ML21098A170

CG-2021-02-FINAL Written Examination

ML21098A170
Person / Time
Site:	Columbia
Issue date:	03/04/2021
From:	Greg Werner Operations Branch IV
To:	Energy Northwest
References
50-397/21-02 50-397/OL-21
Download: ML21098A170 (533)

Text

ey (Amended Post-Examination)

A B C D A B C D A B C D A B C D 1 X 26 X 51 X 76 I X 2 X 27 X 52 X 77 X 3 X 28 X 53 X 78 X 4 X 29 X 54 X 79 X I X- I X 30 X 55 X 80 X 31 X 56 X 81 X 7 X 32 X 57 X 82 X 8 ' X 33 X 58 X 83 X 9 X 34 X 59 X 84 X 10 X 35 X 60 X 85 X I X 11 36 X 61 X 86 _,_ X 12 X X

- I 62 X 87 I X 13 X , 38

_ X 63 X 88 X 14 X ,_39 X 64 X 89 X 15 X 40 X 65 X 90 X 16 X 41 X 66 X 91 X Deleted 17

.~ ,,,,

X 42 43 X X 67 68 X

- X

- -92 93 X X

J.U 19 X 44 X I 69 X 94 X 20 X 45 X 70 X 95 X 21 X 46 X 71 X 96 X 22 X 47 X 72 X 97 X 23 X 48 X 73 I X 98 X 24 X 49 X 74 X I 99 X 25 X 50 X 75 I X 100 X Answer Distribution RO SRO A 17 A 6 B ~ 21 B 7 C 19 C 6 D 17 D 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 Examination Outline Cross-reference: 1 Revision: 2 Date: 1/18/21 Tier: 1 Group: 1 K/A Number: 295001.AK2.04 Level of Difficulty: 3 RO Importance Rating: 3.3 K/A

Description:

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following: Reactor/turbine pressure regulating system CGS is in Mode 1.

A loss of ---

both RRC pumps occurs.

• The crew performs a manual reactor scram.

• The CRS enters PPM 5.1.1 - RPV Control.

Once the plant is stabilized, the CRS enters ABN-RRC-LOSS, Loss of Reactor Recirculation Flow.

• MSIVs are open.

How should RPV pressure be controlled and why is this control method used?

Control RPV pressure A. relatively constant to minimize the onset of thermal stratification .

B. relatively constant to prevent complicating RPV level control actions.

C. in batches (approximately 200 psig) to minimize the onset of thermal stratification.

D. in batches (approximately 200 psig) to prevent complicating RPV level control actions.

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 Answer: C K/A Match:

Requires an understanding of how the RPV/Turbine Pressure Regulating system is operated on a loss of both RRC pumps.

Explanation:

A. Incorrect. Plausible since the strategy for pressure control is consistent with the strategy used in PPM 5.1.1. Plausibility is enhanced since the reason for the method of pressure control is correct.

However, ABN-RRC-LOSS directs controlling RPV pressure in 200 psig batches using SRVs or BPV.

B. Incorrect. Plausible since this is the strategy/reason of pressure control used in PPM 5.1.1.

However, ABN-RRC-LOSS directs controlling RPV pressure in 200 psig batches using SRVs or BPVs to minimize thermal stratifications.

C. Correct. ABN-RRC-LOSS, step 4.1.2.b, directs controlling RPV pressure in 200 psig batches using SRVs or BPVs to reduce thermal stratification.

D. Incorrect. Plausible since the strategy for pressure control is correct. However, the reason for controlling pressure in this manner is to minimize thermal stratification.

Tier 1 Discussion Requires knowledge of abnormal procedure supplemental actions.

Technical Reference(s)

ABN-RRC-LOSS, Loss of Recirculation Flow PPM 5.1.1, RPV Control Attached w/ Revision #

See Comments / Reference 5.0.10, Flowchart Training Manual Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5023 - Predict the effect that a loss or malfunction of the following will have on the Reactor Recirculation System: c. Dual RRC Pump trip Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 Question Source: C Bank #: Bank #

C Modified Bank #: Mod Bank #. (Note changes or attach parent) r New Question History: Last NRC Exam: N/A Question Cognitive Level: r Memory or Fundamental Knowledge

r. Comprehension or Analysis Justification for Cognitive Level Need to synthesize the initial conditions given in the question stem with the requirements of PPM 5.1.1 and ABN-RRC-LOSS 10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

ABN-RRC-LOSS Rev: Major: 16 Minor: N/A 4.1 loss of A[I RRC Flow 4_ 1_1 PERFORM ABN-CO:RE conot1rrently with this procedure..

Wll);.: The following four steps shou d be perf01med ooncurrently _

4_1-2 MIN IMIZE/0 ELAY the onset of them11a stratification by performing! the lol owing actions:

a_ WHEN alI ,oonbd rods are full rn, THEN REDUCE CRD flow to the Reactor as follows:

• CLOSE C!RiD-V-34, OR RESET the SORAMl

• ADJU ST CRD ,oooling flow to minimum 'Mlh CRD-FC-600.

b_ CONTIROt eactor ressure m 15atches ai;i roX1matel:t 200 ps g) to induce water mi>drtg1 in the lov.rer head r,eg ion as follows:

• OPENfCLOS one SRV Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 Number: ABN-RRC-LOSS I Use Category: CONTINUOUS Major Rev: 016 Minor Rev: N/A

Title:

Loss of Reactor Recirculation Flow Page: 21 of 31 4.1 .2b Opening a bypass valve or a safety rel ief valve results in a surge of steam flow. This has been shown to promote mixing in the bottom head region resulting in raising bottom head drain temperatures. Opening of the equivalent of one SRV to control RPV pressure to effect a depressurization in the allowable pressure control band enhances coolant circulation in the RPV bottom head. This can be used to meet the Tech Spec requirements and allow a recircu lation ump to be restarted . Constant depressurization at a un iform rate may not generate this response and may fai l to disrupt the inactive condit ion that promotes thermal stratification . Maximum benefit will be gained by aggressive pressure reduction , with the cooldown rate as close to Technical specification limit of 100°F/hr as possible. If the Techn ical Specificati on temperature limits for start of an idle recirculation pump have been exceeded , or are being approached , if the main condenser is available, open main turbine bypass valves or at least one SRV for a short time. The valve (s) should be opened only long enough to provide the surge of steam flow (typically a reduction in RPV pressure by 200 or 300 psi has been shown to raise bottom head drain temperatures).

Comments /

Reference:

PPM 5.1.1 Rev: Major: 22 Minor: N/A F'V:s 1

• 060 ps*g 1050 ps i Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 Comments /

Reference:

PPM 5.0.10 Rev: Major: 23 Minor: N/A Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1

d. Step P-3:

1) This step specifies the desired RPV pressU1re control band (belov.r the high IRPV presrure scram setpoint) arnd the preferred system (main turbrne IBPVs) rouse when oontrolling IRiPV pressure in this band .. Oxrttro[ling RPV pressure below thiis value av-aids SRV lifting due to a high pressure, and al 'OINS the scram logic to be reset (provided no other scram sigmal exists}.

No minimum value is specified sinoe the IRiPV pressure at whioh the IEOPs are entered cannot be predefined and the instruction mus,t provide appropriate guidance for all events. A rarg,et pressure shou!d be se ected close to the initial value and below the IRPV pressure scram setpaint that permits use of available injection systems. An initial adjustment to es,tablisli an appropriate target pr,essure is permitted, provided the targ,et ,can be r,eached expeditiously and the Techn.iral Specification ooo do"IM"II rate ILCO is not exoeeded. The note attaohed to Step IP-3 specifies best-praoti oe target pressures and the depressurization rate that should be applied to reach the target pressure. The best pradices RPV pressure added to the instrument ioon adjacent to this step is a reminder of the most desirable action level in this step.

2) A pressure reduction from an intermedliate pressure (e.g., 400 ps"g} to a pressure as low as the shutdown oooling RPV p essure intertock is not oons*dered a permissible initial pressure adjustment in this step ,even if the depressurization is within t he LCO oooldown rate. To maintain RPV pressure at intermediate or low pressures, the typical oomplement of RIPV injeotion systems shoulld not require such a preSS1Ure reduction. Refer to the discLESsion of Step P-5 to permission to reduoe RPV pressure below the shlllIDff head of low pressure pumps if injeotioo from these systems is necessary to establish and maintain adequate core cooling.

3) "StaDihze" means to ho d RPV pressure as constant as practicable within the constraints imposed ~ the nature of the event, the degree o con rol afforded by tlhe systems used, and the availarnlity of personnel to perform manual control functions. The intent 1s that pressure be held 3iS oonstant as is practicable. The specific actions required and the degr1ee to v.tliich the ideal of a oonstant pressure can be approo.ohed v.~11 vary aooording to these oonstraints. ror ex.amp e:

• If flow through a pressure control system is automatically regulated and there is no break in the primary system, RPV pressure can Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 usually be he d 'rlllith:in a fairly narrow control band ..,tj1h littte operator acition.

• If flow through a !a.r:ge capacity pressure oon.trol system ,cannot be thrott ed, the IRIPV pressure rontro l band will neoessarily be relatively wide_

• If pressure is d ecreasing, it may be neoessary to d ose S!RVs, main t u:rbrrte bypass va lves, or the IMSIVs. lln some scenarios, however, it may not be possible to ~erminate depressurization_ Inventory loss through a break o r operation ,of steam-driven injection systems may cause pressure to decrease_

• Changing p ant conditions may neoessitate adjustments to the value at 'Which pressure is initially stabiliz.ed_ If the pressure oontrdl system in use becomes urnwailab e, the alternate system used in its place may not be capab e of maintarn_ing the same contr:ol band . Similarly, if an inj ection syste m in use becomes unavailab e,, pressure may need to be reduced to per mit use of lower head systems.

4) Both the rate an.ct the magnitude of RPV p ressure changes must be considered. A pressUJre tha is slowly decreas*ng over a relativejy wide control band may be more stable~ than stlort-penod osallations within a narrower control band. In general, the ad~uac ofsteQS taken to stalfll12e RPV pressure ml.J!St 6e Judged by the effect of aJJY coJ1t1nuing pressure vari ations on RPV water leve l and wfiether additional acl!ioJls are possible or lilrely to afford better control capability. If pressure varial!ioos are not interfenng witti RPV water evel control actions or cannot Be stof:)ped, p:ressu e may be considered stabilized. If rontinuing pressure oscillal!ions aire complicating efforts to control RPV water evel o r ifthe existing J):ress!Ure prevents use of available i[ ljection systems, additional effort is warranted.

5) IMain t u:rbi rte BPVs are preferred beeat..Jse RPV pressure ,can be closeJy controlled, and because discharg,in.g heat fronn the rieactor to the main condenser preserves the heat caJ)acity of the W etwell. Use of the main turbine BIPVs in ies that re-,esrab ishing the man, coodenser as a heat sink is appropriate if it is available but not currently in service. Author ization is not ,given in this step to defeat arny MSIV isolation interlocks_

16) The i 1strument scale icoo associated ,imh thris step visually rndicates the req u:ired RPV pressure ,control bancl applicab e to this step.

Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 Examination Outline Cross-reference: 2 Revision: 1 Date: 12/28/20 Tier: 1 Group: 1 K/A Number: 295003.AK2.04 Level of Difficulty: 3 RO Importance Rating: 3.4 K/A

Description:

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the following: A.C. electrical loads CGS is in Mode 1.

• Transformer TR-B is out of service.

• Steam Tunnel Recirculation Fans, RRC-FN-8 and RRC-FN-9, are running.

A breaker fault causes breaker E-CB-7/1 to open.

• DG-1 starts and repowers SM-7.

Operators are performing actions in accordance with ABN-ELEC-SM1/SM7, Distribution System Failures.

How should the Steam Tunnel Fans be operated?

A. ONLY RRA-FN-8 should be operating. Restart RRA-FN-9 from the control room.

B. ONLY RRA-FN-9 should be operating. Start RRA-FN-21 from the control room.

C. ONLY RRA-FN-8 and RRA-FN-9 should be operating. No further actions are required.

D. ONLY RRA-FN-8 and RRA-FN-9 should be operating. Start RRA-FN-21 from the control room and run 3 fans until Steam Tunnel temperatures return to normal.

Answer: B K/A Match:

Requires knowledge of condition of AC electrical loads following a loss of A.C. power to a bus and the actions required by the ABN.

Explanation:

A. Incorrect. Plausible if believed that RRA-FN-8 is powered by SM-8. However, RRA-FN-8 is powered from SM-7 via SL-71. When E-CB-7/1 opens, SM-7 loses power until it is repowered from DG-1. RRA-FN-8 trips and does not automatically restart when power is restored. It must be restarted from the field. Additionally, RRA-FN-9 will be running and cannot be started from the control room.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 B. Correct. When E-CB-7/1 opens, SM-7 loses power until it is repowered from DG-1. RRA-FN-8 trips and does not automatically restart when power is restored. That leaves RRA-FN-9 as the only Steam Tunnel Fan operating. ABN-ELEC-SM1/SM7, step 4.2.1, states that operators must verify two steam tunnel fans are operating. The operator should start Backup Steam Tunnel Fan RRA-FN-21 from the control room. Note that it would also be correct to restart RRA-FN-8 from the field once DG-1 has repowered SM-7. However, this is not an answer choice.

C. Incorrect. Plausible if it is believed that RRA-FN-8 will automatically restart when SM-7 is repowered from DG-1. However, RRA-FN-8 does not automatically restart and must be restarted from the field. RRA-FN-9 is the only steam tunnel fan in operation. Since ABN-ELEC-SM1/SM7, step 4.2.1 requires operators to verify that two steam tunnel fans are operating, a second fan must be started.

D. Incorrect. Plausible if it is believed that RRA-FN-8 will automatically restart when SM-7 is repowered from DG-1. However, RRA-FN-8 does not automatically restart and must be restarted from the field. Additionally, the procedure caution in ABN-ELEC-SM1/SM7, step 4.2.1 prohibits running 3 fans concurrently in Mode 1.

Tier 1 Discussion Meets requirements for Tier 1 questions since examinees are required to demonstrate knowledge of Abnormal Procedure steps and cautions.

Technical Reference(s)

ABN-ELEC-SM-7, SM-1, SM-7, SM-75, SM-72, SL-71, SL-73 & SL-11 Distribution System Failures Attached w/ Revision #

SD000183, RB HVAC See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 15755 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-ELEC-SM1/SM7.

Question Source: (" Bank #: Bank #

(" Modified Bank #: Mod Bank #. (Note changes or attach parent)

C- New Question History: Last NRC Exam: N/A Question Cognitive Level: r Memory or Fundamental Knowledge

r. Comprehension or Analysis Justification for Cognitive Level Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 Requires candidate to synthesize information given with a knowledge of actions and cautions of the abnormal procedure along with an understanding of the power supplies to steam tunnel fans along with equipment response to a momentary loss of power 10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

ABN-ELEC-SM1/SM7 IRev: Major: 022 Minor: N/A Number: ABN-ELEC-SM 1/SM7 I Use Category: CONT INUOUS Major Rev: 022 Minor Rev: N/A Title : SM-1 , SM-7, SM-75 , SM-72 , SL-7 1, SL-73 & SL-11 Distribution System Page : 7 of 27 Fa ilures 4.2 Momentary or Complete Loss of SM-7 4.2.1 !E E-SM-7 has transferred to either E-TR-B or DG-1 ,

THEN PERFORM the following :

CAUTION Do not run all three Steam Tunnel Supply Fans at the same time in Modes 1, 2, 3.

Running all three fans may prevent the Leak Detection System from performing its design function . All three fans may be operated at the same time in Modes 4 or 5, provided the CRS/Sh ift Manager approves. Momentary operation of 3 fans is allowed during the process of sh ifting fans .

• VERIFY two steam tunnel fans are operating (N/A fan not runn ing ):

• RRA-F N-21 (E-PP-3DAC from SL-31 ) --

• RRA-F N-9 (E-MC -8B) --

• RRA-F N-8 (E-MC-7C ) --

Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 Comments /

Reference:

SD000183 IRev: Major: 12 Minor: 2 COLUMBIA SYSTEMS January 2020 RBHVAC SD000183, r12 mr2 granular, carbon impregnated, activated , coconut shell charcoal. Each charcoal filter contains an integral fire protection system composed of spray nozzles, internal distribution piping and controls to allow for wetting of the charcoal filters in the event that temperatures approach the carbon's iodine desorption temperature.

g) Centrifugal Fans (REA-FN-2A and 2B)

Each filter unit contains a 100 % capacity , direct driven, single inlet centrifugal fan. Each fan is equipped with an automatic, opposed blade , proportional pneumatic, fail closed damper located on the fan discharge.

E . Miscellaneous Systems

1. Main Steam Tunnel Fan Coil Units RRA-FC-8 , 9 & 21 (Figure 7)

The steam tunnel cooling units are recirculation units which maintain a temperature that prevents degradation of the equipment in the steam tunnel.

Each of the main steam tunnel fan coil units consist of water coils and a direct drive centrifugal fan. The water coils are cooled by Plant Service Water (TSW).

During periods of steam operation, two of three fan coil units in the main steam tunnel area are operated to maintain ambient conclitions as designed.

The centrifugal fans RRA-FN-8 (9) are started locally on RBHV-A(B) RB 572 ft. El and RRA-FN-21 is started from H13 -P8r2 (Board R. Main Control Room .

Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 COLUMBIA SYSTEMS January 2020 RBHVAC SD000183, r12 mr2 IX. POWER SUPPLIES REA-FN-lA SL-71 ROA-FN-lA SL-73 REA-FN-1B SL-83 ROA-FN-lB SL-83 REA-FN-2A MC-7C-B ROA-EHC-15 MC-8C-B REA-FN-2B MC-8C-B ROA-EHC-51 MC-8C-B REA-FN-3 E-PP-3DAC RRA-FN-10 MC-8B RRA-FN-1 MC-8B RRA-FN-11 MC-7B RRA-FN-2 MC-7B RRA-FN-12 MC-7B RRA-FN-3 MC-8F RRA-FN-13 MC-7B-B RRA-FN-4 MC-4A RRA-FN-14 MC-8B-B RRA-FN-5 MC-7B RRA-FN-15 MC-7B-B RRA-FN-6 MC-8B RRA-FN-17 MC-8B-B RRA-FN-7 MC-7C-A RRA-FN-19 MC-8B-A RRA-FN-8 MC-7C RRA-FN-20 MC-7B-B RRA-FN-9 MC-8B RRA-FN-21 E-PP-3DAC RRA-EUH-1 E-PP-Fl0 /1 RRA-EUH-5 E-PP-Fl0/ 1 RRA-EUH-2 E-PP-Fl0 /1 RRA-EUH-6 E-PP-Fl0/ 1 RRA-EUH-3 E-PP-Fl0/1 RRA-EUH-7 E-PP-Fl0/ 1 RRA-EUH-4 E-PP-Fl0/1 RRA EUH-8 E-PP-Fl0/ 1 RRA-AHU-8 E-PP-Fl0/ 1 E _ l\ifoireUaneou S~ stems L Main Stean1. 'Tu.rul.el F an Coil. Units RRA-FC-8, 9 1 (Figure 7),

The* stieam. rumrel a :x:ding umts are recu:c:u]a1ticm u nits whicb 11.W:1T1tain a.

ten.1:pernfure dm,t prevents deg:rradaition. of the equipmenl: in the stieam tuoneL Baich of ithe* main sterun ru:rmel fan coil. u nits oonsi:st of wa:ter ooils rum a diced d:rive centrifuga]. fan .. The v. ateir coils are cooled by Pllam Seirvice*

W&Jtec (TSW) .

During pe:riods of ste.mn operation, two o lliiee fan coil units in the nuwi steam hmne-1 area rue erntecl to maintnin am'foeot conditions a s sigoeo.

The oeotrifogru foru RR...\-FN-8 9 are started locally on RBHV-A(B RB 572 ft _E and RRA-FN- l l S stMteo from H]3-P812 BoaroR. Main Control Room) .

Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 Examination Outline Cross-reference: 3 Revision: 1 Date: 12/28/20 Tier: 1 Group: 1 K/A Number: 295004.AK3.03 Level of Difficulty: 2 RO Importance Rating: 3.1 K/A

Description:

Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER : Reactor SCRAM CGS is in Mode 1.

• Reactor power is 100%.

A fault has occurred on DC bus S1-7.

• S1-7 voltage as indicated and lowering at 10 VDC per minute.

0 50 100 150 200 250 300 CHARG ER C1 -7 AMPS DC 200 0 200 400 600 800 1000 0 30 60 90 120 150 BUS S1-7

• 12s voe What action should be taken?

A. Scram the reactor due to imminent loss of both Reactor Recirculation Pumps.

B. Manually transfer SM-1 and SM-3 to TR-S since automatic transfer capability will be lost.

C. Scram the reactor and trip the main turbine to automatically transfer SM-1 and SM-3 to TR-S.

Page 1 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 D. Manually transfer inverter IN-5 to the Alternate AC source since automatic transfer capability will be lost.

Answer: C K/A Match:

Requires understanding of the reason for performing a manual reactor scram when a loss of BOP DC is imminent.

Explanation:

A. Incorrect. Plausible since a reactor scram should be initiated and SH-5 and SH-6 supply breakers lose control power on a loss of S1-7. However, the buses will not be lost and therefore, RRC pumps will remain operating.

B. Incorrect. Plausible since automatic transfer from TR-S to TR-N is lost on a loss of S1-7.

However, automatic transfer of SM-7 and SM-8 to TR-B will still occur and the procedure simplifies the operation of the AC distribution system by requiring an automatic scram and main turbine / main generator trip to ensure that an automatic transfer to TR-S occurs prior to S1-7 voltage going below 105 VDC.

C. Correct. In accordance with ABN-ELEC-125VDC, Plant BOP, DIV1,2, &3 125 VDC Distribution System Failures, a loss of S1-7 (BOP 125 VDC) will cause a loss of breaker control power and the loss of ability to manually or automatically operate breakers remotely. In this condition, a MG trip might not cause an automatic transfer from TR-N to TR-S. A CAUTION prior to the immediate action step details that a loss of control power may occur at DC voltages less than 105 VDC. The indications given in the question stem show that voltage will reach this level in approximately 30 seconds. Therefore, the crew should perform step 4.1.1 to ensure that AC buses automatically transfer to TR-S.

D. Incorrect. Plausible since S1-7 supplies inverter IN-5. However, the normal power supply to IN-5 is MC-8A. Normal operation of IN-5 will continue as long as MC-8A is energized.

Tier 1 Discussion Requires knowledge of abnormal procedure cautions and required actions.

Technical Reference(s)

ABN-ELEC-125VDC, Plant BOP, DIV1,n2, &3 125 VDC Distribution System Failures Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7654 - Predict the effect(s) a failure of 125VDC bus S1-7 will have on:

Page 2 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3

a. IN5

b. RFW

c. Main Turbine

d. SM1,2,3 Control Power

e. SH5,6 Control Power

f. Generator Breaker Control Power

g. CR Annunciators Question Source: OBank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

New Question History: Last NRC Exam: N/A Question Cognitive Level: 0 Memory or Fundamental Knowledge (i:, Comprehension or Analysis Justification for Cognitive Level Requires examinee to synthesize information in the question stem with a knowledge of conditions where S1-7 is considered lost along with an understanding of automatic actions that will occur with a loss of S1-7 and actions that are required to be completed.

10 CFR Part 55 Content: 55.41 10 Page 3 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 Comments /

Reference:

ABN-ELEC-125VDC IRev: Major: 16 Minor: N/A Number: ABN-ELEC-125VDC I Use Category: CONTINUOUS Major Rev: 016 Minor Rev: NIA

Title:

Plant BOP , DIV 1, 2, & 3 125 VDC Distribution System Failures Page: 3 of 35 1.0 ENTRY CONDITIONS Any condition which could result in an unexpected loss of power to the applicable 125 VDC distribution system.

2.0 AUTOMATIC ACTIONS 2.1 125 voe BOP NOTE: H13-P840.A1-1 .1, TURB A TR IP and H13-P840.A1-1 .5, TURB B TRIP wi ll not annunciate on a loss of feed pump turbines due to loss of BOP Alarm power.

• Both Reactor Feed Pump Turbines trip due to loss of power to the RFPT trip circuits

• Normal AC source to E-IN-5 provides complete Inverter load

• Loss of BOP Alarm power (H 13-P820.B1-10.7)

• Loss of TR-S to TR-N transfer Page 4 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 4.0 SUBSEQUENT OPERATOR ACT IONS NOTE: Section 4.0 is subdivided as follows:

Section 4.1 for BOP 125 voe Section 4.2 for Div 1 125 voe Section 4.3 for Div 2 125 voe Section 4.4 for Div 3 125 voe 4.1 BOP 125 voe CAUTION Loss of the ability to trip or close BOP remotely operated circuit breakers may occur at battery voltages of LT 105 VDC. Plant status should be evaluated and appropriate actions taken prior to reach ing th is voltage.

NOTE: A list of loads supplied by 125 VDC BOP distribution system is provided in SOP-ELEC-DC-LU .

4.1.1 lE a loss of power to E-DP-S 1/7 is imminent, THEN PERFORM the following:

a. SCRAM the Reactor per PPM 3.3.1 .

b. TRIP the Main Turbine .

c. TRIP the Main Generator.

Page 5 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 Num bcer: ABN--ELEC-125VOe IUse categol"Y: CONT INUOUS M!ajor Rev: 016 Minor Rev: NIA

Title:

Pliant BOP, DIV 1, 2, & 3 1125 v oe Dis ribution Syste1n Fail1Ures Page: 22 of 15 5.0 BASES 4.1 125 VDC BOP Tllis, procedure an icipates. 3 125 VDC BOP eiharger {E--C1 -7) failure as the most likely reason for entelirng tll is proced:ure, but alisio 1P ro *des.guicfalflce fo r otfiler failure mec1tla11 isrns. Tile guid!ance providecl is centered! aroundl marr1taining1battery availability tor .irs long as IPOSsible with a battery charger out--m-s,ervioe.

l1f ann1Unciatio:n is fost, it is imperative hat tfile affected plant parameters. be continuousl¥ monitored until annunciation is reliumedl to service.

Tile 125 VDC Battery Otlai;ger i,s capa ble of supplying the 1125 v oe bus with th,e 125 VIDC Battery disconnected ffOm the blliS.

Tll is, procedure also p1rOvides direction in the event of a cornprete loss of power to !Jus E--S1-7. If this oc,ours.wh e til e plant is at or near rated power, both RIF'iAI pumps wi II trip due to a loss of cont10I pow.e r trip signal resultir1g1in a RPV low revell scram . The Tu rb ine Gerierator tllen windmills, at 11800 RPM, stil connected to the ,grid. When the 500KV breakers. ar,e tl"ip:pedl,

• tie electric plalllt will OT auto transfer to TR-S becallise of loss of control !POW.e r. Tll is resu lts in deg raded bUiS vol1tage andl frequ:enc.y as the Main Gerierator ,ooasts. down whire sfill ,co:nnectedl o tile electrical !Juses. In additi.on, amrnnciati:on is lost o:n Control !Room Boards A, B and C.

Tile following descri bes ttle basis for the steps taken if til e Reactor has sorammed on low RPV level, due o a loss of power to E--S11--7:

4.1.1 Ua loss of ower o 1=="S1--7 is imm inent, a manual scram ana Ma*n are e sim!;!lest and most ra~ia methodsto transfer IJQWer to R--S.

Page 6 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Examination Outline Cross-reference: 4 Revision: 2 Date: 12/28/20 Tier: 1 Group: 1 K/A Number: 295005.AK1.01 Level of Difficulty: 2 RO Importance Rating: 4.0 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP: Pressure effects on reactor power.

CGS is in Mode 1.

• The reactor has been operating at full power for 335 days since the last refueling outage.

A Main Turbine trip occurs, followed by an automatic reactor scram.

The CRS enters PPM 3.3.1, Reactor Scram, and PPM 5.1.1, RPV Control.

CRO1 is checking status of Reactor Recirculation (RRC) pumps in accordance with the subsequent actions of PPM 3.3.1.

What is the expected RRC pump response to this event and what is the reason for this response?

The RRC pumps are (1) to minimize the effect of (2) .

A. (1) running at 15 Hz (2) low RPV water level B. (1) running at 15 Hz (2) reactor power increase due to rising pressure C. (1) tripped (2) low RPV water level D. (1) tripped (2) reactor power increase due to rising pressure Answer: D Page 1 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 K/A Match:

Requires knowledge of the operational implications s of a main turbine trip on reactor pressure and reactor power.

Explanation:

A. Incorrect. Plausible since a reactor scram will cause a RRC pump runback to 15 Hz due to low RPV water level. However, with the conditions listed in the question stem, RRC pumps will trip when the Main Turbine trips due to the EOC-RPT circuit.

B. Incorrect. Plausible since RRC pump response is based on mitigating a power increase due to an increase in pressure. However, for the conditions given, RRC pumps will trip.

C. Incorrect. Plausible since RRC pumps will trip for the conditions given. However, the reason for this response is not due to low RPV level. It is based on mitigating a power increase due to an increase in pressure.

D. Correct. When the Main Turbine trips, the load reject will cause steam flow to rapidly lower. This will cause reactor pressure to initially rise. The higher reactor pressure will cause reduced voids in the core, adding positive reactivity and causing reactor power to rise. This effect is more pronounced at EOC where rod worth is lower due to fuel depletion. To counteract this effect, RRC pumps automatically trip on a scram with initial power GE 29.5% at EOC.

Tier 1 Discussion This question requires knowledge of expected plant conditions when performing subsequent actions of emergency operating procedures. Therefore, it meets the requirements for a Tier 1 question.

Technical Reference(s)

SD000178, Reactor Recirculation System Description SD000184, Reactor Recirculation Flow Control (ASD) Attached w/ Revision #

See Comments / Reference PPM 3.3.1, Reactor Scram Technical Specifications Bases Proposed references to be provided during examination: N/A Learning Objective: 5023e - Predict the impact on the RRC System of each of the following conditions or events: e. EOC-RPT logic Question Source: 0Bank #: Bank #

@:, Modified Bank #: LO01783 (Note changes or attach parent)

ONew Question History: Last NRC Exam: 2009 Question Cognitive Level: 0 Memory or Fundamental Knowledge Page 2 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4

r. Com prehe nsion or Ana lysis Justification for Cognitive Level Examinee must evaluate question conditions using an understanding of the relationship between reactor pressure and reactor power along with a knowledge of the RRC pump response at EOC, along with expected plant conditions while performing emergency operating procedures.

10 CFR Part 55 Content: 55.41 1 Comments /

Reference:

SD000178 IRev: Major: 17 Minor: 1 COLUMBIA SYSTEMS August 2017 RRC SD000 l 78, rl 7 mrl If an ATWS occurs the "RECIRC A PUMP TRIP ATWS INITIATED "

annunciator alarms in the control room. An annunciator for pump B would also be expected to alarm for all ATWS conditions since both pumps would normally be affected.

12. EOC-RPT (End of Cycle-Recirculation Pump Trip) LO 5023e LO5022n This trip also causes both recirculation pumps to trip in the event of a main turbine trip or load reject while turbine first stage pressure is greater than 14 . psig corres onds to 9. 5 % reactor power . By reducing core flow ,

negative reactivity is added and the increase in reactor power and pressure due to the turbine trip or load reject is minimized. The worst case is at EOC , when the negative reactivity addeo 6y the control rods for the first few inches of travel in a scram is much less than at BOC. This is due to fuel depletion in the bottom of the reactor core and more control rods being further out.

Page 3 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Comments /

Reference:

SD000184 IRev: Major: 20 Minor: 2 VIII. SYSTEM INTERRELATIONSHIPS A. Feedwater Level Control System - A RFPT trip concurrent with ai1 RPV Low LO-ll 787a Level alarm will cause Recirc pumps to run back to 30 Hz. This condition is LO-l l 790d sensed by the Digital Feedwater control system. NLO-12363 Calculation of the Steain Dome temperature (from Steam Dome pressure) and comparison to RRC suction temperature is perfonned by the Digital Feedwater FANUC.

B. Nuclear Boiler Instrumentation - Low RPV level (Level 4 , + 31.5") concurrent LO-l l 787b with a RFPT trip will cause Recirc pumps to run back to 30 Hz.

LO-ll 787c C. Recirculation System - The ASD system controls the speed of the Reactor LO-l l 788d Recirculation Pumps. Malfunctions in the system will affect that ability. LO-l l 790b D. Reactor Power - Once the reactor has been taken to power operation (producing LO-l l 787d steam) by the withdrawal of the control rods it is then possible to c01mol reactor LO-l l 788b power by chai1ging core flow which provides a relatively smooth ai1d evenly distributed power chai1ge at higher power levels when compared to control rod initiated power chai1ges. Raising core flow raises power, lowering core flow lowers power.

E. Reactor Core Flow - Core flow changes are made by chai1ging the speed of the LO-ll 787e RRC pumps . The ASD system does this . If ASD malfunctions , core flow will LO-l l 788a change respectively .

F . Reactor Water Level - Reactor Water Level is an input into the RRC 2ump LO-ll 787f runback circuit. Low RPV level (31.5 inches) with a feed pump trip in will cause LO-l l 788c runback of both RRC pumps to 30 Hz. RPV level lus 13 inches will cause LO-l l 790c runback of both RRC pumps to 15 Hz. Also RRC pump speed chai1ges will change RPV level, when RRC pump speed raises RPV level lowers. When RRC pumps speed lowers RPV level goes up.

Page 4 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Comments /

Reference:

SD000184 IRev: Major: 20 Minor: 2 Number: 3.3.1 I Use Category: CONTINUOUS Major Rev: 066 Minor Rev: 001 Ti tle: Reactor Scram Page: 6 of 26 NOTE: The fo llowing steps may be performed concurrently or as directed by the CRS .

5.2 Subsequent Actions - CRO1 NOTE: Per GE SIL 532 , control rod position FULL-IN indication may be temporarily lost fol lowing a scram due to a temperature excursion , which temporaril y reduces the strength of the magnet for the FULL IN position switch .

NOTE: The reactor is considered shutdown under all cond itions if no more than one control rod is withdrawn past position 02 .

5.2.1 INSERT Source Range and Intermediate Range Monitors. {OE-7 .7}

5.2.2 REPORT control rod status (all rods in/not in) to the CRS .

5.2.3 !E. NOT in an ATWS ,

THEN PERFORM the fol lowing:

a. RESTORE and MAINTAIN reactor water level +13", to +54.5".

See Attachment 8.1 for list of sources. {C-7 .8} {P-7 .10} _ _

b. VERIFY Reactor Recirculation pumps have runback to ~15 Hz.

NOTE: The preferred methods fo r stopping an RRC pump is by use of the STOP pushbuttons or by opening E-CB-RRA (B).

1) !E. the RRC pump(s) is uncontrolled ,

THEN TRIP the affected RRC pump AND REFER to ABN-RRC-LOSS Page 5 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Comments /

Reference:

Parent Question LO01783 IRev: Major: NA Minor: NA 1 ID: 12443 Points: 1.00 CGS is in Mode 1.

• The reactor has been operating 335 days since the last shutdown .

• A Main Turbine trip occurs.

Which of the following describes the basis for the response of the Reactor Recirculation pumps to this event?

The Reactor Recirculation Pumps trip to minimize the effect of the ...

A. increase in reactor power and the increase of reactor water level.

B. decrease in reactor power and the increase of reactor pressure.

C. decrease in reactor pressure and the increase of reactor water level.

D. increase in reactor power due to the increase of reactor pressure.

Answer: D Answer Explanation Ref: S0000178 Reactor Recirculation The RRC pump trip on a MT tri p (EOC-RPT) is designed to mitigate effects of the increase in reactor power and the increase in reactor pressure.

Page 6 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Comments /

Reference:

Technical Specification Bases IRev: 114 EOC~RiPT Instrumentation B 3.3.4.1 B 3.3 IINSTRUMBNTATIO.N B 3.3.4.1 End of Cycle IRecirculrnion Pum p Trip (EOC--IRPT) Instrumentation BASES BACIKGHOUiND The EOC-R PT in5bumentation inmates a reciroulation pump trip (IRPTI to reduce the peak reactor pressurn and power resulting from turbine trip or generator load rejectio n trans ients to provide additional margin to the corn thermal MCPR Safety Limit (SL).

The need for the additiooal negative reactivity in ,excess of that normally inserted o n a scram reflects end -of cycle reaotivity considerntions. Flux shapes at the end of cycle are such fuat t he control rods may not be able to ensure that lhermal limits am maintained by inserting sufficient negative reactivity duri ng ttle first few feet of rod tra.V1el upon a scram caused by Turbine Governo r Valve (TGV} Fast Closure, T ri p OiI Pressure - Low o r Turbine Throttle Valve (TTV) - Closure. The physica l phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity at a faster rate than the contro l rods can add negative reactivity.

The EOC--IRPT instrumentation as described in Reference 1 is comprised of sensors that deteot initiation of c osure of the TTVs, or fast closure of the TGVs, combined wiih relays, log;ic circuits, and fast acting circuit breakers that interrupt the po\',rer to each of the reoircu!ation pump motors. The channels include electronic ,e quipment (e.g., trip relays) that comp0res measured input signals wilh p~ s,t ablished setpoints. W hen the selpoint is e:xoeeded, the channel outputs an EOC-RPT signal to the trip logic. V\llien the drive motor br,eakers trip open, ihe recirculation pumps ooas,t dowt1 under their owt1 *nertia. The EOC-RPT has tv,ro identical t~p systems, either of whim can actuate an IRiPT.

Each EOC-RIPT trip system is a lvilo-oLlt-of..:tvvo log,i c for ea.ch Function ;

thus,. e1ther two TTY - Closure or two TGV Fast Closure, T rip Oil Pressure - Low sig,rmls are required for a trip system to aotuate. If either trip system actuates, both recircula *on pumps will trip. There are two d rive motor breakers in series per recirculation pump. One trip system trips one of the tiNo drive motor breakers for each recirculation pump and the seoond trip system trips the other drive motor breaker for ea.ch recirculation pump .

Cclumbia. Generating Station B 3.3.4. 1-1 Revisio n 73 Page 7 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Page 8 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-5 Examination Outline Cross-reference: 5 Revision: 0 Date: 10/14/20 Tier: 1 Group: 1 K/A Number: 295006.AA1.07 Level of Difficulty: 2 RO Importance Rating: 4.1 K/A

Description:

Ability to operate and/or monitor the following as they apply to SCRAM : Control rod position CGS is in Mode 1.

The crew manually scrams the reactor due to an instrument failure.

The CRS enters PPM 5.1.1, RPV Control.

Operators are performing the following step:

RC-2 IF it is determ ined that existing control rod pattern alone THEN IPP M 5.1.2> 0 does not always assure reactor shutdown Considering control rod position alone, which of the following combinations of control rod positions will provide sufficient shutdown margin to assure that the reactor is shutdown under all conditions?

One control rod at position (1) , an additional control rod at position (2) , and the remaining control rods at position (3) .

A. (1) 08 (2) 04 (3) 02 B. (1) 48 (2) 02 (3) 04 C. (1) 08 (2) 04 (3) 00 D. (1) 48 (2) 02 (3) 00

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-5 Answer: D K/A Match:

Requires knowledge of required rod position for adequate shutdown margin following a scram.

Explanation:

A. Incorrect. Plausible if it is believed that 2 control rods may be greater than position 02 and ensure adequate shutdown margin. However, only 1 control rod may be greater than position 02.

B. Incorrect. Plausible if it is believed that all control rods may be at position 04 (with one greater than position 04) and still demonstrate adequate shutdown margin. However, all control rods must be no greater than position 02 with one greater than position 02.

C. Incorrect. Plausible if it is believed that all control rods may be at position 04 (with one greater than position 04) and still demonstrate adequate shutdown margin. However, all control rods must be no greater than position 02 with one greater than position 02.

D. Correct. In accordance with OI-15, EOP and EAL Clarifications, The reactor is shutdown under all conditions if no more than one controlis withdrawn past position 02. Any number of rods may be partially withdrawn provided they are at position 02.

Tier 1 Discussion Requires knowledge of the basis for emergency operating procedure steps. Meets Tier 1 requirements.

Technical Reference(s)

OI-15, EOP and EAL Clarifications PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7784 - Given a list, identify the criteria that must be met to ensure that the existing rod pattern alone can always assure reactor shutdown.

Question Source: Bank #: LO01786 0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

ONew Question History: Last NRC Exam: 2009 Question Cognitive Level: ' Memory or Fundamental Knowledge 0 Comprehension or Analysis Justification for Cognitive Level

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-5 Requires memory of rod configuration to ensure reactor shutdown following a scram 10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

OI-15 IRev: Major: 032 Minor: N/A Number: 0 1-15 [ Use Category: INFORMATION Major Rev: 032 Minor Rev: N/A Title : EOP and EAL Clarifications Page: 11 of 25

• Step P-3 . If MSIVs are closed and SRVs are being used for pressure control , the normal pressure band direction at this step is 800 to 1050 psig . The CRS should direct the RO to "maintain reactor pressure 800 to 1050 psig with SRVs" . However, if the CRS has not yet given this direction , the CRO should take pressure contro l and report "CRS , MSIVs are closed and I have pressure control with a band of 800 to 1050 psig ."

NOTE: Following a scram if all APRMs are indicating downscale , but not all contro l rods are indicating full in, scram actions should be continued during the approximate 3 minutes to 5 minutes it ta kes for control rods to settle .

{AR-202477}

c. If called upon in PPM 5.1 .1 (override RC-2 ), to determine if the existing rod pattern alone always assures reactor shutdown, make the determination as follows:!

1) The reactor is shutdown under all conditions if no more than one control rod is withdrawn past position 02 . Any number of rods may be partially withdrawn provided they are at position 02.

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-5 Comments /

Reference:

PPM 5.0.10 IRev: Major: 023 Minor: N/A Number: 5.0.10 I Use Category: INFORMATION Major Rev: 023 Minor Rev: N/A

Title:

Flowchart Training Manua l Page: 104 of 370 actions to reset the scram and manual ly insert control rods if prior entry to the Reactor Power Control flowpath of PPM 5.1 .2 had been made.

d. Override RC-2 :

1) This override is applicable to all subsequent steps of PPM 5.1 .1 .

2) This override facilitates transition to PPM 5.1.2, RPV Control - ATWS , for those cond itions in which reactor shutdown cannot be assured under all cond itions on control rod insertion alone .

3) Positive confirmation that the reactor will rema in shutdown under all cond itions is best obtained by determ ining that all control rods are full in.

Green full-in lights on the full core display, GOS , plant process computer, and RW M display on panel H 13-P603 provide ind ication that all rods are full in. It is appropriate to continue non-failure-to-scram RPV control gu idance for this condition .

4) Criteria other than all rods full in may be used to determine that the existing control rod position alone will always assure reactor shutdown. These include:

• Core design basis shutdown marg in with a single control rod fu ll out and all other control rods full in .

• Maximum Subcritical Ban ked W ithdrawal Position (MSBWP): No more than one control rod greater than Notch 02 and all other rods at Notch 02 or less.

• Reactor eng ineer evaluation of shutdown margin.

5) The phrase" ... rod pattern alone always assu re reactor shutdown" requ ires that th is determination not consider negative reactivity which might be present due to Xenon bu ild-in or the injection of sod ium pentaborate. A borated core does not necessari ly behave with a negative temperature coefficient of reactivity. Reactor shutdown must be based on the control rod position exh ibited by the existing rod pattern so that subsequent changes in boron concentration do not return the core to critica lity.

6) If any possibility exists that the reactor may not always rema in shutdown on control rod insertion alone , the actions required for control of RPV parameters differ from those prescribed in PPM 5.1.1. The RPV level, RPV pressure and reactor power con tro l actions that are appropriate under this cond ition are specified in PPM 5.1.2, RPV Control - ATWS .

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-6 Examination Outline Cross-reference: 6 Revision: 1 Date: 12/28/20 Tier: 1 Group: 1 K/A Number: 295016.2.4.31 Level of Difficulty: 2 RO Importance Rating: 4.2 K/A

Description:

Control Room Abandonment: Knowledge of annunciator alarms, indications, or response procedures.

CGS is in Mode 1.

A fire in the control room occurs.

The shift manager directs ABN-CR-EVAC, Control Room Evacuation and Remote Cooldown.

• The control room is evacuated.

• All immediate actions of ABN-CR-EVAC are complete.

RPV level is lowering.

What is the highest RPV level, as indicated at the Remote Shutdown Panel, where Emergency Depressurization is required?

A. -147 inches B. -150 inches C. -161 inches D. -186 inches Answer: A K/A Match:

Requires knowledge of indications that require actions during control room evacuation.

Explanation:

A. Correct. In accordance with the ABN-CR-EVAC flow chart, IF Div 1 125 Battery LE 108 VDC OR RPV Level LE -147 THEN Emergency Depressurization.

B. Incorrect. Plausible since this is the bottom of scale for MS-LI-10, RPV level indication on the Remote Shutdown Panel. However, ABN-CR-EVAC requires Emergency Depressurization LE -

147 inches.

C. Incorrect. Plausible since this is the highest RPV level that requires an ED during non-ATWS conditions in accordance with PPM 5.1.1, RPV Level Control, step L-11. However, ABN-CR-EVAC requires Emergency Depressurization LE - 147 inches.

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-6 D. Incorrect. Plausible since this is the highest RPV level that requires an ED during ATWS conditions in accordance with PPM 5.1.2, RPV Level Control - ATWS, step L-15. However, ABN-CR-EVAC requires Emergency Depressurization LE - 147 inches.

Tier 1 Discussion Requires knowledge of parameters that need to be met to perform a step in the abnormal procedure. Meets Tier 1 requirements.

Technical Reference(s)

ABN-CR-EVAC, Control Room Evacuation and Remote Cooldown PPM 5.1.1, RPV Control Attached w/ Revision #

See Comments / Reference PPM 5.1.2 - RPV Control - ATWS Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 16139 - Upon completion of this class, the student will be able to demonstrate the knowledge required to independently perform the Licensed Operator tasks associated with ABN-CR-EVAC.

Question Source: r- Bank #: LO01585 C Modified Bank #: Mod Bank #. (Note changes or attach parent)

C New Question History: Last NRC Exam: N/A Question Cognitive Level: r- Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Requires memory of required level for ED in ABN-CR-EVAC 10 CFR Part 55 Content: 55.41 10

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-6 Comments /

Reference:

ABN-CR-EVAC IRev: Major: 042 Minor: N/A Nu mber: ABN-CR-EVAC I Use Category: CONTINUOUS Major Rev: 042

>--------------------~----------------< Minor Rev: N/A Title : Control Room Evacuation and Remote Cooldow n Page: 5 of 65 When the Co ntrol Room is habitable THEN I,_ _____

Return to the contro l Room per Attachment 7.2 1

_,I l A-2 J IF Div 1 125 Battery LE 108 voe RPV Level LE-147" THEN OR I EMERGENCY DEPRESSURIZATION I i

Nu mber: ABN-CR-EVAC I Use Category: CONTINUOUS Major Rev: 042 Minor Rev: N/A

Title:

Control Room Evacuation and Remote Cooldown Page: 13 of 65 7.2 When the Remote Shutdown Panel is manned , transfe r switches are actuated to the EMERG position to allow assessment of RPV conditions and to separate equipment control circu its (exposed to the Control Room fi re) from safe shutdown control signals in itiated from the Remote Shutdown Panel , Alternate Remote Shutdown Panel , DG-2 local control panel , an d DIV 2 switchgear cubicles.

Turning the power transfer switches to the EM ERG position results in the loss of automatic functions for applicable components, and components align to the position/status designated by the component control switch. Plant parameters including RPV water level and pressure are then used to determine core coo li ng adequacy and methods of level maintenance .

The equipment listed in the tables following the BASES section 3.0 are the only cred ited SSCs for Post Fire Safe Shutdown.

Control Room fire analysis in accordance with GL 86.10, Q/R 5.3.10 requires plants to take into account loss of offsite power, plus a fire which causes an evacuation of the Control Room , plus one spurious actuation , prior to transferring control to the Remote Shutdown Roo m. After transfer, up to four actuations are considered . The worst case spurious actuation postulated fo r CGS is a spuriously opened re lief valve that requ ires Emergency Depressu rization in 10 minutes (refer to GEH-0000-0075-4920 ana lysis) .

In February of 20 15, GNF2 Fuel Design Analysis (CVI/SPC 981-01 , 134) concluded that RPV level will reach T AF at approximate ly 22 minutes and that for all cases analyzed , the Peak Cladd ing Temperature is maintained well below the limit of 1500 degrees Fahrenheit. Furthermore, the analysis for the GNF2 Fuel Design incorporates a key assumption that depressurization of the Reactor Pressure Vessel will occur at 10 minutes. The refore under worst case cond itions, the Remote Shutdown Room is requ ired to be manned and ready to Emergency Depressurize within 10 minutes from the time the Shift Manager (or designee) orders a reactor scram due to a design basis fire . {AR-6.42}

The SRV control switches are required to be actuated to commence RPV depressurization when the vessel level reaches -161". However the RPV level gauge MS-Ll-10 range is +60 to -150 ". For Control Room evacuation , when level lowers to -1 " Emergency Depressurization is requ ired. {P-6.35}

Bv calculation . a minimum of five SRV's are reauired to be ooened to orevent the

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-6 Comments /

Reference:

PPM 5.1.1 Rev: Major: 22 Minor: N/A r

RPV level drops to

-161 in.

Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A L- 15 WHEN RPV level cannot be re stored and maintained above

-186 in.

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 Examination Outline Cross-reference: 7 Revision: 0 Date: 10/22/20 Tier: 1 Group: 1 K/A Number: 295018.AK1.01 Level of Difficulty: 2 RO Importance Rating: 3.5 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on component/system operation.

CGS is operating in Mode 3.

• RCC-P-1A & 1B are running.

Subsequently, RCC-P-1B trips on overcurrent.

The CRS enters ABN-RCC, Loss of RCC.

What actions should be taken?

A. Verify RCC-P-1C is running.

B. Re-open RCC-V-6, RCC Supply to RW/RB.

C. Stop the running Reactor Water Cleanup (RWCU) pump.

D. Periodically swap running CRD pumps to ensure they do not overheat.

Answer: A K/A Match:

Requires knowledge of system of system response to a partial loss of RCC.

Explanation:

A. Correct. When the breaker of any running RCC pump opens for any reason, either from stopping the pump or from the breaker tripping, the standby pump in Auto will start. RCC-V-6 will remain open since the standby RCC pump will start prior to the 10 second time delay for closing the valve on less than two running pumps. ABN-RCC, step 4.2.1 directs operators to verify that the standby RCC pump has started.

B. Incorrect. Plausible since ABN-RCC, step 4.2.2 directs operators to verify RCC-V-6 is open (RCC-V-6 will close with less than 2 RCC pump breakers closed), and only RCC-P-1A breaker is closed for a short period of time. Plausibility is enhanced if it is believed that RCC pump auto-start interlock is not active out of Mode 1. However, RCC-V-6 will remain open since the standby RCC pump will start prior to the 10 second time delay for closing the valve on less than two running pumps.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 C. Incorrect. Plausible since since ABN-RCC, step 4.2.3 directs operators to secure the running RWCU pump and isolate RWCU if RCC-V-6 is closed. Plausibility is enhanced if it is believed that RCC pump auto-start interlock is not active out of Mode 1. However, RCC-P-1C will start when the running pump breaker opens and RCC-V-6 will remain open.

D. Incorrect. Plausible since ABN-RCC, step 4.2.6 directs operators to swap CRD pumps if they are overheating due to a loss of RCC. However, RCC-P-1C will start when the running pump breaker opens. Additionally, RCC-V-6 will remain open since the standby pump will start in less than 10 seconds. Therefore, adequate CRD pump cooling will remain.

Tier 1 Discussion Requires knowledge of actions required by abnormal procedure for the conditions given in the question. Meets Tier 1 requirements.

Technical Reference(s)

SD000196, RCC System Description ABN-RCC, Loss of RCC Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5706 - Explain the interlocks associated with the following components or system conditions, including setpoints:

a. RCC pump auto start

b. RCC Pump trips.

c. Makeup to Surge Tank RCC-V-48.

d. Radwaste/Rx Bldg Supply RCC-V-6 Question Source: r- Bank #: LO00791

(' Modified Bank #: Mod Bank #. (Note changes or attach parent)

(' New Question History: Last NRC Exam: N/A Question Cognitive Level: (' Memory or Fundamental Knowledge r- Comprehension or Analysis Justification for Cognitive Level Requires an understanding of RCC pump and valve interlocks and a knowledge of abnormal procedures used to mitigate a loss of an RCC pump.

10 CFR Part 55 Content: 55.41 7 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 Comments /

Reference:

SD000196 IRev: Major: 14 Minor: 2 COLUMBIA SYSTEMS Jw1e 2019 RCC SD000196, rl4 1m2 V. CONTROL THEORY AND INTERLOCKS LO-5706 A. Control Room Controls

1. Board B a) PUMP RCC-P-lA (B , C)

Four-position switch: PTL, STOP, AUTO , START. Spring return to AUTO from STOP or START positions.

PTL pump cannot be started and will not auto restart STOP pump stops AUTO the tripping or stopping of either of the rmming pumps, without an "F " or "A" signal present , causes any non-running pump in "AUTO " to start START pump starts PUMP TRIPS:

• Overcurrent

• "F" signal, 1.68 psig drywell pressure

• "A" signal, Rx Wtr Lvl 2 (-50")

RCC pump breakers do not trip (i.e. remain closed) on a loss of voltage, the breaker indicating lights will correctly show the breaker closed, system parameters must be checked to determine pump status.

When power is restored, the pumps will restart because the breakers are closed. RCC pumps will also auto restart on FA reset (if not placed in PTL).

Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 b) RADW ASTE/RX BLDG SUPPLY RCC-V-6 Three-position switch:

NLO-122031 CLOSE, NOR, OPEN, Spring return to NOR CLOSE - the valve closes NOR the valve closes automatically if less than two pumps are on line as sensed by breaker position (plus a 10 second time delay .

OPEN the valve opens if at least two pumps are on line as sensed by breaker position This valve closes if any of the following actions are done:

• Removing the control power fuses from RCC-P-lA or from valve RCC-V -6 itself

• Closing the RCC-V-6 breaker Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 Comments /

Reference:

ABN-RCC IRev: Major: 006 Minor: 004 Number: ABN-RCC I Use Category: CONTINUOUS Major Rev : 006 Minor Rev: 004

Title:

Loss of RCC Page : 5 of 12 4 .2  !.Ea partial loss of RCC flow occurs THEN PERFORM the fo llowing :

NOTE: If RCC flow is lost to the Radwaste Bu ild ing , OG-RF-20A(B)( C) may trip .

4.2.1  !.E an RCC pump has tripped ,

TH EN VERIFY the standby RCC pump has started .

4.2.2 lE two RCC pumps are runn ing ,

TH EN VERIFY RCC-V-6 is OPEN .

CAUTION Closing RWCU-V-4 without throttling open RWCU-V-104 will result in RWCU-RV-3 lifting , if CRD seal purge is not isolated.

4 .2.3  !.E RCC-V-6 (RW/RB Supply) is closed ,

TH EN PERFORM the following .

a. STOP RWCU-P-1A(1B).

b. THROTTLE OPEN RWCU-V-104 .

c. CLOSE RWCU -V-4 (RWCU Suction Outboard Isolation ).

4.2.4 MONITOR RRC-P- 1A(B) pump and motor bearing , wind ing , and seal temperatures (RRC-TRS-601 on H13-P614 and H13-P602.A-6).

NOTE: RWCU -V-4 closes on RWCU non-regenerati ve heat exchanger outlet temperature GT 140°F.

4.2.5  !.E RWCU non-regenerative heat exchanger outlet temperature is GT 130°F (RWCU-T l-607 , point 3, H 13-P602),

TH EN REDUCE RWCU flow, OR SECURE RWCU per SOP-RWCU-SHUTDOWN.

4.2.6 lE CRD pump lube oil housing temperature is hot to the touch ,

TH EN TRANSFER CRD pump cool ing water to the Condensate Transfer header per Attachment 7.1, OR PERIODICALLY SHIFT CRD pumps .

Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Examination Outline Cross-reference: 8 Revision: 2 Date: 12/28/20 Tier: 1 Group: 1 K/A Number: 295019.AK2.14 Level of Difficulty: 3 RO Importance Rating: 3.2 K/A

Description:

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Plant air systems CGS is in Mode 1.

A leak in the Containment Instrument Air (CIA) system occurs.

The crew is taking actions in accordance with ABN-CIA, Containment Instrument Air System Failure.

Current CIA Main Header Pressure (CIA-PI-20) is 136 psig and lowering at 1 psig every 5 minutes.

What actions should be taken?

In accordance with ABN-CIA, operators should (1) and (2) .

A. (1) Place one loop of RHR in Suppression Pool Cooling (2) lineup to supply CIA with Control and Service Air (CAS)

B. (1) Place one loop of RHR in Suppression Pool Cooling (2) start an immediate Reactor shutdown C. (1) Install ADS solenoid keys in H13-P628 or H13-P631 (2) lineup to supply CIA with Control and Service Air (CAS)

D. (1) Install ADS solenoid keys in H13-P628 or H13-P631 (2) start an immediate Reactor shutdown Answer: C K/A Match:

Requires knowledge of actions required to operate plant air systems on a loss of CIA.

Explanation:

A. Incorrect. Plausible since lining up CAS to supply CIA is an action required by ABN-CIA for the conditions given in the question stem. However, lining up RHR for Suppression Pool Cooling is only required after the reactor is manually scrammed due to an imminent closure of MSIVs Page 1 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 (immediate action step 4.1. Closure of MSIVs occur at a header pressure between 50-80 psig, and is not imminent for the conditions given in the question stem B. Incorrect. Plausible since both actions are required for specific conditions in ABN-CIA. However, lining up RHR for Suppression Pool Cooling is only required after the reactor is manually scrammed due to an imminent closure of MSIVs (immediate action step 4.1. Closure of MSIVs occur at a header pressure between 50-80 psig, and is not imminent for the conditions given in the question stem. Additionally, a reactor shutdown is only required if CIA Main Header pressure drops to 80 psig (step 4.4).

C. Correct. In accordance with ABN-CIA, step 4.3, if Main Header pressure (CIA-PI-20) approaches 135 psig, operators should install ADS solenoid keys and supply CIA with CAS.

D. Incorrect. Plausible since installing ADS Solenoid Keys is a required action for the conditions given in the question stem. However, a reactor shutdown is only required if CIA Main Header pressure drops to 80 psig (step 4.4).

Tier 1 Discussion Requires knowledge of ABN actions. Meets Tier 1 requirements.

Technical Reference(s)

ABN-CIA, Containment Instrument Air System Failure Attached w/ Revision #

See Comments / Reference Proposed references to be provided during examination: N/A Learning Objective: 5146 - State the four (4) sources for the CIA and when each source would be used.

Question Source: (' Bank #: Bank #

c;' Modified Bank #: LO01919 (Note changes or attach parent) r New Question History: Last NRC Exam: N/A Question Cognitive Level: r- Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Candidate must have knowledge of the actions required by ABN-CIA for specific plant conditions.

10 CFR Part 55 Content: 55.41 7 Page 2 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Comments /

Reference:

ABN-CIA IRev: I Major: 007 I Minor: 001 Number: ABN-CIA I Use Category: CONTINUOUS Major Rev: 007 Minor Rev: 00 1

Title:

Containment Instrument Air System Failure Page : 4 of 8 4.0 SUBSEQUENT OPERATOR ACTIONS CAUTION INBD MS IV closure will occu r at approximately 50-80 psig.

CAUTION If CIA Main Header pressure drops below 135 psig , SRV Solenoid C control from H13-P60 1 is impaired and will eventually be lost as pressu re drops.

NOTE : Normal Containment Nitrogen Header pressu re is 110 psig and CN-FIT-3 (RB 4 71) normal flow is approximately 1.1 SCFM. A sustained flowrate of GE 12 SCFM will drop Con tainment Nitrogen Header pressure to the alarm setpoint, and ind icates a leak exists.

NOTE : If Containment Nitrogen Header pressure is normal on CIA-Pl-20 (H 13-P840, Bd A) and local Contain ment Nitrogen Header Pressure CIA-Pl-29 (RB 522) is low, Drywell leakage is ind icated.

NOTE : The following steps may be prioritized and performed out of order as determined by the CRS .

4.1 lE MS IV closure is imminent or occurring ,

THEN PERFORM th e following:

4. 1.1 SCRAM the Reactor per PPM 3.3.1.

4 .1.2 lE any inboard MSIV has closed ,

THEN PLACE the control switches for fill MS IVs in the CLOSED position .

4. 1.3 IN ITIATE RCIC for RPV level control.

4. 1.4 PLACE one loop of RHR in Suppression Pool Cooling .

Page 3 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Number: A8N-C IA I Use Category: CONTINUOUS Major Rev: 007 Minor Rev: 001

Title:

Containment Instrument Ai r System Failure Page : 5 of 8 NOTE : The CIA header downstream of CIA-V-39A supplies three ADS SRVs; MS-RV-4A, 48 AND 58 . The CIA header downstream of CIA-V-398 supplies four ADS SRVs; MS-RV-3D , 4D , 4C AND 5C .

4.3 !E. header pressure as indicated on CI.A:-Pl-20 or Cl:A:-Pl-4 approaches 135 psig ,

THEN PERFORM the following :

• INSTALL the ADS solenoid keys (H13-P628 or H13-P631 ) {keys 91-97/ 106-112)

• SUPPLY CIA with CAS per SOP-CIA-OPS .

4.4 lE Containment Nitrogen Header pressure drops to 73 psig (CIA-Pl-29),

OR CIA-Pl-20 (CIA Main Hdr. Press) (H13-P840, 8d A) drops to 80 psig_,

THEN START an immediate controlled Reactor shutdown per PPM 3.2.1.

Page 4 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Comments /

Reference:

Question LO01919 IRev: I Major: Maj I Minor: Min i*1***** ***********. *** *** ******** ** ********** *: ** ** ** **** *** ** *** ***** *** ** **** *** ** ** *** *l o *:***1 2 *1 *3 *0i*** *** ** *** ** * .................. .

i * * **** **** I * *

• o * * * * * * *

• 0 * ** * **** I *** o * * * * * * * * * * ~ * * * * * * * *** * * *** I * ** o * * * * * * *

• 0 * ** * **** I ****** ***** ***** ***** ****** ***** ***** ** * *** 0

• o **** I *** * ** * * *** * * *** * *** ~ * *** * * ** o l * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *~

With Columbia operati ng at rated power, CR02 respond s to a CIA A l R RECEIV ER PRESS LOW allarm (4.84-0 .AS 4-3).

W hich of th,e f oHow*ng is correct?

A* ******-r*rr*crAheader** i:iresisiire*aei;i"rades rLirffier*;*ki:s:s;*o:f *s"F~v*s*oi,eno1 d**.-c *"i::6nfrof If or all SRVs may occur. Enter AB N-OIA and dir,ed inst all ati on of AD S *

......... .....l.?g.l.!=:.. <?i9. ..~E?Y.?.9.rl.. ti t ~:Ptl:?.~.ct.rl:9. P.§~J.:............. ............ . .... ........ .... .............. ......... .................l

B. i If CIA.head er press ure-d eg rad es furthe r, k i,ss of s"i:i v s*o,len oid 'C' co,n tro(

affecting only the ADS SRVs may occur. Enter ABN-CIA and d irect .

.......l. if.1'~:t~I!.cl, i.()n.1.():f_ _,l\Q.$. ..S.(?IE:!*DI(:)itj .~.~Y?.9.'i:i.ti .1..~*_-:P.{3-?~ .~ fl 9.. _p,~ ~.1.:...................................... l

• c_** ... T ff CIA head er pressure degrades f i.frfher , !o,s:s ofSJ:~V Soienoid 'A'for i Divis*on 1 AD S SRVs and SRV Sol!enoid 'B' for Divisi:on .2 ADS SRVs

......... .....l.~ .clY..().c:c;LJ.r.-..E:: r.it.,!=r .J.\13- ~~AP§.. ~ i:i9..:qir~.c:t .c;r.()'.s.~c:9~r.i E? g ir1.g .(?f .QA$. t(:), C:,:I_I'.\,_.. .

ID.  ! If CIA head er press ure d eg rades furthe r, loss of SRV Sol,en oid 'A ' for all

! Divi sion 1 SRVs and SRV Solenoid 'B' for all Division .2 SRVs may occur .

......... .... .l.!::r1~E?*r..A13-.~~AP.$ .. ct.ftq. q.ir.!=.c:~..c:r.():?:.s.:~c;9f!.r1,E,.c:tiD:9. ..CJ{ .C:.A§.. t.() .c:;.1 .A'. .......... ....................

.A nsw er: *A An sw er Explanation A loss of CIA pressure affeds t he G solen oid of alll SRVs. AB N-CIA is ent ered and d irects *nst aHati on of k,ey s for AD S SRVs in backpa nel P6.28 and P631 - A is correct .

Page 5 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Question 1 Info Que-sti on Type: Multiple Choioe Status: Active A mays se!ect on test? No A uthorized f m practice? No Points: 1.00 T ime t o Comp let e: 0 D iffi cu lty: 0.00 Syst em ID: 12730 User-De*fined ID: L00 1919 Cross Referen oe 4.840.A5 4-3 ARP; AB N-GIA Number:

Given plant annunciati on and indicati ons, evaluate T op ic:

conditi ons f or entry int o ABN-CIA. [ABN-CIA)

Num Fie-Id 1: 3.4 Num Fie-Id 2: 3.6 Text Fietct 218000 A2.03 Comment s: 4.840.A5 4-3 ARP; AB N-GIA NRC Exam 2009 218000 A2.03 17Abillity t o (a) predict the i mpacts of the fol owing on th e A u omati c Depres:surizati on System; and (b) based on th ose predictions, use proced ures to corr ect, contml, or mitigate th,e consequences of those abnorma conditi ons or operations: Loss of air supply to ADS valves (3.4 /

3.6) 55.43.5 H2 None Page 6 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-9 Examination Outline Cross-reference: 9 Revision: 1 Date: 12/28/20 Tier: 1 Group: 1 K/A Number: 295021.AA2.07 Level of Difficulty: 4 RO Importance Rating: 2.9 K/A

Description:

Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING:

Reactor recirculation flow CGS is in Mode 4.

• The reactor has been shutdown for 3 days.

• RHR Loop B is in shutdown cooling.

• RHR-P-2A is tagged out for emergent repairs.

• Both RRC pumps are off, but available.

RHR-P-2B trips due to a failure of the pump breaker.

• The CRS enters ABN-RHR-SDC-LOSS, Loss of Shutdown Cooling.

• Maintenance reports that it will take 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to replace the RHR-P-2B breaker.

What is the preferred method for restoring core circulation until RHR-P-2B is restored?

A. Start a RRC pump and monitor RPV temperature.

B. Raise RPV level to GE 60 inches to enhance natural circulation.

C. Makeup to the RPV with the CRD system and letdown to the hotwell through the RWCU system.

D. Bleed steam through the main steam drain lines while maintaining RPV level with Condensate Booster Pumps.

Answer: A K/A Match:

Requires knowledge of requirements for recirculation flow while shutdown and the preferred method for recirculation flow restoration during a Loss of Shutdown Cooling.

Explanation:

A. Correct. With the plant in Mode 4, Technical Specification LCO 3.4.10 requires forced core circulation. A running RRC pump is one method to provide forced circulation. ABN-RHR-SDC-LOSS specifies this as the preferred method when both RHR pumps are unavailable.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-9 B. Incorrect. Plausible since step 4.3 of ABN-RHR-SDC-LOSS directs raising RPV level to GE 60 inches to enhance natural circulation for if no RHR pumps or RRC pumps can be restored.

However, this is only performed if both RHR pumps or a single RRC pump cannot be started.

C. Incorrect. Plausible since this is a method for additional decay heat removal listed in ABN-RHR-SDC-LOSS. However, this is not the preferred method to restore forced circulation.

D. Incorrect. Plausible since this is a method for alternate decay heat removal listed in ABN-ADHR.

However, this is not the preferred method to restore forced circulation.

Technical Reference(s)

ABN-RHR-SDC-LOSS, Loss of Shutdown Cooling Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7728 - Describe the physical connection and/or cause-and-effect relationships between the RHR system and the following: e. RRC Question Source: r Bank #: Bank #

r Modified Bank #: Mod Bank #. (Note changes or attach parent)

r. New Question History: Last NRC Exam: N/A Question Cognitive Level: r Memory or Fundamental Knowledge

r. Comprehension or Analysis Justification for Cognitive Level Requires synthesizing the information given in the question stem with an understanding of the methods available for forced circulation along with the preferred method of ABN-RHR-SDC-LOSS 10 CFR Part 55 Content: 55.41 10 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-9 Comments /

Reference:

ABN-RHR-SDC-LOSS IRev: Major: 7 Minor: N/A 4.1.7 !E SOC restoration is delayed and forced core circulation is required ,

THEN START RRC-P-1A(B) per SOP-RRC-START.

4.1 .8 !E add itional decay heat removal capability is required ,

THEN IMPLEMENT one or more of the following :

• MAKE UP with CRD ,

AND LET DOWN with RWCU per SOP-RWCU-OPS to maintain RPV level and maximize reactor cooling.

Number: ABN-RHR-SDC-LOSS I Use Category: CONTINUOUS Major Rev: 007 Minor Rev: N/A

Title:

Loss of Shutdown Cooling Page: 6 of 14 CAUTION RCC system flows are balanced per SOP-RCC-OPS . Adjusting RCC va lves affects cooling flows to other components.

• !E RWCU is operating THEN THROTTLE OPEN RCC-V-8 , RWCU-HX Outlet, to maximize core cooling ,

AND MONITOR RWCU temperatu res at RWCU-Tl-607 points 1, 4, and 5 hourly.

• !E flooded up with the Fuel Pool gates removed THEN MAXIMIZE Fuel Pool Cooling per SOP-FPC-OPS.

4. 1.9 !E SOC cannot be restored ,

AND additional decay heat removal capabi lity is required ,

THEN REFER to the following:

• ABN-RHR-SDC-AL T

• ABN-ADHR

• SOP-FPC-ASSIST-ALT.

4.2 !E BOTH loops of SOC are out of service, BUT circu lation IS being provided by at least one RRC pump ,

THEN MONITOR RPV temperature at RRC-TR-650 pt. 1(2).

4.3 !Eat least one RRC pump OR two SOC loops are NOT in service, THEN RAISE RPV level to GE 60".

4.4 !E forced core ci rculation IS NOT being provided by at least one RRC pump ,

OR by at least one SOC loop with RPV level GE 60 ",

THEN MONITOR RPV metal temperatures every 30 minutes per OSP-RCS-C103.

Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-9 Number: ABN-RHR-SDC-LOSS I Use Category: CONTINUOUS Major Rev: 007 Minor Rev: N/A

Title:

Loss of Shutdown Cool ing Page: 9 of 14 5.0 BASES 4.1.7 Technical Specifications 3.4.9, 3.4 .1 0, 3.9.8 and 3.9.9 require forced core circulation .

One method of forced circulation besides SOC , and the preferred option in Modes 3 and 4 , is RRC.I When the cavity is flooded up and natu ral circulation can be cred ited for core circulation per PPM 3.4.4, starting an RRC pump is not required .

Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-10 Examination Outline Cross-reference: 10 Revision: 2 Date: 1/18/21 Tier: 1 Group: 1 K/A Number: 295023.AK3.01 Level of Difficulty: 2 RO Importance Rating: 3.6 K/A

Description:

Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS :

Refueling floor evacuation CGS is in Mode 5.

A core shuffle is in progress.

As a fuel bundle is being removed from the core, it impacts a structural component.

• Gas bubbles are visible from the fuel bundle.

• There are no radiation monitors in alarm.

What action should be taken in accordance with ABN-FUEL-HAND?

A. Evacuate the refueling floor to minimize potential dose to personnel.

B. Perform a subcritical check to ensure inadvertent criticality is not occurring.

C. Immediately perform a detailed inspection of the fuel bundle to determine the extent of the damage.

D. Start the standby FPC pump to raise flow through the FPC Demineralizers and lower FP contamination.

Answer: A K/A Match:

Requires knowledge of the reasons for actions in ABN-FUEL-HAND, Damage While Handling Fuel.

Explanation:

A. Correct. One of the entry conditions for ABN-FUEL-HAND, Damage While Handling Fuel, is a fuel bundle impact with another object and visible gas bubbles. Action 4.1 is NOTIFY the Refuel Floor Supervisor/Spent Fuel Pool Supervisor to evacuate the refuel. In accordance with the ABN Bases, this step is performed to minimize the dose to personnel on the refueling floor.

B. Incorrect. Plausible since a subcritical check is performed if an unexpected SRM count rate is encountered during a fuel shuffle (see PPM 6.3.2, section 2.2) However, for the conditions given, ABN-FUEL-HAND should be entered and the refueling floor should be evacuated.

C. Incorrect. Plausible since a detailed inspection of the damaged fuel bundle should be performed in the future. However, this is not an action that should be performed immediately. ABN-FUEL-Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-10 HAND, directs evacuating the refueling floor as soon as possible to minimize personnel exposure.

D. Incorrect. Plausible since running both FPC pumps will increase flow through the FP Demineralizers. However, this action is not directed by ABN-FUEL-HAND.

Tier 1 Discussion Requires knowledge of AOP actions required. Meets Tier 1 requirements.

Technical Reference(s)

ABN-FUEL-HAND, Damage while Handling Fuel PPM-6.3.2, Fuel Shuffling and/or Offloading and Reloading Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8839 - Describe the operator action when fuel damage while refueling is indicated.

Question Source: C Bank #: Bank #

C Modified Bank #: Mod Bank #. (Note changes or attach parent)

C-- New Question History: Last NRC Exam: N/A Question Cognitive Level: r Memory or Fundamental Knowledge r

Comprehension or Analysis Justification for Cognitive Level Examinee must synthesize the conditions given in the question stem with a knowledge ABN actions and reasons for these actions.

10 CFR Part 55 Content: 55.41 13 Comments /

Reference:

ABN-FUEL-HAND Rev: Major: 004 Minor: 002 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-10 Number.: ABN-FUEL-HAN [) IUse category: CONT I uous Major Rev: 004 Minor Rev: 002

Title:

Damage W hHe Handling Fuel Page: 3of 5 LO

• Dropp ing1of a fu el l)undle

• Fuel bundle in pact with anoth er o )ject and visi )le gas bu )bfes

• Refuel1ng1bridge .JJrea radiation monitor al.urn ing (local only}

2.0 AlJTO MA.TIC ACTIONS Possible Secondary Containlillent isolation (Z signal} on Reactor Buifding lligll1rad iation.

3.0 IMMEDIATE OPERATOR ACTIONS None 4.0 SU BSEQUENT OPERATOR ACTIONS 4.11 NOTIFY th e RefOel'""'Floor Supervisor/Spent Fuel Pool Supervisor o evacuate the refu e oor .

5.0 BASES 4.1 Tnis ste!;l acua es tn e refuel floor as soon as possible , in ordelrto minimize the dose to ~ rsonnel on the floor.

Comments /

Reference:

PPM 6.3.2 Rev: Major: 025 Minor: N/A 2.2 Fu:el Shuffl e Source Ran ge Monitor (SRM) cou 111 rate readings maiy lJe reoorded in A *achment 8.4 ait the beginning of the fuel s huffl e sequence . During bundle insenions into core positions, SRM count rates s hould be monitored and ,c ompared to these reference values. Eaclh1bundle insertion will l)e neutrornically coupled to an SRMI. As til e s lluffle progresses, readings should be pelliodtcallly recorded and compared to til e original! va lues recorded the start. of til e sl'luffl'e. If an unex!lectecl clou511n_g m the average SRM coun rate occurs , or tNO u__O_Wec ea cloublings of any single SRM count rate occurs , e fuel shuffle process should be suspended at that point, and a subcrftical c eek s ould be performed o demonstrate adequa e margin o criticality exists. Tille subclitical check is performed by withdraMng ai lligh worth control rod to the full o ut IPOSiliorn , position 48) and verifying the core remains subclitical (PPM 6.3.3 Suboriticall Oheck}.

Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 Examination Outline Cross-reference: 11 Revision: 0 Date: 6/10/20 Tier: 1 Group: 1 K/A Number: 295024.EK3.04 Level of Difficulty: 2 RO Importance Rating: 3.7 K/A

Description:

Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE :

Emergency depressurization CGS is in Mode 1.

An event occurs causing Drywell pressure and Wetwell pressure to rise.

The CRS enters PPM 5.2.1, Primary Containment Control.

Based on Drywell and Wetwell pressure, when is an Emergency Depressurization (ED) required?

An ED is required when Wetwell pressure cannot be restored and maintained below A. Drywell Spray Initiation Limit (DSIL).

B. Pressure Suppression Pressure (PSP).

C. Decay Heat Removal Pressure (DHRP).

D. Primary Containment Pressure Limit (PCPL).

Answer: B K/A Match:

Requires knowledge of the reason for performing an emergency depressurization with high primary containment pressure.

Explanation:

A. Incorrect. Plausible since DSIL is used in the Primary Containment pressure leg to determine if drywell spray may be initiated. However, it is not used to determine if an ED is required.

B. Correct. In accordance with PPM 5.2.1, step P-12, when Wetwell pressure cannot be restored and maintained below PSP an ED is required.

C. Incorrect. Plausible since DHRP is determined by the difference in pressure between the RPV and primary containment and as Drywell and Wetwell pressure rise, the plant is closer to the DHRP limit. However, DHRP is not used to determine when an ED is required.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 D. Incorrect. Plausible since exceeding PCPL requires actions in the primary containment pressure leg of PPM 5.2.1. (step P-13). However, PCPL is not used to determine when an ED is required.

Exceeding PCPL requires venting primary containment.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: Learning Objective.

Question Source: C Bank #: Bank #

[' Modified Bank #: Mod Bank #. (Note changes or attach parent)

('.' New Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level: r. Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Requires knowledge of the conditions that require an emergency depressurization with high primary containment pressure.

10 CFR Part 55 Content: 55.41 5 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A PC PRESSURE PERFORM CONCUF P-1 MAINTAIN PC press ure below 1,68 psig using Primary Co nta inm ent Venting syste m, SO P- CN-CO NT-V ENT 1.68 psig C2:j

5. 14 NO

~-CONT-V ENT

,)

YES P-5 BEFORE 12 psi g WW p ress ure reac hes 12 psig (2 psig)

P.6 SPRAY lh e wetwell with sources llQl req uired for contin uous RPV inj ect ion Extern al spray sources may be used only if PC wa ter ~

level and wetwell press ure B PPM 5.5.2 ca n be res tored and maintained below PC PL ABN-TSG-008 P-7 12 psig Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 P-2 WHEN 1.68 psig PC pressure

~

=sll be NO

' EN T YES P-5 BEFORE 12 psig WW press ure reaches 12 psi g (2 psig)

P-6 SPRAY the wetwell 'vVith sources Jlill.

req uired for cont in uous RPV injection External Sj)fay sources may

~

l>e used only if PC water level and w etiNell press ure can be res!Of'ed and B PPM 5. 5.2 m aW,tained below PC PL AB N-T SG-008 12 psig Is drywelltell'I,)

below YE S DSIL Is WW level YES NO below 51 ft NO P-11 stop drywel 1 . STOP RRC pumps and drywell sprays cooling fans

2. 6 SPRAY the dry.veil with sources run required for continuous RPV iljection External spray sources may be used only if PC water ~

level and wetwe11 pressure can be restored and B

ma l"ltained below PCPL AB N-T SG-008 VENT PCtocontrolWWpre ssure ~

B Page 4 of 5 bebw PCPL, PPM 5.5.14 or PPM 5.5.1 5 (ED concurrence )

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 Comments /

Reference:

PPM 5.0.10 IRev: Major: 28 Minor: N/A

i. SteR P-12:

1) If Wetwell or drywell sprays cou ld not be started or if their operation was not effective in reducing primary conta inment pressure the RPV is depressurized to minimize further release of energy from the RPV to the pri mary containment when Wetwell Qressure cannot be res orea and mainta ined below the Pressure Suppression Pressure (PSP} Th is act7on serves to termi nate or reauce as much as possible any continued primary containment pressure increase.

Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-12 Examination Outline Cross-reference: 12 Revision: 1 Date: 12/28/20 Tier: 1 Group: 1 K/A Number: 295025.EA2.06 Level of Difficulty: 3 RO Importance Rating: 3.7 K/A

Description:

Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor water level CGS is in Mode 1.

The reactor automatically scrams on high RPV pressure.

Current plant conditions:

• Reactor pressure is 1065 psig, up slow.

• RPV level is +40 inches.

• Narrow Range RPV level indication is not available.

The crew is taking actions in accordance with PPM 5.1.1, RPV Control.

The CRS has directed CRO1 to Restore and Maintain RPV water level +13 inches to +54 inches in accordance with PPM 5.1.1, step L-3.

For current plant conditions, which of the following RPV level instruments will indicate the most accurate RPV level.

A. Shutdown (MS-LI-605)

B. Upset (RFW-LR-608)

C. Compensated Fuel Zone (MS-LR-615)

D. Uncompensated Fuel Zone (MS-LI-610)

Answer: B K/A Match:

Requires knowledge of reactor water level accuracy with high reactor pressure.

Explanation:

A. Incorrect. Plausible if it is believed that the Shutdown range is calibrated for higher reactor pressure. However, it is calibrated for 0 psig while other ranges are calibrated for 1000 psig.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-12 B. Correct. The Upset Range is calibrated for 1000 psig, while the other distractors are calibrated for 0 psig.

C. Incorrect. Plausible since this indication is density compensated and is valid for all pressure ranges. However, it is calibrated for 0 psig.

D. Incorrect. Plausible if it is believed that the Shutdown range is calibrated for higher reactor pressure. However, it is calibrated for 0 psig while other ranges are calibrated for 1000 psig.

Tier 1 Discussion Requires knowledge of which RPV level indicator to use while performing EOP actions. Meets Tier 1 requirements.

Technical Reference(s)

PPM 5.1.1, RPV Control SD000126, Nuclear Boiler Instrumentation System Description Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5582 - List the calibration conditions and nominal ranges for each of the five ranges of level instruments.

a. NARROW RANGE

c. WIDE RANGE

d. UPSET RANGE

e. SHUTDOWN RANGE

f. FUEL ZONE RANGE Question Source: r Bank #: Bank #

r Modified Bank #: Mod Bank #. (Note changes or attach parent)

r. New Question History: Last NRC Exam: N/A Question Cognitive Level: r. Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Requires knowledge of RPV level instrumentation calibration conditions.

10 CFR Part 55 Content: 55.41 7 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-12 Comments /

Reference:

PPM 5.1.1 IRev: 22 RPVLevel

~

1 IL-1 t

~

-ti- 13 In .

ENSURE ALL.a ctuatio ns ~50 in .

-129 in.

1.68 psig IL-3 RESTORE and MAINTAIN RPV level +13 in . to +54 in .

8.n with ~ tnj ection systems, Ta ble 1 and , if re,q ui l'ed ,

with Altern ate tnjec Uon Systems , Table 3 I

Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-12 Comments /

Reference:

SD000126 IRev: Major: 13 Minor: 2

1. Narrow range level instruments (Fig 2) indicate over a range of 0 " to 60 "

and are calibrated to be accurate under the following conditions : LO-5582a Reactor pressure: 1000 psig Containment temperature: 135°F Reactor Building temperature : 75 °F

2. Wide range level instruments (Fig 2A) indicate over a range of -150" to LO-5582b

+60 " and are calibrated to be accurate under the following conditions :

Reactor pressure: 1000 psig Containment temperature: 135 °F Reactor Building temperature: 75 °F

3. Upset range level instruments (Fig 2B) indicate over a range of 0 " to 180" and are calibrated to be accurate under the following conditions: LO-5582c Reactor pressure: 1000 psig Containment temperature: 135 °F Reactor Building temperature: 75 °F

4. Shutdown range level instruments (Fig 2C) indicate over a range of O " to 400 " and are calibrated to be accurate under the following conditions: LO-5582d Reactor pressure/temperature: 0 psig , 212 °F Containment temperature: 80 °F Reactor Building temperature: 75 °F

5. Fuel zone range level instruments [compensated and uncompensated] (Fig LO-5582e 2D) indicate over a range of -310 " to -110" and are calibrated to be accurate under the following conditions:

Reactor pressure/temperature: 0 psig , 212 °F Containment temperature: 212 °F Reactor Building temperature: 75 °F Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13 Examination Outline Cross-reference: 13 Revision: 1 Date: 1/5/21 Tier: 1 Group: 1 K/A Number: 295026.EA2.03 Level of Difficulty: 2 RO Importance Rating: 3.9 K/A

Description:

Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor pressure CGS is in Mode 1.

The CRS has entered PPM 5.2.1 due to high Wetwell temperature.

Current conditions:

• Wetwell temperature: 210°F.

• Wetwell level: 22 feet.

HCTL Hea t Capacity Temp Limit LE 39 FT Wetwe ll Le vel

---

------ ------- ------- --- UNSAFE ----------

280 >----1/4-----+-----+-------+----+------+----+

275 ' '

- - - - - - "T - - - - - - - '

260

---

. ..,' . . - - . - - -------r- ' - -- - -- ----*--

~ --- 39

~ 220 1-+-:--+--+--~~-~=-+-""""'1-.:::--+----"""'"""""'=-+--""'1 E

Cl) i 200 f---r-:--+--+--+----+--+---=""""-::-------J---+-----"'""'---t------1 2 7 Y::JJ1::J..

L.e.llel S: --,----- ------ ------- ------- 23 .(ft) 160 t----.-:----+-----+-------+----+------+--------<

14 0 t--:- - - + - - - - - - + - -' '- + - - - ' -'- - + - - ~ - t - - - - - - 1 120  :  :  :

40 200 400 600 800 1000 109 1 RPV Pressure (psig)

What is the lowest RPV pressure which would result in exceeding HCTL?

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13 A. 400 psig B. 500 psig C. 650 psig D. 700 psig Answer: B K/A Match:

Requires ability to interpret reactor pressure with respect to HCTL with high suppression pool temperature.

Explanation:

A. Incorrect. Plausible if it is believed that the unsafe region is below the line. However, the unsafe region is above the line.

B. Correct. Since the given Wetwell level falls between the lines given on the HCTL graph, the most conservative line (i.e. lowest Wetwell level) should be used. With Wetwell temperature at 210°F, the 19.2 foot line intersects at approximately 490 psig. Therefore, 500 psig is the lowest RPV pressure given that exceeds HCTL.

C. Incorrect. Plausible since 650 psig is the value of RPV pressure if attempting to interpolate the value of RPV pressure for a Wetwell level of 22 feet. However, PPM 5.0.10 states When using either HCTL graphs, if WW level is not on one of the lines on the graph, use the next line below (move in the direction to make the unsafe area larger) to determine HCTL.

D. Incorrect. Plausible if the 23 foot line is used. However, the most conservative line (19.2 feet) should be used.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8302 - Given plant conditions and the Heat Capacity Temperature Limit Curve, determine the current operating point on the Curve within 2.5 degrees and 25 psig.

Question Source: Bank #: LO02944 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13

(' Modified Bank #: Mod Bank #. (Note changes or attach parent)

(' New Question History: Last NRC Exam: 2009 Question Cognitive Level: r- Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Requires a knowledge of the usage rules for the HCTL graph and the ability to read the HCTL graph.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A WW TEMP W T-1 MAINTAIN WW temp below 90°f with available WW cooling WT-2 90°F be maintained below 90°F W T-3 MAXIMIZE WW cooling with RHR pumps D.C.1 required for continuous RPV injection WT-4 BEFORE WW temp reaches 110°f Concurrently PPM 5.1 .1 0

W T-5 IF WW temp reaches .a.a.:£ pump NPSH limit Tab le 18 THEN stop and preven t operation of that pump W required to:

• assu re adequate core cooling OR

• prevent PC failure (Notify T SC for assessment of pump NPSH limits)

DT-6 BEFORE dry dro Op Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13 HCTL Hea t Capacity Temp Limit LE 39 FT Wetwel l Level UNSAFE 280 1-+---+----f------'---+---'-------,-----'---+----+

275

~

39 ta. 220 1--+---+-+--t""""".:-l----=:-...::-t-~"!o;;;;;::-t--=--....=-t--=""'1 35 E ----- ...... 31 a,

l-s: 200 f--;----------r-+--1---+--+-------"'t,,,.---+--+-""""1c---+-------l 2 7 '1:£1::L l..e.l£eJ.

s: 23 00

19.2 160 f--,-----+----+-------lf------+--~-+-----1 140 t-+-----+----+-------lf------+--~-+-----1 120 '---------"-----'---'--'---'----'-----'--'

40 200 400 600 800 1000 1091 RPV Pressure (psig)

Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13 Comments /

Reference:

PPM 5.0.10 IRev: Major: 23 Minor: N/A 7.3.9 When using either HCTL graphs, if WW level is not on one of the lines on the graph ,

use the next line below (move in the direction to make the unsafe area larger) to determine HCTL.

7 .3.10 The HCTL is referenced in PPM 5.1.1, PPM 5.1.2 and PPM 5.2.1 with the following identifier:

Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-14 Examination Outline Cross-reference: 14 Revision: 1 Date: 1/11/21 Tier: 1 Group: 1 K/A Number: 295028.2.4.21 Level of Difficulty: 2 RO Importance Rating: 4.0 K/A

Description:

High drywell temperature: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

CGS is in Mode 1.

An event causes a high drywell temperature condition.

The CRS enters PPM 5.2.1, Primary Containment Control.

In accordance with PPM 5.2.1, when is the crew required to perform an Emergency Depressurization (ED) and what is the reason for this requirement?

An ED is required to be performed when drywell temperature cannot be restored and maintained below (1) . This is performed since containment parameters cannot be maintained below (2) design temperature.

A. (1) 330°F (2) Drywell B. (1) 330°F (2) Automatic Depressurization System (ADS)

C. (1) 340°F (2) Drywell D. (1) 340°F (2) Automatic Depressurization System (ADS)

Answer: B K/A Match:

Requires knowledge of parameter that requires action to maintain containment system safety function.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-14 Explanation:

A. Incorrect. Plausible since 330°F is the correct temperature limit. However, in accordance with PPM 5.0.10, section 8.9.5.i, an ED is performed when drywell temperature cannot be maintained below the ADS design temperature.

B. Correct. In accordance with PPM 5.2.1, step DT-8, an ED is required when drywell temperature cannot be restored and maintained below 330°F. PPM 5.0.10, Flowchart Training manual states that this ED is performed when drywell temperature cannot be restored and maintained below the ADS design temperature.

C. Incorrect. Plausible since 340°F is the design temperature for the drywell. However, the ADS design temperature is more restrictive and PPM 5.2.1 requires an ED when drywell temperature cannot be restored and maintained below 330°F.

D. Incorrect. Plausible since the reason for performing an ED due to drywell temperature is correct.

Plausibility is enhanced since 340°F is the design temperature of the drywell. However, PPM 5.2.1 requires an ED when drywell temperature cannot be restored and maintained below 330°F.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

See Comments / Reference SD000127, Primary Containment System Description Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8318 - Given a list, identify the statement that describes the reason for attempting to maintain drywell temperature below 330 °F.

Question Source: OBank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

New Question History: Last NRC Exam: N/A Question Cognitive Level: 0 Memory or Fundamental Knowledge

@, Comprehension or Analysis Justification for Cognitive Level Requires knowledge of the requirements for performing an ED on high drywell temperature and the reason for this requirement.

10 CFR Part 55 Content: 55.41 10 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-14 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A DRYWELL TEMP MAINTAIN drywell temp below 1JS"f PC venting may be use~

with ava ffable drywe n cooling Restore and maintair Lowering RPV an, discharge pressur Di sc harging RCIC Red uce total off site r If PC integrity has 135°F If sig nifica nt fue l c Scrubbing dis.char If further degrada1 personnel resourc I

IF PC pressure BE FORE is required to drywell temp reaches 3300f 330°F and m aintain (28 5"F) core cooling ,

the total offsil Concurrently radiation dos PPM 5.1.1 0 BEFORE WW p DT-4 drops Is 0 psi g drywell temp below YES DSIL DT-5 Is WW level YES below NO 5111 NO

>ress ure THEN stop d rywell 1.STOP RRC pumps a nd dryY-Jell

!IDW sprays coolin g fans 1.68 ps ig) 2. 6 SPRAY the d rywell with sources Q.C!. req ui red for continuous R PV injection External spray sources may be used onl y if PC water revel and wetwell pressure

[;2]

B can be restored and maintained below PCPL ABN-TSG-008 W REN ~

330°F Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-14 Comments /

Reference:

PPM 5.0.10 IRev: Major: 022 Minor: 000

i. Step DT-8 :

1) When drywell temperature cannot be restored and maintained below the ADS design temperature, emergency RPV depressurization is performed.

Th is action minimizes any continuing direct energy release to the drywell through a primary system break and ensures that the SRVs are opened while still operable.

Comments /

Reference:

SD000127 Rev: Major: 16 Minor: 2 COLUMBIA SYSTEMS September 2019 PRllMARY CONTAINJMENT SD000127 r16 mr2 DIC0023.98 B. Primary Containment V:essel The primary c.outaiinm.ent \i

• set is a free standiing steel pressure vessel. It uti.l.izes the pressure suppression

• ochnique through fhe ' lark Il over/under configuration". It is destgned to Fesist all normal operating loads, oads Fesulting from the postulated DBA, as ivell as loads Fe.sult.ing from the destgn earthquake. The destgn also accounts.for stresses.induoed byfhermal expansion. (Figure 1)

The Pfin.lary£_oillainmenfis aes1ed for the follov.ring loads L0-5615

• Internal pteSSit:Fe 4Spsig

• Extem.1il. pressrue 2p *g

• Temperarure 340"'F - Drywell 275 <'f' - Suppress~on Chamber

• Drywell floor dp 2S psid (dO\vnward) 6.4 psid (upward)

Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-15 Examination Outline Cross-reference: 15 Revision: 2 Date: 1/14/21 Tier: 1 Group: 1 K/A Number: 295030.EK1.03 Level of Difficulty: 3 RO Importance Rating: 3.8 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Heat capacity CGS is in Mode 1.

An earthquake occurs causing wetwell level to lower.

The CRS enters PPM 5.2.1, Primary Containment Control.

As wetwell level continues to lower, the CRS enters PPM 5.1.1, RPV Control, and a manual reactor scram is performed.

• Drywell pressure is 0.6 psig and steady.

• Wetwell pressure is 0.1 psig and steady.

• Wetwell temperature is 70°F and steady.

• Wetwell level is 19 feet, 1 inch, down fast.

(1) What action is required?

(2) What is the basis for this action?

The crew must perform an (1) . This action is performed (2) .

A. (1) anticipated RPV depressurization per PPM 5.1.1, RPV Control (2) to ensure equipment in the wetwell necessary for safe shutdown will remain operational B. (1) anticipated RPV depressurization per PPM 5.1.1, RPV Control (2) since the RPV is not allowed to be at pressure when pressure suppression capability is unavailable C. (1) emergency depressurization per PPM 5.1.3, Emergency RPV Depressurization (2) to ensure equipment in the wetwell necessary for safe shutdown will remain operational D. (1) emergency depressurization per PPM 5.1.3, Emergency RPV Depressurization (2) since the RPV is not allowed to be at pressure when pressure suppression capability is unavailable Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-15 Answer: D K/A Match:

Requires knowledge of the operational requirements on a loss or potential loss of heat capacity due to low suppression pool level.

Explanation:

A. Incorrect. Plausible since an ED is required to ensure wetwell equipment remains operational due to wetwell temperature. However, for low wetwell level, an ED is required due to the imminent loss of the pressure suppression function.

B. Incorrect. Plausible since the required action is due to the potential loss of the pressure suppression capability for an emergency depressurization (ED). However, an ED is required since wetwell level is below 19 feet, 2 inches.

C. Incorrect. Plausible since an ED is required. However, an ED is required due to the imminent loss of the pressure suppression function.

D. Correct. In accordance with PPM 5.2.1, Primary Containment Control, an ED is required when wetwell level cannot be maintained above 19 feet, 2 inches. PPM 5.0.10, Flowchart Training Manual states that the reason for ED is that the RPV may not remain at pressure when pressure suppression capability is unavailable.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11150 - Given plant conditions and EOP flowcharts, evaluate plant conditions and determine the appropriate actions according to EOP 5.2.1.

Question Source: Bank #: LO03417 0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

ONew Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-15 Question History: Last NRC Exam: N/A Question Cognitive Level: 0 Memory or Fundamental Knowledge

(.7, Comprehension or Analysis Justification for Cognitive Level Requires synthesizing information in question stem with a knowledge of the actions required along with the reason for performing these actions.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

PPM 5.1.2 Rev: Major: 28 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-15 L-4 MAINTAIN WW level above 19 ft . 2 in .

L-5 BEFORE WW level drops to 19 ft 2 in.

Concurrently PPM 5.1.1 0

L-6 WHEN WW Ieve I ~

be maintained above 19 ft 2 in .

EMERG DEPRESS REQ'D L-7 WHEN WW level ~ be maintained above amt system vortex limit Table 18 L-8 Stop and prevent operation of that system D.QJ. requ ired to:

• assure adequate core cooling OR

• preven t PC failure Comments /

Reference:

PPM 5.0.10 Rev: Major: 22 Minor: 000 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-15 Number: 5.0.10 I Use Category: INFORMATION Major Rev: 022 Minor Rev: 000

Title:

Flowchart Training Manual Page: 321 of 370 depressurization with the reactor at power should therefore be avoided .

• The fourth IF!THEN of PPM 5.1.1 Override P-1 perm its rapid depressurization through the main turbine bypass valves in anticipation of emergency RPV depressurization.

3) Entry to RPV Control is explicitly stated because cond itions requiring entry into this flowchart do not necessarily require entry into RPV Control.

Therefore , a scram may not have yet been initiated. Directing that RPV Control be entered, rather than explicitly stating here "initiate a reactor scram" coordinates actions currently being executed if RPV Control has already been entered. In addition, entry to RPV Control must be made because it is through that flowchart that the transfer to PPM 5.1.3, Emergency RPV Depressurization , (or PPM 5.1 .5 for ATWS events) is effected .

f. Step L-6

1) Wetwell level must be maintained above the elevation of the downcomer vent open ings to ensure that steam discharged from the drywell into the Wetwell following a primary system break will be adequately condensed .

(Results of the Bodega Bay Mark I containment tests indicate 95% steam condensation may be ach ieved from a vertical downcomer vent that discharges at a level six inches above the suppression pool surface.) If Wetwell level cannot be maintained above the specified minimum value ,

steam may not be adequately condensed and primary containment pressure could exceed allowable limits. Since the RPV may not be kept at pressure when pressure suppression capability is unavailable, Emergency RPV Depressurization is required.

Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Examination Outline Cross-reference: 16 Revision: 2 Date: 1/18/21 Tier: 1 Group: 1 K/A Number: 295031.EK1.02 Level of Difficulty: 4 RO Importance Rating: 3.8 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: Natural circulation CGS is in Mode 1.

A LOCA occurs.

• A manual reactor scram is inserted.

• Control rods fail to insert.

The crew is taking actions in accordance with PPM 5.1.2, RPV Control - ATWS.

• The crew is controlling level -80 inches to -140 inches.

..L_-6_ _ _ _ _ _ _ ____.__- _ _-_ _-_ _-_ _-_ _-...., - l in Main Turbine trip lE RPV level is above -65 in. Stop & preven 0

Itlfl:i lower RPV level to below -65 in. by stopping and preventing all injection into RP V except 0

Cond/RF HPCS from boron injection systems , RCIC and CRD, defeatin g interlocks if necessary 0 LPCS r RHR-A I RHR-B 0 RHR-C YES conditions L-8 exist

~

Lower RPV level by stopping and preventing injection into RPV except from boron injection NO systems , RC IC and CRD, defeating interlocks if necessary, until ~

Leveli'Power

• RPV level reaches - 161 in.

• reactor power drops below 5%

• drywell pressure remains below 1.68 psig and all SRVs rem ain closed At step L-7, the CRS determines that Level/Power conditions exist and directs lowering RPV level.

Why is RPV level lowered at step L-8 of the procedure?

RPV level is lowered to reduce core Page 1 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 A. void fraction.

B. boron mixing.

C. level oscillations.

D. natural circulation flow.

Answer: D K/A Match:

Requires understanding of the effect on natural circulation flow when RPV level is lowered.

Explanation:

A. Incorrect. Plausible since void fraction is changed when RPV level is lowered. However, void fraction is increased to lower reactor power.

B. Incorrect. Plausible since actions to promote boron mixing should be taken to ensure that the entire core will shut down. However, lowering RPV level during Level/Power conditions is performed to reduce natural circulation flow, increase void fraction, and reduce reactor power.

Boron stratification is increased when natural circulation flow is lowered. Additionally, boron mixing should be promoted to ensure that the entire core feels the effects of boron addition.

C. Incorrect. Plausible since the initial reduction of RPV level in step L-6 is performed to reduce core inlet subcooling and reduce neutron flux oscillations caused by thermal-hydraulic instabilities.

However, at step L-7, level reduction is performed to reduce natural circulation flow. Additionally, PPM 5.0.10 states that reducing RPV level significantly below normal level may cause level and power oscillations.

D. Correct. If Level/Power conditions exist, heat could be rejected to the wetwell at a rate in excess of that which can be removed by the wetwell cooling system. This could result in containment over pressurization and loss of containment. Lowering RPV level reduces natural circulation flow.

This reduces the amount of steam that is removed from the RPV, which increases void fraction and adds negative reactivity to the core. This will cause reactor power to lower and reduce the amount of energy that is rejected to containment.

Tier 1 discussion Question requires knowledge of reason for performing EOP step. Meets Tier 1 requirements.

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual PPM 5.1.2, RPV Control - ATWS Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Page 2 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Proposed references to be provided during examination: N/A Learning Objective: 8120 - Given a list, identify the statements that describe the purpose of maintaining RPV water level between the Minimum Steam Cooling RPV Water Level and either +54 inches or LL following emergency depressurization during an ATWS.

Question Source: OBank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

(i:, New Question History: Last NRC Exam: N/A Question Cognitive Level: , Memory or Fundamental Knowledge 0 Comprehension or Analysis Justification for Cognitive Level Requires examinee to synthesize information given in the question stem with a knowledge of 10 CFR Part 55 Content: 55.41 1 Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A Page 3 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 LE 5%

5% or unknown V RCIC injectio n wi ll result in Main Turbine trip IE. RPV leve l is above -65 in . Sto p I Ilifl:i lower RPV leve l to below -65 in. by stopping ------------;

and preven ting i.11. injection into RPV exce pt from bo ro n injection syst ems, RCIC and CRD, defeati ng interlocks if necessary r

I

, .13

..el)

YES oond ~io ns L-ll exist Lower RPV level by stop pi ng and preventing injection into RPV except from boro n injection NO systems, RC IC and CRD, defeati ng interlocks if necessary, unt il .6t:£(;

u,

• RPV level reaches -161 in.

5.5.28

• reacto r powe r d rops be low 5%

transfer)

• drywell pressure femains below 1.68 psig and aJL SRVs remain closed L-10 t-11 RECORD LL

-65 in.

Rapid injection may ca use fue l damage Rapid injectio n may ca use fuel damage L-12 L-13 Maintain RPV leve l -186 in. to -65 in. with Outside Maintain RPV level -186 in. to LL with Outside Shroud Injection Syste ms, Table 5 Sh roud Injection Sy.,tems, Table 5, (begi n injecting at LL ) (begin injecting at LL)

Best-practice co ntrol band : -1 40 in. to -BO in.

a bove G

-186 in.

Comments /

Reference:

PPM 5.0.10 Rev: Major: 023 Minor: N/A Page 4 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Number: 5.0.10 I Use Category: INFORMATION Major Rev: 023 Minor Rev: N/A

Title:

Flowchart Training Manual Page : 141 of370 and equipment dictate the relative importance of individual RPV Control steps and the re lative priority with which they shou ld be accomplished .

8.3.4 Level Control

a. The actions specified in PPM 5.1.2 effect control of RPV level under conditions when it cannot be determined that control rod insertion alone wi ll assure that the reactor wi ll remain shutdown under all conditions. The actions to control RPV level in PPM 5.1 .2 differ from those in the RPV level control section of PPM 5.1 .1 to address these basic concerns:

• When boron is injected into the RPV, the systems used for control of RPV water level must be operated to minimize boron dilution and cold water injection and to promote boron mixing .

• If the reactor is not shutdown, RPV level is lowered and controlled below the feedwater spargers to minimize core inlet subcooling , thereby preventing or mitigating the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal -hydraulic instabilities .

• If the reactor cannot be shutdown and Wetwell temperatu re continues to rise , RPV water level must be controlled not on ly to cool the core , but also to minimize Wetwell heatup by minimizing power generation.

• Even if boron has not been injected into the RPV and the reactor is shutdown on control rods under hot conditions, injection of cold water could cause critica lity without negative reactivity feedback occurring until reactor power reaches the heating range .

b. Where necessary, PPM 5.1.2 specifies unique actions to address these concerns . Some of these actions include:

• Reducing core inlet subcooling by deliberately lowering RPV water level sufficiently below the feedwater spargers.

• Reducing reactor power by deliberately lowering RPV water level.

• Ass igning a priority to the use of injection systems .

• Prolonging availabil ity of the main condenser as a heat sink by defeating low RPV water level and high steam tunnel temperature MS IV isolation interlocks.

• Controlling injection of cold water.

• Delaying RPV cooldown until reactor shutdown has been assured by either boron injection or control rod insertion .

Page 5 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 I. Step L-7

1) Level / Power cond itions exist when all of the following are meet:

• RPV water level above -161 in .

• reactor power above 5% OR ca nnot be determined

• WW tempe rature is above 110°F

• drywell pressure above 1.68 psig or any SRV open

2) Level / Power cond itions are listed in Override L-9 (3 rd IF/THEN ). If they exist , Step L-8 gives direction to lower RPV level.

3) The combination of high reactor power (above the APRM downscale tripJ 5%), high Wetwell temperatu re {above the Boron Injection Initiation Temperature , 110°F), and an open SRV or high drywell pressure (above the scram setpoint, 1.68 psig ), are sym 12_toms of heat being rejected to the Wetwell at a rate in excess of that which can be removed by the Wetwell cooling system . Unless mitigated by boron injection or control rod insertion ,

these conditions can result in containment over pressurization and loss of primary containment integrity leading to loss of adequate core cool ing and uncontrolled release of radioactivity to the environment.

Page 6 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16

m. Step L-8

1) Direction is given to lower RPV level by stopping and preventing all injection into the RPV except from boron injection , RC IC and CRD ,

defeating interlocks as necessary until any of the following are meet:

• RPV leve l drops to -161" OR

• reactor power drops below 5%

OR

• drywell pressure remains below 1.68 psig and all SRVs remain closed

2) The direction to "lower RPV level by stopping and preventing all injection into RPV except from boron injection systems, RC IC, and CRD" is not intended to imply that injection from these systems must continue . Rather, the exception simply excludes these systems from actions taken to terminate and prevent injection in th is step, thereby avoiding conflicts with other instructions and permitting continued operation if necessary to accomp lish the objectives of other EOP steps.

3) To facilitate RPV level reduction in an expeditious and contro lled manner, any interlocks which may prevent stopping and preventing injection into the RPV are allowed to be defeated.

4) Further lowering RPV water level reduces natural circu lation driving head and core flow, thereby reducing reactor P,ower and the heat rate to the Wetwell. Th is process occurs as follows:

Page 7 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16

• • The reactor is in a na ur circu fation mode following automatic reactor recIrculatIon pump trip on low RPV water level. f\la ural circula ion cln~ -ng neaa Is a function of tn e flura aensity difference Detween tn e E:!:Q ions inside an outsicle of he sfi roucl {voicl fraction directly affects l he fluia density Insicle the shroud) and the heig of the fluid columns (RPV water level).

• As RPvwateneve Is owe f ffie Hui co umns 1s red uced , thereby reducing ation cfri -ng head.

• As tne natural circu lation drivmg head is reclucecl tne natural c1rcula -on flow througfi e core Is reduced .

• The red uced core flow results In a reclu cecl rate of steam removal from the core_

• The reduced rate of steam removal results in an increased void fraction inside tn e s roucl_

• The increased void fraction adds negative reactivity o the reactor_

• The a -ve reactivity clnves tn e reac or sligfiUy su6critical an power

)egins o decrease_

• The red uce reacto:r power results in a re uced steam generation ra e_

• The reduced s earn generation rate results in a reduced voitrfrac -on.

• v,,rnen e void fraction ro~s o o:riginal va ue (with some slignt adjustment to account for red uced Doppler reactivity), e reacto:r return s o criticality a a rower power leveL

5) ltle above process (fllustrated in Figmes 8 .3-11, 8 .3-2, and 8-3-3), has been observed i111o;peratirng BWRs wi:tlh RPV water level in or near til e nomial operating1range_Com puter analyses and scale model tests have con rmed til e continued validity of these thermal hydrau ~c a11 d reactor physics pnnciptes tor RPV water levels to below U1.e elewtiorn of the steam separators..

6) Level and power oscilla -ans may occur e,n RPVwater lever islowered significantry below th e nom1al operating range with the reactor s -1 at power . [These oscillations have been arnaliyzedl an d determined to resul in tll errnal trallisie11ts welll witlil in the design capabilities of the 1fu eL Page 8 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-17 Examination Outline Cross-reference: 17 Revision: 0 Date: 6/22/20 Tier: 1 Group: 1 K/A Number: 295037.EK3.06 Level of Difficulty: 3 RO Importance Rating: 3.8 K/A

Description:

Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Maintaining heat sinks external to the containment CGS is in Mode 1.

An event occurs that requires a reactor scram.

• Control rods fail to insert.

The crew is taking actions in accordance with PPM 5.1.2, RPV Control - ATWS.

Plant conditions:

• SLC has failed to inject.

• Reactor power is 15% and stable.

• Outboard MSIVs automatically closed due to a loss of CAS pressure.

CAS pressure to MSIVs has been restored.

The CRS directs re-opening the MSIVs.

With the current plant conditions, what limit will first be exceeded if MSIVs are not opened?

MSIVs are opened to discharge heat energy to the main condenser to prevent exceeding...

A. SRV Tail Pipe Level Limit (SRVTPLL).

B. Pressure Suppression Pressure (PSP).

C. Heat Capacity Temperature Limit (HCTL).

D. Primary Containment Pressure Limit (PCPL).

Answer: C K/A Match:

Requires knowledge of reason for restoring an external heat sink during an ATWS Explanation:

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-17 A. Incorrect. Plausible if it is believed that wetwell level will be the most limiting parameter for the conditions given in the stem. However, for the conditions given, an external heat sink is required to maintain the wetwell below HCTL.

B. Incorrect. Plausible since adding heat to primary containment may cause containment pressure to rise. However, for the conditions given, an external heat sink is required to maintain the wetwell below HCTL.

C. Correct. In accordance with PPM 5.1.2, RPV Control - ATWS, override P-4, MSIVs should be opened when boron injection is required. PPM 5.0.10 gives the reason for this: To stabilize and control RPV pressure, the reactor steam generation rate must remain within the capacity of systems designed to remove the steam from the RPV. With the reactor not shutdown, the amount of steam that may have to be released could be substantial. If this heat energy is discharged to the Wetwell, the HCTL could be reached in a very short time. Therefore, utilization of the main condenser as a heat sink for this energy is of sufficient importance to warrant opening the MSIVs D. Incorrect. Plausible since maintaining primary containment is the goal of opening the MSIVs.

However, HCTL will be exceeded first.

Technical Reference(s)

PPM 5.1.2, RPV Control - ATWS PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8169 - Describe why re-opening MSIVs when boron injection is required, is beneficial during an ATWS.

Question Source: OBank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

@:iNew Question History: Last NRC Exam: N/A Question Cognitive Level: (i:, Memory or Fundamental Knowledge 0 Comprehension or Analysis Justification for Cognitive Level Requires knowledge of reason for an action in the EOPs.

10 CFR Part 55 Content: 55.41 10 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-17 Comments /

Reference:

PPM 5.1.2 IRev: Major: 26 Minor: N/A P-4 IF WW tern p can not THEN only if RPV depressurization will .llQi resu lt in loss of injection required for be maintained adequa te core cooling:

below HCTL maintain RPV pressure below HCTL (exceeding 100 'F /hr cooldown rate if necessary)

IF WW level cannot THEN only if RPV depressurization wi ll D.W, be maintained result in loss of injection required fo r adequate core cooling:

below SRVTPL L maintain RPV pre ssure bel ow SRVTPL L (e xce eding 100

• F/hr cooldown rate if necessary)

IF BORON INJECT REQ'D THEN open MS IVs AND main conden ser operable.Table 8 (defeat low RPV level and high steam tunnel temp M SIV isolation interlocks if necessary) n PPM 5.5.7 Comments /

Reference:

PPM 5.0.10 Rev: Major: 023 Minor: N/A Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-17 Number: 5.0.10 I Use Category: INFORMATION Major Rev: 023 Minor Rev: N/A

Title:

Flowcha rt Training Manua l Page : 172 of 370

3) Third IF/THEN:

a) To stabilize and control RPV pressure, the reactor steam generation rate must remain within the capacity of systems designed to remove the steam from the RPV. With the reactor not shutdown, the amount of steam that may have to be released could be substantial. If this heat energy is discharged to the Wetwell the HCTL could be reached in a very short time . Therefore , utilization of the main condenser as a heat sink for this energy is of sufficient importance to warrant open ing the MS IVs even if the valves automatically closed . Such action may be the principal contributor to successful mitigation of a failure-to-scram condition .

Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-18 Examination Outline Cross-reference: 18 Revision: 2 Date: 1/5/21 Tier: 1 Group: 1 K/A Number: 295038.EA1.01 Level of Difficulty: 3 SRO Importance Rating: 3.9 K/A

Description:

Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE: Stack-gas monitoring system.

CGS is in Mode 1.

An event causes the following readings on the Reactor Building Exhaust Plenum Radiation Monitors:

ED REA-RIS-609A: 11 mr/hr, steady REA-RIS-609B: 11 mr/hr, steady REA-RIS-609C: 14 mr/hr, steady ET

• REA-RIS 609D: Downscale The CRS enters PPM 5.3.1, Secondary Containment Control.

EL D

With respect to Reactor Building Ventilation, what should the operator verify?

In accordance with PPM 5.3.1, verify ROA-V-1, Reactor Building Supply Inboard Isolation, is (1) , and ROA-V-2, Reactor Building Supply Outboard Isolation, is (2) .

N A. (1) closed (2) closed TI B. (1) closed O

(2) open ES C. (1) open (2) closed D. (1) open (2) open U

QAnswer: B K/A Match:

Requires knowledge of the expected outcome for stack gas monitoring system conditions when verifying equipment in accordance with PPM 5.3.1, Secondary Containment Control.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-18 Explanation:

A. Incorrect. Plausible since ROA-V-1 will be closed. However, ROA-V-2 will remain open B. Correct. REA-RIS-609C & D will provide a close signal to the RB HVAC inboard isolation valves when both monitors are > Hi-Hi level (13 mr/hr), Downscale, or a combination of either condition.

However, these monitors provide input to the inboard valves.

ED C. Incorrect. Plausible if it is believed that REA-RIS-609C & D operate the outboard valves.

D. Incorrect. Plausible since ROA-V-2 will be open. Plausibility is enhanced if it is believed that a monitor downscale will not provide a close signal. However, for the conditions given, ROA-V-1 will close.

ET Tier 1 Discussion: This question requires knowledge of the expected system response while performing step SC-1 of EOP PPM 5.2.1, Secondary Containment Control.

Technical Reference(s)

SD000147, Process Radiation Monitoring PPM 5.3.1, Secondary Containment Control EL Attached w/ Revision #

See Comments / Reference D

Proposed references to be provided during examination: N/A N

Learning Objective: 5647 State the automatic actions associated with each of the following gaseous and liquid stream Process Radiation Monitors upon sensing high radiation levels: g. Reactor Building Exhaust Plenum RMS Question Source:

TI O CBank #:

(Note changes or attach parent)

C Modified Bank #:

New ES Question History:

Question Cognitive Level:

Last NRC Exam: N/A C Memory or Fundamental Knowledge

@, Comprehension or Analysis UJustification for Cognitive Level QExaminee must demonstrate a knowledge of the conditions required to isolate RB HVAC along with a knowledge of the isolation signals provided by individual radiation monitors.

10 CFR Part 55 Content: 55.41 11 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-18 Comments /

Reference:

SD000147 IRev: Major: 15 Minor: 001

2. Reactor Building Exhaust Plenum RMS (Figure 2)(REA-RIS -609A/B/C/D) a) The range of this RMS is .01 - 102 mR/hr.

b) This radiation monitor has two local alarms as follows:

• LOW (white)

Instrument INOP indication that activates due to any of:

Indication LT .0 1 mR/hr.

Loss of power to the RMS.

ED Control Switch "OUT-OF-OPERATE".

HIGH (red)

ET If sensed on both REA-RIS-609A & B, or C & D causes alarm on H13-P602 AND a "Z " signal trip.

c) RESET Pushbutton Depress to reset alarms .

d) Monitor Function Switch EL D

3-position switch , TRIP TEST /ZERO/OPERA TE

• OPERATE - Monitor operates normally (other positions are for testing; ZERO position generates a trip signal that can be used to LO-5647g N

trip the channel when required by Technical Specifications) .

e) A HI-Hi trip for channels A & B initiates an alarm on H13 -P602 and a O

Grou III containment isolation signal [i.e., closure of the Reactor Building, Containment and Radwaste Building ventilation outboard isolation valves , and initiates startup of Standby Gas Treatment (SGT)

TI Train A Lead Fan and Train B Lag Fan]. Logic is "TWO-OUT-OF-TWO " (trip of both channels in either division are required to cause the isolation. )

ES f) A downscale condition (LE 0.01 mR/hr) sensed on both channels A &

B, causes alarm on Hf3-P602 AND a "Z" signal trip as described in "e" above.

g) The same condition for Channels C & D initiates closure of the corresponding inboara valves and initiates startup of SGT Train B U Lead Fan and Train A Lag Fan (see NS 4 text for details).

h) An additional trip signal for high radiation alarm is provided by Q Recorder REA-RR-603 (on H13 -P600) and actuates a control room anmmdlltor .

Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: N/A Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-18 Secon 5.3.1 Revisio n CONTROl 21 ED

• RB differential pressure at or above O in. of water

• RB area differential temp above alarm level, Table 22

• RB area temp above alarm level, Ta ble 23 ET

• RB exh aust plenum rad iation level above 13 mRl hr

• RB area ra diation level above alarm level, Table 24

• RB area water level abov e alarm level, Table 25 SC-1 EL

• SFP temp above 124°F

• SFP level below 22 ft 4 in.

IF IF FAZ signal exists SG T cannot restore and maintain RB D

diffe rential pressure below O in. of water THEN en sure RB HVAC isolation and SGT init iation T H EN restart RB HVAC A ND N

radioactivity release from RB hinders operation of systems required fo r damage IF by EOPs TI O co ntrol or systems required to be operated RB HVAC is shutdown TH E N reset RB HVAC isolation logic AND operate ava ilable RB HVAC ES U

Q Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Examination Outline Cross-reference: 19 Revision: 3 Date: 1/5/21 Tier: 1 Group: 1 K/A Number: 600000.2.2.44 Level of Difficulty: 3 SRO Importance Rating: 4.2 K/A

Description:

Fire Protection: 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.

CGS is in Mode 1.

The following annunciator alarms in the control room:

I* TG BLDG 471'* 1, SYS 7 WET PIPE ELECT SWGR BAY TG BLDG 471' ELECT SWGR BAY TROUBLE TROUBLE I

FIRE FIRE 4.FCP1.2-2 4.FCP2.4-2 Field operators report a fire in switchgear SH-5.

The crew enters ABN-FIRE.

• Operators deenergize SH-5 in accordance with Pre-Fire Plan PFP-TG-471.

One minute later, operators are checking fire pump operation in accordance with ABN-FIRE, step 4.7.

• Fire Main Pressure is 119 psig, down slow.

With no operator action, which of the following fire pump configurations is expected for the conditions given?

A.

B.

Page 1 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 C.

D.

Answer: B K/A Match:

Requires interpretation of plant indications and determining how the Fire Protection system will respond when SH-5 is isolated.

Explanation:

A. Incorrect. Plausible since the power supply to FP-P-2A is MC-5N, which is powered from SH-5.

Although FP-P-2A should automatically start when Fire Main Pressure is LE 120 psig, it will not start since it is deenergized. However, control power for FP-P-1 (diesel driven) is MC-5N. With a loss of SH-5, MC-5N is lost. FP-P-1 will automatically start 20 seconds after a loss of control power.

B. Correct. The power supply to FP-P-2A is MC-5N, which is powered from SH-5. Although FP-P-2A should automatically start when Fire Main Pressure is LE 120 psig, it will not start since it is deenergized. However, control power for FP-P-1 (diesel driven) is MC-5N. With a loss of SH-5, MC-5N is lost. FP-P-1 will automatically start 20 seconds after a loss of control power. FP-P-2B automatically starts after fire main pressure is LE 110 psig C. Incorrect. Plausible if it is believed that FP-P-2A does not lose power with a loss of SH-5.

Plausibility is enhanced since fire main pressure is below the auto start setpoint for FP-P-2A.

However, FP-P-2A does lose power and cannot start. Additionally, control power for FP-P-1 (diesel driven) is MC-5N. With a loss of SH-5, MC-5N is lost. FP-P-1 will automatically start 20 seconds after a loss of control power.

D. Incorrect. Plausible since control power for FP-P-1 (diesel driven) is MC-5N. With a loss of SH-5, MC-5N is lost. FP-P-1 will automatically start 20 seconds after a loss of control power. Plausibility is enhanced if it is believed that FP-P-2B auto setpoint is at 120 psig. However, the auto start setpoint for FP-P-2B is 110 psig.

Tier 1 Discussion The question requires candidates to know the actions necessary to isolate equipment in accordance with the Pre-Fire Plan and how to validate these actions using pump indications.

Page 2 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Technical Reference(s)

SD000177, Fire Protection System Description SOP-ELEC-SH5-MAINT, Removing SH-5 From Service Attached w/ Revision #

See Comments / Reference ABN-ELEC-SH5, SH-5 Distribution System Failures PFP-TG-471, Turbine Generator 471 Pre-Fire Plan Proposed references to be provided during examination: N/A Learning Objective: 12271 - Explain the function and operation of the following Fire Protection System components, including any automatic features or interlocks:

a. Diesel Fire Pumps, FP-P-1, FP-P-110

b. Electric Fire Pumps, FP-P-2A, FP-P-2B Question Source: OBank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

New Question History: Last NRC Exam: N/A Question Cognitive Level: C, Memory or Fundamental Knowledge

' Comprehension or Analysis Justification for Cognitive Level Requires interpreting plant conditions given in the question stem to determine condition of the electrical distribution system and a knowledge of how plant conditions affect the Fire Protection system.

10 CFR Part 55 Content: 55.41 4 Page 3 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Comments /

Reference:

SD000177 Rev: Major: 16 Minor: 4 C0L.UMBV\ SYSTEMS August2m9' FIRE PROI'EJCTIO 3 D000 177, rl6 nu-4 COl\'i'fROL THEORY AND INTERLOCKS A _ Control Room Gonu-,ols l _ Diese] Fire Pwup FP-P-1 (FCP-3 2 position momentary S?I"ing remn to, AU-TO START Starts Diesel Ptimp, 2, _ Electric Fire Pwnp FP-P- A (FCP-3) 2 position nwmentuy s?I"ifig remn to, AUTO AU10 Auto start at LE 120 ps:i;g w ith no tin-ie delny START Starts pUMp 3 _ Erectr:ic Fire Pwnp FP-P-2B (FCP-3 2 position n1om.ent.uy s?I"ing remn to, A U-TO AU10 Auto start at LE llOllp s*g . , rn secorui tm-iedelay),

START Starts p:wnp 4 _ D iese] Fire Pwup FP-P- llD (FCP-3 2 position mom.entuy s?I"ifig remm to, AUTO AU10 Auto start nt LE 100 p s *g _ with a 30 sec time dela ' (3 5 seconds fm loss of controller powe:r)

Page 4 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 C OLUMBV\ SYSTE MS August 2m9 FIRE PRO/J'ECTIO 3 D000117 , d6, nu-4 IX. P0\\ 1E.R. SUPPLIES FP-P- l D iesel Engine FP-P-2A x;1 -SN FP-P-2B MC-6N FP-P-3 MC-6N FP-P- llD D iesel Engine FP-P-lll PDP-FP-1A FP-TK- 1 Cauto~ Unit MC-3C FP-VZ-1 (CO Vaporizer) MC-3C FCP- 1/-2/-3 E..PP-8AA Page 5 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Comments /

Reference:

SOP-ELEC-SH5-MAINT Rev: Major: 012 Minor: 003 Number: SOP-IEl EC-SIH5-MAINT IUse Category: CONTINUOUS Majo:r Rev: 012 M1no:r Rev: 003

Title:

IRemovinglRes,toring E-SH-5,, IE-Sl-51 , E-SL-52, IE-Sl-53 artd IE-SL-54 Page: n of 8-1 FrornfTo Service NOTE: Unless temporary pcmrer is suppl"ed to various lighting panels in GSB, pcmrer will not be available for routir,;e use to portal monitors, computers, etc.

5.11.33 N:OlllFY PGM Administrative Assistant and Operations Support Specialis,t, SB utility power will be los*t to free~rs. refrigerators m*crowa\'\es and vending machines d uri r,;g bus outage.

5.11.34 N:OlllFY 1he following SB area lighting ar,;d utility power will be lost during bus outage:

• Operations

• Fire Marshal 5.11.35 N:OlllFY HP 1hat power will be lost to 1he following:

All of 1he Argos, SAM's, GEIM-5s, ar,;d the West bank of TES stations will not lose power

• lights in the primary Egress area to 1he RCA

• TES Logon stations (INo:rlhside)

• l ead Tech computer ar,;d l"ghti11g in HP window

• l ighting in the Lead Tech Storage area

• South Supervisory offices r OTE: FD-P- 9B loses power during the bus outage.

5.11.36 VERIFY FD-P-119A is available (SB sur11p pump} (Sump S-1).

5.1 .37 VERIFY the fdllowing RW 437 waste handling area equipment not in use:

• IPVt.lR-GRA-1 (RW cask storage area crane)

• IPV!lR-GRA-2 (RW truck loading crane}

• IPVt.lR-TD-1 (Cask Tmnsfer Ddlly)

• RW 43 (Rollu,p Door C-10 }

NOTE: FP~-2A loses power and FP-P-1 auto starts dunng the bus outage.

Page 6 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Comments /

Reference:

ABN-ELEC-SH5 IRev: Major: 000 Minor: 004 Number: ABN-ELEC-SH5 I Use Category: CONTINUOUS Major Rev : ooo Minor Rev : 004

Title:

SH-5 Distribution System Failures Page: 4 of 9 4.8 RESTART necessary cooling tower fans and loads supplied from SL-51 SL-52 and SL-53 per SOP-ELEC-SH5-MAINT.

4.9 RESTART the idle Reactor Reci rculation Pump per SOP-RRC-START.

5.0 BASES 4.1 If a recirculation pump is lost at full power, the resulting level swell could potentially result in level 8 trips which result in a turbine trip leading to a reactor SCRAM .

4.2 A loss of SH-5 results in a loss of RRC-P-1 A. Th is dynamically alters core flow and is addressed/evaluated by the following :

• SOP-RRC-SINGLE LOOP, for single loop operation

• T.S . 3.4.1 , for single loop operation 4.3 If GT 25% power a loss of SH-5 would result in an unplanned entry into the Region Of Instability, and is addressed/evaluated by the following:

• ABN-RRC-LOSS , Loss Of Reactor Recirculation Flow

• ABN-CORE , Unplanned Core Operating Conditions 4.4 Half of the cooling tower fans are powered via SH-5. The reduction in cooling tower heat transfer abi lities resulting in lowering vacuum and requires adjusting power to match tower heat removal capacity .

4.5 This step and those following return plant systems to normal after corrective actions have been successful. SOP-ELEC-6900V-OPS, SOP-ELEC-480V-OPS and SOP-ELEC-SH5-MAINT provide explicit instructions, including precautions and limitations, for restoration of SH-5 distribution system to service .

Page 7 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Comments /

Reference:

PFP-TG-471 IRev: Major: 004 Minor: 001 Number: PFP-TG-47 1 I Use Category: N/A Major Rev: 004

>------------------~-----------------< Mi nor Rev: 001

Title:

TURBINE GENERATOR 471 Page: 15 of 22 Location: SW SWITCHGEAR AREA I Fire Area : TG-1 /2 Page 2 of 2 Areas covered by th is Pre-Fire Plan :

SW Switchgear Area .. .. .. ...... .. ..... .. .. .... ....... ...... ......... .. .. ..... ... ..... ........ ... .............. .. T207 Post Fire Safe Shutdown Equipment or Associated Cabling:

None Special Entry Considerations:

None Fire Detection :

Ionization detectors, Zone 7, FCP 1 Suppression Systems and Isolation Location :

N/A Electrical Disconnects or Special Shutoffs:

EPN of component Disconnect Location E-TR-5/53 E-SH-5 T 471 GB/16.0 / E-CB-5/53 E-SM-1 N/A E-CB-N 1/1 & E-CB-S1 E-SM-2 N/A E-CB-N 1/2 & E-CB-S2 E-SM-3 N/A E-CB-N 1/3 & E-CB-S3 E-SH-5 N/A E-CB-N2/5 & E-CB-S5 E-SH-6 N/A E-CB-N2/6 & E-CB-S6 Page 8 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-20 Examination Outline Cross-reference: 20 Revision: 0 Date: 7/22/20 Tier: 1 Group: 1 K/A Number: 700000.AA1.05 Level of Difficulty: 3 SRO Importance Rating: 3.9 K/A

Description:

Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Engineered safety features CGS is in Mode 1.

DG-2 is paralleled to the grid to complete OSP-ELEC-S702, DG2 Semi-Annual Operability Test.

The BPA dispatcher informs the control room that the 500 kV system condition is degraded.

The CRS enters ABN-ELEC-GRID, Degraded Off Site Power Grid.

How should the crew operate the emergency diesel generators?

A. Startup DG-1 and parallel it to the grid.

B. Shutdown DG-2 and place it in standby.

C. Unload DG-2 and start DG-1. Run both DGs unloaded.

D. Power SM-7 and SM-8 from their respective DGs and divorce both buses from off site power.

Answer: B K/A Match:

Requires knowledge of how to operate ESF equipment (diesel generators) with a grid disturbance.

Explanation:

A. Incorrect. Plausible since both DGs are picking up ESF bus loads when paralleled to the grid.

However, it is not desirable to parallel the DGs with an unstable grid.

B. Correct. with a DG synchronized to an unstable grid, the possibility of losing the DG as a reliable power source increases. ABN-ELEC-GRID, Degraded Off Site Power Grid, step 4.1.13, directs If any diesel generator is synchronized to the grid, then separate the diesel from the grid, and return the diesel to standby per the applicable procedure.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-20 C. Incorrect. Plausible since DGs running unloaded are capable of picking up ESF bus loads.

However, the DGs should not be run unloaded for long periods, and ABN-ELEC-GRID directs the DGs to be placed in standby D. Incorrect. Plausible since the ESF buses would be powered from reliable power supplies.

However, ABN-ELEC-GRID directs placing both DGs in standby.

Technical Reference(s)

ABN-ELEC-GRID, Degraded Off Site Power Grid Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 15748 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-ELEC-GRID.

Question Source: (' Bank #: Bank #

(' Modified Bank #: Mod Bank #. (Note changes or attach parent) r New Question History: Last NRC Exam: N/A Question Cognitive Level: r Memory or Fundamental Knowledge

r. Comprehension or Analysis Justification for Cognitive Level Requires synthesizing information given in the question stem with a knowledge of action required per ABN-ELEC-GRID along with an understanding that the DGs are not to be run unloaded for a prolonged period of time.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-20 Comments /

Reference:

ABN-ELEC-GRID IRev: Major: 011 Minor: N/A Number: ABN -ELEC-GRID [ Use Category: CONTINUOUS Major Rev: 011 Minor Rev: N/A Title : Degraded Off Site Power Grid Page : 6of 18 4.1.9 l.E. required to maintain Main Generator VARs within al lowable limits ,

THEN PERFORM the following:

CAUTION Do not adjust VARs , if the fu ll load motor Amps will be exceeded.

NOTE : VARs IN (under-excited ) resu lts in a reduction in the 25 KV bus voltage.

With the Plant operating at 100% power, a reduction in the 25 KV bus voltage , results in a rise in fu ll load amps fo r the associated motors fed from 25 KV.

a. MONITOR the Generator Capabil ity Curve, Attachment 7. 1 and 7.2.

b. IF directed by the CRS/SM to pick up positive (OUT) MVAR, THEN PLACE E-RMS-90 (Main Generator Exciter Vo ltage Adjuster) to RAISE (Clockwise ).

c. IF directed by the CRS/SM to pick up negative (I N) MVAR, THEN PLACE E-RMS-90 (Main Generator Exciter Voltage Adjuster) to LOWER (Co unter-Clockwise).

d. VERIFY VARs are within the Generator Capability Cu rve, Attachment 7.1 and 7.2.

4.1.10 SUSPEND all surveillances which have a potential to trip the Reactor or Main Turbine.

4. 1.1 1 REQUEST BPA VALIDATE grid analysis resu lts with alternate method (s).

{OE-6.5}

4. 1.1 2 REQUEST BPA SUSPEND any elective work related activities which could affect E-TR-S or E-TR-B . {OE-6.5}

4. 1.13 l.E any diesel generator is synchron ized to the grid, THEN SEPARATE the diesel from the grid ,

AND RETURN the diesel to standby per the applicable procedure .

t,JU..::>.;:tlU ll llY VI IU.;:)111~ ll l'C.;:)'C .:>UUI vc,;:, UI UI I ..:J I LC tJV VV 'CI .

4.1. 13 If the diesel is synchronized to an unstable grid, the possibility of losing the diesel as a reliable power source increases.

Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 Examination Outline Cross-reference: 21 Revision: 2 Date: 2/1/21 Tier: 1 Group: 2 K/A Number: 295009.AK1.02 Level of Difficulty: 2 RO Importance Rating: 3.0 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to LOW REACTOR WATER LEVEL: Recirculation pump net positive suction head CGS is in Mode 1.

A fault Reactor Feed Water Level Control occurs, causing RPV level to lower.

An automatic reactor scram occurs due to low reactor water level.

Operators are verifying actuations in accordance with PPM 5.1.1, RPV Control, step L-1.

Current plant conditions:

• RPV level is -10 inches, down slow.

• Both RRC pumps are running at 57 Hz.

How should operators control RRC pumps?

A. Runback RRC pumps to 15 Hz.

B. Runback RRC pumps to 30 Hz.

C. Maintain RRC pumps at 57 Hz.

D. Immediately trip both RRC pumps.

Answer: A K/A Match:

Requires knowledge of the required actions to prevent a loss of NPSH to RRC pumps on low RPV water level.

Explanation:

Page 1 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 A. Correct. PPM 5.1.1, step L-1 requires operators to ensure +13 inch actuations. PPM 5.0.10, clarifies that ensure means initiate each actuation which should have initiated but did not. In accordance with SD000126, Nuclear Boiler Instrumentation, RRC pumps runback to 15 Hz when RPV level lowers to +13 inches. Since RRC pumps are running at 57 Hz, operators should manually runback RRC pumps to 15 Hz to prevent pump cavitation due to poor NPSH.

B. Incorrect. Plausible RRC pumps should runback to 30 Hz when RPV level is below 31.5 inches with a trip of a Feed Pump. However, the question stem does not indicate that a Feed Pump has tripped. Additionally, RRC pumps will continue to runback to 15 Hz when RPV level goes below

+13 inches C. Incorrect. Plausible if it is believed that RRC pumps will not runback without a Feed Pump trip.

However, RRC pumps should runback to 15 Hz when RPV level goes below +13 inches.

D. Incorrect. Plausible since RRC pumps will automatically trip on low RPV level ( -50 inches).

However, the question stem indicates that RPV level hasnt lowered to this point. Therefore, for the conditions given in the stem, RRC pumps should have runback to 15 Hz.

Tier 1 Discussion Requires knowledge of actions required by EOPs to prevent cavitation. Meets Tier 1 requirements.

Technical Reference(s)

PPM 5.1.1, RPV Control PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

See Comments / Reference SD000126, Nuclear Boiler Instrumentation PPM 3.3.1, Reactor Scram Proposed references to be provided during examination: N/A Learning Objective: 9686 - Given an initial operating condition, describe the response of the RRFC system to DT LT 10.7°F between RPV Steam Dome and RRC Suction.

Question Source: 0Bank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

New Question History: Last NRC Exam: N/A Question Cognitive Level: 0 Memory or Fundamental Knowledge

' Comprehension or Analysis Justification for Cognitive Level Examinee must have knowledge of the actions required when verifying actuations during a scram and the actions that should be taken if the expected actuations did not occur.

Page 2 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 10 CFR Part 55 Content: 55.41 2 Comments /

Reference:

PPM 5.1.1 IRev: 22 RPV Leve l I, ,

I/ -----

11 l -1 z ~

EN SURE ALL actua tions

+113 in.

-50 in.

-112.9 in.

11.68 psig I

Comments /

Reference:

PPM 5.0.10 Rev: Major: 033 Minor: 002 Number: 5.0.m IUse Category: !INFORMATION Major Rev: 023 Mirtor Rev: NIA T it e: Flowchart Training Manual Page: 106 of 370

f. Definitive advance prioritization of system uses is pre cluded by the symptom-oriented approach to*emergency response u pan which the IEOPs are based.

g. Step L- 1:

1) Step L-1 assures in:itiatioo of f hose automatic actions most important for oontro[ling reactor coolant inventory. Other automatic actions ,o f lesser importance are subordinate to those diredJy affecting RPV water level oontrol, an;d their initiation rteed not be assured at th is point

2) The term 'ensure' means to initiate each actuation which should h av e imfiatea 5u aia not. It encompasses conclifions for wfiich automatic a ion shou d have occurred But failed to. In such a case, manual oQerato r action to initiate such action is required where app ro priate.

Page 3 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 Comments /

Reference:

SD000126 Rev: Major: 033 Minor: 002 COLUMBV\ SYSTE 1:S Februuy 2016

.B I SDCilOO l 6 , d3 m:r2 Table l RPV Water Level Trip Smrunary RPV Watei- Level T tip Summarv LEVEL Level 8 (54 . 5 ")

• TRIP FUNCTION Trip n:win tm-hine .

BASIS The high watei- level trip proteots fue

• RCIC turbine isolation main tu:rooe against nwisture carryovei-

• Tri,p foedwate{ (RFW) pwups and subsequem damage to, turbine

• Cfose HPCS injection vahe

,. ofadi.ng.

The RF\V pumps are tripped to p:revent

,. RPV 0rverfill.

Sterun to l:he RCIC tu11bine is secured, and HPCS iaject:ion valve is closed to prevent RPV oMeefill and flooding of the Level 7 (40..5 ")

• High RPV watei- level ala:rm. .,

RCIC ro:rbine s team line.

The watea- level al.arm. annnncia at the reacroir vi sel water revel abo;ve v.1uch moishir,e canyo,ver in fue, steam is expected no incr,e ~ at a

• ignificant nte.

The afo:rm warns the* opecator of tlus Le,;el 5 , 6

• Upper ll!fld lowe1" bowids ,o f nomw RPV water revel uml:esirnble condition..

The F\\rt,C is nmmal!l.y ,o pernlied based on maintairung wnlier revel near revel 5 (36,,),. Operating at tlus point ntinimizes Level 4 (31.5 ")

• Lo,w RP'V water level alarm RRC runback on RFW tm:hme carryover and ste.mn carryuoieL _

The watea- level al.arm. annnncia

• at the water revel below which stean1 trip car:ryunder in the* water will begin affecting :reciroulation flow rate sigruficam:1 ' at full power due to, RRC

,. pump cavitation.

A water level decrease* no this point, coupled with a trip, of a RF\\1 turbine*,

causes RR!C pumps to runhack to, 3U hertz. This reduces. core flow ll!fld thns reacroir po,w er Jio place it witi.un fue, capacity of the rem.a irung RFW pump .

Page 4 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 COlUMBLI\ SYSTEMS FelmJU)' 2016 BI

' SD0001 2i6 , 1:13 mc2 (cont'1cf I RPV Water Level T:cip Sumnuuv RPV Water Level Ttio Sunu:mi:rv LEVEL TRIP FUNCTION BASIS Le vel 3 {l r

• Reactor scram.

• The scm:m function. occun whil.e water

• NS4 Groups 5 & 6 isolation level is still above tire botiom of tbe

• .I\DS RPV teve] confnmation sterun diyer sk:ict, which precludes high

• RRC runback to 15 hertz moisture carryover dl!le to steam.

b rpassmg the ,drye1" u'fder the seaJI skirt.

Tlus level also resullts m a quantity of rese£Ve coo]ant b et'\.vieen this 1eve1 ,and T AF to, account for evaporation (decay heat boil.- off), fosses, steam v,oid collapse , and othe:r coo am losses from the RPV following a loss of RFW event without water level decreasing to Level 1, which would initiate EOCS.

This quantity ,o f resen,ie, coolant ass11UDes RC IC is capable of pmv:iding adequate makeup -

• Trus level also causes S4 isolations to prevem a loss of ooolant from the,se, potenfa] lealmge paths.

• A Level 3 si~ is used\ by the AIDS log:ic to confum that in deed, a fow water level does ,erust.

• The RRC pui:np& rw:np down to 15 hertz to prevent pump cavitat ion due to poor N PSH .

Level _ (-50l

• Iru.tiate RCIC

• This setpoint was selected so til:tat

• Iru.tiate H P CS RC]C/HPCS wol!l]d not initiate after a

• Start HPCS DG scran1 w ith FW available . It is high

• Isolate NS4 G!"m1ps , 3 4 & 1 enough so that RCIC could prevem: a

• Trip RRC pumps level decrease to Le,rel 1 , - 1 9 " ),

• uutiate A TWS/ ARI folil.ow:ing a loss of feedW&Jter event .

• This level also causes S4 isolations to prevem a loss of ooolant foom these potentia] leaka~, p aths .

• RRC trip at this Iev el adds negative reactivity to the,,oor,e,_

• AT\\i'S/.ARI is a b ackup to the RPS scran1 which should have oco11m ed at L3 ( 13 " },_

Page 5 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 Comments /

Reference:

PPM 3.3.1 IRev: Major: 066 Minor: 001 Number: 3.3. 1 IUse Catego:ry: CONTIIN UOUS Major Rev: 066 1 - - - - - - - - - - - - - - - - - - ~ - - - - - - - - - - - - - 1 Minor Rev: 001

Title:

Reactor Scram Pag,e: 6 of 26 NOTE: The follov.1n;g steps may be pertom1ed ,oonru rrenUy or as directed by the CHS.

5.2 Subsequent Actions - CR01 NOTE: Per GE SIL 532, control rod position FU l l-1N indication may be t empo:rarily lost follov.~n;g a scram due to a te"°'peraru re eXJcursio 1, which temporarily reduces the strength of the magnet for the FUll I position switch .

NOTE: The reactor is considered shuMovm under all .c,onditioos if no more than one contiol rod is withdrawn past position 02.

5.2..11 IINSEiRT Source Range and lntennediate Range Monitors. fOE-7.7}

5.2..2 REPORT ro:ntrol rod status (all rods in/not in), to the CRS.

5.2.3 IF O11n an ATWS, THE PERFORM the following :

a. RESTORE and MAINTA!IN reaotor water leveJ +13" ,. to +54S.

See Attachment -8. 1 fo:r list of sources. {C-7.8} {P-7.10} _ _

b. VERIFY Reactor Recii-cu a *on pump s have runback to~ 15 Hz.

Page 6 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Examination Outline Cross-reference: 22 Revision: 1 Date: 1/19/21 Tier: 1 Group: 2 K/A Number: 295010.AA2.02 Level of Difficulty: 2 SRO Importance Rating: 3.8 K/A

Description:

Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: Drywell Pressure.

CGS is in Mode 1.

A LOCA occurs.

The crew is taking actions in accordance with PPM 5.1.1, RPV Control, and PPM 5.2.1, Primary Containment Control.

Current plant conditions:

• RHR-B is spraying the wetwell.

• RHR-A is spraying the drywell.

• Drywell pressure: 16 psig, down slow.

• Wetwell pressure 13 psig, down slow.

In accordance with OI-15, EOP and EAL Clarifications, when should drywell spray be secured?

Secure drywell spray when drywell pressure A. goes below 0 psig.

B. reaches 1.68 psig.

C. goes below 12 psig.

D. is not within the Drywell Spray Initiation Limit (DSIL).

Answer: B K/A Match:

Requires ability to interpret drywell pressure in relation to securing containment sprays.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Explanation:

A. Incorrect. Plausible since containment sprays are required to be secured prior to going below 0 psig. However, in accordance with PPM 5.2.1, Primary Containment Control, PPM 5.0.10, Flowchart Training Manual, and OI-15, EO and EAL Clarifications, drywell spray is secured when drywell pressure reaches 1.68 psig.

B. Correct. In accordance with PPM 5.2.1, Primary Containment Control, PPM 5.0.10, Flowchart Training Manual, and OI-15, EO and EAL Clarifications, drywell spray is secured when drywell pressure reaches 1.68 psig.

C. Incorrect. Plausible since drywell spray is not initiated until drywell pressure exceeds 12 psig in accordance with PPM 5.2.1, step P-7. This suggests that drywell spray is not needed with drywell pressure less than 12 psig. However, in accordance with PPM 5.2.1, Primary Containment Control, PPM 5.0.10, Flowchart Training Manual, and OI-15, EO and EAL Clarifications, drywell spray is secured when drywell pressure reaches 1.68 psig.

D. Incorrect. Plausible since drywell spray is not initiated if drywell parameters are not within DSIL in accordance with PPM 5.2.1, step P-8. However, in accordance with PPM 5.2.1, Primary Containment Control, PPM 5.0.10, Flowchart Training Manual, and OI-15, EO and EAL Clarifications, drywell spray is secured when drywell pressure reaches 1.68 psig.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

See Comments / Reference OI-15, EOP and EAL Clarifications Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8333 - Given a list, identify the statement that describes the two possible results of continuing to spray the wetwell when wetwell pressure is below 0 psig.

Question Source: OBank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

New Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level: ' Memory or Fundamental Knowledge 0 Comprehension or Analysis Justification for Cognitive Level Requires knowledge of termination point for drywell spray.

10 CFR Part 55 Content: 55.41 8 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A 12 psig P-8 Is drywell temp below YES PPM 5.5.2 DS IL p.g Is WW level Y ES NO below 51 ft NO P-1 1 P-10 BEFORE drywell pressure THEN stop drywell 1. STOP RR C pumps and drywell sprays coo ling fans drops be low 0 ps ig (1.68 psig)

2. 6 SPRAY the drywell with sources w required for continuous RPV injection External spray sources may be used only if PC water ~

level and wetwell pressure can be res to red and B

maintained below PCPL ABN-TSG-008 Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Comments /

Reference:

PPM 5.0.10 IRev: Major: Maj Minor: Min

h. Steps P-8 through P-11:

1) See section DT-4 through DT-7 for discussion of drywell sprays.

g. Override DT-6:

1) Drywell spray operation must be term inated by the tim e drywel l pressure drops below O psig to ensure that primary containment pressure is not reduced below atmospheric. Maintaining a positive pressure precludes air intake through the vacuu m relief system to de inert the pri mary containment and also provides a positive marg in to the negative design pressure of the primary conta inment.

2) Termi nating sprays "before ...O psig" perm its use of the sprays for fission product scrubbing at low pressures or if the containment has failed , yet sti ll avoids negative containment pressures. Consistent with the definition of "before," the actual pressure va lue at which the sprays should be secu red is event-specific:

• Reducing contain ment pressure below the scram setpoint will clear the scram log ic and maximize the marg in to containment pressu re limits.

• If the containment has failed or if pri mary containment venting is anticipated , it may be advisable to continue spray operation at low pressu res to scrub the containme nt atmosphere.

• Reducing primary conta inment pressure will also reduce the NPS H avai lable for pumps drawing suction fro m the suppression pool. If there is no need for continued spray operation , sprays should be terminated at higher pressures if NPSH limits are approached .

Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Comments /

Reference:

OI-15 IRev: Major: 032 Minor: N/A

• Step P-10 , termination of drywell sprays: the expectation is that drywell sprays should occur at 1.68 psig thus ensuring positive pressure maintained.

Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-23 Examination Outline Cross-reference: 23 Revision: 1 Date: 1/11/21 Tier: 1 Group: 2 K/A Number: 295012 AK3.01 Level of Difficulty: 2 RO Importance Rating: 3.5 K/A

Description:

Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE: Increased drywell cooling A transient occurs that results in the following plant conditions:

• The reactor is scrammed.

• CB-S3 failed to automatically close

• Drywell temperature slowly rising

• Drywell pressure slowly rising.

The CRS has entered PPM 5.2.1, Primary Containment Control.

Which of the following actions should be taken now to lower Drywell temperature and pressure?

A. Restart Drywell Cooling Fans.

B. Perform PPM 5.5.15 Emergency Drywell Venting.

C. Open the Weir Gate for the operating TSW Pump 200 turns.

D. Manually close RCC-V-6 to divert all RCC cooling flow to the drywell loads.

Answer: A K/A Match:

Requires an understanding of system inter-relationships between TSW, RCC and Drywell Cooling Loads and the effects on Drywell temperature and pressure.

Explanation:

A. Correct. Failure of S-3 to automatically close following the scram results in power being temporarily lost to SM-8. When power is automatically restored to SM-8 by TR-B, there is no automatic function for restart of the drywell cooling fans. This is a manual action performed by the operator per the direction of the CRS via step DT-1 of PPM 5.2.1 Primary containment control Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-23 DT-1 MAINTAIN drywell t.emp below 1J5°F with available drywell oooling I .

8.9.5 Drvwell Temperature

a. St~, DT-1:

11) Tl\e initial action taken o control drytNelll empera~ure employs the s-ame m etl'lod typi:cally used during1nommll p~ant operat i,ons: mo11itortng its status 1

and placing a\ilailla }le drywell ,coo li11g in operation as required to maintain drywell ernperature below lhe LCO varue. Sep DT- 1 hus provides a smooth transilJion fro m g:ooeral plant procedures to emergency opera ing1 procedures, and assures th the no1m1all me hod of drywe ll temperature control is a empted in advance of i11iUati1111g more oompl:ex actio111JS o termrnate*inoreas ing1d11Y1Nell ternnperatme.

2) As long as average*d'rywe ll temperature remains below no1m1all operatirng limits, no further action is required irn lhis nowpath o~her than continu ing1to mornoor arrnd ,control a .erage drywelll emperarure using1avarla }f,e dli:ywell coolirng .

3) Tl\e applicability of Caution #1 highli:ghts the d1rect effiect tllat high d'rywelll temperature*has on IRPV w*ater level indicatiorn .

B. Incorrect. Plausible While PPM 5.5.15 would certainly lower Drywell pressure, however PPM 5.5.14 provides the direction necessary to vent the drywell during emergencies when the integrity of the Primary Containment is threatened. The stem of the question states that drywell temperature and pressure are slowly rising. PPM 5.5.14/PPM 5.5.15 may be used in PPM 5.2.1.

However, in accordance with step P-4, these procedures are only used if lowering PC pressure is required to maintain adequate core cooling or reduce off site dose.

C. Incorrect. Step DT-1 bases states that the normal method of drywell temperature reduction is attempted in advance of initiating more complex actions to terminate increasing drywell temperature. Opening the Weir Gate for the operating TSW pump will increase the amount of cooler water from TMU to mix with the warmer CW basin water. TSW temperature will lower.

TSW cools the RCC Heat exchangers, a lower TSW temperature will result in more cooling in the RCC heat exchanger and lower RCC temperatures. The affect on RCC cooling is small (compared to restarting drywell cooling fans) and it not a standard action that is taken for this situation.

D. Incorrect. Plausible if the operator believes that RCC flow is not restored due to the failure of CB-S3. The failure of CB-S3 to automatically close will result in TR-B re-energizing SM-8. RCC pump breakers do not trip on a loss of bus voltage (either temporary or complete loss) When SM-8 is re-energized RCC will automatically be restored and no actions are required to manually close RCC-V-6. (RCC-V-6 Noun Name not provided in the distractor; this valve is a required element of the must know RCC drawing for licenses candidates).

Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-23 Technical Reference(s)

PPM 5.0.10 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: a. 8311 Given a list, identify the statement that describes the purpose of attempting to maintain drywell temperature below 135 degrees F. (PPM 5.2.1)

Question Source: OBank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

New Question History: Last NRC Exam: N/A Question Cognitive Level: 0 Memory or Fundamental Knowledge

' Comprehension or Analysis Justification for Cognitive Level Needs solid understanding of system inter-relationships of RCC / TSW / Drywell cooling loads and correct situational mitigation strategies of PPM 5.2.1 based on synthesis of plant conditions.

10 CFR Part 55 Content: 55.41 5 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-23 Comments /

Reference:

PPM 5.0.10 Rev: Major: 14 Minor: 4 8.9.5 Drywell Temmerature

a. Step DT-t :

11) Th.e initial! action taken to contro l drywelll 1en1pera~t.me employs. the same method typically us ed dt1ring1nonnall pf:ant operations: mon itoring its status 1

and p,aci111g available drywell ,coo ling in operaUon as requ ired to mai111tain drywell temperature below the LCO vatue. Step DT- 1 hlus pro'llid'.es a smooth irnnsitJion from ,g:eneral plant procedures to emergency opera ing1 procedu res,. arnd assures th the nom1all ,nethod of dryiNe ll temperature control is attempted in advance of initiating mo re*,oom pte-x actioniS o term inate*inoreas ing1dl)i'W'ell temperatu11re.

2) As long as average*d'rywelll ~empe rature remains below normal operating limils, no further action is required in 1his ftowpath other than contin,uing1to rnonitJor and co nrtrol aive rage drywelll emperatll.lre using ava abte* clrywell cooling.

3) Th.e appli.cab ility of Caution # 1 hrghlighlts thle*d'ireot effecUhiat hig h! drywelll tem perature*has on RPV water level indicati:on.

Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-24 Examination Outline Cross-reference: 24 Revision: 1 Date: 1/19/21 Tier: 1 Group: 2 K/A Number: 295014.AK2.07 Level of Difficulty: 3 RO Importance Rating: 3.9 K/A

Description:

Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following:

Reactor power CGS is in Mode 1.

The plant has just returned to full power following a refueling outage.

Feedwater Heater 5A trips.

How does this event affect the power profile in the core and which procedure should be entered to mitigate this event?

Power peaks at the core (1) . The crew should enter (2) .

A. (1) top (2) ABN-CORE, Unplanned Core Operating Conditions B. (1) top (2) ABN-POWER, Unplanned Reactor Power Change C. (1) bottom (2) ABN-CORE, Unplanned Core Operating Conditions D. (1) bottom (2) ABN-POWER, Unplanned Reactor Power Change Answer: D K/A Match:

Requires knowledge on how an inadvertent reactivity addition (loss of feedwater heater) affects reactor power.

Explanation:

A. Incorrect. Plausible if it is believed that the loss of efficiency when a feedwater heater is lost causes reactor power to lower. Plausibility is enhanced since a condition requiring entry into ABN-CORE may be present if reactor power were to lower. However, a loss of feedwater heating Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-24 causes core inlet temperature to lower, which causes reactor power to peak lower in the core.

Additionally, ABN-POWER is entered to mitigate this event.

B. Incorrect. Plausible since ABN-POWER is entered for this event. However, a loss of feedwater heating causes core inlet temperature to lower, which causes reactor power to peak lower in the core.

C. Incorrect. Plausible since reactor power will peak at the core bottom due to the loss of feedwater heating an subsequent lowering of core inlet temperature. However, ABN-POWER is entered to mitigate this event.

D. Correct. When feedwater heating is reduced, the temperature of the feedwater entering the core is reduced. A lower core inlet temperature causes reactivity at the core inlet to rise, which causes more power production lower in the core (eg - reactor power peaks at the bottom of the core). An entry condition for ABN-POWER is the loss of a #5 or #6 Feedwater Heater. Section 4.3 provides actions to reduce reactor power and stabilize the plant on a loss of a #5 Feedwater Heater.

Technical Reference(s)

ABN-POWER, Unplanned Reactor Power Change ABN-CORE, Unplanned Core Operating Conditions Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 6743 - With the procedures available, describe the reason for the following step associated with ABN-POWER: [ABN-POWER]: The requirement to reduce RRC flow to LE 80 Mlbm/hr, and insert rods to maintain below the 100% rod line when feedwater inlet temperature experiences an unplanned drop of GE 6°F.

Question Source: (' Bank #: Bank #

(' Modified Bank #: Mod Bank #. (Note changes or attach parent) r New Question History: Last NRC Exam: N/A Question Cognitive Level: r Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Requires candidates to synthesize the information given in the stem with an understanding of the effects of a loss of feedwater heating on the core flux profile along with a knowledge of AOP entry conditions..

10 CFR Part 55 Content: 55.41 1 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-24 Comments /

Reference:

ABN-POWER IRev: Major: 016 Minor: 003 CAUTION A loss of reactor feedwater heating causes the core to become more bottom peaked .

This results in the core power being more concentrated toward the botto , creating higher power-to-flow cond itions in these areas of the core. Historically, power oscillation events have occurred when the core was operating with relatively high power-to-flow conditions. Operation with one of the feedwater heaters out of service was considered a contributing cause for the 1992 core oscillation event at Columbia Generating Station.

Number: ABN-POW ER I Use Category: CONT INUOUS Major Rev: 016 Minor Rev: 003 Title : Unplanned Reactor Power Change Page : 3of17 1.0 ENTRY CONDITIONS 1.1 Unplanned Reactor Power Change A noticeable unplanned reactor power change is indicative of changing plant conditions. This procedure should not be entered for normal fluctuations of reactor power observed on the APRMs due to fluctuations in th e boiling boundary. This proced ure should also not be entered for a reacto r power change associated with a reactor scram. This procedure shou ld not be entered for a planned reactor power change di rected by a Reactivity Control Plan (RCP). Typically, small power changes are not identifiable on the APRMs but can be identified on PPCRS. Power changes can be caused by a number of thi ngs including:

• RPV inlet temperature change

• #5 or #6 Feed Water Heater Trip Indication

• Core flow change (Automatic RRC pump speed change)

• Core flow change (Due to operator Core Flow change for unplanned conditions)

• Jet Pump Failure

• Resin intrusion

• RPV pressure change

• Neutron poison chan ges due to reactor power changes Comments /

Reference:

ABN-CORE Rev: Major: Maj Minor: Min Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-24 Number: ABN-CORE I Use Category: CONTINUOUS Major Rev: 017 Minor Rev: 002 Title : Unplanned Core Operating Conditions Page: 4 of 24 1.0 ENTRY CONDITIONS {C-6.2}

1. 1 When any of the followi ng conditions exist:

• The reactor mode switch is in RUN or STARTUP and both RRC pumps have tripped-off (i. e. , neither RRC pump is running ). {C-6.2}

• One RRC pump has tripped-off and reactor power is GT 25% of rated . {C-6.2}

• Automatic runback of both RRC pumps down to 15 hz.

• Any unexplained significant and sustai ned periodic LPRM upscale or downscale alarms .

• Any unexplained significant and sustained oscillations in SRM period , LPRM or APRM levels (characteristic oscillation period is 1-3 seconds). The following are provided as examples and do not represent absolute values:

• Peak to Peak APRM exceeds 10%.

• Any LPRM has oscillations GT 20 % Flux as determined by selecting an interi or control rod (such as 30-43 , 14-35, 42-27, 30-39), and monitoring the RBM ODA.

{C-6.1}

1.2 Entry into Region A on the Power to Flow map.

1.3 Entry into the Area of Increased Awareness (AIA) on the Power to Flow map.

1.4 Unintentional Entry into the OPRM Enabled Region on the Power to Flow map.

1.5 Operation in the OPRM Enabled Region on the Power to Flow map when LCO 3.3. 1. 1 requires an alternate method to detect and suppress thermal hydraulic instability oscillations in accordance with ACTION I.1 .

1.6 Receipt of unexpected OPRM TRIP ENABLED alarm (H13-P603.A7-3.7)

(H13-P603.A8-6.5).

1.7 Receipt OPRM ALARM (H13-P603.A8-1.1) or OPRM TRIP (H13-P603.A7-2.4 )

(H13-P603.A8-2.4) annunciators.

1.8 Control system problems (RRC flow control , DEH , and Feedwater Level Control (FWLC) systems).

Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Examination Outline Cross-reference: 25 Revision: 1 Date: 1/11/21 Tier: 1 Group: 2 K/A Number: 295022.AK1.01 Level of Difficulty: 3 RO Importance Rating: 3.3 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS: Reactor pressure vs. rod insertion capability The plant was operating at 100% power when the following occurs:

Time Event 1159 The operating Control Rod Drive (CRD) pump, CRD-P-1A, trips.

1200 Charging Header Pressure (CRD-PI-15) is 938 psig, down slow.

1203 Control Rod Scram Accumulator CRD-HCU-3027 is declared inoperable due to low accumulator pressure.

1205 Control Rod Scram Accumulator CRD-HCU-4635 is declared inoperable due to low accumulator pressure.

1230 Charging Header Pressure (CRD-PI-15) is 895 psig, down slow.

Based on this timeline, what is the time that a reactor scram is required to be initiated to meet technical specification requirements?

A. 1220 B. 1223 C. 1225 D. 1230 Answer: C K/A Match:

Requires knowledge of the operational requirements in Technical Specifications due to slow rod scram times on a loss of CRD pumps based on reactor pressure.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Explanation:

A. Incorrect. Plausible if it is believed that a reactor scram must be initiated 20 minutes after the CRD pump trips with two inoperable scram accumulators. However, the 20 minute time limit starts when the second scram accumulator becomes inoperable with steam dome pressure 900 psig.

B. Incorrect. Plausible since a reactor scram is required 20 minutes after 2 control rod scram accumulators become inoperable with CRD d/p below 940 psig. However, the second control rod scram accumulator does not become inoperable until 1205.

Therefore, a scram is not required at 1223.

C. Correct. Limiting Condition for Operation (LCO) 3.1.5, Control Rod Scram Accumulators, Condition B states that with Two or more control rod scram accumulators inoperable with reactor steam dome pressure 900 psig, operators are required to Restore charging water header pressure to 940 psig with a required completion time (B.1) of 20 minutes from discovery of Condition B concurrent with charging water header pressure < 940 psig. For the conditions given, the clock for completing the required actions of Condition B starts at 1205, when the second scram accumulator becomes inoperable. Condition D of the LCO requires placing the Reactor Mod Switch in SHUTDOWN immediately if the completion time for required action B.1 is not met.

Therefore, the earliest time when a scram is required is 20 minutes from entering Condition B, or 1225.

D. Incorrect. Plausible since a reactor scram is required if steam dome pressure is < 900 psig and a single scram accumulator is inoperable with charging water pressure < 940 psig. However, the question stem states that at 1230, charging water pressure is < 900 psig, NOT steam dome pressure. With reactor power at 100%, steam dome pressure is well above 900 psig.

Technical Reference(s)

Technical Specifications Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11768 - Discuss when the reactor must be manually scrammed due to Control Rod Drive Hydraulic system malfunctions Question Source: ~Bank #: 64695 (Sys ID) 0 Modified Bank #: Mod Bank #. (Note changes or attach parent) 0New Question History: Last NRC Exam: N/A Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Question Cognitive Level: C Memory or Fundamental Knowledge (e':, Comprehension or .Analysis Justification for Cognitive Level Examinee must synthesize a knowledge of the < 1hour actions required by LCO 3.1.5 along with an understanding of when the plant conditions given in the question stem meet the requirements to take Tech Spec actions.

10 CFR Part 55 Content: 55.41 10 Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Comments /

Reference:

Technical Specification Ammendment: 254 I

Control Rod Scram Accumu lators 3.1. 5 ACTIONS CONDITION REQ UI RED ACTION COMPLETION TI ME B. Two or more control rod B.1 Restore charging water 20 minutes from scram accumulators header pressure to discovery of inoperable with reactor 2 940 psig. Condition B steam dome pressure concurrent w ith 2 900 psig. charg ing w ater header pressure

< 940 psig AND B.2.1 ---------------NOTE-------------

Only applica ble if the associated control rod scram time w as within the limits of Table 3.1.4- 1 during the last scram time Surveillance.

Declare th e associ ated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> co ntrol rod scram time "slow."

OR B.2.2 Declare the associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> co ntrol rod inoperable .

Columbia Generating Station 3.1.5-2 Amendment No.~.~ 225 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Contro l Rod Scram Accum ulators 3.1.5 ACTI ON COND ITION REQUIRED ACTION COMPLETI ON TI ME C. One or more control rod C.1 Verify th e associated Immediately upon scram accumulators control rod is fully inserted. discovery of charging inoperable with reactor water header steam dome pressu re pressure < 940 psig

< 900 psig .

AND C.2 Declare the associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> control rod inoperable.

D. Required Action B.1 or 0 .1 ---------------NOTE--------------

C.1 and associated Not applicable if all Completion Time not inoperable contro l rod met. scram accumulators are associated with ful ly inserted control rods.

Place the reactor mode Immediately switch in the shutdown position .

Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-26 Examination Outline Cross-reference: 26 Revision: 2 Date: 1/20/21 Tier: 1 Group: 2 K/A Number: 295034.EA1.01 Level of Difficulty: 2 RO Importance Rating: 3.8 K/A

Description:

Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Area radiation monitoring system.

CGS is in Mode 1.

An event occurs that causes a reactor scram.

The following conditions exist:

• RPV level lowered to -20 inches. It is now +15 inches, up slow.

• Drywell pressure is 1.59 psig and up slow.

• Suppression Pool level is -1.25 inches and down slow.

• Control rod 30-31 is at position 48. Control rod 18-15 is at position 02. All other control rods are at position 00.

• ARM-RIS-4 East CRD Area is in alarm at 2300 mr/hr.

• Reactor Building differential pressure is -.05 inches of water.

• Reactor Building Exhaust Plenum is 12 mr/hr Based only on these conditions, which procedure(s) should be used to address this event?

Enter PPM 5.1.1, RPV Control...

A. ONLY.

B. and transition to PPM 5.1.2, RPV Control - ATWS.

C. and PPM 5.2.1, Primary Containment Control.

D. and PPM 5.3.1, Secondary Containment Control.

Answer: D K/A Match:

Requires knowledge of EOP entry conditions with respect to Secondary Containment Ventilation radiation levels.

Explanation:

A. Incorrect. Plausible since PPM 5.1.1. is entered on low RPV level. However, this is not the only EOP required to be entered based on conditions given.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-26 B. Incorrect. Plausible since two control rods are not at position 00. However, to enter PPM 5.1.2, at least two control rods must be GE Notch 02.

C. Incorrect. Plausible since entry into PPM 5.2.1 is required when Wetwell lowering. However, PPM 5.2.1 is not entered until Wetwell level is LT -2 inches.

D. Correct. One entry condition for PPM 5.3.1 is RB area radiation level above alarm level. The question stem states that ARM-RIS-4 is in alarm, which is an entry condition for PPM 5.3.1.

Technical Reference(s)

PPM 5.3.1, Secondary Containment Control Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8017 - Given plant conditions, recognize an EOP entry condition(s) and enter the appropriate flow chart.

Question Source: r Bank #: LO01149

(' Modified Bank #: Mod Bank #. (Note changes or attach parent)

(' New Question History: Last NRC Exam: N/A Question Cognitive Level: r Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Examinee must synthesize information in question stem with a knowledge of entry conditions for multiple EOPs.

10 CFR Part 55 Content: 55.41 12 Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: Min Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-26 Seco 5.3.1 Revision CONTRC 2

• RB differential pressure at or above O in. of water

• RB area differential temp above alarm level, Table 22

• RB are a temp above alarm level, Table 23

• RB exhaust plenum rad iation level above 13 mR/hr

• RB area ra diation level above alarm level , Table 24

• RB area water level above alarm level, Table 25

• SFP temp above 124°F

• SFP level below 22 ft 4 in.

I 241 RB Area Radiation Ma~im11m Sale Qoecati Q Alarm ~

A= Instrument

(!!lB/llr}

(mR/hr) (R/hr)

E ast CR D Area ARM-RIS-4 10'! N/A West CRD A rea ARM-RIS-5 10'! N/A H 2 Recomber Area ARM -RIS-6 10'! N/A T IP D ri ve Area ARM-RIS-7 10'! N/A

'12

~

SGT Fi lter A rea ARM -RIS-8 N 10'! N/A Q

R HR P ump A Room ARM-RIS-9 s:i' 101 N/A

"" ~

Q_ (")

RHR P ump B Room ARM-RIS-10 Q_ s: 10'! N/A R H R Pump C Room ARM -RIS-11 2-0 N/A

.§ -~ 10'!

RC IC Pump Room ARM-RIS-12

.9-s N/A

.~ 10'!

HPCS Pump Room ARM-RIS-13 E 10'! N/A CRD P ump Room ARM-RIS-23 <i: 10'! N/A 471 E lev - West A rea ARM -RIS-24 10'! N/A 471 E lev - NE ARM-RIS-32 N/A 10 501 E lev - NW ARM -RIS-33 N/A 10 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 Examination Outline Cross-reference: 27 Revision: 1 Date: 1/5/21 Tier: 1 Group: 2 K/A Number: 500000.2.4.50 Level of Difficulty: 2 RO Importance Rating: 4.2 K/A

Description:

High primary containment hydrogen: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

CGS is in Mode 1.

The following annunciator is in alarm:

814.J2.2-2: CONTAINMENT ATMOSPHERE HYDROGEN LEVEL HIGH Primary Containment H2 indications are:

(CMS-CP-1301) (CMS-CP-1401) 1.7 1.6 0.0 30.0 DRYWELL H2% DRYWELL H2%

R-1 In accordance with the Alarm Response Procedure (ARP) for 814.J2.2-2, what is the first action that should be performed?

A. Place the Hydrogen Water Chemistry Enable switch to SHUTDOWN (H13-P840) in accordance with PPM 5.2.1, Primary Containment Control.

B. Verify channel B operability by aligning channel A (CMS-CP-1301) to match the area being sampled on channel B (CMS-CP-1401).

C. Vent primary containment in accordance with PPM 5.5.21, Emergency Drywell Venting With High Hydrogen and Oxygen Concentrations.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 D. Perform a hydrogen monitor channel calibration in accordance with PPM 5.8.1, Post-LOCA Hydrogen/Oxygen Monitoring.

Answer: B K/A Match:

Requires knowledge of alarm procedure and equipment that needs to be operated for a high containment hydrogen alarm.

Explanation:

A. Incorrect. Plausible since the alarm setpoint for 814.J2.2-2 is 3.56%, which is entry criteria for PPM 5.2.1. However, the ARP directs actions to validate that the reading is correct, especially since the other monitor reads 0%.

B. Correct. In accordance with the ARP, the first action is to validate the hydrogen level by comparing the alarming channel to the other channel and aligning both channels to sample from the same location. Operators should wait for at least 15 minutes after shifting sample points to ensure a valid reading.

C. Incorrect. Plausible since this action would be performed in accordance with PPM 5.2.1, step H-3, if high hydrogen concentration is confirmed and PPM 5.2.1 is entered. However, for the conditions given in the question stem, the crew would not enter PPM 5.2.1 without first confirming hydrogen concentration.

D. Incorrect. Plausible since this calibration is mentioned in the ARP. However, it is performed post-LOCA to correct hydrogen monitor readings for elevated radiation levels. It is not required to be performed for the conditions given in the question stem.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control 4.814.J2.2-2 Containment Atmosphere Hydrogen Level High Alarm Attached w/ Revision #

Response Procedure See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8417 - Given PPM 5.2.1, "Primary Containment Control", identify the two methods of determining hydrogen and oxygen concentrations stated on the flow chart. (PPM 5.2.1)

Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 Question Source: (" Bank #: Bank #

(" Modified Bank #: Mod Bank #. (Note changes or attach parent)

C.- New Question History: Last NRC Exam: N/A Question Cognitive Level: r. Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Requires knowledge of actions for a high containment hydrogen concentration.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

PPM 5.2.1 IRev: Major: 28 Minor: N/A

• WW temp above 90°F

• Drywell temp above 135°F

• Drywell pressure above 1.68 psig

• WW level above +2 in. or below -2 in.

• PC hydrogen above 3.56%

I Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 PC GAS I H-1 IF PC hydrogen and oxygen monitoring systems not available THEN notify chemist to sample PC hyd rogen and oxygen PPM 12.17.3 H-2

-lWHEN

- - - ---7 PPM 5.8.1 provides guidance for operation of hydrogen and oxygen hydrogen is monitoring under post-detected in PC LOCA conditions THEN verify hydrogen injection IF reactor scram has (GE 0.6%)

is secured and isolated been initiated THEN secure vent and purge H-3 IF offsite radioactivity llll.l required by other release reaches EOP steps IF offsite radioactivity release is expected to ODCM RFO limits remain below ODCM RFO limits, Table 27 Table 27 THEN

1. Operate drywell recirculation fans

2. Vent PC , PPM 5.5.20, PPM 5.5.21, PPM 5.5.16, or 3.

ABN-CONT-VENT lE PC can be vented Il::ifN purge PC by injecting nitrogen into PC

4. lCll:lfl:,J. hydrogen is no longer detected CID PPM 5.5.20 PPM 5.5.21 PPM 5.5.16 in PC (LT 0.6%) (not rad Il::ifN secure vent and purge llll.l required interlocks) by other EOP steps l

Comments /

Reference:

4.814.J2.2-2 Rev: Major: 14 Minor: N/A Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 Number: 4.814.J2 I Use Category: CONTINUOU S Major Rev: 014 Minor Rev: N/A Title : 81 4.J2 ANNUNCIATOR PAN EL ALARMS Page : 7 of 14 2-2 CONTAINM ENT ATMOSPH ERE HYDROG EN LEVEL HIGH 2-2 W INDOW SOURC E AUTOMATIC ACTIONS HYDROG EN CMS-CP-1401 (3.56 %) None LEVEL HIGH NOTE: Detectable hydrogen in the primary containment is ind icative of meta l water reaction (fuel clad oxidation ).

1. CHECK hyd rogen level on CMS-O2/H2R-2 (H13-P8 11) and CMS -CP-1 401.

2. CO MPARE hydrogen level on CMS-O2/H2R-2 (H 13-P811) and CMS-CP-1401 with indications on CMS-O2/H2R-1 (H13-P827 ) and CMS-CP-1301 as follows:

a. SH IFT sam Rle oints on CMS -CP-1301 to match the area being sampled on CMS -CP-1401 by following the locally mounted operator aid (wait 15 minutes fo r valid reading ).

b. CO MPARE the indications on CMS -CP- 1301 and CMS-CP-1401 .

3. lE required ,

TH EN PERFORM channel cali bration (l &C ).

4. lE desired ,

TH EN SAMPLE Containment to verify readings (Chemistry).

5. REMOVE the HWC system from service per SOP-HWC-SHUTDOWN , HWC Shutdown.

6. REFER to PPM 5.8.1, Post-LOCA Hydrogen/Oxygen Monitoring .

7. lE Conta inment hydrogen concentration is GT 3.56% ,

TH EN REFER to PPM 5.2.1 , Primary Conta inment Control.

Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-28 Examination Outline Cross-reference: 28 Revision: 1 Date: 1/6/21 Tier: 2 Group: 1 K/A Number: 203000.K6.01 Level of Difficulty: 3 RO Importance Rating: 3.6 K/A

Description:

Knowledge of the effect that a loss or malfunction of the following will have on the RHR/LPCI: INJECTION MODE: A.C. electrical power CGS is in Mode 1.

• Startup Transformer, TR-S is out of service.

A LOCA occurs. An automatic scram occurs on high drywell pressure.

One minute later, a loss of the Normal Transformer (TR-N) occurs.

If SM-7 and SM-8 are reenergized at time t=0, what is the earliest time that RHR-P-2A will be running?

RHR-P-2A will be running at t=

A. 1 second.

B. 6 seconds C. 10 seconds D. 20 seconds Answer: B K/A Match:

Requires knowledge of how a loss of A.C. power supply affects RHR pump operation.

Explanation:

A. Incorrect. Plausible since RHR-P-2C will be running at 1 second. However, RHR-P-2A and RHR-P-2B have a 5 second delay prior to starting.

B. Correct. RHR-P-2A will start 5 seconds after SM-7 is reenergized from either TR-B or DG-1.

C. Incorrect. Plausible since this is the earliest time that RHR-P-2C would be running if TR-S picked up SM-1 and SM-7. However, TR-S is not available and there is a 5 second time delay for starting RHR-P-2A on TR-B.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-28 D. Incorrect. Plausible since this is the earliest time that RHR-P-2A would be running if TR-S picked up SM-1 and SM-7 (19.4 second time delay). However, since TR-B picked up SM-7, there is a 5 second time delay for starting RHR-P-2A.

Technical Reference(s)

SD000198, Residual Heat Removal System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5779 - Describe the expected system response, for any routine lineup, when the initiation logic for the LPCI mode of the RHR system is satisfied.

Question Source: C Bank #: Bank #

C Modified Bank #: Mod Bank #. (Note changes or attach parent) r New Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level: r Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Requires knowledge of start timers for RHR pumps.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-28 Comments /

Reference:

SD000198 IRev: Major: 17 Minor: 1 C. Interlocks 1 . LPCI Initiation L0-5775 A reactor pressure vessel level of LE minus 129 inches , a high drywell pressure of L0-5779 GE 1.65 PSIG or manual initiation* will automatically align the RHR system for the LPCI mode except when the system is initially in the shutdown cooling mode. L0-11811a Each RHR pmnp will be at rated speed and injection valve open within 46 seconds of an initiation signal; including the time necessary for the DG to start and close L0-1181le on the bus . RHR system response to an initiation signal is as follows:

a) RHR Pump starts when not on TR-S:

(1) RHR-P- C - immediately (2) RHR-P-2A & RHR-P- B - 5 second delay b) RHR Pumps start when on TR-S:

(1) RHR-P-2C - 9 .5 second delay (2) RH R-P-2A & RHR-P-2B - 19.4 second delay!

Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-29 Examination Outline Cross-reference: 29 Revision: 1 Date: 1/5/21 Tier: 2 Group: 1 K/A Number: 205000.K4.01 Level of Difficulty: 2 RO Importance Rating: 3.4 K/A

Description:

Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following: High temperature isolation Which of the following signals will cause an isolation of NS4 Group 6?

(1) RPV Level 3 (LE +13 inches)

(2) Drywell high pressure ( GE 1.68 psig)

(3) RHR Area T high (GE 55°F)

A. (1) ONLY B. (2) ONLY C. (1) and (3) ONLY D. (2) and (3) ONLY Answer: C K/A Match:

Requires knowledge of the interlock that isolates RHR Shutdown Cooling on a High RHR Room Temperature.

Explanation:

A. Incorrect. Plausible since (1) will isolate Group 6. However, it is not the only signal listed that will isolate Group 6.

B. Incorrect. Plausible since drywell high pressure will isolate Group 3, 4, and 5. However, it is not an isolation signal for Group 6.

C. Correct. Group 6 is isolated by the following signals:

• RPV Level 3 (LE +13 inches)

• Reactor Pressure (GE 125 psig)

• RHR Area Temp (GE 130°F to 150°F)

• RHR Area T (GE 55°F)

• Excess RHR SDC Flow (GE 6.28 psid)

D. Incorrect. Plausible since (3) will isolate Group 6. However, (2) is not an isolation signal for Group 6.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-29 Technical Reference(s)

SD000173, Nuclear Steam Supply Shutoff System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11811 - Describe the Residual Heat Removal System design features and/or interlocks which provide for the following: j. high temperature isolation in SDC mode.

Question Source: OBank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

New Question History: Last NRC Exam: N/A Question Cognitive Level: ' Memory or Fundamental Knowledge 0 Comprehension or Malysis Justification for Cognitive Level Requires knowledge of the NS4 isolation circuitry for RHR in Shutdown Cooling mode.

10 CFR Part 55 Content: 55.41 7 Comments /

Reference:

SD000173 Rev: Major: 14 Minor: 5 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-29 COLUMBIA SYSTEMS June 2018 NS4 SD000l 73, r14 mr5 NS4LOGIC I PUT PARANIETERS A D ACTUATION SETPOINTS Group 1 Group 2 Group 3 Group 4 Group 5 GroJ!P 6 Group 7 eactor [ evel -129" -50" -50" -50" +13 " 13" -50" Reactor Pressure 25#

Drywell Pressure 1.68# 1.68# 1. 68#

MSLFlow 1 129 psid MS Tunnel Temp 164°F MS T tlllllel I':,. T goop 2.864X MSL Radiation 100%

MSL Pressure 831 #

Mn Cond Vacuum 21.6" Hg RB Vent Exh Rad 13 mr/ln*

ff0°F tO RHR Area Temp 150°F HR Area 1':,.T 55°F RHR SDC Flow 6.28 psid RWCU PumpRm 160°F Temp RWCU HXA..rea 150°F Temp RWCU PumpRm 70°F/60°F HXt:,_T RWCU t:,_ fl ow 2 58 .5 gpm RWCU Pipe .Area 160°F Temp RWCU/RCIC 160°F Pipe Area Temp RWCU BID Flow 3 253.5 gpm NRHXOutlet 140°F Temp 4 SLC start 4 Keyswitch Note 1: MSL Flow high, 129 psid = 140% of nominal (in any main steam line)

Note 2: RWCU t:,_ Flow isolation occurs after a 45 sec. TD Note 3: RWCU BID Flow isolation occurs after a 1.6 sec. TD Note 4: NRHX Outlet Temp high and SLC start - d oses RWCU-V-4 only Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Examination Outline Cross-reference: 30 Revision: 2 Date: 1/19/21 Tier: 2 Group: 1 K/A Number: 209001.K5.05 Level of Difficulty: 3 RO Importance Rating: 2.5 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM : System venting CGS is in Mode 1.

An event occurs which causes the following:

LPCS-Fl-600 LPCS-Pl-3 LPCS LPCS/RHRA LEAK DET PUMP P-1 MAN INITIATION AUX STEAM TRIP SWITCH ARMED TEMP HI-HI

• LPCS/RHRA DRYWELL LPCS PUMP DISCH LEAK DET RCC LEAK RHR EQl PRESS CONT SUPPLY PRESS HIGH HIGH/LOW /:,.THIGH l:.T LPCS/RHRA INIT LPCS OUT OF

• LPCS PUMP RM WATER LEVEL

• LEAK DRYV 60 RPV LEVEL LOW FLOOR SERVICE HIGH

-129" FLOW ANN PANEL P601-A3 40 20 RHRPMP2A RMHVAC WMA-Ftl-53A PWR LOSS RRA-FN-ll PWRLOSS 0

OUTOF SERV CS-RHR2A 0G I BATT SH SW A

• . FLOW PRESSURE OUT OF SERV OUT OF SERV OUT OF SERV OUT OF SERV LPCS-P-1

~~RA ~~~~A MOVNETWORK LPCS-P-2 N TEST STATUS PWR LOSS/Ol PWR LOSS/OL RHR-V-4A MANUAL LAMP TEST NOT FULLY OUTOFSERV OPEN RHRA BYPASS AND INOPERABLE STATUS DISPLAY What action should be performed to mitigate this event for current plant conditions?

A. Manually start LPCS-P-1. Run the pump until the Keep Fill system is restored.

B. Pressurize the LPCS system using the Condensate Storage and Transfer system .

C. Prevent LPCS-P-1 and RHR-P-2C from automatically starting by removing control power fuses.

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 D. Fill the LPCS system with RHR-P-2A in accordance with ABN-LPCS-DEPRESS, LPCS Recovery Following Depressurization from Keep Fill Failure.

Answer: A K/A Match:

Requires knowledge of actions required when system used to keep LPCS system filled and vented fails.

Explanation:

A. Correct. In accordance with Alarm Response Procedure (ARP) 4.601.A3.6-3, LOW PRESSURE CORE SPRAY OUT OF SERVICE, Page 7 for LPCS BYPASS AND INOPERABLE STATUS PANEL (BISP) light for Loss of Power for LPCS-P-2, if the LPCS Pump Discharge Pressure Hi/Low Annunciator (4.601.A3.5-3) is not illuminated, starting LPCS-P-1 in Suppression Pool Mixing should be considered. This is the preferred action for the conditions given since starting the pump will maintain LPCS operability and allow the system to be used immediately while repairs are completed.

B. Incorrect. Plausible since the Condensate Storage and Transfer system is used to fill LPCS following maintenance. However, precaution 4.3 in SOP-LPCS-FILL, LPCS Fill and Vent, prohibits using this system to keep LPCS pressurized.

C. Incorrect. Plausible since RHR-P-2A uses the same keep-fill system as LPCS-P-1 and control power fuses for both of these pumps should be removed when discharge pressure reaches the low pressure alarm setpoint. However, RHR-P-2C uses the other divisions keep-fill system and control power fuses for this pump should not be removed for conditions given in the question stem.

D. Plausible since ABN-LPCS-DEPRESS allows using RHR-A to pressurize LPCS. However, this procedure is entered after LPCS is depressurized due to keep-fill system failure (i.e. the LPCS low pressure alarm is in).

Technical Reference(s) 4.601.A3, Alarm Response Procedure for panel 4.601.A3 SOP-LPCS-FILL, LPCS Fill and Vent Attached w/ Revision #

ABN-LPCS-DEPRESS, LPCS Recovery Following Depressuration See Comments / Reference from Keep Fill Failure Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11590 - Describe the effect that a loss of malfunction of the following will have on the Low Pressure Core Spray System: c. Keep fill system Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Question Source: (' Bank #: Bank #

(' Mod ified Bank #: Mod Bank #. (Note changes or attach parent) r New Question History: Last NRC Exam: N/A Question Cognitive Level: r. Memory or Fundamenta l Knowledge r Com prehens ion or Analys is Justification for Cognitive Level Requires knowledge of actions required when the LPCS keep-fill system fails.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

4.601.A3 IRev: Major: 028 Minor: N/A Number: 4.601 .A3 I Use Category: CONTI NUOUS Major Rev: 028 Minor Rev: N/A

Title:

601 .A3 ANNUNCIATOR PAN EL ALARMS Page : 60 of 82 6-3 LOW PRESSURE COR E SPRAY OOT OF SE RVICE 6-3 WINDOW SOURCE AUTOMATIC ACTIONS LPCS Any of the following alarms on the LPCS Bypass and None OUT OF Inoperable Status Panel :

SERVICE

• WMA-FN-53A PWR LOSS (Pg 1)

• RRA-FN-11 PWR FAIL (Pg 2)

• LPCS PU MP RM HVAC OUT OF SERV (Pg 3)

• LPCS LOG IC PWR FAI L (Pg 4)

• LPCS/RHR A LOG IC IN TEST (Pg 5)

• LPCS-V-1 NOT FULLY OPEN (Pg 6)

• LPCS-P-2 PWR LOSS/OL (Pg 7

• MANUAL OUT OF SERV (Pg 8)

• MOVN ETWORK PWR LOSS/OL (Pg 9)

• CB-LPCS OUT OF SERV (Pg 10)

(LPCS-RL Y-K21 )

Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Number: 4.60 1.A3 I Use Category: CONTINUOUS Major Rev: 028 Minor Rev: N/A Title : 601 .A3 ANNUNCIATOR PANEL ALARMS Page: 67 of 82 LPCS BYPASS AND INOPERABLE STATUS PANEL W INDOW SOURCE AUTOMATIC ACTIONS LPCS-P-2 Loss of power to LPCS-P-2. None PWR LOSS/OL (LPCS-RL Y-49X/P2)

1. !E LPCS and/or RHR Loop A Pump Discharge Pressure Hi/Low Annunciator is not illuminated ,

TREN CONSIDER starting LPCS-P-1 and RHR-P-2A per SOP-LPCS-SP, lPCS Suppression Pool Mixing and SOP-RHR-SPC Suppression Pool Cooling/Spray/Discharge/Mixing to maintain operability.

Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Number: 4.601 .A3 I Use Category: CONTINUOUS Major Rev: 028 Minor Rev: N/A

Title:

601.A3 ANNUNCIATOR PANEL ALARMS Page : 46 of 82 5-3 LPCS PUMP DISCHARGE PRESSURE HIGH/LOW 5-3 WINDOW SOURCE AUTOMATIC ACTIONS LPCS LPCS-P-1 (LPCS-PIS-5):

PUMP DISCH None PRESS High Discharge Pressure GE 453 psig HIGH/LOW (432 psig + 21psig head correction )

or Low Di scharge Pressure LE 75 psig (54 psig + 21psig head correction )

CAUTION If LPCS PUMP DISCHARG E PRESSU RE HIGH/LOW annunciator alarms, LPCS-P-1 may be started if needed for EOP related activities. However if the system is partially voided prior to pump start, a water hammer following pump start could break the LPCS piping or components causing flooding and a reduction in suppression pool inventory.

NOTE: LPCS-PIS-5 indication of 453 psig corresponds to a trip setpoint of 442 psig at sensing line tap elevation (449' 6").

1. CHECK LPCS System pressure at LPCS-Pl-3 (H13-P601 ). LPCS-PIS-5 (H22-P001 , RB 471 ), or TDAS No. X158.

2. lE. a system transient has caused a momentary (clears in LT 10 seconds) low pressure alarm, AND pressure does not drop below 40 psig ,

TH EN EXIT this procedure .

3. lE. LPCS pressure is low, TH EN PERFORM the following :

a. lE. not operating the system per the EOPs, THEN INHIBIT LPCS-P-1 start by pulling the control power fuses .

b. CHECK operation of LPCS-P-2 (Water Leg Pump).

c. VERIFY LPCS-V-12 is CLOSED .

d. lE required ,

THEN REFER to ABN- LPCS-DEPRESS , LPCS Recovery Following Depressurization from Keep Fill Failure.

Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Comments /

Reference:

SOP-LPCS-FILL IRev: Major: 15 Minor: N/A Number: SOP-LPCS-FILL I Use Category: CONTINUOUS Major Rev: 015 Minor Rev: N/A

Title:

LPCS Fill and Vent Page : 6 of 21 NOTE: If the system is partially drained, the System Engineer will determine which points need ultrasonic inspection .

3.9 NOTIFY NOE of requirement to perform ultrasonic inspection of suspect points following system fill , prior to declaring the LPCS system operable . --

3.10 VERIFY a Work Order is in place to install scaffolding and remove insulation, as necessary to perform ultrasonic inspection . --

NOTE: If bringing GT 40 lbs of rubber hose into the Reactor Building, then a temporary combustible permit is requi red per PPM 1.3. 10C.

3.11 !E necessary, THEN OBTAIN a temporary combustible permit per PPM 1.3.10C. --

4.0 PRECAUTIONS AND LIMITATIONS 4 .1 When the LPCS system is drained , either the Condensate Storage and Transfer system or LPCS-P-2 may be used to fill the LPCS system. Do not use both systems to fill the system at the same time. The CRS/Shift Manager will determine which method will be used to fil l the LPCS system .

4.2 If LPCS has only been partially drained , air is likely to be trapped du ring refill. Separate instructions may be necessary to verify LPCS system is acceptably full.

4.3 Due to piping qualifications, do not use the Condensate Storage and Transfer System to keep the LPCS system pressurized .

4.4 Per the M-rule, the unavailability of LPCS-SYS-1, when it is requ ired to be operable, is requ ired to be tracked in the electronic logging system. The time it is unavailable should be minimized .

Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Comments /

Reference:

ABN-LPC-DEPRESS IRev: Major: 002 Minor: 001 Number: ABN-LPCS-DEPRESS I Use Category: CONTINUOUS Major Rev : 002 Minor Rev : 001 Title : LPCS Recovery Following Dep ressuration fro m Keep Fill Fai lure Page : 3 of g 1 .0 ENTRY COND ITIONS LPCS system depressurization due to keep fill system failure .

2.0 AUTOMATIC ACTIONS None 3.0 IMMEDIATE OPERATOR ACTIONS None 4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 lE LPCS keep fill system fails ,

AND cannot be returned to operable status ,

THEN PERFORM either Section 7.1 or 7.2.

5.0 BASES This procedure describes how to return LPCS to operable status following a failure of the keep fil l system (keep fill pump failure). The low pressure alarm (4.601 .A3-5.3) requires LPCS-P-1 start to be inhibited by pull ing the control power fuses until the system can be filled and vented or the pump is ready to be started in a controlled manner. Pressurizing the LPCS system from the Demineralized Water system or RHR A system will limit any potential water hammer when the LPCS pump is subsequently started.

Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 Examination Outline Cross-reference: 31 Revision: 2 Date: 1/19/21 Tier: 2 Group: 1 K/A Number: 209002.K4.02 Level of Difficulty: 2 RO Importance Rating: 3.4 K/A

Description:

Knowledge of HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) design feature(s) and/or interlocks which provide for the following: Prevents over filling reactor vessel CGS is in Mode 1

• A LOCA occurs

• HPCS initiates on low RPV level.

• Drywell Pressure is 1.2 psig, up slow.

• RPV Level is +30 inches, up fast.

The HPCS RPV Injection Valve, HPCS-V-4, automatically closes when RPV level reaches Level-8.

• RPV level is 55 inches, down slow.

When will HPCS-V-4 to automatically re-open?

HPCS-V-4 automatically opens when RPV level reaches...

A. Level 1.

B. Level 2.

C. Level 3.

D. Level 4.

Answer: B K/A Match:

Requires knowledge of the automatic operating setpoint of HPCS-V-4, which prevents over filling the RPV while maintaining RPV level.

Explanation:

A. Incorrect. Plausible since Level 1 is the water level that is high enough bove TAF to allow sufficient time for the low-pressure ECCS systems (RHR, LPCS) to establish adequate core cooling to prevent core damage. However, HPCS-V-4 opens at RPV Level 2 (-50 inches)

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 B. Correct. After HPCS initiates, HPCS-V-4 will automatically close when RPV level reaches Level 8

(+54.5 inches). HPCS-V-4 will remain closed until RPV level goes below Level 2 (-50 inches) to prevent over filling the RPV.

C. Incorrect. Plausible since Level 3 is the automatic scram setpoint to ensure a quantity of water in the RPV above TAF to account for evaporation due to decay heat boil-off, steam void collapse, etc., without water level lowering sufficiently to automatically start low pressure ECCS. However, HPCS-V-4 opens at RPV Level 2 (-50 inches)

D. Incorrect. Plausible since this is the Low RPV water level alarm setpoint. However, HPCS-V-4 opens at RPV Level 2 (-50 inches)

Technical Reference(s)

SD000126 - Nuclear Boiler Instrumentation SD000174 - HPCS Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: Learning Objective.

Question Source: C. Bank #: LX00777

(' Modified Bank #: Mod Bank #. (Note changes or attach parent)

(' New Question History: Last NRC Exam: N/A Question Cognitive Level: (. Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Candidate must recall HPCS-V-4 auto open setpoint 10 CFR Part 55 Content: 55.41 8 Comments /

Reference:

SD000126 Rev: Major: 13 Minor: 2 Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 C OLUMBLA.. SYSTEMS Fehruru:y Oil 6 NiBI SDOOO 1.!!6 , d3 m:r2 Table 1 RPV w*ate.r Leve] Trip Summ.aiy RPV Water Level Trip Swumaiy LEVEL Level 8 (:54..5 ")

• TRIP FUNCTION Trip main m:rbine .

BASIS The high water level trip proteots ire

• RCIC rurbine isolation amin turbine against moisture canyover

• Trip foecfwater {RFW) pwups a.ad subrequem danl.8:ge to, turbine

• Cfose HPCS mject ion valve

,. bfading .

The RF¥l pomps a re tripped to prevent

,. RPV orVerl'ill.

Steain to, the RCIC ru11bme is. secured, a.ad HPC3 iajection valve is closed to prevent RP'V o!Veffill am. fl.oodin.g of the Level 7 (40..5 ")

• High RPV water level alam1 .

RCIC m:rbine s team lme.

The water l evel afocm annunciates at the reactor ve.sse] wate.r leve] a1borVe* which moisture* carry o,ve.r in the stean1 is expected ro increase at a signifi:caat cate .

The al arm wams the operator of th.is Level 5 , 6

• Upper 8!1lrl lower bounds ,o f normal! RP'V wate.r leve]

undesirn.bile condition..

The FWLC is nomia))y ,o perated based on maintaining water Ieve] near leve] 5 (36" ,_ Operating at this point nllllrimiz,es Level 4 (31. 5' )

• Low RPVwater level -alarm .

canyovei and steam can.-yrn:der _

The wnter lev el ruarm annunciates nt the

• RRC lrllUl!back on RFVl hllibme water level below which steam trip C.81" ru.nder in . water will be~

affecting recirc ulation flow rate significantly at fiill power due to RRC

,. pump cavitation.

A wate.r level decrease to this point, coupled with a trip, of a RF\V turbine, causes RRC pumps to rnnhack to, 30 hertz. Tius redures core flow 8!1lrl thns reactm po,we.r lie place* it witihm the*

capacity of the renll.8.Uling RFW p001p.

Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 Taibi,e 1 (mnf ,~),

RP'V Watec Leve]

• RPVWat LEVEL TRIP' FUNCTIO B..A.SJ S Level 3 ( 13" ,

• Reactor scnun

• The scram function OCClu-s while w ater

• NS4 Groups 5

• 6 isolation level is still above the ibotiom of the

• ADS RPV !'.eve] ,c onfum.ation steam dryei- skict , which 1)1"ecluides high

• RRC nm!back to, 15 hertz moisane carryovei- due to steam b rpa:ss:mg the dryer uocler the seal skirt.

'Ilw vel also re~ ts in a quari.h fy of reserve coolant between this level and

'I.'\F to account for evaporatron {decay heat boil-o los ses, steam void collapse , and other coolant losses from the RPV following a os s of RFW event, without water level decreasiflg to Level 1, wliidi v;roulo mitiate E OCS.

This quantity of resenre* coolant asswnes RCIC is capable of providing adeqlllate makeup .

• Trus level also causes NS4 isolations to prevent a loos of ooolant from these*

potentiia] leakage paths.

• A Leve] 3 signa] is wed by the ADS logic to confu1111. fhat in deed a ]ow

,varer level does erust.

• The R.R!C ~ ramp down to, 15 hertz to 1)1"event pmup ,c avitation due to poor N PSH .

Level 2 (-50"

• Initiate R CIC

• This retpoint was selected so that

• Truiiate HPCS RCIC/HPCS would not initiate after a

• Start H P C S DG s.cram with F W avrulable. It i.s. *g.Ji

• Isolate NS4 Gcuups 2 , 3 4

• 1 enough so that RCIC could vent a

• T:rjp RRC puJllJIC, leveloecrease to Level 1 - L g")

• Initiate ATV,lS/ ARI following a loss of feedwater event .

• This level also causes NS4 isolations oo prevent a loos of ooolant foo:m these*

potentia] leakage paths .

• RRC trip at this Level adds negatire react ivity to the* oore .

• ATWS/ARI is a ibackup to the RPS scram which shmild have ocourred .at L3 13" ,_

Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 RPV Wate.r LEV EL TIUP' F UNCTIO Level 1 ( -129'")

• Initiate RHR & LPCS

• This watec level retpoint is selected to

• Initiate AIDS be nigh enoll@ above TAP to allow

• Strut DGl , IDG sufficient time for the low-pressure

• Lsolate S4 Grnup 1 ECCS systems {RHR, I:PCS) to establish adegu.ate core cooling ru:d thus prevent core damage .

• 'T he Level 1 si;gnn] is aho required by

• .i\DS initiation to permit ilese Iow-p res~ure EOCS systen1S to inject into, the RPV .

• .i\ fao this 1eve] setp:)mt is used to sw t IDGl am IDG2, whic.h ,c an ~upp]y elect:ric.ru pov;re:r to fue low-pressure E OCS system buses .

• M SIV s & M:Sl.Ds d ose to, prevent a loss of ooolant from these* p oteruial leaka 0 e <nths .

Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 Comments /

Reference:

SD000174 IRev: Major: 16 Minor: 3 COLUMBIA SYSTEMS January 2020 HIGH PRESSURE CORE SPRAY SD000 174, r16 mr3 The system is designed to pump water into the reactor vessel over a wide range of reactor pressure. For small breaks that do not result in rapid reactor depressurization, the system maintains reactor water level and, due to injection of relatively cool water, aids in depressurizing the vessel. For large breaks the HPCS system cools the core by spraying water directly onto the core.

When the HPCS system initiates_, the HPCS injection valve , HPCS-V-4 , opens and remains open until reactor water level reaches + 54.5" (Level-8) at which point the valve will automatically close to revent over filling the reactor vessel. If reactor water level begins to lower, the injection valve will be signaled to re-open when the system initiation point , -50 "

(Level-2), is reached. The initiation signal will override the previous valve closure signal causing the injection valve to open and restore injection flow to the reactor. To return the injection valve to manual control with the high level closure signal sealed-in and tl1e low level initiation sealed-in, both seal-in logic reset buttons must be depressed while reactor level is between Level-2 and Level-8. Closing HPCS-V-4 , while an initiation signal is present, will prevent it from reopening on a subsequent low level. SOP-HPCS-INJECTION Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 COLUMBIA SYSTEMS January 2020 HIGH PRESSURE CORE SPRAY SD000174 , r16 mr3 V. CONTROL THEORY AND INTERLOCKS HPCS uses separate logic configurations for initiation, suction source transfer and injection valve closure .

HPCS hlitiation on low reactor water level or high drywell pressure occurs on one-out-of-two taken twice logic. (Fig 3) The low water level and high pressure logic trains are independent arrangements requiring the one-out-of-two taken twice logic requirement to be satisfied with either all level channels or all pressure channels, but not a combination of level and pressure channels. Manual initiation of HPCS is one-out-of-one taken once.

Suction source transfer of the HPCS system on high Suppression Pool level(:: : : +5 "), low CST water level (1 '9"), or CST supply line break occurs on one-out-of-two taken once logic.

Closure of the HPCS iitjection valve on high reactor water level occurs on two-out-of-two taken once logic.

A. Control Room Controls

1. HPCS Manual Initiation Pushbutton Pushbutton with 2-position collar DISARMED/ ARMED DISARMED The pushbutton is disabled ARMED Causes P601-Al 3.8 to alann. Depressing pushbutton in this position causes the start of HPCS-P-1, DG-3 , and opens the HPCS injection valve, HPCS-V-4.

2. HPCS pump (HPCS-P-1 )

3-position switch START/AUTO/STOP (spring return to AUTO)

START Pump starts AUTO Starts on initiation signal STOP Pump stops

3. RPV Injection Valve (HPCS-V-4) 3-position switch CLOSE/AUTO/OPEN (spring return to AUTO)

CLOSE The valve closes and overrides the Level-2 (-50 ") open feature .

AUTO Closes at Level-8 ( + 54 .5 ") . 0 ens on initiation and anytime level LE (-50 ") if not over-ridden.

OPEN Opens if LT +54.5" and level 8 closure signal reset.

Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-32 Examination Outline Cross-reference: 32 Revision: 1 Date: 1/6/21 Tier: 2 Group: 1 K/A Number: 211000.A4.03 Level of Difficulty: 2 RO Importance Rating: 4.1 K/A

Description:

Ability to manually operate and/or monitor in the control room: Explosive valves firing circuit status CGS is in Mode 1.

In response to a LOCA the Standby Liquid Control (SLC) system is in the following condition:

CIRCU IT CIRCUIT READY READY TEST TANK OUTLET SLC-V-31 SLC-V-4A SLC-V-4B SQUIB VALVE STORAGE TANK 2 OUTLET SLC SYSTEM A SLC SYSTEM B SLC-V-1A OFF OPER OFF OPER STORAGE TANK OUTLET SLC-V-1B (1) What is the status of the Squib Discharge Valves?

(2) What is the status of the SLC Injection Pumps?

A. (1) SLC-V-4A is OPEN.

(2) ONLY SLC-P-1A is injecting.

B. (1) SLC-V-4A is OPEN.

(2) BOTH SLC-P-1A and SLC-P-1B are injecting.

C. (1) SLC-V-4B is OPEN.

(2) ONLY SLC-P-1B is injecting.

D. (1) SLC-V-4B is OPEN.

(2) BOTH SLC-P-1A and SLC-P-1B are injecting.

Answer: D Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-32 K/A Match:

Requires knowledge of the indications that a SLC Squib Injection Valve is open/closed.

Explanation:

A. Incorrect. Plausible if it is believed that the squib valve Circuit Ready light is ON when the squib has fired and that the squib injection valve must be open for the associated injection pump to inject. However, the SLC-V-4A "Circuit Ready" light ON indicates that the squib for SLC-V-4A did not fire and the valve is closed. Additionally, there is a cross connect on the pump discharge piping and both pumps will inject through the open squib valve (SLC-V-4B).

B. Incorrect. Plausible since there is a cross connect on the pump discharge piping and both pumps will inject through the open squib valve (SLC-V-4B). However, the SLC-V-4A "Circuit Ready" light ON indicates that the squib for SLC-V-4A did not fire and the valve is closed.

C. Incorrect. Plausible since SLC-V-4B is open. However, there is a cross connect on the pump discharge piping and both pumps will inject through the open squib valve (SLC-V-4B).

D. Correct. When SLC System A and B switches are in OPER, SLC-V-1A & B are full open, and SLC-V-31 is closed, SLC-P-1A & B will start. The SLC-V-4B "Circuit Ready" light OFF indicates that the squib for SLC-V-4B fired and the valve is open. There is a cross connect on the pump discharge piping and both pumps will inject through the open squib valve (SLC-V-4B).

Technical Reference(s)

SD000172 - Standby Liquid Control Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5923 - Describe the operation of the following SLC System components:

a. Pumps, d. Explosive Valves Question Source: 0Bank #: Bank #

0 Modified Bank #: Mod Bank #. (Note changes or attach parent)

New Question History: Last NRC Exam: N/A Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-32 Question Cognitive Level: 0 ~mory or Fundamental Knowledge

~ Comprehension or Analysis Justification for Cognitive Level Requires synthesizing system information given in the stem with a knowledge of SLC squib valve indicatons and SLC pump configuration.

10 CFR Part 55 Content: 55.41 8 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-32 Comments /

Reference:

SD000172 IRev: Major: 13 Minor: 1 COLUMBIA SYSTEMS October 2014 STANDBY LIQUID CONTROL SD000172 , r13 mrl Chapter 9.3 of FSAR states that with both pumps running together, the injection rate into the vessel is about 87 to 91.1 gpm (this is with demineralized water). Calculations show that the pumps are capable of full-flow injection at reactor pressures up to 1248 psig without lifting the SLC relief valves (set at about 1400 psig).

3. Ensuring Full-flow Injection The pump discharge piping is crosstied to ensure full flow to the vessel even if one explosive valve fails to open. A check valve in each pump 's discharge pipe prevents potential bypass flow around an idle pump through a failed relief valve.

Starting either SLC pump from P603 causes both Storage Tank outlet L0-5923b,c valves , SLC-V- lA and 1B to open. Suction piping is arranged so that a NL0-12852d failure of either SLC-V- lA or 1B to open does not prevent establishing a flowpath to the vessel. The Storage Tank outlet valves are interlocked such that they will not open if the Test Tank outlet valve, SLC-V-31 , is not fully closed. Valve SLC-V-31 is normally locked closed.

2. SLC Injection (Figure 1)

L0-5922 a To initiate SLC from the main control room , the control operator NL0-12853 manually starts both SLC pumps with the associated keylocked control NL0-1285la switches on panel P603. Once these control switches are rotated to their OPERA TE position, SLC injection occurs from the Storage tank ,

through the SLC pumps , squib valves and into the HPCS sparger ring .

SLC system response to initiation:

• SLC-V- lA & - lB Open (if SLC-V-31 CLOSED)

• SLC-V-4A & -4B fire Open o S uib valve continuity lights will go out o BISI display alarms SLC-V-4A (4B) LOSS OF CONTINUITY flashing Page 18 of 26 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-33 Examination Outline Cross-reference: 33 Revision: 2 Date: 1/19/21 Tier: 2 Group: 1 K/A Number: 212000.A4.01 Level of Difficulty: 2 RO Importance Rating: 4.6 K/A

Description:

Ability to manually operate and/or monitor in the control room: Provide manual SCRAM signal(s)

CGS is in Mode 1.

The crew needs to insert a manual reactor scram.

When considered separately, which of the following combinations of Reactor Scram pushbuttons on H13-P603, when depressed simultaneously, will initiate a reactor scram?

(1) Logic A1 and Logic A2 (2) Logic B1 and Logic B2 (3) Logic A1 and Logic B2 (4) Logic A2 and Logic B2 A. (1) and (2) ONLY B. (1) and (4) ONLY C. (2) and (3) ONLY D. (3) and (4) ONLY Answer: D K/A Match:

Requires knowledge of the method of manually scramming the reactor using Reactor Scram pushbuttons.

Explanation:

A. Incorrect. Plausible if it is believed that both pushbuttons from the same Trip System (A or B) must be depressed to initiate a scram. However, this will only cause a half-scram on Trip System A.

B. Incorrect. Plausible since (4) will cause a scram. However, (1) will not cause a scram.

C. Incorrect. Plausible if it is believed that depressing any two button from different Trip Channels (1 and 2) will cause a scram. However, buttons from different Trip Channels (A and B) must be depressed simultaneously to cause a scram.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-33 D. Correct. Each RPS is divided into two Trip Systems, A and B. Each Trip System is comprised of two redundant Trip Channels, 1 and 2. To initiate a scram, one Trip Channel from each Trip System must be activated. Depressing any combination of either A Trip System button, A1 or A2, and either B Trip System button, B1 or B2, simultaneously causes a full scram. Therefore, (3) and (4) will initiate a reactor scram.

Technical Reference(s)

SD000161, Reactor Protection System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11681 - Describe the function, purpose, and design features of the major RPS components: l. Manual Scram pushbuttons.

Question Source: C Bank #: Bank #

r Modified Bank #: Mod Bank #. (Note changes or attach parent)

(. New Question History: Last NRC Exam: N/A Question Cognitive Level: r- Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Requires knowledge of the function of the Reactor Scram pushbuttons.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-33 Comments /

Reference:

SD000161 Rev: Major: 17 Minor: 5 B. RPS Trip Systems and Logic LO-l 168la

1. The RPS trip logic [Figure 1) consists of two tri systems, RJ>S trip system NL0-12526d "A" and RPS trip system "B , with each trip system comprised of two redundant tri channels:

• RPS "A" trip system consists of "A 1" ana "/L" trip challllels.

• RPS "B" trip system consists of "Bl" and "B2" trip channels -1

2. REACTOR SCRAM pushbuttons - H l3 -P603 L0-5953 L0 -1168 11 Four buttons paired as Al/Bl and A2/B2. One pushbutton is assigned to each RPS trip channel. e ressing any combination of Al or/Land Bl or B2 causes a full scram. However, the pushbuttons are grouped and labeled as fo llows:

• Left side A l and Bl labeled - PUSH BOTH A l /Bl TO SCRAM

• Right side A2 and B2 labeled - PUSH BOTH A2/B2 TO SCRAM Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-34 Examination Outline Cross-reference: 34 Revision: 0 Date: 5/28/20 Tier: 2 Group: 1 K/A Number: 215003.K1.05 Level of Difficulty: 2 RO Importance Rating: 3.3 K/A

Description:

Knowledge of the physical connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following: Display control system Intermediate range monitor (IRM) channel "E" is currently on range 7 reading 10/40 scale.

How does this indication change when the IRM scale is changed?

If channel "E" is placed in range (1) , it will be reading (2) scale.

A. (1) 6 (2) 1/125 B. (1) 6 (2) 10/125 C. (1) 8 (2) 10/125 D. (1) 8 (2) 100/125 Answer: C K/A Match:

Requires understanding of changes in IRM display when the range switch is operated.

Explanation:

A. Incorrect. Plausible if it is believed that the IRM indication would lower when going on a lower scale. However, taking IRM E to range scale would increase the displayed value to 100/125 scale.

B. Incorrect. Plausible if it is believed that the odd ranges are expanded versions of the range below them. However, they are an expanded version of the scale above. Going from an odd scale to the even scale below will cause output to increase by a factor of 10.

C. Correct. IRM odd ranges are an expanded version of the even scale above. The even scale is the same as the 0-40 portion of the odd scale above. Therefore when the IRM reads 10/40scale on range 6, taking the range switch to the next odd range above (range 8) will cause the display to read 10/125 scale.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-34 D. Incorrect. Plausible if it is believed that taking the IRM from an odd range to the next higher even range will cause output to increase by a factor of 10. However, taking the range switch from an odd range to the next lower even range will cause output to increase by a factor of 10. Going to the next higher even range will give an output that is the same, but on a larger scale. 10/40 scale on range 7 will go to 10/125 scale on range 8.

Technical Reference(s)

SD000138, IRM System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5453 - Describe the relationship of readings between IRM even/odd ranges, even/even ranges, and odd/odd ranges.

Question Source: r- Bank #: LO00650 C Modified Bank #: Mod Bank #. (Note changes or attach parent)

C New Question History: Last NRC Exam: N/A Question Cognitive Level: r- Memory or Fundamental Knowledge r Comprehension or Ana lysis Justification for Cognitive Level Requires knowledge of the operation of the IRM range switches.

10 CFR Part 55 Content: 55.41 6 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-34 Comments /

Reference:

SD000138 IRev: Major: 10 Minor: 2 COLUMBIA SYSTEMS August 2019 IRM SD000138 , rlO mr2 A method for determining the relationship between IRM indication and reactor power is to remember that 25 on Range 8 is equivalent to 1 %

reactor power; the point of adding heat.

Range 1, from 0 to 40 indicates 0 to 1. 6 X 10-3 %

Range 2, from Oto 125 indicates0 to 5.0 X 10-3 %

Range 3, from 0 to 40 indicates 0 to 1 .6 X 10-2 %

Range 4, from Oto 125 indicates0 to 5.0 X 10-2 %

Range 10, from Oto 125 indicates 0 to 50 %

c) The 0-40 scale is the same as the 0-40 indication of the 0- _5 scale, but expanded for easier viewing. The 0-40 scale for each IRM recorder allows movement between IRM even ranges while avoiding rod blocks and scrams . The following table illustrates the results of ranging between ranges 7 and 10. This same relationship exists between all ranges.

IRM Indication Scale Actual %

Range Power 10 7.5 0- .%

L5 9 7.5 0 - 40 3%

8 75 0-  %

125 7 OFF SCALE HI 0 - 40 3%

Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-35 Examination Outline Cross-reference: 35 Revision: 1 Date: 1/5/21 Tier: 2 Group: 1 K/A Number: 215004.K2.01 Level of Difficulty: 2 RO Importance Rating: 2.6 K/A

Description:

Knowledge of electrical power supplies to the following: SRM channels/detectors What is the power supply to SRM Channel C?

A. RPS-A B. RPS-B C. DP-SO-A D. DP-SO-B Answer: C K/A Match:

Requires knowledge of SRM power supplies.

Explanation:

A. Incorrect. Plausible since Division 1 Power Range Nuclear Monitoring (PRNM) is powered from RPS-A. However, Division 1 SRMs are powered from DP-SO-A.

B. Incorrect. Plausible since RPS-B is the power supply to one Rod Block Monitor interface panel and Division 2 PRNM. However, the power supply to SRM Channel C is DP-SO-A.

C. Correct. In accordance with the Source Range Monitor System Description (SD000132), the power supply to Division 1 SRM Channels A and C is DP-SO-A.

D. Incorrect. Plausible if it is believed that SRM Channel C is Division 2. However, it is Division 1.

Technical Reference(s)

SD000132, Source Range Monitors Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 12000 - State the power supplies to the following Source Range Monitoring components:

Page 1 of 2

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-35

a. SRM channels/detectors

b. Detector drive modules

c. Detector drive module control Question Source: r. Bank #: LO03583 C Modified Bank #: Mod Bank #. (Note changes or attach parent)

C New Question History: Last NRC Exam: N/A Question Cognitive Level: r Memory or Fundamental Knowledge r Comprehension or Analysis Justification for Cognitive Level Requires a knowledge of the power supply to SRM Channel A.

10 CFR Part 55 Content: 55.41 7 Comments /

Reference:

SD000132 Rev: Major: 12 Minor: 2 IX. PO'WER. SLlPPLIE LO-l [OJ A _ SRl\;f A , B, C mu:l D Detecto:r Drives - PP-SC-AA B _ SRl\-1 ~ s Jii.. and Div_ LO- L 004b C. SRl\;f Chrtnne]s Band D - _4 VDC Dist_ Pue1 DP-.30-B , Div_ 2)

D _ A loss of 2 VDC po,¥ec supply v;rill resuli: in an OP rrip LO- L 005a Page 2 of 2

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-36 Examination Outline Cross-reference: 36 Revision: 0 Date: 5/28/20 Tier: 2 Group: 1 K/A Number: 215005.A3.07 Level of Difficulty: 4 RO Importance Rating: 3.8 K/A

Description:

Ability to monitor automatic operations of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM including: RPS status CGS is in Mode 1.

A transient has resulted in the following:

• Core Flow: 50%

• APRM-1 Reactor Power: 86%

• APRM-2 Reactor Power: 87%

• APRM-3 Reactor Power: 88%

• APRM-4 Reactor Power: 95%

What is the effect of these APRM outputs?

A rod block (1) present. A half-scram (2) present.

A. (1) is (2) is B. (1) is (2) is not C. (1) is not (2) is D. (1) is not (2) is not Answer: B K/A Match:

Requires understanding of the relationship between APRM output and RPS status.

Explanation:

A. Incorrect. Plausible since a rod block is present for the given conditions. However, a half-scram does not occur for any combination of above the scram setpoints. See explanation B below.

B. Correct. Each ARPM channel calculates setpoints based on current core flow as follows:

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-36

• Rod Block setpoint = 0.62(Core Flow %) + 57.1% = 88.1% for the conditions given.

• Scram setpoint = 0.62(Core Flow %) + 60.9% = 91.9% for the conditions given.

Therefore, APRM-4 has exceeded both the rod block and scram setpoint. No other channel has exceeded a setpoint.

A Rod Block can be generated from any single APRM chassis. Rod Blocks DO NOT use 2 out-of-4 logic. The Voter need two similar APRM trips to actuate (e.g. two APRM Upscale Trips or two OPRM Trips). The logic is normally 2 out of 4 unless an APRM is bypassed or inoperative, then the logic will be 2 out of 3. The output of the Voter is arranged such that there are two relay outputs (X and Y) that de-energize to actuate the respective RPS logic. Voter 1 and 3 input into the RPS Channels A1 and A2. Voter 2 and 4 input into the RPS Channels B1 and B2. Since all Voters receive the input from all operable APRMs, when at least two APRM trip signal are generated, a full reactor scram will occur. A half-scram condition will only exist when a single Voter is deenergized or during testing.

C. Incorrect. Plausible if it is believed that the rod block follows two of four logic while the RPS will generate a scram since all APRM Voters receive input from all APRM channels. However, only a rod block will be present. See explanation B above.

D. Incorrect. Plausible since a half-scram is not generated. However, a rod block is generated. See explanation B above.

Technical Reference(s)

SD001819, PRNM System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13711 - Describe PRNM system automatic initiation signals, operation, controls, automatic functions, alarms, setpoints, and or interlocks.

Question Source: r- Bank #: LO03335 C Modified Bank #: Mod Bank #. (Note changes or attach parent)

C New Question History: Last NRC Exam: N/A N/A Question Cognitive Level: r Memory or Fundamental Knowledge r

Comprehension or Analysis Justification for Cognitive Level Requires examinee to synthesize information given in the stem with an understanding of the logic used to create rod blocks, and a knowledge of the logic for scram signals.

10 CFR Part 55 Content: 55.41 6 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-36 Comments /

Reference:

SD001819 Rev: Major: 4 Minor: 0 c) This signal is used to establish the Flow Biased Rod Block and Scram setpoints for the associated APRM channel. (Wd = Recirc Drive Flow)

( 1) 0 .62Wd + 60.9% (Scram) (Capped at 112.9%) (Two Loop Operation (2) 0.62W d + 56.1% (Scram) (Single Loop Operation)

(3) 0 .62Wa + 57 . %"(Rodl3lock') (GE 08%) TLO (4) 0.62Wd + 56% (Rod Block) (SLO)

COLUMBIA SYSTEMS May 2018 PRNM SD001819, r4 mrO a) Each Voter receives input from all 4 APRM Chassis, the Reactor Mode Switch and the APRM Bypass Switch.

b) The relays in the Voter de-energize to actuate the trip function.

c) The Voter detennines if the conditions exist from the APRMs to actuate Rod Blocks or Reactor Scrams.

d) A Rod Block can be generated from an single AP chassis. Rod Blocl

LPSP, RBM A is not bypassed. The primary reference for RBM B is APRM #2. Since APRM #2 is < LPSP, RBM B is bypassed.

D. Incorrect. Plausible since RBM A is NOT bypassed. However, RBM B is bypassed.

Technical Reference(s)

SD001819, PRNM System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5085 - Identify which APRM channels are utilized by the rod block monitor channels.

Question Source: #: Bank #

1. See Attached (Note changes or attach parent)

Question History: Last NRC Exam: 2017 Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information from the question stem with an understanding of the inputs provided to RBM channels from APRMs along with a knowledge of the Low Power Set Point (LPSP) and how the RBMs are affected when an input is above or below LPSP.

10 CFR Part 55 Content: 55.41 7 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-58 Comments /

Reference:

SD001819 Rev: Major: 4 Minor: 0 Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-58 Comments /

Reference:

Parent Question (NRC 2017, Q #59) Rev: Major: N/A Minor: N/A Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-58 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-59 Examination Outline Cross-reference: 59 Revision: 0 Date: 7/29/20 Tier: 2 Group: 2 K/A Number: 233000.A1.03 Level of Difficulty: 2 RO Importance Rating: 3.1 K/A

Description:

Ability to predict and/or monitor changes in parameters associated with operating the FUEL POOL COOLING AND CLEAN-UP controls including: Pool temperature CGS is in Mode 1.

The crew has placed Fuel Pool Heat Exchanger, FPC-HX-1A, out of service for corrective maintenance.

What is the maximum, procedurally allowed, fuel pool temperature under these conditions?

A. 125°F B. 138°F C. 150°F D. 175°F Answer: D K/A Match:

Requires knowledge of the maximum fuel pool temperature allowed when operating controls to place a heat exchanger out of service.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-59 Explanation:

A. Incorrect. Plausible since this is the maximum allowed temperature under normal conditions with both supply water pumps and heat exchangers available. However, fuel pool temperature is allowed to rise to a maximum of 175°F when one supply pump or heat exchanger is unavailable.

B. Incorrect. Plausible since this is the temperature above which SW-V-75AA and SW-V-75BB (SW to FPC Manual Isolation Valves) are required to be open. However, fuel pool temperature is allowed to rise to a maximum of 175°F when one supply pump or heat exchanger is unavailable.

C. Incorrect. Plausible since this is the maximum allowed temperature while draining the Fuel Pool Dryer Separator Pit. However, fuel pool temperature is allowed to rise to a maximum of 175°F when one supply pump or heat exchanger is unavailable.

D. Correct. In accordance with SOP-FPC-OPS, section 4.2, and the FSAR (page 9.1-26), fuel pool temperature is allowed to rise to a maximum of 175°F with one supply pump or heat exchanger unavailable.

Technical Reference(s)

SOP-FPC-OPS, Fuel Pool Cooling and Cleanup Operations Final Safety Analysis Report (FSAR) Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5364 - State the Fuel Pool Cooling and Cleanup System limitation concerning fuel pool temperature.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of fuel pool temperature limits.

10 CFR Part 55 Content: 55.41 10 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-59 Comments /

Reference:

SOP-FPC-OPS Rev: Major: 007 Minor: 008 Comments /

Reference:

FSAR Rev: Major: 007 Minor: 008 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-59 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Examination Outline Cross-reference: 60 Revision: 0 Date: 7/30/20 Tier: 2 Group: 2 K/A Number: 234000.K5.03 Level of Difficulty: 3 RO Importance Rating: 2.9 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING EQUIPMENT: Water as a shield against radiation CGS is in Mode 5.

Fuel Handling is in progress. The following lights are lit on the Refueling Bridge Controller consoles:

How is the operation of the Refueling Bridge Crane affected?

A. The Main Hoist cannot be raised or lowered.

B. The Main Hoist cannot be lowered, but may be raised.

C. The Main Hoist cannot be raised, but may be lowered.

D. All bridge, trolley and Hoist motion is prevented.

Answer: C K/A Match:

Requires knowledge of Refueling Bridge interlocks to ensure sufficient water over the fuel to prevent high radiation.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Explanation:

A. Incorrect. Plausible since the Main Hoist cannot be raised. However, the Main Hoist may be lowered.

B. Incorrect. Plausible since other interlocks prevent the Main Hoist from being lowered (Slack Cable, Grapple Full Down). However, the Grapple Normal Up interlock prevents the Main Hoist from being raised, but allows the hoist to be lowered.

C. Correct. The Normal Grapple Up interlock stops hoist upward travel to ensure that the top of irradiated fuel is not raised to within 76 of the waters surface.

D. Incorrect. Plausible since the Fault Lockout interlock prevents all bridge, trolley, and hoist motion.

However, the Normal Grapple Up interlock only prevents hoist upward motion.

Technical Reference(s)

SD000207, Fuel Handling System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5358 - Explain the following Refueling Bridge indication: a. GRAPPLE NORMAL UP.

Question Source: #: LO02895

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing information in the stem with an understanding of the indications that the NORMAL GRAPPLE UP interlock is active and a knowledge of the effects of this interlock on the operation of the Refueling Bridge Crane.

10 CFR Part 55 Content: 55.41 13 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Comments /

Reference:

SD000207 Rev: Major: 14 Minor: 0 Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-61 Examination Outline Cross-reference: 61 Revision: 0 Date: 7/30/20 Tier: 2 Group: 2 K/A Number: 239001.K3.15 Level of Difficulty: 2 RO Importance Rating: 3.5 K/A

Description:

Knowledge of the effect that a loss or malfunction of the MAIN AND REHEAT STEAM SYSTEM will have on following: Reactor water level control CGS is in Mode 1.

Reactor power is 100% and steady.

A fault causes the total steam flow input to the Feed Water Level Control (FWLC) system to slightly lower.

FWLC remains in 3 Element Control.

How is the plant affected by this fault?

Initially, Reactor Feed Pump turbine speed will (1) . When compared to conditions prior to the fault, RPV level will stabilize at (2) .

A. (1) remain the same (2) a lower level B. (1) remain the same (2) the same level C. (1) lower (2) a lower level D. (1) lower (2) the same level Answer: C K/A Match:

Requires knowledge of the effects of a Main Steam system fault on RPV level control.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-61 Explanation:

A. Incorrect. Plausible since RFP turbine will stabilize at the same speed as before the fault and the RPV will stabilize at a lower level. However, initially, the steam flow/feed flow mismatch will cause turbine speed to lower.

B. Incorrect. Plausible since RFP turbine will stabilize at the same speed as before the fault.

However, the RPV will stabilize at a lower level.

C. Correct. The FWLC system is level dominant. However, steam flow/feed flow mismatch provides faster response to a change in reactor power. The question stem states that the FWLC system remains in 3 element control. Therefore, steam flow/feed flow provides an input. When the indicated steam flow lowered, the RPV level input to FWLC is lowered. This causes RFP turbine speed to initially lower to reduce actual RPV level to the setpoint. Once RPV level input reaches the setpoint, RFP turbine speed is raised to maintain RPV level at the setpoint. Since actual steam flow did not change, feed flow must be maintained at the initial value prior to the fault, which means that RFP turbine speed will stabilize at the same value as before the fault, with actual RPV level lower than initial level to provide an error signal that is equal to the steam flow/feed flow error caused by the fault.

D. Incorrect. Plausible since RFP turbine speed will initially lower. However, RPV level must stabilize at a lower level to provide an error signal to RWLC that is equal to the steam flow/feed flow error caused by the fault.

Technical Reference(s)

SD000157, Feedwater Level Control System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5395 - Describe the response of the FWLC system during steady-state operation and during a change in reactor power in Single Element and Three Element.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-61 Justification for Cognitive Level Requires knowledge of Feed Turbine response to a steam flow fault along with FLWC response to maintain RPV level.

10 CFR Part 55 Content: 55.41 7 Comments /

Reference:

SD000157 Rev: Major: 19 Minor: 1 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-61 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-62 Examination Outline Cross-reference: 62 Revision: 2 Date: 1/20/21 Tier: 2 Group: 2 K/A Number: 245000.A3.10 Level of Difficulty: 2 RO Importance Rating: 2.5 K/A

Description:

Ability to monitor automatic operations of the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS including: Generator output voltage/reactive load CGS is in Mode 1.

A reactor plant startup is in progress.

• Reactor power is 15%.

The operations crew is testing the Main Generator Voltage Adjuster prior to synchronizing the Main Generator to the grid.

• The Voltage Regulator control switch is in ON.

• Main Generator output voltage is 25 kV.

• Main Generator Megavars is 0 MV.

What is the effect if the Main Generator Exciter Voltage Adjuster control switch is taken to RAISE?

Main Generator output voltage will (1) and Main Generator Megavars will (2) .

A. (1) remain the same (2) increase in the IN direction B. (1) remain the same (2) increase in the OUT direction C. (1) increase (2) increase in the OUT direction D. (1) increase (2) remain the same Answer: D K/A Match:

Requires knowledge of the automatic changes in output voltage and reactive load when the Voltage Adjuster is repositioned.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-62 Explanation:

A. Incorrect. Plausible since Main Generator output voltage does not appreciably rise when taken to RAISE with the Main Generator synchronized to the grid. Plausibility is enhanced if candidate believes that Megavars will increase in the opposite direction if voltage adjust is taken to RAISE with the Main Generator not synchronized to the grid. However, Main Generator voltage will increase and Megavars will remain at 0.

B. Incorrect. Plausible since these indications are correct if the Main Generator is synchronized to the grid. However, since the Main Generator is not on the grid, output voltage will increase while reactive load (Megavars) will remain at 0.

C. Incorrect. Plausible since output voltage will increase. However, since the Main Generator is not synchronized to the grid, reactive load will remain at 0.

D. Correct. With the Voltage Regulator ON and the Main Generator not synchronized to the grid, taking the Main Generator Exciter Voltage Adjuster to RAISE will cause Main Generator output voltage to increase. With no load connected to the Main Generator, Megavars (reactive load) will remain constant at 0.

Technical Reference(s)

SD000152 - Main Generator System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7647 - Describe the cause-and-effect relationship between Main Generator MVARs and:

b. Exciter Voltage Adjuster-Raise/Lower Question Source: #: LO03648

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-62 Requires examinee to synthesize information given in the stem with a knowledge of the operation of the Main Generator voltage regulator and an understanding of how voltage and reactive load will respond when the Main Generator is not synchronized to the grid.

10 CFR Part 55 Content: 55.41 5 Comments /

Reference:

SD000152 Rev: Major: 13 Minor: 2 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-62 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-63 Examination Outline Cross-reference: 63 Revision: 0 Date: 5/6/20 Tier: 2 Group: 2 K/A Number: 256000.K2.01 Level of Difficulty: 2 RO Importance Rating: 2.7 K/A

Description:

Knowledge of electrical power supplies to the following: System pumps CGS is in Mode 1.

A Main Generator trip occurs.

• E-SM-2 Startup Power breaker, E-CB-S/2, fails to close.

Which of the following Condensate System pump(s) have power available?

(1) Condensate Pump COND-P-1A (2) Condensate Pump COND-P-1B (3) Condensate Pump COND-P-1C (4) Condensate Booster Pump COND-P-2B A. (1) and (2) ONLY B. (1) and (3) ONLY C. (2) and (4) ONLY D. (3) and (4) ONLY Answer: B K/A Match:

This question requires knowledge of the power supplies to condensate system pumps.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-63 Explanation:

A. Incorrect. Plausible since COND-P-1A is powered from SM-1 and has power available. However, COND-P-1B is powered from SM-2, which is de-energized.

B. Correct. COND-P-1A is powered from SM-1 and COND-P-1C is powered from SM-3. Both of these buses have power.

C. Incorrect. Plausible if it is believed that SM-2 will transfer to the Backup Transformer on a loss of the Startup Transformer. However, for the conditions given in the question stem, SM-2 is de-energized. Therefore, COND-P-1B and COND-P-2B do not have power available.

D. Incorrect. Plausible since COND-P-1C will be available. However, COND-P-2B is powered from E-SM-2. On a loss of E-TR-N1, E-SM-1 through 3 will swap to the Startup Transformer. Since breaker E-CB-S/2 failed to close, E-SM-2 will lose power and COND-P-2B will not be available.

Technical Reference(s)

SD000182, AC Distribution SD000134, Condensate Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5058 - Identify the loads on the following buses: a. SM-1, SM-2, SM-3 Question Source: #: NRC 2019 Exam

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: 2019 Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of power supplies to condensate system pumps.

10 CFR Part 55 Content: 55.41 4 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-63 Comments /

Reference:

SD000182 Rev: Major: 21 Minor: Min Comments /

Reference:

SD000134 Rev: Major: 17 Minor: 3 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-64 Examination Outline Cross-reference: 64 Revision: 1 Date: 1/25/21 Tier: 2 Group: 2 K/A Number: 259001.A1.01 Level of Difficulty: 2 RO Importance Rating: 3.1 K/A

Description:

Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including: Feedwater flow/pressure CGS is in Mode 1.

A reactor plant startup is in progress.

RFW-P-1B is being placed in service as the second Reactor Feed Pump in accordance with SOP-RFT-START, Reactor Feedwater Turbine System Start.

Plant conditions:

• RFW-P-1A is in AUTO.

• RFW-P-1B is in MDEM with Pump Discharge Valve, RFW-W-102B, open.

• RFW-P-1B discharge pressure is at shutoff head.

• RPV level is being maintained with the RPV Master Level Controller, RFW-LIC-600, in AUTO.

• RFW-P-1A and RFW-P-1B discharge pressures are equal and both RFW pumps are feeding.

CRO1 depresses the UP arrow for RFW-P-1B on RFT-COMP-1.

How are both Reactor Feed Pumps affected?

RFW-P-1B speed will A. rise and RFW-P-1A speed will lower.

B. rise and RFW-P-1A speed will remain the same.

C. remain the same and RFW-P-1A speed will rise.

D. remain the same and RFW-P-1A speed will remain the same.

Answer: A K/A Match:

Requires knowledge of the effect of changing RFP settings on speed (i.e. flow) of both pumps.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-64 Explanation:

A. Correct. Pressing the UP arrow on RFT-COMP-1 will cause the speed/flow of RFW-B-1B to rise.

This action does not affect RFW-P-1A directly. The FWLC system will detect the rise in feed flow and signal feed pumps to lower flow by reducing speed. Since RFW-P-1A is in AUTO, its speed/flow will lower. However, since RFW-P-1B is in MDEM, it will not respond to this signal and will remain at a speed/flow higher than the initial value.

B. Incorrect. Plausible since RFW-P-1B speed/flow will increase. However, since RFW-P-1A is in AUTO, it will respond to the FWLC system which sends a signal to slow the feed pumps. See A above.

C. Incorrect. Plausible if it is believed that the UP arrow on RFW-COMP-1 only affects any feed turbine in AUTO, similar to the operation of Master Level Controller controls. However, pressing Pressing the UP arrow on RFT-COMP-1 will cause the speed/flow of RFW-B-1B to rise. See A above.

D. Incorrect. Plausible if it is believed that The UP arrow on RFW-COMP-1 only affects an individual feed turbine if it is in AUTO. However, pressing Pressing the UP arrow on RFT-COMP-1 will cause the speed/flow of RFW-B-1B to rise. See A above.

Technical Reference(s)

SD000157, Feedwater Level Control System Description Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5394 - Describe the function of each of the following controls and how they relate to each other:

a. Startup Level Controller

b. Master Controller

c. Turbine Speed Controllers

d. RFP Minimum Flow Controllers Question Source: #: LO03701

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: 2017 Question Cognitive Level:

Justification for Cognitive Level Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-64 Requires examinee to synthesize information given in the question stem with a knowledge of the operation of the Master Level Controller and an understanding of how adjusting one feed turbine will affect the operation of the other feed turbine in auto.

10 CFR Part 55 Content: 55.41 4 Comments /

Reference:

SD000157 Rev: Major: 19 Minor: 1 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-64 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-65 Examination Outline Cross-reference: 65 Revision: 0 Date: 5/6/20 Tier: 2 Group: 2 K/A Number: 268000.A4.01 Level of Difficulty: 3 RO Importance Rating: 3.4 K/A

Description:

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

CGS is in Mode 1.

A leak in the Reactor Building is filling Sump FDR-SUMP-R1.

Currently, the condition of the sump timer, FDR-TM-605A is as follows:

When will the timer cause an alarm?

The alarm will sound if the sump pump starts before the (1) needle reaches (2) minutes.

A. (1) black (2) 0 B. (1) red Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-65 (2) 0 C. (1) black (2) 20 D. (1) red (2) 30 Answer: B K/A Match:

Requires knowledge of the operation of Reactor Building sump timers.

Explanation:

A. Incorrect. Plausible since the timer moves towards 0 minutes. However, the black needle represents the setpoint and the red needle moves from the setpoint to 0 minutes.

B. Correct. When the timer is reset, the red and black needles are on the same number. The red needle times out by moving towards 0 minutes. The graphic in the question shows that the timer was set at 20 minutes. The timer has been running approximately 10 minutes without the sump pump starting. If the sump pump starts before the red needle times down to 0 minutes, an alarm will sound to signify that the sump s filling faster than expected.

C. Incorrect. Plausible if it is believed that the black needle travels towards 0 minutes when the timer is running and that the red needle represents the alarm setpoint. However, the black needle represents the timer setpoint that is manually inserted. The red needle moves from the setpoint (black needle position) to 0 minutes.

D. Incorrect. Plausible since the red needle moves. however, the red needle moves from the setpoint (black needle position) to 0 minutes. If the sump pump starts prior to the red needle reaching 0 minutes, an alarm will sound.

Technical Reference(s)

SD000167, Leak Detection Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 5466 - Sate the purpose of the leak detection system/s sump pump-out and fill rate timers.

Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-65 Question Source: #: 2019 NRC Exam

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: 2019 Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the operation of the sump timers.

10 CFR Part 55 Content: 55.41 13 Comments /

Reference:

SD000167 Rev: Major: 13 Minor: 2 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-66 Examination Outline Cross-reference: 66 Revision: 0 Date: 6/29/20 Tier: 3 Group: N/A K/A Number: 2.1.2 Level of Difficulty: 2 RO Importance Rating: 4.1 K/A

Description:

Knowledge of operator responsibilities during all modes of plant operation.

CGS is in Mode 1.

The operations crew needs to lower recirculation flow as part of a surveillance.

With CRS/SM permission and under direct supervision of a licensed RO, which of the following individual may operate the RRC pump controls to lower flow?

(1) An unlicensed individual currently in training for a SRO license.

(2) A previously licensed RO that does not currently hold a license.

(3) An operator with an INACTIVE RO license that is standing an under-instruction watch for license activation.

A. (1) ONLY B. (2) ONLY C. (1) and (3) ONLY D. (2) and (3) ONLY Answer: C K/A Match:

Requires knowledge of operators that are allowed to perform reactivity manipulations.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-66 Explanation:

A. Incorrect. Plausible since an individual in license training is authorized to manipulate control room controls under the direct supervision of a licensed operator. However, (3) is also correct.

B. Incorrect. Plausible since a previously licensed individual has performed reactivity manipulations in the past and should possess the requisite knowledge to complete the task correctly. However, non-licensed individuals may not operate control room controls unless in an emergency when directed or if the individual is in a training program leading to a license.

C. Correct. In accordance with PPM 1.3.1, Operating Policies, Programs and Practices, section 4.6.3, individuals in a training program for an operator license may operate control room controls to perform a reactivity manipulation while under the direct supervision of a licensed operator. In addition, inactive licensed SROs and ROs may perform reactivity manipulations while directly supervised by a licensed operator.

D. Incorrect. Plausible (3) is correct. However, an individual not in a training program for a license and without a license (active or inactive) may not perform reactivity manipulations.

Technical Reference(s)

PPM 1.3.1, Operating Policies, Programs and Practices Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 6076 - Identify who can manipulate Control Room controls.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of individuals authorized to perform reactivity manipulations in the control room.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-66 Comments /

Reference:

PPM 1.3.1 Rev: Major: 1281 Minor: N/A Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-67 Examination Outline Cross-reference: 67 Revision: 1 Date: 1/5/21 Tier: 3 Group: N/A K/A Number: 2.1.20 Level of Difficulty: 2 RO Importance Rating: 4.6 K/A

Description:

Ability to interpret and execute procedure steps.

An operator has been directed to lineup a system for startup.

The procedure to be used is designated as Reference Use.

How should the operator placekeep the procedure?

The operator...

A. is not required to placekeep this procedure.

B. must mark each step as complete prior to proceeding to the next step.

C. should placekeep as often as practical (the end of a section, prior to a break, etc.).

D. should print their name and place their initials on the procedure cover sheet to signify that all steps were properly completed.

Answer: C K/A Match:

Requires knowledge of methods to complete Reference Use procedures.

Explanation:

A. Incorrect. Plausible since placekeeping is not required for Information Use procedures. However, placekeeping is required on Reference Use procedures B. Incorrect. Plausible since this method is required when performing most continuous use procedures. However, it is not required to be performed on Reference Use procedures after each step.

C. Correct. In accordance with SWP-PRO-01, Procedure and Work Instruction Use and Adherence, section 4.5.6: IF using a Reference Use procedure or work instructions, THEN PLACEKEEP steps as often as practical (e.g., after each section is complete, prior to a break, etc.).

D. Incorrect. Plausible since some Continuous Use procedures require a printed name and signature on the cover sheet. However, this method is not used on Reference Use procedures.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-67 Technical Reference(s)

SWP-PRO-01, Procedure and Work Instruction Use and Adherence Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 16047 - Upon Completion of this class, the students will be able to demonstrate the ability to perform Licensed Operator tasks while in compliance with procedures to meet management expectations.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information in the question with a knowledge of placekeeping methods along with requirements for Reference Use procedures.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-67 Comments /

Reference:

SWP-PRO-01 Rev: Major: 033 Minor: N/A Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Examination Outline Cross-reference: 68 Revision: 2 Date: 2/18/21 Tier: 3 Group: N/A K/A Number: 2.1.42 Level of Difficulty: 2 RO Importance Rating: 2.5 K/A

Description:

Knowledge of new and spent fuel movement procedures.

CGS is in Mode 5.

Core refueling is in progress in accordance with PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading.

The crew is preparing to move a fuel cell into the core.

Which of the following conditions must be met prior to loading fuel into the core?

(1) No personnel allowed in the entire drywell.

(2) Less than the required number of SRM detectors.

(3) Continuous communications established between the refueling floor and control room.

A. (1) ONLY B. (3) ONLY C. (1) and (2) ONLY D. (2) and (3) ONLY Answer: B K/A Match:

Requires knowledge of requirements to move fuel into the core.

Page 1 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Explanation:

A. Incorrect. Plausible since Drywell entry may be restricted in the upper drywell during certain fuel loading evolutions. However, personnel are always allowed to be in the lower drywell while loading fuel into the core.

B. Correct. In accordance with PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading, Attachment 8.2, Criteria For Stopping Fuel Loading, and the FSAR, continuous communications between the control room and refueling personnel is required during core alterations.

C. Incorrect. See explanation of plausibility for (1) in explanation A above. (2) is plausible since the number of required SRMs is reduced to 1 in some instances. However, at least one SRM is still required for fuel loading.

D. Incorrect. Plausible since (3) is a requirement to load fuel into the core. However, (2) is not a requirement..

Technical Reference(s)

PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading Final Safety Analysis Report (FSAR) Attached w/ Revision #

See Comments / Reference TS LCO 3.3.1.2 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8830 - Discuss what actions are required if communication is lost between the bridge and the control room during fuel shuffling.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires Knowledge of requirements for loading fuel.

10 CFR Part 55 Content: 55.41 13 Page 2 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Comments /

Reference:

PPM 6.3.2 Rev: Major: 25 Minor: N/A Page 3 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Page 4 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Comments /

Reference:

FSAR Ammendment: 60 Page 5 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Comments /

Reference:

LCO 3.3.1.2 Ammendment: 22 5

Page 6 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-69 Examination Outline Cross-reference: 69 Revision: 0 Date: 6/23/20 Tier: 3 Group: N/A K/A Number: 2.2.22 Level of Difficulty: 2 RO Importance Rating: 4.0 K/A

Description:

Knowledge of limiting conditions for operations and safety limits.

Which of the following is the lowest reactor steam dome pressure that results in exceeding Technical Specification Safety Limits?

A. 1310 psig B. 1320 psig C. 1330 psig D. 1340 psig Answer: C K/A Match:

Requires knowledge of Safety Limits.

Explanation:

A. Incorrect. Plausible since this pressure is above the RPS High RPV Pressure Scram setpoint and within 15 psig of the Safety Limit. However, the Safety Limit is 1325 psig, which is above this pressure.

B. Incorrect. Plausible since this pressure is within 5 psig of the Safety Limit. However, the Safety Limit is 1325 psig, which is above this pressure.

C. Correct. 1330 psig is the lowest pressure listed that exceeds the Safety Limit of 1325 psig.

D. Incorrect. Plausible since this pressure exceeds the Safety Limit. However, it is not the lowest pressure listed that exceeds the safety limit.

Technical Reference(s)

Technical Specifications Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-69 Proposed references to be provided during examination: N/A Learning Objective: 10304 - Describe each of the Safety Limits and state what actions are required if a Safety Limit is violated.

Question Source: #: LO02588

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the Safety Limit value for steam dome pressure.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-69 Comments /

Reference:

Technical Specifications Amendment 254 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Examination Outline Cross-reference: 70 Revision: 0 Date: 8/4/20 Tier: 3 Group: N/A K/A Number: 2.2.39 Level of Difficulty: 3 RO Importance Rating: 3.9 K/A

Description:

Knowledge of less than or equal to one hour Technical Specification action statements for systems CGS is in Mode 1.

A seismic event causes the following:

• DG-1 is inoperable.

• The Ashe 230 kV line is lost.

The CRS enters LCO 3.8.1, AC Sources - Operating.

What is the earliest action required by LCO 3.8.1?

A. Restore DG-1 to operable.

B. Restore the Ashe 230 kV line to operable.

C. Enter LCO 3.0.3 and prepare for a reactor shutdown.

D. Complete OSP-ELEC-W101, Offsite Station Power Alignment Check.

Answer: D K/A Match:

Requires knowledge of the one hour LCO action for a loss of an offsite circuit.

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Explanation:

A. Incorrect. Plausible since the conditions in the question requires entry into LCO 3.8.1, Condition D, which requires restoring the DG OR the Offsite Source within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. However, Conditions A and B are also entered. Both of these conditions require that SR 3.8.1.1 (OSP-ELEC-W101) be completed within one hour.

B. Incorrect. Plausible since the conditions in the question requires entry into LCO 3.8.1, Condition D, which requires restoring the DG OR the Offsite Source within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. However, Conditions A and B are also entered. Both of these conditions require that SR 3.8.1.1 (OSP-ELEC-W101) be completed within one hour.

C. Incorrect. Plausible since LCO 3.8.1, Condition G (Three or more required AC sources inoperable) requires entry into LCO 3.0.3 immediately. Plausibility is further enhanced since LCO 3.8.1, Condition D (which should be entered) requires entry into LCO 3.8.7 if any division has no AC sources. LCO 3.8.7, Condition F requires immediate entry into LCO 3.0.3. However, LCO 3.8.7 is not entered and there are only two AC sources lost. For the conditions given, LCO 3.0.3 would not be entered.

D. Correct. For the conditions given, LCO 3.8.1, Conditions A, B, and D are entered. Both Conditions A and B require completing SR 3.8.1.1 (OSP-ELEC-W101) within one hour.

Technical Reference(s)

Technical Specification LCO 3.8.1, AC Sources - Operating OSP-ELEC-W101, Offsite Station Power Alignment Check Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5059 - Referencing Technical Specifications associated with the AC Distribution System and a set of plant conditions, determine as applicable the LCO, the action statement, and the appropriate bases.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Justification for Cognitive Level Requires synthesizing information in the question stem with a knowledge of specific conditions of Technical Specification Action Statements couple with a knowledge of less than or equal to one hour actions.

10 CFR Part 55 Content: 55.41 10 Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Comments /

Reference:

TS LCO 3.8.1 Rev: Major: 225 Minor: N/A Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Comments /

Reference:

OSP-ELEC-W101 Rev: Major: 032 Minor: 004 Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-71 Examination Outline Cross-reference: 71 Revision: 1 Date: 1/5/21 Tier: 3 Group: N/A K/A Number: 2.3.13 Level of Difficulty: 3 RO Importance Rating: 3.4 K/A

Description:

Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc CGS is in Mode 1.

An operator needs to enter an area in the Reactor Building to reposition a valve.

• The highest dose rate in the area is 525 Rad/Hr at 1 meter.

• Emergency conditions DO NOT exist.

Who must authorize this entry?

The entry must be authorized by the A. Shift Manager (SM) ONLY.

B. Radiation Protection Manager (RPM) ONLY.

C. Plant General Manager (PGM) ONLY.

D. Radiation Protection Manager (RPM) and Plant General Manager (PGM).

Answer: D K/A Match:

Requires knowledge of radiological safety procedures for entering Very High Radiation Areas.

Explanation:

A. Incorrect. Plausible since the shift manager may authorize entry into a VHRA during emergency conditions. However, the stem states that emergency conditions do not exist.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-71 B. Incorrect. Plausible since the RPM must authorize the entry. However, this is not the only authorization that is required to enter a VHRA for the conditions given.

C. Incorrect. Plausible since the RPM must authorize the entry. However, this is not the only authorization that is required to enter a VHRA for the conditions given.

D. Correct. In accordance with PPM 11.2.7.3, High Radiation Area, Locked High Radiation Area, and Very High Radiation Area Controls, the RPM and PGM must authorize entry into a VHRA for operational reasons.

Technical Reference(s)

PPM 11.2.7.3, High Radiation Area, Locked High Radiation Area, and Very High Radiation Area Controls Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11257 - Knowledge of 10CFR: 20 and related facility radiation control requirements.

Question Source: #: Bank #

1. RO00072 (Note changes or attach parent)

Question History: Last NRC Exam: 2017 Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information in the question with a knowledge of the definition of a VHRA along with an understanding of the requirements to enter a VHRA.

10 CFR Part 55 Content: 55.41 12 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-71 Comments /

Reference:

PPM 11.2.7.3 Rev: Major: 042 Minor: 001 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-72 Examination Outline Cross-reference: 72 Revision: 0 Date: 6/23/20 Tier: 3 Group: N/A K/A Number: 2.3.11 Level of Difficulty: 2 RO Importance Rating: 3.8 K/A

Description:

Ability to control radiation releases.

CGS is in Mode 1.

A leak occurs from the primary into secondary containment.

The CRS enters PPM 5.3.1, Secondary Containment Control.

• The leak is NOT isolable.

Which of the following combinations of parameters, exceeding Maximum Safe Operating Value (MSOV), will require the crew to perform an emergency depressurization (ED)?

(1) RHR Pump A Room Temperature (2) RHR Pump B Room Level (3) RHR Pump C Room Temperature (4) RHR Pump A Room Radiation Level A. (1) and (2)

B. (1) and (3)

C. (2) and (4)

D. (3) and (4)

Answer: B K/A Match:

Requires knowledge of actions taken to control radiation release.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-72 Explanation:

A. Incorrect. Plausible since 2 areas need to reach MSOV to require an ED. However, to ED there needs to be two areas of the same parameter (temperature, level, radiation level) that reach MSOV.

B. Correct. In accordance with step SC-15 of PPM 5.3.1, when any one parameter exceeds its MSOV in two or more areas, an ED is required.

C. Incorrect. Plausible since two parameters have reached MSOV in two separate areas. However, to ED there needs to be two areas of the same parameter (temperature, level, radiation level) that reach MSOV.

D. Incorrect. Plausible since two parameters have exceeded MSOV in the same area. However, to ED there needs to be two areas of the same parameter (temperature, level, radiation level) that reach MSOV.

Technical Reference(s)

PPM 5.3.1, Secondary Containment Control Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8459 - Given a list, identify the statement that describes the two reasons for emergency depressurizing the RPV if one secondary containment parameter is above Maximum Safe Operating Levels in more than one area and a primary system is discharging reactor coolant into secondary containment.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing information given in the stem with a knowledge of the different paths of PPM 5.3.1 along with an understanding of the requirements to perform an emergency depressurization.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-72 Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: N/A Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Examination Outline Cross-reference: 73 Revision: 0 Date: 8/5/20 Tier: 3 Group: N/A K/A Number: 2.4.4 Level of Difficulty: 2 RO Importance Rating: 4.5 K/A

Description:

Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

CGS is in Mode 1.

A slow increase in drywell unidentified leakage is raising drywell pressure.

• Drywell pressure is 0.7 psig, up slow.

A loss of IN-1 occurs, resulting in a loss of control rod position indication.

10 seconds later, Drywell pressure reaches 1.0 psig and the CRS directs a manual reactor scram.

Current plant conditions:

• Reactor power: 2%, down slow.

• RPV level: -15 inches, up slow.

• Drywell pressure: 1.1 psig, up slow.

Which EOP(s) is/are required to be entered?

Enter PPM 5.1.1, RPV Control A. ONLY.

B. and PPM 5.2.1, Primary Containment Control.

C. and PPM 5.3.1, Secondary Containment Control.

D. and transition to PPM 5.1.2, RPV Control - ATWS.

Answer: D K/A Match:

Requires knowledge of EOP entry requirements.

Page 1 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Explanation:

A. Incorrect. Plausible since PPM 5.1.1. should be entered due to low RPV level. However, a loss of the full core display occurs due to the loss of E-US-PP. Without the display, the operators cannot verify that existing control rod pattern is sufficient to ensure the reactor remains shutdown (step RC-2 of PPM 5.1.1). Therefore PPM 5.1.2 should be entered.

B. Incorrect. Plausible since PPM 5.1.1. should be entered due to low RPV level. Additionally, entry into PPM 5.2.1 is plausible based on rising drywell pressure. However, PPM 5.2.1 is not required to be entered until drywell pressure reaches 1.68 psig.

C. Incorrect. Plausible since PPM 5.1.1. should be entered due to low RPV level. Additionally, entry into PPM 5.3.1 is plausible based on the loss of several radiation monitors with the loss of E-US-PP. However, PPM 5.3.1 is entered on radiation monitors in alarm, not with a loss of power.

D. Correct. PPM 5.1.1 is entered due to low RPV water level. Step RC-2 of PPM 5.1.1 states that if it is determined that existing control rod pattern alone does not always assure reactor shutdown, then a transition to PPM 5.1.2 is required. Once in PPM 5.1.2, operators verify that control rods are sufficient to maintain the reactor shutdown by performing SOP-RXSD-DETERMINATION-QC.

Operators may then exit PPM 5.1.2 (see step RC-2) and re-enter PPM 5.1.1.

Technical Reference(s)

ABN-ELEC-INV, 120 VAC Critical Distribution System Failures PPM 5.1.1, RPV Control Attached w/ Revision #

PPM 5.1.2, RPV Control - ATWS See Comments / Reference SOP-RXSD-DETERMINATION-QC, Reactor Shutdown Determination Quick Card Proposed references to be provided during examination: N/A Learning Objective: 6827 - Given a Loss of IN-1, IN-2A/B, IN-3A/B or IN-5, identify those automatic actions that may have occurred.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires understanding of the consequences of a loss of E-US-PP along with knowledge of conditions requiring entry into EOPs.

Page 2 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

ABN-ELEC-INV Rev: Major: 017 Minor: 001 Page 3 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Comments /

Reference:

PPM 5.1.1 Rev: Major: 22 Minor: N/A Page 4 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A Page 5 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Comments /

Reference:

SOP-RXSD-DETERMINATION-QC Rev: Major: 001 Minor: N/A Page 6 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Examination Outline Cross-reference: 74 Revision: 1 Date: 1/5/21 Tier: 3 Group: N/A K/A Number: 2.4.6 Level of Difficulty: 2 RO Importance Rating: 3.7 K/A

Description:

Knowledge of EOP mitigation strategies.

CGS is in Mode 1.

An automatic scram signal is received. Control Rods did not insert.

The CRS enters PPM 5.1.1, RPV Control, and transitions to PPM 5.1.2, RPV Control - ATWS.

Current plant conditions:

• Reactor power: 16% down slow.

• RPV level: -7 inches, up slow.

• MSIVs are closed.

How should the crew proceed in PPM 5.1.2?

The crew should perform the (1) leg of PPM 5.1.2 first to (2) .

A. (1) level (2) reduce the probability of fuel damaging power oscillations B. (1) level (2) ensure that level does not go below Top of Active Fuel (TAF)

C. (1) pressure (2) reduce the probability of equipment damage due to SRV cycling D. (1) pressure (2) ensure that pressure remains less than the high pressure scram setpoint Answer: A K/A Match:

Requires knowledge of EOP mitigation strategy during an ATWS.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Explanation:

A. Correct. In accordance with OI-15, EOP and EAL Clarifications, Due to the higher rodline associated with MELLA+, a level leg first strategy during ATWS conditions will reduce the probability of fuel damaging power oscillations. Level should be lowered promptly and communications minimized until the order to stop and prevent is issued.

B. Incorrect. Plausible since the level leg of PPM 5.1.2 is the first leg of the EOP to be performed.

However, the level leg does not necessarily maintain RPV level above TAF (-161 inches).

C. Incorrect. Plausible since the question stem states that MSIVs are closed. Therefore, it is important to reduce pressure to prevent equipment damage due to SRVs cycling. However, this is not the first procedure leg to be performed. See A above.

D. Incorrect. Plausible since one of the purposes of the pressure leg of PPM 5.1.2 is to reduce RPV pressure below the high pressure scram setpoint to avoid SRVs lifting. However, this is not the first procedure leg to be performed. See A above.

Technical Reference(s)

OI-15, EOP and EAL Clarifications PPM 5.1.2, RPV Control - ATWS Attached w/ Revision #

See Comments / Reference Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13568 - Given a copy of PPM 5.1.2 RPV Control - ATWS and an event requiring entry into the EOPs, execute the strategies of this PPM in accordance with procedure use standards and expectations without error.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to have knowledge of the first procedure leg in PPM 5.1.2 to be performed, and the reason for performing this leg first.

10 CFR Part 55 Content: 55.41 10 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Comments /

Reference:

OI-15 Rev: Major: 032 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-75 Examination Outline Cross-reference: 75 Revision: 1 Date: 1/25/21 Tier: 3 Group: N/A K/A Number: 2.4.16 Level of Difficulty: 2 RO Importance Rating: 3.5 K/A

Description:

Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

CGS is in Mode 1.

An event occurs that requires control room evacuation.

The CRS enters ABN-CR-EVAC, Control Room Evacuation and Remote Cooldown.

• All immediate actions are completed prior to leaving the control room.

Current plant conditions:

• One control rod is at position 06. All other control rods are fully inserted.

• RPV level is -10 inches, up slow.

• Drywell pressure is 1.6 psig, up slow.

• Reactor Building differential pressure is 0 inches H2O, and stable.

Which procedure should be used for overall control of the plant?

A. PPM 5.1.1, RPV Control.

B. PPM 3.3.1, Reactor Scram.

C. PPM 5.3.1, Secondary Containment Control.

D. ABN-CR-EVAC, Control Room Evacuation and Remote Cooldown.

Answer: D K/A Match:

Requires knowledge of coordination between EOPs and ABN-CR-EVAC Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-75 Explanation:

A. Incorrect. Plausible since entry conditions of PPM 5.1.1 are met (RPV level LE +13 inches).

However, a note in ABN-CR-EVAC states that this procedure supercedes EOP procedures.

B. Incorrect. Plausible since this procedure should be entered for a reactor scram and scramming the reactor is an immediate action of ABN-CR-EVAC. However, a note in ABN-CR-EVAC states that this procedure supercedes PPM 3.3.1.

C. Incorrect. Plausible since the entry condition of PPM 5.3.1 have been met (RB d/p LE 0 inches).

However, a note in ABN-CR-EVAC states that this procedure supercedes PPM 3.3.1.

D. Correct. Normally, EOPs have precedence over ABNs. However, a note in ABN-CR-EVAC states that this procedure supercedes EOPs and PPM 3.3.1 Technical Reference(s)

ABN-CR-EVAC Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 6105 - State which procedures have priority/precedence over all other operating procedures when an emergency exists.

Question Source: #: LO01730

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: 2013 Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information in the question with a knowledge of EOP entry conditions and a knowledge of notes in ABN-CR-EVAC.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-75 Comments /

Reference:

ABN-CR-EVAC Rev: Major: 042 Minor: N/A Page 3 of 3

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Examination Outline Cross-reference: 76 Revision: 0 Date: 11/30/20 Tier: 1 Group: 1 K/A Number: 295003.2.4.20 Level of Difficulty: 3 SRO Importance Rating: 4.3 K/A

Description:

Partial or complete loss of AC power: Knowledge of the operational implications of EOP warnings, cautions, and notes.

CGS is in Mode 1.

A Station Blackout (SBO) occurs.

• DG-1 and DG-2 will not be available for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The CRS enters PPM 5.6.1, SBO/ELAP.

• Station battery loads have been reduced.

• Offsite power will not be available for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

• DG-3 and DG-4 are available.

What is the preferred method for restoring AC buses?

The CRS should direct aligning DG-3 to power (1) and aligning DG-4 to power (2) .

A. (1) Division 1 critical switchgear via SM-7 (2) Division 2 critical switchgear via SM-8 B. (1) Division 1 critical switchgear via SM-7 (2) Division 2 critical switchgear via MC-8A C. (1) Division 2 critical switchgear via SM-8 (2) Division 1 critical switchgear via SM-7 D. (1) Division 2 critical switchgear via SM-8 (2) Division 1 critical switchgear via MC-7A Answer: D K/A Match:

Requires knowledge of notes in EOP supplemental procedure PPM 5.6.2.

Page 1 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 SRO Only:

Page 2 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Explanation:

A. Incorrect. Plausible since PPM 5.6.2, SBO and ELAP Attachments, attachments 8.7 and 8.8 allow DG-3 to power SM-7 and DG-4 to power SM-8 if necessary. However, the note prior to step 8.7.3 directs that it is preferred to align DG-3 to power Div. 2 switchgear ( via SM-8) and the note prior to step 8.8.1 directs that it is preferred to align DG-4 to Div. 1 switchgear (via MC-7A).

B. Incorrect. Plausible since PPM 5.6.2, SBO and ELAP Attachments, attachments 8.7 and 8.8 allow DG-3 to power SM-7 and DG-4 to power SM-8 if necessary. However, the note prior to step 8.7.3 directs that it is preferred to align DG-3 to power Div. 2 switchgear ( via SM-8) and the note prior to step 8.8.1 directs that it is preferred to align DG-4 to Div. 1 switchgear (via MC-7A).

C. Incorrect. Plausible since the note prior to step 8.7.3 directs that it is preferred to align DG-3 to power Div. 2 switchgear (via SM-8) and the note prior to step 8.8.1 directs that it is preferred to align DG-4 to Div. 1 switchgear. However, DG-4 is not connected to SM-7 directly, it must be connected via MC-7A.

D. Correct. Although both DGs are capable of powering either Div. 1 or Div. 2 critical switchgear, the note prior to step 8.7.3 directs that it is preferred to align DG-3 to power Div. 2 switchgear ( via SM-8) and the note prior to step 8.8.1 directs that it is preferred to align DG-4 to Div. 1 switchgear (via MC-7A).

Technical Reference(s)

PPM 5.6.1, SBO/ELAP PPM 5.6.2, SBO and ELAP Attachments Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13570 - Given a copy of PPM 5.6.1, Station Blackout and an event requiring entry into the EOPs, execute the strategies of thisPPM in accordance with procedure use standards and expectations without error.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 3 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Justification for Cognitive Level Requires examinee to synthesize information given in the stem with an understanding of the SBO/ELAP attachments for repowering critical buses and a knowledge of the preferred lineup for repowering buses.

10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

PPM 5.6.1 Rev: Major: 0 Minor: N/A Page 4 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Comments /

Reference:

PPM 5.6.2 Rev: Major: 15 Minor: N/A Page 5 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-77 Examination Outline Cross-reference: 77 Revision: 2 Date: 11/17/20 Tier: 1 Group: 1 K/A Number: 295004.AA2.03 Level of Difficulty: 3 SRO Importance Rating: 2.9 K/A

Description:

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Battery voltage CGS is in Mode 1.

Battery Charger E-C1-7 has failed.

ABN-ELEC-125VDC has been entered.

What is the maximum voltage where ABN-ELEC-125VDC requires that Battery E-B1-7 be removed from service?

A. 115 VDC B. 110 VDC C. 105 VDC D. 100 VDC Answer: C K/A Match:

Requires knowledge of battery voltage require to remove battery from service per subsequent actions of ABN-ELEC-125VDC Page 1 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-77 SRO Only:

Page 2 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-77 Explanation:

A. Incorrect. Plausible since it is less than normal DC battery voltage and requires specific knowledge of the ABN-ELEC-125VDC requirement.

B. Incorrect. Plausible since it is less than normal DC battery voltage and requires specific knowledge of the ABN-ELEC-125VDC requirement.

C. Correct. ABN-ELEC-125VDC requires removing the battery from service if BOP125 VDC battery voltage decreases to 105 VDC.

D. Incorrect. Plausible since it is less than normal DC battery voltage and requires specific knowledge of the ABN-ELEC-125VDC requirement. However in this case the voltage is less than the requirement to remove the battery from service.

Technical Reference(s)

ABN-ELEC-125VDC Attached w/ Revision #

See Comments / Reference Proposed references to be provided during examination: None Learning Objective: 15760 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-ELEC-125VDC.

Question Source: #:

1. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires recalling significant parameter requirements for actions taken in Abnormal procedures.

10 CFR Part 55 Content: 55.43 5 Page 3 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-77 Page 4 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78 Examination Outline Cross-reference: 78 Revision: 1 Date: 1/5/21 Tier: 1 Group: 1 K/A Number: 295006.AA2.06 Level of Difficulty: 2 SRO Importance Rating: 3.8 K/A

Description:

Ability to determine and/or interpret the following as they apply to SCRAM: Cause of reactor SCRAM CGS is in Mode 1.

Reactor power is 100%.

Reactor power lowers to 96%

• Parameters indicate that a Jet Pump failure has occurred.

A Jet Pump Hold Down Beam failure is confirmed.

What action(s) should the CRS direct?

A. Scram the reactor and stop the affected RRC pump.

B. Reduce RRC flow to 74 Mlbm/hour at 5% per minute.

C. Insert control rods per the Fast Shutdown Sequence.

D. Adjust the unaffected RRC pump speed as necessary to balance loop flows.

Answer: A K/A Match:

Requires knowledge of the requirements to scram the reactor.

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78 SRO Only:

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78 Explanation:

A. Correct. In accordance with ABN-POWER, Unplanned Reactor Power Change, section 4.2.4, if a Jet Pump Hold Down Beam failure is confirmed, then SCRAM the reactor and stop the affected RRC pump.

B. Incorrect. Plausible since this action is required by ABN-POWER, section 4.3, for an unplanned feedwater temperature reduction. However, for the conditions given, reducing core flow may result in entering an unstable area of the Power vs. Flow map.

C. Incorrect. Plausible since this action is required in accordance with TS LCO 3.4.2 and section 4.2.5 of ABN-POWER, for a Jet Sensing Line failure. However, a reactor scram is required for the conditions given.

D. Incorrect. Plausible since a failure of a Jet Pump Hold Down Beam will cause changes in flow in the affected loop. However, RRC pump speed should not change.

Technical Reference(s)

ABN-POWER, Unplanned Reactor Change Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 10312 - Evaluate plant conditions associated with a RRC Flow Control System Failure and determine if a reactor scram is required.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of required actions for a Jet Pump Hold Down Beam Failure 10 CFR Part 55 Content: 55.43 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78 Comments /

Reference:

ABN-POWER Rev: Major: 016 Minor: 003

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Examination Outline Cross-reference: 79 Revision: 1 Date: 1/5/21 Tier: 1 Group: 1 K/A Number: 295023.AA2.05 Level of Difficulty: 3 SRO Importance Rating: 4.6 K/A

Description:

Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS: Entry conditions of emergency plan.

CGS is in Mode 5.

Refueling activities are in progress.

• An irradiated fuel bundle is in transit from the Spent Fuel Pool (SFP) to the Reactor Cavity.

SFP level is observed to be lowering.

Just when the FUEL POOL LEVEL HIGH/LOW annunciator alarms (H13.P627.FPC2.2-2),

SFP level stops lowering.

Area Radiation Monitor (ARM) indications have risen and are currently stable as follows:

Using the reference provided, which of the following is correct for this situation?

The CRS should enter (1) , and declare an (2) .

A. (1) ABN-FPC-LOSS, Loss of Fuel Pool Cooling (2) Unusual Event Page 1 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 B. (1) ABN-FPC-LOSS, Loss of Fuel Pool Cooling (2) Alert C. (1) ABN-FUEL-HAND, Damage While Handling Fuel (2) Unusual Event D. (1) ABN-FUEL-HAND, Damage While Handling Fuel (2) Alert Answer: A K/A Match:

Requires interpreting emergency plan entry conditions for a fuel (spent fuel pool) accident and determining emergency classification.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Explanation:

A. Correct. An entry condition for ABN-FPC-LOSS is An unplanned reduction in Fuel Pool level.

Additonally, PPM 13.1.1, RU2.1, requires declaring an Unusual Event for an unplanned loss of SFP level accompanied by an unplanned rise in radiation levels on ARM-RIA-1, ARM-RIS-2, OR ARM-RIS-34.

B. Incorrect. Plausible since ABN-FPC-LOSS should be entered. However, an Alert should not be declared since none of the ARMs are alarming.

C. Incorrect. Plausible since an Unusual Event should be declared for the conditions given in the question stem. Plausibility is enhanced since rising ARM indications could indicate damage to irradiated fuel and require entry into ABN-FUEL-HAND. However, ABN-FUEL-HAND is only entered if ARMs are alarming and no ARMs are in alarm in accordance with the indications given.

D. Incorrect. Plausible since rising ARM indications could indicate damage to irradiated fuel and require entry into ABN-FUEL-HAND. Plausibility is enhanced since rising ARM levels could lead to an Alert declaration. However, indications given in the question stem indicate that ARMs are not in alarm and are currently stable, which precludes an Alert declaration. Additionally, current indications do not meet the conditions for entering ABN-FUEL-HAND.

Technical Reference(s)

PPM 13.1.1, Classifying the Emergency (EAL Charts)

ABN-FPC-LOSS, Loss of Fuel Pool Cooling Attached w/ Revision #

See Comments / Reference ABN-FUEL-HAND, Damage While Handling Fuel Tech Ref 4 Proposed references to be provided during examination: PPM 13.1.1 Learning Objective: 13463 - State when an emergency classification must be made.

Question Source: #: NRC 2011 Exam

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: 2011 Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the use of EAL Charts along with understanding of the entry conditions for Abnormal Procedures.

Page 4 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 10 CFR Part 55 Content: 55.43 1 Page 5 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Comments /

Reference:

PPM 13.1.1 Rev: Major: 49 Minor: 1 Page 6 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Comments /

Reference:

ABN-FPC-LOSS Rev: Major: 16 Minor: N/A Page 7 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Comments /

Reference:

ABN-FUEL-HAND Rev: Major: 004 Minor: 002 Page 8 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Examination Outline Cross-reference: 80 Revision: 2 Date: 1/28/21 Tier: 1 Group: 1 K/A Number: 295025.2.4.4 Level of Difficulty: 3 SRO Importance Rating: 4.7 K/A

Description:

Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

CGS is in Mode 1.

Reactor power is 100%.

• DEH is in automatic.

RPV pressure is 1038 psig, up slow.

What actions should the CRS take?

The CRS should direct lowering pressure using DEH in (1) in accordance with (2) .

A. (1) TP MANUAL (2) PPM 5.1.1, RPV Control B. (1) TP MANUAL (2) ABN-PRESSURE, Unplanned Reactor Pressure Change C. (1) BPV MANUAL (2) PPM 5.1.1, RPV Control D. (1) BPV MANUAL (2) ABN-PRESSURE, Unplanned Reactor Pressure Change Answer: B K/A Match:

Requires knowledge of the entry conditions for ABN-PRESSURE and PPM 5.1.1.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Explanation:

A. Incorrect. Plausible since TP MANUAL is used to lower pressure. However, for the conditions given, ABN-PRESSURE is entered.

B. Correct. ABN-PRESSURE is entered based on an unplanned RPV pressure change. PPM 5.1.1 is only entered if RPV pressure is approaching 1060 psig. In accordance with ABN-PRESSURE, section 4.2, Pressure control is first attempted with DEH in TP MANUAL. If this is not effective in stemming the pressure rise, DEH is placed in BPV MANUAL.

C. Incorrect. Plausible since PPM 5.1.1 is entered when RPV pressure is high (1060 psig) and pressure may be controlled with DEH BPV MANUAL when in PPM 5.1.1. However, for the conditions given, ABN-PRESSURE is entered. Additionally, BPV MANUAL is only used if TP MANUAL is not effective in lowering pressure.

D. Incorrect. Plausible since, for the conditions given, ABN-PRESSURE is entered. However, pressure is controlled in TP MANUAL.

Technical Reference(s)

ABN-PRESSURE, Unplanned Reactor Pressure Change PPM 5.1.1, RPV Control Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 15788 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-PRESSURE.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information given in the question with an understanding of EOP and ABN entry requirements, along with a knowledge of ABN subsequent actions to lower pressure.

10 CFR Part 55 Content: 55.43 5 Page 3 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Comments /

Reference:

ABN-PRESSURE Rev: Major: 015 Minor: 002 Page 4 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Page 5 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Comments /

Reference:

PPM 5.1.1 Rev: Major: 028 Minor: N/A Page 6 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 Examination Outline Cross-reference: 81 Revision: 2 Date: 2/4/21 Tier: 1 Group: 1 K/A Number: 295028.EA2.02 Level of Difficulty: 2 SRO Importance Rating: 3.9 K/A

Description:

Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE:

Reactor pressure CGS is in Mode 1.

An event causes the crew to enter PPM 5.2.1, Primary Containment Control, due to rising Drywell temperature.

As Drywell temperature continues to rise, the crew attempts to manually scram the reactor.

The scram is not successful, and the crew enters PPM 5.1.2, RPV Control, ATWS.

Current plant conditions:

• RPV pressure is 550 psig and stable.

• RPV level is -150 inches and stable.

• RCIC is in operation and injecting into the RPV.

• All other RPV injection sources are unavailable.

Drywell temperature reaches 335°F and cannot be lowered.

What action should be taken?

A. Manually lower RPV pressure to 175 psig to 300 psig. Do not exceed a cooldown rate of 100°F/hr.

B. Perform an Emergency Depressurization and maintain RPV pressure 175 psig to 300 psig.

C. Maintain RPV pressure 500 to 600 psig until rod pattern alone can always ensure the reactor is shutdown.

D. Perform an Emergency Depressurization and maintain RPV pressure LT 40 psig above Wetwell pressure.

Answer: B K/A Match:

Requires knowledge of the with a high Drywell temperature.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 Explanation:

A. Incorrect. Plausible since the plant is in an ATWS condition and lowering pressure will lower RPV temperature which is undesirable during an ATWS (see PPM 5.1.2, step P-7). Maintaining cooldown rate will slow RPV cooldown rate. Plausibility is enhanced since the pressure band is correct for given conditions. However, with the conditions listed in the question stem, the correct action is to ED and maintain RPV high enough to maintain RCIC operation.

B. Correct. In accordance with PPM 5.2.1, step DT-8, when Drywell temperature cannot be restored and maintained below 330°F, an ED is performed in accordance with PPM 5.1.5, Emergency RPV Depressurization, ATWS. Normally, RPV pressure is maintained LT 40 psig above Wetwell pressure following an ED. However, PPM 5.1.5, step P-1 states IF it is anticipated that RPV depressurization will result in loss of injection required for adequate core cooling, 1. Terminate RPV depressurization and 2. Control RPV pressure as low as practicable Since RCIC is the only source of makeup to the RPV, depressurization should be terminated early to ensure sufficient steam pressure to operate RCIC. OI-15, step 4.2.4, clarifies that RPV pressure should be controlled between 175 psig and 300 psig.

C. Incorrect. Plausible since the plant is in an ATWS condition and it is desirable to maintain pressure to minimize positive reactivity addition due to plant cooldown (see PPM 5.1.2, step P-7).

However, with the conditions listed in the question stem, the correct action is to ED and maintain RPV high enough to maintain RCIC operation.

D. Incorrect. Plausible since this is the normal method for controlling RPV pressure following an ED.

However, with the conditions listed in the question stem, the correct action is to ED and maintain RPV high enough to maintain RCIC operation.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control PPM 5.1.2, RPV Control, ATWS Attached w/ Revision #

See Comments / Reference OI-15, EOP and EAL Clarifications PPM 5.1.5, Emergency RPV Depressurization Proposed references to be provided during examination: None Learning Objective: 8303 - Given a list, identify the statement that describes the reason for emergency depressurizing the RPV if wetwell temperature and reactor pressure cannot be maintained below the HCTL.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Page 3 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 Question Cognitive Level:

Justification for Cognitive Level Examinee must synthesize information given in the stem with a knowledge of the parameters that affect HCTL along with the actions required by the EOPs when HCTL is exceeded.

10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

PPM 5.2.1 Rev: 28 Page 4 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A Page 5 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 Comments /

Reference:

OI-15 Rev: Major: 032 Minor: N/A Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 Comments /

Reference:

PPM 5.1.5 Rev: 10 Comment/Reference.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Examination Outline Cross-reference: 82 Revision: 2 Date: 2/4/21 Tier: 1 Group: 1 K/A Number: 295030.2.4.1 Level of Difficulty: 3 SRO Importance Rating: 4.8 K/A

Description:

Low Suppression Pool Water Level: Knowledge of EOP entry conditions and immediate action steps.

CGS is in Mode 1.

The following annunciators are in alarm:

• 4.601.A11.2-3: SUPPRESSION POOL LEVEL HIGH/LOW

• 4.601.A12.2-3: SUPPRESSION POOL LEVEL HIGH/LOW Suppression Pool level is -2.5 inches, down slow.

The crew enters PPM 5.2.1, Primary Containment Control.

What are the negative effects of low Suppression Pool level and when is a reactor scram required?

Low Suppression Pool water level could result in (1) during SRV discharges. A reactor scram is required (2) 19 feet, 2 inches in the wetwell.

A. (1) inability to adequately condense steam (2) before level drops to B. (1) inability to adequately condense steam (2) if level cannot be maintained above C. (1) exceeding code allowable SRV tailpipe limits (2) before level drops to D. (1) exceeding code allowable SRV tailpipe limits (2) if level cannot be maintained above Answer: A K/A Match:

Requires knowledge of the basis behind the entry conditions for PPM 5.2.1, Primary Containment Control, for low suppression pool level, and required actions if level is not restored.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Explanation:

A. Correct. In accordance with TS Bases for LCO 3.6.2.2 and PPM 5.0.10, section 8.9.7, low Suppression Pool water level could inadequately condense steam during SRV discharges. Level at -2.5 inches is approximately 30 feet, 7 inches. PPM 5.2.1, step L-5 states that the crew must enter PPM 5.1.1 (requires a reactor scram) before wetwell level drops to 19 feet, 2 inches.

B. Incorrect. Plausible since the stated negative effect of low Suppression Pool level is correct.

Plausibility is enhanced since an ED is required when wetwell level cannot be maintained above 19 feet, 2 inches. However, a reactor scram is required prior to wetwell level reaching 19 feet, 2 inches.

C. Incorrect. Plausible since a reactor scram is required before wetwell level reaches 19 feet, 2 inches. Plausibility is enhanced since PPM 5.2.1 and TS LCO 3.6.2.2 are entered when wetwell level is below - 2 inches. However, the stated negative effect is for high Suppression Pool water level, not low level.

D. Incorrect. Plausible since PPM 5.2.1 and TS LCO 3.6.2.2 are entered when wetwell level is below

- 2 inches. Plausibility is enhanced since an ED is required when wetwell level cannot be maintained above 19 feet, 2 inches. However, a reactor scram is required prior to wetwell level reaching 19 feet, 2 inches. Additionally, the stated negative effect is for high Suppression Pool water level, not low level.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control LCO 3.6.2.2 bases, Suppression Pool Water Level Attached w/ Revision #

See Comments / Reference PPM 5.0.10, Flowchart Training Manual Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5643 - Referencing Columbia Generating Station Technical Specifications (section 3 only for initial license candidates) associated with the Primary Containment System and a set of plant conditions; determine as applicable the LCO, the action statement and the appropriate bases.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Page 3 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Justification for Cognitive Level Requires Candidate to synthesize information given in the question with a knowledge of the basis behind PPM 5.2.1 entry conditions and when a scram is required for low suppression pool level.

10 CFR Part 55 Content: 55.43 2 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A Page 4 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Page 5 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Comments /

Reference:

LCO 3.6.2.2 Basis Rev: Major: 114 Minor: N/A Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Comments /

Reference:

PPM 5.0.10 Rev: Major: 022 Minor: N/A Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-83 Examination Outline Cross-reference: 83 Revision: 1 Date: 2/1/21 Tier: 1 Group: 2 K/A Number: 295008.2.2.38 Level of Difficulty: 3 SRO Importance Rating: 4.5 K/A

Description:

High Reactor Water Level: Knowledge of conditions and limitations in the facility license.

CGS is in Mode 1.

• Reactor power is 50%

The Feedwater Level Control System has failed, providing maximum feed.

In accordance with Technical Specifications, what is the thermal limit of concern and how does the plant respond to prevent exceeding this limit?

The challenge to (1) is mitigated by a reactor scram initiated by an automatic (2) .

A. (1) Minimum Critical Power Ratio (MCPR)

(2) trip of the main turbine B. (1) Minimum Critical Power Ratio (MCPR)

(2) trip of both Feedwater Pumps C. (1) Linear Heat Generation Rate (LHGR)

(2) trip of the main turbine D. (1) Linear Heat Generation Rate (LHGR)

(2) trip of both Feedwater Pumps Answer: A K/A Match:

Requires knowledge of the effects of high reactor water level on the facility license (technical specificatons).

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-83 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-83 Explanation:

A. Correct. In accordance with the technical specification basis for TS LCO 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation, a reactor scram is indirectly caused by RPV level going greater than Level 8 due to an automatic main turbine trip (with reactor power GE 29.5%). Since the question stem states that reactor power is 50%, this trip will occur. Additionally, the basis states that the reactor scram mitigates the reduction in MCPR.

B. Incorrect. Plausible since MCPR is the limit of concern. Plausibility is enhanced if it is believed that the trip of both feedwater pumps causes a reactor scram, since a trip of both feedwater pumps is automatically inserted when RPV level reaches Level 8. However, tripping the feedwater pumps does not cause a reactor scram. The scram is caused by the automatic main turbine trip.

C. Incorrect. Plausible since the main turbine trip and subsequent reactor scram mitigates the challenge to the limit of concern. However, this limit is MCPR.

D. Incorrect. Plausible if it is believed that the trip of both feedwater pumps causes a reactor scram, since a trip of both feedwater pumps is automatically inserted when RPV level reaches Level 8.

However, tripping the feedwater pumps does not cause a reactor scram. The scram is caused by the automatic main turbine trip. Additionally, the limit of concern is MCPR.

Technical Reference(s)

Technical Specification Bases Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13425 - Describe the bases for the Minimum Critical Power Ratio Safety Limit.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must demonstrate an understanding of the challenge to reactor safety from a high reactor water level event along with a knowledge of the automatic actions that mitigate this challenge.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-83 10 CFR Part 55 Content: 55.43 2 Comments /

Reference:

Technical Specification Bases Rev: Major: 98 Minor: N/A Page 4 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 Examination Outline Cross-reference: 84 Revision: 2 Date: 1/28/21 Tier: 1 Group: 2 K/A Number: 295033.2.4.21 Level of Difficulty: 3 SRO Importance Rating: 4.6 K/A

Description:

High Secondary Containment Area Radiation Levels: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

CGS is in Mode 1.

An event occurs which causes high secondary containment radiation levels.

The crew enters PPM 5.3.1, Secondary Containment Control.

The CRS is evaluating secondary containment area radiation levels against their Maximum Safe Operating Value (MSOV).

What is the basis for the MSOV value?

The MSOV is high enough to (1) and low enough to allow time for shutdown or leak isolation without exceeding (2) .

A. (1) prevent MSOV violations during normal plant operation (2) the allowable dose of the most sensitive safety related equipment B. (1) prevent MSOV violations during normal plant operation (2) operator normal dose limits during event response C. (1) indicate a substantial and immediate problem (2) the allowable dose of the most sensitive safety related equipment D. (1) indicate a substantial and immediate problem (2) operator normal dose limits during event response Answer: C K/A Match:

Requires knowledge of logic used to assess secondary containment radiation levels to determine containment conditions.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 Explanation:

A. Incorrect. Plausible since the MSOV is set low enough to prevent exceeding the total integrated dose allowable for the most sensitive safety related equipment. However, , the MSOV is set high enough to alert operators that there is a condition that represents a substantial risk to elevated dose to the public and requires immediate action.

B. Incorrect. Plausible since the MSOV is much higher than normal radiation levels. Plausibility is enhanced since all actions are designed to minimize dose. However, the MSOV is set high enough to alert operators that there is a condition that represents a substantial risk to elevated dose to the public and requires immediate action.

C. Correct. PPM 5.0.10, Flowchart Training Manual, specifies that the MSOV for secondary containment radiation levels is high enough to be indicative of substantial and immediate problems yet low enough to allow time for shutdown or isolation of a leak without exceeding the total integrated dose allowable for even the most sensitive safety related equipment.

D. Incorrect. Plausible since the MSOV is set high enough to alert operators that there is a condition that represents a substantial risk to elevated dose to the public and requires immediate action.

However, the MSOV is set low enough to prevent exceeding the total integrated dose allowable for the most sensitive safety related equipment.

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8456 - Define Maximum Safe Operating Value for the following secondary containment parameters: c. Area radiation levels Question Source: #: Bank #

1. SRO00010 (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the reasons for the MSOV value.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 10 CFR Part 55 Content: 55.43 2 Page 4 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 Comments /

Reference:

PPM 5.0.10 Rev: Major: 022 Minor: 000 Page 5 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 Comments /

Reference:

Parent Question SRO00010 Rev: Major: N/A Minor: N/A Page 6 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 Page 7 of 9

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 Page 9 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 Examination Outline Cross-reference: 85 Revision: 2 Date: 2/1/21 Tier: 1 Group: 2 K/A Number: 295036.EA2.02 Level of Difficulty: 3 SRO Importance Rating: 3.1 K/A

Description:

Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Water level in the affected area.

CGS is in Mode 1.

HPCS Pump Room water level is above the alarm setpoint.

The CRS enters PPM 5.3.1, Secondary Containment Control.

The HPCS Pump Room door cannot be opened.

If time does not permit removal of floor plugs to determine room water level, which of the following should be met to consider that the HPCS Pump Room water level is above its Maximum Safe Operating Value (MSOV)?

(1) Another areas level exceeds the MSOV.

(2) HPCS Pump Room level remains above the alarm setpoint.

(3) A primary system is discharging into Secondary Containment.

A. (1) ONLY.

B. (2) ONLY.

C. (1) and (3) ONLY.

D. (2) and (3) ONLY.

Answer: B K/A Match:

Requires knowledge of alternate method of determining water level in the HPCS Pump Room.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 Explanation:

A. Incorrect. Plausible since another area with level above the MSOV indicates a severe problem that requires additional action in accordance with PPM 5.3.1, steps SC-11 and SC-15. However, water level in a different area is not used to determine severity of level in an area where entry is not possible.

B. Correct. In accordance with PPM 5.5.27, Reactor Building 422 Max Safe Operating Level Measurement, if the affected room door cannot be opened, and time does not permit removal of floor plugs, then consider the level in the room is above the Max Safe Operating Level. This means that level is considered GT MSOV with a high level alarm only.

C. Incorrect. Plausible since (1) is correct. Plausibility is enhanced since a primary system discharging into Secondary Containment is the most probably reason for HPCS room level rising.

Additionally, a primary system discharging into Secondary Containment is an indication used in PPM 5.3.1 to indicate the severity of the event (step SC-13). However, this indication is not used to determine level in an area where access is not available.

D. Incorrect. Plausible since both (2) and (3) are parameters used to determine the severity of the event and the mitigative actions required. However, neither of these parameters are used to determine the severity of area water level when area access is not available.

Technical Reference(s)

PPM 5.3.1, Secondary Containment Control PPM 5.5.27, RB 422 Max Safe Operating Level Measurement Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 69220 - Given plant conditions identify those symptoms that would indicate Reactor Building 422 Area Flooding.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Page 3 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 Requires knowledge of the method to determine if room level is greater than MSOV when floor plugs cannot be removed.

10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

PPM 5.3.1 Rev: 21 Page 4 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 Page 5 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 Comments /

Reference:

PPM 5.5.27 Rev: Major: 005 Minor: N/A Page 6 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Examination Outline Cross-reference: 86 Revision: 1 Date: 7/8/20 Tier: 2 Group: 1 K/A Number: 209001.2.2.12 Level of Difficulty: 3 SRO Importance Rating: 4.1 K/A

Description:

Low Pressure Core Spray: Knowledge of surveillance procedures.

CGS is in Mode 1.

Surveillance OSP-LPCS/IST-Q702, LPCS System Operability Test, has just been completed.

• The surveillance was performed as an ASME required Comprehensive Pump Test (CPT).

Partial surveillance results for LPCS-P-1:

• Suction Pressure (Fluke Module): 15.0 psig

• Discharge Pressure (Fluke Module): 310.0 psig

• Differential pressure p (Discharge Pressure-Suction Pressure): 295 psig

• Indicated flowrate (TDAS X164): 6410 gpm Surveillance results for LPCS-V-12:

• Opening time: 22.32 seconds

• Closing time: 14.74 seconds Using the references provided, what is the next action that is required to be performed?

A. Perform a second stroke time test of LPCS-V-12 immediately.

B. Close and de-activate LPCS-V-12 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

C. Restore LPCS to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

D. Restore LPCS to operable status within 7 days.

Answer: D K/A Match:

Requires a knowledge of LPCS System Operability surveillance and its relationship to technical specification requirements.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Explanation:

A. Incorrect. Plausible since the raw closing stroke time of LPCS-V-12 is outside the Alert Level in accordance with OSP-LPCS/IST-Q702, Attachment 9.1. Section 4.6.2 states that valves with measurements outside the Alert Value shall be immediately retested or declared inoperable.

However, note (+2) or Attachment 9.1 gives guidance for rounding times to the nearest second, which will put LPCS-V-12 within the Alert Level.

B. Incorrect. Plausible since the raw closing stroke time of LPCS-V-12 is outside the Alert Level in accordance with OSP-LPCS/IST-Q702, Attachment 9.1. Section 4.6.2 states that valves with measurements outside the Alert Value shall be immediately retested or declared inoperable. If declared inoperable, LPCS-V-12 would need to be closed and deenergized in accordance with LCO 3.6.1.3. However, note (+2) or Attachment 9.1 gives guidance for rounding times to the nearest second, which will put LPCS-V-12 within the Alert Level.

C. Incorrect. Based on the surveillance results given, p is in the Action Range on Attachment 9.8.

Section 4.7 states that the LPCS pump must be declared inoperable. If LCO 3.5.1.C is interpreted as one ECCS spray system inoperable then 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore to operable could be interpreted as the right answer.

D. Correct. Based on the surveillance results given, p is in the Action Range on Attachment 9.8.

Section 4.7 states that the LPCS pump must be declared inoperable. This will make LCO 3.5.1.A applicable, which requires the LPCS system to be in operable status within 7 days.

Technical Reference(s)

OSP-LPCS/IST-Q702, LPCS System Operability Test TS LCO 3.5.1, ECCS Operating Attached w/ Revision #

See Comments / Reference Proposed references to be provided during examination: OSP-LPCS/IST-Q702 (Partial)

TS LCO 3.5.1 (Partial)

TS LCO 3.6.1.3 LCS Table 1.6.1.3 (Partial)

Learning Objective: 5487 - Referencing Columbia Generating Station Technical Specifications associated with the Low Pressure Core Spray System and a set of plant conditions, determine as applicable the LSSS, the LCO, the action statement, and the appropriate bases.

Question Source: #:

1. (Note changes or attach parent)

Page 3 of 13

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize the conditions listed in the question using a knowledge of TS LCO 3.5.1 and an understanding to the LPCS System Operability surveillance.

10 CFR Part 55 Content: 55.43 2 Comments /

Reference:

OSP-LPCS/IST-Q702 Rev: Major: 044 Minor: 001 Page 4 of 13

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Page 5 of 13

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Page 9 of 13

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Comments /

Reference:

LCO 3.5.1 Rev: Major: 258 Minor: N/A Page 10 of 13

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Page 13 of 13

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 Examination Outline Cross-reference: 87 Revision: 0 Date: 8/27/20 Tier: 2 Group: 1 K/A Number: 211000.2.1.31 Level of Difficulty: 2 SRO Importance Rating: 4.3 K/A

Description:

Standby Liquid Control: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

CGS is in Mode 1 The following annunciator is in alarm:

• 4.603.A8.6-8: SLC DIV 2 OUT OF SERVICE SLC system indications:

SLC-V-4B LOSS OF CONTINUITY How does this affect the operation and technical specification operability of SLC?

SLC train B (1) operable. If initiated, SLC-P-1B (2) inject into the RPV.

A. (1) is (2) will B. (1) is (2) will not C. (1) is not (2) will D. (1) is not (2) will not Answer: C K/A Match:

Requires understanding of SLC abnormal indications and effect on plant operations.

SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 Explanation:

Page 2 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 A. Incorrect. Plausible since SLC-P-1B will inject to the RPV via SLC-V-4A. Plausiblity is enhanced since the SLC system will still provide 100% flow. However, SLC Train B is considered inoperable.

B. Incorrect. Plausible since the SLC system will still provide 100% flow. Plausibility is enhanced since the Train B discharge squib valve, SLC-V-4B, will not open on an initiation. However, on initiation, SLC-P-1B will inject to the RPV via SLC-V-4A. Additionally, SLC Train B is considered inoperable.

C. Correct. In accordance with Technical Specification Bases for LCO 3.7.1, to be considered operable, each SLC train must have an operable pump, explosive valve and associated piping, valves and instruments and controls. Since the SLC pump discharge paths are cross-connected upstream of the squib valves, SLC-P-1B will still inject to the RPV on SLC initiation.

D. Incorrect. Plausible since SLC Train B is considered inoperable. Plausibility is enhanced since the Train B discharge squib valve, SLC-V-4B, will not open on an initiation. However, on initiation, SLC-P-1B will inject to the RPV via SLC-V-4A.

Technical Reference(s)

Technical Specification Basis for LCO 3.7.1, SLC System SD000172, Standby Liquid Control System Description Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7636 - Explain how the SLC Explosive valve detonator continuity is verified.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires an understanding of the operation of the SLC system with a single squib valve failed closed along with a knowledge of the operability requirements for individual trains of SLC.

10 CFR Part 55 Content: 55.43 2 Page 3 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 Comments /

Reference:

TS Basis for LCO 3.7.1 Rev: Major: 87 Minor: N/A Page 4 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 Comments /

Reference:

SD000172 Rev: Major: 13 Minor: 1 Page 5 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 Examination Outline Cross-reference: 88 Revision: 0 Date: 11/16/20 Tier: 2 Group: 1 K/A Number: 261000.A2.09 Level of Difficulty: 3 SRO Importance Rating: 2.6 K/A

Description:

Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Plant air system failure CGS is in Mode 1.

A complete Loss of CAS occurs.

How should the CRS direct controlling Reactor Building (RB) pressure?

The CRS should direct operators to...

A. verify RB pressure is being controlled LTE -0.25 WG in accordance with SOP-HVAC/RB-OPS, Reactor Building Ventilation System Operation.

B. secure one train of SGT in accordance with SOP-SGT-OPS, since both SGT trains automatically started.

C. restart RB HVAC in accordance with SOP-RBHVAC-RESTART-QC, since both trains of SGT are isolated due to the loss of CAS.

D. start one train of SGT in accordance with SOP-SGT-START-DIV1(2)-QC since SGT trains did not isolate on the loss of CAS.

Answer: D K/A Match:

Requires knowledge of the effects on SGT with a loss of CAS and a knowledge of the procedure used to maintain Reactor Building pressure.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 Explanation:

A. Incorrect. Plausible since RB HVAC fans do not trip directly from a loss of CAS. Plausibility is enhanced if it is believed that SGT-V-2A(B), Reactor Building Intake Isolation Valves, fail closed on a loss of CAS. However, REA/ROA dampers fail closed on a loss of CAS. This causes the REA/ROA fans to trip on differential pressure.

B. Incorrect. Plausible since one train of SGT should be secured per SOP-SGT-OPS, if both trains are operating. Plausibility is enhanced if it is believed that SGT auto-starts on a loss of CAS.

However, SGT does not auto-start and must be manually started.

C. Incorrect. Plausible PPM 5.3.1, step SC-1 directs re-starting RB HVAC if SGT cannot restore RB differential pressure. Plausibility is enhanced since SGT-V-2A and 2B use CAS. However, SGT-V-2A and 2B fail open, which is the position necessary to operate SGT. Therefore one train of SGT is started.

D. Correct. In accordance with ABN-CAS, Control Air System Failure, a complete loss of CAS will cause REA and ROA isolation valves to fail closed, and RB HVAC will be isolated. Step 4.6 directs operators to restore RB differential pressure by starting one train of SGT per SOP-SGT-START-DIV/1(2)- QC, or SOP-SGT-START. The note prior to step 4.6 informs operators that SGT-V-2A(B) fail open on a loss of CAS and do not need to be manually repositioned.

Technical Reference(s)

ABN-CAS, Control Air System Failure PPM 5.3.1, Secondary Containment Control Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 15734 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-CAS.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires candidate to synthesize the information in the question with a knowledge of system response to a loss of CAS along with an understanding of ABN-CAS supplemental actions.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

ABN-CAS Rev: Major: 011 Minor: 001 Page 4 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 Page 5 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: N/A Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Examination Outline Cross-reference: 89 Revision: 1 Date: 2/1/21 Tier: 2 Group: 1 K/A Number: 263000.A2.02 Level of Difficulty: 3 SRO Importance Rating: 3.0 K/A

Description:

Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of ventilation during charging CGS is in Mode 1.

A loss of Div. 1 125 VDC Battery Room ventilation occurs.

• Div. 1 125 VDC Battery Room temperature lowers to 67°F.

In accordance with Licensee Controlled Specification (LCS) 1.8.6.2, 125 and 250 VDC Battery Parameters, what is the first action required to be performed and why is this action performed?

Verify (1) . This action is performed to (2) .

A. (1) electrolyte level is above minimum limit (2) ensure requirements for station blackout are met B. (1) electrolyte level is above minimum limit (2) prevent battery plate dryout and degradation C. (1) battery room temperature is within requirements (2) ensure requirements for station blackout are met D. (1) battery room temperature is within requirements (2) prevent battery plate dryout and degradation Answer: C K/A Match:

Requires knowledge of actions to take if battery room ventilation fails and why those actions are taken.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Explanation:

A. Incorrect. Plausible since the reason for the action taken on a low battery temperature is correct.

Plausibility is enhanced since lowering room temperature could cause electrolyte temperature to lower. However, the correct action for the conditions given is to verify room temperature is within the limit.

B. Incorrect. Plausible since lowering battery temperature could cause cell electrolyte level to lower.

Plausibility is enhanced since a low electrolyte level could cause battery plate dryout and degradation. However, the correct action for the conditions given is to verify room temperature is above the limit. This is done to ensure that the battery will meet the requirements required for a SBO.

C. Correct. In accordance with LCS 1.8.6.2, if battery room ventilation is lost, room temperature is verified to meet requirements within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This action is taken to ensure that the battery will meet station blackout requirements.

D. Incorrect. Plausible since the first required action is to verify temperature is within the limits in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. However, this action is performed to ensure that the battery will meet SBO requirements.

Technical Reference(s)

LCS 1.8.6.2, 125 and 250 VDC Battery Parameters LCS 1.8.6.2 Bases Attached w/ Revision #

See Comments / Reference TS LCO 3.8.6 Bases Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 6925 - Identify the basis for any LCO.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires the examinee to synthesize conditions given in the question stem using a knowledge of LCS compensatory actions along with a knowledge of the reasons for these actions.

10 CFR Part 55 Content: 55.43 2 Page 3 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Comments /

Reference:

LCS 1.8.6.2 Rev: Major: 114 Minor: N/A Page 4 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Comments /

Reference:

LCS 1.8.6.2 Rev: Major: 114 Minor: N/A Page 5 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Comments /

Reference:

TS LCO 3.8.6 Bases Rev: Major: 114 Minor: N/A Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Examination Outline Cross-reference: 90 Revision: 0 Date: 11/13/20 Tier: 2 Group: 1 K/A Number: 400000.A2.01 Level of Difficulty: 2 SRO Importance Rating: 3.4 K/A

Description:

Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Loss of CCW pump CGS is in Mode 1.

RCC-P-1A and RCC-P-1C are running.

The supply breaker to SL-81, CB-8/81, trips.

How is RCC affected and what action(s) should the CRS direct?

There is a (1) . Direct control room operators to (2) .

A. (1) partial loss of RCC flow (2) manually start the standby RCC pump B. (1) complete loss of RCC flow to the Drywell (2) manually SCRAM the reactor C. (1) partial loss of RCC flow (2) place RCC-P-1B and RCC-P-1C in PTL.

D. (1) complete loss of RCC flow to the Radwaste/Rx Building (2) verify that the standby RCC pump has automatically started Answer: C K/A Match:

Requires knowledge of the effects of losing power to a running RCC pump and the actions required.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Explanation:

A. Incorrect. Plausible since a partial loss of RCC flow has occurred. Plausibility is enhanced since the standby RCC pump will not automatically start on a loss of one RCC pump when the lost pumps breaker is still closed (ie - shaft shear, loss of power to the bus, etc.) and ABN-RCC, step 4.2.1 requires operators to manually start the standby RCC pump if not running. However, since RCC-P-1B is also powered from SL-81, it cannot be manually started.

B. Incorrect. Plausible since the RCC Outboard Supply valve (RCC-V-104) and Inboard Return valve (RCC-V-40) are powered from MC-8B-A, which loses power when CB-8/81 trips, and a candidate may believe that these valves will close. Plausibility is enhanced since ABN-RCC, step 4.1, requires a manual scram on a complete loss of RCC flow to the drywell. However, the drywell isolation valves do not reposition on a loss of power. Therefore, a manual scram is not required.

C. Correct. RCC-P-1B and RCC-P-1C are powered from SL-81 while RCC-P-1A is powered from SL-71. Normally, the standby RCC pump will automatically start on a loss of a running RCC pump (less than two RCC pump breakers closed). However, for the conditions given, RCC-P-1C breaker remains closed, so RCC-P-1B will not receive a start signal. Even if a start signal was received, RCC-P-1B could not start since it has no power. RCC-V-6 automatically closes when less than two RCC pump breakers are closed for greater than 10 seconds. It closes to ensure that sufficient RCC flow is directed to vital equipment in the drywell. For the conditions given, RCC-V-6 will not automatically close. On a loss of SL-81, ABN-ELEC-SM3/SM8, step 4.5.1, requires operators to place RCC-P-1B and RCC-P-1C to PTL. This will open the pump breakers and cause RCC-V-6 to automatically close.

D. Incorrect. Plausible since a complete loss of flow to the Radwaste/Rx Building should have occurred with only one RCC pump running (see explanation C above), and ABN-RCC, step 4.2.1, requires operators to verify that the standby RCC pump has started. However, since all RCC pump breakers are closed and RCC-P-1B is not powered, the standby RCC pump will not start and cannot be started manually. Additionally, RCC-V-6 will not automatically close.

Technical Reference(s)

SD000196, RCC ABN-RCC, Loss of RCC Attached w/ Revision #

ABN-ELEC-SM3/SM8, SM-3, SM-8, SM-85, SM-82, SL-81, SL-83 & See Comments / Reference SL-31 Distribution System Failures Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7668 - Predict the plant response to a: a. Partial and a complete loss of the RCC system.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee synthesize information given in the question with an understanding of the requirements for automatic actions of the RCC system along with procedural requirements of ABN-RCC and ABN-ELEC-SM3/SM8 10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

SD000196 Rev: Major: 14 Minor: 2 Page 4 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Page 5 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Comments /

Reference:

ABN-RCC Rev: Major: 006 Minor: 004 Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Comments /

Reference:

ABN-ELEC-SM3/SM8 Rev: Major: 022 Minor: N/A Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Examination Outline Cross-reference: 91 Revision: 1 Date: 2/25/21 Tier: 2 Group: 2 K/A Number: 216000.2.2.22 Level of Difficulty: 3 SRO Importance Rating: 4.7 K/A

Description:

Nuclear Boiler Instrumentation - Knowledge of limiting conditions for operations and safety limits.

CGS is in Mode 1.

A mechanical failure of MS-PS-23A (RPS High RPV Pressure Instrument) occurs.

The CRS enters TS LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation.

• All required TS actions for this condition are complete.

15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> later the crew performs OSP-INST-H101, Shift and Daily Instrument Checks (Modes 1, 2, 3).

• MS-PS-23D (RPS High RPV Pressure Instrument), fails its channel check.

Using the reference provided, what TS action should be taken that has not already been taken?

A. Restore RPS trip capability in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

B. Place a channel in one trip system in trip in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

C. Place MS-PS-23D in trip in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D. Be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Answer: D K/A Match:

Requires knowledge of the LCOs for the Nuclear Boiler Instrumentation system.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Explanation:

A. Incorrect. Plausible since two channels are inoperable. However, When channel A failed, the question states that appropriate TS actions were taken, which means that action A1 was performed and channel A was placed in TRIP. Therefore, trip capability is maintained.

B. Incorrect. Plausible since one channel is inoperable in both trip systems. However, channel A is already in trip. Therefore, required action B.1 is already complete.

C. Incorrect. Plausible since required action A.1 should be completed for channel D. However, this will cause a reactor scram. LCO 3.3.1.1 bases states Alternately, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken.

D. Correct. With a failure of MS-PS-23D, LCO 3.3.1.1 required action A.1 requires the channel to be placed in Trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. However, LCO 3.3.1.1 bases states that if this action will cause a scram, condition D should be entered, which requires the reactor to be placed in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Technical Reference(s)

LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 Bases Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: LCO 3.3.1.1 Actions and Table 3.3.1.1-1 Learning Objective: 11776 - Referencing Columbia Generating Station Licensee Controlled Specifications associated with the Nuclear Boiler Instrumentation System and a set of plant conditions, determine as applicable the LCO, the action statement, and the appropriate bases.

Question Source: #: LO03048

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing conditions given in the question with a knowledge of LCO 3.3.1.1 action statements along with the basis for these actions Page 3 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 10 CFR Part 55 Content: 55.43 2 Comments /

Reference:

LCO 3.3.1.1 Rev: Major: 258 Minor: N/A Page 4 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Page 5 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Comments /

Reference:

LCO 3.3.1.1 Bases Rev: Major: 114 Minor: N/A Page 6 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Comments /

Reference:

Provided Reference Rev: Major: 258 Minor: N/A Page 7 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Page 8 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Page 9 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Page 10 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 Examination Outline Cross-reference: 92 Revision: 1 Date: 2/1/21 Tier: 2 Group: 2 K/A Number: 226001.A2.14 Level of Difficulty: 2 SRO Importance Rating: 3.1 K/A

Description:

Ability to (a) predict the impacts of the following on the RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High suppression pool level.

CGS is in Mode 1.

A LOCA occurs.

The crew is performing actions in accordance with PPM 5.1.1, RPV Control and PPM 5.2.1, Primary Containment Control.

Current plant conditions:

• RHR-A is lined up to inject into the RPV.

• RHR-B- is spraying the wetwell.

• Drywell temperature is 240°F, up slow.

• Drywell pressure is 21 psig, up slow.

• Wetwell pressure is 15 psig, up slow.

• Wetwell level is 52 feet, up slow.

What action should the CRS direct to address Primary Containment parameters and why is this action taken?

The CRS should direct the crew to A. vent Primary Containment in accordance with PPM 5.5.14, Emergency Wetwell Venting, to preclude containment failure.

B. spray the drywell with Service Water in accordance with PPM 5.5.2, RHR/SW Crosstie Lineup, to reduce drywell temperature.

C. emergency depressurize the RPV in accordance with PPM 5.1.3, Emergency RPV Depressurization, to preclude steam in the suppression pool airspace.

D. secure wetwell spray in accordance with SOP-RHR-SPRAY-WW-QC, Initiation of Wetwell Spray - Quick Card, since post-spray drywell vacuum relief may be lost.

Answer: C Page 1 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 K/A Match:

Requires knowledge of the actions required when Wetwell level exceeds the Wetwell-to-Drywell level limit.

SRO Only:

Page 2 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 Explanation:

A. Incorrect. Plausible since PPM 5.2.1, step P-14 requires venting primary containment if the Primary Containment Pressure Limit (PCPL) is exceeded. However, PCPL is not exceeded, Pressure Suppression Pressure (PSP) is exceeded which requires an emergency depressurization in accordance with step P-12.

B. Incorrect. Plausible if it is believed that RHR-A is required to inject into the RPV while Drywell spray is required. However, since wetwell level is greater than 51 feet, drywell spray should not be initiated in accordance with PPM 5.2.1, step DT-5. Additionally, Drywell spray is not required to by initiated until drywell temperature approaches 330°F.

C. Correct. In accordance with the information given in the stem, SRV Tail Pipe Level Limit (SRVTPLL) has been exceeded. PPM 5.2.1, step L-14 requires the performance of an emergency depressurization.

D. Incorrect. Plausible since wetwell level is greater than 51 feet and PPM 5.2.1, step L-17 requires securing drywell spray due to potential loss of drywell vacuum relief. However, wetwell spray is not required to be secured.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8301 - Given a list, identify the statement that describes the failure mode that the Heat Capacity Temperature Limit protects against. (PPM 5.2.1).

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must synthesize information given in the question stem with an understanding of flowchart figures (PCPL, PSP, etc.) along with a knowledge of the procedure steps that are required when a limit is exceeded.

Page 3 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A Page 4 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 Page 5 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 Page 6 of 9

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Examination Outline Cross-reference: 93 Revision: 1 Date: 1/6/21 Tier: 2 Group: 2 K/A Number: 290001.A2.02 Level of Difficulty: 3 SRO Importance Rating: 3.7 K/A

Description:

Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Excessive outleakage CGS is in Mode 1.

Field operators report a breach in the Reactor Building roof.

Plant indications include:

Reactor Building DP is stable.

Which of the following actions should be taken?

1. Enter PPM 5.3.1, Secondary Containment Control, and start one train of SGT.

2. Enter Technical Specification LCO 3.6.4.1, Secondary Containment, and restore secondary containment to operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3. Adjust REA-DPIC-1B in MANUAL and restore secondary containment P in accordance with ARP 4.812.R1.7-3, SEC PRESS CONTR A P HIGH/LOW.

Page 1 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 A. 1 ONLY B. 2 ONLY C. 1 and 3 ONLY D. 2 and 3 ONLY Answer: B K/A Match:

Requires knowledge of the consequences of a leak in secondary containment and the procedures that should be entered to mitigate these consequences.

Page 2 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 SRO Only:

Page 3 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Explanation:

A. Incorrect. Plausible since PPM 5.3.1 is entered on a low vacuum in secondary containment.

Plausibility is enhanced since PPM 5.3.1 contains actions if SGT cannot restore secondary containment P. However, secondary containment P must be 0 inches WG to enter PPM 5.3.1. Additionally, PPM 5.3.1 does not direct starting SGT.

B. Correct. TS 3.6.4.1 requires secondary containment vacuum to be 0.25 inches vacuum WG Additionally. a breach in the Reactor Building roof is also an entry condition for TS 3.6.4.1.

C. Incorrect. Plausible for the reasons listed in explanation A above and since 4.812.R1.7-3 directs operators to place REA-DPIC-1B in MANUAL to restore secondary containment vacuum.

However, this annunciator does not alarm until 0 inches WG. Additionally, REA-DPIC-1B is currently 100% open. Taking the controller to manual will not allow any additional adjustment to restore vacuum.

D. Incorrect. Plausible since (2) is correct. However, (3) is not correct. See explanation C above.

Technical Reference(s)

TS 3.6.4.1, Secondary Containment PPM 5.3.1, Secondary Containment Control Attached w/ Revision #

See Comments / Reference 4.812.R1.7-3, Alarm Response OSP-CONT-M102, Secondary Containment Integrity Verification Proposed references to be provided during examination: N/A Learning Objective: 7007 - With the Technical Specifications provided, locate all Safety Limits and/or LCO's that directly relate to the Secondary Containment.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing conditions given in the question with a knowledge of procedure entry requirements and required actions.

10 CFR Part 55 Content: 55.43 2 Page 4 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Comments /

Reference:

TS 3.6.4.1 Rev: Major: 254 Page 5 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Page 6 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Page 7 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Comments /

Reference:

OSP-CONT-M102 Rev: Major: 12 Minor: N/A Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor:

Page 8 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Comments /

Reference:

4.812.R1.7-3 Rev: Major: 21 Minor: N/A Page 9 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Page 10 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94 Examination Outline Cross-reference: 94 Revision: 2 Date: 1/6/21 Tier: 3 Group: N/A K/A Number: 2.1.7 Level of Difficulty: 3 SRO Importance Rating: 4.7 K/A

Description:

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

CGS is in Mode 2.

A reactor plant startup is in progress in accordance with PPM 3.1.2, Startup Flowchart.

The reactor is critical. The crew is withdrawing control rods to place the reactor in the heating range.

Current plant conditions:

• Reactor period is 85 seconds and stable.

• All SRMs have been fully withdrawn.

• IRMs are on the following ranges:

IRM A: 9 IRM E: 9 IRM B: 8 IRM F: 8 IRM C: 9 IRM G: 9 IRM D: 9 IRM H: 8

• The reactor is currently NOT in the heating range.

What action should the CRS direct?

The CRS should direct A. stopping control rod withdrawal and contacting the SNE.

B. withdrawing control rods to achieve a heatup rate of LE 80°F/hr.

C. adjusting control rod position to maintain a stable period of GT 60 seconds.

D. driving control rods in the reverse order until all control rods are fully inserted.

Answer: A K/A Match:

Requires examinee to understand expected reactor plant response when withdrawing control rods during a reactor startup and the correct response when reactor behavior is not as expected.

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94 SRO Only:

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94 Explanation:

A. Correct. In accordance with PPM 3.1.2, Startup Flowchart, steps Q24/Q25, if nuclear heating does not occur on or before IRMs on range 8, the CRS should stop rod withdrawal and contact the SNE.

B. Incorrect. Plausible since withdrawing control rods to achieve and maintain a heatup rate would be the next step after the reactor is in the heating range. However, since the reactor is not yet in the heating range, a heatup rate cannot be established.

C. Incorrect. Plausible since the direction to adjust control rods to maintain a period GE 60 seconds is applicable while approaching the heating range. However, since IRM D is on a range > range 8, the CRS should direct that control rod withdrawal be suspended.

D. Incorrect. Plausible since fully inserting control rods is the correct action if the reactor isnt critical within the specified maximum and minimum ECP. However, since the reactor is already critical, this requirement does not apply.

Technical Reference(s)

PPM 3.1.2, Startup Flowchart.

Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 6652 - With the procedures available, determine which steps/sections of the procedure must be repeated if the startup is delayed. [PPM 3.1.2] (SRO only)

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information given in the stem with an understanding of expected reactor response during a reactor startup along with a knowledge of the actions required if reactor response is not as expected.

10 CFR Part 55 Content: 55.43 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94 Comments /

Reference:

PPM 3.1.2 Rev: Major: 088 Minor: 000

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 Examination Outline Cross-reference: 95 Revision: 1 Date: 1/6/21 Tier: 3 Group: N/A K/A Number: 2.1.35 Level of Difficulty: 2 SRO Importance Rating: 3.9 K/A

Description:

Knowledge of the fuel-handling responsibilities of SROs.

Which of the following evolutions must be directly supervised by a licensed SRO?

(1) Fuel movement from the core to the Spent Fuel Pool.

(2) Movement of LPRMs within the core.

(3) Fuel movement between locations in the core.

(4) Control Rod blade replacement in a defueled cell.

A. (1) and (2) ONLY B. (1) and (3) ONLY C. (2) and (4) ONLY D. (3) and (4) ONLY Answer: B K/A Match:

Requires knowledge of SRO responsibilities during fuel handling evolutions.

SRO Only:

The question requires knowledge of refuel floor SRO responsibilities which is an SRO-only topic as delineated in NUREG-1021, ES-401, Attachment 2 and 10 CFR 55.43(b)(7)

Page 1 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 Explanation:

A. Incorrect. Plausible since (1) is considered a CORE ALTERATION in accordance with TS and must be supervised by a licensed SRO. However, However, (2) is specifically called out in TS as an evolution that is not considered a CORE ALTERATION, and therefore, licensed SRO supervision is not required.

B. Correct. In accordance with TS, both (1) and (3) are considered a CORE ALTERATION, and must be supervised by a licensed SRO.

C. Incorrect. Plausible since (4) must be supervised by a licensed SRO when a control rod blade is replaced in a fueled cell. However, (2) is specifically called out in TS as an evolution that is not considered a CORE ALTERATION and since the control rod blade is being replaced in a non-fueled cell, it is not considered a CORE ALTERATION. Therefore, licensed SRO supervision is not required for either evolution.

D. Incorrect. Plausible since (3) must be supervised by a licensed SRO and 4) must be supervised by a licensed SRO when a control rod blade is replaced in a fueled cell. However, since the control rod blade is being replaced in a non-fueled cell, it is not considered a CORE ALTERATION. Therefore, licensed SRO supervision is not required for this evolution.

Technical Reference(s)

Technical Specifications PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading Attached w/ Revision #

See Comments / Reference OI-20, Fuel Handling Expectations Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 3051 - Define Core Alteration, including items that are specifically excluded from the definition.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of definition of CORE ALTERATIONS Page 2 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 10 CFR Part 55 Content: 55.43 7 Comments /

Reference:

Technical Specifications Rev: Major: 254 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 Comments /

Reference:

PPM 6.3.2 Rev: Major: 025 Minor: N/A Page 4 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 Comments /

Reference:

OI-20 Rev: Major: 011 Minor: N/A Page 5 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96 Examination Outline Cross-reference: 96 Revision: 1 Date: 11/30/20 Tier: 3 Group: N/A K/A Number: 2.2.25 Level of Difficulty: 2 SRO Importance Rating: 4.2 K/A

Description:

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

What is the basis for the Reactor Coolant System (RCS) Pressure Safety Limit (SL)?

Maintaining RCS pressure below the SL ensures that RPV pressure will remain LE (1) of design pressure at the (2) .

A. (1) 110%

(2) reactor steam dome B. (1) 110%

(2) lowest elevation of the RCS C. (1) 125%

(2) reactor steam dome D. (1) 125%

(2) lowest elevation of the RCS Answer: B K/A Match:

Requires knowledge of the technical specification bases for safety limits.

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96 SRO Only:

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96 Explanation:

A. Incorrect. Plausible since the SL maintains pressure LE 110% of design pressure and the highest pressure in the RPV will be located at the lowest elevation. However, this is based on pressure at the lowest elevation of the RCS.

B. Correct. In accordance with the bases for Technical Specification 2.1.2, the limiting condition for RCS pressure is maintaining the RPV LE 110% of design pressure at the lowest elevation of the RCS.

C. Incorrect. Plausible since 125% of the design pressure is the limit for RCS components outside the RPV and RCS pressure is measured at the reactor steam dome. However, the stem asks for the RPV limit (110%) which is based on pressure at the lowest elevation of the RCS.

D. Incorrect. Plausible since the pressure limit is based on pressure at the lowest elevation of the RCS. Plausibility is enhanced since 125% of the design pressure is the limit for RCS components outside the RPV. However, the limit for the RPV is 110% of design pressure.

Technical Reference(s)

Technical Specification Bases Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13427 - Describe the bases for the Reactor Steam Dome Pressure Safety Limit.

[TS Bases] (SRO-only).

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must know basis for RPV pressure safety limit.

10 CFR Part 55 Content: 55.43 2

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96 Comments /

Reference:

Technical Specification Bases Rev: Major: 114 Minor: N/A

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Examination Outline Cross-reference: 97 Revision: 1 Date: 1/6/21 Tier: 3 Group: N/A K/A Number: 2.2.40 Level of Difficulty: 2 SRO Importance Rating: 4.7 K/A

Description:

Ability to apply Technical Specifications for a system.

Plant conditions are follows:

• All Intermediate Range instruments are on Range 1.

Using the reference provided, what actions are required by Technical Specifications?

Page 1 of 11

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 A. Suspend control rod withdrawals immediately.

B. Suspend CORE ALTERATIONS except control rod insertions immediately.

C. Fully insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. Restore required SRMs to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Answer: D K/A Match:

Requires ability to apply Technical Specifications to SRMs.

Page 2 of 11

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 SRO Only:

Page 3 of 11

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Explanation:

A. Incorrect. Plausible if it is believed that more than one required SRM is inoperable. However, since TS table 3.3.1.2-1 requires only 3 SRM channels to be operable for the conditions given in the stem, only one of the inoperable SRMs is required and TS LCO 3.3.1.2, condition A applies.

B. Incorrect. Plausible if it is believed that the reactor is in Mode 5. However, with the Reactor Mode Switch in START/HOT-STBY, the reactor is in Mode 2 and LCO 3.3.1.2, condition A applies.

C. Incorrect. Plausible since this would be the correct action if the reactor were in Mode 3 or 4.

However, with the Reactor Mode Switch in START/HOT-STBY, the reactor is in Mode 2 and LCO 3.3.1.2, condition A applies.

D. Correct. In accordance with LCO 3.3.1.2, the reactor is in Mode 2 when the Reactor Mode Switch is in START/HOT-STBY. Table 3.3.1.2-1 states that 3 SRMs are required to be operable for the conditions given. Both SRM-B and and SRM-D are inoperable, which leaves two operable channels. Therefore LCO 3.3.1.2, condition A applies which requires that one of the two inoperable SRM channel be returned to operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Technical Reference(s)

Technical Specifications Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: LCO 3.3.1.2 actions and table 3.3.1.2-1 Learning Objective: 5944 - Referencing Columbia Generation Station Technical Specifications (section 3 only for initial license candidates) associated with the Source Range Monitoring System and a set of plant conditions, determine as applicable the LSSS, the LCO, the action statement, and the appropriate bases.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: None Question Cognitive Level:

Justification for Cognitive Level Requires candidate to synthesize information given in the question stem with a knowledge of reactor modes and an understanding of TS application for SRMs.

10 CFR Part 55 Content: 55.43 2 Page 4 of 11

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Comments /

Reference:

Technical Specifications Rev: Major: 254 Minor: N/A Page 5 of 11

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Proposed Reference Page 8 of 11

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Examination Outline Cross-reference: 98 Revision: 0 Date: 8/26/20 Tier: 3 Group: N/A K/A Number: 2.3.6 Level of Difficulty: 2 SRO Importance Rating: 3.8 K/A

Description:

Ability to approve release permits CGS is in Mode 1.

Circulating Water (CW) Blowdown is required to be initiated.

Who is required to approve initiating CW Blowdown?

CW Blowdown initiation must be approved by the CRS/Shift Manager A. ONLY.

B. and Chemistry Manager.

C. and Operations Manager.

D. and Chemistry and Radiological Safety Manager.

Answer: A K/A Match:

Requires knowledge of requirements to approve a release permit.

SRO Only:

From NUREG 1021, ES-401:

Explanation:

A. Correct. In accordance with PPM 12.2.9, Circulating and Plant Service Water Halogenation Surveillance, step 8.5.6, the CRS/Shift Manager approves the CW Blowdown.

Page 1 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 B. Incorrect. Plausible since PPM 16.10.1, Radioactive Liquid Waste Discharge to the River, states that Chemistry Supervision must approve a liquid discharge if projected doses are greater than the limit. Additionally, PPM 12.2.9 requires informing Chemistry Management of certain conditions. However, CW Blowdown is not considered a radioactive release and Chemistry Management is not required to approve the blowdown.

C. Incorrect. Plausible since PPM 16.10.1 states that Liquid discharges to the river are to be avoided if at all possible, even if technically allowable. This suggests that release authorization from senior management is required. However, CW Blowdown is not considered a radioactive release and the Operations Manager is not required to approve the blowdown.

D. Incorrect. Plausible since PPM 16.10.1 states that the Chemistry and Radiological Safety Manager must approve a liquid discharge if the Effluent Concentration is > 3.0 Effluent Concentration Limit (ECL). However, CW Blowdown is not considered a radioactive release and the Chemistry and Radiological Safety Manager is not required to approve the blowdown.

Technical Reference(s)

PPM 12.2.9, Circulating and Plant Service Water Halogenation Surveillance Attached w/ Revision #

PPM 16.10.1, Radioactive Liquid Waste Discharge to the River See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 11260 - Knowledge of the requirements for reviewing and approving release permits.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of individual required to authorize a CW Blowdown release permit.

10 CFR Part 55 Content: 55.43 4 Page 2 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Comments /

Reference:

PPM 12.2.9 Rev: Major: 044 Minor: 002 Page 3 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Page 4 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Page 5 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Comments /

Reference:

PPM 12.2.9 Rev: Major: 008 Minor: N/A Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-99 Examination Outline Cross-reference: 99 Revision: 0 Date: 10/20/20 Tier: 3 Group: N/A K/A Number: 2.4.18 Level of Difficulty: 2 SRO Importance Rating: 4.0 K/A

Description:

Knowledge of the specific bases of the EOPs.

Which of the following is an assumption used to determine Cold Shutdown Boron Weight (CSBW)?

A. Reactor water is at 212°F.

B. All control rods are full out.

C. 50% power equilibrium Xenon in the core.

D. RWCU with filter demineralizers is in service.

Answer: B K/A Match:

Requires knowledge of the bases for determining CSBW which is used in PPM 5.1.2, RPV Control -

ATWS.

Page 1 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-99 SRO Only:

Page 2 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-99 Explanation:

A. Incorrect. Plausible since an assumption for reactor water temperature is used when determining CSBW. However, reactor water temperature is assumed to be at its most reactive temperature (68°F).

B. Correct. In accordance with PPM 5.0.10, Flowchart Training Manual, section 7.1, CSBW is determined assuming all control rods are full out.

C. Incorrect. Plausible since an assumption for Xenon in the core is used when determining CSBW.

However, it is assumed that no Xenon is in the core.

D. Incorrect. Plausible since an assumption for RWCU operation is used when determining CSBW.

However, filter demineralizers are assumed to be bypassed.

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual Tech Ref 2 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 9822 - Using knowledge of the emergency operating procedure bases, discuss the need to predict plant response to recommended actions.

Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of assumptions for determining CSBW.

10 CFR Part 55 Content: 55.43 5 Page 3 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-99 Comments /

Reference:

PPM 5.0.10 Rev: Major: 0232 Minor: N/A Page 4 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 Examination Outline Cross-reference: 100 Revision: 1 Date: 11/17/20 Tier: 3 Group: N/A K/A Number: 2.4.29 Level of Difficulty: 2 SRO Importance Rating: 4.4 K/A

Description:

Knowledge of the emergency plan.

CGS is in Mode 1.

An event occurs that requires entry into the emergency plan and evaluation of Emergency Action Levels (EALs).

The following conditions exist:

• RPV Water Level: -165 inches, down slow. Level cannot be raised.

• Drywell and Wetwell pressures are equal at 18 psig, up slow.

• No drywell or wetwell spray in progress.

• RHR-P-2B room temperature: 145°F, up slow.

Using the reference provided and considering the condition of Fission Product Barriers only, which EAL should be declared?

A. Unusual Event (UE)

B. Alert (A)

C. Site Area Emergency (SAE)

D. General Emergency (GE)

Answer: D K/A Match:

Requires knowledge of the emergency plan with respect to fission product barrier degradation.

SRO Only:

At CGS, EAL event classification is the exclusive responsibility of SRO-licensed individuals. See learning objective 6131 - With the procedures available for reference and plant conditions such that an emergency classification should be declared, correctly classify the event. (SRO only) [PPM 13.1.1]

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 From NUREG-1021, revision 11:

Explanation:

A. Incorrect. Plausible if it is believed that the loss of only one fission product barrier (FCB) is an unusual event. However, for the conditions given, a Site Area Emergency should be declared.

B. Incorrect. Plausible since a loss or potential loss of the fuel clad or RCS constitutes an Alert.

However, for the conditions given, a Site Area Emergency should be declared.

C. Incorrect. Plausible if it is not recognized that with drywell and wetwell pressures equal with no sprays is a PC pressure response not consistent with LOCA, and indicative of a failed drywell floor. The requirement to classify the event as a Site Area Emergency is the Loss or Potential Loss of any two barriers. Conditions given in the stem indicate a loss of two barriers and potential loss of the third barrier, which meets the criteria for GE.

D. Correct. The requirements for a General Emergency require the loss of two barriers and the loss or potential loss of the third barrier. Based on RPV level, there is a loss of the RCS barrier and potential loss of the PC barrier. Drywell and wetwell pressures equal with no sprays is a PC pressure response not consistent with LOCA, and indicative of a failed drywell floor.Therefor, there is a loss of two barriers and potential loss of the third barrier, which meets the criteria o of declaring a GE.

Technical Reference(s)

PPM 13.1.1, Classifying the Emergency PPM 13.1.1A, Classifying the Emergency - Technical Basis Attached w/ Revision #

See Comments / Reference PPM 5.3.1, Secondary Containment Control Tech Ref 4 Proposed references to be provided during examination: PPM 13.1.1 (complete), RB Area Temp data sheet.

Learning Objective: 6131 - With the procedures available for reference and plant conditions such that an emergency classification should be declared, correctly classify the event. (SRO only) [PPM 13.1.1]

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 Question Source: #: Bank #

1. Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires interpreting conditions given in the question stem against EAL requirements and a knowledge of fission barrier conditions that constitute different EAL classifications.

10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

PPM 13.1.1 Rev: Major: 049 Minor: 001

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 Comments /

Reference:

PPM 13.1.1A Rev: Major: 034 Minor: 001

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: N/A

ILC-24 NRC Validation Exam Question SRO-79/100 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Prolonged loss of all offsite and all onsite AC power to emergency Loss of all offsite and all onsite AC power to emergency buses Loss of all but one AC power source to emergency buses Loss of all offsite AC power capability to emergency buses buses for 15 minutes or longer for 15 minutes or longer for 15 minutes or longer MG1.1 1 2 3 MS1.1 1 2 3 MA1.1 1 2 3 MU1.1 1 2 3 Loss of all offsite AND all onsite AC power capability to Loss of all offsite and all onsite AC power capability to AC power capability, Table 2, to emergency buses SM-7 Loss of all offsite AC power capability, Table 2, to emergency emergency buses SM-7 and SM-8 emergency buses SM-7 and SM-8 for GE 15 min. (Note 1) and SM-8 reduced to a single power source for GE 15 min. buses SM-7 and SM-8 for GE 15 min. (Note 1)

AND EITHER: (Note 1) 1 Restoration of emergency bus SM-7 or SM-8 in LT 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)

OR AND Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS Table 2 AC Power Sources Loss of Offsite Emergency RPV level cannot be restored and maintained GT -186 in. Startup Transformer TR-S AC Power Loss of all emergency AC and vital DC power sources for 15 Backup Transformer TR-B minutes or longer Backfeed 500 KV power through Main Transformers (if already MG1.2 1 2 3 aligned in modes 4, 5, def only)

Loss of all offsite AND all onsite AC power capability to Onsite emergency buses SM-7 and SM-8 for GE 15 min. (Note 1) DG1 AND DG2 2 Indicated voltage is LT 108 VDC on both 125 VDC buses Loss of all vital DC power for 15 minutes or longer Main Generator via TR-N1/N2 DP-S1-1 and DP-S1-2 for GE 15 min. (Note 1)

MS2.1 1 2 3 Loss of None None Indicated voltage is LT 108 VDC on both 125 VDC buses Vital DC DP-S1-1 and DP-S1-2 for GE 15 min. (Note 1)

Power UNPLANNED loss of Control Room indications for 15 minutes or UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress longer MA3.1 1 2 3 MU3.1 1 2 3 3 An UNPLANNED event results in the inability to monitor one or more Table 10 parameters from within the Control Room An UNPLANNED event results in the inability to monitor one or more Table 10 parameters from within the Control Loss of None None for GE 15 min. (Note 1) Room for GE 15 min. (Note 1)

Control AND Room Any Table 11 transient event in progress Indications Table 10 Safety System Parameters Reactor coolant activity greater than Technical Specification Reactor power allowable limits Table 5 Plant Structures Containing Safe Shutdown Systems or RPV level Components RPV pressure MU4.1 1 2 3 4

Primary containment pressure SJAE CONDSR OUTLET RAD HI-HI alarm (P602)

Vital portions of the Rad Waste/Control Building:

Wetwell level None None - 467' elevation vital island RCS Wetwell temperature Activity - 487' elevation cable spreading room MU4.2 1 2 3

- Main Control Room and vertical cable chase

- 525' elevation HVAC area Table 11 Transient Events Coolant activity GT 0.2 Ci/gm dose equivalent I-131 Reactor Building Vital portions of the Turbine Building Reactor scram M -

DEH pressure switches RPS switches on turbine throttle valves Runback GT 25% thermal reactor power RCS leakage for 15 minutes or longer System Electrical load rejection GT 25% full MU5.1 1 2 3

- Main steam line radiation monitors 5

Malfunct. electrical load (1) RCS unidentified or pressure boundary leakage

- Turbine Building ventilation radiation monitors ECCS injection GE 10 gpm for GE 15 min.

None - Main steam line piping up to MS-V-146 Noneand the first stop valves Thermal powerNone oscillations GT 10% OR RCS Leakage Standby Service Water Pump Houses (2) RCS identified leakage GT 25 gpm for GE 15 min.

Diesel Generator Building OR (3) Leakage from the RCS to a location outside containment GT 25 gpm for GE 15 min.

Inability to shut down the reactor causing a challenge to RPV Automatic or manual scram fails to shut down the reactor, and Automatic or manual scram fails to shut down the reactor water level or RCS heat removal subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor MS6.1 1 2 MA6.1 1 2 MU6.1 1 2 An automatic OR manual scram fails to shut down the An automatic OR manual scram fails to shut down the An automatic OR manual scram did not shut down the reactor reactor reactor 6 None AND All actions to shut down the reactor are not successful as AND Manual scram actions taken at the reactor control console AND A subsequent automatic scram OR manual scram action RPS indicated by reactor power GT 5% (mode switch in shutdown, manual push buttons or ARI) are taken at the reactor control console (mode switch in Failure AND EITHER: not successful in shutting down the reactor as indicated by shutdown, manual push buttons or ARI) is successful in RPV level cannot be restored and maintained reactor power GT 5% (Note 8) shutting down the reactor as indicated by reactor power LE Table 4 Communication Methods above -186 in. or cannot be determined 5% (APRM downscale) (Note 8)

OR System Onsite ORO NRC WW temperature and RPV pressure cannot be maintained below the HCTL Plant Public Address (PA) System X Loss of all onsite or offsite communications capabilities Plant Telephone System X X MU7.1 1 2 3 7 Plant Radio System Operations and Security Channels X

None None (1) Loss of all Table 4 onsite communication methods OR Loss of (2) Loss of all Table 4 ORO communication methods Comm. Offsite calling capability from the X X OR Control Room via direct telephone (3) Loss of all Table 4 NRC communication methods Long distance calling capability on X X None the commercial phone system Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode MA8.1 1 2 3 The occurrence of any Table 8 hazardous event Table 8 Hazardous Events AND Event damage has caused indications of degraded 8 None Seismic event Internal or external FLOODING event performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

None Hazardous Event High winds Event damage has caused indications of degraded Affecting Tornado strike performance to a second train of a SAFETY SYSTEM Safety FIRE needed for the current operating mode Systems OR EXPLOSION Event damage has resulted in VISIBLE DAMAGE to a Volcanic ash fallout second train of a SAFETY SYSTEM needed for the Other events with similar hazard current operating mode characteristics as determined by the Shift (Notes 9, 10)

Manager F FG1.1 1 2 Loss of any two barriers 3 FS1.1 1 2 3 Loss or potential loss of any two barriers (Table F-1)

FA1.1 1 2 3 Any loss or any potential loss of EITHER Fuel Clad or RCS None Fission Product AND barrier (Table F-1)

Barrier Degradation Loss or potential loss of the third barrier (Table F-1)

Table F-1 Fission Product Barrier Threshold Matrix FC - Fuel Clad Barrier RCS - Reactor Coolant System Barrier PC - Containment Barrier Loss Potential Loss Loss Potential Loss Loss Potential Loss RPV level cannot be restored and RPV level cannot be restored and A SAG entry required maintained GT -161 in. maintained GT -161 in. None None SAG entry required RPV Water Level or cannot be determined. or cannot be determined.

UNISOLABLE break in any of the UNISOLABLE primary system leakage UNISOLABLE primary system leakage following:

that results in exceeding EITHER: that results in exceeding EITHER:

Main Steam Line RCIC Steam Line RB area temperature alarm level (PPM RB area maximum safe operating B None None RWCU 5.3.1 Table 23) temperature (PPM 5.3.1 Table 23) None RCS Leak Rate Feedwater OR OR OR RB area radiation alarm level (PPM RB area maximum safe operating Emergency RPV Depressurization is 5.3.1 Table 24) radiation (PPM 5.3.1 Table 24) required PC pressure GT 45 psig OR UNPLANNED rapid drop in PC pressure following PC pressure rise Explosive mixture exists inside PC C None None PC pressure GT 1.68 psig due to RCS None OR (H2 GE 6% and O2 GE 5%)

PC Conditions leakage OR PC pressure response not consistent with LOCA conditions WW temperature and RPV pressure cannot be maintained below the HCTL Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F D reading GT 3,600 R/hr Containment Radiation Monitor Containment Radiation Monitor None CMS-RIS-27E or CMS-RIS-27F None None CMS-RIS-27E or CMS-RIS-27F PC Rad / OR reading GT 70 R/hr reading GT 14,000 R/hr RCS Activity Primary coolant activity GT 300 µCi/gm Dose Equivalent I-131 UNISOLABLE direct downstream E pathway to the environment exists after PC Integrity or None PC isolation signal None None None None Bypass OR Intentional PC venting per EOPs F Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the Emergency Emergency Director that indicates Emergency Director that indicates loss Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates loss Emergency Director that indicates loss Director potential loss of the Fuel Clad barrier of the RCS barrier potential loss of the RCS barrier potential loss of the Containment of the fuel clad barrier of the Containment barrier Judgment barrier 13.1.1 Rev. 49 MR 1 CLASSIFYING THE EMERGENCY Modes: 1 2 3 HOT CONDITIONS 1/16/2019 Power Operations Startup Hot Shutdown (RCS GT 200°F) 170028

ILC-24 NRC Validation Exam Question SRO-79/100 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Loss of RPV inventory affecting fuel clad integrity with Loss of RPV inventory affecting core decay heat removal Significant loss of RPV inventory Unplanned loss of RPV inventory containment challenged capability CG1.1 4 5 CS1.1 4 5 CA1.1 4 5 CU1.1 4 5 RPV level LT -161 in. for GE 30 min. (Note 1) (1) CONTAINMENT CLOSURE not established (1) Loss of RPV inventory as indicated by RPV level (1) UNPLANNED loss of reactor coolant results in RPV level AND AND LT -50 in. less than a required lower limit for GE 15 min. (Note 1)

OR OR Any of the following indications of containment challenge: RPV level LT -129 in.

(2) RPV level cannot be monitored for GE 15 min. (Note 1) (2) RPV level cannot be monitored CONTAINMENT CLOSURE not established (Note 6) OR AND AND Explosive mixture inside PC (2) CONTAINMENT CLOSURE established UNPLANNED increase in any Table 1 sump or pool levels (H2 GE 6% and O2 GE 5%) UNPLANNED increase in any Table 1 sump or pool AND due to a loss of RPV inventory UNPLANNED rise in PC pressure levels due to a loss of RPV inventory RPV level LT -161 in.

RB area radiation GT any Maximum Safe Operating level (PPM 5.3.1 Table 24)

CS1.2 4 5 CG1.2 4 5 1

RPV level cannot be monitored for GE 30 min. (Note 1)

RPV level cannot be monitored for GE 30 min. (Note 1)

AND AND RPV Core uncovery is indicated by any of the following: Table 1 Sumps/Pool Level Core uncovery is indicated by any of the following:

UNPLANNED wetwell level rise GT 2 inches UNPLANNED wetwell level rise GT 2 inches (PPM 5.2.1 entry condition) Any valid Hi-Hi level alarm on R-1 Table 2 AC Power Sources (PPM 5.2.1 entry condition) VALID indication of RB room flooding as identified by through R-5 sumps VALID indication of RB room flooding as identified by high level alarms (PPM 5.3.1 Table 25) EDR GE 25 GPM high level alarms (PPM 5.3.1 Table 25) Offsite Observation of UNISOLABLE RCS leakage outside FDR GE 10 GPM Observation of UNISOLABLE RCS leakage outside Startup Transformer TR-S primary containment of sufficient magnitude to indicate primary containment of sufficient magnitude to indicate core uncovery Wetwell level rise Backup Transformer TR-B core uncovery Observation of UNISOLABLE RCS Backfeed 500 KV power through Main AND leakage Transformers (if already aligned in Any of the following indications of containment challenge: modes 4, 5, def only)

CONTAINMENT CLOSURE not established (Note 6)

Explosive mixture inside PC Onsite (H2 GE 6% and O2 GE 5%) DG1 UNPLANNED rise in PC pressure DG2 RB area radiation GT any Maximum Safe Operating Main Generator via TR-N1/N2 level (PPM 5.3.1 Table 24)

Loss of all offsite and all onsite AC power to emergency buses Loss of all but one AC power source to emergency buses for 15 for 15 minutes or longer minutes or longer 2 None None CA2.1 4 5 Loss of all offsite and all onsite AC power capability to DEF CU2.1 4 5 AC power capability, Table 2, to emergency buses SM-7 DEF Loss of emergency buses SM-7 and SM-8 for GE 15 min. (Note 1) and SM-8 reduced to a single power source for GE 15 min.

C Emergency AC Power (Note 1)

AND Any additional single power source failure will result in a loss Cold SD/

of all AC power to SAFETY SYSTEMS Refuel System Malfunct. Inability to maintain plant in cold shutdown UNPLANNED increase in RCS temperature Table 7 RCS Reheat Duration Thresholds 4 5 4 5 3

CA3.1 CU3.1 UNPLANNED increase in RCS temperature to GT 200°F UNPLANNED increase in RCS temperature to GT 200°F

• If an RCS heat removal system is in operation within this None for GT Table 7 duration (Note 1)

RCS time frame and RCS temperature is being reduced the EAL is not applicable OR Temp.

UNPLANNED RPV pressure increase GT 10 psig CU3.2 4 5 Containment Heat-up RCS Status Closure Status Duration Loss of all RCS temperature and RPV water level indication for GE 15 min. (Note 1)

Intact N/A 60 min.

• 4 None Not intact established 20 min.

• None CU4.1 Loss of vital DC power for 15 minutes or longer 4 5 Loss of not established 0 min.

Vital DC Indicated voltage LT 108 VDC on required 125 VDC buses Power DP-S1-1 and DP-S1-2 for GE 15 min. (Note 1)

Loss of all onsite or offsite communications capabilities Table 4 Communication Methods CU5.1 4 5 DEF 5 System None Onsite ORO NRC None None Loss of all Table 4 onsite communication methods OR Loss of Comm. Plant Public Address (PA) System X Loss of all Table 4 ORO communication methods OR Plant Telephone System X X Loss of all Table 4 NRC communication methods Hazardous event affecting a SAFETY SYSTEM needed for the Plant Radio System Operations and X current operating mode Security Channels CA6.1 4 5 Offsite calling capability from the X X The occurrence of any Table 8 hazardous event Table 8 Hazardous Events Control Room via direct telephone AND 6

None None Event damage has caused indications of degraded Long distance calling capability on X X Seismic event performance on one train of a SAFETY SYSTEM needed for Hazardous the commercial phone system Internal or external FLOODING event the current operating mode Events High winds AND EITHER:

Affecting Event damage has caused indications of degraded Safety Tornado strike performance to a second train of a SAFETY SYSTEM Systems needed for the current operating mode FIRE EXPLOSION OR Event damage has resulted in VISIBLE DAMAGE to a Volcanic ash fallout second train of a SAFETY SYSTEM needed for the Other events with similar hazard current operating mode characteristics as determined by the Shift (Notes 9, 10)

Manager 13.1.1 Rev. 49 MR 1 CLASSIFYING THE EMERGENCY Modes: 4 5 DEF COLD CONDITIONS 1/16/2019 Cold Shutdown Refueling Defueled (RCS 200°F) 170028

ILC-24 NRC Validation Exam Question SRO-79/100 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite Release of gaseous or liquid radioactivity greater than 2 times than 1,000 mrem TEDE or 5,000 mrem thyroid CDE than 100 mrem TEDE or 500 mrem thyroid CDE dose greater than 10 mrem TEDE or 50 mrem thyroid CDE the ODCM limits for 60 minutes or longer RG1.1 1 2 3 4 5 DEF RS1.1 1 2 3 4 5 DEF RA1.1 1 2 3 4 5 DEF RU1.1 1 2 3 4 5 DEF (1) Reading on any Table 3 effluent radiation monitor (1) Reading on any Table 3 effluent radiation monitor (1) Reading on any Table 3 effluent radiation monitor (1) Reading on any Table 3 effluent radiation monitor GT column "GENERAL" for GE 15 min. GT column "SAE" for GE 15 min. GT column "ALERT" for GE 15 min. GT column "UE" for GE 60 min.

OR OR OR OR (2) Dose assessment using actual meteorology indicates (2) Dose assessment using actual meteorology (2) Dose assessment using actual meteorology indicates (2) Sample analyses for a gaseous or liquid release doses GT 1,000 mrem TEDE or GT 5000 mrem thyroid indicates doses GT 100 mrem TEDE or GT 500 doses GT 10 mrem TEDE or GT 50 mrem thyroid CDE indicates a concentration or release rate 2 x ODCM CDE at or beyond the SITE BOUNDARY mrem thyroid CDE at or beyond the SITE at or beyond the SITE BOUNDARY limits for GE 60 min.

(Notes 1, 2, 3, 4) BOUNDARY (Notes 1, 2, 3, 4) (Notes 1, 2, 3) 1 (Notes 1, 2, 3, 4)

RA1.2 1 2 3 4 5 DEF RG1.2 1 2 3 4 5 DEF RS1.2 1 2 3 4 5 DEF Analysis of a liquid effluent sample indicates a concentration Rad Field survey results indicate EITHER of the following at or Field survey results indicate EITHER of the following at or or release rate that would result in doses GT 10 mrem TEDE Effluent beyond the SITE BOUNDARY: beyond the SITE BOUNDARY: or GT 50 mrem thyroid CDE at or beyond the SITE Closed window dose rates GT 100 mR/hr expected BOUNDARY for 60 min. of exposure (Notes 1, 2)

Closed window dose rates GT 1,000 mR/hr expected to continue for GE 60 min. to continue for GE 60 min.

Analyses of field survey samples indicate thyroid RA1.3 1 2 3 4 5 DEF Analyses of field survey samples indicate thyroid CDE GT 5,000 mrem for 60 min. of inhalation. CDE GT 500 mrem for 60 min. of inhalation.

Field survey results indicate EITHER of the following at or (Notes 1, 2) (Notes 1, 2) beyond the SITE BOUNDARY:

Closed window dose rates GT 10 mR/hr expected to continue for GE 60 min.

Analyses of field survey samples indicate thyroid CDE GT 50 mrem for 60 min. of inhalation.

R (Notes 1, 2)

Abnormal Spent fuel pool level cannot be restored to at least the top of the Spent fuel pool level at the top of the fuel racks Significant lowering of water level above, or damage to, Unplanned loss of water level above irradiated fuel Rad fuel racks for 60 minutes or longer irradiated fuel Levels RG2.1 1 2 3 4 5 DEF RS2.1 1 2 3 4 5 DEF RA2.1 1 2 3 4 5 DEF RU2.1 1 2 3 4 5 DEF

/

Rad Spent fuel pool level cannot be restored to at least 0.5 ft Lowering of spent fuel pool level to 0.5 ft Uncovery of irradiated fuel in the REFUELING PATHWAY UNPLANNED water level drop in the REFUELING PATHWAY Effluent for GE 60 min. (Note 1) as indicated by EITHER of the following:

RA2.2 1 2 3 4 5 DEF 2

SFP level LE 22.3 ft Damage to irradiated fuel resulting in a release of SFP low level alarm Table 3 Effluent Monitor Classification Thresholds radioactivity AND Irradiated AND UNPLANNED rise in area radiation levels as indicated by any Fuel Event Release Point Monitor General SAE Alert UE High alarm on any of the following radiation monitors: of the following radiation monitors:

ARM-RIS-1 Reactor Building Fuel Pool Area ARM-RIS-1 Reactor Building Fuel Pool Area PRM-RE-11 3.05E-03 µCi/cc ARM-RIS-2 Reactor Building Fuel Pool Area ARM-RIS-2 Reactor Building Fuel Pool Area Reactor Building Exhaust PRM-RE-12 ---- ---- 2.82E+1 µCi/cc ---- ARM-RIS-34 Reactor Building Elevation 606 ARM-RIS-34 Reactor Building Elevation 606 Gaseous PRM-RE-13 7.50E+02 µCi/cc 7.50E+1 µCi/cc ---- ---- REA-RIS-609A-D Rx Bldg Vent Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc RA2.3 1 2 3 4 5 DEF Radwaste Building Exhaust WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Lowering of spent fuel pool level to 10 ft Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown 3 Liquid TSW Effluent TSW-RIS-5 ---- ---- ---- 3.00E-05 µCi/cc RA3.1 1 2 3 4 5 DEF Service Water Process A SW-RIS-604 ---- ---- ---- 1.00E+02 cps Area SW-RIS-605 1.00E+02 cps (1) Dose rates GT 15 mR/hr in Control Room Service Water Process B ---- ---- ----

Radiation (ARM-RIS-19) or CAS (by survey)

Levels OR (2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 9 Table 9 Safe Operation & Shutdown Rooms/Areas rooms or areas (Note 5)

Damage to a loaded cask CONFINEMENT BOUNDARY Room/Area Modes Applicability EU1.1 Storage Operations RW 467 Radwaste Control Room (RHR flush to RW tanks) 3 1

Damage to a loaded canister (MPC) CONFINEMENT E Confinement RW 467 Vital Island (RHR-V-9 disconnect)

RB 422 B RHR Pump Rm (local pump temperatures) 3 3

BOUNDARY as indicated by measured dose rates on a loaded overpack GT EITHER:

ISFSI None RB 454 B RHR Pump Rm (operate RHR-V-85B) None 3 None 20 mrem/hr (gamma + neutron) on the top of the Boundary overpack 100 mrem/hr (gamma + neutron) on the side of the overpack, excluding inlet and outlet ducts HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION within the OWNER CONTROLLED AREA or Confirmed SECURITY CONDITION or threat airborne attack threat within 30 minutes HS1.1 1 2 3 4 5 DEF HA1.1 1 2 3 4 5 DEF HU1.1 1 2 3 4 5 DEF A HOSTILE ACTION is occurring or has occurred within (1) A HOSTILE ACTION is occurring or has occurred (1) A SECURITY CONDITION that does not involve a the PROTECTED AREA as reported by the Security within the OWNER CONTROLLED AREA as reported HOSTILE ACTION as reported by the Security 1 None Sergeant or Security Lieutenant by the Security Sergeant or Security Lieutenant OR Sergeant or Security Lieutenant OR Security (2) A validated notification from NRC of an aircraft attack (2) Notification of a credible security threat directed at the threat within 30 min. of the site site OR (3) A validated notification from the NRC providing information of an aircraft threat Seismic event GT OBE levels 2 None None See CA6.1/MA8.1 for potential for upgrade to an AlertNone based on degraded HU2.1 1 2 3 4 5 DEF Seismic safety system performance or damage Seismic event GT Operating Basis Earthquake (OBE) as Event indicated by H13.P851.S1.5-1 (OPERATING BASIS EARTHQUAKE EXCEEDED) activated Notes Hazardous event HU3.1 1 2 3 4 5 DEF 1 The Emergency Director should declare the event promptly upon determining that time limit has been (1) A tornado strike within the PROTECTED AREA exceeded, or will likely be exceeded OR 2 If an ongoing release is detected and the release start (2) Volcanic ash fallout requiring plant shutdown time is unknown, assume that the release duration has exceeded the specified time limit HU3.2 1 2 3 4 5 DEF 3 If the effluent flow past an effluent monitor is known to 3 have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for None See CA6.1/MA8.1 for potential for upgrade to an AlertNone based on degraded safety system performance or damage Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY Natural or classification purposes SYSTEM component needed for the current operating mode Tech. 4 The pre-calculated effluent monitor values presented in Hazard EALs RA1.1, RS1.1 and RG1.1 should be used for HU3.3 1 2 3 4 5 DEF emergency classification assessments until the results (1) Movement of personnel within the PROTECTED AREA is from a dose assessment using actual meteorology are IMPEDED due to an offsite event involving hazardous available materials (e.g., an offsite chemical spill, 618-11 event or 5 If the equipment in the listed room or area was already toxic gas release) inoperable or out-of-service before the event occurred, OR then no emergency classification is warranted (2) A hazardous event that results in on-site conditions 6 If CONTAINMENT CLOSURE is re-established prior to sufficient to prohibit the plant staff from accessing the site exceeding the 30-minute time limit, declaration of a via personal vehicles (Note 7)

General Emergency is not required 7 This EAL does not apply to routine traffic impediments FIRE potentially degrading the level of safety of the plant such as fog, snow, ice, or vehicle breakdowns or Table 5 accidents Plant Structures Containing Safe Shutdown Systems or Components HU4.1 1 2 3 4 5 DEF 8 A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly A FIRE is not extinguished within 15 min. of any of the inserted into the core, and does not include manually Vital portions of the Rad Waste/Control Building: following FIRE detection indications (Note 1):

driving in control rods or implementation of boron injection Report from the field (i.e., visual observation)

- 467' elevation vital island Receipt of multiple (more than 1) fire alarms or strategies H

- 487' elevation cable spreading room indications 9 If the affected SAFETY SYSTEM train was already Field verification of a single fire alarm

- Main Control Room and vertical cable chase inoperable or out of service before the hazardous event AND occurred, then emergency classification is not warranted - 525' elevation HVAC area The FIRE is located within any Table 5 area Hazards 10 If the hazardous event only resulted in VISIBLE Reactor Building DAMAGE, with no indications of degraded performance Vital portions of the Turbine Building HU4.2 1 2 3 4 5 DEF 4 to at least one train of a SAFETY SYSTEM, then this emergency classificationNoneis not warranted None -

DEH pressure switches RPS switches on turbine throttle valves Receipt of a single fire alarm (i.e., no other indications of a FIRE)

Fire AND

- Main steam line radiation monitors The fire alarm is indicating a FIRE within any Table 5 area

- Turbine Building ventilation radiation monitors AND The existence of a FIRE is not verified within 30 min. of alarm

- Main steam line piping up to MS-V-146 and the first stop valves receipt (Note 1)

Standby Service Water Pump Houses Diesel Generator Building HU4.3 1 2 3 4 5 DEF (1) A FIRE within the ISFSI or plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

See CA6.1/MA8.1 for potential for OR upgrade to an Alert based on degraded (2) A FIRE within the ISFSI or plant PROTECTED AREA that safety system performance or damage requires firefighting support by an offsite fire response Table 9 Safe Operation & Shutdown Rooms/Areas agency to extinguish Room/Area Modes Applicability Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown 5

RW 467 Radwaste Control Room (RHR flush to RW tanks) 3 RW 467 Vital Island (RHR-V-9 disconnect) 3 HA5.1 1 2 3 4 5 DEF None None Hazardous RB 422 B RHR Pump Rm (local pump temperatures) 3 Release of a toxic, corrosive, asphyxiant or flammable gas Gases into any Table 9 rooms or areas RB 454 B RHR Pump Rm (operate RHR-V-85B) 3 AND Entry into the room or area is prohibited or IMPEDED (Note 5)

Inability to control a key safety function from outside the Control Control Room evacuation resulting in transfer of plant control to Room alternate locations HS6.1 1 2 3 4 5 HA6.1 1 2 3 4 5 DEF 6 None An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel or Alternate Remote Shutdown Panel An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel or Alternate Remote Shutdown Panel None Control AND Room Control of any of the following key safety functions is not Evacuation reestablished within 15 min. (Note 1):

Reactivity (Modes 1 and 2 only)

RPV water level RCS heat removal Other conditions existing which in the judgment of the Emergency Other conditions existing which in the judgment of the Emergency Other conditions existing which in the judgment of the Other conditions existing which in the judgment of the Director warrant declaration of General Emergency Director warrant declaration of Site Area Emergency Emergency Director warrant declaration of an Alert Emergency Director warrant declaration of a UE HG7.1 1 2 3 4 5 DEF HS7.1 1 2 3 4 5 DEF HA7.1 1 2 3 4 5 DEF HU7.1 1 2 3 4 5 DEF Other conditions exist which, in the judgment of the Other conditions exist which, in the judgment of the Other conditions exist which, in the judgment of the Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or Emergency Director, indicate that events are in progress or Emergency Director, indicate that events are in progress or Emergency Director, indicate that events are in progress or 7 have occurred which involve actual or IMMINENT have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the substantial core degradation or melting with potential for plant functions needed for protection of the public or substantial degradation of the level of safety of the plant or level of safety of the plant or indicate a security threat to loss of containment integrity or HOSTILE ACTION that HOSTILE ACTION that results in intentional damage or a security event that involves probable life threatening risk facility protection has been initiated. No releases of Judgment results in an actual loss of physical control of the facility. malicious acts, (1) toward site personnel or equipment that to site personnel or damage to site equipment because of radioactive material requiring offsite response or monitoring Releases can be reasonably expected to exceed EPA could lead to the likely failure of or, (2) that prevent effective HOSTILE ACTION. Any releases are expected to be limited are expected unless further degradation of SAFETY Protective Action Guideline exposure levels offsite for more access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline SYSTEMS occurs.

than the immediate site area. Any releases are not expected to result in exposure levels exposure levels.

which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

13.1.1 Rev. 49 MR 1 CLASSIFYING THE EMERGENCY Modes: 1 2 3 4 5 DEF 1/16/2019 Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled ALL CONDITIONS 170028

ILC-24 NRC Validation Exam Question: SRO-86 ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS

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

LCO 3.0.4.b is not applicable to High Pressure Core Spray (HPCS).

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days(1) injection/spray ECCS injection/spray subsystem inoperable. subsystem to OPERABLE status.

B High Pressure Core B.1 Verify by administrative Immediately Spray (HPCS) System means RCIC System is inoperable. OPERABLE when RCIC System is required to be OPERABLE.

AND B.2 Restore HPCS System to 14 days OPERABLE status.

(1) The Completion Time that one train of RHR (RHR-A) can be inoperable as specified by Required Action A.1 may be extended beyond the 7 day completion time up to 7 days to support restoration of RHR-A following pump and motor replacement. This footnote will expire at 23:59 PST February 28, 2019.

Columbia Generating Station 3.5.1-1 Amendment No. 187 225 230 245 251 253 1 of 15

ILC-24 NRC Validation Exam Question: SRO-86 ECCS - Operating 3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Two ECCS injection C.1 Restore ECCS 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> subsystems inoperable. injection/spray subsystem to OPERABLE status.

OR One ECCS injection and one ECCS spray subsystem inoperable.

D. Required Action and D.1 --------------NOTE---------------

associated Completion LCO 3.0.4.a is not Time of Condition A, B, applicable when entering or C not met. MODE 3.

Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> E. One required ADS valve E.1 Restore ADS valve to 14 days inoperable. OPERABLE status.

F. One required ADS valve F.1 Restore ADS valve to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

AND OR One low pressure ECCS F.2 Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> injection/spray ECCS injection/spray subsystem inoperable. subsystem to OPERABLE status.

Columbia Generating Station 3.5.1-2 Amendment No. 149,169,225,236 2 of 15

ILC-24 NRC Validation Exam Question: SRO-86 ECCS - Operating 3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 --------------NOTE--------------

associated Completion LCO 3.0.4.a is not Time of Condition E or F applicable when entering not met. MODE 3.

OR Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Two or more required ADS valves inoperable.

H. HPCS and Low H.1 Enter LCO 3.0.3. Immediately Pressure Core Spray (LPCS) Systems inoperable.

OR Three or more ECCS injection/spray subsystems inoperable.

OR HPCS System and one or more required ADS valves inoperable.

OR Two or more ECCS injection/spray subsystems and one or more required ADS valves inoperable.

Columbia Generating Station 3.5.1-3 Amendment No. 149,169,225,236 3 of 15

ILC-24 NRC Validation Exam Question: SRO-86 ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, In accordance locations susceptible to gas accumulation are with the sufficiently filled with water. Surveillance Frequency Control Program SR 3.5.1.2 ------------------------------NOTE---------------------------

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

Verify each ECCS injection/spray subsystem In accordance manual, power operated, and automatic valve in the with the flow path, that is not locked, sealed, or otherwise Surveillance secured in position, is in the correct position. Frequency Control Program SR 3.5.1.3 Verify ADS accumulator backup compressed gas In accordance system average pressure in the required bottles is with the 2200 psig. Surveillance Frequency Control Program SR 3.5.1.4 Verify each ECCS pump develops the specified flow In accordance rate with the specified differential pressure between with the reactor and suction source. INSERVICE TESTING DIFFERENTIAL PROGRAM PRESSURE BETWEEN REACTOR AND SYSTEM FLOW RATE SUCTION SOURCE LPCS 6200 gpm 128 psid LPCI 7200 gpm 26 psid HPCS 6350 gpm 200 psid Columbia Generating Station 3.5.1-4 Amendment No. 169,205,225,229,236 238 243 246 249 4 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 6 of 36 4.0 PRECAUTIONS AND LIMITATIONS 4.1 Prior to starting LPCS-P-1, assure discharge piping is pressurized by verifying H13-P601.A3-5.3, LPCS PUMP DISH PRESS HIGH/LOW alarm is not lit.

4.2 Maintain piping systems pressurized during testing by fully closing LPCS-V-12 prior to stopping pump.

4.3 Do not operate LPCS-P-1 without minimum flow requirements satisfied (770 gpm). {C-9448}

4.4 Watch for pump cavitation during the opening stroke of LPCS-V-12; if fluctuations in pump amperes, system flow or pressure occur, apply a close signal to the valve until fluctuations cease.

4.5 When it is desired to limit the testing to obtain specific test data, exercise care to ensure that the portion(s) of the procedure being performed provides the test data in the same manner as if the entire procedure were performed.

4.6 Alert Range 4.6.1 Pumps Measured test parameters beyond the Alert Value shall have the test frequency increased to once each 46 days until the cause of the deviation is determined and the condition is corrected, at which time the original test frequency may be resumed. Any abnormality or erratic action shall be evaluated per the Corrective Action Program. {R-9653}

4.6.2 Valves Valves with measured stroke times beyond the Alert Value shall be immediately retested or declared inoperable. {R-9653}

If the valve is retested and the second set of data is also beyond the Alert Value, the data shall be analyzed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to verify that the new stroke time represents acceptable valve operation, or the valve shall be declared inoperable. {R-9653}

If the second set of data is within the Acceptance Range, the cause of the initial deviation shall be analyzed. {R-9653}

Any abnormality or erratic actions shall be evaluated per the Corrective Action program. Reference the Condition Report on the procedure cover sheet. {R-9653}

5 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 7 of 36 4.7 Action Range If a valve fails to open or close, or the measured test parameters are beyond the Action Value, declare the pump or valve inoperable. Any abnormality or erratic action shall be evaluated per the Corrective Action program. {R-9653}

4.8 Stroke Timing When stroke timing a valve open or closed, start the stop watch when the control switch is turned to the open or closed position and stopped when the valve indicates full open or closed.

When stroke timing a valve open that has both lights illuminated during its intermediate position, ensure the green light is extinguished prior to stopping the stopwatch and when stroke timing the valve closed, ensure the red light is extinguished prior to stopping the stopwatch.

4.9 Valve motors may be damaged by stroking too frequently or applying repeated start signals to a stalled valve. When possible, investigate malfunctions locally prior to repeating a valve motor energization.

4.10 Valves are to be tested in the As Found condition unless specifically directed otherwise.

Clearly specify the reason for not stroke timing any valve in the As Found Condition in the Comments Section.

4.11 Notify the Shift Manager before performing any step or section of this surveillance which affects the Control Room Operator's ability to control the LPCS system.

4.12 Note all discrepancies encountered during the performance of this surveillance on the cover sheet and report to the Shift Manager.

4.13 Unless specified in this surveillance makes no internal or external adjustments to plant equipment.

4.14 Ensure all trip, alarm, and computer points are clear after a channel is returned to service.

4.15 The unavailability of LPCS when it is required to be operable is required to be recorded.

The time when LPCS is declared unavailable should be minimized. (M-Rule) 4.16 ASME Code requires performance of a Comprehensive Pump Test (CPT) every two years. The CPT will use more accurate pressure instrumentation and has more restrictive acceptance criteria.

4.17 IST acceptance criteria is more limiting than the Technical Specification SR 3.5.1.4 acceptance criteria for LPCS-P-1.

4.18 Only qualified I&C Technicians should attempt to re-zero instruments used in this surveillance.

6 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 8 of 36 4.19 Infrared temperature detectors, like the Fluke laser unit or equivalent, are the preferred device for verification of adequate seal flow. This allows a reduction in dose and gives a larger margin of safety due to distance from rotating equipment.

4.20 When using infrared temperature detectors the target area should not have a reflective surface and the laser gun should be held within 2 feet of the target to ensure accurate readings.

4.21 Per the FSAR, minimum suppression pool temperature should be GT 50 °F per LDCN-14-040 to Table 6.2-19. It is suggested to maintain suppression pool temperatures between 55 - 90°F.

7 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 26 of 36 VALVE STROKE DATA SHEET

1. OPENING TIME IN SECONDS # CLOSING TIME IN SECONDS VALVE ID Ref. Alert Lo Alert Hi Action Hi Ref. Alert Lo Alert Hi Action Hi Measured Value Measured Value Value (+1)(+2) (+1)(+2) (+1)(+2) Value (+1)(+2) (+1)(+2) (+1)(+2)

LPCS-V-1 118 100 136 153 114 97 131 148 H

LPCS-FCV-11 12 10 14 16 10 8 13 H 15 LPCS-V-12 18 N/A N/A N/A 18 15 21 23 (+3)

H LPCS-V-3 NOT NOT N/A N/A N/A N/A N/A N/A OPEN CLOSED

1. (+1) For measured values beyond the Alert Value or Action Value refer to Precaution and Limitations 4.6 or 4.7, respectively.

(+2) When comparing measured values to Alert and Action limits round all measured Stroke Times to the nearest second. Use standard rounding techniques e.g., 17.49 rounds to 17 and 17.5 rounds to 18 seconds.

(+3) Use listed closing stroke time as limiting even though a higher limit is specified in Technical Specification.

H Motor operated valve.

END Attachment 9.1, Valve Stroke Data Sheet 8 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 36 of 36 LPCS-P-1 Acceptance Criteria (CPT)

LPCS-P-1 ACCEPTANCE CRITERIA (Two Year CPT) 340 1

338 336 ACTION RANGE 334 332 330 2

328 326 324 322 320 Differential Pressure in PSID 318 316 REFERENCE CURVE 314 312 310 308 306 304 5

302 300 298 4

296 ALERT RANGE 6 294 292 290 ACTION RANGE 3 288 286 6,375 6,400 6,425 6,450 6,475 6,500 6,525 6,550 6,575 6,600 6,625 Indicated Pump Flowrate in GPM ALERT RANGE = Area Inside 3-4-5-6 ACTION RANGE = Area Outside 1-2-3-4 END Attachment 9.8, LPCS-P-1 ACCEPTANCE CRITERIA (Two Year CPT) 9 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Table 1.6.1.3-1 (page 3 of 19)

Primary Containment Isolation Valves MAXIMUM PEN VALVE ISOLATION TIME VALVE VALVE TYPE NUMBER NUMBER (Seconds) GROUP(a) CODE NOTES 24 FDR-V-3 15 4 AIV 24 FDR-V-4 15 4 AIV 101 FPC-V-149 35 4 AIV 101 FPC-V-156 35 4 AIV 100 FPC-V-153 35 4 AIV (h) 100 FPC-V-154 35 4 AIV (h) 49 HPCS-RV-14 N/A N/A OCIV (g)(j) 49 HPCS-RV-35 N/A N/A OCIV (g)(j) 49 HPCS-V-12 N/A N/A OCIV (p)(k)(m) 49 HPCS-V-23 180 11 AIV (p) 6 HPCS-V-4 N/A N/A OCIV (i)(c)(k)(m) 58 HCV-V-1 N/A N/A OCIV 58 HCV-V-2 N/A N/A OCIV 6 HPCS-V-5 N/A N/A OCIV (i)(c) 31 HPCS-V-15 N/A N/A OCIV (c)(p)(k)(m) 78e HPCS-V-65 N/A N/A MCIV 78e HPCS-V-68 N/A N/A MCIV 63 LPCS-FCV-11 N/A N/A OCIV (p)(k)(m) 63 LPCS-V-12 180 10 AIV (p) 63 LPCS-RV-18 N/A N/A OCIV (g)(j) 63 LPCS-RV-31 N/A N/A OCIV (g)(j)

(a) See Technical Specification Bases 3.3.6.1 for the isolation signal(s) which operate each group.

(c) Valve leakage not included in sum of Type B and C tests.

(g) Not subject to Type C Leak Rate Test.

(h) Hydraulic leak test at 1.10 Pa.

(i) Not subject to Type C test. Test per Technical Specification SR 3.4.6.1.

(j) Tested as part of Type A test.

(k) Automatic open/close instrumentation supports ECCS operability; when not in use for ECCS normal position is closed.

(m) PCIV is operable (per TS 3.6.1.3) with non-functional open/close instrumentation if closed, secured and disabled from automatically opening.

(p) Not subject to leak rate testing (SR 3.6.1.1.1 and SR 3.6.1.3.12).

10 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3 ACTIONS

---

NOTES----------------------------------------------------------

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.

4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment,"

when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.

CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ A.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for Only applicable to penetration flow path by main steam line penetration flow paths use of at least one closed with two PCIVs. and de-activated automatic AND

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valve, closed manual valve, blind flange, or check valve 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main One or more penetration with flow through the valve steam line flow paths with one secured.

PCIV inoperable for reasons other than AND Condition D.

Columbia Generating Station 3.6.1.3-1 Amendment No. 169,208 225 251 11 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 ---------------NOTE--------------

1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected Once per 31 days for penetration flow path is isolation devices isolated. outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment Columbia Generating Station 3.6.1.3-2 Amendment No. 169,208 225 12 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. ------------NOTE------------ B.1 Isolate the affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Only applicable to penetration flow path by penetration flow paths use of at least one closed with two PCIVs. and de-activated automatic

---

valve, closed manual valve, or blind flange.

One or more penetration flow paths with two PCIVs inoperable for reasons other than Condition D.

C. ------------NOTE------------ C.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for Only applicable to penetration flow path by excess flow check penetration flow paths use of at least one closed valves (EFCVs) with only one PCIV. and de-activated automatic

---

valve, closed manual valve, AND or blind flange.

One or more penetration 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for EFCVs flow paths with one PCIV inoperable for AND reasons other than Condition D.

Columbia Generating Station 3.6.1.3-3 Amendment No. 169,208 225 13 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 --------------NOTES-------------

1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected penetration flow path is Once per 31 days for isolated. isolation devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment Columbia Generating Station 3.6.1.3-4 Amendment No. 169,208 225 14 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. One or more secondary D.1 Restore leakage rate to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for containment bypass within limit. hydrostatically tested leakage rate, MSIV line leakage not on a leakage rate, or closed system hydrostatically tested lines leakage rate not AND within limit.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for secondary containment bypass leakage AND 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for MSIV leakage AND 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for hydrostatically tested line leakage on a closed system E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, B, AND C, or D not met.

E.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Columbia Generating Station 3.6.1.3-5 Amendment No. 169,208 225 251 15 of 15

ILC-24 NRC Validation Exam Question: SRO-91 RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1 ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> channels inoperable.

OR

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

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

A.2 Place associated trip 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> system in trip.

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NOTE-------------- B.1 Place channel in one trip 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Not applicable for system in trip.

Functions 2.a, 2.b, 2.c, 2.d, or 2.f. OR B.2 Place one trip system in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B. One or more Functions trip.

with one or more required channels inoperable in both trip systems.

Columbia Generating Station 3.3.1.1-1 Amendment No. 169 225 226 253 1 of 4

ILC-24 NRC Validation Exam Question: SRO-91 RPS Instrumentation 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore RPS trip capability. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Action D.1 and POWER to < 29.5% RTP.

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and all insertable control rods in referenced in core cells containing one or Table 3.3.1.1-1. more fuel assemblies.

Columbia Generating Station 3.3.1.1-2 Amendment No. 169 225 226 253 2 of 4

ILC-24 NRC Validation Exam Question: SRO-91 RPS Instrumentation 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I. As required by Required I.1 Initiate alternate method to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action D.1 and detect and suppress referenced in thermal hydraulic instability Table 3.3.1.1-1. oscillations.

AND

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

LCO 3.0.4 is not applicable.

I.2 Restore required channels 120 days to OPERABLE.

J. Required Action and J.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to less than the Time of Condition I not value specified in the met. COLR.

SURVEILLANCE REQUIREMENTS

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NOTES----------------------------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.1-3 Amendment No. 169 225 226 253 3 of 4

ILC-24 NRC Validation Exam Question: SRO-91 RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

3. Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.8 1079 psig Dome Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

4. Reactor Vessel Water 1,2 2 G SR 3.3.1.1.1 9.5 inches Level - Low, Level 3 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

5. Main Steam Isolation Valve 1 8 F SR 3.3.1.1.8 12.5% closed

- Closure SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

6. Primary Containment 1,2 2 G SR 3.3.1.1.8 1.88 psig Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14

7. Scram Discharge Volume Water Level - High

a. Transmitter/Level 1,2 2 G SR 3.3.1.1.1 529 ft 9 inches Indicating Switch SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.1 529 ft 9 inches SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14

b. Transmitter/Level 1,2 2 G SR 3.3.1.1.8 529 ft 9 inches Switch SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.8 529 ft 9 inches SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(d) If the as-found channel setpoint is outside its predefinded as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and the as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-9 Amendment No.169 225 226 253 4 of 4

ILC-24 NRC Validation Exam Question: SRO-97 SRM Instrumentation 3.3.1.2 3.3 INSTRUMENTATION 3.3.1.2 Source Range Monitor (SRM) Instrumentation LCO 3.3.1.2 The SRM instrumentation in Table 3.3.1.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.2-1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required SRMs to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SRMs inoperable in OPERABLE status.

MODE 2 with intermediate range monitors (IRMs) on Range 2 or below.

B. Three required SRMs B.1 Suspend control rod Immediately inoperable in MODE 2 withdrawal.

with IRMs on Range 2 or below.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A or B not met.

D. One or more required D.1 Fully insert all insertable 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SRMs inoperable in control rods.

MODE 3 or 4.

AND D.2 Place reactor mode switch 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in the shutdown position.

Columbia Generating Station 3.3.1.2-1 Amendment No. 149,169 225 1 of 3

ILC-24 NRC Validation Exam Question: SRO-97 SRM Instrumentation 3.3.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. One or more required E.1 Suspend CORE Immediately SRMs inoperable in ALTERATIONS except for MODE 5. control rod insertion.

AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified conditions.

SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.2-2 Amendment No. 149,169 225 238 2 of 3

ILC-24 NRC Validation Exam Question: SRO-97 SRM Instrumentation 3.3.1.2 Table 3.3.1.2-1 (page 1 of 1)

Source Range Monitor Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS REQUIREMENTS (a)

1. Source Range Monitor 2 3 SR 3.3.1.2.1 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7 3, 4 2 SR 3.3.1.2.3 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7 5 2 (b), (c) SR 3.3.1.2.1 SR 3.3.1.2.2 SR 3.3.1.2.4 SR 3.3.1.2.5 SR 3.3.1.2.7 (a) With IRMs on Range 2 or below.

(b) Only one SRM channel is required to be OPERABLE during spiral offload or reload when the fueled region includes only that SRM detector.

(c) Special movable detectors may be used in place of SRMs if connected to normal SRM circuits.

Columbia Generating Station 3.3.1.2-5 Amendment No. 149,169 225 3 of 3

2021 NRC Exam Question: SRO-100 Page 1 of 1