ML22147A186

04 Draft Written Examination

ML22147A186
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/13/2022
From:	Heather Gepford Operations Branch IV
To:	Entergy Operations
References
Download: ML22147A186 (184)

Text

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1316 Source: New Rev: 1 Rev Date: 1/12/2022 Originator: K. Smith TUOI:

A1LP-RO-EOP01 Objective:

6 System Number: 007 System

Title:

Reactor Trip Section: 4.1 Type:

Generic EPEs

==

Description:==

Ability to determine or interpret the following as they apply to a reactor trip: If reactor should have tripped but has not done so, manually trip the reactor and carry out actions in ATWS EOP K/A Number: EA2.04 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

4.6 Tier:

1 Group:

1 RO Select:

No SRO Select: Yes Question:

Given:

- An ATWS has occurred from 100% power

- The CRD Power Supply Breaker Trip PBs on C03 were depressed but were unsuccessful

- The ATC is inserting control rods in MANUAL

- Local conditions are preventing access to the CRD AC Power Supply Breakers

- Letdown is isolated NOW

- Reactor Power is 7%

- Startup Rate is slightly NEGATIVE For the current conditions, the CRS should commence RT-12, Emergency Boration in accordance with_____(1)_____and the correct Emergency Classification is a(n)_____(2)_____.

A. (1) Step 1, Emergency Boration using Boric Acid Pumps (2) NUE B. (1) Step 1, Emergency Boration using Boric Acid Pumps (2) ALERT C. (1) Step 2, Emergency Boration using HPI Pumps (2) NUE D. (1) Step 2, Emergency Boration using HPI Pumps (2) ALERT Answer:

D. (1) Step 2, Emergency Boration using HPI Pumps (2) ALERT Notes:

Answer "D" is Correct. During an ATWS, the CRS will direct contingency actions from step 1 of OP-1202.001 Reactor Trip. The correct method for Emergency Boration would be with step 2 of RT-12 since letdown is isolated. With the reactor failing to trip and the backup pushbuttons unsuccessful in lowering power to <5%,

then an ALERT should be declared based on EAL SA6.1.

Answer "A" is Incorrect. Step 1 of RT-12 would be true if letdown was not isolated. NUE is plausible if the backup pushbuttons were successful in lowering power to less than 5%.

Answer "B" is Incorrect. Step 1 of RT-12 would be true if letdown was not isolated.

Answer "C" is Incorrect. NUE is plausible if the backup pushbuttons were successful in lowering power to less Difficulty:

3 Taxonomy: H RO:

SRO:

76

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 than 5%.

General Discussion:

This question matches the K/A since the applicant must determine the correct procedural actions to carry out following an ATWS.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. This question requires knowledge of diagnostic steps and decision points in the emergency operating procedures that involve transitions to event-specific sub-procedures or emergency contingency procedures.

History:

Rev. 1 - 1/12/2022 KS

- Based on validation comments, changed part (2) from remaining in RT-12 or transitioning to Overcooling to determine the appropriate EAL classification. Deleted items in stem that supported the original question.

Used in 2022 SRO Exam.

References:

OP-1202.001 Reactor Trip OP-1202.012 Repetitive Tasks, RT-12 Emergency Boration

1202.001 REACTOR TRIP CHANGE 041 PAGE 2 of 30 INSTRUCTIONS CONTINGENCY ACTIONS

1.

DEPRESS Reactor Trip PB.

A. ENSURE ALL rods inserted A. PERFORM the following:

AND reactor power dropping.

1) IF reactor fails to trip, THEN:

a) DEPRESS CRD Power Supply Breaker Trip PBs on C03:

A-501 B-631 b) IF EITHER A-501 or B-631 fails to trip, THEN:

i) Manually INSERT rods at C03.

ii) DISPATCH an operator to open BOTH CRD AC Power Supply breakers.

2) IF more than one rod fails to fully insert OR reactor power is NOT dropping, THEN PERFORM Emergency Boration (RT-12).

3) Do NOT continue until the reactor is shutdown.

RT-12 is performed.

1202.012 REPETITIVE TASKS CHANGE 024 PAGE 54 of 122 1202.012 RT-12 Rev 10-07-20 EMERGENCY BORATION PAGE 1 OF 5

1.

IF Boric Acid pump (P39A or P39B) and Batch Controller are available, THEN:

A.

IF ANY of the following conditions exist:

Both OP and STBY HPI Pumps are off Letdown is isolated An unexpected delay occurs in implementation of Step 1 THEN GO TO step 2.

B.

SET Batch Controller for maximum batch size as follows:

1)

DEPRESS lower DISPLAY.

2)

DEPRESS TOTAL.

3)

DEPRESS TOTAL RESET.

4)

DEPRESS BATCH SET.

5)

DEPRESS 9, six times.

6)

DEPRESS ENTER.

7)

DEPRESS lower DISPLAY.

C.

ENSURE Condensate to Batch Controller (CV-1251) closed.

D.

OPEN Batch Controller Outlet (CV-1250).

E.

ENSURE BOTH Makeup Filters in service:

F3A F3B F.

RECORD initial BAAT (T-6) level ____________ in.

G.

START available Boric Acid pump(s) (P39A or P39B or both).

(1. CONTINUED ON NEXT PAGE)

With letdown isolated, step 2 will be performed.

PROC./WORK PLAN NO.

1903.010 PROCEDURE/WORK PLAN TITLE:

EMERGENCY ACTION LEVEL CLASSIFICATION PAGE: 230 of 243 CHANGE:

059 - Emergency Action Level Technical Bases Category:

S - System Malfunction Subcategory:

6 - RPS Failure Initiating Condition:

Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:

SA6.1 Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power > 5%

AND Manual trip actions taken at the reactor control console (C03[2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) are not successful in shutting down the reactor as indicated by reactor power > 5% (Note 8)

Note 8:

A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation Definition(s):

None Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.

A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).

This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at back panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control console.

Correct.

PROC./WORK PLAN NO.

1903.010 PROCEDURE/WORK PLAN TITLE:

EMERGENCY ACTION LEVEL CLASSIFICATION PAGE: 225 of 243 CHANGE:

059 - Emergency Action Level Technical Bases Category:

S - System Malfunction Subcategory:

6 - RPS Failure Initiating Condition:

Automatic or manual trip fails to shut down the reactor EAL:

SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power > 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (C03

[2C03/2C14]) (manual reactor trip pushbuttons or DROPS[DSS]) is successful in shutting down the reactor as indicated by reactor power 5% (Note 8)

Note 8:

A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Mode Applicability:

1 - Power Operation Definition(s):

IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

Basis:

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

In the event that the operator identifies a reactor trip is IMMINENT and initiates a successful manual reactor trip before the automatic trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss.

Plausible if actions taken at C03 were successful in lowering power to <5%.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1331 Source: New Rev: 0 Rev Date: 8/25/21 Originator: K. Smith TUOI:

A1SPG-SRO-EAL Objective:

2 System Number: 009 System

Title:

Emergency Procedures / Plan Section: 2 Type:

Generic Knowledge And Abilities

==

Description:==

Emergency Procedures / Plan Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

K/A Number: 2.4.30 CFR

Reference:

41.10 / 43.5 / 45.11 Point Value: 1 RO Imp:

SRO Imp:

4.1 Tier:

1 Group:

1 RO Select:

No SRO Select: Yes Question:

Given:

- Unit 1 has tripped from 100% power due to a 100 gpm RCS leak into containment

- Letdown is isolated

- Pressurizer level is being maintained with P-36A HPI pump in accordance with RT-2 Initiate HPI

- CETs are stable

- SCM is being controlled by the ATC operator between 40°F - 50°F

- The SM has just declared an emergency classification based on the RCS leak and is beginning the notification process in accordance with EN-EP-603, Emergency Notifications Based on the guidance in OP-1203.039, Excess RCS Leakage, the CRS should direct a plant cooldown in accordance with_____(1)_____and the order of offsite notifications shall be_____(2)_____.

A. (1) OP-1203.040, Forced Flow Cooldown (2) State and Local Agencies then the NRC B. (1) OP-1203.040, Forced Flow Cooldown (2) NRC then the State and Local Agencies C. (1) OP-1203.041, Small Break LOCA Cooldown (2) State and Local Agencies then the NRC D. (1) OP-1203.041, Small Break LOCA Cooldown (2) NRC then the State and Local Agencies Answer:

A. (1) OP-1203.040, Forced Flow Cooldown (2) State and Local Agencies then the NRC Notes:

Answer "A" is Correct. OP-1203.039, Excess RCS Leakage states if any RCP is running then perform 1203.040, Forced Flow Cooldown. EN-EP-603 states to notify the NRC immediately after notifying state and local authorities but no later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the declaration.

Answer "B" is Incorrect. Plausible since both notifications to the NRC and to state and local authorities must be performed, however EN-EP-603 states to notify the NRC after notifying state and local authorities.

Answer "C" is Incorrect. Plausible since the name of the procedure is Small Break LOCA cooldown and a small break LOCA exists, however OP-1203.039 states to perform 1203.041 if the RCS is cooling down due to HPI/break flow. This is incorrect since the stem states that CETs are stable, and SCM is being controlled by the ATC operator.

Difficulty:

3 Taxonomy: H RO:

SRO:

77

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 Answer "D" is Incorrect. Plausible since the name of the procedure is Small Break LOCA cooldown and a small break LOCA exists, however OP-1203.039 states to perform 1203.041 if the RCS is cooling down due to HPI/break flow. This is incorrect since the stem states that CETs are stable, and SCM is being controlled by the ATC operator. Both notifications to the NRC and to state and local authorities must be performed, however EN-EP-603 states to notify the NRC after notifying state and local authorities.

General Discussion:

This question matches the K/A since it requires knowledge of notification requirements to the State and the NRC during a small break LOCA event.

SRO Justification:

10 CFR 55.43(b)(5) Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures. This question requires knowledge of the requirements in emergency planning procedures related to notifications to offsite agencies.

History:

Used in 2022 SRO Exam.

References:

EN-EP-603 Emergency Notifications

1203.039 EXCESS RCS LEAKAGE PAGE:

17 of 22 CHANGE: 021 INSTRUCTIONS CONTINGENCY ACTIONS

14. (Continued)

C. CHECK reactor is critical.

C. PERFORM RCS cooldown by ONE of the following:

1) IF RCS is cooling down due to HPI/break flow, independent of SG cooling, THEN PERFORM 1203.041, Small Break LOCA Cooldown in conjunction with this procedure.

2) IF any RCP is running, THEN PERFORM 1203.040, Forced Flow Cooldown in conjunction with this procedure.

3) IF all RCPs are off, THEN PERFORM 1203.013, Natural Circulation Cooldown in conjunction with this procedure.

4) GO TO step 15.

NOTE Recommended shutdown rates for RCS leaks inside containment with no additional complications are as follows:

Less than 50 gpm -- 0.5 to 5% per minute Greater than or equal to 50 gpm -- 5 to 10% per minute D. COMMENCE plant shutdown per 1203.045, Rapid Plant Shutdown.

15. PROCEED as directed by Senior Manager, Operations.

END Correct. SCM is adequate and nothing in the stem implies RCPs were secured.

Incorrect but plausible since the name of the procedure is Small Break LOCA Cooldown which implies the procedure is used during any small break LOCA event.

EN-EP-603 Rev. 1 Page 15 of 29 Emergency Notifications Page 1 of 2 Initial Notification to State and Local Authorities NOTES The Initial Notification Message Form (INMF) shall be used for initial notifications to the State and local authorities within 15 minutes of the emergency declaration or a change in Protective Action Recommendation (PAR) using the information contained in the Notification Message Form.

The Initial Notification for a General Emergency classification must include PARs.

Release is the release of radioactive material to the environment attributable to the event.

1.0 INSTRUCTIONS 1.1 Initial Notifications NOTE The Initial Notification Message Form shall include the following information:

Class of emergency Wind direction Protective Action Recommendation (PAR)

Whether a release is taking place Time of declaration of emergency Whether the event is a drill or actual event Plant name A hand-written copy of the Initial Notification Message Form should be completed and provided to the Control Room Communicator EXCEPT when completion of this hand-written form would challenge the 15-minute notification requirement.

1.

OBTAIN the hand-written copy of the Initial Notification Message Form.

2.

IF a higher Emergency Classification Level (ECL) is declared before the notification begins for the lesser ECL, THEN PERFORM the following:

a.

IF possible, THEN update the NMF to reflect the higher ECL and complete the notification within 15 minutes of the lesser ECL.

b.

IF it is not possible to update the NMF within 15 minutes of the lesser classification, THEN ADD a comment that explains a change in classification is forthcoming AND continue notification for lesser classification to meet the 15-minute requirement.

1)

COMPLETE an Initial NMF for the higher classification level AND perform the notification within 15 minutes of declaration of the higher classification level.

EN-EP-603 Rev. 1 Page 27 of 29 Emergency Notifications Page 1 of 3 Guidelines for Emergency Notifications to the NRC / NRC Form 361 NOTES Entergy sites notify the NRC Operations Center immediately after notifying the State and local authorities and in all cases within one hour of declaration of a classified emergency.

The Initial Notification Form for the Emergency Classification should be referenced when contacting the NRC.

NRC Form-361 is a two-page form used by the NRC Duty Officer. This form is reproduced on the next two pages as an aid. It may be used for reference prior to and during calls. Use of NRC Form-361 should NOT delay notification to the NRC.

All notifications to the NRC are logged in the Station Log including the name of the NRC Duty Officer receiving the call.

1.0 Emergency Notifications to the NRC 1.1 INSTRUCTIONS

1.

Using the Initial Notification Form for the Emergency Classification, NOTIFY the NRC of the emergency using ENS immediately after notifying the State and local authorities and no later than one hour after the declaration of the emergency.

a.

The NRC Operations Center phone numbers are 1-301-816-5100 (main) and 1-301-951-0550 or 1-301-415-0550 (backups).

2.

COMPLETE the NRC Event Notification Worksheet (NRC Form 361) using the forms located in the Control Room (Pages 2 & 3 of this Attachment).

3.

RECORD Event Notification number (obtained from NRC Operations Center) and person contacted on NRC Form 361.

4.

IF requested, THEN TRANSMIT NRC Form 361 to the NRC Operations Center via fax at *9-1-301-816-5151* or alternate method.

5.

WHEN the event is terminated, THEN SEND copies of Notification Forms, Checklists and other related documentation to Emergency Planning.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1319 Source: Bank Rev: 0 Rev Date: 8/25/21 Originator: K. Smith TUOI:

A1LP-RO-EOP08 Objective:

14 System Number: 055 System

Title:

Emergency Procedures / Plan Section: 2 Type:

Generic Knowledge And Abilities

==

Description:==

Knowledge of the operational implications of EOP warnings, cautions, and notes.

K/A Number: 2.4.20 CFR

Reference:

41.10 / 43.5 Point Value: 1 RO Imp:

SRO Imp:

4.3 Tier:

1 Group:

1 RO Select:

No SRO Select: Yes Question:

The plant was operating at rated power when a loss of offsite power occurred.

The following conditions exist 5 minutes after the loss of offsite power:

- P-7A EFW pump is in service

- Bus A3 has a lockout trip

- #2 EDG was secured due to a large oil leak

- The AAC generator is unavailable

- The ERV lifted and failed to seat creating a LOCA

- ERV isolation (CV-1000) failed to close 30 minutes later the following conditions exist:

- SU1 voltage 15.3 KV

- SU2 voltage 60.1 KV

- RCS pressure is 1400 psig and lowering

- SCM is 25°F

- RVLMS Level 1 and 2 indicate "Dry" Which of the following procedures and/or actions has the highest priority?

A. Initiate HPI Cooling (RT-4).

B. Loss of Subcooling Margin (1202.002).

C. Unit 1 Extended Loss of AC Power (FDS-001).

D. Rapid Cooldown Section in Blackout (1202.008).

Answer:

D. Rapid Cooldown Section in Blackout (1202.008).

Notes:

Answer "D" is Correct. With inadequate subcooling margin and head voids indicated, a rapid plant cooldown is required per the Note in 1202.008 Blackout.

Answer "A" is Incorrect. Plausible since step 9 in RT-4 provides for continuing the RT without HPI pumps but 1202.008 only requires performing RT-4 when P-7A is not available.

Answer "B" is Incorrect. Plausible since a Loss of Subcooling Margin is indicated however with no 4160 vital bus energized, 1202.008, Blackout takes priority.

Answer "C" is Incorrect. Plausible because FDS-001 is entered from 1202.008 Blackout EOP, but only if it can be assumed that power cannot be restored within an hour. The SU1 and SU2 voltages given are too low to be considered normal which makes this choice more plausible, however there is guidance within 1202.008 Difficulty:

3 Taxonomy: H RO:

SRO:

78

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 Blackout to defeat undervoltage relays and restore power in a low voltage condition using Attachment 1.

General Discussion:

This question matches the K/A since the applicant must diagnose the given conditions and apply the note in the Blackout EOP to perform a rapid cooldown.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. This question is SRO Only because it tests the ability to apply a note within the Blackout EOP which will require a transition to the section in the Blackout EOP that will perform a rapid cooldown of the RCS History:

Similar to QID 1026 Reworded QID 1026 to better fit KA and updated revision OP-1202.008 Blackout.

Used in 2022 SRO Exam.

References:

1202.008, Blackout

1202.008 BLACKOUT CHANGE 020 PAGE 30 of 43 INSTRUCTIONS CONTINGENCY ACTIONS NOTE This section is used for rapid RCS cooldown if adequate SCM is lost and HPI flow is less than full flow from one HPI pump and RV Head void is indicated.

Cooldown rate limits do not apply.

During this cooldown, primary to secondary heat transfer will be temporarily lost as the primary level drops from the bottom of the hot leg bend to below the secondary level or below EFW nozzles if EFW flow exists.

RV Head voids can be identified during a Blackout using SPDS ICC2 display.

55. IF SCM is NOT adequate AND RV Head void is indicated, THEN BEGIN rapid cooldown as follows:

A. DISPATCH operator to fully open the following ATM Dump valves (REFER TO 1203.002, Alternate Shutdown, Exhibit A):

CV-2619 CV-2676 CV-2618 CV-2668

1) CHECK a SG feed source other than EFW Pump (P7A) is operating and maintaining proper SG levels.

1) IF EFW Pump P7A is the only available feed source, THEN:

a) MAINTAIN sufficient SG press for P7A operation while P7A is the only source of feed.

b) INITIATE action to restore P7B or another SG feed source to operation prior to depressurizing SGs.

B. PLACE all EFW CNTRL valves in VECTOR OVERRIDE:

SG A SG B CV-2645 CV-2646 CV-2647 CV-2648 (55. CONTINUED ON NEXT PAGE)

1202.008 BLACKOUT CHANGE 020 PAGE 14 of 43 INSTRUCTIONS CONTINGENCY ACTIONS

15. (Continued)

C. START P7B.

C. IF EFW Pump P-7A is NOT available, THEN INITIATE HPI Cooling (RT-4).

1) IF NO HPI pumps are available, THEN PERFORM the following to manually cycle ERV, while continuing with this procedure:

a) IF SCM is NOT adequate, THEN:

(1)

OPEN ERV until RCS press less than or equal to 1650 psig.

(2)

WHEN RCS press less than or equal to 1650 psig, THEN RETURN ERV to AUTO.

(3)

REPEAT this step as necessary to maintain RCS press less than 2400 psig.

b) IF SCM is adequate, THEN:

(1)

OPEN ERV.

(2)

WHEN SCM approaches minimum adequate, THEN CLOSE ERV.

(3)

REPEAT this step as necessary to maintain RCS press less than 2400 psig.

c) IF ERV fails open, THEN CLOSE Electromatic Relief ERV Isolation (CV-1000).

d) GO TO step 18.

(15. CONTINUED ON NEXT PAGE)

1202.012 REPETITIVE TASKS CHANGE 024 PAGE 20 of 122 1202.012 RT-4 Rev 10-07-20 INITIATE HPI COOLING PAGE 4 OF 7

8.

IF OP and STBY HPI pumps are both OFF, THEN PLACE OP or STBY HPI pump in service as follows:

A.

IF HPI Pump (P36B) will be used, THEN ENSURE the following selected to energized bus:

P36B/P64B Bus Select MOD Control P64B Transfer Switch B.

START AUX Lube Oil pump for OP or STBY HPI pump.

C.

WHEN associated BWST T3 Outlet is open, THEN START OP or STBY HPI pump.

D.

STOP AUX Lube Oil pump.

E.

Fully OPEN ALL associated HPI Block valves:

P36A/P36B (RED)

P36C/P36B (GREEN)

CV-1219 CV-1220 CV-1278 CV-1279 CV-1227 CV-1228 CV-1284 CV-1285

9.

IF NO HPI pumps are available, THEN:

A.

NOTIFY CRS that NO HPI pumps are available and contingency actions apply.

B.

GO TO step 13.

1202.012 REPETITIVE TASKS CHANGE 024 PAGE 22 of 122 1202.012 RT-4 Rev 10-07-20 INITIATE HPI COOLING PAGE 6 OF 7

13. PLACE ALL Pressurizer Heaters in OFF:

Pressurizer Heater Bank 1 (HS-1005)

Pressurizer Heater Bank 2 (HS-1004)

Pressurizer Heater Bank 4 (HS-1006)

Pressurizer Heater Bank 3 (HS-1010)

Pressurizer Heater Bank 5 (HS-1007)

Pressurizer Heater Group 5 (RUB-13, 14 & 15) (HS-1007A)

14. For RCPs:

A.

IF RCPs are in service, THEN ENSURE only ONE RCP is operating.

B.

IF RCPs are OFF, THEN MAINTAIN RCPs OFF.

15. MAXIMIZE RB cooling as follows:

A.

ENSURE ALL four RB Cooling Fans running:

VSF1A VSF1C VSF1B VSF1D B.

OPEN RB Cooling Coils Service Water Inlet/Outlet valves:

CV-3812/CV-3814 CV-3813/CV-3815 C.

UNLATCH key-locked Chiller Bypass Dampers:

SV-7410 SV-7412 SV-7411 SV-7413

1202.012 REPETITIVE TASKS CHANGE 024 PAGE 23 of 122 1202.012 RT-4 Rev 10-07-20 INITIATE HPI COOLING PAGE 7 OF 7

16. ISOLATE possible RB leak paths as follows:

A.

IF RB Sump draining is in progress, THEN CLOSE RB Sump to AUX Sump valves:

CV-4400 CV-4446 B.

Securing RB Leak Detector RX-7460 at C25:

1)

At EITHER RDU (RI-7460 or RI-7461) Command screen, SELECT PumpCmd.

2)

SELECT 0 (password:0000/confirm).

C.

On C26, PLACE RB Leak Detector Isolations (HS-7454) in CLOSE BOTH.

END

1202.008 BLACKOUT CHANGE 020 PAGE 3 of 43 INSTRUCTIONS CONTINGENCY ACTIONS

2. (Continued).

E. IF ANY of the following apply:

one hour has elapsed since Blackout occurred it is evident that Blackout will exceed one hour when directed by TSC, EOF, or Shift Manager THEN DECLARE that an Extended Loss of AC Power has occurred AND GO TO FDS-001, Unit 1 Extended Loss of AC Power to address loss of feedwater.

F. WHEN feedwater becomes available, THEN RESTORE feedwater per RT-16.

3.

NOTIFY Unit 2 of need for AAC Gen (2K9)

AND ATTEMPT to restore EDG using 1203.012A, Annunciator K01 Corrective Action and 1104.036, Emergency Diesel Generator Operation, while continuing with this procedure.

A. IF AAC Gen becomes available, THEN ENERGIZE a vital bus using 1107.002, ES Electrical System Operation, Placing Alternate AC Generator on bus A3 (A4) section.

B. IF a vital bus becomes energized by EDG or AAC Gen, THEN GO TO 1202.007, DEGRADED POWER procedure.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1320 Source: New Rev: 0 Rev Date: 8/26/21 Originator: K. Smith TUOI:

A1LP-RO-EOP07 Objective:

14 System Number: 056 System

Title:

Loss of Offsite Power Section: 4.2 Type:

Generic APEs

==

Description:==

Ability to determine and interpret the following as they apply to the Loss of Offsite Power: T-cold and T-hot indicators (wide range)

K/A Number: AA2.19 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

4.2 Tier:

1 Group:

1 RO Select:

No SRO Select: Yes Question:

Unit 1 is currently recovering from a Loss of Offsite Power.

During the event the following was noted:

- P-7A Steam Driven EFW pump could not be started

- #1 EDG output breaker could not be closed After Startup Transformer #1 was energized to 22.3kV the following conditions exist:

- CBOT energized buses H1,H2,A1,A2,A3,A4

- RCS Pressure is 1450 psig

- Pressurizer level is 150 inches

- CETs are 583°F and rising

- T-hot is 580°F and rising

- T-cold is 545°F and stable Which of the following procedures and/or actions has the highest priority?

A. ESAS (1202.010).

B. Overheating (1202.004).

C. HPI Cooldown (1202.011).

D. Inadequate Core Cooling (1202.005).

Answer:

B. Overheating (1202.004).

Notes:

Answer "B" is Correct. Once power is restored to buses in step 75 of Degraded Power, primary to secondary heat transfer is checked in step 81 which looks at T-hot/T-cold T stable or dropping. If it is not stable or dropping then the contingency column is performed which states GO TO 1202.004, Overheating.

Answer "A" is Incorrect. Plausible because RCS pressure is below ESAS actuation setpoint, however ESAS actions are handled using RT-10 during these conditions. There is a floating step in Degraded Power to Go to ESAS when RCS pressure is less than 150 psig.

Answer "C" is Incorrect. Plausible since HPI Cooldown would be performed within the Overheating section of the Degraded Power EOP, but only if EFW restoration is not imminent. In this case, motor driven EFW pump P-7B can be started on A3, powered from offsite, so HPI Cooling would not be necessary.

Answer "D" is Incorrect. Plausible since CET temperatures are rising and a loss of subcooling margin exists, however CET temperatures are not superheated, therefore transition to ICC is not correct.

Difficulty:

3 Taxonomy: H RO:

SRO:

79

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 General Discussion:

This question matches the K/A since the applicant must interpret the T-hot and T-cold indication during a Degraded Power event to determine which procedure transition is required.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. This question is SRO Only because it tests the ability to assess T-hot and T-cold indications once in the Degraded Power EOP and select the correct procedure transition that will mitigate the event.

History:

Used in 2022 SRO Exam

References:

OP-1202.007 Degraded Power

1202.007 DEGRADED POWER CHANGE 017 PAGE 57 of 85 INSTRUCTIONS CONTINGENCY ACTIONS

80. IF HPI Cooling is in service, THEN GO TO step 81.

80. GO TO step 84.

81. Check primary to secondary heat transfer in progress indicated by all of the following:

T-cold tracking associated SG T-sat (Fig. 2)

T-hot tracking CET temps T-hot/T-cold T stable or dropping

81. GO TO 1202.004, "OVERHEATING" procedure.

A. Check adequate SCM.

A. IF overheating caused inadequate SCM, THEN perform the following:

1) Place ERV in AUTO.

2) IF ERV fails open, THEN close Electromatic Relief ERV Isolation (CV-1000)..

3) WHEN SCM is adequate, THEN continue.

B. Control RCS press within limits of Figure 3 (RT-14).

1202.007 DEGRADED POWER CHANGE 017 PAGE 84 of 85 Major Recovery Strategies Ensure proper DG operation and swing component alignment Ensure SGs available as heat sinks Control RCS pressure and inventory (including RCP seals)

Establish conditions for power restoration Address existing heat transfer upsets if present (loss of SCM, overheating, overcooling)

Restore off-site power if available or establish conditions for long-term DG operation Floating Steps RCS Inventory/Press Control RCS press within limits of Figure 3 (RT-14).

IF SCM is less than adequate AND not caused by overheating, THEN GO TO step 24.

IF SCM is less than adequate, AND caused by overheating, THEN GO TO step 55.

IF PZR level drops below 30" OR RCS press drops below 1700 psig, THEN initiate HPI (RT-2).

IF RCS press < 150 psig, THEN GO TO 1202.010, "ESAS" procedure.

IF MU Tank level drops below 18",

THEN close Makeup Tank Outlet (CV-1275).

RCS Temp IF CET temps are > 610 F AND EFW is not available, THEN GO TO step 55.

IF RCS T-cold is < 540 F AND dropping AND SCM is adequate, THEN GO TO step 40.

IF CET temps are superheated AND moving away from the saturation line, THEN GO TO 1202.005, "INADEQUATE CORE COOLING" procedure.

1202.007 DEGRADED POWER CHANGE 017 PAGE 85 of 85 Floating Steps Electrical IF off-site power becomes available AND no abnormal conditions exist, THEN verify steps 2 through 22 have been completed AND GO TO step 74.

IF two DGs are operating AND one DG is lost while performing this procedure, THEN RETURN TO step 2.

IF all DGs are lost, THEN GO TO 1202.008, "BLACKOUT" procedure.

IF A3 and A4 are cross-connected, AND another DG or off-site power becomes available, THEN restore buses to normal using Electrical System Operation (11 ng Paralleled Buses A3 IF A3 and A4 are not cross-connected AND another DG becomes available, THEN RETURN TO step 2.

ESAS IF ESAS actuates, THEN verify proper ESAS actuation (RT-10).

Instrument Air IF Instrument Air to ATM Dump CNTRL valves is lost, THEN perform step 5.B.

Secondary IF EFW is lost AND SCM is adequate, THEN GO TO step 55.

IF SG press 900 psig, except when SG is being allowed to boil dry AND SCM is adequate, THEN GO TO step 40.

IF SG tube leakage is greater than 1 GPM, THEN PERFORM 1202.006, "TUBE RUPTURE" in conjunction with this procedure.

SF Pool Cooling IF Spent Fuel Pool cooling is not in service, THEN perform Unit 1 Spent Fuel Pool Emergencies (1203.050) in conjunction with this procedure.

1202.013 EOP FIGURES CHANGE 005 PAGE 1 of 6 0

500 1000 1500 2000 2500 200 250 300 350 400 450 500 550 600 650 700 RCS Pressure (100 psig Increments)

RCS Temperature (10oF Increments)

FIGURE 1 Saturation and Adequate SCM Saturation Line 50oF Subcooled 70oF Subcooled 30oF Subcooled RCS Pressure Adequate SCM

>1000 psig 30 F 350 to 1000 psig 50 F

<350 psig 70 F

1202.007 DEGRADED POWER CHANGE 017 PAGE 84 of 85 Major Recovery Strategies Ensure proper DG operation and swing component alignment Ensure SGs available as heat sinks Control RCS pressure and inventory (including RCP seals)

Establish conditions for power restoration Address existing heat transfer upsets if present (loss of SCM, overheating, overcooling)

Restore off-site power if available or establish conditions for long-term DG operation Floating Steps RCS Inventory/Press Control RCS press within limits of Figure 3 (RT-14).

IF SCM is less than adequate AND not caused by overheating, THEN GO TO step 24.

IF SCM is less than adequate, AND caused by overheating, THEN GO TO step 55.

IF PZR level drops below 30" OR RCS press drops below 1700 psig, THEN initiate HPI (RT-2).

IF RCS press < 150 psig, THEN GO TO 1202.010, "ESAS" procedure.

IF MU Tank level drops below 18",

THEN close Makeup Tank Outlet (CV-1275).

RCS Temp IF CET temps are > 610 F AND EFW is not available, THEN GO TO step 55.

IF RCS T-cold is < 540 F AND dropping AND SCM is adequate, THEN GO TO step 40.

IF CET temps are superheated AND moving away from the saturation line, THEN GO TO 1202.005, "INADEQUATE CORE COOLING" procedure.

1202.007 DEGRADED POWER CHANGE 017 PAGE 40 of 85 INSTRUCTIONS CONTINGENCY ACTIONS

62. Check CET temps stable or dropping.

62. Perform one of the following:

A. IF HPI flow is < full flow from one HPI

pump, THEN GO TO step 68.

B. Hold at this point until one of the following conditions is met, while attempting to restore buses per step 74:

IF EFW becomes available, THEN GO TO step 65.

IF CET temps begin to drop, THEN GO TO step 63.

IF 120 minutes on HPI cooling elapse AND CET temps are still rising, THEN GO TO step 68.

IF CET temps are superheated AND moving away from the saturation line, THEN GO TO 1202.005, "INADEQUATE CORE COOLING" procedure while attempting to restore buses per step 74 of this procedure.

63. WHEN CET temps begin to drop, THEN perform the following:

A. IF SCM is adequate, THEN maintain RCS cooldown rate 100F/hr by throttling HPI.

64. IF EFW restoration is imminent, THEN continue with this procedure.

64. GO TO 1202.011, "HPI COOLDOWN" while attempting to restore buses per step 74 of this procedure.

65. WHEN EFW becomes available, THEN refill available SG(s) using RT-16.

65. RETURN TO step 64.

1202.011 would be entered PRIOR to buses being energized from offsite power, within the Overheating section of degraded power if EFW cannot be restored.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1386 Source: New Rev: 0 Rev Date: 12/7/21 Originator: K. Smith TUOI:

Objective:

System Number: 057 System

Title:

Loss of Vital AC Electrical Instrument Bus Section: 4.2 Type:

Generic Abnormal Plant Evolutions

==

Description:==

Knowledge of EOP mitigation strategies.

K/A Number: 2.4.6 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

4.7 Tier:

1 Group:

1 RO Select:

No SRO Select: Yes Question:

Given:

U1 is at 75% power when the ATC notes the following indications for Instrument Power Supply Status lights on C13:

- NNI X AC lights are OFF

- NNI X DC lights are OFF

- NNI Y AC lights are OFF

- NNI Y DC lights are ON In accordance with OP-1203.047, Loss of NNI Power, the CRS should direct the board operator to perform which of the following procedures to ensure the secondary heat removal safety function is maintained?

A. RT-5 Proper EFW Actuation and control verification.

B. RT-6 Proper MSLI and EFW Actuation and control verification.

C. RT-16A feeding intact SG with EFW nozzles.

D. RT-16B feeding intact SG with MFW nozzles.

Answer:

A. RT-5 Proper EFW Actuation and control verification.

Notes:

Answer "A" is Correct. The given indications show that both NNI X AC/DC and NNI Y AC power was lost.

1203.047 states for a loss of both NNI X AC/DC and NNI Y power unavailable then to trip the reactor, actuate EFW and perform RT-5 EFW actuation and control. This is because control of main feedwater is lost, so MFW is secured and EFW actuated.

Answer "B" is Incorrect. Plausible since RT-6 will also align EFW to inject in automatic like RT-5, however RT-6 also actuates MSLI which is not required to be performed in accordance with OP-1203.047.

Answer "C" is Incorrect. Plausible since RT-16A will align EFW to inject, however this is not directed from OP-1203.047. RT-16A would be used in a situation where feedwater was lost to a SG for some time. The steps in this RT slowly restore EFW in manual while maintaining SG tube/shell delta T limits until the SG level is in its associated level band, then placed in automatic.

Answer "D" is Incorrect. Plausible since RT-16B is an alternate method to inject feedwater to a SG via Main Feedwater Pumps or Aux Feed Pump P-75. This RT is not directed out of OP-1203.047.

General Discussion:

This question matches the K/A since it requires knowledge of the strategy for controlling the secondary heat Difficulty:

3 Taxonomy: H RO:

SRO:

80

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 removal safety function during a loss of vital AC power situation.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Repetitive Tasks do not contain stand alone entry conditions and must be directed as a transition within an existing AOP or EOP.

History:

Used in 2022 SRO Exam.

References:

OP-1203.047 RT-5

1203.047 LOSS OF NNI POWER PAGE:

2 of 16 CHANGE: 008 INSTRUCTIONS CONTINGENCY ACTIONS

1.

CHECK ICS and NNI Instrument Power Supply Status lights on C13.

2.

CHECK ICS and NNI indications are normal other than C13 ICS and NNI Instrument Power Supply Status lights out.

2.

PERFORM the following:

A. CHECK ALL NNI X power available.

1) IF ALL NNI X power is unavailable AND ANY NNI Y power is also unavailable, THEN GO TO step 4.

2) IF ONLY NNI X power is unavailable, THEN GO TO step 23.

B. IF ANY NNI Y power is unavailable, THEN GO TO step 47.

NOTE D11 breaker 25 supplies power to ICS and NNI Instrument Power Supply Status lights on C13.

3.

RESET D11 breaker 25 using 1107.001, Electrical System Operations, Reclosing Tripped Individual Load Supply Breakers section.

END

1203.047 LOSS OF NNI POWER PAGE:

3 of 16 CHANGE: 008 INSTRUCTIONS CONTINGENCY ACTIONS NOTE Makeup Tank Level (LR-1248) is unavailable when NNI X AC is de-energized.

Pressurizer Level Control valve (CV-1235) and RC Pump Seals Total INJ Flow valve (CV-1207) fail as follows:

both fail to 50% when NNI X AC and NNI X DC are de-energized both fail to 50% when NNI X DC only is de-energized CV-1207 fails closed when NNI X AC only is de-energized CV-1235 failure position is indeterminate when NNI X AC only is de-energized Pressurizer Heater, Spray, and ERV controls respond as follows for loss of NNI X power:

automatic controls fail when NNI X AC and NNI X DC are de-energized automatic controls fail when NNI X DC only is de-energized Proportional Pressurizer Heaters fail off when NNI X AC only is de-energized manual controls remain available Letdown Flow indication is unavailable when any combination of BOTH NNI X and NNI Y power is lost.

If NNI Y AC power is de-energized, the following occurs:

Letdown Orifice Bypass (CV-1223) fails to 50%

Letdown Pressure indication is unavailable

4.

CHECK ANY combination of BOTH NNI X and NNI Y power unavailable.

4.

RETURN TO step 1.

5.

PERFORM 1202.001, Reactor Trip, in conjunction with this procedure.

6.

ACTUATE EFW and PERFORM RT-5, EFW Actuation and Control.

7.

TRIP BOTH Main Feed Pumps:

A Main Feed Pump B Main Feed Pump

8.

OPEN BWST T3 Outlet (CV-1407 or CV-1408) to OP HPI pump.

Correct.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1114 Source: Modified Rev: 4 Rev Date: 8/26/21 Originator: Cork TUOI:

A1LP-RO-ALOIA Objective:

2 System Number: 065 System

Title:

Loss of Instrument Air Section: 4.2 Type:

Generic APEs

==

Description:==

Ability to determine and interpret the following as they apply to the Loss of Instrument Air: When to commence plant shutdown if instrument air pressure is decreasing.

K/A Number: AA2.05 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

4.1 Tier:

1 Group:

1 RO Select:

No SRO Select: Yes Question:

Given:

- Unit 1 100% power

- INST AIR HEADER PRESS LO (K12-B3) alarms

- Loud hissing sounds on mezzanine in Turbine Building

- Loss of Instrument Air (1203.024) is entered

- Pressurizer level is 175" and lowering slowly

- Instrument Air pressure is 55 psig and lowering slowly

- M-1/F-8 P has risen to 3.6 psid and is stable Which of the following is the correct mitigating action and procedure?

A. Trip the reactor and perform Reactor Trip (1202.001).

B. Commence plant shutdown using Rapid Plant Shutdown (1203.045).

C. Bypass Instrument Air Dryers using Loss of Instrument Air (1203.024).

D. Take manual control of Pressurizer level using Makeup and Purification System Operation (1104.002)

Answer:

B. Commence plant shutdown using Rapid Plant Shutdown (1203.045).

Notes:

Answer "B" is Correct, step 16 of 1203.024 contingency action will direct CRS to step 20 which directs a plant shutdown per Rapid Plant Shutdown AOP when IA pressure lowers to < 60 psig.

Answer "A" is Incorrect. Plausible since step 27 of 1203.024 directs a Reactor Trip when Instrument Air pressure drops to 35 psig or if system degradation warrants a Reactor Trip. The given conditions indicate the overall plant impact to the loss of instrument air isnt severe enough to warrant a Reactor Trip. For example, there is no RPS trip or manual trip setpoint that is being approached, and no system is indicated as being impacted enough to require tripping the reactor.

Answer "C" is Incorrect. Its plausible to bypass the IA Dryer if the loss of instrument air was due to the IA dryer not operating properly. DP is elevated at 3.6 psid but is expected to be with the higher system demand due to the leak indicated by stating the loud hissing sounds on the turbine mezzanine. With DP at less than 5psid, it can be assumed the dryer is operating normally.

Answer "D" is Incorrect. Plausible because normal pressurizer level is 220 inches, so pressurizer level at 175 inches indicates the system is not maintaining normal level control in automatic and may require some manual intervention. 1203.024 refers to pressurizer level and has multiple contingency actions for controlling pressurizer level in manual in accordance with 1104.002; the most restrictive being 100 inches, less than 175 Difficulty:

3 Taxonomy: H RO:

SRO:

81

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 inches.

General Discussion:

This question matches the K/A since it requires assessing the given conditions during a loss of instrument air to determine a plant shutdown is required due to decreasing instrument air pressure.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. This question is SRO Only because it tests the ability to assess instrument air indications once in the loss of instrument air AOP and select the correct procedure transition that will mitigate the event.

History:

New question for 2017 SRO Re-exam Rev. 1

1. Removed "all previous action completed" statement

2. Changed distractor" so x-connect should not be an issue 3 4 & 5. Provided steps as requested

6. 1102.016 is provided Rev. 2 5/4/17

1. Made grammatical changes based on NRC comment Rev. 2 5/4/17

1. Made grammatical changes based on NRC comment

7. Changed distractor "D" Rev. 3, 5/16/17 Editorial changes Rev. 4 8/26/21 - KS Changed correct answer to remove 10%/min since new rev. of LOIA AOP doesn't specify a downpower rate anymore.

Changed distractor from RPSD at 5%/min to bypass the IA Dryer.

Added a bullet that M-1/F-8 DP has risen to 3.6 psid and stable to make new distractor plausible.

Re-worded justification.

Used in 2022 SRO Exam.

References:

1203.024, Loss of Instrument Air

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1114 Source: Repeat Rev: 3 Rev Date: 5/16/17 Originator: Cork TUOI:

A1LP-RO-ALOIA Objective:

2 System Number: 065 System

Title:

Loss of Instrument Air Section: 4.2 Type:

Generic APEs

==

Description:==

Ability to determine and interpret the following as they apply to the Loss of Instrument Air: When to commence plant shutdown if instrument air pressure is decreasing.

K/A Number: AA2.05 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

4.1 Tier:

1 Group:

1 RO Select:

No SRO Select: No Question:

Given:

• Unit 1 100% power

• INST AIR HEADER PRESS LO (K12-B3) alarms

• Loud hissing sounds on mezzanine in Turbine Building

• Loss of Instrument Air (1203.024) is entered

• Pressurizer level 175" and lowering slowly

• Instrument Air pressure 55 psig and lowering slowly Which of the following is the correct mitigating action and procedure?

A. Trip the reactor and perform Reactor Trip (1202.001).

B. Commence plant shutdown at 10% per minute using Rapid Plant Shutdown (1203.045).

C. Commence plant shutdown at 5% per minute using Power Reduction and Plant Shutdown (1102.016).

D, Take manual control of Pressurizer level using Makeup and Purification System Operation (1104.002)

Answer:

B. Commence plant shutdown at 10% per minute using Rapid Plant Shutdown (1203.045).

Notes:

"B" is correct, step 14 of 1203.024 contingency action will direct CRS to step 18 which directs a plant shutdown per Rapid Plant Shutdown AOP at 10% per minute. FYI, low instrument air pressure alarm comes in at 75 psig.

"A" is wrong but plausible step 25 directs a Reactor Trip when Instrument Air pressure drops to 35 psig as checked in step 19 with a contingency to transition to step 25.

"C" is wrong but plausible, for most plant shutdowns the normal shutdown procedure is used but if Instrument Air pressure drops to less than 60 psig, then Rapid Plant Shutdown is used.

"D" is wrong but plausible because this procedure refers to pressurizer level and has multiple contingency actions in regards to level, this requires the applicant to consider this condition and if an action should be taken.

This question matches the K/A since conditions are given for a loss of instrument air which requires commencement of a plant shutdown.

Difficulty:

3 Taxonomy: H

References:

RO:

SRO:

PARENT to QID 1114

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 History:

New question for 2017 SRO Re-exam Rev. 1

1. Removed "all previous action completed" statement

2. Changed distractor" so x-connect should not be an issue 3 4 & 5. Provided steps as requested

6. 1102.016 is provided Rev. 2 5/4/17

1. Made grammatical changes based on NRC comment Rev. 2 5/4/17

1. Made grammatical changes based on NRC comment

7. Changed distractor "D" Rev. 3, 5/16/17 Editorial changes

References:

1203.024, Loss of Instrument Air 1102.016, Power Reduction and Plant Shutdown

1203.024 LOSS OF INSTRUMENT AIR PAGE:

9 of 38 CHANGE: 018 INSTRUCTIONS CONTINGENCY ACTIONS

15. CHECK Decay Heat Removal system is idle.

CAUTION Throttling Service Water flow less than 1600 gpm through DH coolers when RCS temperatures are greater than 200°F can result in unacceptably high Service Water piping temperatures.

NOTE Decay Heat Cooler Bypass valves (CV-1432/CV-1433) fail CLOSED on loss of air.

15. IF necessary for RCS temperature control, THEN:

A. UNLOCK and THROTTLE applicable DH Cooler SW Outlet valve as required to control temperature:

SW-22A SW-22B B. LOCK throttled DH Cooler SW Outlet valve (SW-22A/B).

C. IF above Mode 5, THEN:

ENSURE condition report written to evaluate SW system operability.

REFER TO Technical Specification 3.7.7 for an inoperable service water loop.

16. CHECK Instrument Air Header PRESS remains greater than 60 psig.

16. PERFORM the following:

A. IF Instrument Air Header PRESS recovers (greater than 60 psig),

THEN:

1) TERMINATE power reduction

2) GO TO step 17 B. GO TO step 20.

1203.024 LOSS OF INSTRUMENT AIR PAGE:

11 of 38 CHANGE: 018 INSTRUCTIONS CONTINGENCY ACTIONS

20. COMMENCE plant shutdown at CRS-directed rate using 1203.045, Rapid Plant Shutdown and CONTINUE with this procedure.

21. CHECK BOTH of the following:

Instrument Air Header PRESS remains greater than 35 psig.

NO system degradation occurs that would require a reactor trip.

21. GO TO step 27.

22. PLACE RCP Seal INJ Block (CV-1206) in OVRD.

NOTE Using HPI Block valve CV-1220 or CV-1285 minimizes nozzle stress cycles since these valves share the normal makeup path to the RCS.

Pressurizer Level Control valve (CV-1235) fails AS IS when Instrument Air pressure drops to approximately 45 psig 23.

CHECK PZR level remains greater than 100.

23.

Using ONE of the following, MAINTAIN PZR level greater than 100:

OPEN HPI Block valve associated with OP HPI pump (CV-1220 or CV-1285).

USE 1104.002, Makeup & Purification System Operation, Exhibit B, Manual (Handwheel) Control of Pressurizer Level Control Valve CV-1235 to establish manual control and THROTTLE CV-1235.

Correct.

Reactor trip would be required at 35 psig or if any system degradation requires a trip plausible if pressurizer level lowered to 100"

1203.024 LOSS OF INSTRUMENT AIR PAGE:

14 of 38 CHANGE: 018 INSTRUCTIONS CONTINGENCY ACTIONS

27. Tripping the reactor:

A. TRIP the reactor.

B. PERFORM 1202.001, Reactor Trip in conjunction with this procedure.

28. Isolating SGs:

A. ACTUATE MSLI for BOTH SGs.

B. PERFORM MSLI and EFW actuation and control verification (RT-6).

29. CLOSE EITHER of the following to isolate Letdown:

Letdown Coolers Outlet (RCS) (CV-1221)

BOTH of the following Letdown Coolers Outlets (RCS):

CV-1214 CV-1216

30. ENSURE RCP Seal INJ Block (CV-1206) in OVRD.

NOTE Using HPI Block valve CV-1220 or CV-1285 minimizes nozzle stress cycles since these valves share the normal makeup path to the RCS Pressurizer Level Control valve (CV-1235) fails AS IS when Instrument Air pressure drops to approximately 45 psig 31.

CHECK PZR level remains greater than 55.

31.

Using ONE of the following, MAINTAIN PZR level greater than 55:

OPEN HPI Block valve associated with OP HPI pump (CV-1220 or CV-1285).

USE 1104.002, Makeup & Purification System Operation, Exhibit B, Manual (Handwheel) Control of Pressurizer Level Control Valve CV-1235 to establish manual control and THROTTLE CV-1235.

1203.024 LOSS OF INSTRUMENT AIR PAGE:

3 of 38 CHANGE: 018 INSTRUCTIONS CONTINGENCY ACTIONS NOTE Some control board indicators are air-operated. These indicators can be erratic or inaccurate due to low IA pressure. SPDS diagnostic displays can provide system status and parameter indication.

IA pressure is lowest near a leak causing air-operated components to exhibit abnormal indication or response to control signals.

7.

PERFORM the following:

REFER TO Attachment A for critical air-operated components and component failure modes and CONTINUE with this procedure.

MONITOR control panels for abnormal indications indicating leak location.

(INPO IER 15-16 REC 1)

8.

CHECK NO IA compressor is available or BOTH C-28A/B are already in service.

8.

IF an IA compressor is available and is NOT running THEN ATTEMPT to start available C-28.

NOTE If the in-service IA Dryer (M-14/M-15) or C-28A/B IA Disch Filter (F-44A/B) becomes clogged or isolated, IA Dryer (M-1) is expected to provide dry air for approximately 55 minutes.

9.

CHECK in-service IA Dryer (M-14/M-15) operating properly.

9.

PERFORM the following:

A. Slowly OPEN associated Inlet and Outlet ISOL valves to place standby IA Dryer into service:

M-14 M-15 IA-799 Inlet ISOL IA-801 IA-798 Outlet ISOL IA-800 B. As time permits, PERFORM 1104.024, Instrument Air System, Placing IA Dryer M-14 (M-15) into Service and Removing M-15 (M-14) From Service section.

C. IF standby IA Dryer is NOT available, THEN OPEN IA Dryer M-14/15 Bypass (IA-797).

1203.012K ANNUNCIATOR K12 CORRECTIVE ACTION PAGE:

61 of 97 CHANGE:

054 ATTACHMENT A PAGE 1 OF 2 M-1/F-8 P ALARM ACTIONS Location: C131 Device and Setpoint: (EITHER of the following):

M-1/F-8A Diff Press (PDIS-5400): 5 psid M-1/F-8B Diff Press (PDIS-5401): 5 psid M-1/F-8 P

Alarm: K21-5 1.0 OPERATOR ACTIONS NOTE Normally, C-2A/B IA FLTR F-8A/B Diff Press (PDIS-5400/1) are aligned to indicate if system conditions have forced M-1/F-8 train into service. When M-1 is in-service, the PDIS can be aligned to check the associated F-8 P.

This alarm has a four second time delay.

This alarm inputs to INST AIR COMPRESSOR TROUBLE (K12-C3).

CAUTION Air Dryer M-1 can saturate and pass moisture after ~55 minutes of service.

1.

ENSURE an IA Dryer (M-14 or M-15) is in-service per 1104.024, Instrument Air System, System Startup section.

2.

REFER TO 1203.024, Loss of Instrument Air.

3.

IF directed by CRS/SM to check F-8 P, THEN:

A.

CLOSE M-1 P Upstrm Isols:

IA-348A IA-348B B.

OPEN C2-A/B F-8 P PDIS-5400/01 Upstream Root Isols:

IA-5400A IA-5401A C.

CHECK F-8 P.

Step 3. continued on next page

1203.024 LOSS OF INSTRUMENT AIR PAGE:

11 of 38 CHANGE: 017 INSTRUCTIONS CONTINGENCY ACTIONS

20. COMMENCE plant shutdown at CRS-directed rate using 1203.045, Rapid Plant Shutdown and CONTINUE with this procedure.

21. CHECK BOTH of the following:

Instrument Air Header PRESS remains greater than 35 psig.

NO system degradation occurs that would require a reactor trip.

21. GO TO step 27.

22. PLACE RCP Seal INJ Block (CV-1206) in OVRD.

NOTE Using HPI Block valve CV-1220 or CV-1285 minimizes nozzle stress cycles since these valves share the normal makeup path to the RCS.

Pressurizer Level Control valve (CV-1235) fails AS IS when Instrument Air pressure drops to approximately 45 psig 23.

CHECK PZR level remains greater than 100.

23.

Using ONE of the following, MAINTAIN PZR level greater than 100:

OPEN HPI Block valve associated with OP HPI pump (CV-1220 or CV-1285).

USE 1104.002, Makeup & Purification System Operation, Exhibit B, Manual (Handwheel) Control of Pressurizer Level Control Valve CV-1235 to establish manual control and THROTTLE CV-1235.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 590 Source: Bank Rev: 1 Rev Date: 12/23/21 Originator: S.Pullin TUOI:

A1LP-RO-NI Objective:

2 System Number: 033 System

Title:

Loss of Intermediate Range Nuclear Instrumentation Section: 4.2 Type:

Generic Abnormal Plant Evolutions

==

Description:==

Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Satisfactory overlap between source-range, intermediate-range and power-range instrumentation K/A Number: AA2.04 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

3.6 Tier:

1 Group:

2 RO Select:

No SRO Select: Yes Question:

Given:

- Plant startup in progress

- Channel 1 Source Range, NI-501 at 9 E4 cps

- Channel 2 Source Range, NI-502 at 1 E5 cps

- Reactor power Wide Range recorder, NR-502, is operable and at 5 E -2% power

- Intermediate Range Channel NI-3 at 2 E -11 amps

- Intermediate Range Channel NI-4 at 3 E -11 amps

- Power Range Channels NI-5 thru 8 at 0%

Which of the following actions are REQUIRED in accordance with OP-1203.021 Loss of Neutron Flux?

A. Trip the reactor immediately and refer to 1202.001, Reactor Trip.

B. Suspend positive reactivity additions and open all CRD breakers.

C. Verify SDM to be within the limit provided in the COLR.

D. Since NR-502 is operable, continue with plant operations until Power Range channels come on-scale.

Answer:

B. Suspend positive reactivity additions and open all CRD breakers.

Notes:

Answer "B" is Correct. It would be expected that with the current Source Range Monitor counts, that Intermediate Range Counts would be approximately 5 E-9 amps. With the current IRM reading approximately 2 decades lower, it can be concluded that SRM to IRM overlap has failed. TS 3.3.10 bases states that during a failure of overlap between SRMs and IRMs, the IRMs are assumed to be inoperable. OP-1203.021 Loss of Neutron Flux indication states for two inoperable IRM channels, to refer to TS 3.3.10 for required actions.

Answer "A" is Incorrect. Plausible since this would be required if no on-scale neutron flux existed.

Answer "C" is Incorrect. Plausible since this action is performed when Source Range (not intermediate range) instruments are inoperable.

Answer "D" is Incorrect. Plausible this would be allowed if only one IRM channel was inoperable.

General Discussion:

This question matches the K/A since it requires the ability to interpret Nuclear Instrumentation indications and determine that failed overlap exists between SRMs and IRMs.

Difficulty:

3 Taxonomy: H RO:

SRO:

82

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. Specific TS bases knowledge is required to determine what Nuclear Instruments are inoperable after failing overlap.

History:

New for 2005 SRO exam.

Selected for 2011 SRO Exam Rev. 1 - 12/23/21 - KS

- Changed KA from 033 AA2.10 to 033 AA2.04 to fit 2022 Exam Outline.

- re-worded correct answer to be consistent with procedure and TS.

- changed distractor C since no longer performed in the procedure to an action if Source Range instruments were declared inoperable.

- Removed TS 3.3.10 from stem to avoid asking directly what actions are required in TS since it would make it RO only knowledge.

- Updated Answer Justification Used in 2022 SRO Exam

References:

T.S. 3.3.10 1203.021 STM 1-67

Nuclear Instrumentation STM 1-67 Rev 16 4

STM 1-67 Nuclear Instrumentation 1.0 Introduction This STM contains information on the Excore (Out of Core)

Nuclear Instrument System (NIs) for ANO Unit 1. It includes operational theory of detectors, component locations in the plant and normal and abnormal operations and equipment conditions. The effect nuclear instruments have on plant operation, and the effect plant operations have on the Nuclear Instruments is discussed.

Additional information on theory of detector operation is found in STM 1-62, Radiation Monitoring.

The Nuclear Instrumentation (NI) System is designed to measure over twelve decades of neutron flux using ten channels of out of core neutron detectors and instrumentation. (Refer to Figure 67.01) The full range of indications are displayed to the Reactor Operator and are supplied to the Reactor Protection and Integrated Control systems.

Measurement ranges are designed to overlap to provide complete and continuous information of the full operating range of the reactor.

10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 1

1

-2

-2

-1

-1

-3

-4

-5

-6

-7

-8

-9 2

2 3

3 4

5 6

7 8

9 10 DETECTOR NEUTRON FLUX, nV REACTOR POWER, %

10 10 10 1

2 3

10 10 10 10 10 10 10 10 10 10 1

-2

-1

-3

-4

-5

-6

-7

-8 2

10 1

10 10 10 10 10 10 10-3

-4

-5

-6

-7

-8

-9 COUNTS PER SECOND LOG POWER (%)

LINEAR POWER, %

RATED POWER, %

POWER RANGE 0.1 10 10 4

5 2x10 2 100 125 10 10

-10

-11 INTER-MEDIATE RANGE 10 1

100 125 GAMMA METRICS GAMMA METRICS GAMMA METRICS LOG ION CURRENT AMPERES FIGURE 67.01: NUCLEAR INSTRUMENTATION FLUX RANGES NR-502 WIDE RANGE SPDS NI-3 & 4 NI-5, 6, 7 & 8 NI-1

& 2 SOURCE RANGE 1.1 System Function Red line shows correlation between multiple NIs. When SRMs read ~1E5 cps then IRMs should read ~5E-9amps. With IRMs reading 2-5E-11 amps, it can be concluded that IRMs have failed overlap.

Intermediate Range Neutron Flux B 3.3.10 ANO-1 B 3.3.10-2 Amendment No. 215 Rev. 67 APPLICABILITY (continued)

The intermediate range instrumentation is designed to detect power changes when the power range and source range instrumentation cannot provide reliable indications, e.g., during initial criticality and power escalation. Since those conditions can exist in, or propagate from, all of these MODES, the intermediate range instrumentation must be OPERABLE.

ACTIONS A.1 and A.2 With the required intermediate range neutron flux channel inoperable when THERMAL POWER is 5% RTP, the operators must place the reactor in the next lowest condition for which the intermediate range instrumentation is not required. This involves providing power level indication on the source range instrumentation by immediately suspending operations involving positive reactivity changes and, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, placing the reactor in the tripped condition with the CRD trip breakers open. RCS temperature changes are permitted provided the effects of such changes are accounted for in the SDM calculations. The Completion Times are based on unit operating experience and allow the operators sufficient time to manually insert the CONTROL RODS prior to opening the CRD breakers.

SURVEILLANCE REQUIREMENTS SR 3.3.10.1 Performance of the CHANNEL CHECK provides reasonable assurance of prompt identification that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to the same parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff based on a combination of factors including channel instrument uncertainties. If a channel is outside the criteria, it may be an indication that the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. Off scale low current loop channels are verified, where practical to be reading at the bottom of the range and not failed low.

The agreement criteria include an expectation of one decade of overlap when transitioning between neutron flux instrumentation. For example, during a power increase near the top of the scale for the source range monitors, an intermediate range monitor reading is expected with at least one decade overlap. Without such an overlap, the intermediate range monitors are considered inoperable unless it is clear that a source range monitor inoperability is responsible for the lack of the expected overlap.

1203.021 LOSS OF NEUTRON FLUX INDICATION PAGE:

8 of 12 CHANGE: 015 SECTION 2 - LOSS OF ONE OR MORE INTERMEDIATE RANGE NI CHANNELS IN MODE 2 INSTRUCTIONS

1.

IF ALL of the following apply (no on-scale indication of neutron flux is available):

Three of four power range instruments are less than or equal to 5% power BOTH intermediate range instruments are less than or equal to 10-10 amps BOTH source range instruments are greater than 105 cps Reactor Power Wide Range Recorder (NR-502) is inoperable THEN:

A.

TRIP the Reactor.

B.

PERFORM 1202.001, Reactor Trip, in conjunction with this procedure.

2.

CHECK for normal voltage on failed detector as follows:

In associated RPS cabinet, CHECK Power Supply module reads approximately 600V (590 V to 610 V).

CHECK PDS/PMS Computer Point at 590 V to 620 V:

RPS-C NI-3 HV POWER SUPPLY (E0547)

RPS-D NI-4 HV POWER SUPPLY (E0548)

NOTE Polarity of the IR Compensating Voltage is negative and the polarity is not identified on the indicator in RPS but is indicated on the computer point.

Auxiliary power supply in associated RPS Cabinet 17 V to 23 V PDS/PMS Computer Point ( - 17 V to - 23 V)

RPS-C NI-3 COMP POWER SUPPLY (E0553)

RPS-D NI-4 COMP POWER SUPPLY (E0554)

Answer A. Plausible

1203.021 LOSS OF NEUTRON FLUX INDICATION PAGE:

9 of 12 CHANGE: 015 SECTION 2 - LOSS OF ONE OR MORE INTERMEDIATE RANGE NI CHANNELS IN MODE 2

3.

IF BOTH intermediate range channels have failed, THEN:

A.

REFER TO TS 3.3.10 for required actions.

B.

IF reactor power is greater than or equal to 2%,

THEN PERFORM plant shutdown using applicable steps of 1102.016, Power Reduction and Plant Shutdown.

C.

During shutdown:

1)

IF Reactor Power Wide Range Recorder (NR-502) is available, THEN MONITOR reactor power.

2)

WHEN ~1E-4 log reactor power on NR-502 is reached, THEN OBSERVE that source range indicators come on scale.

4.

IF one intermediate range channel only is operable, THEN CONTINUE plant operations (TS 3.3.10).

NOTE RPS modules are typically left in the tripped state following a failure to preserve any information for engineering, I&C, etc. to aid in troubleshooting.

5.

IF normal voltage is NOT indicated, THEN:

A.

WHEN directed by SM, THEN RESET power supplies as follows:

1)

CHECK detector power supply toggle switch ON.

2)

CHECK auxiliary power supply toggle switch ON.

NOTE Resetting intermediate range channel power supplies will cause voltage spikes and neutron power signal spikes which can cause CRD WITHDRAWAL INHIBITED (K08-A2) and HI SUR ROD HOLD (K08-B2).

3)

DEPRESS BOTH the detector power supply AND auxiliary power supply toggle switches to RESET position.

4)

IF intermediate range indication returns to normal, THEN CONTINUE plant operations.

END Must interpret that IRM channels are both inoperable and know the transition to perform TS 3.3.10 Continued operations allowed if only 1 IRM channel is inoperable.

Intermediate Range Neutron Flux 3.3.10 ANO-1 3.3.10-1 Amendment No. 215,264 3.3 INSTRUMENTATION 3.3.10 Intermediate Range Neutron Flux LCO 3.3.10 One intermediate range neutron flux channel shall be OPERABLE.

APPLICABILITY:

MODE 2 MODES 3, 4, and 5 with any control rod drive (CRD) trip breaker in the closed position and the CRD System capable of rod withdrawal.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required channel inoperable.

---

NOTE-------------------

Plant temperature changes are allowed provided the temperature change is accounted for in the SDM calculations.

A.1 Suspend operations involving positive reactivity changes.

AND A.2 Open CRD trip breakers.

Immediately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.10.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.10.2 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program

Source Range Neutron Flux 3.3.9 ANO-1 3.3.9-1 Amendment No. 215 3.3 INSTRUMENTATION 3.3.9 Source Range Neutron Flux LCO 3.3.9 One source range neutron flux channel shall be OPERABLE.

APPLICABILITY:

MODES 2, 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Required source range neutron flux channel inoperable with 1E-10 amp on the intermediate range neutron flux channel.

---

NOTE-------------------

Plant temperature changes are allowed provided the temperature change is accounted for in the SDM calculations.

A.1 Suspend operations involving positive reactivity changes.

AND A.2 Initiate action to insert all CONTROL RODS.

AND A.3 Open control rod drive trip breakers.

AND A.4 Verify SDM to be within the limit provided in the COLR.

Immediately Immediately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 hour AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter B.

Required source range neutron flux channel inoperable with

> 1E-10 amp on the intermediate range neutron flux channel.

B.1 Initiate action to restore required channel to OPERABLE status.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Plausible if applicant believed source range instruments were inoperable.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1086 Source: Bank Rev: 2 Rev Date: 11/3/21 Originator: Cork TUOI:

A1LP-RO-TS Objective:

10 System Number: 059 System

Title:

Accidental Liquid Radwaste Release Section: 4.2 Type:

Generic APEs

==

Description:==

Knowledge of conditions and limitations in the facility license.

K/A Number: 2.2.38 CFR

Reference:

41.7 / 41.10 / 43.1 Point Value: 1 RO Imp:

3.6 SRO Imp:

4.5 Tier:

1 Group:

2 RO Select:

No SRO Select: Yes Question:

Given:

- Unit 1 is at 100% power

- Chemistry reports the "A" OTSG secondary leak rate is 35 gpd

- A report from the Inside AO reveals a Turbine Building Trench continuous release is in progress

- Discharge Flume (RI-3618) subsequently goes into high alarm Which of the following actions is required for the above conditions?

A. Obtain a grab sample and perform gamma and I-131 analysis per ODCM L2.3.1.

B. Suspend the release and initiate a special report per Technical Specifications 3.7.4, Secondary Activity.

C. Obtain a grab sample and perform gamma and I-131 analysis per Technical Specifications 3.7.4, Secondary Activity.

D. Suspend the release and initiate a condition report per ODCM L2.3.1.

Answer:

D. Suspend the release and initiate a condition report per ODCM L2.3.1.

Notes:

Answer "D" is Correct. Per ODCM L2.3.1 B.1 and B.2. This action is also consistent with guidance in 1104.044, Turbine Building Draining System, and 1203.014, Control of Secondary System Contamination when a primary to secondary tube leak is in progress (also refer to 1203.023, Small Steam Generator Tube Leaks).

During tube leaks only batch releases are allowed. The ODCM spec for continuous releases is not applicable since the release should be terminated immediately. Only batch releases are allowed and the sampling requirements are not met for a batch release so the release should be suspended and a condition report initiated.

Answer "A" is Incorrect. Plausible as this is the correct action (and correct licensing basis document) for exceeding I-131 limits for continuous releases of secondary coolant but continuous releases should be isolated per 1203.014, Control of Secondary System Contamination when a primary to secondary tube leak >30 gpd is in progress.

Answer "B" is Incorrect. Plausible as this is the correct action but the incorrect licensing basis document.

Technical Specifications adds to the plausibility since during a SG tube leak TS LCO 3.4.13 for primary-to-secondary leakage and 3.7.4 for secondary system activity could apply.

Answer "C" is Incorrect. Plausible correct action for exceeding I-131 limits for continuous releases of secondary coolant but this distractor gives the wrong licensing basis document.

Difficulty:

2 Taxonomy: H RO:

SRO:

83

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 General Discussion:

This question meets the K/A since it concerns accidental liquid radwaste releases (an alarming radiation monitor and tube leak constitute an accidental liquid release since the sampling and analysis requirements are not met) and the question concerns the knowledge of the actions required in the facility license.

SRO Justification:

10 CFR 55.43(b)(4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

History:

New question for 2016 SRO exam Revised question per NRC examiner suggestions. JWC 7/15/16 Rev. 2 - 11/3/21 KS

- Changed K/A from 2.3.11 to APE 059 2.2.38 to match 2022 Exam Outline

- Changed Leak Rate from 15 gpd to 35 gpd to meet new requirement in 1203.023.

- Reformatted answer justification Used in 2022 SRO Exam.

References:

Offsite Dose Calculations Manual, L2.3.1 1203.023, Small Steam Generator Tube Leaks 1203.014, Control of Secondary System Contamination

ARKANSAS NUCLEAR ONE ODCM Revision 30 53 L 2.3 RADIOACTIVE LIQUID EFFLUENTS L 2.3.1 Radioactive material released to the discharge canal shall:

a.

For dissolved or entrained noble gases, be limited to a total concentration of 2 x 10-4 µCi/ml.

b.

For radioactive nuclides other than dissolved or entrained noble gases, be limited to the concentration specified in 10 CFR 20, Appendix B, Table II, Column 2.

c.

During any calendar quarter, result in a dose commitment to a MEMBER OF THE PUBLIC of 1.5 mrem to the total body and 5 mrem to any organ.

d.

During any calendar year, result in a dose commitment to a MEMBER OF THE PUBLIC of 3 mrem to the total body and 10 mrem to any organ.

e.

Be processed by a LIQUID RADWASTE TREATMENT SYSTEM when accumulative dose during a calendar quarter is projected to exceed 0.18 mrem to the total body and/or 0.625 mrem to any organ.

APPLICABILITY:

At all times.

ACTIONS

---

NOTE------------------------------------------------------------

Separate Condition entry is allowed for each Limitation L 2.3.1.a through L 2.3.1.e above and for each BATCH RELEASE and CONTINUOUS RELEASE Surveillance requirement not met.

CONDITION REQUIRED ACTION COMPLETION TIME A. Any limit listed in L 2.3.1.a through L 2.3.1.e not met.

A.1 Initiate action to restore to within limit.

AND A.2 Initiate a condition report to document the condition, determine any limitations for the continued effluent release operations, and track the condition for inclusion in the Radioactive Effluent Release Report pursuant to TS 5.6.3 (ANO-1) / TS 6.6.3 (ANO-2).

Immediately Immediately

ARKANSAS NUCLEAR ONE ODCM Revision 30 54 L 2.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. --------------NOTE--------------

Only applicable to BATCH RELEASE.

Sampling and/or analysis requirements not met.

B.1 Verify associated effluent release suspended.

AND B.2 Initiate a condition report to document the condition and determine any limitations for the continued effluent release operations.

Immediately Immediately C. --------------NOTE--------------

Only applicable to CONTINUOUS RELEASE of secondary coolant.

Secondary coolant dose equivalent I-131 (DEI)

> 0.01 µCi/ml.

C.1 Obtain a grab sample of the associated secondary coolant.

AND C.2 Perform gamma isotopic and I-131 analysis of sample.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 12 hours following sample acquisition D. Annual dose limits of L 2.3.1.d projected to exceed 40 CFR 190 limits.

D.1 Apply for a variance from the NRC to permit releases in excess of 40 CFR 190 limits.

Prior to exceed 40 CFR 190 limits E. Required Action(s) and/or Completion Time(s) of Conditions C and/or D not met.

OR Sampling and/or analysis requirements not met.

E.1 Initiate a condition report to document the condition and determine any limitations for the continued effluent release operations.

Immediately

1203.023 SMALL STEAM GENERATOR TUBE LEAKS PAGE:

3 of 12 CHANGE: 026

4.

IF total SG tube leakage (both SGs) is greater than or equal to 1 gpm (1,440 gpd),

THEN:

A.

IF turbine trips, THEN:

1)

TRIP the reactor.

2)

GO TO 1202.006, Tube Rupture.

B.

IF reactor trips, THEN GO TO 1202.006, Tube Rupture.

5.

IF SG Leak Rate is greater than or equal to 75 gpd AND rate-of-change is greater than or equal to 30 gpd/hour, THEN:

A.

INITIATE a shutdown to less than 50% within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1.0%/minute or higher per 1203.045, Rapid Plant Shutdown, Rapid Plant Shutdown With Small SG Tube Leak section in conjunction with this procedure.

B.

PLACE unit in Mode 3 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by continuing shutdown at 0.5%/minute or higher per 1203.045, Rapid Plant Shutdown With Small SG Tube Leak section in conjunction with this procedure.

C.

GO TO step 8 in conjunction with plant shutdown and cooldown.

6.

IF SG leak rate is greater than or equal to 30 gpd or confirmed leakage spikes exceed 30 gpd AND rate of change is less than 30 gpd/hr, THEN:

A.

INITIATE a controlled shutdown per 1203.045, Rapid Plant Shutdown With Small SG Tube Leak section at 0.5%/minute or higher to be in Mode 3 in less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> without delay.

B.

GO TO step 8 in conjunction with plant shutdown and cooldown.

7.

IF SG leak rate is greater than or equal to 5 gpd, THEN:

A.

MONITOR N-16 detectors and secondary activity levels.

B.

PROCEED as directed by Senior Manager, Operations in conjunction with this procedure.

C.

GO TO step 9.

8.

PERFORM 1203.014, Control of Secondary System Contamination.

1203.014 CONTROL OF SECONDARY SYSTEM CONTAMINATION PAGE:

3 of 8 CHANGE: 019 NOTE The remainder of the steps in this procedure are typically performed by non-Control Room personnel.

5.

IF trench dump is in progress, THEN PLACE the following handswitches in OFF to stop trench dump:

Trench Sump Pump (P-122A) (HS-5635)

Trench Sump Pump (P-122B) (HS-5636)

Emergency Trench Sump Pump (P-97) (HS-3613)

6.

Securing systems:

PERFORM 1106.031, MSR Drain Demineralizer Operation, "Removing MSR DI from Service" section.

IF the plant is being shut down, THEN PERFORM 1104.003, Chemical Addition, "Securing Zinc Injection" section.

7.

IF plant shutdown is required, THEN:

A.

Align Condensate Polishers to prevent wide spread contamination of polisher resin and reduce secondary system activity level as follows:

1)

INFORM Control Room personnel of intent to remove all but two polishers from service.

NOTE To minimize radiation exposure to personnel at the polisher controls and in the train bay, C & D polishers are preferred to remain in service.

2)

IF only one polisher is in service, AND flow can be maintained above 1500 gpm/polisher with two polishers, THEN PLACE an idle polisher in service per 1106.024, Condensate Demineralizer System Operation, applicable Placing Standby Condensate Demin in Service (Without Recycle) section.

ES-401, Page 21 of 52 ES-401 5

Figure 2-1 Screening for SRO-Only Linked to 10 CFR 55.43(b)(2)

(Technical Specifications)

Can the question be answered solely by knowing 1-hour TS/TRM Action?

RO question Yes No Can the question be answered solely by knowing the LCO/TRM information listed above the line?

Yes RO question No Can the question be answered solely by knowing the TS safety limits?

Yes RO question No Does the question involve one or more of the following for the TS, TRM, or ODCM:

application of required actions (TS Section 3) and SRs (TS Section 4) in accordance with rules of application requirements (TS Section 1) application of generic LCO requirements (LCO 3.0.1 through 3.0.7 and SR 4.0.1 through 4.0.4) knowledge of TS bases that is required to analyze TS-required actions and terminology SRO-only question Yes No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1116 Source: Bank Rev: 2 Rev Date: 9/08/21 Originator: Cork TUOI:

A1LP-RO-AOP Objective:

3 System Number: 067 System

Title:

Plant fire on site Section: 4.2 Type:

Generic APEs

==

Description:==

Ability to determine and interpret the following as they apply to the Plant Fire on Site: Vital equipment and control systems to be maintained and operated during a fire K/A Number: AA2.16 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

4.0 Tier:

1 Group:

2 RO Select:

No SRO Select: Yes Question:

Given:

- Unit 1 100% power

- Unit 2 has a fire in their Control Room

- Heavy smoke has accumulated in the Unit 1 Control Room The CRS will direct actions in_____(1)_____to control SG pressures 950 to 1020 psig using_____(2)_____.

A. (1) Alternate Shutdown (1203.002)

(2) Atmospheric Dump Valves B. (1) Alternate Shutdown (1203.002)

(2) Turbine Bypass Valves C. (1) Remote Shutdown (1203.029)

(2) Atmospheric Dump Valves D. (1) Remote Shutdown (1203.029)

(2) Turbine Bypass Valves Answer:

D. (1) Remote Shutdown (1203.029)

(2) Turbine Bypass Valves Notes:

Answer "D" is Correct. Since a fire is not forcing evacuation of the Control Room, MSIVs will remain open and the Turbine Bypass Valves will be used to control SG pressures.

Answer "A" is Incorrect. Plausible because Alternate Shutdown is a CR evacuation procedure and local manual operation of the ADVs is available.

Answer "B" is Incorrect. Plausible because Alternate Shutdown is a CR evacuation procedure and correctly identifies the use Turbine Bypass Valves.

Answer "C" is Incorrect. Plausible because it identifies the correct procedure and local manual operation of the ADVs is available.

General Discussion:

This question matches the KA because it requires knowledge of what equipment will be used to control steam generator pressures during a plant fire on site. Given that U1 and U2 control rooms are in such close proximity that they share the same control room envelope, and that the stem states there is smoke accumulating in the U1 control room, this fire is directly affecting U1 therefore the applicant is responsible for directing any actions Difficulty:

3 Taxonomy: H RO:

SRO:

84

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 contained in U1 procedures for a U2 fire.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

The question isn't asking what procedure is required to be entered for this event, but rather what procedure contains the guidance to control steam generator pressures. The guidance to control SG pressures using ADVs in 1203.002 Alternate Shutdown Exhibit A is directed from several other procedures such as 1202.008, Blackout and 1202.007, Degraded Power, even when entry conditions to 1203.002 have not been met. Therefore this question is SRO Only since it is asking what procedure section will be used not entry conditions to the AOP.

History:

New question for 2017 SRO Re-exam Rev. 1, 5/16/17 Editorial changes only Rev. 2 9/8/21 Changed KA from Sys. 068 KA 2.1.23 to Sys. 067 AA2.16 to fit 2022 SRO Exam Outline Re-ordered distractors to fit normal 2x2 convention.

Updated distractor analysis.

Re-worded stem so not stating entry conditions of procedure.

Used in 2022 SRO Exam

References:

1203.029, Remote Shutdown 1203.002, Alternate Shutdown

1203.029 REMOTE SHUTDOWN PAGE:

8 of 27 CHANGE: 014 SECTION 1 - REMOTE SHUTDOWN WITH AFW PUMP AVAILABLE PAGE 6 OF 11 SECTION 1B - CRS FOLLOW-UP ACTIONS (continued)

D.

MONITOR Makeup Tank (T-4) level.

1)

IF level continues to drop, THEN:

a)

CONFIRM with RO #1/RO #2 that BWST Outlet to running Makeup Pump (CV-1407, CV-1408) is open.

b)

DIRECT RO #2, at Makeup Tank Outlet CV-1275 breaker (B5652), to close CV-1275.

2)

IF Makeup Tank level rises above 86",

THEN DIRECT RO #2, at Makeup Tank Outlet CV-1275 breaker (B5652), to open CV-1275.

NOTE

~8 psig steam pressure change = ~1°F T-Hot change = ~5" PZR level change.

7.

CHECK SG pressure.

A.

IF Turbine Bypass Valves fail to maintain SG pressures at 950 to 1020 psig, THEN DIRECT RO #1/RO #2 to take manual control of affected valve(s) and maintain SG pressure at 950 to 1020 psig (REFER TO 1106.016, Condensate, Feedwater and Steam System Operation, Exhibit B, Turbine Bypass Valve (TBV) Manual Operation):

SGB to E-11A Turb BYP Valves (CV-6687 and CV-6688)

SGA to E-11B Turb BYP Valves (CV-6689 and CV-6690)

8.

STABILIZE plant at Mode 3, greater than 525°F conditions:

RCS temperature 540 to 560°F RCS pressure 2050 to 2250 psig SG levels at low level limit, 20 to 40 inches PZR level 90 to 110 inches 1203.029 directs use of TBVs to control steam generator pressure.

1203.002 ALTERNATE SHUTDOWN PAGE:

40 of 85 CHANGE: 031 SECTION 2 PAGE 1 OF 7 MAINTAINING PLANT AT MODE 3, GREATER THAN 525°F INITIAL CONDITIONS

1.

CHECK BOTH of the following:

Recovery Actions {RA} are completed Section 1A Shift Manager Actions are completed to the extent that plant is stable in Mode 3, greater than 525°F natural circulation conditions.

2.

IF DG2 is operating, THEN ENSURE the following:

DC control power de-energized DG2 Output breaker (A-408) closed.

3.

IF DG1 is operating, THEN ENSURE the following:

DC control power de-energized DG1 Output breaker (A-308) closed

4.

IF vital MCCs were de-energized, THEN CHECK vital MCCs are energized per the following:

• , Motor Control Centers B55 and B56 Recovery

• , Motor Control Centers B51, B52 and B53 Recovery

• , Motor Control Centers B61, B62 and B63 Recovery.

5.

CHECK SG levels are 300" to 340" and being maintained by ANY of the following:

EFW CFW

6.

CHECK SG pressure is being maintained by local manual operation of ADVs.

7.

CHECK RCS pressure and PZR level are being maintained by intermittent manual operation of an HPI pump.

INITIAL CONDITIONS Continued on next page Alternate Shutdown directs use of ADVs.

Page 3 of 22.

SYSTEMS: The following are safe shutdown system components used by this procedure and a brief description of how they are used.

1.

Emergency Feedwater (EFW) or Common Feedwater (CFW)

If EFW is affected by the fire to the extent that neither P-7A nor P-7B is capable of providing water to the SGs automatically, CFW will be started and operated to maintain SG level. This is a Recovery Action for Fire Area G with a Fire PRA assumed availability time of 27.5 minutes.

Local control of EFW Pump (P-TA) is established with the EFW Turb K3 Trip/Throt Valve (CV-6601A). EFW P-7A to SG-B lsol (CV-2620) is manually controlled to control SG-B level.

Manual control of EFW P-7A to SG-A lsol (CV2627) is established to control SG-A level.

Steam release will be initially controlled by Main Steam Safeties and, as additional operators become available, by manual control of the ADVs.

CFW - CST (T-41) will provide the initial source of water to the CFW system with Unit 2 CST3 (2T-41A/B) as the backup source.

EFW CST (T-41 B) will provide the initial source of water to the EFW system with the CST (T-41) and Service Water, the backup sources, if necessary.

As time permits, P-TB and its associated train of EFW will be made available for backup purposes.

2.

Electrical Power If offsite power is lost, EDGs will be started, tied to their respective buses and overridden to prevent spurious shutdown. A3 and A4 breakers will have their control power removed to prevent spurious actuations (A3 and A4 energized with one Service Water pump on each bus, all other loads secured).

480-volt buses (B5 and B6) will be deenergized to prevent spurious actuations. Vital 480-volt load centers will later be re-energized after certain breakers have been opened to prevent undesirable spurious actuations. These buses are required to support extended DG operations, HPI and to power the battery chargers.

CHANGE 029 Alternate Shutdown Basis document cites use of ADVs to control steam release.

1202.007 DEGRADED POWER CHANGE 017 PAGE 3 of 85 INSTRUCTIONS CONTINGENCY ACTIONS

2. (Continued)

E. IF AAC Gen is available, THEN perform the following:

1) Energize vital bus using ES Electrical System Operation (1107.002), "Placing Alternate AC Generator on bus A3 (A4)" section, while continuing with this procedure.

2) GO TO step 3.

F. IF AAC Gen is not available, THEN perform the following:

1) Verify P36B Bus Select MOD Control selected to bus with operating EDG.

a) IF P36B MOD was transferred, THEN dispatch an operator to perform Attachment 3, Makeup Pump P36B Alignment to Operating EDG, while continuing with this procedure.

2) Verify P4B Bus Select MOD Control selected to bus with operating EDG.

3) IF SG press associated with de-energized ATM Dump ISOL is

> 1040 psig, THEN as time permits, dispatch an operator with a radio to hand jack associated ATM Dump CNTRL to maintain SG press 1000 to 1040 psig.

(Refer to Alternate Shutdown (1203.002), Exhibit A).

Example where 1203.002 guidance is used to control SG pressure even though 1203.002 entry conditions have not been met.

1202.008 BLACKOUT CHANGE 020 PAGE 24 of 43 INSTRUCTIONS CONTINGENCY ACTIONS

43. IF additional manpower becomes available, THEN DISPATCH operator with a radio to hand jack ATM Dump CNTRL valves (CV-2618 and CV-2668) to maintain stable CET temp using 1203.002, Alternate Shutdown, Exhibit A, Local Operation of ADVs.

44. CHECK EFW CST (T-41B) level remains greater than or equal to 5.1 feet.

44. OPERATE valves manually to shift EFW pump suction using 1106.006, Emergency Feedwater Pump Operation, EFW Pump (P-7A or P-7B) Suction Transfer section.

45. PERFORM 1203.050, Unit 1 Spent Fuel Pool Emergencies, while continuing with this procedure. [INPO IER 11-2]

46. MAINTAIN Mode 3, greater than 525F, until directed otherwise by Senior Manager, Operations AND PERFORM the following:

A. IF a DG is placed in service, THEN GO TO 1202.007, DEGRADED POWER.

B. IF off-site power becomes available, THEN RETURN TO step 34.

END Example where 1203.002 guidance is used to control SG pressure even though 1203.002 entry conditions do not exist.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1387 Source: New Rev: 0 Rev Date: 12/9/21 Originator: K. Smith TUOI:

A1LP-RO-ANE.1 Objective:

2 System Number: A07 System

Title:

Flooding Section: 4.3 Type:

Babcock and Wilcox

==

Description:==

Ability to use plant computers to evaluate system or component status.

K/A Number: 2.1.19 CFR

Reference:

41.10 / 43.5 Point Value: 1 RO Imp:

SRO Imp:

3.8 Tier:

1 Group:

2 RO Select:

No SRO Select: Yes Question:

Given:

- U1 is at 100% power

- Sustained heavy rains upstream in the Arkansas River Basin have caused the Arkansas River level to rise

- The highest predicted level is expected to be 352 ft

- Service Water pumps P-4A and P-4C are in service aligned to the lake

- The CRS has entered OP-1203.025, Natural Emergencies section 6 - Flood Based on the attached SPDS trend, what procedure is required to be performed by OP-1203.025, Natural Emergencies?

A. OP-1203.054, Internal Flooding - Att. 6 Emergency Pumping Diesel Fuel Vault.

B. OP-1203.045, Rapid Plant Shutdown - Section 1 RPSD without SG Tube Leak.

C. OP-1107.001, Electrical System Operations - Placing Startup 2 Transformer in Service.

D. OP-1104.029, Service Water and Auxiliary Cooling Water - Transfer Service Water Bays from Lake to Emergency Cooling Pond.

Answer:

B. OP-1203.045, Rapid Plant Shutdown - Section 1 RPSD without SG Tube Leak Notes:

Answer "B" is Correct. 1203.045 states to monitor lake level using the given SPDS screen. At 345 ft. lake level, a rapid plant shutdown is required. Since the given lake level is 348 ft. a rapid plant shutdown should be performed.

Answer "A" is Incorrect. Plausible since at 345 ft. Att. B Local Flooding Actions directs securing the Diesel Fuel Vault, however emergency pumping per 1203.054 Att. 6 isn't performed until indications show that flooding is present. With the flood occurring from the river and not due to localized rain, it can be assumed the site will not experience flooding until river level exceeds local site elevation. Local site elevation is 354', much less than the 348' given in the SPDS graph.

Answer "C" is Incorrect. This action is not performed until 354 ft. in accordance with 1203.025.

Answer "D" is Incorrect. Plausible since lake level is rising it can be assumed that swapping to the ECP would be a safer configuration. This action isn't performed in the flooding section of OP-1203.025, but is performed in section 5 for a loss of lake.

General Discussion:

This question matches the K/A since it requires the ability to interpret the SPDS trend to evaluate the abnormal condition of high level.

Difficulty:

3 Taxonomy: H RO:

SRO:

85

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

History:

Used in 2022 SRO Exam.

References:

• • • • • • • • • • • REFERENCE REQUIRED************

-show P-4A SPDS screen 1203.025

1203.025 NATURAL EMERGENCIES PAGE:

70 of 89 CHANGE: 077 SECTION 6 - FLOOD INSTRUCTIONS

1.

NOTIFY Unit 2 Control Room.

2.

ESTABLISH contact with US Army Corps of Engineers, Dardanelle Lock and Dam (e.g., 479-970-8827, 501-340-1229, Emergency Telephone Directory) for peak flood condition forecasts and updates.

3.

NOTIFY Transmission Control Center-North (TCC-N).

NOTE SPDS displays level in feet. PMS/PDS displays level in inches above 324' (reference level).

Instructions that follow provide level in feet and corresponding level from PMS/PDS in brackets, e.g.,

340' [PMS 192 in.].

At flood levels above 349' [PMS 300 in.], SPDS and PMS/PDS are off-scale above sensor span.

4.

MONITOR lake level using ONE or more of the following methods:

SPDS MONITOR SW pump aligned to lake (P-4A, P-4B1, P-4B2, P-4C)

PMS/PDS MONITOR SW or Circ bay aligned to lake (SW bays L3664, L3666, L3668, and B & C Circ Bays L3601, L3602)

WHEN lake level exceeds 349' [PMS 300 in.],

THEN MONITOR lake level locally on at least an hourly basis.

5.

REVIEW out of service or degraded flood barriers using Operations Home Page, Maintenance, Flood Deficiency Database link (if available) or equivalent access to the applicable spreadsheet.

6.

INITIATE required compensatory actions as necessary for out of service or degraded flood barriers.

NOTE After flooding begins, BWST level indication can be affected depending upon flooding level at the transmitter. If this is the case, physical verification of level could be required. This could include visual verification via manways or non-intrusive methods such as thermal imaging (EC-62823).

7.

MAINTAIN BWST (T-3) level greater than 4.5 ft (LIS-1411/L1411, LIS-1421/L1421).

8.

CONSULT Operations Management to determine need for plant shutdown.

P-4A is the given SPDS trend.

1203.025 NATURAL EMERGENCIES PAGE:

71 of 89 CHANGE: 077 SECTION 6 - FLOOD

9.

EVALUATE the need to call out additional personnel for support.

10.

COORDINATE with Unit 2 and CONTACT appropriate personnel (work management, etc.) to seal bus duct drains associated with Startup Transformer (SU-2) using WO#424627-01 or equivalent (PMID#50010669-18).

11.

REQUEST Duty Manager initiate Model WO#402438 to install and stage FLEX Phase II equipment in preparation for a potential Beyond Design Basis External Event resulting in an Extended Loss of AC Power (ELAP).

12.

WHEN Lake Dardanelle level greater than 345 ft. (PMS 252 in),

THEN:

A.

BEGIN plant shutdown per 1203.045, Rapid Plant Shutdown.

B.

DISPATCH operators to perform Local Flooding Actions, Attachment B.

C.

PERFORM the following:

CONSIDER relocating the B.5.b pump to higher elevation (e.g., RERTC parking lot).

IF notified that a railcar or a vehicle which has the potential to weigh 10,000 lbs or greater is present in the Train Bay, THEN DIRECT TSC, OCC or other site management group to remove railcar/vehicle and CONFIRM removal of railcar/vehicle.

NOTE The backflow preventers for the Unit 1 Void Area Drains are typically stored in a toolbox near the ladder to the void area in Room 47.

Access to the void area (Room 83) requires RP and Security assistance.

IF flood level at the site is expected to exceed 354, THEN DIRECT appropriate personnel (TSC, if manned; Work Week Manager, etc.) to install Unit 1 void area backflow preventers per EC-50090 or equivalent.

DIRECT maintenance to commence WO #00374472-02 for installation of temporary pumps in Unit 1 Aux building.

NOTE Blind flange with gasket are located inside the flood mitigation toolbox located on floor in room 72 (Solid Waste Filler Storage Room).

DIRECT maintenance to install blind flange in discharge of Drumming Station and Hot Machine Shop Supply Fan (2VSF-38) per MWO 00404581.

DIRECT maintenance to commence WO #00374472-03 for installation of temporary pumps in Diesel Fuel vault.

Correct since lake level is 348 ft.

Att. B is performed at this level as well which directs securing the DG fuel vault.

However this does not direct performing emergency pumping

1203.025 NATURAL EMERGENCIES PAGE:

75 of 89 CHANGE: 077 SECTION 6 - FLOOD

20.

PRIOR to flood waters exceeding elevation 354', PERFORM the following:

A.

SECURE non-essential electrical loads.

B.

ENSURE ALL necessary work is completed on SU 2 Xfmr.

NOTE SU XFMR #2 load limits with no fans or oil pumps running apply:

161 KV winding: 95 amps at 161 KV (27 MVA) 6900V winding: 1255 amps at 6900V (15 MVA) 4160V winding: 1745 amps at 4160V (12.6 MVA)

C.

COORDINATE with Unit 2 Control Room to transfer plant auxiliaries to SU 2 Xfmr using 1107.001, Electrical System Operation.

D.

COORDINATE with Transmission Control Center-North (TCC-N) to de-energize SU 1 Xfmr.

21.

IF a potential or actual threat exists that could affect SF Pool integrity, level or cooling capability, THEN PERFORM 1203.050, Unit 1 Spent Fuel Pool Emergencies in conjunction with this procedure. (INPO IER 11-2 Rec 4)

22.

For each component confirmed in position by Attachment B, INSTALL a Caution Tag stating, This component is positioned for Unit 1 flooding concerns. Contact the Unit 1 Control Room prior to repositioning.

23.

ANNOTATE on the Shift Turnover Sheet that performance of Attachment B of 1203.025 is required daily while Lake Dardanelle is greater than 345 ft.

24.

IF T-25 Diesel Oil Storage Tank is NOT available, THEN REFER TO 1104.023, Diesel Oil Transfer Procedure, Filling Diesel Fuel Oil Tanks When Bulk FO Storage Tank (T-25) is Unavailable section.

25.

CONDUCT further operations as directed by plant management.

END SU2 transformer placed in service prior to exceeding 354', however highest predicted level is 352' so this action would not be required.

1203.025 NATURAL EMERGENCIES PAGE:

68 of 89 CHANGE: 077 SECTION 5A/5B - LOSS OF DARDANELLE RESERVOIR DISCUSSION Circ Water Pumps are stopped when pump performance indicates a loss of suction. It is desirable to complete a cooldown while Circ Water is in service and the condenser is available. If lake level is forecasted to reach 334', an immediate plant shutdown at a safe rate commensurate with lake level loss is performed. The required submergence for the Service Water Pumps makes it necessary to swap SW suction to the Emergency Cooling Pond at a bay level of 332'. ACW is secured because ACW cannot be returned to the ECP and the safety analysis for the ECP does not account for ACW losses.

If Dardanelle Reservoir is unavailable due to a chemical/oil spill, non-functional Intake Structure components or any other reason, the Circ Water Pumps will be stopped. This causes a loss of the condenser, normal feedwater and steam condensing capability. SW is transferred to the ECP. If both units are affected, makeup to the ECP will be severely limited.

The ECP accident analysis assumes that during a loss of lake event, no more than 160,000 gallons of ECP volume (16 feet) is retained in EFW CST (T-41B) from EFW pump recirc. Therefore, if SW is aligned to EFW suction, action is taken to shift EFW suction back to T-41B in order to restore ECP inventory to analyzed volume.

REFERENCES Calculation No. 91-E-0099-10, Rev. 1, ECP Peak Temperature and Inventory Loss Analysis Summary.

Condition Report C-96-0245 Reduction of Site Sensitivity to ECP Boundary Valve and Sluice Gate Degradation.

Condition Report CR-C-ANO-2005-1206, Evaluation of Differences in ECP Design Versus Existing ECP Configuration.

Condition Report CR-1-ANO-2007-1508, EFW Pumps P-7A/P-7B recirc line is not seismically supported, and ECP inventory analysis does not account for EFW recirc routed to T-41B.

SW is transferred to the ECP in the same procedure but only in the section for a loss of lake, not for flooding.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1324 Source: New Rev: 0 Rev Date: 8/31/21 Originator: K. Smith TUOI:

A1LP-RO-POISN Objective:

5 System Number: 004 System

Title:

Chemical and Volume Control System (CVCS)

Section: 3.1 Type:

Reactivity Control

==

Description:==

Ability to interpret reference materials, such as graphs, curves, tables, etc.

K/A Number: 2.1.25 CFR

Reference:

43.2 / 43.5 Point Value: 1 RO Imp:

SRO Imp:

4.2 Tier:

2 Group:

1 RO Select:

No SRO Select: Yes Question:

• • • • • • • • • • • • • • • • • • • • • • • • • • • REFERENCE PROVIDED********************************

U1 has shutdown for a refueling outage with the following conditions present:

- RCS Average Temperature is 500°F

- Boric Acid Addition Tank (BAAT T-6) volume is 5,000 gallons

- Boric Acid Addition Tank (BAAT T-6) boron concentration is 10,000 ppm

- Boric Acid Addition Tank (BAAT T-6) temperature is 85°F Based on these parameters,_____(1)_____to provide sufficient boric acid can be delivered to the RCS to ensure 1% subcritical margin at_____(2)_____at the worst time in core life with a stuck control rod assembly and after xenon decay.

A. (1) The BAAT T-6 boron concentration is too low (2) 280°F B. (1) The BAAT T-6 boron concentration is too low (2) 200°F C. (1) Boric acid flow is not assured (2) 280°F D. (1) Boric acid flow is not assured (2) 200°F Answer:

D. (1) Boric acid flow is not assured (2) 200°F Notes:

Answer "D" is Correct. TRM 3.5.1 bases state that the BAAT T-6 and its associated piping are maintained at least 10°F above crystallization temperature to assure a flow of boric acid. The applicant should analyze both OP-1103.004 Att E and Att F and determine that BAAT volume and boron concentration are acceptable per Att E, but that BAAT temperature per Att F is not at least 10F above crystallization temperature. TRM 3.5.1 bases also states the quantity of boric acid in storage is sufficient to borate the reactor coolant system to a 1%

subcritical margin in the cold condition (200°F) at the worst time in core life with a stuck control rod assembly and after xenon decay.

Answer "A" is Incorrect. Plausible if the applicant mis-reads Att E. This is probable since the applicant should analyze Att E. which is acceptable when above and to the left of the curve, and Att F which is acceptable when below and to the right of the curve. 280F is plausible since this represents the upper limit of Mode 4, Hot Shutdown but is incorrect since the TRM bases states the analysis assumes reaching Cold Shutdown (200F).

Difficulty:

3 Taxonomy: H RO:

SRO:

86

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 Answer "B" is Incorrect. Plausible if the applicant mis-reads Att E. This is probable since the applicant should analyze Att E. which is acceptable when above and to the left of the curve, and Att F which is acceptable when below and to the right of the curve.

Answer "C" is Incorrect. 280F is plausible since this represents the upper limit of Mode 4, Hot Shutdown but is incorrect since the TRM bases states the analysis assumes reaching Cold Shutdown (200F).

General Discussion:

This question matches the KA since it requires the ability to interpret OP-1103.004 Att E and Att F to determine if assumptions in the TRM bases are met.

SRO Justification:

Requires application of SR requirements to determine that the 10F margin to crystallization temperature is not met, therefore the assumptions in the TRM Bases are not met.

History:

Used in 2022 SRO Exam.

References:

• • • • • Handout 1103.004 Att. E, Att. F, and TRM 3.5.1*****

OP-1103.004 TRM 3.5.1 Bases

1103.004 Rev. 036 Page 87 of 111 SOLUBLE POISON CONCENTRATION CONTROL Attachment E Page 1 of 1 BAAT VOLUME AND CONCENTRATION vs. RCS T-ave (Ref. TRM Figure 3.5.1-1)

BAAT Volume and Concentration Vs. RCS T-ave (Ref. TRM Figure 3.5.1-1) 0 1000 2000 3000 4000 5000 6000 7000 8000 200 300 400 500 600 RCS Average Temp, °F Boric Acid Addition Tank Volume, gal.

8700 ppm 9500 ppm 10,000 ppm 12,000 ppm Operation above and to the left of the applicable curve is acceptable.

Intersection of given Volume and RCS Tave is above and to the left of the 10,000 ppm line which is acceptable

1103.004 Rev. 036 Page 88 of 111 SOLUBLE POISON CONCENTRATION CONTROL Attachment F Page 1 of 1 SOLUBILITY OF BORIC ACID IN WATER (Crystallization Temperature) 3,000 5,000 7,000 9,000 11,000 13,000 15,000 17,000 19,000 21,000 23,000 20 40 60 80 100 120 140 160 TEMPERATURE (5° F increments)

PPM BORON (500 ppm increments)

Crystallization Temperature 10°F Above Crystallization Operation is Allowed Below and to Right of Solid Curve Intersection of the given boron concentration and the BAAT temperature is NOT below and to the right of the Solid curve, therefore the temperature is less than 10F above crystallization temperature.

Makeup and Chemical Addition Systems B 3.5.1 ANO-1 TRM B 3.5.1-1 Rev. 5,15,16,39,62 TRM B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

TRM B 3.5.1 Makeup and Chemical Addition Systems BASES BACKGROUND The Makeup and Chemical Addition systems provide control of the reactor coolant system (RCS) boron concentration (Ref. 1). This is normally accomplished by using any of the three makeup pumps in series with a boric acid pump associated with the boric acid addition tank (BAAT).

The quantity of boric acid in storage from either of the two above mentioned sources is sufficient to borate the reactor coolant system to a 1% subcritical margin in the cold condition (200 °F) at the worst time in core life with a stuck control rod assembly and after xenon decay.

Minimum volumes (including a 10% safety factor) as specified by Figure 3.5.1-1 for the BAAT or an operable BWST (Ref. 3) will each satisfy this requirement. The specification assures that adequate supplies are available whenever the reactor is heated above 200 °F so that a single failure will not prevent boration to a cold condition. The minimum volumes of boric acid solution given include the boron necessary to account for xenon decay.

The principal method of adding boron to the primary system is to pump the concentrated boric acid solution (8700 ppm boron, minimum) into the Makeup Tank using the 25 gpm boric acid pump.

The alternate method of addition is to inject boric acid from the BWST using the makeup pumps (Ref. 2).

Concentration of boron in the boric acid addition tank may be higher than the concentration which would crystallize at ambient conditions. For this reason and to assure a flow of boric acid is available when needed, the BAAT and its associated piping are maintained at least 10 °F above the crystallization temperature for the concentration present. Once in the makeup system, the concentrate is sufficiently well mixed and diluted such that normal system temperatures assure boric acid solubility.

ACTIONS The Required Actions are modified by a Note exempting entry into applicable Conditions when the flow path from the boric acid pump to the Makeup Tank is unavailable during procedurally controlled activities such as alternate Makeup Tank dilution methods (requiring isolation of the boric acid pump from the Makeup Tank), boric acid pump testing, and boric acid makeup to the spent fuel pool (SFP), BWST, and core flood tanks (CFTs). Exemptions are not permitted for degraded or non-functional components. Condition entry is not required in such instances since the Chemical Addition System remains FUNCTIONAL with system re-alignment achievable in a minimal period of time.

Assumes 200F Cold Shutdown is achieved.

10F above crystallization temperature assures flow.

ES-401, Page 21 of 52 ES-401 5

Figure 2-1 Screening for SRO-Only Linked to 10 CFR 55.43(b)(2)

(Technical Specifications)

Can the question be answered solely by knowing 1-hour TS/TRM Action?

RO question Yes No Can the question be answered solely by knowing the LCO/TRM information listed above the line?

Yes RO question No Can the question be answered solely by knowing the TS safety limits?

Yes RO question No Does the question involve one or more of the following for the TS, TRM, or ODCM:

application of required actions (TS Section 3) and SRs (TS Section 4) in accordance with rules of application requirements (TS Section 1) application of generic LCO requirements (LCO 3.0.1 through 3.0.7 and SR 4.0.1 through 4.0.4) knowledge of TS bases that is required to analyze TS-required actions and terminology SRO-only question Yes No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1391 Source: New Rev: 0 Rev Date: 1/16/22 Originator: K. Smith TUOI:

A1SPG-SRO-EAL Objective:

1 System Number: 2.4 System

Title:

Component Cooling Water System (CCWS)

Section: 3.8 Type:

Plant Service Systems

==

Description:==

Knowledge of the emergency action level thresholds and classifications.

K/A Number: 2.4.41 CFR

Reference:

41.10 / 43.5 Point Value: 1 RO Imp:

SRO Imp:

4.6 Tier:

2 Group:

1 RO Select:

No SRO Select: Yes Question:

Given:

- U1 has been shutdown due to high RCS activity 20 minutes after shutdown:

- Chemistry reports RCS activity is 35 µCi/gm dose equivalent I-131

- Pressurizer Level and Makeup Tank Level are both lowering slowly

- P-32A Reactor Coolant Pump Seal Injection Flow indicates 15 gpm

- All other Reactor Coolant Pump Seal Injection Flows indicate approximately 5 gpm

- ICW Surge Tank T-37A level is stable over the last 2 minutes

- ICW Surge Tank T-37B level has risen 0.4 psid over the last 2 minutes In accordance with OP-1903.010 Emergency Action Level Classification, the SM should declare a(n)__________.

A. Unusual Event B. Alert C. Site Area Emergency D.General Emergency Answer:

B. Alert Notes:

Answer "B" is Correct. 1203.039 estimates 333 gallons per psid in T-37B. 333(gallons/psid) x 0.4(psid/2 minutes) = 66.6 gallons/min. 66.6gpm is greater than the 50 gpm threshold for a potential loss of the RCS barrier RCB2. With a potential loss of RCS barrier, an Alert FA1.1 should be declared.

Answer "A" is Incorrect. Plausible since an Unusual Event would be declared due to Identified RCS leakage above 25 gpm per SU5.1. Since the RCS leakage is > 50 gpm the higher classification (FA1.1) should be declared.

Answer "C" is Incorrect. Plausible since a > 50 gpm Steam Generator Tube Rupture exists which indicates a potential Loss of the Reactor Coolant System barrier. If the SG was also Faulted outside containment then a loss of Containment barrier would exist simultaneously which would be a SAE on Loss or Potential Loss of any two barriers.

Answer "D" is Incorrect. Plausible because the applicant could mistaken the SG tube rupture has a Loss of Containment and a potential loss of the RCS barrier. They could also mistake the RCS Activity as a loss of Fuel Clad barrier, however the threshold is at 300 µCi/gm not 30 µCi/gm for a Loss of Fuel Clad.

Difficulty:

3 Taxonomy: H RO:

SRO:

87

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 General Discussion:

This question matches the KA since it requires knowledge of EAL thresholds and classifications associated with the ICW system. Specifically being able to identify changing parameters within the ICW system and translate them into the correct EAL classification.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

History:

Used in 2022 SRO Exam.

References:

OP-1203.039 Excess RCS Leakage OP-1903.010 Emergency Action Level Classification

INSTRUCTIONS CONTINGENCY ACTIONS

3. (Continued)

B. IF leakage into RB Sump is indicated, THEN GO TO step 14.

C. MONITOR ANY of the following for indications of RCS leakage into ICW system:

Nuclear Loop ICW activity rising Indication of Letdown Cooler RCS leak into ICW [Letdown Cooler ICW Outlet temp rising on PMS (T2214 for E29A; T2215 for E29B)]

Indication of RCP Seal Cooler RCS leak into ICW:

RCP Seal Temp rising RCP Seal Bleedoff Temp rising Skewed RCP Seal Injection Flows NOTE ICW Surge Tank T-37B Level (PDIS 2229) 0.5 to 2.7 psid (1 psid = 333 gallons).

D. DISPATCH an operator to determine Nuclear Loop ICW Surge Tank (T37B) level trend.

E. IF leakage into ICW is indicated, THEN GO TO step 5.

(3. CONTINUED ON NEXT PAGE)

Conversion of T-37B level to gallons. (gallons per psid).

PROC./WORK PLAN NO.

1903.010 PROCEDURE/WORK PLAN TITLE:

EMERGENCY ACTION LEVEL CLASSIFICATION PAGE: 133 of 243 CHANGE:

059 - Emergency Action Level Technical Bases Barrier:

Reactor Coolant System Category:

A - RCS or S/G Tube Leakage Degradation Threat:

Potential Loss Threshold:

RCB2 UNISOLABLE RCS leakage or S/G tube leakage > 50[44] gpm excluding normal reductions in RCS inventory (e.g., letdown, RCP seal leakoff)

Definition(s):

FAULTED - The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

Basis:

Failure to isolate the leak (from the Control Room or locally), within 15 minutes or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used makeup [charging]

pump, but an ESAS [ESFAS] actuation has not occurred.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system.

The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.

If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold CNB1 will also be met.

Reference(s):

1.

1SAR 9.1 Makeup and Purification System

2.

2SAR 9.3.4 Chemical and Volume Control System

3.

NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A

PROC./WORK PLAN NO.

1903.010 PROCEDURE/WORK PLAN TITLE:

EMERGENCY ACTION LEVEL CLASSIFICATION PAGE: 112 of 243 CHANGE:

059 - Emergency Action Level Technical Bases Category:

F - Fission Product Barrier Degradation Subcategory:

N/A Initiating Condition:

Any loss or any potential loss of either Fuel Clad or RCS EAL:

FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table 1[2]F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table 1[2]F-1 lists the fission product barrier thresholds, bases and references.

At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1.

Reference(s):

1.

NEI 99-01 FA1

PROC./WORK PLAN NO.

1903.010 PROCEDURE/WORK PLAN TITLE:

EMERGENCY ACTION LEVEL CLASSIFICATION PAGE: 223 of 243 CHANGE:

059 - Emergency Action Level Technical Bases Category:

S - System Malfunction Subcategory:

5 - RCS Leakage Initiating Condition:

RCS leakage for 15 minutes or longer EAL:

SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for 15 min. (Note 1)

OR RCS identified leakage > 25 gpm for 15 min. (Note 1)

OR Reactor coolant leakage to a location outside containment > 25 gpm for 15 min. (Note 1)

Note 1:

The Emergency Director should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded. The Emergency Director is not allowed an additional 15 minutes to declare after the time limit is exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

Basis:

Failure to isolate the leak (from the Control Room or locally) within 15 minutes, or if known that the leak cannot be isolated within 15 minutes, from the start of the leak requires immediate classification.

Steam generator tube leakage is identified RCS leakage.

This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

The first and second EAL conditions are focused on a loss of mass from the RCS due to unidentified leakage", "pressure boundary leakage" or "identified leakage (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage) or a location outside of containment.

Plausible to meet FCB6 Plausible to confuse pri-mary to secondary leak as CNB1 Only applicable EAL in table 1F-1 Correct

PROC./WORK PLAN NO.

1903.010 PROCEDURE/WORK PLAN TITLE:

EMERGENCY ACTION LEVEL CLASSIFICATION PAGE: 10 of 243 CHANGE:

059 chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

3.1.6 Emergency Director Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition.

A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.

3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process clock starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process clock started.

When assessing an EAL that specifies a time duration for the off-normal condition, the clock for the EAL time duration runs concurrently with the emergency classification process clock.

For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 4.1.8).

3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two units, a Site Area Emergency should be declared.

There is no additive effect from multiple EALs meeting the same ECL. For example:

If two Alert EALs are met, whether at one unit or at two units, an Alert should be declared.

3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1327 Source: New Rev: 1 Rev Date: 1/12/22 Originator: K. Smith TUOI:

A1LP-RO-AICS Objective:

6 System Number: 039 System

Title:

Main and Reheat Steam System (MRSS)

Section: 3.4 Type:

Heat Removal From Reactor Core Secondary System

==

Description:==

Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Malfunctioning steam dump K/A Number: A2.04 CFR

Reference:

41.5 / 43.5 Point Value: 1 RO Imp:

SRO Imp:

3.7 Tier:

2 Group:

1 RO Select:

No SRO Select: Yes Question:

Given:

- U1 is at 100% power

- A control signal failure causes CV-6689 and CV-6690, Steam Generator 'A' turbine bypass valves (TBVs) to fail OPEN

- Generated Megawatts are 840 MWe and LOWERING Assuming no action is taken by the crew, 15 seconds after the TBVs fail open, Governor Valve Demand will be_____(1)_____the pre-transient value.

The CRS should direct_____(2)_____to mitigate the consequences of this malfunction.

A. (1) above (2) 1203.001, ICS Abnormal Operation Section 2 - Undesired Power Change B. (1) above (2) 1106.016, Condensate, Feedwater and Steam System Operation Exhibit B - Turbine Bypass Valve Manual Operation C. (1) below (2) 1203.001, ICS Abnormal Operation Section 2 - Undesired Power Change D. (1) below (2) 1106.016, Condensate, Feedwater and Steam System Operation Exhibit B - Turbine Bypass Valve Manual Operation Answer:

A. (1) above (2) 1203.001, ICS Abnormal Operation Section 2 - Undesired Power Change Notes:

Answer "A" is Correct. ICS will see generated MW lower due to steam short cycling to the condenser and will raise governor valve demand in order to attempt to maintain MW. The governor valve demand will rise and ultimately stop and settle out at the built in valve limiter. This condition meets the definition of an "uncorrected upset to controlling steam pressure" which requires application of Section 2 - Undesired Power Change of 1203.001 ICS Abnormal Operations. Section 2 will result in placing the affected TBVs in Hand and close since the Main Generator would still be on line.

Answer "B" is Incorrect. 1106.016 Exhibit B is plausible since the stem states it is a control signal failure which may lead the applicant to believe that attempting to close the TBVs remotely would be unsuccessful. This action is taken as a contingency in 1202.003, Overcooling after an attempt to close the TBVs was made in the control room but was unsuccessful.

Difficulty:

4 Taxonomy: H RO:

SRO:

88

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 Answer "C" is Incorrect. Governor valve demand below the pre-transient value is plausible since governor valves also operate to maintain Steam Header Pressure. With TBVs failing open, Steam Header pressure will lower and stabilize at a lower value, but the demand for generated MW overrides the demand for steam header pressure and the governor valve demand will be above the pre-transient value. Also if the applicant believes the reactor will trip due to this failure, they could select "below" as an option.

Answer "D" is Incorrect. Governor valve demand below the pre-transient value is plausible since governor valves also operate to maintain Steam Header Pressure. With TBVs failing open, Steam Header pressure will lower and stabilize at a lower value, but the demand for generated MW overrides the demand for steam header pressure and the governor valve demand will be above the pre-transient value. Also if the applicant believes the reactor will trip due to this failure, they could select "below" as an option. 1106.016 Exhibit B is plausible since the stem states it is a control signal failure which may lead the applicant to believe that attempting to close the TBVs remotely would be unsuccessful. This action is taken as a contingency in 1202.003, Overcooling after an attempt to close the TBVs was made in the control room but was unsuccessful.

General Discussion:

This question matches the KA since it requires the ability to predict the impact of failed open TBVs on plant operations and then to select the correct procedure to correct the malfunction.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

History:

Rev. 1 - 1/12/22 KS

- Based on validation comments, deleted bullet providing SG pressure and added bullet to provide information on MWe.

Used in 2022 SRO Exam

References:

STM-1-64 OP-1203.001

1203.001 ICS ABNORMAL OPERATION PAGE:

6 of 26 CHANGE: 018 SECTION 2-- UNDESIRED POWER CHANGE ENTRY CONDITIONS One or more of the following:

Undesired power change with ALL of the following:

Feedwater flows trending together Reactor controls responding with feedwater Crosslimits NOT in effect SASS MISMATCH (K07-B4) for Controlling Steam Pressure Uncorrected upset to Controlling Steam Pressure

1203.001 ICS ABNORMAL OPERATION PAGE:

7 of 26 CHANGE: 018 SECTION 2 - UNDESIRED POWER CHANGE INSTRUCTIONS (REFERENCE USE)

NOTE SG/RX Master output changes ~3% per second when toggle is depressed.

Feedwater flow is a leading indicator of heat balance power. When feedwater flows are less than pre-transient values, heat balance power is expected to update less than 100%.

1.

PLACE SG/RX Master in HAND.

2.

CHECK steam pressure control is NOT impacted as indicated by:

SASS MISMATCH (K07-B4) NOT alarmed for Controlling SG Pressure Controlling SG Pressure(s) controlled between 870 and 930 psig Throttle Pressure Error (IC08) greater than +/- 10 psi is corrected by Turbine (when IC08 is displayed)

3.

IF steam pressure control is impacted, THEN:

A.

ENSURE Turbine is in TURB MAN control.

B.

IF TURB BYP Valves affected, THEN PLACE affected TURB BYP Valves H/A station(s) in HAND.

4.

ADJUST components in HAND/TURB MAN as necessary to achieve the following:

Reactor Power less than 100%

SG Pressures stable

5.

IF ICS control parameters are NOT stabilized, THEN GO TO Section 1, Heat Transfer Upset.

6.

ADJUST components in HAND/TURB MAN as necessary to achieve the following prescribed bands:

Reactor Power less than 100% at band provided by CRS Controlling SG Pressure(s) 880 to 910 psig (with TURB BYP Valves closed if Main Generator is on line)

Steam pressure control is impacted TBVs taken to HAND TBVs are closed since the Main Generator is online.

1202.003 OVERCOOLING CHANGE 013 PAGE 14 of 36 INSTRUCTIONS CONTINGENCY ACTIONS

19. CHECK ATM Dump ISOL valves closed:

19. PERFORM the following:

CV-2619 CV-2676 A. IF EITHER ATM Dump Control System is being used for SG press control, THEN ENSURE proper operation:

SG A SG B CV-2676 ATM Dump ISOL CV-2619 CV-2668 ATM Dump CNTRL CV-2618 B. IF ATM Dump Control System is NOT being used for SG press control, THEN CLOSE applicable ATM Dump ISOL valve(s) (CV-2619 or CV-2676).

20. CHECK TURB BYP Valves closed.

20. PERFORM the following:

A. PLACE affected TURB BYP Valve H/A Station(s) in HAND:

TURB BYP VALVES LOOP A TURB BYP VALVES LOOP B B. CLOSE affected TURB BYP Valve(s).

C. IF turbine bypass valve(s) fail to close, THEN DISPATCH an operator to close affected valve(s) locally using 1106.016, Condensate, Feedwater and Steam System Operation, Turbine Bypass Valve (TBV) Manual Operation (Exhibit B)

(located on column between TBVs), while continuing with this procedure.

(20. CONTINUED ON NEXT PAGE)

TBVs are manually closed using 1106.016 after an attempt to close in remote is unsuccessful.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1328 Source: Bank Rev: 0 Rev Date: 9/06/21 Originator: K. Smith TUOI:

A1LP-RO-EOP08 Objective:

14 System Number: 062 System

Title:

A.C. Electrical Distribution Section: 3.6 Type:

Electrical

==

Description:==

Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Degraded system voltages K/A Number: A2.16 CFR

Reference:

41.5 / 43.5 Point Value: 1 RO Imp:

SRO Imp:

2.9 Tier:

2 Group:

1 RO Select:

No SRO Select: Yes Question:

Given:

- Unit 1 is in a Blackout condition

- Voltage has been recovered on SU#2 and is 144 kV on C10 In order to mitigate this event, the CRS must direct the performance of_____(1)_____in OP-1202.008 Blackout PRIOR to restoration of offsite power to A-3 and A-4 to ensure_____(2)_____.

A. (1) Attachment 1, Blackout Breaker Alignment and UV Relay Defeat (2) automatic trips of equipment necessary to protect the core are disabled B. (1) Attachment 1, Blackout Breaker Alignment and UV Relay Defeat (2) excess current during starting of the motors will NOT occur C. (1) Attachment 2, Recovery from Blackout Breaker Alignment and UV Relay Defeat (2) automatic trips of equipment necessary to protect the core are disabled D. (1) Attachment 2, Recovery from Blackout Breaker Alignment and UV Relay Defeat (2) excess current during starting of the motors will NOT occur Answer:

A. (1) Attachment 1, Blackout Breaker Alignment and UV Relay Defeat (2) automatic trips of equipment necessary to protect the core are disabled Notes:

Answer "A" is Correct. With degraded voltage on SU#2, Att. 1 is required to defeat the UV interlocks. With UV trips active, the equipment could trip due to the low SU#2 voltage.

Answer "B" is Incorrect. Plausible because low voltage conditions can cause higher currents, however Att. 1 would have no effect on the actual starting current for motors.

Answers "C" & "D" are Incorrect. Att. 2 will only be performed to restore the UV interlocks that were bypassed with Att. 1 when SU#2 voltage is restored to normal.

General Discussion:

This question matches the KA since it requires knowledge of the impacts that low SU#2 transformer voltage would have on safety related systems and how to defeat the UV interlocks to be able to restore those systems to service.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Difficulty:

3 Taxonomy: H RO:

SRO:

89

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 History:

Updated Bank QID 790 to fit 2022 SRO Exam.

Re-worded stem and answers to fit 2022 KA selection and fit SRO criteria.

Used in 2022 SRO Exam.

References:

OP-1202.008

1202.008 BLACKOUT CHANGE 020 PAGE 21 of 43 INSTRUCTIONS CONTINGENCY ACTIONS

34. (Continued).

NOTE Off-site power is considered degraded if both conditions are true:

SU1 voltage is less than 22KV (SPDS: E22ATU1)

AND Any of the following conditions exist:

SU XFMR #2 Voltage Regulator:

out-of-service with input voltage less than 162.5KV (C10 indication or SPDS1: E161ST21)

OR in service with the following conditions, NOT in Automatic 3% reduction enabled Input voltage less than 155.4KV (C10 indication or SPDS1: E161ST21)

Unit 2 SU XFMR #2 loads are:

Non-Vital loads Greater than 4.4MVA Neither 161KV transmission line in service (Russellville East or Pleasant Hill)

B. IF voltage is indicated on either Startup XFMR and is degraded, THEN GO TO step 47.

C. IF voltage on both Startup XFMRs is zero, THEN GO TO step 35.

END SU2 voltage is 144KV therefore off-site power is degraded.

Since there is voltage on SU2 go to step 47.

1202.008 BLACKOUT CHANGE 020 PAGE 25 of 43 INSTRUCTIONS CONTINGENCY ACTIONS

47. DISPATCH operator to perform, "Blackout Breaker Alignment and UV Relay Defeat".

NOTE Off-site power is considered restored to normal if either of the following conditions exists:

SU1 is available with greater than or equal to 22KV (SPDS: E22ATU1)

OR All of the following conditions met for SU2:

SU XFMR #2 Voltage Regulator:

out-of-service with input voltage greater than or equal to 162.5KV (C10 indication or SPDS1: E161ST21)

OR in service with the following conditions, In Automatic 3% reduction disabled Input voltage greater than or equal to 155.4KV (C10 indication or SPDS1: E161ST21)

ANY Unit 2 SU XFMR #2 loads are BOTH:

Vital loads ONLY Less than or equal to 4.4MVA ONE or BOTH 161KV transmission lines in service (Russellville East or Pleasant Hill)

A. IF off-site power is restored to normal, THEN DISPATCH operator to perform, "Recovery From Blackout Breaker Alignment and UV Relay Defeat" AND RETURN TO step 9.

would be performed if Attachment 1 was performed then off-site power was restored to normal.

1202.008 BLACKOUT CHANGE 020 PAGE 26 of 43 INSTRUCTIONS CONTINGENCY ACTIONS

48. WHEN Attachment 1 is complete, THEN RE-ENERGIZE A1, A2, H1 and H2 by performing the following for each bus:

A. CHECK associated bus L.O. RELAY TRIP alarm clear on K02.

Bus A1 Bus A2 Bus H1 Bus H2 K02-A6 K02-A7 K02-A4 K02-A5 A. DETERMINE AND CORRECT cause of L.O. RELAY TRIP before energizing bus, while continuing with this procedure (REFER TO 1107.001, Electrical System Operation, "Reclosing Tripped Bus or MCC Feeder Breakers" section).

B. IF buses are to be energized from SU2, THEN NOTIFY Unit 2.

C. TURN SYNC switch for associated bus feeder breaker ON:

Bus A1 Bus A2 Bus H1 Bus H2 A-113 or A-111 A-213 or A-211 H-15 or H-13 H-25 or H-23

1) CLOSE associated breaker from handswitch.

1) RESET breaker anti-pump feature by taking associated handswitch to PULL-TO-LOCK AND releasing.

2) IF neither A1 or A2 is energized, THEN RETURN TO step 34.

1202.008 BLACKOUT CHANGE 020 PAGE 27 of 43 INSTRUCTIONS CONTINGENCY ACTIONS

49. RE-ENERGIZE A3 and A4 by performing the following for each bus.

A. CHECK associated bus L.O. RELAY TRIP alarm clear on K02:

Bus A3 Bus A4 K02-B6 K02-B7 A. DETERMINE AND CORRECT cause of L.O. RELAY TRIP before energizing bus, while continuing with this procedure (REFER TO 1107.001, Electrical System Operation, "Reclosing Tripped Bus or MCC Feeder Breakers" section).

B. TURN SYNC switch for associated bus feeder breaker ON:

Bus A3 Bus A4 A-309 A-409 C. CLOSE associated breaker from handswitch.

C. IF non-vital bus voltage is less than 3160 V, THEN DISPATCH operator to close the associated breaker(s) in LOCAL (to override sync-check relays):

A-309 (A1 Feed to A3)

A-409 (A2 Feed to A4)

1202.008 BLACKOUT CHANGE 020 PAGE 28 of 43 INSTRUCTIONS CONTINGENCY ACTIONS CAUTION During degraded voltage conditions the following problems might occur:

Motors might trip on overload, overheat due to high running currents, or stall.

MCC starter might not pick up to energize loads.

AC auxiliary relays might not pick up to provide interlock or load energization features.

50. RESTART only equipment absolutely necessary to protect the core as follows:

A. ENSURE suction and discharge flow path aligned.

B. REVIEW system operating procedure to ensure essential pump services available.

C. CONSIDER closing centrifugal pump discharge valve before starting to reduce starting current.

D. For motors selected to be started, DISPATCH operator to PERFORM the following at the breaker:

1) CLOSE DC control power breaker

2) PLACE breaker control in REMOTE E. IF both Unit 1 and Unit 2 are aligned to SU2, THEN COORDINATE starting loads with Unit 2.

F. Start loads necessary to protect the core as follows:

1) START one motor at a time.

2) ALLOW motor to reach run speed.

G. MONITOR motor winding temp if available.

51. CHECK EFW CST (T-41B) level remains greater than or equal to 5.1 feet.

51. OPERATE valves manually to shift EFW pump suction using 1106.006, Emergency Feedwater Pump Operation, EFW Pump (P-7A or P-7B) Suction Transfer section.

After Attachment 1 is complete and power restored to buses, only start equipment necessary to protect the core.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1329 Source: New Rev: 0 Rev Date: 9/07/21 Originator: K. Smith TUOI:

A1LP-RO-TS Objective:

11 System Number: 076 System

Title:

Service Water System (SWS)

Section: 3.4 Type:

Heat Removal From Reactor Core Secondary System

==

Description:==

Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of SWS K/A Number: A2.01 CFR

Reference:

41.5 / 43.5 Point Value: 1 RO Imp:

3.5 SRO Imp:

3.1 Tier:

2 Group:

1 RO Select:

No SRO Select: Yes Question:

Given:

- U1 is at 100% power

- Service Water Pumps P-4A and P-4C are in service

- Service Water Pump P-4B is aligned to A4

- Debris has been building in the intake canal for several shifts

- Intake canal water level is rising The following alarms are received:

- K05-F1, TRAV SCREEN SYSTEM TROUBLE

- K10-A4, SW BAY LEVEL LOW NOW

- Service Water Bay P-4A level is 331.8 ft and slowly lowering

- Service Water Bay P-4C level is 331.6 ft and slowly lowering What action is required to ensure Service Water cooling is maintained to vital systems?

A. Start P-4B per OP-1203.030, Loss of Service Water Attachment B, Placing Standby Service Water Pump Into Service.

B. Divert Service Water return to the Emergency Cooling Pond per OP-1203.025 Section 5A Loss of Dardanelle Reservoir Due to a Loss of Water Inventory.

C. Swap Service Water Suctions to the Emergency Cooling Pond per OP-1104.029 Service Water and Auxiliary Cooling System, Transferring Service Water Bays from Lake to Emergency Cooling Pond.

D. Backwash Service Water Bay Strainers per OP-1104.029 Service Water and Auxiliary Cooling System Backwashing Service Water Bay Strainers section.

Answer:

C. Swap Service Water Suctions to the Emergency Cooling Pond per OP-1104.029 Service Water and Auxiliary Cooling System, Transferring Service Water Bays from Lake to Emergency Cooling Pond.

Notes:

Answer "C" is Correct. Debris has been building in the intake canal for several shifts causing clogging of the traveling screens and service water bay strainers. This has resulted in low service water bay levels which are continuing to lower. The applicant should enter 1203.025 section 5B due to a loss of lake that is not due to a loss of inventory. Section 5B states to swap to the ECP iaw 1104.029.

Answer "A" is Incorrect. Plausible since this action would be required if bay levels were greater than 332 ft.

Also the question asks what action is required to ensure Service Water cooling is maintained to vital systems.

Difficulty:

3 Taxonomy: H RO:

SRO:

90

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 Even if bay levels were greater than 332 ft but still lowering, starting P-4B would only provide additional cooling for a short period of time since the service water system is losing its suction source and needs to be aligned to the ECP.

Answer "B" is Incorrect. Plausible since this would be required if the cause of the loss of service water bay level was due to a loss of lake Dardanelle, however the stem states there is excessive debris in the intake canal which is clogging the intake screens. If section 5A was entered, the 2nd step states to transition to 5B if the lake is unavailable for any other reason besides a loss of water inventory which makes this answer incorrect.

Answer "D" is Incorrect. Plausible since this action is performed as directed by K10-A4 ACA. Since the stem states bay levels are less than 332 ft and lowering, and intake canal level is rising with excessive debris, this action will not ensure service water cooling is maintained. The suction source must be aligned to the ECP prior to 332 ft per 1203.025.

General Discussion:

This question matches the K/A since it requires the ability to predict the impact of excessive debris in the intake canal on the service water system and select the correct procedure to mitigate this event.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

History:

Used in 2022 SRO Exam

References:

OP-1203.025 OP-1203.030

1203.030 LOSS OF SERVICE WATER PAGE:

3 of 27 CHANGE: 029 INSTRUCTIONS CONTINGENCY ACTIONS NOTE Failure of traveling screens can clog SW Bay Strainers and Pump Discharge Strainers.

SW pump operation with bay level less than 332 feet (8 feet or 96 inches indicated bay level locally or by PDS/PMS) can damage pump.

4.

CHECK operating SW Pump discharge strainers are clogged.

4.

IF SW Bay Levels are greater than 332 feet, THEN:

IF operating SW Pump trips, AND this is an emergency, THEN:

A. At CRS/SM discretion, ATTEMPT to restart one time.

B. As time permits, REFER TO 1107.001, Electrical System Operation, Reclosing Tripped Individual Load Supply Breakers" section.

IF a standby SW Pump is available, THEN START standby pump (P-4A, B, or C) per Attachment B, Placing Standby Service Water Pump Into Service.

CAUTION Maximum Service Water Pump flow is 8000 gpm.

5.

CHECK BOTH of the following exist:

5.

GO TO step 24.

A single SW Pump is operating.

A second SW Pump can NOT be restored within 5 minutes.

6.

PERFORM 1202.001, Reactor Trip in conjunction with this procedure.

7.

ENSURE the following valves are closed:

Generator H2 Temp Control (CV-4018)

Main Lube Oil Temp Control (CV-4026)

Answer A plausible since this would be performed if Bay levels were > 332 ft.

1203.012I ANNUNCIATOR K10 CORRECTIVE ACTION PAGE:

42 of 94 CHANGE: 062 PAGE 1 OF 3 Location: C16 Device and Setpoint:

SW P-4A Bay Level (LITS-3664) less than 336 ft, reset 336 ft 6 in SW P-4B Bay Level (LITS-3666) less than 336 ft, reset 336 ft 6 in SW BAY LEVEL LO SW P-4C Bay Level (LITS-3668) less than 336 ft, reset 336 ft 6 in Alarm: K10-A4 1.0 OPERATOR ACTIONS NOTE This alarm can coincide with operation of Diesel or Electric Fire Pumps (P-6A/B) due to proximity of relief discharge to level detector.

1.

IF alarm is due to operation of Diesel or Electric Fire Pumps (P-6A/B),

THEN:

A.

CHECK alarm is spurious.

B.

EXIT this procedure.

2.

DETERMINE which bay level is low using SPDS Service Water Diagnostic Instrumentation display or local indication.

3.

IF ALL bay levels are low due to loss of Dardanelle Reservoir, THEN:

A.

NOTIFY Unit 2 Control Room of low Lake Dardanelle level. {1}

B.

GO TO 1203.025, Natural Emergencies, Loss of Dardanelle Reservoir section.

4.

REFER TO 1203.030, Loss of Service Water.

5.

IF A or C SW Bay level low due to clogging SW Bay Strainer, THEN PERFORM 1104.029, Service Water and Auxiliary Cooling System, Backwashing Service Water Bay Strainers section.

6.

CHECK for proper sluice gate alignment.

All procedures in the answer choices are plausible as they are mentioned in the ACA for SW bay level low. 1104.029 for back washing strainers would be performed, however since bay level is less than 332 ft, SW pumps must be aligned to the ECP iaw 1203.025.

1203.025 NATURAL EMERGENCIES PAGE:

37 of 89 CHANGE: 077 SECTION 5A - LOSS OF DARDANELLE RESERVOIR DUE TO A LOSS OF WATER INVENTORY ENTRY CONDITIONS ONE or more of the following alarms:

DARDANELLE RESERVOIR LEVEL LO (K15-B5)

SW BAY LEVEL LOW (K10-A4)

Visual observation of intake and discharge canal levels dropping rapidly.

Report by U.S. Army Corps of Engineers at Dardanelle Lock and Dam of ONE of the following:

Dardanelle dam failure Lake level will be drawn down below elevation 335' Section 5A would not be performed as intake canal levels are rising due to the excessive debris.

1203.025 NATURAL EMERGENCIES PAGE:

38 of 89 CHANGE: 077 SECTION 5A - LOSS OF DARDANELLE RESERVOIR DUE TO A LOSS OF WATER INVENTORY INSTRUCTIONS

1.

NOTIFY Unit 2 Control Room.

2.

IF Dardanelle Reservoir is unavailable due to:

Chemical/oil spill Inoperable Intake Structure components ANY other reason besides loss of water inventory THEN GO TO SECTION 5B - Loss of Dardanelle Reservoir Due to Impending or Immediate Unavailability of Lake.

3.

IF an earthquake caused the loss of Dardanelle Reservoir, THEN PERFORM Earthquake section in conjunction with this section.

4.

MONITOR bay/lake level using ONE of the following methods:

NOTE SPDS displays level in feet. PMS/PDS displays level in inches above 324' reference level. Instructions that follow give level in feet and corresponding level from PMS/PDS in brackets, e.g., 332' [PMS 96 in.].

SPDS MONITOR SW pump aligned to lake (P-4A, P-4B1, P-4B2, P-4C)

PMS/PDS MONITOR SW or Circ bay aligned to lake (SW bays L3664, L3666, L3668, and B & C Circ Bays L3601, L3602)

COORDINATE with Unit 2 and STATION individual with radio at Intake to monitor lake level.

5.

CONTACT US Army Corps of Engineers, Dardanelle Lock and Dam (e.g., 479-970-8827, 501-340-1229, Emergency Telephone Directory) to determine forecasted lake level.

6.

IF expected to lose the lake for SW pump suction source, THEN REQUEST Security to set appropriate security measures at ECP.

7.

IF unit is at power, THEN:

A.

COMMENCE 1203.045, Rapid Plant Shutdown to bring unit to Mode 3 in conjunction with this procedure.

B.

IF lake level is forecasted to reach 334',

THEN PERFORM a plant shutdown at the maximum safe rate.

8.

ADVISE Shift Manager to implement 1903.010, Emergency Action Level Classification.

Even if Section 5A was mistakenly entered, the 2nd step ensures transi-tion to section 5B since the cause is not a loss of water inventory.

1203.025 NATURAL EMERGENCIES PAGE:

58 of 89 CHANGE: 077 SECTION 5B - LOSS OF DARDANELLE RESERVOIR DUE TO IMPENDING OR IMMEDIATE UNAVAILABILITY OF LAKE ENTRY CONDITIONS ONE or more of the following alarms:

DARDANELLE RESERVOIR LEVEL LO (K15-B5)

SW BAY LEVEL LOW (K10-A4)

Dardanelle Reservoir unavailable due to a chemical/oil spill at the intake canal.

Dardanelle Reservoir unavailable due to non-functional Intake Structure components.

1203.025 section 5B would be entered.

1203.025 NATURAL EMERGENCIES PAGE:

60 of 89 CHANGE: 077 SECTION 5B - LOSS OF DARDANELLE RESERVOIR DUE TO IMPENDING OR IMMEDIATE UNAVAILABILITY OF LAKE

3. (Continued).

H.

TRIP ALL Circ Water Pumps:

Circ Water Pump (P-3A)

Circ Water Pump (P-3B)

Circ Water Pump (P-3C)

Circ Water Pump (P-3D)

I.

PERFORM the following security measures:

NOTIFY security that all Circ Water Pumps are off.

REQUEST security set security measures at ECP.

J.

NOTIFY Unit 2 to secure cooling tower blowdown.

K.

ENSURE NO liquid radwaste releases in progress on Unit 1 or Unit 2.

L.

ALIGN the following SW bays to the ECP per 1104.029, Service Water and Auxiliary Cooling System, Transferring Service Water Bays from Lake to Emergency Cooling Pond section:

P-4A SW Bay P-4B SW Bay P-4C SW Bay M.

IF 1104.028, ICW System Operating Procedure, Temporary Installation of a Service Water Outlet at ICW Cooler E-28C Attachment is installed AND ICW Cooler (E-28C) is in service on T-mod, THEN:

1)

CLOSE Temporary Outlet Valve (T-1).

2)

CLOSE Loop II Supply to ICW Coolers (CV-3811).

3)

CLOSE Loop I Supply to ICW Coolers (CV-3820).

4)

ENSURE ICW Cooler E-28C isolated by danger tags per EN-OP-102, Protective and Caution Tagging.

5)

IF desired to return ICW Cooler E-28C to service, THEN INITIATE actions to remove T-Mod.

(3. CONTINUED ON NEXT PAGE)

Section 5B is entered and SW suctions are aligned to the ECP iaw 1104.029.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1332 Source: New Rev: 0 Rev Date: 10/18/21 Originator: K. Smith TUOI:

A1LP-RO-ALNF Objective:

4 System Number: 015 System

Title:

Nuclear Instrumentation System Section: 3.7 Type:

Instrumentation

==

Description:==

Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Power supply loss or erratic operation.

K/A Number: A2.01 CFR

Reference:

41.5 / 43.5 Point Value: 1 RO Imp:

SRO Imp:

3.9 Tier:

2 Group:

2 RO Select:

No SRO Select: Yes Question:

Given:

- U1 is maintaining Reactor Power less than 80% per dispatcher request

- All SASS selector switches are in their normal lineup

- RPS Channel 'A' is de-energized

- Power Range Nuclear Instruments indicate the following:

- NI-5 = 0 %

- NI-6 = 79.7 %

- NI-7 = 80.3 %

- NI-8 = 80.2%

NOW:

- NI-6 unexpectedly begins to drift to 125% over 10 minutes If no action is taken, the NI input signal to ICS will_____(1)_____.

To stop the power rise, the CRS should FIRST direct actions in_____(2)_____.

A. (1) rise to 80.3% power and stabilize (2) ICS Abnormal Operation Section 1 - Heat Transfer Upset B. (1) rise to 80.3% power and stabilize (2) K07-B4 SASS Mismatch Annunciator Corrective Action C. (1) rise to 125% power over 10 minutes (2) ICS Abnormal Operation Section 1 - Heat Transfer Upset D. (1) rise to 125% power over 10 minutes (2) K07-B4 SASS Mismatch Annunciator Corrective Action Answer:

C. (1) rise to 125% power over 10 minutes.

(2) ICS Abnormal Operation Section 1 - Heat Transfer Upset Notes:

Answer "C" is Correct. SASS selects the high channel of either NI-5 or NI-6. While SASS can detect a fast failure of either NI-5 or NI-6 and select the good channel, a slow failure cannot be detected and the highest power will input to ICS. In this case, NI-6 will continue to input to ICS and affect the plant. OP-1203.021 states that if NI failure is affecting reactor power, then enter 1203.001 section 1 - heat transfer upset. This will ultimately place the diamond station to hand and stop the failing NI signal from inputting to ICS.

Answer "A" is Incorrect. The first part is plausible since SASS would automatically select the higher channel of NI-7 or NI-8 upon a fast failure of NI-6. Since the failure is slow (over 10 minutes), this would not occur and Difficulty:

2 Taxonomy: H RO:

SRO:

91

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 plant power would continue to rise.

Answer "B" is Incorrect. The first part is plausible since SASS would automatically select the higher channel of NI-7 or NI-8 upon a fast failure of NI-6. Since the failure is slow (over 10 minutes), this would not occur and plant power would continue to rise. Second part is plausible since NI-6 failing high will cause K07-B4 SASS Mismatch to alarm. The guidance to manually select NNI-Y is in this ACA, however it states if the ICS controlling input is impacted, then ensure affected stations are in manual. After ICS stations are placed in manual, this section will be performed to return ICS back in automatic, but not to stop the power rise. Stopping the power rise will be performed in ICS Abnormal Operation Section 1 - Heat Transfer Upset.

Answer "D" is Incorrect. Plausible since NI-6 failing high will cause K07-B4 SASS Mismatch to alarm. The guidance to manually select NNI-Y is in this ACA, however it states if the ICS controlling input is impacted, then ensure affected stations are in manual. After ICS stations are placed in manual, this section will be performed to return ICS back in automatic, but not to stop the power rise. Stopping the power rise will be performed in ICS Abnormal Operation Section 1 - Heat Transfer Upset.

General Discussion:

This question matches the KA since it requires the ability to predict the impacts of a controlling NI input to ICS on the plant and to select the correct procedure to minimize the impact.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

History:

Used in 2022 SRO Exam.

References:

OP-1203.012F K07 ACA OP-1203.021 OP-1203.001

Nuclear Instrumentation STM 1-67 Rev 16 20 power, an indicator that shows the setting of an internal link which sets the Coarse Range Gain, a 10 position switch for fine gain adjustment and a 10 turn dial for extra fine gain adjustment. There is also a test jack and a meter zero-adjust potentiometer on the front plate.

Refer to Figure 67.30 There are two Sum/Difference Amplifiers in each power range channel. One Sum/Difference Amplifier module sums the 2 linear amplifier signals. The other is used to take the difference between the two linear amplifiers and indicates the difference in flux between the top and bottom of the core. The outputs of this summing amplifier are:

a. Remote indications of total power (0-125%) on CO-3, (see Table 67.1) and to the plant computer.

b. Power level to auctioneer unit for ICS.

c. Power level signals to the following bistables:

1) Reactor high power trip.

2) Power to pumps trip.

3) Power at 10% which feeds Source Range rod withdrawal interlock c/o and EFIC functions

4) ARTS bypass bistables for main turbine and main feed pumps.

5) Power, imbalance and flow trip.

6) Shutdown bypass hi power trip.

The outputs of the difference amplifier are:

a. Remote indications (-62.5% to +62.5%) on CO-3 (see Table 67.5) and the plant computer for axial power distribution.

b. Difference signal to function generator. The function generator provides the delta flux portion of the trip signal to the flux /delta flux/flow trip bistable.

Refer to Figure 67.32 for face plate view.

The Function Generator is a Bailey E92358. It has two inputs and one output. One input is the difference amplifier output (Delta Flux) the other is RCS flow. The output of the Function Generator is the set point signal for the Power/Imbalance/Flow Bistable. (see STM-1-63 RPS for further explanation).

The power level of NI-5 and NI-6 are high auctioneered (highest signal is passed) in B RPS cabinet and sent as the X input to SASS for automatic signal selection for the ICS total power signal.

2.4.5 Summing and Difference Amplifiers 2.4.6 Function Generator 2.4.7 Auctioneer Unit How NI signals input to ICS.

Nuclear Instrumentation STM 1-67 Rev 16 21 Power level of NI-7 and NI-8 are high auctioned in C RPS cabinet and sent as the Y input to SASS for automatic signal selection for the ICS total power signal.

A power range test module, Bailey E923231, provides on line testing of the power range instrumentation. (See Figure 67.33) The module has a front panel rotary switch that provides manually variable test signals for testing or calibrating the 2 linear amplifiers, function generator, and the bistable modules associated with that channel.

When the power range test module is placed in any test mode, (not in Operate), the input to the auctioneer is placed at ground potential for that channel. Thus, since the auctioneer is a high auctioneer, its output will not be affected by the channel in test.

The SUR Rod Withdrawal Interlock provided by the Source and Intermediate Range instruments prevent withdrawal of the control rods when SUR exceeds 2 Decades Per Minute (DPM) in the source range or 3 DPM in the intermediate range. A bistable circuit trips to stop rod withdrawal until the SUR decreases to 0.5 DPM. It then automatically resets, and rod withdrawal can start again. The Power Range NIs bypass this interlock from both the Source and Intermediate Range channels when either channels 5 or 6 and channels 7 or 8 reach 10% reactor power. These interlocks and electrical block diagrams are discussed in detail in STM 1-2, "Control Rod Drive".

3.0 Associated Procedures and Technical Specifications Nuclear Instrumentation is monitored continuously during plant operation by observation of control board indicators on panel C03.

The nuclear instruments are used to monitor total power and radial and axial flux distribution. Additionally, operator logs require certain channel checks for each shift. For information on instruments checked and operational limits, see 1015.03A, "Operator Log Taking", and form 1015.03A-7, Reactor Protection System Logs.

Operators are directed by T.S. surveillances (logs) to check operation of the nuclear instrument channels in service, by observation, a minimum of once each shift above cold shutdown to verify proper response to existing plant conditions. This includes comparison of the output and/or state of independent channels measuring the same variable. The instrumentation section of Limiting Conditions for Operation in Tech Specs provides for operability requirements on NIs.

NI readings can be compared to heat balance and incore detectors for system operational checks and calibrations. The sum of the outputs of the two sections of each power range detector will be calibrated to a heat balance. The sum will be calibrated whenever it disagrees with the heat balance by 2% or more. The difference signal 2.4.8 Power Range Test Module 2.4.9 SUR Rod Withdrawal Interlock 3.1 Normal Operations.

Non-Nuclear Instrumentation System STM 1-69 Rev. 17 7

The NNIX and the NNIY signals for each parameter (i.e., NNIX and NNIY signal for OTSG S/U level) are supplied to the SASS module and the transfer relay. Within the SASS module the signals input to the signal conditioning board.

Signal conditioning board filters the input signal to remove noise and converts the signal to a 0 to +5 volt signal. Each signal conditioning board can receive up to 8 input signals (four pairs). A low pass filter (4 hertz) removes noise from the NNI input signal. A voltage converter changes the +/-10 volts input signal to a 0 to +5 volt signal that is required by the computer board. The resulting NNIX and NNIY signals are output to the computer board.

The computer board periodically (at a set scan rate) samples each input that is supplied from the signal conditioning board. When the computer is initialized (power on or reset), all inputs are set to zero and the program starts. Each input value is stored in memory on the computer board (provided no mismatch condition exists). The computer board determines if a mismatch or a trip condition is present. A mismatch condition is determined by comparing the NNIX and the NNIY signal. If the difference between the two signals exceeds a preset value, then a mismatch condition is triggered. When a mismatch condition exists, the computer performs the mismatch calculation but does not store new values of the input signal. A trip condition is determined by comparing the stored value of the NNIX signal (last value stored prior to the mismatch) to the present value of the NNIX signal. If the difference between the stored and present NNIX signals exceeds a preset value then a trip condition is triggered. The computer board outputs the trip signal, mismatch signal, and an Auto signal to the relay board. A reset button is located under the front panel. The reset switch is used to initialize the SASS modules. It is important to note that all inputs are reset and the SASS program starts. This will select all signals to the NNIX channel unless the NNIY signal is manually selected.

The relay board provides the relay interface between the transfer relay, the mismatch alarm and SASS automatic operation. The signals generated by the computer board are routed through the relay board to the front panel. Indications for mismatch, auto, trip X and trip Y are located on the front panel.

An alarm bypass module allows bypassing the mismatch alarm.

Placing the switch to the bypass position will defeat the mismatch signal from the associated channel to the control room annunciator.

All other functions of the SASS module remain functional. This allows another channel to actuate the control room mismatch annunciator.

The NNI signals can fail in two general ways. The input signal can exhibit a step change or the NNIX and NNIY signals can slowly diverge. When the NNIX and NNIY signals slowly diverge, the SASS system cannot determine which signal is at fault. For this condition a mismatch alarm is generated and the operator will have to take the appropriate actions. When an input signals exhibits a step change, the SASS system will determine which signal is at fault and 3.1.1 SASS Module Operation 3.1.2 SASS Logic SASS Logic describes the system response to a slow failure. In this case, NNI-X corresponds to NI-5 and 6.

NNI-Y corresponds to NI-7 and 8.

Non-Nuclear Instrumentation System STM 1-69 Rev. 17 8

transfer, if necessary. In either case, the mismatch condition will occur.

For a slow deviation between the NNIX and NNIY signals, the following sequence occurs:

NNIX and NNIY signals diverge generating a mismatch Computer generates a mismatch signal (front panel indication and annunciator alarm)

The computer checks for a trip condition (should not be present for a slow signal failure).

The computer is taken out of automatic for that signal pair.

When the NNIX and the NNIY signals are within the mismatch setpoint, the computer will return to automatic operation (capable of initiating a signal transfer).

For a step change or a rapid signal failure, the following sequence occurs:

NNIX and NNIY signals diverge generating a mismatch Computer generates a mismatch signal (front panel indication and annunciator alarm)

The computer checks for a trip condition (should be present for a step change of the input signal)

The computer scans the inputs for a second time and checks for a trip condition (the new input values are compared to the stored value prior to the mismatch).

If the first and second scan result in a trip condition the computer will send a trip signal to the relay board energizing the trip relay.

The transfer relay will select the opposite signal (all signals with the exception to Tc can transfer only from X to Y).

When the signal is returned to the normal value (the same as the Y signal or within the mismatch setpoint), the SASS channel can be returned to automatic by depressing the Auto push-button on the SASS module.

The following controls and indications are associated with the SASS system.

SASS module front panel controls and indications:

Reset switch Switch is located under the front cover and may be used to reset (initialize the computer and start the data gathering).

Auto push-button When depressed, the associated SASS channel will return to automatic if the NNIX and NNIY signals are 3.1.3 Controls and indications

1203.021 LOSS OF NEUTRON FLUX INDICATION PAGE:

3 of 12 CHANGE: 015 SECTION 1 LOSS OF ONE OR MORE POWER RANGE NI CHANNELS INSTRUCTIONS

1.

IF NI failure affects reactor power input to ICS, THEN PERFORM Heat Transfer Upset section of 1203.001 ICS Abnormal Operation.

2.

CHECK for normal voltage on failed detector as follows:

In associated RPS cabinet, CHECK Power Supply module reads approximately 600V (590 V to 610 V).

CHECK PDS/PMS Computer Points at 590 V to 620 V:

RPS-A NI-5 HV POWER SUPPLY (E0549)

RPS-B NI-6 HV POWER SUPPLY (E0550)

RPS-C NI-7 HV POWER SUPPLY (E0551)

RPS-D NI-8 HV POWER SUPPLY (E0552)

3.

IF ALL power range channels have failed, AND there is NO on-scale indication of neutron flux available, THEN:

A.

TRIP the Reactor.

B.

PERFORM 1202.001, Reactor Trip, in conjunction with this procedure.

4.

IF three or more power range channels have failed AND on-scale indication of neutron flux is available, THEN REFER TO the following:

TS 3.3.1 Conditions C and D for required actions TS 3.3.11 for failed NI power input to the respective EFIC channels NI-6 slow failing high would affect reactor power input to ICS so Section 1 Heat Transfer Upset would be correct.

1203.001 ICS ABNORMAL OPERATION PAGE:

3 of 26 CHANGE: 018 SECTION 1 - HEAT TRANSFER UPSET INSTRUCTIONS (REFERENCE USE)

1.

IF parameters can NOT be controlled within Reactor Trip setpoints, THEN GO TO 1202.001, Reactor Trip.

2.

PLACE BOTH MFW Pump H/A Stations in HAND:

MFW Pump A H/A MFW Pump B H/A

3.

IF Main Feedwater Block Valve(s) closed, THEN PLACE associated Startup and Low Load Control Valves in HAND.

4.

PLACE Diamond Panel in MANUAL.

5.

CHECK Controlling Steam Pressure(s) controlled between 870 and 930 psig.

6.

IF Controlling Steam Pressure(s) is NOT controlled between 870 and 930 psig, THEN:

PLACE Turbine in TURB MAN control.

IF TURB BYP Valves affected, THEN PLACE affected TURB BYP Valves H/A station(s) in HAND.

NOTE Due to the dead band associated with Low Load Valves, further than normal operation of these valves could be required to obtain a valve response. As much as 10% demand can be required for valve movement. [CAPR CR-ANO-1-2015-4178]

7.

IF Low Load Control Valve(s) is/are adjusted and NOT responsive, THEN MODULATE Low Load Block if needed to control feedwater flow.

8.

ADJUST components in HAND and INSERT Control Rods as necessary to achieve the following:

Reactor Power less than 100%

T-ave stable Delta-T-cold less than +/- 5 F SG pressures stable

9.

IF Startup or Low Load Control Valve(s) placed in HAND, THEN ENTER applicable condition per TS 3.7.3.

Placing the Diamond Panel in Manual will ultimately stop the power rise.

1203.012F ANNUNCIATOR K07 CORRECTIVE ACTION PAGE:

26 of 57 CHANGE: 033 K07-B4 Page 3 of 4 C.

IF desired by CRS/SM, THEN RESTORE pressure controls:

1) ENSURE ERV is closed.

2) OPEN CV-1000.

3) ENSURE Pressurizer Spray (CV-1008) is closed.

4) PLACE Pressurizer Spray Control Mode in AUTO.

5) ENSURE heater controls in AUTO:

PZR Heater Bank 1 PZR Heater Bank 4 PZR Heater Bank 2 PZR Heater Bank 3 PZR Heater Bank 5 6.

IF a transfer has occurred, THEN PLACE the transferred instrument selector switch in the position of the automatically selected input signal.

7.

IF transfer has NOT occurred AND desired by CRS/SM, THEN to select X or Y signal:

A.

CHECK signal to be selected is a usable signal.

B.

IF ICS controlling input is impacted, THEN ENSURE affected stations are in manual (Ref. 1203.001, ICS Abnormal Operation).

C.

IF FW Temperature or FW Flow input is impacted, THEN ENSURE selection for BOTH FW Temperature and FW Flow is maintained on same NNI-X or NNI-Y.

D.

MAKE selection.

E.

CHECK that new signal source is providing proper control signal.

8.

For failed input signals or persistent mismatch indications, INITIATE corrective action for repair.

The SASS Mismatch ACA contains the guidance to select the good signal. This would stop the power change however step B states if ICS controlling input is impacted, then place affected stations in manual. This would be required to be performed prior to selecting the Y signal.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1330 Source: Bank Rev: 0 Rev Date: 9/07/21 Originator: K. Smith TUOI:

A1LP-RO-RBVEN Objective:

2 System Number: 029 System

Title:

Containment Purge System (CPS)

Section: 3.8 Type:

Plant Service Systems

==

Description:==

Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

K/A Number: 2.2.44 CFR

Reference:

41.5 / 43.5 Point Value: 1 RO Imp:

SRO Imp:

4.4 Tier:

2 Group:

2 RO Select:

No SRO Select: Yes Question:

Given:

- Plant shutdown and cooldown in progress

- RCS Tave 185°F

- RB Purge in progress to lower RB atmospheric activity

- RB Purge projected release duration is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

- RB Purge release commenced two (2) hours ago

- Nuclear Chemistry stated gaseous releases projected to exceed quarterly objectives by 30%

- Release report preliminary release rate 2.16 E4 cfm

- Design flow rate 1.44 E4 cfm to 2.16 E4 cfm Which of the following would violate the RB Purge permit and require RB Purge termination?

A. Actual (stable) RB purge flow rate of 1.5 E4 cfm.

B. RB Atmosphere Gaseous Detector slowly trending upward.

C. Loss of Decay Heat Removal results in RCS temperature at 220°F.

D. Maintenance personnel defeat the Personnel Hatch door interlocks.

Answer:

C. Loss of Decay Heat Removal results in RCS temperature at 220°F.

Notes:

Answer "C" is Correct. RB Purge may only be performed at Cold Shutdown, RCS temp >200°F requires purge termination.

Answer "A" is Incorrect. This only requires notifying Nuclear Chemistry to adjust release data. Plausible since the actual flow rate is less than the preliminary release flow rate which would be non-conservative relative to the assumptions made when generating the release permit. Termination would only be required if not within the design flow rate Answer "B" is Incorrect. Only significant changes or values approaching an NUE criteria require termination.

Answer "D" is Incorrect. This affects containment closure criteria and does not impact the release permit.

General Discussion:

This question matches the KA since it requires interpreting RCS temperature indications to determine that plant conditions have changed such that purge is no longer allowed by Tech Specs to be operated in the current mode.

SRO Justification:

Difficulty:

3 Taxonomy: F RO:

SRO:

92

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. This question requires knowledge of administrative requirements of performing a reactor building purge which is can only be authorized by an SRO.

History:

Edited QID123 Updated KA to fit 2022 SRO Outline Updated Justification Used in 2022 SRO Outline

References:

OP-1104.033

1203.028 LOSS OF DECAY HEAT REMOVAL CHANGE 032 PAGE 45 of 90 ATTACHMENT 2 PAGE 1 OF 3 CONTROL ROOM ACTIONS FOR CONTAINMENT CLOSURE AND EVACUATION NOTE This attachment is typically performed by a non-Control Room watchstanders (e.g., Extra CRS, Third RO, Look-ahead SRO, Test Team, etc.).

The steps in this attachment can be performed in any order with sub-steps performed in order.

1.

ADVISE Shift Manager to implement 1903.010, Emergency Action Level Classification.

2.

MAKE the following announcement TWICE:

Attention all personnel. Attention all personnel. Due to a loss of Decay Heat Removal on Unit 1, responsible personnel set containment closure for Unit 1.

NOTE Postponing evacuation is prudent in situations such as refueling canal filled and Decay Heat Removal system expected to be returned to service without delay.

WARNING If RCS is open to atmosphere, RCS expansion due to heatup could result in spillage into containment.

Areas near RCS opening(s) could become hazardous to personnel during evacuation.

3.

IF ANY of the following conditions exist:

Time to boil is initially less than ONE hour OR becomes less than ONE hour due to RCS heatup.

Reactor Building conditions degrade (examples: RB temperature rise, RB humidity rise, RCS spillage, rise in airborne contaminants, etc.)

RCS temperature rises 5°F RCS pressure rises 5 psig Maintenance activities in Reactor Building could be affected by RCS level rise (expansion or during refill)

THEN:

A.

PERFORM a localized evacuation of Unit 1 Reactor Building per 1903.030, Evacuation.

B.

NOTIFY Radiation Protection and Reactor Building Coordinator to evacuate containment of all personnel except those involved in containment closure, avoiding ANY areas where RCS is open to atmosphere.

In 1203.028 LODHR AOP, purge would be required to be secured since RCS temperature rose from 185F to 220F.

1203.028 LOSS OF DECAY HEAT REMOVAL CHANGE 032 PAGE 46 of 90 ATTACHMENT 2 PAGE 2 OF 3

4.

Control room actions to set containment closure:

A.

For Reactor Building Purge/Ventilation:

1)

ENSURE the following fans are stopped:

RB Purge Supply Fan (VSF-2)

RB Purge Exhaust Fan (VEF-15)

2)

ENSURE at least ONE RB Purge Inlet is closed:

CV-7404 CV-7402

3)

ENSURE at least ONE RB Purge Outlet is closed:

CV-7403 CV-7401 B.

ENSURE ONE of the following RB sump drains is closed:

CV-4400 CV-4446 C.

CLOSE AVAILABLE containment isolation valves for idle systems on the following panels:

C16 C18 C26 D.

DIRECT 1015.002, Decay Heat Removal and LTOP System Control, Attachment K, Setting Containment Closure (USE CRS Admin and Outage Control Center when manned).

E.

IF Reactor Building atmosphere degrades due to rise in temperature or humidity, THEN MAXIMIZE RB cooling as follows:

1)

START ALL available RB Cooling Fans:

VSF-1A VSF-1B VSF-1C VSF-1D

Reactor Building Isolation Valves B 3.6.3 ANO-1 B 3.6.3-2 Amendment No. 215 Rev. 39,50 BACKGROUND (continued)

Failure of the purge valves to close following a design basis event would cause a significant increase in the radioactive release because of the large reactor building leakage path introduced by these 24-inch purge lines. Failure of the purge valves to close would result in leakage considerably in excess of the reactor building design leakage rate of 0.2% of reactor building air weight per day (La) (Ref. 1). The 24-inch purge valves are not tested for automatic closure from their open position under DBA conditions. Therefore, the 24-inch supply and exhaust purge valves are maintained closed with the handswitch keys removed (SR 3.6.3.1) in MODES 1, 2, 3, and 4 to ensure the reactor building boundary is maintained.

APPLICABLE SAFETY ANALYSES The reactor building isolation valve LCO was derived from the assumptions related to minimizing the loss of reactor coolant inventory and establishing the reactor building boundary during major accidents. As part of the reactor building boundary, the reactor building isolation valve OPERABILITY supports leak tightness of the reactor building. Therefore, the safety analysis of any event requiring isolation of the reactor building is applicable to this LCO.

The DBAs that result in a release of radioactive material within the reactor building are a loss of coolant accident (LOCA) and a main steam line break (Ref. 2). In the analysis for each of these accidents, it is assumed that the reactor building isolation valves are either closed or function to close. This ensures that potential paths to the environment through the reactor building isolation valves (including reactor building purge valves) are minimized. The safety analysis assumes that the 24 inch purge valves are closed at event initiation.

The LOCA analysis assumes a fixed amount of core inventory escapes. No mechanistic scenario is evaluated to determine what portion of the inventory is released prior to closure of the reactor building isolation valves. Industry standards for sizing valve operators govern the closure times of the reactor building isolation valves.

ANO-1 does not currently allow reactor building purging in MODES 1, 2, 3, and 4. Therefore, each of the reactor building purge valves is required to remain closed with its handswitch key removed during MODES 1, 2, 3, and 4. This prevents inadvertent actuation of the reactor building purge valves while in MODES 1, 2, 3, and 4. The purge system valve design prevents a single failure from compromising the reactor building boundary as long as the system is operated in accordance with the subject LCO.

In MODES 1 and 2, the reactor building isolation valves satisfy Criterion 3 of 10 CFR 50.36 (Ref. 3). In MODES 3 and 4, the reactor building isolation valves satisfy Criterion 4 of 10 CFR 50.36.

Purge not allowed in modes 1-4. In the stem, RCS temperature was 185F which is Mode 5 (<200F). If a loss of DHR was to occur and RCS temperature rose above 200F, then Mode 4 would be entered and purge would be required to be secured.

1104.033 Rev. 087 Page 61 of 120 REACTOR BUILDING VENTILATION Attachment B Page 8 of 12 Reactor Building Purge Gaseous Release Permit Section 5.0 Step 9 (Continued)

NOTE Maintaining the prescribed stack flow ensures isokinetic sampling conditions for noble gas, particulate, charcoal analysis streams remain functional. During non-isokinetic conditions, sampling accuracy is invalidated for SPING 1 and alternate sampling per ODCM L 2.2.1.

d.

IF VEF-15 flow is NOT within 1.44E4 to 2.16E4 cfm, THEN:

1)

PLACE HS-7421 for VSF-2 to STOP.................................................... _____

2)

PLACE HS-7422 to OFF....................................................................... _____

3)

CLOSE the following valves:

RB Purge Outlet (CV-7403)........................................................ _____

RB Purge Inlet (CV-7404)........................................................... _____

4)

TAKE action to have fan flow adjusted to within limits........................... _____

10.

RECORD the following:........................................................................................... _____

Release start date/time ______________ /__________________

VEF-15 flow ______________________ E4 cfm.

11.

NOTIFY Chemistry of Release start time................................................................. _____

12.

MAKE a Station Log entry....................................................................................... _____

NOTE TRM 3.6.1 is satisfied by the performance of 5120.416, In-Place Testing Of The Unit 1 Rx Building Purge Filtration System. SPING 1 Channel 10 flow rates are not accurate enough to confirm flow within 10% of design flow. SPING 1 Channel 10 flow is monitored to detect adverse trends in fan performance and to bound the release criteria by observing flow within 20% of design flow.

13.

IF SPING 1 Channel 10 is available AND notification is received from Chemistry that flow is out of specification per Form 1604.051A, Unit 1 SPING Monitor Log, THEN:

a.

CHECK actual (stable) SPING 1 Channel 10 flow rate 1.44E4 to 2.16E4 cfm..................................................................................................... _____

Required to terminate purge if flow rate is outside of the design flow rate. In this case, flow rate is within the design flow rates so termination is not required based on flow rate.

1104.033 Rev. 087 Page 62 of 120 REACTOR BUILDING VENTILATION Attachment B Page 9 of 12 Reactor Building Purge Gaseous Release Permit Section 5.0 Step 13 (Continued)

b.

IF SPING 1 Channel 10 flow is NOT within 1.44E4 to 2.16E4 cfm, THEN TERMINATE release as follows:

1)

NOTIFY SM.......................................................................................... _____

2)

PLACE RB Purge Supply Fan VSF-2 (HS-7421) to STOP.................... _____

3)

PLACE RB Purge Exhaust Fan VEF-15 (HS-7422) to OFF.................. _____

4)

CLOSE the following valves:

RB Purge Outlet (CV-7403)........................................................ _____

RB Purge Inlet (CV-7404)........................................................... _____

5)

MAKE a Station Log entry..................................................................... _____

6)

IF the release is a fractional-hour release, THEN NOTIFY Chemistry of problem to determine if a new preliminary report needs to be prepared................................................ _____

7)

TAKE action to have fan flow adjusted to within limits........................... _____

8)

WHEN fan flow is adjusted, THEN RESTART the purge.................................................................. _____

c.

RECORD the following:.................................................................................. _____

Preliminary report flow rate ____________ E4 cfm Actual (stable) SPING 1 Channel 10 flow rate after purge started

____________ E4 cfm

d.

DETERMINE whether the actual (stable) SPING 1 Channel 10 flow rate is less than the preliminary report flow rate........................................................ _____

e.

IF flow rate is less than the preliminary release rate AND the release is a fractional-hour release, THEN NOTIFY Chemistry of the actual (stable) flow rate so that the appropriate correction can be made to the hours fan run time to account for the difference between the predicted and actual flow rates..................... _____

14.

IF SPING 1 Channel 10 effluent flow indication is non-functional OR SPING 1 Channel 14 and 15 sample vacuum/flow indications are non-functional, THEN:

REFER TO ODCM L2.2.1.............................................................................. _____

ESTIMATE flow rate every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />................................................................ _____

Not required to terminate just because actual purge flow is low so long as it is within the design flow rate. In this case, chemistry will make a correction to the run time.

1104.033 Rev. 087 Page 63 of 120 REACTOR BUILDING VENTILATION Attachment B Page 10 of 12 Reactor Building Purge Gaseous Release Permit Section 5.0 (Continued)

CRITICAL STEP

15.

IF ANY of the following conditions occur during purge:

SPING 1 indicates stack activity approaches NUE criteria at site boundary over ANY one-hour period.

SPING 1 Channel 5 two-minute average value exceeds setpoint.

RB Atmos Particulate Monitor (RI-7460) or RB Atmos Gaseous Monitor (RI-7461) detects a significant rise in activity after purge initiation.

THEN:

a.

NOTIFY CRS/SM.......................................................................................... _____

b.

PLACE RB Purge Supply Fan VSF-2 handswitch (HS-7421) in STOP........... _____

c.

PLACE RB Purge Exhaust Fan VEF-15 handswitch (HS-7422) in OFF......... _____

d.

CLOSE the following dampers:

RB Purge Outlet (CV-7401).................................................................. _____

RB Purge Inlet (CV-7402).................................................................... _____

RB Purge Outlet (CV-7403).................................................................. _____

RB Purge Inlet (CV-7404).................................................................... _____

e.

RECORD release stop time ______________ Date _____________............ _____

f.

ENSURE Station Log entry is made............................................................... _____

g.

NOTIFY Chemistry that purge was terminated............................................... _____

16.

WHEN RB activity reduced to background (expect about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of purge)

OR purge is otherwise deemed complete, THEN COORDINATE with Chemistry for the following:

a.

PLACE RB Purge Supply Fan VSF-2 handswitch (HS-7421) in STOP........... _____

b.

PLACE RB Purge Exhaust Fan VEF-15 handswitch (HS-7422) in OFF......... _____

Not required to terminate purge for slowly trending upward rad levels.

Must meet these criteria.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1389 Source: New Rev: 0 Rev Date: 12/13/21 Originator: K. Smith TUOI:

Objective:

System Number: 075 System

Title:

Circulating Water System Section: 3.8 Type:

Plant Service Systems

==

Description:==

Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

K/A Number: 2.4.4 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

4.7 Tier:

2 Group:

2 RO Select:

No SRO Select: Yes Question:

Given:

- U1 is at 95% power

- Sustained high temperatures have caused Lake Dardanelle temperature to rise

- Condenser Vacuum is ranging between 26.6" Hg and 27" Hg from day to night

- Condenser Circ Water Outlet Temperatures are rising as follows

- Day 1 at 1500 - 102°F

- Day 2 at 1500 - 105°F

- Day 3 at 1500 - 111°F (1) In what section of OP-1203.008, Exceeding Thermal Limits on Condenser Discharge Water should the CRS direct actions?

(2) What procedure should be transitioned to from the initial procedure?

A. (1) Section 1 - Instantaneous Maximum.

(2) OP-1203.016, Loss of Condenser Vacuum.

B. (1) Section 1 - Instantaneous Maximum.

(2) OP-1102.016, Power Reduction and Plant Shutdown.

C. (1) Section 2 - Maximum Daily Average.

(2) OP-1203.016, Loss of Condenser Vacuum.

D. (1) Section 2 - Maximum Daily Average.

(2) OP-1102.016, Power Reduction and Plant Shutdown.

Answer:

B. (1) Section 1 - Instantaneous Maximum.

(2) OP-1102.016, Power Reduction and Plant Shutdown.

Notes:

Answer "B" is Correct. Section 1 for instantaneous maximum entry requirements is greater than 110°F. Step 4 states to perform a power reduction and plant shutdown Answer "A" is Incorrect. Plausible because vacuum is lower than normal however vacuum is not currently degrading therefore entry to 1203.016 is not required.

Answer "C" is Incorrect. This would be required if circ water outlet temperatures were above 105°F for 48 hrs.

Plausible because vacuum is lower than normal however vacuum is not currently degrading therefore entry to 1203.016 is not required.

Answer "D" is Incorrect. This would be required if circ water outlet temperatures were above 105°F for 48 hrs.

Difficulty:

3 Taxonomy: H RO:

SRO:

93

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 General Discussion:

This question matches the K/A since it requires the ability to recognize conditions that require specific procedures to be entered.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

History:

Used in 2022 SRO Exam.

References:

OP-1203.008

1203.008 EXCEEDING THERMAL LIMITS ON CONDENSER DISCHARGE WATER PAGE:

2 of 6 CHANGE: 012 SECTION 1INSTANTANEOUS MAXIMUM ENTRY CONDITIONS One of more of the following:

ANY of the following PMS points exceed 110°F for greater than 1 hour:

- Cond E-11A CW Outlet Temp (T3648)

- Cond E-11B CW Outlet Temp (T3655)

- Average Circ Water Outlet Temp (T1651)

EITHER of the following points on Cond Waterbox Inlet & Disch Temp Recorder (TR-3640) exceed 110°F for greater than 1 hour:

- PT. 5 E-11B Combine Disch TE-3654

- PT. 6 E-11A Combine Disch TE-3649

1203.008 EXCEEDING THERMAL LIMITS ON CONDENSER DISCHARGE WATER PAGE:

3 of 6 CHANGE: 012 SECTION 1 -- INSTANTANEOUS MAXIMUM INSTRUCTIONS

1.

NOTIFY the following:

Chemistry Manager Senior Manager, Operations

2.

RECORD the time and value of both (E-11A and E-11B) CW outlet temperatures in the Unit 1 Station Log.

3.

INITIATE a Condition Report documenting exceeding 110°F for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

NOTE A violation of the NPDES permit is not necessarily created by instantaneous average Circ Water outlet temp exceeding 110°F.

If the Daily Average of average Circ Water outlet temp exceeds 110°F, it is a violation of the NPDES permit. Part I, Section A of the permit lists compliance requirements. Chemistry can provide clarification of the permit.

4.

Prior to the Daily Average of average Circ Water outlet temp reaching 110°F, PERFORM the following:

A.

CONVENE a Bridge call with site management to discuss options for reducing Circ Water outlet temperature.

B.

IF management deems it necessary to reduce plant load, THEN PERFORM ONE of the following to maintain the Daily Average of average Circ Water outlet temp below 110°F:

1102.004, Power Operations 1102.016, Power Reduction and Plant Shutdown 1203.045, Rapid Plant Shutdown END

1203.008 EXCEEDING THERMAL LIMITS ON CONDENSER DISCHARGE WATER PAGE:

4 of 6 CHANGE: 012 SECTION 2 -- MAXIMUM DAILY AVERAGE ENTRY CONDITIONS One of more of the following:

ANY of the following PMS points exceed 105°F for greater than 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />s:

- Cond E-11A CW Outlet Temp (T3648)

- Cond E-11B CW Outlet Temp (T3655)

- Average Circ Water Outlet Temp (T1651)

EITHER of the following points on Cond Waterbox Inlet & Disch Temp Recorder (TR-3640) exceed 105°F for greater than 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />s:

- PT. 5 E-11B Combine Disch TE-3654

- PT. 6 E-11A Combine Disch TE-3649

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1372 Source: Bank Rev: 0 Rev Date: 11/3/21 Originator: K. Smith TUOI:

ASLP-RO-EPLAN Objective:

4 System Number: 2.1 System

Title:

Conduct of Operations Section: 2 Type:

Generic Knowledge And Abilities

==

Description:==

Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc.

K/A Number: 2.1.14 CFR

Reference:

41.10 / 43.5 Point Value: 1 RO Imp:

SRO Imp:

3.1 Tier:

3 Group:

G RO Select:

No SRO Select: Yes Question:

In accordance with OP-1903.011 Emergency Response/Notifications, a plant evacuation and personnel accountability are REQUIRED for a_____(1)_____.

Personnel are notified of the initiation of a plant evacuation and accountability by_____(2)_____.

A. (1) General Emergency classification only (2) Everbridge B. (1) General Emergency classification only (2) dialing 198 and making a plant announcement C. (1) Site Area Emergency classification or higher (2) Everbridge D. (1) Site Area Emergency classification or higher (2) dialing 198 and making a plant announcement Answer:

D. (1) Site Area Emergency classification or higher (2) dialing 198 and making a plant announcement Notes:

Answer "D" is Correct. Per 1903.011, a plant evacuation and personnel accountability are initiated at a Site Area Emergency classification using the plant paging system.

Answer "A" is Incorrect. General Emergency is plausible since this is when Protective Action Recommendations are required to be performed for the public. Everbridge is plausible as this is the method used to activate the ERO during the initial emergency class declaration. Everbridge is also used to notify Entergy management and the NRC resident of plant status changes during non-emergency events.

Answer "B" is Incorrect. General Emergency is plausible since this is when Protective Action Recommendations are required to be performed for the public.

Answer "C" is Incorrect. Everbridge is plausible as this is the method used to activate the ERO during the initial emergency class declaration. Everbridge is also used to notify Entergy management and the NRC resident of plant status changes during non-emergency events.

General Discussion:

This question matches the K/A since it requires knowledge of when and how a plant announcement is made during implementation of the emergency plan.

SRO Justification:

10 CFR 55.43(b)(1) Conditions and limitations in the facility license.

Difficulty:

3 Taxonomy: F RO:

SRO:

94

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 History:

Used Question from Diablo Canyon 2019 SRO Exam Used in 2022 SRO Exam

References:

OP-1903.030 OP-1903.011

1903.011 Rev. 062 Page 28 of 84 EMERGENCY RESPONSE/NOTIFICATIONS Form 1903.011-P Page 3 of 5 SAE Emergency Direction and Control Checklist - Shift Manager Step 8 (Continued) d.

INSTRUCT Emergency Response Personnel and operators from BOTH units in the field to do the following:

1)

LOG into the nearest accountability card reader.

2)

LOG onto the emergency RWP as time allows.

9.

IF performing a Plant Evacuation, THEN:

a.

DETERMINE if a radiological or toxic gas release exist or is a release suspected which is originating from the plant.

o o

Yes No b.

IF YES, THEN DETERMINE the available routes from the chart below using wind direction:

Wind Direction Evacuation Route 0 to 60 degrees

¨ USE Route 1 and 2 61 to 90 degrees

¨ USE Route 2 91 to 260 degrees

¨ USE Route 1 261 to 360 degrees

¨ USE Route 1 and 2 c.

IF NO, THEN USE Route 1, Route 2 or BOTH Routes 1 and 2.

d.

IF there are known physical impediments blocking a site evacuation

route, THEN ADJUST the evacuation route as necessary.

e.

CHECK the appropriate routes in the plant announcement below.

f.

CONTACT Security (ext. 3388, 3108 or 3109).

A site area emergency is the lowest classification where the action to perform a plant evacuation is required.

1903.011 Rev. 062 Page 29 of 84 EMERGENCY RESPONSE/NOTIFICATIONS Form 1903.011-P Page 4 of 5 SAE Emergency Direction and Control Checklist - Shift Manager Step 9 (continued) g.

Interfacing with Security:

• REQUEST Security to perform initial accountability before

___________ hours (30 minutes from SAE declaration or after Hostile Action Event has ended).

• INFORM Security of Evacuation Routes.

h.

CONTACT Radiation Protection (CA1 - 5166 or OSC - 6614).

i.

Interfacing with Radiation Protection:

• REQUEST RP coverage at the plant exit portal monitors.

• INSTRUCT RP personnel at the controlled access exit point to relax decontamination and radiation protection measures as necessary in order to expedite evacuation of the controlled access area.

j.

IF there are ANY plant areas to avoid OR special protective actions are to be taken by plant personnel due to a release in progress or other hazards, THEN RECORD areas/actions here and INCLUDE in announcement:

k.

DIAL 198 and MAKE the following plant announcement:

"Attention all personnel. Attention all personnel. This is (state name and title). A Site Area Emergency has been declared on (state unit one, unit two, or both units) based upon (state EAL condition). Emergency response personnel report to your designated assembly areas and perform initial accountability. All other personnel evacuate the plant using evacuation route(s) 1 2 and proceed to the Atkins Emergency Worker Center.

l.

SOUND the evacuation alarm for approximately 10 seconds.

m.

REPEAT the announcement and the subsequent plant evacuation alarm.

n.

INSTRUCT Emergency Response Personnel and operators from BOTH units in the field to do the following:

1)

LOG into the nearest accountability card reader.

2)

LOG onto the emergency RWP as time allows.

Action to perform accountability and plant evacuation is to make a plant announcement.

1903.011 Rev. 062 Page 65 of 84 EMERGENCY RESPONSE/NOTIFICATIONS Page 2 of 7 Emergency Response Organization Notification System 2.0 Non-Emergency/Off-Normal Notification Using Everbridge NOTE The following steps will notify the following positions/personnel for Non-Emergency / Off-Normal events:

Vice President, Operations General Manager, Plant Operations Operations Manager Manager, Emergency Planning Emergency Planners Emergency Directors Emergency Plant Managers JIC Managers Company Spokesperson NRC Resident Inspector 1.

LOG onto Everbridge Mass Notification system by accessing the users site from Internet Explorer: https://manager.everbridge.net/login and ENTER the following:

User name: ANOSTA1 (case sensitive)

Password: 1ano#sta 2.

CLICK New Notification button.

Everbridge plausible

1903.011 Rev. 062 Page 15 of 84 EMERGENCY RESPONSE/NOTIFICATIONS 5.0 RESPONSIBILITIES 1.

SHIFT MANAGER (SM / ED)

The SM will initially have Emergency Direction and Control responsibility for implementation of response actions described in this procedure until relieved by the Emergency Director (ED) following Emergency Operations Facility activation.

2.

EMERGENCY DIRECTOR (ED)

Upon assumption of responsibility for Emergency Direction and Control the Emergency Director is responsible for implementation of the response actions described in this procedure.

3.

CONTROL ROOM COMMUNICATORS Control Room Communicators are responsible for performing emergency response notifications/communications.

4.

EMERGENCY RESPONSE ORGANIZATION (ERO)

Members of the ERO are responsible to ensure completion of notifications as denoted on Attachment 9, Emergency Response Organization Notification System Section 4.0if the ERO cannot be activated by the Emergency Response Organization Notification System (Everbridge).

Everbridge plausible

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1250 Source: Repeat Rev: 0 Rev Date: 08/27/19 Originator: K. Smith TUOI:

A1LP-RO-FH Objective:

4 System Number: 2.1 System

Title:

Conduct of Operations Section: 2.0 Type:

Generic K/As

==

Description:==

Knowledge of procedures and limitations involved in core alterations.

K/A Number: 2.1.36 CFR

Reference:

41.10 / 43.6 / 45.7 Point Value: 1 RO Imp:

3.0 SRO Imp:

4.1 Tier:

3 Group:

G RO Select:

No SRO Select: Yes Question:

Which one of the following completes the statement below?

Assuming all other prerequisites are met for refueling IAW 1502.004 Control of Unit 1 refueling, (1) how long must the reactor be subcritical before you can start the refueling process, and (2) who grants permission?

A. (1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

(2) the Shift Manager.

B. (1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

(2) both the Reactor Engineer and SRO in Charge of Fuel Handling.

C. (1) 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

(2) the Shift Manager.

D. (1) 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

(2) both the Reactor Engineer and SRO in Charge of Fuel Handling.

Answer:

D. (1) 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

(2) both the Reactor Engineer and SRO in Charge of Fuel Handling.

Notes:

Answer "D" is Correct. (1) Correct. TRM 3.9.3 requires the reactor to be subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to handling irradiated fuel from the reactor. (2) Correct. 1502.004 Control of Unit 1 Refueling states permission to commence fuel movement must come from both the Reactor Engineer and SRO in Charge of Refueling.

Answer "A" is Incorrect. (1) Incorrect. TRM 3.9.3 requires the reactor to be subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to handling irradiated fuel from the reactor. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> represents 3 days which could seem plausible to an applicant who is not familiar with the TRM requirements. (2) Incorrect. Plausible because the Shift Manager grants permission to perform many activities, however in accordance with 1502.004 Control of Unit 1 Refueling, the SM does not grant permission. Permission must come from both the Reactor Engineer and SRO in Charge of Refueling.

Answer "B" is Incorrect. (1) Incorrect. TRM 3.9.3 requires the reactor to be subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to handling irradiated fuel from the reactor. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> represents 3 days which could seem plausible to an applicant who is not familiar with the TRM requirements. (2) Correct. 1502.004 Control of Unit 1 Refueling states permission to commence fuel movement must come from both the Reactor Engineer and SRO in Charge of Refueling.

Answer "C" is Incorrect. (1) Correct. TRM 3.9.3 requires the reactor to be subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to handling irradiated fuel from the reactor. (2) Incorrect. Plausible because the Shift Manager grants permission to perform many activities, however in accordance with 1502.004 Control of Unit 1 Refueling, the SM does not Difficulty:

2 Taxonomy: F RO:

SRO:

95

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 grant permission. Permission must come from both the Reactor Engineer and SRO in Charge of Refueling.

General Discussion:

This question matches the K/A since the applicant must have the knowledge of the information contained within 1502.004 Control of Unit 1 Refueling to apply the limitations involved in core alterations.

SRO Justification:

10 CFR 55.43(b)(7) Fuel handling facilities and procedures. This question requires knowledge of the responsibilities of the SRO in Charge of Fuel Handling.

History:

New for 2020 SRO Exam.

Used in 2022 SRO Exam.

Updated formatting of distractor analysis.

References:

1502.004 Control of Unit 1 Refueling

1502.004 Rev. 066 Page 26 of 75 CONTROL OF UNIT 1 REFUELING (Section 5.1, Final Preparations For Fuel Handling continued)

10.

Control Room Communicator:

a.

OBTAIN permission from BOTH of the following to commence refueling:

SRO in Charge of Fuel Handling Reactor Engineer

b.

LOG permission obtained from BOTH individuals in Attachment C, Chronological Log.

5.2 Continuous Actions During Fuel Handling

1.

IF Core Alterations are suspended, THEN NOTIFY Radiation Protection (CR-ANO-2-2003-1405).

2.

IF Core Alterations are suspended, THEN PRIOR to commencing Core Alterations:

a.

ENSURE steps as directed by SRO in Charge of Fuel Handling of the following are performed:

Section 4.0, Prerequisites Section 5.1, Final Preparations For Fuel Handling

b.

DOCUMENT the following in Attachment C, Chronological Log:

Actions taken to allow restart of Core Alterations SRO in Charge of Fuel Handling permission to continue

c.

NOTIFY Radiation Protection of Core Alteration commencement.

3.

Concurrently with refueling operations, PERFORM the following attachments:

Attachment A, Dilution Prevention Valve Check Attachment B, Refueling Boron, Temperature, Tritium, and Level Check Attachment D, Refueling Housekeeping and Access Control (Informational Use) 1015.002, Decay Heat Removal And LTOP System Control, Attachment I, Refueling Penetration Control section Permission from both are required

Irradiated Fuel Handling - Reactor Building 3.9.3 ANO-1 TRM 3.9.3-1 Rev. 5,39 TRM 3.9 REFUELING OPERATIONS TRM 3.9.3 Irradiated Fuel Handling - Reactor Building TRO 3.9.3 Handling of irradiated fuel shall be subject to the following:

a.

Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, and

b.

A minimum of 10 feet separation shall be maintained between fuel assemblies when two assemblies are moved simultaneously in the fuel transfer canal.

APPLICABILITY:

During movement of irradiated fuel assemblies in the reactor building.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. TRO not met.

A.1 Suspend fuel movement in the reactor building.

Immediately TEST REQUIREMENTS TEST FREQUENCY None.

Must be subcritical for 100 hrs.

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1326 Source: New Rev: 0 Rev Date: 9/1/21 Originator: K. Smith TUOI:

ASLP-SRO-MNTC Objective:

1 System Number: 2.2 System

Title:

Equipment Control Section: 2 Type:

Generic Knowledge And Abilities

==

Description:==

Knowledge of the process for managing troubleshooting activities.

K/A Number: 2.2.20 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

3.8 Tier:

3 Group:

G RO Select:

No SRO Select: Yes Question:

A troubleshooting plan is being developed in accordance with EN-MA-125 Troubleshooting Control of Maintenance Activities.

Per EN-MA-125, what is the Shift Managers responsibility for this activity?

A. Acts as the single point of contact for status and communications.

B. Determines if a Failure Modes Analysis (FMA) is required for troubleshooting.

C. Determines the Troubleshooting Determination Level based on risk significance.

D. Ensures that the Troubleshooting Plan is followed and documentation is accurate and thorough.

Answer:

C. Determines the Troubleshooting Determination Level based on risk significance.

Notes:

Answer "C" is Correct. Per EN-MA-125 the SM is responsible for determining the troubleshooting level in conjunction with the Responsible Maintenance Supervisor/Specialist.

Answer "A" is Incorrect. This is the responsibility of the Troubleshooting Team Leader. Plausible since the SM leads the site and has overall authority over operations of the plant. Since troubleshooting activities require a multi disciplined team, it's plausible for an applicant to believe the SM will be the single point of contact.

Answer "B" is Incorrect. This is the responsibility of Engineering. Plausible since the distractor states "determines" not "performs".

Answer "D" is Incorrect. This is the responsibility of the Responsible Maintenance Supervisor. Plausible since the SM is usually in an oversight role with operations of the plant. This distractor aligns with what an individual in an oversight role would perform, however it is not performed by the SM.

General Discussion:

This question matches the KA since it requires knowledge of the process for managing troubleshooting activities by knowing the responsibilities of those involved while performing troubleshooting activities.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. This question specifically asks what the responsibilities of an SRO are during performance of troubleshooting activities.

Difficulty:

3 Taxonomy: F

References:

EN-MA-125 RO:

SRO:

96

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 History:

Used in 2022 SRO Exam

EN-MA-125 Rev. 26 Page 11 of 58 Troubleshooting Control of Maintenance Activities

21.

Structured Teardown Process: A guide developed to preserve and document physical evidence and testing activities during disassembly of equipment.

Reference EN-FAP-DC-015, the need for a structured teardown is determined after the initial troubleshooting is complete and the failed component has been removed from service. It should be used for the following reasons:

Repeat failures of the same component occur in the same plant operating cycle. (i.e. a pump mechanical seal is replaced then fails again in the same operating cycle.)

Repeat failures of a component occur after troubleshooting has determined a confirmed cause, and a subsequent failure occurs with the same confirmed cause, after all corrective actions for the previous failure are complete.

As directed by Station Leadership.

22.

Removed From Service: Equipment that is isolated either by tagout (per EN-OP-102), LCO action statement or by installing or using devices to bypass functions associated with the equipment. (i.e. bypass APRM or jumpering trips associated with radial walls.)

23.

Temporary Plant Configuration Change: The installation, removal, or modification of equipment or components which results in a departure from plant configured design. Such changes are intended to be minor in nature and of short duration to allow operation or maintenance troubleshooting. This process is controlled through use of EN-DC-136.

24.

Troubleshooting: Troubleshooting is a systematic method of isolating faults and analyzing abnormal conditions that occur in operating equipment and components. If the failed component is known, and replacement and retest is the only activity, then the activity should not be considered troubleshooting and is not governed by this procedure.

25.

Troubleshooting Boundary: An area which the troubleshooting activities are intended to be conducted and which has been addressed to prevent unknown or unwanted plant, system or component transients.

26.

Troubleshooting Control Form (TCF): An Attachment to this procedure that the originator is required to complete. The form is then processed for use in troubleshooting activities.

4.0 RESPONSIBILITIES

1.

Responsible Maintenance Supervisor / Specialist

a.

Ensures that personnel performing the troubleshooting have reviewed and understand the activities to be performed and the expected response.

b.

Ensures that the personnel performing the troubleshooting are qualified to perform the activity.

EN-MA-125 Rev. 26 Page 12 of 58 Troubleshooting Control of Maintenance Activities

c.

In conjunction with the Shift Manager, determines the Troubleshooting Determination Level in accordance with Section 7.3 being mindful that equipment problems involving single point vulnerable and risk sensitive structures, systems, and components and complex problems typically require focused controls to manage the risk and ensure timely resolution.

d.

Assists in obtaining information relative to observations of equipment behavior, equipment status and performance data immediately before, during and after equipment failure when this information is readily available.

e.

Acts as the Field Lead on the Formal Troubleshooting Team when established in accordance with Section 7.4.

f.

Works to ensure that the Troubleshooting Plan is followed and documentation is accurate and thorough.

g.

Ensures cyber security controls are in place and are utilized during performance of troubleshooting activities.

1)

Ensures personnel making a digital connection to a CDA are members of the critical group per EN-NS-101 Unescorted Access Authorization Program.

2)

Determines Critical Digital Asset (CDA) and the component Defensive In Depth (DID) level.

2.

Craft

a.

Assists in obtaining information relative to observations of equipment behavior, equipment status and performance data immediately before, during and after equipment failure when this information is readily available.

b.

Works with the development of the troubleshooting plan and field execution.

c.

Ensures that all necessary drawings, materials, and test equipment that will be used are available.

d.

Ensures nonconforming parts removed during troubleshooting and repair are tagged and handled in accordance with EN-MA-101 and EN-MA-101-02.

e.

Stops when unexpected conditions occur.

NOTE EN-IT-103-01 provides the steps and actions needed when using Media and connecting to a CDA.

f.

Ensures cyber security controls have been incorporated into troubleshooting plan/work order per EN-IT-103 (CYBER SECURITY PROGRAM) and EN-IT-103-01 (CONTROL OF PORTABLE DIGITAL MEDIA CONNECTED TO CRITICAL DIGITAL ASSETS).

Incorrect.

EN-MA-125 Rev. 26 Page 14 of 58 Troubleshooting Control of Maintenance Activities

3.

Planning

a.

Assists in developing the troubleshooting plan and incorporating the appropriate guidance as required.

b.

Assists in developing Post Maintenance Testing (PMT) according to EN-WM-107 after the completion of Troubleshooting/Maintenance as required.

c.

Ensures cyber security controls are incorporated into the work package for troubleshooting CDA components per EN-IT-103-01.

4.

Engineering

a.

Assists in developing troubleshooting plan and troubleshooting decisions.

b.

Assists in determining if component is a CDA.

c.

Engaged and notified of all levels of troubleshooting. Determines if a Failure Modes Analysis (FMA) is required for troubleshooting.

d.

Performs FMA when required.

e.

Acts as the Analysis Lead on the Formal Troubleshooting Team when established in accordance with Section 7.4.

5.

Shift Manager

a.

Provides oversight of operations personnel performing initial troubleshooting for system performance issues and documenting equipment problems, the initial troubleshooting performed, and working with maintenance to address the causes and resolutions of equipment issues. Operators are the first line of defense in recognizing degraded trends and proactively engaging the organization in timely resolution.

b.

Assists in performing Initial Investigation per Section 1.1 and in obtaining information relative to observations of equipment behavior, equipment status and performance data immediately before, during and after equipment failure when this information is readily available.

c.

In conjunction with Responsible Maintenance Supervisor/Specialist, determines the troubleshooting level in accordance with Section 7.3 being mindful that equipment problems involving single point vulnerable and risk sensitive structures, systems, and components and complex problems typically require formal controls to manage the risk and ensure timely resolution.

d.

Evaluates the need for, or revision to, EN-OP-104 Operability Risk and Functionality Assessments as new information becomes available through troubleshooting corrective maintenance, etc.

Correct.

Incorrect.

EN-MA-125 Rev. 26 Page 15 of 58 Troubleshooting Control of Maintenance Activities

e.

Evaluates the need for the use of the EN-FAP-OM-012, Prompt Investigation, Notifications and Duty Manager Responsibilities for significant troubleshooting activities that may lead to further degraded plant status.

f.

Concurs with exiting the troubleshooting when a direct cause is not found in accordance with Section 7.7.

6.

Troubleshooting Team Leader

a.

Appoints appropriate troubleshooting team members (i.e. Engineering, Operation, Maintenance, Planning) and work location(s).

b.

For formal troubleshooting, ensures the Analysis Team provides technical support for the development of the troubleshooting plan using problem analysis techniques.

c.

Ensures the Field Team performs troubleshooting actions safely, efficiently, and in accordance with the requirements of this procedure.

d.

For formal troubleshooting, ensures required actions are incorporated into the troubleshooting plan from the cause analysis and the troubleshooting field results are incorporated into the problem analysis techniques.

e.

Acts as the single point of contact for status and communications.

f.

Ensures cyber security controls are in place and are utilized during performance of troubleshooting activities.

1)

Ensures personnel making a digital connection to a CDA are members of the critical group per EN-NS-101 Unescorted Access Authorization Program.

7.

Personnel Assigned To Determine Apparent Or Root Cause

a.

Are involved in the development and implementation of the troubleshooting process.

b.

Determines if the components they will be troubleshooting are Critical Digital Assets (CDAs).

5.0 PRECAUTIONS AND LIMITATIONS

1.

Shift Manager is required to be informed of any changes to operational status of equipment/component being worked on by this procedure.

a.

Operations should contact Engineering for direction when the current condition requires repeated breakers resets or replacement of blown fuses. One attempt to close a breaker after a trip and one fuse replacement is allowed without entering into troubleshooting provided there is no indication of abnormality. Shift Manager has discretion to take this action. [Industry Benchmark Performed CR-HQN-2019-2353]

Incorrect.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1323 Source: New Rev: 0 Rev Date: 8/30/21 Originator: K. Smith TUOI:

ASLP-RO-COPD Objective:

4 System Number: 2.2 System

Title:

Equipment Control Section: 2 Type:

Generic Knowledge And Abilities

==

Description:==

Knowledge of the process used to track inoperable alarms.

K/A Number: 2.2.43 CFR

Reference:

41.10 / 43.3 Point Value: 1 RO Imp:

SRO Imp:

3.3 Tier:

3 Group:

G RO Select:

No SRO Select: Yes Question:

Given:

- A nuisance alarm on panel K06 has been spuriously alarming for a shift.

- The SM authorizes disabling the alarm input in accordance with OP-3305.002, Ops Equipment Removal From Service, Testing and Return to Service.

- There is no other affect to plant equipment as a result of this inoperable alarm.

In accordance with the Standardized Operations Metrics, this inoperable alarm will be tracked as a(n)_____(1)_____.

_____(2)_____is the MAXIMUM time the alarm may be out of service prior to performing a PAD review.

A. (1) Operator Burden (2) 60 days B. (1) Operator Burden (2) 90 days C. (1) Main Control Room Deficiency (2) 60 days D. (1) Main Control Room Deficiency (2) 90 days Answer:

D. (1) Main Control Room Deficiency (2) 90 days Notes:

Answer "D" is Correct. Problems with annunicators and alarms are included as Main Control Room Deficiencies. A PAD must be completed prior to the annunciator exceeding 90 days out of service time.

Answer "A" is Incorrect. Plausible because the alarm being a nuisance and taking action to remove it from service could appear to meet the definition as a burden, however an Operator Burden would be a deficiency that would require operators to take some significant compensatory action that could potentially place the plant in a transient. 60 days is plausible since at 60 days a CR is written to perform a PAD if one has not been performed yet.

Answer "B" is Incorrect. Plausible because the alarm being a nuisance and taking action to remove it from service could appear to meet the definition as a burden, however an Operator Burden would be a deficiency that would require operators to take some significant compensatory action that could potentially place the plant in a transient.

Difficulty:

3 Taxonomy: F RO:

SRO:

97

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 Answer "C" is Incorrect. 60 days is plausible since at 60 days a CR is written to perform a PAD if one has not been performed yet.

General Discussion:

This question meets the KA since it requires knowledge of the process used to track inoperable alarms such as the administrative classification assigned to track the inoperable alarm, and how long the alarm can be out of service prior to requiring a PAD.

SRO Justification:

10 CFR 55.43(b)(3) Facility licensee procedures required to obtain authority for design and operating changes in the facility. The question focuses on aspects of the disabled annunciator administrative process that would be performed by an SRO.

History:

Used in 2022 SRO Exam

References:

OP-3305.002 Standardized Operations Metrics INPO 19-002

Page 17 of 53 4

Main Control Room Deficiencies Definition:

The number of equipment deficiencies (on the last day of the month) that degrade the performance of an indication, switch, or controller in the control room. Problems with annunciators and alarms would be included, as would detectors or sensing elements that are remotely located but feed equipment in the control room.

A pump that would not start because of a problem with its motor (not its controller) would not be included.

Computer alarms would not be included.

Control room indications of switchyard components would be included.

Deficiencies affecting indications, etc., on back panels in the control room would be included.

Deficiencies that affect more than one indication, etc., are only counted as one deficiency.

Source of Definition: Data Element Manual 11 INPO 04-004 Revision 14 Calculation: Simple count of number on last day of month. This is a total number and includes both outage and on-line deficiencies. This value is for each unit. If a deficiency affects more than one unit, each unit should count the deficiency.

Source of Calculation: Data Element Manual 11 INPO 04-004 Revision 14

Page 19 of 53 6

Operator Burdens Definition:

includes; A deficiency that could place an unreasonable burden on the operators that could result in an initiation of an abnormal or emergency plant condition A deficiency that requires significant compensatory actions that could detract from the A deficiency that requires significant compensatory actions that potentially challenges operator human performance where an error during these actions could result in placing the plant in a transient Significant Compensatory Action:

compensat An action that takes more than five minutes to perform An action that is performed more than once per shift An action whose performance represents a personnel safety challenge (e.g., involves climbing, working in close proximity to hot piping, etc.)

Source of Definition: PWROG Calculation: The number of burdens the last day of each month. This value is provided for each unit and both outage and online burdens are counted. If the deficiency may affect actions on both units, it should be counted for each unit. The value provided is the number of deficiencies that existed on the last day of the month. This is a per unit indicator.

Source of Calculation: PWR Owners Group document PA-OSC-0384, Standard Operations Key Performance Indicators.

3305.002 Rev. 003 Page 41 of 45 OPS EQUIPMENT REMOVAL FROM SERVICE, TESTING AND RETURN TO SERVICE Form 3305.002B Page 1 of 2 Annunciator Control Periodic Review This review is completed by a Licensed Operator or STA to ensure the continued need for each annunciator that has been removed from service or modified. [PMRQ (U1) 9670, (U2) 9671]

1.

CHECK applicable unit for annunciator review.

Unit 1 Unit 2 2.

PERFORM the following actions for Annunciator Removal from Service or Modification sheets:

IF window marker is identified as being installed, THEN ENSURE window marker remains in place.

IF Condition Report was initiated, THEN:

ENSURE an active Work Order exists for the condition.

ENSURE Work Order is appropriately coded and prioritized.

IF alternative method of monitoring is listed, THEN ENSURE method valid for present plant condition.

IF annunciator is no longer needed to be disabled, THEN INITIATE action to restore annunciator.

ENSURE Annunciator Out of Service Index is current.

3.

RECORD corrective actions taken in the Comments section.

4.

IF annunciator has been out of service greater than 60 days AND PAD has NOT been performed, THEN:

a.

INITIATE a Condition Report that includes the following:

A PAD is needed for the annunciator out of service by a qualified individual prior to the annunciator exceeding 90 days of out of service time.

Completed PAD form be included with Annunciator Removal from Service or Modification form in Annunciator Out of Service binder.

Date that the annunciator out of service will exceed 90 days.

b.

IF annunciator out of service greater than 90 days with PAD review NOT completed, THEN:

1)

INITIATE a Condition Report documenting that the PAD has NOT been completed.

2)

INSERT CR number on Form 3305.002A, Annunciator Removal from Service or Modification.

60 days plausible since at 60 days a CR must be written to perform a PAD PAD requirement is 90 days

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1322 Source: New Rev: 0 Rev Date: 8/30/21 Originator: K. Smith TUOI:

A1LP-RO-TS Objective:

10 System Number: 2.3 System

Title:

Radiation Control Section: 2 Type:

Generic Knowledge And Abilities

==

Description:==

Ability to approve release permits.

K/A Number: 2.3.6 CFR

Reference:

41.13 / 43.4 Point Value: 1 RO Imp:

SRO Imp:

3.8 Tier:

3 Group:

G RO Select:

No SRO Select: Yes Question:

Given:

- A Gaseous Release Permit is being developed in accordance with OP-1104.022 Gaseous Radwaste System, Attachment C for T-18A Waste Gas Decay Tank

- RI-4830, Gaseous Radwaste Process Monitor failed its source check The SM/CRS may approve the Gaseous Release Permit provided that at a MINIMUM, two independent verifications are made for which combination of the following:

1. Sample of T-18A

2. Analysis of T-18A

3. Sping 2 setpoint adjustment

4. Computer Input Data

5. Valve lineup A. ONLY 2, 3, 4.

B. ONLY 2, 4, 5.

C. ONLY 1, 2, 3, 4.

D. ONLY 1, 2, 4, 5.

Answer:

D. ONLY 1, 2, 4, 5.

Notes:

Answer "D" is Correct. Per OP-1104.022, Att. C and ODCM L2.2.1, if RI-4830 is non-functional, a release may be performed provided that 2 independent verifications are made on Analyzing samples of T-18A, Computer Input Data, AND the valve lineup.

Answer "A" is Incorrect. Plausible that a valve lineup is not required to be independently verified since it is not required for a release associated with the Reactor Building Purge system. Also plausible that a sample isn't required to be independently verified since the analysis will be.

Answer "B" is Incorrect. Plausible that a sample isn't required to be independently verified since the analysis will be.

Answer "C" is Incorrect. Plausible that a valve lineup is not required to be independently verified since it is not required for a release associated with the Reactor Building Purge system. Also the Sping 2 setpoint adjustment is not required to be independently verified.

General Discussion:

Difficulty:

2 Taxonomy: F RO:

SRO:

98

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 This question matches the KA since the applicant must know what is required to approve a gaseous release when RI-4830 is non-functional.

SRO Justification:

10 CFR 55.43(b)(4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

History:

Used in 2022 SRO Exam

References:

ODCM L 2.2.1 OP-1104.022 Attachment C Gaseous Release Permit

1104.022 GASEOUS RADWASTE SYSTEM PAGE:

60 of 72 CHANGE: 044 ATTACHMENT C PAGE 3 OF 15 2.0 ANALYSIS (Chemistry) 2.1 SAMPLE Waste Gas Decay Tank circled in REQUEST section, step 1.1 for gamma spectroscopy. (Circle Tank)

Waste Gas Decay Tank: T-18A T-18B T-18C T-18D Sample obtained by: ____________________________

2.1.1 RECORD M&TE number ___________

2.1.2 RECORD Cal Expiration Date ___________

NOTE If an independent sample and analysis is needed per step 2.4, independent sampling and analysis may be performed concurrently with the following steps.

2.2 PERFORM Gamma spectroscopy.

Gamma spectroscopy performed by ________________________________

2.3 REVIEW Gamma spectroscopy report.

Gamma spectroscopy report reviewed by: ____________________________

2.4 IF Gaseous Radwaste Process Monitor (RI-4830) is non-functional, THEN:

2.4.1 PERFORM additional independent sample and analysis.

Performed by:________________________ Date ________

2.4.2 PERFORM additional independent verification of computer input data Performed by:________________________ Date ________

1104.022 GASEOUS RADWASTE SYSTEM PAGE:

67 of 72 CHANGE: 044 ATTACHMENT C PAGE 10 OF 15 4.3.3 REMOVE tag from the OUTLET valve of the Waste Gas Decay Tank (T-18) to be released.

NOTE The following step can alarm WASTE GAS DECAY TK HDR PRESS HI (K-115 C5) and is expected for this alignment. Alarm should clear when release starts.

4.3.4 OPEN outlet valve of Waste Gas Decay Tank to be released:

T-18A Outlet Isolation (GZ-13A)

T-18B Outlet Isolation (GZ-13B)

T-18C Outlet Isolation (GZ-13C)

T-18D Outlet Isolation (GZ-13D) 4.4 IF Gaseous Radwaste Process Monitor (RI-4830) is non-functional, THEN PERFORM Independent Verification of step 4.3 alignment.

(ODCM App.1, L2.2.1.a).

IV performed by _______________________

4.5 OPEN CV-4820 Outlet Isol (GZ-15).

4.6 NOTIFY Control Room of intent to begin release.

4.7 RECORD the following data:

Release Permit Number _____________________________

Release Start Time _________________________________

Date _____________________________________________

Tank being released _________________________________

4.8 RECORD above data on Gaseous Waste Disch Flow (FR-4831).

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1321 Source: Bank Rev: 0 Rev Date: 8/30/21 Originator: K. Smith TUOI:

Objective:

System Number: 2.4 System

Title:

Emergency Procedures / Plan Section: 2 Type:

Generic Knowledge And Abilities

==

Description:==

Knowledge of emergency response facilities.

K/A Number: 2.4.42 CFR

Reference:

43.1 Point Value: 1 RO Imp:

SRO Imp:

3.8 Tier:

3 Group:

G RO Select:

No SRO Select: Yes Question:

Per EN-EP-801, Emergency Response Organization, the lowest level of event classification that requires activation of the Technical Support Center (TSC) is an_____(1)_____. It is the responsibility of the_____(2)_____to assume command and control of the TSC and the onsite mitigation efforts.

A. (1) Unusual Event (2) Emergency Director B. (1) Unusual Event (2) Emergency Plant Manager C. (1) Alert (2) Emergency Director D. (1) Alert (2) Emergency Plant Manager Answer:

D. (1) Alert (2) Emergency Plant Manager Notes:

Answer "D" is Correct. The TSC must be activated at an Alert Condition or higher. The Emergency Plant Manager assumes command and control of the TSC, OSC, and all onsite mitigation efforts.

Answer "A" is Incorrect. An Unusual Event represents a lower classification than an Alert which is the lowest level classification that requires ERO activation. It is plausible to activate the ERO at an Unusual Event level since the SM can activate the ERO using discretion, however the question asks the lowest event level that requires activation. Emergency Director is plausible since they have command and control in the Emergency Operations Facility, not in the TSC.

Answer "B" is Incorrect. An Unusual Event represents a lower classification than an Alert which is the lowest level classification that requires ERO activation. It is plausible to activate the ERO at an Unusual Event level since the SM can activate the ERO using discretion, however the question asks the lowest event level that requires activation.

Answer "C" is Incorrect. Emergency Director is plausible since they have command and control in the Emergency Operations Facility, not in the TSC.

General Discussion:

This question matches the KA since the applicant must have knowledge of when ERFs are activated as well as the responsibilities of personnel within the ERFs.

SRO Justification:

Difficulty:

3 Taxonomy: F RO:

SRO:

99

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 10 CFR 55.43(b)(1) Conditions and limitations in the facility license.

History:

Bank QID58264 from WF3. Last used in WF3 2014 SRO Exam.

Updated to reference new procedure EN-EP-801.

Used in 2022 SRO Exam.

References:

EN-EP-801

EN-EP-801 Rev. 18 Page 10 of 67 Emergency Response Organization

c.

Pool ERO Position - Shift Augmentation ERO positions (e.g.,

Maintenance, Chemistry, Radiation Protection technicians and Operations staff) that are NOT assigned an On-Duty status.

6.

Minimum Staffing: The on-shift and augmented staff shown in the emergency plan table of minimum staffing (NUREG 0654, Table B-1 equivalent) required to provide facility accident response capability in key functional areas.

7.

On-Duty: Period of time (usually a week) that an ERO member is expected to be able to respond to their emergency response facility within their response time when notified of an emergency condition at the site and specifically designated to:

a.

Remain Fit-For-Duty

b.

Remain within the required response time of assigned ERF

c.

Ensure reliable transportation is available to respond to assigned ERF

d.

Ensure child/elder care arrangements have been made, if required

e.

Ensure proper attire is available, should ERF response be required

f.

Ensure required access badging is brought to ERF if required to respond

g.

Know the location of alternate and off-site ERF locations

h.

Fulfill ERO On-Duty Muster Meeting (if applicable)

8.

Operational: Status of an emergency facility declared by the appropriate facility manager upon determining that the facility is adequately staffed and equipment is setup and available to assume/perform the emergency functions assigned to that facility.

9.

Performance Indicator (PI): Objective data regarding licensee performance in the Reactor Oversight Process cornerstones of safety and security

10.

Response Time Goals: The emergency response facilities may be activated at any time and shall be activated at an Alert, Site Area Emergency, or General Emergency declaration. Once activated, the facility shall become operational as soon as possible (without delay) after declaration of any of these emergency classifications. Otherwise, it is the goal to be operational within the response time goals established by the site emergency plan or procedure. Response time goals for ERO positions are included in Attachment 7 - 13 of this procedure.

11.

Site-Specific ERO Positions: ERO Positions assigned at an Entergy facility that are in addition to the fleet standard ERO.

4.0 RESPONSIBILITIES

1.

All employees who may respond to an emergency are responsible to:

a.

Maintain ERO qualifications current.

EN-EP-801 Rev. 18 Page 25 of 67 Emergency Response Organization Page 2 of 5 TSC ERO Responsibilities Standard ERO Position Responsibilities Emergency Plant Manager (EPM)

The EPM reports to the EOF ED. The EPM has the responsibility for the command and control of all accident mitigation actions at the site and performs these duties from the Technical Support Center (TSC). Responsibilities include:

1.

DIRECT the activation, operation and deactivation of the TSC

2.

ASSUME command and control of the TSC and OSC and the onsite mitigation efforts

3.

PROVIDE information and recommendations to the ED regarding the classification of an emergency

4.

PREPARE and FACILITATE facility briefings.

5.

PROVIDE PEER check for event classifications when requested.

6.

ENSURE timely ENS notifications

7.

PERFORM accident assessment to prioritize mitigation actions.

8.

COORDINATE the activities of the CR, TSC and OSC

9.

DIRECT personnel evacuation, assembly and accountability of non-essential personnel

10. PROVIDE information and recommendations to the ED regarding plant activities

11. ADVISE the ED on core damage and plant conditions for classification and PAR determination.

12. DIRECT the organization, coordination, and prioritization of repair and corrective action teams

13. DIRECT onsite protective actions

14. AUTHORIZE emergency radiation exposure and issuance of KI to recommended personnel in the CR, TSC or OSC or to Security personnel.

15. MAKE operational decisions involving the safety of the plant and its personnel and MAKE recommendations to the Control Room Personnel

16. INITIATE immediate corrective actions to limit or contain the emergency invoking the provisions of 10 CFR 50.54(x) if appropriate1

17. IMPLEMENT severe accident management procedure strategies

18. DIRECT relocation to an alternate location.

19. INTEGRATE offsite responders with on-site response efforts when required

20. PERFORM emergency termination duties TSC Manager The TSC Manager reports to the EPM. Responsibilities include:

1. ASSURE staffing/timely activation of the TSC.

2. NOTIFY EPM when operational conditions exist.

3. RECOGNIZE and IMPLEMENT all technical aspects of accident mitigation for the emergency.

4. PERFORM technical assessments and COMMUNICATE the conclusions to the EPM.

5. SET priorities for the TSC personnel/OSC Teams.

6. ASSIST the EPM to make operational decisions concerning the safety of the plant.

7. OVERSEE the activities for relocation to an alternate location.

8. DIRECT the tracking of plant configuration changes.

9. DEACTIVATE the TSC when the emergency is terminated.

1The decision to depart from the license or a technical specification in an emergency shall be approved, as a minimum, by a licensed senior operator. If more senior licensee personnel are available, the decision to depart from the license in an emergency would pass to them as higher authorities in the chain of command. The rule does not specify that the senior licensee personnel be licensed senior operators or that they obtain the concurrence of a licensed senior operator to make such a decision. [RIS 2008-26]

Correct.

EN-EP-801 Rev. 18 Page 34 of 67 Emergency Response Organization Page 2 of 6 EOF ERO Responsibilities Standard ERO Position Responsibilities Emergency Director PROVIDE overall command and control of the emergency response. Responsibilities include:

1. RECEIVE turnover from the ED and ASSUME command/control of EOF and activities outside the area controlled by the TSC

2. DIRECT the activation, operation and deactivation of the EOF.

3. PREPARE and FACILITATE facility briefings

4. UPGRADE the emergency classification level. (cannot delegate)

5. MAKE protective action recommendations (PAR) to offsite agencies (cannot delegate)

6. DIRECT and APPROVE offsite notification to State and local agencies (cannot delegate)

7. COMMUNICATE within and between the emergency response facilities

8. ENSURE event information is communicated to other organizations (NRC, Entergy Corp, etc.) to keep them informed of the emergency situation.

9. DIRECT the activities of the EOF organization in support of the TSC and offsite response agencies. (County, Parish and State)

10. DIRECT protective actions for offsite monitoring teams, EOF ERO and offsite resources.

11. REQUEST assistance from offsite agencies, excluding requests for offsite medical/fire, security assistance. COORDINATE request for Federal assistance through the State.

12. INTEGRATE off-site responders with site response efforts when required.

13. AUTHORIZE issuance of KI and radiation exposure in excess of 10CFR 20 limits for ERO members outside of the protected area.

14. AUTHORIZE press releases (cannot delegate)

15. DIRECT facility relocation to the alternate EOF (where applicable)

16. DETERMINE reportability actions for non-emergency reportable events during an emergency (hazardous material spills, contaminated injured personnel, etc.).

17. CONDUCT turnover of command and control to relief ED.

18. TERMINATE the event in accordance with procedures (cannot delegate)

19. ESTABLISH and DIRECT recovery actions Plausible but in EOF not TSC.

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 QID: 1162 Source: Bank Rev: 4 Rev Date: 9/1/2021 Originator: Burton TUOI:

ASLP-RO-EPLAN Objective:

11 System Number: 2.4 System

Title:

Emergency Procedures/Plan Section: 2 Type:

Generic Knowledge and Abilities

==

Description:==

Knowledge of emergency plan protective action recommendations.

K/A Number: 2.4.44 CFR

Reference:

43.5 Point Value: 1 RO Imp:

SRO Imp:

4.4 Tier:

3 Group:

G RO Select:

No SRO Select: Yes Question:

Per Emergency Response/Notifications (1903.011):

(1) Protective Action Recommendations (PARs) are required to be provided after a declaration of_____(1)_____.

(2) If there has been a change in wind direction, then PAR must be re-assessed within a MINIMUM of_____(2)_____.

A. (1) ONLY General Emergencies (2) 15 minutes B. (1) ONLY General Emergencies (2) 30 minutes C. (1) BOTH General and Site Area Emergencies (2) 15 minutes D. (1) BOTH General and Site Area Emergencies (2) 30 minutes Answer:

A. (1) ONLY General Emergencies (2) 15 minutes Notes:

Answer "A" is Correct. 1903.011 states that PARs are required for GEs but not for NUE, Alert or SAE classifications. 15 minutes is correct requirement for making the notification for the GE which includes the PAR.

Answer "B" is Incorrect. Only General Emergency is correct, 30 minutes is wrong but plausible because this is the maximum allowed time to complete notifications from the time conditions are available in the Control Room.

(15 minutes + 15 to complete notifications)

Answer "C" is Incorrect. PARs are not required for SAEs but plausible as plant or localized evacuations could be required for GEs or SAEs based on the plant conditions. 15 minutes is correct.

Answer "D" is Incorrect. PARs are not required for SAEs as stated in "C". 30 minutes is plausible as stated in "B".

General Discussion:

This question matches the K/A since the applicant must have knowledge of when to provide Emergency Plan PARs and how often PARs are re-assessed.

SRO Justification:

10 CFR 55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Difficulty:

3 Taxonomy: F RO:

SRO:

100

INITIAL RO/SRO EXAM BANK QUESTION DATA ARKANSAS NUCLEAR ONE - UNIT 1 History:

Selected for 2017 SRO Retake exam.

Rev. 2 5/6/17

1. Revised stem and Notes based on NRC comment

2. Added "BOTH" in front of Distractors C & D to make them appear more like A & B which state "ONLY" Rev. 3, 5/17/17 Editorial changes.

Rev. 4 9/1/2021 Updated Objective Reference Updated procedure references and distractor analysis.

Used in 2022 SRO Exam.

References:

1903.011, Emergency Response/Notifications

1903.011 Rev. 062 Page 53 of 84 EMERGENCY RESPONSE/NOTIFICATIONS Page 1 of 11 Protective Action Recommendations (PARs) For General Emergency DISCUSSION: This attachment provides instructions for the assessment and initiation of Protective Action Recommendations (PARs) following the declaration of a General Emergency classification.

Offsite response agencies are notified of Protective Action Recommendation within 15 minutes.

Revisions to Protective Action Recommendations can be based upon:

Current plant conditions Projected offsite dose assessment Forecasted/actual wind shifts Evacuation is the preferred method for protecting the public within the ANO 10-mile Emergency Planning Zone (EPZ) as a result of a radiological emergency event at ANO. However, some circumstances might warrant a protective action of SHELTER when evacuation cannot be performed due to impediments or severe weather conditions. Sheltering might also be recommended when a PUFF RELEASE (less than one hour) is occurring. Individuals responsible for determining PARs at ANO consider all circumstances when developing protective actions.

In the event of a SHELTER PAR, Arkansas Department of Health (ADH) coordination is necessary to develop a plan for transitioning out of this protective action as soon as possible. This is especially of concern during weather extremes since the public is advised to shut down ventilation systems.

ADH is notified of the ANO protective action recommendations and is responsible for determining and issuing a Protective Action Advisory (PAA) to the County Judges (Conway, Johnson, Logan, Pope and Yell counties). Arkansas law places the responsibility for issuing protective actions to the public with the County Judges which will have both a Protective Action Recommendation and a Protective Action Advisory available for decision making. At a General Emergency classification, the Arkansas Department of Health, at a minimum, will issue a default Protective Action Advisory of evacuate a 5-mile radius and evacuate 5-10 miles downwind and the remaining EPZ to remain indoors and listen to emergency broadcasts. At a General Emergency classification, ANO, at a minimum, will issue a default Protective Action Recommendation (PAR) of evacuate a 2-mile radius and evacuate 2-5 miles downwind and the remaining EPZ to remain indoors and listen to emergency broadcasts. The ADH Protective Action Advisory encompasses a larger area than that recommended by federal guidance and the ANO General Emergency classification PAR. Awareness of this difference between the ANO protective action recommendation and the ADH protective action advisory is needed should a question arise. ANO PARs meet all of the EPA/NRC recommended regulatory guidance and are consistent with the rest of the nuclear industry.

WIND SHIFT discussion: If wind shifts are occurring or are predicted to occur within the 10-mile EPZ, guidance is provided on PAR No. 6 within this attachment.

USE OF FLOWCHART discussion: A PAR Flowchart is included in this attachment. This flowchart is used to help determine the correct PAR to issue based on plant conditions, release status, evacuation impediments and offsite dose assessment.

PAR CHANGES discussion: It should be noted that the numbers associated with the PAR (1-7) are for reference only. A PAR change and subsequent notification to offsite agencies via Form 1903.011-Y, Emergency Class Initial Notification Message, is only applicable when the actual recommended actions change. For Example: If PAR 1 is issued at the initial GE notification, and upon review of plant conditions PAR 3 becomes applicable; if the recommended actions (Evacuate, Shelter, Go Indoors) or Zones (A, B, C etc.) do not change there is NO CHANGE in the PAR and a new initial notification form is not required.

PARs required only after General Emergency declaration

1903.011 Rev. 062 Page 40 of 84 EMERGENCY RESPONSE/NOTIFICATIONS Form 1903.011-S Page 1 of 6 GE Emergency Direction and Control Checklist - Shift Manager GE NOTE This form is intended to be used by the SHIFT MANAGER when a General Emergency has been declared and the Shift Manager has the responsibility for Emergency Direction and Control (ED&C).

Steps can be performed by another individual if directed by the person with ED&C unless noted.

1.

IF this is a security event AND entry conditions for 1203.048, Security Event, are met, THEN MAKE ERO and offsite notifications per 1203.048.

2.

DETERMINE if a radiological release is occurring using 1, Release in Progress.

3.

IF the ERO has NOT been activated, THEN DIRECT a Control Room Communicator to activate Everbridge per, Emergency Response Organization Notification System Section 1.0. (SOER 99-1) 4.

Performing Protective Action Recommendations (NON-DELEGABLE):

a.

DETERMINE the appropriate Protective Action Recommendation using, Protective Action Recommendations (PARs) For General Emergency.

b.

REVIEW PAR criteria every 15 minutes for effects of wind shifts or offsite dose assessment or BOTH.

c.

Per Form 1903.011-Y, Emergency Class Initial Notification Message, NOTIFY the NRC, State, and local governments of any changes to PARs.

5.

COMPLETE page 1 of Form 1903.011-Y, Emergency Class Initial Notification Message for turnover to Control Room Communicator. Refer to Attachment 12 for detailed instructions if needed.

6.

DIRECT a Control Room Communicator to initiate notifications per EN-EP-603,, Initial Notification to State and Local Authorities using the completed Form 1903.011Y.

7.

TRACK personnel dispatched to plant until ERO Facilities are operational.

8.

ENSURE Chemistry personnel (Initial Dose Assessor) have been directed to the control room to implement 1904.002, Offsite Dose Projections.

15 minute time limit for wind shifts

ES-401, Page 24 of 52 ES-401 8

Figure 2-2 Screening for SRO-Only Linked to 10 CFR 55.43(b)(5)

(Assessment and Selection of Procedures)

Can the question be answered solely by knowing systems knowledge (i.e., how the system works, flowpath, logic, component location)?

RO question Yes No Can the question be answered solely by knowing immediate operator actions?

Yes Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs?

Yes No Does the question require one or more of the following:

assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event-specific sub-procedures or emergency contingency procedures knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

No No Yes Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only No RO question RO question RO question