ML13003A118

LaSalle County Station Initial License Examination Written Examination Post Examination Comments and Documents

ML13003A118
Person / Time
Site:	LaSalle
Issue date:	10/22/2012
From:	McNeil D R NRC/RGN-III/DRS/OLB
To:	Exelon Generation Co
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ML11354A201	List: ML13003A102 ML13003A118 ML13003A120
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Download: ML13003A118 (42)

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LaSalle County Station Initial License Written Post Examination Comments and 1 LaSalle Station ; Exelon 10CFR50.4 RA12-056 November 08, 2012 U. S. Nuclear Regulatory Commission Attention:

NRC Region III Administrator 2443 Warrenville Rd. Suite 210 Lisle, IL 60532-4352 laSalle County Station, Units 1 and Facility Operating License Nos. NPF-11 and NRC Docket Nos. 50-373 and Comments on NRC Initial License Examination administered the weeks of October 22 and October 29. 2012 In accordance with NUREG-1021 , "Operator Licensing Examination Standards for Power Reactors," Revision 9, Supplement

1. Exelon Generation Company, LLC, (EGC) submits comments for your review on the examination administrated during the weeks of October 22 and October 29, 2012. Attachment I provides the examination comments in accordance with the guidelines in revision 9, Supplement 1, of NUREG-1021 , "Operator Licensing Examination Standards for Power Reactors,*

E5-402, "Administering Initial Written Examinations*.

Attachment II provides the marked up contended questions along with supporting reference documentation.

Should you have any questions concerning this letter, please contact Mr. Guy V. Ford, Regulatory Assurance Manager, at (815) 415-2800.

Respectfully, ,n U* Harold T. Vinyard Plant Manager laSalle County Station 9 Chief, NRC Operator LicenSing Branch (with enclosures)

Senior Resident Inspector

-laSalle County Station (w/o enclosures) -t:IOI.c.:CT H t&--Z")ftD"""12

-';;')L Docket No.12-056 Test Number 2012301 Attachment 1 Page 1 of 6 LaSalle County Station NRC Initial License Examination Comments A review of the NRC IL T examination administered at LaSalle County Station during the weeks of 10/22 and 10/29,2012 was conducted on Friday, 11/2/12 with all examinees participating in the review. During the administration of the written examination, eight (8) different license applicants asked questions related to the exam. The exam question number, license applicant's question, and the facility's reply are included below. During the Post-Exam review, two (2) candidates commented on questions.

Docket Number 55-3648 has challenges on five (5) questions having two possible answers utilizing TQ-AA-151-F11 (see attached).

The challenge and station's responses to this TQ-AA-151-F11 comment sheet begin on page 4 of this Attachment.

Docket Number 55-3649 commented on four (4) questions, but did not specify any challenges.

WRITTEN EXAM Questions asked! Proctor Question Examinee Docket Number 55-33642 Distractor C states, Increasing TBCCW Expansion Tank Levels. Should this question answered based on plant The answer provided was: Per Nureg 1021 Appendix E B.7 all questions should answered based on plant Question Examinee Docket Number 55-33650 Is the synchroscope in the 12 O'clock position on Unit 1 or Unit The answer provided was: Based on Unit 1 unless stated Question Examinee Docket Number 55-33642 Is the indication given stable or a snapshot The answer provided was: Answer based on given Question Examinee Docket Number 55-33643 Should I assume the pumps are going to trip when I put water in The answer provided was: Answer based on LGA-002 Docket No.12-056 Test Number 2012301 Attachment 1 Page 2 of 6 Question 13: Examinee Docket Number 55-33643 asked: Is the RPS MG Set Fine? The answer provided was: You have all the information you Question Examinee Docket Number 55-33648 Are the DG room CO 2 timers timed The answer provided was: You have all the information you Question Examinee Docket Number 55-33650 Is the term FIRST in the question stem based on lapse in time or right The answer provided was: No time specified.

If we wanted a delay in time would provided Question Examinee Docket Number 55-33641 Are we supposed to know which switch is associated with the alarms The answer provided was: Knowledge you need to Question Examinee Docket Number 55-33642 Is the second refuel floor ARM C or D The answer provided was: You have all the information you Question Examinee Docket Number 55-33649 Does the term stable mean no The answer provided was: Answer based on your Question Examinee Docket Number 55-33641 CRD Flow Controller Image: Which controller is this in the simulator? (The simulator currently has a different controller than the plant) The answer provided was: Identified controller in the simulator with Manual and Auto on the bottom.

Docket No.12-056 Test Number 2012301 Attachment 1 Page 3 of 6 Question 37: Examinee Docket Number 55-33640 asked: Can you tell me what PT 1 and PT 2 says in the circle? The answer provided was: PT 1 is PS and PT 2 is PTS. (Based on the poor picture clarity candidates could not clearly identify the terms in the circle.) Question 58: Examinee Docket Number 55-33644 asked: Does the term re-entry imply you had a transient?

The answer provided was: Transient does not matter to answer the question.

Question 79: Examinee Docket Number 55-33641 asked: Is there an alarm indication?

The answer provided was: There are no alarms up based on the picture. Question 97: Examinee Docket Number 55-33640 asked: Are EAL bases required to be memorized?

The answer provided was: You do not need EAL bases to answer the question.

Docket No.12-056 Test Number 2012301 Attachment 1 Page 4 of 6 Post-Exam Review Comments made by Docket # 55-33648.

1. RO Question #10 Question #10 on the RO exam has the correct answer as (C), which is correct per LOP-AA-03 Table 2 Sheet 1. Candidate Comment: The comment made was when a Flow Unit fails to 0% a Downscale/lNOP condition would cause a Rod Block ONLY, which is answer (B). Station's Response:

Based on the Lesson Plan provided and LOP-AA-03 Table 2 Sheet 1 you would not receive an INOP condition without making assumptions that the Flow Unit mode switch is not in operate, module was unplugged or the power supply out of specified range. A Downscale condition does exist, which would cause a Rod Block and a Half-Scram.

The Half-Scram is caused by exceeding the Flow Biased setpoint.

STATION RECOMMENDATION:

ACCEPT ONLY (C) as the correct answer. 2. RO Question #35 Question #35 on the RO exam has the correct answer as (C), which is correct per System Description 047. Candidate Comment: The comment made was when "??" is displayed it represents 4 rods that have Data Faults, which is answer (A) and is also correct based on new technical information.

Station's Response:

The rod position text that is displayed by RCMS is based on the output value from the rod position indication system (RPIS) portion of the system. Per system description 047, a Probe Multiplexer (MUX) card not responding will cause an RPiS '255' Signal to be generated which will then display"??" for the 4 rods controlled by that Probe MUX card. This new technical information (GE Propriety Class III Information data sheet DO NOT RELEASE), shows that when a code 255 is present, a data fault bit is present (OF) and thus a data fault would be displayed on RCMS screens. This would make (A) a correct answer as well due to the limited information in the stem. STATION RECOMMENDATION:

ACCEPT BOTH (A and C) as correct answers.

Docket No.12-056 Test Number 2012301 Attachment 1 Page 5 of 6 3. RO Question #48 Question #48 on the RO exam has the correct answer as (A), which is correct per LOA-WR-101.

Candidate Comment: The comment made was that a loss of WR also has an impact on RWCU (RT) Pump motor and should be monitored making answer (D) correct. Station's Response:

RWCU (RT) pumps are cooled by WR, but are not required to be monitored in LOA-WR-101.

Steps C.2.2, of the discussion section, and B.1.5, of the Reduced Cooling Capacity section, both specify monitoring the RR Pump Seal temperatures.

STATION RECOMMENDATION:

ACCEPT ONLY (A) as the correct answer. 4. RO Question #60 Question #60 on the RO exam has the correct answer as (B), which is correct per LGA-003 and LGA-011 Emergency procedures.

Candidate Comment: The comment made was that the plant is below 2% Hydrogen indication, therefore the Hydrogen leg of LGA-003 would not be entered making answer (A) correct as well. Station's Response:

LGA-003lesson plan states if LGA-003 is entered than parallel execution is also required because of the symptomatic approach to emergency response precludes the prioritization of anyone action path since independence for initiating events and transients must be maintained.

Therefore the Hydrogen leg of LGA-003 is entered because the stem of the question states Hydrogen is 1% and rising slowly, which leads to entering LGA-011 and starting the Hydrogen Recornbiner.

STATION RECOMMENDATION:

ACCEPT ONLY (B) as the correct answer. 5. RO Question #74 Question #74 on the RO exam has the correct answer as (B), which is correct per LOP-NR-06.

Candidate Comment: The comment made was based on when the RB TIP ROOM RAD HI/DOWNSCALE alarm comes in. It is possible the alarm was caused by the second TIP run and the TIP run should be placed in a safe condition until the cause of the alarm is verified thus making answer (D) also correct. Station's Response:

The stem of this question indicates that the "B" TIP has just been completed and has been returned to its shield. The "A" TIP has just been positioned at 0001. At this point, the question states the TIP Room Area Radiation Monitor alarms and it is verified to be "HI". To answer the question based only on the data provided without assumptions, the following logical approach is expected.

The first operator response to an alarm is to follow the panel alarm procedure, in this case LOR 1 H13-P601 B211.

Docket No.12-056 Test Number 2012301 Attachment 1 Page 6 of 6 This procedure has the operator read the corresponding Area Radiation Monitor (ARM) to determine actual radiation level and also refer to the ARM abnormal response procedure, LOA-AR-101.

The magnitude of the dose rate level is not provided and it is not stated whether this is due to known reasons, thus the steps of LOR 1 H13-P601 B211 and LOA-AR-101 properly direct actions. These procedural steps direct appropriate actions to place the equipment in a safe condition.

To place a TIP in a safe condition is to withdraw it into its shield and close the TIP ball valve. While execution of a TIP trace set typically does lead to receipt of alarm 1 P601 B211, the question does not contain enough details that assure that this is an expected alarm condition.

Proper panel alarm and abnormal response procedure actions lead to placing the equipment in a safe condition.

This response leads to answer (D) also being a correct answer. STATION RECOMMENDATION:

ACCEPT BOTH (B and D) as correct answers.

TQ-AA-1S1-F11 Revision 0 Page 1 of 1 Exam Comment Sheet Z. Date: 11-2-2012 Submitted By:

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Docket No.12-056 Test Number 2012301 Attachment 2 Page 1 of 1 Marked-Up contended Questions and Supporting Reference Documentation (33 pages)

EXAM1NATION****.*ANSWER**KEY 11-1 NRC ROExam Unit 1 is operating at 100% of rated thermal power operating on the 100% Flow Control Line

• All nuclear instruments are operable

• The C Flow Unit fails to Based on the above conditions, which of the following correctly states the effect on the A. A Full Scram. B. A Rod Block ONLY. C. A Rod Block and a Half Scram. D. A Flow Comparator alarm ONLY. Answer: C Answer Explanation:

A Control Rod Block will be generated if the following conditions exist while the mode switch is in -Flow Unit Upscale -Flow Unit INOP (unplugged, switch not in

-Comparator Trip (10% difference in output flow A Half-Scram will be generated based on APRM Flow Bias Failing such that power is greater than flow. A Full Scram signal will be generated is incorrect because you are below the scram setpoint and only have one trip unit in RPS channel A failing. If a Flow Unit in RPS Channel B were to fail to 0% then you would receive a Full Scram. A Flow Comparator alarm ONLY is incorrect because you have exceeded the Flow Unit upscale setpoint although you would receive the alarm the Rod Block is in. A Rod Block ONLY signal will be generated on the A RPS channel is incorrect because you have passed the scram setpoint.

Reference:

LOR 1 H13-P603-A209.

LOP-AA-03 Table 2 Sheet Reference provided during examination: Cognitive level: High Level (RO/SRO): Tier: 2 Group: KIA: 215005 Average Power Range MonitorlLocal Power Range Monitor System KS.05 Knowledge of the operational implications of the following concepts as they apply to AVERAGE POWER RANGE MONITORILOCAL POWER RANGE MONITOR SYSTEM : Core flow effects on APRM trip setpoints 10 CFR Part 55 Content: SRO Justification: LAS LICENSED OPS Page: 19 of 136 31 October 2012 EXAMINATION ANSWER 11*1 NRC AO Exam Question Source: Question History:

Associated Given various plant conditions, predict the response of the following supported systems to a loss of the APRM System while operating the system, or on an exam in accordance with student text: a. Rod Control Management System b. Reactor Protection System c. Process Computer d. Rod Block Monitor e. OSCillation Power Range Monitor LAS LICENSED OPS Page: 20 of 136 31 October 2012 Content/Skills Activities/Notes IV. Interlocks A. APRM TripslRod Blocks/Alarms OBJ 044.00.15

T'DTPi> .**.**.**.REFEREN(JI**.}

...*.*

gO%RTP Neutron Flux -High TS -Allowable Will cause a 14.5% LIS -Cal Scram if mode 15% LOR -

NOT in Neutron Flux -High s14% TRM -Allowable Value Setdown: Will cause 11.5% LIS Cal Block if mode 12% LOR -Setpoint ! NOT in RUN i Fixed Neutron Flux s120%RTP TS -Allowable (LCO) I Will cause a Reactor 117.5% LIS -Cal 118% LOR -Setpoint I Flow Biased sO.61W + 68.2% RTP and TS -Allowable Value I Simulated Thermal sI15.5%RTP (LCO):Two Power-sO.54W + 55.9% RTP and TS -Allowable Will cause a SI 12.3% RTP (LCO) Single 0.61W + 62.6% RTP and LIS -Cal Setpoint 113%RTP Two Loop 0.54W + 50.5% RTP and LIS -Cal Setpoint:

107.5%RTP Single Loop 0.61W + 62.6%RTP and LOR -Setpoint Two 113.5%RTP Loop 0.54W + 50.5% RTP and LOR -Setpoint Single 108.l%RTP Flow Biased sO.61W + 56.9% RTP TRM -

Flux -

Value (LCO): Will cause Block if mode s0.54W + 44.7% RTP TRM -

in Value (LCO): Single I Loop O.61W + 50.8% RTP LIS Cal Setpoint:

Two Loop 0.54W + 38.7% RTP LIS Cal Setpoint:

Single Loop O.61W + 51.3%RTP LOR -Setpoint Two Loop 0.54W + 39.2% RTP LOR Setpoint Inoperative TS, TRM, LIS, Will cause 1. Too Few Inputs (SI4) or, 2. Module Unplugged or, TS requires at least Scram!

3. Function Switch Not in LPRM inputs, LIS Operate trip at 2:3% TRM -

Will cause Value Block if mode 5.5% LIS -Cal in 5% LOR Setpoint P:\PROCUPGD\APPROVED\TRAIN-LP\Ops\System LPs\044, Average Power Range Monitoring\044 APRM.doc Page 13 of45

..

10 Content/Skills Activities/Notes

• ..**....*.****** ... ************.*

• • • • • • . W.If.*

Recirculation Flow :::; 1111125 of full scale TRM -Allowable Unit Upscale: Value (LCO) Will cause AlarmlRod Block 107.8751125 offull scale LIS -Setpoint ! I 108/125 offull scale LOR -Setpoint I Recirculation Flow :::; II% flow deviation TRM -Allowable Unit Comparator Value (LCO) Trip: 10% flow deviation LIS -Setpoint Will cause AlarmlRod LOR Setpoint Block Recirculation Flow I. Mode Selector Switch TRM, LIS, LOR Unit Inoperative not in OPERATE Trip: 2. Internal power supply Will cause AlarmlRod out of specified range Block 3. A card is removed from the flow limit card cage P:\PROCUPGD\APPROVED\TRAIN-LP\Ops\System LPs\044, Average Power Range Monitoring\044 APRM.doc Page 14 of 45 EXAMINATION ANSWER 11-1 NRC RO Exam Unit 1 is at rated power when the operator observes the following:

Jl 21 19 15 48 48 48 12 48 48 17 17 48 48 48 48 48 48 ?? ?? 48 48 48 48 48 00 48 48 48 48 48 48 48 48 48 48 --48 48 16 48 48 48 12 48 48 48 48 48 48 48 48 48 48 48 I I 48 48 48 48 lD 22 Which of the following is the cause of these indication for the four control rods in the red box? A. Data Faults B. Rods are at the overtravel position C. Probe MUX Card not responding D. No position switches are CLOSED Answer: C Answer Explanation:

The loss of the Probe MUX Card will cause an indication of ?? on the Full Core Display. The PIP cables feed to the Probe MUX cards. The Probe MUX cards feed to the File Control processer which feed to the RCMS Controller card. The RCMS Controller then feeds the MCR Controller card which sends data to all RCMS Displays in the control room. When the data from the PIP cable is lost due to Probe MUX Card not responding it will cause the?? Data Fault will cause a text box telling you have lost data and the rod indication on the Full Core Display will be "Xx." If a rod has overtraveled there will be a text box stating Withdraw Error and the rod indication on the Full Core Display will be "OT." If there were no position switches CLOSED the rod indication on the Full Core Display will be blank.

Reference:

system description 047 Reference provided during examination:

N/A Cognitive level: memory LAS LICENSED OPS Page: 66 of 136 31 October 2012 EXAMINATION ANSWER 11*1 NRC ROE:xam Level (ROISRO):

RO Tier: 2 Group: 2 KIA: 214000 Rod Position Infonnation System A3.01 Ability to monitor automatic operations of the ROD POSITION INFORMATION SYSTEM including:

Fun core display 10 CFR Part 55 Content: 41.7 SRO Justification:

N/A Question Source: New Question History: N/A Comments:

Associated objective(s):

Given various plant conditions, predict the Rod Control Management System response to a loss of the major power supplies, RD system, or RL system while operating the system, or on an exam in accordance with station procedures.

LAS LICENSED OPS Page: 67 of 136 31 October 2012 Reference material intentionally withheld.

Proprietary information returned to the facility licensee.

EXAMINATION

• .A.NSWERKEY NBC HQExam Which of the following is directed to be monitored per LOA-WR-1 01, "Loss of Reactor Building Closed Cooling Water RBCCW" during reduced cooling capacity conditions in the RBCCW system? A. RR Pump seal cavity outlet temperature B. Offgas Refrigeration Machine glycol outlet temperature C. IN Compressor intercooler discharge temperature D. RWCU Pump Motor Cooler outlet temperature Answer: A Answer Explanation:

RR Pump seal cavity outlet temperature per LOA-WR-1 01 section B.1.5 specifically calls for monitoring of RR pump seal temperatures, and no other equipment is listed. Offgas Refrigeration Machine, IN Compressor and RWCU Pump Motor cooler are all loads of RBCCW and would be required to be shutdown if you lost all RBCCW cooling.

Reference:

LOA-WR-101 Reference provided during examination: Cognitive level: Level Tier: 1 Group: KIA: 295018, Partial or Total Loss of Component Cooling AA 1.02 -Ability to operate and/or monitor the following as they apply to PARTIAL COMPLETE LOSS OF COMPONENT COOLING WATER: System 10 CFR Part 55 Content: SRO Justification: Question Source: Question History:

Associated Given a Service Water/chemical feed system lineup and various plant evaluate system indications/responses and determine in the indications/responses are expected and normal while operating the system, or on an exam in accordance with the student text. LAS LICENSED OPS Page: 90 of 136 31 October 2012 LaSalle UNIT OPERATING ABNORMAL LOSS OF REACTOR BUILDING CLOSED COOLING Revision March 23, Procedure ResponsibilitylReviewl Approval Requirements Responsible Department Head: SOS Minimum Review Type: TR Restuired Cross-Discipline Review(s):

N/A Approval Position Required:

SOS Specific Requirements:

1. Review/Approval requirements apply to non-editorial procedure revisions.

Level of Use Continuous I of 13 LOSS OF REACTOR BUILDING CLOSED COOLING WATER (RBCCW) TABLE OF CONTENTS A. SYMPTOMSIENTRY CONDITIONS

...............................................................................

.B. ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ...............................

.B.l Reduced Cooling Capacity (Potential Manual Scram) ...........................................

.B.2 Loss ofRBCCW System (Potential Manual Scram) ............................................

.C. DISCUSSION

....................................................................................................................Level of 20f13 Revision 10 LOA-WR-lOl March 23,2012 LOSS OF REACTOR BUILDING CLOSED COOLING WATER (RBCCW) SYMPTOMSIENTRY CONDITIONS A.I RBCCW System Reduced Cooling Capacity.

A.2 Alarms at panel] PMlOJ: AIOl, Reactor Building Closed Cooling Water Pump Automatic Trip A201, Reactor Building Closed Cooling Water Pump Discharge Header Pressure Low A202, Reactor Building Closed Cooling Water Pump Suction Temperature Hi Level of Use Continuous LOA-WR-IOI 30fl3 Revision 10 March 23, 2012 MANUAL SCRAMS Loss ofRBCCW.f!if!::o ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED Reduced Cooling Capacity "'"Immediate CHECK RBCCW header 1.1 START standby RBCCW Action discharge pressure -pump. GREATER mAN 57 psig. If NO RBCCW pumps can be operated, EXIT to Section B.2. CHECK Service water pressure -2.1 START standby Service Water GREATER THAN 80 psig. pump (4 WS Pump Maximum). If Service Water pressure can NOT be maintained above 40 psig, EXIT to Section B.2. MONITOR RBCCW discharge

3.1 COMMENCE

an orderly unit header shutdown per LGP-2-1, Normal Unit Shutdown. Temperature remains LESS THAN Level of Use Continuous LOA-WR-101 4 of 13 Revision 10 March 23, 2012 MANUAL SCRAMS Loss ofRBCCW. B. ACTIONIEXPECTED RESPONSE RESPONSE NOT Reduced Cooling Capacity (continued) MONITOR Drywell for potential

4.1 RBCCW

leak -NO leak in DW. DWFDS inputs have NOT increased 4.2 DW temperature NOT 4.2.1 increasing

• DW pressure LESS THAN 4.2.2 1.69 # 4.2.3 4.3 SMIUS EVALUATE operation can continue based RBCCW leakage If unit operation CANNOT continue.

MANUALL Y SCRAM reactor. PLACE RR Pmp breakers in PTL: lARRPmp:

• BkrlA

• Bkr2A

• Bkr3A 1BRRPmp:

• Bkr IB

• Bkr2B Bkr3B ISOLATE RBCCW to DW by CLOSING: lWR029 lWR040 lWR179 lWR180 IfRBCCW system parameters restored to normal, exit this procedure.

Level of Use Continuous LOA-WR-IOI 5 of 13 Revision 10 March 23,2012 MANUAL SCRAMS Loss ofRBCCW. So ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED Sol Reduced Cooling Capacity (continued)

CAUTIONS RR Pump Operation with Seal temperature( s) greater than 185°F may result in seal damage. RR Pump shutdown from rated temperature without seal purge will robably result in complete failure of a damaged seal. MONITOR RR Pump lA and IB Seal Temperatures: Seal Temperatures LESS THAN 185°F: (T 2) Seal Injection CHECK operating RBCCW heat 6.1 exchanger(s) temperature controller setpoint(s)

-75°F to 90°F. CHECK operating RBCCW heat 7.1 exchanger(s) outlet 80 TO 95°F. JiRR Pump Seal temperature(s) reach 200°F: PLACE breakers for affected RR Pump(s) in PTL: ENTER LOA-RR-lOl for RR Pump Trip. ADJUST operating temperature controller(s) to 75°F to 90°F: lTIC-WR032/33. OTIC-WR007. THROTTLE 1 WS087 AlB to maintain RBCCW outlet header temperature 80 to 95°F. If Unit 0 RBCCW Hx is aligned to Unit 1, THROTTLE lWS087C to maintain RBCCW outlet header temperature 80 to 95°F. 6. 7. Level of LOA-WR-lOI 60f13 Revision 10 March 23, 2012 MANUAL SCRAMS -".... -.::>-..,-0 Loss ofRBCCW.-' . ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED Reduced Cooling Capacity (continued) CHECK RBCCW discharge 8.1 If available, PLACE the off-line header temperature

-LESS lAIB RBCCW Heat Exchanger TIIAN lOO°F. in parallel operation: OPEN 1 WS085AIB, WS to RBCCW Hx IAIB Inlet Stop. OPEN I WS088A1B, WS From RBCCW Hx IAIB FCV Downstream Stop. OPEN 1 WS086AIB, WS From RBCCW Hx lA FCV Upstream Stop. VERIFY lTIC-WR032/33 set to 75°F to 90°F. OPEN IWR041A/B, lAIB RBCCW Heat Exchanger RBCCW Inlet Stop. OPEN 1 WR042AIB, lAIB RBCCW Heat Exchanger RBCCW Outlet Stop. Level of LOA-WR-lOl 7 of 13 Revision 10 March 23, 2012 MANUAL SCRAMS .....*....-, -' a Loss ofRBCCW. . , R ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED Rl Reduced Cooling Capacity (continued)

CAUTION To prevent loss of RBCCW, the Unit 0 RBCCW Heat Exchanger must NOT be simultaneously aligned to both Units. If I AlB RBCCW Hx is NOT available and 0 RBCCW Hx is available to Unit 1, PLACE o RBCCW Heat Exchanger in parallel operation: VERIFY 0 RBCCW Hx is NOT required operating for Unit 2. OPEN 1 WS085C, WS to o RBCCW Hx Inlet Stop. OPEN 1 WS086C, WS From o RBCCW Hx Upstream Stop. OPEN 1 WS088C, WS From o RBCCW Hx Downstream Stop. VERIFY Temperature Controller OTIC-WR007 set at 75°F to 90°F. Level of Use Continuous LOA-WR-lOl 80fl3 Revision 10 March 23, 2012 MANUAL SCRAMS Loss of RBCCW. ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED Reduced Cooling Capacity (continued) CLOSE 2WR042C, Tie From RBCCW Hx OA to Unit 2 Stop. CLOSE 2WR041C, U2 Cross-Tie to RBCCW Hx OA Stop. OPEN 1WR041C, o RBCCW Heat Exchanger RBCCW Inlet Stop. OPEN 1WR042C, o RBCCW Heat Exchanger Outlet Stop. CHECK RBCCW discharge

9.1 CONSIDER

shutting down header temperature

-LESS RWCU system per LOP-RT-03 THAN 105 0 F. to reduce heat load on the WR system. Level of LOA-WR-101 9 of 13 Revision 10 March 23, 2012 MANUAL SCRAMS Loss ofRBCCW. B. ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED Loss of RBCCW System CAUTION WR cooling is lost to both RR pumps and other plant equipment. MANUALLY SCRAM reactor. PLACE RR Pmp breakers in PTL:

• IARRPmp:

• Bkr IA

• Bkr2A

• Bkr3A

• IB RRPmp:

• Bkr IB

• Bkr2B

• Bkr 3B STOP running RWCU Pump(s): o IARWCUPmp o IB RWCU Pmp MONITOR for Reactor Vessel stratification per LOA-RR-101 while continuing here. Level of Use Continuous LOA-WR-101 10 of 13 Revision 10 March 23, 2012 MANUAL SCRAMS Loss of RBCCW. B. ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED B.2 Loss ofRBCCW System (continued)

S. SHUTDOWN RBCCW loads: RWCU System Off Gas machines lAIB OffOas Building HVAC per LOP-VO-02. Reactor Building Storage Room Drywell Compressors CRD Feed pumps Level of Use Continuous LOA-WR-lOl 11 of 13 Revision 10 March 23, 2012 C. DISCUSSION C.l C.2 Co2. I Co2.2 C.3 Procedure is written to address reduced cooling capacity and loss of the system. Reduced cooling capacity section directs the user to exit to Section B.2, (loss ofRBCCW) in the event no RBCCW pumps can be started, Service Water pressure can NOT be improved or RBCCW system temperature has risen to 110°F (FSAR 9.2.3.2).

Otherwise:

The Drywell temperature, pressure and Unidentified Leakage are monitored for indications of an RBCCW leak inside the DrywelL These parameters are suggested for monitoring because 1) the increased DWFDS inputs identify water leakage, 2) the low DW pressure and temperature indicate that the water is from a cold water source and not from condensed steam, and 3) stable DW temperature also indicates that VP is probably NOT the source of the water. Obviously, the Control Room team should use all available indications.

If the Control Room team identifies that the RBCCW degradation is due to an RBCCW leak in the DryweU and unit operation cannot continue based on size of the leak, the Reactor is scrammed, the RR pumps are shutdown and RBCCW is isolated to the DrywelL This action is intended to restore RBCCW cooling capacity to maintain cooling to other plant loads. RR Pump Seal temperatures are monitored.

If sea peratures are at or above 185° F, the pump seal faces are probably being damaged and a shutdown of the pump should be commenced within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the CRD seal purge flow entering the seal (T2) or seal staging outflow (T3) reach 200° F, then seal failure is imminent and the pump must be shutdown. (Ref. EC 339538) The RBCCW system Temperature control valve setpoint is checked/adjusted as necessary.

Additional system heat exchangers are placed in service as necessary.

The loss ofRBCCW section is written assuming that actions of supporting annunciator procedures have already been taken, (pump trip, expansion tank level, etc.). A complete loss of the RBCCW system requires immediate action to minimize damage to plant components.

The most limiting concern is the Recirculation Pump Motor Windings.

Level of Use Continuous LOA-WR-IOI 12 of 13 Revision 10 March 23, 2012 Personnel are directed to consider the shutdown of the Reactor Water Cleanup (RWCU) System if the heat load on the RBCCW system cannot be controlled below 105° F. The RWCU system is the primary heat load for the RBCCW system and temporarily removing this heat load will gain time for personnel to take additional compensatory actions to regain temperature control prior to the system reaching its 110° F design limit. LOP-RT-03 for removing RWCU contains actions for the Chemistry Department to monitor reactor water chemistry while the system is shutdown. The following systems are affected by loss ofRBCCW:

• Reactor Bldg Drain Tank Heat Exchanger, lREOIA.

• Off Gas Refrigeration Machines 1 AlB.

• Off Gas Bldg HV AC A/C Unit A.

• Reactor Bldg fist Storage Room HV AC A/C I VROl S.

• Reactor Recirc Pump lAiB Seal, Bearing and Motor Coolers.

• Reactor Bldg Process Sample Cooler PnIIPL14J.

• Drywell Penetration Cooling Coils.

• Drywell Equip Dm Sump Heat Exchanger lRE02A.

• CRD Feed Pumps IAIB.

• RWCU Non-Regen Heat Exchangers 1 AlB.

• RWCU Recirc Pumps IAIB.

• Drywell Pneumatic Compressors lIN02CA, IIN02CB.

• Drywell Pneumatic Compressor Aftercoolers lINI6AA, IlNI6AB.

• Drywell Sump Sample Pump lREI4P. Level of Use Continuous LOA-WR-lOl 13 of 13 Revision 10 March 23,2012

..

1 t RO Exam With Unit 2 at rated power, a large LOCA occurred in which fuel became temporarily uncovered.

LGA-001 and LGA-003 have been entered. 20 Minutes after operators started the Post-LOCA H2/02 monitoring system, the following readings are taken:

• Drywell 02 concentration is 0.5% by volume and stable.

• Drywell H2 concentration is 1% by volume and rising slowly. 1) Would these Post-LOCA H2/02 monitoring system readings be RELIABLE or would they still NEED MORE TIME to warm up and stabilize?

2) Based solely on the H2/02 content, for these post-LOCA conditions, would operation of the H2 Recombiners per LGA-HG-101 be DESIRABLE or NOT DESIRABLE? 1)

2) NOT 1)

2) 1) NEED MORE 2) 1) NEED MORE 2) NOT Answer: B Answer Explanation:

Per LGP 2-1, Drywell de-inerting can begin while at power, making it plausible to have elevated 02 levels. Post-LOCA H2I02 monitoring system is started as part of LGA-003. Per LOP-CM-02, these analyzers take 15 minutes to warm up and stabilize.

After 20 minutes these readings would be RELIABLE.

When hydrogen is detected (0.5% is minimum detectable) then LGA-011 Hydrogen Control is entered. The 1 st step is to place H2 recombiners in service as a mixing system per LGA-HG-201.

They would not be shutdown unless H2 concentrations exceed 5% at the given 02 level, so their operation is DESIRABLE.

RELIABLE, NOT DESIRABLE:

Selected if the candidate does not recognize

1) that the Lower Explosive Limit of Hydrogen is 4%, 2) that the LGA-011 breakpoint for recombiner shutdown is 5%, or 3) that does not include recombiner operation with the electric heaters (Le.: the recombiner) in service. NEED MORE TIME, DESIRABLE:

Any time over 20 minutes is plausible if the 15-minute warmup time in not known. NEED MORE TIME, NOT DESIRABLE:

A combination of Distractor 1 &2 explanations.

Also selected if the candidate thinks the analyzers NEED MORE TIME because, per LGA-HG-101, unknown readings are interpreted as high readings, which would require H2 Recombiner shutdown.

LAS LICENSED OPS Page: 109 of 31 October 2012 EXAf\IIlNATION ANSVVER f NRC ROExam

Reference:

LGA-011, LGA-HG-201, LOP-CM-02 Reference provided during examination:

N/A Cognitive level: High Level (RO/SRO):RO Tier: 1 Group: 2 KJ A: 500000 High Containment Hydrogen Concentration EK2.01Knowledge of the interrelations between fiGH CONTAINMENT HYDROGEN CONCENTRATIONS the following:

Containment hydrogen monitoring systems 10 CFR Part 55 Content:41.7 SRO Justification:

N/A Question Source: New Question History:N/A Comments:

Associated objective(s):

Given LGA-003, Primary Containment Control, in progress, mitigate the consequences of detectable or unknown Hydrogen concentration in the Suppression Chamber or Drywell, while operating the plant or on an exam, lAW LGA-003. LAS LICENSED OPS Page: 110 of 136 31 October 2012

( POOl lEVa ( HYDROGEN , Hold 5OPJ"MSion pooI __-4.5 in. ..... 3 ". (_rang.o) rCOtlltJlllfttlllCiftQl..OfIooAH.1L

.....lfteIt__

F THEIl LoN ......: Cannot hold _ Go ..@ ........._5 in.

___ Go"'@_3in. _..,..,..

.... __

• ... 0IC .....

F THEIl

__<:<lOIfMIOL;

"'_a__ ....t.OH1f.,.. .j.I j f----. *.... -1 & LGA-011 HYDROGEN START a.( ) 1. Operate recombiner 3S mi)(inq system (LGA-HG-1011201).

2. Monitor the follOWin9 While continuinq here: J,* Hyd rogen concentrations in dryWeU ilIl!1 suppression chamber.

• Oxygen concentrations in drywell and suppression chamber

• Both WRGMs for high 1'3(\ release rates: ... Slation Vent Stack WRGM Red lEO and Sia Vent $tad< Wide R""Il" Rad H. alarm at I N62 -POOO-B31M.

... SBGT WRGM Red LED and SBGT Wide R""Il" Gas Mon T"""ble alarm at 1 (2)PM07 J..A304. IF THEN Hydrogen de1RcUi!d or u.w-in "'" dryIIrf!Il m; !llJPPf"5sion ctwnber Geto@

LGA-003, Primary Containment Control Content/Skills Activities/Notes Parallel execution is also required because the symptomatic approach to emergency response precludes the prioritization 0 Objective 769.00.01 anyone action path since independence from initiating events and transients must be maintained.

• *While this procedure is structured along five parallel paths, perfonning actions simultaneously is most times not possible mainly because there are only so many people at your command. Therefore you must prioritize your actions to the degree you have manpower but this prioritization should not force you to tunnel in on one particular path at the expense of checking other paths. Generally, your first actions should be along the path of the specific entry condition(s) that caused you to enter LGA-003. Since many paths have you ultimately depressurize or ADS, suppression pool cooling should be started as soon as possible.

Once pool cooling is started and you are carrying out actions for the entry conditions of concern, scan across the flowchart to ensure other paths are under control. P:\PROCUPGD\APPROVED\TRAIN-LP\Ops\LGA*s\007-LGA-003.doc Page 3 of 45

ROExam Unit 1 is at rated conditions with TIP traces in progress.

TIP area is posted and verified clear. * 'B' TIP has just completed a trace and has been returned to the shield. * 'A' TIP is at position 0001. The 1 H13-P601-B211, RB TIP ROOM RAD HI/DOWNSCALE, has come in is verified to be Which of the following is the NEXT expected action(s) for the operator running the TIP A. close the 'A' TIP ball valve ONLY B. continue with the next TIP trace C. close the 'A' TIP ball valve and shutdown the TIP machine D. withdraw the 'A' TIP to the in-shield position and then close the 'A' TIP ball valve Answer: B Answer Explanation:

The TIP room HIGH Rad alarm is a normal occurrence while operating TIPs. So therefore the operator should continue performing TIP traces. If one were to assume this was an abnormal condition you would withdraw the TIPs to the in-shield position and close the ball valve. Per LOP-NR-06 you are to return the remaining TIPS not just 'A' TIP. so therefore closing the TIP ball valve only is incorrect.

Shutdown the TIP machine would normally be done after one withdraws the TIPs but in this case the operator is to continue on with the next TIP trace.

Reference:

LOR-1 H13-P601*B211, Reference provided during examination: Cognitive level: memory Level (RO/SRO): Tier: 3 Group: KIA: 2.3.13 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. 10 CFR Part 55 Content:41

.12 SRO Justification:

N/A Question Source: Question Comments:

LAS LICENSED OPS Page: 133 of 136 31 October 2012 DescriptionComputer Sensor Alarm DrawingActivating DeviceLOR-1 H 13-P601-B211 Revision 1 May 5, 2008 Reactor Building TIP Room Radiation High or Downscale NA008 (No HI -Per LRP-5800-3 as documented Attachment A of last LlP-AR-501A performed 1 1 1

RB TIP R O O M RAD H IIDOWNSCAL E AUTOMATIC ACTIONS 1Local Audible Alarm on Hi Rad. OPERATOR ACTIONS 1CHECK Rx Building TIP Room ARM on Panel 1 D21-P600 to determine radiation level. If radiation level is high due to 0 eration of TIP System VERIFY TIP Room is evacuated . bREFER to LOA-AR-1 01, Unit 1 Area Radiation Monitoring System Abnormal.

cMONITOR TIP Room dose rates until detector decays below alarm setpoint. If radiation level is high due to unknown reasons: EVACUATE TIP Room. DIRECT Radiation Protection to check validity of alarm to determine its source if possible.

cCHECK fuse on right-hand side on back of indicator and trip unit at 1 D21-P600, if indicator is pegged high. CHECK if 120 VAC MCC 136Y-2 Ckt.#19 has been turned off. of Use 10f2 Continuous LOR-1H13-P601-B211 Revision 1 May 5,2008 REFER to LOA-AR-101, Unit 1 Area Radiation Monitoring System Abnormal. If radiation level is downscale: CHECK associated power supply is ON. DEPRESS RESET Pushbutton. If monitor is still downscale:

1) NOTIFY Unit Supervisor.

2) NOTIFY Radiation Protection Supervisor.

3) CHECK fuse on left-hand side on back of indicator and trip unit at 1D21-P600, if indicator is pegged downscale.

4) CHECK if detector has been disconnected.

5) INITIATE Action Request to have monitor repaired.

Level of Use Continuous 20f2 ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED Area High Radiation INOTIFYIEVACUATE from affected RESTRICT access to I w RP 1 SURVEY and SAMPLE affected area. REFER to LGA-002, Containment I W SM 1 5REVIEW EALs and IMPLEMENT Emergency Plan as appropriate. CHE CK cause of hig h r adi ation -DETERMINE cause of high ETERMINED.

radiation. CHECK radiation monitors for abnonnal readings. CHECK stack gas release rate. CHECK Off-Gas System release rate and flow. CHECK area temperatures and Leak Detection System temperatures. CHECK continuous air monitors.

6.2 Ifradiation

levels peITIlit, 1 Eo""l CHECK affected areas for visible system leakage or loss of shielding. Level of Use Continuous LOA-AR-IOI 4 of 10 Revision 2 April 30, 2007 ACTIONIEXPECTED RESPONSE RESPONSE NOT OBTAINED Area High Radiation (continued) CHECK c ause of hi r adiation -7.1 ISOLA TE leak or SHUTDOWN STOPPED. affected system. 8CHECK Controlled Area -8.1 ESTABLISH a Controlled Area. I ESTABLISHED.

9CHECK exposure of any 9.1 READ dosimeters/TLD of I RP<" I personnel in affected area -personnel who may have been DETERMINED.

exposed. Level of LOA-AR-lOl 5 of 10 Revision 2 April 30,2007 E.6 If a TIP does inadvertently retract beyond shield, and TIP DRIVE PRM K60 I E reaches u scale value, PERFORM followin : REQUEST Rad Protection to confinn that personnel have been evacuated from TIP area and radiation levels at es tablished boundaries and floor of. 761' above do NOT exceed 40Rlhr. RETURN remaining TIPs to IN-Shield 'position.

CHECK all TIP ball valves close and VALVE OPEN Ii ts on Drive Control Units go dim. CHECK there are no indic a tions of system leakage into Containment.

Level of Use Continuous LOP-NR-06 200f28 Revision 26 June 10, 2008