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10-CW-2023-09 Post-Exam Submittal
	ML23305A094
Person / Time
Site:	Callaway
Issue date:	10/02/2023
From:	Chilton J; Ameren Missouri
To:	Thomas Farina; NRC/RGN-IV/DORS/OB
References
	Download: ML23305A094 (1)
	v • d • e

APPLICANT SUBMITTAL Queson 56: The keyed answer is that the Pressurizer level High deviaon, 32D, would be present aer the 20% down power at 3%/min with the assumpon that all controllers are in auto and no operator acon beyond the turbine controls load reducon. The annunciator alarms at a 5% or greater deviaon from actual level to programed setpoint. The programed pressurizer level is a setpoint from 25% to 56.17% that varies linearly with auconeered high Tave from 557 to 585.25 degrees F. This equates to 1.1% level change per degree which is designed to match the level changes resulng from the coolant temperature changes (FSAR 7.7.1.6). Therefore, Pressurizer actual level and program level should match when level is changed due to temperature only. Addionally, the rod control system automacally responds to a temperature dierence of 1.5 degrees between Tref and Tave and is capable of controlling Tave during a load ramp decrease of up to 5%/min (FSAR 7.7.1.1).

At 100% Tave will be at 585.25 degrees F and Pressurizer level will be a 56.17%. As turbine load is lowered at 3% per minute, Tref would begin to lower and Tave would begin to rise which raises Pressurizer level. The Rod Control System would respond to the deviaon within 1.5 degrees F (1.65%

Pressurizer level) to insert rods unl Tave stops rising (occurs around 586.75 degrees F Tave and 57.82%

Pressurizer level). Tave will then begin to lower as the 3% load reducon is within the 5% design capability of the Control Rods. Tave and Pressurizer level will lower back to 585.25 degrees and 56.17%

respecvely. Pressurizer program level and actual level will then both begin to lower at approximately the same rate due to charging "ow responding to the mismatch. Pressurizer level is maintained within 5% of programmed level and the 32D annunciator will not alarm. Aer the turbine load reaches set load, the rods will connue to step in to maintain Tave and Tref deviaon. The pressurizer pressure will connue to drop unl annunciator 33C, PZR Press LO HTRS ON, comes in. Of the four answers provided, Annunciator 33C is the only annunciator present as demonstrated on the simulator.

FACILITY ENDORSEMENT Queson 56. The applicant provided a writeup based on their desktop simulator results which show a dierent plant response than what was keyed. Training sta recreated stem condions on the full scope simulator which showed the keyed answer; Annunciator 32D (choice B), did NOT alarm during or aer the down power from 100% to 80% at 3%/minute. Note: Inially at 100% power, control rod is in automac control. However, aer the power maneuver is complete but while rods are stepping in to maintain Tavg vs Tref deviaon within limits and with PZR pressure sll lowering, Annunciator 33C, PZR Press Lo Htrs On, (choice D) alarms. These results were duplicated by rerunning the stem condions and recorded via Scenario Based Tesng (see atached report). Due to the full scope simulator results, Callaway recommends changing the correct answer from choice B to choice D.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 PRESSURIZER LEVEL CONTROL SYSTEM Group # 2 K/A # 011 A1.06 Importance Rating 3.3 Ability to predict and/or monitor changes in parameters associated with operation of the (SF2 PZR LCS) PRESSURIZER LEVEL CONTROL SYSTEM, including: PZR temperature Question #56 Reactor power was 100% when the load dispatcher directed Callaway perform a rapid downpower to 80% at 3%/minute.

With no additional operator action, what is the response of the Pressurizer Level Control System to this rapid downpower?

A. Annunciator 32C, PZR Lo Lev Dev, alarms and charging flow automatically rises B. Annunciator 32D, PZR Hi Lev Dev Htrs On, alarms and PZR backup heaters turn on C. Annunciator 32F, PZR Surge Temp Lo, alarms and the PZR variable heaters energize fully D. Annunciator 33C, PZR Press Lo Htrs On, alarms and PZR backup heaters turn on Answer: B Explanation:

Background:

PZR level is programmed to respond to reactor coolant volume changes caused by changes in Tavg. The intent is to have an essentially constant mass of reactor coolant regardless of power. A constant mass will reduce the required capacity of letdown and charging systems.

To achieve this constant mass, the PZR level program is a function of auctioneered high Tavg.

For no-load Tavg (557 °F), program level is 25 percent of span. This is a minimum program level no matter how much lower Tavg goes. This level ensures adequate capacity to absorb insurges and outsurges caused by load changes, while covering the heaters, and maintaining pressure.

If the initial PZR insurge following a downpower transient results in actual level exceeding programmed level by 5 percent, the backup heaters are energized and a "PZR HIGH LEVEL DEV HTRS ON" alarm is received. The heaters heat the relatively cool influx of water to prevent a pressure drop due to cooling of the PZR. Furthermore, when actual RCS Tavg subsequently decreases to the programmed Tavg for the reduced power level, an outsurge from the PZR occurs. Heating of the initial insurge minimizes any RCS pressure drop resulting from the outsurge.

The PZR surge line is connect to the RCS via the D hot leg.

NRC Written Examination Callaway Plant Reactor Operator Annunciators relevant:

32D - "PZR High Level Deviation Heaters ON" - 5% above program. (Upper Selected Channel/Controlling channel) 32C - "PZR Low Level Deviation" - 5% Below Program. (Upper Selected Channel/Controlling channel) 33F - "PZR Surge Temp Lo" - <520F 33C - Pressurizer Pressure Low - Heaters On, Setpoint: >25 psig below reference pressure on BB PK-455A Pressurizer program level is 25% at no load and 56.17% at 100%. Linearly this equation is 25% + power x 0.3117%. Once stable at 80% power, PZR programmed level will be 50%.

As less heat is rejected to the main turbine generator, RCS Tavg will initially rise (and associate density will lower) resulting in an insurge of relatively cooler water into the PZR. As shown by the PZR level calculation above, the insurge combined with the lower program level setpoint will be enough to cause a 5% level deviation alarm and the backup heaters to turn on to heat in cooler insurge.

A. Incorrect - plausible if the effect on insurge is confused / not determined correctly and an outsurge is believed to occur resulting in a level less than programmed and the backup heaters must full energize to prevent a pressure transient.

B. Correct - See above explanation C. Incorrect - plausible as the correct insurge response of relatively cooler water is correctly determined and the need to use heaters to warm the insurge is required but incorrect as the is alarm setpoint will not be reached (520F) nor is the PZR heater response limited to the variable heaters.

D. Incorrect - plausible as the response of the backup heater is correct but the reason and alarm response are wrong. The level deviation and backup heater response occurs such that a pressure transient (and this annunciator) does not occur Technical Reference(s):

1. OTA-RK-00018, Addendum 33F, Pressurizer Surge Line Temperature Low, rev 0

2. OTA-RK-00018, Addendum 32D, Pressurizer High Level Deviation Heaters On, rev 1

3. OTA-RK-00018, Addendum 32C, Pressurizer Low Level Deviation, rev 1 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#30, Objective G &J:

G. LIST the Pressurizer Pressure/Level Control System alarms, controls and protective function setpoints including coincidence and interlocks.

J. LIST the systems that interface with the Reactor Instrumentation and Pressurizer Pressure/Level Control Systems and EXPLAIN how a failure or setpoint change of a Reactor Instrumentation or Pressurizer Pressure/Level Control component affects the interfacing system.

Question Source: Vision System ID # ______

Modified Bank # ______

New ___X____

NRC Written Examination Callaway Plant Reactor Operator Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3___

10 CFR Part 55 Content:

10 CFR 55.41(b)(5)

Comments:

K/A match as the question examines the PZR level control response to subcooled insurge of RCS coolant into a saturated system i.e. the system response to temperature change

Table of Contents SBT Checklist.................................................................................................................................................................... 3 SBT Test Crew Personnel .................................................................................................................................................. 5 Scenario Summary ........................................................................................................................................................... 6 Recorded Data List ........................................................................................................................................................... 6 Recorded Data ................................................................................................................................................................. 8 Critical Parameters List ...................................................................................................................................................19 Attachments ...................................................................................................................................................................22 Alarm/Annunciator Output .........................................................................................................................................23 Simulator Action Monitor Log .....................................................................................................................................24 2

SBT Checklist Nuclear Power Plant-Referenced Simulator Scenario Based Testing Methodology Checklist Configuration: Trng2302 Workspace: ws2302 Scenario Name: Q56 Test Run Data Recorder: Q56_test.drb IC : 10 IC

Description:

MOC 26 100% FPSS (10000 MWD/MTU)

Date Validated: 10/10/2023 Revision: 0 Start Time: 12:35 PM End Time: 12:52 PM Duration: 00:16:58 Item Simulator Performance Circle Initial 1 Simulator performance supported scenario objectives. YES NO 2 Simulator initial conditions (IC) agreed with reference plant with respect to YES NO reactor status, plant configuration and system operation.

3 Simulator operated in real time during conduct of SBT.

YES NO Note: Use of Freeze allowed when evaluating specific performance.

4 Simulator demonstrated expected plant response to operator input and to normal, transient, and accident conditions to which the simulator has been designed to YES NO respond.

5 Simulator permitted use of the reference plants procedures so that the scenario was completed without procedural exceptions, simulator performance YES NO exceptions, or deviation from the scenario sequence.

6 Simulator did not fail to cause an expected alarm or automatic action and did not TRUE cause an unexpected alarm or automatic action. Note: Attach simulator alarm FALSE summary (versus time) to SBT Documentation Package.

7 Observable change in simulated parameters corresponded in trend and direction to those expected from actual or best estimate response of the reference plant.

YES NO Note: Attach predetermined Monitored Parameter List (versus time) to SBT Test Results record.

8 Reference plant design limitations were NOT exceeded. TRUE FALSE 9 Each scenario malfunction demonstrated expected plant response to its initiating YES NO cause.

10 SBT conducted in a manner sufficient (i.e. meets requirements of ANSI/ANS-3.5 2009) to ensure that simulator fidelity has been demonstrated and met for this YES NO scenario. Note: Attach relevant as-run marked-up plant procedures and or procedure portions/pages utilized to support assertion.

3

11 Modeling and hardware discrepancies identified during the conduct of SBT are documented and entered in accordance with the site simulator configuration management procedures. Note: Discrepancies that directly affect operator YES NO response (or action) or expected plant response must be resolved before the SBT test results can be judged as satisfactory.

12 Simulator SBT performance test results: Initial

____ SATISFACTORY/ ____ UNSATISFACTORY Note: attach list of SBT personnel including name, job title, and role.

Technical comments attached: Yes/No (circle one: if Yes, attach comments)

___________________________________/_______________________________ _____/____/____

SBT LEAD NAME SIGNATURE DATE 4

SBT Test Crew Personnel Name Job Title Role Steve Jones SRO Cert, STS Simulator and Lead Instructor Exam Groups Clay Cottingham SRO Cert, STS Initial License CRS Training Luke Williams SRO Cert ATC RO Eric Bussick SRO Cert BOP RO 5

Recorded Data Figure 1 RATPW 8

Figure 2 REP0499A Figure 3 REF0128A 9

Figure 4 REF0134A Figure 5 REL0112A 10

Figure 6 REL0480A Figure 7 RET0406A 11

Figure 8 RET0426A Figure 9 RET0446A 12

Figure 10 RET0466A Figure 11 RET0419A 13

Figure 12 RET0439A Figure 13 RET0459A 14

Figure 14 RET0479A Figure 15 MAJ0001 15

Figure 16 REP0480A Figure 17 REP0481A 16

Figure 18 SFAPOUTPUTR635920OUT Figure 19 RET0496A 17

Figure 20 RET0499A 18

Attachments 22

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Inadequate Heat Transfer - Loss of Group # 1 Secondary Heat Sink K/A # W/E05 EA2.10 Importance Rating 3.7 Ability to determine and/or interpret the following as they apply to (W E05) Loss of Secondary Heat Sink: High head SI flow Question #81 The crew has just entered FR-H.1, Response to Loss of Secondary Heat Sink.

Current plant conditions are:

All SG WR levels are at approximately 25%, lowering slowly

Core exit TCs are rising

Auxiliary Feedwater flow is 20,000 lbm/hr to each SG

CST to AFP suction header pressure is 6 psig, lowering slowly (1) Per FR-H.1 and based on the above conditions, what action(s) should the crew take NEXT?

And (2) Per FR-H.1, what procedure should the CRS direct to be performed in parallel with the above action(s)?

A. (1) initiate safety injection and open both PZR PORVs (2) EOP Addendum 19, Aligning ESW To AFW Suction B. (1) initiate safety injection and open both PZR PORVs (2) EOP Addendum 38, Non Safety Auxiliary Feedwater Pump C. (1) feed all SG at maximum rate until core exit TCs lower (2) EOP Addendum 19, Aligning ESW To AFW Suction D. (1) feed all SG at maximum rate until core exit TCs lower (2) EOP Addendum 38, Non Safety Auxiliary Feedwater Pump Answer: B Explanation:

Per FR-H.1, step #2 the criteria for establishing high head injection flow is SG WIDE RANGE level in any three SGs - LESS THAN 27% [42%] at which time step #12 actuate safety injection

NRC Written Examination Callaway Plant Senior Reactor Operator with step #13 and #14 verifying SI flow and establishing a bleed path via both PZR PORVs.

The distractor is from FR-H.1 foldout page item #3 - SG FEED FLOW RESTRICTIONS FOLLOWING RCS BLEED AND FEED CRITERIA:

IF core exit TCs are rising, THEN RESTORE feed flow as follows:

a. FEED any SG(s) that are NOT dry at maximum rate until core exit TCs lower.

Note: NO SGs are dry (SG WIDE RANGE level is greater than 10% - no adverse CTMT exists)

Per the foldout page of FR-H.1, EOP Addendum 19 (ESW to AFP) should not be directed until 2.75 psig. However, with the current CST to AFP suction pressure, EOP Addendum 42 (HCST to AFP) would be in progress or completed. Therefore, if the setpoints and/or EOP addendum are confused, it is plausible to believe EOP Addendum 19 should be directed next or in parallel with the above actions. Per FR-H-1, step #3e "ESTABLISH Non Safety Auxiliary Feedwater flow:"

directs the crew to "Perform EOP Addendum 38, Non Safety Auxiliary Feedwater Pump while continuing with this Procedure". Therefore, directing the EOP ADD 38 is correct.

A. Incorrect - see above explanations - part 2) is wrong B. Correct -

C. Incorrect - see above explanations - both are wrong D. Incorrect - see above explanations - part 1) is wrong Technical Reference(s):

1. FR-H.1, Response to Loss of Secondary Heat Sink, rev 20 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP#26, Objective G &I:

G. STATE and EXPLAIN the parameters which are evaluated, including their Criteria and Basis, to transition from the following procedures to another procedure.

1. FR-H.1, Response To Loss Of Secondary Heat Sink.

I. EXPLAIN the procedural flow path including major system and equipment operation in accomplishing the goal of:

1. FR-H.1, Response To Loss Of Secondary Heat Sink.

Question Source: Vision System ID # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

NRC Written Examination Callaway Plant Senior Reactor Operator LOD ____3___

10 CFR Part 55 Content:

10 CFR 55.43(b)(5)

Comments:

SRO ONLY due to ES-4.2 (Figure 4.2-3) of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely using fundamental knowledge of immediate operator actions of a procedure? NO Can the question be answered solely using fundamental knowledge of any AOP entry condition?

NO Can the question be answered solely using fundamental knowledge of major EOPs for Westinghouse: E0, E1, E2, E3, ECA-0.0, and red/orange functional restoration procedures? NO Can the question be answered solely using fundamental knowledge of the basic purpose, overall sequence of events, or the overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

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

FACILITY SUBMITTAL Queson 81. Due to an applicant queson regarding the "ow rate to each SG; was it operator induced or due to equipment malfuncons, it was determined that a further analysis of this queson is required.

The intent of part 1) of the queson was for the applicant to determine that RCS Bleed and Feed is required per connuous acon Step #2 when asked Per FR-H.1 and based on the above condions, what acon(s) should the crew take NEXT? The keyed answer of iniate safety injecon and open both PZR PORVs are acons of steps #12 and #13 which is directed from Step #2.c. The next acon would be to Stop all RCPs per Step #2.b prior to the keyed answer. Furthermore, when comparing part 1) of the keyed answer to the distractor, feed all SG at maximum rate unl core exit TCs lower there is litle discriminatory value. This is due to the fact that the distractor will always be done aer keyed answer.

Speci"cally, iniate safety injecon and open both PZR PORVs (RCS Bleed and Feed) will always be done before the distractor from the fold out page is applicable. E.g., the distractor is an acon from foldout page item #3, SG Feed Flow Restricons FOLLOWING RCS Bleed and Feed Criteria, therefore the distractor oers litle or no discriminatory value except knowing the procedure order. Due to part 1) of the queson having no correct answer and "awed psychometrics, Callaway recommends that this queson be removed from the exam.

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Pressurizer (PZR) Level Control Group # 2 Malfunction K/A # 000028 (APE28) AA2.13 Importance Rating 3.3 Ability to determine and/or interpret the following as they apply to (APE 28) PRESSURIZER (PZR) Level Control Malfunction: The actual PZR level, given an uncompensated level with an appropriate graph (REFERENCE PROVIDED)

Question #83 The plant is in Mode 1.

Various control room Pressurizer (PZR) level indications and controls have been lost.

Pertinent plant conditions are:

Loop 1 Tavg = 571F, stable

Loop 2 Tavg = 572F, stable

Loop 3 Tavg = 570F, stable

Loop 4 Tavg = 571F, stable (1) Based on the information above, what should PZR level be?

And (2) In order to restore these control room functions, a Temporary Modification (TM) is required. What action is required BEFORE the Shift Manager can authorize installation of the TM?

A. (1) 40%

(2) SRO review complete and documented on CA4691, Comment/Impact Review Only B. (1) 40%

(2) SRO review complete and documented on CA4691, Comment/Impact Review Form and Onsite Review Committee (ORC) approval C. (1) 42%

(2) SRO review complete and documented on CA4691, Comment/Impact Review Only

NRC Written Examination Callaway Plant Senior Reactor Operator D. (1) 42%

(2) SRO review complete and documented on CA4691, Comment/Impact Review Form and Onsite Review Committee (ORC) approval Answer: C Explanation:

Per 7250D64 Sheet 11, Pressurizer Level control uses the auctioneered high Tavg. Using the reference provided and the HIGHEST Tavg, 572F, PZR level should be ~42%. The distractor comes from using the average Tavg of 571F, which would be ~40%.

Per APA-ZZ-00605, step 4.4.2, "A TM requiring ORC approval may be installed prior to ORC approval provided the TM is ORC approved by the close of the next working day" therefore ORC approval is not required before installation just desired which makes it a plausible distractor.

Per EDP-ZZ-04600, Attachment 8, "Temporary Modification and Temporary Alterations Supporting Maintenance Special Instructions" Step 3.3 states "ENSURE a Senior Reactor Operator (SRO) is identified as a stakeholder in the package, as applicable. COPY the bulleted list below into the Feedback/Comments section of the CA4691, Comment/Impact Review and forward to the SRO.

SRO review must be documented (completed CA4691) in the Engineering Approval Package prior to TM installation."

A. Incorrect - See above explanation - wrong part 1)

B. Incorrect - See above explanation - both are wrong C. Correct D. Incorrect - See above explanation - wrong part 2)

Technical Reference(s):

1. OTO-BG-00001, Pressurizer Level Control Malfunctions, rev 25

2. 7250D64, Sheet 11, Functional Diagram Pressurizer Pressure & Level Control, rev 8

3. EDP-ZZ-04600, Engineering Change control, rev 16, Attachment 8 References to be provided to applicants during examination:

1. OTO-BG-00001, Figure 1, Programmed PZR Level (edited)

Learning Objective: T61.0110, Systems, LP# 9, Objective B: DESCRIBE the purpose and operation of the following RCS components to include interlocks, controller operations and power supply:

1. Reactor Vessel

2. Steam Generators (Primary Side)

3. Reactor Coolant Pumps (RCPs)

4. Pressurizer (Pzr)

Question Source: Vision System ID # ______

Modified Bank # ______

New ___X____

NRC Written Examination Callaway Plant Senior Reactor Operator Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ___4____

10 CFR Part 55 Content:

10 CFR 55.43(b)(3)

Comments:

SRO ONLY due to Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Figure 1 PROGRAMMED PZR LEVEL PZR LEVEL VS. PLANT POWER PZR LEVEL VS. RCS TAVG

APPLICANT SUBMITTAL:

Queson 83: The keyed answer is that the SRO review and compleon of comment/impact review form CA4691 is the ONLY item required before the Shi Manager can authorize the installaon of a Temporary Modi"caon. This is incorrect when ulizing guiding documents EDP-ZZ-04600 and IP-ENG-001 which are required per step 4.4.1 in APA-ZZ-00605.

Per EDP-ZZ-04600 step 4.12.11, Plant Review Board and Plant Manager approval is required because the temporary modi"caon is on a safety related system. This step references APA-ZZ-00091 which states in the ORC responsibility secon and step 4.2.3.a.18 "Review of Design Changes and Temporary Modi"caons that are: Safety Related, [Ref: 5.2.6], Special Scope, or classi"ed as Important to Safety Independent Spent Fuel Storage Installaon (Dry Cask Storage). (Engineering)

Safety Related Design changes shall be reviewed by the ORC and approved by the Senior Director, Nuclear Operaons."

The ORC approval is also required in IP-ENG-001 on Atachment 5 step 3.10.13 as part of the approval process for a safety related temp mod. Once Atachment 5 and the approval process for the temp mod is complete the modi"caon will move into the planning and implementaon phases, which IP-ENG-001 will direct use of ulity speci"c procedures and APA-ZZ-00605 will be entered to allow the Shi Manager to approve Authorizaon of installaon. The Standard Design Process (EDP-ZZ-04600 and IP-ENG-001),

APA-ZZ-00605 At 2, and APA-ZZ-00091 all state that ORC approval is required for Temporary Modi"caons of Safety Related systems and SRO only approval would not be correct.

Addionally, recent modi"caon (MP 23-0020) for D RCP seal leako was recently (July 2023) processed in accordance with staon procedures for a safety related Temp Modi"caon and DID require ORC approval prior to installaon.

FACILITY ENDORSEMENT:

Queson 83. The applicant challenged that the keyed answer for part 2) of the queson is incorrect. Part

2) asks What acon is required BEFORE the Shi Manager can authorize installaon of the TM? with the keyed answer of SRO review complete and documented on CA4691, Comment/Impact Review Only. The distractor adds an addional requirement that the Onsite Review Commitee (ORC) approval is required. While APA-ZZ-00091, On-Site review Commitee, step 4.2.3.a.18 states that ORC Safety Related Design changes shall be reviewed by the ORC and approved by the Senior Director, Nuclear Operaons. It does not speci"c that this approval is before or aer TM installaon, this informaon is in APA-ZZ-00605. Speci"cally, APA-ZZ-00605 step 4.4.2 states A TM requiring ORC approval may be installed prior to CRC approval provided the TM is ORC approved by the close of the next working day, as indicated in the queson explanaon. However, APA-ZZ-00605 Step 4.4.1 does direct the implementaon of IP-ENG-01, Standard Design Process. IP-ENG-01 de"nes temporary modi"caon as A short-term alteraon made to systems, structures, or components that is not controlled by procedure or work order instrucons, and is evaluated via a temporary Commercial Change, Design Equivalent Change or Design Change. Therefore, the Design Equivalent Change process "ow, Atachment 5, will apply. Per the "owchart on Atachment 5 page 1, Steps 3.3.13 and 3.3.14, review by Plant Review Board then Plant Manager or Site VP approval occur before the approved package is ready for WO planning and implementaon. It therefore appears that IP-ENG-01 requires ORC approval before implementaon. Note per APA-ZZ-00091, the plant manager is the ORC chairperson which implies the plant review board is the ORC.

Due to the con"icng procedural guidance of APA-ZZ-00605 vice APA-ZZ-00091 with IP ENG-Ol, Callaway recommends that this queson be removed from the exam. CR 202306570 was generated to revise APA-ZZ-00605 to remove the con"ict.

I2R\Incident Forms\Condion Report 202306570 All Details History CR Number 202306570 Title Procedure con"ict / clari"caon needed between APA-ZZ-00605 and IP-ENG-01 Descripon See Extended Descripon Extended Descripon Con"ict and the need for clari"caon was discovered while researching queson challenges for inial license training writen exam.

The speci"c issue is if On Site Review commitee approval is needed before installaon of a safety related temp modi"caon. APA-ZZ-00605 step 4.4.2 states " A TM requiring ORC approval may be installed prior to ORC approval provided the TM is ORC approved by the close of the next working day. However, APA-ZZ-00605 Step 4.4.1 does direct the implementaon of IP-ENG-01, Standard Design Process. IP-ENG-01 de"nes temporary modi"caon as "A short-term alteraon made to systems, structures, or components that is not controlled by procedure or work order instrucons, and is evaluated via a temporary Commercial Change, Design Equivalent Change or Design Change." Therefore, the Design Equivalent Change process "ow, Atachment 5, will apply. Per the "owchart on Atachment 5 page 1, Steps 3.3.13 and 3.3.14, "review by Plant Review Board then Plant Manager or Site VP approval" occur before the approved package is ready for WO planning and implementaon. It therefore appears that IP-ENG-01 requires ORC approval before implementaon hence the con"ict between the 2 procedures.

Due Date 10/27/2023 10:26:18 AM Immediate Acon Contacted Engineering, Generated CR Recommended Acon Evaluate if the direcon in APA-ZZ-00605 is correct - i.e. can a temp mod be implemented before ORC approval and revise procedure(s) as appropriate. This is an administrave issue.

SM No"ed? No Discovery Date 9/28/2023 10:26:18 AM Nofy Originator? Yes Business Tracking No

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 CONTAINMENT SPRAY SYSTEM Group # 1 K/A # 026 A2.04 Importance Rating 4.0 Ability to (a) predict the impacts of the following on the (SF5 CSS) CONTAINMENT SPRAY SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Failure of spray pump (REFERENCE PROVIDED)

Question #89 Reactor power is 100% with the 'A' Containment Spray Pump, PEN01A, in service for a full flow test.

A transient occurs which results in PEN01A tripping on overcurrent.

Simultaneously, the 'A' CTMT Cooler Fan, GN-HIS-7, locks out.

Per Technical Specifications and assuming no repairs can be completed, what is the LONGEST time allowed before the plant is required to enter Mode 3?

(Note: Risk Informed Technical Specifications are not available)

A. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months C. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months D. 174 hours0.00201 days 0.0483 hours 2.876984e-4 weeks 6.6207e-5 months Answer: B Explanation:

Per Technical Specification basis for 3.6.6,

The Containment Spray System consists of two separate trains of equal capacity, each capable of meeting the design bases. Each train includes a containment spray pump, spray headers, nozzles, valves, and piping.

Two trains of containment cooling, each of sufficient capacity to supply 100% of the design cooling requirement, are provided. Each train of two fan units is supplied with cooling water from a separate train of essential service water (ESW).

Therefore, with the conditions provided in the stem, 1 train of CTMT Spray and 1 train of CTMT Cooling is inoperable. TS 3.6.6 does not provide for this situation, and based on Section 3,

NRC Written Examination Callaway Plant Senior Reactor Operator LCO 3.0.3 is required because "When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit, as applicable, in: MODE 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

If only the CTMT spray pump is evaluated LCO 3.6.6 A & B should be applied which would result in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months + 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months = 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months .

If it is believed that Condition E could be applied due to the failures in the stem, then 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is plausible. Additionally if the applicant applied either LCO 3.6.6 B or D by itself, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is plausible.

If only the CTMT cooling train is evaluated LCO 3.6.6 C & D should be applied which would result in 7 days ( 7x24 = 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months ) + 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months = 174 hours0.00201 days 0.0483 hours 2.876984e-4 weeks 6.6207e-5 months A. Incorrect - See above explanation B. Correct - LCO 3.0.3 and its associated time requirements are applicable.

C. Incorrect - See above explanation D. Incorrect - See above explanation Technical Reference(s):

1. M-22GN01(Q), P&ID Containment Cooling System, rev 30

2. Technical Specification 3.6.6, Containment Spray and Cooling Systems, and its basis References to be provided to applicants during examination:

1. Technical Specification 3.6.6 Learning Objective: T61.0110, Systems, LP#18, Objective K: STATE the Limiting Condition for Operation (LCO) and Bases associated with the Containment Spray System Technical Specifications.

Question Source: Vision System ID # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

LOD ___3____

10 CFR Part 55 Content:

NRC Written Examination Callaway Plant Senior Reactor Operator 10 CFR 55.43(b)(2)

Comments:

SRO ONLY due to ES-4.2 (Figure 4.2-2) of NUREG 1021 as follows:

Can the question be answered solely by knowing 1-hour TS/TRM Action? NO Can the question be answered solely by knowing the LCO/TRM information listed above the line? NO Can the question be answered solely by knowing the TS safety limits? NO Can the question be answered solely by knowing the TS bases information associated with the above-the-line LCO information or general systems knowledge? NO Does the question involve one or more of the following for the TS, TRM, or ODCM? YES

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 (LCOs 3.0.1 through 3.0.7 and LCOs 4.0.1 through 4.0.4) YES

knowledge of TS bases that is required to analyze TS-required actions and terminology

Containment Spray and Cooling Systems 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray and Cooling Systems LCO 3.6.6 Two containment spray trains and two containment cooling trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One containment spray A.1 Restore containment 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months train inoperable. spray train to OPERABLE status. OR In accordance with the Risk Informed Completion Time Program B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months C. One containment cooling C.1 Restore containment 7 days train inoperable. cooling train to OPERABLE status. OR In accordance with the Risk Informed Completion Time Program (continued)

CALLAWAY PLANT 3.6-18 Amendment No. 229

Containment Spray and Cooling Systems 3.6.6 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition C not AND met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. Two containment spray E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months trains inoperable.

AND OR E.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Two containment cooling trains inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 --------------------------------- NOTE ----------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each containment spray manual, power In accordance operated, and automatic valve in the flow path that is with the not locked, sealed, or otherwise secured in position is Surveillance in the correct position. Frequency Control Program SR 3.6.6.2 Operate each containment cooling train fan unit for In accordance 15 minutes. with the Surveillance Frequency Control Program (continued)

CALLAWAY PLANT 3.6-19 Amendment No. 213

APPLICANT SUBMITTAL:

Queson 89: The correct keyed answer is that Mode 3 will be required in 7 Hours. The stem of the queson has the A Containment spray pump trip at the same me as the A Containment Cooler trips with that leading to entry into TS 3.0.3. If a containment spray train is inoperable, you enter Condion A.

If a containment cooling train is inoperable, you enter Condion C. This is consistent with the Secon 1.3 example, which states, When one Funcon X train and one Funcon Y train are inoperable, Condion A and Condion B are concurrently applicable. The Compleon Times for Condion A and Condion B are tracked separately for each train starng from the me each train was declared inoperable and the Condion was entered. LCO 3.0.3 is NOT applicable. LCO 3.0.3 is only applicable in three speci"c circumstances:

When an LCO is not met and the associated ACTIONS are not met

An associated ACTION is not provided, or

if directed by the associated ACTIONS There was a conversaon about an amendment for TS 3.6.6, Amendment 171, in 2006 with the following informaon provided:

"In reviewing the applicaon, the NRC sta had a conference call with the licensee on January 27, 2006, to "nd out if one containment spray train and one containment cooling train were sucient in themselves to provide 100 percent of the required design basis containment atmosphere cooling and iodine removal. The licensee stated that the case of only one containment spray train and one containment cooling train being operable is addressed in the Callaway Final Safety Analysis Report (FSAR). FSAR Secons 6.2.1.3 and 6.2.1.4 discuss the containment response to a postulated loss-of-coolant accident (LOCA) and to a postulated secondary pipe (i.e., a steam line) rupture, respecvely, inside containment. This is also addressed in FSAR Table 6.1.1-3, "Engineered Safety Features Design Parameters for Containment Analysis." The most severe single acve failure is loss of one diesel generator, which is the same as loss of one containment spray train and one containment cooling train.

With two containment spray and cooling trains provided, this means that for the worst single acve failure there is only one containment spray train and one containment cooling train that is operable. For this case, by the FSAR, there is 100 percent of the required design basis containment atmosphere cooling and iodine removal" In the case cited, the LCO is not met but the Acons are met, Acons are provided for the situaon, and the Acons dont direct entry into LCO 3.0.3. The Correct answer for this Queson would be 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months for entry into mode 3.

FACILITY ENDORSEMENT Queson 89. The applicant challenged that LCO 3.0.3 is not applicable to the situaon provided in the stem and that the correct answer is choice C, 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months not the LCO 3.0.3 answer of 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months (choice B). Addionally, the applicant references technical speci"caons example 1.3-3 (provided for reference) as a clear example or not implemenng LCO 3.0.3 for the situaon in the stem. Regulatory Aairs was contacted, provided the queson, and asked if LCO 3.0.3 was applicable. Regulatory Aairs then contacted an industry expert for guidance. In an email from Brian D. Mann to Thomas Elwood

dated September 22, 2023 (enclosed for reference), it was determined that LCO 3.0.3 is NOT applicable.

Therefore, the keyed answer, B, of 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months is incorrect. The correct answer should be LCO 3.6.6 Condions As compleon me plus condion Bs compleon me of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in Mode 3, i.e. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months which is choice C. Per Operaons Management and supported by Regulatory Aairs, Callaway recommends changing the correct answer from choice B (7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ) to choice C (78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months ).

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One A.1 Restore Function X 7 days Function X train to OPERABLE train status.

inoperable.

B. One B.1 Restore Function Y 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Function Y train to OPERABLE train status.

inoperable.

C. One C.1 Restore Function X 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Function X train to OPERABLE train status.

inoperable.

OR AND C.2 Restore Function Y 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months One train to OPERABLE Function Y status.

train inoperable.

(continued)

CALLAWAY PLANT 1.3-6 Amendment 229

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)

When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered).

If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A.

It is possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. However, doing so would be inconsistent with the basis of the Completion Times. Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended.

(continued)

CALLAWAY PLANT 1.3-7 Amendment 229

ROs Student Name ID Number % Score # Correct Q1 Q2 Q3 Q4 Q5 Q6 Q7 Q8 Q9 Key 75 75 C A A A C A B A A D A A A C A B B A C A A A C A B D A D A A A C A B B A SROs Student Name ID Number % Score # Correct Q 1 Q2 Q3 Q4 Q5 Q6 Q7 Q8 Q9 Key 100 100 C A A A C A B A A C A A A C A B B A C A A A C A B D A C A A A C A B A A D A A A C A B A A C A A A C A B B A C A A A C A B B A Total Missed 3 7

% incorrect 33.3 77.8 High miss question =

Q 10 Q 11 Q 12 Q 13 Q 14 Q 15 Q 16 Q 17 Q 18 Q 19 Q 20 Q 21 Q 22 Q 23 Q 24 Q 25 Q 26 Q 27 D B B B C C C D B B C C C D B A A B D B B B C C C D D B C C C D B A A A D B B B D C C D B B C C C D B B A B D B B B D C C D D B C C C D B B A B Q 10 Q 11 Q 12 Q 13 Q 14 Q 15 Q 16 Q 17 Q 18 Q 19 Q 20 Q 21 Q 22 Q 23 Q 24 Q 25 Q 26 Q 27 D B B B C C C D B B C C C D B A A B D B B B D C C D B B D C C D A A C B D B B B D C C D D B C C C D B A A B D B B B D C C D B B C C C D B B A B D B B B C C C D B B C C C A B B A B D B D B D C C D D D C C C D B B C B D B B B D C C D D B C C C D B A A B 1 7 5 1 1 1 1 5 2 1 11.1 77.8 55.6 11.1 11.1 11.1 11.1 55.6 22.2 11.1

Q 28 Q 29 Q 30 Q 31 Q 32 Q 33 Q 34 Q 35 Q 36 Q 37 Q 38 Q 39 Q 40 Q 41 Q 42 Q 43 Q 44 Q 45 D C C B D B C C C C B A A A A C B D D C C B D B C A D C B A A A A C B D D C C B D B C C C C B A A A A C B D D C C B D B C C A C B A A A A C B D Q 28 Q 29 Q 30 Q 31 Q 32 Q 33 Q 34 Q 35 Q 36 Q 37 Q 38 Q 39 Q 40 Q 41 Q 42 Q 43 Q 44 Q 45 D C C B D B C C C C B A A A A C B D C C C B D B C B C C B A A A A C B D B D C B D B C C C D B A A A A C B D D C C B D B C C C C B A A A A C B D D D C A D B C C C C B A B A A C B D D C C B D B C C D C B A A A A C B D D C C B D B C C C C B A A A A C B D 2 2 1 2 3 1 1 22.2 22.2 11.1 22.2 33.3 11.1 11.1

Q 46 Q 47 Q 48 Q 49 Q 50 Q 51 Q 52 Q 53 Q 54 Q 55 Q 56 Q 57 Q 58 Q 59 Q 60 Q 61 Q 62 Q 63 B D C A D D B C A D B A D A A C B D B D C A D D B C A D A A B A A C C D B D C A D B B C A B B A D A A C B D B D C A D C B C A D B A D A D C B D Q 46 Q 47 Q 48 Q 49 Q 50 Q 51 Q 52 Q 53 Q 54 Q 55 Q 56 Q 57 Q 58 Q 59 Q 60 Q 61 Q 62 Q 63 B D C A D D B C A D B A D A A C B D B D C A D B D C A D B A B A D C B D B D C A D D B C A D B A D A A C C D B D C A C D B C A D B A B A A C B D B A C A D D B C A D D A B A A D B C B D C A D D B C A D B A B A A C B D B D C A D B B C A D D A B A A C B D 1 1 4 1 1 3 6 2 1 2 1 11.1 11.1 44.4 11.1 11.1 33.3 66.7 22.2 11.1 22.2 11.1

Q 64 Q 65 Q 66 Q 67 Q 68 Q 69 Q 70 Q 71 Q 72 Q 73 Q 74 Q 75 D A A D B C B A A A C B D A A D B A A A A B D A D A A D B C B A A A C B D A A D B D B A A A D A Q 64 Q 65 Q 66 Q 67 Q 68 Q 69 Q 70 Q 71 Q 72 Q 73 Q 74 Q 75 Q 76 Q 77 Q 78 Q 79 D A A D B C B A A A C B C A D B D A A D B C B A A A C B D B C B D A A D B C B A A A C B C A D B D A A D B C A A A A C B C A D B D A A D D D B A A A D B C A D A D A A D B C B A A A C B D A C B D A A D B C B A A A D B D A C B 1 3 2 1 4 2 3 1 3 1 11.1 33.3 22.2 11.1 44.4 22.2 50.0 16.7 50.0 16.7

Q 80 Q 81 Q 82 Q 83 Q 84 Q 85 Q 86 Q 87 Q 88 Q 89 Q 90 Q 91 Q 92 B B B C C B A D B B A A B B B B D C B A D A C A A B B B B D C B A D A B A A B B B B C C B A D B C C A B B D B D C B A B A C A B B B B B C C B A D A C A A B B B B C C B A D B C A A B 1 3 1 4 5 1 1 16.7 50.0 16.7 66.7 83.3 16.7 16.7

Q 93 Q 94 Q 95 Q 96 Q 97 Q 98 Q 99 Q 100 D D C B C D C A D D C B C D C A D D C B A C C A D D C B C D C A D B C B C D C C D A C B C D C A D D C B C D C C 2 1 1 2 33.3 16.7 16.7 33.3

Thomas Farina From: Cottingham, Clay H <CCottingham@ameren.com>

Sent: Thursday, October 12, 2023 7:29 AM To: Thomas Farina Cc: Bussick, Eric; Swan, Phillip B; Turner, Robert G; Jones, Stephen M

Subject:

[External_Sender] RE: Additional TRs created Attachments: Callaway 2023 ILT NRC Exam TRs.pdf TJ -

Please find attached a PDF file containing the following Training Requests that have been generated from the 2023 ILT Class exam.

TR202300155 Evaluate reinforcing / incorporating the generic class weakness of application of Technical Specifications and FSAR (TRM) from Op Test TR202300156 Evaluate ILT training for CVCS lineups TR202300158 ILT Critical Safety Function training TR202300159 ILT GFES Refresher TR202300160 Evaluate use of EAL wallcharts for closed book examinations TR202300162 Foldout page of ECA2.1 (High missed exam question #8)

TR202300163 CCW systems loads and prioritization of abnormal (OTO) actions (High missed exam questions #14, 78)

TR202300164 RHR SDC flowpath and procedural requirements (High missed exam question #18)

TR202300165 Reportability Training for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months requirements (High missed exam question #88)

TR202300166 Functional Restoration Yellow path procedure entry and actions (High missed exam questions #25, 76)

Please let me know if you need any further information.

Thanks, Clay From: Thomas Farina <Thomas.Farina@nrc.gov>

Sent: Wednesday, October 11, 2023 10:57 AM To: Swan, Phillip B <PSwan@ameren.com>; Cottingham, Clay H <CCottingham@ameren.com>

Cc: Bussick, Eric <EBussick@ameren.com>

Subject:

[EXTERNAL] RE: Additional TRs created EXTERNAL SENDER STOP.THINK.QUESTION.

Verify unexpected requests before opening links or attachments.

Phil or Clay, please provide copies of the TRs generated for the exam, to include the ones highlighted below.

Also please provide a TR for the op test generic weakness that was debriefed onsite, related to applicants missing the TS calls for Scenario 1, Event 4: Loss of B Coolant Charging Pump (CCP).

Thanks, TJ Thomas J. Farina 1

ML23305A094

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