 Recirc Fans and ONE Vent Fan start Proposed Answer: D Explanation (Optional): D. Correct - Loss of power lo "A Bus will prevent the "A and "E" Recirc fan from starting. The "F Recirc fan will not start until 30 seconds after the LOCA sequencer actuates.

The "A vent fan has no power and the "B" will start after 45 second time delay. The B fan has a flow sensor in the A ductwork.

IAWHC.OP-SO.GU-0001 - FRVS Recirculation Fans AV213 through FV213 in AUTO and FRVS Vent Fan in AUTO LEAD will automatically start under any of the following conditions: High Drywell Pressure (1.68 psig). Low RPV Water Level (Level 2, - 38"). Refueling Floor Exhaust Duct High Radiation Reactor Building Exhaust Air High Radiation D. Correct. Page 53 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet A. B. Incorrect.

4 recirc fans start. C. Incorrect. B vent fan will be running. Incorrect.

Only 3 recirc fans running and B recirc fan Technical Reference( s) H C. OP-SO. G U-000 1 LPNOHOI RBVENTC-00 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

RBVENTEOOG (As available) Question Source:

Question History: Question Cognitive Level:

10 CFR Part 55 Content: -. Bank # Modified Bank # -. (Note changes or attach parent)

New Last NRC Exam X -. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Comments:

Page 54 of 205 I ' ' LESSON NAME: REACTOR BUILDING VENTILATION NOH01 RBVENTC 08/14/06 # Indicates Bezel alarms.

yellow flashing backlights

• Indicates no audible alarm I OBJ. 3cJ 7,24c,28 ___. I c. Auto Features 1) The FRVS vent fan selecied for Auto Lead operation will auto start on the same signals that start the FRVS recirc fans.

The LOCA Sequencers will start the Auto Lead fan 19 seconds after initiation.

The vent fan selected for Auto operation will auto start under the following conditions:

a) 2) 3) Automatic start signal present as described in 1 and 2 above and I4EC-3665 I b) A low flow condition, (6950 scfm), exists on the Auto Lead fan far more than 45 seconds. A low fiow condition, (6950 scfm), for 30 seconds will trip the associated fan. Deluge flow, (9.0 gpm), will trip the associated fan.

4) 5) Table 4 M-57-1 9. Purge and Vent Valves

a. All valves are pneumatically operated and fail closed on a loss of power or air. The 26" Outboard Primary Containment to CPCS valve and the 24" Outboard Torus to CPCS valve are each bypassed by 2" motor operated vent valves that can be used to control containment pressure or supply a backup flowpath to the HZ recombiners.

The RBVS supplies purge air through HD-9372C to the torus and drywell. Suppression Chamber and Drywell atmospheres are vented or purged to RBVS via HD-9372A.

The Containment Atmosphere Control Vent and Purge Valves will automatically close on the following:

1) 1.68 psig Drywell pressure

b. c. d. -~ -__ i [ Deleted: 71 Page 33 of72 C \Documents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\Keferences\LE SSON PLANS\Reactor Building Ventilation\Master\Lesson Plans\NOHOl RBVENTC-DO REACTOR BUILDING VENTILATION DOC Pb $& LESSON NAME: REACTOR BUILDING VENTILATION NOHOIRBVENTC 08/14/06 IOBJ. 6,27 -___. I c. Automatic features 1) The following signals will start all six recirc fans:

a) -38" Rx Lvl. b) 1.68 psig Drywell 13ressure.

c) d) Reactor Building L'ent. Exh. hi-hi rad 1x10" pci/cc. Refuel floor Exh. hi-hi rad 2~10'~ pci/cc. I Obj. 3c, 24c .I The LOCA Sequencers start recirc fans A,B,C and D 19 seconds after initiation. Fans E and F start 30 seconds after initiation.

After all six fans are started in response to an auto (LOCA) signal, E and F may be stopped and placed back in Auto. Then, if one of the running recirc fans (A,B,C or D) has a low flow condition, both E and F will restart with no time delay. Fans E and F are the only fans with this automatic start feature. A low flow condition, (18,000 scfm for A-E and 15,000 for "F") on any operating fan for 15 seconds will trip that fan and provide a low flow alarm.

A low flow condition of 17,200 scfm for the A-D fans, if started due to an automatic signal, will auto start the E & F fans if they are in auto. Signals are from two independent flow elements, transmitters and switches one is sent to echo fan, (FSL-9377AA thru DA), and one is sent to the foxtrot fan, (FSL-9377AB-DB). Deluge flow, (9.0gpm), will trip the associated fan.

M-84-1 1 Figure 1 Obj. 3c,24c 7. FRVS Vent System a. The system consists of two 1OO?h capacity units whose purpose is to maintain negative reactor building pressure during accident conditions. Each unit is identical and consists of: 7 -~ I Deleted: 71 i__ Page 31 of72 C:\Docurnents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\References\LEISSON PLANSReactor Building Ventilation\Master\Lesson Plans\NOHOl RBVENTC-00 REACTOR BUILDING VENTILATIC)N.DOC I I Y 7L LESSON NAME: REACTOR BUILDING VENTILATION NOH01 RBVENTC 08/14/06 ELECTRIC POWER SUPPLIES FOR RBVS AND FRVS FANS

~ ~~ ~ FAN RBVS AVH300 BVH300 CVH300 I RBVE AVH301 BVH301 CVH301 Recirc Unit AVH213 Recirc Unit BVH213 Recirc Unit CVH213 Recirc Unit DVH213 Recirc Unit EVH213 Recirc Unit FVH213 FRVS Vent Fan A Vent Fan B FRVS TABLE 2 POWER SUPPLY 1 OB440-1 OB448 1 OB41 01 OB4 15 1 OB430-1 OB437 1 OB440-10B448 1 OB420-1 OB426 10B410-10B415 10B410 1 OB420 1 OB430 1 OB440 1 OB450 1 OB460 10B212 1 OB222 CHANNEL NOTE1 D NOTE1 A NOTE1 C NOTE1 D NOTE1 B NOTE1 A A,, B NOTE 1 : The RBVS and RBVE FANS are supplied Power thru 1 E and Non-I E Breakers. The associated channel 1 E breakers trip on -38 Rx LVL, 1.68# Drywell Pressure, Reactor Bldg. Vent Exh. Hi-Hi rad IxlO"pci/cc, and Refuel Floor Vent Exh. Hi-Hi rad 2x1 O-31~ci/cc.

The Non-1 E breakers trip on undervoltage.

I I Deleted: 71 Page 69 ofJ2 C \Documents and Settings\sdennis\My Documents\Hope Creek 2008-2009\References\L ESSON PLANS\Reactor Building Ventilation\Master\Lesson Plans\NOHOl RBVENTC-00 REACTOR BUILDING VENTILATION DOC PSEG Internal Use Only HC.OP-SO.GU-OOOl(Q) 3.2.6 With respect to SACS cooling and EDG power supply, the following is required for FRVS Recirculation Fan Operability:

-- In OPERATIONAL CONDITION 1,2, and 3, SACS and EDG backing are required support systems for all fans. In OPERATIONAL CONDITION

"*" (OPDRVs), SACS and EDG backing are required support systems for only three recirc fans (additional fans may he considered operable without SACSl cooling when powered from an offsite source). OPERATIONAL CONDITION It* (handling recently irradiated fuel in the secondary containment) is prohibited by administrative procedures.

During fuel handling of fuel that is not recently irradiated and CORE ALTERATIONS, either the RBVS or one FRVS vent fan is required to be operating and capable of drawing air into the building and exhausting through a monitored pathway. REFER TO SH.OP-AP.ZZ-01 OS(Q) for details.

[80005080,70020722]

3.3 Interlocks

3.3.1 FRVS RecirculationFans AV213 through FV213 in AUTO and FRVS Vent Fan in AUTO LEAD will automatically start under any of the following conditions

High Drywell Pressure (1.68 psig). 0 Low RPV Water Level (Level 2, - 38"). Refueling Floor Exhaust Duct High Radiation (2 2x10 -' micro Ci/cc). Reactor Building Exhaust Air High Radiation (2 1 xl0 -3 micro Cilcc).

3.3.2 With the FRVS Recirculation Fans EV213 and FV213 in AUTO, and the A, B, C, and D fans running due to receiving a LOCA a low flow signal &om a running FRVS Recirculation Fan will automatically start the EV2 13 and FV2 1 3 fans. High Rad signal, 3.3.3 The PCIS signal is locked into the FRVS logic and resets after the PCIS is reset and the A, B, C, and D FRVS Recirculation fans are manually secured and all four fans are below the low flow setpoint. Placing the E and F Fans in "Lockout" will prevent an Auto Start while securing the A, B, C, and D fans. This action will hop 2 FRVS Fans and place the plant in an Action Statement. Tech Spec 3.6.5.3.2 should be reviewed to ensure compliance.

Hope Creek Page 7 of 27 Rev. 23 I I I I ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 KIA # 201003 K1.01 Importance Rating

3.2 Group

1. 2 Knowledge of the physical connections and/or cause-effect relationship!; between CONTROL ROD AND DRIVE MECHANISM and the following: CRD Hydraulic system Proposed Question: Common 27 Which one of the following describes scram valve response and indication on the full core display following a reactor scram? A. The scram inlet valve opens faster than the scram outlet valve.

The blue scram light will be illuminated as soon as the scram inlet valve is fully open. B. The scram outlet valve opens faster than the scram inlet valve.

The blue scram light will be illuminated as soon as the scram outlet valve is fully open. C. The scram inlet valve opens faster than the scram outlet valve.

The blue scram light will be illuminated when both scram inlet AND outlet valves are fully open. D. The scram outlet valve opens faster than the scram inlet valve.

The blue scram light will be illuminated when both scram inlet AND outlet valves are fully open. Proposed Answer:

D Explanation (Optional): IAW CRD Lesson Plan NOH04CRDHYD-04 Section 10.h. & 10.9.3) - Scram Outlet Valve (XV-127) - fast acting globe valve that is opened by an internal spring and which exhausts water from the top of the drive piston. The scram outlet valve opens faster than the scram inlet valve because of a stronger spring and more rapid venting.

A position indicator switch on the valve energizes a blue light on the Control Room panel 10C650C when the scram inlet and outlet valves are open. D. Correct.

A. Incorrect - The scram light requires both valves open. The outlet valve opens faster. B. Incorrect - The scram light requires both valves open C. Incorrect - The outlet valve opens faster. The light will not illuminate unless both valves are open Page 55 of 205 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s) Lesson plan NOH04CRDHYD- (Attach if not previously provided) 04 Proposed references to be provided to applicants during examination:

none Learning Objective: Question Source:

Question History: CRDHYDE025 (As available)

Bank # New X (Note changes or attach parent) Modified Bank

1. -. Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 56 of 205 LESSON NAME: NOH04CRDHYD-04 CONTROL ROD DRIVE HYDRAULICS - 1/15/08 b) In OPERATIONAL CONDITIONS 1 and 2, control rod speeds are determined by measuring the stroke time between positions 04 and 08 (in both the insert and withdraw directions), then ensuring that the stroke time IS 2 3.'7 seconds and 5 4.4 seconds. Additional requirements for controlling the position of the integral flow control devices are also provided following maintenance activities on the respective CRDWI. 6) g. Scram Inlet Valve (XV-126) - Opens to supply pressurized water to the bottom of the drive piston. I) 2) 3) Fast acting globe valve operated by an internal spring and system pressure. Held closed by air pressure applied to the top of a diaphragm operator.

A position indicator switch on the valve energizes a blue light on the Control Room panel 1OC65OC when the scram inlet and outlet valves are open. Scram Outlet Valve (XV-127) - fast acting globe valve that is opened by an internal spring and which exhausts water from the top of the drive piston. The scram outlet valve opens faster than the scram inlet valve because of a stronger spring and more rapid venting. Scram Accumulator - stores sufficient energy to fully insert a control rod at lower reactor vessel pressures (less than 600 psig). At higher reactor vessel pressure, scram accumulator pressure is supplemented by reactor pressure.

1) h. I. The scram accumulator is a hydraulic cylinder with a free floating piston.

The piston separates water from nitrogen gas.

A check valve in the scram accumulator charging line prevents loss of water pressure if supply pressure in the charging header is lost. 2) Page 27 of 67 I I ES-401 Sample Written E.xamination Form ES-40 1-5 Question -. Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 2 KIA # 226001 A3.01 I m porta rice Rating

3.0 Ability

to monitor automatic operations of the RHWLPCI: CONTAINMENT SPRAY SYSTEM MODE including: Valve operation Proposed Question: Common 28 A LOCA has occurred and the following conditions exist.

0 Reactor is scrammed 0 0 Drywell Pressure is 10.4 psig and rising RHR Loop "A is injecting to the RPV RHR Loop "B" is in Torus Spray Which one of the following describes how RHR Containment Spray Isolation Valves BC-HV-FO16A(B) and BC-HV-F021A(B) logic would function under these conditions?

A. NO valves could be opened simultaneously B. ONLY the F016A & F021A could be opened simultaneously C. ONLY the F016B & F021B could be opened simultaneously D. BOTH the F016A & F021A and the F016B & F021B could be opened simultaneously Proposed Answer:

C Explanation (Optional):

C. Correct IAW RHR Lesson Plan NOH01 RHRSYSC-06,Section IV.A.14.b.

Page 43 - FO16A(B) and F021A(B) are interlocked such that both valves can only be opened simultaneously when: There is a LPCl initiation signal present AND High drywell pressure condition exists AND F017A(B)

IS 100% CLOSED. In this case, a LPCl initiation signal is present and high drywell pressure exists. With the F017A is open and therefore the F016A &F021A cannot be opened simultaneously. With RHR loop "B" in Torus Cooling, the F017B is closed, therefore, the F016B & F021 B can be opened simultaneously.

A. 6. D. Incorrect. - The B loop valves can be opened simultaneously Incorrect. - ONLY the B loop valves can be opened simultaneously Incorrect - ONLY the B loop valves can be opened simultaneously Page 57 of 205

, I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s)

RHR Lesson Plan NOH01 RHRSYSC-06 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

RHRSYSEOI 1 (As available)

-. Question Source: Bank

1. Modified Bank # -. (Note changes or attach parent)

New X Question History: Last NRC Exam _. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 58 of 205 I , I 1 LESSON NAME: NOH01 RHRSYSC-06 I RESIDUAL HEAT REMOVAL SYSTEM - 02/27/08 - _I 2) OVLD/PWR FAIL (AMBER) status indication for all/valves adjacent to control switch on 1 OC650. (See IV.A.2.a.2))

NOTE: Containment Spray flow indication on 1OC65OC, PAMS section RHR Loop A and B Containment Spray flow meter indication FI-4462A,B.

(0 - 12,000 GPM)

I Obj. EO77 - _I b. Control

1) OPEN/CLOSE push-button controls are provided on 1 OC65C for FO16A(B) and FO21A(B).

can only be opened simultaneously when:

There is a LPCl initiation signal present High drywell pressure condition exists and FC121A(B) are interlocked such that both valve; AND AN! FO17A(B) IS 100% CLOSED. (Obj. EO75 _I 3) F016B and F021A(B) automatically close when their associated RSP transfer switch is placed in the EMERGENC position a) Channel A - F021A b) Channel B - F016B, F021B

15. RHR Shutdown Cooling Return Isolation Valves (HV-F015A,B)

I Obj. *004d, Obj. *071 - J a. instrumentation

1) 2) b. Control 1) 2) OPEN/CLOSE valve position indication is available for each valve on 10C650, and for HV-FO15B on 1OC399. OVLD/PWR FAIL status indication is available adjacent to ea :h valve control switch on 10C650. (See IV.A.2.a.2). INCR/DECR push-button controls are provided on 1 OC650 fo FO I 5A( B). Torque switch pratection can be bypassed by the operator when positioning the valve if the "OVLD BYP IN OPEN/OVLD BYP IN CLOSE" push buttons are depressed.

Page 43 of 85 C:\Docurnents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\F:eferences\LESSON PLANS\Residual Heat Rernoval\Master\Les:

in ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 2 KIA # 234000 K3.03 Importance Rating

3.1 Knowledge

of the effect that a loss or malfunction of the FUEL HANDLING EQUIPMENT will have on following: Fuel handling operations. Proposed Question: Common 29 The plant is in a refueling outage with the following conditions.

0 Mode Switch in REFUEL 0 0 0 All Control Rods are full in Refuel Platform over the core Fuel grapple is being raised carrying a fuel bundle Then, the Refuel Bridge System "Rods Out Relay" contact fails indicating all rods are NOT full in. Which one of the following describes the effect on the Main Hoist, if any, of the failed contact on f ue I hand I i ng operations?

A. NO effect. B. C. D. ONLY Main Hoist raise motion is prevented. ONLY Main Hoist lower motion is prevented.

Main Hoist raise and lower motion is prevented. Proposed Answer:

D Explanation (Optional):

IAW Refueling lesson plan NOH01 REFUEL-03 Table 2, first page. D. Correct -with the system "seeing" one rod out with the platform over the vessel, no motion is permitted A. B. C. Incorrect - ALL motion is prevented due to the system "seeing" one rod out. Incorrect - ALL motion is prevented due to the system "seeing" one rod out. Incorrect - ALL motion is prevented due to the system "seeing" one rod out. Page 59 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s)

NOH01 REFUEL-03 Table 2 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

REFUELE005 (As available)

-. Question Source:

Bank # Modified Bank # _. (Note changes or attach parent)

New X Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 60 of 205

, 1. 2. 3. 4. 5. 6. 7. -r'w )$ L'LESSON NAME: NOHOIREFUEL-02 REFUELING PLATFORM - 10/05/06 Hoist motion is not allowed if a fault condition exists unles.~fault is overridden by activating the Fault Override Keyswitch. Hoist motion is not allowed manually when in the semi-automatic or automatic mode. (Note:

Movement of the joysticl, will stop semi-auto or auto operalion.)

Hoist motion is not allowed if not in a permissive secure travel zone unless bypassed by activating the Travel Override Push-button. If fuel loaded, the hoist may only be raised to the Up Limit when in Travel Override. {Fuel Hoist Interlock) Hoist motion is not allowed: When the Rods Out Relay contact (ROR) is opened indicating at least one control rod is withdrawn.

AND The refueling platform is over the reactor vessel (RS I contat.:t is open) When the Interlock Override Key Switch is turned to the ovm-ide position the hoist motion outputs from the PLC are disabled. (Note: the hoist is externally limited to slow speed in this condition.) When hoist motion is commanded the hoist screen will be di.splayed on the touchscreen.

When hoist motion is Commanded and all the interlocks are

\atisfied the output for the hoist inhibit is turned on. /Table 2 Motion Coqtrol (Mdal Mode) Common Hoist Motion Requirements 4 ----I 8. 9. 10. 11. Following are the requirements for either hoist raise or hoist lower motion. When hoist motion is commanded and all the interlocks are satisfied the output for the motor brake and redundant brake is turned on to energize the hoist brakes. Whenever the Hoist is in motion, the encoder is monitored to insure that the hoist position is changing in the direction expected for the direction of travel being commanded.

Failure of the encoder value to change in the predicted manner will result in a Hoist Motion Fault. This error condition can be cleared by cycling power (Off, On) on the Stadstop Station. When the hoist motion command is turned off, the output for.hoist inhibit is turned off after the speed command decays below a preset value or a two second timer elapses. When the hoist motion command is turned off and one second after the output for hoist inhibit is turned off and the output for the hoist brakes is then turned off.

I ' LESSON NAME: NOH01 REFUEL-02 REFUELING PLATFORM - 10/05106 Bridge Reverse Stop Interlock

1. 1 Bridge Reverse Stop Interlock

1. I prevents reverse bridge motion and illuminates the Bridge Reverse Stop Interlock

1. I light when:

a. The Rod Out Relay (ROR) contact is open indicating at least one control rod is withdrawn in the reactor. AND A contact of bridge refueling switch RSl opens indicating that the bridge is over the reactor vessel.

AND An internal setpoint of the main hoist load weighing s,ysteni is operated indicating a hoist fuel loaded condition (550 pounds or more on grapple.)

b. 4 Bridge Reverse Stop Interlock

1. 2 3ridge Reverse Stop Interlock

1. 2 prevents reverse bridge motion and illuminates the Bridge ieverse Stop Interlock

1. 2 light when:

a. A contact of RS2 opens indicating that the bridge is over the reactor. AND A signal from the Reactor Manual Control System indicating that the Reactor mode switch is in Start Up. Rod Block Interlock

1. 1 <od Block Interlock

1. I prevents withdrawal of control rods from the reactor by opening output :ontact Rod Block #I Interlock when the bridge is over the reactor vessel. This illuminates the

<od Block Interlock

1. I light and occurs when:

1 contact of RSI opens indicating that the bridge is over the reactor vessel. IND 1. r, IND 4 signal from the reactor manual control system indicating the Reactor Mode Switch in Rehel. Rod Block Interlock

1. 2 Cod Block Interlock

1. 2 prevents withdrawal of control rods from the reactor when Reactor node switch is in Startup and the bridge is over the vessel.

7 his is indicated on the machine by lpening contacts to the PLC. The Rod Block Interlock

1. 2 light illuminates when: i contact of RS2 opens indicating that the bridge is over the reactor vessel.

iND i signal from the Reactor Manual Control System indicating that the Reactor mode switch is in )tart Up. b. An internal setpoint of the main hoist load weighing system is operated indicating a hoist fuel loaded condition (535 pounds or more on grapple) vp LESSON NAME: NOHOlREFUEL-02 REFUELING PLATFORM - 10/05/06 Table 5 Refueling Interlocks Fuel Hoist Interlock The main fuel hoist motion will be inhibited and the Fuel Hoist Interlock light will be illuminated when the following conditions exist.

The Rod Out Relay (ROR) contact is opened indicating at It:ast one control rod is withdrawn in the reactor. AND The refueling platform is over the reactor vessel. Le. (RSl contact is open) Frame Hoist Interlock The frame hoist motion will be inhibited by opening the output contact (FRI) and the Frame Hoist Interlock light will be illuminated when the following conditions exist. The Rod Out Relay (ROR) contact is opened indicating at least one control rod is withdrawn in the reactor. AND There is a 400 pound load or greater detected on the Frame hoist. (fuel loaded)

AND The refueling platform is over the reactor vessel. i.e. (RSI contact is open) Monorail Hoist Interlock The Monorail hoist motion will be inhibited by opening the output contact (MRI) and the Monorail Hoist Interlock light will be illuminated when the following conditions exist.

The Rod Out Relay (ROR) contact is opened indicating at least one control rod is withdrawn in the reactor.

4ND a 400 pound load or greater detected on the rhe refueling platform is over the reactor vessel.

.e. (RSl contact is open)

I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group #: 2 WA # 214000 K4.01 I m porta rice Rating 3.0 Knowledge of (ROD POSITION INFORMATION SYSTEM) design feaiure(s) and or interlocks which provide for the following: Reed switch locations Proposed Question:

Common 30 The control rod reed switches are designed to illuminate a Full Core Display amber DRIFT indicator under which one of the following conditions?

NO rod motion command present and.. , A. ONLY an odd reed switch closed.

B. ONLY an even reed switch closed.

C. an odd AND even reed switch closed.

D. the ROD DRIFT TEST PB is depressed.

Proposed Answer: A Explanation (Optional): IAW HC.OP-SO.SF-0001, Att.6 Section B.2. - ROD DRIFT a control rod is changing position with no command signal or the ROD DRIFT TEST PB pressed while a rod is being moved (amber).

A. Correct -see HC.OP-SO.SF-0001 attachment If 6 B. C. D. Incorrect - requires odd reed switch without rod motion command signal. Incorrect - requires odd reed switch without rod motion command signal. Incorrect - Rod motion must be present Technical Reference(s)

HC. OP-S0.S F-000 1 (Attach if not previously provided) Proposed references to be provided to applicants during examination: none Learning Objective:

MANCONE002 (As available)

Question Source: Bank

1. 54400 Page 61 of 205 I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

_. Modified Bank

1. __ (Note changes or attach parent)

New _. Question History: Question Cognitive Level: 10 CFR Part 55 Content: Last NRC Exam _. Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 X Comments:

Page 62 of 205 ATTAC x HC.OP-SO.SF-0001 (Q) ENT 6 (Page 3 of 3) REACTOR MANUAL CONTROL SYSTEM OPERATION REACTOR MANUAL CONTROL SYSTEM FUNCTIONAL DESCRIPTION CONTROL/INDICATION - FUNCTION B. Full Core Displav 1. Rod Select Indilcates the selected control rod (white light)

2. Rod DRIFT Indicates a control rod is changing position with no command signal or the ROD DRIFT TEST PB pressed while a rod is being moved (amber). ) 3. Rod Position Indicates FULL-OUT (red) and FULL-IN (green) position of control rods.

4. ACCUM Trouble 5. SCRAM Indicates either low N2 press (940 +85 -0 PSIG) or high H20 level (>37cc) in that control rod's HCIi accumulator (orange light). Indicates each individual rod's inlet and outlet scram valves are both open (blue light).

Hope Creek Page 44 of 45 Rev. 24

, I ES-40 1 Sam pie Written Exam inat ion Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 2 KIA # 245000 K5.07 importance Rating

2.6 Knowledge

of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR AND AUXILIARIES SYTEMS: Generator operations and limitations Proposed Question:

Common 31 Page 63 of 205 I I ES-401 Sample Written Examination Form ES-401-5 Question ~. Worksheet GENERATOR REACTIVE CAPABILITY CURVE ATE 4 POLE 1373100 KVA 1800 RPM 25000 VOLTS 0.94 PF 0 50 SCR 75 PSIG HYDROGEN PRESSURE 530 VOLTS EXCITATION CURVE AB LIMITED BY FIELD HEATING CURVE BC LIMITED BY ARMATUR.E HEATING CURVE CD LIMITED BY ARMATURE CORE END HEATING GENERATOR REACTIVE CURRENT CAPABILITY CURVE VDrive\Drawings\l 000L4vl561 I.vsd Rev 9 Page 64 of 205 1 I ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet Given the attached Generator Capability Curve and the following information:

0 Hydrogen Pressure 65 psig 0 MWatts = 1290 Generator MVARS LAG

= 200 The load dispatcher requests increasing MVARS from 200 to 400. Determine what, if any, curve limitation will be exceeded if the MVARS are increased as requested.

A. NO curve limitation will be exceeded.

B. The curve limitation for Field heating will be exceeded.

C. The curve limitation for Armature heating will be exceeded.

D. The curve limitation for Armature Core End heating will be exceeded. Proposed Answer:

C. Explanation (Optional):

Per the curve (see markup), the region for Armature Heating would be exceeded (BC region) Technical Reference(s) Gen Capability curve (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

MNGENOEOOS (As available)

-. Question Source: Bank

1. Modified Bank # New X (Note changes or attach parent)

Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 65 of 205

__-__ ~1- ES-401 ~ "d-r-=)T Sample Written Exariination Form ES-401-5 Question Workst ieet _-__ ,/' ( %'\ ,I GENERATOR REACTIVE CAPABILITY CURVE ATE 4 POLE 137310L KVA IYOO RPM 25000 VOLTS 0.94 PF 0 50 SCR 75 PSIG HYDROGEN PRESSURE 530 VOLTS EXCITATION cn U > 0 z 0 0 0 d a zf 1600 1400 1200 1000 600 400 a -1 a 20 0 0 -200 u: Minimum Limit -1000 SlZIz CURVE AB LIMITED BY FIELD HEATING CURVE BC LIMITED BY ARMATURE HEATING CURVE CD LIMITED BY ARMATURE CORE END HEATING GENERATOR REACTIVE CURRENT CAPABILITY CURVE VDnve\Drawings\l 000\Avl561 I.vsd Rev. 9 Page 65 of 200 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination 0 ut I i ne Cross-reference

Level Tier # Group fi: KIA # Importance Rating RO SRO 2 204000 K6.08 3.5 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER CLEANUP SYSTEM:

PCIS/NSSSS.

Proposed Question:

Common 32 The plant is operating at full power when an I&C surveillance inadvertently causes a loss of power to the 12OVAC RPS Bus "A. Which one of the following describes the effect of the loss, if any, on the Reactor Water Cleanup system? A. NO effect. B. ONLY RWCU PMP SUCT CONT INBD ISOLATION VALVE BG-HV-F001 will close.

C. ONLY RWCU PMP SUCT CONT OUTBD ISOLATION VALVE BG-HV-F004 will close.

D. RWCU PMP SUCT CONT INBD ISOLATION VALVE BG-HV-F001 will close. RWCU PMP SUCT CONT OUTBD ISOLATION VALVE BG-HV-F004 will close. Proposed Answer:

B Explanation (Optional): IAW RWCU Lesson Plan NOH04RWCUOOC Section lll.B.3.b.5) - A loss of power to a channel of the (NS4) Leak Detection System will cause the respective valve to isolate. The power supplies are 120 VAC RPS Bus A for channel A (NS4) (shuts Fool) and 120 VAC RPS Bus B for channel D (NS4) (shuts F004).

B. Correct A. Incorrect. - The Fool will shut.

C. D. Incorrect. - The F004 will remain open Incorrect - The F004 will remain open Technical Reference(s) RWCU Lesson Plan (Attach if not previously provided)

NOH04RWCUOOC-01 Proposed references to be provided to applicants during examination:

none Page 66 of 205 ES-401 Sample Written Eitamination Form ES-40 1-5 Question Worksheet Learning Objective: Question Source:

RWCUOOEOI 3 (As available)

Bank # Question History: Question Cognitive Level:

10 CFR Part 55 Content: Modified Bank # __. (Note changes or attach parent)

New X -. Last NRC Exam -. Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 X Comments:

Page 67 of 205

&%kON NAME: REACTOR WATER CLEANUP SYSTEM NOH04RWCUOOC 01121108 RWCU Pipe Chase Areas RWCU high differential flow 56.3 gpm after 45 second time delay (LCE via NS4) Redundant Reactivity Control System actuation.

[either "A" or "B" RRCS channel] g) h) The HV-F001 will also auto isolate (in addition to the common) on:

a) b) "A* SLC pump start The HV-F004 will also auto isolate (in addition to the common) on: a) "D" NS4 manual isolation pushbutton armed and depressed b) "B" SLC pump start c) Non-regenerative heal exchanger high outlet (F/D inlet) temperature (14OoF) (NS4) A loss of power to a channel of the (NS4) Leak Detection System will cause the respective valve to isolate. The power supplies are 120 VAC RPS Bus A for channel A (Ids4) (shuts FOOI) and 120 'VAC RPS Bus B for channel D (NS4) (shuts F004). The power monitor relay for each channel has a test pushbutton on Panels HI 1-Pf309 and HI 1-P611 for A and D respectively. Depressing a test pb will simulate a loss of power to that channel and cause the respective RWCU valve to isolate. Both valves have status indicating lights on the NS4/PCIS status board at 1 OC650D (OPENKLOSED)

In order to ensure long term compliance with NRC GL 89- 10, a DCP was completed to modify the FOOI and F004 valves. The motor size for the valves was raised,. The breakers for the valves were replaced with larger amp breakers.

The magnetic trip for the breakers was set at higher and the overload heaters were also raised.. "A' NS4 manual isolation pushbutton armed and depressed 15 of 62 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008-2009\References\LESSON PLANSReactor Water Cleanup\Master\Lesson Plans\NOH04RWCUOOC-01 Reactor Water Cleanup Systerwdoc

___. ES-401 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 WA # 201001 A1.10 Importance Rating

2.8 Group

1. 2 Ability to predict and/or monitor changes in parameters associated with operating the CONTROL ROD DRIVE HYDRAULIC SYSTEM controls including: CRD cooling water flow. Proposed Question: Common 33 Given the following conditions:

0 Reactor operating at 100% rated power 0 CRD flow control valve in automatic The RO throttles closed on the Pressure Control Valve (BF-HV-F003) for two seconds Which one of the following describes how parameters will stabilize when this CRD system transient is over? A. Differential pressure between the Drive Water Header and the RPV will lower.

Cooling water flow will lower.

6. Differential pressure between the Drive Water Header and the RPV will rise. Cooling water flow will lower.

C. Differential pressure between the Drive Water Header and the RPV will lower.

Cooling water flow will remain the same.

D. Differential pressure between the Drive Water Header and the RPV will rise. Cooling water flow will remain the same.

Proposed Answer: D Explanation (Optional):

IAW CRD Lesson Plan NOH04CRDHYD-04, Section 11.8.6, page 21 - The motor-operated PCV is positioned to maintain approximately 265 psid between drive water header pressure and reactor pressure (senses above core plate pressure).

To RAISE the differential pressure between the drive water header and the RPV, press the DECREASE PB to cause the motor-operated PCV to travel in the closed direction Throttling the drive water pressure control valve closed will increase pressure in the line. Cooling water flow will be unaffected due to a different flowpath from the system pumps(see P&ID M-46-1)

Page 68 of 2!05 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet D. Correct. A. B. C. Incorrect.

D/P will rise. Cooling water flow will be constant Incorrect. Cooling water flow will be constant Incorrect. - D/P will rise Technical Reference(s) P&ID M-46-1 (Attach if not previously provided) CRD Lesson Plan NOH04CRDHYD-04 Proposed references to be provided to applicants during examination:

None Learning Objective:

CRDHYDOEOOG (As available) 56276 -. Question Source: Bank

1. Modified Bank # New (Note changes or attach parent)

-. Question History: Last NRC Exam _. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 69 of 205 LESSON NAME: NOH04CRDHYD-04 CONTROL ROD DRIVE HYDRAULICS - Ill 5/08 ~~ 2) One manually operated PCV in parallel, which allows continued operation if the motor-operated PCV is out of service. C. Instrumentation and Control I) DIP between the drive water header and reactor pressure is continuously monitored by: a) Local indication b) Reactor control panel indication on CRIDS IOBJ. 7 I 2) The motor-operated PCV is positioned to maintain approximately 265 psid between drive water header pressure and reactor pressure (senses above core plate pressure).

Pressure control is accomplished by the use of four PBs; 3) IFIG. 7 .j ~ ~~ - a) Two PBs (OPENiCLOSE) will cause the motor- operated PCV to travel either full open or full closed, depending on which PB is depressed.

Two PHs (INCREASEiDECREASE) will cause the motor-operated PCV to throttle open or closed for as long as the PB is depressed. To RAISE the differential pressure between the drive water header and the RPV, press the DECREASE PB to cause the motor- operated PCV to travel in the closed direction.

To LOWER the differential pressure between the drive water header and the RPV, press the INCREASE PB to cause the valve to travel in the open direction.

b) 4) The OPEN PEI is backlit to indicate the motor- operated PCV (HV-F003) is not full closed and the CLOSED PB is "backlit" to indicate the valve is not full open. d. Automatic Actuations and Interlocks - None - Page 21 of 67

, I ES-401 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 2 KIA # 239001 A2.12 Importance Rating

4.2 Ability

to (a) predict the impacts of the following on the MAIN AND REHEAT STEAM SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

PClSlNSSSS actuation. Proposed Question: Common 34 The plant is in OPCON 4 with the following conditions:

0 0 0 Main Turbine is tripped 0 All MSlVs are open Mode Switch is in SHUTDOWN All RPS Channels are reset Main Condenser Pressure is 30" HgA Then, NSSSS channels A and B inadvertently actuate. The cause of the isolation signal has been cleared. Which one of the following describes the response of the MSlVs when NSSSS actuated and what actions must now be taken to reopen an MSlV that had closed IAW HC.OP-SO.SM-0001?

A. ONLY the Inboard MSlVs CLOSE.

The MSlV control switches must be placed in the CLOSED position then the NSSSS LOGIC RESET PBs must be depressed.

8. ONLY the Inboard MSlVs CLOSE, The NSSSS LOGIC RESET PBs must be depressed then the MSlV control switches must be placed in the CLOSED position.

C. All MSlVs CLOSE.

The MSlV control switches must be placed in the CLOSED position then the NSSSS LOGIC RESET PBs must be depressed.

D. All MSlVs CLOSE. The NSSSS LOGIC RESET PBs must be depressed then the MSlV control switches must be placed in the CLOSED position. Proposed Answer: C Explanation (Optional): C.

Correct IAW NSSSS Lesson Plan NOH04NSSSSOC All MSlV control switches must be in Page 70 of 205 I I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet "CLOSE" to enable reset of the isolation. This prevents automatic reopening of the MSlVs following isolation reset.

The initiating condition must have cleared to reset the isolation logic. NSSSS Channels A or C and 6 or D must trip to clctse the MSIV. A. Incorrect. All MSlVs close 6. A. Incorrect.

All MSlVs close. The MSlV control switches must be placed in CLOSE first Incorrect.

The MSlV control switches must be placed in CLOSE first Technical Reference(s) NOH04NSSSSOC-02 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: MSSTEAME012 (As available)

-. Question Source: Bank

1. Modified Bank

1. __. (Note changes or attach parent)

New X Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 71 of 205 I 1 I LESSON NAME: NOH04NSSSSOC-02 NUCLEAR STEAM SUPPLY SHUTOFF SYSTEM 1/4/2007 3. Main steam isolation valve logic I Figure 6 a. All four NSSSS channels provide input to the MSlV trip logic.

Each trip channel is single coincidence, requiring only one of the initiating conditions to be present to de-energize its respective initiation relay.

Each MSlV has two normally energized , solenoid-actuated pilot valves controlling air to the valve operator.

Channels A and C control the position of two series arranged contacts in the power supply to one of the two pilot valves. Channels B and D control the position of similar contacts to the other. The contacts are normally closed, maintaining the solenoids energized and allowing normal operation of the MSIVs. An MSlV logic trip will de-energize the associated initiation relay by opening its, associated contact and interrupting power to the solenoid. Tripping of either channel associated with that solenoid will interrupt power. 1) 2) When the solenoid is de-energized, the pilot valve will reposition, blocking the air supply to the valve operator.

Due to the air supply piping arrangement, bofh solenoids must de- energize to block the air supply and vent the operator of a MSIV. Both inboard and outboard MSlVs operate in the same manner. This logic arrangement is referred to as "one-out-of-two-twice". 1) 2) 3) b. c. d. A E C for one solenoid B E D for the other solenoid

e. f. One of the two channels controlling the power to each solenoid must trip (one-out-of-two). Both solenoid operated pilot valves must reposition to effect MSlV closure (twice).

Channels A or C and B or D must trip to close the MSIV. g. Manual isolation All MSlVs (both inboard and outboard) will close by depression of A or C and B or D manual initiation push buttons. Page 15 of 52 C \Documents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\References\LESSON PLANS\Nuclear Steam Supply Chlltnff Cwctnm fNCCCC\\hIIactnr\l eccnn Planc\NnUndNlSCCCnr-n~

IUI Irl FAR CTFALII CI lPPl V CUI ITnFF CVCTFhA LESSON NAME: NOHO4NSSSSOC-02 NUCLEAR STEAM SUPPLY SHUTOFF SYSTEM 1 /4/200 7 - 1 Objecfive 5c C. Reset I) 2) 3) Accomplished via RESET pushbutton on Control Room Panel 1 OC651 C. Each "tripped" channel logic RESET must be pressed Indication of the main steam line drain valve isolation logic status is available on Control Room Panel 10C609 (inboard) and 10C611 (outboard). d. Interlocks

1) Isolation function, once initiated, is sealed-in ensuring isolation goes to completion, even if the initiating condition clears. To effect a manual isolation, the stop collar on the associated pushbutton must be rotated clockwise and the pushbutton must be pressed. This interlock ensures no single operator action can effect an isolation.

2) .K Objective 5c Figure 7b 3) All MSlV control switches must be in "CLOSE" to enable reset of the isolation. This prevents automatic reopening of the MSlVs following isolation reset.

The initiating condition must have cleared to reset the isolation logic. Main condenser low vacuum bypass

4) PN1-B21-1090-0062 SHT 11, 13; PNl-C71-1020-0006 SHT 8, 10 Figure 7a a) Four keylock NORM-BYPASS maintained co n ta cf switches 0 One for each channel, B21 H-SZ5A - D 1 Objective 5c e e Key removable in NORM 0 t9 and C channels 1OC609 E3 and D channels 1 OC611 0 Page 31 of 52 C:\Documents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\References\LESSON PLANSNuclear Steam Supply chlItn(C cIlrtnm /t,icccc\\r,~~rtn.u neefin Dinnr\hinunnhiccccnr n3 ~II IPI CAD c-rCAfin CI IDDI v CUI i-rncr cvc-rcki ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 2 KIA # 215002 A3.01 I m porta rice Rating -____ 3.1 Ability to monitor automatic operations of the ROD BLOCK MONITOR !SYSTEM including: Four rod display: BWR-3,4,5. Proposed Question: Common 35 Given the following conditions:

0 0 NO LPRMs are bypassed 0 The plant is operating at 100% power Control Rod 30-31 is selected and is at notch position IO. Rod Block Monitor Channel B is then bypassed with the joystick.

Which of the following describes the effect on the FOUR ROD DISPLAY indication?

A. B. C. D. NO LPRM BYPASSED lights are illuminated.

All LPRM BYPASSED lights are illuminated.

ONLY the B and D level LPRM BYPASSED lights are illuminated.

ONLY the B level LPRM BYPASSED lights are illuminated.

Proposed Answer: C Explanation (Optional):

C - Correct IAW RBM Lesson plan NOH04RBMSYS-00, Page 29,30 - Section IV.A.2.b. - The amber LPRM level A/B/C/D BYPASSED lights (on the Four Rod Display) illuminate when any of the following conditions exist: The RBM BYPASS switch is in either the CH. A CH. B position (only the BYPASSED lights for the LPRM detectors associated with the RBM channel will illuminate. RBM channel A: only the A channel B: only the B A. 9. D. C level LPRM detector BYPASSED lights will illuminate.

RBM D level LPRM detector BYPASSED lights will illuminate)

Incorrect.

The B and D lights will illuminate Incorrect.

ONLY the 6 and D lights will illuminate Incorrect.

The D lights will also illuminate Technical Reference(s) NOH04RBMSYS-00 (Attach if not previously provided)

Page 72 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

none Learning Objective:

LPRMOOE002 (As available) Question Source:

Question History: Question Cognitive Level:

10 CFR Part 55 Content: Bank # -. Modified Bank # _. (Note changes or attach parent)

New X - Last NRC Exam -. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Comments:

Page 73 of 205

/ P 4 LESSON NAME: ROD BLOCK MONITOR IRBM) SYSTENL, 'NOH04RBMSYSC4i3, ,0714 7/05 d. e. f. 9. h. RBM channel output trips should be bypassed if average core power (as sensed by the reference APRM channel) is less than 30% of rated. BYPASS light The white BYPASS light illuminates when any of the three RBM channel BYPASS conditions exist HIGH light The white HIGH light illuminates when the "HIGH" UPSC Trip Unit trip setpoint (0.66(W-AW)

+ @%) is being enforced. INTMD light The white INTMD light illuminates when the INTERMEDIATE UPSC Trip Unit trip setpoint (0.66(W-AW)

+s%) is being enforced.

LOW light The white LOW light illuminates when the LOW UPSC Trip Unit trip setpoint (0.66(W-AW)

+ 9%) 1s being enforced. SETUP ALARM SETPOINT light The white SETUP ALARM SETPOINT light illuminates when conditions permit the UPSC Trip Unit trip setpoint to be "SETUP" to the next higher value (INTERMEDIATE E HIGH). 2. Control Room Panel 10C650C OBJ. 2 OBJ. 3a, b Fig. 8 and 9 a. IRM/APRM/RBM recorders (2) RBM channel Averaging Circuit output (0 to 125%) will be recorded on the IRM/APRM/RBM recorders (RED pen) when RBM A and/or RBM B is chosen on the RECORDER SELECT bezel. LPRM Level A/B/C/D BYPASSED lights (16) The amber LPRM level A/B/C/D BYPASSED lights (on the Four Rod Display) illuminate when any of the following canditions exist:

1) A peripheral control rod has been selected for movement (all BYPASSED lights will illuminate), E 2) An LPRM detector(s) has been bypassed a control rod adjacent to that LPRM detector has been selected for b. Page 29 of 45- C:\Docurnenls and Setttnqs\sdennis\My Docurnenls!li~~Creek 2008-2009',Refti~eiices1,LES!,ON PLANS\Rod Block Monitorcnp

{RBM)\Master\Lesson Plans\NOH04RBMSYSC-OO Roo Bioch Monitonnq (RBM] Svstemdoc

__ -- - Deleted ROD BLOCK 4ONITOR (RBM) SYSTEM Formatted:

Font 11 pt I Deleted: ROD BLOCK AONITOR I (RBM) SYSTEM ___ - Formatted:

Font 12 Dt __ - _==- _--_=_ _____ __- ____- ' Deleted: NOH04RBMS SGO 1 1 Formatted:

Font 12 3t _____- - I Deleted: 07/17/06 ~ L - _____-I , Deleted: 24 -___ -~ Deleted: 45 Deleted: 48 Deleted: S \VDnve\TFWI IING DOCUMENTS\Operattons Training\Hope Creek\Plan Technology\Systerns\Rud Nock Monitonng (RBM)\Master\

?sson Plans\NOH04RBMSYSC-r 0 Rod Block Monitonnq (REM)

S stem doc _____-.- ~ -~ ________.--

__- -___ ~~ Deleted: S \UDnve\TRAll ING DOCUMENTS - WORKIN FOLDERS\Operattons\HoI e Creek\Plant Technology'S stem\Rod Block Monitonng (RBM)\NOH04RBMSYSz-0 Rod Block Monitonng (RBM)

S item doc _______ ____

-~ . -_ 1 - Deleted: ROD BLOCK IONITOR LESSON NAME: ROD BLOCK, MONITOR (RBM! SYSTEM!! (RBM) SYSTEM - - - - ~- JIOHO4R5MSYSC-OGi-

,0711 7/06, 1 Deleted: ROD BLOCK ONITOR (RBM) SYSTEM movement (only the BYPASSED light for the associated LPRM detector(s) will illuminate), &r Average core power IS less than 30% of rated, as sensed by the reference APRM channel (all BYPASSED lights will illuminate), The REM BYPASS switch is in either the CH. A &r CH. B position (only the BYPASSE:D lights for the LPRM detedurs associated with the RBM channel will illuminate. RBM channel A: onJ the A am C level LPRM detector BYPASSED lights will illuminate. RBM channel B: on_ly the B and D level LPRM detector BYPASSED lights will illuminate), &r An LPRM detector output signal is downscale (onJ the BYPASSED light for the associated LPRM detector will illuminate), A 3-string or 2-string control rod has been selected for movement (depending on the selected control rods location in the core, the BYPASSED lights for all absent LPRM detectors will illuminate).

c. LPRM meters (16) 1) The LPRM meters provide indication (0 to 125%) of the local power levels surrounding the control rod selected for movement. These indications are the signals supplied to the RBM channel Averaging Circuit.

When a 3-string or 2-string control rod is selected for movement, the LPRM meters for all absent LPRM detectors will indicate "zero" (fully downscale).

2) 3. Control Room Panel 10'308 Formatted:

Font. 11 iX Formatted:

Font 12 pt Deleted: NOH04RBMS SC-OO Formatted:

Font 12 [it - - __ - - - - __- ~ - -~ ___ I Deleted: 07/17/06 ---__ - _- I - ____ _- I Deleted: 45 _- _____ [LGZ48 Deleted: S \VDnve\TRAlt ING DOCUMENTS\Operations Training\Hope Creek\Plan Technology\Systems\Rod jlock --.-.-A Fig. 7 OBJ. 2 OBJ. 3a, b NOTE: The following indications on Panel 1 OC608 are identical for both RBM channels.

a. BYPASS light The white BYPASS light illuminates when any of the three RBM channel BYPASS conditions exist.

All output trip signals (control rod withdrawal block signals) from the respective RBM channel should be bypassed when this light IS illuminated Page 30 of 4!5- C \Documents and Seltinqs\sdennis\Mv Docms\Hooc' Creek 200U-200YiRel~r~~nces\LESL ON PLANSRod Block Monirorinq jRBM)\Master\Lesson Pla~s\NOH04RBMSYSC-00 Hot1 Block Monilorino (Rl?bJ2 Svstem do5 I I I ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 2 KIA # 241 000 A4.07 Importance Rating

3.5 Ability

to manually operate and/or monitor in the control room: Main stop/throttle valves (operation). Proposed Question:

Common 36 Given the following conditions:

0 The Main Turbine is reset. 0 As part of the Main Turbine Startup, the RO has depressed the -1, I PRE-WARMING], SHELL WARMING m, on the DEHC HMI. Which of the following directly occurs as a result of these actions? (Assume VPL Limiter is at its normal setting of 1 OOc%) A. Intercept Valves OPEN 6. Turbine Stop Valves OPEN C. Turbine Control Valves OPEN D. Intermediate Stop Valves OPEN Proposed Answer: C Explanation (Optional):

C. Correct IAW HC.OP-AC-0001 step 5.2.7, The Control Valves OPEN, all others close or go closed. Shell warming is initiated by depressing the CHEST-SHELL WARMING SHELL pushbutton.

This opens the Turbine Control Valves and the pilot valve to the

1. 2 Turbine Stop Valve. The stop valves remain closed. Intermediate Stop Valves are open on reset and go closed. The Intercept Valves remain closed. A. Incorrect. IVs remain closed B. D. Incorrect. lSVs go closed Incorrect. Turbine Stop Valves remain closed Page 74 of 205 ES-40 1 Sample Written Examination Form ES-40 1 -5 Question Worksheet Technical Reference(s) HC.OP-AC-0001 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: MNTURBE024 (As available) Question Source:

Question History: Question Cognitive Level:

10 CFR Part 55 Content: 62008 -. Bank # Modified Bank # -. (Note changes or attach parent)

New ~. Last NRC Exam _. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Comments:

Page 75 of 205 5.2.7 (continued)

HC.OP-SO.AC-0001 (a) NOTE 1 While chest shell warming, the temperature limits of Att. 2 should be observed.

K. SELECT m. EW- L. SELECT Shell Warming i:a N. ADJUST Valve Position Limiter, VPL Setpoint to 100.0%. 0. OBSERVE the following:

AU Cantrat Vatves open fully. 0 Alt tntzrrmediate Stop Valves (tw go closed. All Intercept Valves (IV) remain closed. 0 All Main Stop Valves (MSV) remain closed.

P. VERIFY proper valve posi,tions utilizing Attachment

7. Q. DETERMINE required time period for pre-warming using Attachment

5. (continued on next page)

Hope Creek Page 25 of 56 Rev. 60 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 2 WA ## 223001 G2.2.39 Importance Rating 3.4 -___ Equipment Control: Knowledge of less than or equal to one hour Technical Specifications for a system (Primary Cont) Proposed Question:

Common 37 The plant is operating at 75% power and you note the following readings while taking logs at the start of your shift: 0 0 0 Drywell average temperature is 137 degrees F. Drywell Pressure is 1.2 psig. Suppression Pool water level is 74 inches. Which of the following must be restored to within Temchnical Specifications limits within ONE hour to preclude further actions? A. Suppression Pool Level ONLY. B. Suppression Pool Level AND Drywell Pressure ONLY. C. Drywell Average Temperature AND Suppression Pool Level ONLY. D. Drywell Average Temperature AND Suppression Pool Level AND Drywell Pressure.

Proposed Answer: A Explanation (Optional):

A. Correct IAW TS 3.6.2 - The suppression chamber shall be OPERABLE with: With an indicated water level between 74.5" and 78.5". Per Action a)

With the suppression chamber water level outside the above limits, restore the water level to within the limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. B. Incorrect. - Action is required within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if DW pressure exceeds 1.5 psig Page 76 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet C. D. Incorrect. - Action is required in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for exceeding Drywell average temperature limits. Incorrect. - Action is required in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for exlceeding Drywell average temperature limits. Technical Reference(s)

TS 3.6.2.1, 3.6.1.7, 3.6.1.6 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

PRI CO N E009 (As available) Question Source:

Bank # Modified Bank # -. (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 77 of 205

@:;7 CONTAINMENT SYSTEMS DRYWELL AVERAGE AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.7 Drywell average air temperatLre sriall not exceed 135°F. APPLICABILITY: OPERATIONAL CONDITIONS 1 I ,> and 3. ACTION : With the drywell average air temperature clreater than 135"F, reduce the average air temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. SURVEILLANCE REQUIREMENTS 4.6.1.7 The drywell average air temperature shall be the volumetric average of the temperatures at the following locations and shall be determined to be within the limit at least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: Elevation Zone ~ Approximate Azimuth*

a. 8 6 11 "-1 12 ' 8 I' (under vessel) b . 8 6 11 "-1 11 10 I' (outside of pedestal) 90", 225", 90", 270" 135", 300", 100", 190" C. 111'10"-139'2" 55", 240", 155", 315" d. 139'2"-168'0" e . 1 6 8 ' 0 'I- 1 9 2 ' 7 I' 45", 215", O", 90", 180", 270" 95", 130", 300", 355", 45", 225"

• At least one reading from each elevation zone is required for a volumetric average calculation.

HOPE CREEK 314 6-Z.0 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION SYSTEMS SUPPRESSION CHAMBER LIMITING CONDITION FOR OPERATION 3.6.2.1 The suppression chamber shall be OPERABLE with: a. The pool water: 1. With ;in indicated water level between 74.5" and 78.5" and a 2. Maximum average temperal..ure of 95°F during OPERATIONAL CONDI'I'ION 1 or 2, except that the maximum average temperature may be permitted to incr-ease to: a) 105°F during testr.ng which adds heat. to the suppression chamber. b) 110°F with THERMAI, POWER less than or equal to 1% of RATED THERMAL P0WI:P.. 3. Maximum average te1perat:ure of 95°F during OPERATIONAL CONDITION 3 , except that: the maximum average temperature may be permitted to increase to 120°F with the main steam line isolat.ion valves closed following a scram. b. A total leakage between the suppression chamber and drywell of less than the equivalent leakage through a 1-inch diameter orifice at a differential pressure of 0.80 psig.

APPLICABILITY:

OPERATIOICAL CONDITIONS 1, 2' and 3 ACTION: a. With the suppression chamber water level outside the above limits, restore the water level to within the limfts within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. b. With the suppression chamber average water temperature greater than 95°F and THERMAL POWER greater than 1% of RATED THERMAI, POWER, restore the average temperature to less than or equal to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, except, as permitted above: 1. With the suppression chamber average water temperature greater than 105°F during testing which adds heat to thz suppression chamber, stop all testing which adds heat t2 the suppression chamber and restore the average temperature to less than 95" within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 2. With the suppression chamber average water temperature greater than llO"F, placl? the reactor mode switch in the Shutdown position arid operate at least one residual hea-t removal loop in the supp.ression pool cooling mode. HOPE CREEK 3/4 6-12 Amendment No. 110 LIMITING CONDITION FOR OPERATION 3.6.5.1 SECONDARY CONTAINMENT 1NTE:GRITY shall be maintained.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 21, 3 and *. ACTION: Without SECONDARY CONTAl NMENT 1NTE:GRJ'I'Y

a. In OPEFWTIONAJd CONDITION 1, 2 OL 3, restore SECONDARY CONTAINMENT INTEGRITY wittiin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. b. in Operational Condition

• , suspend handling of recently irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

I SURVEILLANCE REQUIREMENTS 4.6.5.1 SECONDARY CONTAlNMENT INTEGRITY sh3ll be demonstrated by: a. Verifying at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the reactor building is b. Verifying at least once per 31 days that:

at a negative pressure.

1. All secondary containment equipment hatches and blowout panels are closed and sealed. 2. a. For double door arrang:?ments, at least one door in each access to the secondarir containment is closed. b. For single door arrangl?ments, the door in each access to the secondary containm1:nt is closed except for routine entry and exit. 3. All secondary containment penetrations not capable of being closed by OPERABLE secondary containment automatic isolation dampers/valves and reqJired to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic dampers/valves secured in position.

c. At least once per 18 months: 1. Verifying that four filtrat'on recirculation and ventilation system (FRVS) recirculation units and one ventilation unit of the filtration recirculation and ventilation system will draw down the secondary containment to greater than or equal to 0.25 inches of vacuum water gauge in less than or equal to 375 seconds , and *When recently irradiated fuel is being handled in the secondary containment and during operations with a potential for draining the reactor vessel.

I I HOPE CREEK 3/4 6-47 Amendment No. 146 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level Tier # Group ## KIA # Importance Rating RO SRO 2 2 259001 A1.06

2.7 -- Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including: Feedwater heater level.

Proposed Question:

Common 38 The plant is at full power when the operators notice an increasing shell level in the 2A Feedwater Heater (FWH). An operator is sent out to the local panel to operate FWH 2A drain valve manually when the Hi- Hi shell level setpoint is reached for the 2A FWH. Which one of the following describes how the 2A FWH is affected?

A. Extraction steam to the heater is isolated AND Condensate flow through the heater tube side is isolated.

B. Condensate flow through the heater tube side is isolated.

NO other automatic actions occur. C. Extraction steam to the heater is isolated AND the cascading drain flow from the 3A FWH is isolated.

D. Condensate flow through the heater tube side is isolated AND the cascading drain flow from the 3A FWH is isolated. Proposed Answer: D Explanation (Optional):

D. Correct IAW FWH Lesson Plan NOH04FWHEATC-01 Sectian lll.C.2.b)-

Page 32, If the "Hi-Hi" setpoint is reached, the following actions will occur: [For the '1,2 (A, B, C) only, the isolation will occur after a ten second time delay once the Hi-Hi level is reached]

0 FWHTRs 1,2 (A, B, C) - Condensate flow through the heater tube side is isolated (this reduces the extraction flow to that heater)

AND Cascading drains from heater 3 (A, B, C) are isolated A. B. D. Incorrect - no extraction steam isolation Incorrect - Also, cascading drain flow from the 3A FWH is isolated.

Incorrect - no extraction steam isolation Page 78 of 205 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference( s) N 0 H 04- FW H EATC-0 1 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: F W H EAT E 0 0 8 (As available) Question Source:

Bank # -. Modified Bank # -. (Note changes or attach parent)

New X -. Question History: Last NRC Exam Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 79 of 205 NAME: FEED WATER HEATER EXTRACTION,, VENT AND DRAIN SYSTEW +NOH84FWHEATC-O1--

.'I 01'23107, The level control valve is positioned as a function of heater 3 (A, B, C) level. (5) Extraction steam to FWHTR 2! (A, B, C) is also condensed by secondary condensate pump flow before being mixed with the incoming drain flow. The condensate flows directly to the external drain cooler where it is subcooled by secondary condensate pump flow through the drain cooler.

The drain flow is then directed to the main condenser via a motor operated isolation valve (HV-1469 A, B, C) and a level control valve (LV-1464 A, B, C). of FWHTR 2 level. normal operation. (a) The level control valve is positioned as a function (b) The drain cooler is maintained full of water during (6) Secondary condensate pump flow through FWHTR 1 (A, B, C) condenses its extraction steam which is then drained to the main condenser via a loop seal.

1 Obj. 4d I b) Abnormal (1) An increasing heater shell level (in FWHTR's 2-6 A, B, C) will open the associated alternate drain (dump) valve to restore water level. (a) The dump valve is modulated as a function of the (b) The drains are directed to the main condenser (2) If heater level reaches the "Hi" setpoint, the alternate drain valve will fail open.

(3) If the "Hi-Hi" setpoint is reached, the following actions will occur: [For the 1,2 (A, 6, C) only, the isolation will occur after a ten second time dekqmrrce t'he-Wldk-reached] high level in its respective heater.

shell. (a) FWHTR's 3-6 (A, B, C) (i) Extraction steam to that heater is isolated. (ii) Cascading drain flow from the upstream heater is isolated (level control valve fails closed). Page 32 of C \Documents and Settinqs\sdennis\My Documents\HopaCleek 2008 2009\Refererices\LESSON PLANSFeeUwater Healer I- Extraction, Vent and Drain\Masler\Lesson Plans\NOH04FWHEATC-O3~

__ -- - ~ Formatted:

Font 12 pt Deleted: FEED WATER t EATER ' EXTRACTION, VENT ANI DRAIN

' Deleted: FEED WATER t EATER I EXTRACTION, VENT ANI DRAIN 1 SYSTEM _-~ - _-_ __ - - ___ - - ~ , j i -1 EXTRACTION, VENT ANC DRAIN 1 ~- SYSTEM - __ I ____- -- -_ - - -~ -- ____-~ - 1 Deleted: FEED WATER t EATER 1 SYSTEM -_ -- __ c__~~ __._- Deleted: 10/23/07 Formatted:

Font 12 pt I Deleted: NOH04FWHEAl

-01 1 Formatted:

Font 12 pt _________ - -~ __-_ _ ~ - ---=_ - -- _____-___- - _ - -__ Deleted: 53 Deleted S \VDrive\TRAINIf 3 DOCUMENTS\Operations Training\Hope Creek\Plant Technology\Systerns\Feedw ter Heater Extraction, Vent and Drain\Master\Lesson Plans\NOH04FWHEATC-O1 bOC Deleted: S \VDrive\UDnve\l iAlNlN G DOCUMENTS - WORKINC FOLDERS\Operations\Hope Creek\Plant Technology\Systern\Feedwat r Heater Extraction, Vent and Drain\Lesson Plans\NOHO4FWHEATC-O1 OC -- _____ __

I 1 I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # Group # 1 I m porta nce Rating 29501 8 AKI .01 3.5 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on componentkystem operations.

Proposed Question: Common 39 Given the following:

0 The plant is operating at 100% reactor power. "A" SACS Loop is supplying TACS with "'A" & "C" SACS Pumps running. 'ID" SACS Pump is running supplying the "B" SACS Loop loads. A trip of the "A" SACS Pump has occurred.

The "C" SACS Pump remains running. The idle "B" SACS Pump has auto started. Which one of the following describes the affects, if any, on the SACSlTACS system? (assume NO operator action) SACSRACS Isolation valves, EG-HV-2522A

-( 1 )-.- and EG-HV-2522C

-(2)- A. (1) CLOSES (2) CLOSES B. (1 ) remains OPEN (2) remains OPEN C. (1) remains OPEN (2) CLOSES D. (1) CLOSES (2) remains OPEN Proposed Answer:

D Explanation (Optional):

D. Correct IAW HC.OP-SO.EG-0001 interlocks section D. Correct - "A valve receives a close signal from the "A pump stop input. The "C" valve does not close because the "C" pump is still running.

A. Incorrect - The "C" valve does not close because the "C" pump is still running.

Page 80 of ,205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet B. C. Incorrect - "AI valve receives a close signal from the "A" pump stop input Incorrect - "A valve receives a close signal from the "A pump stop input.

The "C" valve does not close because the "C" pump is still running. Technical Reference(s) HC.OP-SO.EG-0001 in,terlocks (Attach if not previously provided) section Proposed references to be provided to applicants during examination:

none Learning Objective: Question Source:

Question History: Question Cognitive Level:

10 CFR Part 55 Content: STACSOEOI 8 (As available)

Bank # 55907 Modified Bank # -. (Note changes or attach parent)

New Last NRC Exam _. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Comments:

Page 81 of 205 IMMEDIATE OPERATOR ACTIONS NONE AUTOMATIC ACTIONS HC.OP-AB.COOL-QQQZ(( 2) SAFETYlTURBlNE AUXILIARIES COOLING SYSTE Vi I IF Expansion tank Low-Low-Low level (33%) (See NOTE

1) Low SACS to TACS flow from In-Service loop (9,900 gpm) Low Pump d/p (34.3 psid) Pump Trip (Automatic or Manual) TACS Override NOT enabled. AND SACS Supply Temperature High (88OF) Low TACS pressure at supply accumulators (22 psig).

return - THEN 1. TACS supply return valves from affected SACS loop close. (HV-252212496)

2. The affected loop SACS cross-connecting valves for the Fuel Pool Heat Exchanger AND Containment Instrument Gas Compressors close.

I 1. In-service loop SACS pumps transfer to MAN. 2. Standby Loop SACS Pumps Start in AUTO. 3. TACS supply return valves on standby SACS loop open respective pump AND valve control is in AUTO. Respective pump trips (Equates to 17,000 gpm pump flow) Associated TACS supply and return valve shut. (H\J-2522/2496)

Associated Heat Exchanger Bypass Valves trip closed. (HV-2457/2517) TACS HV-2522E/F close -. i NOTES: 1. SACS Expansion Tank Values: Low Level-65.0% Hope Creek Low Low Level- 44.0% Low Low Low Level- 33.0% Page 3 of 43 Rev 3 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 295006 AKI .03 Group # 1 Importance Rating

3.7 Knowledge

of the operational implications of the following concepts as they apply to SCRAM : Reactivity control. Proposed Question:

Common 40 What are the operational implications of RPS having a built in time delay of 10 seconds before allowing a SCRAM to be manually reset? A. To ensure all the control rods fully insert.

B. To allow the Scram Air header to repressurize.

C. To allow reactor water level to recover above the scram setpoint.

D. To ensure the Scram Discharge Volume veni and drain valves are fully closed. Proposed Answer: A Explanation (Optional):

IAW RPS Lesson Plan NOH01 RPSOOC-05, Page 37, Section E.6. - Manual reset of a full scram is inhibited for 10 seconds after initiation to ensure control rods drive to FULL IN position.

A. Correct. B. C. D. Incorrect.

Not the reason cited in Lesson plan Incorrect.

Not the reason cited in Lesson plan Incorrect.

Not the reason cited in Lesson plan Technical Reference(s) NOH01 RPSOOC-05 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

RPSOOOE007 (As available) Question Source:

Bank # 68852 Modified Bank

1. New (Note changes or attach parent)

-. Page 82 of 205 ES-401 Sample Written Examination Form ES-40 1-5 Question Worksheet Question History: Question Cognitive Level:

10 CFR Part 55 Content: Comments:

Last NRC Exam 2002 Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 X Page 83 of 2!05 I I 1 I - bq LESSON NAME: NOH01 RPSOQC-05 REACTOR PROTECTION SYSTEM - 01/10/2008 E. Operating Procedures

1. Review the following procedures

a. HC.OP-S0,SB-0001, RPS Operation

b. HC.OP-SO.SM-0001 , Isolation System Operation

2. For RPS operation, seven separate subsections are used as appropriate.

a. 5.1 System Operability Observation

b. 5.2 RPS Scram c. 5.3 Resetting RPS Trips d. 5.4 Transfer of RPS Bus A (B) Power from RPS MG Set A (B) -0 RPS Alternate Transformer A (B) e. 5.5 Transfer of RPS Bus A (B) Power from RPS Alternate Transformer A (B) to RPS MG set A (B) f. 5.6 Bypassing Reactor Mode Switch Scram

g. 5.7 Resetting the Reactor Protection EPA Breaker Trips h. 5.8 Moving Reactor Mode Switch out of Shutdown in OPCON 3 or 4 3. In sections 5.1, 5.4, and 5.5, local operations including stopping or starting the RPS MG sets. 4. For scram.

confirming "All Rods In" is the most important action using various indicators. Placing the Reactor Mode Switch in the SHUTDOWN position is a backup trip action.

5. Resetting a Full Scram as soon as reasonable is necessary to reduce CRD internal seal damage from excessive flow, and to minimize vesscl thermal stratification if the Recirculation Pumps are out of service. [Resetting the scram will also depressurize the scram discharge volume. The RCPB is extended out to the scram discharge volume after a scram and any leakage past the scram discharge volume vent and drain valves could result in a LOCA outside of containment.]

6- Mmual reset of a full scrarn is inhibited for 10 seconds after initiation t 2 mre controt rods drive to FULL IN posithn. Page 37 of 48 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008-2009\References\LESSON I I I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 295001 AKI .02 Group # 1 importance Rating

3.3 Knowledge

of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Power/flow distribution. Proposed Question: Common 41 Plant conditions are as follows: 0 0 OPRMs are inoperable. Initially the reactor is at 96.9% power with both Recirculation loops in operation.

Following a lightning strike event and various power losses, core flow and power are reduced.

0 0 0 Core flow is at 40% of rated and steady. Reactor power is 40% and slowly rising.

APRM Recorders are currently reading

INSTRUCTIONAL CONTENT:

.--- Obj. I Fig 18-AV6208 Ill. INTRODUCTION/PURPOSE A. The High Pressure Coolant Injection (HPCI) System provides adequate make-up water during a small break loss of coolant accident (LOCA) to ensure adequate core cooling. The HPCl System may be used for reactor vessel inventory or pressure control whenever the vessel is pressurized and isolated from the main condenser.

B. DESIGN BASES Formatted:

Bullets and Numbering The HPCl System shall be able to maintain reactor vessel inventory during a loss' - of coolant accident originating from a 1" or smaller break under the following conditions.

A Reactor pressure - 165 psia to 1 156 psia B Developed head - 71 0 feet to 2950 feet C Flow rate - 5690 gpm within 15 seconds ctf ir?itia:ior!

~- D Independent of AC power DESCRIPTION A. General lM-55-1, M-56-1 1 1. The HPCl System consists of the following major components:

a. HPCl turbine

b. HPCl pump c. HPCl booster pump d. HPCl lubricatingkontrol oil system

e. HPCl turbine governor control system f. Barometric condenser Paae 11 of 144 I I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 295037 AK3.01 Group # 1 I m porta rice Rating

-~ 3.4 Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Recirc Pump TriplRunback Proposed Question: Common 45 A failure to scram has occurred and reactor power i:s 65%. The main turbine is on line. The recirculation pumps are required to be runback to minimum speed before tripping the pumps to.. . A. maintain the largest margin to the MCPR limit. B. prevent an RPV high level trip to ensure HPWCore Spray injection flowpath.

C. prevent power instabilities due to operating at high power without adequate core flow. D. prevent additional heat loading of the torus if power remains above the bypass valve capacity. Proposed Answer:

D Explanation (Optional): EOP-101 A RCIQ-8 bases The most rapid flow rate reduction and, consequently, the most rapid power reduction, is achieved by tripping the recirculation pumps. However, if the recirculation pump trip is initiated from a high power level, the resulting rapid changes in steam flow, RPV pressure, and RPV water level may cause a trip of the main turbine-generator and a trip of RPV injection systems. If the main turbine-generator trips and reactor power exceeds the turbine bypass valve capacity, RPV pressure will increase until one or more SRVs open.

Heatup of the suppression pool then begins and RPV level lowering may be required. If RPV injection systems trip, the resultant RPV water level transient may require emergency depressurization of the RPV and operation of less desirable RPV injection sources. To effect a more controlled reduction in reactor power and thereby avoid main turbine-generator and RPV injection system trips and their associated complications, a recirculation flow runback is performed prior to tripping the recirculation pumps. If an automatic runback has occurred, the operator need only confirm the action.

D. Correct - prevent additional heat loading of the torus if power remains above the bypass valve capacity A. Incorrect - maintain the largest margin to the MCPR limit. Removing RPV flow will rely Page 92 of 205 on natural circulation to prevent approaching the MCPR limit during an ATWS, it will certainly not lessen it Incorrect - In an ATWS condition, HPCl injection through Core Spray flowpath is not desired. Incorrect - prevent power instabilities due to operating at high power without adequate core flow. Actions taken will remove all forced circulation, and lower RPV level to lower power, power takes precedent over instabilities.

B. C. Technical Reference(s) EOP-101A bases (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: Question Source:

E01 01 AE006 (As available)

Bank # 56604 Modified Bank # _. (Note changes or attach parent)

New Question History: Last NRC Exam none Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 93 of 205 I I HC.OP-EO.=-0101 A BASES considered. However, if power is less than 4%, the basis for BllT will be met as long as boron injection commences prior to the BllT temperature associated with 4% power. A conservative value of 140°F has been selected based on supp pool temperature instrument readability.

Since failure-to-scram conditions may present severe plant safety consequences, the requirement to initiate boron injection is independent of any anticipated success of control rod insertion. When attempts to insert control rods satisfactorily achieve reactor shutdown, the requirement for boron injection no longer exists. (Control rod insertion is directed by Step SLC initiation generates a RWCU isolation signal.

This isolation prevents RWCU FDs from removing injected boron. RC/Q-5) RClQ-8 VERIFY recirc runback to minimum Discussion An immediate and rapid reactor power reduction may be effected by reducing reactor coolant recirculation flow rate.

This action may place plant operations in a high power-to-flow condition where, under certain circumstances, neutronidthermal-hydraulic instabilities are possible.

However, the action to reduce recirculation flow rate remains appropriate because: Severe suppression pool heating and potential containment failure result if recirculation flow is not reduced. Neutronic/thermal-hydraulic instabilities can be accommodated without fuel failure in most cases.

• RPV water level control actions which are being performed in Contingency

1. 5 concurrently with the actions of this subsection will prevent or mitigate the occurrence of extremely large neutronic/thermal-hydraulic instabilities which may damage the fuel. Reactor power control actions to inject soluble boron and insert control rods, if successful, will terminate the failure-to-scram condition, thus preventing prolonged exposure to instabilities.

The most rapid flow rate reduction and, consequently, the most rapid power reduction, is ---? achieved by tripping the recirculation pumps. However, if the recirculation pump trip is initiated from a high power level, the resulting rapid changes in steam flow, RPV pressure, and RPV water level may cause a trip of the main turbine-generator and a trip of RPV injection systems. If the main turbine-generator trips and reactor power exceeds the turbine bypass valve capacity, RPV pressure will increase until one or more SRVs open. Heatup of the suppression pool then begins and RPV level lowering may be required.

If RPV injection systems trip, the resultant RPV water level transient may require emergency depressurization of the RPV and operation of less desirable RPV injection sources. To effect a more controlled reduction in reactor power and thereby avoid main turbine- generator and RPV injection system trips and their associated complications, a recirculation flow runback is performed prior to tripping the recirculation pumps. If an automatic runback has occurred, the operator need only confirm the action. 10 Of 53 Rev 3 C \Documents and Settinqs\sdennis\My Documents\Hope Creek 2008-2309\References\I_ESSON PLANS\Ememency Operatinq I Procedures\EOP HopeCreek Bases\Master\iesson Plans\EOP 5 01Abases03 do5 Deleted: S \UDnve\TRAl JING FOLDERS\Operations\Ho

!e Creek\Operating Expenence\Emergency 0 erating Procedures\EOP Bases\EOPl OlAbasesC3 ioc DOCUMENTS -WORK N I I I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 1 WA # 295026 EK3.04 Importance Rating

3.7 Knowledge

of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

SBLC injection. Proposed Question: Common 46 EOP 101 A steps RC/Q-10 and RC/Q -1 1 state: Before Suppression Pool temperature reaches 140 degrees F, initiate SLC and verify RWCU isolates.

What is the reason for these EOP steps? A. It ensures the reactor will be in Cold Shutdown before the Suppression Pool reaches the Heat Capacity Temperature Limit. B. It ensures the reactor will be in Hot Shutdown before the Suppression Pool reaches the Heat Capacity Temperature Limit. C. It ensures the reactor will be in Cold Shutdown before the Suppression Pool reaches the Boron Injection Initiation Temperature.

D. It ensures the reactor will be in Hot Shutdown before the Suppression Pool reaches the Boron Injection Initiation Temperature. Proposed Answer:

B. Explanation (Optional):

B. Correct IAW EOP-1 OIA, step RC/Q-10 bases A. C. D. Incorrect - Hot S/D is the bases Incorrect - The (BIIT) is not the bases. HCTL in Hot S/D is the reason. Incorrect - The (BIIT) is not the bases. Technical Reference(s)

EOP-IOIA bases page 11 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

E0101AE004 (As available)

Page 94 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Question Source:

53439 Bank # _. Modified Bank # _. (Note changes or attach parent)

New -. Question History: Question Cognitive Level:

10 CFR Part 55 Content: Comments: Last NRC Exam -. Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 X Page 95 of 205

, I HC.OP-EO.ZZ-O101 A Dyl.7 BASES RCIQ-9 Trip the recirc pumps. Discussion If reactor power remains above the APRM downscale setpoint, Ihe recirculation pumps are tripped to effect a prompt reduction in power. While tripping the pumps may place the plant in a high power-to-flow condition and thereby contribute to neutronic and thermal-hydraulic instabilities, continued recirculation pump operation may not be desirable or even possible:

If RPV water level is lowered in accordance with the LP leg, the pumps will trip automatically when the low RPV water level trip setpoint is reached. Allowing reactor power to remain high would increase the flow demand on RPV injection systems and the heat load on the primary containment.

Tripping the recirculation pumps may also reduce the boron mixing efficiency if boron injection is required. However, three-dimensional scale model tests have demonstrated that natural circulation provides adequate mixing as long as RPV water level is above the elevation at which a natural circulation flowpath can be established.

If reactor power is below the APRM downscale trip setpoint, tripping the recirculation pumps results in little, if any, reduction in reactor power since power is already near the decay heat level. In this case, forced recirculation flow is continued, if possible, to enhance boron mixing if boron injection is later required. RCIQ-10 and I1 Before suppression pool temperature reaches ISO'F, INITIATE SLC AND VERIFY RWCU isolates Discussion The Boron Injection Initiation Temperature (BIIT) is the greater of:

The highest suppression pool temperature at which initiation of boron injection will permit injection of the Hot Shutdown Boron Weight of boron before suppression pool temperature exceeds the Heat Capacity Temperature Limit. - The suppression pool temperature at which a reactor scram is required by plant Technical Specifications.

The BllT is a function of reactor power. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the Hot Shutdown Boron Weight of boron cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Refer to HC.OP-EO.ZZ-LIMITS-CONV, EOP Limit Curves and Cautions, for a detailed discussion of the BIIT. Reactor power above 4% results in immediate SLC injection.

At 4% or less, there is no major benefit to injecting SLC if control rods can be inserted as long as supp pool temperature remains below the Boron Injection Initiation Temperature (BIIT). The BllT for power levels at or below 4% is approximately 144°F (See Appendix C, WS-1). The value has been conservatively rounded to 140°F. SLC initiation generates a RWCU isolation signal.

This isolation prevents boron from plating out on the RWCU piping, and eliminates the RWCU water volume not considered in the SLC injection total RPV volume. Hope Creek 11 of 53 Rev 3 C \Documents and Settmqs\sdennis\My Documents\HoDe Creek 2008-2009\Reference~\LESSON PLANSEmerqency Operatinq Procedures\EOP Bases\Master\Lesson Plans\EO~lO1AbasesO3 doG I - ____ Deleted: S \UDnve\TRP YlNG DOCUMENTS - WORKlt G FOLDERS\Operations\H pe Creek\Operating Expenence\Ernergency ( oerating Procedures\EOP Bases\EOPlOlAbasesO.

doc ~___~- _____

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Out I ine Cross-reference:

Level RO SRO Tier # 1 WA # 295025 EK3.05 Importance Rating

3.6 Group

1. 1 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE:

RClC operation:

Plant- Specific. Proposed Question: Common 47 Ten minutes after a scram, an MSlV Isolation occuris and RClC is in pressure control augmented by SRVs. RClC speed is then observed to oscillate with the flow controller in AUTO. Which of the following explains the RClC speed oscAlations? (assume NO other operator actions) A. B. C. D. Swings in RPV pressure are occurring due to the methods being used for pressure control. This requires the RClC speed to change as the controller maintains a constant flow. In pressure control, the RClC controller attempts to maintain a constant speed, but CANNOT respond fast enough to maintain speed as the RPV pressure changes.

Using RClC for pressure control is inherently less stable than using it for level control due to the lower pressure in the CST compared to the RPV. The greater instability is seen as an increase in oscillations. The comparatively small CST volume results in the RClC suction and discharge points in the CST being close together, and at high flow the turbulence causes oscillations. Proposed Answer:

A. Explanation (Optional):

A. B. C. D. Correct -As the SRVs cycle the reactor pressure will change, this changes the head the RClC pump must operate against to maintain the constant flow the speed must change.

Incorrect - In auto the controller maintains flow and in manual it maintains speed. The candidate may reverse the methods of RClC control. Incorrect - The controller is equally stable in the pressure and level control modes. The candidate may believe that the operation of RClC in other than its design function of injecting to the core is less stable Incorrect - The CST has a relatively small volume compared to the SP but it does not result in oscillations. The operator may accept that the smaller flow volume results in Page 96 of 205 ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet suction/discharge interaction Technical Reference(s)

LP NOH04RCIC00-05 FClC (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective:

RCICOOE012 (As available) Question Source: Bank

1. -. Modified Bank # -. (Note changes or attach parent)

New X Question History: Last NRC Exam -. Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 97 of 205 I / /Jw- ' q?.LESSON PLAN: NOH04RCIC00-05 REACTOR CORE ISOLATION COOLING - 10/29/07 Obj. 2, 23a, 23d Fig. 7 & 6 k. Turbine Governor Valve (FV-4283)

I) 2) The purpose of this valve is to control steam flow to the RClC turbine. It is a 3" poppet valve which is hydraulically throttled closed using RClC turbine lube oil as the hydraulic fluid. With no hydraulic pressure, the valve is maintained open by spring pressure. Bailey Flow Indicating Controller (FIC-RGOO).

a) 3) 4) MAN(UAL) - when depressed, the operator controis turbine speed via the (up arrow) (down arrow) pushbuttons.

AUTO - when depressed, the operator controls RClC pump flow via the RAISE SETPOINT, LOWER SETPOINT pushbuttons. Actual flow, which is determined by RClC turbine speed, is adjusted to match demanded flow. See Turbine Control System. b) Obj. 2, 3a Fig. I, 9, 10 2. RCIC Turbine and Governor Control System a. RClC Turbine 1) Purpose - Coriverts heat energy of reactor vessel steam into mechanical rotational energy of the RClC turbine shaft in order to provide the motive force to drive the RClC pump.

2) Characteristics a) b) Manufacturer - Terry Steam Turbine Company Variable speed, single wheel, impulse turbine - Selected because of its suitability for rapid start service and the ruggedness of the solid wheel const ru ction .

I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 WA # 295031 EK3.05 Group # 1 Importance Rating

4.2 Knowledge

of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL: Emergency depressurization Proposed Question: Common 48 Actions for Steam Cooling are being performed in accordance with HC.OP-EO.ZZ-0101, RPV Control, and RPV level has dropped to -200" on Fuel Zone indication. The procedure requires emergency depressurization.

Which of the following is the reason for emergency depressurizing?

A. Maintain peak cladding temperature below 1500 degrees F. B. Maintain peak cladding temperature below 1800 degrees F. C. Maintain total oxidation of the cladding less than 0.1 7 of the total cladding thickness.

D. Maintain the maximum H2 generation less than 0.01 times the hypothetical maximum. Proposed Answer:

B Explanation (Optional):

IAW EOP-1 01 Bases Discussion Steam cooling is effected by allowing RPV water level to decrease through boil-off until it drops to the Minimum Zero-Injection RPV Water Level (MZIRWL). During this period the fuel temperatures in the uncovered portion of the core increase, and heat is transferred from the fuel rods to the steam. The MZIRWL is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1800°F. When RPV water level drops below the MZIRWL, steam cooling may no longer be sufficient to preclude the peak clad temperature from exceeding 1800°F. Emergency RPV depressurization is then performed in accordance with EOP-202. Unless the RPV is already depressurized, it is expected that the resulting swell will be sufficient to quench the uncovered portion of the fuel and reduce PCT almost to the value that would exist if the core were submerged.

As the swell subsides and steam flow through the open SRVs decreases, however, PCT turns and again rises. Opening the SRVs before RPV water level reaches the MZIRWL would reduce the time over which the core remains adequately cooled with no injection. Waiting much after RPV water level reaches the MZIRWL could result in significant core damage due to excessive fuel temperatures. Page 98 of 205 1 I I ES-40 1 Sample Written Examination Form ES-401-5 Question ___. Worksheet B. Correct - Maintain peak cladding temperature below 1800 degrees F. A. Incorrect - Maintain peak cladding temperature!

below 1500 degrees F. - this is the MSCRWL the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F Incorrect - Maintain total oxidation of the cladding less than 0.17 of the total cladding Thickness. - This is an ECCS criteria based on c 2200 degrees F PCT Incorrect - Maintain the maximum H2 generation less than 0.01 times the hypothetical maximum. - This is ECCS criteria based on c 2200 degrees F PCT C. D. Technical Reference(s)

EOP 101 Bases (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective:

E01 01 LE008 (As available) 56'1 26 -. Question Source:

Bank # Modified Bank # -. (Note changes or attach parent)

New -. Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 99 of 205 I I &R %fP 0 N S I B L E FO E V E R 1 FYI Id G RE V IS ION , :,TAT LIZ A id D C HA NG ES RINTED 2008 1209 PSEG Internal Use Only HC.OP-EO.ZZ-0101 XONV EOP STEPIS): sc-2 e JUSTIFICATION The part of the step referring to pressure not stabilized and increasing has been implemented as part of the SC-I override.

EPG STEP: c3-3 When RPV water level drops to [-208 in. (Minimum Zero-Injection RPV Water Level)], EMERGENCY RPV DEPRESSURIZATION IS REQUIRED; enter [procedure developed from Emergency Procedure Guideline Contingency

1. 2]. .$ PSTG STEP: sc-3 When RPV water level drops to -205 in. (Minimum Zero-Injection RPV Water Level), EMERGENCY RPV DEPRESSURIZATION IS REQUIRED; enter HC.OP-EO.ZZ-0202.

JUSTIFICATION 0 0 322.5 IN is the Minimum Zero-Injection RPV Water Level as referenced to the bottom of the RPV (Hope Creek Appendix C Calculation, Work Sheet IO). RPV water level instrumentation is referenced to 527.5 IN (instrument zero, Technical Specifications, Bases Figure B 3/4 3-8). Therefore, the 322.5 IN value has been converted to an indicated value of -204.8 IN which can be read on the RPV Fuel Zone level instrumentation. HC.0P-EO.ZZ-0202, Emergency Depressurization, is the plant specific procedure developed from Contingency

1. 2. 4 0 EOP STEP6): sc-3 & 4 JUSTIFICATION The value of -204.8 IN has been conservatively rounded to -200 IN in order to provide a value that can be readily determined using Control Room instrumentation. -Hope Creek Page 61 of 61 Rev. 7 USER RESPONSIBLE FOR VERIFYING REVISION, STATiJC AND CHANGES 'RINTED ZOOS1209 HC . 0 P-EO.ZZ-PS TG 0 -- PLANT SPECIFIC TECHNICAL GUIDELINE HC.OP-EO.ZZ-O203(Q)

STEAM COOLING sc-1 If while executing the following steps: Emergency RPV Depressurrzation is required, enter OP-EO.ZZ-202.

RPV water level cannot be determined, enter OP-E6.ZZ-206 Any system, injection subsystem, or alternate Injectton subsystem is lined up for injection with at least one. pump running and RPV water level cannot be restored . and maintained above

-186.0 in. (MSCRWL), EMERGENCY RPV DEPRESSURIZATION IS REQUIRED; enter OP-E0.Z-202.

RPV water level IS increasing, enter OP-EO.ZZ-IO1 at Step RCIP-2. One or more SRVs are being used to stabilize RPV pressure and the continuous SRV pneumatic supply is or becomes unavailable, EMERGENCY RPV DEPRESSURIZATION IS REQUIRED; enter OP-EOZZ-202.

SC-2 Stabilize RPV pressure using one or moro of the following systems:

@+ x 0 SRVs, only when suppression pool water level is above 0 in., defeating pneumatic supply isolation interlocks and restoring the pneumatic supply if necessary; open SRVs in a sequence that will uniformly distribute heat in the Suppression Pool. 0 HPCI, defeating interlocks if necessary.

RCIC, defeating interlocks if necessary.

When RPV water level drops to -205 in. (Iflinimum Zero-Injection RPV Water Level), EMERGENCY RPV DEPRESSURIZATION IS REQUIRED; enter OP-E0.Z-202.

-? . I PS'I'G , , 35 of 47 Rev. 8 I I I USER RESPONSIBLE FOR VERIFYING REVISION, STATUS AND CHANGES 'RINJED 2008 1209 0 MINIMUM ZERO-INJECTION RPV WATER LEVEL [Reference EPG/SAG Appendix B section 17.251 H C . OP-EO .ZZ-LI M ITS -CO NV The Minimum Zero-Injection RPV Water Level (MZIRWL) is the lowest RPV water level at whish the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered parf of the core from exceeding 1800°F. This water level is utiiized to preclude significant fuel damage and hydrogen generation for as long as possible.

The derivation of the MZIRWL is shown graphically irr Figure 8-17-19. Curve @ graphs the steam flow required to maintain clad temperature less than 1800°F. Curve @ graphs the actual steam flow generated by the reactor core. At the intersection of Curves @ and 0, the steam flow generated by the partially submerged core equals the steam flow needed for cooling the uncovered part of the core. The corresponding level is the MZIRWL. The MZIRWL is determined assuming:

1, The reactor has been shutdown from rated power For ten minutes. 2. The reactor axial power shape was the most limiting top-peaked power shape prior to reactor shutdown.

3. No water is injected into the RPV. .. Planf-specific data required to calculate the MZIRWL are: 0 1. Minimum active fuel length fraction which must be covered to maintain peak clad temperature below 1800°F without injection.

[70.83 %.] 2. Active fuel length. Il50 in.] 3. Water level at the bottom of active fuel. [-31 I in.] The MZIRWL differs from the Minimum Steam Cooling RPV Water Level (MSCRWL) in two respects:

1, The MZIRWL Is based upon maintaining peak clad temperature below 1800°F. The MSCRWL is based upon maintaining peak clad temperature below 1500°F. 2, Water in the lower plenum is assumed to be saturated in the MZIRWL calculation.

It is assumed to be subcooled in the MSCRWL calculation.

[The Hope Creek MZ/RWL is -205 in. rounded to The MZJRWL is referenced in EPG Step C3-3. :' HopeCreek Page 38 of 56 Rev. 4 _.

ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 29501 9 AA1.03 Importance Rating

3.0 Group

1. 1 Ability to operate andlor monitor the following as they apply to PARTIAL. OR COMPLETE LOSS OF INSTRUMENT AIR:

Instrument air compressor power supplies.

Proposed Question:

Common 49 Given the following conditions:

0 A loss of coolant accident has previously occurred The LOCA signal has cleared Instrument air header pressure is lowering Which of the following describes requirements to manually start the Emergency Instrument Air Compressor (EIAC) before depressing the START pushbutton?

A. The LOCA signal must be reset then the 1 E breaker closed. B. The LOCA signal must be reset then the Non-1E breaker must be closed. C. The LOCA signal must be reset then the 1E breaker closed AND Instrument air header pressure must then drop below 85 psig. D. The LOCA signal must be reset then the Non-I E breaker must be closed AND Instrument air header pressure must then drop below 85 psig. Proposed Answer:

A Explanation (Optional):

A. Emergency Instrument Air Compressor following a LOCA, the feeder breaker on Class 1 E Unit Substation 108450 must be reclosed. The Compressor can then be started from either the Control Room OR Local Panel 1 OC189. The Emergency Instrument Air Compressor will start anytime the MANUAL pushbutton is pressed.

Correct - IAW SO.KB-0001 Steps 3.3.4 & 3.3.5 - To restart the

9. C. D. Incorrect - ElAC is powered by a 1 E supply. Incorrect - IA header pressure is not a restraint.

Incorrect - ElAC is powered by a 1 E supply. Ik header pressure is not a restraint. Technical Reference(s)

HC. OP-SO. KB-000 1 (Attach if not previously provided) Page 100 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

none Learning Objective:

INSAIREOI 5 (As available) Question Source:

Bank # 53430 Modified Bank # New (Note changes or attach parent) - Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 Comments: Page 101 of 205 HC.0P-SO .KB-OOO 1 (Q) 3.3.4. 3.3.5. 3.3.6. 3.3.7. 3.3.8. 3.3.9. The Emergency Instrument Air Cornprcssor will start anytime the MANUAL pushbutton is pressed. - To stop the Emergency Instrument Air Compressor after swapping to manual the STOP push-button must be pressed AND held for 3 seconds. To restart the Emergency Instrument Axr Compressor following a LOCA, the feeder breaker on Class 1E Unit Substation 1 OB450 must be reclosed.

The Compressor can then be started fiom either the Control Room - OR Local Panel 1 OC 189. - IF R4CS cooling water is less than 45 psig, THEN the Emergency Instrument Air Compressor Will NOT stat. AND has been inhibited fiom start due to low RACS pressure, THEN the compressor will AUTO starr AND run normally once RACS pressure has been restored.

the compressor is in AUTO - WHEN the Emerg Inst Air Cornpressor is placed in-service THEN the RACS System Demineralizers will isolate. the Emergency Air Compressor breaker is racked out, THEN you will get a one time compressor AUTO-START WHEN you go fiom MANUAL to AUTO position (once the breaker is racked in). - The Emergency Instrument Air Compressor will run for a minimum of 45 minutes following an AUTO-START signal. (This assumes the compressor never loaded - Step 3.3.2) - 3.3.10. 1AF104, Standby Instrument Air Dryer, will automatically be in-service with power failure or air pressure dropping below 85 psig. - 4.0. EOU IPMENT REXIUIRED None Hope Creek Page 5 of 21 Rev. 18 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 1 KIA # 295038 EA1.03

-~ Importance Rating

3.7 Ability

to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE: Process liquid radiation monitoring system. Proposed Question: Common 50 Given the following:

0 A discharge of the Equipment Drain Sample Tank is in progress to the River The Liquid Radwaste Discharge Isolation Valve (HV-5377A) to the Cooling Tower Blowdown automatically closes Which one of the following condition(s) would cause this termination? (Assume NO operator action) (1) Liquid Radwaste Effluent High radiation setpoint is reached (2) Cooling Tower Blowdown dilution flow low flow setpoint is reached (3) Liquid Radwaste Effluent sample flow rate HI setpoint is reached (4) Cooling Tower Blowdown RMS High radiation setpoint is reached (5) Liquid Radwaste Effluent High discharge flow setpoint is reached A. (1) and (3) ONLY B. (2), (4) and (5) ONLY C. (2), (3) and (4) ONLY D. (I), (2) and (5) ONLY Proposed Answer: D Explanation (Optional): D. Correct IAW HC.OP-AR.SP-0001 Rev.

19 Alarm Point 9RX508 (page 23) AUTOMATIC ACTION Isolation of HV-5377A&B due to any one of the following:

High radiation (HIGH LED on OSP-RI-4861) High Disch Flow ( setpoint determined by Liquid Effluent Permit ) Page 102 of 205 I i ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Low Dilution Flow ( setpoint determined by Liquid Effluent Permit ) Low Sample Flow (OHBFIS-4861)

Monitor Failure D. Correct. A. B. C. Incorrect.

(3) is incorrect.

(5) is also correct Incorrect.

(4) is incorrect.

(1) is also correct Incorrect.

(3) is incorrect.

(1) is also correct. Technical Reference(s)

HC.OP-AR.SP-OOO1 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: RWOVERE005 (As available) 68906 ___. Question Source: Bank

1. Modified Bank # New (Note changes or attach parent)

___. Question History: Last NRC Exam 2002 Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 103 of 205 I LESSON NAME: NOH04RWOVER-02 w RADWASTE SYSTEM OVERVIEW - 4/8/08 $!\q b- :d&c" ' c 10. Dilution of liquid radwaste discharges shall be accomplished by introducing the effluent inti:, the cooling tower blowdown line. The discharge connection with the circulating water system shall be outsid 3 the power block. -7 , f." 105J.4 - -. 11. 12 Waste discharge from the liquid radwaste system shall be sampled before discharge, shall be monitored during discharge, and shall be automatically terminated when the instantaneous radioactivity concentration would reach 1 OCFR20 limits for an unrestricted area after dilution.

The liquid radwaste release system automatically isolates on the following: -r --+c FJ. 5 -1 12. 13. 14. 15. 16. 17. 18. 19. a. Radwaste effluent high radiation b. Radwaste effluent high flow c. Cooling Tower blowdown low flow d. Loss of operate (Liquid Radwaste Effluent Radiation Monitor Downscale)

e. Loss of sample flow High purity wastes should have a Suspended solids concentration

<2( PPm- Lower purity wastes shouiid have a suspended solids concentration

<500 ppm. Liquid radwaste discharges should not exceed 1 % of the 1 OCFR20 limits on an annual average basis at the point discharge.

Class 1 E power shall be provided for primary and secondary con ta i n m en t is0 lat io n on I y . Environment qualification shall be provided for primary and secondap containment isolation valves only.

Equipment and components of the liquid radwaste system shall be ncn- seismic Category I with the exception of the containment isolation valves and the associated piping. Foundations and walls of the structure that house the liquid radwaste system shall be designed to the seismic criteria stipulated in Regulatc ry Guide 1.143. The equipment and associated piping of liquid radwastf.

system is non-seismic Category I. The liquid radwaste tanks rooms shall be designed to contain the maximum liquid inventory in case the tank ruptures.

Page 12 of 29 f?.\Dnri irnants and Sattinns\cdannic\hAv nnrl tmPnts\Hnno CropPk 3nnl~-3nnq\R~farpnrpc\i FSSSnN PI ANS\Radwasta Svctem I I I ES-401 Sample Written Eiamination Form ES-401-5 Question Worksheet Examination Outline Cross-reference

Level RO SRO Tier # 1 WA # 295003 AA2.04 Group # 1 Importance Rating

3.5 Ability

to determine andlor interpret the following as they apply to PARVIAL OR COMPLETE LOSS OF A.C. POWER : System Lineups Proposed Question: Common 51 The plant was at full power with all systems operable and in their normal alignment.

0 'A' and 'C' RACS pumps were in service Then a Loss of Offsite Power occurred.

0 All four EDG's started and their loads were sequenced on as designed. Which one of the following describes the response of RACS and Chilled Water system to this transient?

The 'A' and 'C' RACS pumps trip,.. A. the 'A' and 'B' RACS pumps ONLY are automatically started by the LOP sequencers and Chilled Water is aligned to the Drywell coolers.

B. the 'A', 'B' and 'C' RACS pumps are automatically started by the LOP sequencers and Chilled Water is aligned to the Drywell coolers.

C. the 'A' and 'B' RACS pumps ONLY are automatically started by the LOP sequencers and RACS is aligned to the Drywell coolers.

D. the 'A', 'B' and 'C' RACS pumps are automatically started by the LOP sequencers and RACS is aligned to the Drywell coolers.

Proposed Answer: C Explanation (Optional):

C. Correct - IAW HC.OP-SC>.ED-0001, Section 3.2.10 - A and B RACS Pump Motors are connected to Class 1 E buses AND upon Loss of Power (LOP) without occurrence of a Loss of Coolant Accident (LOCA), A and B RACS pumps restart automatically (in 85 seconds) after the sequencer permissive is received. CHILLED WATER CONTAINMENT CLG SPLY SELECT GB-HV-9530 AI/A3 AND B1/B3 LOOP A and B SPLY/RTN CHW will close AND GB-HV-9530 A2/A4 and B2/B4 LOOP A and B SPLY/RTN RACS will open and if in AUTO, Page 104 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet and not in REMOTE, HV-2537 A and B HX INLET VLVS 1AE217 and 1 BE217 INLET will open. C RACS Pump is connected to a non-I E bus (10B250) and upon loss of power, is de-energized with no restoration capabilities.

A. B. D. Incorrect.

RACs is aligned to the DW Coolers Incorrect. C RACs pump will have no power. RACs is aligned to the DW Coolers. Incorrect. C RACs pump will have no power. Technical Reference( s) HC. OP-SO.

ED-000 1 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

RACSOOE009

___. (As available) Question Source:

Bank # Modified Bank # -. 64579 (Note changes or attach parent)

New -. Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments: Page 105 of 205 I 3.2. HC.OP-SO.ED-000 1 (Q) 0. A and B RACS Pump Motors are connected to Class 1E buses AND upon Loss of Power (LOP) without occurrence of a Loss of Coolant Accident (LOCA), A and B RACS pumps restart automatically (in 85 seconds) after the sequencer permissive is received.

CHILLED AND BUB3 LOOP A and B SPLY/E<TN CHW will close AND GB-HV-9530 A2/A4 and B2/B4 LOOP A and B SPLY/RTN RACS will open and if in AUTO, and not in REMOTE, HV-2537 A and B HX INLET VLVS 1AE217 and 113E217 INLET will open.

C RACS Pump is connected to a non-1 E3 bus ( 1 OB250) and upon loss of power. is de-energized with no restoration capabilities.

WATER CONTAINMENT CLG SFLY SELECT GB-HV-9530 A1 /A3 ___ 3.2.1 I. All pump operations shall be performed IAW OP-HC-108-106-1001, Equipment Operational Control.

~ 3.2.12. The below listed valves will isolate the Service Water supply to RACS Heat Exchangers on either a LOCA signal signal: RACS room flooded EA-HV-2203 SERVICE WATER LOOP A RACS HX MDR SPLY - 0 EA-HV-2204 SERVICE WATE:R LOOP B RACS HX HDR SPLY - EA-HV-2207 RACS HX COOLING INLET VALVE - 0 EA-HV-2346 RACS HX COOLING OUTLET VALVE __ 3.2.13. Operations of RACS in restricted operation, without Service Water System, shall be limited to a maximrim RACS System Temperature of 95 OF (as indicated by TE-2563). - 3.2.14. RACS must be removed from service _. OR operated in the restricted mode, INE"QRM Chemistry to isolate the Crack Monitoring System. 3.2.15. MAINTAIN RACS Pump(s) suction temperature between 45 "F and 95 OF for normal operation. It may be necessary to provide artificial heat load and/or isolate SSWS flow to one RACS heat exchanger and throttle SSWS flow to < 4000 gpm in the remaining RACS Heat Exchanger to maintain temperature

> 45°F. Limit flow of KACS through any one (1 A-E-2 17 or I B-E-2 17) Heat Exchanger to 33 84 gpm, maximum. For normal 2 pump operation this means 2 Heat Exchangers are required.

[CD-929G]

~ 3.2.1 6. RACS allowable operating range is 45°F to 95"F, RACS UFSAR temperature range is 40°F to 95°F. Hope Creek Page 6 of 27 Rev. 22 The plant was at 100% power with the 'B' and 'C' RACS pumps in service when a Loss of Offsite Power occurred.

All four EDG's started and loaded properly What is the expected response of the RACS pumps to this transient?

A. The '6' and 'C' RACS pumps tnp. The 'A' and 'B' RACS pumps are automatically started by the LOP sequencers.

B. The 'B' and 'C' RACS pumps trip. The 'A' & 'B' RACS pumps must be manually restarted.

C. The 'B' RACS pumps trips and rs automatically restarted by the LOP sequencer The 'C RACS pump remains running.

D. The 'C' RACS pump trips. The

'B' RACS pump remains running. The

'A RACS pump is automatically started by the LOP sequencer.

Answer: A I I I ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO -~ Tier # 1 WA # 295005 AA2.03 Importance Rating 3.1 Group # 1 Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP

Turbine valve position. Proposed Question: Common 52 Given the following conditions:

0 Reactor power is 50% 0 The plant is operating normally with Main Turbine First Stage Pressure at 243 psig. Then, 0 A main generator load reject has just occurred resulting in a power/load unbalance trip. Which one of the following is the immediate response of the Turbine Control Valves (TCVs), Intercept Valves (IVs) and the Reactor Protection System (RPS)?

A. The TCVs and IVs Fast Close. RPS will trip.

B. The TCVs and IVs Fast Close. RPS will NOT trip. C. The TCVs and IVs Throttle Close.

RPS will trip.

D. The TCVs and IVs Throttle Close.

RPS will NOT trip. Proposed Answer: A Explanation (Optional):

A. Correct IAW EHC Lesson plan NOH01 EHC LOG-04, Page

'I 8, If a power to load unbalance occurs, the control valve and intercept valve fast acting solenoids are actuated. IAW Turbine Lesson plan NOH01 NMTURB-04, page 66, RPS is automatically bypassed at <24% power which is equal to approximately 104.2 psig first stage turbine pressure.

B. C. Incorrect - RPS will trip Incorrect - Valves will fast close Page 106 of 205 I I ES-401 Sample Written Examination Form ES-40 1-5 Question Worksheet D. Incorrect - valves will fast close.

RPS will trip Technical Reference(s) NOH01 EHC LOG-04 (Attach if not previously provided) NOH01 NMTURB-04 Proposed references to be provided to applicants during examination:

none Learning Objective:

EHCLOGEOOS (As available) Question Source:

Bank # Modified Bank # New -. NRC 2005 (Note changes or attach parent)

Question History: Last NRC Exam -. 2005 Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments: Page 107 of 205 MAIN TURBINE CONSTRUCTION AND COMPONENT:09/18/07 This annunciator will be received if the hydraulic fluid pressure to any of the turbine CVs is 630 psig. It is an expected alarm on any main turbine trip event and

\Mill also be received on a main generator load rejection and/or on a loss of EHC Control Oil pressure.

MAIN STOP VALVE CLOSURE (Window C5/B2)

This annunciator will be received if any MSV is <95% open. It is an expected alarm on any main turbine trip event and will also be received on a loss of EHC Control Oil pressure.

TCV FAST CLOSE AND MSV TRIP BYP (Window C5/C2) The TCV fast closure and MSV closure inputs to the Reactor Protection System are automatically bypassed when reactor power is < 30% (as sensed by main turbine first stage pressure). This annunciator will Ire received if main turbine first stage pressure is 435.7 psig. 6. 7. C. Control Functions and Interlocks

1. Quill Shaft Failure I Obj. 4a,14 I a. Failure of the quill shaft with the main turbine in operation could result in a false indication of a stationary rotor from the low speed switch. Quill shaft failure is sensed by the presence of both of the following signals simultaneously:

1) Turbine speed wr than 100 RPM as sensed by the magnetic speed pickups located on the driving side of the quill shaft. (Turbine Side) Turbine stopped signal as sensed by the low speed switch located on the driven side of the quill shaft. (Pump shaft) AND 2) b. The presence of both of these signals together indicates that the main turbine rotor is still rolling at high speed but the pump shaft has stopped. This signifies the failure of the quill shaft, and will cause the TURBINE GENERATOR TROUBLE annunciator to be received.

If main turbine speed was initially greater than 1350 RPM, the main turbine will trip on low main shaft oil pump discharge pressure (1 05 psig:,. This trip is bypassed if main turbine speed was initially less than 1350 rpm. c. I -66of102 C:\Docurnents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\References\LESSON PLANS\Main Turbine 1 Construction B Components\Master\Lesson Plans\NOHOl MNTURB-04 Main Turbine Construction and Cornponents.doq~ ~ ~ ~' -~ Deleted: Page 103 of 107 Y \VDnve\TRAINING DOCUMENTS\Operations Training\Hope Creek\Plant Technology\Systems\Main urbine Construction B Cornponents\MasterVesso Plans\NOHOl MNTURB-02 Main Turbine Cons & Componer s doc -~

I I Question # RO SRO d5005 AK2 04 Question 20 Hope Creek RO Exam - Nov 2005 Tier # 1 Group # 1 HC Ubj: Importance 3 3 Main Turbine Generator Trip 13 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following Main generator protection (CFR 41 7/45 8) Given the following conditions: - The plant is operating at 20% power - A main generator load reject has just occurred -A fault in the control circuit causes a power/load unbalance trip during the load reject Which of the following is the immediate expected response of the Turbine Contrcd Valves (TCVs) and the Reactor Protection System (RPS)?

A B TCVs throttle close, RPS trips TCVs throttle close, RPS does NOT trip c TCVs fast close, RPS trips r) TCVs fast close, RPS does NOT trip Answer D References actions Hope Creek Question - Qbl307, and notes COMPONENTS, p. 66 HC.OP-AB.BOP-0002 Additional Information /Automat

NOH01 MNTURB-02, MAIN TURBINE CONSTRUCTION AND Justification References during Exam None CORRECT - TCVs fast close, RPS does NOT trip. The load reject causes the TCVs to fast close.

The fast closure does NOT initiate a RPS trip because turbine load is <30%. Since power is within the capacity of the BPVs, NO pressure transient will trip RPS.

INCORRECT - TCVs throttle close, RPS does trip. The load reject causes the TCVs to fast close. The fast closure does NOT initiate a RPS trip because turbine load is <30%. Since power is within the capacity of the BPVs, NO pressure transient will trip RPS. INCORRECT - TCVs fast close, RPS does trip. The fast closure does NOT initiate a RPS trip because turbine load is <30%. Since power is within the capacity of the BPVs, NO pressure transient will trip RPS. INCORRECT - TCVs throttle close, RPS does NOT trip.

The load reject causes the TCVs to fast close Question Source Bank Memory Level Comprehension Level Question History:

SXD review - 7/21 - OK AF - 10/13 possible WA mismatch AF - 8/23 - OK MB - OK 1 l/8

, I' LESSON NAME: NOHOIEHC LOG-04 EHC CONTROL LOGIC - 03/14/07 _____. NOTE: Loss of SWC signals are: 0 Temp 79°C = Hi stator winding cooling outlet temp 0 Press 64 psis = Lo stator winding cooling inlet press 0 Flow setpoint - 690 gpm. These are a 2/3 logic. If stator cooling for the generator is lost, the load set is reducetl at 26.8%/min. Turbine load will be reduced only when the load set signal is less than the pressure setpoint signal. The load set will continue to be reduced until either stator coolant system status returns to normal or generator load is reduced to less than the stator amperage of N 26.8% of rated turbine load(7055 ,amps). Below this value the generator has the capability to operate safely without stator cooling. Therefore, further reduction of generator load is not required. The runback progress is checked after 2 minutes. If generator loading is not below

= 95% of rated turbine load (24683 amps) at the end of this period, protective circuitry indicates a failure of the runback circuitry and initiates a turbine trip. After 3.5 minutes, if generator loading has not been reduced to the minimum limit (26.8%) of rated turbine load (6814 amps), the circuitry will initiate a turbine trip.

Restoration of stator cooling during this period will stop the turbine runback.

Currently, following1 RF12, the 100% turbine load will be when the generator is carrying 26314 amps.

IObj. 9a;b;c .- f) If a power-to-load unbalance (PLU) (also called a load reject) occurs:

0 0 0 The load signal is immediately set to minimum. The control valve and intercept valve fast acting solenoids are actuated A direct turbine trip is generated.

Page 18 of 47 C:\Docurnents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\References\LESSON PLANSEHC Control Logic\Master\Lesson Plans\NOHOl EHCLOG-04 EHC Control Logic.doc I I ES-40 1 Sample Written Exam inat ion Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 1 KIA # 295037 EA2.02 Importance Rating 4.1 Ability to determine andlor interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Reactor water level. Proposed Question:

Common 53 While operating at 100% Reactor Power, an MSlV (IVSSSS) isolation occurs and the reactor fails to scram; all rods remain at their pre-trip conditions. The operators are in the process of deliberately lowering RPV water level. The current plant conditions are:

0 RX Power 4.5% 0 RPV Pressure 900 psig 0 RPV Level -120 inches 0 0 0 Drywell Pressure 4.5 psig 0 0 Suppression Pool Level 79 inches and rising Suppression Pool Temp 175 F and rising SLC Injecting with 3000 gallons remaining in SLC Tank Control Rods are being inserted What action(s) are required to be performed IAW EClPs? A. B. C. D. Open SRVs to depressurize.

Continue to lower RPV level. Open SRVs to Emergency Depressurize. Restore and maintain RPV water level between +12.5 inches and

+54 inches. Proposed Answer:

A. Explanation (Optional):

A. Correct - Continue to lower RPV level because power is above 4.5 YO B. Incorrect - HCTL action required and RPV water level do not meet conditions for ATWS Page 108 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Emergency de-pressurization.

Incorrect - Can not restore RPV water level until the Reactor is shutdown under all conditions without boron, exit EOP-101A and enter EOP-101. Incorrect - Suppression pool temperature can be maintained below HCTL action required area. SRVs not required C. D. Technical Reference(s) EOP-101 A (Attach if not previously provided) Proposed references to be provided to applicants during examination:

EOP-IOIA - no entry conditions Learning Objective: Question Source:

E01 01AE008 (As available)

~. Bank # Modified Bank # ID: Q56142 (Note changes or attach New parent) Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments: Page 109 of 205 I I ~. ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 WA # 295021 G2.1.28 Importance Rating

4.1 Group

1. 1 Conduct of Operations: Knowledge of the purpose and function of major system components and controls. (Loss of Shutdown Cooling) Proposed Question:

Common 54 Plant Conditions are as follows: Alternate Shutdown Cooling is being implemented by using the C to A RHR Loop Cross-Tie IAW HC.OP-AB.

RPV-0009.

RHR Loop C is the only means of removing heat from the RPV. If the operator opens HV-F007C, C RHR PMP MIN F'L MOV during this operation, how will the plant initially respond? A. RHR Pump C will lose NPSH. B. The RPV will drain to the Suppression Pool. C. Flow through the A RHR Heat Exchanger will rise. D. SACS outlet temperature from A RHR Heat Eixchanger will rise. Proposed Answer:

B. Explanation (Optional):

B. Correct - Opening HV-F007 C will establish a drain path from the B Recirculation Pump Loop to the Torus via C RHR Pump Suction and HV-F007 A. C. D. Incorrect - C RHR Pump would eventually lose NPSH. The stem stipulates the selection of the first consequence Incorrect - The flow which existed initially in the A RHR Heat Exchanger will lower due to a drain path being opened to the Torus.

Incorrect - The loss of RHR flow to the A RHR Heat Exchanger will lower the heat burden on SACS and hence the SACS outlet temperature will not rise. Technical Reference(s) HC.OP-AB.RPV-0009, Rev.5 (Attach if not previously provided)

Caution 1.6 Page 110 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

none Learning Objective:

ABRPVSE004 (As available)

Question Source: Bank ## ID: Q61858 Modified Bank # New (Note changes or attach parent) ~. Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 11 1 of 205 I I I E R SPONSIBLE FOP L'ERIF'ilNG REL'ISION. STATU(;

AND ::Hkl\lGES

~~22ooalrr:.'

lJrcL PSEG Internal Use ~nlv HC.OP-AB.RPV-O009(Q)

SHUTDOWN COOLING ATTACHMENT 2 (Page 2 of 8) ALTERNATE DECAY HEAT REMOVAL USING C TO A CROSS-TIE

/-? CAUTION! I .6 Manual or automatic opening of HV-F007 A(C) RHR PMP A(C) MIN FL MOV will drain the Reactor Vessel to the Suppression Pool. 1.6 1.7 1.8 1.9 1.10 1.11 Hope Creek ENSURE the following valves are CLOSED: HV-FO07A RHR PMP A MIN FL VLV. HV-FO07C RHR PMP C MIN FL 'VLV. VERIFY the following LPCI Injection Valves are CLOSED: HV-FO17A RHR LOOP A LPCI INJ MOV HV-FO17C RHR LOOP C LPCI DJJ MOV CLOSE the following valves: HV-FO1 OA RHR LOOP C TEST RET MOV HV-F024A RHR LOOP A TEST PET MOV HV-F021A RHR LOOP A SPRAY ISLN MOV HV-F027A RHR LOOP A SUPP CHAMBER SPRAY HDR ISLN MOV Fully OPEN 1BC-V133 RHR Pmp C Siict Frm Recir Loop B (Rm 4227E). ENSURE F077 RECIRC LOOP B TO IWR SUP MAN VLV is open. - IF the Shutdown Cooling suction line war; isolated, THEN PERFORM a fiIl and vent lAW *IC.OP-SO.BC-O002(Q), Decay Heat Removal Operation.

Page 27 of 55 ~- -- -I Rev. 5 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 1 K/A # 295024 G2.4.1 Importance Rating

4.6 Emergency

Procedures

/ Plan: Knowledge of EOP entry conditions and immediate action steps. (High DW pressure)

Proposed Question:

Common 55 A small break LOCA occurs and the following conditions exist:

Drywell temperature is 150°F 0 Drywell pressure is 2.5 psig Reactor Water Level is +I5 inches Suppression pool temperature is 93 degrees F Suppression Pool pressure is 1.6 psig Suppression Pool level is 77 inches Which of the following correctly describes the sections of Emergency Operating Procedures which have been entered and the initial step(s) required?

A. ALL sections of OP-EO.ZZ-101, Reactor/Pressure Vessel (RPV)

Control, and ALL sections of OP-EO.ZZ-102, Primary Containment Control. Lock the Mode Switch in Shutdown.

B. ALL sections of OP-EO.ZZ-101, Reactor/Pressure Vessel (RPV)

Control, and ALL sections of OP-EO.ZZ-102, Primary Containment Control. Lock the Mode Switch in Shutdown and place Drywell Spray in service.

C. Drywell Pressure (DW/P) and Drywell Temperature (DWn) sections of OP- EO.ZZ-I 02, Primary Containment Control only, and ALL sections of OP-EO.ZZ-101, ReactodPressure Vessel (RPV)

Control. Lock the Mode Switch in Shutdown.

D. Drywell Pressure (DW/P) and Drywell Temperature (DW/T) sections of OP- EO.ZZ-I 02, Primary Containment Control only, and ALL sections of OP-EO.ZZ-101, Reactor/Pressure Vessel (RPV)

Control. Lock the Mode Switch in Shutdown and place Drywell Spray in service. Page 112 of 205

, I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed Answer: A Explanation (Optional):

A. Correct - IAW EOP entry conditions and initial steps. Per EOP usage guidance all legs of each EOP are performed concurrently.

Conditions are not met for Drywell spray (see attached EOPs)

A. Correct. B. C. D. Incorrect. Drywell spray conditions not met Incorrect. All legs of both EOPs 101 and 102 must be entered Incorrect.

All legs of both EOPs 101 and 102 must be entered. Drywell spray conditions not met Technical Reference(s) EOP 101 & 102 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

DSlL curve Learning Objective:

E01 01 LE003 (As available) 56092 - Question Source:

Bank # Modified Bank # -. (Note changes or attach parent)

New - Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments: Page 113 of 205 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 295016 G2.4.46 I m portaiice Rating 4.2 Group # 1 Emergency Procedures

/ Plan: Ability to verify that the alarms are consistent with the plant conditions. Proposed Question: Common 56 Plant conditions are as follows: 0 0 0 0 0 A fire causes an MSlV closure resulting in a scram HC.OP-AB.HVAC-0002, Control Room Environment is complete Control is being established at the Remote Shutdown Panel (RSP) IAW HC.OP-1O.ZZ-0008, Shutdown from Outside the Control Room All Transfer Switches are in the Emergency position While placing RClC in service at the RSP the following indication is received:

0 Turbine Tripped and Low Bearing Oil Pressure alarm indicating lights illuminated Which one of the following would cause this response?

A. RClC System trip on high RPV water level. B. Trip of the RClC Turbine Mechanical Overspeed device. C. RClC System Steam Line break causing an automatic system Isolation.

D. Consequences of the fire because there are NO automatic actions associated with the RClC system with control from the RSP. Proposed Answer:

B Explanation (Optional): IAW HC.OP-IO.ZZ-0008, Note 3.1.8.D B. CORRECT - Trip of the RClC Turbine Mechanical Overspeed device. With control at the RSP, all automatic trips and interlocks are disabled.

One exception is the RClC Overspeed Trip. Since it is a mechanical device, it will perform its function even with control at the RSP. A. INCORRECT - RClC System trip on high RPV water level. With control at the RSP, all automatic trips and interlocks are disabled.

Page 114 of 205 I I ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet C. D. INCORRECT - RClC System Steam Line break causing an automatic system Isolation.

With control at the RSP, all automatic trips and interlocks are disabled.

INCORRECT - Consequences of the fire, there are NO automatic actions associated with the RClC system with control from the RSP. One exception is the RClC Overspeed Trip. Since it is a mechanical device, it will perform its function even with control at the RSP. Technical Reference(s) HC.oP-I0.ZZ-0008 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective:

IOP008E004

____. (As available) 62224 ~. Question Source:

Bank # Modified Bank # New (Note changes or attach parent) Question History: Last NRC Exam Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 11 5 of 205 I I I LESSON NAME: NOH01 REMSID-02 REMOTE SHUTDOWN SYSTEMS - 08/13/08 ~ ~~ ~~ ~~ IO) Indication for the following valves: 1) 2) 3) 4) 5) 6) 7) 8) Control/indication for the following pumps:

1) 2) Indication for ECCS (RCIC) jockey pump BP228 Indication for RClC turbine speed (SI-4280-2) Control/indication for RClC system injection flow (FIC-4158) Indication for RClC turbine trip (ZA-4275)

Indication for the following alarm conditions:

1) 2) 3) 4) RClC steam line inboard isolation valve bypass valve (HV-F076)

b. RClC pump discharge valve (HV-F012) Test return valve to condensate storage tank (HV-F022) RClC turbine exhaust to suppression pool valve (HV- F059) RClC condenser vacuum pump discharge valve (HV-F060) RClC turbine exhaust outboard vacuum breaker isolation valve (HV-F062)

RClC turbine exhaust inboard vacuum breaker isolation valve (HV-F084)

RClC condensate pot drain to main condenser valve RClC vacuum tank condensate pump discharge to clean radwaste valve (HV-F004) RClC vacuum tank condensate pump OP220 RClC gland seal condenser vacuum pump OP219 (HV-F025)

c. d. e. f. g. h. RClC turbine lbearing oil pressure low (PAL-4276)

RClC high pressure turbine bearing temperature high RClC low pressure turbine bearing temperature high Condensate storage tank level low-low (LALL-NO61 ) c (TAH-4277) (TAH-4278)

2. Residual Heat Removal (RHR)

System a. ControMndication for the following valves: Page 16 of 27 S:\VDrive\TRAINING DOCUMENTS\Operations Training\Hope Creek\Plant Technology\Systems\Rernote . Shujdown\Master\Lesson Plans\NOHOl REMSD-02 Remote ShLitdown.doc d 5% LESSON NAME: NOH01 REMS/D-02 REMOTE SHUTDOWN SYSTEMS - 08/13/08 B . M ai o r S u b s y s t e m s/C o m p o n en t s I Obi. 3c .I 1. Non-1 E Emergency Transfer Switch The NON 1 E TRANSFER SWITCH, when positioned to EMER will initiate closure of the B reactor recirculation pump discharge valve, HV-F031 B. This action prevents "short cycling" of shutdown cooling flow when established.

HV-F031 B going closed provides a trip signal and a runback signal to the "B'l recirc pump; thus, if the pump is running when the Emergency Transfer Switch1 is taken to emergency, the pump will trip. 2. The Channel "A" 1 E Transfer Switch being placed in emergency causes the following:

a. RHR valves HV-F021A (Inboard Cont Spray), HV-FOOGA (Pump A suction from Recirc) auto-close.

The following actions will occur but will not be observed at the RSP: 1) RHR valves HV-F021A, HV-FOOGA become inoperative from the control room.

2) RHR valves HV-FOO9 (SDC Inboard Isolation) and HV-F049 (B Loop discharge to RW) become INOP from the Main Control Room. 3) SACS valves HV-231 7A and HV-7922A (SACS FPC X-Conn) auto-[:lose signals are disabled and valve control becomes INOP from Main Control Room. Supply/Return)become INOP from Main Control Room. The OVERLOADIPOWER FAIL MONITOR (OPF) is disabled for all valves mentioned in steps 1, 2, 3 & 4. Fuel Pool Cooling Pump AP211 becomes INOP from Main Control Room. 4) SACS valves t-IV-2314A HV-7921A (SACS FPC 5) 6) Page 11 of 27 S:\VDrive\TRAINING DOCUMENTS\Operations Training\Hope Creek\Plant Technology\Systerns\Remote Shutdown\Master\Lesson Plans\NOHOl REMSD-02 Remote Shutdown.doc I I I LESSON NAME: NOH01 REMSID-02 REMOTE SHUTDOWN SYSTEMS - 08113108 7) The RSP/RSS TAKEOVER and the BOP SAFETY SYS OUT OF SVCE alarm annunciates.

3. The Channel "B" 1 E Transfer Switch being placed in emergency causes the following:

a. The following will be observed at the RSP: 1) RHR valves I-iV-FOIGB (DW Spray), HV-F017B(LPCl Injection) , HV-F021 B(DW Spray), & HV-F027B (SP Spray)auto-close.

RClC valves I-lV-F012( Pump Discharge), HV- F059( Exhaust Isolation), HV-F06O(Vacuum Pump discharge), H'V-F062( Exhaust Vacuum breaker isolation) auto-o pen.

RClC valve HV-F022 (Min Flow valve) auto-closes.

The RClC Jockey Pump BP-228 auto-starts.

@ RClC turbine alarms are enabled at the RSP (may not The following actions will occur but will not be observed at the RSP: 1) RHR valves HV-FOIGB, HV-F017B, HV-F021B, HV- F027, become inoperative.

2) RHR valves HV-F003B, HV-F004B, HV-FOOGB, HV- F007B, HV-F024B, HV-F047B, and HV-F048B become INOP from the Main Control Room. RClC valves HV-FO12, HV-F059, HV-FOGO, HV-FOG2 become INOP from Main Control Room. The RClC Vacuum Pump OP-219

& RClC Condenser Pump OP-220 auto-start signals are disabled.

RClC turbine alarms are enabled at the RSP (may not be on). RClC valves HV-F008, HV-FOIO, HV-FO13, HV-FO19, HV-F022, HV-F031, & HV-4282 become INOP from the Main Control Room. 2) 3) 4) 4 be on). b. 3) 4) 5) 6) Page 12 of 27 S:\VDrive\TRAINING DOCUMENTS\Operations Training\Hope C:reek\Plant Technology\Systems\Remote Shutdown\MasteALesson Plans\NOHOl REMSD-02 Remote Shutdown.doc I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level Tier # Group # KIA # RO SRO 1 1 295028 EK2.01 Importance Rating

3.7 Knowledge

of the interrelations between HIGH DRYWELL TEMPERATURE and the following: Drywell spray: Mark-l&ll. Proposed Question:

Common 57 Which one of the following describes the consequences of spraying the drywell when drywell temperature is ABOVE the drywell temperature limit of the Drywell Spray Initiation Limit Curve? A. External pressure limits on the Secondary Containment would be exceeded.

B. Automatic Depressurization System instruments would NO longer be qualified for these conditions . C. The relief capacity of the Suppression Chamber to Drywell vacuum breakers would be exceeded.

D. RPV water level instrumentation would become inaccurate due to rapidly lowering drywell temperatures. Proposed Answer:

C. Explanation (Optional):

C. Correct - If unrestricted, the evaporative cooling affect of Drywell spray could result in an immediate, rapid and large reduction in Drywell pressure at a rate much faster than can be compensated for by the Primary Containment Vacuum Relief System and thus result in a negative Drywell-to-Suppression Chamber differential pressure large enough to cause a loss of Primary Containment integrity.

A. B. D. Incorrect - The limit of concern is on the Primary Containment.

Incorrect - There is no relationship between the DSlL Curve and ADS. The DSlL Curve permits Spray for a wide range of pressures with Drywell temperature above 340°F. Incorrect - Inaccurate level indication may occur when drywell temperature exceeds the saturation temperature for the existing RPV pressure, since drywell sprays lower drywell temperature they will have the opposite affect. Technical Reference(s)

Bases for DW Spray Curve- (Attach if not previously provided)

EOP-102 Page 11 6 of 205 I I ES-40 1 Sample Written Eiamination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

none Learning Objective:

E01 02PE006 (As available) - ___. Question Source:

Bank # Modified Bank # ID: Q53337 (Note changes or attach New parent) ___. Question History: Question Cognitive Level:

Last NRC Exam -. Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 11 7 of 205 I I HC.OP-EO.ZZ-0102 LJ BASES initiated, the resulting pressure reduction opens the wetwell-to-drywell vacuum breakers, drawing noncondensibles from the suppression chamber back into the drywell.

As an added conservatism, drywell sprays are initiated when the drywell noncondensible content reaches 5%. This condition defines the Suppression Chamber Spray Initiation Pressure (SCSIP). As the drywell atmosphere is purged to the suppression chamber and replaced by steam, suppression chamber pressure increases. The SCSIP is the lowest suppression chamber pressure which can occur when 95% of the noncondensibles in the drywell have been transferred to the suppression chamber. Refer to HC.OP-E0.Z-LIMITS-CONV, EOP Limit Curves and Cautions for detailed discussion of the SCSIP. The restriction regarding suppression pool water level applies to the internal suppression chamber-to-drywell vacuum breakers. These vacuum breakers will not function as designed if any portion of the valve is covered with water. The specified suppression pool water level assures that no portion of the drywell side of the valve is submerged for any drywell below wetwell differential pressure less than or equal to the valve opening differential pressure. Spray operation with vacuum breakers inoperable (i.e., with no drywell vacuum relief capability) may cause the containment differential pressure capabiiity to be exceeded and therefore is not permitted.

Spray operation effects a drywell pressure and temperature reduction through the effects of r evaporative cooling.

In evaporative cooling the water spray undergoes a change of state, liquid to vapor. Evaporative cooling refers to spray droplet heat and mass transfer which occurs when water is sprayed into a superheated atmosphere. The water in each droplet is assumed to instantaneously heat and flash to steam until the surrounding atmosphere saturates, absorbing heat energy from the atmosphere. For bounding calculations with typical drywell spray flowrates, this cooling process results in an immediate, rapid, and large reduction in drywell pressure at a rate much faster than can be comperisated for by the primary containment vacuum relief system. Unrestricted operation of drywell sprays could thus result in a negative drywell-to-suppression chamber differential pressure large enough to cause a loss of primary containment integrity. Considering the pressure drop concerns described above, the Drywell Spray Initiation Limit (DWSIL) is the highest drywell temperature at which initiation of drywell sprays will not result in an evaporative cooling pressure drop to below either: The drywell-below-wetwell differential pressure?

capability, or The high drywell pressure scram setpoint.

The DWSIL is a function of drywell pressure. It is utilized to preclude containment failure or de- inertion following initiation of drywell sprays.

Hope Creek Page 10 OX 36 Rev. 01 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008-2009\References\LESSON PLANS\Emergency Operating Procedures\EOP Bases\Master\Lesson Plans\EOP102bases.doc ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Out I i ne Cross-reference

Level RO SRO Tier # 1 Group #i 1 KIA # 600000 AA1.06 Importance Rating

3.0 Ability

to operate and

/or monitor the following as they apply to PLANT FIRE ON SITE: Fire Alarm Proposed Question: Common 58 Given the following conditions:

0 0 The Diesel Generator Room Carbon Dioxide Fire protection system is aligned for automatic operation A valid EDG room high temperature alarmi has just occurred Which of the following describes how the Diesel Generator Room Carbon Dioxide Fire protection system responds IAW HC.OP-SO. KC-0002 "C02 Fire Protection System Operation"?

A. A discharge alarm occurs and CO2 with a wintergreen scent is discharged into the room immediately.

B. A pre-discharge alarm is activated.

After a time delay, C02 with a wintergreen scent is discharged into the room. C. A pre-discharge alarm is activated.

NO C02 is discharged into the room. D. A pre-discharge alarm is activated and a wintergreen scent is discharged into the room. After a time delay, C02 is discharged into the room. Proposed Answer:

B Explanation (Optional):

IAW LP NOH01 FIRPRO-03 B. CORRECT: A pre-discharge alarm is activated. After a time delay C02 with a wintergreen scent is discharged into the room. A. C. D. INCORRECT: A discharge alarm occurs, C0:2 with a wintergreen scent is discharged into the room immediately.

No there is a warning time delay. INCORRECT: A pre-discharge alarm is activated.

No C02 is discharged into the room. EDG room actuates on Hi Temp INCORRECT: A pre-discharge alarm is activated and a wintergreen scent is discharged into the room. After a time delay, C02 is discharged into the room. Wintergreen comes in with the CO@, not before Page 11 8 of 205 I I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

___. Technical Reference(s) LP NOH01 FIRPRO-03 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: FlRPROEOl5 (As available) 53407 Question Source: Bank

1. -. (Note changes or attach parent)

-. Modified Bank # New Question History: 1998 Last NRC Exam _. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 11 9 of 205 NOH01 FIRPRO-04 FIRE PROTECTION SYSTEM - 07/23/2008 FA" LESSON NAME: applied to the port of the valve piston chamber forcing the valve disc downward, opening the valve.

When flow is to be terminated, the 3-way pilot valve repositions to allow a flow path to release the CO, pressure from the piston chamber so the valve disc repositions to close the valve.

valve and selector valve is the 3-way pilot valve. For the mater valve, when the solenoid that controls the 3-way pilot valve is energized, the flow path created maintains pressure beneath the valve disc so the master valve remains closed. De-energizing the pilot solenoid valve repositions the 3-way valve so that CO, pressure is applied to open the valve For the selector valves, a normally de- energized solenoid valve provides the necessary flow path to keep the selector valve closed.

Energizing the pilot solenoid valve repositions the 3-way valve to apply pressure to open the selector valve. e A major difference between the master e e In the event of a loss of power to the system, the master valve would fail open, allowing the system discharge piping to become pressurized with CO, up to the selector valves. However, the selector valves fail closed so that CO, is not inadvertently discharged.

This configuration also supports fire brigade response to a single location, the location of the selector valve, in the event that CO, is required to be discharged during a loss of power. d) System Operation e Each of the CO, systems is provided with a Chemetron control panel (white). When a CO, system actuates, the Page 57 of 85 S:\VDrive\TRAINING DOCUMENTS\Operations Training\Hope Creek\Plan:

Technology\Systems\Fire Protection\Master\Lesson Plans\NOHOl FIREPRO-04 FIRE PROTECTION SYSTEM.DOC Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 2 K/A # 295010 AKI .01 Importance Rating 3.0 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE: Downcomer submergence:

Mark-l&l I. Proposed Question: Common 59 Which of the following events would result in compromising the pressure suppression function of p ri ma ry con ta i n men t? (1) Uncovering the SRV T-Quenchers (2) Downcomer openings becoming uncovered (3) Torus to Drywell Vacuum Breakers failing closed (4) Reactor Building to Torus Vacuum Breakers failing open A. (1) and (2) ONLY B. (2) and (3) ONLY C. (1) and (4) ONLY D. (3) and (4) ONLY Proposed Answer:

A. Explanation (Optional):

IAW EOPI 02 bases SP/L-7 A. Correct - (1) and (2) would result in pressurizng directly the containment atmosphere therby potentially compromising high pressure limit in the containment B. Incorrect - SRV t-quenchers correct C. Incorrect - downcomer openings correct, Rx bldg to torus vac bkrs failing open would relieve torus pressure. D. Incorrect - would not compromise Pressure Supression function, suppression pool would still be available. Technical Reference(s)

EOP 102 bases (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Page 120 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet (As available)

~. Learning Objective:

EOPI 02E009 Question Source: Bank # Question History: Question Cognitive Level:

10 CFR Part 55 Content: (Note changes or attach parent)

Modified Bank # -. X - New Last NRC Exam -. Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 X Comments:

Page 121 of 205 HC.OP-EO.ZZ-0102 d lLL SPIL-6 I BASES RUNBACK Recirc AND INITIATE a manual scram Enter and Execute Concurrently EOP-101 Discussion Entering RPV Control at Step RC-1 assures that, if possible, the reactor is scrammed and shutdown is assured by control rod insertion before RPV depressurization is initiated. Entry into RPV Control must be explicitly stated because conditions requiring entry into Primary Containment Control do not necessarily require entry into RPV Control . Therefore, a scram may not yet have been initiated. Directing that RPV Control be entered, rather than explicitly stating here "Initiate a reactor scram," coordinates actions currently being executed if RPV Control has already been entered. In addition, entry to the RPV Control guideline must be made because it is through this guideline that the transfer to "Emergency RPV Depressurization,"

is effected.

SPIL-7 Can supp pool level be maintained above 38.5 in. YES 2 Return to SP/L-1 NO GO to SP/L-8 Discussion 6-- When suppression pool water level decreases to the elevation of the downcomer openings, any further drop in water level could result in direct exposure of the drywell atmosphere to the sippression chamber airspace thus compromising the pressure suppression function of the primary containmerQuppression - pool water levei should therefore be maintained above this eEva t ion. - . SPIL-8 EMERGENCY RPV DEPRESSURIZATION IS REQUIRED Enter and Execute Concurrently EOP-202 Discussion The RPV is not permitted to remain at pressure if suppression of steam discharged from the RPV into the drywell cannot be assured. When the downcomer vent openings are not adequately submerged, any steam discharged from the RPV into the drywell may not condense Hope Creek Page 26 of 36 Rev. 01 C:\Docurnents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\References\LESSON PLANS\Emergency Operating Procedures\EOP BasesWlaster\Lesson Plans\EOP102bases.doc I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO ~~ Tier # 1 Group # 2 KIA # 295022 AK2.03 Importance Rating

3.4 ______

Knowledge of the interrelations between LOSS OF CRD PUMPS the following: Accumulator pressure Proposed Question: Common 60 Given the following:

0 0 0 0 The plant is at 37% power Both CRD pumps are tripped on low suction pressure The Reactor Building Operator is swapping CRD suction filters CRD ACCUM TROUBLE Overhead Annuciator C6-D4 is clear For the two minutes following the CRD pump trip, what will be the response of HCU Accumulator Gas pressure? (Assume NO other operator actions) HCU Accumulator Gas pressure A. stays the same because reactor pressure holds the charging water check valve closed. B. stays the same because cooling water pressure holds the charging water check valve closed. C. lowers because the accumulator piston moves when charging water header pressure is lost. D. lowers because the cooling water pressure lowers when charging water header pressure is lost. Proposed Answer: C Explanation (Optional):

HC.OP-IS.BF-0103 Charging water check valve 1 15 maintains water volume on a loss of charging pressure from the CRD pumps initially. However, Accumulator gas pressures will begin to lower immediately after pump trip depending on the leak rate of the check valves. Actual plant experience demonstrated that the first alarm comes in at 2.1 minutes. N2 gas pressure will remain the same as long as the check valve holds.

When the check valve begins to leak, the piston will stroke and N2 pressure will drop causing low accumulator pressure alarm. C. Correct. Page 122 of 205 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet A. B. D. Incorrect. Lowers due to check valve leak by. Also reactor pressure does not hold the check valve closed Incorrect. Lowers due to check valve leak by. Also water pressure does not hold the check valve closed.

Incorrect. The pressure lowers due to check valve leak by. Technical Reference( s) H C. OP- IS. B F-0 1 03 (Attach if not previously provided) Proposed references to be provided to applicants during examination: None Learning Objective:

CRDHYDEOI 7 (As available) Question Source:

Question History: Question Cognitive Level:

Bank # 6891 4 Modified Bank # -. (Note changes or attach parent)

New NRC2002 -. Last NRC Exam -. Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 123 of 205 HC.OP-IS.BF-O103(Q)

@'URPOSE u The purpose of this procedure is to demonstrate during plant refuel, the operability of the Control Rod Drive Accumulator Charging Water Check Valve, 1-BF-VI 15, as required by Technical Specification 4.0.5. This is performed by verifying each individual accumulator check valve maintains the associated accumulator pressure above the alarm setpoint for greater than or equal to 2 minutes, starting at normal operating pressure, with no Control Rod Drive Pump operating.

[Ct)-270A]

2.0 PREREQUISITES

2.1 Charging

Water Check Valve Exercise Test

2.1.1. Permission

to perform this test has been obtained from the SM/CRS as indicated by the completion of Attachment 1, Section 1 .o. 2.1.2. All personnel involved in the performance of this procedure, should complete Attachment 1, Section 3.0, prior to performing any part of this procedure.

2.1.3. No other testing OR maintenance is in progress that would adversely effect the performance of this test. NOTE All Control Rod Drive Scram Accumulators need not be OPERABLE to perform this test provided INOPERABLE accumulators are tracked IAW OP-HC-108-115-1001, Removal and Return of Equipment to Service, and this surveillance is listed in Part B of Attachment 1, Action Statement Log Sheet as required to restore the equipment to operability.

2.1.4. The Control Rod Drive system is in service AND all OPERABLE Hydraulic Control Units are charged to normal operating pressure IAW HC.OP-SO.BF-0001 (a), CRD Hydraulic System Operation.

2.1 5. All insertable control rods are inserted except for rods removed IAW Technical Specifications 3.9. IO. 1 and/or 3.9.10.2.

2.1.6. The plant is in Condition 3, 4 or 5. 2.1.7. Radiation Protection should be contacted prior to performing venting - OR draining in this procedure.

The individuals performing the venting OR draining shall obtain instructions - OR Radiation Protection Supervisor. approval from the Radiation Protection Shift Technician Hope Creek Page 2 of 21 Rev. 3 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level Tier # Group # KIA # RO SRO 1 2 295032 EK3.02 Importance Rating

3.6 Knowledge

of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Reactor SCRAM. Proposed Question: Common 61 Given the following conditions:

0 An unisolable steam line leak has occurred in the RClC room RClC Equipment Room Area Temperature is 207°F and rising Which of the following is the reason for initiating a Reactor Scram with the above conditions?

A. Emergency Depressurization is anticipated.

B. The scram will begin to reduce the energy that the RPV will discharge to the RClC room. C. A scram will reduce the driving head and flow through the break in the RClC room to prevent the blowout panel from opening.

D. Failure of Secondary Containment due to high temperatures must be assumed and the scram will stop the radioactive re#ease. Proposed Answer:

Explanation (Optional): From EOP-103 bases:

B. If temperatures or floor aels in any one o the ROOMS listel in Table 1 or 2 o Reactor Building Control approach their maximum safe operating value, adequate core cooling, containment integrity, safety of personnel, or continued operability of equipment required to perform EOP actions can no longer be assured. EOP-101 must be entered to make certain the reactor is scrammed.

Scramming the reactor reduces to decay heat levels the energy that the RPV may be discharging to the reactor building.

Page 124 of 205 I , I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet A. B. Correct C. D. Incorrect - the scram is not performed for anticipating an ED. Levels may not reach the ED requirement Incorrect. - The blowout panel is not a concern per the bases. Incorrect - The failure of Secondary Containment is not a concern at this point in the event with the conditions stated. Technical Reference(s) EOP-103 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

EOPI 03E006 (As available) Question Source:

Question History: Question Cognitive Level: 10 CFR Part 55 Content: Bank # Modified Bank # -. (Note changes or attach parent)

X -. New Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X ~- Comments:

Page 125 of 205 HC .OP-EO.=-0 1 03/4 system is the source of radioactivity, the action that is directed in terminate any further increase in reactor building radiation levels BASES RB-14 should be adequate to If temperatures or floor levels in any one of the ROOMS listed in Table 1 or 2 of Reactor Building Control approach their maximum safe operating value, adequate core cooling, containment integrity, safety of personnel, or continued operability of equipment required to perform EOP actions can no longer be assured. EOP-101 mus't be entered to make certain the reactor is scrammed. Scramming the reactor reduces to decay heat levels the energy that the RPV may be discharging to the reactor building. However, an explicit direction to scram the reactor is not provided in this step. Rather, entry (or re-entry) to EOP-101 is required. This accomrnodates the parallel execution of the reactor power control steps of RPV Control. if needed, and at the same time avoids unnecessarily cycling the control rod drive hydraulic system. RB-18 and I9 WHEN the same parameter exceeds Max Safe Op Limit in 2 or more areas, EMERGENCY RPV DEPRESSURIZATION IS REQUIRED.

Discussion Should reactor building temperatures radiation levels or floor levels exceed their maximum safe operating values in more than one area, the RPV must be depressurized to preclude further temperature, increases.

RPV depressurization places the primary system in its lowest possible energy state, rejects heat to the suppression pool in preference to outside the containment, and reduces the driving head and flow of primary systems that are misolated and discharging into the reactor building.

The criteria of "more than one area" specified in this step identifies the rise in reactor building temperature, reactor building radiation level or reactor building water level as a wide-spread problem which may pose a direct and immediate threat to reactor building integrity, equipment located in the reactor building, and continued safe operation of the plant.

One parameter (e.g., temperature) above its maximum safe operating value in one area [example "A & "C" Core Spray rooms] and a different parameter (e.g., radiation or water level) above its maximum safe operating value in the same or another area [example "A" and "C" RHR rooms] is not a condition which requires emergency RPV depressurization.

A combination of parameters exceeding maximum safe operating values in one area does not necessarily indicate that control of a given parameter cannot be maintaineal or that previous actions have not been effective in confining the trouble to one area. Expanding the application of "more than one area" to encompass multiple parameters might lead to depressurization of the RPV when such action is not, in fact, appropriate or needed. Hope Creek Page 10 of 17 Rev. I C:\Documenk and Settings\sdennis\My Documenk\Hope Creek 2008-2009\Refemnces\LESSON PLANS\Emergency Operating Procedures\EOP Bases\MasteflLesson Plans\EOP1034bases01 .doc I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 295029 EA1.01 Group # 2 I m portance Rating

3.4 Ability

to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: HPCI: Plant-Specific. Proposed Question: Common 62 HPCI and RCIC both started and are injecting in response to a valid low reactor water level. Current plant conditions are as follows: 0 0 0 Reactor water level is +25 inches, steady Reactor pressure is 845 psig, rising slowly Drywell pressure is 1.1 psig, steady RClC has been aligned to Full Flow Recirc operation (CST to CST) for pressure control HPCl is injecting to the reactor for level control After 10 minutes of operation, suppression pool level reaches 78.5" Which of the following would be the response of HPCI & RClC for the given conditions?

A. HPCl will continue to inject and RClC will operate on minimum flow. B. HPCl will continue to inject and RClC will trip on low suction pressure.

C. HPCl will trip on low suction pressure and RClC will operate on minimum flow. D. HPCl will trip on low suction pressure and RCX will trip on low suction pressure. Proposed Answer:

A. A. Correct - The FOI 1 closes on the HPCl Suppression Pool Suction Valve (F042) opening.

HPCl will continue to inject, RClC has no discharge path, Min. flow opens. B. C. D. Incorrect - RClC Suction flow path will remain on the CST. Incorrect - HPCl will continue to inject, AP-HV-FO1 1 closes in the return line to the CST. Incorrect - HPCl will continue to inject, RClC has no discharge path, Min. flow opens. Page 126 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s) HC.OP-SO.BJ-0001 (Qh Sect (Attach if not previously provided)

3.3 Interlocks

Proposed references to be provided to applicants during examination:

None Learning Objective: (As available)

HPCIOOE012

-. Question Source:

Bank # X Modified Bank

1. New (Note changes or attach parent) Question History: Last NRC Exam Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments: Page 127 of 205 HC.OY-SO.BJ-OOOl(Q) 3.3.4 (continued) e BJ-HV-8278 (HV-F105) PMP IISCH TO FW ISLN MOV Auto closes E FD-HV-F001 HPCI TURB STM SPLY is full closed (K57) AND

• LINE FILL keyswitch is in NORM (K100) (AND Test Jack installed) (K78) AND

• LINE FILL keyswitch is in NORM (K 100) AND

• LINE FILL keyswitch is in NORM (KIOO). Auto opens on HPCI Initiation (K80) AND FD-FV-4880 HPCI TURB STOP VLV is not full closed (K79) AND FD-HV-F001 HPCI TURB STM SPLY not full closed (K57) AND HV-F006/HV-8278 not in test (K78). Test Selector switch is in HV-F006/HV-8278 (HV-FI

05) FD-FV-4880 HPCI TURB STOP is full closed (K79) ___ 0 LINE FILL keyswitch interlock. requires Reactor steam pressure

< 100 psig on PT E41-NO58A 8; N058E (K16A & K17A) AND keyswitch in BYPASS to defeat auto closure. - 0 BJ-HV-F008 TEST BYP TO CST MOV Auto closes on HPCI Initiation (K34) Q& BJ-HV-F(l06 PMP DSCH TO CS ISLN not full closed (limit switch) BJ-HV-8278 (HV-F105)

PMP DSCH TO FW ISLN MOV not full closed (limit switch). Opens manually E no HPCI Initiation (K34) present A,ND both BJ-HV-8278 (HV-F105)

CS ISLN are full closed (limit switches).

PMP DSCH TO FW ISLN MOV AND BJ-HV-F006 PMP DSCH TO e AP-HV-FOl 1 HPCI/RCIC RET. TO CST Auto closes BD-HV-FO3 1 RCIC OR HPCI Initiation Initiation (K34) AND BJ-HV-IF042 PMP SUCT FROM SUPP CHB is not full open.

VLV is fd1 open (K76) PMP SUCT FROM -7 SYPP CHB is full - Rev. 34 Continued on next page Hope Creek Page 13 of57 I I HCOP-SO.BJ-OOOl(Q)

PCI valve interlocks are as follows: 0 FD-HV-FOO2 INBD HPCl STM lSLN MOV Auto Closes on HPCI Div 3 Isolation signal (K50C). Opens manually if no HPCI Div 3 Isolation signal (K5OC). IF; Isolation signal was sealed in (K92) with FD-HV-F002 handswitch in OPEN THEN handswitch must be placed in CLOSE, THEN OPEN. 0 FD-HV-F003 OUTB EIPCI STM.ISLN MOV Auto closes on HPCI Div 1 Isolation signal (K50A). Opens manually if no HPCI Div 1 Isolation signal (K50A). E isolation signal was sealed in (K93) with FD-HV-F003 handswitch in OPEN THEN handswitch must be placed in CLOSE THEN OPEN. e FD-HV-FOO L HPCI TURB STM SPLY Auto opens on WCI Initiation (K33), FD-HV-F071 TURB EXH ISLN is full open. Opens manually E FD-HV-F071 TIIRE3 EX13 ISLN is full open.

0 FD-HV-FI 00 HPCI WAJ VLV Auto closes on HPCi Div 3 Isolation signal (K50C Jz K52C). Opens nianually no HPCI Div 3 Isolation signal (KSOC). B3-HV-F004 PMP SLTCT FROM= Auto closes BJ-HV-FO42 PMP SUCT FROM SUPP CHB is full open (K43). Auto opens BJ-HV-F042 PMF SUCT FROM SUPP CHI3 is not full open (K43) AND WCI Initiation (K33).

0 BJ-HV-F042 PMP SUCT FROM SUPP CHB Auto closes on HPCI Div I Isolation signal (K51A). Auto opens on CST low level Suppression Chamber high level (K42), E BJ-HV-FO42 handswitch is not in AUTO OPEN OVRD AND no HPCI Div 1 Isolation signal (K5 1 A). Opens manually no HPCI Div 1 Isolation signal (K5 IA). 0 BJ-HV-F007 PMP DSCH ISLN Auto closes Test Selector switch is in HV-F007 position AND Test Jack installed (K66). Auto opens on HPCI Initiation o(32) AND not in test (K66). BJ-HV-F006 PMP DSCH TO CS ISLN Auto closes E FD-IFV-FOOl HPCI TURB STM SPLY is full closed (K44) (AND Test Jack installed) (K65) - OR FD-FV4880 HPCI TLRl3 STOP VLV is full closed. Auto opens on HPCI Initiation (K90)

AND FD-FV4880 HPCI TURB STOP VLV is not full closed (K39) AND FD-HV-F001 HPCI TW3 STM SPLY is not full closed (K44) AND not in test (K65). Test Selector switch is in W-F006/HV-8278 (HV-F 105) Continued on next page Hope Creek Page 12 of 57 I I PSEG Internal Use Onlv SUBSEOUENT OPERATOR ACTION CONDITION L A. Unexpected rise in Drywell Pressure.

DatelTime

B. Turbine Bldg. Chill Water System is lost to the Drywell. Date/Time:

Hope Creek HCOP-AB-CONT-0001 (Q) DRYWELL PRESSURE n A.1 TERMINATE Containment Makeup AND , -T hierting.

/ A.2 )MAXIMIZE Drywell Cooling by ENSURING:

Cl a AI1 Drywell Fan Cooling Coils are Open. 0 a All Drywell Fans are running in Fast Speed. n I **NOTE 1** I 0 a Proper TBCW system operation 0 a Check Reactor Recirc. Pump Seals.

0 a Check SRV Tailpipe Temperatures.

U a Drywell Leakage Source Detection A.3 PERFORM the following:

IAW GP.ZZ-0005.

0 B. 1 7 3 7 3 1 1 1 1 1 1 7k CAUTION1 & ALIGN RACS to the Chill Water System for Drywell Cooling as follows: a, ENSURE RACS to the out of service Off- Gas Train is ISOLATED as follows: e the Common Off-Gas Train is in service, 0 E Unit 1 Off-Gas Train is in service, THEN CLOSE HV-2577. THEN CLOSE HY-77 12A1. b. CLOSE HV-9532-1 AND HV-9532-2.

c I PRESS LOOP A SPLYRTN OPEN RACS PB. d. PRESS LOOP B SPLYRTN OPEN RACS PB. e OBSERVE the following indications:

HV-953OAl!A3 CLOSED HV-9530Blh33 CLOSED HV-953OA2lA4 OPEN HV-9530B2434 OPEN f. OPEN HV-9532-1 AND HV-9532-2.

~ Page 7 of 116 Rev. I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 29501 2 AA2.02 Group # 2 Importance Rating 3.9 ~~ Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: Drywell pressure. Proposed Question: Common 63 Given the following conditions:

0 0 0 0 The plant is operating at 100% power. The COMPUTER PT IN ALARM A4-F5 alarm is received.

Drywell pressure is 1.1 psig.

HC.OP-AB.CONT-0001, Drywell Pressure abnormal is entered. Which of the following events, BY ITSELF, could be the cause of the pressure rise? A. Failure of the "A" Reactor Recirculation Pump #2 Seal B. FV-4971 Nitrogen Flow Control Valve fails open C. Torus Vent Valve Isolation Valve HV-11541 fails open D. Loss of power to multiple Drywell Fans Proposed Answer: D. Explanation (Optional):

D. Correct: A reduction in cooling will raise temperature and therefore pressure in the Drywell. A. B. C. Incorrect:

Assuming the #I seal is intact, no change in DW conditions will occur.

Incorrect: During normal operation, the nitrogen FCV is isolated from the DW. Incorrect: A vent valve opening would result in a reduction in pressure.

However the rupture disk downstream of HV-11541 should be intact resulting in no effect on DW pressure. Technical Reference( s) HC. OP-AB.

CONT-000 1 (Attach if not previously provided) ~~ ~~ Proposed references to be provided to applicants during examination:

none Page 128 of 205 ES-401 Sample Written Examination Form ES-40 1-5 Question Worksheet Learning Objective:

ABCNTI E004 (As available) 61761 ~. Question Source: Bank

1. Modified Bank # New (Note changes or attach parent) ___. ___. Question History: 2007 Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 129 of 205 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 295009 G2.1.20 Importance Rating

4.6 Group

1. 2 Conduct of Operations: Ability to interpret and execute procedure steps. (Low reactor water level) Proposed Question: Common 64 Given the following conditions:

0 A startup following a refueling outage was in progress when a loss of offsite AC Power occurred.

Only A, C, and D Emergency Diesel Generators are running. HPCl and RClC are NOT available.

All control rods are at 00. RPV water level is stable at (-35) inches. RPV pressure is stable at 910 psig. NO operator actions have been taken. 0 0 0 0 0 0 Which of the following statements describes the actions required for the conditions above? A. Restore and maintain level to +12.5 to +54 inches by maximizing CRD flow. B. Lower reactor pressure to 600 psig and restore level using the Secondary Condensate Pumps. C. Emergency Depressurize the reactor and restore level using the low pressure ECCS systems. D. Override 1 E Breakers and restore RFPTs to raise RPV level to between +I 2.5 and + 54 inches. Proposed Answer: A Explanation (Optional):

A. Correct - No ATWS exists and Level is not lowering.

A, C and D diesels allows use of RACS and 2 CRD pumps.

CRD is a Preferred Table 1 system 0-1500 psig. B. C. D. Incorrect - This would be the normal method, but condensate has lost power. Incorrect - ED not required.

RPV level still too high.

Incorrect - No power for secondary and primary condensate pumps Page 130 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s) HC.OP-EO.ZZ-0101, Steps (Attach if not previously provided)

RCIL-2 thru RC/L-5 Proposed references to be provided to applicants during examination:

None Learning Objective:

E01 01 LE006 (As available) Question Source:

Bank # ID: Q76668 Modified Bank # New (Note changes or attach parent) Question History: Last NRC Exam Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments: Page 131 of 205 I I I ____. ES-40 1 Sample Written Examination Form ES-401-5 Quest ion Wo r k.s hee t Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 29501 3 AK3.02 Importance Rating

3.6 Group

1. 2 Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Limiting heat additions. Proposed Question:

Common 65 The plant is operating at 80% power with RClC quarterly testing in progress. Which one of the following is the reason for the Technical Specification temperature limitation for the suppression pool during the test?

A. B. C. D. To assure primary containment integrity following a stuck open Safety Relief Valve. To assure that excessive steam condensing loading does NOT occur during the test. To assure that sufficient RHR and Core Spray NPSH exists following a containment blowdown in conjunction wit ti containment overpressure.

To assure sufficient RHR and Core Spray NPSH exists during LOCA conditions without overpressure.

Proposed Answer: D. Explanation (Optional):

D. Correct - TS requires any testing that adds heat to the SP be terminated at 105°F the TS bases is PC integrity during a LOCA and NPSH without overpressure for RHR and Core Spray. A. B. C. Incorrect - the bases is for a LOCA. Incorrect - not a concern during the test Incorrect - NPSH without overpressure for RHF! and Core Spray. Technical Reference(s)

T.S. 3.6.2 Bases (Attach if not previously provided) Page 132 of 205 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

None Learning Objective: (As available)

-. Question Source:

Bank # Modified Bank # -. (Note changes or attach parent)

New X -. Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 133 of 205 I fl" CONTAINMENT SYSTEMS DEPRESSURIZATION SYSTEMS (Continued 1 - Under full power operating conditions, blowdown from an initial suppressi~on chamber water temperature of 95°F resul.ts in a water temperature of approximately 135°F immediately f ollow.ing blowdown which is below the ZOOOF used for complete condensation via mitered T-quencher devices. At this temperature and atmospheric pressure, thc available NPSH exceeds that containment over ressure durin the accic- . If both RHR thus there is no dependency on oops are 1 LvilL.uL~pn cn&a+-L nt /overpressure for post-LOCA operations.

/ Experimental data indicates that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained below 200°F during any period of relief v,iilve operation. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

Because of the large volume anc therinal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.

Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest streiss. In addition to the limits on temperat:ure of the suppression chamber pool water, operating procedures define the ac-i:ion to be taken in the event a safety-relief valve inadvertently opens 0.1: sticks open.

As a minimum this action shall include:

(1) use of all avail-able means to close the valve, (2) initiate suppression pool water cooling, 113) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from th3t of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool. In conjunction with the Mark I containment Long Term Program, a plant unique analysis was performed which demonstrated that the containment, the attached piping and internal structures meet the applicable structural and mechanical acceptance criteria for Hope Creek. The evaluation followed t:he design basis loads defined in the Mark I Load Definition Report, NEDO-21888, December 1978, as modified by NRC SER NUREG 0661, July 1980 and Supplement 1, August 1982, to ensure that hydrodynamic

.oads, appropriate for the life of the plant, were applied.

I HOPE CREEK H 3/4 6-4a Amendment No. 128 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 Group # 1 KIA # G2.1.30 Importance Rating 4.4 -~ Ability to locate and operate components, including local controls. Proposed Question: Common 66 Given the following conditions:

RPV level is +IO" Control Room has been abandoned Control has been transferred to the Remote Shutdown Panel (RSP) RPV pressure is 80 psig Which one of the following describes the ability to operate the BC-HV-FOO9 SDC Suction Isolation valve and whether the valve would close automatically if reactor pressure exceeded 82 psig? A. The valve CAN be opened at the RSP. If already open, the valve would automatically close.

B. The valve CAN be opened at the RSP. If already open, the valve would NOT automatically close.

C. The valve CANNOT be opened at the RSP. If already open, the valve would NOT automatically close.

D. The valve CANNOT be opened at the RSP. If already open, the valve would automatically close.

Proposed Answer: B Explanation (Optional): Note and a Caution in HC.OP-IO.ZZ-0008 (5.9.6 of Rev 28) - CAUTION WHEN the RSP Transfer Switch is placed in EMER, RHR S/D Cooling interlocks for overpressure AND low Reactor level are inoperable.

RX pressure of 80 psig should NOT be exceeded WITH Suction Valves F008

& FOO9 open. B. Correct. When control is transferred to the RSP, both the Low RPV Water Level AND high RPV Pressure isolations for the BC-HV-FOO9 are defeated. There remains a pressure switch permissive in series with the opening contactor that requires reactor pressure to be below 82 psig to open the valve (this is NOT a function of NSSSS). This is identified in a Page 134 of 205 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Note and a Caution in HC.OP-IO.ZZ-0008 (5.9.6 of Rev 23). A. C. D. Incorrect - The valve will NOT isolate if reactor pressure exceeds 82 psig. Incorrect. The valve CAN be opened, since the RPV Low Water Level isolation is defeated and there is NO Low Water Level opening permissive.

Incorrect.

The valve can be opened.

The valve will NOT isolate if reactor pressure exceeds 82 psig. Technical Reference( s) H C . OP-l0.ZZ-0008 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective:

Question Source: IOP008E006 (As available)

Bank # 5391 4 -. Modified Bank # -. (Note changes or attach parent)

New -. Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 X Comments:

Page 135 of 205 HC.OP-IO.ZZ-0008( Qi 5.9.5. BEFORE Reactor Pressure decreases to < 660 psig, THEN VERIFY all Primary ANB Secondary Condensate Pumps are secured. (Local)

-_ NOTE! - IF the A RHR Loop is in Suppression Pool Cooling: 0 The A RHR Loop should be maintained in operation while the B RHR Loop is placed in Shutdown Cooling operation.

0 - IF time AND personnel are available, the B RMR Loop can be drained and filled with CST water (locally) prior to reaching 65 psig to 80 psig. High Reactor pressure will prevent opening of the shutdown cooling valves but will NOT isolate the valve E pressure rises above setpoint.

-- i i WHEN the RSP Transfer Switch is placed in EMER, RHR S/D Cooling interlocks for overpressure-ASCD low Reactor level are inoperable.

RX pressure of 80 psig should NOT be exceeded WITH Suction Valves F008 & FUO9 open. 5.9.6. At < 80 psig, PERFORM the following:

[CD-987X, CD-370XI A. E B RHR Loop is operating in Suppression Pool Cooling, THEN REMOVE it from operation per Step 5.8.2.

B. a LOP has occurred AND continued cooldown is required INITIATE alternate S/D Cooling IAW Att.

IO. [PR 981 2281 741 -- PLACE B RHR in Shutdown Cooling Operation IAW Attachment

4. -- IF DESIRED to place "A RHR Loop in Shutdown Cooling, PLACE A RHR in Shutdown Cooling Operation IAW Attachment

11. C. D. 5.9.7. MAINTAIN a cooldown rate 90°F/hr by throttling open/closed HV-F048B RHR HX B SHELL- SIDE BYP MOV. -- Hope Creek Page 22 of 85 Rev. 28 I I ES-40 1 Sample Written Examination Question Worksheet Form ES-401-5 Examination Outline Cross-reference:

Level RO SRO ~~ Tier # 3 Group # 1 KIA # G2.1.4 Importance Rating

3.3 Knowledge

of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no-solo" operation, maintenance of active license status, 10CFR55, etc. Proposed Question: Common 67 You are a licensed Reactor Operator.

Due to illness, you have worked the following schedule over the past quarter (July thru September). July 1 - Off July 2 - Off July 3 - 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> day shift as RO July 4 -12 hour day shift as RO July 8 -12 hour night shift as RO July 9 - 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> night shift as RO July 10 Through September 30 - Off Shift due to illness. All licensed operator training is up to date. You're received medical clearance to stand watch. Which one of the following describes the status of your license and additional requirements, if any, to stand watch on October 1 st IAW OP-AA-105-'102 "NRC ACTIVE LICENSE MA1 NTENANCE?

A. Your license is Active because you stood watch for at least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> the previous quarter, no additional requirements are needed to stand watch on 10/1. B. Your license is Inactive.

You must reactivate

your license by performing shift functions under the sole direct supervision of an active licensed RO or SRO for at least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. C. Your license is Inactive. You must reactivate your license by performing shift functions under the sole direct supervision of ONLY an active licensed SRO for at least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. D. Your license is Inactive.

You must reactivate your license by performing shift functions under the sole direct supervision of ONLY an active licensed RO for one additional 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift. Page 136 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed Answer:

B Explanation (Optional):

Explanation (Optional):

IAW OP-AA-105-102 "NRC ACTIVE LICENSE MAINTENANCE, Steps 4.1.1. &4.2.1 MAINTAIN an active license by actively performing the functions of RO, SRO, or LSRO. 1. RO licenses by performing the duties of the Unit RO and/or Unit Assist RO for a minimum of seven 8-hour or five 12-hour shifts per calendar quarter, including turnover to the next shift.

REACTIVATE an RO or SRO license to an "active status" by performing 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of shift functions in the presence and under the sole direct supervision of an active RO or SRO, as appropriate and in the position to which the individual will be assigned.

A. B. Correct.

C. D. Incorrect. License is inactive.

Previous quarter requirements not met Incorrect.

An RO is required Incorrect.

Previous quarter requirements not met with one additional shift on 1011. One day too late. Technical Reference(s) OP-AA-105-102 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

NOH04ADM062C-0 1 (As available)

Question Source: Bank # Modified Bank # -. (Note changes or attach parent)

New X Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 137 of 205 I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Page 138 of 205

4.2. License

Reactivation OP-AA-I 05-1 02 Revision 9 Page 4 of 7 NOTE: If more than 3 months have passed between the time of NRC examination results being issued and issuance of the NRC license, then the license must be activated in accordance with the requirements of this procedure and 1 OCFR55.53(f).

4.2.1. REACTIVATE

an RO or SRO license to "active statu efiorming 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of shift functions in the presence and under the sole di -7 a- s appropriate andtin the position to which the individual will be assigned.

The 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> will include completion of a position-specific activation guide containing specific and detailed activation requirements (if required), a plant tour in the presence and under the sole direct supervision of an active RO LSRO's (or SRO licenses that will be activated for fuel handling only) need to complete only one 8-hour shift under the sole direction of an active SRO or LSRO. All parts of the reactivation for LSRO, or SRO licenses activated for fuel handling only, will be performed with the accompaniment of an active SRO or LSRO. L .____ or-SRO;asappropriate;anda-review-~f-~~e-pasr~eedufe~- - 4.2.2. DOCUMENT the reactivation on Attachment 2, Reactivation of License Log . I. 2. The Shift Manager shall signify that the required OJT hours were Completed.

The Operations Training Manager shall signify that the position-specific activation guide was completed (if required).

NOTIFY the Security Shift Supervisor to update the Main Control Room SRO tracking program, if appropriate.

3. . 4. If an SRO license is reactivated as LSRO, then RESTRICT LSRO from performance of the SRO duties. 4.3. NOTIFY the Shift Operations Superintendent in the event an individual fails to meet the Active License requirements defined in Section 4. I. REMOVE the individual from the "Active License List". NOTIFY the Security Shift Supervisor to remove the individual from the Main Control Room SRO Tracking Program, if appropriate.

4.3.1. 4.3.2. . -. - . . .- ... -. . . .. . . -

ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 Group # 2 K/A # G2.2.6 Importance Rating 3.0 -~ Knowledge of the process for making changes to procedures Proposed Question: Common 68 Given the following conditions:

0 The plant is operating at rated power 0 The Common Offgas train is experiencing problems and must be swapped before vacuum starts degrading. While performing the evolution brief, a critical procedure step was found to be missing. Which of the following describes the requirement, if any, to continue the evolution?

A. Complete the evolution as written then perform a permanent revision change after the evolution is complete.

B. A procedure change request is required and an on-the-spot-change can be made.

C. Obtain verbal concurrence from the CRS to change the sequence of steps and continue.

D. A full procedure revision to the Offgas system operating procedure is required. An on- the-spot change CANNOT be performed. Proposed Answer:

B Explanation (Optional): IAW AD-AA-101-101 B. Correct. An on-the-spot-change (OTSC) may used.

A. C. D. Incorrect.

An on-the-spot change may be used Incorrect.

If an error is found in the procedure actions must be taken to correct the issue before proceeding. Incorrect. Written documentation is required (On-the-spot change) Technical Reference(s) AD-AA-101-101 (Attach if not previously provided)

Page 139 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

none Learning Objective: ADMPROE002 (As available)

-. Question Source: Bank

1. Modified Bank # -. (Note changes or attach parent)

New X __. Question History: Last NRC Exam

-. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 140 01 205 AD-AA-I 01 -1 01 Revision 1 Page 1 of 11 Level 2 - Reference Use PSEG IMPLEMENTING PROCEDURE ON THE SPOT CHANGE (OTSC) PROCESS 1. PURPOSE 1.1. This procedure provides instructions for performing On-The-Spot-Changes (OTSCs) to PSEG Nuclear's Department Implementing Procedures (DIPS). OTSCs are allowed for PSEG Station Implementing Procedures ONLY.

On-The-Spot-Change (OTSC) process inay be used to revise a procedure when the change is of an urgent nature support resolution of condition in a timely manner. routine procedure revision process will not 1.3. OTSC process is subject to restrictions listed in 1.3.1 and 1.3.2, for conditions found during pre-job review, or during field or ishop use of a DIP in conjunction with a work activity that would require a procedure revision be made before beginning activity or prior to continuing activity because procedure cannot be performed as written (cannot use procedure as-is). 1.3.1. Permitted Uses - OTSC allowed to modify procedure for temporary conditions which would not require a permanent change to procedure. - OTSC allowed to modify procedure in accordance with an approved calculation, engineering procedure or engineering evaluation that does not change intent of procedure as described in Attachment 1, Change of Intent Criteria. - OTSC to Q, F and R designated procedure is subject to same level of review and approval as required for procedure being revised. 1.3.2. Prohibited Uses

[80073759] - OTSC shall not be used for Administrative Procedures. - OTSC shall not change intent of procedure as described in Attachment 1, Change of Intent Criteria. - OTSC shall not be used to change procedures: - - For rewording or clarification of statements based on personnel preferences For reordering of steps, except when failure to do so will prevent task corn pletion I I i ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 KIA # G2.2.12 Group # 2 Importance Rating

3.7 Knowledge

of surveillance procedures Proposed Question: Common 69 The CRS has directed you to perform a RClC inservice test following maintenance using HC.OP-IS.BD-0001, Reactor Core Isolation Cooling (RCIC) Pump - OP203 - lnservice Test. Which one of the following describes a surveillance procedure requirement due to the performance of the test?

A. Suppression Pool Temperature Monitoring prior to starting the RClC pump. B. Suppression Pool Cooling when Suppression Pool temperature exceeds 95 degrees F. C. RClC must be secured when Suppression Pool temperature reaches 110 degrees F. D. Remote Shutdown System Suppression Pool Temperature Instrumentation Channel Check. Proposed Answer: A Explanation (Optional):

A. Correct B. C. D. Incorrect - SPC is required prior to placing the pump in service Incorrect - RClC must be secured when temperature reaches 105 dgrees F. Incorrect - not required by the procedure. Technical Reference(s)

HC.OP-IS.BD-0001 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective: (As available) Question Source: Bank

1. INPO 191 32 Page 141 of 205 ES-401 Sample Written Examination Form ES-401-5 Question ~. Worksheet Question Question History: Cognitive Level:

10 CFR Part 55 Content: Comments: (Note changes or attach parent)

-. Modified Bank # New Last NRC Exam Memory or Fundamental Knowledge X Comprehension or Anal,ysis 55.41 X Page 142 of 205 HC.OP-IS.BD-0001 (a) NOTE Installation in the following step must be performed by qualified personnel.

5.12 DIRECT qualified personnel to perform the following:

5.12.1. 5.12.2. INSTALL the Measuring and Test Equipment (M&TE), specified I ENTER test equipment data on Attachment 3, lnplant Data Sheet. in Section 4.0, IAW Attachment 4, M&TE Installation and Removal. 5.12.3. INSTALL Key Phasor Tach. at Turbine coupling end I AND POSITION RPM indicator near page for communication with the Control Room. INITIAL Attachment 3, lnplant Data Sheet performerherifier.

5.12.4. -I 5.13 GETARS is to be used, REQUEST STA or System Engineer START GETARS. 5.14 PERFORM pre-start checks and alignment as follows: I 5.14.1. RECORD RClC Pump lubrication level on Attachment 3. -I NOTE 7 ' The following step requires establishment of communication between the Control Room and operator stationed locally at valve to observe movement.

5.14.2. WHILE verifying approximately 11 - 13 seconds of valve I movement locally THROTTLE OPEN HV-F022 RClC TEST BYP TO CST ISLN VLV approximately 11 - 13 seconds (Control Room)

AND INITIAL Attachment 3.

NOTE Tech Spec 3.6.2.1 addresses the Suppression Pool Average Water temperature requirements.

5.1 5 PERFORM Attachment 3m, Containment Systems TS 4.6.2.1 .b. 1, of I HC. OP-DL.ZZ-0026( Q), Surveillance Log.

[TS 3.6.2.1 ] PIr Is ~fi~~~~~ MCJ&\V/\.\/JL Hope Creek Page 11 of 46 Rev. 43 H .OP-l 5.16 E the following Suppression Pool Average Water Temperature(s) are 9 reached during the performance of this test the indicated actions shall be taken: 5.16.1. 95"F, ENSURE RHR is in the Suppression Pool Cooling mode. (Entry into HC.OP-EO.ZZ-O102(Q) is not required at this time).

5.16.2. 105°F; IMMEDIATELY TERMINATE this test AND STOP all testing that adds heat to the Suppression Pool.

5 16.3. E Suppression Pool Average VVater Temperature can not be maintained below 105°F THEN ENTER HC.OP-EO.ZZ-O102(Q). Tech Spec 3.6.2.1 and

3.5.3 address

the Suppression Pool Water Level requirements.

A 5.1 7 MAINTAIN Suppression Pool level at 75-77 inches during the performance of this test using either one of the following procedures as appropriate:

0 HC.OP-SO.EE-0001 (Q), Torus Water Cleanup System Operation.

0 HC.OP-SO.BC-0001 (Q), RHR System Operation 5.18 E any of the following occur during operation of RClC THEN TRIP the Turbine immediately:

0 Excessive vibration 0 Excessive oil temperature 0 Sudden drop in oil pressure 0 Other unusual operating conditions are noted. 0 Pump discharge pressure reaches the indicated maximum range of 1500 psig. Hope Creek Page 12 of 46 Rev. 43 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 Group # 3 WA # G2.3.4 Importance Rating

3.2 Knowledge

of radiation exposure limits under normal or emergency conditions. Proposed Question: Common 70 A Hope Creek operator has received the following dose: 0 November 1, 2008 thru November 21, 2008 - 350 mrem - while visiting a foreign nuclear plant as part of a Technical Exchange Program.

0 July 1, 2008 thru December 31, 2008 - 175 nirem - while working at Hope Creek.

0 January 1, 2009 thru January 31, 2009 - 125 mrem while working at Hope Creek.

Which of the following describes the MAXIMUM additional non-emergency Total Effective Dose Equivalent (TEDE) that this individual could receive at Hope Creek through October 31, 2009? A. 1350 mrem B. 1700 mrem C. 1875 mrem D. 2375mrem Proposed Answer:

C Explanation (Optional):

IAW C. Correct - 2000 - 125 = 1875 A. B. D. Incorrect. - prior year is not included Incorrect. - prior year is not included Incorrect. - limit is 2000 not 2500 per year Technical Reference(s) RP-AA-203 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Page 143 of 205 ES-40 1 Sam p I e Written Exam i nation Question Worksheet Form ES-40 1-5 Learning Objective:

NOH04ADM024C-01 (As available)

ES-40 1 Sam p I e Written Exam i nation Question Worksheet Form ES-40 1-5 Learning Objective:

NOH04ADM024C-01 (As available)

Question Source: -. Bank # Modified Bank # -. 77351 (Note changes or attach parent)

New Question History: Question Cognitive Level:

10 CFR Part 55 Content: Last NRC Exam -. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 X Comments:

Page 144 of 205 RP-AA-203 Revision 3 Page 3 of 10 NOTE: Any request to raise the administrative dose control level for a minor shall be approved and documented by the Radiation Protection Manager. Administrative dose control levels have been established for Total Effective Dose Equivalent Limits as follows: - - 2000 mrem routine cumulative TEDE/yr.

100 mrem TEDE for minors.

4.1.3. The Radiation Protection Manger shall review an individual's occupational exposure when the dose equivalent reaches 80% of the NRC Limits for Lens Dose Equivalent (LDE), Shallow Dose Equivalent (SDE), and Total Organ Dose Equivalent (TODE).

The 80% threshold values are as follows: - 12 rem LDE. - 40remSDE. - 40remTODE 4.1.4. If an individual's current year dose history documentation includes an absent/

no record (A) dose type, then REDUCE the individual's allowable exposure (normally 2000 mrem TEDE for the year) by 1250 mrem TEDE for each quarter of the current year for which dose history documentation is absent/ no record (A), until all of that dose is resolved. All reductions require RPM approval.

4.1 5. If an individual is suspected of exceeding any of the NRC exposure limits in Table 1, then PROHIBIT the individual from entering the RCA until a detailed evaluation of the individual's actual dose equivalent has been conducted.

Future access will depend upon the results of the evaluation.

1. If an exposure in excess of the applicable exposure limit has occurred, then PROHIBIT the individual from entering the RCA until the end of the current calendar year. 2. If an exposure in excess of the applicable exposure limit has not occurred, then the individual may be permitted to re-enter the RCA. 4.1.6. During a condition where the Generating Stations Emergency Plan has been initiated, emergency exposure authorizations shall be performed in accordance with the station's Emergency Plan Implementing Procedures.

33658 Time to Complete:

Point Value: Cross

Reference:

RO Value 12.6 In one calendar year, a NE0 received 453 mrem while visiting a foreign nuclear plant as part of a Technical Exchange Program The Operator's prior exposure at Hope Creek was 175 mrem for that same year Q77351 3 SRO Value 13.0 If no exposure limit extensions have been authorized, which of the following list the MAXIMUM additional non-emergency Total Effective Dose Equivalent (TEDE) that this individual could receive at Hope Creek for the remainder of that year7 A. 1375 mrem B. 1825 mrem C. 3375 mrem D. 3825 mrem Answer: A Associated objective(s):

NOH04ADM024C-From Memory Describe what the worker is acknowledging when signing a RWP 01 prior to use. IAW NC.NA-AP.ZZ-0024, Radiation Protection Program License I ES-40 I Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 WA # G2.1.3 Group # 3 Importance Rating

3.7 Knowledge

of shift or short-term relief turnover practices Proposed Question: Common 71 Which one of the following describes Reactor Operator pre-and post-shift relief actions that should be implemented by the oncoming operator IAW OP-AA-112-101 "Shift Relief and Turnover"?

A. B. C. D. PRIOR to relief, read the Control Room logs through the last previous date on shift, or the preceding four (4) days logs, whichever is less. PRIOR to relief, review the Daily Orders.

POST relief, tour the main control room back panels. PRIOR to relief, read the Control Room logs through the last previous date on shift, or the preceding seven (7) days logs, whichever is less. PRIOR to relief, tour the main control room back panels. POST relief, review the Daily Orders. PRIOR to relief, read the Control Room logs through the last previous date on shift, or the preceding four (4) days logs, whichever is less. PRIOR to relief, tour the main control room back panels. POST relief, review the Daily Orders. PRIOR to relief, read the Control Room logs through the last previous date on shift, or the preceding seven (7) days logs, whichever is less. PRIOR to relief, review the Daily Orders. POST relief, tour the main control room back panels. Proposed Answer:

C Explanation (Optional):

C. Correct IAW OP-AA-112-101 "Shift Relief and Turnover" - Section 4.8.3. - Prior to relief, the on-coming Reactor Operators should PERFORM the following: - READ the Control Room logs through the last previous date on shift, or the preceding four days logs, whichever is less. - DISCUSS with the off-going Reactor Operator all items listed on the turnover sheet, Shiftly and Daily Surveillance, and any other information pertinent to Page 145 of 205 I I ES-401 Sample Written Examination Form ES-40 1-5 Question Worksheet proper continuity of operations. - TOUR Main Control Room back panel areas After relief, the on-coming RO's should PERFORM tile following: - ANNOUNCE shift turnover and relief to the Unit Supervisor. - CONFER with the Unit Supervisor to determine the scope of planned shift activities and their responsibilities for that shift. - REVIEW Daily Orders. - REVIEW Standing Orders for new entries. A. B. D. Incorrect - the back panels are toured prior to relief, the daily orders are reviewed post re I ief. Incorrect - only the preceeding 4 days logs should be reviewed Incorrect - the back panels are toured prior to relief, the daily orders are reviewed post relief. Only the preceeding 4 days logs should be reviewed. Technical Reference(s) IAW OP-AA-112-101 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: ADMPROI 02E004 (As available)

-. Question Source:

Bank # Modified Bank # New X (Note changes or attach parent)

Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments: Page 146 of ,205 11, I I I q.71 OP-AA-112-101 Revision 2 Page 7 of 11 4.7. Field Supervisor Shift Turnover Responsibilities:

4.7.1. INITIATE

a turnover sheet and ENTER the current status for their respective responsibilities (Off-going Field Supervisor).

Examples of items that may be included are: - Systems status in plant. - Alarms in plant. - - Abnormal controllers in plant. - Abnormal instrumentation in plant. - Ground & Test Devices installed. Special chemistry or radwaste evolutions in progress or pending. - Releases - Switchyard evolutions in progress or pending switching orders. - Radiation Precautions. - Secondary/Balance of Plant (BOP) steam leaks. - Miscellaneous Secondary/BOP items. - Fire Brigade Leader Duties - Fire Protection System status and impairments.

4.7.2. CONFER

with the Shift Manager concerning the scope of the shift and applicable responsibility for that shift.

4.8. Reactor

Operator Shift Turnover Responsibilities:

4.8.1. PERFORM

turnover in the Main Control F!oom. The Reactor Operator position must be filled by a qualified licensed reactor operator.

4.8.2. INITIATE

a turnover sheet and ENTER the current status for their respective Unit (Off-going Reactor Operator).

Examples of items that may be included are: - Unit status, - LCOAR / LCO status, - Major Out-Of-Services, - Abnormal positioned components, - Daily orders, - Su rvei Ila nces, - Major evolutions in progress or planned. 4.8.3. Prior to relief, the on-coming Reactor Operators should PERFORM the following: - READ the Control Room logs through the last previous date on shift, or the preceding four days logs, whichever is less. DISCUSS with the off-going Reactor Operator all items listed on the turnover sheet, Shiftly and Daily Surveillance, and any other information pertinent to proper continuity of operations. -

OP-AA-I 12-101 Revision 2 Page 8 of 11 4.8.4. 4.9. 4.9.1. 4.9.2. - TOUR the Main Control Boards with the off-going Reactor Operators and DISCUSS the following as a minimum: - Status of safety-related systems. - - Running equipment and safety train alignments. Inoperable equipment, including instrumentation, and Limiting Conditions of Operations, including surveillance requirements.

Contingency actions / abnormal procedures when redundant pieces of equipment are unavailable (e.g., stator cooling pump) Reasons for new annunciator alarms. Out-of-service equipment including any surveillance or equipment work in progress at time of the shift relief.

Abnormal events occurring during the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. - - - - TOUR Main Control Room back panel areas - After relief, the on-coming RO's should PERFORM the following: - - - REVIEW Daily Orders. - ANNOUNCE shift turnover and relief to the Unit Supervisor.

CONFER with the Unit Supervisor to determine the scope of planned shift activities and their responsibilities for that shift. REVIEW Standing Orders for new entries. Radwaste Panel ODerator Turnover Responsibilities PERFORM turnover at the radwaste panel as appropriate.

INITIATE a turnover sheet and ENTER the current status for their respective responsibilities (Off-going Radwaste panel operator).

Examples of items that may be included are: - Liquid Systems. - Gaseous Systems. - Waste Water Treatment status. - Sewage Treatment status. - Makeup Demins status. - CP Demins status. - Shipments status. - - Need to Order status. - Gas Decay Tanks status. - Abnormal Radwaste Controllers. - Abnormal Radwaste Instrumentation. - Water Inventory status in Radwasle.

LCO / TRM / RETS / DEL status.

, ES-40 1 Sample Written Examination Form ES-40 1-5 Question -. Worksheet Examination Outline Cross-reference:

Knowledge of em rgency communications systems d Level RO SRO Tier # 3 KIA # G2.4.43 I m po rta rice Rating Group # 4 3.2 schniq ues. Proposed Question:

Common 72 An Alert has been declared at Hope Creek.

You have been designated as the Secondary Communicator (CM2).

IAW the ECG Secondary Communicator Log, you are required to A. activate ERDS within 30 minutes from the Shift Manager OR Control Room SPDS terminal.

B. activate ERDS within 60 minutes from the Shift Manager OR Control Room SPDS terminal.

C. establish communications with state and local organizations using ERDS within 30 minutes. D. establish communications with state and local organizations using ERDS within 60 minutes. Proposed Answer:

B Explanation (Optional):

B. Correct - IAW ECG Att.8 section A.4.b.

A. C. D. Incorrect - required within 60 minutes Incorrect - ERDS is not used for this purpose Incorrect - ERDS is not used for this purpose Technical Reference(s)

ECG, ATT 8. Section A.4.b. (Attach if not previously provided) Proposed references to be provided to applicants during examination: none Learning Objective: (As available) Question Source: Bank

1. Page 147 of 205 ES-40 1 Sample Written Eitamination Form ES-401-5 Question Worksheet Question Question His tory: Cognitive Level:

10 CFR Part 55 Content: Comments: Modified Bank

1. -. (Note changes or attach parent)

New X -. Last NRC Exam -. Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 X Page 148 of 205 ECG ATT 8 P&. 3 of 13 YO TPFICATIONS (con t ' d) ACTIVATE EIWS withm 60 minutes from EITHER the SM Office or the CR SPDS terminal;

1) PRESS <TOP> button. 3) 3) Under Displays, PRESS .<Ems> button. Under ACTION, PRESS <START ERDS LINK> button NOTE See pages 11 8L 12 of this attachment a!; needed for fax machine information.

-. "Link Denied" may be displayed until NRC accepts transmission of data, which may take Initials 4) ERDS LINK STATUS w.ill read "TRANSMITTING DATA".

CM2 5. OBTAIN a copy of the ICMF and FAX [he ICMF to Group A (EOF2 - FAX to Group C). CM2iTSC2EOF2

6. COIvfPLETE a Station Status Checklist (SSCL) Form, Pg. 8 or Common Site UNUSUAL EVENT Statim Status Checklist (SSCL) Form, Pg. 10; ( ) a. OBTAIN SM (TSS/SSM) assistance, as needed for Pg. 1. ( ) b. OBTAIN SWT (RACIRSM) assistance, as needed for Pg. 2. ( ) c. FAX to Group B. (EOF2 - FAX to Group D) ( ) d. fax transmission of the SSCL is incomplete, (N/A for Common Site) THEN CONTACT the State Agencies listed below, READ the data, AND DOCUMENT on SSCL, Pg. 2. DEMA Delaware Emergency Management Agency 302-659-2290 BNE NJ Bureau of Nuclear Ehgineering 609-984-7700 ( ) e. REPEAT Steps 6a - d approximately every half hour Q& I IMMEDIATELY for significant changes in Station status, until either Turnover or relief. CM2/TS C2EOF2 7. OBTAIN a completed NRC Data Sheet &$om the Primary Communicator (CMlITSClIEOFI) and FAX form to Group B (EOF2 -FAX to Group D) HCGS Rev. 18 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 KIA # G2.4.18 Group # 4 Importar\ce Rating 3.3 Knowledge of the specific bases for EOPs. Proposed Question: Common 73 EOP 102 PRIMARY CONTAINMENT CONTROL, contains the following retainment override.

THEM. I EXIT this procedure I BEFORE Dw'l press reaches 0 prig. TERMINATE dwI sprays I BEFORE suppression chamber press reaches 0 psig. TERMINATE supp chamber sprays EXIT this procedure and ENTER SAG I Which one of the following statements describes the bases for terminating drywell spray before drywell pressure reaches 0 psig? A. It makes one more RHR loop available as soon as possible for injection into the reactor pressure vessel. B. This action ensures that the drywell structure will NOT endure excessive thermal stresses due to rapid cooldown.

C. It ensures a drywell temperature below 212 degrees F, therefore there is NO need to continue drywell sprays.

D. It prevents drawing a negative pressure in the containment, which would open the vacuum breakers and draw air into the containment.

Proposed Answer: D Explanation (Optional): IAW EOP 102 Bases for step PCC It prevents drawing a negative pressure in the Page 149 of 205 I I I I ES-401 Sample Written Eiamination Form ES-401-5 Question Worksheet containment, which would open the vacuum breakers and draw air into the containment.

D. Correct. A. B. C. Incorrect - Concern is de-inerting containment.

Incorrect - a negative pressure will open the SC to RB vacuum breakers and de-inert containment. Thermal stress is not a concern.

Incorrect - Concern is de-inerting containment. Technical Reference(s) (Attach if not previously provided)

EOP 102 Bases for step PCC-1 Proposed references to be provided to applicants during examination:

None Learning Objective:

E01 01 PE008 (As available) 80632 -. Question Source: Bank

1. Modified Bank # New (Note changes or attach parent) 2003 Question History: Last NRC Exam - Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 150 of 205 HC.OP-EO.=-0102 73 PCC-I I BASES 1 All entry conditions have cleared 1 EXIT this procedure

-? Discussion If actions taken to control reactor building conditions are successful, steps in this EOP have performed their intended mitigation function and exit to normal procedures is appropriate.

This step helps to promote consistent EOP implementaiion.

Drwl sprays have been initiated BEFORE Drwl Press reaches 0 psig, 'TERMINATE Drwl Sprays Discussion The operation of drywell sprays must be terminated by the time drywell pressure decreases to 0 psig to ensure that primary containment pressure i:j not reduced below atmospheric.

Terminating the sprays "before ... 0 psi$ permits use of the sprays for fission product scrubbing at low pressures or if the containment has failed, yet still avoids negative containment pressures.

Consistent with the definition of "before," the actual pressure value at which the sprays should be secured is event-specific:

Reducing containment pressure below the scram setpoint will clear the scram logic and maximize the margin to containment pressure limits. If the containment has failed or if primary containment venting is anticipated, it may be advisable to continue spray operation at low pressures to scrub the containment atmosphere.

Reducing primary containment pressure will also reduce the NPSH available for pumps drawing suction from the suppression pool. As stated in the discussion of Caution #2, NPSH limits should be observed if possible, but may be exceeded if warranted by event- specific conditions.

If there is no need for conlinued spray operation, however, sprays may be terminated at higher pressures if NPSH limits are approached. The reference to Caution #6 emphasizes the relationship between primary containment pressure and available NPSH. Hope Creek Page 4 of 36 Rev. 01 C:\Documents and Settings\sdennis\My Documents\Hope Creek 2008,-2009\References\LESSON PLANSEmergency Operating Procedures\EOP BasesWaster\Lesson Plans\EOP102bases.doc I I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Out I i ne Cross-reference

Level Tier # Group # KIA # RO SRO 3 2 G2.2.13 Importance Rating

4.1 Knowledge

of tagging an clearance procedures Proposed Question: Common 74 IAW the Safety Tagging Program procedures, which of the following statements is correct regarding Worker's Blocking Tags (WBTs)?

A. The Clearing Agent will place a label designating the Worker and Job Technician on the WBT. B. A WBT may be used to isolate a high voltage energy source

(>600 volts). C. A Work Clearance Document (WCD) containing WBTs may also contain Yellow Permissive Tags (YPTs).

D. Two WBTs may be simultaneously installed on the same blocking point. Proposed Answer: C Explanation (Optional):

SH.OP-AP.ZZ-0015, rev 20 Att. 3 C. CORRECT- A Work Control Document (WCD) containing WBTs may also contain Yellow Permissive Tags (YPTs).

WCDs containing WBTs may contain other tag types such as RBTs and YPTs.

A. B. D. INCORRECT - A label designating the Worker and the Clearing Agent shall be placed on the WBT by the Worker. INCORRECT - A WBT may not be used to isolate a high voltage energy source. INCORRECT - The WBT shall not be installed on any blocking point that is already tagged with any safety tag except for a WCT. Technical Reference(s)

SH.OP-AP.ZZ-0015, rev 20 Att. 3 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Page 151 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Learning Objective: Question Source:

NAOOI 5E004 (As available)

Bank # 62253 Question History: Question Cognitive Level:

10 CFR Part 55 Content: Modified Bank # -. (Note changes or attach parent)

-. New Last NRC Exam - Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 X Comments: Page 152 of 205 (continued)

WORKER'S BLOCKING TAG (WBT) \ - / - -i SH.OP-AP.ZZ-O015(Q)

ATTACHMENT 3 TAGGING RULES (Page 4 of 6) WBTs shall not be used if work scope changes from that which the WBT use was initially intended.

Work shall not be periormed on the component to which the WBT is affixed. WBTs shall not be used to isolate equipment from a high Voltage (600 volts and above) energy source. WBTs shall not be placed on any blocking point that is tagged with any other Safety Tag except a White Caution Tag. WCDs containing WBTs will only allow activation of one Job Supervisor and one Worker at a time; only one worker shall be designated to operate the equipment (or hold responsibility for the tag) at a time. > The individual worker specified on WCD shall be the lead worker for the job (the lead worker may designate another worker directly int olved with the job to operate the equipment).

9 After the Clearing Agent has signed onto the request, the Worlca shall place a label designating the Worker and the Clearing Agent on the WBT, P %%en work is completed, components tagged with WI3Ts shall be returned to the tagged position as specified on the WCD. conditions warrant, THEN the SWCRWWCCS may grant permission to leave components in an cdternate position. This permission shall be documented in the long text or the mtagging text of the WCD. A WCD containing WBTs can be turned over from one Clearing Agentlworker to another. This allows jobs to continue from one shift to another without the necessity of swapping tags. WCDs containing wB'T(s) may contain other tag types such as red blocking tags and yellow permissive tags.

Electrical T&D (Substation Mechanics) personnel performing work on station switchyard equipment under the tagging jurisdiction of ESO may apply T&D WBT(s)provided tags are within tagged boundaries and approvied by SM prior to application.

Electrical T&D (Relay Department) personnel may apply T&D WBT(s) if all of the following conditions are met: > The circuits are useid for relay protection (ex. breaker failure, generator protection).

> The SM/CRS/WCCS and/or ESO approve manipulation prior to hanging tag@), 9 Relay Department shall use an approved procedure that has specific steps to log all tag(s) applied and released.

i Salem/Hope Creek Page 60 of 75 Rev 20 I I I ES-401 Sample Written Examination Form ES-401-5 Quest ion Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 KIA # G2.3.13 Importance Rating

3.4 Knowledge

of Radiological Safety procedures pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high -radiation areas, aligning filters, etc. Proposed Question: Common 75 The following conditions exist for a job to be performed on a system.

0 0 0 The general area radiation levels are 10 mrem/hr in the room.

The hot spot in the room is a pipe elbow that has a radiation level of 100 mrem/hr.

The job will be performed near the hot spot area. (Assumptions: ALL 4 cases below have the same transition time to and from destinations. All shielding placement and removal is at 100 mrem/hr. The hot spot with shielding in place is 10 mremlhr) Which one of the following methods would comply with ALARA procedural requirements for performance of the task?

A. The job is performed by using 2 operators for 3 hrs each on the job at the hot spot.

B The job is performed by 3 operators for 1 hr each on the job at the hot spot and a fourth operator reading instructions in the general room area for 1 hr. C. The job is performed by 2 operators for 2 hrs each on the job at the hot spot and a third operator reading instructions in the general room area for 2 hrs. D. Two Radiation Protection personnel hang and remove 1 tenth thickness of lead shielding on the hot spot in 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the job. The job is performed after the lead shielding is in place by using 2 operators for 3 hrs each on the job. Proposed Answer: B Explanation (Optional):

B. Correct: The job is performed by 3 operators for 1 hr each on the job at the hot spot and a fourth operator reading instructions in the general room area for 1 hr.( 3 operators X 100 mrem/hr x 1 hr) + (I operators X 10 mremlhr x I hr) = 310 mremlhr. A. Incorrect:

The job is performed by using 2 operators, for 3 hrs each on the job at the hot spot. ( 2 Page 153 of 205 ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet operators X 100 mremlhr x 3hrs) = 600 mrem/hr. Incorrect:

The job is performed by 2 operators for 2 hrs each on the job at the hot spot and a third Operator reading instructions in the general room area for 2 hrs. ( 2 operators X 100 mrem/hr x 2hr) + (1 operators X 10 mrem/hr x 2hrs) = 420 mrem/hr. Incorrect:

Two Radiation Protection personnel hang and remove 1 tenth thickness of lead shielding on the hot spot in 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the job. The job is performed after the lead shielding is in place by using 2 operators for 3 hrs each on the job. ( 2 rad techs X 100 mremlhr x 1 Shrs) + (2 operators X 10 mrem/hr x 3hr) = 360 mrem/hr. C. D. Technical Reference(s) RP-AA-400 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective: Question Source: (As available)

Bank # WTS Bank Modified Bank

1. -. (Note changes or attach parent)

-. New Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X Comments:

Page 154 of 205 iW' Exe Ian jM 1. Nuclear PURPOSE R P -AA-400 Revision 4 Page 1 of 13 Level 2 - Reference Use ALARA PROGRAM 1.1. This procedure establishes the requirements and responsibilities for the effective implementation of the ALARA Program.

The objective of the ALARA program is to ensure that occupational radiation exposure, both individually and collectively, is maintained ALARA.

2. TERMS AND DEFINITIONS 2.1. Q ALARA: cronym for "as low as reasonably achievable." ALARA means making limits, as defined in 10 CFR 20, as is practiical consistent with the purpose for which the licensed activity is undertaken, taking into account the state of technology, the economics of improvements in relation to benefits to the public health and safety, and other societal and socioeconomic considerations, and in relation to utilization of nuclear energy and licensed material in the public interest.

ALARA Plan: A documented job evaluation. The evaluation will consider the radiological conditions expected during each phase of the job and the methods and controls to minimize contamination and collective radiation exposure (person-rem).

The term "ALARA Plan" is equivalent to "AI-ARA Action Review" and "ALARA Review." easonable effort to maintain exposure to radiation as far below the dose 2.2. 2.3. Station ALARA Committee (SAC): Station committee responsible for the overall coordination of the Station ALARA Program and for advising site management in matters relating to ALARA and other pertinent Radiation Protection programs.

3. RESPONSIBILITIES NOTE: Sites may utilize position titles that are not identical to those outlined in this procedure.

In those cases, the responsibilities should be performed by the site individual with responsibilities closest to those described in this procedure.

3.1. Plant

ManaqerISite Vice President 3.1.1. ACT as the Chairperson for the Station ALARA Committee under normal circumstances.

If the Plant Manager or Site Vice President is pJ available to fulfill this duty, a senior member of station management staff may be appointed as an alternate to chair the meeting.

I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group lib 1 KIA # 295021 AA2.01 Importance Rating

3.6 Ability

to determine andlor interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor water heatuplcooldown rate. Proposed Question:

SRO 76 The plant is shutting down for a refuel outage. Shutdown Cooling has been in service for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. T= 12:OO - RPV temperature is 149 degrees F. Then, a complete loss of Shutdown Cooling occurs. After 20 minutes, the operators determine that RPV temperature is rising at 16 degrees every 10 minutes. T= 12:20 - RPV temperature is 186 degrees F. Which one of the following describes how the heatup, if it continues at the rate stated above, will affect the plant Operational Condition and Technical Specification (TS) heatup limits? A. After T=12:30, a mode change will occur.

At T=13:00, the TS heatup rate limit will be exceeded.

B. Before T=12:30, a mode change would occiir. At T=13:00, the TS heatup rate limit will be exceeded.

C. After T=l2:30, a mode change will occur.

At T=13:00, the TS heatup rate limit will NOT be exceeded.

D. Before T=12:30, a mode change would occur. At T=13:00, the TS heatup rate limit will NOT be exceeded.

Proposed Answer: B Explanation (Optional): Mode change occurs at >200 dgrees F. per TS definitions.

The TS heatup limit is 100 degrees in a one hour period. Although the rate is >IO0 degrees per hour the limit is not exceeded until the one hour time period has been met (1 200 - 1300) TS 3.4.6.1 .a. & TS definitions of operational conditions B. Correct. Mode change has already occurred.

(202 F @12:30) Limit has been exceeded.

Page 155 of 205 ES-401 Sample Written Examination Form ES-40 1-5 Question Worksheet 101 degrees F at 1300. (1 49 @ 12:OO to 250 @ 13:OO) A. C. D. Incorrect. Mode change has already occurred. (202 F @12:30) Incorrect.

The TS limit has been exceeded. 101 degrees @13:00 Incorrect. A mode change would occur prior to 1230 (202 F @12:30) The TS limit has been exceeded.

1 Oldegrees

@I 3:OO Technical Reference(s)

TS 3.4.6.1 .a.

& TS definitions (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

RXVESS E EO0 7 (As available)

-. Question Source:

Bank # (Note changes or attach parent)

_. Modified Bank

1. X -. New Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments: Page 156 of 205 REACTOR COOLANT SYSTEM

((L 3/4.4. 6 PRESSURE/TEMPXRATURE LIMITS - REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be li-mited in accordance with the limit lines shown on Figure 3.4.6.1-1 (hydrostati-c or leak testing), and Figure 3.4.6.1-2 (heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS), and Figure 3.4.6.1-3 (operations with a critical core other than low power PHYSICS TESTS) , with: A maximum heatup of 100°F in any one hour period, A maximum cooLdown of 100°F in any one hour period, 6 b. c. A maximum temperature change of less than or equal to 20°F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d. The reactor vessel flange and head flange metal temperature shall be maintained greater than or equal to 79°F when reactor vessel head bolting studs are under tension.

APPLICABILITY:

At all times ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for contir.ued o3erations or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. SURVEILLANCE REQUIREMENTS 4.4.6.1.1 During system heatup, cocldown and inservice leak and hydrostatic testing operations, the reactor coolant s:y.stem temperature and pressure shall be determined to be within the above required heatup and cooldown limits and as applicable, at least once per 30 minutr?s.

to the right of the limit lines of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 I HOPE CREEK 3/4 4-211 Amendment No. 88 I CONDITIOK TABLE 1.2 OPERRTIONAL C 3NDITIONS MODE SWITCH POSITION - 1. POWER. OPERATION Run AVERAGE REACTOR COOLANT TEMPERATURE Any temperature

2. STARTUP Startup/Hot Stancby Any temperature

-7. HOT SHUTDOWN Shutdown', '.* > 200°F . COLD SHUTDOWN Shutdown#l##f

' ** 2 200""F' -7 5. REFUELING*

Shutdown OK Refuel**r#

I 140°F #The reactor mode switch may be p.laced in the Run, Startup/Hot Standby, or Refuel pi3sition to test the switch interlock functions and related instrumentation provided that -t:he control rods are verified to remain fully inserted by a secor-d licerised operator or other technically qualified member of the unit technical staff. If the reactor mode switch is placed in the Refuel position, the one-rod-out interlock shall be OPERABLE.

1. 1. The reactor mode switch may be placed in the Refuel position while a single control rod drive is bei-ng removed from the reactor pressure vessel per Specification 3.9.13.1.

• Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the !lead removed. ** See Special Test Exceptions 3.10.1 and 3.10.3. The reactor mode switch may be placed i.1 the Refuel position *** while a single control rod is seicg re-oupled or withdrawn provided that the one-rod-out interlock is 0PE:RABLE.

'See Special Test Exception 3.10.8 HOPE CREEK 1-1 1 Amendment No.

69 I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 295023 AA2.05 Importance Rating

4.6 Group

1. $ 1 Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS

Entry conditions of Emergency plan Proposed Question:

SRO 77 Given the following:

0 0 0 The plant is in OPCON 5. Core offload is in progress.

A spent fuel bundle is full up on the main hoist over the core. The refuel bridge spotter notices the fuel bundle has become unlatched and has fallen into the vessel.

A short time later, the following Refuel Floor Rad Monitors are in HIGH alarm: Spent Fuel Pool ARM 0 0 0 Refuel Floor Exhaust Channels A, B, C General Area radiation surveys have NOT been performed All other plant systems are operating as designed Which one of the following describes actions required IAW AB-CONT-0005 "Irradiated Fuel Damage" and the Emergency Plan?

Suspend the handling of Irradiated FueKomponents.. . A. re-establish Secondary Containment and declare an Alert. 6. re-establish Secondary Containment and perform a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report to the NRC. C. evacuate all unnecessary personnel from the refuel floor and declare an Alert. D. evacuate all unnecessary personnel from the refuel floor and perform a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report to the NRC. Proposed Answer: C Explanation (Optional): ECG Section 6.4.2.b. -with the alarms noted in stem an Alert declaration is warranted.

C. Correct. A. Incorrect. Secondary Containment was not lost based on stem conditions Page 157 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet B. D. Incorrect. Secondary Containment was not lost based on stem conditions Incorrect.

May be correct if Alert is believed tal be incorrect. Based on voluntarykourtesy report IAW ECG Section 1 1.10.2 Technical Reference(s) AB-CONT-0005 (Attach if not previously provided) ECG Section 6.4.2.b Proposed references to be provided to applicants during examination:

ECG - not the attachments Learning Objective: ABCNT5E007 (As available)

Question Source: Question History: Question Cognitive Level:

10 CFR Part 55 Content: Bank # Modified Bank # (Note changes or attach parent)

X -. New Last NRC Exam Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 55.43 1 Comments: Page 158 of 205 Initiating Condition OPCON EAL # I EI MI RI GI NI cf AI TI II NI EI LI El El YI I I CI 01 I I El El SI Action Required I IICG 7 ECG Rev 02 rag? 2 of) 6.0 Radiological ReleasedOccurrences

6.4 Irradiated

Fuel Event Events that have or may result in uncovering Irradiated Fuel outside the Reactor Vessel Irradiated Fuel m 6.4.2.a IF Major Damage to Irradiated Fuel has occurred AND Valid High Alarm received hz ANY one of the following RMS channels:

' Refuel Floor Exhaust Channel A (9RX627) ' Refuel Floor Exhaust Channel B (9RX628) Refuel Floor Exhaust Channel C (9RX629) ----r-- /i*= \ IF/ \ /' ---- Unplanned rise on ANY one of the Area Rad h4onitors or by general area rad survey indicates - > 2000 mRem/hr: ' Spent Fuel Storage Pool Area (9RX707) ' New Fuel Criticality Storage Chatlliel A (9RXG 12) ' New Fuel Criticality Storage Cliaimel B (9RX613) *allowing (X] -- 6.4.2.c IF Visual observation or Irradiated Fuel uncovered I THEN I v 1 I Refer to Attachment 2 31 Err

, I rJ lLG SUBSEOUENT OPERATOR ACTIONS CONDITION A. Irradiated Fuel Damage. DateRime:

B. Damaged irradiated fuel is attached to the fuel handling grapple. Datenime:

C. Damaged irradiated fuel is in the Spent Fuel Pool, Reactor Cavity - OR Cask Loading Pit. Datemime:

HC.OP-AB.CONT-0005( < 1) IRRADIATED FUEL DAMAG S - ACTION - A.1 PERFORM the following:

0 E,NSURE either RBVS - OR FRVS is in service PERFORM ONE of the following:

0 R 1. ENSURE Secondary Containment Integrity is 0 2. WITHIN 30 minutes, in effect. IMPLEMENT SH.OP-AP.ZZ-0108 Attachment 10 to seal secondary containment breaches.

P 0 DIRECT Radiation Protection to take air samples AND control access to the Reactor Bldg and Refuel floor. EVACUATE all UNNECESSARY personnel from the affected area.

0 0 B.l E Radiation levels permit, PLACE the damaged fuel in the Spent Fuel Pool. 3 C. 1 DIRECT Radiation Protection to survey the affected Spent Fuel Pool Cooling -- AND Residual Heat Removal systems piping as appropriate.

Hope Creek Page 5 of 10 Rev. 2 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 1 KIA # 29501 9 AA2.02 Importance Rating

3.7 Ability

to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR : Status of safety-related instrument air system loads Proposed Question: SRO 78 Hope Creek is operating at 100% power when an Irlstrument Air line in the Turbine Building ruptures.

The air compressors are unable to keep up with the loss of air and Instrument Air pressure is lowering. The operators insert a manual scram.

What will the Reactor Pressure Vessel (RPV) level control and pressure control strategy be for the loss of Instrument Air? A. IAW EOP-101 "RPV Control", SRVs for pressure control, HPCVRCIC for level control.

B. IAW EOP-IO1 "RPV Control", SRVs for pressure control, Maximize CRD for level control . C. IAW AB.ZZ-0000 "Reactor SCRAM", Bypass; Valves for pressure control, HPCllRClC for level control.

D. IAW AB.ZZ-0000 "Reactor SCRAM", Bypass valves for pressure control, Maximize CRD for level control.

Proposed Answer: A Page 159 of 205 ES-40 1 Sample Written Ekaminat ion Form ES-401-5 Question Worksheet Explanation (Optional):

A CORRECT - Outboard MSlVs will go closed on a loss of air, therefore NO steam for feedpumps or use of the main condenser for decay heat. Condensate will be unavailable due to NO feedpath on a loss of air. B C D INCORRECT - CRD flow control valves fail closed on a loss of air INCORRECT - Condenser is NOT available for pressure control INCORRECT - Condenser is NOT available and NO condensate line up is possible due to level control valves fail closed on a loss of air. (Attach if not previously provided)

___. Technical Reference(s):

EOP-l o1 Proposed references to be provided to applicants during examination: None Learning Objective:

I NSAl REO 16 (As available) Question Source:

Bank # INPO 25895 Modified Bank # New (Note changes or attach parent)

-. Question History: Last NRC Exam 2005 Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Page 160 of 205 I I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 1 KIA # 295025 G2.1.23 Importance Rating 4.4 (K&A Statement) Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of operation. (High Reactor Pressure) Proposed Question:

SRO 79 The plant is operating at 22% power when an EHC failure raises RPV pressure to 1052 psig. The following events occur: 0 Main Turbine trips 0 0 0 NO control rod motion Mode Switch locked in the Shut Down position Scram Air header pressure lowers to 72 psig Which one of the following EOP entries correctly describes the required operator action(s) and the basis for the action(s)?

A. B. C. D. Enter EOP-IOIA ATWS-RPV Control because the SDV is full. Enter EOP-IOIA ATWS-RPV Control because manual ARI is required. Enter EOP-101 Reactor Pressure Vessel Control because a scram reset is required. Enter EOP-101 Reactor Pressure Vessel Control because the Main Turbine is tripped. Proposed Answer:

B. Page 161 of 205 I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):

B. Correct - EOP-101 is initially entered on RPV press >I037 psig, but with the Mode Switch in SD and all rods not in the SRO must transition to EOP-IOIA, where the verification of ARI is the next step (the Turbine has already tripped).

A. Incorrect - EOP-IOIA is entered because the rods did not fully insert, additionally scram air header pressure has not lowered therefore the SDV Vents and Drains are open, there is no confirmation in the question stem that the SDV is full. Incorrect - A scram reset is not required at this time because scram air pressure has not lowered (failure to scram), additionally EOP-101 is exited and EOP-IOIA is entered. Incorrect - The turbine tripping is not an entry condition into EOP-101, additionally EOP- 101 is exited and EOP-IOIA is entered. C. D. Technical Reference(s):

EOP-l (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective:

EA1 01 AE002 (As available)

Question Source: Bank

1. ID: Q56465 Question History: Question Cognitive Level: 10 CFR Part 55 Content: Modified Bank # New (Note changes or attach parent)

___. ~. Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 55.43 5 Comments: Page 162 of 205 ES-40 1 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 WA # 295026 G2.1.7 Group # 1 I m porta rice Rating 4.7 Conduct of Operations: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (Suppression Pool High Water Temperature) Proposed Question: SRO 80 While operating at 60% Reactor power, a Reactor Scram on low reactor water level occurs but all rods remain at their pre-trip conditions. Plant conditions thirty minutes after the transient start are: SLC tank level Rx power RPV pressure RPV level Suppression pool level Suppression pool temp Drywell pressure Main steam tunnel temperature 2600 gal <4 Yo 900 psig being controlled using SRVs Intentionally lowered to -1 35 inches and steady 79 inches 185°F and rising at 1 "F/5 min 4.5 psig 170°F and rising at 1 "F/2 min Which one of the following is required for the conditions above? A. Maintain RPV water level between +54" and -1 85". B. Bypass interlocks to open the MSIV's and reduce RPV pressure.

C. Reduce RPV pressure to prevent exceeding the Heat Capacity Temperature Limit curve.

D. Emergency Depressurize to prevent exceeding the Pressure Suppression Pressure curve. Proposed Answer: C Explanation (Optional):

C. Correct - HCTL limit is being approached and will reach Action Required line within 15 minutes. RPV pressure must be reduced to move away from limit. A. Incorrect - Wrong level band. Upper end of band limited to - 50 inches with an ATWS. Page 163 of 205 I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet B. D. Incorrect - Incorrect action based on evidence of leak in the Main Steam Tunnel.

Incorrect - Action to be taken if RPV pressure cannot be lowered. Technical Reference(s) EOP-102, 101A _._. (Attach if not previously provided) Proposed references to be provided to applicants during examination: SCP-L, SPT-P Learning Objective:

Question Source: Question History: Quest ion Cognitive Leve I: 10 CFR Part 55 Content: E01 01 AE008 (As available) 55997 -. Bank # Modified Bank

1. -. (Note changes or attach parent)

New _. Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 55.43 5 Comments:

Page 164 of 205 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 1 KIA # 295031 G2.4.3 3.9- ~ -- Importance Rating (K&A Statement) Emergency Procedures

/ Pian: Ability to identify post-accident instrumentation. (Reactor Low Water Level). Proposed Question:

SRO 81 A plant event has occurred.

RPV level is +20" and lowering.

You are informed by I & C that the following level instruments are inoperable.

0 Fuel Zone Range Level Recorder LR-R615 Upset Range Level Recorder LR-R608 Which one of the following sections of Technical Specifications must be entered for these inoperable instruments once plant conditions stabilize?

A. 3.3.7.5, Accident Monitoring Instrumentation ONLY B. 3.3.4, Recirculation Pump Trip Actuation Instrumentation ONLY C. 3.2.1, Isolation Actuation Instrumentation and 3.3.3 Emergency Core Cooling Actuation I ns t ru men t a t io n 0 N LY D. 3.3.7.5 Accident Monitoring Instrumentation and 3.3.3 Emergency Core Cooling Actuation Instrumentation ONLY. Proposed Answer:

A. Explanation (Optional):

A. Correct - The fuel zone instruments provide past accident indication only B. C. D. Incorrect - These TS instruments apply to normal and wide range instruments.

Incorrect - These TS instruments apply to normal and wide range instruments.

incorrect - These TS instruments apply to normal and wide range instruments.

Technical Reference(s):

T.S. 3.3.7.5 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Page 165 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question -. Worksheet Learning Objective: RXI NSTE02 1 (As available)

____. Question Source:

Bank # Modified Bank # (Note changes or attach parent)

Question History: Question Cognitive Level:

10 CFR Part 55 Content: New X Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 55.43 2 Comments: Page 166 of 205 INSTRUMENT TABLE 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION

& 3. 4. 5. 6. 7. 8. 9. 10. 11. 12. 13. 14. Reactor Vessel Pressure Reactor Vessel Water level Suppression Chamber Water level Suppression Chamber Water Temperature* Suppression Chamber Pressure Dryweii Pressure Drywell Air Temperature Deleted Safety/Relief Valve Position Indicators Drywell Atmosphere Post-Accident Radiation Monitor North Plant Vent Radiation Monitor

1. South Plant Vent Radiation Monitor

1. FRVS Vent Radiation Monitor # Primary Containment Isolation Valve Position InUicariun

$# MINIMUM APPLICABLE REQUIRED NUMBER CHANNELS OPERATIONAL OF CHANNELS OPERABLE CONDITIONS 2/valve**

2 1 3 1 2ivaive i ivaive i, 2,3 ACTION 80 80 80 80 80 80 80 80 80 81 81 81

• Average bulk pool temperature.

1. High range noble gas monitors.

• • Acoustic monitoring and tail pipe temperature.

1. 1. One channel consists of the open limit switch, and the other channel consists of the closed limit switch. HOPE CREEK 3/4 3-a5 Amendment No. 160 INSTRUMENTATION rJ+ ACCIDENT MONITORING INSTRUMENTATI(E LIMITING CONDITION FOR OPERATION 3.3.7.5 The accident monitoring Insirurnerltation channels shown in Table 3.3.7.5-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3.7.5-1.

ACTION: With one or more accident monitoring instr-urnentation channels inoperable, take the ACTION required by Table 3.3.7.5-1.

SURVEILLANCE REQUIREMENTS 4.3.7.5 Each of the above required accident monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at: the frequencies shown in Table 4.3.7.5-1.

HOPE CREEK 3/4 3-84 LESSON NAME: NUCLEAR BOILER INSTRUMENTATION NOH04RXINSTC 07/12/06 NOTE: I & C training has an optical isolator assembly that may be available to Figure 14 /vw :___.__1 demonstrate the following points, 7. Optical Isolator (GE Systems) a. Optical isolators are used to transfer signals between 1 E and non- 1 E electronic systems to maintain electrical isolation. This prevents an electrical fault in a non-1 E system from propagating to a 1 E system. There are two general types of optical isolators: analog and digital.

The optical isolator (designated AT on GE elementary drawings) has two circuits. The first is an infrared light emitting diode with its own power supply. The second circuit has a phototransistor and its own power supply. The digital and ,analog isolators are functionally equivalent, although the analog isolator converts the data input into a serialized digital format for transmission via the LED. The infrared light output of the input isolator travels through a 'light pipe'-a % inch cylindrical quartz about one inch in length-to the phototransistor in the output isolator module. The digital and analog output isolators provide an output equivalent to the input. The digital outputs are on-off and the analog isolator converts the serialized data into an analog output.

Both the input and output isolator cards require power. The digital isolators are typically powered by the intetfacing system. The analog isolators require an external source of power.

b. c. d. 8. Post Accident Monitoring System (PAMS)

I Objective 14a I a. The PAM System utilizes existing reactor vessel level and pressure instrumentation for post accident monitoring.

1) -7 Six level and two pressure transmitters supply signals to indicators and recorders in tht? control room on 1 OC650A and C. a) Two channels each (A & B) of fuel zone level, wide range level, shutdown range level and wide range pressure Fuel zone indicator LI-R610 and recorder LR-R615 Wide range recorders LR-R623A and B Shutdown range recorders LR-3622A and B Wide range reactor pressure recorders PR- 3684A and B -?(I) (2) (3) (4) Page 25 of 55 C:\Docurnents and Settings\sdennis\My Docuinents\Hope Creek 2008-2009\F!eferences\LESSON PLANS\Nuclear Boiler Instrurnentation\Master\Lesson Plans\NOH04RXINSTC-OO Nuclear Boiler 1nstrurnentation.DOC I 1 LESSON NAME: NUCLEAR BOILER INSTRUMENTATION NOH04RXINSTC 07/12/06 FSAR Sections 1.8.1.97 and 7.5,1.3.4 Tech Spec 3.3.7.5 HC. OP-ST.SH-0001 Although other level and pressure instrumentation channels are provided, only the above instruments meet the requirements of Reg Guide 1.97 (FSAR Table 7.5-1) and Technical Specification 3.3.7.5, Accident Monitoring Instrumentation.

NOTE: Both the indicating and recording functions are required for these instruments to be considered operable.

They are channel-checked IAW HC.OP- ST.SH-0001.

Figure 15 LObjective 14b b. The PAMS instruments utilize the 'A' and 'B' channel reference legs. These reference legs have a Short elevation drop within the drywell, making them less susceptible to the effects of elevated d rywell tem pera tu re. LR-R623A and B have two modes for trending parameters:

1) 2) c. Normal Operation - Slow (typically 20 mm/hour) Accident Conditions - High-speed (1200 mmlhour) 3) They automatically shift to the high-speed trend to provide increased resolution of the recorded traces of wide range reactor level and pressure.

a) The recorders will shift to the fast-speed trend when level drops to 12.5" or pressure reaches 1037 psig.

b) A 'HIGH SPEED CHART DRIVE RESET' pushbutton is provided to return the chart drives to the low-speed trend after the initiating1 conditions have cleared.

.= Figure 8 Objectives 7c, 15a, 15b, 15c, 15d, 16 IV. INSTRUMENTATION AND CONTROL A. Instrumentation

1. Main Control Room I Table I .1 a. Reactor Vessel Level See Table 1 Page 26 of 55 C:\Documents and Settings\sdennis\My Docurnents\Hope Creek 2008-2009\l?eferences\LESSON PLANS\Nuclear Boiler Instrurnentation\MasteALesson Plans\NOH04RXINSTC-O0 Nuclear Boiler 1nstrurnentation.DOC ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 1 KIA # 295028 EA2.03 Importarice Rating 3.9 ~- Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE

Reactor water level Proposed Question:

SRO 82 Given the following conditions: A Large Break LOCA has occurred in the Drywell concurrent with a LOP Only "C" EDG is running All control rods are fully inserted Wide Range RPV level indicator LR-623A is reading +20 inches Wide Range RPV level indicator LR-623B is reading -55 inches Drywell pressure is 29 psig and rising Drywell temperature is 300 F and rising Reactor pressure is 25 psig and steady Suppression Pool Level is 80 inches and rising Suppression Chamber pressure is 30 psig and rising "C" RHR Pump has been injecting LPCl Flow for 3 minutes Based on the above conditions, which one of the following actions is required?

A. Continue LPCl injection and enter EOP-206 "RPV Flooding".

B. Continue LPCl injection and continue in EOP-101 "RPV Control" in all control legs.

C. Stop LPCl injection, Emergency Depressurize IAW EOP-202, and then resume LPCl injection.

D. Continue LPCl injection, Emergency Depressurize IAW EOP-202, and then enter EOP- 206 "RPV Flooding".

Proposed Answer: A Explanation (Optional): with High drywell temps and low RPV pressure, per EOP caution 1 level is unreliable. Therefore it is not known and RPV flooding is required A. Correct. Level is unknown due to unreliability of level instruments with high drywell temperature. RPV flooding is required.

LPCl injection would continue Page 167 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet

6. C. D. Incorrect.

RPV flooding is required.

Pressure Control Leg of 101 is exited when 206 is required.

Incorrect.

Would not stop LPCl injection Incorrect. ED already performed, RPV flooding required Technical Reference(s) EOP-102 retainment step (Attach if not previously provided)

EOP caution 1 Proposed references to be provided to applicants during examination: Caution 1 Learning Objective: EOP206E008 (As available) Question Source:

Question History: Question Cognitive Level:

10 CFR Part 55 Content: 56161 -. Bank # Modified Bank # -. (Note changes or attach parent)

New -. Last NRC Exam -. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 55.43 5 Comments:

Page 168 of 205 EOP CAUTION 1 Dml' SPDS Points A2266 A2274 I &mated RPV water me1 lnatmments LR-RBZW-BZI Wds Range A (-150 to +BO in.)

• LR-368W Namm Rango A (0 to +EO in ) LR-35838 Narmw Range B (0 to +Bo in.)

• LRP615-BZI FuelZoneA(-3ll 10-111 in.) * *

• LR-R623A-821 Wlde Range A (-150 to +60 in.) LR-3683A Namm Range A (0 to +Bo in.) LR-PS16B21 Fuel Zone A (-31 1 to -11 1 in.) A Under mndltions of eiavated drywell temperature channels A and B of the wide namm and upset RW water level instruments provlde the most relIablB IndmUon:, If any dlywell temperature SPDS point exceeds the RW saturation temperelure ttie assodatad level inatrumenla may be unmliabb B A2280 A2281 .- RPV Sntuntlon Temperature

• LR-PS15-BZl FuelZDneA(-311 to-I11 in.)

• LI-R610-B21 Fuel Zone B (-311 10-111 in) 559 500 A2287 0 IO0 200 3w 400 500 WO 700 800 9W 1000 1100 880 RW Pre~ure (Pcrn) (Dlywsll lemperatum SPDS wlnta indicate the local temperatures near the sasocistsd inslrumenr.

runs) .- Drywsll ternpeaturn lndicalkm IMX nnga ~ 500'F; I lac can ehiend thmunh uw of mistsnw tsbkn LI-R61O-BZ1 Fuel Zone B (-311 to -111 in.) - *

• LI-R610-B21 FwlZoneB(-311 lo-111 in) LR-RS23BE21 Wide Range B (-150 to -60 n) LR-3683B Nanow Range B (0 to +60 in J 1 A2283 LR-RSZM-E21 Wide Range A (-150 lo +60 in ) LR-RS23B-EI21 Wlde Range B (-150 to 60 in )

• LRJBBW Namm Range A (0 to +60 in )

• LR-P616B21 FuelZonsA(-311 to-111 in) A2284 * - - LI-RSlD-BZ1 FualZona B (-311 10.111 In.) LR-Rt3235821 We Range B (-150 lo +EO In.) LR-36BJB Narrow Range B (0 to +60 In.) EOP C4t.ITION 1

ES-401 Sample Written ijxamination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 Group # 2 WA # 295033 EA2.01 Importance Rating 3.9 EA2.01 -Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS:

Area Radiation levels Proposed Question:

SRO 83 Page 169 of 205

, 1 I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet The plant is operating at rated power. A Backwash of Clean-up Filter Demineralizer AF-203 has just been completed.

Transfer of the RWCU Backwash Receiver Tank to Radwaste is in progress.

A catastrophic failure of Backwash Transfer pump IAP-214 suction line causes a spill into the Reactor Building.

All attempts to isolate the leak have been unsuccessful. Reactor Building Area Radiation conditions are as follows: Reactor Building Area Radiation 9RX706 Reactor Cleanup Monitor Demin. Sys. Equipment 9RX723 Outside Reactor Bldg. Sample Station 9RX708 Sample Station Other Reactor Building Area Radiation Monitors Beginning of Shift 2 mr/hr 3 mr/hr 3.5 mr/hr 2 to 5 rnr/hr Current Conditions 2400 mrlhr - In Alarm 1100 mr/hr - In Alarm 4500 mr/hr - In Alarm 3 to 7mr/hr - NOT In Alarm Which one of the following is the required action?

A. Commence a normal reactor shutdown to cold shutdown IAW 1O.ZZ-0004.

B. Continue reactor operation and attempt to stop the tank drain line leakage IAW SO.BG-0001.

C. IAW EOP-O103/4, Runback Recirc, Initiate a Manual Scram and Emergency Depressurize the RPV. D. IAW EOP-010314, Runback Recirc and Initiate a Manual Scram. Emergency Depressurization is NOT required.

Proposed Answer: A Explanation (Optional):

RWCU Backwash Receiving tank is not a primary System, with 2 areas > Max Safe Operating Limit, Plant shutdown and cooldown per 10-004 is applicable.

A. Correct. Page 170 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet B. C. D. Incorrect. Per EOP 10314 and since the leak is not from a primary system, plant shutdown and cooldown applies Incorrect.

Only applicable for a primary system leak Incorrect.

Only applicable for a primary system leak Technical Reference(s)

EOP 103/4 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

EOP 103 no entrys no retainment stem Learning Objective:

EOPI 03E006 (As available) Question Source:

Question History: Question Cognitive Level: 10 CFR Part 55 Content: 54264 __ Bank # Modified Bank # _. (Note changes or attach parent)

New _. Last NRC Exam Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 55.43 5 Comments: Page 171 of 205 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # KIA # 29501 7 G2.4.21 Importance Rating 4.6 1 ___~ Group # 2 Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. Proposed Question:

SRO 84 Given the following conditions:

A High Off-site release event is in progress 0 Various RMS points are indicating elevated values but less than the RM-11 ALERT values The Iodine release rate must be determined Which of the following describes the procedure to be entered and the method used to determine the total release rate?

A. Enter AB.CONT-0004 "Radioactive Gaseous Release".

Use RMS values for all inputs to the release rate formula.

B. Enter AB.CONT-0004 "Radioactive Gaseous Release". Use RMS values for FRVS and HTV. An Iodine sample must be taken for the NPV and SPV. C. Enter OP-AR.SP-0001 "Radiation Monitoring System Alarm Response". Samples must be taken for all inputs to release rate formula.

D. Enter OP-AR.SP-0001 "Radiation Monitoring System Alarm Response". Use RMS values for the NPV, SPV, and FRVS. An Iodine sample must be taken for the HTV. Proposed Answer:

A Explanation (Optional):

IAW AB.CONT-0004 , OP-AR.SP-0001 may be entered but the parameters are specified in the abnormal A. CORRECT - The FRVS, NPV, SPV & HTV sample skids all have Iodine Monitors that can be used for the calculation.

B. INCORRECT - The NPV and SPV have iodine monitors and sampling is not specified in Page 172 of 205 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet the procedure.

INCORRECT - The NPV, SPV & HTV sample skids all have Iodine Monitors that can be used for the calculation. Samples are not required for those values. INCORRECT - The FRVS sample skid does not have an Iodine Monitor so a sample must be taken. The HTV sample skid has an Iodine Monitor.

C. D. Technical Reference(s) AB.CONT-0004 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

ABCNTE005 (As available) Question Source:

Bank # 6441 6 Question History: Question Cognitive Level:

10 CFR Part 55 Content: Modified Bank # ~. (Note changes or attach parent)

-. New Last NRC Exam -. Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 55.43 4 Comments: Page 173 of 205 H C . 0 P -A B . CO NT-0004( Q) RADIOACTIVE GASEOUS RELEASE ADDITIONAL INFORMATION:

Procedures:

0 HC.RP-AR.SP-0001 (a), Radiation Monitoring System Alarm Response 0 HC.OP-SO.GU-0001 (a), Filtration Recirculation and Ventilation System Operation TABLE "fill (Iodine release rates can be calculated by substituting Iodine values for Noble Gas) CALCULATE the total Noble Gas release rate from Hope Creek Generating Station by adding all gaseous effluent channels:

__~. - + + + - _____.- pCi/sec pCi/sec pCi/sec pCi/sec pCi/sec SPV NPV FRVS HTV Total (9RX580) (9RX590) (9RX680)

(9RX518) - IF the effluent (pCi/sec) channel on the RM-11 is Nf3T operating for a specific plant vent, THEN CALCULATE the Noble Gas release rate for that vent using the following:

• 472 =

• pCi/cc (n.g.) Plant Vent Exh pCi/sec (n.g.) Flow in cfm Where: pCi/cc (n.g.) The concentration of Noble Gas obtained from the RM-11 (the operable channel will be highlighted in GREEN) - OR from an actual sample of the plant vent 472 The conversion factor in units of cc/sec/cfm pCi/sec (n.g.)

The calculated release rate from the specified plant vent (Noble Gas)

Hope Creek Page 6 of 19 Rev. 2 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 1 KIA # 29501 5 G2.4.30 I m porta rice Rating Group # 2 4.1 Emergency Procedures

/ Plan: Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as State, the NRC, or the transm ission system operator (incomplete scram) Proposed Question:

SRO 85 The plant was operating at 100% when an existing leak in the drywell worsened. The operators scrammed the plant prior to reaching the High Drywell scram setpoint.

The following conditions now exist: 0 0 0 0 0 Two peripheral control rods are at position 48 ALL APRM downscale lights are lit Drywell Pressure is 1.80 psig and slowly rising RPV level is +20 inches and slowly rising RPV pressure is 910 psig Which one of the following describes the NRC notification required?

A. 50.72 - 1 Hour Report B. 50.72 - 4 Hour Report C. Alert D. SAE Proposed Answer: C Explanation (Optional): ECG Section 5.1.2.b manually initiated scram unsuccessful

= Alert C. Correct A. B. D. Incorrect.

Does not meet ECG Section 11 .I for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reports Incorrect.

Correct for actuation of RPS ONLY - ECG 11.3.2 Incorrect.

Correct only if power remained about 4% Technical Reference(s)

ECG section 5.1.2. b (Attach if not previously provided)

Page 174 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

ECG, no attachments Learning Objective:

Question Source: (As available)

Bank # Question History: Question Cognitive Level:

10 CFR Part 55 Content: Modified Bank # ___ (Note changes or attach parent)

New X Last NRC Exam ~ Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 55.43 1 Comments: Page 175 of 205 Initiating Condition OPCON EAL # 1 E M E R G E N C Y A C T I 0 N L E Y E L S 1 I I I I I I I I I I I I I i I I I I I 1 I I I I I I I I I I I 1 I I I I I 5.0 Failure to Scram 5.1 ATWS System (RF'S) to Successfuliy Complete a Reactor Scram (Automatic and Manual) Failure of the Reactor Protection System OCpS) to Successfully Complete a Reactor Scram (Automatic or Manual) (T-) 5.4 2.a IF AnAutamatic i Reactor Scram Condition exists An Automatic

'i Reactor Scram c1p\fsj IS NOT successful I I AND (7) 5.1.3 ANY Manually 1 Initiated Reactor Scram (RPS) from the Control Room IS NOT silccessfrrl I ALL Reactor Scram attempts (RPS and ARI) DID NOT Reactor Power to 5 4% AND from the Control Room HCGS ECG Rev. 02 Pqc 1 of 1 / Failure of the Reactor Protection System (RFS) \ AND I-GZ to Successfully Complete a Reactor Scram \ (Automatic and hlanual) and there is indication of an Extreme ChaUcnge to the Ability to Cool the Core/ 5.1.4 ~ EITHER one of the following:

Reactor Water Level CANNOT BE RESTORED and MAINTAINED above -185" The combination of Suppression Pool Temperature and RPV Pressure CANNOTBE h.IAINTAJNED belowthe HCTL Curve 1 THEN I I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # KIA # 206000 2.2.37 1 -~ Group # 2 I m portanee Rating -___ 4.6 Equipment Control: Ability to determine operability and/or availability of safety related [equipment. (HPCI) Proposed Question: SRO 86 Given the following:

0 0 0 0 0 The plant is operating at 100% reactor power. HPCl Pump IS1 test is in progress at rated flow.

HPCl discharge pressure is 1150 psig While attempting to adjust pump flow, the flow controller setpoint remains stationary at 4000 gpm in AUTO The PO reports the HPCl flow controller works in MANUAL and develops rated flow.

What effect does this have on HPCl Operability at the PRESENT time? A. HPCl is operable because it can develop rated flow.

B. HPCl is "operable but degraded" because it has lost its testing capacity.

C. HPCl is inoperable because it is NOT capable of meeting all surveillance requirements.

D. HPCl is "operable but non-conforming" because it is NOT capable of meeting all surveillance requirements. Proposed Answer: C Explanation (Optional):

TS 3.5.1 C. Correct - HPCl must be in AUTO with a setpoint of 5600 gpm and capable of rated flow and discharge pressure. C. Correct.

A. B. D. Incorrect - It must develop rated flow in AUTO Incorrect - The case could be made if flow in AUTO remained stationary at 5600 gpm. Incorrect - operable but non-conforming does not apply with flow at 4000 gpm in AUTO. Technical Reference(s)

TS 3.5.1 (Attach if not previously provided) Page 176 of 205 ES-40 1 Sample Written Examination Form ES-40 1

-5 Question Worksheet Proposed references to be provided to applicants during examination:

None Learning Objective: Question Source:

HPCIOOE018 (As available)

Bank # Modified Bank # -. 55949 (Note changes or attach parent)

New Question History: Last NRC Exam -. 2003 Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments:

Page 177 of 205 EMERGENCY CORE COOLING SYSTEMS .~ SURVEILLANCE REQUIREMENTS 4.5.1 The emergency core cooling sysyems shall be demonstrated OPEWLE bl7: a. At least once per 31 days: 1. For the core spray system, ::he LPCI system, and the HPCI system: a) Verifying by venthg al: the high point vents that the system piping from the pump discharge valve to che system isolation valve is fil Led with water. b) Veri-fying that each va:Lve, manual, power operated or automatic, in the flow path that is not l.ocked, sealed, or otherwise secured in position, is in its correct* position.

c) Verjfy the RHR System cross tie valves on the discharge side of the pumps are closed and power, if any, is rerroved from the valve operatocs.

2. For the HPCI system, verifymg that the HPCI pump flow controller is in the correcl: position.

b. Verifying that, when tested pursl:.ant to Specification 4.0.5: 1. The two core spray system pamps in each subsystem together develop a flow of at least 6150 gpm against a test line pressure corresponding to a reactor vessel pressure of 1105 psi above suppression pool pressure.

2. Each LPCI pump in each subsvstem develops a flow of at least 10,000 gpm against a test lme pressure corresponding to a reactor vessel to pri-mary cmtainment differential pressure of - >20 psid. . The HPCI pump develops a flow of at least 5600 gpm against a test line pressure corresponding to a reactor vessel pressure of 1000 psig when steam is beiag supplied to the turbine at 1000, +20, -80 psig.** c. At least once per 18 months: 1. For the core spray system, Lhe LPCI system, and the HPCI system, performing a system functional test which includes simulated automatic actuation of the <system throughout its emergency operating sequence and veriEying that each automatic valve in the flow path actuates to its correct position.

Actual injection of coolant into the reactor vessel may be excluded from this test. *Except that an automatic valve capable 0:: automatic return to its ECCS position when an ECCS signal is present :nay be in position for another of operation.

• • The provisions of Specification 4.0.4 arc? not applicable provided the surveillance is performed within 12 hour::: after reactor steam pressure adequate to perform the test. mode is HnPF PRFFU ?/A 5-A nmendmenf 1\Tn 126 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 KIA # 209001 G2.2.40 Group # 1 Importance Rating 4.7 Equipment Control: - Ability to apply Technical Specifications for a system. Proposed Question:

SRO 87 The plant is operating at rated power. The following alarm is then received:

B3-C1 "CORE SPRAY LOOP A TROUBLE" I&C Technicians report that 1 -BE-PT-N054A, Core Spray Loop A Header Pressure transmitter is failed. Which one of the following describes the Technical Specification action time(s) of this failure? A. Declare Core Spray Loop A inoperable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 6. Declare Core Spray Loop A inoperable within seven days. C. Restore the transmitter to operable status within thirty days and verify Core Spray Loop A pressure is less than 475 psig every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. D. Restore the transmitter to operable status within seven days or verify Core Spray Loop A pressure is less than 475 psig every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Proposed Answer: D Explanation (Optional):

D Correct: Restore to operable status within seven days or verify Core Spray Loop A pressure less than 475 psig every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for thirty days. T.S. 3.5.1, Action f. including

• . Then T.S. 3.4.3.2, Action d. The stated transmitter feeds both Hi-Lo pressure interface alarm and Keepfill low pressure alarm. A. Incorrect.

Declare CS loop A inoperable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Not required if pressure verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. Incorrect: Declare CS loop A inoperable within 7 days. 30 days to restore provided Page 178 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet pressure verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Incorrect:

Restore to operable within 30 days or verify pressure less than 475 every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Verify every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. C. Technical Reference(s) TS3.5.1.

& 3.4.3.2 (Attach if not previously provided) Proposed references to be provided to applicants during examination: TS 3.5.1 & 3.4.3.2 M-52 Sht.1 Learning Objective:

CSSYSOE014 (As available)

Question Source: Bank

1. 551 39 (Note changes or attach parent)

Modified Bank # -. __. New Question History: Last NRC Exam _I Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2 Comments: Page 179 of 205 B7 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems shall be OPERABLE with: a. The core spray system (CSS) consisting of two subsystems with ezch subsystem comprised of: 1. Two OPEWLBLE core sp.ra;y pumos, and 2. An OPERAHLE flow path capable of taking suction from the suppression chamber and tra9sferring the water through the spray sprirger to the reactor vessel.

b. The low pressure coolant irljectilm (LPCI) system of the residual heat removal system consist-ing of four subsystems with each subsystem comprised of: 1. One OPERABLE LPCI pump, and 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel. c. The high pressure coolant injection (HPCI) system consisting of: 1. One OPERABLE HPCI pmp, and 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel. d. The automatic depressurizati-on system (ADS) with five OPERABLE ADS valves. APPLICABILITY:

OPERATIONAL CONDITION 1, Zr, ** #, and 3*, **, ##. *The HPCI system is not required to be OP3RABLE when reactor steam dome pressure is less than or equal to 200 ps-g. **The ADS is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig. #See Special Test Exception 3.10.6. ##Two LPCI subsystems of the RHR system ma:y be inoperable in that they are aligned in the shutdown cooling mode when the reactor vessel pressure is less than the RHR shutdown cooling permissive setpoint.

HOPE CREEK 3/4 5-1

, I REAETOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to: a. No PRESSURE BOUNDARY LEAXAGE. b. 5 gpm UNIDENTIFIED LEAKAGE. C. 25 gpm IDENTIFIED LEAKAGE aveyaged over any 24-hour period. d. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm from any reactor coolant- system pressure isolation valve specified in Table 3.4.3.2-1, at rated pressure.

e. 2 gpm increase in UNIDENTIFIEI:)

LEAKAGE within any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less. APPLICABILITY:

OPERATIONAL CONDITIONS 11, 2 and 3. ACTION : a. b. C. 0 e. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SI-!UTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at 1e;ist HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With any reactor coolant systcm pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one other closed manual or deactivated automatic or check* valves, 01 be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br />. With one or more of the high/low pressure interface valve leakage pressure monitors shown insle 3.4.3 7 -2, inoperable, restore the inoperable monitor(s) to OPERTiBLE status within 7 dacor verify the pressure to be less than the alarm setpoint at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; restore the inoperakle monitor(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With any increase in UNIDENTIFIED LEAKAGE exceeding the limit in e above, implement preplanned leak location and isolation actions and either veriiy that the source of the leakage is not service-sensitive type 304 or 316 stainless steel or reduce the leakage rate-of-change to less than the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• Which have been verified not to exceed the allowable leakage limit at the last refueling outage or after last time the valve was disturbed, whichever is more recent. HOPE CREEK 3/4 4-1.1 Amendment No.

51 SERVICE INSTRUMENT --Fore Spray 1-BE-PISH-N654A Core Spray l-BE-PISH-N654B LPCI/RHR l-BC-PISH-N653A LPCI /RHR l-BC-PISH-N653B LPCI /RHR 1 -BC-PTSH-N653C LPCI /RHR l-BC-PISH-N653D RHR l-BC-PISH-N657 TABLE 3.4.3.2-2 REACTOR COOLANT SYSTEM INTERFACE VALVES LEAKAGE PRESSURE MONITORS ALARM ALARM (psig) VALUE (psig) SETPOINT ALLOWABLE 2 500 2 500 @ 475 380 I 410 380 < 410 380 5 410 380 5 410 130 2 155 I , I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 1 K/A # 21 1000 G2.4.47 Importance Rating

4.2 Emergency

Procedures

/ Plan: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.(SLC) Proposed Question:

SRO 88 Given the following: A feedwater line break in the drywell has occurred Reactor Power is 3.5% and steady Reactor Pressure is cycling between 800 to 1000 psig RPV level is -1 00 inches and lowering at 1" per minute HPCl is in service Drywell Pressure is 3.8 psig and rising at 0.1 psig per minute Drywell Temperature is 168 degrees F and rising at 3 degrees per minute Suppression Pool Temperature is 112 degrees and rising at 4 degrees per minute Assuming the trends continue as above and all systems are operable, which one of the following idwill be required?

A. Immediately lower RPV Level until it reaches -129" IAW EOP-IOIA "ATWS-RPV Control" B. Inject SLC before 7 minutes has elapsed IAW EOP-IOIA "ATWS-RPV Control" C. Emergency Depressurize in 5 minutes IAW EOP-202 "Emergency Depressurization" D. Terminate

& Prevent Injection in 6 minutes IAW EOP-IOIA "ATWS-RPV Control", once that is complete Emergency Depressurize IAW EOP-202 "Emergency Depressurization" Proposed Answer: B Explanation (Optional): IAW EOP-101A B. Correct - in 7 minutes, SP temperature will be '140 degrees. IAW the EOPIOIA, Step RC/Q-10, SLC must be injected before SP temp reaching 140 degrees A. C. Incorrect. With power

<4%, level is not lowered Incorrect.

No parameters will have met the ED requirement Page 180 of 205 I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet D. Incorrect.

Terminate

& Prevent would occur only if level could not be maintained above

-- -185 inches Technical Reference(s) EOP-101 A (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: E0101AE006

-. (As available)

Question Source: Bank

1. -. Modified Bank # New X (Note changes or attach parent)

Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Page 181 of 205 1 I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO 2 -___ Tier # Group # 1 KIA # 205000. 2.2.22 Importance Rating 4.7 (K&A Statement) Equipment Control: Knowledge of limiting conditions for operations and safety limits. (Shutdown Cooling) Proposed Question: SRO 89 The plant is 3 days into a shutdown. Reactor cavity level is 23 feet above the RPV head flange. Which one of the following is the minimum Shutdown Cooling systemslcomponents required by Technical Specifications for these conditions.

A. One loop of Shutdown Cooling consisting of one OPERABLE RHR pump and one OPERABLE RHR Heat Exchanger in operalion.

B. Two loops of Shutdown Cooling each consisting of one OPERABLE RHR pump and one OPERABLE RHR Heat Exchanger in operation.

C. Two loops of Shutdown Cooling OPERABLE and one Recirculation Pump in operation.

D. NO Shutdown Cooling loops OPERABLE arid one Recirculation Pump in operation.

Proposed Answer: A Explanation (Optional):

A. Correct - IAW TIS 3.9.1 1 .I B. C. D. Incorrect - this would be correct for low water level TS 3.9.1 1.2 Incorrect - not correct for given conditions, not the minimum.

Incorrect - this is the action requirements for TS 3.4.9.2 Technical Reference(s): T.S. 3.9.1 1 .I (Attach if not previously provided) Proposed references to be provided to applicants during examination: none Learning Objective:

IOP009E006 (As available)

Question Source: Bank # Page 182 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Modified Bank # (Note changes or attach parent)

New X Question History: -. Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments: Page 183 of 205 I P+ REFUELING OPERATIONS 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLAN-C CIRCULATION

_____- HIGH WATER LEVEL LIMITING CONDITION FOR 3PERATION 3.9.11.1 At least one shutdown cool.ing mode loop of the residual heat removal (RHR) system shall be OPERABLE and in operation*

with: a. One OPERABLE RHR pump, and b. One OPERABLE RHR heat exchanger.

PPLICABILITY:

OPERATIONAL CONDITION 5, when irradiated fuel is in the reactor the top of the reactor pressure vessel flange and heat Tosses to ambient**

are not sufficient to maintain OPERATIONAL CONDITION

5. 4 vessel and the water level is greater than or equal to 22 feet 2 inches above

a. With no RHR shutdown cooling mode locip OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal. Otherwise, suspend all operations involving an increase in the reactor decay heat load and establish SECONDARY CONTAINYENT INTEGRITY within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. b. With no RHR shutdown cooling mode loop in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour. SURVEILLANCE REQUIREMENTS 4.9.11.1 At least one shutdown cool.ing mode loop of the residual heat removal system or alternate method shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• The shutdown cooling pump may k'e removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8-hour period. **Ambient losses must be such that no increase in reactor vessel water temperature will occur (even though REFUELING conditions are being maintained) . HOPE CREEK 3/4 9-17 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group Xt 1 WA # 218000 A2.05 Importance Rating 3.6 (K&A Statement) Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM : and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of A.C. or D.C. power to ADS valves Proposed Question: SRO 90 The reactor scrammed from 100%

power due to a loss of offsite power with a LOCA. The following conditions exist:

0 0 0 0 0 0 All EDGs started, however busses 10A401, 1OA402, 1OA403, and 10A404 all have bus lockouts (due to ground faults). Reactor pressure is 400 psig and lowering Reactor water level has reached -185 inches and is lowering at 1 inch/minute.

RClC is injecting at rated flow.

HPCl is injecting and reaching rated flow. Drywell pressure is 12.5 psig and slowly rising. Which one of the following is the result of these conditions and what actions are required?

A. The ADS valves are available, enter EOP-206 "RPV Flooding" and open five SRVs and flood the RPV to the Main Steam Lines. 6. The ADS valves are available, enter EOP-202, "Emergency RPV Depressurization", and open five SRVs and depressurize the RPV. C. The ADS valves are NOT available, enter EOP-202 "Emergency RPV Depressurization", and use Alternate Depressurization Systems to depressurize.

D. The ADS valves are NOT available, enter EOP-206 "RPV Flooding" and use Alternate Depressurization Systems to depressurize and flood the RPV to the Main Steam Lines. Proposed Answer: B Page 184 of 205 I I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Explanation (Optional):

B. Correct - although no RHR or Core Spray systems are available, HPCI and RClC are both preferred systems and although they will go away when the RPV is depressurized EOP-101 directs a blowdown when an injection system is available and RPV level cannot be maintained

>-I 85 inches. A. C. D. Incorrect - There is no need to enter EOP-206 at this time. Incorrect - The ADS valves are available (the auto ADS function is NOT available but the SRVs can be manually opened.

Incorrect - The ADS valves are available (the auto ADS function is NOT available but the SRVs can be manually opened. There is no need to enter EOP-206 at this time. Technical Reference(s):

EOP-101 step ALE-9,10,11 (Attach if not previously provided) Proposed references to be provided to applicants during examination: none Learning Objective:

E01 01 LE006 (As available) Question Source: Bank

1. Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Page 185 of 205 I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 KIA # 245000 A2.02 Group # 2 Importance Rating 3.5 Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS

and (b) based on those predictions, use procedures to correct, control, or mitigate the mnsequences of those abnormal conditions or operations

Loss of lube oil Proposed Question:

SRO 91 The plant was operating at 100% power in a normal system alignment when a steam leak in the Drywell causes pressure to rise. The Reactor is shutdown prior to Drywell pressure reaching 1.68 psig and all SCRAM actions are carried out by the RO. Ten minutes after the scram the following conditions exist: Drywell Pressure is 2.1 psig and slowly rising RPV level is -40" and rising, with HPCl and RClC injecting RPV pressure is 800 psig and lowering slowly Turbine Speed is 200 RPM and lowering Further investigation identifies all Lift Pumps and the Turning Gear Oil Pump are NOT running. Which one of the following describes the plant response and actions required, if any? A. B. C. D. The TGOP failed to AUTO start and must be manually started to allow the Lift Pump to start IAW HC.OP-AB.BOP-0002 "Main Turbine". The Lube Oil system has an apparent leak causing the TGOP and lift pumps to trip. The NE0 should be sent to investigate IAW HC.OP-SO.CB-0001 "Main Turbine and Generator Lube Oil System Operation".

The plant responded as designed due to the High Drywell pressure.

The operator must verify the EBOP is running IAW the overhead alarm response for window C8-F3, Digital Point D5573, "Turning Gear Oil Pump Trouble".

The plant responded as designed due to the Main Turbine speed coasting down. The operator must verify the TGOP and Lift Pumps start when Turbine speed reaches -c 100 RPM IAW HC.OP-SO.CB-10001 "Main Turbine and Generator Lube Oil System Operation".

Page 186 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed Answer: C Explanation (Optional):

C. Correct-: IAW HC.OP-SO.SM-0001 the 108323 load center is stripped on a LOCA 1 signal the TGOP is de-energized resulting in the Lift pumps tripping. Therefore the EBOP must be verified as running for the turbine lube oil system IAW the alarm response procedure.

A. Incorrect- IAW HC.OP-SO.SM-0001 the 1 OB323 load center is stripped on a LOCA 1 signal the TGOP is de-energized.

B. Incorrect-without the TGOP running and suction pressure

> 1 psig the Lift pumps trip. D. Incorrect- the LIFT pumps and TGOP due not rely on Turbine speed for a start signal, the TGOP starts on pressure and the lift pump starts on the TGOP and > 1 psig suction pressure. Technical Reference(s)

HC. OP-AB. BOP-0002 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

MTLOOOEOI 1 (As available) Question Source:

Question History: Question Cognitive Level:

10 CFR Part 55 Content: 56902 -. Bank # Modified Bank # New (Note changes or attach parent)

-. Last NRC Exam -. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 55.43 5 Comments: Page 187 of 205 PSEG Internal Use Only pL SUBSEQUENT OPERATOR ACTIONS (continued)

.___ CONDITION E. Hydrogen Seal Oil System Malfunction.

Datenime:

.\Generator H2 pressure drops fh 1 below 63 psig. 'L ' Datenime:

G. Both the Emergency Seal Oil Pump AND The Main Seal Oil Pump CANNOT be started, - OR BEFORE Generator H2 pressure drops below 30 psig. Datemime: (Continued on Page 19) H C . 0 P-A B . BO P-0002 ( C, ) MAIN TURBIN E - - ________ ______ ACTION - - **NOTE 9** the Emergency Seal Oil Pump fails to auto start, - E.l I- THEN START the Emergency Seal Oil Pump.

F.l Reactor Power 224% -- THEN PERFORM the following: - a. REDUCE Recirc. Pump speed to MINIMUM. - b. LOCK the Mode Switch in Shutdown. - F.2 TRIP the Turbine -- AND LOCK OUT the Generator.

$k CAUTION2 %k PUCE BOTH Stator Water Cooling Pump Control Switches to PULL TO LOCK. [CD-I97G]

OPEN Breaker 52-1 121 13 Stator Coolant Water Heater 1 OEl13 power supply. - - F.3 - F.4 G. 1 E Reactor Power 224% -- THEN PERFORM the following: - a. REDUCE Recirc. Pump speed to MINIMUM. - b. LOCK the Mode Switch in Shutdown. - G.2 TRIP the Turbine LOCK OUT the Generator.

G.3 PERFORM the following: - 0 EVACUATE unnecessary personnel from - 0 ENSURE Turbine Building Ventilation in Purge - 0 START all available Turbine Building Exhaust / Supply Fans. Turbine Generator Area (Continued on Page 19) Hope Creek Page 17 (of 26 Rev. I1

, r'LC H C .O P-A R.ZZ-0 029 (Q) DIGITAL ALARM POINT D5573 NOMENCLATURE TURNING GEAR OIL PUMP SETPOINT 10 psig TROUBLE DESCRIPTION TGOP auto start failure troubie

.- ORIGIN PSH-3 I 09 AUTOMATIC ACTION:

I. 2. 3. OPERATOR ACTION:

1. E Main Turbine is on Turning Gear, ENSURE Turning Gear trips (TURN GEAR MOT AMPS indicates 0). 2. ENSURE LIFT BEARING PUMPS STOP is ON.

ENSURE Motor SuctipnPump

4. E no bearing oil headet; pressure is available, STOP Main Turbine shaft rotation as quickly as possible. Turning Gear Oil Pump auto start has occurred. Loss of Lift Pumps running signal causes Turning Gear to trip. Loss of discharge pressure signal from TGOP causes Lift Pumps to trip. 9' -0J Emergencv Bearina bil PumD is running and supplying bearing header CAUSE 1. Discharge pressure low WHEN a start signal is present

a. b. TGOP LOCKOUT c. Loss of bus power

d. Loss of control power Failure of TGOP to start on auto start signal

1. Blown control power fuse Associated with Annunciator C8 F3

REFERENCES:

J-200-Q-181-10 J-19-0 Sht. 4 CORRECTIVE ACTION 1 A. 1 B. I C. 1 D. PRESS START on Lube Oil Reservoir F'umps Turning Gear. (1 OC651 C) CHECK Turning Gear Oil Pump is not in LOCKOUT. -- IF TGOP starts, PRESS LOW DISCH PRESS to clear amber light. (10C651C)

DETERMINE cause of auto start failure. REQUEST SM/CRS to initiate corrective action. E-0252-0 Sht. 1 Hope Creek Page 8 of 228 Rev. 8 CAUSE H C . 0 P-A R.ZZ-0 029 (a) DIGITAL ALARM POINT D5573 Discharge pressure low WHEN a start signal is present (continued)

Associated with Annunciator C8 F3

REFERENCES:

J-200-Q-181-10 J-19-0 Sht. 4 CORRECTIVE ACTION 1 E. 1 F. DISPATCH an operator to MCC 10B323. Check TGOP breaker on (52-323074).

RESET thermal overload ENSURE all Lift Pumps are on. CHECK bearing oil is supplied by visually inspecting Turbine Bearing Sightglasses.

RESTART TGOP. E-0252-0 Sht. 1 Hope Creek Page 9 of 228 Rev. 8 H C.OP-SO.SM-000 I (a) -, OBSERVE I, ,e Equipment or Valves listed in Table SM-020 have met their required Action under the Manual or Automatic Isolation Signals "52-44014 CRD PMP BP207 BRKR TRIP D CS X X "52-41014 MCC 10B313 BRKR TRIP A CS X X TRIP B CS X X "52-42014 MCC 10B323 BRKR "52-4701 1 1 "52-48011 I MCC 10B282 BRKR I TRIP 1 DCS 1 X I 1 X 1 I MCC 1 OB272 BRKR TRIP c cs X X I "52-4501 1 I MCC 10B252 BRKR 1 TRIP I ACS I X I 1 X 1 I 1 "52-46011 1 :CC lOBr2BRKR 1 TRIP 1 BCS 1 1 E 1 1 "52-47031 MCC 00B474 BRKR TRIP c cs A CS52-451 023 PA SYST INVERTER 1 OD496 BIU PWR TRli13 ISOLATION S ETPOl NT A - REACTOR VESSEL WATER LEVEL 1 -1 29" C - DRYWELL PRESSURE - HIGH D - REACTOR BUILDING EXHAUST RADIATION - HIGH 1.68 psig I x 10" uCi/cc + - Also Isolates from the corresponding Core Spray Manual Initiation PB

• - Can receive a Half Isolation from the corresponding NSSSS Manual Isolation " - Isolation can be bypassed by TRIP OVRD PB Hope Creek Page 30 of 4.2 Rev. 17 I I I ES-40 I Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 KIA # 202001 G2.4.49 Group ## 2 Importance Rating

4.4 Emergency

Procedures

/ Plan: Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (recirculation) Proposed Question:

SRO 92 The plant is operating at 100% power, when a lighting strike caused the "A Reactor Recirc pump to trip, the "B" Reactor Recirc Pump to runback to it's intermediate speed and isolation of 6A FW Heater. All other plant equipment is operable.

Currently Rx power is 56% and Core flow is 41%. Which of the following describes the action required, if any, to ensure core stability?

A. Unintentional operation in this region is allowed, NO further actions are required.

B. IAW OP-SO.BB-0001 "Recirculation System Operation", Start the "A Reactor Recirc Pump to exit this region.

C. Exit Region 1 using Enhanced Stability Guidance IAW OP-AB.RPV-0003 "Recirculation SystemlPower Oscillations".

D. Immediately lock the mode switch in shutdown IAW OP-AB.RPV-0003 "Recirculation System/Power Oscillations". Proposed Answer:

C Explanation (Optional):

HC. OP-AB. RPV-0003 C. Correct - Insert control rods IAW RE guidance to exit this region.

A. Incorrect Unintentional operation in this region is not allowed, immediately Lock the Mode Switch in Shutdown.

Incorrect - Mode switch is not required Locked in Shutdown and rod insertion is method to exit this region Incorrect - Start the "A Reactor Recirc pump to exit this region - Starting an idle Recirc pump to exit this region is NOT IAW with the AB. Incorrect - Unintentional operation in this region is allowed, no further actions are required. Unintentional operations in this region are not allowed and actions to exit this region are required.

B. D. Page 188 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Technical Reference(s) HC.OP-AB.RPV-0003 (Attach if not previously provided) Proposed references to be provided to applicants during examination: Power-to-flow map OPRMs operable/inoperable - reference does not state the action required but is needed to determine the action Learning Objective:

Question Source: Question History: Question Cognitive Level:

10 CFR Part 55 Content: IOP003E005 (As available) 56978 -. Bank # Modified Bank # -. (Note changes or attach parent)

New -. Last NRC Exam -- Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 55.43 5 Comments: Page 189 of 205 I I I PSEG Internal Use Only H C .OP-A B. RPV-0003( 6 ) RECIRCULATION SYSTEMlPOWER OSCILLATION S CONDITION

.- B. Unplanned entry resulting in Reactor Power 2 26.1 % - AND Core Flow I 68% (nominal) . Datemime:

Continued on Page 11 ACTION 0 B.l MONITOR for Power Oscillations. (Therm a I H y d ra u I i c I n s t a bi I i ti es) rT1 **NOTE 3** 0 8.2 VERIFY OPRM TRIP ENABLE (C3-F1) alarm.

0 $f CAUTION2

• 8.3 E OPRM's are OPERABLE, THEN PERFORM the following:

n-. /'A. JF in REGION 1 of Figure 2 0 i-' -'??HEN -- EXIT REGION 1 IAW Enhanced Stability Guidance. (RE.ZZ-0001)

B. E OPRM ALARM (C3-F2) is received -- THEN PERFORM the following:

0 SUSPEND Control Rod withdrawal AND Recirculation Pump speed reduction.

0 applicable, CONTINUE operation WITH Enhanced Stability Guidance. (RE.ZZ-0001) 0 C. OPRM ALARM (C3-F2) will NOT remain Clear, - OR MORE THAN one OPRM Channel has an "OF'RM Alarm" (REFER to CRIDS Page 248) THEN CONTINUE Rod insertion IAVJ Enhanced Stability Guidance

-- UNTIL the alarm is clear. (RE.ZZ-0001)

• • NOTE 4** D. WHEN OPRM ALARM (C3-F2) is CLEAR, PRESS the OPRM LED reset at Panel 10C608.

Continued on Page 11 Hope Creek Page 9 of 41 ~~ Rev. 19 PSEG Internal Use Only HC . 0 P-AB. RPV-0003( Q) RECIRCULATION SYSTEM/POWER OSCILLATIONS RETAINMEN' CONDITION

1. DANGER Limit is reached on Recirc Pump/Motor Vibrations analog points:

0 A(B) Radial PMP A2601 (A2603) - 16.0 mils 0 A Radial MTR A2602 - 4.5 mils 0 B Radial MTR A2604 - 8.0 mils DateiTime: Affected Recirc Pump seal cavity temperature greater than 200 degrees F DateiTime:

II. WERRIDE =__ ACTION 0 1.a TRIP the affected Recirc Pump. c;I 1.b ENTER Condition A.

0 1l.a REDUCE the affected Recirc pump speed to minimum, TRIP the affected Recirc pump and enter Condition A __. -. -. CAUTIONS:

2. NOTES: 3. DO NOT intentionally enter the REGION 1 while changing power. Passing through this region or lowering flow once in REGION I is acceptable if the intent is to TAKE the Mode Switch to shutdown.

[PR 971229163]

68% core flow is the nominal value to enable the alarm. To ensure the OFRM enables during all conditions, e.g.

SLO, the value has been set at 68% core flow. It is possible to receive the Trip Enable Alarm at a higher value, and may require up to 73% core flow to clear the alarm. The OPRM ALARM overhead alarm (C3-FZ) and associated CRIDS alarms will clear automatically when the condition clears. The OPRM Alarm LED on Local Panel 1OC608 is a "lock in" indication and requires physically resetting when the condition clears.

4. ADDITIONAL INFORMATION:

Indications:

CRIDS PAGE 248 OPRM Alarm, Red LED @ Panel 10C608 Figure 2 -POWER TO FLOW MAP (OPRM OPERABLE) 110 - , 11,1111 I,, 105 ~~~ -2 ~ ~ i ~ -I-. -1. i. L - J~~-!. _'.--'-~I - ~ L--i- -J-- L - ~ L -L- ~ ,Illlllll~Il!I/l/Ill Ill// /I// 0- , , ~ ,,,,,I, 0 5 10 15 20 25 30 35 40 45 SO 55 60 65 70 75 80 85 90 95 100 105 110 YO Rated Core Flow Power Oscillations (Thermal Hydraulic Instabilities) are defined as any of the follcwing:

Periodic Upscale 10% peak to peak Power Oscillations on APRM Recorders Strong positive/negative swings on the Period Meters Downscale LPRM Alarms (on the Full Core Display) Hope Creek Page 8 of 41 Rev. 19 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 2 Group # 2 WA # 215001 A2.07 Importance Rating

3.7 Ability

to (a) predict the impacts of the following on the TIPS: and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Failure to retract during accident cond itions Proposed Question:

SRO 93 A Traversing lncore Probe (TIP) trace is in progress when a high drywell pressure event

(> 1.68 psig) occurs due to a leak in the recirculatiori system. Three minutes following the event, the Reactor Operator reports the following indications on the TIP Valve Control Monitor:

0 "SQUIB MONITOR lights .- both illuminated "SHEAR VALVE MONITOR lights .- both extinguished "BALL VALVE OPEN" lights .- both illuminated "BALL VALVE CLOSED" lights .- both extinguished Which of the following describes the status of the TIP system, the next required operator action(s), if any, and status of Primary Containment Isolation (in regard to TIPS ONLY)? (Assume NO operator actions have been taken) A. B. C. D. The system has responded as designed. Operator action is required to close the ball valves IAW OP-SO.SE-0002 "TIP System Operation", to ensure Primary Containment isolation. The system has responded as designed.

IALV OP-SO.SE-0002 "TIP System Operation", direct the operators to fire the shear valves to ensure Primary Containment has been isolated. The TIP detectors may NOT have withdrawn. IAW OP-AB.CONT-0002, Withdraw the detectors and verify the ball valves close. Primary Containment will be isolated once the ball valves are closed. The TIP detectors may NOT have withdrawn. IAW OP.AB-CONT-0002, Fire the shear valves, withdraw the detectors and then close the ball valves. Primary Containment is considered isolated ONLY after the ball valve has closed and shear valve has fired. Proposed Answer:

C Explanation (Optional):

C. Correct: The ball valve should be closed. A failure to auto retract could be the problem. The next action would be to attempt to withdraw the detectors and verify the ball valve closes. Page 190 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet A. B. D. Incorrect: Did not respond as designed.

The ball valve should be closed and the squib monitor light extinguished.

Incorrect: The system did not respond as designed. The ball valve should be closed and the squib monitor light extinguished.

Primary Containment is not isolated.

Incorrect: The next action is to attempt a manual withdrawal and close the ball valve.

The squib valves have lost continuity and will not fire. Primary Containment Isolation will occur once the ball valves are closed. Technical Reference(s) IAW 0P.AB-CONT-000% (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective:

ABPRCUEOO3 (As available) Question Source: Bank

1. -. Modified Bank # New -. NRC 2007 (Note changes or attach parent)

-. Question History: Last NRC Exam 2007 Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Page 191 of 205 I I I I CONDITION F. TIP CANNOT be isolated i.l& PSEG Internal Use Only ACTION I - ** NOTE 2** H C. 0 P-AB. CO NT-0002 (Q) PRIMARY CONTAINMENT manually AND it is a source of leakage [T/S 3.6.31 [CD-l18X] SUBSEQUENT OPERATOR ACTIONS (continued)

F.l F.2 At the discretion of the SM/CRS, FIRE the shear valve. VERIFY the following indications lit:

G. TIP fails to retract.

[TIS 3.6.31 Date/Time: - TIP SHEAR VALVE CLOSED/ INOP __ G.l RETRACTtheTIP AND CLOSE the Ball Valve manually - OR DIRECT I&C to manually retract the TIP ~ I AW HC. I C-GP.SE-0005( Q). Datenime: - 0 SQUIB MONITOR - 0 SHEAR VALVE MONITOR Hope Creek Page I1 of 15 Rev. 7 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # KIA # G2.1.35 Group # 3 Importance Rating 3.9 Knowledge of refueling responsibilities of SRO Proposed Question:

SRO 94 Plant conditions are as follows: The Reactor Vessel is prepared for refueling operations IAW "Cold Shutdown to Refueling", HC.OP-IO.ZZ-0005. Prerequisite plant conditions have been verified IAW "Refueling Operations" HC.OP- 1o.zz-0009.

Spiral Fuel offload is in progress per directions of Reactor Engineers and Fuel Handling Control Core Alteration forms HC.RE-FR.ZZ-0001.

Multiple Control Rod blades and drive mechanisms are being removed IAW Technical Specification 3.9.10.2 Then, Reactor Engineering reports Shutdown Margin CANNOT be demonstrated.

Which of the following are required?

A. B. C. D. Stop fuel handling in the fuel pool and return the Control Rod Blades to the reactor vessel. Then, remove the shorting links prior to resuming any fuel or Control Blade movement. Stop Control Rod Blade removal from the reactor vessel. Fuel handling in the fuel pool may continue. Control Rod Blade removal from the reactor vessel may continue once the shorting links are removed. Stop fuel handling in the fuel pool and return the Control Rod Blades to the reactor vessel. Then, install the shorting links prior to resuming either of the above activities. Stop Control Rod Blade removal from the reactor vessel. Fuel handling in the fuel pool may continue. Control Rod Blade removal from the reactor vessel may continue once the shorting links are installed.

Proposed Answer: B Explanation (Optional):

B. Correct Page 192 of 205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet B. Correct. T.S. action 3.1.1 c requires suspension of Core Alterations and other activities that could reduce the Shutdown margin, and insert all insert able control rods.

Returning CRBs to the RPV would be Core Alterations (as defined in TS) TS 3.9.10.2.c both require SDM demonstration or suspend control rod removal T.S. 3.9.2.c allows RPS shorting Link removal in place of SDM demonstration, or suspend all Core Alterations and insert all insert able control rods.

A. C. D. Incorrect- Fuel Pool fuel handling is allowed. Incorrect. Shorting links must be removed. Fuel Pool fuel handling is allowed Incorrect. Shorting links must be removed. Technical Reference(s) HC.OP-IO-ZZ-0009 (Attach if not previously provided) TS 3.9.10.2, 3.1.1 Proposed references to be provided to applicants during examination:

None Learning Objective:

Question Source: IOP009E006 (As available)

Bank # 76882 __. Modified Bank

1. -. (Note changes or attach parent)

New Question History: Last NRC Exam -. Question Cognitive Level:

Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 2. 7 Comments: Page 193 of 205 hr NOTE If adequate Shutdown Margin has been demonstrated IAW TIS 3.1 .I, then removal of the RPS shorting links is NOT required.

5.1.3. WHEN within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the start of CORE ALTERATIONS requiring control rod withdrawal (E.ucept for control rods removed IAW T/S 3.9.10.1 3.9.10.2), THEN PERFORM the following:

[T/S 4.9.2.d] A. VERIFY EITHER Reactor Protection System shorting links are removed OR Adequate Shutdown Margin has been demonstrated IA W T/S 3.1.1. __ B. RECORD the time, and date, the applicable surveillance requirement was satisfied AND W/A" for other ENTER initials on Attachment I. - Hope Creek Page 7 of 40 Rev. 32 3/4.1 REACTIVITY CONTFtOL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN - LIMITING CONDITION FOR OPERATION - 3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than: a. 0.38% delta E:/k with the highest worth rod analytically determinec, or b. 0.28% delta !s/k with the highest worth rod determined by test. APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5. ACTION: With the SHUTDOWN MARGIN less than specified:

a. In OPERATIONAL CONDITION

1. or E, reestablish the required SHUTDOWN MARGIN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> lor be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. b. In OPERATIONAL CONDITION 11 or 4, immediately verify all insertable control rods t c be inserted arid suspend all activit Les that could reduce the SHUTDOWN PARGIN.

In OPERATIONAL CONDITION 4, establish SEC3NDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. cr In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS and other activities that could reduce tke SHUTDOWN MARGIN and insert all insertable control rods wichin 1 hour- Establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. -- SURVEILLANCE REQUIREMEKTS 4.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater than specified at any time during the fue.1 cycle: a. By measurement, prior to or during the first startup after each refueling.

b. By measurement, within 500 MWD/T prior to the core average exposure at which the predicted SHUTDOWN MARGIN, including uncertainties and calculation biases, is equal to the specified limit.

c. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after oetectior of a withdrawn control rod that is immovable, as a result of exlzessive friction or mechanical interference, or is untrippable, except that the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withorawn worth of the immovable or untrippable control rod. HOPE CREEK 3/i 1-1 DEFINITIONS CORE ALTERATION I .7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement), and

b. Control rod movement, provided there are no fuel assemblies in the associated core cell Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

1.8 DELETED

CORE OPERATING LIMITS REPORT 1.9 The CORE OPERATING LIMITS REPORT is the unit-sipecific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Plant operation within these limits is addressed in individual specifications.

CRITICAL POWER RATIO 1.10 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the applicable NRC- approved critical power correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power. DOSE EQUIVALENT 1-1 3 1 1. 1 1 DOSE EQUIVALENT I- 13 1 shall be that concentration ( if I- 13 1, microcuries per gram, which alone would produce the szame thyroid dose as the quantity and isotopic mixture of 1-13 1,1-132,1-133,1-134, and 1-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table I11 of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites." - E-AVERAGE DISINTEGRATION ENERGY 1.15 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. HOPE CREEK 1-2 Amendment No. 163 I REFUELING OPERATIONS MULTIPLE CONTROL ROD REMOVAL LIMITING CONDITION FOR OPERATION 3.9.10.2 Any number of control rods and/cr control rod drive mechanisms may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control roc drive mechanisms are reinstalled and all control rods are inserted in the core. a. The reactor mode switch i.3 0PEE:ABLE and locked in the Shutdown position or in the Refuel posii:i.on per Specification 3.9.1, except that the Refuel position "one-rod-out

interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed, aftei: the fuel assemblies have been removed as specified below. b. The source range monitors SRM ice OPERABLE per Specification 2'.9.2. c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are sat is f ied . d. All other corltrol rods are eitk.er inserted or have the surrounding four fuel assemblies removed from the core cell. e. The four fuel assemblies siirroLnding each control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed frorr the core cell. f. All fuel loading operations shall be suspended.

APPLICABILITY:

OPERATIONAL CONDITION

5. ACTION: With the requirements of the above specif-cation not satisfied, suspend removal of control rods &nd/or cantrol rock drivemechanisms from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.

HOPE CREEK 3/4 9-15 1 I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 WA # G2.2.7 Importance Rating

3.6 Knowledge

of the process for conducting special or infrequent tests.

Proposed Question:

SRO 95 Given the following conditions:

The plant is in Operational Condition

1. It's been determined that work must be performed on the 6A Feedwater Heater level controller.

It has been determined that the work could result in an unplanned load reduction of 40 MWe. This evolution is.

A. a production risk activity and an IPTE brief is required.

B. a production risk activity and a HLNIPA brief is required.

C. NOT a production risk activity but an IPTE brief is still required.

D. NOT a production risk activity but an HLNIPA brief is still required.

Proposed Answer: B Explanation (Optional): IAW WC-AA-104 - Step 2.4. defines production risk activity as

>20 MWe. Then an HLNIPA B. Correct A. C. D. Incorrect - an HLNIPA briefing is required.

IPTE briefings are no longer performed for these evolutions.

Incorrect - this is a production risk activity, an HLNIPA briefing is required.

Incorrect - this is a production risk activity. IPTE briefings are no longer performed for these evolutions. Technical Reference(s)

OP-AA-108-110 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Page 194 of ,205 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Learning Objective:

OPSBRIEFE005 (As available) Question Source:

-. Bank # New X (Note changes or attach parent)

Modified Bank # -. Question History: Question Cognitive Level:

10 CFR Part 55 Content: Last NRC Exam Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 55.43 3 Com men ts : Page 195 of 205

&F- WC-AA-104 Revision 12 Page 1 of 15 REVIEW AND SCREENING FOR PRODUCTION RISK 1. PURPOSE 1.1. This procedure provides the proper steps required to screen, plan and oversee work activities to manage production risk. 1.2. Applicability 1.2.1. This procedure applies to all on-line work activities pre-outage activities, and common unit outage work unless excluded by this procedure.

1.2.2. It is Management's expectation that perscinnel understand their work activities and the effects of those activities on plantkyslem operation.

This is best accomplished through a combined effort of involved parties.

1.2.3. Work activities screened prior to the implementation of this revision, are not required to be revised to the new forms/process.

2 TERMS AND DEFINITIONS

2.1. Critical

Step or Task - Any action that when performed improperly will lead to an unintentional change that adversely impacts plant, systems, or personnel status.

2.2. Power

Block - Includes all structures, systems and components which perform a direct function in the production of, transport of, or storage of heat energy, electrical energy or radioactive wastes. Also included are structures, systems and components that monitor, control, protect or otherwise support the above equipment. Hope Creek - Reactor Building, Turbine Building, Control and DIG areas, Service Water Intake, Circulating Water and Cooling Tower, and Switchyard.

Salem (1&2) - Containment Building, Aux Building, Turbine Building, Control and D/G areas, Fuel Handling Building, Service Water Intake, Circulating Water Intake, and Switchyard.

Salem 3 - all areas.

2.3. Procedure

Risk Matrix - A listing of frequently performed procedures that have been evaluated for Production Risk. Production Risk Activity - Any activity that has the potential to derate the plant. A derate constitutes a > 20 MWE load reduction from the power level(s) committed to the electric operations dispatcher.

2.5. Production

Risk Svstems -A list of systems determined by each site. Work activities on those systems could cause a > 20 MWE load reduction from the power level(s) committed to the electric operations dispatcher, see Attachment 4, Hope Creek Production Risk System Matrix and Attachment 5, Salem Production Risk System Matrix..

OP-AA-I 08-1 IO Revision 0 Page 3 of 10 4.1.3. The individual evaluating an activity ENSURES the SLM concurs with the results of their evaluation.

If an activity is determined to be a special test or evolution, then the SLM shall ASSIGN a Special Test or Evolution Coordinator to perform sections 4.2 and 4.3. If - not, then no further action is required under this procedure. The SLM shall ENSURE proper plant management approval has been received to perform the special test or evolution.

4.1.4. 4.1.5. 4.2. 4.2.1. lmplementinq Special Test Or Evolution Controls If an evolution is determined to be a special test or evolution and no procedure g u idance . The SQR shall DETERMINE the cross discipline review requirements of a special test or evolution procedure for technical adequacy.

1. exists, then PREPARE a special procedure in accordance with approved plant I 4.2.2. I If the special test or evolution has the potential to affect Reactivity Management, then this review shall be performed by Reactor Engineering (0 P-AA-300-1 540). 4.3. Special Test Or Evolution Performance I 4.3.1. CONSIDER Just in Time training to ENSLIRE plant conditions and the activity are understood by those involved.

ENSURE the special test or Evolution is reviewed by senior management prior to the start of the evolution.

The SLM or his designee shall CONDUCT a HLNIPA briefing prior to performing the special test or evolution.

This briefing shall cover all the applicable fopics from HU-AA-1211 and should be attended by all personnel performing the activity.

If the duration of the special test or evolution is greater than one shift, PERFORM the HLNIPA briefing for each shift. The SLM and the Special Test Or Evolution Coordinator shall ENSURE appropriate station management involvement is present prior to and during special test or evolution performance.

Operations Shift Management shall ENSURE required plant conditions are maintained as required for the special test or evolution.

4.3.2. 4.3.3. -7 4.3.4. 0 4.3.5. 4.3.6. 5. DOCUMENTATION - None

, I OP-AA-I 08-1 IO Revision 0 Page 2 of 10 3.3. 3.3.1. Special Test Or Evolution Coordinator Utilizes the guidance provided in Attachment 1 and Attachment 2 to determine if the activity requires special test or evolution c.ontrols. Ensures Senior Line Manager and Shift Management are aware of the status of a special test or evolution.

3.3.2. 3.4. Operations Shift Manaqement 3.4.1. 3.4.2. Approves implementation of the special test or evolution.

Ensures required plant conditions are maintained during performance of the special test or evolution.

I 3.5. Site Reactor Enqineerinq

3.5.1. Reviews

each special test or evolution that involves plant systems potentially affecting reactivity or changes to the core or fuel pool (OP-AA-300-1540).

I 3.6. Traininq 3.6.1. Develops and validates appropriate training material to support the evolution or test as necessary.

Provides challenging, relevant, and effective training necessary to perform the special evolution or test. Reinforces line management expectations and department fundamentals in areas applicable to the training setting.

3.6.2. 3.6.3. 4. MAIN BODY 4.1. Identification of Special Test or Evolutions, 4.1 .I. REVIEW the following types of activities to determine if screening under this procedure is warranted:

I Work activities screened as "high-risk" in the work control process Activities similar in nature to those "example" activities discussed in Attachment

1. -49 (WC-AA- 1 04). I 2. 4.1.2. The individual evaluating an activity to determine if it is a special test or evolution shall USE Attachment 1 and Attachment 2 to determine if special test or evolution controls are required.

I WC-AA-104 Revision 12 Page 3 of 15 4. MAIN BODY Top Level -___~__ \ , Any On-Lin? Activity 1 1 i NO YES --2. , Perform Risk i Screenng ' I using Work Planning Process 1 '1; work Procuction

<,, Rtsk:' 7- .., Designate as Production Risk Screening I I ES-401 Sample Written Examination Form ES-40 1-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 Group # 3 K/A # G2.2.18 Importance Rating

3.9 Knowledge

of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc. Proposed Question: SRO 96 The plant is in Cold Shutdown for a forced outage. You are approving work to be performed during the outage. Which one of the following describes an example of systems/components which would become an "Operation With Potential To Drain The Reactor Vessel" (OPDRV) if NOT isolated by a barrier IAW HC.OM-AP.ZZ-0001 "Shutdown Safety Management Program"?

A. SRVs B. HPCl C. RPV Instrumentation D. Reactor Water Cleanup Proposed Answer: D Explanation (Optional): HC.OM-AP.ZZ-0001 Step 7.2 D. Correct.

A. B. C. Incorrect. This would be considered an Operation with a Potential for Draining the Reactor Cavity as defined in step 7.3 of the procedure Incorrect. This would be considered an Operation with a Potential for Draining the Reactor Cavity as defined in step 7.3 of the procedure Incorrect.

This would be considered an Operation with a Potential for Draining the Reactor Cavity as defined in step 7.3 of the procedure Tech n ica I Reference( s ) H C . OM -AP . ZZ-000 1 (Attach if not previously provided) Page 196 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Proposed references to be provided to applicants during examination:

none Learning Objective: Question Source: (As available)

-. Bank # Modified Bank # New X (Note changes or attach parent)

Question History: Question Cognitive Level:

10 CFR Part 55 Content: Comments:

Last NRC Exam Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 55.43 2 Page 197 of 205 I t Old5ldPFOP VERIFYING REVISION, STATUS AND CHANGES HC.OM-AY.ZL-UUU1 (Q) Involves opening piping with an inner diameter of greater than one (1 ) inch or multiple pipes which, when summed, exceed an inner cross-sectional area equivalent to one (I) inch diameter pipe.

Exceptions to this are equipment that:

a) Is protected by at least one isolation valve (and associated functional instrumentation and logic) capable of automatically closing on a low reactor water level signal; gr required position. Approved devices are closed manual valves, back seated valve, or a deactivated automatic valve secured in the closed position. Any other device must be recommended for approval by SORC; c) Has a MOV capable of being operated from the control room: d) Has a valve capable of being closed locally by a dedicated equipment b) is isolated from the RPV by at least one approved device controlled in the operator or technician, as appropriate?, with no concurrent duties who are maintaining communications with the control room. The following systemskomponents are examples of those, which could become subject to OPORVs if not isolated by a barrier: Reactor Water Cleanup 0 Reactor Recirculation Residual Heat Removal (RH'R) Control Rod Drive (CRD) Reactor Recirculation Sample 2. Involves control rod drive mechanism (CRDM) removal unless: a) Prior to complete removal of the CRDM, communications between the under vessel area and the control room are established and it is verified that the control rod is back seated against the CRDM guide tube, Following removal of the CRDM (if a CRDM is not immediately placed in the guide tube), a blank flange is placed over the removed CRDM guide tube; and b) Any Control Rod Blade Exchanges performed concurrently with CRDM Removal shall be performed in accordance with HCOP-SP.BF-0001 (Q), Control Rod Drive Mechanism/Blade Simuttaneous Removal and shall be considered an Infrequently Performed Activity or Evolution in accordance with HU-AA-I 21 1 Conduct of Infrequently Performed Activities.

Page 10 of 39 Rev. 1 I I 46 PAC- USER ESPONSIBLE FOR VERIFYING RCVISION, STPTIJZ AND CHANGES PRINTED 20081113 *iC.OM-AP.ZZ-0001 (Q) 7.3 7.4 Hope Creek Operations with a Potential for Draininn the Reactor Cavity lOPDRCl- An OPDRC is any operations or maintenance activity that: I. Has the potential to uncover irradiated fuel when the Fuel Pool Reactor Cavity Gates are removed, or Core Alterations are in progress; and involves opening piping with an inner diameter of greater than one (I) inch or multiple pipes which, when summed, exceed an inner cross-sectional area equivalent to one (1 ) inch diameter pipe. Exceptions to this are: Is protected by at least one isolation valve (and associated functional instrumentation and logic) capable of automatically closing on a low reactor water level signal; Is isolated from the RPV cavity by at least one approved device controlled in the required position.

Approved devices are closed manual valves, back seated valve, or a deactivated automatic valve secured in the closed position.

Any other device must be recommended for approval by SORC; Has a motor operated valve capable of being operated from the control room; Has a valve capable of being closed locally by a dedicated equipment operator or technician, as appropriate, with no concurrent duties who are maintaining communications with the control room. The following systems/components are examples of those, which could become subject to OPDRCs if not isolated by a barrier: 0 All systems/components for 0PDRV:s MSlVs SRVs Main Steam Feedwater Core Spray HPCl RClC RPV Instrumentation Fuel Pool Cooling ECCS lniection Source - An RHR or Core Spray subsystem which consists of a minimum of one AVAILABLE pump; and AVAILABLE flow path capable of being lined up from the Torus or the CST to the vessel. Page 11 of39 Rev. 1 I I ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 Group # 4 WA # G2.4.6 I m portance Rating 4.7 -- Knowledge of EOP mitigation strategies. Proposed Question:

SRO 97 Given the following conditions: Reactor Power is at 6% Reactor Pressure is being controlled by SRVs at 950 psig Reactor Water Level is (-1 0) inches, slowly lowering Drywell Temperature is 355"F, and rising Drywell Pressure is 23 psig, and rising Suppression Pool Temperature is 11 5"F, and rising Suppression Pool Level is 85 inches, steady Suppression Chamber Pressure is 21.7 psig, and rising NO operator actions have been taken Which one of the following action(s) is(are) are requared?

A. ONLY initiate Drywell Sprays IAW EOP-102. B. ONLY initiate Drywell Sprays and Suppression Pool Cooling/Sprays IAW EOP-102.

C. Enter EOP-202 and Emergency Depressurize.

D. Place Suppression Pool Cooling/Sprays in service then Emergency Depressurize IAW EOP-202. Proposed Answer: C Explanation (Optional): C.

Correct IAW EOP-102 Step DWT-8, If DW temp cannot be maintained below 340 degrees F., ED is required.

C. Correct. A. B. D. Incorrect. - ED is required Incorrect. - ED is required Incorrect. -All RPV injection must be secured prior to ED (EOP-202, step ED-3) Page 198 of 205 ES-401 Sample Written Ex ami nat ion Form ES-401-5 Question Worksheet Technical Reference(s) EOP-102 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: Question Source:

Question History: Question Cognitive Level:

10 CFR Part 55 Content: E01 02PE007 (As available)

-. Bank # Modified Bank # -. (Note changes or attach parent)

New X -. Last NRC Exam -. Memory or Fundamental Knowledge Comprehension or Analysis X 55.41 55.43 5 Comments: Page 199 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 KIA # G2.4.40 Group # 4 Importance Rating

4.5 Knowledge

of SRO responsibilities in emergency plan implementation Proposed Question:

SRO 98 An event has occurred at the plant. The TSC and EOF are manned and NOT activated. IAW NC.EP-EP.ZZ-0102 "Emergency Coordinator Response", which one of the following describes the individual responsible for escalating an emergency event level from a SAE to a GE? A. The Shift Manager. 6. The Emergency Duty Officer. C. The Emergency Response Manager.

D. The Site Vice President. Proposed Answer:

A Explanation:

The SM is the Emergency Coordinator until the TSC is ACTIVATED.

Until then the SM (as the EC) is responsible for escalating an event A. Correct. 6. C. D. Incorrect. Correct if the TSC was activated and the EOF manned Incorrect.

Correct if the EOF was activated Incorrect. Site VP is not a designated EC Technical Reference(s) NC.EP-EP.ZZ-0102 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: (As available)

Question Source: Bank # Modified Bank

1. -. (Note changes or attach parent)

New X Page 200 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Question Question History: Cognitive Level:

10 CFR Part 55 Content: Comments:

Last NRC Exam -. Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 55.43 1 Page 201 of 205 I NC.EP-EP.ZZ-0102 Q) ATTACHMENT 8 Page 2 of 2 2. Direct RAC to "take turnover" of dose assessment from the SRPT -- 3. Contact the Shift Manager and complete EC turnover.

-- a VERIFY current plant/emergency status -7 a DETERMINE which steps in Section 5, (page 3) of this procedure are completed and which steps will be turned over for completion.

0 ENSURE timing and responsibility for next SSCL and/or NRC data sheet are coordinated - --T With SM concurrence, ASSUME the EC function and declare the TSC activated.

2.0 Uoon Assuming Emergency Coordinator Duties', the ED0 Should: ANNOUNCE to the staff that the TSC is activated, the effective time and that you are the emergency coordinator. ANNOUNCE (over the plant page) "THE TSC IS ACTIVATED AND IS THE EMERGENCY COORDINATOR" NOTIFY ERM of TSC activation and provide a brief status update.

IMPLEMENT Sections 5.2, 5.3, & 5.4 of this procedure.

I Nuclear Common Page 20 of 29 Rev. 14

, I I NC.EP-EP.ZZ-0102 Q) ATTACHMENT 8 Page 1 of 2 ACTIVATION af the TSC lnit ais -- I .O Prior To TSC Activation (i.e., Before Assuming Emergency Coordinator Duties):

0 ESTABLISH and MAINTAIN a chronological log of activity and events. (If needed, obtain TSC Administrative staff member to maintain logbook)

-- 0 If not already done, OBTAIN a briefing on the status of the emergency from the Shift Manager (SM). Refer to Attachment 2, EC Emergency Status Briefing Form for turnover points of discussion.

0 DlRECTlENSURE TSC section leads are making preparations to assume emergency response functions while ensuring adequate staffing:

0 Radiological Assessment Coordinator (RAC) 0 Technical Support Supervisor (TSS) 0 Administrative Support Supervisor 0 EPAorCM-1 0 Security 0 PERFORM initial briefing to the TSC staff on emergency conditions and the following issues: (Full TSC staff briefing should be done after TSC has activated) 0 Planffemergency conditions - Why are we here? 0 Establish TSC Activation target time - Should be within 75 minutes of Alert or higher classification.

0 PREPARE to activate the TSC and ASSUME the duties and responsibilities of the Emergency Coordinator as follows:

Ensure key roles for TSC activation are covered: TSC has either an ED0 or TSS to provide command

& control 0 TSC has at least one qualified communicator 0 TSC can support Radiological and Engineering assessment 0 ACTIVATE the TSC as follows: (Communicator, RP and EC turnovers should be done coincident with each other) 4 1. Direct TSC communicators to "take turnover" from the Control Room Communicators. Communications turnover does not have to be completed prior to TSC activation but should be in progress. (continued on next paqe) Nuclear Common Page 19 of 29 Rev. 14 ES-40 1 Sample Written Eiamination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level Tier # Group # KIA # Importance Rating RO SRO 3 3 G2.3.12 3.7 Knowledge of Radialogical Safely Principles pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc. Proposed Question:

SRO 99 A Planned Special Exposure (PSE) is required today for a containment entry. The individual involved and their supervisor have agreed to the PSE. The individual has never performed a PSE. Their lifetime dose is 2 Rem and has been fully documented. Which one of the following describes the additional approvals required and the maximum dose that may be received IAW RP-AA-203 "Exposure Control and Authorization"?

A. The maximum dose permitted for the PSE is 3 Rem. It requires additional approvals by the RPM and Plant Manager ONLY.

B. The maximum dose permitted for the PSE is 3 Rem. It requires additional approvals by the RPM, Plant Manager and Site VP. C. The maximum dose permitted for the PSE is 25 Rem. It requires additional approvals by the RPM and Plant Manager ONLY. D. The maximum dose permitted for the PSE is 25 Rem.

It requires additional approvals by the RPM, Plant Manager and Site VP. Proposed Answer: D Explanation (0ptional):IAW RP-AA-203, Section 4.3 For a PSE, the max annual NRC limit is treated separately. Therefore 25 rem is permissible. It requires additional approvals by the RPM, Plant Manager and Site VP. D. Correct. A. Incorrect. If subtracting the lifetime dose (2 R) from the annual limit (5R) this would be correct. However the PSE is treated separately with a 25 R limit. Site VP is also required for approval . Incorrect. If subtracting the lifetime dose (2 R) from the annual limit (5R) this would be B. Page 202 of 205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet correct. However the PSE is treated separately with a 25 R limit C. Incorrect. Site VP is also required for approval. Technical Reference(s) RP-AA-203, Section 4.3 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

none Learning Objective: NOH04ADM024C-0 1 (As available) Question Source:

Question History: Question Cognitive Level: 10 CFR Part 55 Content: Bank # Modified Bank # -. (Note changes or attach parent)

X -. New __ Last NRC Exam Memory or Fundamental Knowledge X Comprehension or Analysis 55.41 55.43 4 ~- Comments: Page 203 of 205 RP -AA-2 0 3 Revision 3 Page 4 of 10 4.2. 4.2.1. 4.2.2. 4.2.3. 4.2.4. 4.2.5. 4.2.6. 4.2.7. D Authorization To Raise Administrative Dose Control Levels (ADCLs) USE Attachment 1, Dose Control Level Extension Form, or a computerized equivalent, to authorize exposures for adult individuals in excess of 2000 mrem routine TEDE in a year. A supervisor from the department requesting approval shall complete Section I of Attachment 1 and submit the request to the Radiation Protection Department indicating: - The name, identification number, and signature of the individual for whom a dose extension is being requested.

Whether or not other qualified individuals with lower dose are available to perform the work. A detailed explanation of why the dose extension is necessary.

The requested annual TEDE limit for the individual (expressed in 500 mrem increments, i.e. 2500 mrem, 3000 rnrem, etc.) - - - The Radiation Protection Department shall complete Section II Attachment 1 or com p ut e rized equivalent . Pending investigations or calculations of internal exposure shall be reviewed and evaluated to determine the individuals TEDE. NOTE: An individual shall not be approved to receive greater than 2000 mrem TEDE if that person has any absent/

no record (A) dose equivalent for the year. To raise the ADCL to 3000 mrem TEDE in a calendar year, written approval is required by the Radiation Protection Manager and the work group supervisor.

To raise the ADCL to 4000 mrem TEDE in a calendar year, written approval is required by the Radiation Protection Manger, a work group supervisor, and the Station/Plant Manager. NOTE: An individual being considered for approval greater than 4000 mrem TEDE for the year should not have significant estimated dose equivalent.

To raise the ADCL above 4000 mrem, pJ to exceed 5000 mrem, written approval is required by the Site Vice President.

Planned Special Exposures (PSEs) NOTE: PSEs are not the same as emergency doses.

PSEs only apply to adult workers.

, I R P -AA-2 0 3 Revision 3 Page 5 of 10 4.3.1. 4.3.2. 4.3.3. 4.3.4. PSEs are to be authorized only in exceptional situations when alternatives that might avoid the dose estimated to result from the PSE are poJ available or are deemed impractical.

A manager from the department requesting a PSE shall submit a request to the Radiation Protection Department, indicating: - The name and identification number of each individual for whom a PSE is being requested, and The nature of the task for which a F'SE is being requested, and A detailed explanation of why the PSE is necessary. - - Prior written approval is required before the PSE occurs.

Prior to participating in a PSE, individuals involved shall be: 1. Informed of the purpose of the PSE. 2. Informed of the estimated dose and associated potential risk or conditions involved in performing the PSE.

3. Instructed in dose reduction measures and techniques for the PSE. PSE ap roval is granted when the PSE document is signed (by hand) and dated by: - &e individual(s) for whom a PSE is being requested, and - dhe work group manager or other level of supervisory authority for the individual as chosen by the RPM or designee, and - The RPM or designee, and The Plant Manager, and The Site Vice President.

4.3.6. 4.3.7. NOTE: If there are any periods of exposure during the life of the monitoring individual that have not been determined or documented (i.e., Absent/ No record), then participation in a PSE is not permitted.

All of the individual's previous PSE dose equivalents and previous doses in excess of routine occupational limits must be determined from records for each individual who will participate in the PSE. Doses received in excess of the routine occupational dose limits in effect at the time of exposures during accidents and emergencies must also be determined and subtracted from the limits for PSEs.

DOCUMENT each individual's current year and previous years: - PSE dose equivalents, and - Dose equivalents in excess of the exposure limits in effect at the time of the exposures (rows (1) and (2) of Table 1, "NRC Exposure Limits," and the former 10 CFR 20.101), and RP -AA-20 3 Revision 3 Page 2 of 10 5 rem 5 rem 3.1. Responsibility for approval of exposures in excess of administrative dose control levels resides with the Radiation Protection Manager (RPM), the Station/Plant Manager, and the Site Vice President.

15 rem 50 rem 50 rem 15 rem 50 rem 50 rem 75 rem 250 rem 250 rem 4. MAIN BODY 4.1. Lim itat ions 4.1 .I. Exposures shall pJ exceed the 1 OCFR20 Exposure Limits as described in Table 1, "NRC Exposure Limits." Individual and Limit Type (1) Occupational Worker, Minor, Routine Annual (2) Occupational Worker, Adult, Routine Annual (3) Occupational Worker, Adult, PSE Annual (4) Occupational Worker, Adult, PSE Lifetime (5) Declared Pregnant Woman (6) Member of The Public TABLE 1 - NRC EXPOSURE LIMITS 0.5 rem 1.5 rem 5 rem I I ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference:

Level RO SRO Tier # 3 WA # G2.1.34 Group # 1 Importance Rating

3.5 Knowledge

of primary and secondary plant chemistry limits. Proposed Question:

SRO 100 Given the following conditions:

0 Power is currently 20%. 0 0 0 A reactor shutdown is in progress. Hydrogen Water Chemistry Injection (HWCI) is out of service. Main Steam Line RMS Setpoints are set High.

2 Condensate Demineralizers are in service at 3000 gpm each. Plant chemistry parameters are as follows: 0 Condensate demin influent conductivity is 0.21 uS/cm 0 Condensate demin effluent conductivity is 0.08 uS/cm 0 Reactor Water Cleanup conductivity is 0.07 uS/cm 0 Reactor coolant sample conductivity is 0.07 uS/cm 0 Reactor coolant specific activity is 1 .O XI O'3 uci/gm Dose Equivalent Iodine Based on these conditions, which one of the following would cause these indications and what actions must be taken IAW AB-RPV-0007 "Reactor Coolant Conductivity"?

A. Crud burst due to removing HWCl from service; restore HWCl to service.

B. Main Condenser tube leak; isolate the affected condenser waterbox.

C. Reactor fuel pin cladding leak; continue power reduction at normal rate. D. Condensate Demineralizer channeling; remove one demineralizer from service. Proposed Answer: B Explanation (Optional):

B. Conductivity into the Cond Demins is high. This is a symptom of a Condenser tube Leak.

Required action would be to remove the waterbox IAW AB-RPV-0008.

Correct. Main Condenser tube leak; isolate the affected condenser waterbox. Page 204 of ,205 ES-40 1 Sample Written Examination Form ES-401-5 Question Worksheet A. C. Incorrect.

Crud burst from removing HWCl from service; restore HWCl to service.

RWCU and Reactor coolant conductivity levels are normal. Incorrect.

Reactor fuel pin cladding leak; continue power reduction at normal rate. Power reduction at normal rate not permitted due to hASL RMS setpoints are set high. Indications are not cause for emergency power reduction.

Incorrect. Condensate Demineralizer channeling due to low flow; remove one demineralizer from service. Demineralizer outlet conductivity is normal. Would have low inlet and high outlet conductivity.

D. Technical Reference(s) HC.OP-AB.RPV-0007 (Attach if not previously provided) Proposed references to be provided to applicants during examination:

None Learning Objective:

HWCIOOE006 (As available) 80628 Question Source:

Bank ## -. Modified Bank # (Note changes or attach parent)

-. New Question History: Last NRC Exam 2003 Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 6 Comments: Page 205 of 205 PSEG Internal Use Onlv SUBSEQUENT OPERATOR ACTIONS A. Rising Condensate Conductivity Time: Condensate cond: Hope Creek HC.OP-AB.RPV-OQ07(Q)

REACTOR COOLANT CONDUCTIVITY ACTION 1 111 A.l A.2 n 0 a A.3 111 0 A.4 0 0 0 0 AS MONITOR Reactor Coolant Conductivity.

VERIFY proper operation of the Condensate Demineralizers (CRIDS Page 24). 0 Demin flow 3000-6000 gpm 0 0 - IFCD.LC~@iSICU crsfctzm -- THEN PERFORM the following 0 MONITOR Main Condenser Drip Tray AND hotwell conductivities for indications of a hfain Condenser Tube Leak. (CRIDS Pg. 19) Clutlet Conductivity 5 0.060 pS/cm Ciond. Demineralizer Inlet Temp. 4359. 0 DIRECT Chemistry to implement IF CDI Conductivity is > 0.1 pWcm -- THEN PEFWORM the following:

HC.CH-GP.ZZ-0009.

• • NOTE I** **NOTE 2** a. VERIFY via Main Condenser Drip Tray -- AND Hotwell Conductivities in suspected waterbox.

b. ImAm- AND DRAIN the suspected waterbox per Condition D. - IF Main Condenser in-leakage is not indicated, -- THEN DIFlECT Chemistry to Sample - ANC, Analyze condensate to determine the source of the rising conductivity.

DIRlECT Radwaste to verify normal Glycol Cooler Conductivity.

A.6 Page 7 of I3 Rev. 4